ML20196F593
| ML20196F593 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/22/1999 |
| From: | James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20196F584 | List: |
| References | |
| NUDOCS 9906290253 | |
| Download: ML20196F593 (23) | |
Text
1 BEFORE THE UNITED STATES ,
NUCLEAR REGULATORY COMMISSION In the Matter of NEW YORK POWER AUTHORITY
) )
i ) Docket No. 50-333 '
I James A. FitzPatrick Nuclear Power Plant )
APPLICATION FOR AMENDMENT TO OPERATING LICENSE The New York Power Authority requests an amendment to the Technical Specifications (TS) contained in Appendix A to Facility Operating License DPR-59 for the James A.
FitzPatrick Nuclear Power Plant. This application is filed in accordance with Section 10 CFR 50.90 of the Nuclear Regulatory Commission's regulations.
This application for an amendment to the James A. FitzPatrick TS proposes changes to the Pressure-Temperature (P-T) limits, in compliance with the requirements of 10 CFR 50 Appendix H, a surveillance capsule was removed from the FitzPatrick reactor in November 1996 after 13.4 EFPY, and subsequently )
tested. The Authority forwarded the reactor vessel material surveillance program summary {
report and implementation schedule to the NRC (Reference 1). The Reference 1 letter included a report from General Electric (Reference 2) documenting the surveillance capsule test results. Based on the test results, P-T curves were prepared based on the new projected fluence levels for both 24 EFPY and 32 EFPY. The P-T curves and associated Limiting Condition for Operation and Bases changes are included in this proposed amendment.
As part of this submittal, the Authority is proposing to add separate bottom head curves Aos and Bas for in-service hydrostatic and leak tests and non-nuclear heatup and cooldown, respectively. The GE Report (Reference 2) provides the basis fcr the new bottom head curves. In addition, a non-beltline curve (i.e., Ana) for in-service hydrostatic and leak tests l is being added. GE has provided Errata and Addenda dated June 1999 which supplements the Reference 2 report. This Errata and Addenda provides the basis for the non-beltline ;
curve and is included in this submittal as Attachment IV. l The signed original of the Application for Amendment to the Operating License is enclosed for filing. Attachment I contains the proposed new TS pages and Attachment 11 is the Safety Evaluation for the proposed change. A markup of the affected TS pages is included as Attachment 111.
New York Power Authority STATE OF NEW YORK COUNTY OF WESTCHESTER Subscribed and sworn to before me thisA?* day of L4 1999.
1 J. Knubel C-Senior Vice President and Notary Public /
Chief Nuclear Officer OERALDINE STRAND 9906290253 990622 PDR hasary Pubhc. Stete of New tuli P ADOCK 05000333 No 4991272 PDR Ousedm westcheswr ComnussenEmpiresJan 37, f)O
r I
(- t g Attachment I to JPN-99-021 REVISED TECHNICAL SPECIFICATION PAGES PRESSURE-TEMPERATURE LIMITS (JPTS-99-003) l 1
l l
l l
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l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l DPR-59 l
l
I . ,
. . 1 JAFNPP LIST OF FIGURES Fiaures Title Paae 4.1-1 (Deleted) 4.2-1 (Deleted) 3.4-1 Sodium Pentaborate Solution (Minimum 34.7 B-10 Atom % Enriched) 110 Volume-Concentration Requirements i i
3.4-2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111 3.5-1 (Deleted) 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 24 EFPY 163 Part 1 3.6-1 Reactor Vessel Pressu're - Temperature Limits Through 32 EFPY 163a Part 2 3.6-1 (Deleted) 163b Part 3 4.6-1 (Deleted) 6.1 1 (Deleted) !
6.2-1 (Deleted)
Amendment No. 44r2-2, 4 3, S4, 72, 71, SS, DS,109,113,11 S,117,134,137,158,162, 227, 236, 247,-
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0I ---+- - - - - - - - -
0 50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature ( F)
Figure 3.6-1 Part 1 Reactor Vessel Pressure-Temperature Limits Through 24 EFPY Amendment No. 444468, 163 l l
1
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l i
0-- -- .-- . -- - . -
4 O 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 32 EFPY Amendment No. 468, 163a 1
F .. .
- JAFNPP 1
I THIS PAGE INTENTIONALLY LEFT BLANK 1
I 163b Amendment No. MB,
r? , .
, ' Attachment 11 to JPN-99-021 SAFETY EVALUATION PRESSURE-TEMPERATURE LIMITS (JPTS-99-003) l l
i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
p , ..
. Attachment 11 to JPN-99-021 l SAFETY EVALUATION Page 1 of 11
- 1. DESCRIPTION This application for an amendment to the James A. FitzPatrick Technical Specifications (TS) revises Specification 3.6.A, " Pressurization and Thermal Limits,"
and the associated Bases. Specifically, the Pressure-Temperature (P-T) curves in Figure 3.6-1 are replaced with new curves for operation up to 24 and up to 32 Effective Full Power Years (EFPY). The associated Table of Contents, Limiting Condition for Operation (LCO), and the Bases Section are revised to reflect the new P-T curves.
The specific changes to the TS are:
Paae vii Replace:
"3.6-1 Reactor Vessel Pressure - Temperature Limits Through 12 EFPY
? art 1 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 14 EFPY Part 2 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 16 EFPY Part 3" With:
"3.6-1 Reactor Vessel Pressure - Temperature Limits Through 24 EFPY Part 1 3.6-1 Reactor Vessel Pressure - Temperature Limits Through 32 EFPY Part 2 3.6-1 (DELETED)
Part 3" Paae 136. Specification 3.6.A.2 (In-Service Hydrostatic and Leak Testsl Replace:
...the Reactor Coolant System pressure and temperature shall be on or to the right of curve A shown in Figure 3.6-1 Part 1, 2, or 3 and the maximum temperature change during any one hour period shall be:..."
with- I i
1
"...the Reactor Coolant System pressure and temperature shall be on or to the right l of curve A shown in Figure 3.61 Part 1 or 2 for the flange and the beltline region, !
and on or to the right of curve Aus for the non-beltline regions, and on or to the right !
of curve Ass for the bottom head region. The maximum temperature change during l any one hour period shall be:..."
l L --
j
p , , )
, , Attachment Il to JPN-99-021 SAFETY EVALUATION Page 2 of 11 Paae 137, Specification 3.6.A.3 (Non-Nuclear Heatuo and Cooldown)
Replace:
l l "...the Reactor Coolant System pressure and temperature shall be on or to the right I of the curve B shown in Figure 3.6-1 Part 1,2, or 3 and the maximum temperature change during any one hour shall be s100 F."
with:
"...the Reactor Coolant System pressure and temperature shall be on or to the right of the curve B shown in Figure 3.6-1 Part 1 or 2 for the flange, upper vessel and beltline regions, and on or to the right of curve Beg for the bottom head region. The maximum temperature change during any one hour shall be s100 F."
Paae 137 Specification 3.6.A.4 (Core Critical Ooeration) l Replace- -
l i
"...the Reactor Coolant System pressure and temperature shall be at or to the right i l of the curve C shown in Figure 3.6-1 Part 1,2, or 3 and the maximum temperature l change during any one hour shall be s100 F."
With:
"...the Reactor Coolant System pressure and temperature shall be at or to the right of the curve C shown in Figure 3.6-1 Part 1 or 2. The maximum temperature change during any one hour shall be s100 F."
l 1 i Paae 146, Bases 3.6.A/ 4.6. A. Second Column, Second Paraaraoh, Last Sentence l Replace:
1.7 X 10'8 n/cm 2 j I With:
1.8 X 10'8 n/cm 2 Paae 147, Bases 3.6. A/ 4.6. A i
- 1. First Column, Second Paragraph, First Sentence l Revise to state that the date of Regulatory Guide 1.99, Revision 2 is May i 1988.
I l
)
[
l
. , Attachment 11 to JPN-99-021 SAFETY EVALUATION Page 3 of 11 l 2. First Column, Second Paragraph, Second Sentence Revise to state that an evaluation of two sets of irradiated surveillance j specimens, which were withdrawn from the reactor in April,1985 (6 EFPY) l and November 1996(13.4 EFPY), respectively, shows a shift in RTuor less than that predicted by Regulatory Guide 1.99, Revision 2.
- 3. First Column, Third Paragraph, First Sentence Replace:
l
"...were established using 10 CFR 50 Appendix G, May,1983 and Appendix G of the Summer 1984 Addenda to Section Ill of the ASME Boiler and Pressure Vessel Code."
With:
l "...were established using 10 CFR 50 Appendix G, December 1995, and l Appendix G of Section XI of the ASME Boiler and Pressure Vessel Code (1989 Edition)."
- 4. First Column, Third Paragraph, Third Sentence Replace:
"...the reference temperature, RTwor, of the vessel material was estimated from impact test data taken in accordance with the requirements of the Code to which the vessel was designed ;ad manufactured (1965 Edition including Winter 1966 addenda)
With:
"...the reference temperature, RTnor, of the vessel material was estimated from impact test data taken in accordance with the requirements of Section lll of the ASME Boiler and Pressure Vessel Code to which the vessel was designed and manufactured (1965 Edition including Winter 1966 addenda)."
- 5. First Column, Third Paragraph, Fourth and Fifth Sentences Replace:
"The RTwor values for the reactor vessel flange region and for the reactor vessel shell beltline region are 30 F, based on fabrication test reports. The RTuor for the remainder of the vessel is 40 F."
p -.- .
1
, ' Attachment 11 to JPN-99-021 ,
SAFETY EVALUATION !
Page 4 of 11 l With:
i "The RTuor values of the reactor. vessel materials are listed on Table 3-2 of l General Electric Report GE-NE-B1100732-01, " Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Capsule at 13.4 EFPY," ;
Revision 1 (February 1998), including Errata and Addenda dated June l 1999." I i 6. Second Column, First Paragraph After the fourth sentence, add the following two sentences:
l "The second surveillance capsule containing test specimens was withdrawn i in November,1996 after 13.4 EFPY. The test specimens removed were !
l tested according to ASTM E 185-82 and the results are in General Electric Report GE-NE-B1100732-01, Revision 1 (February 1998), including Errata and Addenda dated June 1999." i
- 7. Second Column, Second Paragraph Replace the second paragraph in its entirety with the following paragraph:
" Figure 3.6-1 is comprised of two parts: Part 1 and Part 2. Part 1
- establishes the pressure-temperature limits for the bottom head, flange, l upper vessel and beltline regions for plant operations through 24 Effective 1
Full Power Years (EFPY). Part 2 establishes the pressure-temperature limits for plant operations through 32 EFPY. The curves contained in Figure 3.6-1 are developed from the General Electric Report GE-NE-B1100732-01, l Revision 1 (February 1998), including Errata and Addenda dated June l
1999."
- 8. Second Column, Third Paragraph, First Sentence
! Replace:
l l " Figure 3.6-1 curve A establishes the minimum...."
1 With:
" Figure 3.6-1 curves A, Aos, and Aus establish the minimum..."
l L
f ..
Attachmsnt 11 to JPN-99-021 SAFETY EVALUATION Page 5 of 11 Paae 148, Bases 3.6. A/ 4.6. A
- 1. First Column, First Paragraph, First Sentence Replace:
" Fig. 3.6-1, curve B, provides limitations..."
With:
" Fig. 3.6-1, curves B and B,g, provide limitations..."
- 2. First Column, Third Paragraph, l
Replace: j 1
"The requirements for cold boltup of the reactor vessel closure are based on I NDT temperature plus a 60 F factor of safety. This factor is based on the ,
requirements of the ASME Code to which the vessel was built. For Fig.
3.6-1, curves A, B and C, margins are only added to the low temperature portion of the curve where non ductile failure is a concern. The closure flanges have an NDT temperature not greater than 30 F and are not subject ,
to any appreciable neutron radiation exposure. Therefore, the minimum '
temperature of the flanges when the studs are in tension is 30 F plus 60*F, or 90 F."
With:
"The requirements for cold boltup of the reactor vessel closure region are established based on RTuor plus 60 F. This factor is based on the original requirements of the ASME Code to which the vessel was built, as well as additional, conservative requirements developed by General Electric that are typically applied to most BWRs. For Fig. 3.6-1, curves A, B, and C, this factor leads to the 90 F lower temperature limit. This limit is based on the closure flange materials maximum RTwor of 30 F, and the fact that the closure flange materials are not subjected to any appreciable neutron radiation exposure. Therefore, the minimum temperature for cold boltup is 30 F plus 60 F, or 90 F."
- 3. Second Column, Second Paragraph, Second Sentence Replace:
"This temperature limit is based on providing a 50 F band for operating flexibility between the 300 F limit and the highest estimated minimum testing temperature at 32 EFPY (approximately 250 F).*
4~ , .
s
, Attachment 11 to JPN-99-021 SAFETY EVALUATION Page 6 of 11 With:
"This temperature limit is based on providing at least a 50 F band for operating flexibility between the 300 F limit and the highest estimated minimum testing temperature at 32 EFPY (originally approximated as 250 F from testing of the first surveillance capsule)."
- 4. Second Column, Second Paragraph ;
l Add the following two sentences at the end of the paragraph:
" Based on the latest surveillance capsule test results, the minimum temperature required to stay on or to the right of curve A at the maximum test pressure is 212 F for 32 EFPY. The previously established hydrostatic test limitation of 300 F continues to provide adequate operating flexibility above this minimum temperature."
Pace 163. Fiaure 3.6-1. Part 1 Replace :
Figure containing reactor vessel Pressure -Temperature limits through 12 EFPY, With:
l Figure containing reactor vessel Pressure -Temperature limits through 24 EFPY.
Paae 163a. Fiaure 3.6-1. Part 2 Replace :
Figure containing reactor vessel Pressure -Temperature limits through 14 EFPY.
With:
Figure containing reactor vessel Pressure -Temperature limits through 32 EFPY.
Paae 163b Fiaure 3.61, Part 3
, Replace :
l l Figure containing reactor vessel Pressure -Temperature limits through 16 EFPY.
l With the following words:
"This Page Intentionally Left Blank"
[ .- .
l .
, Attachment ll to JPN-99-021 i SAFETY EVALUATION Page 7 of 11
- 11. PURPOSE OF THE PROPOSED CHANGE in compliance with the requirements of 10 CFR 50 Appendix H, a surveillance capsule was removed from the FitzPatrick reactor in November 1996 after 13.4 EFPY, and subsequently tested. The Authority forwarded the reactor vessel material surveillance program summary report and implementation schedule to the NRC (Reference 1). The Reference 1 letter included a report from General Electric (Reference 2) evaluating the surveillance capsule test results. Based on this evaluation, new operating limit curves were prepared for both 24 EFPY and 32 ,
EFPY. The proposed changes reflect the new limits. '
As part of this submittal, the Authority is proposing to add separate bottom head curves Aes and B,s for in-service hydrostatic and leak tests and non-nuclear heatup and cooldown, respectively. The GE Report (Reference 2) provides the basis for the new bottom head curves. In addition, a non-beltline curve (i.e., A ns) for in-service :
hydrostatic and leak tests is being added. GE has provided Errata and Addenda l dated June 1999 which supplements the Reference 2 report. This Errata and '
Addenda provides the basis for the non-beltline curve and is included in this submittal as Attachment IV.
The purpose of the proposed changes are to establish operating limits that provide a wide margin to conditions which could lead to brittle fracture of the Reactor Pressure Vessel (RPV).
111. SAFETY IMPLICATIONS OF THE PROPOSED CHANGE As stated above, the proposed changes include replacement of the existing P-T curves, revisions to TS 3.6.A, as well as changes to the associated Bases section.
The changes to the P-T curves are being proposed to preclude brittle fracture of RPV materials for up to 32 EFPY. In addition to the P-T curve developed for up to 32 i EFPY, a P-T curve has been prepared for exposures up to 24 EFPY to shorten outage I time for startups conducted prior to reaching this exposure.
I The Authority forwarded the reactor vessel material surveillance program summary report and implementation schedule to the NRC (Reference 1). The Reference 1 i letter included a report from General Electric (Reference 2) documenting the j surveillance capsule test results. The Authority has performed an independent i review (Reference 3) of the Reference 2 General Electric Report and the Errata and Addenda to the Reference 2 report included in Attachment IV of this submittal, j Based on these reviews, the Authority has concluded that the P-T curves are 4 acceptable, and the methods employed are appropriate. The NRC has approved the l Authority's surveillance capsule withdrawal schedule (Reference 4). In the Reference 4 letter the NRC states the following:
I "The NRC staff has reviewed the information provided, and has determined that the ;
proposed JAF surveillance program is capable of monitoring the effects of neutron l irradiation and the thermal environment on the fracture toughness of ferritic reactor l Vessel beltline materials. The staff verified that your proposed surveillance capsule withdrawal schedule satisfies the American Society for Testing and Materials (ASTM) E 185-70 standard, and that the withdrawal of the third surveillance capsule will also be in accordance with ASTM E 185-82. Therefore, as discussed in
. , Attachment ll to JPN-99-021 SAFETY EVALUATION Page 8 of 11 the enclosed safety evaluation, the staff found the proposed JAF surveillance
! capsule withdrawal schedule to be acceptable. "
The proposed changes to TS Section 3.6.A and the P-T curves describe requirements concerning the control of reactor vessel temperature and pressure when performing in-service hydrostatic and leak tests, non-nuclear heatup and cooldown, and core critical operation. These requirements are consistent with the General Electric (GE) Report (Reference 2), including the Errata and Addenda dated June 1999, and with the current methods of performing in-service hydrostatic and leak tests, non-nuclear heatup and cooldown, and core critical operation. The Bases changes: (1) reflect that the new P-T curves are valid up to 32 EFPY, (2) provide evaluation results of the second surveillance capsule, (3) clarify requirements regarding cold boltup of the reactor pressure vessel closure region, and (4) delete information applicable to the existing curves.
As part of this submittal, the Authority is proposing to add separate bottom head curves Aanand B sa for in-service hydrostatic and leak tests and non-nuclear heatup and cooldown, respectively. The GE Report (Reference 2) provides the basis for the new bottom head curves. In addition, a non-beltline curve (i.e., Aus) for in-service hydrostatic and leak tests is being added. GE has provided Errata and Addenda dated June 1999 which supplements the Reference 2 report. This Errata and Addenda provides the basis for the non-beltline curve and is included in this submittal as Attachment IV.
Under certain conditions, portions of the reactor vessel in the non-beltline region can be significantly cooler relative to other portions of the reactor vessel. These conditions can result from recirculation pump operation at low speed or off, during water injection through the control rod drives, or when Reactor Water Cleanup is used to make up to the vessel through feedwater nozzle. To account for these circumstances, individual limits for the bottom head and non-beltline regions, as described above, are being proposed.
The non-beltline regions, and the bottom head are not subject to a fluence high enough to cause a shift in the nil-ductility transition temperature (RTuoy). Therefore, it is not necessary to subject these regions to the higher temperature limits imposed on beltline materials due to irradiation embrittlement. Based on this, the proposed change includes separate curves for the bottom head and non-beltline region with less restrictive minimum temperatures and in compliance with 10 CFR 50 Appendix G. This proposed change provides additional operational flexibility, and adequate margin against brittle fracture.
The methods and results of testing the second surveillance capsule are presented in the Sections of the Reference 2 report as follows:
- 1. Section 3: Surveillance Program Background
- RPV Materials and Fabrication
- Materials Properties
- Surveillance Specimen Chemical Composition
- Specimen Description
. l l
. Attachment 11 to JPN-99-021 SAFETY EVALUATION Page 9 of 11
- 2. - Section 4: Peak RPV Fluence Evaluation
- 3. Section 5: Charpy V-Notch Impact Testing
- 4. Section 6: Tensile Testing
)
- 5. Section 7: Adjusted Reference Temperature and Upper Shelf Energy
- 6. Section 8: Pressure-Temperature curves in addition, Attachment IV provides the basis for the non-beltline curve (Aus) to be utilized during in-service hydrostatic and leak tests.
This GE report and the information contained in Attachment IV provides the bases for the Authority's responses to the No Significant Hazards Consideration Analysis provided in Section IV of this submittal. Safety margins specified in 10 CFR 50, Appendix G and Appendix G to Section XI of the ASME Code will continue to be met. The Authority has performed an independent review (Reference 3) of the Reference 2 General Electric Report and the Errata and Addenda to the Reference 2 I report included in Attachment IV of this submittal and agrees with the documented conclusions. Therefore, operation of FitzPatrick in accordance with the proposed changes will not endanger the health and safety of the public.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION ;
Operation of the FitzPatrick plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated.
The changes to the P-T curves are being proposed to preclude brittle fracture of RPV materials for up to 32 EFPY. In addition to the P-T curve for up to 32 EFPY, a P-T curve has been prepared for exposures up to 24 EFPY to shorten outage time for startups conducted prior to reaching this exposure.
Safety margins specified in 10 CFR 50, Appendix G and Appendix G to Section XI of the ASME Code will continue to be met for each of these curves. Therefore, there is not a significant increase in the probability of an accident previously evaluated.
The RPV, as part of the reactor coolant system, provides a barrier to the release of reactor coolant. Operation in accordance with the proposed amendment will preclude brittle fracture of the RPV consistent with current requirements, and consequently, does not significantly increase the consequences of an accident previously evaluated.
l Based on the above, operation of the FitzPatrick plant in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
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l
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e
, Attachment ll to JPN-99-021 SAFETY EVALUATION Page 10 of 11
)
- 2. create the possibility of a new or different kind of accident from any l<
accident previously evaluated.
The propc' sed change does not involve any physical alterations to plant configurations or introduce any new accident precursors which could initiate a new or different kind of accident. The proposed change does not affect the inter >ded function of the RPV nor does it affect the operation of the RPV in a way which would create a new or different kind of accident. The changes to the P-T curves are being proposed to preclude brittle fracture of RPV materials for up to 32 EFPY. Safety margins specified in 10 CFR 50, Appendix G and Appendix G to Section XI of the ASME Code will continue to be met. Therefore, operation of the FitzPatrick plant in accordance with
{
the proposed amendment will not create the possibility of a new or different i kind of accident from any accident previously evaluated.
- 3. involve a significant reduction in a margin of safety.
The existing FitzPatrick P-T curves were developed using safety margins for brittle fracture found in 10 CFR 50 Appendix G. The proposed FitzPatrick P-T curves, which are valid for up to 32 EFPY of operation, were also .
developed using safety margins for brittle fracture found in 10 CFR 50 l Appendix G. Based on this, operation of the FitzPatrick plant in accordance with the proposed amendment will continue to preclude brittle fracture of the RPV materials during in-service hydrostatic and leak tests, non-nuclear heatup and cooldown, and core critical operation without a significant reduction in a margin of safety. Therefore, operation of the FitzPatrick plant in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.
V. IMPLEMENTATION OF THE PROPOSED CHANGE This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:
(i) the amendment involves no significant hazards consideration.
As described in Section IV of this evaluation, the proposed change involves no significant hazards consideration.
(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
The proposed change does not involve any physical alterations to plant configurations. The proposed change does not affect the intended function of the RPV nor does it affect the operation of the RPV in a way which could significantly change the types or l significantly increase the amounts of any effluents that may be released offsite.
, Attschment 11 to JPN-99-021 l SAFETY EVALUATION l Page 11 of 11 l
(iii) there is no significant increase in individual or cumulative occupational i radiation exposure.
l l The RPV, as part of the reactor coolant system, provides a barrier to the release of reactor coolant. Operation in accordance with the proposed amendment will preclude brittle fracture of the RPV consistent with current requirements, and consequently will not significantly increase individual or cumulative occupational radiation exposure.
Based on the above, the Authority concludes that the proposed changes meet the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements
! of 10 CFR 51.21 relative to requiring a specific environmental assessment by the l Commission.
VI. CONCLUSION The changes to the P-T curves are being proposed to preclude brittle fracture of RPV materials for up to 32 EFPY. In addition to the P-T curve for up to 32 EFPY, a P-T curve has been prepared for exposures up to 24 EFPY to shorten outage time for j startups conducted prior to reaching this exposure. The proposed changes provide l adequate protection against brittle fracture during in-service hydrostatic and leak tests, non-nuclear heatup and cooldown, and core critical operation. As detailed in 4 the GE Report (Reference 2), including Errata and Addenda dated June 1999, safety margins specified in 10 CFR 50, Appendix G and Appendix G to Section XI of the l ASME Code will continue to be met. Therefore, operation of the FitzPatrick plant in I
accordance with the proposed change will not endanger the health or safety of the l public.
The Plant Operating Review Committee (PORC) and Safety Review Committee (SRC) have reviewed this proposed change to the TS and agree with this conclusion.
l Vll. REFERENCES
- 1. NYPA Letter, J. Knubel to the NRC, " Revised Reactor Pressure Vessel Material Surveillance Program Summary report and Implementation Schedule," (JPN-98-00m, dated March 9,1998 l
- 2. General Electric Report, " Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120* Capsule at 13.4 EFPY," Revision 1, (GE-NE-B1100732-01), dated February 1998 l 3. Structural Integrity Report, "P-T Curve and Materials Review for the James A. FitzPatrick Reactor Pressure Vessel," (Report No. SIR-99-031, Revision l' A), dated March 1999
- 4. NRC Letter to J. Knubel, " James A. FitzPatrick Nuclear Power Plant - Reactor Vessel Surveillance Capsule Withdrawal Schedule (TAC NO. MA1233),"
dated July 10,1998