ACCELERATED D> UTION DEMONS TION SYSTEMI

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ACCESSION NBR!9003290079 DOC.DATE: 89~~ NOTARIZED: NOFACIL:50-275 Diablo Canyon Nuclear Power Plant, Unit .1, Pacific Ga

50-323 Diablo Canyon Nuclear Power Plant, Unit 2, Pacific GaAUTH.NAME AUTHOR AFFILIATION

SHIFFER,J.D. Pacific Gas & Electric Co.RECIP.NAME RECIPIENT AFFILIATION

DOCKET0500027505000323

SUBJECT: "10CFR50.59 Annual Rept of Facility ChangesChanges, Tests 6 Experiments 880323-890322. '/900321 tr.

DISTRIBUTXON CODE: IE47D COPIES RECEIVED:LTR ENCL : L'ITLE:50.59 Annual Report of Changes, Tests or xperiments Mage W/out pprov

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NLII

)

Pacific Gas and Electric Company 77 Beale Street

San Francisco, CA 94106

415I972 7000415/973-4684

James 0. ShifferSenior Vice President andGeneral HanagerNuclear Power Generation

Harch 21, 1990

PGLE Letter No. DCL-90-080

U.S. Nuclear Regulatory CommissionATTN: Document Control DeskHashington, D.C. 20555

Re: Docket No. 50-275, OL-DPR-80Docket No. 50-323, OL-DPR-82Diablo Canyon Units 1 and 210 CFR 50.59 Annual Report of Facility Changes, ProcedureChanges, Tests, and Experiments Harch 23, 1988 — Harch 22, 1989

Gentlemen:

Pursuant to 10 CFR 50.59, enclosed is the Annual Report of Changes,Tests, and Experiments for Diablo Canyon Power Plant (DCPP) Units 1

and 2 for the reporting interval of Harch 23, 1988, toMarch 22, 1989. This is the same as the reporting interval of theannual Final Safety Analysis Report (FSAR) Update revision.

Licensees are allowed by 10 CFR 50.59 to make changes in thefacility as described in the FSAR, make changes in procedures asdescribed in the FSAR, and conduct tests and experiments notdescribed in the FSAR without prior NRC approval, provided that thechanges, tests, or experiments do not involve changes in theTechnical Specifications or unrevi ewed safety questions. Suchchanges are required to be reported on an annual basis by10 CFR 50.59(b)(2).

Proposed facility changes, procedure changes, tests, or experimentsthat are determined to involve unrevi ewed safety questions orchanges to the Technical Specifications are submitted separately tothe NRC for prior approval under the provisions of 10 CFR 50.90.

h n in h F ili A D ri in h F AR. The enclosedAnnual Report provides a brief description of the 10 CFR 50.59facility design changes, including a summary of each safetyevaluation. Each change was reviewed and accepted by the PlantStaff Review Committee (PSRC). Each change is complete and has beenaccepted by the plant staff.

None of the design changes involved an unreviewed safety question ora change to the DCPP Technical Specifications, as determined by thePSRC. The changes do not (a) increase the probability orconsequences of an accident or malfunction of equipment important tosafety as previously analyzed in the FSAR; (b) create thepossibility of an accident or malfunction of equipment notpreviously analyzed in the FSAR; or (c) reduce the margin of safetyas defined in the bases of the Technical Specifications.

'~003290079 890=-2".POD ADCiCK C~.":i00027.5R FDC

0

Document Control DeskPG&E Letter No. DCL-90-080

Harch 21, 1990

Design changes resulting in minor changes to FSAR Update figures (e.g., pipingand instrumentation drawings and schematics) that do not impact the safety ofoperations are not included in the Annual Report.

h n i r r A D ri in h F AR . The enclosed AnnualReport provides a brief description of a 10 CFR 50.59 procedure change,including a safety evaluation summary. The procedure change does not involvean unreviewed safety question or a change to the DCPP TechnicalSpecifications, as determined by the PSRC.

N Tests and Experimentsare performed to procedures. The enclosed Annual Report also provides a briefdescription of tests or experiments not described in the FSAR Update togetherwith a safety evaluation summary. None of the tests or experiments involvedan unreviewed safety question or a change to the DCPP TechnicalSpecifications, as determined by the PSRC.

m r /Lif L . Several Jumpers installed during the time periodcovered by the Annual Report were determined to result in a change to afunction of a system operation as described in the FSAR Update. A briefdiscussion of each of the jumpers and a safety evaluation summary is includedin the Annual Report. None of the jumpers involved an unreviewed safetyquestion or a change to the Technical Specifications, as determined by thePSRC.

A setpoint that was determined to be described or referencedin the FSAR Update was changed during this 10 CFR 50.59 reporting period. Abrief discussion of this change and a safety evaluation summary is presentedin the Annual Report. It was determined by the PSRC that this setpoint changewas not an unreviewed safety question or a change to the TechnicalSpecifications.

Kindly acknowledge receipt of this material on the enclosed copy ofthis letter and return it in the enclosed addressed envelope.

incerely,

J. D. Sh f er

cc: A. P. kodgdonJ. B. HartinM. H. HendoncaP. P. Narbut

'H. RoodCPUCDiablo Distribution

Enclosure

3014S/0081K/RNH/1392

PGhE etter No. DCL-90-080

ENCLOSURE

10 CFR 50.59 ANNUAL REPORT OF FACILITY CHANGES, PROCEDURE CHANGES,TESTS, AND EXPERIMENTS

MARCH 23, 1988 — MARCH 22, 1989

DIABLO CANYON UNITS 1 AND 2DOCKET NOS. 50-275 AND 50-323

Pacific Gas 5 Electric Company

3014S/0081K

SUMMARY OF 10 CFR 50.59 FACILITY CHANGES, PROCEDURE CHANGES, AND TESTSFOR THE REPORT PERIOD

March 23, 1988 - Harch 22, 1989

A.

1. Installation of AMSAC

2. Setpoint Revision for RHR PumpMiniflow Control 1 ers (FIC-641A and641B)

Uni til

1,2

ifi i n

DCP H-40065 R.l

DCP J-41647 R.lDCP J-42647 R.2

3.

4.

5.

Removal from Service Four SteamGenerator Tubes in Response to NRCBulletin No. 88-02

Removal from Service Four SteamGenerator Tubes in Response to NRCBulletin No. 88-02

Replacement of Unit 2 SafetyInjection Pump 2-2

DCP M-42312 R.O

DCP H-41601 R.l

DCP H-42453 R.O

6. Replacement of Emergency FlowTransmitter (FT-113)

7. Replace RHR Suction ValvesAutoclosure Interlocks withAnnunciator on Valve(s) Open andHigh Pressure

8. Replacement of the Auxiliary SaltWater Pump Impeller

9. Hodifications to the Reactor CavityManipulator Crane

10. Hodifications to the Fuel TransferSystem

ll. Change Position of the Spent FuelPool Bridge Crane

12. Removal of Clarifier and Installationof Makeup Water Pretreatment Systemand Supply of Domestic Water

13. Installation of a Seawater ReverseOsmosi s System

1,2

1,2

1,2

1,2

1,2

1,2

DCP J-37477 R.lDCP J-38477 R.l

DCP H-35469 R.2DCP H-36469 RE 1

DCP M-'39834 R.O

DCP H-40515 R.O

DCP M-39441 R.2

DCP M-39779 R.ODCP H-40779 R.O

DCP H-37800 R.O

DCP M-31394 R.7

3014S/0081K

SUHMARY OF 10 CFR 50.59 FACILITY CHANGES, PROCEDURE CHANGES, AND TESTSFOR THE REPORT PERIOD

Harch 23, 1988 - Harch 22, 1989

(Continued)

A. (c t.)Uni t

1 i i 1

14. Change to the Volume Control TankHydrogen Inlet Pressure Setoint

15 ~

16.

Install New Secondary Hater ChemistryLab

Relocation of the Non-Safety-RelatedAir Compressors

17. Abandon Control Room Electric DuctHeaters

1,2

1,2

1,2

1,2

DCP 3-41078 R.OKP 3-42078 R.O

DCP H-31753 R.10DCP H-32753 R.2

DCP H-35366 R.O

DCP H-35020 R.O

18. Installation of a Spare Step-UpTransformer

DCP E-40367 RE 1

19. Installation of a Secondary BoricAcid System

20. Replacement of Equipment Drain andFloor Drain Receiver Pumps

*26. Replace Reactor Coolant Resistan'ceTemperature Detectors

21. Construction of New Radwaste StorageBuilding

*22. Removal of Drumming Station RadiationHonitor RE-8

*23. Replace Pressurizer Spray Control'alves

*24. Replace Control Room Lighting,Ceiling and Hall and Floor Coverings

*25. Remove High Pressure Turbine Cover

1,2

1,2

1,2

1,2

DCP H-39881 R.3DCP H-40881 R.O

DCP H-37759 R.O

DCP M-29953 R.ODCP H-29953 R.3

DCP E-11868 R,OKP E-25781 R.O

DCP J-40053 R.l

DCP E40333 R. 1

DCP H-39126 R.O

DCP J-32961 R.l

*27. Replace Steam Generator FeedwaterRing 3-Tubes

DCP H-36476 R.l

3014S/0081K

SUMMARY OF 10 CFR 50.59 FACILITY CHANGES, PROCEDURE CHANGES, AND TESTSFOR THE REPORT PERIOD

March 23, 1988 — March 22, 1989

(Continued)

A. <C t)Unit

f i i n

*28. Modify the Solid'tate ProtectionSystem

*29. Add Redundant CVCS Letdown Filter*30. Add Backup Overcurrent Protection for

Containment Electrical Penetrations

DCP E-32853 R.O

DCP M-28132 R.3

DCP E-4459 R.8DCP E-4941 R.2DCP E-6410 R.ODCP E-6415 R.5DCP E 6504 R.lDCP E-6546 R.3DCP E-8140 R.ODCP E-10010 R.lDCP E-12426 R.ODCP E-12981 R.ODCP E-14651 R.ODCP E-16206 R.O

*These changes were completed in previous reporting periods and documentchanges simi liar to both Diablo Canyon units. The changes to the firstof the two Diablo Canyon Units were previously reported. However, theselisted changes were not included in previous Annual 50.59 Reports and areincluded in this report to update -the reporting record.

B. r h n

ri i nUni t

1 ifi i n

1. Procedure NPAP B-101 revised toincorporate the requirements ofthe May 26, 1987 Revision to10 CFR 50.55.

1,2 NPAP B-101

3014S/0081K

SUMMARY OF 10 CFR 50.59 FACILITY CHANGES, PROCEDURE CHANGES, AND TESTSFOR THE REPORT PERIOD

March 23, 1988 - March 22, 1989

(Continued)

C. Pr r T

2.

3.

4.

A special Leak Rate Test of theContainment Spray AdditiveTank Valves.

A test of the Component CoolingHater (CCW) Heat Exchanger InletHaterbox Level.

Residual Heat Removal (RHR)Pipe Vibration Test. This testallows different RHR flows andvalve positions. Data will berecorded at each different-configuration.

Test Procedure for reducingCarbon Monoxide (CO) incontainment. This procedureutilizes the service air forpurging CO out of Unit 2Containment.

Uni t

1,2

n fTemporaryProcedureTB-8809, Rev. '0

TemporaryProcedureTB-8816, Rev. 0

TemporaryProcedureT0-8805, Rev. 0

TemporaryProcedureTB-8829, Rev. 0

5.

6.

Test Procedure for measuring BITmanway leakage at operatingpressure

On-The-Spot-Changes to theProcedures allow the SafetyIn]ection portions to beperformed in Mode 5 rather thanMode 6. The ob)ective of theSafety In)ection portions ofSTP V-15 is to set cold leg andhot leg in)ection flows byad)usting runout. throttle valves.The objective of STP P-1A is torunout flow in order to obtain apump characteristic curve.

1,2

TemporaryProcedureTB-8833, Rev. 0

On-The-Spot-Changesfor proceduresSTP V-15, Rev. 1 andSTP P-1A, Rev. 1

3014S/0081K

SUMMARY OF 10 CFR 50.59 FACILITY CHANGES, PROCEDURE CHANGES, AND TESTSFOR THE REPORT PERIOD

March 23, 1988 — March 22, 1989

(Continued),

D. m r f

ri in Unit1 i i 1 i

1. Temporary )umper to disable theauto-initiated "ContainmentEvacuation On High Flux At Shutdown"alarm.

2. Temporary Jumper allows thecomparator for TE-4238 which isbroken to be tripped. Once Unit 2is in Mode 4 and <333'F, thejumper will make up the temperaturesignal to the PORV-456 circuit.

1,2 Jumper Log No. 88-034Unit 1

Jumper Log No. 88-024Unit 2

Jumper Log No. 88-026

2. Temporary jumper allows thecomparator for TE-423B which isbroken to be tripped. Once Unit 2is in Mode 4 and <333'F, the)umper will make up the temperaturesignal to the PORV-456 circuit.

Jumper Log No. 88-026

E.

D ri i nUni tli il n ifi i n

1. Change nominal high alarm setpointfor area radiation monitors toreduce the number of auto-initiatediodine removal modes resulting infewer hours elapsed on charcoalbanks, better equipment reliability.

Fuel HandlingBuilding AreaMonitors from7.4 mR/hr to10 mR/hr

3014S/0081K

SUMMARY OF 10 CFR 50.59 SAFETY EVALUATIONS FOR FACILITY CHANGES,'PROCEDURE CHANGES AND TESTS FOR THE REPORT PERIOD

March 23, 1988 - Harch 22, 1989

A. F ili h n

1. 11 i f AM

DCP H-40065 Rev.l (Unit 2)

PG&E has installed the Hestinghouse, Option 1, Anticipated TransientHlthout Scram (ATHS) Hitigation System Actuation Circuitry (AHSAC)design in DCPP Unit 2. The DCPP design is based on the sensing oflow steam generator (SG) level using a three out of four logicwhenever the unit load is above 4'as measured by turbine impulsechamber pressure). The system consists of a Design Class II AHSAC

package, powered from a battery-backed, Class II inverter. Thepackage receives the analog input signal through qualified, Class IEisolation amplifiers from the existing Class IE level and pressureinstrument loops. Three independent microprocessors each monitor allinput signals and route them to two independent logic networks whichvote to initiate a final AHSAC output signal. The output (trip)signal actuates qualified, Class IE isolation relays which initiate aturbine trip, an AFH pump start and, terminates steam generatorblowdown.

1 i n r

The NRC Staff approved the DCPP AMSAC design in NRC letter datedAugust 15, 1988, "Safety Evaluation Of the AMSAC System, PG&E'sProposed Hethod of Implementing the Requirements of 10 CFR 50.62(ATHS) for Diablo Canyon (TAC NOS. 59088 and 59089)."

The DCPP AHSAC provides an alternate (and diverse) means of trippingthe main turbine and actuating AFH flow apart from the reactorprotection system (RPS). As such it interfaces with safety relatedinstrumentation loops for the detection of adverse plant conditionsand with safety-related plant control systems for the mitigatoryresponses. Since Class II portions of the AHSAC interface withsafety related systems these interfaces were examined and adequatelyisolated so as to not compromise their integrity, redundancy orfunction. This was accomplished by the use of qualified isolationamplifiers and relays on the input and output sides, respectively, ofthe AHSAC system.

2. P m ini w r ll rHUB.DCP 3-41647 Rev. 1 (Unit 1) and 3-42647 Rev. 2 (Unit 2)

-4 n

This change involved the revision of RHR pump miniflow controller(FCI-641A and,6418) setpoints. The purpose of the controller is to

3014S/0081K

control motor-operated RHR pump rec1rculation valves to ensure thatthe pump frict1onal heat is d1ss1pated during low system flowconditions. As a result of Barton flow switch qualificat1on tests,the original setpoints (based on an accuracy of z2.55'L of span) forFIC-641A and 64lB were revised to account for seismic drift of up to

, zl(C of span. The new setpoints reflect a revised accuracy, of+12.55K of span. FIC-64lA and 641B are used to control FCV-64lA and64lB.

The setpo1nt methodology is that the "open" setpoint is chosen toprotect RHR pumps at low system flow conditions by ensuring that theminiflow valve will open before cold water flow decreases below500 gpm. The "close" setpoint is chosen to ensure that the valvewill close within the instrument range considering the most adverseapplication of the 12.55'L 1nstrument error.

The more critical setpoint 1s that which opens the miniflow valve, as500 gpm is the minimum required flow to protect the RHR pumps fromoverheating. The "open" setpoint of 17.57 inches water columncorresponds to a true flow of 728 gpm across the orifice plate at68 F water temperature. If this instrument setpoint drifts to anegative 12.55'i of span, it will still open the miniflow valve at 500gpm. The less critical setpoint 1s that which closes the miniflowvalve. The only requ1rements, while considering instrument error,are that the valve will close within the instrument range and thatthe "open" and "close" setpoints will not cross over one another.The "close" setpoint of 1398 gpm combined with the worse caseinstrument error of +12.55'L of span results in the required 1500 gpmsetpoint within instrument range.

3. R v fr rvi F m

1 1 N. -2DCP M-42312 Rev. 0 (Unit 2)

As a result of the July 15, 1987, steam generator tube rupture eventat Virginia Power's North Anna Unit l, the NRC issued NRC BulletinNo. 88-02, "Rapidly Propagating Fatigue Cracks 1n Steam GeneratorTubes." In response to th1s NRC Bullet1n an evaluation was performedby Hestinghouse which identified four tubes (RlOC43, RllC43, R12C43,and RlOC37) 1n Unit 2 steam generator 2-3 as being susceptible tofatigue cracking. To preclude the potential for a tube rupture dueto fatigue cracking PGhE removed the four tubes from serv1ceutilizing a combination of a cable-type tube damper and solidmechanical plugs. In addition the tubes 1mmediately ad)acent tothose w1th the dampers were removed from serv1ce by the installationof a mechanical tube plug 1n the Hot Leg and a sentinel tube plug inthe Cold Leg. (In the case of one tube (RllC42) a mechanical tubeplug was installed on the Cold Leg end in error.)

3014S/0081K

f v i n mmr

The structural integrity of a tube with a cable-type tube damperinserted in its bore is expected to be, at a minimum, equivalent to aplugged tube. It is anticipated that installing the cable-. type tubedampers and removing ad]acent surrounding active tubes from servicethrough the installation of, sentinel, mechanical plugs will

eliminate'he

potential for a steam generator tube rupture event similar tothat which occurred at North Anna Unit 1. Should a rapidlypropagating fatigue crack develop in a tube in which a cable-typetube damper has been inserted, the worst case scenario would bepossible interaction with an ad]acent tube which has been removedfrom service using a sentinel plug. The expected primary tosecondary leakage should this ad]acent tube wear through-wall isapproximately 270 gpd, which ts significantly below Diablo CanyonUnit 2 Technical Specification allowable primary to secondary leakagelimit of 500 gpd per steam generator. Additionally an evaluation was

.performed and determined that there is no safety impact on the DiabloCanyon FSAR Chapter 15 LOCA and Non-LOCA safety analyses as a resultof the removal of steam generator tubes from service as madenecessary by the installation of the cable- type tube vibrationdampers (and associated sentinel plugs).

4. R v 1 frm rvi F rB ll in . - 2

DCP H-41601 Rev. 1 (Unit 1)

n r r NR

As a result of the July 15, 1987, steam generator tube rupture eventat Virginia Power' North Anna Unit 1, the NRC issued NRC BulletinNo. 88-02, "Rapidly Propagating Fatigue Cracks in Steam GeneratorTubes." In response to this NRC Bulletin an evaluation was performedby Westinghouse which identified four tubes in Unit 1, SteamGenerator 1-1 (R9C41, RlOC47, RllC47 and R9C60) as being susceptibleto fatigue cracking. To preclude the potential for a tube rupturedue to fatigue cracking PGhE removed the four tubes from serviceutilizing a combi nation of a cable-type tube damper and solidmechanical plugs.

f v i r

The structural integrity of a tube with a cable-type tube damperinserted in its bore is expected to be, at a minimum, equivalent to aplugged tube. It is anticipated that installing the cable-type tubedampers will eliminate the potential for a steam generator tuberupture event similar to that which occurred at North Anna Unit l.It is concluded that the presence of the as-installed cable-type tubedamper will not adversely impact plug retentivity and/or primarypressure boundary integrity. Operation of Diablo Canyon Unit 1,following insertion of the cable-type tube damper in potentiallysusceptible tubes, is not expected to result in a previouslyunanalyzed accident.

3014S/0081j'

r

5.

Since use of a cable-type damper is not expected to permit asusceptible tube to interact with an adjacent tube the safety marginof the plant as defined in the bases of the DCPP TechnicalSpecifications for the maintenance of steam generator tube integrityremains comnensurate with Regulatory Guide 1.121 criteria and is notreduced . Additionally an evaluation was performed and determinedthat there is no safety impact on the Diablo Canyon FSAR Chapter 15LOCA and Non-LOCA safety analyses as a result of the removal of steamgenerator tubes from service as made necessary by the installation ofthe cable-type tube dampers.

DCP H-42453 Rev. 0 (Unit 2)

Safety Injection pump 2-2, manufactured by Pacific Pump (Hodel JTCH,Serial No. 45500) was replaced with a Hodel 3TCH, Serial Ko. 51890,from the same manufacturer. This change was required because pumpSerial No. 45500 had become inoperable.

f v 1 i n mm r

6.

The new pump is from the same manufacturer and has the same modelnumber as the old pump. There are only minor differences between thenew and the old pumps due to product line evolution. The old

pump'as

manufactured in 1968 and the new pump was manufactured in 1982.The head capacity curve and pressure retaining parts of the new pumpmeet or exceed the requirements of the old pump and those shown onFigure 6.3-3 of the SAR. The new pump is seismically qualified tothe Diablo Canyon plant specific seismic spectra.

1 m n f m r n w mi F -11DCP J-37477 Rev. 1 (Unit 1) and 3-38477 Rev.l (Unit 2)

This design change replaced the existing emergency borate flowmeter(Taylor Hagnetic Flowmeter) (FT-113) with Controlotron flowtransducers environmentally qualified for post-LOCA conditions. Thischange was performed to meet Regulatory Guide 1.97 ceanitments to theNRC.

f v n

The replacement of the Taylor Magnetic flowmeter with environmentallyqualified Controlotron flow transducers involved no functional ordesign changes to the system. The new Controlotron flow transducerswere environmentally qualified by Hyle Laboratories and meetRegulatory Guide 1.97 requirements. The new flow transducers do notchange the function or ability of the reactor makeup system or thechemical and volume control system to perform their functions.

3014S/0081K

/

7.v n p

DCP M-35469 Rev. 2 (Unit 1) and M-36469 Rev. 1 (Unit 2)

This change removed the autoclosure interlocks (ACI) on residual heatremoval syste~ (RHRS) suction valves 8701 and 8702 and added anindependent control room alarm.

v 1 i mm

The NRC Staff approved the removal of the ACI in NRC letter datedFebruary 17, 1988, "Safety Evaluation of Removal of RHR AutoclosureInterlock Function and Installation of an Alarm at Diablo CanyonUnits 1 and 2'(TAC NOS. 66030 and 66031)."

Hith removal of the ACI the open permissive interlock remainedintact. An alarm was added to each valve, which will actuate if thevalve is open and if RCS pressure is above the RHR valve openpermissive setpoint. In addition, the status lights. on theoperator's panel, which indicate that these valves are open or

. closed, wi 11 remain functional after power has been removed formthese valves.

Removal of the ACI was based on the results of Hestinghouse reportHCAP-11117, Rev. 2, that applies probabilistic risk assessmentmethodology to the original design, the proposed design, and analternative design. Based on,this probabi listic risk assessment thefrequency of an interfacing system LOCA is decreased (from6.2E-07/year to 5.8E-07/year) with removal of the ACI and theaddition of an independent control room alarm to alert operators for

. prompt corrective action.

8. 1 m n f h ili 1 H r m

DCP M-39834 Rev. 0 (Units 1 and 2)

This design change replaced the Auxiliary Salt Hater (ASH) pumpimpeller with one of a slightly larger diameter. The replacementimpeller diameter is 25-1/4 inches as compared to the originalimpeller diameter of 24-1/2 inches. Both impellers were manufacturedby the same manufacturer and constructed from the same type material.

Other than the impeller diameter change, there is no change inmaterial, fit, function or other characteristics. Per themanufacturer, the new impeller diameter is within the original designlimits of the pump. A review of the potential effects this increasein impeller diameter would have on the auxiliary saltwater systemsystem, pump room ventilation, pump motor, bus, and diesel generatorsloadings was also performed. This review determined that the

3014S/0081K -10-

9.

increased impeller diameter would have no adverse effect on plantoperation. An evaluation to determine the impact of increasedhorsepower upon the usage of fuel oil by the emergency dieselgenerators was also performed. This evaluation determined that theincrease 1n impeller diameter resulted in a small increase in dieselfuel oil consumption. However, this small increase was eas11yaccomnodated within the margin of the fuel consumption calculation.

DCP H-40515 Rev. 0 (Unit 2)

This design change performed the following modifications to thereactor cavity manipulator crane: l) The Thyristor control systemwas replaced with a variable frequency control, 2) The existingpilot generators on the bridge and trolley drive motors were replacedwith inching motors, and 3) A bridge camera/monitor positionindication system was added.

f v l i n mm r

The above modifications do not directly impact the handling of fuelassemblies or control rods. The changes are used for assisting thecrane operator in the exact positioning of the bridge and trolley,and to control the current to the bridge, trolley and hoist motors.

The new variable frequency system controls the current to the bridge,trolley and hoist motors. Should the variable frequency controlfail, the bridge and trolley would merely stop in place, not caus1ngany safety concern.

The addition of the video indicat1ng system for the bridge and theinching system for the trolley and bridge improves the operations ofthe crane by allowing the operator to locate the precise assemblylocation with greater ease and less chance of impacting the fuelassembly. All existing 1nterlocks of the manipulator crane, such as,the existing interlock which prevents movement of the bridge ortrolley until the hoist is in the full up pos1tion assuring a fuelassembly is fully removed from the core are unaffected by thesechanges.

The manipulator bridge crane is seismically qualif1ed for structuralintegrity per HCAP 9496. A review of HCAP 9496 indicated that theadded weight to the manipulator crane as a result of this designchange 1s m1nor and w1ll not impact the existing structural seismicqualification of the crane.

10. h 1 TrnfrDCP H-39441 Rev. 2 (Unit l)This change replaced the fuel transfer system air motor andassociated chains, sprockets, air lines, underwater limit switches

3014S/0081K — ll—

and control panel with an electric motor and winch with control panelincluding programmable limit switches. The electrical motor andwinch, as with the existing air motor, is a drive mechanism used tomove the transfer'art from the containment, through the transfertube to the fuel handling building and return during refuelingoutages. The transfer cart carries one spent or new fuel assembly.

f i n

The air motor and chain drive, including the associated limitswi tches, were located on the bottom of the reactor cavity. The newelectric motor and winch are mounted on a platform, over the fueltransfer tube side of the reactor cavity, on the 140 foot elevationinside containment. The reactor vessel and the CRDHs are located asufficient distance away that, should the structure fail, it wouldnot impact the reactor. The failure of the platform was alsopostulated during the movement of fuel assemblies. Based on thisevaluation a complete release of radioactive isotopes from one spentfuel assembly was assumed. This evaluation determined that even witha complete release of the radioactive isotopes from a spent fuelassembly the resultant dose would still be below 10 CFR 100 limitsand in accordance with existing fuel handling accident analyses inthe FSAR. The fuel transfer system is not a safety-related systemand will not effect any previously evaluated accidents or equipmentimportant to safety. This system is classified as Class III.

11. h n P i i n f h n F 1 P 1 Bri nDCP H-39779 Rev. 0 (Unit 1) and M-40779 Rev. 0 (Unit 2)

This change swaps the hoists on the spent fuel pool bridge crane toallow access to all of the cells in the new high density spent fuelstorage racks. In the old configuration, one row of cells on theeastern edge of the pool and one row on the western edge were notaccessible to either hoist on the bridge crane. In the newconfiguration the long hoist and the short hoist are swapped so thatthe long hoist is on the north monorail and the short hoist on thesouth monorail. This new configuration allows access to all cells inthe new high density spent fuel storage racks.

The spent fuel pool bridge crane is not safety related. However,because of its proximity with spent and new fuel the crane wasseismically evaluated to ensure no system interaction. Thisevaluation was reviewed with respect to the above modification andfound to have no impact on the crane's qualification.

The weight lifting capacity of the crane has not been changed withthe above modification so the weight limits addressed in the bases offor the Technical Specifications are not affected.

3014S/0081K - 12-

12.R vl f rf n n r r mn

DCP M-37800 Rev.0 (Units 1 and 2)

In order to provide less costly creek and well water to the raw waterreservoirs the clarifier was removed and a pretreatment system wasinstalled in its place. The pretreatment system will provide 100 gpm

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high quality domestic water to the raw water reservoirs. Thereservoir water will then be treated by the makeup water system.This system is capable of producing up to 600 gpm ofdeoxygenated/demineralized water for makeup to the condensate storagetank or primary water storage tank.

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The installation of the pretreatment system is located near the rawwater reservoirs. The pretreatment system does not interface with anysafety related equipment and was installed as 'a non-safety relatedsystem.

13. In 1 in f w rRvr m

DCP H-31394 Rev. 7 (Units 1 and 2)

A reverse osmosis system was added to provide a reliable source ofmakeup water to support Unit 2 startup while Unit 1 is in operation.

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14.

The addition of the reverse osmosis system does not interface withany safety related equipment and was installed as a non-safetyrel'ated system. During operation, the reverse osmosis systemproduces, as a byproduct, brine having twice the salt content ofstandard seawater. This brine will be adequately diluted by thescreen wash discharge. Hhen the screen wash system is not in use,ocean wave action occurring at the discharge point is more thansufficient to properly dilute the brine. Hazardous chemicals used bythe reverse osmosis system will be stored and used in a metalbuilding to house the system. The building will have a separate,self-contained sump, and storage system. Any spills or system leaksof hazardous chemicals will be contained by the building sumpsystem.

n r 1 k nl r rDCP J-41078 Rev. 0 (Unit 1) and J-42078 Rev. 0 (Unit 2)

This change modifies the setpoint of the Volume Control Tank (VCT)hydrogen inlet pressure regulator, PC-955, from 18 psig to 23 psig.This change allows PGhE to keep the required equilibrium hydrogenconcentration in the reactor coolant system between 25-50 cc/kg water.

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Allowing the hydrogen pressure to 1ncrease in the VCT from 18 to 23psig increases the charging pump suct1on piping pressure and thevarious systems feeding into it (RCP seal water return, SIS from theRWST, LPSI from the RHR HX outlet header, bor1c ac1d pump discharge,and rel1ef valve discharge from the letdown HX) by 5 ps1. Th1sincrease in pressure results 1n the potential for back flow of theseal water return system during plant shutdown. Th1s back flow couldpotentially cause crud from the letdown filter to flow into the RCPseals. To eliminate this from happening DCPP operating procedure OPL-5, requires closure of RCP seal leak off valves 8141A, B, C, and D

when RCS pressure goes below 100 psig.

The VCT was also evaluated to ensure that the increase in pressure to23 psig was within the design pressure of the VCT. The designpressure of the associated CVCS piping and the various CVCScomponents were also examined and found to have design pressuresgreater then 23 psig. The high alarm set point of the VCT was alsomodified from 65 psig to 60 psig to allow for a greater margin ofsafety between a high pressure alarm signal and the VCT designpressure of 75 psig. Also, due to the pressure in the VCT beingraised the setpoint of the VCT vent pressure controller PC-190 isallowed to be ra1sed from 20 psig to 25 psig.

15. ll N w rDCP M-31753 Rev. 10 (Unit l) and M-32753 Rev. 2 (Unit 2)

This change provides a new chemistry lab in the west buttress of theUnit l turbine building. The new lab is sound proofed and designedto be used for the analysis of secondary side water fornon-radiological contamination.

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16.

The new chemistry lab is strictly for the analysis of non-radiologically contaminated secondary water. The lab 1s physicallyisolated from all safety related systems. Increased buttressstructure loading was evaluated and determ1ned to have a negligibleeffect on the se1smic analysis of the safety related and important tosafety equipment. The plant fire protection system was extended tocover the new lab facility. In the event of s1gnif1cant primary tosecondary leakage, radiation elements are located in the blowdownlines to cause the 11nes to be 1solated, thereby conta1ning thecontaminated samples within the radiologically controlled area. Anevaluation of the tox1c chemicals used 1n the lab was performed andfound to have no impact on control room habitability analyses.

f n- fDCP M-35366 Rev. 0 (Units l and 2)

This modification made permanent the two rotary screw compressorsbeing used at DCPP. One of the compressors was temporarily installed

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17.

near the permanent reciprocating air compressors and the second wastemporarily placed in the CCH heat exchanger tube pull space. Thischange permanently installs the two screw compressors along with theinstallation of a new air dryer and after filters in the vicinity ofthe permanent reciprocating air compressors.

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The two rotary screw compressors provides compressed air for process*control systems and for station service throughout Units 1 and 2,under normal operating conditions. They are not required for reactorprotection, containment isolation, or engineered safety features.The permanent location of the two rotary screw compressors to thevicinity of the permanent reciprocating air compressors has beenevaluated with respect to high energy line breaks, fire protection,floor loading, seismic interaction, and heavy loads handling anddetermined to have no impact on safety related systems or components.

r 1 m 1 rDCP H-35020 Rev. 0 (Units 1 and 2)

This modification abandons in-place, four electrical duct heatersprovided for personnel comfort in the control room HYAC system. Heatgenerated by equipment within the control room and heat emanatingfrom the cable spreading room below have proven adequate to maintainthe control room within a temperature range judged adequate by humanfactors standards.,

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The heating coils for the Electric duct heaters, abandoned in- place,EH-21, 22, 24, and 25, are Design Class II. However, to preserve theintegrity of the control room HVAC system the heater casings aredesigned Class I. Abandonment of the heaters in-place preserves theintegrity of the HVAC system flow boundaries.

n f r f r rDCP E-40367 Rev. 1 (Unit 2)

This change added a spare 25KV/500KV step-up transformer just southof the three existing Unit 2, 500 KV main transformers. The sparetransformer was placed on an existing concrete pad and provided with480 V non-vital power to the control cabinet.

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The spare transformer was installed as a back-up transformer toUnit 2 and will not be energized at the rated voltage of 25 KV. Afire protection deluge system is provided in case of a fire to any ofthe oil filled equipment and to prevent fire damage from the adjacentin-service step-up transformers. The spare step-up transformer isnot safety related and does not physically or electrically affect anysafety related equipment or systems.

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19. n l 1 n* f r m

KP H-39881 Rev. 3 (Unit 1) and M-40881 Rev. 0 (Unit 2)

This change 1nstalled a secondary bor1c acid system. The systemconsists of two boric acid mixing/feed tanks, three 5(C feed pumpsand a screw feeder for loading boric acid to the mixing/feed tanks.The system 1s used to prepare a 4X bor1c acid solution and ingect thesolution into the steam generators via the condensate/feed system.This system was added to reduce steam generator tube denting.

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The secondary boric acid system is not safety related nor is itrequired to ma1ntai n the reactor in safe shutdown. Hone of theequipment is powered by vital buses or has direct interface withsafety related equipment or piping. Impact of the secondary waterchemistry on the potential for corrosion of steam generator tubes andturbine components has been evaluated by Westinghouse. Severalmaterials (Alloy 600, Alloy 800, SA 285 Carbon Steel, Type 304Sta1nless Steel, 90:10 Copper Nickel and Admiralty Brass) have been

. tested by Westinghouse in AVT and faulted AVT with 100 ppm boron asboric acid at feedwater temperatures. The corrosion rates of allmaterials was very low, up to 0.1 mil/year. Add1tionally, materialstypical of those used in piping systems, carbon and low alloy steel,have been tested and found to have an acceptably low corrosion ratein AVT chemistry with boric ac1d added. Based on the above1nformation the previously analyzed consequences of potentialexcessive corrosion, e.g., tube rupture, feedline break, turbinemiss1les have not been increased nor have the probab1lities of suchpostulated events been increased.

20. R l n r in v rDCP H-37759 Rev. 0 (Units l and 2)

This change replaces the existing equipment drain and floor drainreceiver pumps with larger capacity pumps to satisfy increaseddemands on the liquid radwaste system.. This change increases thesystem throughput and, therefore, its capacity, but the basic systemoperation remains the same.

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The liquid radwaste system is non-safety related and 1s not reliedupon to mitigate an accident. With the exception of 1ncreasing theequipment drain and floor drain receiver pumps head capacity froml22 feet to 300 feet and minor piping and valve changes associatedwith the pump replacements no other s1gnificant modif1cations weremade to the liquid radwaste system. Th1s change did not crossconnect the liquid radwaste system with others systems that haveaccident mitigating functions, increase the. chance for flooding or

3014S/0081K. -16-

create new SISI issues from previous designs. This modificationmeets the seismic, quality group and qual1ty assurance criteria ofRegulatory Guide 1.143, and has no impact on the consequences of anunexpected or uncontrolled release of radioactive material from thesystem. The new pump capacities do not exceed the flow requirementslisted in the FSAR Update.

2l. n w rKP M-29953 Rev. 0, KP H-29953 Rev. 3 (Un1ts l and 2)

A new radwaste storage building (RHSB) was constructed to provide atemporary,ons1te storage facil1ty for packaged low level radioactivewaste. The new RHSB consists of a control room, HVAC room, rail carand truck bay, two solidification waste liner storage areas, a dryactive waste storage area, and a liner inspection and decontaminationarea. The RHSB will provide temporary storage for low levelradioactive waste in the event bur1al s1tes are closed or 1n theevent of exclusion of Oiablo Canyon waste form burial sites.

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The new RHSB does not support any safety functions and the structuresand systems are designed as Class II. Building conf1guration,radiation monitoring, isokineti c sampling and HVAC filters areprovided to ensure personnel exposure is maintained as low asreasonably achievable (ALARA). The dose rate in the continuouslyoccup1ed,control room is less than 0.5 mr/hr. Adequate shielding 1sprovided for restricted access areas to keep the dose rates ALARA.Radioactivity inside a container cannot escape from the containerduring normal storage. The dry active waste will be packaged in55-gallon drums and/or 96 cu-ft carbon steel boxes (LSA containers).

The new RHSB and its associated systems and components are designedto control a1rborne and effluent releases and to facilitate access,operation, inspect1on, testing and ma1ntenance. Offs1te radiationdoses from the onsite storage are controlled and kept at a smallfraction of the 40 CFR 190 limits.

22. v l rmm niKP E-ll868 Rev. 0 and E-25781 Rev.0 (Units 1 and 2)

This change removed area rad1ation monitor RE-B, which was located inthe area previously used as a liquid radioactive waste drummingstation. Airborne act1vity could be created during the drummingprocess. The drumming station was removed and a dry waste compactorand liqu1d radwaste filters were installed in this area.

The new equipment in this area is normally not capable of releasingradioactive materials that produce airborne radiation. The filters

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part of a sealed piping system, the dry waste (LSA) compactor hasan internal fan and HEPA filter and is connected to the exhaust(vacuum) side of the building ventilation system. There are nocomnitments/requirements for such an area to be monitored as RE-8provided.

DCP J-40053 Rev. 1 (Unit 2)

Due to the former globe valves leaking, the pressurizer spray valveswere replaced with new, ball valves. To facil1tate their1nstallation, two inact1ve pipe whip restraints were removed.

These valves, though forming a portion of the Reactor CoolantPressure Boundary (RCPB), are not act1ve in the mitigation of anydesign basis events. Their ability to maintain the 1ntegrity of theRCPB has been demonstrated by their design to require loadingconditions (including thermal transients and seismic) and assuranceof materials compatibi 11ty. The changes to the valve and pipingconfigurations have been evaluated and shown to not adversely affectsystem hydraulics, integrity or control or 1nservice inspection andtesting.

Removal of the 1nactive whip restraints does not compromise piperupture mitigation; containment heat sink calculations were notcompromised by the removal of this minor amount of steel insidecontainment.

24. R l nr 1 R mL1hin ilin n Wll n F r vrinDCP E-40333 Rev. l (Unit 2)

The Unit 2 side of the control room was g1ven a facelift per PG&Eceanltments to NUREG 0700 for control room design rev1ew. This workconsisted of: replacing floor cover1ng (carpet); - 1nstalling a newuni nterruptable power supply for the emergency AC lights; andrepainting the control room walls.

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The following 1ssues were addressed and resolved to assure that nounrevi ewed safety question was created; l) Se1smic integrity of thece111ng and 11ght fixtures was verified; 2) Fire resistance of- thenew pa1nt, cei 11ng tile and fixtures meet appropriate requirements;3) Actual combustibles loading 1n the room was reduced; 4) Protectionfor vital electrical power systems and changes to vital electricalpower loads were reviewed and were found acceptable; 5) Hab1tabi lityand integr1ty of the control room envelope was not adversely impacted.

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25. R v Hi h P r in vDCP H-39126 Rev. 0 (Unit 1)

The cover over the high pressure turbine has been removed to astorage area outside the main power block to provide better operationand maintenance accessibility.

No credit for the presence of the cover was taken in the turbine-generator missile analysis for the plant. Per the FSAR, the coverdid provide protection from seismically-generated, falling materialsfor the turbine overspeed trip system and high pressure stop valves.

Further review of the potential for this falling material revealedthat either: 1). the postulated dislodging of roof panels due to aseismic event is not credible based on calculation; or 2) thepostulated, smaller debris (rivets, etc.) could either not impact theoverspeed trip system components or could not render them incapableof performing their safety (trip) function.

26. R 1 R r 1 n R i n T m r rDCP 3-32961 Rev. 1 (Unit, 2)

This change replaced the reactor coolant RTDs. The RTDs werereplaced because their environmental qualification was only good forone fuel cycle. The replacement RTDs are qualified to function,within the expected radiation field for 40 years.

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The replacement RTDs comply with specifications of the original RTDsand are qualified for 40 years. The temperature range, response timeand accuracy of the replacement RTDs is equal to or better than theoriginal RTDs.

27. R 1 m n rF w r i -TDCP H-36476 Rev. 1 (Unit 2)

This change replaced the original carbon steel steam generatorfeedwater ring J-tubes with inconel J-tubes. Inconel provides asuitable combination of resistance to corrosion and to stresscorrosion cracking.

The inconel J-tubes increase the reliability of-the steam generatorsby providing increased resistance to erosion/corrosion of the J-tubenozzles. Thus, enhancing the safety related function of theauxiliary feedwater system. The new J-tubes have the same geometryas the original J-tubes and are compatible with the serviceconditions.

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28. i f h li Pr inDCP E-32853 Rev. 0 (Unit 2)

This change modifies the Solid state protection system (SSPS) fromone-out-of-four (1/4) logic to two-out-of-four (2/4) for auxiliarycontact input signals from the reactor coolant pump circuit breakerscoincident with permissive P7.

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Reactor coolant low flow reactor trip logic is not changed by thismodification. This design change only removes an anticipatoryreactor trip for low reactor coolant flow as noted in LAR 85-08.

29. n n wn FiDCP M-28132 Rev. 3 (Uni't 2)

This change adds a parallel, redundant filter for reactor coolantletdown to the volume control tank to allow removing demineralizeddebris even with one filter out of service.

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To maintain piping integrity, the new piping and supports aredesigned and constructed to equivalent codes as the original filter.ALARA considerations are incorporated into the design. Ko netincrease in radwaste system inputs is expected.

30. A k v r rr n Pr i n imn 1 ri 1Pn r inDCPs E-4459 Rev. 8, E-4941 Rev. 2, E-6410 Rev. 0, E 6415 Rev. 5,E-6504 Rev. 1, E-6546 Rev. 3, E-8140 Rev. 0, E-10010 Rev. 1,E-12426 Rev. 0, E-12981 Rev. 0, E-14651 Rev. 0, and E-16206 Rev. 0,(Unit 2)

Redundant, appropriately qualified, in-series overcurrent protectionwas installed on electrical circuits penetrating contai nment toassure single electrical fai lure protection for the penetration'sphysical integrity.

This change was requested by the NRC Staff, as documented in theOctober 18, 1978, meeting minutes, during a meeting with them onOctober 12, 1987, and as noted in SSER No. 8, dated November 14, 1978.

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1. Procedure NPAP B-101 (Units 152). Revised to incorporate therequirements of the Hay 26, 1987 Revision to 10 CFR 55

"Operators'icenses".

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The probability of an accident previously evaluated in the FSARUpdate will not increase as the result of the implementation of thisprocedure change because:

The amount of training time has not been reduced.Operators will receive training based on an analysis of theirtraining needs.Operators will receive training on all abnormal and emergencyprocedures every two years.

Consequences of an accident previously evaluated in the FSAR Updatewill not be increased as a result of this change to the procedurebecause:

~ The training is more focused on plant needs and provides addedassurance that operators are trained on infrequently performednormal and emergency tasks.

The possibility of a malfunction or accident of a different type thanany previously evaluated in the FSAR Update will not be created asthe result of the implementing the changes to the procedures because:

~ The intensity of operator training has not been changed. It hasonly been focused more specifically on a )ob analysis. If newabnormal or emergency procedures are added to the plant, thesewill be added to the )ob analysis. The level of training inabnormal and emergency procedure training has not been reduced.

The Hargin of Safety as defined in the bases for any TechnicalSpecification will not be reduced as a result of this change to theprocedure. The procedure change does not. reduce the training hoursreceived by the licensed operators.

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Tm rr Pr T— vA special Leak Rate Test of Containment Spray Additive Tank Valves.

The Containment Spray System is not required to be operable in theMode of operation in which the test will be performed.

The test will not impact other'ystems required to be operable inHodes 5 & 6.

2.

The Technical Specifications require a Boric Acid Storage System(BAST) or the Raw Hater Storage Tank (RHST) to be in operation inHodes 5 5 6. This test will not render inoperable the BAST or theRHST, or their associated injection systems.

m r P r 8- 1 v n 1

A test of the Component Cooling Hater (CCH) Meat Exchanger InletHaterbox Level.

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The CCH outlet temperature wi 11 be monitored and the test terminatedif the CCH temperature reaches 110 degrees F. This will preventreaching the maximum CCH temperature of, 125 degrees F as described inthe FSAR Update, or the alarm setpoint of 120 degrees F. The CCHSystem wi 1 1 be maintained wi thin its design basis temperature duringthe test, so there will be no impact on the FSAR Update accident or

-malfunction analysis.

By maintaining the CCH System within its design limits, and restoringthe Auxiliary Saltwater System if the design limits are approached,the possibility of a different type of accident wi 11 not be created.

The margin of safety will not be affected, since the CCH Systemoperated within its design basis.

3.Residual Meat Removal (RMR) Pipe Vibration Test. This test allowsdifferent RMR flows and valve positions. Data will be recorded ateach different configuration.

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The RMR system will be aligned and operated per design, therefore,there is no increase in the probability of an occurrence or theconsequences of an accident or malfunction as previously evaluated inthe FSAR Update. Because the RMR System wil.l be operating asdesigned there is no possibility for an accident or malfunction of adifferent type than previously evaluated in the FSAR Update.

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This test does not violate the Technical Specifications and willprovide greater than 1000 GPH flow through the RHR Heat Exchangers.

4. T r r Pr r v. niTest Procedure for reducing Carbon Monoxide (CO) in containment;This procedure utilizes the service air for purging CO out of Unit 2Containment.

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Prior to using the Service Air System (which is not safety related)to purge containment of CO, the Instrument Air System will beisolated from the Service Air System, thereby not increasing theprobability or the consequences of a malfunction of equipmentImportant to Safety. The containment pressure will be closelymonitored in the control room and will not be allowed to exceed 0.7psig. This is well within the Technical Specification range of -1.0to +1.2 psig. During the test an operator will be stationed at theService Air Containment Isolation Valve to'solate the service air ifnecessary.

5. T m r r Pr r - R v. ni 1

Test Procedure for measuring BIT manway leakage at operating pressure.

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All valves required for automatic 'BIT injection remain operable andno valves precluding automatic BIT injection will be closed duringthe performance of this test.

6. n r P P -1 R v. 1

1 n T i 1 TPrfr n T n i Pm ni 1 nThe changes to the procedures allow the Safety Injection portions tobe performed in Hode 5 rather than Hode 6. The objective of theSafety Injection portions of STP V-15 is to set cold leg and hot leginjection flows by adjusting runout throttle valves. The objectiveof STP P-lA is to runout flow in order to obtain a pumpcharacteristic curve.

The test will be conducted entirely in the cold shutdown condition,the tests do not alter the probability of occurrence or theconsequences of any accident analyzed in.the FSAR Update.

The procedures do not disable or modify the ability of any safetyequipment to function. The performance of the tests in the coldshutdown condition and the normal use of equipment and source ofwater conclude that no new types of accidents or malfunctions wouldbe created by the performance of these tests in Hode 5.

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Adequate core cooling, reactor vessel level and reactor vesseloverpressurization are controlled procedurally. Hith all safetyequipment functions maintained the Technical Specifications basesthat the pressurizer vent path will protect the reactor coolantsystem for all anticipated low temperature transient is notcompromised.

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1.4 ni r

This temporary jumper disables the auto-initiated "Containment, Evacuation On High Flux At Shutdown" alarm (Source Range NIS)before/after Core alterations.

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The auto-initiated Containment Evacuation alarm is mentioned only inthe Fuel Loading and Initial Operation section (section 14.1.4.1) ofthe FSAR Update. The High Count Rate alarm is discussed in the"Dilution During Refueling" and the "Dilution During Startup"Sections of the FSAR Update with no mention of the auto-initiatedContainment Evacuation Alarm. The analysis of the dilution accidentduring refueling and startup indicates there will be at least 40minutes for operator action after receiving the High Count Rate Alarm.

The manual actuation of the alarm is not affected. The jumper isremoved before leaving Mode 4.

m r n wTm r r vr- m

m r N . - iThis temporary jumper allows the comparator for TE-423B which isbroken to be tripped. Once Unit 2 is in Mode 4 and <330'F, theJumper will make up the temperature signal to the PORV-456 circuit.

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This is a temporary jumper used until the RTD (TE-423B) is repairedor replaced.

With the RCS <330'F and with the jumper in place, PORV 456 actionwill still be automatic on Loop 3 Wide Range Pressure signal (PT 405).

The brittle fracture protection of the reactor pressure vesselprovided by the LTOP System is not affected by this jumper.

The LTOP System (PORV's 456 and 455c) remain operable.

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il in1 i h 1

This change will reduce the number of auto-initiated iodine removalIodes resulting in, fewer hours elapsed on charcoal banks, betterequipment reliability.

Reevaluation of accuracy tolerances show that High=- Alarm Setpoint canbe set less conservative and still meet the Technical Specificationrequirements for the Fuel Handling Building Radiation Honltors.

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0'