OECD LOFT-T-3804, 'OECD LOFT Project, Quick-Look Report on ... · OECD LOFT-T--3804 "WW OECD LOFT...

298
OECD LOFT-T--3804 "WW OECD LOFT Project Quick-Look Report on OECD LOFT Experiment LP-FP-2 September 1985 NOTICE This report is for the benefit of OECD LOFT participants and their designees only This report has been prepared pursuant to the "Agreement on an OECD Project of the LOFT Experimental Programme." It is the policy of the Management Board that the Information contained in this report be used only for the benefit of the participants and the participants' designees. The contents of this report should not be disclosed to others or reproduced wholly or partially unless authorized in accordance with the laws, regulations, policies, or written permission of the appropriate project participant. Prepared by EG&G Idaho, Inc. under the direction of the U.S. Department of Energy, Idaho National Engineering Laboratory

Transcript of OECD LOFT-T-3804, 'OECD LOFT Project, Quick-Look Report on ... · OECD LOFT-T--3804 "WW OECD LOFT...

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OECD LOFT-T--3804

"WW

OECD LOFT Project

Quick-Look Report on OECD LOFT

Experiment LP-FP-2

September 1985

NOTICE

This report is for the benefit of OECD LOFTparticipants and their designees only

This report has been prepared pursuant to the "Agreement on anOECD Project of the LOFT Experimental Programme." It is thepolicy of the Management Board that the Information contained inthis report be used only for the benefit of the participants and theparticipants' designees. The contents of this report should not bedisclosed to others or reproduced wholly or partially unlessauthorized in accordance with the laws, regulations, policies, orwritten permission of the appropriate project participant.

Prepared by EG&G Idaho, Inc.under the direction of the U.S. Department of Energy,Idaho National Engineering Laboratory

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0

E0•

-ri 0

QUICK-LOOK REPORT ON OECD LOFTEXPERIMENT LP-FP-2

Authors: - .

J. P. Adams 0J. C. Birchley z

N. Newman N

E. W. Coryell ) MiM. L. Carboneau x ( -<

S. GuntayL. J. Siefken M O

Contributors: X (.

Y. Anoda C)J. Bagues

V. T. BertaE. Borioli0. Briney

J. EstebanD. L. Hagrman

R. HesbolK. J. McKennaD. C. Mecham

S. M. Modro

D. L. Batt, SupervisorFP-2 Experiment Section

D. W. Croucher, ManagerLOFT Experiment Planning and Analysis Branch

P. North, ManagerOECD LOFT Project

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QUICK-LOOK REPORT ON OECD LOFTEXPERIMENT LP-FP-2

Authors:

J. P. AdamsJ. C. Birchley

N. NewmanE. W. Coryell

M. L. CarboneauS. Guntay

L. J. Siefken

Contributors:

Y. AnodaJ. Bagues

V. T. BertaE. Borioli0. Briney

J. Esteban0. L. Hagrman

R. HesbolK. J. McKennaD. C. Mecham

S. M. Modro

Published September 1985

EG&G Idaho, Inc.Idaho Falls, Idaho 83415

Prepared for theU. S. Department of Energy

Idaho Operations OfficeUnder DOE Contract No. OE-ACO7-761DO1570

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ABSTRACT

Experiment LP-FP-2 was conducted on July 9, 1985, in the Loss-of-Fluid

Test (LOFT) facility at the Idaho National Engineering Laboratory under the

auspices of the Organization for Economic Cooperation and Development

(OECD). The objectives of this experiment were to obtain information on

the release of fission products from the fuel and transport of these

fission products in a vapor and aerosol environment from the primarycoolant system. The thermal/hydraulic boundary conditions during the

release and transport of fission products were based on a V-Sequence

accident. The emergency core cooling (ECC) injection was delayed until

specified temperature limits on the thermal shroud were reached, by which

time the desired time and conditions for fission product release and

transport were achieved. The plant was then brought to a safe condition

with full ECC injection. Specially designed fission product measurements

were made in the primary coolant system and blowdown suppression systemduring the transient and also up to 44 days thereafter, during which time

the plant was maintained in a quiescent state and the two systems were

individually isolated. This document provides an initial assessment of the

experiment that covers the initial conditions, sequence of events,

preliminary results of the fission product behavior within the

thermal/hydraulic boundary conditions, and comparisons of the results with

preexperiment calculations.

ii

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SUMMARY

Experiment LP-FP-2, conducted on July 9, 1985, was the second fission

product release and transport experiment and the eighth (and last)

experiment conducted in the Loss-of-Fluid Test (LOFT) facility at the Idaho

National Engineering Laboratory under the auspices of the Organization for

Economic Cooperation and Development. The principal objectives for this

experiment were to determine the fission product release from the fuel and

the subsequent transport of those fission products (in a predominantly

vapor/aerosol environment) from the primary coolant system. The initial

conditions were representative of commercial pressurized water reactor

(PWR) operations. The thermal/hydraulic boundary conditions during fission

product release and transport were based on a V-Sequence accident wherein a

low-pressure injection system (LPIS) line ruptures and the emergency core

cooling (ECC) injection is delayed until fuel rod cladding and control rod

melting and material relocation occurs. The transient was initiated by

scramming the reactor, inserting the center fuel module control rods, and,

after a specific delay, opening a break in the intact loop cold leg. A

second break (simulated LPIS line) was opened 222 s after reactor scram.

The first break was closed prior to fuel rod failure to provide a well

defined path for fission product transport. The transient continued until

control rod and fuel rod cladding melting and fission product release fromthe fuel occurred. The experiment was terminated by injection from both

ECC lines into the reactor vessel downcomer and lower plenum.

The initial assessment of data from instruments monitoring the upper

plenum (Fl) and the reactor vessel outlet (F2) sampling lines indicates

that fission products were sampled and the lines operated as expected.

Since the gamma spectrometer located on the Fl line (G6) failed prior to

the experiment, a gross gamma detector (remote area monitor) was placed on

the top of the reactor vessel and detected fission products from both the

fuel/cladding gap and the fuel as they were transported through the Fl

line. The spectrometer (G2) that monitored the combined effluent from the

Fl and F2 lines during the transient measured several isotopes of xenon and

krypton.

iii

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Radiation scans of the simulated LPIS line at the time of the first

postexperiment containment entry indicate that fission products were

collected by the deposition coupons in this line. Metal temperatures at

the deposition coupon locations in both the reactor vessel and simulated

LPIS line were approximately 100 K (180'F) higher than saturation, which

was sufficiently high to ensure fission product deposition in steam.

The gamma spectrometer in the simulated LPIS line (G5) sampled various

isotopes. In order of volatility, these were xenon, iodine, cesium,

tellurium, and rubidium.

The G5 gamma spectrometer, the F2 and F3 aerosol sampling systems, and

the D2 and D3 deposition spool pieces appear to have operated as designed.

High background may limit the applicability of the GI (primary system),

G2 [blowdown suppression tank (BST) vapor], and G3 (BST liquid) gamma

spectrometers during the early part of the posttransient phase, and one or

more of the Dl protected coupons may have been exposed to reflood. There

was a loss of data from the G6 gamma spectrometer, which was partially

recovered with data from a remote area monitor. Also, significant data

were obtained from grab samples and Health and Safety instrumentation.

Experiment predictions indicated that, in order to produce the desired

fission product release and transport boundary conditions, the thermal

transient should produce cladding temperatures of 2100 K (3320°F) or higher

for a minimum of three minutes. During the experiment, cladding

temperatures exceeded 2100 K (3320*F) for at least 4-1/2 min, which was 50%

longer than the minimum 3 min identified prior to the experiment. As a

result, the final fission product concentrations in the primary coolant

system and blowdown suppression tank are expected to be higher than those

which were predicted, thus enhancing the detectability of low-yield fission

products.

Comparison with the measured thermal/hydraulic response showed that

the predictions were very adequate as a planning tool for this experiment.

The timing and extent of the core thermal response was closely predicted

iv

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with the exception of the lack of steam starvation in the upper parts of

the center fuel module. This discrepancy resulted from a

larger-than-predicted center fuel module steam flow which, in turn, is

judged to have been caused by greater-than-calculated depressurization rate

during the high temperature period of the transient. The resistance in the

simulated LPIS line was much greater than modeled. This led to a higher

primary system pressure at the start of the high temperature period and a

continued depressurization during the high temperature period as opposed to

the nearly flat pressure response that was predicted. Inability to

accurately predict the flow resistance in this line was recognized prior to

the experiment as an area of experimental uncertainty, and adequate

contingency measures were included in the Experiment Operating Procedure.

Based on the preliminary information presented herein, the data

obtained from this experiment are considered adequate to meet the fission

product measurement objectives and, ultimately, the overall experiment

objectives, which were to provide data to assess the fission product

release and transport during the early phases of a risk dominant accident

and the capability of computer codes to predict the same.

v

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ACKNOWLEDGEMENTS

Since this is the Quick Look Report documenting the final LOFT

experiment, it is appropriate to acknowledge and thank all who have

contributed to the LOFT program, either during the original NRC test

program or during the OECD sponsored program. In a very real sense, the

success of this, the final experiment conducted in LOFT, was the direct

result of all their efforts. Without their professional contributions,

LOFT would not have been prepared for the rigors that were imposed on both

the facility and the staff in preparing for and carrying out LP-FP-2. To

all these, present and past associates, the authors express their thanks.

Those who contributed directly to the production of this report fall intotwo basic categories: contributors--those who contributed ideas, analyses,

time, and support to determine the response of the system during the

experiment--and authors--those who, in addition to the above, also

contributed to the writing found in this report. These are all

acknowledged on the title page. In addition, the authors wish to

acknowledge Darwin Grigg, who provided an excellent service as technical

editor; without his efforts, this report could not have been completed on

the schedule achieved.

vi

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CONTENTS

ABSTRACT .............................................................. ii

SUMMARY ............................................................... iii

ACKNOWLEDGEMENTS ...................................................... vi

1. INTRODUCTION ..................................................... I

1.1 Objectives ................................................. 4

1.2 Experiment Description ..................................... 5

1.3 Systems Description ........................................ 8

2. EXPERIMENT CONDUCT ............................................... 12

2.1 Initial Conditions and Operational Setpoints ............... 12

2.2 Chronology of Events ....................................... 17

3. PCS THERMAL/HYDRAULIC RESULTS .................................... 23

3.1 Blowdown Hydraulics ........................................ 23

3.2 Core Thermal Response ...................................... 27

3.3 Comparison with Calculations ............................... 33

3.4 Metal/Fluid Conditions Near the FPMS ....................... 40

3.5 Summary .................................................... 40

4. FISSION PRODUCT RESULTS .......................................... 43

4.1 ORIGEN 2 Results for the LP-FP-2 Experiment ................ 43

4.2 Results of the Elemental Release Calculations .............. 46

4.3 FPMS Performance ........................................... 51

4.4 Instrument Operation ....................................... 53

4.5 Preliminary Results ........................................ 56

4.5.1 Fl and F2 Sample Lines ............................. 574.5.2 Deposition Measurements ............................ 574.5.3 G5 Gamma Spectrometer .............................. 60

vii

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4.5.4 GI, G2, and G3 Gamma Spectrometers ................. 694.5.5 Grab Samples ....................................... 69

4.6 Preliminary Analysis of the BST and G5 Data ................ 69

4.7 Potential for Meeting Fission Product MeasurementObjectives ................................................ 76

4.8 Future PIE Plans ..................................... ...... 77

5. CONCLUSIONS ...................................................... 79

6. REFERENCES ...................................................... 81

APPENDIX A--REVISIONS TO THE EXPERIMENT SPECIFICATION DOCUMENT FOREXPERIMENT LP-FP-2 ......................................... A-1

APPENDIX B--FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2 ............ B-1

APPENDIX C--DESCRIPTION OF THE LOFT SYSTEM AND INSTRUMENTATION ......... C-i

APPENDIX D--PCS THERMAL/HYDRAULIC RESPONSE .......................... D-l

APPENDIX E--CORE THERMAL RESPONSE ..................................... E-l

APPENDIX F--COMPARISON OF THERMAL/HYDRAULIC DATA WITH PREEXPERIMENTCALCULATIONS................................... F-i

APPENDIX G--SPECIAL INSTRUMENTATION ................................... G-l

APPENDIX H-- SCDAP/RELAP5/TRAP-MELT CODE CALCULATION AND DATACOMPARISONS ............................. H-1

APPENDIX I--QUALIFIED TRANSIENT DATA PLOTS ............................ I-1

APPENDIX J--ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT ................ J-l

FIGURES

1. Preexperiment core power history .................................. 7

2. FPMS schematic ..................................................... 9

3. Axonometric representation of the LOFT primary coolant system 11

4. Primary system pressure (short term) .............................. 22

5. Primary system pressure (full term) .............................. 22

viii

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6. Intact loop hot leg density ....................................... 24

7. Conductivity level probe response above Fuel Assembly 3 ........... 25

8. Comparison of cladding temperatures at the 1.14-, 0.38-, and0.28-m (45-, 15-, and 11-in.) elevations in Fuel Assembly 2with saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................. 28

9. Comparison of cladding temperatures at the 1.07-, 0.69-, and0.25-m (42-, 27-, and 10-in.) elevations in Fuel Assembly 5with saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................. 28

10. Comparison of three guide tube temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits) ............................... 29

11. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits) ............................... 31

12. Comparison of four external wall temperatures at the 1.07-,0.81-, 0.69-, and 0.25-m (42-, 32-, 27-, and 10-in.)elevations on the south side of the flow shroud. (SeeAppendix I for thermocouple qualification limits) ................ 31

13. Comparison of cladding temperatures at the 1.24-, 0.99-,0.71-, and 0.28-m (49-, 39-, 28-, and 11-in.) elevations inFuel Assembly 2. (See Appendix I for thermocouplequalification limits) ............................................ 32

14. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5 with saturation temperature.(See Appendix I for thermocouple qualification limits) ........... 34

15. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (short term) ................. 36

16. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (full term) .................. 36

17. Comparison of the measured cladding temperature at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BDI.(Thermocouple qualified throughout) .............................. 38

18. Comparison of the measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BDI.(Thermocouple qualified to 1720 s) ............................... 38

ix

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19. Comparison of the measured cladding temperature at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BOl.(Thermocouple qualified to 1510.3) .............................. 39

20. Comparison of metal and fluid temperatures at the lower 01deposition coupon location with saturation temperature. (SeeAppendix I for thermocouple qualification limits) ................ 41

21. Comparison of fluid temperature in the Fl aerosol sample linewith saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................ 41

22. Release rate functions taken from NUREG-0772 ..................... 47

23. Comparison of measured cladding temperatures at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 47

24. Comparison of measured cladding temperatures at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGER1 calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 48

25. Comparison of measured cladding temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 48

26. Input model of the center fuel assembly for the CORSOR/TIGERIcalculation ...................................................... 49

27. Elemental cumulative fractional release inventories calculatedusing TIGERI ..................................................... 49

28. Release rate functions for Cs, I, Sb, Xe, Kr, and Tecalculated using TIGERI .......................................... 50

29. Comparison of the radiation area monitor response on the Flaerosol sample line with fuel centerline temperature(TC-5108-027). (See Appendix I for thermocouple qualificationlimits) .......................................................... 58

30. Measured pressure upstream of the critical orifice in theFl aerosol sample line ........................................... 58

31. Measured pressure upstream of the critical orifice in theF2 aerosol sample line ........................................... 59

x

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32. Measured metal temperatures at the Dl deposition couponlocations. (See Appendix I for thermocouple qualificationlimits) ..........................................................

33. Measured metal temperatures at the 02 and D3 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits) ..........................................................

63

63

34. Measured 1311corrected for

35. Measured 1321corrected for

36. Measured 1331corrected for

37. Measured 1341corrected for

38. Measured 1351corrected for

39. Measured 8 8 Rbcorrected for

concentration in the simulated LPIS line (notplateout) ..........................................

concentration in the simulated LPIS line (notplateout) ..........................................

concentration in the simulated LPIS line (notplateout) ..........................................

concentration in the simulated LPIS line (notplateout) ..........................................

concentration in the simulated LPIS line (notplateout) ..........................................

concentration in the simulated LPIS line (notplateout) ..........................................

40. Measured 13 5 Xe concentration in the simulated LPIS line (notcorrected for plateout) ..........................................

41. Measured 13 2Te concentration in the simulated LPIS line (notcorrected for plateout) ..........................................

42. Measured 138 Cs concentration in the simulated LPIS line (notcorrected for plateout) ..........................................

43. Estimates of the elemental Cs, I, and Rb mass concentrationsin the simulated LPIS line based on isotopic activitiesmeasured by the G5 gamma spectrometer ............................

B-l. FPMS Schematic ..................................................

B-2. Schematic of Fl and F2 aerosol sample systems ...................

B-3. Sample line probe ...............................................

B-4. Cyclone separator/isolation valve ...............................

B-5. Impactor and filter train .......................................

B-6. Three stage virtual impactor ....................................

64

64

65

65

66

66

67

67

68

75

B-2

B-4

B-6

B-7

B-7

B-8

xi

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B-7. Simulated LPIS line components .................................. B-10

C-I. Axonometric representation of the LOFT system for ExperimentLP-FP-2 ......................................................... C-2

C-2. Schematic of the LOFT primary and emergency core cooling

systems .......................................................... C-3

C-3. Simulated LPIS breakline instrumentation ........................ C-4

C-4. LOFT reactor vessel ............................................. C-5

C-5. LOFT cladding and guide tube thermocouple locations ............. C-8

C-6. Center fuel assembly instrumentation locations .................. C-9

C-7. LOFT steam generator instrumentation ............................ C-10

C-8. Reactor vessel upper plenum instrumentation locations ........... C-l1

D-1. Intact loop hot leg densities ................................... D-2

D-2. Intact loop cold leg densities .................................. D-2

D-3. Broken loop hot leg densities ................................... D-3

D-4. Broken loop cold leg densities ................................... D-3

0-5. Comparison of upper plenum fluid temperature with saturationtemperature. (See Appendix I for thermocouple qualificationlimits) ........................................................... D-4

0-6. Conductivity level probe response above Fuel Assembly 3 ......... D-6

D-7. Conductivity level probe response in Fuel Assembly I ............ D-7

D-8. Conductivity level probe response in Fuel Assembly 3 ............ D-8

D-9. Response of SPND in Fuel Assembly 2 to core uncovery ............ D-10

D-10. Response of SPND in Fuel Assembly 4 to core uncovery ........... D-10

V-li. Response of SPND in Fuel Assembly 6 to core uncovery ........... D-l1

D-12. Averaged BST liquid level ...................................... D-11

D-13. Primary coolant system mass inventory .......................... D-12

D-14. Comparison of fluid temperature in the Fl aerosol sampleline with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-16

xii

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D-15. Comparison of fluid temperature in the F2 aerosol sampleline with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-16

D-16. Comparison of metal and fluid temperatures at theD2 deposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-17

0-17. Comparison of fluid temperature at the 03 deposition couponlocation with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-17

D-18. Comparison of metal and fluid temperatures at the DI upperdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... 0-18

D-19. Comparison of metal and fluid temperatures at the Dl middledeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-18

D-20. Comparison of metal and fluid temperatures at the Dl lowerdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-19

D-21. Comparison of primary system pressure with pressuresmeasured in the accumulators ................................... D-19

D-22. Primary system pressure during the posttransient phase ......... 0-21

D-23. Primary system fluid temperature during the posttransientphase .......................................................... D-21

E-1. Comparison of cladding temperatures at the 1.14-, 0.38- and0.28-m (45-, 15- and 11-inch) elevations in Fuel Assembly 2with saturation temperature during early stages of heatup(600 to 1000 s). (See Appendix I for thermocouplequalification limits) ........................................... E-2

E-2. Comparison of cladding temperatures at the 1.07-, 0.69- and0.25-m (42-, 27- and 10-inch) elevations in the center fuelassembly with saturation temperature during early stages ofheatup (600 to 1000 s). (See Appendix I for thermocouplequalification limits) ........................................... E-2

E-3. Effect of flow changes on rate of temperature increasemeasured at 0.69-m (27-inch) elevation on fuel rod claddingin Fuel Assembly 4 (700 to 1400 s). (See Appendix I forthermocouple qualification limits) .............................. E-3

xiii

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E-4. Effect of presence of control rods on guide tube temperatureincrease at the 0.69-m (27-inch) elevation in the center fuelassembly (1100 to 1600 s). (See Appendix I for thermocouplequalification limits) ........................................... E-5

E-5. Comparison of fluid temperatures at upper tie plate aboveFuel Assembly 4 with saturation temperature (600 to 2000 s).(See Appendix I for thermocouple qualification limits) .......... E-5

E-6. Comparison of fluid temperatures at lower tie plate belowFuel Assembly 4 with saturation temperature (600 to 2000 s).(See Appendix I for thermocouple qualification limits) .......... E-6

E-7. Effect of metal-water reaction on guide tube temperatureincrease at 0.69-m (27-inch) elevation in center fuelassembly (600 to 1600 s). (See Appendix I for thermocouplequalification limits) ........................................... E-6

E-8. Measurement of gross gamma activity near reactor vessel head(600 to 2000 s) ................................................. E-9

E-9. Comparison of fluid temperatures at different horizontallocations on center fuel assembly upper tie plate(600 to 1800 s). (See Appendix I for thermocouplequalification limits) ........................................... E-9

E-l0. Momentum flux in reactor vessel downcomer (-200 to 2000 s) ..... E-10

E-11. Cladding temperatures at 0.25-m (10-inch) elevation incenter fuel assembly during high temperature stage oftransient (1200 to 2000 s). (See Appendix I forthermocouple qualification limits) ............................. E-12

E-12. Comparison of cladding temperature at 0.38-m (15-inch)elevation in Fuel Assembly 4 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouplequalification limits) .......................................... E-12

E-13. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 8 with saturation temperature (600 to 1900 s).(See Appendix I for thermocouple qualification limits) ......... E-14

xiv

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E-14. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 4 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ...................... k ...... E-14

E-15. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 2 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................ E-15

E-16. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-n(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 6 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................. E-15

E-17. Comparison of cladding temperatures at 0.28-, 0.71-, 0.99-and 1.24-m (11-, 28-, 39- and 49-inch) elevations in FuelAssembly 2 (1400 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................ E-16

E-18. Comparison of cladding temperatures at 0.28-m (11-inch)elevation on fuel rods in peripheral assemblies 2, 4, and6 close to shroud (600 to 2000 s). (See Appendix I forthermocouple qualification limits) ............................. E-16

E-19. Comparison of fluid temperature at upper tie plate abovecenter fuel assembly with saturation temperature duringreflood (1750 to 1850 s). (See Appendix I for thermocouplequalification limits) .......................................... E-18

E-20. Effect of material relocation on cladding temperatures at0.69-m (27-inch) elevation in center bundle during reflood(600 to 2100 s). (See Appendix I for thermocouplequalification limits) .......................................... E-18

F-1. RELAP5/MOD2 nodalization diagram ................................ F-4

F-2. TRAC-LOFT nodalization .......................................... F-1l

F-3. TRAC-LOFT center assembly rod grouping .......................... F-13

F-4. TRAC-LOFT peripheral assembly rod grouping .................... F-14

F-5. Secondary system pressure (0 to 400 s) .......................... F-16

F-6. Secondary system pressure (0 to 2000 s) ......................... F-16

F-7. Secondary system liquid level ................................... F-17

F-8. Primary system hot leg pressure (0 to 400 s) .................... F-19

xv

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F-9. Primary system hot leg pressure (0 to 2000 s) ................... F-19

F-10. Pressure drop along LPIS line .................................. F-21

F-1l. LPIS line mass flow rate ....................................... F-21

F-12. Progression of core uncovery ................................... F-24

F-13. Fuel rod cladding temperature in center fuel assembly,0.25-m (10-in.) elevation. (Thermocouple qualifiedthroughout) .................................................... F-24

F-14. Fuel rod cladding temperature in center fuel assembly,0.69-m (27-in.) elevation. (Thermocouple qualified to1720 s) ........................................................ F-26

F-15. Fuel rod cladding temperature in center fuel assembly,1.07-m (42-in.) elevation. (Thermocouple qualified to1510 s) ........................................................ F-26

F-16. Fuel rod cladding temperature in peripheral fuel assembly,0.66-m (26-in.) elevation. (Thermocouple showed possibleshunting after 1700 s) ......................................... F-28

F-li. Shroud outer wall temperature at 0.69-m (27 in.) elevation.(Thermocouple qualified to 1790 s) .......................... F-28

G-l. Response of SPND at the 0.69-m (27-in.) elevation in FuelAssembly 5 ...... .......................................... G-3

G-2. Response of SPND at the 0.28-m (11-in.) elevation in Fuel

Assembly 5 ---..................................... . ....... G-5

G-3. Iodine species sampler ...................................... G-9

G-4. Cutaway of the LOFT reactor vessel illustrating the locationof PSU detectors ... o...................... ... G-12

G-5. Planar view of the LOFT reactor vessel illustrating thelocation of PSU detectors ...... ... .................... G-13

G-6. Normalized current response of PSU detectors (0 to 120 s) ...... G-14

G-7. Normalized current response of PSU detectors (0 to 1800 s) ...... G-16

G-8. Normalized pulse height response of PSU detectors(0 to 1800 s) .............................. G-17

G-9. Normalized pulse height response of PSU detectors(1800 to 3600 s) ...... .......... ..... ............... G-18

xvi

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H-I. Comparison of measured hot leg pressure with pressurecalculated using the integrated code ............................ H-6

H-2. Comparison of measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits) ........................................... H-6

H-3. Calculated cladding temperature at the 1.26-m (50-in.)elevation in Fuel Assembly 5 made using the integrated code.(See Appendix I for thermocouple qualification limits) .......... H-8

H-4. Comparison of measured cladding temperature at the 0.94-m(37-in.) elevation in Fuel Assembly 4 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits) ........................................... H-8

TABLES

1. Initial conditions for Experiment LP-FP-2 ........................ 13

2. Operational setpoints for Experiment LP-FP-2 ..................... 16

3. Chronology of events for Experiment LP-FP-2 ...................... 18

4. Selected ORIGEN2 inventory results for the Experiment LP-FP-2center fuel bundle at 430 MWD/MTU burnup ......................... 45

5. Comparison of TIGERl and CORSOR cumulative release fractions forExperiment LP-FP-2 (at 1800s) .................................... 52

6. Experiment LP-FP-2 fission product measurement system sequence

of events ........................................................ 54

7. Iodine species identified by G5 spectrometer ..................... 61

8. Non-iodine species identified by G5 spectrometer ................. 62

9. BST liquid grab sample preliminary results ....................... 70

10. 8ST vapor grab sample gamma spectroscopy results ................. 71

11. BST vapor grab sample mass spectroscopy results .................. 71

12. Cumulative release fractions to the BST ........................ 73

13. Planned postirradiation examination for Experiment LP-FP-2 ....... 78

C-l. Initial conditions for Experiment LP-FP-2 ....................... C-7

xvii

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E-1. Times for the center fuel module thermocouples to reach1800 K (2780°F) ................................................. E-8

F-1. Initial conditions for Experiment LP-FP-2 ....................... F-8

G-1. Measured 131I concentrations .................................... G-7

G-2. Measured 1311 species admixture ................................. G-8

H-l. Description of modeling of reactor core ......................... H-3

H-2. Conduct of Experiment LP-FP-2 assumed in preexperimentprediction ...................................................... H-4

1-1. Listing of qualified data fiche ................................. 1-2

xviii

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QUICK-LOOK REPORT ON OECD LOFT

EXPERIMENT LP-FP-2

1. INTRODUCTION

This report documents the preliminary results and analyses of

Experiment LP-FP-2, which was conducted on July 9, 1985, in the

Loss-of-Fluid Test (LOFT) facility located at the Idaho National

Engineering Laboratory (INEL). This fission product release and transport

experiment was the eighth and final experiment conducted under the auspices

of the Organization for Economic Cooperation and Development (OECD).

Experiment LP-FP-2 provides information on the release and transport

of fission products and aerosols in a severe core damage scenario. The

nature of the phenomena governing fission product and aerosol release and

transport can be linked to potential pressurized water reactor (PWR) system

thermal hydraulics and core thermal response leading to fuel failure and

fission product transport behavior. The fuel rod cladding temperatures in

the center fuel module (CFM) exceeded 2100 K (3320'F) for an estimated

4-1/2 min before temperature limits for experiment termination were

reached. The 4-1/2 min fission product release and transport transient

simulates the initial part of a severe damage transient without emergency

core cooling (ECC) system operation wherein the core thermal induced damage

originated from a V-Sequence scenario.

Probabilistic Risk Assessment (PRA) studies 1 have shown that the

interfacing systems loss-of-coolant accident (LOCA), a hypothetical event

first postulated in the Reactor Safety Study2 and labeled the V sequence,

has a significant contribution to the risk associated with PWR operation.

Therefore, this risk dominant accident sequence was selected as the

thermal/hydraulic environment in which fission product release and

transport would be measured in Experiment LP-FP-2. The specific

interfacing systems LOCA associated with the significant operational risk

is a pipe break in the low pressure injection system (LPIS), also called

the residual heat removal system (RHRS). This system typically serves two

I

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functions in a commercial PWR: (a) it provides emergency coolant injection

for core recovery during intermediate and large break LOCAs and (b) it

provides for decay heat removal during normal shutdown. The RHRS

represents a potential path for release of primary coolant from the

containment. If core cooling were not maintained during such an event and

failure of the fuel rods were to occur, fission product release to the

environment could occur through concomitant failure of the auxiliary

building.

Experiment LP-FP-2 simulated the system thermal/hydraulic and core

uncovery conditions during fission product release and transport that are

calculated to occur in a four-loop PWR from rupture of an RHRS pipe as a

result of a V sequence accident. The initial conditions were

representative of commercial PWR operations. The break size resulted in a

depressurization that was bounded by previously conducted LOFT

Experiments L8-2 and L5-1 3 ' 4 on the upper end and by

Experiments L3-1, 5 ' 6 L3-5/3-5A,7, 8 and L3-6/L8-1 9 ' 10 on the lower

end. The transient was initiated by a reactor scram followed by insertion

of the central fuel assembly control rods (designed to provide typical

control rod behavior during the transient). A break line in the intact

loop cold leg was then opened to start the depressurization. A second

break path, which simulated the LPIS line, was opened in the broken loop

cold leg. The intact loop cold leg break was closed in accordance with

procedure, but the subsequent depressurization was too slow and the

pressure remained too high for FPMS operation. Therefore, both this break

and the power operated relief valve were opened for a brief period to

depressurize the primary system prior to fission product release. The core

was allowed to uncover and to heat up until a high temperature trip on the

shroud outside wall was reached. By that time, the central fuel assembly

had reached an estimated maximum temperature in excess of 2400 K (3860°F)

and had been above 2100 K (3320'F) for at least 4-1/2 min. The emergency

core cooling (ECC) system was then activated to reflood the reactor vessel

and recover the plant.

2

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The requirements imposed on Experiment LP-FP-2 from the standpoint of11

facility decontamination and recovery were:

1. Experiment LP-FP-2 must be conducted with peripheral assembly

fuel rod cladding temperatures limited to 1533 K (23000 F).

2. The structural integrity of the center fuel assembly must be

maintained to facilitate removal from the reactor vessel.

To meet the above facility requirements, a center fuel assembly was

designed and fabricated specifically for this experiment. This fuel

assembly contained 9.74-wt%-enriched, 2.41-MPa-(350-psia)-prepressurized

fuel rods and was separated from the peripheral fuel rods by a 2.5-cm

(l-in)-thick, canned, zirconium-oxide thermal shroud. The center fuel

assembly was designed to enable the 9.74-wt%-enriched fuel rods to heat up

above 2100 K (3320°F) while maintaining the peripheral fuel rods below

their temperature limit sufficiently long to allow fission product release

and transport.

Section 2 presents an evaluation of the plant performance during

Experiment LP-FP-2, and includes a summary of specified11 and measured

initial conditions, a list of operational setpoints, a chronological

listing of identifiable significant events, and a description of the LOFT

system geometry. Section 3 presents a summary of the PCS thermal/hydraulic

boundary conditions measured during Experiment LP-FP-2, including summariesof the core thermal response and the comparison between the measured data

and the preexperiment calculations. Section 4 presents the fission product

measurement results, including a preliminary assessment of the

achievability of the measurement objectives. Section 5 presentsconclusions based on a preliminary examination of the results discussed in

Section 3 and relates those conclusions to the experiment objectives.

Appendix A contains revisions to the Experiment Specification Document

(ESD), which were made subsequent to its issue. That appendix may be

removed from this report for insertion of the replacement pages into the

ESO. Appendix B contains a description of the FPMS, and Appendix C, a

3

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description of the LOFT system and thermal/hydraulic instrumentation.

Appendix D contains the detailed primary system thermal/hydraulic

response. Appendix E contains a discussion of the detailed core thermal

response, and Appendix F contains the detailed comparison of data with

predictions. Appendix G contains the results from three special instrument

systems which were installed specifically for this experiment and which

were beyond the scope of the experiment design. Appendix H presents a

brief summary of the comparison of the data with predictions made using a

combined RELAP5/MOD2, SCDAP, and TRAP-MELT computer code. Appendix I

contains plots of all the qualified data recorded during the transient

phase, and Appendix J contains the source term calculations. These last

two appendices are contained on microfiche inside the back cover.

1.1 Objectives

The governing objective for Experiment LP-FP-2 was:

Obtain fission product release, transport, and deposition data during

the early phases of a risk dominant reactor transient to establish a

benchmark data base for:

1. Assessing the understanding of the physical phenomena controlling

reactor system fission product behavior.

2. Assessing the capability of computer models to predict reactor

system fission product release and transport.

To support that governing objective, the following two

thermal/hydraulic and four fission product objectives were defined:

Thermal/hydraulic

1. Provide LPIS interfacing system LOCA thermal/hydraulic conditions

from the initiation of the LPIS pipe break through the early

phases of severe core damage.

4

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2. Provide transient fuel rod cladding temperatures in the center

fuel assembly up to the rapid metal-water reaction temperature of

2100 K (3320 0 F) and concomitant aerosol generation from the

(Ag-In-Cd) control rods.

Fission Product

1. Determine the fraction of the volatile fission products (Cs, I,

Te, Xe, Kr) and aerosols released to and from the upper plenum

region.

2. Determine the fraction of volatile fission products and aerosols

transported out of the primary coolant system.

3. Determine the retention of volatile fission products on

representative primary coolant system surfaces in the plenum and

piping.

4. Determine the general mass balance of volatile fission products

in the fuel, primary coolant system and blowdown tank.

Due to the preliminary nature of the data analysis for this

experiment, the governing objective will not be discussed. However, a

preliminary assessment of each of the specific measurement objectives is

presented.

1.2 Experiment Description

As with OECD LOFT Experiment LP-FP-1 12 (a previously conducted

fission product experiment), Experiment LP-FP-2 consisted of four phases,

designated (1) fuel preconditioning, (2) pretransient, (3) transient, and

(4) posttransient. The four phases were contiguous and had specificphenomenologically defined beginnings and endings.

5

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The purpose of the fuel preconditioning phase, in conjunction with the

pretransient phase, was to subject the 9.72-wt%-enriched fuel rods to a

minimum burnup of 325 MWD/MTU. This was done by operating the core at a

thermal power of 32 MW for 80 h, shutting down for 75 h, and operating at

26.5 MW for a period of 80 h. This phase started when the plant was heated

up just prior to power operations and ended after the required burnup in

this phase (at least 252 MWD/MTU) was reached.

The pretransient phase had, as its purposes, the completion of the

burnup in the 9.72-wt%-enriched fuel and the establishment of the initial

conditions for the experiment. Figure I shows the pretransient power

history for this experiment. The initial condition requirements included

short-lived decay heat buildup (685 kW at 200 s after reactor scram),

pressure, temperature, flow, etc., which simulated typical operation of

commercial PWRs. This phase began upon termination of the preconditioning

phase and ended upon initiation of the transient phase after 30 h at 31 MW

followed by 15 h at 26.5 MW.

The transient phase started with a reactor scram and ended when the

simulated LPIS line was closed. Plant actions taken during this phase

comprised turning off the primary coolant pumps and inserting the central

fuel assembly control rods within 20 s of reactor scram, opening first the

intact loop cold leg and then the broken loop hot leg (simulated LPIS line)

breaks, closing the intact loop cold leg break, and then recycling the

intact cold leg break and cycling the PORV prior to fission product

release. This phase was terminated when the shroud external temperature

reached 1517 K (2271*F), at which time the reflood of the reactor vessel

was initiated. The maximum cladding temperature measured during this

experiment exceeded 2400 K (3860°F), and the time at temperature [time with

cladding temperatures in excess of 2100 K (3320°F)] was 4-1/2 min.

The final, or posttransient, phase consisted of a time interval of

44 d during which time the redistribution of fission products in the gas

and liquid volumes in the blowdown suppression tank and the leaching of

fission products from the damaged fuel rods in the reactor vessel were

6

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40 I I

30

20

0~1010 r

1 :

-500 -400 -300 -200 -100Time (hr)

Figure 1. Preexperiment core power history.

0

7

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measured. This phase initiated upon closure of the simulated LPIS line,

which terminated the blowdown and initiated the reflood of the reactor

vessel, and ended 44 d later.

1.3 Systems Description

A fission product measurement system (FPMS) was designed and

fabricated for use during Experiment LP-FP-2. This system (illustrated inFigure 2) consisted of three basic subsystems: the gamma spectrometer

system, the deposition coupons, and the filter sampling system. Each of

these is briefly described in turn in this section. Appendix B contains a

more detailed description of the FPMS.

The gamma detection sampling system included five different sample

locations: Gl (spectrometer), which sampled from the reactor vessel lower

plenum or, alternately, from the intact loop hot leg; G2 (spectrometer),

which sampled from the blowdown suppression tank vapor spaces;

G3 (spectrometer), which sampled from the blowdown suppression tank liquid

spaces; G5 (spectrometer), which sampled from the simulated LPIS line; and

G6 (gross gamma monitor), which sampled from the upper plenum. (G4 was

used during Experiment LP-FP-l and was not used in this experiment.) Each

gamma spectrometer was designed to operate remotely and could be calibrated

using a 2 28 Th source mounted on a collimator wheel. With the exception

of G5 and G6, this system operated only during the posttransient phase.

(Additionally, the G-2 spectrometer measured the activity from the combined

Fl and F2 aerosol sample line during the transient.) G5 and G6 operated

during the transient and posttransient phases.

The deposition sampling system consisted of six stainless steel

coupons and two deposition spool pieces. Two of these were located at each

of three elevations above the central fuel assembly (for a total of six

coupons, collectively designated DI). At each elevation, both coupons were

exposed to the fluid stream during the transient phase. One coupon at each

elevation was to be isolated from the fluid prior to initiation of reflood

while the other coupon remained exposed. This system design, if

successfully operated, would have allowed differentiation between the

8

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F2

5 4066

Figure 2. FPMS schematic.

9

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leaching and deposition during the reflood and the deposition during the

heatup phase. During the experiment, the cover did not seal around the dry

coupons and contact with the reflood water may have occurred. Two

deposition spool pieces, located at the inlet and outlet of the simulated

LPIS line header, were designated D2 and D3, respectively. These coupons

were designed to provide a measurement of the primary coolant system

surface deposition of volatile fission products during the heatup or

transient phase. Since this line was isolated prior to reflood, these

coupons did not experience any deposition or leaching subsequent to the

transient phase.

The final FPMS subsystem was the aerosol and steam sampling system.

This system was designed to provide a sample of the vapor and aerosols

generated during the heatup phase of the experiment. The Fl filter

sampling line consisted of the following major components, in order:

sample line probe, dilution gas supply, cyclone separator and isolation

valve, dilution filter, virtual impactor, collection filters, infrared

moisture detectors, and hydrogen recombiner. The F2 sampling line was

similar except there were no dilution gas supply and moisture detectors.

The F3 filter sampling line consisted of a filter, flow venturi, and Dl

and D3 deposition coupons. The three sample locations were: Fl, 180 cm

(70.75 in.) above the top of the lower tie plate and located directly above

the center fuel assembly; F2, the broken loop hot leg spool piece just

outside of the upper plenum; and F3, the exit of the simulated LPIS line

header.

Figure 3 is an axonometric representation of the LOFT primary coolant

system. The system consists of the reactor vessel, which houses the 1.68-m

(5.5-ft) nuclear core; an intact loop, which represents three of four loops

of a four-loop PWR and which contains active components (steam generator,

pumps, pressurizer, etc.); a broken loop, which represents the fourth loop;

and the blowdown suppression tank, which collected the effluent from the

primary coolant system. The LOFT PCS is volumetrically scaled to a

commercial PWR, using the ratio of core thermal powers (LOFT/PWR) as the

scaling constant. Additional details on the LOFT system and the scaling

basis used in its design are available in Appendix C and in References 13

and 14.

10

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C C C

Intact loop Broken loop

Figure 3. Axonometric representation of the LOFT primary coolant system.

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2. EXPERIMENT CONDUCT

The experiment conduct is described in this section. The initial

conditions and the operational setpoints for the experiment are presented

in Section 2.1. Section 2.2 briefly describes the sequence of events that

occurred during the experiment.

2.1 Initial Conditions and Operational Setpoints

A summary of the specified and measured system conditions immediately

prior to Experiment LP-FP-2 is given in Table 1. All initial conditions

were within the limits specified by Reference 11 except for the liquid

level in the blowdown suppression tank, which was <1% low. Since no

attempt had been made to simulate the containment with the blowdown

suppression tank, this single out-of-specification value is judged not to

have affected the experiment outcome.

The operational setpoints specified in the ESD for this experiment are

listed in Table 2, together with the measured values. Two of the operator

actions occurred prior to reaching the respective operational setpoint:

isolation of the gamma densitometer sources (683 s early) and initiation of

ECCS flow (1.1 s early). The latter is judged to have no effect on the

data since all FPMS lines were closed prior to injection of the

ECCS water. There are two effects caused by the early isolation of the

gamma densitometer sources: (a) loss of loop fluid density information

subsequent to isolation of the sources and (b) increase in uncertainty of

the fluid density data prior to isolation because the in situ calibration

must now be based on one (as opposed to two) known fluid density. While

the impact of the loss of density information should not be minimized, it

does not affect other data and has only a minimal impact on the

understanding of the experiment because the loops were already partially

voided prior to this time. Thus, the initiation of voiding in the loops

was measured; the major loss is knowledge of when the loops were completely

voided. The latter can be bounded using thermocouple and level information

in the upper plenum as discussed in Section 3.

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TABLE I. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2

SpecifiedaValueParameter

Primary Coolant System

Core delta T (K)(OF)

Primary system pressure(hot leg) (MPa)

(psia)14.95 + 0.12168 + 15.0

MeasuredValue

11.7 + 1.421.1 T 2.5

14.98 + 0.12173 F 15

571.6 + 0.8569.2 T 1.4

559.9 + 1.1548.2 + 2

475 + 2.53.77 + 0.02

Hot leg temperature (K)(OF)

571 + 1.1569 T 2

Cold leg temperature (K)(OF)

Loop mass flow (kg/s)

(lbm/h x 106)

Boron concentration (ppm)

Primary coolant pump injection(both pumps) (L/s)

(gpm)

479 + 193.8 ; 0.15

499 + 15

0.1272.0

+ 0.0167 0.25

0.128 + 0.0031.98 T 0.02

Reactor Vessel

Power level (MW)

Decay heat (200 s)

26.5 + 0.5

685 + 10

26.8 + 1.4

(kW) 684.6

Maximum linear heat generationrate (kW/m)

(kW/ft),4AO--12

Control rod position(above full-in position)

42.6 + 3.612.97 T 1.1

1.38 + 0.0154.3 7 2.0

Wm)(in.)

1.37 + 0.0154.0 7 2.0

13

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TABLE 1. (continued)

ParameterSteam Generator

Secondary system pressure (MPa)(psia)

Water levelb (m)(in.)

Pressurizer

Liquid volume (m3(ft])

Steam volume (m3(ftl)

Water temperature (K)(OF)

Pressure (MPa)(psia)

Liquid level (m)

(in.)

Suppression Tank

Liquid level (m)

(in.)

Gas volume (m3

Water temperature (K)(OF)

Pressure (gas space) kPa)t (pma)

Boron concentration (ppm)

SpecifiedaValue

1.12 + 0.144 4

1.19 + 0.051- 0.0

47.0 + 2- 0.0

<311cl100

100 + 2014.7 W 3

MeasuredValue

6.38 + 0.08925 T 12

0.17 + 0.066.7 ; 2.4

0.57 + 0.0320.13 " 1.06

0.37 + 0.0313.07 + 1.06

616.9 + 2.1650.8 T 3.8

15.1 + 0.12190 " 14.5

1.06 + 0.0644.4 + 2.4

1.18

46.5

59.112087

295.672

9513.8

3710

+ 0.06c

+ 2.4

+ 2.02+ 71

+ 0.5Ti1

+3T 0.4

+ 15

14

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TABLE 1. (continued)

SpecifiedaValueParameter

MeasuredValue

Emergency Core Cooling System

Borated waterTemperature

Accumulator A

Accumulator A

storage tank(K)(OF)

liquid level (m)(in.)

pressure (MPa)(psia)

Accumulatortemperature

Accumulator

Accumulator

Accumulatortemperature

A liquid(K)(OF)

303 +85 :

<2.17<86

>4.21>611

303 +85 +

<2.16<86

>4.21>611

303 +85 T

35

35

301.3 + 382 7 5

1.81 + 0.0271.3 ; 0.8

5.1 + 0.06740 ; 9

303.1 + 0.786 ; 1.3

1.81 + 0.0271 T 0.8

4.95 + 0.06718 T 9

305.6 + 0.790.4 T 1.3

B liquid level (m)(in.)

B pressure (MPa)(psia)

6 liquid(K)(OF)

35

a. If no value is listed, none was specified

b. Steam generator liquid level referenced to 2.95 m (116 in.) above thetop of the tube sheet.

c. This value is out of specification.

15

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TABLE 2. OPERATIONAL SETPOINTS FOR EXPERIMENT LP-FP-2

Event Specified Measured

Scram reactor (s) 0.0 0.0

Turn off primary pumps (s) 8 + 2 9.7 + 0.1

Insert CFM control rodsa (s) 20 22.4 + 0.1

ILCL break opened (s)b 23 32.9 + 0.1

LPIS break opened (s) 220 + 5 221.6 + 0.1

ILCL break closed (s)d 721 735.5 + 0.1

Fl and F2 opened (s)c 905 1015.7 + 0.1

Align LPIS line filter (s)c 945 950.8 + 0.1

Isolate gamma densitometer sources (s)e 945 262 + 2

Close FPMS lines (s)f 1766 1777.1 + 0.1

Close the LPIS line (s)f 1766 1777.6 + 0.1

ECCS flow initiated (s)g 1783.6 + 0.5 1782.6 + 0.1

a. Insertion of the CFM control rods was initiated when the primarycoolant flow decreased to 189 kg/s (1.5 x 106 Ibm/h), as specified.

b. The ILCL break was opened upon verification that the CFM control rodswere fully inserted.

c. The F3 filter and the FPMS line isolation valves were opened whencladding temperatures reached 840K (1052°F), as specified.

d. The ILCL break was closed when cladding temperatures reached 566 K(560°F) or PCS pressure reached 1.2 MPa (160 psig).

e. The gamma densitometer sources were to have been isolated from thedetectors when the cladding temperatures reached 840 K (1052°F).

f. The FPMS sampling line and LPIS line isolation valves were closed whenshroud temperatures reached 1517 K (2272°F).

g. ECC flow was initiated 6 s after initiation of closure of the LPIS lineisolation valves.

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2.2 Chronology of Events

Identifiable significant events for Experiment LP-FP-2 are listed in

Table 3, which compares the times of occurrence with the times predicted bythe pre-experimental calculations. Annotated intact loop hot leg pressure

histories are shown in Figures 4 and 5.

The experiment was initiated by scramming the reactor using the

peripheral control rods. The primary coolant pumps were then turned off at

approximately 10 s. After the PCS flow had decreased to 190 kg/s

(1.5 x 105 lbm/h) at 22 s, the center fuel assembly control rods were

inserted. The intact loop cold leg break was opened at 33 s to initiate

the blowdown. This was followed by the opening of the simulated LPIS line

at 222 s. The core started heating up when the liquid level decreased intothe core at 662 s and 689 s in the peripheral and central fuel modules,

respectively. The intact loop cold leg break was closed at 736 s but was

reopened to increase the depressurization rate at 878 s. In addition, the

PORV was opened at 882 s, also to increase the depressurization rate;

however, no effect was measured due to opening the PORV. After a

sufficient depressurization had been achieved, the intact loop cold leg andPORV lines were closed at 1022 and 1162 s, respectively. Fission product

release was first measured in the Fl and F2 lines at approximately 1200 s.The hottest cladding temperatures reached 2100 K (3320°F) by 1504 s. The

transient continued without intervention until the outer shroud walltemperature limitation of 1517 K (2272'F) was reached at 1766 s.

Subsequently, the FPMS lines were isolated at 1777 s and ECC injection was

initiated at 1783 s. The core was quenched at 1795 s (although a few,

isolated thermocouples indicated temperatures in excess of saturation for

several minutes thereafter), and the plant was maintained in a quiescent

state for 14 days while fission product measurements were taken using the

on-line measurements systems. In addition, batch samples were taken from

the BST and PCS for several days thereafter: BST liquid samples (21 d),BST vapor samples (28 d), and PCS liquid samples (44 d). During the early

part of the cooldown, the PORV was cycled twice (see Table 3) to preventthe PCS from overpressurizing, and a feed-and-bleed operation on the steam

generator was initiated.

17

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TABLE 3. CHRONOLOGY OF EVENTS FOR EXPERIMENT LP-FP-2

Time After ExperimentInitiation

(s)

Event

Scram

Control rods fully inserted

PCP coastdown initiated

CFM control rods fully inserted

ILCL break initiated

PCP coastdown completeb

End of subcooled blowdownd

Secondary relief valve cycle

Pressurizer empty

LPIS line break initiated

Secondary pressure exceeds primary systempressure

Earliest coolant thermocouple deviationfrom saturation (voidage at that location)

upper plenumhot leg pipedowncomerlower plenum

Fuel rod cladding heatup initiated in

peripheral fuel assembly

Fuel rod cladding heatup initiated in CFM

ILCL break closed

ILCL break opened

PredictedValue

0.0

0.0

25

--a

20

27c

42

51

45

220

220

52.55464

MeasuredData

0.0

2.4 + 0.1

9.7 + 0.1

23.4 + 0.5

32.9 + 0.1

25.1 + 0.1

53 + 1

56 + I

60 + 5

221.6 + 0.1

260 + 10

300390730800

662665

+

+

+

+

+

÷

+

10101020

2

2

0.1

0.1

713.5

1007

-- a

689

735.5

877.6

18

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TABLE 3. (continued)

Time After ExperimentInitiation

(s)

Event

PORV opened

F3 filter on line

LPIS bypass closed

FPMS lines opened

ILCL closed

PORV closed

First indication of (gap) fissionproducts at Fl

First indication of (gap) fissionproducts at F2

First indication of (gap) fissionproducts at F3

Peripheral fuel cladding reaches1460 K (2172-F)

Maximum upper plenum coolanttemperature reachedf

First indication of (fuel) fissionfission products at Fl, F2, and F3

Cladding temperatures reach2100 K (33200)

Shroud temperature reachestrip setpoint

1st thermocouple2nd thermocouple

Maximum cladding temperature reached

PredictedValue

-- a

-- a

MeasuredData

882.0 + 0.1

950.8 + 0.1

951.9 + 0.1

1013.1 + 0.1

1021.5 + 0.1

1162.0 + 0.1

1200 + 20

1200 + 20

1249 + 60

-- e

1495 + 5

1500 + 10

1504 + 11721.6

1743 +1766 T

__g

11

19

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TABLE 3. (continued)

Time After ExperimentInitiation

(s)

Predicted MeasuredEvent Value Data

LPIS line break closed __h 1777.6 + 0.1

FPMS lines closed ._h 1778.1 + 0.1

Maximum upper plenum metal temperature __h 1780 + 5reachedf

ECCS initiated ._h 1782.6 + 0.1

Accumulator flow stopped --h 1795 + 2

Maximum LPIS line coolant _.h 1800 + 5temperature reached

Core quenched __h 1795 + 5i

Cooldown initiated ._h _.i

Steam generator --h 2600 + 10feed-and-bleed started

PORV opened --h 3350 + 10

PORV closed --h 3380 + 10

PORV opened .-h 3680 + 10

PORV closed --h 3690 + 10

Experiment terminated --h __j

a. This value was not calculated.

b. The pumps were allowed to coastdown under the influence of the motorgenerator flywheel until the pump speed reached 750 rpm. At that time, theflywheel was disconnected from the motor generator and the pumps quicklystopped adding energy to the fluid. The time at which the flywheel wasdisconnected is defined as the time the PCP coastdown was complete.

c. Due to an error in the version of RELAP5/MOD2 that was used for thiscalculation, the initial prediction indicated a sharp pump coastdown.However, a later partial calculation made using a corrected version of thecode indicated completion of the pump coastdown (pump speed below 750 rpm)at 43.5 s.

20

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TABLE 3. (continued)

d. End of subcooled blowdown is defined as the time when the first measuredfluid temperature outside of the pressurizer reaches saturation conditions.

e. None of the cladding thermocouples in the peripheral fuel bundlemeasured validated temperatures above the setpoint. The two which gavereadings above this setpoint were failed prior to reaching the setpoint.

f. These temperatures represent the maximum measured temperatures prior toreflood at these locations. The thermocouple output during reflood couldnot be interpreted.

g. Due to the large number of cladding thermocouples in the central fuelmodule that failed at high temperatures during the transient, it is notpossible to determine the precise maximum temperature or the time at whichit occurred. The time is estimated to be between 1782 and 1795 s. Themaximum temperature exceeded 2400 K (3860°F) based on valid temperaturereadings prior to thermocouple failure.

h. The calculations were terminated prior to this event.

i. The peripheral fuel modules were quenched by 1793 s. Most of thecentral fuel module cladding thermocouples were quenched by 1795s. Someisolated thermocouples indicated persistent high (superheated) temperaturesa few minutes longer. Interpretation of the temperature data is complicatedby the large number of thermocouples in the center fuel module that failedduring or just prior to reflood (see Appendix I).

j. Due to the high background in the area surrounding the G1, G2, and G3spectrometers, data were collected for several weeks subsequent totermination of the thermal transient.

21

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Scram PCP trip

15

CL

a-

12.5

10

7.5

CL

50 50 100 150 200

Time (s)

Figure 4. Primary system pressure (short term).I

00~

0L:3InIna)La-

15

12.5

10

7.5

5

2.5

0

0

(a

0.

a)L:3(a(aa)I.-0~

0 500 1000 1500 2000Time (s)

Figure 5. Primary system pressure (full term).

22

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3. PCS THERMAL/HYDRAULIC RESULTS

This section summarizes the thermal/hydraulic boundary conditions of

the PCS prior to and during fission product release and transport.

Included are the hydraulic response during the blowdown, the fluid and

metal temperatures during fission product release and transport, and the

fuel rod cladding response during the blowdown and heatup phases. Also

included is a brief comparison of the thermal/hydraulic response with

preexperiment calculations made using the RELAP5/MOD2 1 5 and TRAC-BD1 16

computer codes. An additional comparison is made in Appendix J, which

compares the experimental results with calculations made using a unified

code made up of SC)AP, RELAP5/MOD2, and TRAP-MELT. The detailed

discussions of the PCS hydraulics, core thermal response, and comparison of

data with RELAP5/MOD2 and TRAC-BDI calculations are included in

Appendices D, E, and F.

3.1 Blowdown Hydraulics

This section discusses the reactor vessel liquid level, PCS mass

inventory, center fuel module mass flow rate, and PCS reflood.

The experiment hydraulics resulted in a gradual PCS level decrease

and, ultimately, in a slow core boil-off. The loops began to void at

approximately 50 s (intact loop hot leg) as shown in Figure 6, which

compares the individual average chordal densities measured by the gamma

densitometer in this leg. The level decreased until the loops were

completely voided by 470 s (based on dryout of thermocouples in the upper

plenum). The upper plenum was voided by approximately 600 s and the level

continued to drop, entering the top of the core by 700 s. The entire core

was voided by approximately 1355 s as indicated by the level probe in the

#3 fuel module. The data from this probe is shown in Figure 7. As

discussed below, the completion of voiding as indicated by the level probe

occurred more than 300 s after all the cladding thermocouples in the core

indicated heatup.

23

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I

__ JUL-r-U-UUZti-- DE-PC-002C

50E 0.75IN E

" I" 40 .,'II

" 0.50 30t30

C II

-20

0 0.25 ' ,,

010 A,-

0 50 100 150 200 250 300

T i me (s)

Figure 6. Intact loop hot leg density.

24

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*** BLOBLE * VERSICN 301 * 1OD 0C2 0 2/23/31 DATE OF RUN e5/Oq/Ilo PACE 30C3

RUBBLE PLOT OF FILE 'LE3F10 .

* CHARACTER RANGE TABLE *

{ c 1.000 • N < 999.30C ** (0) .010 •N f 1.000 ** MX) -999.00C 4 1 • .C10 *

LEVEL(M ABOVE

CORE BOTTOM)

1.78 *XXXXOXCOOOOOOCCOOCOCOOCO)XXXXXXXXXXXXXOCCOOOOXCCOCGOCO OXXXYXXXYXX1.08 *XXXXOCCOOOOOOOCOCOOCCCCCCOO000000000000C0000000 CCCOCOXXXXXXXXXXXXXXXYXYXXXXXYXXX XXX XXXXxXxxXXXXX.98 *xxxcoooooOO3COOGooccccocoCooocOo320009cooc000 ccooGooxxxxxxxXoOOcOOcOOOcccccocc3c.ý4xxxxxyxxXXXxxxU, .89 *XXXXOOCOOOOOOCOCOOCoOCCoo0O3OOO0OOOOOOXX0o0000 0OOOOOCCOOOOOOCOCCCCCcOCOOCXXXXXXXx•x.71 *XXOOOCC00OOOOOOOCO3G300COOOOOOOOOOOOOOOOOGCOOOOO OOocCXXXXXXXXXXYXXX•XXX YXXXYXXX)XXXXXXXIXXXI.61 *xXXOOCGO000000CCCOOCCCOCCOOO003OO0)OO0OOOOO0 OCCc3ooGCAXXXXCCCCCCC3c CCCCCCCOCCC CXXXxx)y Xxx.51 *XXXOOOCOOOOOOOCOOCOOCOCC OOOOOOOOOOOO3O030xXXOXOCO0 CCOOOcXX XXXXXXXXxxxXXXXX xwxCCIOO0CXxxxx xXX.41 *XXXXCXCjoo0OO3OCC33OC3COCOOOOOOOOOOOCOJOOOOXXxXXXO0 COXXCOC CCCCOCOCrCCC3 Oxx)wxxxxxxxxx.20 *XXXXXXXXCOOaOO0CCO33COCCCOCO0OOOOOXOXOXXXXXXXXXXXXXXXXXXXXXXXOO OxxxXXxXXXxxx.1c *xxxxxxxxoooooocoocoocccccccoooocoo.oooocoocooccocooccxccccccoxxxxxxxxxxxxxoccýCoo OOCCCCC5c^Z03C0xXX).•XXXA

**----------- *------- -- --- ---------------- *------ *---------*--------.053 31ý8.447 S25.'.47 492.447 54,.447 826.447 993-.4'? 1160.447 1327.447 14r4.147 t'.L7I? 1e~' j~P . 47 'ýC.4

TIM~E (5)

Figure 7. Conductivity level probe response above Fuel Assembly 3.

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The PCS mass inventory declined to a minimum of approximately 500 kg

(based on the blowdown suppression tank level increase) by 1300 s. At that

time, the center fuel module mass flow had decreased to approximately

0.04 kg/s (this mass flow rate was calculated from the measured cladding

temperature response. The details of the calculation are in Appendix D)

and the flow out the LPIS line, to approximately 0.3 kg/s. This mass flow,

though very small, was sufficient to sustain a rapid metal-water reaction

in much of the central fuel module as the temperatures increased above

1700 K (2600°F). The center fuel module mass flow results in an average of

0.4 gm/s/fuel rod (0.04 kg/s per 100 fuel rods). Data from the Power Burst

Facility indicate that flows as little as 0.1 g/s/fuel rod are sufficient

to sustain the metal-water reaction without steam starvation. 1 7

When the shroud temperatures reached the experiment termination

setpoint of 1517 K (2272°F), the FPMS and LPIS lines were closed and

reflood of the plant was initiated using both ECC systems. Rapid injection

of approximately 1000 kg (2200 ibm) of water from the accumulators resulted

in a PCS repressurization from 1.2 to approximately 3 MPa (174 to

435 psia). This caused the accumulator flow to momentarily cease.

Additional cycles of-accumulator flow and PCS repressurization were

required before all of the damaged core could be quenched; the ECCS was

fully capable of accomplishing this and the plant was in a safe shutdown

condition within a few hundred seconds of ECCS injection initiation. The

peripheral fuel rods quenched rapidly, in a manner similar to previous LOFT

core uncovery experiments. Most of the center fuel module also quenched

rapidly, though more slowly than in previous experiments. A small fraction

of the center fuel module, however, took much longer to quench, indicating

the disruption of the fuel rod geometry in part of this module. Additional

details on the thermal response of the core during reflood is located in

Section 3.2 and in Appendix E.

26

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3.2 Core Thermal Response

This section summarizes the fuel rod cladding temperature response,

including the initiation of dryout at various core locations, the effect of

control rod melting on the thermal response, the occurrence and propagation

of a rapid metal-water reaction, and the quench of the core during reflood.

The temperature excursion began in the upper part of the peripheral

fuel modules at 662 s and moved downwards as the coolant boiled away. The

propagation of the core heatup was generally top-to-bottom in the

peripheral module, with the dryout reaching elevations of 1.14, 0.38, and

0.28 m (45, 15, and 11 in.) above the core bottom at 662, 730, and 930 s,

respectively. This is illustrated in Figure 8, which compares cladding and

saturation temperatures at these elevations in the #2 fuel module. The

quench at the 10-in, elevation associated with the opening of the PORV is

also seen. Figure 9 is a similar figure for the central fuel module, with

temperatures shown from the 1.07-, 0.69-, and 0.25-m (42-, 27-, and 10-in.)

elevations. The dryout started a little later in this module, with the

corresponding times being 689, 740, and 938 s, respectively.

At approximately 1050 K (1430*F), the guide tube temperatures

responded to a phenomena that is thought to be connected with melting of

the absorber material (Ag-In-Cd) at the 0.69-m (27-in.) elevation. The

temperatures on guide tubes 5J13 and 5K05 both show a definite decrease in

the heatup rate (from 1.2 K/s down to 0.7 K/s) which is interpreted as

resulting from the melting of the control rod material in these guide

tubes. The argument is that the latent heat of melting absorbed some of

the decay heat, causing a decrease in the heatup rate. This is consistent

with the observation that the heatup rate of guide tube 5H08, which does

not contain a control rod, was not similarly affected. Figure 10 compares

these three temperatures. The latent heat associated with the melting of

the control rods could account for a temperature shift of up to 280 K

(504 0 F). The difference between this value and the 50 K (90 0 F) measured

shift could be explained by the metal-water reaction, which was occurring

at that time.

27

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600 I I I I I I II

I / ,

I I i

.4

-. 9.It •

"- 550

CL.

E 500I--

450600

I .1

* /1/

$ . *

/.7/

I ' I

* ~I,.., * /'I,,

-'-I

-600

-550

Soo

500

~450 0

TE-2E08-045- TE-2F07-015- TE-2G14-011X Saturation temrp

I 400

I.-

eraturetI I I ~

650 700 750 800 850Time (s)

900 950 1000

Figure 8. Comparison of cladding temperatures at the 1.14-, 0.38-, and0.28-m (45-, 15-, and 11-in.) elevations in Fuel Assembly 2 withsaturation temperature. (See Appendix I for thermocouplequalification limits).

600

%--1550

E o

4..L3

E 500II--

CDCL

EQ,

I--

450 L600 650 700 750 800 850 900 950 1000

Time (s)

Figure 9. Comparison of cladding temperatures at the 1.07-, 0.69-, and0.25-m (42-, 27-, and 10-in.) elevations in Fuel Assembly 5 withsaturation temperature. (See Appendix I for thermocouplequalification limits).

28

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1600

'- 1300

,L9 1200

800

I100.

E 1000

I-,900

800-

700

110.

Figure 10.

J ll"

-- 00

(2-i. elvtoni ul seby .(e Apni I for

It q

- -- 1000" II

I I I I I I I '

1150 1200 1250 1300 1350 1400 1450 1500 1550 1600Time (s)

Comparison of three guide tube temperatures at the 0.69-rn(27-in.) elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits).

I.

29

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At about 1500 s, several control rod guide tube thermocouples at the

27-inch elevation show a small discontinuity that is thought to be

associated with the failure of the rod (see, for example, Figure 10). This

occurred at approximately 1200 K (1700'F). Once again, the effect is

absent from thermocouple TE-5H08-027, which is in an empty guide tube.

The first recorded and qualified rapid temperature rise associated

with the rapid reaction between zircaloy and water occurred at about 1430 s

and 1400 K on a guide tube at the 0.69-m (27-in.) elevation. This

temperature is shown in Figure 11. A cladding thermocouple at the same

elevation (see Figure 11) reacted earlier, but was judged to have failed

after 1310 s, prior to the rapid temperature increase. Note that, due to

the limited number of measured cladding temperature locations, the precise

location of the initiation of metal water reaction on any given fuel rod or

guide tube is not likely to coincide with the location of a thermocouple.

Thus, the temperature rises are probably associated with precursory heating

as the metal-water reaction propagates away from the initiation point.

Care must be taken in determining the temperature at which the metal water

reaction initiates, since the precursory heating can occur at a much lower

temperature. It can be concluded from examination of the recorded

temperatures that the oxidation of zircaloy by steam becomes rapid at

temperatures in excess of 1400 K (2060°F).

The temperatures in the center fuel module reached the target

temperature of 2100 K (3320°F) due to the rapid reaction between the

zircaloy and the steam, and remained above this temperature for

four-and-a-half minutes. The maximum temperature reached is difficult to

determine because of the failure of the thermocouples at the high

temperatures experienced, but it was certainly in excess of 2400 K (3860°F).

During the transient, the temperatures on the outside of the shroud

increased steadily from 740 to about 1700 s. This is illustrated in

Figure 12, which compares the temperatures on the south side of the

shroud. At approximately 1700 s, the heatup rate increases. At about the

same time, the thermocouples near the outside of the shroud also start to

heat up more rapidly. Figure 13 illustrates this by comparing the

30

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2500 ý 4000

.- 2000

U,)lid

4- 1500IL

E

1000

500600

Figure 11.

1700

1500

2 1300

L- 1100.

0 900C-E

I. 700

--- s -- - I I.. - -_ 1700 800 90o 1000 1100

Time

Comparison of two claddingelevation in Fuel Assemblyqualification limits).

-.3000

-2000 L-

100

I a.', E

I I 0

- - -L1000

1200 1300 1400 1500 1600(s)

temperatures at the 0.69-m (27-in.)5. (See Appendix I for thermocouple

2500~r- i -. -/1

/1// 2000/ /

- U). *" .1500

0- -°-1000

E

.50050101

300600

I I I I I I I I L I / J700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900

Time (s)

Figure 12. Comparison of four external wall temperatures at'the 1.07-,0.81-, 0.69-, and 0.25-m (42-, 32-, 27-, and 10-in.) elevationson the south side of the flow shroud. (See Appendix I forthermocouple qualification limits).

31

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1500 I I I I I I T I I

-2000

1300o . ,

-- " 1100 -"1500 %.

L L.

9001) 1000

E 700 E•D •ph- I--

TE-2G14-011

500 " TE-2H14-028 500TE-2114-039TE-2H13-049

.1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 1900Time (s)

Figure 13. Comparison of cladding temperatures at the 1.24-, 0.99-, 0.71-,and 0.28-m (49-, 39-, 28-, and 11-in.) elevations in FuelAssembly 2. (See Appendix I for thermocouple qualificationlimits).

32

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temperatures at various elevations in the #2 fuel module, just adjacent to

the shroud south wall. By the time the reflood turns the temperatures

around (1785 s), all of these temperatures exceed 1400 K (2060°F). In most

cases, the fuel rod cladding temperatures exceed the shroud temperatures at

the same elevation. The cause of this rapid heatup is not presently known,

but it may be an effect caused by the thermocouple leads passing through a

hot area as they exit from the top of the core (shunting) rather than a

true local effect.

The cooling of the core took much longer than any previously measured

quench in LOFT. This was in part due to the much higher temperatures that

existed prior to quench (>2400 K [38600 F] for this experiment comparedwith the previous maximum of 1261 K [181 0 F] measured during Experiment

LP-LB-l 8). More important, however, is the geometry of the core during

reflood. Relocation of the core undoubtedly resulted in masses of core

material much thicker than normal. These masses would require much more

time to cool than would the regular fuel rod geometry. This is postulated

to be the cause of the slow cooldown manifested by thermocouple TE-5JO7.-027

(failed), shown in Figure 14. (Even though this thermocouple failed, it is

believed that the failure mode is a junction relocation and that the

thermocouple is indicating a temperature at some location in the center

fuel module.) That thermocouple was slowly cooling towards saturation

until 2010 s, when the junction apparently broke. Thus, even though the

core had been essentially quenched for more than 200 s, the temperature was

only slowly decreasing, indicating the insulating effect of a large mass ofmaterial surrounding the thermocouple. Additional evidence that the center

fuel module experienced significant control rod fuel relocation is

discussed in Appendix E.

3.3 Comparison with Calculations

This section summarizes the results of the comparison of the data from

Experiment LP-FP-2 with preexperiment calculations19 made using the

RELAP5/MOD215 and TRAC-BO116 computer codes. Appendix F contains a

more detailed presentation of these comparisons, along with a description

of the input models.

33

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2500 4000

2000

0L

4-

0L00~E0I-

1500

1000

3000

2000 .

0-

05001

0600 0oo 1200 1500 1800 2100

Time (s)

Figure 14. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5 with saturation temperature. (SeeAppendix I for thermocouple qualification limits).

34

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The predicted pressure response agreed well with the data except for

the time after initiation of the LPIS line break at approximately 220 s.

This is illustrated in Figure 15, which compares the measured and predicted

PCS pressure response for the first 400 s of the transient. The relatively

minor differences between measured and predicted pressure response can be

explained on the basis of differences between the initial conditions and

the experiment sequence that were assumed in the calculations and the

actual conditions of the experiment. However, the opening of the LPIS line

break did not have as great an impact on the depressurization rate as was

predicted. This is emphasized in Figure 16, which compares measured and

predicted pressures for the first 2000 s of the transient. The PCS

depressurized much more slowly than predicted from 220 until 735 s, when

the intact loop cold leg break was closed. This indicates that the flow

resistance in the LPIS line was much greater than was modeled in theprediction. The LPIS line was modeled as a straight line pipe segment with

a flow resistance based on the length and pipe bends. It is evident that

this modeling was not adequate. This inadequacy is again emphasized during

the time subsequent to 735 s. Closure of the intact loop cold leg break

was predicted to have little impact on the depressurization rate, whereas

the measured depressurization almost stopped (see Figure 16). In fact, the

intact loop cold leg break and plant PORV were cycled subsequent to this

time in the experiment in order to reduce the PCS pressure to the point

that the FPMS lines could be opened. While these actions were notpredicted to be necessary, the depressurization rate was foreseen as an

area of uncertainty and these actions were designated as a contingency

action should the rate of depressurization be too small. The higher than

predicted pressure may have had an indirect impact on the fluid

temperatures. Although the depressurization rate prior to the onset of

rapid core heatup (metal-water reaction discussed below) was lower than was

predicted, the reverse was true during the period of rapid heatup. This

higher depressurization rate resulted in a larger-than-predicted steam

flow, which prevented the expected steam limitation in the upper parts of

the core.

35

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CL

(f)(n

EL

15 I I i

10 -

5

00 50 100 150 200 250

Time (s)

0(0

0~

0I.-

InIn0L.

a-

300 350 400

Figure 15. Comparison of primary system pressure withcalculations made using RELAP5/MOD2 (short

preexperimentterm).

15

PE-PC-002RELAP5/MOD2(25 MW, 100 pct break flow area)RELAP5/MOD2(33 MW, 70 pct break flow area)

0-

10

5

-2000

-1500 0(/}0L

-1000 L(50(0

fL.

-500

v -

0 500 1000Time (s)

1500 2000

Figure 16. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (full term).

36

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Figure 17 presents the measured cladding temperatures at the 0.25-m

(10-in.) elevation in the center fuel assembly with the prediction for the

nearest modeled location. The observed initial temperature rise rate was

1.3 K/s (2.3 0 F/s), which was predicted exactly by TRAC-BD1, whereas

RELAP5/MOD2 predicted 2.4 K/s (4.3 0 F/s). The average temperature rise rate

until 1700 s was observed to be about 0.5 K/s (0.9 0 F/s). The rise rate was

overpredicted for most of this period in both calculations, TRAC-BDl

providing the better agreement with 0.7 K/s (1.3 0 F/s) compared with

RELAPS/MOD2 with 1.0 K/s (1.8 0 F/s). The overprediction was contrary to the

fact that the modeled decay heat level in this node was lower than that

which existed at the measurement location. The underprediction of mass

flow of steam through the core is believed to have resulted in an

underprediction of the heat transfer coefficient. The observed increase in

temperature rise rate at 1700 s occurred at too low a temperature [about

900 K (1161°F)] to be the result of metal-water reaction locally and was

not predicted. The observed behavior may be the result of thermal

radiation from high temperature material at a higher elevation or to

material relocation. Neither thermal radiation in the axial direction nor

the desired effect of material location on local temperature is modeled.

The initial heat up rate at the 0.69-m (27-in.) elevation prior to the

time when the PORV and intact loop cold leg break were opened was well

calculated by both codes. This is illustrated in Figure 18, which compares

the temperatures at this elevation in the center fuel module. After that

time, however, the measured temperature rise rate decreased and the

predicted rise rate was higher. The temperature rise rate due to metal

water reaction was well predicted by TRAC-BD1, even though it initiated

from a lower-than-predicted temperature. This lower initiation temperature

may be due to precursory heating discussed above and in Appendix E.

The initial heat up rate at the 1.07-m (42-in.) elevation was

accurately predicted by both codes. After approximately 1450 s, however,

the cladding experienced the rapid heat up due to metal water reaction, in

contrast to the prediction, as illustrated in Figure 19, which compares the

temperatures at this elevation in the center fuel module. The metal water

reaction was not predicted to occur at this elevation due to steam

37

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4)

4-

1400

1200

1000

800

600

40060(

Figure 17.

2500

2000

--1500

L

1000 C

E

800 1000 1200 1400 1600 1800 2000Time (s)

Comparison of the measured cladding temperature at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BD1.(Thermocouple qualified throughout).

I I I I 1 " 4000

TE-5J07-027 -

RELAP5/MOD2. TRAC-BD1 .. . .0

- 3000

- .- - 2000 .-

- L..

1000 E

a,I..

4-0I.Sa-ES

I-

1500

1000

500

0 L600 800 1000 1200 1400 1600 1800 2000

Ti me (s)

Figure 18. Comparison of the measured cladding(27-in.) elevation in Fuel Assemblycalculations made using RELAP5/MOD2(Thermocouple qualified to 172.0 s).

temperature at the 0.69-m5 with preexperimentand TRAC-BDI.

38

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3000

2500-

-S

0LJ

-4-01.~00~E0

I-

2000

1500

1000

500

1800 2000

0.

21000

-0

1800 2000

temperature at the 1.07-m5 with preexperimentand TRAC-BD1.

00600 800 1000 1200 1400 1600

Time (s)

Figure 19. Comparison of the measured cladding(42-in.) elevation in Fuel Assemblycalculations made using RELAP5/MOD2(Thermocouple qualified to 1510.3).

39

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limitation. As discussed above, the estimated center fuel module steam

flow was much higher than was predicted; thus, the metal water reaction was

sustained. This had the effect of shifting the location of maximum

cladding temperatures from the predicted 0.56-to 0.84-m (22-to 23-in.)

elevation to the estimated 0.81-to 1.07-m (32-to 42-in.) elevation. Lack

of instrumentation prevents a more precise determination of the actual

level.

In general, the calculations closely predicted both the general and

specific experimental results. Despite the differences noted above and in

Appendix F, these calculations formed a very adequate basis for experiment

planning, including the need for contingencies in the experiment operating

procedure.

3.4 Metal/Fluid Conditions Near the FPMS

The fluid conditions near the FPMS sample locations were generally

superheated. This is illustrated by Figure 20, which compares the fluid

and metal temperatures at the lower Dl deposition coupon. The other

deposition coupon environments were similar. The fluid conditions near the

locations from which fission products and aerosols were drawn into the

Fl and F2 sample lines were similarly superheated except for two temporary

quenches in the Fl line. This is illustrated in Figure 21, which compares

the fluid temperature with saturation temperature in this line. There was

no measured perturbation on the pressure at this location, indicating that

very little steam generation occurred. It is possible that a small water

droplet quenched only the thermocouple.

3.5 Summary

The core boiled dry and heated up to temperatures in excess of 2400 K

(3860°F) due initially to decay heat and ultimately to a rapid metal-water

reaction. The center fuel module control rods melted, as did a substantial

fraction of the adjacent fuel rods. Much of the center fuel module between

the 0.69-m and 1.07-m (27-and 42-in.) elevations relocated to the bottom of

the fuel module. The thermal shroud was able to adequately shield the

40

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1000

"- 750

L

E soo0p

I-,

--

L-

0

E

2500 500 1000 1500 2000

T i me (s)

Figure 20. Comparison of metal and fluid temperatures at the lower D1deposition coupon location with saturation temperature. (SeeAppendix I for thermocouple qualification limits).

600

550

500

a 450

E 400I.-

450

350

0

E0)

0 500 1000 1500 2000T i me (s)

Figure 21. Comparison of fluid temperature in the F1 aerosol sample linewith saturation temperature. (See Appendix I for thermocouplequalification limits).

41

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peripheral fuel rods during this time, and no fuel rod failure occurred in

the peripheral modules. The prediction calculations corresponded well with

the data. The principal discrepancies between data and prediction were

dominated by inadequate modeling of the simulated LPIS line. In general,

the thermal/hydraulic boundary conditions for fission product release and

transport adequately simulated a V-sequence accident.

42

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4.0 FISSION PRODUCT RESULTS

This section describes the source term calculations, the behavior of

the Fission Product Measurement System (FPMS), and the preliminary results

of the measurements concerning fission product transport.

4.1 ORIGEN2 Results for the LP-FP-2 Experiment

The pretransient isotopic mass and activity inventories in the center

bunale were calculated using the ORIGEN220 computer code and using the

measured irradiation history. The ORIGEN2 computer code is designed to

calculate the nuclide composition in a nuclear reactor as a function of

time. This code accounts for several forms of nuclide decay, neutron

activation events, or other changes induced by time-dependent fuel cycle

operations.

The center fuel bundle for Experiment LP-FP-2 consists of an

11 x 11 fuel rod geometry surrounded by a 25.4-mm (1.0-in.) thick thermal

shroud. The 11 x 11 rod geometry contains 100 fuel rods [1.67 m (5.5 ft)

in length] and 21 zircaloy guide tubes, of which 11 contain

stainless-steel-clad control rods. The fuel rods contain 1136.7 grams of

UO2 (1001.4 grams of uranium per rod or 0.10014 MTU for the center

bundle) enriched to 9.744% U-235 (97.57 grams of U-235 per rod). The

control rods contain 1270 grams of a 80% Ag-15% In-5% Cd alloy.

For calculational purposes, the preconditioning phase of the LP-FP-2

experiment was modeled as follows: (a) the reactor was assumed to be

operating for 3.5 days at a constant power of 32 MW (52.5 kW/m or

16 kW/ft), and then was shut down for 3.12 days; (b) 3.333 days of

additional operation at 26.5 MW, followed by a 4.0 day interval of

down-time; and (c) a final irradiation period consisting of running the

reactor at 32 MW for 24 hours followed by an irradiation at 26.5 MW for

16 hours immediately preceding the initiation of the experiment. This

irradiation history approximates the actual irradiation history shown in

Figure 1.

43

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Since the power generated in the LP-FP-2 center bundle accounts for

approximately 17.2% of the total core power,21 it follows that a core

power of 32 MW corresponds to a bundle power of 5.5 MW (Note: the input

values listed in this section are approximate. For exact values, see

Appendix J), and that a core power of 26.5 MW corresponds to a bundle power

of 4.6 MW. Consequently, for the above irradiation history, the center

bundle burnup is calculated to be 429.4 MWD/MTU. Another calculation of

the center bundle burnup for the LP-FP-2 experiment, based on the detailed

measured irradiation history, indicates a burnup of 430 MWD/MTU, which is

in good agreement with the previous burnup result.

The input powers (e.g. 5.7 and 4.7 MW) used in the ORIGEN2 analysis

(see Appendix J) were obtained by multiplying the calculated center bundle

powers (e.g. 5.5 and 4.6 MW) by the factor 1.041. Since the ORIGEN2 code

assumes a total fission energy of 202 MeV/fission to compute the fission

rate (based on U-235), the factor 1.041 (202/194) was used to adjust the

ORIGEN2 power so that the fission rate would be based on 194 MeV/fission,

which represents the recoverable or thermal fission energy (total released

energy minus neutrino energy), instead of 202 MeV/fission.

Selected results of the ORIGEN2 analysis for several important

nuclides are shown in Table 4. Detailed results of the ORIGEN2

calculation, showing fuel bundle activities, masses, thermal power, and

estimated coolant inventories, are identified in Appendix I. The input to

the ORIGEN2 code is also presented with the ORIGEN2 output listing.

Based on the ORIGEN2 data shown in Table 4, the cesium-to-iodine mass

ratio for the center bundle, just prior to the LP-FP-2 experiment, is

calculated to be 4.00 (the atom ratio is 3.88). The ORIGEN2 code was also

used to calculate the decay heat of the center bundle at 200 s into the

transient. The result of the calculation indicates that the center bundle

decay heat is about 115.3 kW at 200 s. Based on a peripheral bundle

44

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TABLE 4. SELECTED ORIGEN2 INVENTORY RESULTS FOR THE EXPERIMENT LP-FP-2

CENTER FUEL BUNDLE AT 430 MWD/MTU BURNUP

Initial Fuel Inventorya

Material

Kr-85Total Kr

Rb-88Total Rb

Xe-131MXe-133Xe-l 33MTotal Xe

1-1311-1321-1331-134Total I

Cs-136Cs-137Cs-138Total Cs

Te-129MTe-132Total Te

Ru-103Total Ru

Ba-140Total Ba

La-140Total La

(grams)

4.408 x 10-26.654 x 10-1

1.216 x 10-36.051 x 10-1

(curies)

1.730 x 1014.039 x 105

1.461 x 1054.663 x 105

SpecificActivity

(curies/gram)b

2.600 x 101

2.414 x 105

2.8305.8977.4966.493

3.8358.6571.8951.1288.588

1.3671.6616.2373.440

2.1792.9928.564

xxxx

xxxxx

xxxx

xxx

10- 3

10-110- 3

100

10-110-310-110-210-1

10- 3

10010-3100

10-210-110-1

2.3721.1043.3626.231

4.7578.9392.1473.0119.472

1.0021.4462.6405.091

6.5699.0897.363

xxxx

xxxxx

xxxx

xxx

102105103105

104104105105105

102102105105

102104105

3.6531.7005.178

5.5391.0412.5003.506

2.9134.2037.674

xxx

x

x

101104102

104105105105

101101104

xxx

5.610 x 10-12.955 x 100

1.100 x 1002.977 x 100

1.243 x 10-11.947 x 100

.811 x 1047.095 x 104

8.025 x l047.227 X 105

6.922 x 1047.355 x 105

7.670 x 1021.061 x 105

6.128 x 103

2.696 x 104

3.555 x 104

a. Center bundle inventory at 200 s into the FP-2 experiment.

b. Curies of the nuclide per gram of the element (all nuclides).

45

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calculation, the total core decay heat is approximately 5.885 a times the

decay heat of the center bundle. Hence, the total core decay heat at 200 s

is estimated to be about 678.5 kW at 200 s into the transient. Detailed

reactor physics calculations based on the actual irradiation history

indicate a core decay heat of 684.8 kW at 200 s. Both results fall within

the 675 to 695 kW range set as a pretest objective for the experiment.

4.2 Results of the Elemental Release Calculations

The elemental source terms to the primary coolant system during the

LP-FP-2 transient were calculated with the CORSOR 2 2 and TIGERl 2 3

computer programs. TIGERI is an INEL-developed program that uses the

release rate data base published in NUREG-077224 (shown in Figure 22) to

predict fission product release as a function of time, similar to the

CORSOR program. Besides the NUREG-0772 data base, the CORSOR and TIGERI

programs require time-dependent axial fuel temperatures, the axial

distribution of the fission products, and the initial fission product

inventory as computed from ORIGEN2. The time-dependent axial fuel

temperatures, shown in Figures 23, 24, and 25, indicate the measured

thermocouple temperatures during the experiment and the averaged

temperatures (the "LEVEL" data) used in the CORSOR/TIGERI calculations.

The axial fission product distribution in the center bundle was

assumed to follow the axial power distribution. A schematic of the

CORSOR/TIGERI model for the LOFT center bundle, depicting the axial fission

product distribution, axial nodalization, and thermocouple locations, is

shown in Figure 26. The time-dependent output of the TIGERI program are

shown in Figure 27 and 28.

a. The factor 5.885 that was used in the decay heat calculation is notequal to the ratio of the core to center bundle powers (e.g. 5.8139 =1/0.172). This occurs because the peripheral bundles have a differentU-235 enrichment (4% versus 9.7%), a different amount of burnup (because ofprevious experiments conducted with the peripheral bundles), andconsequently a different Pu-239 concentration.

46 1,1 - ,

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Temperature (*F)

3000200010

ior

4000

0

0

L.

0cc

(M10"

larg

ioroI. i.1000 1500 2000 2500

Temperature (K)3000

L87-KMI12-00

Figure 22.

3000

Release rate functions taken from NUREG-0772.

4)

:3.6-

4)aE4)

2500

2000

1500

1000

500

0

LL..

CL

ECD

0 500 1000 1500 2000Time (s)

Figure 23. Comparison of measured cladding temperatures at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).

47

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3000

2500

"- 2000

L.

1500

E ioooa..

E

500

00 500 1000 1500 2000

Time (s)

Figure 24.

3000

2500

a,L.

4-01...a,0~Ea,

I-

2000

1500

1000

Comparison of measured cladding temperatures at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).

I I I

-- TC-5108-27--- TC-SM08-27

--- TC-5M04-27 -- 4000

X LEVEL 27

,3000

-20000

E-1000 I

-0

500 1000 1500 2000Time (s)

Comparison of measured cladding temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGER1 calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).

500

0

Figure 25.

1, "- , e

48

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Axialnode Thermocouple

Elevation FP density number location

167.0 cm

42 in. at0.305 112, C07

88.0 cm

27 in. at0.359 2 108, MOB, M04

47.0 cm

0.336 110 in. atM07, L07

0.00 cmL11O]1KM13:5-11

Figure 26. Input model of the center fuel assembly for the CORSOR/TIGERIcalculation.

Kr, Xe, I, Cs

00".• 10 "

W

,,Sb*~10"z/

-I / Ba,> Te

F Sr

E/

o Zri ii1 U 4 I

u

0 500 1000 1500 2000imne (s) Ll10 KM135-12

Figure 27. Elemental cumulative fractional release inventories calculatedusing TIGERI.

49

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10

E

0

4-C.)

W0U

10"

1 0-2

10*

1•0 500 1000

Time (s)1500 2000

L110-KM135-13

I, Sb, Xe, Kr, and Te calculatedFigure 28. Release rate functions for Cs,using TIGERI.

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The CORSOR calculation25 was performed with the same input used in

the TIGER] analysis. The cumulative release results (at 1800 s) of TIGERl

and CURSOR are listed in Table 5. Since the TIGERI results only include

the transient fission product release and not the initial fuel rod gap

inventory, the CORSOR results (shown in Table 5) were modified (by

subtracting the steady state WASH-1400 2 assumed gap inventories) to

produce a net transient release fraction.

Using the TIGERI release rate results, an ORIGEN2 calculation was

performed to estimate the fission product activity released from the center

bundle to the coolant. Results of this calculation are presented in

Appendix J.

4.3 FPMS Performance

The Fission Product Measurement System (FPMS) consisted of three types

of measurement devices: the steam and aerosol sampling systems, the

deposition devices, and the gamma spectrometer systems. The aerosol

sampling system consisted of the Fl sampling line, which extracted a steam

sample from the center fuel module; the F2 sampling line, which extracted

an aerosol steam sample from the broken loop hot leg; and the F3 filter,

which extracted a sample from the LPIS line. The deposition rod (Dl) and

the deposition spool pieces (02 and D3) presented a representative

stainless steel surface to coolant flow in the center fuel assembly upper

structure and the LPIS line, respectively, for fission product plateout.

The gamma spectrometer systems can be divided into two groups, those that

provided isotopic measurements during the transient and those that provided

measurements during the posttransient period. The gamma spectrometer on

the Fl sample line (G6) and on the LPIS line (G5) were intended to provide

data during the transient, while the Gl, G2, and G3 gamma spectrometers

were intended to provide isotopic concentrations in the primary coolant

system (PCS), in the BST liquid space, and in the BST vapor space,

respectively, during the posttransient period. In addition, the G2 gamma

spectrometer, which measured the posttransient isotopic concentrations in

the BST vapor space, was used to measure the concentrations from the

51

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TABLE 5. COMPARISON OF TIGERI AND CORSOR CUMULATIVE RELEASEFRACTIONS FOR EXPERIMENT LP-FP-2 (at 1800 s)

Element

Xe

TIGERI Results

Net Release

0.206

0.206

0.206

0.206

Kr

Cs

CORSOR Results

Code Result

0.275

0.275

0.290

0.265

Gapa

0.030

0.030

0.050

0.017

0.0001

Net Release

0.245

0.245

0.240

0.248

0.0079

I

Te

Sb

Ba

Ru

Sr

La

0.057b

0.020

0.006

0.00014

0.0021

0.0000001

0.020

0.007

0.0002

0.002

<0.0001

a. WASH-1400 fuel rodVII-13, WASH-1400.)

b. The CORSOR resultsthat of NUREG-0772 butoxidation holdup model.tellurium-zircaloy modE

gap inventory fractions.

for Te are based on a resomewhat similar to the

The TIGERI result corr!I is 0.006.

(See Table VII 1-2 page

lease model different than)RNL tellurium-zircaloyesponding to the

52

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combined Fl and F2 sample line flows during the transient. The design and

planned operation of these systems is described in more detail in

Appendix B.

The sequence of events controlling the FPMS is shown in Table 6. All

operator actions were performed as specified.

In addition to the data obtained from the FPMS, a number of

measurements were taken by the plant operators and by Health Physics

technicians to characterize plant conditions during the posttransient

period of the experiment. Of particular interest are remote area monitors

(RAMs) placed at strategic locations inside the containment building and

the stack monitors, which extract an air sample from the exhaust of the

containment building and pass that sample through filters equipped with

radiation monitors. While these measurements do not characterize the

release from the primary coolant system, they are useful in characterizing

BST and PCS leakage during the posttransient phase of the experiment.

4.4 Instrument Operation

The FPMS instruments were operated as specified and, for the most

part, collected valid data. However, a few problems were encountered.

These problems are categorized into three distinct time frames: failure of

instruments during the preconditioning phase, difficulties during the

transient and reflood phase, and difficulties during the posttransient

phase.

During the preconditioning phase of Experiment LP-FP-2, gamma

spectrometer G6, which monitored the Fl sample line, was determined to have

irreparably failed. In its place, a remote area monitor was placed on top

of the reactor vessel to view the Fl line. While this instrument cannot be

used to quantitatively measure the activity in the Fl sample line, it can

be used as a gross gamma instrument to determine the time-dependence of the

Fl activity. The Programmatic Risk Assessment Document (PRAD) 2 6

identifies the impact of the failure of the G6 spectrometer as diminishing

53

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TABLE 6. EXPERIMENT LP-FP-2 FISSION PRODUCT MEASUREMENT SYSTEMSEQUENCE OF EVENTS

EVENT

Fl DILUTION GAS LINE OPENEDF1 AND F2 VENT LINE CLOSEDREACTOR SCRAMDl MOVED (TO DROP CONTROL RODS)Dl WITHDRAWN (TO CLEAR COUPONS)LPIS LINE OPENEDD1 INITIAL PURGE STARTD1 INITIAL PURGE STOPFl STEAM ANALYZER EXTERNAL PURGE STARTFl STEAM ANALYZER EXTERNAL PURGE STOPFl ANNULUS GAS LINE OPENEDF3 LINE OPENEDF3 BYPASS LINE CLOSEDFl AND F2 SAMPLE LINES OPENEDSTEAM DETECTED IN Fl LINEFl LINE FISSION PRODUCTS DETECTEDBLHL FISSION PRODUCTS DETECTEDLPIS LINE CLOSEDFl LINE CLOSEDF2 LINE CLOSEDDl "CLOSED"D0 NITROGEN BACKUP ONFl AND F2 VENT LINE OPENEDFl DILUTION GAS LINE CLOSEDD0 "OPENED"D1 NITROGEN BACKUP BYPASS OPENEDD1 NITROGEN BACKUP BYPASS CLOSEDDl NITROGEN BACKUP BYPASS OPENEDD1 NITROGEN BACKUP BYPASS CLOSEDFl ANNULUS GAS LINE CLOSED

TIME (s)

-199.4-146.8

0.020.648.9

221.6750.6763.1878.1883.0883.1950.8951.9

1013.11013.1

-,1198.0-.1201.0

1777.61778.01778.11780.61808.01823.01833.12085.62143.12148.02933.12968.23401.6

•,• ij•

54

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the number and types of radionuclides measured. Noble gas data might still

be provided by G-5 and/or G2, but potential trapping of gaseous species in

the top of the vessel and in dead spots in the piping can increase the

uncertainties of the results.

Also during the preconditioning phase of the experiment, the

calibration pulsers on the G5 spectrometer failed. While this latter

failure has delayed the processing of the G5 data, it does not impact the

quality of the data.

Just prior to reflood, the deposition coupon device was to have been

closed (to protect a coupon at each elevation from reflood water) and

placed on a nitrogen purge (to maintain a positive pressure differential).

After Experiment LP-FP-2, those actions were performed, but a positive

pressure differential could not be maintained on the deposition rod. This

indicates that some of the protected coupons may have been exposed to

either reflood or PCS water, and that some of the distinction between

protected and unprotected coupons may have been lost. However, significantdeposition data has still been collected: the PRAD identifies data from D2

and D3 as an alternative means of estimating the plateout in the upper

structure.

During the transient portion of the experiment, the Fl sample line was

heated with argon gas. During the reflood phase of the experiment,

however, the sheath through which this gas was fed to the sample line was

sealed, possibly due to high temperatures causing warpage of the line.

Since the measurement of H2 in the BST was calibrated to account for the

presence of this argon, the plugging of this line has caused the BST H2

measurement to be out of calibration and may cause some additional

uncertainty in the H2 measurement.

The third area of difficulty with the FPMS involved high background

activity for the Gl, G2, and G3 gamma spectrometers during the

posttransient phase of the experiment. These instruments were intended to

measure the posttransient isotopic concentrations in the reactor vessel

lower plenum, in the BST vapor space, and in the BST liquid space,

55

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respectively. Because of background problems during Experiment LP-FP-l, it

was decided that a sealed 'tent' would be placed around the Gl, G2, and G3

gamma spectrometers, with a slight purge to maintain a positive pressure

differential. Unfortunately, it appears that there were one or more leaks

either from the G2 sampling system or from the BST directly into the tent,

and that the tent was, in reality, causing a higher Gi, G2, and

G3 background activity than would have been present without it. Therefore,

the posttransient period was extended in order to allow decay and cleanup

of fission products inside the containment building and acquisition of

reliable data. The online FPMS measurements were recorded for 14 d. A

liquid sample was taken from the BST at 21 days after the experiment.

Vapor samples were taken from the BST up to 28 days after the experiment.

Liquid samples from the PCS were taken up to 44 d after the transient.

Because of this extension, the Gl and G3 systems were able to quantify

fission product concentrations in the PCS and BST liquid. There is,

however, currently some concern that posttransient data from the

G2 spectrometer for quantifying BST vapor concentrations will be highly

uncertain. This uncertainty will primarily impact noble gas release data;

I and Cs measurements will be little affected.

The final problem concerns the F2 sample line. Measurements planned

for the F2 line are intended to include both reversible and irreversible

plateout. Therefore, the line will be kept dry until postirradiation

examination in the hot cells. However, during the posttransient phase,

FPMS personnel became aware that the isolation valve on this sample line

was leaking PCS water into the sample line. To preserve the sample line

data, a slight N2 purge was applied; all expectations are that data from

this sample line will not be affected.

4.5 Preliminary Results

This section presents preliminary results (from both FPMS and non-FPMS

instrumentation) that will give insight to fission product behavior and

transport. Due to the short time available for data analysis and due to

the incompleteness of the data, no attempt is made to present a

comprehensive picture of fission product behavior during

56

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Experiment LP-FP-2. Instead, the limited data available are presented both

as an indication of instrument performance and as an example of the type of

data forthcoming from the experiment.

It should be stressed that the information presented herein Is

preliminary In nature and may change as additional data are collected and

inconsistencies are resolved.

4.5.1 Fl and F2 Sample Lines

No Fl or F2 sample line components were examined in time for inclusion

in this document. However, successful operation of the sample line can be

surmised from data now available. Figure 29 shows the response of the RAM

that was placed on top of the reactor vessel to view the Fl sample line.

As seen in this figure, the RAM begins to respond at approximately

1200 seconds to what is believed to be the fission product release from the

gap. By approximately 1500 seconds, the release from the fuel can be seen.

Figures 30 and 31 show the pressure measured upstream of the critical

orifices in the Fl and F2 sample lines, respectively. As seen in these

figures, steam did flow through the Fl and F2 sample lines during the time

of fission product transport. An additional indication of successful

operation of the Fl and F2 sample lines is given by the G2 spectrometer.

During the transient phase of the experiment, this spectrometer measured

the combined effluent from both the Fl and F2 sample lines and identified

isotopes of xenon and krypton.

4.5.2 Deposition Measurements

All information available at the time of this document indicate that

significant quantities of fission products were collected on each of the

deposition pieces. However, as discussed In Section 4.4, some of the

57

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id 3000

'id

L..-c id

10

10"10

- 2500

- 2000 2C)

- 1500 TCD0.

E- 10o

- 500

- 02500

L105-KM130-04

500 1000 1500 2000

Time (s)

Figure 29. Comparison of the radiation area monitor response on the F1aerosol sample line with fuel centerline temperature(TC-5108-027). (See Appendix I for thermocouple qualificationlimits).

1.50

1.25

"" 1

_ 0.75a,o

a- 0.50

0.25

0750

5-F1-8B - 200

- 150

0.

- 100($)

- 50

2000 2250

L105-KMI30-09A

1000 1250 1500 1750

Time (s)

Figure 30. Measured pressure upstream of the critical orifice in the F1aerosol sample line.

58

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1.50 1 I I 1

PTP165-F2-43 - 200

1.25

1- 150

.= 0 .7 5 -6P=~- 100

CL 0.50 a.

50

0=0.75 -10

750 1000 1250 1500 1750 2000 2250

Timne (s) L106i-KM130-08A

Figure 31. Measured pre!ssure upstream of the critical orifice in the F2aerosol sample line.

59

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distinction between the protected and unprotected coupons of the Dl device

may have been lost, in that all of the protected coupons may not have been

sealed from reflood water.

Significant data is expected from the unprotected Dl coupons with

respect to irreversible plateout of fission products at each of the three

elevations. Figure 32 shows the metal temperatures that were measured at

each of the three elevations of the D0 deposition coupons. The deposition

data can be correlated to the measured temperatures and interpolated over

the length of the upper structure to characterize volatile fission product

plateout in the upper plenum.

Radiation scans of the LPIS line at the time of the first containment

entry indicate that fission products were collected on deposition

devices 02 and D3. Figure 33 shows the temperatures of each of the LPIS

line deposition spool pieces. As seen in this figure, the D2 coupon was

exposed to two-phase coolant when the LPIS line was opened, but during the

time of fission product release, was exposed to dry steam. The D3 coupon

was protected from coolant flow while the LPIS line bypass was open, but

after bypass closure the coupon was exposed to steam and fission products.

4.5.3 G5 Gamma Spectrometer

The G5 gamma spectrometer operated as expected, with the exception of

the loss of the G5 calibration pulsers. Isotopes of iodine, rubidium,

xenon, tellurium, and cesium have been identified. Tables 7 and 8 list the

indicated concentrations at various times during the test. Other isotopes

may be identified as data processing continues. While the background

activity for this spectrometer is negligible, plateout has not been

accounted for, although plateout is probably a significant fraction of the

signal by the end of the transient.

Figures 34 through 42 show the concentrations measured by the

G5 spectrometer for each of the isotopes identified in Tables 7 and 8. The

60

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TABLE 7. IODINE SPECIES IDENTIFIED BY G5 SPECTROMETER

All measurements are in microCurles/cm3

Time(s)

179239300361422483544605666727788849910

1024105411151189124913101371143214931554160816681729179018512372267229733274357538764478

1-131 1-132

0.60 + 7%

1-133 1-134 1-135

0.36 17%

0.25 + 16%

0.400.500.50

4.86.8

1526483924

104365

135719452540262028002900302529153100286029002650

+ 12%+ 11%+ 17%

+

+

+

+

+

+

+

+

+

+

+

+

+4.

+

+

+

+

+

~1*

+

10%10%11%11%10%10%12%5%3%6%7%8%7%9%3%3%3%4%3%4%10%

0.41)1.32.22.42.12.72.53.13

14.1522.15

21362

111313151271012513871)

29004400655066506001)664065006481)610059005701)5700

10%5%10%10%10%10%10%15%15%10%10%10%10%10%10%10%10%10%10%10%5%6%8%10%5%3%3%4%4%3%

0.601.01.6

+ 11%. 9r+ 5X

2.2 + 20%

2.2 + 10%

16 + 10%

35 + 10%66 + 10%79 + 10%

114 + 10%

88 + 13%355 + 12%

2.214222951

÷

÷

+

÷1

0.4 + 25%

1.6 + 15%

20%10%15%10%10%

45007600

113001120010700

12800128501330013500

4.

+

+

+4.

+

+

+

4.

1)13%12%5%6%

69 + 10%65 + 10%

10600 + 10%14850 + 10%

14200 + 10%

2.22.22.84.3

22

4182

106121106

50020906760

12200105001750017400

+

+

4.

+

+

+

+

+

4.

+

4.

+

+

4.

+

4.4.

18%20%16%12%12%

10%18%18%10%30%

10%106%10%6%6%8%

5%5%5%6%

16500 + 4%

16400 + 4%

61

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TABLE 8. NON-IODINE SPECIES IDENTIFIED BY 65 SPECTROMETER

All Measurements are in microCuries/cm3

Time(s)

179239300361422483544605666727788849910

10241054111511891249131013711432149315b4160816681729179018512372267229733274357538764478

Rb-88

0.5 + 34%

1.6 + 21%1.6 + 12%1.6 _ 30%

Xe-135 Te-132

1 .400.860.90

0.670.300.20

++

7%6%7%

+ 7%7 19%+ 18%

Cs-138

0.18 + 15%

0.50 + 17%0.50 T 12%

0.40 + 14%0.49 + 11%

0.60 + 10%

0.60 + 12%

0.23 + 16%1.81.81.92.4

+

++

10%11%13%10%

0.150.300.59

+ 20%+ 13%+ 12%

5.4 + 21%4.2 ; 26%

2943474751

+

7+

7

10%8%8%6%9%

21 + 40% 7.4 + 23%16 + 20%81 T 15%

3700 + 8%3800 + 10%

2000 + 10%

6.1 + 34%

8 + 30%78 + 10%

473 + 5%2395 + 10%4350 + 10%5800 + 8%5760 + 5%5500 + 5%4600 + 5%4200 + 5%3700 F 3%3600 + 2%3200 + 2%2800 7 2%2370 7 5%

1900 + 2%3360 + 3%4900 7 3%2400

25903260

+ 7%T 7%: 7%

62

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CL

Ea)

750

700

650

600

550

500

450

- Boo

- 700 0

W

-600 CD0.E

500

400

2000 24000 400 800 1200 1600

Time (s) LIO5-KM130-18M

Figure 32. Measured metal temperatures at. the D1 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits).

4-

CLE9

800

700

600

500

400

300

200

800

L-600 2

400

E

200

0 400 800 1200 1600

Time (s)2000 2400

L105-KM130-17M

Figure 33. Measured metal temperatures at the D2 and D3 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits).

63

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3500

3000

2500

2000

1500

1000

0

E)

-5(

0 -I I

500 1000 1500 2000 2500 3000 3500 4000 4500

Time (s) L1O-KM135-02

34. Measured 1311 concentration in the simulated LPIS line (notcorrected for plateout).

Figure

E

8000

7000

6000

5000

4000

3000

2000

1000

0

-10000 500 1000 1500 2000 2500 3000

Time (s)3500 4000 4500

LI1O-KM135-03

Figure. 35. Measured. 132Icorrected for

concentration in the simulated LPIS line (notplateout).

64

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a

14000

12000

10000

8000E

6000

4000

2000

0

-2000

Figure 36

F

Li0 500 1000 1500 2000 2500 3000 3500 4000 4500

Time (sW L11O-KM135-04

5. Measured 133I concentration in the simulated LPIS line (notcorrected for plateout).

18000

16000

14000

12000

. 10000E

8000

6000

4000

20000 e E) O00 6

-2000 I __ _

500 1000 1500 2000 2500Time (s) L110-KM135-05

Figure 37. Measured 13;4I concentration in the simulated LPIS line (notcorrected for plateout).

65

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0ES

20000

18000

16000

14000

12000

10000

8000

6000

4000

2000

0

-2000500 1000 1500 2000 2500 3000 3500 4000 4500

Time (s) L110-KM135-08

8. Measured 135I concentration in the simulated LPIS line (notcorrected for plateout).

Figure 34

4500 -

4000 -

3500 -

3000 -

2500 -E

2000 -

4- 1500 -

V)

1000

500

0

-5000 500 1000 1500 2000 2500 3000 3500 4000 4500

Time (S) L110-KM135-07

Figure 39. Measured 88Rbcorrected for

concentration in the simulated LPIS line (notplateout).

66

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10 T --- r - ---- -- T --.- I

E0L

8

6

4

2

0 --L

0 250 500 750 1000 1250

Time (s'1500

L11O-KM135-OSA

Figure 40. Measured 13 5 Xecorrected for p

concentration inlateout).

the simulated LPIS line (not

60

50

40E

30

201200 1250 1300 1350 1400 1450

Time (s)

1500

L110-KM135-09

Figure 41. Measured 132 Te concentration irvcorrected for plateout).

the simulated LPIS line (not

67

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,." p

E0

65006000

550050004500

40003500

30002500

20001500

1000

5000

-5000 500 1000 1500 2000 2500 3000 3500 4000 4500

Time (s) L11O-KM135-iO

Figure 42. Measured 13 8 Cs concentrationcorrected for plateout).

in the simulated LPIS line (not

68

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LPIS line was isolated at: 1777.6 seconds and after that time these figures

show decay processes only. The xenon data shown in Figure 40 has been

truncated because of significant background beyond 1477 seconds.

4.5.4 Gl, G2, and G3 Gamma Spectrometers

The 62 gamma spectrometer was used to m2asure isotopic concentrations

in the combined effluent from the Fl and F2 sample lines. As yet, only

isotopes of krypton and xenon have been identified. Other isotopes may be

identified as data interpretation continues. Results from the GI and

G3 spectrometers are not yet available.

4.5.5 Grab Samples

Grab samples of the primary coolant system (PCS) and of the blowdown

suppression tank (BST) liquid and vapor space were taken both preexperiment

and postexperiment. Since postexperiment saniples were not expected prior

to the test, they represent a potential for enhanced understanding of the

experiment. The postexperiment isotopic concentrations of samples taken

21 days after the experiment are shown in TaIles 9 and 10. Mass

spectroscopy results are shown in Table 11. Preliminary analyses of tne

H2 measurements, shown in Table 11, indicate that the total mass of H2

in the BST was 237 g. This compares reasona)ly well with a postexperiment

estimate of the total H2 release of 320 g.

4.6 Preliminary Analysis of the BST and G5 Data

The purpose of this section is to present a general and preliminary

analysis concerning the 8ST grab samples and 65 spectrometer data.

The first part of the analysis concerns the BST data and is dedicated

to calculating the cumulative release rates to the BST for Xe, Kr, Cs, I,

Ba, Te, and Ru. This calculation is dependelt upon the measured data

presented in Tables 9 and 10, the postexperiment liquid and gas volumes in

the BST, and the ORIGEN2.-calculated center bundle inventory shown in

Table 4.

69

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TABLE 9. BST LIQUID GRAB SAMPLE PRELIMINARY RESULTS

and have

Isotope Sample 1A

1-131 9.6 x 10-1Te-132 2.8 x 10-1Cs-137 1.2 x 10-2Cs-136 9.6 x 10-3Ba-140 7.2 x 10-1Ru-103 ND

a. Vecay correction is simple e

All Measurements are in microCuries/cm3been decay corrected to the time of the experimenta

Sample lB Sample 2A Sample 26

9.4 x 10-" 9.8 x 10"I 1.022.6 x 10-1 2.7 x 101- 2.8 x 10-11.2 x 10-2 1.1 x 10-2 1.2 x 10-29.6 x 1o-3 9.1 x 10-3 9.1 x i037.1 x 101- 6.8 x 101- 7.3 x 10-18.6 x 10-4 6.0 x 10-4 6.2 x 1(04

Uncertainty

0.6 x 10-10.2 x 10-10.1 x 10-20.9 x 10-30.6 x 10-10.9 x 10-4

xponential decay with no branching effects.

( (

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TABLE 10. BST VAPOR GRAB SAMPLE GAMMA SPECTROSCOPY RESULTS

All Measurements are in microCuries/cm3

and have been decay corrected to the time of the experimenta

Isotope

Xe-133Xe-133mXe-131mKr-851-131

Sample 1

3.7 x 109.1 x 10-17.3 x 10-25.9 x 10-31.2 x 10-4

Sample 2

3.58.77.15.61.2

XXXXx

1010-110-210-310-4

a. Decay correction is simple exponential decay with no branching effects.

TABLE 11. BST VAPOR GRAB SAMPLE MASS SPECTROSCOPY RESULTS

Results are percent by volume

Element Sample 1 Sample 2

H2 3.7 3.6N2 78 7802 4.0 3.8Ar 14.2 14.3CO2 0.06 0.05

71

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Prior to the LP-FP-2 experiment, the BST liquid volume was 25.48 m3

and the gas volume was 59.11 m3 (total = 84.59 mi3 ). After the

experiment, the liquid volume increased by 5.25 m 3 to 30.73 mi3, and the

gas volume decreased by 5.25 m3 to 53.86 m3 .

Assuming that the grab sample data represent average conditions

existing in the BST following the experiment, the average sample

concentrations shown in either Table 9 or Table 10 are multiplied by the

respective liquid or vapor volumes (presented above) to obtain the total

amount of curies of each nuclide in the BST. This calculation assumes no

adjustment due to deposition or plateout of fission products in the BST,

which are undetected by the current measurement process. The results of

the calculation are shown in Table 12, along with the ORIGEN2 calculated

fuel inventories and the ratio of the BST activities to the respective

center bundle fuel activities.

The release fraction data presented in Table 12 should be interpreted

as only an initial estimate of the cumulative fraction of the center bundle

inventory that reaches the BST via one of four paths: (a) LPIS, (b) Fl

sample line, (c) F2 sample line, or (d) the PORV (which was opened briefly

during the reflood portion of the experiment). Although the release rate

data in Table 12 should not be interpreted as a cumulative source term for

the experiment, it does represent a minimum, or lower bound, estimate of

the source term. Finally, notice the consistency in the release rates for

the noble gases and the two independent cesium release numbers shown in

Table 12.

The second part of the analysis in this section involves the

determination of the mass concentrations in the LPIS (near the

65 spectrometer) for cesium, iodine, and rubidium. The calculation is

based on the activity data presented in Tables 7 and 8 (or Figures 34

through 42) and the specific activity information shown in Table 4. By

dividing the measured isotopic activity concentration by its specific

activity (see Table 4 for definition), an estimate can be made for the

amount of that element that should be present near the G5 detector. For

72

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C (

TABLE 12. CUMULATIVE RELEASE FRACTIONS TO THE BST

BST Data (decay corrected) CumulativeRelease FractiORIGEN2 calculated

Nuclide Fuel Inventory (Ci) Liquid (Ci) Gas (Ci) lotal Ni) to tne bmI1-131 47570.0 29.96 0.0065 29.97 0.00063

Cs-lJb 100.2 0.287 0.287 0.0029Cs-137 144.6 0.361 0.361 0.0U25

Kr-85 17.3 0.310 0.310 0.0179

Xe-131M 237.2 3.88 3.88 0.0164Xe-133 1104UO.O 1939.0 1939.0 0.0176Xe-133M 3362.0 47.9 47.9 0.0142

Te-132 90890.0 8.37 8.37 0.000092

ba-140 80250.0 21.82 21.82 0.00027

Ru-103 18110.0 0.0213 0.0213 0.0000012

a. The cumulative release fraction is defined as the ratio of the total number of curies of a nuclide inthe BST to the number of curies of this nuclide in the center bundle.

rns

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example, the specific activity of 1-131 is 55,370 Ci per gram of I. The

1-131 measured activity concentration at 1790 s is

2620 micro-curies/cm 3. Therefore, the mass of I that should exist in theLPIS at 1790 s is calculated to be 0.0473 micrograms/cm3

(0.0473 = 2620/55,370). The results of these calculations, for the iodine,

cesium, and rubidium isotopes measured at the G5 detector location, are

shown in Figure 43.

From Figure 43 it is clear that the concentration of iodine, as

indicated by the five separate calculations based on the 1-131, 1-132,

1-133, 1-134, and 1-135 gamma spectrometer data and the ORIGEN2 specific

activities, are in good agreement. This tends to indicate that the G5 data

are consistent and probably good. In addition to the iodine calculation,

an estimate of the cesium mass concentration, based on the Cs-138 gamma

spectrometer data (which were decay corrected), and an estimate of the

rubidium concentration are also shown in Figure 43. Since Cs-138 has a

relatively short half-life, the gamma spectrometer data for Cs-138 was

decay corrected before the cesium mass concentration was calculated.

The above analyses assume that the relative proportions of the

isotopes within one elemental group are the same for the fuel and the

LPIS. This condition should be met for most measured isotopes that have a

long enough half-life relative to the sampling time (or at least can be

decay corrected as in the case of Cs-138), and whose concentrations do not

depend greatly upon the concentrations of more populous parent nuclides.

For the purposes of generating the results shown in Figure 43,

parent-daughter relationships have been ignored.

From the data presented in Figure 43, it is estimated that the Cs

concentration in the LPIS near the G5 detector is about 0.078 x 10-6

3 -6 3g/cm and the I concentration is approximately 0.055 x 106 g/cm

If it is assumed that all of the I is transported as CsI and that the Cs is

transported as either CsI or CsOH, the mass concentrations of CsI and CsOH

necessary to produce the above calculated concentrations of Cs and I are

computed to be: 0.1127 x 10-6 g/cm3 of CsI and 0.0229 x 10-6 g/cm3

of CsOH. These mass concentrations translate into molecular concentrations

74

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E

E

0.09

0.08

0.07

0.06

0.05

0.04

0.03

0.02

0.01

0.00

-0.010 500 1000 1500 2000 2500

Time (s)3000 3500 4000 4500

LI10-KM125-15A

Figure 43. Estimates of the elemental Cs, I, and Rb mass concentrations inthe simulated LPIS line based on isotopic activities measuredby the G5 gamma spectrometer.

75

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of 2.613 x 1014 molecules per cm3 for CsI and 9.201 x 1013 molecules

per cm3 for CsOH. Based on this information, it appears that there may

be as much as 2.8 times as many CsI molecules as CsOH molecules in theLPIS. Because of the low wall temperature of the LPIS, it is expected that

much of the Cs and I (in the form of CsI) is probably condensed on thewalls of the LPIS (or F3 filter) or on deposited aerosols. In either case,

much of the Cs and I probably deposited early during the experiment and

only a small amount reached the BST. This conclusion appears to be

supported by the fact that only a small fraction of the center bundleinventory of Cs and I (compared to the noble gases) was detected in the

BST, as shown by the information contained in Table 12.

4.7 Potential for Meeting Fission Product Measurement Objectives

The fission product measurement objectives were derived from the

governing objectives for Experiment LP-FP-2. Achievement of the fission

product measurement objectives depends upon the thermal/hydraulic

instrumentation, as well as the FPMS and associated PIE. Since PIE is

incomplete and FPMS data have not been fully analyzed (particularly the

gamma spectrometer data), it is not possible to conclusively state that the

fission product measurement objectives have been achieved. However, on the

assumption that the data obtained, or to be obtained, are satisfactory, thepotential for achieving the objectives can be assessed.

Objective 1

Determine the fraction of volatile fission products (Cs, I, Te, Xe,

Kr) and aerosols released to and from the upper plenum region. Achievement

of this objective will require data from the Fl sample line, the F2 sample

line, and the Dl deposition coupon device. Sufficient thermal/hydraulic

data to determine flow conditions through the center fuel module and upper

structure would also support the achievement of this objective.Achievement of this objective appears likely at this time, although having

a gross gamma monitor instead of a spectrometer at the G6 location may

affect the uncertainty of the results.

76

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Objective 2

Determine the fraction of volatile fission products and aerosols

transported out of the primary coolant systen. Achievement of this

objective will require an adequate fission product mass balance as well as

data on the volatile fission product inventory in the BST. Achievement of

this objective at this time appears likely.

Objective 3

Determine the retention of volatile fission products on representative

primary coolant surfaces in the plenum and piping. Resolution of this

objective will require measurements of volatile fission product retention

on the deposition coupons in the upper structure and on deposition spool

pieces in the LPIS line. Achievement of this objective appears likely,

although if all three protected coupons on the upper plenum deposition

coupon device have been exposed to reflood, there may be some additional

uncertainty.

Objective 4

Determine the general mass balance of volatile fission products in the

fuel, primary coolant system and blowdown suppression tank. This objective

requires data from all of the fission product and PIE measurements. In

addition, some supporting thermal/hydraulic data are required for

estimating the fission product plateout on PCS surfaces. Achievement of

this objective appears likely.

4.8 Future PIE Plans

After the experiment, a variety of samp'es (fluid and FPMS components)

will be examined. These samples and the planned analyses for these samples

are identified in Table 13. Results of these analyses will provide fission

product release, transport, and deposition dzlta during the early phases of

a risk dominant reactor transient. These data will aid the understanding

of fission product behavior and will be used to assess the capability ofcomputer models to predict fission product rElease and transport.

77

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TABLE 13. PLANNED POSTIRRADIATION EXAMINATION FOR EXPERIMENT LP-FP-2

Analyses

Samples

AerosolBST/PCS BST Deposition Deposition LPIS Cyclone DilutionLiquids Vapor Device Spool Piece Filter Separator Filters

AerosolAerosol Collection Hydrogen Steam

Impactors Filter Recombiner Condenser

Gammaspectrometer

Sr-89, -90

Fissile

Total mass-10o Elemental

Particle size

Compounds

x x x x x x x x x x x

X x

x

xx

x x

x

x

x

x

K

K

xK

K

K

K

x

K

K

K

( (

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5. CONCLUSION:S

The conduct and results of this experiment are considered sufficient

to meet the experiment objectives. Specifically, the thermal/hydraulic

conditions during the release and transport of fission products adequately

simulated those expected to occur during a V.-Sequence accident. It should

be noted that much of the information, especially fission product data,

needed to confirm this general conclusion is not yet available. Thus, the

realization of the specific experiment objec:ives depends on data that are

currently being collected. Specific conclusions and observations based on

the limited analysis completed thus far are:

1. The thermal and hydraulic boundary conditions existing during

fission product. release and transport adequately simulated the

early phases of a V-sequence accident and were as desired for

fission product transport and depo;ition.

2. The fuel rod cladding temperatures exceeded 2100 K (3320 0 F) for

at least 4-1/2 min. This exceeded by 50% the experiment goal of

at least 3 min at those temperatures.

3. The flow resistance in the simulated LPIS line was much higher

than was expected. This delayed the depressurization into the

metal-water reaction time frame and resulted in a higher than

expected steam flow rate in the center fuel module.

4. A higher-than-predicted steam flow rate in the center fuel module

resulted in adequate steam to sustain a rapid metal-water

reaction in the upper region of the core and in the elevation of

the maximum temperature being much higher than was expected.

5. The center fuel module control rods melted and a significant

fraction of the module is judged to have been relocated to the

bottom of the module.

79

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6. Results of the postexperiment ORIGEN2 analysis indicate that the

Cs to I mass ratio in the center fuel module at the beginning of

the experiment is 4.00. The fission gas release calculations,

based on the measured thermocouple data and the NUREG-0072 data

base, indicate that between 20.6 and 25% of the initial

inventories of Xe, Kr, Cs, and I were released to the coolantduring the course of the experiment.

7. Based on the limited amount of fission product data available at

the time of this report, it is judged that all experiment

objectives are achievable.

80

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6.. REFERENCES

1. P. R. Davis et al., "The Risk Significant of Transient Accidents fromPRA Studies," ANS Topical Meeting on Anticipated and AbnormalTransients in Light Water Reactors, Jackson, WY, September 1983.

2. Reactor Safety Study--An Assessment of Accident Risks in U.S.Commercial Nuclear Power Plants, WASH-1400, USNRC, October 1975.

3. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiments L5-1 andL8-2, EGG-LOFT-5625, October 1981.

4. D. B. Jarrell and J. M. Divine, Experiment Data Report for LOFTIntermediate Break Experiment L5-1 and Severe Core TransientExperiment L8-2, NUREG/CR-2398, EGG-2136, November 1981.

5. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiment L3-1,EGG-LOFT-5057, November 1979.

6. P. 0. Bayless, J. B. Marlow, R. H. Aver'11, Experiment Data Report forLOFT Nuclear Small Break Experiment L3-1, NUREG/CR-l145, EGG-2007,January 1980.

7. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiment L3-5/L3-5A,EGG-LOFT-5242, October 1980.

8. L. T. L. Dao and J. M. Carpenter, Experiment Data Report for LOFTNuclear Small Break Experiment L3-5/L3-EA, NUREG/CR-1695, EGG-2060,November 1980.

9. G. E. McCreery, Quick-Look Report on LOFT Nuclear ExperimentL3-6/L8-1, EGG-LOFT-5318, December 1980.

10. P. 0. Bayless and J. M. Carpenter, Experiment Data Report for LOFTNuclear Small Break Experiment L3-6 and Severe Core TransientExperiment L8-1, NUREG/CR-1868, EGG-207E, January 1981.

11. V. T. Berta, OECD LOFT Project Experiment Specification DocumentFission Product Experiment LP-FP-2, OEC[' LOFT-T-3802, Rev. 1, May 1985.

12. J. P. Adams et al., Quick-Look Report on OECD LOFT Experiment LP-FP-l,OECD LOFT-T-3704, March 1985.

13. D. L. Reeder, LOFT System and Test Description (5.5 ft Nuclear Core ILOCES), Change 1, NUREG/CR-0247, TREE-1208, July 1978.

14. L. J. Ybarrondo et a]., "Examination of LOFT Scaling," 74-WA-HT-53,Proceedings of the Winter Meetino of the American Society ofMechanical Engineers, New York, November 17 - 22, 1974, CUN-741104.

81

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15. V. H. Ransom et al., REALP5/MOD2 Code Manual, EGG-SAAM-6377, April1984.

16. J. W. Spore et al., TRAC/BOI: An Advanced Best Estimate ComputerProgram for Boiling Water Reactor Loss-of-Coolant Accident Analysis,NUREG/CR-2178, EGG-2109, uctober 1981.

17. 0. J. Usetek et al., "Fission Product Behavior during the First TwoPBF Severe Fuel Damage Tests," ANS Topical Meeting on Fission ProductBehavior and Source Term Research, Snowbird, Utah, July l5-19, 1984.

18. J. P. Adams and J. C. Birchley, Quick-Look Report on OECD LOFTExperiment LP-LB-l, OECO LOFT-T-3504, February 1984.

19. S. Guntay, M. Carboneau, Y. Anoda, Best Estimate Prediction for OECDLOFT Project Fission Product Experiment LI-l-P-Z, ULGU LUI-I-i8UJ,June 1985.

20. A. G. Croff, A Users Manual for the ORIGEN2 Computer Code,ORNL-TM-7175, July 1985.

21. B. L. Rushton and J. B. Briggs, PDQ Calculated Results for SafetyAnalyses Evaluations for the FP-2 Reload Core at Beginning of Life,OECD LOFT-I-08-5118, December 21, 1984.

22. M. R. Kuhlman et al., CORSOR User's Manual, NUREG/CR-4173, BMI-2122,March 1985.

23. M. L. Carboneau, A Report on the Transient Isotope Generation andElemental Release (TIGER) Program, to be published.

24. USNRC, Technical Bases for Estimating Fission Product Behavior DuringLWR Accidents, NUREG-0772, June 1982.

25. R. R. Sherry letter to 0. L. Batt, "Letter Report on the LOFT FP-2CORSOR Analysis," CD-CAS-85-175, August 9, 1985.

26. P. North letter to J. E. Solecki, "Potential Programmatic RiskAssessment for LP-FP-2 Experiment," PN-106-85, May 31, 1985.

82

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APPENDIX AREVISIONS TO THE EXPERIMENT SPECIFICATION

DOCUMENT FOR EXPERIMENT LF'-FP-2

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( OPER. NO.

PAGE I...OF -- 3-FORM EG&G-1844R na- DOCUMENT REVISION BEQUEST

( ) REQUESTER I OAR DATE , OAR NO.V.T.Berta 5-31-85 L-7423

(5.) DOCUMENT NO. (IF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE DATE

OECD LOFT-T-3802 Rev.1 OECD LOFT Project Experiment Specification May 19851nm mpi F'nm.4 P-rnt'lt t.xeXrm TLP-FPý2 Na 18

(D) CHECK APPLICABLE BLANK (-) MAAGEFP APPR9VAL .r DATE

PERMANENT CHANGE X TEMPORARY CHANGE __ BULLETIN 7A ,/,(,/L" " . -2.- ,AJPRINT OR TYPE PROPOSED CHANGE -- NUMBER EACH CHANGE SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER FORFOR EACH CHAeNGE. WATR'S USE

STEP OR INSTRUCTIONS: REWRIrEPARAGRAPH(S) OR FOR EXTENSIVE CIIAIGFS ATTACH REVISED COPY AND STATE "REVISE PERITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT ANE STATE "PREPARE NEW (SP. DOP. ETC.) PER ATTACHEDDRAFT'.

1 8 2nd Section 3.4 2nd paragraph let sentence. Delete "At 6+-is".Begin the sentence with the word "Following".

2 7 1st Section 3.3 1st paragraph. Replace the first paragraph with thefollowing paragraph.

The transient phase of the experiment will be initiated bya reactor shutdown as specified in Section 4.3. Time zero

corresponds to the action taken to drop the center fuel module(CFM) control rods. The 1.16 in. inner diameter simulated break

in the intact loop cold leg will bE opened at 20+2s. The primarycoolant pumps will be tripped at 25i±5s and will undergo a normal

coastdown. At 220*5s the simulatec. LPIS line (also 1.16 in.diameter) will be opened. (The LPIS line filter shall be bypassedto prevent plugging prior to the fission product release.) This

line is connected to the broken loop hot leg and to a blowdown

suppression tank (BST) inlet vent. The intact loop cold legsimulated break will be closed at either a CFM cladding temperature

(CONTINUED) USE CONTINUATIIN SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT

®)JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. AROVE): (1V OTHER DOCUMENTATION AFFECTED:

DOC. NO. ORR NO. DATE COMPLETEDApproved experiment planning revisions

(12) ORIGINATING OAR NO:

@ REVIEW_

NAME/SIGNATURE ORG. DATE NAMEISIGNATURE OCG _ RATE NAME/SIGNATURE ORG. DATE

-- 11 _ 4 ____ f•____ __-_ _ PUACTY

cpr 24-10 s/i______I ________ SHS

7 1, o 1ii~J ______________

(9COMMENTS: * I f(P.C 4 ,A,.-i s RADDITIONALRS IN THISDOCUMENT

REVISION

DOCUMENT CONTROLLER

Q-Mýl_) DOCMEN CTR L. 6-" ... D cM.

nMPI ETED DATE:

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DOCUMENT REVISION REQUEST OPER. NO.(CONTINUATION SHEET) PAGE OF 3

FORM EG&G-1544A(A" 5-"7) E .- -74;

DOCUMENT NO. (OF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE DATEBECD LO TPro ect Ex~erimntSeii tinOECD LOFT-T-3802 Rev.1 •ocUMen7Fsson Pro uc• Epterimen• ,f -•-E May 1985

PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR

WRITER'S

STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPr. FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP. ETC.) PER ATTACHED DRAFT".

2 7 lst of 566 + 5K (560 + 90F) or a system pressure of 1.2 + 0.03 MPa

(160 + 5 psig). The fuel rod cladding is calculated to beginheatup at approximately 700s and to reach 2100K at approximately 17(0s.Fission product and aerosol release will occur from fuel rod

failure until the center assembly shroud outer wall reaches

1473K (21920F) or the peripheral fuel rod cladding reaches1417K (20920F), at which time the fission product filter samplingsystems and the upper plenum deposition coupons will be closedin the primary coolant system coincident with the closure of the

simulated LPIS line. Within 6 + O.5s of these actions core reflood

will commence with both accumulators.

3 13 13 Step 13 is to'read as follows:13. Close the FPMS sampling system lines, the deposition coupon

device, and the simulated LPIS line in the broken loop hot

leg at a cladding temperature of 1417K (2092OF) or a

thermal shroud outer wall temperature of 1473K (21920F).

4 13 14 Step 14 is to read as follows:14. Initiate core reflood 6 + 0.5s after initiation of the

system closures in Step 13.

5 20 6 2nd paragraph of Section 4.7 is to read as follows:

Sequence Step 7 begins the transient phase of the experiment.Reactor shutdown with all control rods in as specified in Step 7

is to be completed before proceeding to Step 8. The eleven control

rods in the CFM are the aerosol source for the experiment.

If these control rods cannot be inserted the experiment sequence-mu t be

(the remaining part of this paragraph on page 21 is unchanged)

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DOCUMENT REVISION FREQUEST OPER. NO.(CONTINUATION SHEET) PAGE 3 OF 3

FORM EG&G.1i44A(Rev. 5-77) DRR NO. -__74 2 3__"_

DOCUMENT NO. (IF APPLICABLE) OECD LOFT PrO TntNW riren t S peci fi cation DOCUMENT ISSUE DATE

OECD LOFT-T-3802 Rev.1 Document Fission Product ExpEriment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER

FOR EACH CHANGE.FOR

WRITER'SSTEP OR I INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHAIGES ATTACH REVISED COPY AND STATE "REVISE PER USE

ITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP, ETC.) PER ATTACHED DRAFT".

6 22 table Delete the second sentence in the course of action for item 1. The

course of action is to read as follows:

If the F1 and F2 isolation valvesdo not open within 5s terminate theexperiment and commence recoveryoperations with the ECCS. Do not openthe LPIS line filter and do not store thegamma densitometer sources.

7 22 table Add the following item to the table on page 22:

All criteria met and F1 and The experiment is to proceed.

F2 open but LPIS line filter Continue attempts to completecannot be opened and/or these actions.gamma densitometer sourcescannot be stored.

8 22 Add the following sentences after tie table on page 22. Thisaddition is riot a new paragraph.

If the condition occurs where the s'stem pressure is above themaximum allowable for F1 and F2 operation when the CFM claddingtemperature is 840K (10520F), commence actions to lower thesystem pressure. These actions may be, but not limited to, openingthe PR x xr A Terminatea othese actions before the CFM cladding temperature reaches150K 4310 F). The experiment is to continue in the event thatthe system pressure cannot be lowered below the F1 and F2operati.ng pressure limit.

opening the intact loop break path.

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DRR-L-7423

Rev. 1 Chg. I

of an effective 40 hours will establish the required minimum decay heat

level of 675 kW at 200 s and will complete the burnup required on the

9.72 wt% enriched fuel rods. The plant should be operated at a steady

state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment

initiation.

3.3 Transient Phase

The transient phase of the experiment will be initiated by a reactor

shutdown as specified in Section 4.3. Time zero corresponds to the action

taken to drop the center fuel module (CFM) controls rods. The 1.16 in.

inner diameter simulated break in the intact loop cold leg will be opened

at 20 ± 2 s. The primary coolant pumps will be tripped at 25 ± 5 s and

will undergo a normal coastdown. At 220 ± 5 s the simulated LPIS line

(alos 1.16 in. diameter) will be opened. (The LPIS line filter shall be

bypassed to prevent plugging prior to the fission product release.) This

line is connected to the broken loop hot leg and to a blowdown suppression

tank (BST) inlet vent. The intact loop cold leg simulated break will be

closed at either a CFM cladding temperature of 566 ± 5 K (560 ± 91F) or a

system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is

calculated to begin heatup at approximately 700 s and to reach 2100 K at

approximately 1700 s. Fission product and aerosol release will occur from

fuel rod failure until the center assembly shroud outer wall reaches 1473 K

(2192 0 F) or the peripheral fuel rod cladding reaches 1417 K (20921F), at

which time the fission product filter sampling systems and the upper plenum

deposition coupons will be closed in the primary coolant system coincident

with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these

actions core reflood will commence with both accumulators.

The beginning and end of the transient phase of the experiment are

defined as follows:

7

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DRR-L-7423

Rev. 1 Chg. I

Beginning initiation of the transient hy a reactor scram

End initiation of the closure of the simulated LPIS line in

the broken loop hot leg.

3.4 Posttransient Phase

The posttransient phase consists of a time interval of 12 hr for

measurement of (a) the redistribution of fission product-inventory in the

gas and liquid volumes in the blowdown suppression tank (BST), and (b) the

leaching of fission products from the damaged fuel rods in the primary

coolant system (PCS). The beginning and end of the posttransient phase of

the experiment are defined as follows:

Beginning initiation of the closure of the simulated LPIS line in

the broken loop hot leg

End completion of the time interval specified for fission

product measurements.

Following closure initiation of the simulated LPIS line in the primary

coolant system, reflood operations will commence with initiation of both

accumulators. System refill will continue as required with the high

pressure injection system (HPIS). The governing requirements for plant

operation involving the PCS in this phase are:

1. Mass transfer from the PCS is to be minimized. The purification

system can be used in the decay heat removal mode (bypassing the

ion exchanger) for temperature control along with steam generator

feed and bleed.

2. Forced coolant circulation with the primary coolant pumps is

prohibited.

3. PCS temperature is to be reduced and maintained below 449 K

(350*F) as soon as possible after simulated LPIS line closure.

8

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DRR-L-7423

Rev. 1 Chg. I

b. Center fuel assembly cladding temperature increases to

566 ± 5K (5601 ± 90 F).

The cladding temperature following time zero will decrease and

correspond to saturation temperature until core uncovery occurs.

Then, cladding heatup will commence at some temperature below the

value specified in (b). Therefore, cladding temperature must be

increasing in order for criterion (b) to be valid.

12. Open the FPMS F1 and F2 sampling systems at a center fuel

assembly cladding temperature of 840 ± 5K (1052 ± 91F) and open

only if the system pressure is less than 1.43 ± 0.03 MPa

(195 ± 5 psig). )if the pressure criterion is met, for the F1 and

F2 sampling systems, also open the LPIS break line filter and

store the gamma densitometer sources. Thermocouples to be used

in determining the temperature are listed in Table 8.

13. Close the FPMS sampling system lines, the deposition coupon

device, and the simulated LPIS line in the broken loop hot leg at

a CFM cladding temperature of 1417 K (20921F) or a thermal shroud

outer wall temperature of 1473 K (2192°F).

14. Initiate core reflood 6 ± 0.5 s after initiation of the system

closures in Step 13.

15. Continue system refill with the HPIS and maintain the PCS

temperature below 449 K (350'F) as soon as possible for the

remainder of a 12 hr minimum time after initiation of core

reflood. Maintain the PCS pressure below 8.96 MPa (1300 psig).

Mass transfer to and from the PCS, excluding system leakage and

replacement, is to be minimized. PCS energy control is to be

accomplished with steam generator feed/bleed operations and/or

13

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with the purification system in the decay heat removal mode (ion

exchanger bypassed). PORV venting is permitted but is to be

minimized to the extent possible.

16. Maintain isolation of the BST (no mass transfer into or out of

the BST) for the same time interval as in Item 15 subject to

plant limiting conditions which may require operator action.

Within 10 minutes of PCS isolation purge the BST downcomer with

60 SCF of N2.2

17. Maintain PLSS data acquisition continuously over the time

interval for Items 15 and 16.

4.4 System Configuration

The general system and component configuration of the LOFT PWR for

Experiment LP-FP-2 is shown in Figure 1. Specific details are given inthe

following sections on the reactor core, primary coolant system, secondary

coolant system, blowdown system, and emergency core coolant system.

4.4.1 Reactor Core

Experiment LP-FP-2 will be conducted with a specially constructed

center fuel assembly. The cross section of this fuel assembly is shown ini

Figure 2. The fuel rods are 350 psia prepressurized and 9.72 wt%

enriched. The outer two rows of fuel rods have been replaced with a

thermal shield of zircaloy with zirconium oxide ceramic internal

insulation. The purpose of the higher than normal enrichment and the

thermal shield is to provide at least 3 minutes at peak cladding

temperatures above 2100 K (3321*F) in the center fuel assembly before the

peripheral assembly fuel cladding reaches the transient termination

temperature. This time at temperature will provide a sufficiently large

fission product and aerosol release fraction.

The aerosol release will occur from (Ag-In-Cd) control rods which will

be inserted at reactor scram in the guide tubes shown in Figure 2.

14 I

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4.5.3 Simulated LPIS Pipe

Thermal-hydraulic measurements in the simulated LPIS pipe will consist

of steam flow, steam temperature, and wall temperature.

4.5.4 FPMS

The identification and location of fission product measurements are

shown in Figure 4. These measurements are described in the FPMS Functional4

and Operational Requirements (F&OR) document.

4.5.5 Postirradiation Examination

An integral part of the measurements which are necessary to meet the

experiment objectives are the postirradlation examination (PIE)

measurements. A summary of' items specified for postirradiation examination

is contained in Table 5. Examination of these items will be in accordance

with the postirradiation plan for LP-FP-2. 5

4.5.6 Critical Measurements

Sets of critical measurements, required during the transient and

posttransient phases of the experiment, have been identified and are listed

in Tables 6 and 7, respectively. The transient phase of the experiment

should not be initiated without these measurements since the experiment

objectives may be jeopardized. Appendix A lists by instrument identifier

all critical measurements which are considered necessary for the successful

conduct of the experiment. The measurement uncertainties will be equal to

or less than those specified in the document "LOFT Experimental Measurement

Uncertainty Analysis," NUREG/CR-0169.

A complete list of measurements required for Experiment LP-FP-2 is

provided on the Data Acquisition Requirements list to be published prior to

the experiment.

19 I

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DRR-L-7423

Rev. 1 Chg. 1

The digital data acquisition and processing system, and analog and

digital data acquisition recording is required to begin no later than 1 min

before initiation of the transient phase of the experiment. Continuous

PLSS recording is required through the posttransient phase of the

experiment.

Measurements identified on the Data Acquisition Requirements List that

fail prior to experiment initiation should be repaired or replaced if

possible. If a failed instrument(s) cannot be repaired or replaced, the

Joint Experiment Group shall determine the course of action.

Process instruments requiring calibration prior to Experiment LP-FP-2

are listed in Table 1.

4.6 Experiment Termination

Experiment LP-FP-2 will be terminated at the end of the posttransient

phase of the experiment. The posttransient phase ends with the completion

of the time interval required for monitoring the redistribution of fission

products in the vapor and liquid volumes in the blowdown suppression

system. The time interval is specified in Section 4.3 to be 12 hr minimum

after closure of the simulated LPIS line.

4.7 Abnormal Experiment Sequence

If instrumentation, hardware components, or operating systems fail

prior to or during any of the four phases of the experiment, every effort

should be made to substitute, repair, or provide alternate actions to

safely continue the experiment and to meet the programmatic objectives.

Sequence Step 7 begins the transient phase of the experiment.

Reactor shutdown with all control rods in as specified in Step 7 is to

be completed before proceeding to Step 8. The eleven control rods in

the CFM are the aerosol source for the experiment. If these control

rods cannot be inserted the experiment sequency must be

20

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stopped. On completion of the corrective actions the experiment is to be

resumed as approved by the LOFT Operations Manager, DOE Site Program

Manager and LOFT Program Division Manager.

The decision matrix on center assembly control rod insertion is as

follows:

Condition Decision

Reactivity change and one rod bottom Continue experiment sequencelight indication

Reactivity change and no rod bottom Terminate experimentlight indication

No reactivity change and no rod bottom Terminate experimentlight indication

Sequence Step 8 opens the break path in the intact loop cold leg.

This flow path is the primary blowdown path and is intended to be the path

for venting the major part of the primary coolant system fluid. High

quality steam flow only is desired for venting through the simulated LPIS

pipe in the broken loop hot leg. If the intact loop cold leg break cannot

be opened, the experiment sequence must be halted. On completion of the

corrective actions the experiment is to be resumed as approved by the LOFT

Operations Manager, DOE Site Program Manager and LOFT Program Division

Manager.

The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.

This flow path is the designed fission product and aerosol vent path to the

BST. If this line cannot be opened within 50 s of the time specified,

close the intact loop cold leg break and commence recovery operations with

the ECCS. Resume the experiment as approved by the LOFT Operations

Manager, DOE Site Program Manager and LOFT Program Division Manager.

The intact loop cold leg break is to be closed on either a cladding

temperature value or a system pressure value as specified in Sequence

Step 11. If this break is not closed then another flow path to the BST

will exist for fission product and aerosol venting. Two vent paths are not

21 I

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DRR-L-7423

Rev. 1 Chg.

provided for in the experiment plan. If this break cannot be closed before

the CFM cladding reaches 840 ± 5 K (1052 ± 90 F) then terminate the

experiment and commence recovery operations with the ECCS. If this break

has not been closed after 566 K (5601F) and the system pressure decreases

below 1.2 ± 0.03 MPa (160 ± 5 psig) before the cladding reaches 840 ± 5K

(1052 ± 91F) then, the Operations Branch is to consult with the JEG to

decide if attempts to close this break should continue or if the experiment

should be terminated. If experiment termination and plant recovery

operations commence with the ECCS, return to Sequence Step I or as approved

by the LOFT Operations Manager, DOE Site Program Manager, and LOFT Program

Division Manager after repairs are made.

The FPMS F1 and F2 sampling systems are to be opened at a cladding

temperature of 840 ± 5K (1052 ± 90 F) if the system pressure has decreased

below 1.43 ± 0.03 MPa (195 ± 5 psig). The pressure criterion corresponds

to the F1 and F2 design pressure limit of 200 psig. The LPIS line filter

is to be valved in and the gamma densitometer sources are to be stored in

conjunction with the opening of the F1 and F2 sampling systems. The

following courses of action are defined in the event that abnormal

conditions occur:

1

Abnormal Condition Course of Action

System pressure below criterionbut F1 and F2 valves fail to open

Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.

All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.

If the F1 and F2 isolation valvesdo not open within 5s terminate theexperiment and commence recoveryoperations with the ECCS. Do notopen the LPIS line filter and do notstore the gamma densitometer sources.

Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.

The experiment is to proceed.Continue attempts to completethese actions.

N '-, I

22

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DRR-L-7423Rev. I Chg. 1

If the condition occurs where the system pressure is above the maximum

allowable for F1 and F2 operation when the CFM cladding temperature is

840 K (1052°F), commence actions to lower the system pressure. These

actions may be, but not limited to, opening the PORV or opening the intact

loop break path. Terminate all of these actions before the CFM cladding

temperature reaches 1050 K (1431°F). The experiment is to continue in

the event that the system pressure cannot be lowered belwo the F1 and

F2 operating pressure limit.

The LPIS break line filter is to be valved in when the peak cladding

temperature is 840 + 90 F). As the filter loads up the differential pres-

sure may increase. At a filter differential pressure of

22A

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(59 OPER. NO

FORM EG&G-1844 ,DOCUMENT REVISION iFUUEIST PAGEfle ,. Wg- 2) P. IOF

(D REQUESTER (1 ORR DATE DAII NO

V. T. Berta June 26, 1985 L - 79 37C5 DOCUMENT NO. (IF APPLICABLE) DOCUMENT In E DOCUMENT ISSUE DATE

OECD LOFT Project Experiment Specification DocumentOECD LOFT-T-3802 Rev. 1 Fission-Iroduct-Experiment LP-FP-2 .May 1985

® CHECK APPLICABLE BLANK b(,1) M-ANAR APPROVAL/, DATE

PERMANENT CHANGE __ TEMPORARY CHANGE .-- II,.,.ETIN

(1 PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEQUENTIALLY IN IST COLUMN AND RFCORD PAGE AND STEP OR PARAGRAPH NUMBER 1 9 FORFOR EACH CHANGE. WR TER'S USE

v 4,%

ITEM PAGESTEP OR

PARA.INSTRUCTIONS: REWRI[E PARAGRAPH(S) ORR EX TENSIVE CHANGES AT TACH REVISED COPY AND STATE 'REVISE PERATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGHOIIA.FTANID STAI E PREFARE NEW (SP. DOP. ETC.) PER ATTACHEDDRAFr.

-4-4-4

1 5 3.1s

Place an asterisk after the sentence, "In termw of reactoroperation...."

Add the following footnote at the bottom of the page:

Any combination of preconditioning and pretransient power

operation that provides the specified initial conditions(Section 4.2) is permissable. The specific combinationdescribed is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation.

USE CONTINUATION SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATEIEVENT

0

(• JUSTIFICATION: (REASON FOR CHANGE - NUMBEFi TO CORRESPOND TO ITEM NO. ABOVE) (@1 OTIIER DOCUMENTATION AFFECTED:

Provides the Operations Branch with flexibility in DDCNO. ORRNO. DATECOMPLETEDattaining the specified initial conditions and allows for OECD-LOET-I_-I_1515- ] . h!5!/..power operation variations that may be needed to account or---factors that may arise which affect power operation.

T172); 0RIGINAIING DRR NO:

@9 REVIEW

NAMEISIGNATURE ORG DATE NAMFISIGNATIUInE ORG. RATE NAME/SIGNATIURE ORG. DATE

. m-i.cy 10 &Wjp ed•", _ QUALITY

__,_,_____/,j____,o_____.-__ ASHS

Iazc Z ~ i 0~ it _ _~ Ii_ _ _ _ _ _ PRAC __

(94 COMEWS: I q 5ADDITIONALRRS IN THISDOCUMENTREVISION

Q) DOCUMENT CONTROLLER J), RELEASE DTE: \- • .'-IR" 1 I. :OMPLEIEO DATE:JT:•A--• ", we- '\ .

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)RR-L-7437Rev. 1 Chg. 2

3. EXPERIMENT DESCRIPTION

Experiment LP-FP-2 consists of four distinct phases. These phases are

designated as (a) fuel preconditioning, (b) pretransient, (c) transient,

and (d) posttransient. The four phases together represent a continuous

process and have specific beginning and ending definitions. Each phase is

described in the following sections.

3.1 Preconditioning Phase

The purpose of the preconditioning phase is to subject the fuel rods

in the new center assembly [19.72 wt% enriched, 2.41 MPa (350 psia)

prepressurization] to a burnup which, in combination with the burnup

corresponding to the EFPH requirement at the experiment initial conditions,

provides the required minimum burnup for the experiment. The minimum

burnup required for these fuel rods is 325 MWD/MTU. The minimum burnup

required in the preconditioning phase is 252 MWD/MTU. In terms of reactor

operation, the preconditioning burnup is equivalent, as an example, to

power operation at a maximum linear heat generation rate (MLHGR) of

52.2 kW/m (16 kW/ft) for 111.5 hours on the 9.72 wt% enriched fuel rods* A

core power level of 32.0 MW ± 0.5 is calculated to provide a MLHGR of

-16 kW/ft on the 9.72 wt% enriched fuel rods. Power profile data using

the traversing incore probe system (TIPS) must be obtained during the

preconditioning phase. The TIPS locations in the center assembly will be

capped during the reactor shutdown time in the pretransient phase.

The beginning and end of the preconditioning phase are defined as

follows:

Beginning the start of plant heatup prior to power operation to

establish fuel burnup

End termination of power operation after the required

burnup in this phase has been achieved.

Any combination of preconditioning and pretransient power operation that provides

the specified initial conditions (Section 4.2) is permissable. The specificcombination described is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation .

5

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The preconditioning phase does not require special procedures relative

to experiment specification and is to be conducted with established plant

operating procedures.

3.2 Pretranslent Phase

The pretransient phase consists of a reactor shutdown interval of

2-5 days followed by a power operation interval. The final plant

preparations are to be completed during the reactor shutdown interval. The

power operation interval is to complete the required central assembly fuel

burnup, establish the required minimum decay heat level, and establish the

required initial conditions for conduct of the experiment. The beginning

and end of the pretransient phase are defined as follows:

Beginning termination of the power operation after the required

burnup is achieved in the preconditioning phase

End initiation of the transient by a reactor scram.

Specifically, the final plant preparations required by this

specification are:

1. Operational readiness of thermal hydraulic and nuclear

measurements (experimental and process) required by this

specification for this experiment and approval by the LOFT

Operations Branch to proceed into power operation andsubsequently into the transient phase of the experiment.

2. Operational readiness of the fission product measurement system

(FPMS) in accordance with FPMS operational documents.

The power operation prior to experiment initiation is to be an

effective 40 hours at a core power level of 26.5 ± 0.5 MW (21.2 EFPH*).

This core power is the required initial condition value. Power operation

*EFPH is defined as 50 MW-hr.

6

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FORM EGAG.Bd44(Ile-. 09-82)

DOCUMENT REVISION REQUEST() OPER. NO. -

PAGE 1_L OF § 5

(D REQUESTER 39 OAR DATE 49 DRR NO.

V. T. Berta June 17, 1985 L-7435() DOCUMENT NO. (IF APPLICABLE) OECD LOFT Prq95ftNT P erimen t Speci fi cation DocumPfEMENT ISSUE DATE

OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 198569 CHECK APPLICABLE BLANK (0) MAGERfPPROV DATE

PERMANENT CHANGE X TEMPORARY CHANGE - BULLETIN /t"

@9 PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEDUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER (i FORFOR EACH CHANGE. WRTER'S USE

STEP OR INSTRUCTIONS: REWRITE' PAGRAPHIS) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE REVISE PERI TEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP. ETC I PER ATTACHED DRAFT'.

1 10 Section 4.1 item 3 Change to read:

3. Complete the pre-experiment calibration requirements for the FPM!instruments and the pre-experiment calibration requirementsspecified in DOP-87-005, "DAVDS Experimental Measurements andTest Procedure," and in DOP-87-008, "Pre LOCE Data Verification"

2 71 Tbl 2 Change the first entry to read:

Reactor shutdown CFM CR drop t = 0Peripheral CR scram 5 s

- 3W I add "or other experiment termination events." under time/setpoint

for FPMS, Broken loop and deposition closure.

Change time/se-tpoint for FPMS sampling system isolation valvea"Open" from A89i 840 + 5 K (1052 + 9 F) to 811 + 5 K (1000 + 9 F)

USE CONTINUATION SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT

6•JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVEI: (© OHER DOCUMENTATION AFFECTED:

DOC. NO. DRR NO. DATE COMPLETED

Approved experiment changes. OECD LOFT-I-11-5150 6/5/85

r(l2 ORIGINATING DAR NO:

(D3 REVIEW

NAME/SIGNATURE ORG. DATE NAME/SIGNATURE ORG. RATE NAME/SIGNATURE ORG. DATE

16o j2 QUALITY

~21 _ _SHS

PRAC

1(9 COMMENTS: 0 ADDITIONALARS IN THIS

DOUMN

'i) DOCUMENT CONTROLLER

1 0 ;Z-11

EEASE .,- , ORBCOMM ETED DATE:

K I _--_ AR

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DOCUMENT REVISION REQUEST OPER. NO.(CONTINUATION SHEET) PAGE. 2_ OF -, 5

rOnRM EG&O-11M4AMe,. 11.70 1 DRR NO _L-7435

DOCUMENT NO. (IF APPLICABLE) _T S DOCUMENT ISSUE DATEDOCMENNO IF PPLCABE)OECD LOFT 'roje~c txperiment Specification uocumenTu

OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985

PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR

WRITER'S

STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE -PREPARE NEW (SP. fOP. ETC.) PER ATTACHED DRAFT"

Add item and action to list immediately above last item listed.

Deposition coupon device Close

The time/setpoint for close actions on the FPMS sampling system isolationvalves, the simulated LPIS pipe, and the deposition coupon deviceis as follows:

1462 217ZT4 K (2M•eF) on 1517peripheral fuel assembly 0 O2272cladding or• K (ew F)on center fuel assembly'....shroud outer wall

Change time/setpoint for core reflood to 6 + 0.5 s.

22a Replace wording at top of page 22a with the following (retain thelast paragraph on page 22a):

If the condition occurs where the system pressure is above themaximum allowable for Fl gnd F2 operation when the CFM claddingtemperature is 800 K (981 F) commence actions to lower the systempressure. These actions may be, but not limited to, opening thePORV or opening the intact loop break path. Terminate all ofthese actions before the CFM cladding temperature reaches 1050 K(1431UF). The experiment is to continue in the event that thesystem pressure cannot be lowered below the Fl and F2 operating press relimit. If the actions taken to lower the pressure are successful,terminate all actions before the system pressure decreases belowan indicated value of 1.2 MPa (160 psig).

(The replacement pages will not contain a page 22a).

ORIGINAI "

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DOCUMENT REVISION REQUEST OPER. NO. _

(CONTINUATION SHEET) PAGE 3 OF _ 5FORM EG&G-,•1A i -7435

IROv. 11-T9g non unL

DOCUMENT NO. (IF APPUCABLE) OECD LOFT - -- --t periment Specification Documen IOCUMENT ISSUE DATE

OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 'ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER

FOR EACH CHANCE. FORWRITER'S

I I STEP OR I INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPY". FOR NI-W DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP. ETC.) PER AT TACHED DRAFT".

4 7 Ist Correct typo made in first DRR on primary coolant pump uncertainty.

Change 25 + El s to 25 + 2 s

/5 13 13 Change step 13 to read as follows:

13. Close the FPMS sampling system lines, the deposition coupon

2172 device, and the simulated LPIS line in the broken loop hotleg a per~pheral fuel assembly cladding temperature of

1462- K9 M F) or a thermal shroud outer wall temperature of1517 I K (M rF)

22726 13 11 Change (b) in step 11 to read as follows:

b. Cel assembly cladding temperature increases to

566 + 5 .K (560 +9 F). This criterion effective only after 300v7 6 3.2 Replace Be first s"ntence with the following:

The pretransient phase consists of a reactor shutdown interval(minimum of 2 days) followed by a power operation interval.

8 7 Add the following paragraph before the start of section 3.3:

The length of the shutdown interval shall be dependent on plantoperations tasks to be completed prior to the final power operation,and on project decision on Cs/I ratio.

, I

ORIGINAl

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DOCUMENT REVISION REQUEST(CONTINUATION SHEET)

OPER. NO .

PAGE 4 OFFORM EG&G-o844A

nn•I D' I - 7A"Vrun,, flU - ______-,-___-"

DOCUMENT NO. OF APPLICABLE) DOCUMENT ISSUE DATEOECD LOFT roject Lxperiment Specification Document

OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1945PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER

FOR EACH CHANGE. FORWRITER'S

STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE 'REVISE PER USEITEM I PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND S ATE "PREPARE NEW (SP. DOP. ETC) PER ATTACHED DRAFT".

'1 9 21 Tble Replace the 3 conditions/decisions with the following 2 conditions/decisions:

Condition

Reactivity change

No reactivity change

Decision

Continue experiment sequence

Terminate experiment

10

11

12

-'13

22

2

8

23a

Thle

3.3

23a

Revise first abnormal condition and course of action to read asfollows:

System pressure below criterion If either the Fl or F2 isolatibut either Fl or F2 isolation valves fail to open w4"4h4.valves fail to open terminate the experiment

and commence recovery operatio0 with the ECCS. Do not open

prior to 840 K (1052°F) the LPIS line filter and do(indicated) not store the gamma densitomet

sources.

in

Is

1' Change temperature in item 1 to: 1533 t (2300 0 F)

Change centeg assembly shroud ogter wall temperature from1473 K (2192 F) to 1573 K (2272 F)

Change peripheral fuel rod cladding temperature from

1417 K (2092 F) to 1462 K (2172 F).

Add last two paragraphs:

The last situatign to be considered is if the CFM cladding hasexceeded 2100 K (3321uF) and if the temperature indications on elthethe CFM shroud outer wall or the peripheral fuel assembly cladding Lto fail before the temperature limits are reached. In the case ofthe peripheral fuel assembly cladding, a minimum of two valid temperindications are required to maintain the experiment termination crilin effect. (Two valid indications of the temperature limit willactivate experiment termination.) If the minimum number oftemperature indications is not maintained before the limit is reachEa clock function defining the time remaining to experiment terminatlwill be utilized. The specifics of the operation of the clockfunction are the responsibility of the Operations Branch.

in the hot region

regin

atureerion

d,on

ORIWNAIORIGINAl

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DOCUMENT REVISION REQUEST OPER. NO.(CON INUATION SHEET) PAG __ OF 5

FORM E040-1844A(Rev. I.g NRR nE L-743

DOCUMENT NO. (IF APPLICABLE) OECD LOFT p r3WJ ý11ýT .Jeriment Specification Document DOCUMENT ISSUE DATE

OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN IST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER

FOR EACH CHANGE.FOR

S WRITER'SI STEP OR I NSTRUCTIONS: REWRITE-PARAGRAPH(S1 OR FOR EXTENSIVE CHANGES ATTACH REVISErD COPY AND STATE REVISE PER USE

ITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP. ETC.) PER A TTACHED DRAFrT.

In the case of the CFM shroud outer wall, a minimum of twovalid temperature indications (of the total of 16 indications) arerequired to maintain this experiment termination criterion in effec:.(Two valid indications of the temperature limit will activate exper menttermination.) In the event that 15 of the 16 indications fail, thi,experiment termination criterion is deleted. Experimenttermination reverts to the criterion on the peripheral fuel assemblcladding. If one valid temperature indication exceeds thetemperature limit and supporting information on adjacentperipheral temperatures indicate a shroud heatup, the experimentwill also be terminated.

14 23 last Delete the last paragraph on page 23 (reference ESD Rev. 1)

15 22 1st Delete sentence, "If this break has not been closed ..... should beterminated."

16 22 2nd Revise to read as follows:

The FPMS Fl and F2 sampling systems arS to be opened at a CFMcladding temperature of 811 + 5 K (1000 + 9 F) if the system pressuehas decreased below 1.43 + 0:03 MPa (195-+ 5 psig). The pressurecriterion corresponds to the Fl and F2 design pressure limit of200 psig. The LPIS line filter is to be valved in and the gammadensitometer sources are to be stored following the opening of theFl and F2 sampling 8ystems at a CFM cladding temperature of840 + 5 K (1052 + 9 F). The following courses of action aredefined in the event that abnormal conditions occur:

A7 13 12 Revise to read:Open the FPMS Fl and F2 sampling systems at a center fuel assemblycladding temperature of 811 + 5 K (1000 + 9°F) and open only if thesystem pressure is less than 1.43 XAE)N + 0.03 MPa (195 + 5 psig).If the Fl and F2 sampling systems are opened, open the L-EPIS breakline filter and store the gamma densitometer sources at a center fu lassembly cladding temperature of 840 + 5 K (1052 + 9 F). Thermocou lesto be used in determining the temperature are listed in Table 8.

18 7 3.3 Change censer assembly shroud outer wall temperature from1473 (2192 F) to 1573 K (2272 0 F). Cha g ge the peripheral guel rodcladding temperature from 1417 K (2092 F) to 1462 K (2172 F).

nflIGNAI

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1. INTRODUCTION

OECD LOFT Experiment LF-FP-2 is intended to provide information on the

release and transport of fission products and aerosols in a severe core

damage scenario wherein the fuel rod cladding would exceed a temperature of

2100 K for at least 3 minutes, resulting in rapid metal-water reaction.

Aerosols from the Ag-In-Cd ccntrol rods would provide the principal

environmental constituent for the transport of fission products. This

severe core damage scenario criginates from a risk dominant accident

sequence postulated for comrmercial PWR plants. Within this framework, the

nature of the phenomena governing fission product and aerosol release and

transport can be linked to pctential PWR system thermal hydraulics and core

thermal response leading to fuel failure and fission product transport

behavior.

PRA studies 1 revealed trat the interfacing systems LOCA, a

hypothetical event first postulated in the Reactor Safety Study2 and

labeled the V sequence, has a significant potential contribution to risk

from the operation of nuclear power plants. This risk dominant accident

sequence was selected as the mechanism under which severe core damage

phenomena would be studied in Experiment LP-FP-2. The specific interfacing

systems LOCA associated with the significant operational risk is a pipe

break In the low pressure injection system (LPIS), also called the residual

heat removal system. This system typically serves two functions: (a) it

provides emergency coolant injection for core recovery during intermediate

and large LOCAs, and (b) it provides for decay heat removal during normal

shutdown. The LPIS represents a potential path by which a LOCA outside

containment could occur, discharging primary system coolant external to the

containment. If core cooling cannot be maintained during such an event,

fission product release to the environment could occur through failure of

the auxiliary building.

Experiment LP-FP-2 will simulate the system thermal-hydraulic and core

uncovery conditions that are calculated to occur in commercial PWRs from

rupture of an LPIS pipe as a result of a V-sequence. The facility

I

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DRR-L-7435Rev. I Chg. 3

configuration will include simulation of typical commercial PWR LPIS piping

in an experimental configuration in addition to the current LOFT LPIS. The

experimental configuration will be connected to the blowdown suppression

system.

Experiment LP-FP-2, the last of a series of eight experiments in the

OECD LOFT Program, may cause significant contamination of the Primary

Coolant System (PCS) and blowdown suppression system. The requirements

imposed on experiment LP-FP-2 from the standpoint of facility

decontamination and recovery are:

1. Experiment LP-FP-2 must be conducted with peripheral assembly

fuel rod cladding temperature limited to 1533 K (2300°F).

2. The structural integrity of the center fuel assembly must be

maintained to facilitate removal from the reactor vessel.

To meet the above project requirements, a center fuel assembly

specifically for the LP-FP-2 experiment has been designed with higher fuel

enrichment than the peripheral assemblies, and with prepressurized fuel

rods. The center assembly design also includes considerations of:

(a) providing a fuel rod-to-control rod (Ag-In-Cd) mass ratio similar to

that of a commercial PWR fuel assembly with control rods; and (b) providing

a well-defined geometry for the release and transport of fission products

and aerosols to facilitate interpretation of the results and code

assessment studies; and (c) thermally insulating the peripheral fuel

assemblies from the center fuel assembly.

2

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DRR-L-7437Rev. 1 Chg. 2

3. EXPERIMENT DEs:.IPTION

Experieret LP-FP-2 consists of four discinct phases. These phases are

designated as (a) fuel preconditioning, (b) Dretransient, (c) transient,

and (d) postt-ansientl The four phases tocether represent a continuous

process and rave specific beginning and enc'ng definitions. Each phase is

described in the following sections.

3.1 Preconditioninc Phase

The purpose of the preconditioning phase is to subject the fuel rods

in the new center assembly [9.72 wt*, enriched, 2.41 MPa (350 psia)

prepressurization] to a burnup which, in co-bination with the burnup

corresponding to the EFPH requirement at the experiment initial conditions,

provides the required minimum burnup for the experiment. The minimum

burnup required for these fuel rods is 325 MYD/MTU. The minimum burnup

required in the preconditioning phase is 252 MWD/MTU. In terms of reactor

operation, the preconditioning burnup is equivalent, as an example, to

power operation at a maximum linear heat generation rate (MLHGR) of

52.2 kW/m (16 kW/ft) for 111.5 hours on the 9.72 wt% enriched fuel rods* A

core power level of 32.0 MW ± 0.5 is calculated to provide a MLHGR of

-16 kW/ft on the 9.72 wt% enriched fuel rods. Power profile data using

the traversing incore probe system (TIPS) must be obtained during the

preconditioning phase. The TIPS locations in the center assembly will be

capped during the reactor shutdown time in the pretransient phase.

The beginning and end of the preconditioning phase are defined as

follows:

Beginning the start of plant heatup prior to power operation to

establish fuel burnup

End termination of power operation after the required

burnup in this phase has been achieved.* Any combination of preconditioning and pretransient power operation that provides

the specified initial conditions (Section 4.2) is permissable. The specificcombination described is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation .

5

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DRR-L-7435Rev. 1 Chg. 3

The preconditioning phase does not require special procedures relative

to experiment specification and is to be conducted with established plant

operating procedures.

3.2 Pretransient Phase

The pretransient phase consists of a reactor shutdown interval

(minimum of 2 days) followed by a power operation interval. The final

plant preparations are to be completed during the reactor shutdown

interval. The power operation interval is to complete the required central

assembly fuel burnup, establish the required minimum decay heat level, and

establish the required initial conditions for conduct of the experiment.

The beginning and end of the pretransient phase are defined as follows:

Beginning termination of the power operation after the required

burnup is achieved in the preconditioning phase

End initiation of the transient by a reactor scram.

Specifically, the final plant preparations required by this

specification are:

1. Operational readiness of thermal hydraulic and nuclear

measurements (experimental and process) required by this

specification for this experiment and approval by the LOFT

Operations Branch to proceed into power operation and

subsequently into the transient phase of the experiment.

2. Operational readiness of the fission product measurement system

(FPMS) in accordance with FPMS operational documents.

The power operation prior to experiment initiation is to be an

effective 40 hours at a core power level of 26.5 ± 0.5 MW (21.2 EFPH*).

This core power is the required initial condition value. Power operation

*EFPH is defined as 50 MW-hr.

6

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DRR-L-7435Rev. 1 Chg. 3

of an effective 40 hours will establish the required minimum decay heat

level of 675 kW at 200 s and will complete the burnup required on the

9.72 wt% enriched fuel rods. The plant should be operated at a steady

state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment

initiation.

The length of the shutdown interval shall be dependent on plant

operations tasks to be completed prior to the final power operation, and on

project decision on Cs/I ratio.

3.3 Transient Phase

The transient phase of the experiment will be initiated by a reactor

shutdown as specified in Section 4.3. Time zero corresponds to the action

taken to drop the center fuel module (CFM) controls rods. The 1.16 in.

inner diameter simulated break in the intact loop cold leg will be opened

at 20 ± 2 s. The primary coolant pumps will be tripped at 25 ± 2 s and

will undergo a normal coastdown. At 220 ± 5 s the simulated LPIS line

(also 1.16 in. diameter) will be opened. (The LPIS line filter shall be

bypassed to prevent plugging prior to the fission product release.) This

line is connected to the broken loop hot leg and to a blowdown suppression

tank (BST) inlet vent. The intact loop cold leg simulated break will be

closed at either a CFM cladding temperature of 566 ± 5 K (560 ± 91F) or a

system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is

calculated to begin heatup at approximately 700 s and to reach 2100 K at

approximately 1700 s. Fission product and aerosol release will occur from

fuel rod failure until the center assembly shroud outer wall reaches 1573 K

(22721F) or the peripheral fuel rod cladding reaches 1462 K (21721F), at

which time the fission product filter sampling systems and the upper plenum

deposition coupons will be closed in the primary coolant system coincident

with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these

actions core reflood will commence with both accumulators.

The beginning and end of the transient phase of the experiment are

defined as follows:

7

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DRR-L-7423

Rev. I Chg. 1

Beginning initiation of the transiert by a reactor scram

End initiation of the closure of the simulated LPIS line in

the broken loop hot leg.

3.4 Posttransient Phase

The posttransient phase consists of a tire interval of 12 hr for

measurement of (a) the redistribution of fission product inventory in the

gas and liquid volumes in the blowdown suppression tank (BST), and (b) the

leaching of fission products from the damaged fuel rods in the primary

coolant system (PCS). The beginning and end of the posttransient phase of

the experiment are defined as follows:

Beginning initiation of the closure of the simulated LPIS line in

the broken loop hot leg

End completion of the time interval specified for fission

product measurement).

Following closure initiation of the simulated LPIS line in the primary

coolant system, reflood operations will commence with initiation of both

accumulators. System refill will continue as required with the high

pressure injection system (HPIS). The governing requirements for plant

operation involving the PCS in this phase are:

1. Mass transfer from the PCS is to be minimized. The purification

system can be used in the decay heat removal mode (bypassing the

ion exchanger) for temperature control along with steam generator

feed and bleed.

2. Forced coolant circulation with the primary coolant pumps is

prohibited.

3. PCS temperature is to be reduced and maintained below 449 K

(350*F) as soon as possible after simulated LPIS line closure.

8

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4. PCS pressure is to be maintained below E.96 MPa (1300 psig).

PORV venting is permitted but must be mirmized. The preferred

operation, if the PORV is needed, is to c::n it once for a long

enough time to bring the pressure down. This should eliminate

multiple openings which could effect the £ST fission product

measurements.

The degree of system refill with HPIS is not s:ecified by this

document but is to be utilized as required by plant operations to meet the

above four governing requirements.

Following closure of the simulated LPIS line, :he spray systems in the

BST vapor volume are not to be actuated. Mass transfer to and from the BST

is prohibited in the posttransient phase subject to the occurrence of PCS

conditions requiring operator or automatic plant actions. Measurements of

the fission product concentration in the gas and li:uid volumes, together

with temperature and pressure boundary conditions, will be used to

determine the fission product gas/liquid partition coefficients and fission

product redistribution and settling phenomena.

9

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DRR-L-7435Rev. 1 Chg. 3

4. EXPERIMENT REQUIREMENTS

This section specifies experiment requirements for OECD LOFT

Experiment LP-FP-2. Included in this section are requirements for

experiment prerequisites, initial conditions, operational sequence, plant

configuration, and experimental measurements for the transient and

posttransient phases of the experiment.

4.1 Experiment Prerequisites

The following prerequisites are of programmatic concern and must be

completed prior to initiating Experiment LP-FP-2.

1. Perform a one point end-to-end check of the process instruments

identified In Table 1, within 90 days of the experiment. If a

problem is indicated, recalibrate, repair or replace the

instrument.

2. Establishment of event time procedures and parameter setpoints

for the items listed in Table 2.

3. Complete the pre-experiment calibration requirements for the FPMS

instruments and the pre-experiment calibration requirements

specified in DOP-87-O05, "DAVDS Experimental Measurements and

Test Procedure," and in DOP-87-008, "Pre LOCE Data Verification."

4. Complete testing of the center fuel assembly control rod

insertion function.

4.2 Initial Conditions

Experiment LP-FP-2 initial conditions will approximate commercial PWR

pressures and temperatures, and will be consistent with the LOFT plant

safety analysis. Specifically, three initial conditions are required for

the primary coolant system. These are intact loop hot leg temperature,

571 ± 1.1 K (569 ± 20 F), decay heat level between 675 kW and 695 kW at

10

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DRR-L-7435Rev. 1 Chg. 3

b. Center fuel assembly cladding temperature increases to

566 ± 5K (560 ± 90 F). This criterion effective only after

300 s.

The cladding temperature following time zero will decrease and

correspond to saturation temperature until core uncovery occurs.

Then, cladding heatup will commence at some temperature below the

value specified in (b). Therefore, cladding temperature must be

increasing in order for criterion (b) to be valid.

12. Open the FPMS Fl and F2 sampling systems at a center fuel assembly

cladding temperature of 811 + 5 K (1000 + 9°F) and open only

if the system pressure is less than 1.43 + 0.03 MPa (195 + 5 psig).

If the Fl and F2 sampling systems are opened, open the LPIS

break line filter and store the gamma densitometer sources at a center

fuel assembly cladding temperature of 840 -I- 5 K (1052 + 90 F). Thermocouples

to be used in determinig the temperature are listed in Table 8.

13. Close the FPMS sampling system lines, the deposition coupon

device, and the simulated LPIS line in the broken loop hot leg at

a peripheral fuel assembly cladding temperature of 1462 K

( 2172'F) or a thermal shroud outer wall temperature of 1517 K

(2272 0 F).

14. Initiate core reflood 6 ± 0.5 s after initiation of the system

closures in Step 13.

15. Continue system refill with the IIPIS and maintain the PCS

temperature below 449 K (350'F) as soon as possible for the

remainder of a 12 hr minimum time after initiation of core

reflood. Maintain the PCS pressure below 8.96 MPa (1300 psig).

Mass transfer to and from the PCS, excluding system leakage and

replacement, is to be minimized. PCS energy control is to be

accomplished with steam generator feed/bleed operations and/or

with the purification system in the decay heat removal mode (ion

exchanger bypassed). PORV venting is permitted but is to be

minimized to the extent possible.

13

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with the purification system in the decay heat removal mode (ion

exchanger bypassed). PORV venting is permitted but is to be

minimized to the extent possible.

16. Maintain isolation of the BST (no mass trans'er into or out of

the BST) for the same time interval as in Ite- 15 subject to

plant limitin; conditions which may require cerator action.

Within 10 minutes of PCS isolation purge the SST downcomer with

60 SCF of N2.I

17. Maintain PLSS data acquisition continuously ever the time

interval for Items 15 and 16.

4.4 System Configuration

The general syster and component configuration of the LOFT PWR for

Experiment LP-FP-2 is shown in Figure 1. Specific details are given in the

following sections on the reactor core, primary coolant system, secondary

coolant system, blowdowr system, and emergency core coclant system.

4.4.1 Reactor Core

Experiment LP-FP-2 will be conducted with a specially constructed

center fuel assembly.. The cross section of this fuel assembly is shown in

Figure 2. The fuel rods are 350 psia prepressurized ard 9.72 wt%

enriched. The outer two rows of fuel rods have been replaced with a

thermal shield of zircaloy with zirconium oxide ceramic internal

insulation. The purpose of the higher than normal enrichment and the

thermal shield is to provide at least 3 minutes at peak cladding

temperatures above 2100 K (3321*F) in the center fuel assembly before the

peripheral assembly fuel cladding reaches the transient termination

temperature. This time at temperature will provide a sufficiently large

fission product and aerosol release fraction.

The aerosol release will occur from (Ag-In-Cd) control rods which will

be inserted at reactor scram in the guide tubes shown in Figure 2.

14 I

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DRR-L-7435Rev. 1 Chg. 3

inserted the experiment sequence must be stopped. On completion of the

corrective actions the experiment is to be resumed as approved by the LOFT

Operations Manager, DOE Site Program Manager and LOFT Program Division

Manager.

The decision matrix on center assembly control rod insertion is as

follows:

Condition Decision

Reactivity change Continue experiment sequence

No reactivity change Terminate experiment

Sequence Step 8 opens the break path in the intact loop cold leg.

This flow path is the primary blowdown path and is intended to be the path

for venting the major part of the primary coolant system fluid. High

quality steam flow only is desired for venting through the simulated LPIS

pipe in the broken loop hot leg. If the intact loop cold leg break cannot

be opened, the experiment sequence must be halted. On completion of the

corrective actions the experiment is to be resumed as approved by the LOFT

Operations Manager, DOE Site Program Manager and LOFT Program Division

Manager.

The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.

This flow path is the designed fission product and aerosol vent path to the

BST. If this line cannot be opened within 50 s of the time specified,

close the intact loop cold leg break and commence recovery operations with

the ECCS. Resume the experiment as approved by the LOFT Operations

Manager, DOE Site Program Manager and LOFT Program Division Manager.

21

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DRR-L-7435Rev. 1 Chg. 3

The intact loop cold leg break is to bp closed on either a cladding

temperature value or a system pressure value as specified in Sequence

Step 11. If this break is not closed then another flow path to the BST

will exist for fission product and aerosol venting. Two vent paths are not

provided for in the experiment plan. If this break cannot be closed before

the CFM cladding reaches 840 + 5 K (1052 + 9°F) then terminate the

experiment and commence recovery operations with the ECCS. If experiment

termination and plant recovery operations commence with the ECCS, return to

Sequence Step 1 or as approved by the LOF1 Operations Manager, DOE Site

Program Manager, and LOFT Program Division Manager after repairs are made.

The FPMS Fl and F2 sampling systems are to be opened at a CFM cladding

temperature of 811 + 5 K (1000+ 9gF) if the system pressure has decreased

below 1.43 + 0.03 MPa (195 + 5 psig). ihe pressure criterion corresponds

to the Fl and F2 design pressure limit of 200 psig. The LPIS line filter

is to be valved in and the gamma densitometer sources are to be stored

following the opening of the Fl and F2 sampling systems at a CFM cladding

temperature of 840 + 5 K (1052 + 9OF). The following courses of action

are defined in the event that abnormal conditions occur:

22

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DRR-L-7435Rev. I Chg. 3

Abnormal Condition

System pressure below criterionbut either F1 or F2 isolationvalves fall to open

Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.

All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.

Course of. Action

If either the Fl or F2 isolation galvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand commence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.

Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.

The experiment is to proceed.Continue attempts to completethese actions.

If the condition occurs where the system pressure is above the maximum

allowable for F1 and F2 operation when the CFM cladding temperature is

800 K (981 0 F), commence actions to lower the system pressure. These

•, actions may be, but not limited to, opening the PORV or injecting full HPIS

for 10-20 s durations. Terminate all of these actions before the CFM

cladding temperature reaches 1050 K (14311F). The experiment is to

continue in the event that the system pressure cannot be lowered below the

F1 and F2 operating pressure limit. If the actions taken to lower the

pressure are successful, terminate all actions before the system pressure

decreases below an indicated value of 1.2 MPa (160 psig).

The LPIS break line filter is to be valved in when the peak cladding

temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the

differential pressure may increase. At a filter differential pressure of

150 psi, the filter should be bypassed to maintain the ability to measure a

sample in the aerosol collection lines. If this occurs, the time shall be

recorded.

23

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DRR-L-7435Rev. 1 Chg. 3

The FPMS sampling lines are closed, the deposition coupon devices are

closed, and the simulated LPIS line is closed in Sequence Step 13. If any

of these actions cannot be completed the experiment sequence is to

continue. Core reflood must commence in time to prevent temperatures from

exceeding 1588 K (2400 0 F) on the center fuel assembly shroud outer wall

and 1533 K (2300 0 F) on the peripheral assembly fuel rod cladding.

Continue attempts to complete Sequence Step 13 but continue through

Sequence Steps 14-17.

The possibility exists that the time interval between the occurrence

of 2100 K (3321 0 F) cladding temperature in the center assembly and the

peripheral assembly cladding temperature or the shroud outer wall

temperature limit may be excessive relative to core damage and source

release limitations. Therefore, a maximum time of 4.0 minutes is

specified, after the 2100 K (3321 0 F) temperature is reached, for

continuing the experiment sequence with Steps 13 and 14.

The last situation to be considered is if the CFM cladding has

exceeded 2100 K (3321 0 F) and if the temperature indications on either

the; CFM shroud outer wall or the peripheral fuel assembly cladding begin

to fail before the temperature limits are reached. In the case of

the peripheral fuel assembly cladding, a minimum of two valid temperature

indications in the hot region are required to maintain the experiment termination

criterion in effect. (Two valid indications ot the temperature limit will

activate experiment termination.) If the minimum number of temperature

indications is not maintained before the limit is reached, a clock function

defining the time remaining to experiment termination will be utilized. The

specifics of the operation of the clock function are the responsibility of

the Operations Branch.

23A

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DRR-L-7435Rev. I Chg. 3

In the case of the CFM shroud outer wall, a minimum of two valid

temperature indications (of the total of 16 indications) are required to

maintain this experiment termination criterion in effect. (Two valid indications

of the temperature limit will activate experiment termination.) In the event

that 15 of the 16 indications fail, this experiment termination criterion is

deleted. Experiment termination reverts to the criterion on the peripheral

fuel assembly cladding. If one valid temperature indication exceeds the

temperature limit and supporting information on adjacent peri-pheral temperatures

indicate a shroud heatup, the experiment will also be terminated.

23B

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5. PLANNING ANALYSIS

This section discusses scaling considerations for the LOFT system for

Experiment LP-FP-2 compared with commercial PWRs, and pressents results of

computer code calculations on which this specification do:.-ent is based.

5.1 Scaling Consideration

Scaling considerations for Experiment LP-FP-2 center c the V-sequence

phenomena that are calculated to occur in commercial PWRs. The V-sequence

scenarios include little or no operation of the PWR ECCS. The LOFT

V-sequence experiment does not include ECCS operation. Therefore, the

specifications and/or recommendations for the operation of the LOFT ECCS

contained in Section 4 are based on plant recovery plannin: specific to

LOFT. The required scaling considerations for Experiment -P-FP-2 include

only the sizing of the break path flow areas (for both break paths for the

experiment) and also sizing the length of the simulated LF:S flow path.

Break area scaling provides representative thermal-hydraulics and, in

particular, similar coolant velocities that are required f:r the transport

of fission products and aerosols. Simulated LPIS pipe ler:th scaling is

necessary to provide similar residence times for transport and retention

phenomena in the LPIS piping.

PWR LPIS pipe sizes range from 6 in. Sch 160 to 10 in. Sch 160. The

inside diameter range is 0.13-0.22 m (5.19-8.5 in.). The s:aling basis

used to size the break area in LOFT is break flow area/sys:em volume.

Using the following values for system volume:

V (PWR) = 355.10 m3 (12,540 ft 3 ) (Reference 6)

V (LOFT) = 7.36 m3 (260 ft 3) (Reference 7)

gives the LPIS pipe diameter range in LOFT of 0.019 - 0.03: m

(0.747 - 1.224 in.). The pipe size selected for the LPIS ripe simulation

in LOFT is 1-1/4 in. nominal Sch 160 which has an inside diameter of

0.0295 m (1.16 in.). The scaled flow area equivalent in a commercial PWR

is 10%1 larger than an 8 in. Sch 160 pipe and 13% less than a 10 in. Sch 160

24 I

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DRR-L-7435Rev. 1 Chg. 3

TABLE 2. EXPERIMENT LP-FP-2 EVENT TIMES AND PARAMETER SETPOINTS

Item Action Time/Setpoint

Reactor shutdown CFM CR t = 0dropPeripheral 5 sCR scram

Intact loop cold leg simulated break Open 20 ± 2 s

Primary coolant pumps Tripped 25 ± 2 s

Broken loop hot leg simulated LPIS Open 220 ± 5 spipe

Intact loop cold leg simulated break Close 566 ± 5 K (560 ± 90 F)1.2 ± 0.03 MPa(160 ± 5 psig)

FPMS sampling system isolation Open 811 K ± 5 Kvalves (I1000OF ± 90 F)

LPIS Break Line Filter Open 840 K ± 5 K(10520 F ± 90 F)

Gamma densitometer sources Store 840 K ± 5 K(10520 F ± 90 F)

FPMS PCS sampling system isolation Close 1462 K (2172 0 F) onvalves peripheral. fuel assembly

cladding or 1517 K(2272°F) on CFM shroud

Broken loop hot leg simulated LPIS Close outer wall or other experimentpipe termination events.

Deposition coupon device Close

Core reflood Initiation 6 ± 0.5 s from simulatedLPIS line closureinitiation

71

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TABLE 3. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2

Parameter

Primary Coolant System

Core AT

Hot leg pressure (MPa). (psia)

Hot leg temperature (K)(OF)

Primary coolant flow (Ibm/hr)(kg/s)

Boron concentration (ppm)

Power level (MW)

Decay heat at 200 s minimummaximum

Maximum linear heat

generation ratec (kW/m)(kW/ft)

Control rod position(above full-in position) (m)

(in.)

Steam Generator Secondary Side

Water level

Pressurizer

Liquid level (m)(in.)

Specified Value a

As requiredb

14.95 ± 0.12168 ± 15.0

571 ± 1.1569 ± 2

3.8 ± 0.15 x 106

479 ± 19

As requiredb

26.5 _ 0.5

675 kW695 kW

-40-12.26

1.37 _ 0.0154.0 ± 2.0

As requiredb

1.12± 0.144 ± 4

I

72 I

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(® OPER. NO.

rOflM EG&G-I844 DOCUMENT REVISION REOUEST

(D REQUESTSA(® OTR NO.V. T. BErta July 2, 1985 L-7448

05 DOCUMENT NO. (IF APPCABLE) OECD LOFT PrMWRIT EJerimen t Specification DocumLPIfMENT ISSUE DATE

OECD LOFT-T-3802 Rev. 1 Fisslon Product Experiment LP-FP-2 May 1985

@ CHECK APPLICABLE BLANK ______® 7 GER.PR0 I DATE

PERMANENT CHANGE X TEMPORARY CHANGE - BULLETIN 4%/~(®) PRINT OR TYPE PROPOSED CHANGE - NUMBE EACH CHANGE SEQUENTIALLY IN IST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER FOR

FOR EACH CHANGE. WRTER'S USE

STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER

ITEM PAGE PAR. ATTACHED COPY. FOR NEW DOCUMENT. ATTACH ROUGHDRAFT ANDSTATE"PREPARE NEW (SP..DOP. ETC.) PERATTACHEDORAFT".

1 23A 2nd Change 4.0 minutes to 7.0 minutes in the last sentence.

2 79 Tbl E Change the wording at the bottom of the list of thermocouples to th(following:

Any valid thermocouple may be used. All of the listedthermocouples are to be monitored. If any of the listed'thermocouples fail prior to the experiment, they may be replacedwith similar thermocouples. If replacement thermocouples arenot available, the failed thermocouples may be deleted fromthe monitors.

USE CONTINUATION SHEET AS REQUIRED

NEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATEIEVENT

(.JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVE): (9 OTHER DOCUMENTATION AFFECTED:

1. Increase detectability of fission products at all IDOC.N. DARNO. DATECOMPLETED

measurement locations. OECD LOFT-1I-I1-510 675785

2. Allows for flexibility in the event of instrumentfailures.

"1__ ORIGINATING ORR NO:

@ REVIEW

NAMEISIGNATURE ORG. DATE NAMEISIGNATURE ORG. DATE NAMEISIGNATURE ORG. DATE

24 k/ I 0v QUALITY

PIIAC(I IcbMMENTS: ADDITIONAL

DARS IN THISDOCUMENT

REVISION

(J3 DOCUMENT CONTROLLER Qq RELEADE PCTFR OLLER 8t) ORR COMPLETED DATE:

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DRR-L-7435Rev. I Chg. 3

Abnormal Condition

System pressure below criterionbut either F1 or F2 isolationvalves fail to open

Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.

All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.

Course of Action

If either the Fl'.or F2 isolation galvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand coninence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.

Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.

The experiment is to proceed.Continue attempts to completethese actions.

If the condition occurs where the system pressure is above the maximum

allowable for F1 and F2 operation when the CFM cladding temperature is

800 K (981'F), commence actions to lower the system pressure. These

actions may be, but not limited to, opening the PORV or injecting full HPIS

for 10-20 s durations. Terminate all of these actions before the CFM

cladding temperature reaches 1050 K (1431*F). The experiment is to

continue in the event that the system pressure cannot be lowered below the

F1 and F2 operating pressure limit. If the actions taken to lower the

pressure are successful, terminate all actions before the system pressure

decreases below an indicated value of 1.2 MPa (160 psig).

The LPIS break line filter is to bevalved in when the peak cladding

temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the

differential pressure may increase. At a filter differential pressure of

150 psi, the filter should be bypassed to maintain the ability to measure a

sample in the aerosol collection lines. If this occurs, the time shall be

recorded.

23

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DRR-L-7448Rev. I Chg. 4

The FPMS sampling lines are closed, the deposition coupon devices are

closed, and the simulated LPIS line is closed in Sequence Step 13. If any

of these actions cannot be completed the experiment sequence is to

continue. Core reflood must commence in time to prevent temperatures from

exceeding 1588 K (2400 0 F) on the center fuel assembly shroud outer wall

and 1533 K (2300 0 F) on the peripheral assembly fuel roa cladding.

Continue attempts to complete SeQuence Step 13 but continue through

Sequence Steps 14-17.

The possibility exists that the time interval between the occurrence

of 2100 K (3321 0 F) cladding temperature in the center assembly and the

peripheral assembly cladding temperature or the shroud outer wall

temperature limit may be excessive relative to core damage and source

release limitations. Therefore, a maximum time of 7.0 minutes is

specified, after the 2100 K (3321 0 F) temperature is reached, for

continuing the experiment sequence with Steps 13 and 14.

The last situation to be considered is if the CFM claddinc has

exceeded 2100 K (3321OF) and if the temperature indications on either

the CFM shroud outer wall or the peripheral fuel assembly cladding begin

to fail before the temperature limits are reached. In the case of

the peripheral fuel assembly cladding, a minimum of two valid temperature

indications in the hot region are required to maintain the experiment termination

criterion in effect. (Two valid indications ot the temperature limit will

activate experiment termination.) If the minimum number of temperature

indications is not maintained before the limit is reached, a clock function

defining the time remaining to experiment termination will be utilized. The

specifics of the operation of the clock function are the responsibility of

the Operations Branch.

23A

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DRR-L-7448Rev. I Chg. 4

TABLE 8. LP-FP-2 CENTER MODULE THERMOCOUPLES FOR FPMS AND LPIS LINEFILTER - OPENING

At a peak cladding temperature of 840 K (10521F) open the valves to the F-1

and F-2 sample lines and to the F-3 filter in the simulated LPIS line.

The thermocouples to be used in this decision are present below:

TE-5C7-42TE-5C12-27TE-5D9-27TE-5DI3-42TE-5F9-27TE-5G4-27TE-5G12-27TE-5G13-27TE-5H6-27TE-5H10-27TE-5103-27TE-5104-42TE-5112-42TE-5J7-27TE-5J9-42TE-5L7-10TE-5L7-27TE-5L9-42TE-5M7-27TE-5M9-42

Any valid thermocouple may be used. All of the listed thermocouples areto be monitored. If any of the listed thermocouples fail prior to theexperiment, they may be replaced with similar thermocouples. If replace-ment thermocouples are not available, the failed thermocouples may bedeleted from the monitors.

79

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ropm EG&G.1-44Jltev. •0%6 DOCUMENT REVISION REQUEST

( OPER. NO. ____

PAGE L OF 3,(® REQUESTER (I ® RR DATE (•) DOR ND.V. T. Berta July 5, 1985 L-7450

®. DOCUMENT NO. (IF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE pATE

OECD LOFT-T-3802 Rev. 1 OECD LOFT Project Experiment SpecificationDocument Fission Product.ExpjerimientjLP-FP-2 , May 1985

@ CHECK APPLICABLE BLANK . M APP A LV DATE

PERMANENT CHANGE X TEMPORARY CHANGE -___ BULLETIN____PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER () FOR

FOR EACH CHANGE. WRTTER'S USE

STEP OR INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE REVISE PERITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP, ETC ) PER ATTACHED DRAFT".

1 7 Add the following sentence to the wording at the top of the page:

Refer to Section 4.2 for further initial conditionspecifications.

2 7 3.3 Insert the following in place of the 2nd, 3rd, and 4th sentences:

Time zero corresponds to the reactor scram action. Theprimary coolant pumps are tripped at 8 ± 2 s. The centerfuel module (CFM) control rods are dropped whehithe'loop flow hisThe intact loop cold leg simulated break is opened Idecreased toimmediately after verification that the CFM control 11.5.x 106 ibm/hr.rods are in. This action is estimated to occur atapproximately 20-25 s.

Delete the following in the next sentence:

(also 1.16 in. diameter)USE CONTINUATION SHEET AS REQUIRED

NEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT 7//5/L85

(:0) JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVEI: (5 OTHER DOCUMENTATION AFFECTED:

All items 'reflect sequence Step 7, 8, 9 revisions to DOC. NO. DRR NO. DATE COMPLETED

account for CFM control rod drop equipment character- 6EEDiFi:4-ULJ-51L 6/5/85istics. New sequence steps are 7, 8, 9A, and 9B.

-_ 2 ORIGINATING DRR NO:.

(03 REVIEW

NAME/SIGNATURE ORG. DATE NAME/SIGNATURE ORG DATE NAME/SIGNATURE ORG. DATE

~~~j~ 6I V'~ 1 I2-IZ MLI4Ž Q__ _ __ _ _ UALITY17 SHS

Q)) dOMMENTS: f15 ADDITIONALIRRS IN THIS

DOCUMENTREVISION

6_ý DOCUMENT CONTROLLER

0-b L(ý.

D C RELEASE DATE J -I PnR CMPLETED DATE

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DOCUMENT REVISION REQUEST OPER NO,

(CONTINUATION SHEET) PAGE OF

FORM EG&G-8I4AA. 1R NO _L-7450

DOCUMENT NO. (IF APPLICABLE) DOCMEDOCUMENT ISSUE DATE

OECD LOFT-T-3802 Rev. I OECD LOFT Project lxperiment Specification DocumentFission Product Experiment LP-FP-2 May 1985

PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEGUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND SFP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR

. . . ...... ... _WRITER'S

ITEM PAGE STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGfe ATTACH REVISED COPY AND STATE 'REVISE PER USEPARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND SIATE PREPARE NEW (SP.DOP. ETC.I PER A1IACHEDDRAFT,.

3 11 1 Revise the last sentence and add one sentence to the last paragraph

of Section 4.2 as follows:

Power operation in the pretransient phase will establish theinitial conditions specified in Table 3 and the requiredminimum burnup on the center assembly fuel rods prior totransient initiation. In the event that reactor shutdownintervals occur after the initial conditions and fuelburnup are reached, subsequent power operation will bebased on establishing only the initial conditionsspecified in Table 3.

4 12 7,8,9 Replace Sequence Steps 7,8, and 9 with the following Steps 7, 8,9A and 9B:

7. Scram the reactor. Reactor scram does not includedropping the CFM control rods (refer to Step 9A).Reactor scram establishes time zero.

8. Trip the primary coolant pumps at 8 t 2 s. The pumpsare to undergo a normal coastdown.

when the loop flow has decreased lo

9A. Drop the CFM control rodsAIwI . Verify that 1.5 x 106 bm/hy

the CFM control rods are in before proceeding toStep 9B.

9B. Open the simulated break in the intact loop cold legimmediately after verification of CFM control rods in.

This action is estimated to occur at approximately20-25 s.

5 20 last Revise 2nd paragraph of Section 4.7 to read as follows:

The CFM control rods are to be inserted in Sequence Step 9A.

Verification of the insertion of these control rods is to beobtained from proximity switch indications before proceedingto Step 9B. If insertion of the CFM control rods cannot beverified the experiment may be either (1) placed on hold for

further decisions or actions to drop the CFM control rods,or (2) terminated. If a hold is placed on the experiment,the experiment sequence must be resumed before the decayheat loss jeopardizes the experiment. If too much decay heat

is lost during the hold, the experiment must be terminated.The minimum acceptable decay heat is 641 kW at time zeroplus the hold interval plus 200 s. After termination, oncompletion of the corrective actions the experiment is tobe resumed as approved by the LOFT Operations Manager, DOESite Program Manager and LOFT Program Division Manager.

ORIGINAt

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DOCUMENT REVISION REQUEST(CONTINUATION SHEET)

OPER NO. -

PAGE 3 OF .

FORM EG&G-1844AfRAv. It-.7 I n• f' I -74!fl•

L-7450V- ... - -

DOCUMENT NO. (IF APPUCABLE) Do1UMETJITLE DOCUMENT ISSUE DATEOECD LOFT Project Experiment Specification Document

)ECD LOFT-T-3802 Rev. 1 Fission Product Exoeriment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOIJENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER

FOR EACH CHANGE. FOR

WRITERSSTEP OR INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHANGF-, ATTACH REVISED COPY AND STATE 'REVISE PER USE

ITEM PAGE PARA. ATTACHED COPY" FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND S I A IE 'PREPARE NEW (SP, DOP, ETC.) PER A TTACHED DRAFT".

6

7

8

21

21

71

1st

2nd

Tabl e2

Delete the paragraph on the decision matrix.

Change Sequence Step 8 to Sequence Step 9B in the first sentence.

Replace the first three items with the following four items:

Item Action Time/Setpoint

Reactor scram Peripheral 0CR scram

Primary coolant pumps Tripped t 2 s

CFM control rods Dropped ' t t l oop flow1.5 x 106 lbm/hr

Intact loop cold leg Open Verification of

0 f

simulated break CFM control rods ir

UNIGINA1

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DRR-L-7450Rev. 1 Chg. 5

of an effective 40 hours will establish the required minimum decay heat

level of 675 kW at 200 s and will complete the burnup required on the

9.72 wt% enriched fuel rods. The plant should be operated at a steady

state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment

initiation. Refer to Section 4.2 for further initial condition specifications.

The length of the shutdown interval shall be dependent on plant

operations tasks to be completed prior to the final power operation, and on

project decision on Cs/I ratio.

3.3 Transient Phase

The transient phase of the experiment will be initiated by a reactor

shutdown as specified in Section 4.3. Time zero corresponds to the reactor

scram action. The primary coolant pumps are tripped at 8 t 2s. The

center fuel module (CFM) control rods are dropped when the loop

flow has decreased to 1.5 x 106 Ibm/hr. The intact loop

cold leg timulated break is opened immediately after veri-

fication that the CFM control rods are in. This action is estimated to

occur at approximately 20-25 s. At 220 ± 5 s the simulated LPIS line

will be opened. (The LPIS line filter shall be bypassed to prevent

plugging prior to the fission product release.) This line is connected

to the broken loop hot leg and to a blowdown suppression tank (BST)

inlet vent. The intact loop cold leg simulated break will be closed

at either a CFM cladding temperature of 566 ± 5 K (560 ± 90 F) or a

system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is

calculated to begin heatup at approximately 700 s and to reach 2100 K at

approximately 1700 s. Fission product and aerosol release will occur from

fuel rod failure until the center assembly shroud outer wall reaches 1573 K

(2272*F) or the peripheral fuel rod cladding reaches 1462K (2172'F), at

which time the fission product filter sampling systems and the upper plenum

deposition coupons will be closed in the primary coolant system coincident

with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these

actions core reflood will commence with both accumulators.

The beginning and end of the transient phase of the experiment are

defined as follows:

7

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DRR-L-7423

Rev. 1 Chg.

Beginning initiation of the transient by a reactor scram

End initiation of the closure of the simulated LPIS line in

the broken loop hot leg.

3.4 Posttransient Phase

The posttransient phase consists of a time interval of 12 hr for

measurement of (a) the redistribution of fission product inventory in the

gas and liquid volumes in the blowdown suppression tank (BST), and (b) the

leaching of fission products from the damaged fuel rods in the primary

coolant system (PCS). The beginning and end of the posttransient phase of

the experiment are defined as follows:

Beginning initiation of the closure of the simulated LPIS line in

the broken loop hot leg

End completion of the time interval specified for fission

product measurements.

Following closure initiation of the simulated LPIS line in the primary

coolant system, reflood operations will commence with initiation of both

accumulators. System refill will continue as required with the high

pressure injection system (HPIS). The governing requirements for plant

operation involving the PCS in this phase are:

1. Mass transfer from the PCS is to be minimized. The purification

system can be used in the decay heat removal mode (bypassing the

ion exchanger) for temperature control along with steam generator

feed and bleed.

2. Forced coolant circulation with the primary coolant pumps is

prohibited.

3. PCS temperature is to be reduced and maintained below 449 K

(350 0 F) as soon as possible after simulated LPIS line closure.

B

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DRR L-7450Rev. 1 Chg. 5

200 s, and maximum linear heat generation rate, 40.0 kW/m (12.26 kW/ft), in

the high power fuel rods. The latter initial condition (not measurable)

can be met, according to core physics calculations, by a reactor power

level of 26.5 ± 0.5 MW. Table 3 contains the initial condition

specifications for the experiment. Systems or controllable parameters not

identified in Table 3 shall be operated as specified in the Plant Operating

Manual.

Prior to the preconditioning phase of the experiment, the necessary

hardware configurations will be established and the required prerequisites

identified in Section 4.1 will be completed. The high power fuel rod

burnup will be achieved in the preconditioning phase and pretransient

phase. The fission product measurement systems operational readiness will

be completed during the reactor shutdown interval in the pretransient

phase. Also, the TIPS locations in the center assembly will be capped.

Power operation in the pretransient phase will establish the initial

conditions specified in Table 3 and the required minimum burnup on the

center assembly fuel rods prior to transient initiation. In the event

that reactor shutdown intervals occur after the initial conditions and

fuel burnup are reached, subsequent power operation will be based on

establishing only the initial conditions specified in Table 3.

4.3 Experiment Sequence

The following items will be completed prior to experiment initiation

but have no requirement to be done in the order listed:

1. Complete prerequisites established in Section 4.1.

2. Complete acquisition of pre-transient primary coolant system and

blowdown suppression tank liquid samples. These pre-transient

liquid samples will be analyzed by Chemical Sciences Branch and

LOFT Operations to establish a baseline data set.

3. Establish the initial conditions specified in Table 3.

The following actions should be performed in sequence:

11

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DRR-L-7450Rev. 1 Chg. 5

NOTE: Refer to Section 4.7 for abnormal event procedures, as

required, for the following actions.

4. Start the gross gamma detector systems not later than 300 s

before experiment initiation. Start the data acquisition and

visual display system (DAVDS) and the FPMS data system not later

than 60 s before experiment initiation.

5. Secure the pressurizer cycling and backup heaters.

6. Initiate primary coolant pump injection flow. Terminate primary

coolant pump injection flow at to +60 s ± 5 s.

7. Scram the reactor. Reactor scram does not include dropping the

CFM control rods (refer to Step 9A). Reactor scram establishes

time zero.

8. Trip the primary coolant pumps at 8 t 2s. The pumps are toundergo a normal coastdown.

9A. Drop the CFM control rods when the loop flow has decreased to1.5 x 106 lbm/hr. Verify that the CFM control rods are inbefore proceeding to Step 9B.

9B. Open the simulated break in the intact loop cold leg imnediately

after verification of CFM control rods in. This action is estimated

to occur at approximately 20-25 s.

10. Open the simulated LPIS line in the broken loop hot leg at

220 ± 5 s. The filter in this line shall be bypassed to prevent

plugging until the core reaches a temperature of 840 ± 5K

(1052 ± 9*F). Thermocouples to be used in determining valve

opening are listed in Table 8.

11. Close the simulated break in the intact loop cold leg when either

of the following occurs:

a. System pressure decreases to 1.2 ± 0.03 MPa (160 ± 5 psig).

1?

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4.5.3 Simulated LPIS Pipe

Thermal-hydraulic measurements in the simulated LPIS pipe will consist

of steam flow, steam temperature, and wall temperature.

4.5.4 FPMS

The identification and location of fission product measurements are

shown in Figure 4. These measurements are described in the FPMS Functional

and Operational Requirements (F&OR) document. 4

4.5.5 Postirradiation Examination

An integral part of the measurements which are necessary to meet the

experiment objectives are the postirradiation examination (PIE)

measurements. A summary of items specified for postirradiation examination

is contained in Table 5. Examination of these items will be in accordance

with the postirradiation plan for LP-FP-2. 5

4.5.6 Critical Measurements

Sets of critical measurements, required during the transient and

posttransient phases of the experiment, have been identified and are listed

in Tables 6 and 7, respectively. The transient phase of the experiment

should not be initiated without these measurements since the experiment

objectives may be Jeopardized. Appendix A lists by instrument identifier

all critical measurements which are considered necessary for the successful

conduct of the experiment. The measurement uncertainties will be equal to

or less than those specified in the document "LOFT Experimental Measurement

Uncertainty Analysis,* NUREG/CR-0169.

A complete list of measurements required for Experiment LP-FP-2 is

provided on the Data Acquisition Requirements list to be published prior to

the experiment.

.19 I

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DRR-L-7450

Rev. 1 Chg. 5

The digital data acquisition and processing system, and analog and

) digital data acquisition recording is required to begin no later than i min

before initiation of the transient phase of the experiment. Continuous

PLSS recording is required through the posttransient phase of the

experiment.

Measurements identified on the Data Acquisition Requirements List that

fail prior to experiment initiation should be repaired or replaced if

possible. If a failed instrument(s) cannot be repaired or replaced, the

Joint Experiment Group shall determine the course of action.

Process instruments requiring calibration prior to Experiment LP-FP-2

are listed in Table 1.

4.6 Experiment Termination

Experiment LP-FP-2 will be terminated at the end of the posttransient

phase of the experiment. The posttransient phase ends with the completion

of the. time interval required for monitoring the redistribution of fission

products in the vapor and liquid volumes in the blowdown suppression

system. The time interval is specified in Section 4.3 to be 12 hr minimum

after closure of the simulated LPIS line.

4.7 Abnormal Experiment Sequence

If instrumentation, hardware components, or operating systems fail

prior to or during any of the four phases of the experiment, every effort

should be made to substitute, repair, or provide alternate actions to

safely continue the experiment and to meet the programmatic objectives.

The CFM control rods are to be inserted in Sequence Step 9A.

Verification of the insertion of these control rods is to be obtained

from proximity switch indications before proceeding to Step 9B. If

insertion of the CFM control rods cannot be verified the experiment may

be either (1) placed on hold for further decisions or actions to drop

20

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DRR-L-7450Rev. 1 Chg. 5

the CFM control rods, or (2) terminated. If a hold is placed on the

experiment, the experiment sequence must be resume(] before the decay

heat loss jeopardizes the experiment. If too much decay heat is lost

during the hold, the experiment must be terminated. The minimum

acceptable decay heat is 641 kW at time zero plus the hold interval

plus 200 s. After termination, on completion of the corrective actions

the experiment is to be resumed as approved by the LOFT Operations

Manager, DOE Site Program Manager and LOFT Program Division Manager.

Sequence Step 9B opens the break path in the intact loop cold leg.

This flow path is the primary blowdown path and is intended to be the path

for venting the major part of the primary coolant system fluid. High

quality steam flow only is desired for venting through the simulated LPIS

pipe in the broken loop hot leg. If the intact loop cold leg break cannot

be opened, the experiment sequence must be halted. On completion of the

corrective actions the experiment is to be resumed as approved by the LOFT

Operations Manager, DOE Site Program Manager and LOFT Program Division

Manager.

The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.

This flow path is the designed fission product and aerosol vent path to the

BST. If this line cannot be opened within 50 s of the time specified,

close the intact loop cold leg break and commence recovery operations with

the ECCS. Resume the experiment as approved by the LOFT Operations

Manager, DOE Site Program Manager and LOFT Program Division Manager.

21

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DRR-L-7435Rev. 1 Chg. 3

The intact loop cold leg break is to he closed on either a cladding

temperature value or a system pressure value as specified in Sequence

Step 11. If this break is not closed then another flow path to the BSTwill exist for fission product and aerosol venting. Two vent paths are not

provided for in the experiment plan. If thi- break cannot be closed before

the CFM cladding reaches 840 + 5 K (1052 1 90F) then terminate the

experiment and commence recovery operations with the ECCS. If experiment

termination and plant recovery operations commence with the ECCS, return to

Sequence Step 1 or as approved by the LOFT Operations Manager, DOE Site

Program Manager, and LOFT Program Division Manager after repairs are made.

The FPMS Fl and F2 sampling systems are to be opened at a CFM cladding

temperature of 811 + 5 K (1000+ 90 F) if the system pressure has decreased

below 1.43 + 0.03 MPa (195 + 5 psig). The pressure criterion corresponds

to the Fl and F2 design pressure limit of 200 psio. The LPIS line filter

is to be valved in and the gamma densitometer sources are to be stored

following the opening of the Fl and F2 sampling systems at a CFM cladding

temperature of 840 + 5 K (1052 + 9OF). The following courses of action

are defined in the event that abnormal conditions occur:

22

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DRR-L-7435Rev. 1 Chg. 3

Abnormal Condition

System pressure below criterionbut either F1 or F2 isolationvalves fail to open

Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.

All criteria met ana F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.

Course of Action

If either the Fl or F2 isolation nalvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand comTmence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.

Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.

The experiment is to proceed.Continue attempts to completethese actions.

If the condition occurs where the system pressure is above the maximum

allowable for F1 and F2 operation when the CFM cladding temperature is

800 K (981°F), commence actions to lower the system pressure. These

actions may be, but not limited to, opening the PORV or opening the

intact loop break path. Terminate all of these actions before the CFM

cladding temperature reaches 1050 K (1431'F). The experiment is to

continue in the event that the system pressure cannot be lowered below the

F1 and F2 operating pressure limit. If the actions taken to lower the

pressure are successful, terminate all actions before the system pressure

decreases below an indicated value of 1.2 MPa (160 psig).

The LPIS break line filter is to be valved in when the peak cladding

temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the

differential pressure may increase. At a filter differential pressure of

150 psi, the filter should be bypassed to maintain the ability to measure a

sample in the aerosol collection lines. If this occurs, the time shall be

recorded.

23

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DRR-L-7448Rev. 1 Chg. 4

The FPMS sampling lines are closed, the denosition counon devices are

closed, and the simulated LPIS line is Closed in Seauence SteD 13. U any

of these actions cannot be comoleted the experiment seauence is to

continue. Core refiood must commence in time to orevent tomperatures frcm.

exceeding 1583 K (2400 0 F) on tne center fuel assetflo sn, oud outer wa,.I

and 1533 K (2300 0 F) on tne peripneral assemoly fuel rca cladding.

Continue attempts to complete Sequence Step 13 but continue tnrough

Sequence Steps 14-17.

The possibility exists that the time interval between the occurrence

of 2100 K (3321 0 F) cladding temperature in the center assembly and. tne

peripheral assembly cladding temperature or the shroud outer wall

temperature limit may be excessive relative to core damage and source

release limitations. Therefore, a maximum time of 7.0 minutes is,

specified, after the 2100 K (3321 0 F) temperature is reached, for

continuing the experiment sequence with Steps 13 and 14.

The last situation to be considered is if the CFM cladding has

exceeded 2100 K (3321 0 F) and if the temperature indications on either

the CFM shroud outer wall or the peripheral fuel assembly claddina begin

to fail before the temperature limits are reached. In the case of

the peripheral fuel assembly cladding, a minimum of two valid temperature

indications in the hot region are required to maintain the experiment termination

criterion in effect. (Two valid indications ot the temperature limit will

activate exper.ime'nt termination.) If the minimum number of temperature

indications is not maintained before the limit is reached, a clock function

defining the time remaining to experiment termination will be utilized. The

specifics of the operation of the clock function are the responsibility of

the Operations Branch.

23A

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DRR-L- 7450Rev. I Chg. 5

TABLE 2. EXPERIMENT LP-FP-2 EVENT TIMES AND PARAMETER SETPOINTS

Item Action Time/Setpoint

Reactor scram Peripheral t 0

CR scram

Primary coolant pumps Tripped 8 ± 2s

CFM control rods Dropped 1.5 x 106 lbm/hr

Intact loop cold leg simulated break Open Verification of CFMcontrol rods in

Broken loop hot leg simulated LPIS Open 220 ± 5 spipe

Intact loop cold leg simulated break Close 566 ± 5 K (560 ± 90 F)

FPMS sampling system isolationvalves

LPIS Break Line Filter

Gamma densitometer sources

FPMS PCS sampling system isolationvalves

Broken loop hot leg simulated LPISpipe

Deposition coupon device

Core reflood

Open

Open

Store

Close

Close

Close

Initiation

1.2 ± 0.03 MPa(160 ± 5 psig)

811 K ± 5 K(10000 F ± 90 F)

840 K ± 5 K(1052 0 F ± 90 F)

840 K ± 5 K(1052 0 F ± 90 F)

1462 K (2172'F) onperipheral fuel assemblycladding or 1517 K(2272°F) on CFM shroudouter wall or other experimenttermination events.

6 ± 0.5 s from simulatedLPIS line closureinitiation

71

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TABLE 3. INITIAL CONDITIONS FOR EXPERIMENI LP-FP-2

Parameter

Primary Coolant System

Core 6T

Hot leg pressure (MPa)(psia)

Hot leg temperature (K)(OF)

Primary coolant flow (Ibm/hr)(kg/s)

Boron concentration (ppm)

Power level (IMW)

Decay heat at 200 s minimummaximum

Maximum linear heat

generation ratec (kW/m)(kW/ft)

Control rod position(above full-in position) (m)

(in.)

Steam Generator Secondary Side

Water level

Pressurizer

Liquid level (m)(in.)

Specified Value a

As requiredb

14.95 ± 0.12168 t 15.0

571 ± 1.1569 ± 2

3.8 ± 0.15 x 106

479 ± 19

As requlredb

26.5 ± 0.5

675 kW695 kW

-40-12.26

1.37 _ 0.0154.0 ± 2.0

As requiredb

1.12 1 0.144 1 4

I

I72

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APPENDIX B

FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2

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APPENDIX B

FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2

This appendix describes the design and operation of the Fission

Product Measurement System (FPMS) that was used during Experiment LP-FP-2.

The FPMS consisted of three basic systems: the aerosol sampling system,

which was operated during the transient phase only, and the gamma detection

and deposition coupon systems, which collected data during both transient

and post-transient phases. Each of these systems is described herein.Figure B-l is a schematic of the FPMS showing the location of each of the

individual measurements.

Deposition Coupon System

Stainless steel deposition coupons were positioned in the reactor

vessel upper plenum region to provide postexperiment information on fission

product plateout. These are designated D-l on Figure B-l. Two coupons

were located at each of three axial elevations, corresponding to 0.152,

0.61, and 1.65 m (6, 24, and 65 in.) above the upper tie plate. Both

coupons at each elevation were exposed to the reactor environment during

the heatup. One coupon at each elevation was isolated and sealed prior to

initiation of reflood, while the second coupon remained exposed. Thus, the

plateout during the heatup phase was to have been distinguished from the

plateout/leaching during the reflood phase.

The D-l deposition device is a hollow rod containing deposition

coupons. At experiment initiation, the 0-1 deposition device was full of

liquid water. A nitrogen purge gas system was connected to the rod to

ensure dry coupons for fission product plateout. The hollow rod was pusheddown before reflood to isolate the protected coupons. At that time, the

nitrogen gas purge was restarted to remove steam, which could condense onto

the coupons. The nitrogen gas supply to the rod was then to have been

controlled at 1.4 MPa (200 psia) above reactor pressure to ensure that anyleakage of the deposition rod seals was outward, thereby maintaining a dry

atmosphere for the protected coupons.

8-1

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F2Aerosol saper..

%A I'

SuppressioveslG G2

G3

5 4066

Figure B-i. FPMS Schematic.

B-2

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The D-2 and D-3 coupons were located upstream and downstream of the

simulated LPIS header, respectively. To allow only high quality steam toflow in the line, it was not opened until the primary system mass inventory

had decreased. In addition, the line was isolated prior to reflood so that

these deposition coupons were protected from water flow.

Filter Sampling System

There were three filter sampling systems installed for this

experiment. These systems provided samples of the vapor and aerosols

generated during the heatup phase of the experiment. Both of these

constituents were expected to combine to provide the medium for transportof the fission products. Figure B-2 is a schematic representation of thedesign of the Fl and F2 sample lines. The filter sample locations were:

0 Fl--in the reactor vessel upper plenum at 1.80 m (70.75 in.)above the top of the lower tie plate

o F2--in the broken loop hot leg spool piece just outside of the

upper plenum

o F3--in the exit of the broken loop hot leg

The Fl system consisted of the following major components:

o sample line probe

o cyclone separator/isolation valve

o dilution filter

o virtual impactor

o collection filters

6--3

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F1 sample line F2 sample line

I f iI

I I j filter

Collectionfilters

co

Figure B-2. Schematic of F1 and F2 aerosol sample systems.

( (

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0 infrared moisture detectors

o recombiner

o critical flow orifice

o gamma spectrometer.

The sample line probe, shown in Figure B-3, diluted the vapor/aerosol

sample with an inert gas to minimize sample line deposition and to inhibit

interactions within the sample.

The cyclone separator/isolation valve, shown in Figure B-4, isolated

the filter assembly before and after the heatup phase and removed particleswith an aerodynamic diameter larger than 20 to 30 micrometers.

The dilution filter reduced the mass loading of the aerosols to

prevent plugging of the virtual impactor.

The filter train, shown in Figure B-5, consisted of the dilution

filter, the three stage virtual impactor, shown in Figure B-6, and the

collection filters. The train separated the aerosols into size ranges of6 to 20 micrometers, 1.7 to 6 micrometers, and less than 1.7 micrometers

with each size range being collected on a separate filter.

The recombiner contained cupric oxide, which converted the hydrogen to

water. The infrared moisture detectors then provided quantitative data onthe amount of argon, hydrogen, and steam entering and exiting the

recombiner. These data provide the necessary input for calculating the

amount of hydrogen and the dilution ratio of argon/hydrogen sampled during

the transient.

The critical flow orifice provided a mass flow out of the line during

the transient.

B-5

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V

Sheath gas in

Mixture out

iling

Heating/coolinggas out

I-

gas in

I1F,4= Il

Sample flow INEL 4 4962

Figure B-3. Samiple line probe.

9w

B-6

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Pneumatic operator

rator/'

From sampleprobe

5 4060

Figure B-4. Cyclone separator/isolation valve.

Three-stagevirtual Collectionimpactor filters

Dilution -- @- 6-20 ýJm H

filter

Sample in ""• b "w 1.7-6 pm

F--O[< <1.7 j•m I

Flow-controlorifices

I , Efi

I ou

I

fluentt

5 4064

Figure B-5. Impactor and filter train.

B-7

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.113 in.

0.040 in.

1.600 in.

/ 0.048 In.

0.110 in.t

id = 0.071 In. Id - 0.025 in. dia 0.026 in.

IW - --o.325 in. 0 .500 in. .I II

5 4065

Figure B-6. Three stage virtual impactor.

8-8

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The F2 line was similar to the Fl line except for the deletion of the

moisture analyzers and inert dilution gas. The F3 line, also designatedthe simulated LPIS line and shown in Figure B-7, contained the following

components:

o Ueposition samples upstream and downstream of the gamma

spectrometer (Dl & D2)

0 Gamma spectrometer (G5)

o Filter (F3)

o Flow venturi.

FPMS Gamma Detection Sampling System

Four gamma spectrometers and one gross gamma monitor were used in the

FPMS to provide a real time quantitative measurement of the radioisotopes

present in the LOFT system during the 12-hour post-transient sampling

phase. Two of the five were operated during the transient phase. The

sample points are shown on Figure 8-1 and are as follows:

o Gl--spectrometer operated only during postexperiment: reactor

vessel lower plenum at 0.584 m (23 in.) below the core, or

alternately from the primary coolant hot leg in the horizontal

PC-3 flange.

o G2--spectrometer operated during transient (combined Fl and

F2 sample lines effluent) and postexperiment: vapor space of the

BST.

o G3--spectrometer operated only during postexperiment: liquid

space of the BST.

o G5--spectrometer operated during transient: upstream of the

filter in the simulated LPIS line.

B-9

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Removablespool

y-spectrometer

Removablespool

Filter Venturispool spool

Broken loophot leg

Blowdownheader

5 4061

Figure B-7. Simulated LPIS line components.

B-10

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o G6--gross gamma monitor operated during transient: viewed sample

being drawn by Fl sample line located in the reactor vessel upper

plenum, which is 1.80 m (70.75 in.) above the lower tie plate.

The gamma spectrometer sample systems included valves for isolation

and sample point selection, pumps to provide flow, and pressure and

temperature instruments. The samples were returned to the same source that

was being sampled. The Gl, G2, and G3 spectrometers were enclosed in a

tent, to which an inert gas purge was applied to minimize the buildup of

background contamination that occurred during Experiment LP-FP-l.Additionally, the liquid and gas sample lines were purged with clean water

and inert gas, respectively, to measure plateout.

Each gamma spectrometer was designed to operate remotely over a broad

range of sample intensities. To improve accuracy, the spectrometer was

calibrated during the experiment (also remotely) using a thorium 228 source

mounted on the collimator wheel and background radiation levels were

recorded.

B-11

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APPENDIX C

DESCRIPTION OF THE LOFT SYSTEM

AND INSTRUMENTATION

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APPENDIX C

DESCRIPTION OF THE LOFT SYSTEM

AN) INSTRUMENTATION

The LOFT facility was designed to simulate the major components and

system responses of a commercial pressurized water reactor (PWR) during a

LOCA. The experimental assembly includes five major subsystems that have

been instrumented such that system variables can be measured and recorded

during a LOCA simulation. The subsystems include the reactor vessel, the

intact loop, the broken loop, the blowdown suppression system (BST), and

the ECC systems. Complete information on the LOFT system is provided in

Reference C-l, and a discussion of the LOFT scaling philosophy is provided

in Reference C-2.

The arrangement of the major LOFT components is shown in Figures C-l

and C-2. The intact loop simulated three loops of a commercial four-loop

PWR and contains a steam generator, two primary coolant pumps in parallel,

a pressurizer, a Venturi flowmeter, and connecting piping. A spool piece

was connected to the intact loop cold leg downstream of the pump

discharge. This provided the initial break path during the blowdown. This

line was closed prior to fission product release so the fission product

transport would be solely in the simulated LPIS line.

The broken loop consists of a hot leg and a cold leg. For this

experiment, the broken loop cold leg was flanged off and the broken loop

hot leg pump and steam generator simulators were removed. The simulated

LPIS line was connected to the end of the broken loop hot leg and provided

the path for fission product transport from the primary system to the BST.

The simulated LPIS line and associated flow instrumentation are illustrated

in Figure C-3.

The LOFT reactor vessel, shown in Figure C-4, has an annular

downcomer, a lower plenum, lower core support plates, a nuclear core, and

an upper plenum. The downcomer is connected to the cold legs of the intact

and broken loops, and the upper plenum, to the hot legs. The core consists

C-1

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Intact loop Broken loop

I .4 1• Vapor monitoro ©(gamma spectrometer)

~' Pumps

ReactorveslSuppression Downcomer

Lower plenum

Reactor vessel5 0962

Figure C-1. Axonometric representation of the LOFT system for ExperimentLP-FP-2.

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( C

Morded Wket.,e

3 LPIS

HPIS 4 LpwPSe B

t1,• LJ.' -4argeo B

Aoolitpcantmpnto

.way fron TE-SG-3. 4. 5pwlhcWt n

.yeSt.. PT-P139-2, 3. 4

generat P ary ,Coolan

TE-SG-2-2A TE-SG-I-1A ventut, FT-P13-271.2, 3 T.PT139- 4

Spray I..C• 1 3 pE.PC.2 EL C-2 ME-PC-2A. 8, C - r

.old g 314 D.-PC-2. BD. DD inje.l1o 1 PE -L-IAcO-d le DTE-PC--2A. . P0,

P139 1w-6. 7.8 TE-PI139 -- -- -

Pd Cor. CV-PC38-O7A

TE-PI39-20

439- Upper

14PIT- platn..

Pre, iP13en30d1

Cydltg - - - -IB I r-L -b a c k u p2 T E -B L H . ? T E -B L ) . T E -B L H -5 5• . L -

reaterST 14 PE-11-2

no TE-BLPm2 •2 TE-B- PE-BL-I

p oPI3Pmp8 TE-Pt 39- S. C. 0, 6o 0a HL SL. .H C.D0 2 .9E.L2A B. C. B 1

BI down supeso tankCV.TI 33-57 2noze

to PCP-2

PT-Pt20-43"• -- cnetd tkprSue

.

I- CV-PE-P503-153 0 00-P 3e-1 PE.P33- F-Zt39-139EI -

Accueruia'o PuPA .e opunt~,_n,

I

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il (epetent •r==n

L ! q I ~~~~~3~~~~---

kU ýG...ACrAC1

Croseonn fine - 8-Pipe diameten in IncthesD Rd.UC.

380.94

iLi

III'

59673

Figure C-2. Schematic of the LOFT primary and emergency core coolingsystems.

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Steam temperature PE-BLH-003Pressure

Shield tank N.penetration

N.-

CV-P-138-192

Gammaspectrometerspool

K

Wall temperature TE-BLH-004

D2 (deposition sample) spool

!

out offilter

5 0961

Figure C-3. Simulated LPIS breakline instrumentation.

( ( (

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ft

DowncomerLLTLE-1 ST-1LE-1ST-2

-Upper plenum LLTLE-3UP-1

-Core LLTLE-11F10LE-3F10

INEL-MCL-3000

Lower plenumLLT-

Figure C-4. LOFT reactor vessel.

C-5

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of approximately 1200 enriched uranium fuel rods arranged in five square

and four triangular (corner) fuel assemblies. The fuel rods were designed

to commercial PWR specifications except that they are only 1.68 m (5.5 ft)

long and several fuel rods have special instrumentation. All 100 fuel rods

in the central fuel assembly were enriched to 9.74 wt% 235U and were

prepressurized at cold conditions to 2.41 MPa (350 psia). All fuel rods in

the peripheral fuel assemblies were unpressurized and were enriched to

4 wt% 235U. Figure C-5 shows the fuel cladding thermocouple locations

and Figure C-6 shows all the central fuel assembly instrumentation.

Figure C-6 also illustrates the insulating flow shroud which took the space

of the outer two rows of fuel rods. This shroud protected the peripheral

fuel assemblies from excessive temperatures while allowing the central fuel

assembly to reach the requisite temperatures for fuel rod failure and

fission product release.

The two LOFT ECC systems are capable of simulating the emergency

injection of a commercial PWR. They each consist of an accumulator, a

high-pressure injection system, and a low-pressure injection system. There

were no programmatic considerations inherent in ECC operation; therefore,

the ECC injection was not scaled to represent commercial PWR operations

during Experiment LP-FP-2.

The LOFT steam generator, located in the intact loop, is a vertical

U-tube design steam generator. Operation of the secondary coolant system

during Experiment LP-FP-2 approximated that of a commercial PWR.

Figure C-7 is an illustration of the steam generator and its

instrumentation.

Figure C-8 illustrates the thermocouples in the upper plenum.

Table C-1 lists the instrument nomenclature.

C-6

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TABLE C-1. NOMENCLATURE FOR LOFT INSTRUMENTATION

Designations for the Different Types of Transducers a

RPE - Pump speedPE - Pressure transducerPdE - Differential pressure

transducerLE - Coolant level transducerPS - Pressure switch

FE - Coolant flow transducerDE - DensitometerME - Momentum flux transducerFT - Flow rate transducerTC - Fuel centerline

thermocoupleTE - Thermocouple

Designations for the Different Systems, Except the Nuclear Core

PCBLRVSvUP

Primary coolant intact loopBroken loopReactor vesselSuppression tankUpper plenum

LP - Lower plenumST - Downcomer stalkP120 - Emergency core coolant

systemP128 - Primary coolant

addition and control

Designations for Nuclear Core Instrumentation

Transducer location (inches from bottom of fuel rod)

Fuel assembly row

Fuel assembly column

Fuel assembly number

Transducer typeIIITE-3B 1-28

a. Includes only instruments discussed in this report.

C-7

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Broken 1oo0cold leg

Intact loophot leg

1

2@@@ZI( r@ 2 ®20@@@2 &

®@®@'A

3 ri(D@@P®A

135" 225'

Q-- Identification number ?SF-45" 0 315 Height of thermocouple above

\core bottom (in.)

Thermocouple

G5 TO® © •@42

/c77©WC©®EN

\._ @1®©7

W ©2

©

@2®424(241

0

6Z

N3@ar32\r-'Typicalrod guide

tubeQ

X 9JSD79

7.l2 12 I 1 J

®©®5 0970

Figure C-5. LOFT cladding and guide tube thermocouple locations.

( ( (

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LP-FP-2 Center Fuel BundleInstrumentation

A IB JC ID I E IF IG IH III J IK IL IM IN 107

TZ-58-

10 27 32 42m

1

2

3

4

5

6

7TZ-SE- 10

278 32

42

9

10

11

12

13

14

-1N15

I

- 0_C227

00 ®Q2®

27 G) 6

©QQO N ® 27

Thermal sh.edflow shroud

I

TZ-5W-11027

3242

I10 2" 32 42

TZ-SN-

Instrumented guide tube

Instrumented fuel pin

Q Neutron flux scan tube (tip)Note: Thermocouple at location F7-42 failed prior to bundle installation

5 0963

Figure C-6. Center fuel assembly instrumentation locations.

C-9

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HF-SG-001

Differential pressureFeedwater transducer for feedwaterinlet• liquid level LT-P004-8A

& 8AA

Secondary sidecoolant temperaturethermocouple, TE-SG-5

pressurefor feedwaterLT-P004-BB

Secondary sidecoolant temperaturethermocouple, TE-SG-4

FRC-202le Secondary side

le 'coolant temperaturenperature thermocouple, TE-SG-3p ie s , -,IA • Primary side

coolant temperaturethermocouples

~TE-SG-2 & 2A

inlet Primary coolant outlet

INEL-MCL-1803

Figure C-7. LOFT steam generator instrumentation.

C-10

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( C CFuel

Assembly1

FuelAssembly

2

Fuel FuelAssembly % Assembly

3 4

FuelAssembly

5

FuelAssembly

6Station*

290-

280 -

270 -

260 -

250 -

240 -

0- TE-1UP-4

("3I 230 -

220

210 -

200 -

190 -

180-

0- TE-1UP-6

0- TE-1UP-7DTT

DL4TE.1UP.50

TTE-iUP-i"•TE-lUP-2

--A TE-3U.1P-48 • TE-3UP-8

0- LE-3UP-1-1

I 0 LE-3UP-1-2

Io -TE-3UP-10 I

LE-3UP-I-3 I0 'xTE-3UP-11 0 TE-5L

1 - LE-3UP-1-4 TE-5LI TE-3UP-12 O TE-5L

O- LE-3UP-1-5 - TE-5LI - TE-3UP-13

0- TE-2UP-4 0 LE-3UP-1-6 -TE-4UP-4

LýE-3UP-i-7I •TE-3UP-14 K TE-5L

U XLE-3UP-1-8 7 E5

LE-3UP-1- 110 TE-51.TE-3UP-15LE-3UP-1-8 0O TE-5ULE-3UP-1-9 OTE-5L

O---TE-2UP-250 I o31 TE-4UP-005. TE'5UI DTT 1 0OTE-5U

-TE-2UP-1 1 TE-4UP-10 E50 --,TE-2UP-210 0<TE-4UP-2 1I TE-5L•TE-2UP-3 I `'TE-4UP-3 i?¶ TE-5I

TE-3UP-1 I 1t4S/rTE-sII. TE-5[

P-31 A, B"

JP-30A, B1

IP-215B1, B2IP-29A, B & 250G1, G2

JP-33A,IBI

JP-32A, B!JP-251B1, B2JP-250G1, G2

I,. r."1- n*I I~ A

Nozzlecenterline

Top of uppertie plate

IP-197B1 ,B20TE 6 UPIP-194G1,G2 T.....IP-28A, B I'l TE-6UP-

P4" 0 TE-6UP-2JP-4, 17, TE-6U"P-

19,23JP-24, 25, 26, 27JP-188A, B, C, D

T

*Station numbers are a dimensionless measure ofrelative elevation within the reactor vessel. Theyare assigned in increments of 25.4 mm withstation 300.00 defined at the core barrel supportledge inside the reactor vessel flange.

5 0971

Figure C-8. Reactor vessel upper plenum instrumentation locations.

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REFERENCES

C-1 0. L. Reeder, LOFT System and Test Description (5.5-ft Nuclear Core ILOCEs), NUREG/CR-0247, TREE-1208, Change 1, Sept. 1980.

C-2 L. J. Ybarrando et al., "Examination of LOFT Scaling," 74-WA/HT-53,Proc. 95th Annual Winter Meeting of the ASME, New York,November 17-22, 1974.

C-12

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APPENDIX D

PCS THERMAL/HYDRAULIC RESPONSE

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APPENDIX 0

PCS THERMAL/HYDRAULIC RESPONSE

This appendix describes the thermal/hydraulic boundary conditions in

the primary coolant system (PCS) during the transient and posttransient

phases of the experiment. Particular emphasis is placed on the estimated

core hydraulics and on the fluid conditions in the vicinity of the fission

product measurement system (FPMS) measurement locations during the time of

fission product release and transport. This appendix is broken into two

parts, the first describing the PCS response during the transient and the

second, the PCS response during the posttransient phase.

1. RESPONSE DURING BLOWDOWN AND REFLOOD

Experiment LP-FP-2 was dominated early by a gradual PCS level decrease

leading to a slow core boil-off. The level in the intact and broken loops

can be inferred by examining the densitometer data. Figures D-1 to D-4

show the average chordal densities measured by the individual gamma

densitometer beams in the intact loop hot leg, intact loop cold leg, broken

loop hot leg, and broken loop cold leg, respectively. In each case, the

three chordal densities are shown except for the A-beam at the intact loop

hot leg location. The data from this detector were failed due to

inadequate background correction. As explained in Section 2, the 6 0 Co

source was prematurely isolated. Thus, density data are only available for

about the first 260 s of the transient. These data show that the loops

started voiding at approximately 50 s in the intact loop hot leg, 85 s in

the intact loop cold leg and broken loop hot leg, and 120 s in the broken

loop cold leg. The voiding continued but was not complete when the

densitometer sources were isolated at 262 s. While the level decrease in

the loops could not be directly monitored beyond this time, it is clear

from thermocouple data in the upper plenum that the loop voidage was

complete by 470 s. This is illustrated in Figure D-5, which compares the

response of an upper plenum thermocouple with saturation temperature. The

time at which the thermocouple response deviates from saturated (470 s) is

an indication of the time when the level had decreased below that

0-1

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_0

'0

LA-

0.75

0.50

0.25

'4-

(nC0

~0

-I,

U-

0 50 100 150 200 250 300Time (s)

Figure D-1. Intact loop hot leg densities.

I

N"0.75

0.50

0.25

.00 50 100 150

Time (s)

if)4-.4-NE

>5

to

0-o0

U-

200 250 300

Figure D-2. Intact loop cold leg densities.

0-2

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I

C,,C0,-D

-o:3

LL

0.75

0.50

0.25

-I-

E

"3CU,-o

-D

:3

LI-

00 50 100 150 200 250 300

Time (s)

Figure D-3. Broken loop hot leg densities.

0,

C0-o-o:3

0.75

0.50

0.25

I')

E.0

U,

U,

-U

:3

0 50 100 150 200 250 300Time (s)

Figure D-4. Broken loop cold leg densities.

0-3

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625 I , I

FFTEi-011Oll 650I --- T-PC-02

600

.4-0

-55

CL550E

525

0 50 100 150 200 250 300 350 400 450 500 550 600Time (s)

Figure D-5. Comparison of upper plenum fluid temperature with saturationtemperature. (See Appendix I for thermocouple qualificationlimits).I

D-4

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elevation. Since this thermocouple is located at an axial elevation just

below the bottom of the loops, this time represents an approximation for

the time the loops were completely voided, with the exception of the loop

seals.

Figure D-6 shows the upper plenum level and indicates that the upper

plenum was voided at least down to within 1.1 m (43 in.) of the top of the

core by approximately 600 s. (This liquid level probe has been in the

reactor vessel since the initial core fuel load was installed in 1978. The

conductivity probes at several of the elevations had already failed prior

to conduct of Experiment LP-FP-2.) The level dropped into the core region

by about 700 s. This can be inferred from Figures D-7 and D-8, which show

the level in the Fuel Modules 1 and 3, respectively. The heatup, which

initiated in Fuel Modules 4 and 5 as discussed in Appendix E, preceeded

this measured level drop by approximately 1 min. This apparent discrepancy

could have been caused by a combination of two phenomena. First, the level

probes are located in the two instrumented corner fuel modules; the radial

decay heat distribution could have resulted in a depressed level near the

center of the core with a higher level near the edge. Based on0-1

calculations using the traversing in-core probeD, the ratio of initial

linear heat generation rates between the hot fuel rod in the center fuel

module and the fuel rod adjacent to the level probe is 2.3. The ratio

between the hottest part of the peripheral fuel assembly and the fuel rod

adjacent to the level probe is 1.4. The same values should exist for the

ratios of decay heat levels at these locations. Second, a liquid level

probe operation depends on the fluid between the closely spaced probes

being at the same void fraction as that of the bulk fluid. In the

environment of lower decay heat in the corner assemblies, it could be

expected that the response of the probes lags the voiding of the bulk

fluid. The boil-off continued until the entire center fuel module was

dry. Figure D-8 indicates that Fuel Module 3 was essentially uncovered by

approximately 1000 s. However, Figure 0-7 indicates the persistence of

fluid in Fuel Module I throughout the transient. Again, interpretation of

the data must take into account the limitations discussed above.

0-5

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*** BUBBLE * VERSION. 001 * MOO 002 * OZ/Z3/81 *** DATE OF RUN 85/09/10. PAGE 00C2

BUBBLE PLOT OF FILE ILE3UP 0.

* CHARACTER RANGE TABLE *

* ( ) 1.000 K5 < 999.000 ** (0) .010 5 -C 1.CCC ** (X) -999.000 5 N ..010 *

4=• LEVELI (CM ABCVESCORE TOP)

19?.00 *XXXXXXXXXXXXXXXXXXXXXXXXXXX0132aOC *XXOOOOCOOOOOO0000000COCO00000000000o112.00 *XO000CO0000000000000COOOOCO

-. 053 165.147 339*147 513.147 687.147 861.147 1C35.147

oXXXXXXXXXXXXXXxOXXXXXXXXXXXXXXXOXXXXXXXXXXXXXXXx

1209.147 1363.147 1557.147 1731.147 1905.147 2079.147

TIME (S)

Figure D-6. Conductivity level probe response above Fuel Assembly 3.

I ( (

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C C

J

* BUBBLE * VERSION 3001 M OD 002 * 02/73/81 C DATE CF RUN 85109/10. PAGF COCZ

BUBBLE PLOT OF FILE 'LE1FIO '.

* CHARACTER RANGE TABLE

* C ) 1.000' • N 999.000 *• (0) .010 N H 1.000 *

( CX) -999.000 • N • .010 *

LEVEL(M ABOVE

CORE BOTTOM)

1-68 *XO0000CO000aO00300O0COCC000 DOOCCO00000 coooo000x~xxxxx~xxxxxyxxxxxxxxxxx

1.67 *XOOOOOCOOOO3OOOOOOOOCOCOCOOOOOOOOOOOOOOOO 00 000C00CCCCxOxx XXXXXXXXXXXx.71 *XXXGC0COO003030000C0CCOOOCOOO000OOOXXOOO oootcXXXXXXx 00xxoc0coCCCOC~CC~COOCcc GyxxYxxxxxxxx.51 *XXXXXXXXXXXXOCOcoca cooxoxooxxxxxxxxxxxxxxxxxxxxxxx XXXXX • x•xx•cc•ccxocc xxxyx • X.41 *XXXXOXCXoX•CXCXCXCYC0 00 X xoCXx xXXXXxxX X COx•×•x X oxוCocoxxxxYx xx xXXXK.20 *XXXXXXOOOOOOOOOOOOOOcocooooooooooooooOooooooxxxxxxOOCOooCOCCOoO GCOOCCCCCCCcCCCCOCCXXXXXXXXXX

-. 053 158.447 325.447 492.447 659.447 826.447 993.447 1160.447 1327.447 1494.447 1661.447 IF26.447 1995.447

TItE (S)

Figure D-7. Conductivity level probe response in Fuel Assembly 1.

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*** BUBBLE * VERSICH 001 * MOO 002 *' 02/23/81 **E DATE OF RUN e5/09/10, PACE 0003

RUBBLE PLOT OF FILE 'LE3F1O '.

* CHARACTER RANGE TABLE *

* ( ) 1.000 1 N 4 q99.000** (0) .010 i H 4 1.000 *

M (Xl -999.000 i N ' .010 *

LEVELtM ABOVE

CORE BOTTOM)

1.78 *XXXXOX000000000O0000COOCOYXXXXXXXXXXXXX000000OXCC000OCO OxxxxxxxxxXxx0o 1.08 *XXXXOO0OOOOOOOOOCOCOCCOCOCOOOooooooo0000000 0 0 0 0 cOcO0OxxxxxxxxxXXXXXXXXXxXXXXXXXXXXXXXXxxxxXXxxXXXXxxXXX.98 *XXXXGOOoooooo0ooooooCOC00COoo000000000000000000 oCO0000XXXXXXXXOOOOOOOCOOOOOOOcOOCOOOOXXxXXXXX XXXXX.89 *XXXXOOOOOOOOOOOOOOOQOOOOOCOOOOOOOOOOOOOOCXXOOOOOO OOOOOOCCOOOO0OOCOCCCOCCOCOOCXXXxxxXXxxxx.71 *XX00000C00000000C000000C0000O000000000000C0O000 OOOCXXXXXXXXXXXXXXXXXXXXXXXXXKXXXXXXXXXXXXXXXXXXXk.61 *XXXOOCOOOOOOOOCOC0ooCCOCCCOOO000000000000000000 OCC0C0OGCcXXXXxOCCCCCO000CCCCCCOCCC Coxxxxxxxxxx.51 *XXXOOOOO0OOOOOOOOCOOCOOOOO0000000O003000XXXOXOCO0 OcooooxxXxxxxxxxxxxxxxxxxxxxcoooocxxxXxxixxxx.41 *XXXXOXCOOOOOOOOOCOOO0CCOCOOOOOOOOOOOOOOOOOOXXXXXXOO OOXXCOC ccccoC00coCCCo oxxxxxxxxxxxx.20 *XXXXXXXXOOOOOOOCCOOOCOOCCOOO00000000XXXXXXXXXXXXXXXXXXXXXXXXOo oxxxxxxxxxXXX.10 * oocc00ccoccoooc • 0xx•00xxxxcoxCxxxxxxxxx•c• OOCCCCCCCCO00Cxxxxxxxxxx

-.053 1!8.447 325.447 492.447 b5Q.447 826.447 993.447 1160.447 1327.447 14q4.447 letl.447 1e2P..147 IC,.447

TIME (S)

Figure 0-8. Conductivity level probe response in Fuel Assembly 3.

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The in-core self powered neutron detectors (SPNDs) have previously

been shown to be sensitive to changes in liquid level.D-2 Figures 0-9

to 0-11 show the response of the SPNDs in the peripheral fuel modules.

These detectors are centered at an elevation of 0.66 m (26 in.) above the

bottom of the core and each indicates that the local void fraction

increased to near 1.0 at that elevation between 690 and 740 s. Cladding

temperatures in the vicinity of the SPNDs indicates heatup initiation at

approximately 730 s, or 40 s after initiation of the SPND response to the

density decrease. This time lag between SPND response to density decrease

and heatup under decay heat conditions is consistent with previous

experiments.D-2 The SPND output exhibited a sharp decrease at

approximately 735 s, or approximately when the intact loop cold leg break

was isolated. This could be indicative of a rapid collapse in liquid level

as the depressurization and steaming rates suddenly decreased. The SPNDs

which were installed in the peripheral fuel modules are all prompt

detectors utilizing a Co emitter. The SPNDs which were installed in the

center fuel module in Experiment LP-FP-2 are delayed detectors and utilize

a Rh emitter. The response of these latter detectors to the transient is

discussed in Appendix G.

The PCS mass inventory was derived from the mass increase in the

Dlowdown suppression tank. The suppression tank level is shown in

Figure D-12. This level is an average of two independently measured levels

in the suppression tank. It should be noted that due to the large

oscillations in the data and the nonphysical offset between the two level

measurements, these data are not qualified during the transient. Thus, the

derived PCS mass inventory is useful for trend information though not for

absolute magnitudes during the transient. The trends are reasonable,

however, and a single point check of the mass inventory can be made since

the levels were qualified both for initial conditions and for the time

after isolation of the PCS. The derived mass inventory is shown in

Figure D-13 and indicates that the inventory decreased from an initial

value of 4700 kg (10360 Ibm) to a minimum of just over 500 kg (1100 Ibm) by

13uO s. After this time, the mass flow was high quality steam and the

overall mass inventory remained approximately constant (at least within the

D-9

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E

Y0.6

0-

~O5°•0 .

L

0.4

ED

0.7

0.3aU

-.J

0.2600

Figure 0-9.

• 0.9C

"- 0.8L0CS

S0.7

0

-- 0.6-00

-J

NE-2H08-26 700

N-

•600

C

0-500 :-a

$..

Q)

400 o

-300

0

:5 650 675 700 725 750 775 800

Time (s)

Response of SPND in Fuel Assembly 2 to core uncovery.

62

-1000 I.. B

9D

-900 cn

C

0

600

C

'700 0)•

4-

600 _

0C.,.

0.5 L600 625 650 675 700 725

Time (s)750 775 800

Figure D-1O. Response of SPND in Fuel Assembly 4 to core uncovery.

0-10

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0.6

E

"0.5

0

-04c

0) 0.3

0l-0.

0

0

0.1600 6

Figure D-11.

1.35 r

25

NE-6H08-26 600I...

- C

500 M

0-400

0L.

-300 ,

-- 200

0o

5 650 675 700 725 750 775 800T i me (s)

Response of SPND in Fuel Assembly 6 to core uncovery.

4.4

4.3

-.4-

4.2 v

4.1 -0

4.0 '

-j

3.9

3.8

E

OD

-)

1.30

1.25

1.20

1.150 500 1000 1500 2000

T i me (s)

Figure D-12. Averaged BST liquid level.

D-1l

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I. Mor1

5000 I

- '4500-401000000

4000

03500 -8000 0

3 3000 U-,

E -60002500 - E

4-2000

C1000_ -2000 -

500

0 500 1000 1500 2000Time (s)

Figure D-13. Primary coolant system mass inventory.

0-12

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sensitivity of the suppression tank level transducers). The final mass

inventory, based on the (qualified) suppression tank level after isolation

of the PCS, was approximately 200 kg (440 Ibm), which is 300 kg (660 Ibm)

lower than the minimum value calculated using the transient calculation

method.

A key parameter affecting the core thermal response is the steam mass

flow in the center fuel, module during the period of metal-water reaction.

One of the concerns prior to the experiment was that blockage of the center

fuel module inlet might result in a steam-starved environment which would

limit the metal-water reaction. That, in turn, woula have limited the core

heatup and subsequent fission product release and transport. Although

there was no direct measure of the mass flow, there were two independent

indirect indications of the flow. The first was the thermal response

itself, which indicated that no steam starvation occurred (see

Appendix E). The second was provided by an analysis of the heat flux from

the cladding to the fluid. That heat flux is given by:

"(z) - h(Tclad - Tfluid)

Tfluid = TSAT f 2T a Q"(z) dz

where

a = fuel rod radius

cp steam thermal capacity

z axial elevation

S= mass flow

D-13

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h h= heat transfer coefficient

-" surface heat flux

T clad cladding temperature

T fluid fluid temperature

T SAT saturation temperature

Data for heat generation rates and fuel rod parameters furnish the

adiabatic heatup rate and the surface heat flux under isothermal

conditions. Data for cladding temperatures, differentiated with respect to

time and compared with the adiabatic heatup rate and isothermal surface

heat flux, provide an estimate of the actual surface heat flux. Combining

the two equations to eliminate the fluid temperature results in one

equation with two unknowns, the heat transfer coefficient and the mass

flow. An assumption was made that these two parameters did not vary with

elevation at any given time, and data from the three thermocouple

elevations in the center fuel module (0.25, 0.69, and 1.07 m, 10, 27, and

42 in.) were used to solve for these two unknowns with one check for

consistency. The calculation was performed using the data at 1300 s, or

just prior to initiation of the metal-water reaction. The resulting heat

transfer coefficient did not vary more than 25% at the three elevations,

which is considered reasonable for this analysis. The resulting total mass

flow rate for the center fuel module was 0.04 kg/s (0.09 lbm/s) or 0.4 g/s

(9 x lU"4 Ibm/s) per fuel rod which is approximately 3 times the value

calculated prior to the experiment. The mass flow rate was sufficient to

allow the metal-water reaction to proceed without steam starvation. This

result is consistent with testing performed at the Power Burst Facility

wherein steam starvation was not measured for comparable steam flow

rates. D-3

The mass flow rate in the simulated LPIS line was derived from the

pressure differential across the calibrated flow orifice. During the time

D-14

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of fission product release and transport, the flow rate was approximately

constant at 0.2 kg/s. This is approximately twice the flow rate predicted

prior to the experiment and, again, is consistent with the

higher-than-predicted flow rate in the center fuel module and with the lack

of steam starvation.

The fluid in the vicinity of the FPMS measurement locations generally

was superheated steam. Figures 0-14 through 0-17 compare the fluid and

saturation temperatures at the Fl, F2, D2, and D3 locations, respectively.

Figures D-18 through D-20 compare the fluid, metal, and saturation

temperatures for the upper, middle, and lower coupons in the Dl deposition

coupon rod, respectively. The fluid temperatures at all three deposition

coupon locations as well as the metal temperatures at the Dl coupon

locations were superheated throughout the period of fission product release

and transport. The fluid temperature at the F2 location is superheated

from approximately 1300 to 1800 s. The fluid temperature at theFl location indicated general superheat with two periods of quench. There

were no perturbations measured by a pressure transducer at the same

location. This implies that the amount of water present in the line was

probably very small, such that steam generation was not measurable. It is

possible that a small droplet of water (in, perhaps, a mixed flow regime)

quenched the thermocouple without affecting the steam or metal temperature.

When the shroud external temperature trip setpoint of 1517 K (2270*F)

was reached, the FPMS and LPIS lines were isolated and the ECCS injection

was initiated. Performance of the ECCS was excellent and the PCS was

rapidly brought to a stable shutdown condition. The low pressure pumps

injected approximately 25 kg of water during the initial 5 s of ECCS

operation. The accumulators injected over 1000 kg of water over the same

time interval. The resulting generation of steam caused the pressure to

increase from 1.2 to over 3 MPa (174 to over 435 psia). This

repressurization caused a temporary cessation of accumulator flow as the

PCS pressure exceeded that of the accumulator. Flow restarted at 1910 and

1940 s from Accumulators B and A, respectively, again causing a PCSrepressurization. This is illustrated in Figure D-21 which compares

D-15

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600

550

v- 500

S-

-- 450La,a-

E 4001--

350

3000

Figure D-14.

500 1000 1500

Time (s)

Comparison of fluid temperature inline with saturation temperature.thermocouple qualification limits)

-400 qj

" 300p--

2O00

100

2000

the F1 aerosol sample(See Appendix I for

-400

-300

2 oE

300 a"

1--

-100

2000

the F2 aerosol sample(See Appendix I for

500

, 450

0

400La)a-

E

350

3000 500 1000 1500

Time (s)

Figure D-15. Comparison of fluid temperature inline with saturation temperature.thermocouple qualification limits).

D-16

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750

v

0

EVD

I.--

625

500

375.

TE-LH00

TE-OLH-004

-. ST-OLH-003

-- -- - -- -- -

-800

-600

0-400 L

a-

EI-

-200

.02500 500 1000

Time (s)1500 2000

Figure D-16.

550

500

450

4-0L,

a)400

EI-.

Comparison of metal and fluid temperatures at the D2deposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).

500

400 ~~

300 oL

0.

E(D

200 1-

100

350

3000 500 1000 1500 2000

T i me (s)

Figure D-17. Comparison of fluid temperature at the D3 deposition couponlocation with saturation temperature. (See Appendix I forthermocouple qualification limits).

0-17

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"- 600

550

L

E soo

450

4000

Figure D-18.a

Soo

.•.-700'--

E 6000

0-

E4500

4000

Figure D-19.

-600

I "I

. 500

E

-300

500 1000 1500 2000Time (s)

Comparison of metal and fluid temperatures at the D1 upperdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).

I I

- TE-5UP-21581TE-5UP-029A

X Saturation temperature

,-800

JA-

,,•0.

L --

E

-400

I !

500 1000 1500 2000Time (s)

Comparison of metal and fluid temperatures at the D1 middledeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).

0-18

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1000

8000

L

5. 700L.S

E oo

0

I-

500

400-0

Figure 0-20.

5.5

57-

4.5

V.' 40~

a. S3.5

.. 3

u 2.5

D. 2

100

I • L

500 1000 1500 2000Time (s)

Comparison of metal and fluid temperatures at the D1 lowerdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).

|I I

"_- PE-PC-002--- PT-P120-029

PT-P120-043

600 ,-,0

Co

a.0.

4200

I !

1800 1900 2000 2100Time (s)

• Comparison of primary system pressure with pressuresmeasured in the accumulators.

1.5

0.51700

Figure 0-21

0-19

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primary system pressure with the pressures measured in the two

accumulators. The damaged core required several hundred seconds of reflood

before all the metal had quenched, but the continued ECCS injection was

able to overcome the core thermal inertia and cool the core. The core

thermal response during reflood is discussed in more detail in Appendix E.

The PCS was recovered to cold shutdown conditions by secondary feed

and bleed operations. The times of significant events during this phase

are indicated in Section 3 of this report. The PCS was filled with

subcooled water and was solid throughout most of this phase. Figures D-22

and D-23 show the primary system pressure and upper plenum fluid

temperature during this phase. There were no unexpected thermal/hydraulic

events which were measured during the posttransient phase.

2. RESPONSE DURING THE POSTTRANSIENT PHASE

There was excessive background contamination at the G1, G2, and G3

spectrometer stations during the early part of the posttransient phase.

For that reason, the plant was maintained in a quiescent state for 44 d (to

allow for background reduction). During this time, data were collected

from the on-line gamma spectrometer systems (14 d) as well as from grab

samples taken from the PCS (44 d) and blowdown suppression tank (liquid

sample, 21 d; vapor sample, 28 d). Those data are discussed in Section 4

of this report.

A calculation of the mass flow in the PCS during the first 40 h of

this phase was made, using the decay heat, reactor vessel temperature rise,

and known fluid conditions. The resultant mass flow was in the range of

2 to 3 kg/s (4.4 to 6.6 lbm/s). This resulted in a loop transit time of

1900 to 2800 s. There were no unexpected events during this phase and thefluid boundary conditions for operation of the data collection systems were

as expected.

1ý .- ý IV

0-20

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15

V)(L

12.5

10

7.5

5

2.5

0

-2000

-1500 0

(j-

a)-1000 L

(0)

a,,

CL

-500

0 500 1000 1500 2000 2500T-i me (s)

3000 3500 4000

Figure D-22. Primary system pressure during the posttransient phase.

800

750

,1-700

CD 650

-O 600L

o-550E

500

450

400

LL-60

40-I--

- 40

0 500 1000 1500 2000 2500Time (s)

3000 3500 4000

Figure D-23. Primary systemphase.

fluid temperature during the posttransient

D-21

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REFERENCES

D-1. G. E. Putnam and B. L. Rushton, Three-Dimensional Pin Power from LOFTCore-I TIP Data, LTR-l1l-87, RE-Z-77-OOI, Sept. 1977.

D-2. J. P. Adams and V. T. Berta, "Monitoring Reactor Vessel Liquid Levelwith a Vertical String of Self-Powered Neutron Detectors," NuclearScience and Engineering, 88, 1984, pp. 367-375.

0-3. D. J. Osetek et al., "Fission Product Behavior During the First TwoPBF Severe Fuel Damage Tests," EGG-M-11284, ANS Topical Meeting onFission Product Behavior and Source Term Research, Snowbird, Utah,July 15-19, 1984.

D-22

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APPENDIX E

CORE THERMAL RESPONSE

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APPENDIX E

CORE THERMAL RESPONSE

The measured temperature excursion began in the upper part of the

peripheral fuel modules at 662 s (on TE-2E08-045) and moved downwards as

the coolant boiled away; the fifteen-inch elevation dried out at about

730 s (on TE-2F07-015); and the ten-inch elevation dried out at about 930 s

(on TE-2G14-Oll). The departure of these temperatures from saturation is

shown in Figure E-1. (The quench of the ten-inch elevation thermocouple

caused by the closure of the PORV can also be seen.) The center fuel

module remained at saturation for a longer time. This is believed to be

due to higher decay heat levels and higher steam flows in this module,

which caused the froth level inside the shroud to be higher than that

outside. The first recorded departure from saturation is at 689 s (on

TE-5M09-042); the 27-inch elevation dried out at about 740 s; and the

thermocouples at the ten-inch elevation in the center fuel module have a

minor [about 10 K (18°F)] departure from saturation starting at 780 s,

followed by a much more significant excursion at 940 s. Thesethermocouples are shown in Figure E-2. The dryout was therefore top-down,

although a few guide tube thermocouples did not follow this pattern.

In order to control the system pressure, the intact loop cold leg

break and PORV were cycled during the early part of the temperature

excursion. The effects of all of these changes (except the opening of the

intact loop cold leg break at 877.6 s, which is followed too closely by the

opening of the PORV at 882 s) can be seen on the heatup rates of some

thermocouples in the core. The greatest effect by far was caused by the

opening of the PORV at 882 s, which caused cooling throughout the core, and

quench in the lower parts of the peripheral modules as the level responded

to the changing pressure gradient in the reactor vessel. In order tobetter see the small changes in temperature gradient caused by the various

events, Figure E-3 shows TE-4G08-021 data after subtraction of the average

rate of increase of 0.538 K/s (O.9684°F/s),

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600

550

450

600

Figure E-1.

600

55-

E 5oo

450 -600

Figure E-2.0L

D.-

I-

I'-,

450O600

Figure E-2.

I•TE-2EO8-045I--- TE-2F07-015 r• -400

m- TE-2G14-011-X Saturation tenperature

5/ II I I..

650 700 750 800 850 900 950 1000Time (s)

Comparison of cladding temperatures at the 1.14-, 0.38- and0.28-m (45-, 15- and 11-inch) elevations in Fuel Assembly 2with saturation temperature during early stages of heatup(600 to 1000 s). (See Appendix I for thermocouplequalification limits).

-600

-550// -500

•, .-. • •'/ 450 C#E

TE-5MO9-042

--- TE-5D09-027 -400. TE-5C12-010

X Saturation temperatureI I I I I I I

650 700 750 800 850 900 950 1000Time (s)

Comparison of cladding temperatures at the 1.07-, 0.69- and0.25-m (42-, 27- and 10-inch) elevations in the center fuelassembly with saturation temperature during early stages ofheatup (600 to 10001s). (See Appendix I for thermocouplequalification limits).

E-2

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160

150

-' 140

L

1300

0.

E 12oI-

110-

100 -700

Figure E-3.

-150

-200

L.-220

0

-240 0LE

- -260

800 900 1000 1100 1200 1300 1400Time (s)

Effect of flow changes on rate of temperature increasemeasured at 0.69-m (27-inch) elevation on fuel rod claddingin Fuel Assembly 4 (700 to 1400 s). (See Appendix I forthermocouple qualification limits).

E-3

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When the guide tube thermocouples at the 27-inch elevation reach about

1050 K (1430°F), which occurred at about 1300 s, the temperature gradient

decreased from about 1.2 K/s to about 0.7 K/s, as is shown in Figure E-4.

This is judged to have been caused by melting of the control rod material

and the resulting adsorption of latent heat. This hypothesis is supported

by the fact that TE-5H08-027 is the only 27-inch thermocouple which does

not exhibit this effect, and this guide tube does not contain a control

rod. An approximate calculation indicates that the latent heat could cause

a temperature shift of 280 K (504°F), which is considerably more than the

observed shift of about 50 K (90°F). That difference can be explained by

the heat input from metal-water reaction, which initiated during this time.

At 1330 s some water appears at the top of module 4, as shown by the

upper plenum thermocouples in Figure E-5. At 1360 s, as shown in

Figure E-6, the lower plenum thermocouples in this module are cooled. Close

examination of the fuel rod cladding thermocouples in module 4 show a very

small effect at about 1360 s, as can be seen on Figure E-3. The cause of

this is not known, but may be due to water, from a presently unknown

source, either dripping from the upper tie plate to the lower tie plate

without touching the fuel rods or running down guide tubes between the tie

plates. Of these two possibilities, the latter is probably more reasonable.

The first recorded (and qualified) rapid temperature rise caused by

the exothermic reaction between the steam and the zircaloy is at about

1430 s on guide tube thermocouple TE-5H08-027. (Thermocouple TE-5EIl-027

was judged to have failed at 1311 s, but the mode of failure suggests that

temperatures reached 1800 K (2780 0 F) at some location in the core by

1381 s.) The rapid temperature rise began from approximately 1400 K

(2060°F). (These figures must necessarily be approximate because there is

no abrupt change to characterize the start of the reaction). The data from

these two thermocouples are shown in Figure E-7.

Because of the relatively few locations in the center fuel module at

which the fuel rod cladding temperatures are measured, it is considered

unlikely that the first occurrence of the rapid reaction between steam and

E-4

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1600

"-" 13006,L.: 1200

,4--0L 1100,6L

E 1oo04)

I--900

800'-,

7001100

Figure E-4.

g00

750

700

* 650L

4o 600L

550E

P-500

450

400 L600

Figure E-5.

/-2000

*1 Li.. 1500 L

E/ ° ._..- *° ,, f•

115 120 125 1300 13010E4010 5010

/~~ Tm (s)' ' I

,'|,I-

fuel assembly (1100 to 1600 s). (See Appendix I forthermocouple qualification limits).

I..

-600 0

CL0 )

4O00

8oo 1000 1200 1400 1600 1800 2000Time (s)

Comparison of fluid temperatures at upper tie plate aboveFuel Assembly 4 with saturation temperature(600 to 2000 s). (See Appendix I for thermocouplequalification limits).

E-5

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550

"- 500

L

E 450I-.

400 -600

Figure E-6.

2500 -

- TE-4LP-003 500Saturation temperature

,450

0

400

0.

35o E

300

I I I I l. .

800 1000 1200 1400 1600 1800 2000Time (s)

Comparison of fluid temperatures at lower tie plate belowFuel Assembly 4 with saturation temperature(600 to 2000 s). (See Appendix I for thermocouplequalification limits).

CL

0

2000

1500

1000

4000

3000

2000 LID.

EQD

10-

-1000

I I I I I 1700 800 900 1000 1100 1200 1300 1400 1500 1800

Time (s)600

Figure E-7. Effect of metal-water reaction on guide tube temperatureincrease at 0.69-m (27-inch) elevation in center fuelassembly (600 to 1600 s). (See Appendix I for thermocouplequalification limits).

E-6

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the zircaloy occurred at a thermocouple location. Once the reaction has

begun in some location, however, the resultant high temperatures will

influence nearby surfaces by radiation and conduction, thus causing an

increase in the rate of temperature rise. It is almost certainly this

effect that is being measured in the center fuel module.

The course of the rapid reaction between the zircaloy and the steam

can be tracked by noting the times at which indicated cladding temperatures

exceed 1800 K (2780 0 F), that being a reasonable indication that the rapid

reaction has occurred. The results, for those thermocouples which had not

failed by that temperature, are shown in Table E-l. The reaction probably

started between the 0.64-m (27-in.) and the 1.07-m (42-in.) elevations.

The reaction then spread across the entire center fuel module at the 1.07-m

(42-in.) elevation between 1480 and 1530 s, and then across the 0.69-m

(27-in.) elevation. The few thermocouples at the 0.69-m (27-in.) elevation

that reacted early (TE-5H08-027 and TE-5H06-027) seem to be exceptions to

the pattern. There is no evidence of a rapid temperature rise due to the

reaction between steam and zircaloy at the ten-inch elevation.

At about 1500 s several instruments show effects of some event that

has taken place. These instruments include the gross gamma monitor (shown

in Figure E-8), upper plenum thermocouples (shown in Figure E-9), the

momentum flux meter in the downcomer (shown in Figure E-l0), and guide tube

thermocouples. A possible initiating event for these effects is the

rupture of the control rod cladding: the sudden release of aerosols could

explain the rapid increase in gamma activity; the flow redistribution

caused by blockage would affect upper plenum temperatures; and the flashing

of water (caused by molten absorber material falling into the water below

the lower tie plate) could cause the downcomer flow.

A more direct indication of the guide tube behavior can be obtained

from an examination of the guide tube thermocouples. Several of the guide

tube thermocouples at the 0.69-m (27-in.) elevation show a discontinuity in

temperature at about 1500 s, as can be seen in Figure E-4. This may be an

indication of their rupture, perhaps caused by absorber material from a

E-7

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TABLE E-1. TIMES FOR THE CENTER FUEL MODULE1800 K (2780-F)

THERMOCOUPLES TO REACH

Identifier

TE-5H08-027TE-5H06-027TE-5112-042TE-5104-042TE-5L09-042TE-5C07-042TE-5M09-042TE-5Dl3-042TE-5H12-027TE-5K05-027TE-5KlI-027TE-5103-027TE-5M06-027TE-5F03-027TE-5C09-027TE-5J07-027TE-5L07-027TE-5C12-027TE-5G12-027

Time to 1800 K Type of thermocouple

1451147514871488149114951513152915361538154515491564158016711674168616861695

S

SS

S .S

S

S

SS

S

SS

S

SS

SSS

Guide TubeInternal CladInternal CladInternal CladInternal CladInternal CladInternal CladInternal CladGuide TubeGuide TubeGuide TubeInternal CladGuide TubeGuide TubeInternal CladInternal CladInternal CladInternal CladInternal Clad

This table lists the times at which the center fuel modulethermocouples reach 1800 K (2780 0 F). Only thermocouples which reach 1800 K(2780*F) before they are judged to have failed are listed.

E-8

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12

10

8

6

40

02600 800 1000 1200 t400 1600 1800 2000

Time (s)

ement of gross gamma activity near reactor vessel headFigure E-8.

1000

'-" 8006)

L

04--

0LU,r

E 600I--

4O0600

Measur(600 to 2000 s).

0U,L-

U,.0-

EI-

Figure E-9.'

800 1000 1200 1400 1600 1800Time (s)

Comparison of fluid temperatures at different horizontallocations on center fuel assembly upper tie plate(600 to 1800 s). (See Appendix I for thermocouplequalification limits).

E-9

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2 E

1000X

0 E

CC

0E0

-21 -1000

-200 0 200 400 600 800 1000 1200 1400 1600 1800 2000

Time (s)

Figure E-IO. Momentum flux in reactor vessel downcomer (-200 to 2000 s).

E-10

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hotter region flowing past the thermocouple. Once again, the effect is

absent from TE-5H08-027, the thermocouple on the guide tube without a

control rod. The temperature at the 0.69-m (27-in.) elevation is

approximately 1200 K (1700°F) at 1500 s, which is lower than the

temperature at which the control rod material release was expected [1400 K

(2060*F)]. However, the temperatures were probably higher at higher

elevations.

Between 1520 and 1680 s some of the thermocouples at the ten-inch

elevation in the center fuel module measured small temperature increases,

some examples of which are shown in Figure E-11. Those may be due to

molten material running down the rod or down a nearby rod.

The peripheral fuel rods exhibited some cooling around the time of

1550 s (as shown in Figure E-12); it was particularly strong further away

from the center fuel module, and also at lower elevations. This may be an

effect of the steam flow that is diverted from the center fuel module,

which is now partially blocked. Added to this steam flow may be steam

produced from hot materials from the center fuel module falling into the

water in the lower plenum.

At 1640 s there was apparently another sudden flow change in the

peripheral modules, and the cooling becomes more severe. The cooling is

strong enough to make some thermocouples exhibit a drop in temperature (see

Figure E-12).

It is possible that there is a connection between the small

temperature increases noted at the ten-inch level in the center fuel module

and the cooling of the peripheral modules, which occurred at the same time.

This would be consistent if the control rod material was running past the

ten-inch level, through the tie plate, and into the water, thus creating

steam, which cools the peripheral modules.

During most of the transient, the thermocouples on the outside of the

shroud increased steadily in temperature from saturation at about 740 s

(940 s for the ten-inch elevation) until about 1700 s. All of the shroud

E-11

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1100

L

E

4..

1000

900

800

TE-5G04-010 1400

1200 L

4W

. IL

1000

300 1400 1500 1600 1700 1800 1900Time (s)

Cladding temperatures at 0.25-m (10-inch) elevation incenter fuel assembly during high temperature stage oftransient (1200 to 2000 s). (See Appendix I forthermocouple qualification limits).

7W01200

Figure E-11.

I

1000

900

I,

.4-

UL

a.E4)

I-

600

700-

600

TE 4I I I I 7 IF

I

1200

1000

0 0 "

BoL.8600 e4"

400500

V- I

600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)

Figure E-12. Comparison of cladding temperature at 0.38-m (15-inch)elevation in Fuel Assembly 4 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouplequalification limits).

E-12

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thermocouples are shown in Figures E-13 to E-16. The thermocouples at the

1.07-m (42-in.) elevation deviated from this a little in that their

temperature rise rates increased at about 1540 s (TE-5E-042 and TE-5S-042)

or 1620 s (TE-5N-042 and TE-5W-042). These changes in rise rate may be

another facet of the effect that caused cooling of the peripheral rods at

about these times, or it may be due to the shroud becoming less efficient

as an insulator. The temperature of the shroud at 1700 s varied from about

8UO K (980°F) at the ten-inch elevation to about 1400 K (2060 0 F) at the

42-inch elevation. At 1700 s several of the shroud thermocouples exhibited

an increase in temperature rise rate, the effect being strongest on the

south side (next to module 2) and particularly on thermocouple TE-5S-0lO.

Also at about 1700 s, the thermocouples near the outside of the

shroud, particularly at lower elevations, began an extraordinary

temperature excursion. By 1780 s (just before reflood) all of the

thermocouples in module 2 near the south wall of the shroud, with

elevations ranging from 0.28 to 1.24 m (11 to 49 in.) reached approximately

1400 K (2060'F). A selection of these thermocouples are shown in

Figure E-17. The shroud wall thermocouples at or below 0.81 m (32-in.)

were cooler than 1400 K (2060°F). The thermocouples near the shroud in

modules 4 and 6 behaved similarly, but not in such an extreme manner. A

comparison of modules 2, 4 and 6 are shown in Figure E-18. (The

thermocouples in the peripheral modules away from the shroud see a little

of a similar effect just before reflood, as can be seen in Figure E-12.

However the time of the temperature rise in Figure E-12 corresponds closely

to the closure of the LPIS line break and the isolation of the FPMS system,

so the effect may be unrelated.)

The cause of the rapid temperature rise is somewhat uncertain. The

exothermic reaction between zircaloy and water is not a possibility because

the initiation temperatures for the rapid rise are too low; nor is

radiation from the shroud wall a possible cause because the wall

temperature is less than that reached by the fuel rod thermocouples at this

elevation. It is judged that the rapid temperature rise is caused by

E-13

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1700

1500

, 13000

L 11004-I

0~-900

E,. 700

-2500TE-5N-010TE-5N-027 ,1TE-5N-032 /TE-5N-042 •20Saturation temperature

-1500 '

-1000 E0

50

800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900T ime (s)

Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-rn (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 8 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouple

qualification limits).

500

300 -I-

600 700

Figure E-13.

1700

1500

IiL.

4-a

U)a.EU)

I-

1300

tIo 40

I-TE-5E--0104--TE-5E-027A

X Sauraiontemperature

4-

X XKX

-2500

-2000

0

- 1500

L

-1000 CLE

900

700

500 -500

300

600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)

Figure E-14. Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-m (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 4 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).

E-14

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1700

1500

1300

CL

E4)

I-

11004

I I I I I I I I ! I !

TE-5S-010TE-5S-027

-- TE-5S-032- TE-5S-042

X Saturation temperature /

/ X XI

-2500

-2000

-1500 L

-1000 0-

I--

- 500

900

700

500,

.0U,600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900

Time (s)

Figure E-15. Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-m (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 2 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).

1700

1500

2 1300

1 100

900-.

,~700

-2500

-2000

-1500 .

-1000 .

E

-500500

300 1600 700

Figure E-16.

800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)

Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-n (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 6 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).

E-15

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1500

1300

-" 1100

L

E 900L.4)a-E 700

I,-

500

3001400

Figure E-17.

2000

f - ).... ...- "

-1500

0CD

L

1000 L0.-EI.-

- TE-2G4-011Soo--- TE-2H14-028 500

TE-2114-039-TE-2H13-049

I I I I I I I I J

1450 1500 1550 1600 1650 1700 1750 1800 1850 1900Time (s)

Comparison of cladding temperatures at 0.28-, 0.71-, 0.99-and 1.24-m (11-, 28-, 39- and 49-inch) elevations in FuelAssembly 2 (1400 to 1900 s). (See Appendix I forthermocouple qualification limits).

I I I I I I

TE-2G14-011- TE-4G14-011 -2000

-- TE-6G14-011

-0

"".1500 0 L

S-I,=

1000 L

E0I--

500

, I I I I II ,

800 1000 1200 1400 1600 1800 2000T I me (s)

Comparison of cladding temperatures at 0.28-m (11-Inch)elevation on fuel rods in peripheral assemblies 2, 4, and6 close to shroud (600 to 2000 s). (See Appendix I forthermocouple qualification limits).

1500

I300

v 1100

900L

E 700F--

500

300 i600

Figure E-18.

E-16

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shunting of the thermocouple leads where they pass through an area of high

temperature (near the top of the core). This is reflected in the

qualification statements for those transducers in Appendix J.

Reflood was initiated at 1782.6 s, and the peripheral modules quenched

very rapidly between 1789 s (on TE-4F07-015) and 1793 s (on TE-2HI5-032 andTE-4114-021). The shroud outer wall also quenched during this period; all

thermocouples on the south side of the shroud failed simultaneously at

1790 s, presumably due to some mechanical strain associated with the

reflood process.

In the center fuel module, only the ten-inch elevation thermocouples

were qualified as accurate through the reflood, and the last of these to

quench was TE-5J07-0lO at 1795 s. The other center fuel module

thermocouples, including those on the upper tie plate, failed at or before

reflood. The nearest thermocouple above the center fuel module that

survived the reflood was TE-5UP-028B, which is shown in Figure E-19. This

thermocouple quenched at 1793 s, then heated rapidly to 1900 K (2960*F)

before quenching again at 1801 s. This gives some evidence that the center

fuel module remained hot for at least a short period after the rest of the

core quenched.

More information on the cooling of the center fuel module can be

inferred from the thermocouples which failed during the transient. One

mode of failure that is expected is for the thermocouples to melt and

rejunction at some hot location that the thermocouple lead passes through.

If this happens, then the thermocouple may still give qualitative

temperature information about the location where the new junction is

formed. Two examples of this appear to be TE-5K1l-027 and TE-5J07-027,

both of which are shown in Figure E-20. Thermocouple TE-5K11-027 was

evidently reading a true temperature because it-quenched at 1880 s and then

remained subcooled, as might be expected after reflood. This is therefore

fairly reliable evidence that some location in the center fuel module was

hot until at least 1880 s. Thermocouple TE-5J07-027 was cooling towardssaturation until 2010 s, at which time the junction apparently broke

E-17

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2000

1500

vL

CD

0

EI,-

!i

I I

TE-5UP-028BSaturationtemperature

1000

3000

-2000

1000 0

E

'D0uu [-- - -

0L1750

I I I I I I I I I

1760 1770 1780 1790 1800 1810 1820Time (s)

1830 1840 1850

Figure E-19. Comparison of fluid temperature at upper tie plate abovecenter fuel assembly with saturation temperature duringreflood (1750 to 1850 s). (See Appendix I for thermocouplequalification limits).

0L.idCD

E.

0

I.-

2500

2000

1500

1000

500

- 4000

-3000

CD-2000

-0a.-

-1000 E"

I--

-0

0 LI600

Figure E-20.

900 1200 1500 1800 2100Time (s)

Effect of material relocation on cladding temperatures at0.69-m (27-inch) elevation in center bundle during reflood(600 to 2100 s). (See Appendix I for thermocouplequalification limits).

E-18

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again. In order to cool so slowly, the thermocouple junction must be well

insulated from the reflood water. It is therefore reasonable to assume

that there is a mass of material in the center of the center fuel module

which is in a compact, and thus difficult to cool, geometry. This stayed

above the saturation temperature for several hundred seconds.

Additional information concerning the configuration of the center fuel

module after the transient is provided by experiments conducted at

Kernforshungs Karlsruhe (KfK) in the Federal Republic of Germany. The rate

of temperature rise in Experiment LP-FP-2 is bounded by the rates measured

in KfK Experiments ESBU-l [2K/s (3.6 0F/s)] and ESBU-2A [0.5/K/s

u.9 0F/s)]. E- The maximum temperatures reached in these experiments were

approximately 2525 K (4086°F) and 2450 K (3951 0 F), respectively. These

maximum temperatures approximate those that are believed to have occurred

in Experiment LP-FP-2, so it is reasonable to believe that some aspects of

the final configuration may be similar. In the KfK experiments, the center

part of the core melted and flowed downwards, liquidifying some of the fuel

pellets; the test bundle was almost completely blocked by the resolidifiedmelt. A similar configuration is expected to have resulted in this LOFT

experiment.

E-19

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REFERENCES

E-1 S. Hagen and B. J. Buescher, "Out-of-Pile Experiments on PWR FuelBehavior Under Severe Accident Conditions," BNES Conference on NuclearFuel Performance, Stratford-on Avon, England, March 25-29, 1985.

E-20

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APPENDIX F

COMPARISON OF THERMAL/HYDRAULIC DATA

WITH PREEXPERIMENT CALCULATIONS

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APPENDIX F

COMPARISON OF THERMAL/HYDRAULIC DATA

WITH PREEXPERIMENT CALCULATIONS

This appendix presents comparisons of the data with preexperiment

thermal/hydraulic calculations for the entire system using the computer

code RELAP5/MOD2,F-I'a and for the core thermal response using a special

version of TRAC-BDi. F 2 ,b The major features of the codes and the

respective input model descriptions are summarized in Section 1. Several

sets of calculations, employing a range of values for initial and boundary

conditions, were performed as part of the planning analyses and are

described in some detail in the Experiment Prediction Document (EPD).F- 3

None of the calculations employ initial and boundary conditions identical

to those that applied in the experiment itself, since the experiment

operation was finally specified after consideration of the results of the

planning analyses. The comparisons described in this section relate to the

most recent calculations, on which the experiment operation was based and

which provide the closest agreement (among the calculations) with the

initial and boundary conditions. The effect of variations in initial and

boundary conditions is discussed in the EPD in relation to sensitivity

studies based on a range of operating conditions. The comparisons between

predictions and measured data are described in Section 2.

a. This analysis was performed using RELAP5/MOD2 Cycle 36, a productionversion of the RELAP5/MOD2 code which is filed under INEL Computer CodeConfiguration Management (CCCM) Archival Number A05844.

b. TRAC/BDI modified Version 8, INEL CCCM Archival Number F01498.

F-l

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1. DESCRIPTION OF CALCULATIONAL METHODS

This section presents comparisons of the experiment data with

preexperiment thermal/hydraulic calculations using RELAP5/MOD2 and TRAC-BDI.

1.1 RELAP5/MOD2 Computer Code

RELAP5/MOD2 is an advanced, best-estimate computer program developed

at the Idaho National Engineering Laboratory (INEL) for the analysis of

Loss-of-Coolant Accident (LOCA) and other PWR transients. The specific

application of the code to the Experiment LP-FP-2 prediction is discussed

in this section.

1.1.1 RELAP5/MOD2 Description

RELAP5/MOD2 employs a finite-difference fluid cell representation of

the primary and secondary coolant systems. The six-equation hydrodynamic

formulation employs separate equations to describe the conservation of

mass, momentum, and energy for liquid and steam within each fluid cell.

The description of the hydrodynamics is essentially one dimensional within

each fluid cell. The inclusion of a simplified treatment of the

conservation of momentum in the direction perpendicular to the main stream

flow, where cross flow occurs between parallel volumes and in branches,

brings a special treatment of two-dimensional effects.

Descriptions of the hydrodynamics of choked flow, stratified flow, and

abrupt area changes are carried out with special process models. Special

models are included for simulation of particular components, such as pumps

and accumulators. Flow-regime-dependent constitutive equation and heat

transfer packages are incorporated to complement the hydrodynamic

description. Conduction of heat within metalwork and fuel rods is

calculated with a one-dimensional (two-dimensional in fuel cladding for a

reflooding simulation) finite difference formulation. An extensive control

and trip logic capability is built into the code.

F-2

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1.1.2 RELAP5/MOD2 Input model for Experiment LP-FP-2

The nodalization used in RELAP5/MOD2 for this calculation is based ona

a standard LOFT nodalization, with changes that were necessary to

represent the particular system configuration for Experiment LP-FP-2.

Several updates were made to overcome code difficulties encountered or to

better represent phenomena expected to occur during Experiment LP-FP-2.

The final version of the nodalization model used for the calculations

presented in this report is shown in Figure F-l.

The nodalization differs from the standard RELAP5 LOFT model in the

following aspects:

1. The broken loop hot leg pump and steam generator simulators and

the quick-opening blowdown valve were replaced by a pipe

simulating the LPIS break line with two valves attached at both

ends.

2. The quick-opening valve on the broken loop cold leg and its

connection piping to the cold leg were deleted. The broken loop

cold leg is a dead end volume.

3. The reactor vessel was extensively remodeled to represent the

special core configuration and to better simulate the flow

splitting and mixing. Special emphases were given to peak

cladding temperature behavior in the center and peripheral fuel

assemblies and also to the thermal responses of the guide tubes,

control rods and thermal shroud surrounding the center fuel

assembly.

a. The standard LOFT input model Version 131 was used as the basis for theinput deck for Experiment LP-FP-2. The model is continually being updatedand improved. However, complete traceability of each version is maintainedin the model and by the LOFT Program Division.

F-3

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Steam valve leak--,I A

r I - 116 114 11--2 Intact loop hot leg 720 261.€= Broken

1 Pum Intact loop cold log 03 22Boe

- 13S•T 14i4 7 32 118 P16 p ,10 1 1 2180 185 2 230 72236

Fir 160. 182 d"[-1120 • .. BSB-3 break -7

Accumulator 222 23

q-1- 600 214 215

1 1zI zII I

Figure F-1. RELAPS/MOD2 nodalization diagram.

( ( (

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4. The cross flow model was applied to the junctions connecting the

cold legs to the vessel and to the junction connecting the

pressurizer to the intact loop hot leg.

5. Although the reflood phase was not simulated, the emergency core

coolant (ECC) system and its injection locations (one into the

lower plenum and the other into the downcomer) were shown for a

complete system nodalization.

6. The blowdown piping was attached to the ILCL leg with a

nodalization similar to that used in the Experiment LP-SB-3

prediction calculation.F4

The input model contains a total of 43 fluid cells for the vessel and

95 cells for the remainder of the primary and secondary systems. Detailed

models were developedF-5 to better simulate the flow splitting (from the

lower plenum into the channels representing the peripheral and center

assemblies) and flow mixing (in the upper plenum). Seven fluid cells are

used to represent the lower plenum and lower core support structure. The

flow splitting is represented by cross flow junctions between the cells at

three elevations below the core. The diffuser plate is specifically

modeled by the junctions connecting Volumes 222 and 224 and Volumes 223

and 225. The detailed upper plenum model specifically considers the mixing

in the upper end box represented by Volumes 240 and 241 with a cross flow

junction between these volumes. The mixing between the flows from the

center assembly and the peripheral assemblies below the 5.69-m (224-in.)

elevation, as referenced to the bottom of the reactor vessel, is also

modeled by the cross flow junction between Volumes 245 and 246. No mixing

is allowed between Volumes 252 and 253 due to the geometry of the upper

plenum between the 5.69-m (224-in.) elevation and the nozzle level.

The nozzle area is modeled by four fluid cells. The hot legs are

connected to Volume 250. The split downcomer approach was chosen

especially to simulate the effect of liquid level in the downcomer on the

void distribution in the core at the time of ILCL break valve closure. The

two downcomer channels are horizontally connected with cross flow junctions

F-5

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at five different axial elevations. The core is divided into two channels,

each containing six axial fluid cells of equal length. The channels are

hydraulically isolated. The thermal shroud, which is represented by a heat

structure, is the thermal link between the two core channels. The leak

path between the upper downcomer annulus and the upper plenum is modeled by

a cross flow junction connecting Volumes 700 (upper downcomer annulus)

and 256 (upper portion of the nozzle area above the peripheral bundles).

The eight hot rods in the center assembly and the remaining 9.74%

enriched fuel rods are represented by two heat structures. The 10 guide

tubes and 11 control rods are separately represented by two heat

structures. The fuel rods in the peripheral assemblies are represented by

two heat structures. One structure represents the four rows of rod groups

surrounding the thermal shroud outer surface. The remaining fuel rods are

represented by the second heat structure. The control rods in the

peripheral fuel assemblies are not simulated.

Other input features include the containment modeled as a

time-dependent volume. The decay heat power was based on ORIGEN2

calculations specifically performed for the peripheral and center fuel

bundles. A burnup of 500 MWD/MTU was used for the calculations. The

RELAP5/MOD2 code does not include a metal-water reaction model. However,

as the fuel rod cladding temperature rises, metal-water reaction becomes an

increasingly important, and eventually the dominant heat source.

Therefore, a metal-water reaction model was included using the RELAP5

control system.

Heat generation was calculated using the Cathcart-PawelF-5 model for

cladding temperature in the range 1273 to 1853 K (1832 to 2876°F) and the

UrbanicF-5 model for cladding temperatures above 1853 K (2876°F). A

steam limitation model was included to account for the steam availability

for the reaction. Weaknesses as of the model are (a) the energy generated

by the metal-water reaction is deposited in the body of the fuel rods

rather than in the cladding surface, (b) no hydrogen is generated, and

(c) the center assembly flow is required to be positive. The metal-water

reaction was also calculated on the cladding of the guide tubes and the

F-6

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inner surface of the thermal shroud. These models were included in the

input deck and can be seen in Appendix B of the LP-FP-2 Experiment

Prediction Document. F-3

The calculation was initiated from a power level of 25 MW, which

corresponds to a calculated peak power density in the 9.74 wt% enriched

fuel rods of approximately 38.8 kW/m (11.8 kW/ft). The important initial

conditions are presented in Table F-1. All of the initial conditions used

in the calculation are close to the experimental values, with the exception

of core bypass. The modeled bypass results in a calculated initial bypass

flow of about half the estimated value for the experiment; however, this

exception is judged to have only a minor effect on the LP-FP-2 transient.

1.2 TRAC-BDI Computer Code

The TRAC (Transient Reactor Analysis Code) is an advanced best

estimate system analysis computer program designed for the analysis of

postulated accidents in light water reactors. TRAC-BDl is designed

primarily for the simulation of design basis LOCAs and transients in

boiling water reactors (BWR). This version of the code is suitable for

analysis of the thermal behavior of the LOFT core with two separated flow

channels during Experiment LP-FP-2.

1.2.1 TRAC-BOl Description

Unique features of the code include (a) a full nonhomogeneous,

nonequilibrium two-fluid thermal-hydraulic model of the two-phase flow in

all portions of the BWR system and (b) a detailed model of BWR fuel

assemblies, which includes a radiation heat transfer model for thermal

radiation between multiple fuel rod groups, inner surface of the fuel

channel wall, and liquid and steam phases within the bundle.

Extensive development work was carried out at the INEL to improve the

TRAC-BDI computational capabilities to more closely predict the LOFT core

thermal response during Experiment LP-FP-2. The detailed description of

F-7

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TABLE F-1. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2

Parameter

Core power

Maximum linear heatgeneration rate

Core AT

Primary systempressure - hot leg

Intact looptemperature

hot leg

cold leg

Primary coolant massflow rate

Units

MW

kW/mkW/ft

KOF

MPapsia

KOF

KOF

kg/salbm/h

Measured Calculated

26.8

42.612.97

11.721.1

14.982173

571.6569.2

559.9548.2

25.0

38.811.8

9.917.8

14.902161

567.8562.4

557.9544.5

475.03.77x106

476.83.78xi06

Total bypass flow kg/s a 28.0Ibm/h 2.2xi05

Pressurizer liquid m 1.12 1.12level in 44.1 44.1

Steam generator MPa 6.38 6.06secondary pressure psia 925 879

a. Not measured but estimated to be approximately 63 kg/s (5 x 1O5 ibm/h).

F-8

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the models (the first five items indicated below) developed has already

been published in Reference F-6. The models which are important for this

calculation are:

o LOFT full core radiation model

o Thermal shroud model, which thermally links the two separated

flow channels

o Extended metal-water reaction model

o Steam limitation model

o Conductance across the gap between the guide tube and control rod

0 Capability of starting post dryout heat transfer at a given time

and axial elevation, as defined in input.

The sixth model indicated above is built to start the postdryout heat

transfer calculation at a given axial elevation and a time. This is a

necessity due to (a) deletion of several code models (as indicated below)

which must be used for a boil-off simulation and (b) the simple input model

developed for the simulation which allows only a heatup calculation. The

model employs a 20,000 W/m2K liquid heat transfer coefficient for the

nucleate boiling heat transfer regime so that the initial temperature at

that elevation can be maintained steady until the transient time passes the

given time. The code is then allowed to pass into the postdryout heat

transfer calculation.

The metal-water reaction model developed considers the reaction

occurring on the rod groups simulating fuel or control rods and on the

inner surface of the thermal shroud. The lack of heat source due to

metal-water reaction on the outer surface of the shroud will yield a

slightly lower temperature excursion on this surface and also on the

peripheral bundle fuel rods surrounding the thermal shroud because of the

radiation heat loss to the shroud.

F-9

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In addition to the models above, numerical stability of the

calculation was further enhanced by improving the calculational strategy of

the code.

All of the code models that were not of use for this analysis, such as

jet pump model, valve model, etc., were removed to offset the extra

dimensions required for the radiation model. New material properties were

added to analyze the control rods, guide tubes, and the thermal shroud

responses. This special version of the code, called TRAC-LOFT, was used to

calculate the thermal behavior of the LOFT core.

1.2.2 TRAC-LOFT Nodalization for Experiment LP-FP-2

The TRAC-LOFT nodalization used for this calculation is presented in

Figure F-2. The LOFT core is represented by two CHAN components (a special

TRAC model for the simulation of a BWR assembly), one representing the

center assembly, and the other representing the peripheral assembly. The

thermal shroud is modeled by the wall of the CHAN component representing

the center assembly. Boundary conditions at the outer surface of the

thermal shroud wall are taken from the CHAN component representing the

peripheral assembly. Both CHAN components are divided into six equally

spaced axial volumes for consistency with the RELAP5/MOD2 nodalization.

Two flow boundary conditions (FILL components) simulating the steam flow

entering the center and peripheral assemblies are separately modeled. The

outlet flows from the CHAN components are mixed in a TEE component. A

pressure boundary condition at the core exit is provided with a BREAK

component attached to the TEE component. The wall of the peripheral

assembly CHAN component represents the core filler plates and the flow

shroud.

The fuel rods, control rods, and guide tubes are represented by rod

groups arranged by rod powers and geometry. The individual members of a

rod group are assumed to be equivalent to all other members of the rod

group. The model used for this analysis employs three groups for the fuel

rods in the center assembly. These additional groups are used for the

center assembly guide tubes and control rods as follows: the center guide

F-1O

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LOI-KMI116-03

Figure F-2. TRAC-LOFT nodalization.

F-li

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tube, by one group;the other 10 guide tubes, by one group; and the guide

tubes with the control rods, by one group. The peripheral assembly rods

are grouped into nine rod groups according to their radial peaking

factors. Due to excessive computer memory requirements, the guide tubes

with and without the control rods in the peripheral assemblies are not

modeled. The rod grouping established for the center and the peripheral

assemblies are presented in Figures F-3 and F-4, respectively. In addition

to the thermal/hydraulic boundary conditions required from the RELAP5

calculations, an adiabatic boundary condition is assumed on the outside

surface of the wall belonging to the peripheral assembly CHAN component.

This assumption added some conservatism to the predicted temperature

transient in the peripheral assemblies. A listing of the TRAC-LOFT input

model is provided in Appendix C of the LP-FP-2 Experiment Prediction

Document. F-3

F-12

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C (

S66 6666X66633333_ 3333663 3 3333 3 6

63311111 336331 1T1 13366 _311 :iT3 6

66331 1 11 13363311111 1 33

63 33333 3663333 33336

66 666 M 66

l W E-6] Fuel rods(24) (48) (28)

MM Guide tubes(1) (9)

• Control rods(l1)

"11

(A

LOI-KM116-04

Figure F-3. TRAC-LOFT center assembly rod grouping.

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Figure F-4. TRAC-LOFT peripheral assembly rod grouping.

F-14

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2. COMPARISON BETWEEN PREDICTION AND EXPERIMENT DATA

This section presents a comparison between pre-experiment predictions

and experiment data. The RELAP5/MOD2 and TRAC-BOI calculations, described

in some detail in the Experiment Prediction Document (EPD), were terminated

shortly prior to reaching the criterion for initiation of reflood.

The initial power of 25 MW used in the calculations is sufficiently

close to the experimental value of 26.8 MW to provide meaningful comparison

with the data. Other differences between the calculated and actual

operating conditions include a slightly lower primary system initial

temperature and secondary side pressure, a different experiment initiation

sequence and trip times for the calculation , and reopening of the cold leg

break and operation of the PORV to assist the depressurization. The

probable effect of these differences on the calculations are considered in

the discussion that follow. Comparison between the predicted and measured

data for system hydraulic response is presented in Section 2.1. A

discussion of the comparison for core thermal response is presented in

Section 2.2.

2.1 Comparison of System Hydraulics

The measured and predicted secondary system pressure during the first

400 s are shown in Figure F-5. The measured steam generator secondary

pressure, after termination of feedwater and steam flows, increased to the

main steam valve cycling setpoint of 7.11 MPa (1031 psia) at 56 s compared

with 50 s predicted. The pressure increase and time of steam valve cycling

were correctly predicted, with some minor differences due, possibly, to the

slightly different initial conditions and experiment initiation sequence.

The secondary system continued to act as a heat sink until the primary

pressure had dropped below the secondary pressure. This was predicted at

224 s compared with the observed time of 260 s. Figures F-6 and F-7 show

the measured and predicted secondary side pressure for the entire

transient, and the collapsed level in the steam generator secondary. The

rates of depressurization and liquid depletion were overpredicted due to an

F-15

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8 I I I I I I

7

00.v

L

C-5

-- PT-PO04-O1OA--- RELAP5/MOD2

# I I I I I

-1000 -"0Ut)-.7

-900:3'I,

-800 6

a)

-700

-600

-1100

40 50 100 150 200 250

Time (s)300 350 400

Figure F-5. Secondary system pressure (0 to 400 s).

8

7a

C-,

6

:3_

5

4

PTPO4-10

RELAP5/MODLJ 1100

-1000

-900

-800

-700

-600

aU,

0.

4)L.:30U,U)1~

0~

0 400 800 1200Time (s)

1600 2000

Figure F-6. Secondary system pressure (0 to 2000 s).

F-16

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C" 2.5 o

.7

3 .5I I I I I I I I I

-200 0 200 400 600 800 1000 1200 1400 160o 1800Time (S)

Figure F-7. Secondary system liquid level.

F-17

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overestimate, based on data from Experiment LP-SB-3, of the leak rate

through the main steam valve. The leak rate tends to differ each time the

the valve is cycled and can vary by a factor of two or more.

Figure F-8 shows the measured and predicted primary system pressures

for the first 400 s of the transient. Both curves show a slight drop in

pressure following scram and a subsequent rapid decrease down to saturation

pressure following break initiation. A minor interruption in the early

depressurization observed in the data at 37s does not correspond to any

operational or measured event. Examination of the predicted cold leg breakfluid conditions indicates the possibility that liquid that was subcooled

in the cold leg flashed to steam as it depressurized while accelerating

through the break line. The end of subcooled blowdown was predicted at

42 s, compared with 53 s indicated from measurements. A slightly lower

pressure was reached in the calculation due to the slightly lower initial

fluid temperature. The predicted pressure response agreed well with data

for the period until initiation of the LPIS line break at 220 s, except for

minor differences that were a direct result of different initial conditions

and initiation sequence. Figure F-9 compares the measured and predicted

pressure for the transient to 2000 s. In contrast with the good agreement

for the period prior to LPIS line break initiation, the subsequent

depressurization rate was considerably overpredicted. The LPIS line and

break characteristics had previously been considered to be a major source

of uncertainty. An attempt was made to estimate the effect of the

uncertainty by performing a sensitivity calculation with the break flow

areas reduced by 30%. This provided slightly better agreement, but still

overpredicted the depressurization rate. Because the cold leg break area

was also reduced, and the decay heat levels were considerably higher (due

to the initial power of 33 MW used), the depressurization rate should have

been underpredicted given reasonably representative modeling of the LPISline and break. From this it is concluded that the straight pipe volume

used to represent the LPIS line, which in reality contains numerous bends,

does not provide a realistic representation of the line. The closure of

the cold leg break resulted in almost no change in the predicted pressure

response, yet the data show that this action terminated the

F-18

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15

a-

a.

10

5

0 I I

0 50 100 150 200 250Time (s)

0

0.

'I,1...:3

In0La-

300 350 400

Figure F-8.

15

a-

CL

Primary system hot leg pressure (0 to 400 s).

II I

PE-PC-002 2000--- RELAP5/MOD2

(25 MW, 100 pct break flow area)-- RELAP5/MOD2

(33 MW, 70 pct break flow area)1500 0

-1000

-50L

• \ -500

500 1000 1500 2000Time (s)

Primary system hot leg pressure (0 to 2000 s).

10

5

00

Figure F-9.

F-19

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depressurization until the cold leg break was reopened and the PORV was

operated. It is postulated that the complicated network of bends in the

LPIS line resulted in a higher flow resistance under single phase

conditions and also inhibited the draining of liquid from the line under

two phase conditions. There is indication from measurements of fluid

temperature that the LPIS line was not completely drained until after about

1200 s. The latter effect differs from the prediction in which the LPIS

line was completely void after about 350 s. The venting of steam,

calculated by the code, would not readily take place with liquid remaining

in the line. The higher system pressure observed and the use of the PORV

and cold leg break to effect depressurization affects all the comparisons

of system hydraulics and core thermal response beyond 300 s.

The fact that the primary system pressure was much higher during the

heat up and core damage phase means that there was a much greater driving

head to sustain the break flow. The measured and predicted differential

pressures between the PCS and BST are compared in Figure F-l0 for the

period from 1200 to 1700 s. The measured values for the pressure drop were

in the range 1.0 to 1.3 MPa (145 to 189 psia), compared with predicted

values of 0.15 to 0.25 MPa (22 to 36 psia). The LPIS line flow calculated

from measured variables and the predicted flow are compared in Figure F-ll

for the same time period, for which the flow of single phase vapor was both

predicted and indicated by measurement. The experimental values for mass

flow were approximately twice the predicted values. This result is

consistent, qualitatively at any rate, with the fact that the measured PCS

depressurization rate was about 2.5 times the predicted rate during this

time frame. The rate of vapor generation is approximately proportional to

the depressurization rate for a given liquid mass, provided the heat input

from the fuel and metalwork is small, as was the case with all or most of

the core uncovered. Also, the estimated steam flow rate in the center fuel

assembly of 0.04 kg/s, obtained from analysis of the core thermal response

presented in Section 3 of this report, exceeds the predicted center

assembly flow by about a factor of 2.5. Despite the mass flow in the LPIS

line being higher than predicted at this time, the momentum flux, as

F-20

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Ký1 A 2

1.75

1.50

T

0

CL

1.25

1

0.75

0.50

0.25

III I

RELAP5/MOD2 -2:

21

-5-

i I i 1 10

5o0

00

0.

0~

01200 1300 1400 1500

Time (s)1600 1700

Figure F-1O. Pressure drop along LPIS line.

0.5

0.4 1-

(#7

0

M

0.3

- -- RELAP5/MOD2)-1.0

L0.8

E0.6

0

0.4

0.2

0.0

0.2

0.1 t

0.01000 1200 1400 1600

Time (S)1800 2000

Figure F-11. LPIS line mass flow rate.

F-21

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calculated from measured variables was similar. Therefore, the effective

flow resistance between the PCS and BST was greater than the predicted

value by a factor of about six.

The areas of agreement and disagreement between the predicted and

observed system hydraulic response indicate that most of the important

features of the primary and secondary systems were correctly modeled, the

major exception to this being the LPIS line flow characteristics. Some

differences between predicted and measured data are a result of differences

in operational initial and boundary conditions.

2.2 Comparison of Core Thermal Response

The major objectives in performing the pre-experiment analyses were

(a) the assessment of the achievability of the required hydraulics and core

thermal response and (b) of maximizing the likelihood of achieving them.

In order to make as good an assessment as possible of the core thermal

response, system calculations were carried out using RELAP5/MOD2, in

conjunction with estimates for blockage in the center fuel assembly. Known

limitations in the capability of RELAP5/MOD2 to model the core thermal

response with the necessary accuracy and detail meant that these objectives

could not be met by using RELAP5/MOD2 alone. Therefore the core hydraulic

conditions calculated by RELAP5/MOD2 were input into a special version of

TRAC-BDI. Certain aspects of the RELAP5/MOD2 calculated hydraulics

(negative core flow at certain times) were judged to be unphysical and were

suppressed in the TRAC-BDI input in order to seek a more realistic

representation. The comparisons presented in this section include both the

RELAP5/MOD2 and TRAC-BDI calculations. However, the comparisons with

RELAP5/MOD2 are considered to be less meaningful and of limited

usefulness. Differences between experiment and calculation in respect of

operation, and in the primary system hydraulics have a bearing on any

conclusions that may be drawn from the comparisons and should be borne in

mind by the reader. The reader should also recall that many of the

cladding thermocouple readings are not qualified during the portion of the

transient at which high temperatures occurred. The time after which the

data is no longer qualified is indicated on the figures.

F-22

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Figure F-12 shows the progression of core uncovery in the center and

peripheral fuel assemblies, as indicated by the observed initiation of heat

up and as predicted by RELAP5/MOD2. The times of initiation of uncovery

and for uncovery of the whole core were quite well predicted, remarkably

so in view of the degree of disagreement in the system hydraulics. The

data did not, however, exhibit the significantly later uncovery of the

center fuel assembly (compared with the peripheral assemblies) that was

predicted. The observed progression of uncovery also differed from

prediction in the following respect. The cold led break was closed in the

experiment at 735 s, compared with 1007 s in the calculation. Although the

the predicted and observed rates of uncovery are similar up to 735 s, the

data show an increase in the core uncovery rate after 735 s down to the

0.38 m (15 inch) elevation, followed by a very slow uncovery shortly

thereafter. It is hypothesized that closure of the cold leg break, in

terminating the system depressurization, caused a sharp reduction in the

rate of vapor generation and thereby brought about a total or partial

collapse of the froth level in the vessel. The system pressure then

remained almost constant until the cold leg break was reopened at 878 s, sothat the continuation of core uncovery until then was solely dependent on

heat input from the fuel and metalwork. As a result, the remainder of the

core uncovered extremely slowly, 180 s elapsing between the initiation of

heat up at the 0.38- and 0.28-m (15- and 11-inch elevations).

Figure F-13 presents the measured cladding temperatures at the 0.25 m

(10 inch) elevation in the center fuel assembly with the prediction at the

nearest modeled location. The average temperature rise rate until 1700 s

was observed to be about 0.5 K/s (0.9°F/s). The rise rate was

overpredicted for most of this period in both calculations, TRAC-BDI

providing the better agreement with 0.7 K/s (l.3 0 F/s) compared with the

RELAP5/MOD2 prediction of 1.0 K/s (l.8 0 F/s). The better agreement using

the TRAC-BDI was due in part to the effect of finer (13-node) axial

nodalization on fluid conditions. The local power factors were similar in

the two calculations. The overprediction was contrary to the fact that the

modeled decay heat level in this node was lower than that which existed at

the measurement location because of the lower elevation of the node used in

the model. The underprediction of mass flow of steam through the core is

F-23

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ILCL closed (736 a)

0W

w

160

140

120

100

80

60

40

20

0

N( 4"

>0(

.

I II IILCL open (878 $) PORV open (882 s)

&I

x

x

v Center measured* Center calculatedx Peripheral measured* Peripheral calculated

60

50

40

30

20

10

0)00

0

S,W,

10600 700 800Time (s)

900

LIIO-KM135-01

Figure F-12. Progression of core uncovery.

.- I-

E E'0

400 ' I -600 800 1000 1200 1400 1600 1800 2000

T i me (s)

Figure F-13. Fuel rod cladding temperature in center fuel assembly,0.25-m (10-in.) elevation. (Thermocouple qualifiedthroughout). ,

F-24

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believed to have resulted in an underprediction of heat transfer

coefficient. The observed increase in temperature rise rate at 1700 s

occurred at too low a temperature, about 900 K (1161°F), to be the result

of rapid metal-water reaction at this location and was not predicted. The

observed behavior may be the result of thermal radiation from high

temperature material at higher elevation or to material relocation.

Neither thermal radiation in the axial direction nor the direct effect of

material relocation on local temperature is modeled.

Figure F-14 compares the fuel rod cladding temperature measured at the

0.69 m (27 inch) elevation in the center assembly with the corresponding

predicted temperatures. Good agreement with the initial heat up rate of

2.2 K/s (4.0 F/s) was obtained in both calculations, during the period

prior to PORV operation and reopening of the cold leg break. The observed

temperature rise rate then decreased due, apparently, to flashing of liquid

in the lower plenum induced by the depressurization. The rise rate was

then overpredicted for the remaining period until about 1630 s when the

metal-water reaction initiated at this elevation. The TRAC-BDI calculation

achieved good agreement with the rate of temperature rise during the time

of metal water reaction, despite the fact that the heat up rate was

observed to increase from a lower temperature than predicted. This, again,

may be due to processes not calculated.

Figure F-15 shows the measured fuel rod cladding temperature at the

1.07 m (42 inch) in the center assembly with the nearest corresponding

calculated temperatures (0.84- to 1.12-m (33- to 44-inches) elevation).

The average rate of temperature rise was observed to be about 1.3 K/s

(2.3 F/s) until 1450 s ( after which the temperature increased very rapidly

due to the metal-water reaction). To this'point in time the temperature

increase was fairly accurately predicted in both calculations despite the

difference in system hydraulic conditions between calculation and

experiment. The observed temperature rise rate increased rapidly after

1450 s, at which time the cladding temperature was about 1500 K (2240°F),

to about 22 K/s (40°F/s). Only a small increase in temperature rise rate

was predicted because the comparatively low mass flow rate in the center

assembly, about 0.015 kg/s (0.033 Ibm/s), resulted in steam limitation at

F-25

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2500

1500

.4,-

L

CD 10000.

EIV

500

0600

Figure F-14.

3000

2500 L- 2000CD

i 1500

E 1000I--

CD

-2000 =

-1000 E

0.0

II I ,I I I

800 1000 1200 1400 1600 1800 2000Time (s)

Fuel rod cladding temperature in center fuel assembly,0.69-m (27-in.) elevation. (Thermocouple qualified to1720 s).

w

a,

-4-

0L.a,a.2a,I-

500

0Boo

Figure F-15.

800 1000 1200 1400 1600 1800 2000Time (s)

Fuel rod cladding temperature in center fuel assembly,1.07-m (42-in.) elevation. (Thermocouple qualified to1510 s).,

, F-26

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this elevation. The maximum cladding temperatures that were measured

occurred at this elevation, whereas the maximum predicted temperatures

occurred at the 0.56- to 0.84-m (22- to 33-inch) elevation.

The maximum cladding temperatures measured in the peripheral fuel

assemblies occurred at the 0.66 m (26 inch) elevation on fuel rods adjacent

to the insulating shroud. Figure F-16 shows the temperature history

recorded by thermocouple TE-2H15-026 together with the corresponding

temperatures predicted by RELAP5/MOD2 and TRAC-PDI. The predictions were

in good agreement with the data until about 1600 s, after which the

cladding temperature increased more rapidly. A similar temperature rise

was observed on many of the peripheral fuel rods and was particularly"

noticeable at the lower elevations. This phenomenon was not calculated,

and a completely satisfactory explanation has not, as yet, been found. It

is possible that the observed behavior is the result of the thermocouple

forming a new junction at a hotter location. The temperature measured on

the outer wall of the shroud at the location close to TE-2H15-026 and the

temperature calculated by TRAC-BDI are shown in Figure F-17. The

calculation using RELAP5/MOD2 underpredicted the shroud temperature

measured at this location and also those measured at the 0.81 m (32 inch)

1.07 m (42 inch) elevations due to the lack of a model for thermal

radiation, an important mechanism controlling the temperature rise of

unheated structures. TRAC-BDl gave much better agreement, with a slight

overprediction in temperature rise rate for most of the transient. The

increase in measured temperature rise rate may have been due to

deterioration in shroud insulation and was not calculated. Both data and

calculation with TRAC-BOl show that the peripheral fuel rod temperatures

tended to be somewhat higher than the shroud temperatures, except at the

elevations at which the highest temperatures occurred in the center fuel

assembly. Heat conduction through the shroud raised the outer wall

temperatures at these elevations to above those of the nearby peripheral

assembly fuel rods. The predicted relationship between the center and

peripheral fuel rod temperatures and the shroud temperatures was in good

agreement with the data. As a result, the time above 2100 K (3321'F) in

the center bundle (about 270 s) was very close to expected time of

F-27

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2000

-S

5-

V

L4-

L.4~a-EV

I-

1600

1200

800

400 1--600

Figure F-16.

-. 2000

-1500 LW0.

E1000

"500

800 1000 1200 1400 1600 1800 2000Time (s)

Fuel rod cladding temperature in peripheral fuel assembly,0.66-m (26-in.) elevation. (Thermocouple showed possibleshunting after 1700 s).

/,/•2000- TE-5S-027 /- RELAP5/MOD2 /

TRAC-BD1 -

1500

1000

CL•- -1000 0, oa.

-. E

500

I I I I I _ _ _ _

800 1000 1200 1400 1600 1800 2000T i me (s)

17. Shroud outer wall temperature at 0.69-m (27 in.)elevation. (Thermocouple qualified to 1790 s).

-S

VI.-

.6-a1~00~E0

I-

1400

1200

1000

80

I.

600

400600

Figure F-

F -28

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280-300 s based on the predictions. The programmatic experiment

termination criterion was reached on the shroud outer wall temperature, as

had been predicted.

The core thermal response during Experiment LP-FP-2 was, in general,

fairly accurately predicted by TRAC-BOI for the period prior to metal-water

reaction. Fairly good agreement was also obtained for the period during

the metal-water reaction, within the limited capability of the codes to

model the processes that take place at the high temperatures. However, it

should be remembered that many of the cladding thermocouple readings are

not qualified for this latter portion of the transient. In view of

discrepancies between calculated and measured system hydraulics, the extent

of the agreement is somewhat surprising and may be to some extent

fortuitous. Several aspects of the predictions that differ from the data

can be traced to known limitations in the modeling, for example thermal

radiation in the axial direction and effects of material relocation. Other

areas of disagreement result from differences between the predicted and

measured system hydraulics. Of these, the most significant effect was the

prediction of steam limitation at the higher elevation, in contrast to the

observed behavior. Also, the temperature rise rate prior to metal water

reaction tended to be slightly overpredicted at many locations due,

apparently, to the predicted steam flows being lower than those obtained in

the experiment.

F-29

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REFERENCES

F-1. V.H. Ransom and R.J. Wagner, RELAP5/MOD2 Code Manual Volume 1:Code Structure, System Models, and Solution Methods, EGG-SAAM-6377,April 1984.

F-2. J. W. Spore et al., TRAC-BDl: An Advanced Best.Estimate Computer Codefor Boiling Water Reactor Loss-ot-Goolant Accident Analysis,NUREK/CR-2I/8, EGG-2109, October 1981.

F-3. S. K. Guntay et al., Best Estimate Prediction for OECD LOFT ProjectFission Product Experiment LP-tP-Z, UtLU LUFI-I-JdU3, June I98b.

F-4. M. Tanaka et al., Quick Look Report on OECD LOFT Experiment LP-SB-3,OECD LOFT-T-36U4, March 1984.

F-5. D. L. Hagrman et al., MATPRO-Version 11 (Revision 2), A Handbook ofMaterials Properties for use in the Analysis of Light Water ReactorFuel Rod Behavior, NUREG/CR-0497, TREE-1280, Rev. 2, August 1981.

F-6. G. A. Dineen et al., LP-FP-2 Supplement to the LOFT Integral TestSystem Final Safety Analysis Report, OECD LOFT-I-ll-5113,December 1984.

F-30

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APPENDIX G

SPECIAL INSTRUMENTATION

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APPENDIX G

SPECIAL INSTRUMENTATION

Experiment LP-FP-2 represented a unique opportunity to study

PWR response to a very severe nuclear accident. Because of this, specially

designed instrumentation was installed in the reactor vessel and in the

containment to take advantage of this opportunity. This instrumentation

was in addition to and independent of that installed by the project for

measuring the thermal, hydraulic, and fission product transport response of

the system. The instruments were: Three Mile Island (TMI)-design

rhodium-emitter self powered neutron detectors (SPNDs) installed in the

central fuel module; a multi-stage iodine species sampler; and nuclear

detectors installed in the reactor vessel shield tank. The information

contained in this appendix came from References G-l, G-2, and G-3 for the

SPN~s, iodine species sampler, and nuclear detectors, respectively. As is

the case with all data from this experiment, these results are preliminary;

in-depth analyses of these data will be reported in the future.

TMI Type SPNDs

The TMI core contained 364 SPNDs which were installed in 52 guide

tubes to measure the local power densities. All of these transducers were

monitored by the plant computer with data being recorded only when they

changed states. In addition to the plant computer, signals from 36 of

these detectors were recorded on two multipoint back-up recorders. The

computer data were lost during the time interval corresponding to the

initial core heatup of the November 1979 accident. Thus, the only SPND

data for this time interval came from the two back-up recorders.

The SPNDs responded normally to the reactor scram early in the

transient. At 2 h 15 min into the accident, some of the SPNDs produced a

negative signal (of unmeasured magnitude), indicating that they were being

heated. Fifteen minutes later, the polarity of several of these detectors

switched to positive. Eventually the output current exceeded the upper

G-1

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recording range of 1500 nanoamps. Since the nuclear reaction had been

terminated at the beginning of the transient (more than 2 h prior to

generation of the large positive signals), the residual neutron flux was

much too low to account for this large positive signal. Thus, the signals

are generally judged to have been caused by a combination of high

temperatures and the physical and chemical environment surrounding the

SPNDs. In an effort to determine after the fact what the core temperatures

were during the accident, attempts were made in the laboratory to reproduce

these large positive signals. Those attempts, reported in References G-4

through G-7, failed to reproduce the phenomena even though temperatures up

to 1700 K (2600*F) and gamma radiation fields up to 2 x l05 R/h were

produced in laboratory testing.

A vertical string of 4 TMI-design SPNDs were installed in the LOFT

center fuel module for Experiment LP-FP-2. These detectors contained a

rhodium emitter and were encapsulated in a zirconium alloy sheath. They

were located in the center guide tube (location H-8) and their sensitive

lengths were centered at 0.28-, 0.69-, 1.11-, and 1.55-m (11-, 27-, 44-,

and 61-in.) above the bottom of the core. The detectors at the 1.11 and

1.55-m (44- and 61-in.) elevations were damaged during installation of the

fuel module into the core and did not operate during the experiment. The

two lower SPNDs remained intact and functional during the transient.

The response of the SPND at the 0.69-m (27-in.) elevation to

Experiment LP-FP-2 is shown in Figure G-l. The detector output was

initially at 340 nanoamps and decayed to near zero within 500 s of scram.

Starting at 1125 s, the output became negative, increasing in magnitude

until 1325 s, when it was -10,O00 nanoamps. The detector output remained

negative at this value until 1425 s, when it suddenly increased to more

than 10,000 nanoamps with positive polarity, remaining at this value for

approximately 20 s. A second large positive pulse occurred at 1450 s,

though for a shorter time. The fuel rod cladding temperatures at the

0.69-m (27-in.) elevation during this time interval were approximately

2200 K (3500°F). This temperature is close to that at which the zirconium

alloy sheath material melts. The output from the SPND at the 0.28-m

G-2

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15000 iL. NE-5H8-7 -

o 12000 o

o 10000 0

9000

6000-0.. 6000 " ,----

r < 5000 c<L 3000 L

0-- 0 -

CD -3000 - D-5000

-6000 -0 0

- -900010000

ch -12000 f w

0 500 1000 1500 2000Time (s)

Figure G-1. Response of SPND at the 0.69-m (27-in.) elevation in FuelAssembly 5.

G-3

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(11-in.) elevation, shown in Figure G-2, remained negative throughout the

transient (except for a brief positive "spike" during reflood). The

difference in responses between these two detectors is believed to be due

to the much lower cladding temperatures [approximately 1000 K (18000 F)

lower than at the 0.69-m (27-in.) elevation], which persisted at this

elevation.

The following preliminary conclusions can be drawn from these

observations. First, the conditions which existed in the LOFT core during

Experiment LP-FP-2 approximated those of the TMI core during the November

1979 accident. (Previous attempts to duplicate these conditions in the

laboratory had failed.) Second, the temperature at which the SPND output

became positive corresponds approximately to the melting point of the

zirconium sheath material.

Iodine Species Sampler

Samples of reactor containment air were taken and analyzed for

radioactive iodine species using an iodine species sampler. The samples

were taken from the Heating and Ventilating Systems (H&V) 8 and 9. It

should be noted that the repository for most of the fission products

released during Experiment LP-FP-2 was the blowdown suppression tank. The

fission products measured by this iodine species sampler were those which

were leaked from the blowdown suppression tank and primary coolant system

into the LOFT containment. Thus, the fission product concentrations

measured by this sampler were not intended to be representative of those

that would be expected in the containment of a commercial PWR during a

V-Sequence accident. A two-in, diameter line is used to vent the

containment into H&V 8, which is a 10-in, line. The exhaust from H&V 8 runs

into a silver zeolite cleanup filter before it is vented to the

atmosphere. H&V 9 consists of a 24-in. diameter line that exhausts into a

silver zeolite cleanup filter and then is either vented back into thecontainment (recirculation mode) or to the atmosphere (vent mode). The

venting rate was not constant with time. In each case, the samples

described in this appendix were taken from locations upstream of the

cleanup filters.

G-4

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3000

L

0-I-()

a,1,

L

L-D ja, 0L0,

0Q.

a,C,,

0

-3000

-6000

-g000

-120000

L.

0"4--

0 C.2D.,-o

z<•

0)

-10000

o )'CA

500 1000 1500 2000Time (s)

Figure G-2. Response ofAssembly 5.

SPND at the 0.28-m (11-in.) elevation in Fuel

G-5

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The iodine species samplers used in this experiment were those

described in Reference G-8 and shown in Figure G-3. These consist of five

media: a particulate filter to retain particulate material; cadmium iodide -

on Chromosorb to adsorb elemental iodine; 4-iodophenol on alumina to adsorb

hypoiodous acid; silver zeolite to adsorb organic iodine species; and a

backup filter to adsorb any species possibly escaping the other media.

After each sample was drawn through the iodine species sampler, the sampler

was removed and taken to the laboratory for analysis. A new sampler was

installed prior to taking the next sample. Two samples were taken from

H&V 9 during the power operations prior to initiation of the experiment.

Six samples were taken from H&V 8 and four, from H&V 9 after experiment

initiation. Table G-1 lists the samples and indicates the location and

time of sampling. Table G-l also lists the total radioactive iodine

concentrations for each sample.

Preliminary results for 1311 are shown in Tables G-l and G-2.

Cursory examination of the results indicates the following. Just prior to

initiation of the experiment, when H&V 9 was rapidly purging the131

containment, more than 50% of I was in the form of 12 and 18% was

organic. The sample taken at 3 h 39 min was 70% 129 12% HOI, 12%

organic, and 7% particulate. The particulate and 12 fractions then

decreased with time; the organic fraction increased with time; and the HOI

fraction first increased, peaked, and then decreased with time. All this

occurred as the H&V 8 system was slowly venting the containment.

Ventilation of the containment was changed from H&V 8 to H&V 9 between

samples 6 and 7. The iodine species admixture in these samples are very

similar, indicating that the smaller diameter H&V 8 line was taking a

representative sample and that there was no excessive plate-out. Results

from samples 7 through 10 (taken with H&V 9) indicate that the relative

fractions tended toward the preexperiment sample fractions as the rapid

purge of the containment continued.

Other radioactive iodine species were also measured. The iodine

species admixtures for 1331 and 1351 were similar to that measured

G-6

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TABLE G-1. MEASURED 1311 CONCENTRATIONS

Sample

Pre-1

Pre-2

1

2

3

4

5

6

7

8

9

10

H&VSystem

9

9

8

8

8

8

8

8

9

9

9

9

Time After Initiation(h, min)

3

5

7

31

42

68

75

77

138

211

39

50

33

17

55

19

49

40

45

11

Concentration a

(microCi/cc)

1.38 + 0.08 x 10 9

9.7 + 0.8 x 10-10

4.18 + 0.03 x 10-7

1.37 + 0.02 x 1O"5

1.55 + 0.02 x 1O-5

2.84 + 0.02 x 1O-5

7.85 + 0.04 x 10-6

6.62 + 0.03 x 10-6

6.33 + 0.03 x 10-5

1.43 + 0.008 x 10-5

9.85 + 0.05 x 10-7

5.79 + 0.04 x 1O- 7

a. Concentration measured in the H&V system. A dilution factor must beapplied to correct these values to the concentration in the containment air.

G-7

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TABLE G-2. MEASURED 1311 SPECIES ADMIXTURE

Species Percent "" V

Sample Particulate12 HOI Organic

Pre-1 6 + 2 55 + 7Pre-2 4 T 2 58 121 6.6 + 0.2 70 +12 2.9T0.1 64 23 2.2 + 0.] 57 + 14 1.04 + 0.01 45.2 + 0.75a 8.0 + 0.1 22.1 + 0.26 0.26 + 0.04 24.1 + 0.27 0.64 + 0.01 23.5 + 0.28 2.13 T 0.04 30.2 + 0.49 3.62 + 0.09 57.0 + 0.7

10 9.2 • 0.2 60.0 + 1.0

a. Results from this sample are suspect.

520

11.816

21.016.55.7

13.79.89.1

21.015.2

+

T++

T++

240.210.40.20.10.20.10.30.30.3

3418

12.017.020.037.364.261.966.158.618.415.1

+5+4

+0.3+ 0.4+0.3* 0.5+ 0.8¥0.6

70.8+0.8+ 0.3¥0.3

G-81, IV

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.0 - Particle filter

Elemental Iodine (12)adsorber

..- Hypoiodous acid (HOIW)edsorber

Organic iodineadeorber

Backup charcoaladsorber

P-ST-OO6O-18

Figure G-3. Iodine species sampler.

G-9

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for 131I. Results for 1321, however, differed. The predominate

fraction for this isotope was elemental 12 (56 to 71% for all samplestaken after experiment initiation) and the organic fraction remained

relatively low (10 to 28%). Thus, this isotope did not exhibit a reduction

in the elemental fraction with time with a corresponding increase in the

organic fraction, as was the case with the other three isotopes.

Non-Intrusive Level Transducers

The final part of this appendix deals with a string of nuclear

detectors that were installed in the reactor vessel shield tank. These

detectors, which are very sensitive to neutron flux, were used to determine

the reactor vessel liquid level in a nonintrusive manner. They have been

used in several previous LOFT experimentsG-9 and have been shown to be

sensitive to changes in liquid level, correlating very well with level

determinations made using in-core instruments such as in-core SPNDs and

cladding temperatures as well as in-core conductivity probe level

transducers.

A detailed description of the measurement system is found inReference G-1O and is briefly summarized here. The detectors are Reuter

Stokes Model R/SP6-0805-135 fission chambers with a 235U core. Each

fission chamber has a 13-cm (5-in.) sensitive length and is surrounded by a

15-mm (0.59-in.) thick blanket of polyethylene and encapsulated in a 0.5-mm(0.02-in.) thick can of cadmium. The cadmium allows only epithermal

neutrons to be transmitted to the polyethylene, which then thermalizes

them, thus enhancing their detectability by the 2 3 5U. Each detector was

recorded using three different discriminator modes. These are:

1. Current mode--this is the straight current out of the amplifier

and has no discrimination between neutron and gamma events;

G-l0

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2. Pulse mode--in this mode, a pulse height discriminator is used to

discriminate between neutrons which result in fission products

with 160 - 179 MeV energy and gamma rays which have only a few

MeV energy; and

3. Rise time mode--this mode takes advantage of the observation that

gamma ray events have a shorter rise time (approximately 1/3)

than neutron events.

The current mode is typically used for the first 400 to 500 s after

scram, when the radiation fields are high and the events are too closely

spaced to allow individual counting. The pulse and rise time modes are

then used to individually count the events and to discriminate between

neutrons and gamma rays.

Figure G-4 is a schematic of the LOFT reactor vessel and shows the

relative location of the Pennsylvania State University (PSU) detectors to

the core and the power range detectors. Figure G-5 is a plane view of the

reactor vessel, illustrating the azimuthal and radial location of the PSU

detectors relative to other reactor vessel instrumentation. The

instrumentation tube used for these detectors lies between the intact loop

hot and cold legs, or closest to Fuel Module 7. The closest instrumented

fuel module is #4. The approximate axial locations are: Detector A--just

above the top of the core; Detector B--just below the top of the core;

Detector C--just above the bottom of the core; Detector D--just below the

bottom of the core; and Detector E--l.O m (40 in.) above the top of the

core. Thus, Detectors B and C respond to the top and bottom approximately

20 cm (8 in.) of the core, respectively and Detectors A and D, to the

approximately 20 cm (8 in.) above and below the core, respectively.

Detector E views a section in the middle of the upper plenum.

Figure G-6 shows the normalized current output from these five

detectors for the first 120 s of Experiment LP-FP-2. The initial drop

corresponds to the reactor scram, after which the detectors followed a

normal scram curve for approximately 20 s. The four detectors near the

core responded to the insertion of the center fuel module control rods at

G-11

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r

57 7 inreactor vessel inside diameter

57 2 Incore fillet outside diameter

PSU

DETECTORS -I Top Of fuel 4XusemblitsCore barrel and

•,%• ~flow skirll•J

LOFT 2 In.anularSOURCE & downcomerIR RANGE •.•Center fuel

DETECTORS module-- -- Corner luel modules•._ •\Lows,..-,.e

.. , , support s|rucluffPSU C _

DETECTORS

INSTRUMENTTUBE

Figure G-4. Cutaway of the LOFT reactor vessel illustrating the locationof PSU detectors.

G-12

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PSU LEVEL GAUGE DETECTORS

RE-T-77-1A2AE-T-7-IA2 Sied lank

hot leg ECC inlet

I C) RE-TRM-86-5

R~E-RC-66-5

Drag disk E"6

turbine Ilowmeter 10

ECC inlet-L__ ' nloRE-TRM-86-6 t oken loop

RE-T-86-6HE-RC-86-6• A• E-T-87-4AI

Reacloto vesse, RE-T-87-4A2Reacor vessel V-2166support bracket Reactor head

Thermocoupt31RE-T-77-2A1 4 ,._" RE'-7-8-2

•E77-A RE-T-86-3ECC inlet •

locatiooken loop) " hOt leg

Inla *c, loop- /•RE-T-77-3AI

cold leg j(. A E-T-77.3A2

RE-T- B5-1 M~anway

RE-T-86-4E CC inletO

Figure G-5. Planar view of the LOFT reactor vessel illustrating thelocation of PSU detectors.

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PSU LEVEL GAUGES

1 .0000.

0.1000

0.0100

9 JULY 85

I

0.0010

0.0001

0 20 40 60 80

TIME (SEC.)

LEGEND : CHANNEL 1 2_4. ------- 4 5

1-DETECTOR E (ABOVE ACTIVE CORE)2-DET. A 3-DET. B (UPPER CORE DETS.) 4-DET. C 5-DET. D (LOWER CORE DETS.)

Figure G-6. Normalized current response of PSU detectors (0 to 120 s).

100

------------ 3

120

( (

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21 s, and all five detectors continued to follow a normal shutdown curve,

indicating the absence of any voiding during this time. Figure G-7 shows

the normalized current output for the first 1800 s. Detectors A, B, D,

and E were all relatively insensitive to the density fluctuations after

approximately 400 s, as was expected using this current mode. However,

Detector C continued to be sensitive to density fluctuations in this

recording mode, responding to decreasing density and boiling near the

bottom of the core during this time frame.

Figure G-8 shows the output from these detectors recorded in the

Pulse-mode for the first 1800 s of the transient. The pulse mode recording

for Detector C failed, so only Detectors A, B, D, and E are shown.

Detector E indicated dryout at the 1-m (40-in.) elevation in the upper

plenum at approximately 500 s into the transient. This detector also

apparently responded to the release of fission products from the gap and

fuel at approximately 1150 s and 1550 s, respectively. Detector Aresponded to the boiling in the upper plenum, also. Detector B clearly

showed the onset of dryout in the top of the core at approximately 600 s,

after which this detector continued to follow a shutdown curve, offset to

indicate continued dryout. Detector 0, however, did not indicate a similar

dryout below the bottom of the core. While boiling occurred at this

elevation, there was a continued indication of a recognizable level near

the bottom of the core throughout the transient.

Figure G-9 shows the pulse mode data for the four detectors for the

period from 1800 to 3600 s. The reflood was detected by the upper three

detectors, but was not seen as distinctly by Detector B. This gives

further credence to the conclusion that the level remained near the core

bottom throughout the transient. There was a large offset between the

post-reflood detector response and that measured in previous LOFT LOCA

experiments. This offset is perhaps indicative of fuel relocation, though

the magnitude of the relocation cannot be determined on the basis of these

detectors alone.

G- 15

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;SU LEVEL CJUGES

1.0000

0.1000

0.0100

9 JULY 85

!

0.0010

0.0001

0 600 1200 1800

L

TIME (SEC.)

EGEND CHANNEL 1 245

1-DETECTOR E (ABOVE ACTIVE CORE)2-DET. A 3-DET. B (UPPER CORE DETS.) 4-DET. C 5-DET. 0 (LOWER CORE DETS.)

Figure G-7. Normalized current response of PSU detectors (0 to 1800 s).

------------ 3

( ( (

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a

(PSU LEVEL GAUGE

S1 00000.

0 100000

0 01000

0 00100

C9 JULY 85

0 0 000 1 0-

0 00001

LEGEND:

0 600 1200 1800

TIME (SEC.)

DETECTO R 1.2 ------ 3 .. 4

I-DETECTOR E (ABOVE ACTIVE CORE) 2-DEFECTOR A (AT ACTIVE CORE TOP)3-DETECTOR B (4 INCHES BELOW DEL. A) 4-DETECTOR D (4 INCHES BELOW ACTIVE CORE)

Figure G-8. Normalized pulse height response of PSU detectors (0 to 1800 s).

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9 JULY 85PSU LEVEL GAUGES

1 .00000

0. 10000

0.01000

0.00100

I•o

0.00010--

"-

0 .00001- 0

18002400 3000 3600

TIME (SEC.)

LEGEND: DETECTOR 12 ---------2- 3

1-DETECTOR E (ABOVE ACTIVE CORE) 2-DETECTOR A (AT ACTIVE CORE TOP)3-DETECTOR B (4 INCHES BELOW DET. A) 4-DETECTOR D (4 INCHES BELOW ACTIVE CORE)

Figure G-9. Normalized pulse height response of PSU detectors(1800 to 3600 s).

4

( ( (

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To summarize, the PSU fission chambers responded to the Experiment

LP-FP-2 transient as was expected based on previous data. Deviations from

the behavior during this and previous transients are consistent with the

postulations that the level remained near the bottom of the core throughout

the transient and that fuel relocation took place during the transient.

G-19

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REFERENCES

G-1. D. J. N. Taylor, "Testing of TMI-2 Type SPNDs in LOFT," privatecommunication, received August 1, 1985.

G-2. J. W. Mandler, J. T. Case, J..W. Tkachyk, "Airborne RadioiodineMeasurements - LOFT FP-2 Test Preliminary Results," privatecommunication, received August 1, 1985.

U-3. A. Baratta, private communication August 20, 1985.

G-4. H. 0. Warren, "SPND Thermal Currents in a Furnace," Appendix B,Interpretation of TMI-2 Instrument Data, NSAC 28, May 1982.

G-5. M. N. Baldwin and H. D. Warren, "SPND Thermal Currents in aFurnace and Gamma Ray Field," Appendix C., Interpretation ofTMI-2 Instrument Data, NSAC 28, May 1982.

G-6. 0. J. N. Taylor, "The Results of Separate Effects Testing onTMI-2 Type SPNDs," to be published.

G-7. C. P. Cannon, D. P. Brown, S. C. Wilkins, R. D. Meininger,Mechanisms for Anamolous Signal Outputs for Self Powered NeutronDetectors During the TMI-2 Accident, October 1984.

G-8. N. C. Dyer et al., Procedures: Source Term Management Program,NUREG-0384, October 1977.

G-9. W. A. Jester et al., Final Report, 1983 LOFT Reactor Testing ofthe Penn State Non-Invasive Liquid Level/Density Gauge,non-published report, June 1984.

G-10. A. J. Baratta et al., Feasibility Study on the Development of aNon-Invasive Liquid Level Gauge for Nuclear Power Reactors,NUREG/CR-3290, May 1983.

G-20

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APPENDIX H

SCDAP/RELAP5/TRAP-MELT CODE CALCULATION AND DATA COMPARISONS

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APPENDIX H

SCDAP/RELAP5/TRAP-MELT.CODE CALCULATION AND.DATA COMPARISONS

The SCDAP/RELAP5/TRAP-MELT code is an integrated code designed to

predict the damage to a reactor coolant system and the transport of fission

products and hydrogen in the event of a severe accident. The code

integrates models for calculating thermal/hydraulic response, reactor core

oxidation and meltdown, and transport of fission products released from

fuel rods and does this in a fully coupled manner. The oxidation and

meltdown models were obtained from the SCDAP codeHl, the

thermal/hydraulic models were obtained from the RELAP5 code H-2, and the

fission product transport models were obtained from the TRAP-MELTH- 3 code.

The integrated code has the capability of calculating the strong

interaction between core damage progression, reactor coolant system

thermal-hydraulics and radionuclide behavior. For example, fission product

release and the transport of iodine and cesium are strongly influenced by

the presence of steam and hydrogen. For high steam flow rates, iodine and

cesium can be transported as free iodine and as CsOH, while for low flow

rate conditions with high hydrogen concentrations, the predominant forms

are CsI, CsOH, and HI. During a severe accident, steam and hydrogen

flowrates can change from a steam-rich environment during initial heatup to

a hydrogen-rich environment during the period of maximum heatup. The

integrated code takes into account the effect of these changing mixtures of

steam and hydrogen on fission product transport.

The entire LOFT primary coolant system and a portion of the secondary

coolant system were modeled with the integrated code. The primary and

secondary systems were divided into 136 fluid cells. The nodalization is

similar to that for the RELAP5 calculations shown in Figure F-1. The heat

transfer into or from the structural components contacted by the fluid was

also modeled. A total of 167 heat structures were used to model this heat

transfer.

H-1

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The reactor core was modeled by a combination of SCDAP and RELAP5 heat

structures. The rods in the center assembly were modeled by SCDAP heat

structures and the rods in the peripheral assemblies were modeled by RELAP5

heat structures. The modeling is described in Table H-l. All of the SCDAP

and RELAP5 heat structures were divided into six evenly spaced axial

nodes. The SCDAP heat structures are capable of modeling severe damage and

of calculating ballooning, oxidation, meltdown and fission product

release. Rod-to-rod and rod-to-steam radiation heat transfer is also

calculated. A total of five SCDAP heat structures were used to model the

center assembly. Two groups of RELAP5 heat structures were used to model

the rods in the peripheral assemblies: the first group modeled the 220 hot

rods and the second group modeled the 876 average rods.

Operator control of the reactor system was assumed to proceed as shown

in Table H-2. The transient was initiated by scramming the reactor. The

intact loop cold leg break valve was opened 20 s after the scram and was

closed 620 s after the scram. The LPIS line valves were opened 220 s after

scram. The assumed operator control differed somewhat from the actual

conduct of the experiment. The cold leg valve was open 115 s longer in the

actual experiment than in the preexperiment calculation. The cold leg

valve was reopened from 878 to 1022 s during the actual experiment, but was

not reopened in the preexperiment calculation. The PORV at the top of the

pressurizer was open from 882 to 1162 s during the actual experiment, but

was never open in the preexperiment calculation.

Results of the preexperiment calculation using the integrated code are

next presented. The calculations are divided into three areas: system

pressure response, fuel rod temperature response, and fuel rod ballooning.

The calculation was performed through the beginning of the high temperature

period of the transient. The end time of the calculation was 1200 s. The

preexperiment calculation was not extended beyond 1200 s because, beyond

that point, deviations from planned test conditions were expected to render

calculation/data comparisons of little value.

H-2

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TABLE H-I. DESCRIPTION OF MODELING OF REACTOR CORE

Relative Power(Fraction of

Average Power inPeripheral AssembliesReactor Core Component

Heat StructureType

24 center fuel rods incenter assembly

76 outer fuel rods incenter assembly

Control rods in centerassembly (total of 11)

Hollow guide tubes incenter assembly(total of 2)

Insulated flow shroudfor center assembly

220 hot fuel rods inperipheral assemblies

876 average fuel rodsin peripheral assemblies

2.04

2.38

0

0

0

SCDAP

SCDAP

SCOAP

SCDAP

SCDAP

RELAP5

RELAP5

1.28

0.93

H-3

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TABLE H-2. CONDUCT OF EXPERIMENT LP-FP-2 ASSUMED IN PREEXPERIMENTPREDICTION

Time FromReactor Scram

0.0

20.0

25.0

220

620

Reactor Operator Action

Scram reactor by inserting fuel rods incenter and peripheral assemblies

Open intact loop cold leg break valve

Turn off power to pumps in primary coolantsystem

Open valves to allow flow throughLPIS line

Close intact cold loop leg break valve inresponse to system pressure dropping to lessthan 1.2 MPa.

H-4

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System Pressure Response

The calculated and measured system pressure response are compared in

Figure H-i. The calculated and measured end of subcooled blowdown occurred

at 60 and 53 s, respectively. The depressurization rates continued to be

in excellent agreement until flow through the LPIS line was actuated at

220 s. After 220 s, the calculation overpredicts the rate of flow through

the LPIS line and as a result overpredicts the rate of system

depressurization.

The difficulty in accurately modeling the LPIS line is due to its

complex configuration, with many bends and changes in flow area. No

experiments were performed prior to the LP-FP-2 Experiment to provide data

for determining loss coefficients. The preexperiment calculation used loss

coefficients ranging from 0.82 to 0.97. The results of the FP-2 experiment

indicate that, to accurately model flow through the LPIS line, the loss

factors need to be increased by a factor of six.

Reactor Core Temperature Response

Measurements show that heatup of the center fuel assembly began at

662 s and that the heatup accelerated when the intact loop cold leg break

valve was closed at 735 s. The closure of this valve reduced the rate of

depressurization and precipitated a collapse of the two-phase mixture level

in the core due to reduced flashing. The calculation also showed that

heatup of the center fuel assembly began with closure of the intact loop

cold leg break valve, but at 620 s. Thus, heatup began 42 s earlier in the

calculation than in the experiment.

The calculated and measured temperature responses of rod J07 in the

center assembly at an elevation 0.69 m (27 in.) above the bottom of the rod

are compared in Figure H-2. The measured temperature response flattened

out from 880 to 1162 s due to the opening of the PORV and the reopening of

the break valve. Oxidation began when the cladding temperature exceeded

1000 K (1340-F). The calculated and measured rates of temperature increase

prior to oxidation were both about I K/s (1.8°F/s).

H-5

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17.5

--- PE-BL-002ACalculated

15

0~

S.-

VL:3C,,

0~

12.5

10

7.5

5

2.5

0

2500

-2000

-1500 O.

L

-1000

L

-500

LA

0 250 500 750 1000 1250Time (s)

1500

Figure H-i. Comparison of measured hot leg pressurecalculated using the integrated code.

with pressure

w2500- 4000

5-

VL:3

0LV0~EV

I-

2000

1500

1000

TE-5J07-027Calculated

-3000

-2000

L

lOOO0 E

I-

500

00 250 500 750 1000

Time (s)1250 1500

Figure H-2. Comparison of measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits).

H-6

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The calculated response of rod J07 in the center assembly at an

elevation of 1.26 m (50 in.) is shown in Figure H-3. A measured

temperature response is not available at this elevation. Rapid oxidation

is calculated to begin at 1180 s and cause a cladding temperature increase

of 35 K/s (63°F/s).

The calculated and measured response of rod H13 in a peripheral

assembly at an elevation of 0.94 m (37 in.) are compared in Figure H-4 and

are seen to be in good agreement. Both also show dryout and fuel rod

heatup beginning when the intact loop cold leg break valve is closed. Both

calculation and measurement show a heatup rate of 0.8 K/s (l.4*F/s). The

measurement shows a significant decrease in the rate of heatup beginning at

1400 s. This decrease in heatup may be due to an increase in steam flow

caused by liquefied material in the center bundle falling into the pool of

water in the lower plenum of the vessel.

Fuel Rod Ballooning

The preexperiment calculation predicted that the fuel rods in the

center bundle would balloon and rupture. The ballooning was predicted to

occur in the elevation span of 0.84 to 1.12 m (33 to 44. in.) and to cause a

78% reduction in flow area. The rods were predicted to rupture in the

period of 1000 to 1100 s at cladding temperatures of 1140 to 1160 K

(1593 to 1629 0 F). A radiation measurement at the top of the vessel

indicates that fuel rod rupture began at 1200 s.

Conclusions

The SCDAP/RELAP5/TRAP-MELT code was successfully used to perform a

preexperiment calculation of the blowdown and heatup phases of

Experiment LP-FP-2. The code accurately predicted that the time of

initiation of core heatup would begin when the intact loop cold leg break

valve was closed. The code also predicted that the initial rate of heatup

of the center bundle would be I K/s (l.8 0 F/s) and that the rate of heatup

during rapid oridation would exceed 35 K/s (63 0 F/s). The code appears to

have correctly predicted the time at which fuel rods in the center bundle

would rupture and release fission products. Experiment results clearly

indicate that modeling of the LPIS line needs improvement.

H-7

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I-N

V,

E

2500

2000

1500

1000

500

0

LL

E

0 250 500 750 1000 1250 1500Time (s)

Figure H-3.

1500

1250

I 1000

-4--a

• 750

E

Calculated cladding temperature at the 1.26-m (50-in.)"elevation in Fuel Assembly 5 made using the integratedcode. (See Appendix I for thermocouple qualification limits).

U)L:3

.9-

aLU)0.EU)I-

500

2500 250 500 750 1000 1250 1500

Time (s)

Figure H-4. Comparison of measured cladding temperature at the 0.94-m(37-in.) elevation in Fuel Assembly 4 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits).

H-8

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REFERENCES

H-1. C. M. Allison, E. R. Carlson, R. H. Smith, Proceedings of anInternational Meeting on Light Water Severe Accident Evaluation,August 1983, 5.1.1 to 5.1.5.

H-2. V. H. Ransom and R. J. Wagner, RELAP5/MOD2 Code Manual, EG&GReport EGG-SAAM-6377, April 1984.

H-3. H. Jordan, J. A. Gieseke, P. Baybutt, TRAP-MELT User's Manual,NUREG/CR-0632, February 1979.

H-9

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APPENDIX I

QUALIFIED TRANSIENT DATA PLOTS

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APPENDIX I

QUALIFIED TRANSIENT DATA PLOTS

This section contains qualified transient plots of all data. The

plots are on fields on the inside of the back cover of this report.

Additional files are included which contain the first report from the Data

Integrity Review Committee. A list of all the parameters on this file is

included in Table I-1.

I-1

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TABLE I-1. LISTING OF QUALIFIED DATA FISCHE

Column

AAAAAAAAAAAAAAA

BB

BBBB

8

fi

BaC

BBBB

CCCCCCCCCCCCCCC

PlotNumber

001OJ2003004005036007008009010Oil012013014015

016017018019020021C2202302402502602?028029030

0310320J33(34035036037038C3 9C40C41042043

C45

MeasurementIdentification

AH2E-T75-O01AH2E-Tt 5-U02AH2E-T 5.-003CR-5UP-ACR-5UP-BCV-PC04-008CV-P004-01 0CV-PO04-09CCV-P 004-091CV-P13e-070AC V-P 13u-07 1ADE-KL-OIADE-t-1 L-OC IBOE-dL-O01CDE-8 L-C02A

DE-IlL-002BDE-BL-002CDE-B L-105DE-flt-205JE-PC-CO1ADE-PC- 016OE-PC-0G1COE-PC-002AOE-PC-0028DE-PC-GOZC)E -PC-i 05DE-PC-2C5F EP165-F1-22FE-PC-CC2OFE-IST-GO1

FE-I ST-CC2FR-PC-201FR-PC-205FR-PC-20bFT-PO04-012F T-PC04-72-2FT-PIZE-065FT-P 139-27-1FT-P 139-27-2FT-P 139-27-3L EPO T-P 139-007LE-FCC-01ALI T-Pi20-013LIT-P120-C14LIT-PlýC-08(i

Column

DDDDD0DDDDDD

D

E

EEEEEEEEEEEEEEE

FFFFFFFFFFFFF

FF

PlotNumber

C46047048049050051052053G54055056057058059Cb0

061062063064065066067068069470071072073CUT 3('74

C75

076077078C79080£8108208308408508608708809G90

MeasurementIdentification

LT-PO04-OObALT-PO04-008BLT-PO04-042LT-PO04-OBAALT-P 138-033LT-Pl3E-058ME-PC-002AME-PC-0028ME-PC-G02ME-IST-001N E-2HO-26NE-4HO-26NE-6HO0-26POE-BLH-001POE-0Li-002

POE-BLIt-003POE-OLH-004PU)E-BLH-005POT-PI39-OC6POT-P 139-0C7PDT-P139-030PDT-P13 9-30AP0T-P139-3C8PE-BLH-OC1PE-BiLH-002PE-BLh-003PE-BL-OOZAPE-PC-002PE-PC-GOSPE-PC-CC6

PTPI6S-F--bBPTPl5-F2-43PT-POC4-O1CAPT-P 004-022PT-P 004-034PT-PuO4-05PT-P12C-029PT-P 120-04 3PT-P138-05OPT-P 13b-057PT-P13S-004PT-P139-042PT-P 139-05-1RE-T-77-lA1RE-T-77-1A2

Column

GGGGGGGGGGG6GGG

HHHHHHHHHH

HHH

IH

IIIIIIIIIIIII

PlotNumber

091C92093094095096097098099100101102103104135

106107108109110ill112113114115116117118119120

121122123124125126127128129130131132133134135

MeasurementIdentification

RE-T-77-?A 1RE-T-77-3A1RE-T-77-3A2RE-T-85-1RE-T-85-2RE-T-86-3RE-T-86-4RE-T-87-4A1RE-r-87-4A2RE-T4-096RP E-PC-001RPE-PC-002RP-CROM2-PTRP-CRDh2-TCRP-CPDP4-PT

RP-CkDfl4-TCRP-C RDM6-P TRP-C RDMb-TCRP-CRDMB-PTRP-C PDM8-1CSP-8LH-CO1SP-BLH-002SP-BLH-CC4SP-LLH-305SP-3LH-006SP-OLH-007ASP-BLH-008SP-PC-002BS P-P139-019SP-P13q-02C

SP-SG-C£3SO-SG-CO4SP-lST-005ST-BLH-001ST-BLH-002ST-BLH-003ST-BL-002AST-PC-002ST-PC-005ST-P139-05-1TC-5108-27TC-5M04-27TC-5M08-27TE-BLH-001TE-1LH-002

( ,( (

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t tj I A

C ( CTABLE 1-1. (continued)

PlotColumn Number

KKKKKKKK

KK

LLLLL

JJJJJJ

L

K-o K

KKKKKKKKKK

LL

LtLLLLLLLLL

13713813914U14114214314414514b14714814915C

151152153154155156157158159160161

1631641b5

166167168169170171172173174175176177178179180

MeasurementIdentification

TE-BLIIt-C04TE-BLh-605TE-8LIi-CG7AT E-i3Li- 008TE-PC-OC2ATE-PC-CC28TE-PC-OC2C-TE-P CC,4-0,4FE-Pi2c.-Jo11E-t0120-027TE-PI26-ý41TE-tI 2C-132

"E-P 134-01 'TE-P139-'2 CYE-PYo2

TE-11 3q--2-2TE-P134-32-1TE-P 141-C94TE-P 141-095tE-SG-OCIATE-SG-t.O2ATE-SC-GO3TE-SG-O6CTE-SG-6C5

TE-SV-0O01TE-SV-002TE-SV-003TE-SV-OC5TE-SV-OC6TE-SV-007

TE-SV-GCETE-SV-00qTE-SV-GIG

TE-SV-0.11TE-SV-C12TE-TOS5-002TE-LA11-030TE-IB1C-037T1-111-02eTE-1B11-O3ZTE-1C1l-021TE-LC11-034TE-IF07-015TE-IF 07-026TE-IST-001

PlotColumn Number

18118218318416518618718818919019 1192193194195

1961971981992002012022032042052J 6207208209210

211212213214215216217218219220221222223224225

MeasurementIdentification

YE-1SI-C02TE-IST-003TE-1 ST-004TE-LST-O05TE-i ST-OObTE-1SI-O0eTE-iST-009TE-17ST-CITE-iST-O1ITE-1 ST-012TE-IST-1I3TE-1ST-15TE-IUP-001TE-1UP-002TE-1UP-O05

TE-I UP-006TE-IUP-OG7TE-2E08-045TE-ZFOT-Ol0T E-2 FOO-03 2TE-ZFOr--026TE-2G14-011TE-2G14-030TE-ZG14-04TE-2H14-026TE-2HG2-028TE-ZH13-021TE-2H13-049TE-2H14-032r E-2H1 -02 6

TE-2HI5-04 1TE-2114-021TE-2114-039TE-20P-001TE-ZLP-002TE-2LP-003TE-ZUP-C01TE-2UP-O02TE-2UP-C03TE-2UP-004TE-2UP-005TE-3A1-030TE-3B11-028TE-32!1-032TE-3CII-02 1

PlotColumn Number

pPPPPPPPPPPPPPP

226227228229230231232233234235236237238239240

241242243244245246247248249250251252253

2s0

255

2562372582592b0261262263264265266267268269273

MeasurementIdentification

TE-3C 11-039TE-3F07-02CTE-3UP-C01TE-3UP-CO6TE-3UP-CO0TE-3UP-010TE -3 UP-O11TE-3UP-012TE-3UP-013TE-3bP-014TE-3UP-015TE-3UP-C16TE-4ECB-04TE-4FO7-015TE-4FOb-032

TE-IGO-02 1TE-4GI4-0O 1TE-4GI4-030TE-4G14-045TE-4H13-01 5TE-4H13-037TE-41114-02 8TE-4 K15-02 tTE-4H115-04 1TE-4 114-021TE-4 114-039TE-41LP-001TE-40LP-003TE-4UP-CC1TE-4UP-C02

TE-4UP-003TE-4UP-004TE-,UP-O05TE-5C06-027TE-5CO6-066TE-5CG7-04 2TE-5COY-010TE-5CO-02 7TE-5C10-G27TE-5C12-010TE-5CI 2-027TE-5D09-027TE-5D13-042TE-5EO5-027TE-5E11-027

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TABLE 1-1. (continued)

Column

AAAAAAAAAAAAAAA

BBBBBBBBBBBB

8

BB

CCCC

CCCCCCCCCC

PlotNumber

27127227327427527627727b279280281282283284285

286287Z8828 929029129229329421)29629729829q300

3u3323J33043033u6337308309310311312313314315

MeasurementIdentification

TE-5f-010TE-5E-027TE-5E-C42TE-5FG3-027TE-5F13-066TE-3GO4-O10TE-5GC4-027TE-5L12-0101 E-5(,12-027TE-5tiCt-C271iE-Ai~b-027TE-,H12-027TE-ý163-C27TE-,I'4-042TE-5• l2-L42

TE-,JC3-GCLTE-,J07-OI uTE-5J07-027TE-,.j13-027TE-5KO5-02 7TE-iK11-027TE-5L C7-CI 0TE-5L07-027TE-5 L09-04 2TE-,MG6-02 7TE-5MC7-010TE-5hO",-04 2TE-5 PIO-G-bTE-54-0ICTE-N-Cj27

SE -5U- 032T E -51-042TE-5S-010TE-5S-027FE-5S-032TE-iS-G42TE-SUP-004TE-5UP-017TE-5UP-019TE-5UP-023TE-5UP-024TE-5UP-025TE-SUP-0261E-SUP-627r E - U P--028 A

Column

D

DD0

0DDDD0DD

E

EEEEEEEEEEEE

F

FFFFFFFFFFFFFFF

F

PlotNumber

316317318319320321322323324325326327328329330

331332333334335336337338339340341342343344345

3463473483493503513453523533,4355356357358

359360

MeasurementIdentification

rE-sUP-028BTE-5UP-029AT E-5IP-a 296TE-SUP-030ATE-5IP-G30BTE-5 LP-031ATE-5 UP-031BTE-SUP-C32ATE-SUP-32 BTE-5 LP-033 BrE-5uP-188ATE-SUF-1 888TE-SUP-IhBCTE-,bP-16860TE-5UP-194C1

TE-3LP-1l4G2IF-5UP-i c#7FlTE-5UP-197B2TE-SUP-21262TE-5UP-2 1G2TF-SbP-2152B1IE-5UP-21500

TE-jaUP-2 50G2TE-SUP-2 5181TE-5UP-251B2TE -W-CIOFE-S W-0 27TE-SW-L32TE-5W-C42

TE-6EG8-04,T E-6F7-037TE-b6FOI-041TE-bGGS-03csTE-bG14-011TE-6G14-030TE-5W-042TE-6G14-04 5TE-6H13-01 STE-6H13-037TE-6H14-028TE-6H14-032TE-b5,15-02bTE-SI14-GZ1TE-6114-039TE-6LP-001

Column

GGGG6GGGGGGGG

GG

HHHHHH

HHhH

HH

I-H

IIIIIIIIIIIII

PlotNumber

36136Z363364365366367368

370370371372373374375

37 b377378379380

301382383384385386387388389390

391392393394395396397398399400

41140243340440 i

MeasurementIdentification

TE-bLP-002TE-6LP-003TE-bUP-001TE-6UP-002TE-61P-003TE-6UP-004TE-6UP-005TT-PCC4-004FT-P13%-03 2TT-P139-03 3Tt-P139-G34XEP16 -Fl-42XE P165-F1-44CVP165-F 1-348CVP165-F 1-12

CVPlt5-F2-36C VP 165-Fl- 48ECVP1f5-F1-13CVP165-FI-34 AC VPI65-F 1-20CVP165-F1-14IL P165-D1-5ACVPI65-DI-4ACVP165-D1-3ACVPI .5-F 2-34ACVP1(5-FI-2PCVP165-F2-48CVP165-DI-15CVP165-F2-34BCVPI65-F1-316.

PFP165-FI-40FTP165-F 1-22PTP165-FI-5PTP165-U1-20PrP165-Fl-BAPTP165-Dl-2PT P165-D1-19TEP165-DI-21BTE P165-F I-30ATEP165-F2-45TEP165-F1-tCTEPI65-F2-38TEP165-F1-8AFEP1b5-F1-BBTEPlbS-F1-38

I, / ((

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mmmm-mý

APPENDIX J

ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT

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APPENDIX J

ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT

A microfiche copy of the ORIGEN2 calculations reported in Section 4 of

this report are attached at the end of this report. The two microfiche

cards that contain the ORIGEN2 results are titled: "FP-2 ORIGEN2

CALCULATION FOR QLR." This calculation lists the center bundle fuel and

coolant inventories in terms of grams, gram-atoms, and curies. The decayheat of the center bundle is also listed on these cards. Notice that the

listed fuel inventory results do not assume any fission product loss during

the experiment. That is, to obtain the end of experiment fuel inventory,

subtract the coolant inventory from the listed fuel inventory. Also, for

this calculation, it is assumed that the entire fission product loss to the

coolant occurs between 1740 and 1800 s.

The input deck that was used to create the ORIGEN2 calculation is

shown near the end of the microfiche listing.

J-1