ML19332C364
| ML19332C364 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/24/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332C361 | List: |
| References | |
| GL-88-11, NUDOCS 8911280025 | |
| Download: ML19332C364 (6) | |
Text
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NUCLEAR REGULATORY COMM!SSION i
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
NO. DPR-66 i
DUQUESNE LIGHT COMPANY l
BEAVER VALLEY POWER STATION UNIT 1 DOCKET NO.60-334
1.0 INTRODUCTION
j In response to Generic Letter 88 11, "NRC Position on Radiation Embrittlement of Feactor Vessel Materials and Its Effect on Plant Operations,"
Duquesne Light Company (the licensee) proposed to revise the pressure / temperature (P/T: limits in the Beaver Valley Power Station Unit 1 Technical Specifications. The proposal was documented in a letter from the licensee dated November 23,1938. This proposal also changes the effectiveness of the P/T limits for 9.5 effective full power years (EFPY). The proposed P/T limits were based on Re' ulatory Guide (RG)1.99, Revision 2.
The proposed revision provides up-to-date I/T limits for the operation of the reactor coolant system during heatup, cooldown, criticality, l
and hydrottst.
To evdluate the P/T limits, the staff uses the following NRC regulations and guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code,whicharereferencedinAppendicesGandH;10CFR50.36(c)(2);RG1.99, l
Rev. 2; Standard Review Plant (SRP) Section 5.3.2; and Generic _ Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by l
10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular,10 CFR 50.36(c)(2) requires that limiting conditions of o >eration l
be included in the Technical Specifications. The P/T limits are among tie limit-ing conditions of operation in the Technical Specifications for all commercial nuclear plants in the U.S.
Ap)endices G and H of 10 CFR Part 50 describe specific requirements for fracture toug sness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards.
These tests define the extent of vessel embrittlement at the time of capsule with-drawal in terms of the increase in reference temperature. Appendix G also re-8911280025 891324 PDR ADOCK 05000334 P
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i 1 quires the liter.see to predict the effects of neutren irradiation on vessel em.
brittlement by calculating the adjusted reference temperature (ART) and Charpy l
uppershelfenergy(U5E). Generic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron 1
irradiation on reactor vessel materials. This guide defines the ART as the sum l
of unitradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to recount for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) uterials of the reactor beltline.
2.0 EVALUATION The st6ff evaluated the effect of neutron irraciation embrittlement on each belt-line material in the Leaver Valley Unit I reactor vessel.
The amount of neutron irradiation cmbrittleunt was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 9.5 EFPY for Beaver Valley Unit I was the lower shell plate (B6607-2) with 0.14% copper (Cu), 0.621 nickel (Ni), and an initial Ri t 73 F.
ndt The licensee has removed three surveillance capsules (V, U, and W) from Beaser Valley Unit 1.
The results from capsule V were publishec in Westinghouse Report WCAP-9860; the results from capsule b in WCAP-10867; and the results from capsule W in WCAP-12005.
All surveillance capsules contoined Charpy impact specimens and l
tensile specimens made from base metal, weld metal, and HA2 metal.
For the limiting beltline r.aterial, plate B6607-2, the staff calculated the ART at 9.5 EFP) at 1/4T (T = reactor vessel beltline thicknes)) to be 201.5'F and 175.6'F at 3/4T.
The stcff used a fluence of 8.1E18 n/cm for 1/4T and 3.15E18 n/cm' for 3/4T. The ART was determined by Section 1 of RG 1.99, Rev. 2.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 202 F at j.
9.5 EFPY at 1/4T and 176'F at 3/4T for the same limiting plate material. The staff judges that a difference of less than l'F between tie licensee's ART of 202*F and the staff's ART of 201.5'F is acceptable. Substituting the ART of 202*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Appendix G states that when the pressure exceeds 20%
of the preservice system hydrostatic test pressure, the temperature of the closure i
flange regions highly stressed by the bolt prcload must exceed the reference temperature of the materiel in those regions by at least 120*F for normal opera-
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3-tion and by 90*F for hydrostatic pressure tests and leak tests. Based on the flan 9e reference temperature of 60'F, the staff has determined that the proposed P/T lindts satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life 2
be above 50 ft-lb. Based on data from a surveillance capsule withdrawn at 5.89 EFPY, the measured transverse Charpy USE is 59 ft-lb for the Mr shell plate i
material, 86903 1.
This is a 26.21 reduction from the unirrasided value of 80 ft-lb. Using the method in RG 1.99 Rev. 2 the predicted Charpy USE of the lower shellplatematerialattheendoflifewillbebelow50ft-lb. The staff will monitor the weld metal Charpy USE from future surveillance capsules. The surveil-lance capsule data will provide early warning of the decreese in Charpy USE, because the surveillance capsult lead factors are greater than 1.0.
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3.0 CONCLUSION
L The staff concludes that the proposed P/T limits for the reactor coolant system for heatup cooldown, leak test, and criticality are v611d through 9.5 EFPY be-cause the limits conform to the requirenents of Appendices G and H of 10 CFF, Part
- 50. The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2 to calculate the ART. Hence, the proposed P/T limits may be incorporated into the Beaver Valley Unit 1 Technical Specifications via a future amendment.
l
4.0 REFERENCES
1.
Regulatory Guide 1.99 Radiation Embrittlement of Reactor Vessel Materials, Revision 2 May 1988 2.
NUREG-0800, Standard Review Plan, Section S.3.2 Pressure-Temperature Limits 3.
November 23, 1988, Letter f rom J. D. Sieber (DL) to USNRC Document Control Desk, subject:
Beaver Valley Power Station, Units 1 and 2 Response to Generic Letter 88-11 4.
January 24, 1989 LetterfromJ.D.Sieber(DL)toUSNRCDocumentControl l
Desk, subject:
Beaver Valley Power Station, Unit 1, Reactor Vessel Capsule WTestResultsReport(WCAP-12005) 5.
R. S. Boggs et al, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit Reactor Vessel R6diation Surveillance Program, WCAP-10867,"
Westinghouse Electric Corporation, September 1985 6.
.0ctober 15, 1981, Letter from J. J. Carey (DL) to S. A. Varga (USNRC),
t subject: Beaver Valle Specimen Test Report (y Power Station, Unit 1, Reactor Vessel Irradiation WCAP-9860) 7.
July 21, 1977, Letter from C. N. Dunn (DL) to R. W. Reid (USNRC), subject:
l Beaver Valley Power Station, Unit 1, Reactor Vessel Material Surveillance Program 8.
Beaver Valley Final Safety Analysis Report 5.0 PRINCIPAL CONTRIBUTORS John Tsao, with contractual assistance from the Idaho National Engineering Laboratory. Report completed in November,1989.
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}AFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION NO. NPF-73 DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 2 DOCKET h0. 50-412
1.0 INTRODUCTION
In response to Generic Letter 88-11, *NRC Position on Radiation En6rittlement of Reactor Vessel Materials and Its Effect on Plant Operations,"
Duquesne Light t
Company (the licensee) proposed to revise the pressure / temperature (P/T) limits in the Beaver Valley Power Station Unit 2 Technical Specifications, Section 3.4. The proposal was documented in a letter from the licensee dated November E3,1938. This proposal also changes the effectiveness of the P/T limits from 10 to 5 effective full power years (EFPY). The i
developed based on Section 1 of Regulatory Guide (RG) proposed P/T limits were 1.99, Revision 2.
The proposed revision provides up-to-date P/T limits for the operation of the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and Hs 10 CFR 50.36(c)(2); RG 1.99, Rev. 2; Standard Review Plant (SRP) Section 5.3.2; and Generic Letter 88-11.
Each ifcensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
.Inparticular,10CFR50.36(c)(2)'requiresthatlimitingconditionsofo>eration be included in the Technical Specifications. The P/T limits are among tie limit-ing conditions of operation in the Technical Specifications for all connercial nuclear plants in the U.S.
Ap>endices 6 and H of 10 CFR Part 50 describe specific 1
requirements for fracture touginess and reactor vessel material surveillance that aust be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards.
l These tests define the extent of vessel embrittlement at the time of capsule with-l drarsi in terms of the increase in reference temperature. Appendix G also re-
l
.,, quires the licensee to predict the effects of neutron irradiation en vessel em-brittlenent by calculating the adjusted reference temperature (ART) and Charpy uppershelfenergy(USE),
Generic Letter 88-11 requested that licensees and i
permittees use the methods in RG 1.99 irradiation on reactor vessel materials.ev. 2 to pred R
This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licer.see to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens i
made from plate, weld, and heat-affected-rcre (HAZ) matcrials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each belt-line materiel in the Beaver Valley 2 reactor vessel.
The amount of neutron irradiation embrittlement was cllculated in accordance with RG 1.90, Rev. 2.
The staff has determined that the material with the highest ART at 5 EFPY was the intermediatg shell plate with 0.07% copper (Cu), 0.53t nickel (Hi), and an initial RT of 60 F.
ndt The licensee has not withdrawn any surveillence capsules from the Beaver Valley 2 reactor vessel.
All six surveillance capsules contain Charpy impact specimens and tensile specimens made from b6se metal, weld metal, and HAZ metal.
Fortheljmitingbeltlinematerial,plateB9004-1,thesteffcalcul6tedtheART to be 132 F et 1/4T (T = reactor vessel beltline thickness) for 5 EFPY. The neu-tron fluence at 1/4T was estimated to be 6.3E18 n/cm2 for 5 EFPY, The ART was determined by Section 1 of RG 1.99, Rev. 2, because no surveillence capsule data are available.
U The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 13? F at 5 EgPY at 1/41 for the same limiting plate material. Substituting the ART of 132 F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Appendix G states that when the pressure exceeds 20%
of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must egceed the reference tem-peraturegfthematerialinthoseregionsbyatlecst120Ffornormaloperation I
and by 90 F for hydrostatig pressure tests and leak tests. Based on the flange reference temperature of 0 F, the staff has 6ctermined that the proposed P/T limits setisfy Section IV.2 of Appendix G.
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.,_ Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb. The material with the lowest unitradiated USE is intenaediate shell plate 89004 2 with 75.b tt-1b.
Based on its 0.075 Cu and Figure 2 of RG 1.99,Rev.2,theEOL(32EFPY)USEispredictedtobe53.6ft-1b. This is greater than 50 ft-lb and, therefore, is acceptable.
3.0 CONCLUSION
The staff concludes that the proposed P/T lir.its for the reactor coolant system for heatup, cooldown, leak test, ano criticality are valid through 5 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the aethod in RG !.99, Rev. 2 to calculate the ART. Hence, the proposed P/T limits may be incorporated into the Beaver Valley Unit 2 Technical Specifications via a future amendment.
4.0 REFERENCES
1.
Regulatory Guide 1.99 Radiation Ent>rittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2 Pressure-Temperature Limits 2.
hovember 23, 1988, Letter from J. D. Sieber (DL) to USNRC Docurent Control Desk, subject:
Beaver Valley Power Station, Units 1 and 2 Response to Generic Letter 88-11 4.
beaver Valley Final Safety Analysis Report 5.0 PRINCIPAL CONTRIBUTORS John Tsao, with contractual assistance from the Idaho National Engineering Laboratory.
Report completed in November,1989, l