ML20045F955

Proposed Tech Specs Requesting Deletion of Surveillance Requirement Containing Withdrawal Schedule for Removing Rv Flux Monitoring Surveillance Capsules from TS

ML20045F955
Person / Time
Site:	FitzPatrick
Issue date:	06/24/1993
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20045F879	List: ML20045F878 ML20045F953 ML20045F955
References
NUDOCS 9307090225
Download: ML20045F955 (12)

Text

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. Attachment I to JPN-93-043 1 i

PROPOSED TECHNICAL SPECIFICATION CHANGES ,

REMOVAL OF WITHDRAWAL SCHEDULE FOR REACTOR VESSEL MATERIAL SURVEILLANCE CAPSULES i

(JPTS-91-003)

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4 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT

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Docket No. 50-333 DPR-59 9307090225 930624 E PDR ADOCK 05000333  ? .

P PDR j

JAFNPP '

3.G (cont'd) 4.6 (cont'd) i B. Deleted B. Deleted C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system radioactivity concentration in 1. a. A samne of reactor coolant shall be taken at least water shall not exceed the equilibrium value of 3.1 every 96 hr and analyzed for gross gamma activity.

pCi/gm of dose equivalent 1-131. This limit may be exceeded, following a power transient, for a maximum of b. Isotopic analysis of a sample of reactor coolant shall 48 hr. During this iodine activity transient the iodine be made at least once/ month.

concentrations shall not exceed the equilibrium limits by more than a factor of 10 whenever the main steamline c. A sample of reactor coolant shall be taken prior to isolation valves are open. The reactor shall not be startup and at 4 hr intervals during startup and operated more than 5 percent of its annual power analyzed for gross gamma activity.

operation under this exception to the equilibrium limits. If the iodine concentration exceeds the equilibrium limit by d. During plant steady state operation and following an more than a factor of 10, the reactor shall be placed in a offgas activity increase (at the Steam Jet Air cold condition within 24 hr. Ejectors) of 10,000 pCi/sec within a 48 hr. period or a power level change of 2:20 percent of full rated power /hr reactor coolant samples shall be taken and analyzed for qross gamma activity. At least three samples will be taken at 4 hr intervals. These sampling requirements may be omitted whenever the equilibrium l-131 concentration in the reactor coolant is less than 0.007 pCi/ml.

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Amendment No.1[9' 139

JAFNPP 3.6 and 4.6 BASES (cont'd)

The expected neutron fluence at the reactor vessel wall can be The RTmr for the remainder of the vesselis 40 F.

determined at any point during plant life based on the linear relationship between the reactor thermal power output and the The actual shift in the RTmr of the vessel material will be curresponding number of neutrons produced. Accordingly, established periodically by removing and evaluating flux neutron flux wires were removed from the reactor vessel with monitoring surveillance capsules in accordance with ASTM E the surveillance specimens to establish the correlation at the 185-82 and 10 CFR 50, Appendix H. The evaluation findings capsulo location by experimental methods. The flux and recommendations of Regulatory Guide 1.99 Revision 2 will distribution at the vessel wall and 1/4 thickness (1/4T) depth provide the basis for revising Figure 3.6-1 curves A, B and C was analvtically determined as a function of core height and for operation of the plant. The first surveillance capsule azimuth to establish the peak flux location in the vessel and containing test specimens was withdrawn in April,1985 after the lead factor of the surveillance specimens. 6 EFPY. The test specimens removed were tested according to ASTM E 185-82 and the results are in GE report MDE Regulatory Guide 1.99, Revision 2 is used to predict the shift 0386. The NRC approved schedule for subsequent specimen [

in RTer as a function of fluence in the reactor vessel beltline withdrawal is located in the updated FSAR (Section 4.2.7). l region. An evaluation of the irradiated surveillance specimens, which were withdrawn from the reactor in April,1985 (6 Figure 3.6-1 is comprised of three parts: Part 1, Part 2, and EFPY), shows a shift in RTmr less than that predicted by Part 3. Parts 1,2, and 3 establish the pressure-temperature Regulatory Guide 1.99, Revision 2. limits for plant operations through 12,14, and 16 Effective Full Power Years (EFPY) respectively. The appropriate figure Operating limits for the reactor vessel pressure and and the pressure-temperature curves are dependent on the temperature during normal heatup and cooldown, and during number of accumulated EFPY. Figure 3.6-1, Part 1 is for in-service hydrostatic and leak testing were established using operation through 12 EFPY, Figure 3.6-1, Part 2 is for 10 CFR 50 Appendix G, May,1983 and Appendix G of the operation at greater than 12 EFPY through 14 EFPY, and Summer 1984 Addenda to Section lli of the ASME Boiler and Figure 3.6-1, Part 3 is for operation at greater than 14 EFPY Pressure Vessel Code. These operating limits assure that the through 16 EFPY. The curves contained in Figure 3.6-1 are vessel could safely accommodate a postulated surface flaw developed from the Genatal Electric Report DRF 137-0010, having a depth of 0.24 inch at the flange-to-vessel junction, " Implementation of Regetatory Guide 1.99, Revision 2 for the and one-quarter of the material thickness at all other reactor James A. FitzPatrick Nuclear Power Plant," dated June,1989.

vessellocations and discontinuity regions. For the purpose of setting these operating limits, the reference temperature, Figure 3.6-1 curve A establishes the minimum temperature for RTer, of the vessel material was estimated from impact test hydrostatic and leak testing required by the ASME Boiler and data taken in accordance with the requirements of the Code to Pressure Vessel Code,Section XI. Test pressures for which the vessel was designed and manufactured (1965 in-service hydrostatic and leak testing are a function of the Edition including Winter 1966 addenda). The RT er values for testing temperature and the component material. Accordingly, the reactor vessel flange region and for the reactor vessel shell the maximum hydrostatic test pressure will be 1.1 times the beltline region are 30'F, based on fabrication test reports. operating pressure of about 1105 psig.

Amendment No.1 ,1/18, 147

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Attachment ll to'JPN-93-043 i

. SAFETY EVALUATION FOR l

PROPOSED TECHNICAL SPECIFICATION CHANGES

REMOVAL OF WITHDRAWAL SCHEDULE FOR REACTOR VESSEL MATERIAL SURVEILLANCE CAPSULES (JPTS-91-003) l i

i 1. DESCRIPTION OF THE PROPOSED CHANGES

! This application for an amendment to the James A. FitzPatrick Technical

Specifications removes the withdrawal schedule for reactor vessel flux monitoring i surveillance capsules.

j Minor changes in format, such as type font, margins or hyphenation, are not j described in this submittal. These changes are typographical in nature and do not 1 affect the content of the Technical Specifications.

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j Pace 139. Soecification 4.6.A.7 j Delete the Specification which reads:

" Reactor Vessel Flux Monitoring l The reactor vessel Flux Monitoring Surveillance Program complies with the intent j of the May,1983 revision to 10 CFR 50, Appendices G and H. The next flux 1 monitoring surveillance capsule shall be removed after 15 effective full power years (EFPYs) and the test procedures and reporting requirements shall meet the requirements of ASTM E 185-82."

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l Paae 147 Bases 3.6 and 4.6 Section A 3

in the beginning of the fourth paragraph add the following sentences:

"The actual shift in the RTa of the vessel material will be established periodically l by removing and evaluating flux monitoring surveillance capsules in accordance j with ASTM E 185-82 and 10 CFR 50, Appendix H. The evaluation findings and i recommendations of Regulatory Guide 1.99 Revision 2 will provide the basis for revising Figure 3.6-1 curves A, B and C for operation of the plant."

l At the end of the fourth paragraph replace the last sentence:

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"The next surveillance capsule will be removed after 15 EFPYs of operation and

the results of the examination used as a basis for revision of Figure 3.6-1 curves j A, B and C for operation of the plant after 16 EFPYs."

with the sentence:

i 4 "The NRC approved schedule for subsequent specimen withdrawal is located in i the updated FSAR (Section 4.2.7)."

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Attachrnent 11 to JPiV-93-043 i SAFETY EVALUATION Page 2 of 5 ll. PURPOSE OF THE PROPOSED CHANGES The proposed changes delete the Surveillance Requirement containing the withdrawal schedule for removing reactor vessel flux monitoring surveillance capsules from the Technical Specifications. This change is made using the guidance provided in NRC Generic Letter 91-01 (Reference 1). Future control of this schedule will be in accordance with 10 CFR 50, Appendix H, Section ll.B.3. The NRC approved version l of the specimen withdrawal scladule (References 2 and 3) will be maintained in the updated Final Safety Analysis Report (FSAR). This schedule will be incorporated j during the July,1993 annual revision of the updated FSAR.

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111. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES The schedule and requirements for the reactor vessel Flux Monitoring Program in Technical Specification 4.6.A.7 and Bases Section 3.6 and 4.6 A are duplicative of the requirements of 10 CFR 50, Appendix H. The removal of the schedule and the surveillance requirement from the Technical Specifications will have no effect on safety because it does not affect the Reactor Vessel Material Surveillance Program or technical requirements, results in no change to any equipment or operations, has no effect on analyses documented in the updated FSAR and Safety Evaluation Report (SER), and results in no change to regulatory control.

The proposed Technical Specification change is in accordance with the guidance contained in NRC Generic Letter 91-01. Generic Letter 91-01 allows removal of the schedule from the Surveillance Requirement or removal of the complete Surveillance 3 Requirement if it does not specify that the results of the examination are to be used

to update the pressure and temperature limits of limiting condition for operation (LCO) 3.6.A. This change is consistent with the improved Standard Technical Specification (Reference 4) which has no Surveillance Requirement for specirnen removal and updating of the LCO. Generic Letter 91-01 also allows removal of the schedule from the Bases if a commitment is made to include the schedule in the updated FSAR.

The Authority is making this commitment as part of this amendment application.

. The proposed changes to remove the schedule for the withdrawal of reactor vessel material will not result in any loss of clarity related to the regulatory requirements of l Appendix H to 10 CFR Part 50. The LCO in Technical Specification 3.6.A includes pressure and temperature limits for operation during heatup, cooldown, criticality, idle recirculation loop startup, stud tensioning and inservice leak and hydrostatic testing.

Surveillance Requirement 4.6.A specifies the frequency of verifying that the pressure and temperature limits are met. Surveillance Requirement 4.6.A.7 contains the requirement and schedule for removal and evaluation of reactor vessel material surveillance capsules in accordance with 10 CFR 50, Appendices G and H. Bases Section 3.6 and 4.6 A provide a detailed description of the bases for LCO 3.6.A as well as Surveillance Requirement 4.6.A.7. The Bases includes a description of how the pressure and temperature lirnits were derived, the relationship to regulations and regulatory guidance, and the relationship to the American Society of Mechanical Engineers (ASME) code. The proposed revision removes the requirement and schedule for removal and evaluation of the reactor vessel material surveillance capsules from Surveillance Requirement 4.6.A.7 and the scheduta from Bases Section

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I Attachment 11 to JPN-93-043 i -

SAFETY EVALUATION Page 3 of 5 l

3.6 and 4.6 A. The Bases continue to refer to 10 CFR 50, Appendix H, for the surveillance requirement and the updated FSAR for the withdrawal schedule. The Authority commits to incorporate the approved schedule in the updated FSAR.

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:

l l 1. involve a significant increase in the probability or consequences of an accident l previously evaluated.

The proposed changes will not increase the probability or consequences of an accident previously evaluated because the plant's accident analyses is not affected by the Technical Specification change. The proposed changes will not affect the Reactor Vessel Material Surveillance Program nor the requirements to update pressure and temperature operating limits resulting from reactor vessel flux monitoring surveillance capsule examination. Although the Surveillance Requirement and the withdrawal schedule will be removed,10 CFR 50 Appendix H requires that reactor vessel flux monitoring surveillance capsules be periodically removed and examined to determine changes in their material properties. The NRC approved schedule will be in the updated FSAR. Therefore, no reduction in the overall effectiveness of the Reactor Vessel Material Surveillance Program will result from the proposed changes.

2. create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not create the possibility of a new or different kind of accident from those previously evaluated because they will not require i modification to any plant structures, systems, components or practices.

10 CFR 50, Appendix H, will continue to require reactor vessel flux monitoring surveillance capsules be periodically removed and examined to determine changes to pressure and temperature operating limits. The absence of any changes to plant hardware or to the withdrawal schedule ensures that accident initiators are i unaffected.

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3. involve a significant reduction in the margin of safety.  !

The proposed changes will not cause a reduction in the margin of safety. The results of the plant accident analyses continue to bound operation under the proposed changes so there is no reduction in the margin of safety. Removal of the Surveillance Requirement containing the withdrawal schedule for reactor vessel flux monitoring surveillance capsules from the Technical Specifications will not result in any loss of regulatory control. Changes to the withdrawal schedule are controlled by the requirements of Appendix H to 10 CFR Part 50. The actual

l f . Attachment il to JPN-93-043 i* SAFETY EVALUATION Page 4 of 5 j .

l f withdrawal schedule will be added to the updated FSAR in the next revision made in accordance with 10 CFR 50.71(e). Removal of the Surveillance Requirement

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1 containing the withdrawal schedule for reactor vessel flux monitoring surveillance capsules will not result in any loss of clarity related to the regulatory requirements l

j of Appendix H to 10 CFR Part 50. The Bases for Section 3.6 and 4.6 A provide l

background information on the use of the data obtained from material specimen j examination.

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l V. IMPLEMENTATION OF THE PROPOSED CHANGES '  :

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i implementation of the proposed changes removes the withdrawal scheduie for

reactor vessel material surveillance capsules. This removal will not affect the ALARA  ;

{ or Fire Protection Programs at the FitzPatrick plant, nor will the changes affect the l i environment. )

l I VI. CONCLUSION i

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

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1. will not change the probability nor the consequences of an accident or  ;

malfunction of equipment important to safety as previously evaluated in the I Safety Analysis Report;

2. will not increase the possibility of an accident or malfunction of a type different

. from any previously evaluated in the Safety Analysis Report; and 4

l 3. will not reduce the margin of safety as defined in the basis for any technical specification.

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i The changes involve no significant hazards consideration, as defined in i 10 CFR 50.92.

! Vll. REFERENCES

1. NRC Generic Letter 91-01, " Removal of the Schedule for the Withdrawal of 3

Reactor Vessel Material Specimens from Technical Specifications," dated January 4,1991.

2. NYPA letter, J. C. Brons to the NRC dated March 16,1987 (JPN-87-013),

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" Proposed Changes to the Technical Specifications Regarding Pressure-l Temperature Limits."

3. NRC letter, H. Abelson to J. C. Brons dated October 22,1987 ([[::JAF-87-252|JAF-87-252]]),

, Transmits Amendment 113 to the Technical Specifications.

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Attachment il to JPN-93-043 SAFETY EVALUATION Page 5 of 5

4. NRC NUREG-1433 " Standard Technical Specifications in General Electric Boiling Water Reactors (BWR/4)," Revision 0, dated September 1992.

5. American Society of Testing Materials (ASTM) " Conducting Surveillance Tests For Light-Water Cooled Nuclear Power Reactor Vessels," ASTM 185-82.

6. Code of Federal Regulations 10 CFR 50, Appendix H " Reactor Vessel Material Surveillance Program Requirements," dated January 1,1993.

7. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 4.2.7 " Inspection and Testing," through Revision 5, dated January 1992.

8. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.

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- Attachment lll to JPN-93-043 PROPOSED TECHNICAL SPECIFICATION CHANGES

- REMOVAL OF WITHDRAWAL SCHEDULE FOR REACTOR VESSEL MATERIAL SURVEILLANCE CAPSULES MARKUP OF TECHNICAL SPECIFICATION PAGES (JPTS 91-003) i l

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i l New York Power Authority l

JAMES A. FITZPATRICK NUCLEAR POWER PLANT .

Docket No. 50-333 DPR-59

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

r. Reactor Vessel FluYonitoring The reactor vessel Flux Monitoring Surveillance Program complies with the intent of the May,1983 revision to 10 CFR 50. Appendices G and H. The next flux monitoring surveillance capsu!c shall be removed after 15 effective full ,

power years (EFPYs) and the test procedures and reporting requirements shall meet the requirements of ASTM E 185-82.

18. Deleted B. Deleted J

C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system iadioactivity concentration in 1. a. A sample of reactor coolant shall be taken at least water shall not exceed the equilibrium value of 3.1 pCi/gm every 96 hr and analyzed for gross gamma activity.

of dose equivalent 1-131. This limit may be cxceeded b. Isotopic analysis of a sample of reactor coolant shall following a power transient, for a maximum cf 48 hr.'

the iodine be made at least once/ month.

During this iodine activity transient concentrations shall not exceed the equilibrium limits by c. A sample of reactor coolant shall be taken prior to more than a factor of 10 whenever the main steamline startup and at 4 hr intervals during startup and isolation valves are open. The reactor shall not be analyzed for gross gamma activity.

operated more than 5 percent of its annual power d. During plant steady state operation and following an operation under this exception to the equilibrium limits. If offgas activity increase (at the Steam Jet Air the rodine concentration exceeds the equilibrium limit by Ejectors) of 10,000 pCi/sec within a 48 hr. period or more than a factor of 10, the reactor shall be placed in a a power level change of >20 percent of full rated cold condition within 24 hr. power /hr reactor coolant samples shall be taken and analyzed for gross gamma activity. At least three samples will be taken at 4 hr intervals. These sampling requirements may be omitted whenever the equilibrium I-131 concentration in the reactor coolant is less than 0.007 pCi/mt.

Amendment No.1 139

1 JAFNPP ,

3.6 and 4.6 BASES (cont'd)

The expected noutron fluenco at the reactor vessel wall can be vessel flange region and for the reactor vossel shell bottline dotormined at any point during plant lifo based on the linear region are 307, based on fabrication test reports. The RT t4DT relationship betwoon the reactor thermal power output and the for the remainder of the vossol is 40 F.

corresponding number of noutrons produced. Accordingly, 9 The first survoillance capsufo containing test specimens was neutron flux wires woro removed from the reactor vessel with the survolitanco specimens to establish the correlation at the withdrawn in April,1985 after 6 EFPY. The test specimens capsulo location by experimental methods. Tha flux removed were testod according to ASTM E 135-82 and the  :

results are in GE report MDE-494386./The next surveillance i distribution at the vossol wall and 1/4 thickness (1/41) depth d was analytically determined as a (unction of core height and apsuio will bo romoved after 15 EFPYs of operation and the azimuth to establish the peak flux location in the vessel and the results of the examination used as a basis for revision of Figure load factor of the surveillanco specimens. 3.6-1 curves A, B and C for operation of the plant after 16; -

Regulatory Guido 1.99, Revision 2 is usod to predict the shift in 1(c[4cd RTNOT Figure 3.6-1 is comprised of throo parts: Part 1, Part 2, and as a function of fluence in the roactor vessel bottline wiO Part 3. Parts 1,2, and 3 establish the pressuro-temperature region. An ovaluation of the irradiated survoillanco specimens, which woro withdrawn from the reactor in April,1

• d limits for plant operations through 12,14, and 16 Effective Full Power Years (EFPY) respectively. The appropriato figure and 1985 (6 EFPY), shows a shift in RTNOT less than that predicted ,,f5 the pressuro-tomporaturo curves are dopondent on the number by Regulatory Guido 1.99, Revision 2.

of accumulated EFPY. Figuro 3.6-1, Part 1 is for operation Operating limits for the reactor vessel pressure and through 12 EFPY, Figure 3.6-1, Part 2 is for operation at greater temperature during normal heatup and cooldown, and during than 12 EFPY through 14 EFPY, and Figuro 3.6-1, Part 3 is for in-service hydrostatic and leak testing woro established using operation at greator than 14 EFPY through 16 EFPY. The 10 CFR 50 Appendix G, May,1983 and Appendix G of the curves contained in Figuro 3.6-1 are developed from the Summor 1984 Addenda to Section ill of the ASME Boiler and General Electric Report DRF 137-0010, *lmplomontation of Pressuro Vessel Code. These operating limits assure that the Regulatory Guido 1.99, Revision 2 for the James A. FitzPatrick j vossol could safely accommodato a postulated surface flaw Nuclear Power Plant," dated June,1989. U having a depth of 0.24 inch at the flango-to-vessel junction, and Figure 3.6-1 curvo A establishes the minimum temperature for ono-quartor of the material thicknoss at all other roactor vessel hydrostatic and leak testing required by the ASME Boiler and locations and discontinuity regions. For the purpose of setting Pressuro Vessel Codo,Section XI. Test pressures for in-service these oporating limits, the reference temperaturo, RTuoy , of the vossol material was estimated from impact test data taken hydrostatic and leak testing are a function of the testing in accordanco with the requirements of the Codo to which the temperaturo and the component material. Accordingly, the maximum hydrostatic test pressure will be 1.1 times the vessel was designed and manufactured (1965 Edition including

' operating pressure or about 1105 psig.

Wintor 1966 addenda). The RTNor values for the reactor Amendment No.Jd , J/1k 147

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INSERT "A" The actual shift in the RTer of the vessel material will be established periodically by removing and evaluating flux monitoring surveillance capsules in accordance with ASTM E 185-82 and 10 CFR 50, Appendix H. The evaluation findings and recommendations of Regulatory Guide 1.99 Revision 2 will provide the basis for revising Figure 3.6-1 curves A, B and C for operation of the plant.

INSERT "B" The NRC approved schedule for subsequent specimen withdrawal is located in the updated FSAR (Section 4.2.7).

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