ML20059N408
| ML20059N408 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/03/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059N407 | List: |
| References | |
| GL-88-11, NUDOCS 9010120203 | |
| Download: ML20059N408 (3) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PRESSURE-TEMPERATURE LIMTS RELATED TO GENERIC LETTER 88-11 NORTHEAST UTILITIES MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-336-
1.0 INTRODUCTION
In response to Generic letter 88-11,-"NRC Position on Radiation Embrittlement-of Reactor. Vessel Materials and Its Effect on Plant Operations," the Northeast Utilities (the licensee) presented the current pressure / temperature (P/T) limits in the Millstone 2 Technical. Specifications, Section 3.4.- The response was documented in a letter: from the licensee dated-Noverrber 1,:1988, with-two-supplements dated April 6, 1990 and July 18,-1990.
The P/T limits are valid for 12 effective full power years (EFPY). 'The. current P/T' limits were>
developed using Regulatory Guide (RG) 1.99, Revision 2 (draft).
Generic-Letter 88 11 recommends RG 1.99, Rev. 2,' be used in calculating P/T limits, unless the use of different method; can be justified.
The P/T= limits provide 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 H;-10 CFR'50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plan (SRP) Section 5.3.2;:and Generic-Letter 88-11.
Each licensee authorized to operate a nuclear power reactorfis required by 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
. operation be included in the Technical Specifications. The P/T limits are among 'the limiting: conditions of operation in the Technical Specifications for all; commercial nuclear plants in the U.S.
Appendices:G and:H of -10 CFR Part 50 describe specific requirements-for fracture toughness'andcreactor vessel material s;urveillance 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,1in turn, refers to ASTM Standards. These tests define the extent of ~ vessel t
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embrittlement at the time of capsule withdrawal in terms.of the increase in-reference: temperature. Appendix G also requires the licensee.to predict the effects of. neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor
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vessel materials. -This guide defines the ART as the sum of unirradiated.
reference temperature, the increase in referer.ce terperature resulting from neutron irradiation, and a margin to account for: uncertainties in the prediction method.
l Appendix H.of 10 CFR Part 50 requires the licensee to establish a surveillance l
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program to periodically withdraw surveillance capsules from the reactor vessel.: Appendix H refers to the ASTM Standards which, i.. 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) materials-of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline r.aterial -in the Millstone 2 reactor vessels.
The amount of irradiation embrittlement was calculated in accordance with RG.1.99, Rev.-2.
The: staff has determined that the material with the highest ART at 12 EFPY was the' intermediate shell plate C-505-2 with 0.13% copper (Cu), 0.64% nickel (Ni), and.an initial RT f 259.
ndt The licensee has removed one surveillance capsule from Millstone 2.
The results from capsule W-97 were published in Combustion-Engineering Report TR-N-MCM-008.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material,- intermediate shell plate C-505-2, the staff calculated the ART to be 149.7'F at 1/4T (T = reactor vessel beltline thickness)andj23.5'Ffor3/4 Tat 12-EFP of 9.53E18 n/cm at 1/4T and 3.38E18 n/cm,Y.The staff used a neutron fluence at 3/4T.
The ART was determined.by-Section 1 of RG 1.99, Rev. 2, because only one surveillance capsule has been removed from Millstone 2.
The licensee used the method in RG 1.99, Rev. 2 (draft), to calculate an ART 4
of 152'F at 12 EFPY at 1/4T for the same limiting plate. The staff judges that the licensee's ART of 152*F is more conservative than the staff's ART of 149.7'F, and it. is acceptable.
Substituting the. ART of 152*F into equations in SRP 5.3.2, the staff verified that the current P/T limits for heatup, J
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 i
temperature of the closure flange regions highly stressed by the bolt preload i
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least 120'F for normal operation and by 90'F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 10'F, the staff has determined that the current P/T limits satisfy 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 J
life be above 50 ft-1b. The beltline u.aterial with the lowest predicted USE l
at EOL is intermediate shell plate C 5051 with an initial USE of 75.8 ft-lb.
Using Figure 2 of RG 1.99, Rev. 2, the staff calculated that the EOL USE is i
51.5 ft-lb.
This is greater than 50 ft-1b and, therefore, is acceptable, j
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3.0 CONCLUSION
The staff concludes that the current P/T limits for the reactor coulant system i
for heatus, cooldown, leak test, and criticality are valid through 12: EFPY because tie limits conform to the requirements of Appendices G and H of 10 i
CFR Part 50.
The licensee's submittal also satisfies Generic Letter 8811 :
because the adjusted RT conservative than the RY, used in the current P/T limits are more att i
calculated by RG 1.99, Rev. 2.
Hence, the current F/T limits may be kept iIthe Millstone 2 Technical Specifications, c.
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4.0 REFERENCES
' :r 1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 l
2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature Limitt
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11 G,
3.
November 1,128, Letter from :. J. Mroczka (NUCo) to USNRC Document i
i Control Desk,
Subject:
Hadd ni Weck Plant. I'111st e Nuclear Power Station, Units 1, 2, and 3: Response tr O neric Lettei' 88-11 4.
April 6,1990, Letter from E. J. Mroczka (NUCo) to USNF.C Document CoMrol Desk,
Subject:
Response to Generic Letter 88 11 w
5.
July 18, 1990, Letter from E. J. Mroczka (NUCo) to USNRC Document Control Desk,
Subject:
Request for Additional to Generic Lettere88-ll 6.
S. T. Byrne, " Post-irradiation Evaluation of Reactor Vessel Surveillance e
Capsule W 97," TR N MCM 008, Combustion-Engineering, April 1982 l
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Principal Contributor: J. 'sao
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Dated: October 3, 1990 l
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