ML19354E059

Proposed Tech Spec 3.6.A, Pressurization & Thermal Limits.

ML19354E059
Person / Time
Site:	FitzPatrick
Issue date:	01/12/1990
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML19354E054	List: ML19354E053 ML19354E058 ML19354E059
References
NUDOCS 9001250305
Download: ML19354E059 (15)

Text

._-_ -

s .. ,

.v_

ATTACHMENT 1:

PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING PRESSURE-TEMPERATURE UMITS (JPTS-89 024) .

m

~l l

i i

j

.I New York Power Author.ity l q

~

JAMES A. FITZPATRICK NUCLEAR POWER PLANT  !

Docket No. 50-333 ' ...

]

DPR-59 '

1 e '

fjgg12gggg{j@M 3 ]'

p .

1

h JAFNPP' e UST OF FIGURES l 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 Test Interval vs. Probability of System Unavailability- 87 .;

3.4 14 Sodium Pentaborate Solution 34.7 B 10 Atom % Enriched Volume- 110 Concentration Requirements 3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution '111-  !

3.5 Thermal Power and Core Flow Umits of Specifications 3.5.J.1,3.5.J.2 and .134 3.5J.3 ' y 3.5-6 (Deleted) 135d:

3.5-7 (Deleted) 135e 3.5-8 (Deleted) - 135f =

3.5-9 (Deleted) 135g 3.5-10 (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 l- .. Part 1 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 163b Part 3 4.6-1 Chloride Stress Corrosion Test Results at 500 F 164 q 6.1 1 (Deleted) 259 6.2 1 (Deleted) 260  ;

i 1

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

JAFNPP .,

^

4.6 SURVEILLANCE REQUIREMENTS ^.

- 3.6 UMITING CONDITIONS FOR OPERATION 4 6 REACTOR COOLANTSYSTEM

-3.6 REACTOR COOLANTSYSTEM Applicability:

W D ,.

Applies to the operating status of the Reactor Coolant System.

Wes to h p exWh W testing Wrm fm W

- 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 Tensoning

1. Reactor Vessel Head Stud Tensioning Vfwi in the cold conditon, the reactor vessel head Range The reactor vessel head botting studs shall not be under ~

and the . reactor vessel flange temperatures shall be tension unless the temperatures of the reactor vessel mcw M

- flange and the reactor head flange are at least 90 F.

a. Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor vessel head Range

' is <1207 and the studs are tensioned.

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

' During in-service hydrostatic or. leak' testing the Reactor t% dew W W W

- 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 every 30 minutes until two consecutwe. temperature readings are within 5*F of each other.

- Amendment No. M, tFJ ~

136

,_a.- n _. ,__ ,, w_,-. , _,__ _ ____ ,___ ,, _ __

%= n- 1o.- - n- . ,

1 -_:_--___2____1__

JAFNPP' -

~

~ : 3.6 (cont'd) 4.6 (cont'd)

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

3. ~ Non-Nuclear Heatup and Cooldown During - heatup - by : non-nuclear rneans (mechanical),

by non- means, h W cooldown following nuclear shutdown and low power Mear Wh W W poww % tests,the rMw physics tests the Reactor Coolant System pressure and-ant @mn pmsswe W tanpwatwei M; k temperature shall be on or to the right of the curve B re m may 30- W es. W h6 tanpwatwe rMngs are m WW each dhw.

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

4. Core Critical Operation

4. Core Critical. Operation During all modes of operation with a critical core (except ng an e d wh 2 a crh cwe (ex$.

for low power physics tests) the reactor Coolant System fa poww @ tests) h e W W 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 maximum temperature change during any one hour shall.

mactw W W may 30 Mes dunng W W -

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

~

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

. condition on the structural integnty of the_ reactor -

l coolant system, and : determine that the; reactor l- coolant system remains acceptable for continued i

operations; or .

.~

i

(. Amendment No. jHf,126 o 137~

j. a.

L -.

v

=

JAFNPP -

3.6 and 4.6 BASES (cont'd) ~~

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

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 - ,

capsule iocauen by expeiiniental methods. removed were tested according to ASTM E 18542 and the The flux '

results are in GE report MDE-49-0386.. The next surveillance distribution'at the vessel wall and 1/4 thickness (1/4T) depth 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 Rgure lead 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 x i the reactor vessel belthneI Figure 3.6-1'is congnsed of three parts: Part 1, Part 2, and -

RTNTO as a function of fluence ,n Part 3. Parts 1,2, and 3 establish the pressure-temperature region. An evaluation of the irradiated suryeillance '

specimens, which were w,thdrawn i

from the reactor in April, limits for plant operations through 12,14, and 16 Effective Fulli 1985 (6 EFPY), shows a shift in RTNm less than that predicted Yews (EFW r%tW. The weide @ and

j. by Regulatory Guide 1.99, Revision 2. esseteme m we W m h W of 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 for

in-service hydrostatic and leak testing were established using ~ operation at greater than 14 EFPY through'16 EFPY. The l 10 CFR 50 Appendix-G, May,' 1983 and Appendix G of the . curves ' contained in Rgure 3.6-1 are developed from the t '

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 atrick -

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, and j 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 and these operating . limits, the reference temperature, RTNOT, Of Preswe Vessel % M'CJ@ msW' n m the vessel material was estimated from impact test data taken -Wostdic W M M'q 'we a Wion d the teW'm-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 i including maximum hydrostatic test pressure will be 1.1 times the '

ope ng pressme d M N 4 Winter 1966 add nida). The' RTNOT values for the reactor Amendment No.1 147 u - - = .. , _ _ ,. - -  :. ~ , , . - . .. . ~ . .- .

- a ,

JAFNPP

l. VALID TO 12 EFPY i A SYSTEM HYOROTEST LIMIT , , , , , , , , , , , , , ,

WITH FUEL IN VESSEL A * *s ** B esa *r C su *r I

1400 B - NON N'0 CLEAR HEATING LIMIT f C NUCLEAR (CORE CRIT 4C AL) LIMIT / f

/ / /

{ 1200 ) / /

& l l l 0 '

W ~_

a.

$ 1000

/

[ [ D BELTLINE ART 106 *F d /

L i / / f j

l *** / / /

1 < / ) /

) E  ;;;W / /

1 F-l l h 600 s

-w 400 N A. NON BELTLINE

[

na esa> n .

/

f. /

j

.RTdia 3 0 *F -

10 *F 120 'F r i

/

200 BO'. TUP 90 7 ) [/

/

SATURATICN r i 0

'0- 60 100 150' 200 250 .300 350 MINIMUM REACTOR YESSEL METAL TEMP' E RATURE (*F)_

Figure 3.6-1 Part1 Reactor Vessel Pressure-Temperature Limits Through 12 EFPY  !

l 163 Am. nam.niNo.1/3 j

4 4

JAFNPP-VALID 12 TO 14 EFPY 5: A SYSTEM HYDROTEST LIMIT +

A $[I'[" B SO'[" C IO'["

~

4 WITH FUEL IN VESSEL 1400 B NON NUCLEAR HEATINO LIMIT f f ,

C NUCLEAR (CORE CRITICAL) LIMIT f f ,

~

{ 1200 l /

l s , l I

4x AF=

R

j 1000

/

/

/

/ BELTLINE .

ART s 111 *F 1

800

/

/

/ /

l

/  ;

E  :::;[

/. /

j eoo f f i

g 400

/

Abw /

NM NON BILTLINE '

m emo m es.4 / - m em RT : 3 0 *F '

,,. 37 j .., ,

t / /

200 BOLTUP 90*F / [/ /

SATURATidN

/ ,

0 0 50 100 150 200 250' 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 3.6-1 Part 2 ' Reactor. Vessel Pressure-Temperature Limits Through 14 EFPY 163a Amendment No. ,

g S e JAFNPP k:

b VALID 14.TO 16 EFPY A

• SYSTEM HYDROTEST Ll!.;lT WITH FUEL IN VESSEL A m 'r. - B a.a 'r C saa ',

1400 B NON NUCLEAR HEATINO LIMIT -I f

C NUCLEAR (CORE CRITICAL) LIMIT f )

3 .1200

// I I

l

/~'

$ f, l O

IN

n. N' d

1000

/

/

/

/ BELTLINE ART s 116 *F

$ 600

/ / /

N J /

f "*4[ / [

!" / /

i

[ 400

[

PA N k NON BELTLINE sia nio na n.a / - na n. RTuois 30 *F

., , , . , m, l

200 BOLTUP /-  !

90*F /

/

/ SATURATICN r

~

0

/

0 50 100 150' 200 260 300 350 .

1 '

MINIMUM REACTOR VESSEL METAL TEMPERATURE OF)

Figure 3.6-1 Part 3 Reactor Vessel Pressure-Temperature Limits Through 16 EFPY 163b Amendment No. i 1

ATTACHMENT 11 -

SAFETY EVALUATION FOR PROPOSED RE EIIIWilTS 2 1 (JPTS-89 024)

, l .

.i 1

i l

New York Power Authority-  ;

i JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 '

DPR-59 -j

'-_3 y I.' i '* .

Attachment ll -

. SAFETY EVALUATION

. Page 1 of 6 '

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 with Generic 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 for operation 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

,i

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

B. Associated Wording Changes 1

-t

1. 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:

e '

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

with:

Regulatory Guide 1.99, Revisl.on 2 is used to predict the shift in RT NTD as a function of fluence in the reactor vessel beltline region. An evaluation of the irrac'iated surveillance specimens, which were withdrawn from the reactor in April,1985 (6 EFPY), shows a shift in RTNTD l ess than that predicted by Regulatory Guide 1.99, Revision 2.

I

. Attachment II.

SAFETY EVALUATION -

Page 2 of 6 e Delete the third sentence in the fourth paragraph on page 147. The sentence to be

' deleted reads as follows:

The curves of Figure 3.6-1, A through C, reflect findings in the report j related to copper phosphorus content of the reactor vessel shell beltline, 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. -

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 plant operations through 12,14, and 16 Effective Full Power Years (EFPY) respectively. The appropriate figure and the pressure-temperature curves 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. The curves contained in Figure 3.61 are developed from the General Electric Report DRF 137 0010, " implementation of Regulatory Guide 1.99, Revision 2 for the James A. Fitzpatrick Nuclear Power Plant," 1

- dated June,1989. l I$ki

• 3. Ust of Figures, page vil: replace " Figure 3.61, Reactor Vessel Pressure - Temperature

~

Umits," page 163 with 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 Umits Through 14 EFPY," page 163a, Figure 3.6-1, Part 3, " Reactor Vessel Pressure - Temperature Umits ,

Through 16 EFPY," page 163b :

II. PURPOSE OF THE PROPOSED CHANGES Regulatory Guide 1.99, Revision 2 (Referenco 2) revised the methodology used to evaluate -

i neutron 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 -

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

Fitzpatrick Technict.1 Specifications. This proposed change is consistent with the requirements of Section V of 10 CFR 50, Appendix G..

{

Attachment ll SAFETY EVALUATION Page 3 of 6

Background

Specification 3.6.A,

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

Amendment 113 (Reference 4) revised the pressure-temperature limits to be consistent with test results and analyses performed on the irradiated surveillance capsule removed from the Fitzpatrick reactor in April,1985. (Surveillance capsules are installed in the reactor vessel before startup and contain test specimens that are made from the plate, weld, and heat affected zone materials of the reactor beltline.) Radiation embrittlement was calculated using the surveillance data and adjusting the nil-ductility reference temperature (RTNDT ) in accordance with Regulatory Guide 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 resultant shift in RT NTO bounds the previously calculated results for the beltline region of the core.

New beltline Pressure-Temperature curves were developed for operation to 12,14, and 16 Effective Full Power Years (EFPY). The non-beltline region curves (recirculation inlet nozzles and head flanges) are not affected by the changes to RTNTD shift associated with Regulatory Guide 1.99, Revision 2.

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

\

ll1. IMPACT OF THE PROPOSED CHANGES The purpose of Specification 3.6.A, " Pressurization and Thermal Umits," is to establish operating limits that provide a wide margin to brittle failure of the reactor pressure vessel. The basis of the Pressure-Temperature (P-T) limits is found in Appendix G to 10 CFR 50 and in Section 4.2 of the updated FSAR. The limits are not derived from the design basis accident analyses, but are prescribed to avoid encountering pressures, temperatures, and temperature rate-of-changes which might cause undetected flaws to propagate.

The first technical specification change lowers the P-T curves (i.e., a higher temperature is l required for a given pressure) which in turn " narrows" the reactor coolant system operating window and lengthens the time required for hydrostatic testing. This change is consistent with Generic 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 5 16 EFPYs. The use of three curves instead of one lessens operational impacts by phasing the increases in minimum temperature over three distinct service periods. Each set of P T curves is I

.' l

~ Attachment il .

SAFETY EVALUATION Page 4 of 6 conservative, because the conditions at the end of each servise period (12,14,' or 16 EFPY)' yield the 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 also updates 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 alter the 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 proposed amendment 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 previously evaluated. The effect of neutron radiation on reactor vessel materials has been recalculated using 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 .  ;

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 l pressure-temperature curves. These changes are editorial in nature and , as such, can not involve a significant increase in the probability or consequences of an accident previously evaluated.

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_ do not involve any physical modification to the plant, nor do they introduce any new failure-modes.

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.

, 3. involve a significant reduction in the margin of safety. The safety margins are increased because the new Pressure Temperature limitations are more conservative (restrictive) and a more' 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 can not involve a significant reduction in the margin of safety.

-.  : AttachmentII:

SAFETY EVALUATION-Page 5 o'6 in the April 6,1983 Federal Register (48FR14870), NRC published examples of license amendments 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 technical specifications, correction of an error, or a change in nomenclature. g p

(vil) A change to make a license conform to the change in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

V. - IMPLEMENTATION OF THE PROPOSED CHANGES f

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

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

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

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

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

and '

4

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

i

-l Vll. REFERENCES j

.1

1. 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 Generic Letter 88-11. Radiation Embrittlement of Reactor Vessel Materials." t

Attachment 11

- SAFETY EVALUATION

- Page 6 of 6 -

- 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 Power Plant" (SER), dated October 4,1974.

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

.L -

4 I

.i l

i E'

. _ . . . . . . . . . .