HNP-96-206, Application for Amend to License NPF-63,requesting Rev to Chemistry Data (Nickel Content) in TS 3/4.4.9 Re Pressure/ Temp Limits

Application for Amend to License NPF-63,requesting Rev to Chemistry Data (Nickel Content) in TS 3/4.4.9 Re Pressure/ Temp Limits

ML18012A449
Person / Time
Site:	Harris
Issue date:	12/30/1996
From:	Robinson W CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18012A451	List: HNP-96-206, Application for Amend to License NPF-63,requesting Rev to Chemistry Data (Nickel Content) in TS 3/4.4.9 Re Pressure/ Temp Limits ML18012A453
References
HNP-96-206, NUDOCS 9701060058
Download: ML18012A449 (23)

Text

CATEGORY 1

.It'EGUIAT INFORMATION DISTRIBUTIONSTEM (RIDE)

ACCESSiON NBR:9701060058 DOC.DATE': 96/12/30 NOTARIZED: YES FACIL:50-400 Shearon Harris Nuclear Power.Plant, Unit 1, Carolina AUTH.NAME<

AUTHOR AFFILIATION-ROBINSON,W.R.

Carolina Power

!E Light Co.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Application for amend to license NPF-63,requesting rev to chemist y data (nickel content) in TS 3/4.4.9 re pressure/

temp limits.

DISTRIBUTION CODE:

A001D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution NOTES:Application for permit renewal filed.

DOCKET ¹ 05000400 T

05000400 E

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E NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!

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Carolina Power & Light Company PO Box 165 New Hill NC 27562 DEC 30 1996 William R. Robinson Vice President Harris Nuclear Plant SERIAL: HNP-96-206 10 CFR 50.90 10 CFR 50 App. G 10 CFR 50 App. H United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEARPOWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/I EMPERATURE LIMITS

Dear Sir or Madam:

In accordance with the Code ofFederal Regulations, Title 10, Part 50.90, Carolina Power &

Light Company (CP&L) requests a revision to the Technical Specifications (TS) for the IIarris Nuclear Plant (HNP). The proposed amendment to Technical Specification (TS) 3/4.4.9, "Pressure/Temperature Limits,"revises chemistry data (nickel content) shown on TS Figures 3.4-2 and 3.4-3. In addition, the associated Bases 3/4.4.9, is revised to reflect changes to chemistry and material properties and changes to comply with recent NRC rule changes to 10 CFR 50, Appendix G.

Also attached to this letter is a corresponding revision to the Reactor Vessel Surveillance Capsule Report, originally submitted on April2, 1992, in compliance with 10 CFR 50, Appendix H. This license amendment request and revision to the Reactor Vessel Surveillance Capsule Report are being submitted in accordance with commitments contained in the HNP response to NRC Generic Letter 92-01, Revision 1, Supplement 1, dated November 16, 1995.

Enclosure 1 provides a description ofthe proposed changes and the basis for the changes. details, in accordance with 10 CFR 50.91(a), the basis for the Company's determination that the proposed changes do not involve a significant hazards consideration. provides an environmental evaluation which demonstrates that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment needs to be prepared in connection with the issuance ofthe amendment.

970i060058 9'hi230

PDR, ADGCK 05000400 p

PDR State Road 1134 New Hill NC Tel 919 362-2502 Fax 919 362-2095

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Document Control Desk HNP-96-206 / Page 2 provides page change instructions for incorporating the proposed revisions. provides the proposed Technical Specification pages. contains the Reactor Vessel Surveillance Capsule Report Revision.

CP&L requests that the proposed amendment be issued such that implementation willoccur within 60 days ofissuance to allow time for procedure revision and orderly incorporation into copies ofthe Technical Specifications.

Please refer any questions regarding this submittal to Ms. D. B. Alexander at (919) 362-3190.

Sincerely, W. R. Robinson JHE/jhe

Enclosures:

1. Basis for Change Request 2.

10 CFR 50.92 Evaluation

3. Environmental Considerations

4. Page Change Instructions

5. Technical Specification Pages

6. Reactor Vessel Surveillance Capsule Report Revisions W. R. Robinson, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best ofhis information, knowledge and belief; and the sources of his information are employees, contractors, and agents ofCarolina Power 8: Light Company.

My commission expires: Q- (p -+c g g Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Dayne H. Brown, N.C. DEHNR Mr. S. D. Ebneter, NRC Regional Administrator Mr. N. B. Le, NRC Project Manager Notary al)

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Document Control Desk HNP-96-206 / Page 3 bc:

Ms. P. B. Brannan Mr. H. K. Chernoff (RNP)

Mr. G. W. Davis Mr. J. W. Donahue Ms. S. F. Flynn Mr. H. W. Habermeyer, Jr.

Mr. M. D. Hill Mr. W. J. Hindman Ms. W. C. Langston (PEEcRAS File)

Mr. R. D. Martin Mr. W. S. Orser Mr. G. A. Rolfson Mr. R. S. Stancil Mr. M. A. Turkal (BNP)

Mr. T. D. Walt Nuclear Records Harris Licensing File File: H-X-0511

N'

ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE 1

SHEARON HARRIS NUCLEARPOWER PLANT NRC DOCKETNO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/TEMPERATURE LIMITS HN R

FT Qack~rgnnl Title 10 ofthe Code ofFederal Regulations, Part 50, Appendix A, General Design Criterion 31, "Fracture Prevention ofReactor Coolant Pressure Boundary", requires that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions: (1) the boundary behaves in a nonbrittle manner, and (2) the probability ofrapidly propagating fracture is minimized.

Appendix G to 10 CFR 50, "Fracture Toughness Requirements", describes specific requirements for fracture toughness and reactor vessel operation to meet the 10 CFR 50.60 criteria regarding prevention ofbrittle fracture. In addition, Appendix G requires changes in &acture toughness of reactor vessel materials caused by neutron radiation throughout the service lifeofnuclear reactors to be considered in the limits on operation.

Regulatory Guide 1.99 contains procedures for calculating the effects ofneutron radiation embrittlement oflow-alloy steels used for light water-cooled reactor vessels.

In accordance with 10 CFR 50.36(c)(2), limitingconditions for operation are to be included in a plant's Technical Specification (TS). Technical Specifications 3.4.9.1 and 3.4.9.2, "REACTOR COOLANTSYSTEM PRESSURE/TEMPERATURE LIMITS",provide Reactor Coolant System (RCS) Pressure-Temperature Limits to protect the reactor pressure vessel from brittle fracture by clearly separating the region ofnormal operations from the region where the reactor vessel may be subject to brittle fracture. The current RCS Pressure-Temperature Limitations for the Harris Nuclear Plant (HNP) were developed in accordance with 10 CFR 50 Appendix G criteria and the calculative procedure for determining the adjusted reference temperature in Regulatory Guide 1.99, Revision 2, for a predicted reactor vessel neutron irradiation equivalent to eleven Effective Full Power Years (EFPY). These Pressure-Temperature Limits were approved by the NRC on August 20, 1993, as Amendment 38 to the Operating License.

NRC Generic Letter 92-01, Revision 1, "Reactor Vessel Structural Integrity, 10 CFR 50.54 (f),"

requested licensees to provide the NRC with specific information relative to reactor vessel integrity. Carolina Power &Light Company (CP&L)provided a response for HNP in a letter dated July 6, 1992. By letter dated May 13, 1994, the NRC requested CP&L to verify certain information contained in the NRC's Reactor Vessel Integrity Database (RVID)relative to HNP Pressurized Thermal Shock (PTS) and Upper Shelf energy (USE) parameters.

HNP provided a Page E1-1

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ENCLOSURE TO SERIAL: HNP-96-206 response in a letter dated June 10, 1994. The NRC issued Supplement 1 to Generic Letter 92-01, Revision 1, dated May 19, 1995. This Supplement requested licensees to identify, collect and report any new data pertinent to the analysis ofstructural integrity and to assess the impact ofthat data on their Reactor Pressure Vessel's (RPV's) integrity analyses relative to the requirements of 10 CFR 50.60, 10 CFR 50.61 and 10 CFR Part 50 Appendices G and H. These requirements relate to PTS, USE, Pressure-Temperature (P-T) Limits and Low Temperature Overpressure Protection System (LTOPS) setpoints.

HNP provided a 90-day response to Part 1 ofGL 92-01, Revision 1, Supplement 1, in a letter dated August 17, 1995 and a 6-month response to Parts 2, 3

&4 ofGL 92-01, Revision 1, Supplement 1, in a letter dated November 16, 1995. The NRC issued a closeout letter for GL 92-01, Revision 1, Supplement 1 in regard to HNP, dated August 7, 1996.

In response to GL 92-01, Revision 1, Supplement 1, HNP reviewed available reactor pressure vessel beltline material data and identified plants possessing the same beltline weld heats as those contained within the HNP vessel.

Also, CP&Lparticipated in a cooperative data sharing activity with these plants and the Westinghouse Owners Group in an effort to establish "best estimate chemistry" for the beltline materials in the HNP vessel.

As a result ofthe assessment, the best estimate chemistry has been revised for some ofthe beltline materials; and there were some minor material property changes (T>>r and USE) regarding weld heat 5P6771.

The response also indicated that these changes did not adversely impact reactor vessel integrity, however, some documentation changes were necessary, specifically to the chemistry data contained in the inset to Technical Specification Figures 3.4-2 and 3.4-3, Reactor Coolant System Cooldown and Heatup Limitations. Also, the response indicated that these changes would be submitted to the NRC during 1996.

A Rule change to 10 CFR 50, Appendix G was published in the Federal Register on December 19, 1995 and was made effective January 18, 1996. In part, this amended rule described the conditions pertaining to In-Service Leak &Hydrotests (ISLH) and changed the American Society ofMechanical Engineers (ASME) Code Section to be used for the methodology ofdeveloping the Pressure-Temperature Limits referred to above.

The purpose ofthis Technical Specification Change Request is to revise Figures 3.4-2 and 3.4-3 as stated above and to implement the chemistry and material property changes with'respect to the Technical Specification Bases.

Further, as a result ofRule changes related to 10 CFR 50, Appendix G, effective January 18, 1996, a revision to the Bases ofthe HNP Technical Specifications is being implemented to comply with the new Rule changes.

This change revises the chemistry data in the inset to TS Figures 3.4-2 and 3.4-3 (TS 3/4.4.9).

Specifically, the nickel content for the controlling material plate ofthe reactor vessel beltline region, A9153-1, is revised from 0.45% to 0.46%. In addition, the proposed amendment revises the associated Bases, to reflect: (i) chemistry data changes (nickel and copper) for reactor vessel Page E1-2

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ENCLOSURE TO SERIAL: HNP-96-206 beltline materials; (ii)dropweight temperature (T~~) and Upper Shelf Energy (USE) changes for the reactor vessel circumferential weld 5P6771; (iii)the provision to perform In-Service Leak &

Hydrotests (ISLH) using the ISLH Pressure-Temperature Limits whenever fuel is in the reactor vessel; (iv) to require that ISLH tests required by the ASME Code to be completed before the core is made critical and (v) to revise the reference for the Pressure-Temperature Limits requirements from ASME Section III,Appendix G to ASME Section XI,Appendix G.

~Basi As indicated in the November 16, 1995, letter to the NRC, the changes in the copper and nickel values for the reactor vessel beltline materials were due to an assessment ofthe best estimate chemistry made in response to NRC Generic Letter 92-01, Revision 1, Supplement

1. For the affected materials, the changes reflect a reduction in copper and nickel content, except for base metal plates A9153-1, C9924-1 and C9924-2 which experienced a slight increase in nickel content.

The best estimate chemistry for controlling plate A9153-1 has been determined to be 0.09%

copper and 0.46% nickel resulting in a chemistry factor of58 (Table 2 - Regulatory Guide 1.99, Revision 2). Although this is a very slight increase in the current nickel content value, (0.45%),

the chemistry factor is not affected. Therefore the Pressure-Temperature Limits in Figures 3.4-2 and 3.4-3, which are based upon the controlling plate material, A9153-1, remain unaffected by the slight increase in nickel content.

Since the Pressure-Temperature Limits remain unaffected, there is no adverse impact on the integrity ofthe Reactor Coolant System by the proposed changes.

However, the chemistry data in the inset to TS Figures 3.4-2 and 3.4-3 and in the BASES Table B 3/4.4-1 must be revised.

None ofthe remaining base metal plate nickel increases (for plates C9924-1 and C9924-,2) affected the chemistry factors used in the pressurized thermal shock reference temperature (RT~) or in the nil-ductilitytransition reference temperature (RT~r) determination.

For reactor vessel plate C9924-1, the best estimate chemistry was determined to be 0.08% copper and 0.47%

nickel representing a very slight increase in nickel content over the value, 0.45%, currently stated in the BASES Table B 3/4.4-1. However, the chemistry factor remains the same at 51. For reactor vessel plate C9924-2, the best estimate chemistry was determined to be 0.08% copper and 0.47% nickel representing a very slight increase in nickel content over the value, 0.44%,

currently stated in the BASES Table B 3/4.4-1. However, the chemistry factor remains the same at 51. Those beltline materials which experienced a reduction in copper or nickel content resulted in a chemistry factor that remained the same or which could have been reduced.

For example, for reactor vessel plate B4197-2, the best estimate chemistry has been determined to be 0.09% copper and 0.50% nickel representing a slight reduction in copper content over the value, 0.10%, currently stated in the BASES Table B 3/4.4-1. For the reactor vessel intermediate and lower axial (longitudinal) welds, 4P4784, the best estimate chemistry has been determined to be 0.05% copper and 0.91% nickel representing a slight reduction in copper content over the value, 0.06%, currently stated in the BASES Table B 3/4.4-1. For the reactor vessel circumferential Page El-3

ENCLOSURE TO SERIAL: HNP-96-206 (girth) weld, 5P6771, the best estimate chemistry has been determined to be 0.03% copper and 0.94% nickel representing a slight reduction in the copper and nickel contents over the values, 0.04% &0.95% respectively, currently stated in the BASES Table B 3/4.4-1. However, for the reasons stated in the November 16, 1995, letter, CP&L has elected at this time to retain the existing docketed chemistry factors as previously reported to you. As stated above, the best estimate copper content for reactor vessel beltline materials, plate B4197-2, weld 4P4784 and weld 5P6771 were reduced slightly. A lower copper content results in a smaller percentage reduction in Upper Shelf Energy (USE) when using the Regulatory Guide 1.99, Revision 2, Position 1.2 method, which is the method applied to the HNP predicted End-Of-Life (EOL) USE.

However, the percentage reduction in USE due to the reduction in copper content for these beltline materials was conservatively maintained the same, (i.e. unchanged).

Since the chemistry data for the reactor vessel beltline materials is contained in the BASES Table B 3/4.4-1, it must therefore be revised.

CP&L'sNovember 16, 1995 letter stated that a review of newly acquired and available data also indicated a minor change to material properties with respect to weld material heat number 5P6771.

This weld material was used in the reactor vessel circumferential weld joining the intermediate and lower shell base metal plates in the beltline region. The dropweight temperature, T~~, has been determined to'be -80'F based on unirradiated surveillance weldment test data, versus -20'F currently stated in the BASES Table B 3/4.4-1. However, this does not impact the initial or unirradiated nil-ductilitytransition reference temperature, RT~~, as stated in BASES Table B 3/4.4-1 at -20'F, since it is now based on the surveillance weldment temperature for the Charpy 50 ft-lb value, less 60'F, which results in the same RT~r value,

-20'F. Further, based on the additional test data, the initial or unirradiated USE for the reactor vessel circumferential weld has been determined to be 80 ft-lb, versus 88 ft-lbcurrently stated in the BASES Table B 3/4.4-'1. The initial USE remains above the 75 ft-Ibs value prescribed in 10 CFR 50, Appendix G, paragraph IV.A.1. Although this adversely affects the irradiated EOL USE, the value willremain greater than the 50 ft-lblimitprescribed by 10 CFR 50, Appendix G, paragraph IV.A.1. Specifically, the EOL USE is predicted to be 60 ft-lbs at the inside surface and 62 ft-ibs at the quarter thickness (T/4) location. This prediction is based on the Regulatory Guide 1.99, Revision 2, Position 1.2 method, using a conservative percentage reduction in USE, as described above. This weld material is included in the surveillance capsule program, therefore it is expected that the percent reduction in irradiated USE would not be as great ifthe benefit of the surveillance capsule results were applied as allowed by RG 1.99, Revision 2, Position 2.2.

Since the T~~ and USE for the circumferential weld material heat number 5P6771 are contained in the BASES Table B 3/4.4-1, they must therefore be revised.

The existing BASES 3/4.4.9 only allows ISLH testing using the ISLH Pressure-Temperature Limits, provided fuel is removed from the reactor vessel. ISLH testing with fuel in the reactor vessel is allowed using the normal Pressure-Temperature Limits. This was implemented as part ofa change to the BASES 3/4.4.9 in Amendment 38 to the Operating Licence proposed in CP&L letter dated February 26, 1993. The change was made to provide consistency with the recommendations ofWelding Research Council Bulletin 175, the source input document for the Page E1-4

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ENCLOSURE TO SERIAL: HNP-96-206 ASME Appendix G Pressure-Temperature Limits criteria/methodology, in the absence ofclear guidance or criteria in 10 CFR 50, Appendix G. As a result ofthe amended Rule for 10 CFR 50, Appendix G, effective January 18, 1996, the rule clarified, as described in'paragraph IV.A.2and Table 1, the acceptability ofallowing ISLH tests to be performed with fuel in the reactor vessel using the ISLH Pressure-Temperature Limits. Therefore the proposed change to BASES 3/4.4.9 in this submittal does not conflict with any regulation. Also, the amended Rule for 10 CFR 50, Appendix G, clarified, as described in paragraph IV.A.2.d, that any ISLH tests required by ASME Section XImust be completed prior to allowing the core to go critical. This is not explicitly stated in the existing BASES 3/4.4.9, therefore the proposed change to BASES 3/4.4.9 makes this clear.

In addition, the amended Rule for 10 CFR 50, Appendix G, revised the particular Section ofthe ASME Code to be used for the development ofthe Pressure-Temperature Limits. Specifically, the change referenced ASME Section XI,Appendix G versus ASME Section III,Appendix G currently referenced in BASES 3/4.4.9. The NRC has stated in the Federal Register Notice for the amended Rule that changing ofthe reference from ASME Section III,Appendix G to ASME Section XI,Appendix G has no impact because the requirements in the Appendices are identical.

Therefore, BASES Table B 3/4.4-1, is being revised to be consistent with the new Rule.

~i~el ~n The previously reported chemistry factors for the reactor vessel beltline materials are not adversely affected by the minor changes in chemistry composition described above. With this consideration and the fact that the initialRT>>r for weld heat 5P6771 has not changed due to the reduction in T>>r value, there is no impact to previously reported values for the adjusted nil-ductilitytransition reference temperature (ART>>~) or pressurized thermal shock reference temperature (RT~). Consequently, there is no adverse impact on the operating Pressure-Temperature (P-T) Limits described in the HNP Technical Specifications, Low Temperature Overpressure Protection System (LTOPS) setpoints, or Emergency Operating procedures (EOP's). The reduction in USE at EOL for weld 5P6771 remains above the criterion specified in 10 CFR 50, Appendix G and therefore does not adversely affect reactor vessel integrity. Other beltline material USE values at EOL remain unaffected.

The proposed ISLH test condition is in compliance with the amended Rule, 10 CFR 50, Appendix G, and is therefore acceptable.

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ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE 2 SHEARON HARRIS NUCLEARPOWER PLANT NRC DOCKETNO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/TEMPERATURE LIMITS V

The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation ofthe facilityin accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences ofan accident previously evaluated, (2) create the possibility ofa new or different kind ofaccident from any accident previously evaluated, or (3) involve a significant reduction in a margin ofsafety. Carolina Power &Light Company has reviewed this proposed license amendment request and determined that its adoption would not involve a significant hazards determination.

The bases for this determination are as follows:

e a

e This Technical Specification change revises the REACTOR COOLANTSYSTEM Specification 3/4.4.9, PRESSURE/TEMPERATURE LIMITS,by revising the chemistry data in the inset to Figures 3.4-2 and 3.4-3. Specifically, the nickel content for the controlling material plate ofthe reactor vessel beltline region, A9153-1, is revised from 0.45% to 0.46%. In addition, the proposed amendment revises the BASES 3/4.4.9, PRESSURE/TEMPERATURE LIMITS,to reflect: (i) chemistry data changes (nickel and copper) for reactor vessel beltline materials; (ii) dropweight temperature (T~~) and Upper Shelf Energy (USE) changes for the reactor vessel circumferential weld 5P6771; (iii)the provision to perform In-Service Leak &Hydrotests (ISLH) using the ISLH Pressure-Temperature Limits whenever fuel is in the reactor vessel; (iv) to require that ISLH tests required by the ASME Code to be completed before the core is made critical and (v) to revise the reference for the Pressure-Temperature Limits requirements from ASME Section III,Appendix G to ASME Section XI,Appendix G.

~Basi This change does not involve a significant hazards consideration for the followingreasons:

The proposed amendment does not involve a significant increase in the probability or consequences ofan accident previously evaluated.

There are no physical changes to any plant equipment created by the proposed changes.

The chemistry and material property changes do not impact the ability ofthe reactor Page E2-1

h ENCLOSURE TO SERIAL: HNP-96-206 vessel to maintain it's pressure boundary integrity as previously evaluated.

The decrease in EOL USE for weld heat 5P6771 is relatively minor and remains above the required value that has been prescribed by the NRC to provide the necessary level ofductility assumed for reactor vessel integrity evaluations.

Therefore, the accident initiating and mitigating aspects ofthe pressure vessel are not affected. In addition, neither the proposed change requiring the ISLH test to be complete before the core is critical nor the proposed change allowing fuel in the reactor vessel during ISLH affects any accident initiating mechanisms. The proposed change requiring the ISLH test to be completed before the core is critical willnot increase the consequences ofpreviously evaluated accidents because it conservatively assures the core is subcritical. Although the proposed change allows fuel in the vessel during ISLH utilizing the ISLH Pressure-Temperature (P-T) limits, the consequences ofa pressure boundary leak have not changed because ISLH testing is already allowed using the normal plant P-T limits. In addition, the ISLH willbe required to be completed before the core is allowed to go critical. The consequences ofa leak with fuel in the vessel during ISLH are the same using either the normal P-T limits or the ISLH limits.

Therefore, there would be no increase in the probability or consequences ofan accident previously evaluated.

The proposed amendment does not create the possibility ofa new or different kind of accident from any accident previously evaluated.

There are no physical changes to any plant equipment or new components created by the proposed changes.

The chemistry and material property changes do not impact the pressure boundary integrity ofthe reactor vessel.

The decrease in EOL USE for weld heat 5P6771 is relatively minor and remains above the required value that has been prescribed by the NRC to provide the necessary level ofductility assumed for reactor vessel integrity evaluations.

Therefore, the accident initiating aspects ofthe pressure vessel are not affected. In addition, neither the proposed change requiring the ISLH test to be complete before the core is critical nor the proposed change allowing fuel in the reactor vessel during ISLH creates any new accident initiating mechanisms.

Therefore, the proposed change does not create the possibility ofa new or different kind ofaccident from any accident previously evaluated.

The proposed amendment does not involve a significant reduction in the margin ofsafety.

The changes in chemical and material properties do not adversely affect any reactor vessel integrity evaluations, such as PTS or P-T limits. The USE for weld heat 5P6771 does decrease slightly as described in TS Bases Table B 3/4.4-1. However, the predicted EOL USE remains above the value prescribed in 10 CFR 50, Appendix G and is not a significant reduction in the margin ofsafety. With regard to the proposed changes Page E2-2

ENCLOSURE TO SERIAL: HNP-96-206 allowing fuel in the reactor vessel during ISLH, the existing TS Bases specifically state that fuel is not to be in the reactor vessel when the ISLH P-T curve is utilized. However, this change is consistent with the revised 10 CFR 50, Appendix G rule and as such, is not a significant reduction in the margin ofsafety.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE3 SHEARON HARRIS NUCLEARPOWER PLANT NRC DOCKETNO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/TEMPERATURE LIMITS ENVIR N NT I R TI 10 CFR 51.22(c)(9) provides criterion for and identification oflicensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.

A proposed amendment to an operating license for a facilityrequires no environmental assessment if operation ofthe facilityin accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts ofany effluents that may be released offsite; (3) result in a significant increase in individual or cumulative occupational radiation exposure.

Carolina Power &Light Company has reviewed this request and determined that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance ofthe amendment.

The basis for this determination follows:

hne This Technical Specification change revises the REACTOR COOLANTSYSTEM Specification 3/4.4.9, PRESSURE/TEMPERATURE LIMITS,by revising the chemistry data in the inset to Figures 3.4-2 and 3.4-3. Specifically, the nickel content for the controlling material plate ofthe reactor vessel beltline region, A9153-1, is revised from 0.45% to 0.46%. In addition, the proposed amendment revises the BASES 3/4.4.9, PRESSURE/TEMPERATURE LIMITS,to reflect: (i) chemistry data changes (nickel and copper) for reactor vessel beltline materials; (ii) dropweight temperature (TND~) and Upper Shelf Energy (USE) changes for the reactor vessel circumferential weld 5P6771; (iii)the provision to perform In-Service Leak &, Hydrotests (ISLH) using the ISLH Pressure-Temperature Limits whenever fuel is in the reactor vessel; (iv) to require that ISLH tests required by the ASME Code to be completed before the core is made critical and (v) to revise the reference for the Pressure-Temperature Limits requirements from ASME Section III,Appendix G to ASME Section XI,Appendix G.

ai The change meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1.

As demonstrated in Enclosure 2, the proposed amendment does not involve a significant hazards consideration.

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ENCLOSURE TO SERIAL: HNP-96-206 2.

The proposed amendment does not result in a significant change in the types or increase in the amounts ofany effluents that may be released offsite.

The proposed change does not involve any new equipment or require existing systems to perform a different type offunction than they are currently designed to perform. The change does not introduce any new effluents or increase the quantities ofexisting effluents. As such, the change cannot affect the types or amounts ofany effluents that may be released offsite.

3.

The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure.

The proposed change does not result in any physical plant changes or new surveillances which would require additional personnel entry into radiation controlled areas.

Therefore, the amendment has no affect on either individual or cumulative occupational radiation exposure.

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ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE4 SHEARON HARRIS NUCLEARPOWER PLANT NRC DOCKETNO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/TEMPERATURE LIMITS A F T

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