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ATTACHMENT 1:

PROPOSED TECHNICAL SPECIFICATION CHANGESREGARDING PRESSURE-TEMPERATURE UMITS

(JPTS-89 024) .

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New York Power Author.ity |'

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JAMES A. FITZPATRICK NUCLEAR POWER PLANT~

!Docket No. 50-333 ' ]. . .

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UST OF FIGURES |

Figure Title ' Paje ,

3.11 Manual Flow Control .47a;

3.12 Operating Umit MCPR versus t 47b

4.1 1 ' Graphic Aid in the Selection of an Adequate Interval Between Tests 48.,

'4.2-1- Test Interval vs. Probability of System Unavailability- 87 .;

3.4 14 Sodium Pentaborate Solution 34.7 B 10 Atom % Enriched Volume- 110Concentration Requirements

3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution '111- !

3.5-1- Thermal Power and Core Flow Umits of Specifications 3.5.J.1,3.5.J.2 and .1343.5J.3 ' y

3.5-6 (Deleted) 135d:3.5-7 (Deleted) 135e3.5-8 (Deleted) - 135f =

3.5-9 (Deleted) 135g3.5-10 (Deleted) :135h- -

3.5-11 MAPLHGR Versus Planar Average Exposure Reloads 6 &7, BP8DRB299, 1351

QUAD + ' .

3.5 12 MAPLHGR Versus Planar Average Expohure Reload 7, BD319A i :135]

3.5-13 MAPLHGR Versus Planar Average Exposure Reload 8, BD336A .135k

3.5-14 MAPLHGR Versus Planar Average Exposure Reload 8, BD339A 1351.

3.6-1 Reactor Vessel Pressure - Temperature Umits Through 12 EFPY 163

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3.6 1 Reactor Vessel Pressure - Temperature Umits Through 14 EFPY 163a' Part 2 '

3.6-1 Reactor Vessel Pressure - Temperature Umits Through 16 EFPY 163bPart 3

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4.6-1 Chloride Stress Corrosion Test Results at 500 F 164 q

'6.1 1 (Deleted) 2596.2 1 (Deleted) 260 ;

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Amendment No.14, g, g,64',74', '/4, Nf, Stf,196,1)6,1M,1FT, Mrt

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4.6 SURVEILLANCE REQUIREMENTS ^ .- 3.6 UMITING CONDITIONS FOR OPERATION

4 6 REACTOR COOLANTSYSTEM-3.6 REACTOR COOLANTSYSTEM

W DApplicability: ,..

Wes to h p exWh W testing Wrm fm WApplies to the operating status of the Reactor Coolant System. - Reactor Coolant System.

ectim Objectwe:

T assure the integrity and safe operation of the. Reactor Coolant To Me h w&'mWhhwW @mWh' operation of the safety devices related to it.:

Specification:Specification:

A. Pressurization and Thermal Umits- A. Pressurization and Thermal Limits

1. Reactor Vessel Head Sud Tensoning1. Reactor Vessel Head Stud Tensioning

Vfwi in the cold conditon, the reactor vessel head RangeThe reactor vessel head botting studs shall not be under and the . reactor vessel flange temperatures shall be

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tension unless the temperatures of the reactor vessel mcw M- flange and the reactor head flange are at least 90 F.

a. Every 12 hours when the reactor vessel head Range' is <1207 and the studs are tensioned.

b. - Every. 30 mmutes when the reactor vessel head' '

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flange is <1007 and the studs are tenseoned.

2. ' in-Service Hydrostatic and Leak Tests __ c. Within 30 minutes prior to and every 30 minutes' t% dew W W W

' During in-service hydrostatic or. leak' testing the Reactor'

- Coolant System pressure and temperature shall be on or;I to the right of curve A shown in Figure 3.6-1 Part 1,~ 2, or 3 2. ' Imh @wtatic and M Tests

and the. maximum temperature change during any one -- hour penod shall bei Dunng hydrostatic and leak testog the Reactor C(mlant

System pressure and temperature shall be recorded every30 minutes until two consecutwe. temperature readings arewithin 5*F of each other.

- Amendment No. M, tFJ ~136

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a. <20Pwhen to the left of curve C. ,

b. <100 Fwhen on or to the right of curve C.

3. Non-Nuclear Heatup and Cooldown t

3. ~ Non-Nuclear Heatup and Cooldownby non- means, h WDuring - heatup - by : non-nuclear rneans (mechanical), Mear Wh W W poww % tests,the rMw .

cooldown following nuclear shutdown and low power ant @mn pmsswe W tanpwatwei M; kphysics tests the Reactor Coolant System pressure and- .

re m may 30- W es. W h6temperature shall be on or to the right of the curve Btanpwatwe rMngs are m WW each dhw.

| shown in Figure 3.6-1 Part 1,2, or 3 and the maximumtemperature change during any one hour shall be < 100*F.

4. Core Critical Operation4. Core Critical. Operation

During all modes of operation with a critical core (except ng an e d wh 2 a crh cwe (ex$.fa poww @ tests) h e W Wfor low power physics tests) the reactor Coolant System

pressure and temperature shall be at or to the right of the F.esswe W tempwatwe M be M N 30'l- curve C shown in Figure 3.6-1 Part 1, 2, or 3 and the inMes W to @awal d W % 2 % h '

mactw W W may 30 Mes dunng W W -

maximum temperature change during any one hour shall.be <1007. . . two e tempwam medings am & W d

- each other.5. ..With. any:of the limits of 3.6.A.1: through 3.6.A.4 above

exceeded, either-

a.- restore the temperature and/or pressure.to within- the limits within 30 minutes, pa16 n an engineering

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evaluahon to determine the effects of the out-of-limit .~

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. condition on the structural integnty of the_ reactor -| coolant system, and : determine that the; reactor|- coolant system remains acceptable for continuedi

operations; or .: .~

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3.6 and 4.6 BASES (cont'd)- ..

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The expected neutron fluence at the reactor vessel wall can be vessel flange region and for the reactor vessel shell_belthnedeterrnined at any point during plant life based on the linear ' region are 30*F, based on fabrication test reports. The RTNOT _

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relationship between the reactor thermal power output and the for the remainder of the vessel is 40*F.corresponding number of neutrons produced. Accordingly,-neutron flux wires were removed from the reactor vessel with . The first surveillance capsule containing test spm,inws was'-

the surveillance specimens to establish the correlation at the withdrawn in April,1985 after 6 EFPY. The test spm.iiYws -'

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removed were tested according to ASTM E 18542 and thecapsule iocauen by expeiiniental methods. The flux 'distribution'at the vessel wall and 1/4 thickness (1/4T) depth results are in GE report MDE-49-0386.. The next surveillance

was analytically determined as a function of core height and capsule will be removed after 15 EFPYs of operation and the ~

azimuth to establish the peak flux location in the vessel and the resWs d the exWioW as a Wm&M Rgurelead factor of the surveillance specimens. :3.6-1 curves A, B and C for operation of the plant after 16 '

EFPYs.Regulatory Guide 1.99, Revision 2 is used to predict the shift in

Figure 3.6-1'is congnsed of three parts: Part 1, Part 2, and -RTNTO as a function of fluence ,n the reactor vessel belthneI Part 3. Parts 1,2, and 3 establish the pressure-temperature

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ixregion. An evaluation of the irradiated suryeillance

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specimens, which were w,thdrawn from the reactor in April, limits for plant operations through 12,14, and 16 Effective Fullii,

Nm less than that predicted Yews (EFW r%tW. The weide @ and1985 (6 EFPY), shows a shift in RT

j. by Regulatory Guide 1.99, Revision 2. esseteme m we W m h Wof accumulated EFPY. Figure 3.6-1, Part 1 is for operation

i Operating ' limits for - the reactor vessel pressure and through 12 EFPY, Rgure 3.6-1, Part 2 is for operation at greater~

i temperature during normal heatup and cooldown, and during ' than 12 EFPY through 14 EFPY,'and Figure 3.6-1, Part 3.is forin-service hydrostatic and leak testing were established using ~ operation at greater than 14 EFPY through'16 EFPY. The:

| 10 CFR 50 Appendix-G, May,' 1983 and Appendix G of the . curves ' contained in Rgure 3.6-1 are developed from thet '

Summer 1984 Addenda to Section lli of the ASME Boiler 'and General Electric Report DRF 137-0010, "Imniementation of; Pressure Vessel Code. These operating limits assure that the Regulatory Guide 1.99, Revision 2 for the JamA A. Fitw trick -a

vessel could safely accuYiiviodate a postulated surface flaw- Nuclear Power Plant," dated June,1989..

having a depth of 0.24 inch at the flange-to-vessel junction, andj one-quarter of the matenal thickness at all other reactor vessel - Rgure 3.6-1 curve A establishes the minimum temperature for

locations and discuiunuity regions. - For the purpose of setting - hydrostatic and leak testing required by the ASME Boiler andPreswe Vessel % M'CJ@ msW' mn

these operating . limits, the reference temperature, RTNOT, -Wostdic W M M'q 'we a Wion d the teW'm-Of

the vessel material was estimated from impact test data taken,

in accordance with the requirements of the Code to which the - tWMure W the wgst Mwial. hQ the,

vessel was designed and manufactured (1965 Edit,on including maximum hydrostatic test pressure will be 1.1 times the 'i

ope ng pressme d M N 4Winter 1966 add nida). The' RTNOT values for the reactor

Amendment No.1'

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Amendment No. ,,

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''VALID 14.TO 16 EFPY

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163bAmendment No. i

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ATTACHMENT 11 -

SAFETY EVALUATION FOR PROPOSED

RE EIIIWilTS2 1

(JPTS-89 024)

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New York Power Authority- ;

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JAMES A. FITZPATRICK NUCLEAR POWER PLANTDocket No. 50-333 '

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l. DESCRIPTION OF THE PROPOSED CHANGES

This application for an amendment to the James A. FitzPatrick Technical Specifications revises 'Specification 3.6.A, " Pressurization and Thermal Umits," and its associated bases to comply withGeneric Letter 88-11 (Reference 1) and Regulatory Guide 1.99, Revision 2 (Reference 2).'

. Specifically, the pressure-temperature curves in Figure 3.6-1 are replaced with new curves foroperation to 12,14, and 16 Effective Full Power Years (EFPY). The associated Umiting .Condition for Operation (LCO) and the Bases Section are revised to reflect the new pressure-temperature curves.-

The specific changes to the Technical Specifications are:

A. Pressure Temperature Umit Changes.

Replace existing Figure 3.6-1, " Reactor Vessel Pressure - Temperature Umits," on page 163:with the following new figures:

Figure 3.6-1 Part 1, " Reactor Vessel Pressure Temperature Umits Through 12 -EFPY," on page 163

Figure 3.6-1 Part 2, " Reactor Vessel Pressure Temperature Umits Through 14. EFPY," on page 163a

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- Figure 3.61 Part 3, " Reactor Vessel Pressure Temperature Umits Through 16 - |EFPY," on page _163b

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B. Associated Wording Changes1

-t1. Section 3.6.A.2, "In-service Hydrostatic and Leak Tests," page 136;

- Section 3.6.A.3, "Non-nuclear heatup and Cooldown," page 137;Section 3.6.A.4, " Core Critical Operation," page 137:

Replace " Figure 3.61" with " Figure 3.6-1 Part 1,2, or 3"

2. Bases Section 3.6, " Pressurization and Thermal Umits," page 147:

Replace second paragraph on page 147 (begins with "A method of relating ...") 'e

with:

Regulatory Guide 1.99, Revisl.on 2 is used to predict the shift in RTNTD

as a function of fluence in the reactor vessel beltline region. Anevaluation of the irrac'iated surveillance specimens, which werewithdrawn from the reactor in April,1985 (6 EFPY), shows a shift inRT lNTD ess than that predicted by Regulatory Guide 1.99, Revision 2.

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SAFETY EVALUATION -Page 2 of 6

Delete the third sentence in the fourth paragraph on page 147. The sentence to bee

' deleted reads as follows:'

The curves of Figure 3.6-1, A through C, reflect findings in the report jrelated to copper phosphorus content of the reactor vessel shellbeltline, flux wire testing fluence distribution analysis, and Charpy V-

- Notch specimen testing.

Add the following paragraph between the fourth and fifth paragraphs on page 147. -.

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Figure 3.6-1 is comprised of three parts: Part 1, Part 2, and Part 3.Parts 1,2, and 3 establish the pressure temperature limits for plantoperations through 12,14, and 16 Effective Full Power Years (EFPY),

respectively. The appropriate figure and the pressure-temperaturecurves are dependent on the number of accumulated EFPY.~ Figure:3.6-1, Part 1 is for operation through 12 EFPY, Figure 3.61, Part 2 is for

. operation at greater than 12 EFPY through 14 EFPY, and Figure 3.6-1,Part 3 is for operation at greater than 14 EFPY through 16 EFPY. Thecurves contained in Figure 3.61 are developed from the GeneralElectric Report DRF 137 0010, " implementation of Regulatory Guide

11.99, Revision 2 for the James A. Fitzpatrick Nuclear Power Plant,"dated June,1989. |

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I$ki* 3. Ust of Figures, page vil: replace " Figure 3.61, Reactor Vessel Pressure - TemperatureUmits," page 163 with

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Figure 3.6-1, Part 1, " Reactor Vessel Pressure -' Temperature Umits 1-

Through 12 EFPY," page 163

Figure 3.61, Part 2, " Reactor Vessel Pressure Temperature UmitsThrough 14 EFPY," page 163a, '

Figure 3.6-1, Part 3, " Reactor Vessel Pressure - Temperature UmitsThrough 16 EFPY," page 163b :

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II. PURPOSE OF THE PROPOSED CHANGES

Regulatory Guide 1.99, Revision 2 (Referenco 2) revised the methodology used to evaluate - ineutron radiation embrittlement of reactor vessel beltline materials. Generic Letter 88-11(Reference 1) requests licensees to use Revision 2 of the regulatory guide to evaluate predicted -

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embrittlement.- The Authority has reevaluated the effect of neutron radiation on reactor vesselmaterials (Reference 3) and is changing the pressure temperature limits contained in the -

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Fitzpatrick Technict.1 Specifications. This proposed change is consistent with the requirements'

of Section V of 10 CFR 50, Appendix G..{

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Background

Specification 3.6.A, * Pressurization and Thermal Umits," establishes, in part, pressure-temperature curves which define the minimum pressure and temperature for three reactoroperating conditions: 1) system hydrostatic and leakage tests,2) heatup and cooldown, and 3)core critical operation. These pressure-temperature curves protect the reactor pressure vesselfrom brittle failure by clearly identifying the regions where the vessel is subject to brittle fracturefailure modes.

' Amendment 113 (Reference 4) revised the pressure-temperature limits to be consistent with testresults and analyses performed on the irradiated surveillance capsule removed from theFitzpatrick reactor in April,1985. (Surveillance capsules are installed in the reactor vessel beforestartup and contain test specimens that are made from the plate, weld, and heat affected zonematerials of the reactor beltline.) Radiation embrittlement was calculated using the surveillancedata and adjusting the nil-ductility reference temperature (RTNDT ) in accordance with RegulatoryGuide 1.99, Revision 1 methodology.

The effect of neutron radiation on reactor vessel materials has been recalculated (Reference 3)in accordance with Generic Letter 88-11 and Regulatory Guide 1.99, Revision 2. The resultantshift in RT bounds the previously calculated results for the beltline region of the core.NTO

New beltline Pressure-Temperature curves were developed for operation to 12,14, and 16Effective Full Power Years (EFPY). The non-beltline region curves (recirculation inlet nozzles andhead flanges) are not affected by the changes to RTNTD shift associated with Regulatory Guide1.99, Revision 2.

The beltline curves apply to the vessel plates and welds and are limiting above 500 psig. Forexample, at a 1000 psig on the leak test curve, the required test temperature is 192 F for 16EFPY compared to the current limitation of 157 F.

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||1. IMPACT OF THE PROPOSED CHANGES

The purpose of Specification 3.6.A, " Pressurization and Thermal Umits," is to establish operatinglimits that provide a wide margin to brittle failure of the reactor pressure vessel. The basis of thePressure-Temperature (P-T) limits is found in Appendix G to 10 CFR 50 and in Section 4.2 of theupdated FSAR. The limits are not derived from the design basis accident analyses, but areprescribed to avoid encountering pressures, temperatures, and temperature rate-of-changeswhich might cause undetected flaws to propagate.

The first technical specification change lowers the P-T curves (i.e., a higher temperature is |required for a given pressure) which in turn " narrows" the reactor coolant system operatingwindow and lengthens the time required for hydrostatic testing. This change is consistent withGeneric Letter 88-11 to ensure a conservative margin to non-ductile failure.

The new P-T curves were developed for three service periods: 512 EFPYs,514 EFPYs, and 516 EFPYs. The use of three curves instead of one lessens operational impacts by phasing theincreases in minimum temperature over three distinct service periods. Each set of P T curves is I

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SAFETY EVALUATION*

Page 4 of 6'

conservative, because the conditions at the end of each servise period (12,14,' or 16 EFPY)' yieldthe highest fluence and, therefore, the largest predicted shift in RTNTD-

The second change revises the text of Sections 3.6.A and its associated Bases. The change alsoupdates the Ust of Figures provided at the beginning of the Fitzpatrick Technical Specifications.These changes are editorial in nature and reflect the new limits on pressure and temperature.

~ Both proposed changes are administrative in nature. They do not involve any physical.

modification to the plant, nor do they introduce any new failure modes. The changes do not alterthe conclusions of the safety analyses contained in the FSAR and the NRC staff's SER.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION

Operation of the James A(RtzPatrick Nuclear Power Plant in accordance.with the proposedamendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, 4

since it would not:

1. involve a significant increase in the probability or consequences of an accident previouslyevaluated. The effect of neutron radiation on reactor vessel materials has been recalculatedusing the latest NRC approved guidance (Regulatory Guide 1.99, Revision 2 methodology).The resultant changes to the pressure temperature limits contained in Specification 3.6.A -

- will preclude brittle fracture failure of the reactor vessel. The requirements on pressure .~

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temperature limitations contained in FSAR Section 4.2 are unaffected.

Changes are also proposed to Section 3.6.A and its associated Bases to reflect the new lpressure-temperature curves. These changes are editorial in nature and , as such, can notinvolve a significant increase in the probability or consequences of an accident previouslyevaluated.

2. create the possibility of a new or different kind of accident from those previously evaluated.The proposed changes revise existing limitations and are administrative in nature. They_ donot involve any physical modification to the plant, nor do they introduce any new failure-modes.

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The changes to Section 3.6.A and its Bases Section are editorial in nature; thus, they can >

not create the possibility of a new or different kind of accident from those previously 'evaluated.

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, involve a significant reduction in the margin of safety. The safety margins are increased3.

because the new Pressure Temperature limitations are more conservative (restrictive) and amore' accurate method is used to predict radiation embrittlement.-

The changes to Section 3.6.A and its Bases Section are editorial in nature; thus, they cannot involve a significant reduction in the margin of safety.

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: SAFETY EVALUATION-Page 5 o'6

in the April 6,1983 Federal Register (48FR14870), NRC published examples of licenseamendments that are not likely to involve a significant hazards consideration.' Examples (1) and -

' (vil) are applicable to these changes,

(i) ' A purely administrative change to technical specifications: for example, .a change to achieve consistency throughout the technicalspecifications, correction of an error, or a change in nomenclature. g

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(vil) A change to make a license conform to the change in the regulations, '

where the license change results in very minor changes to facilityoperations clearly in keeping with the regulations.

V. - IMPLEMENTATION OF THE PROPOSED CHANGESf

implementation of the proposed changes do not impact the Fire Protection Program at the |FitzPatrick plant, nor.will the change impact the environment. :

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VI. CONCLUSION!

The changes, as proposed, do not constitute an unreviewed safety question as defined in 10CFR 50.59. That is, they:

a. will not increase the probability of occurrence or the consequences of an accident ormalfunction of equipment important to safety previously evaluated in the safety analysisreport;

b. will not increase the possibility for an accident or malfunction of a different type than anyevaluated previously in the safety analysis report;

c. will not reduce the margin of safety as defined in the basis for any technical specification; -and '

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d. Involves no significant hazards consideration, as defined in 10 CFR 50.92.

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-lVll. REFERENCES j

. .11. USNRC Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel- i

Materials and its impact on Plant Operations," dated July 12,1988.

2. USNRC Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel -Materials," dated May 1988.

3. NYPA letter, J. C. Brons to NRC, dated June 30,1989, JPN-89-044, " Response to GenericLetter 88-11. Radiation Embrittlement of Reactor Vessel Materials." t

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Attachment 11.

'- - - SAFETY EVALUATION

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- 4. . Amendrnent 113 to the James A. Fitzpatrick Operating Ucense, October 22,1987.

5. USAEC " Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated. November 20,1972. - j

6. USAEC " Supplement 1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power

__

Plant" (SER), dated February 1,1973.

7. USAEC " Supplement 2 to the Safety Evaluation of the James A. FitzPatrick Nuclear PowerPlant" (SER), dated October 4,1974.

8. - James A. Fitzpatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section- 4.2.

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