ML19325E348
| ML19325E348 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/25/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19325E347 | List: |
| References | |
| NUDOCS 8911060301 | |
| Download: ML19325E348 (5) | |
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M UNITEO STATES NUCLEAR REGULATORY COMMIS860N q
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- IdFETY EVALUATION BY THE OFFICE OF NUCLEAR PFACTOR REGULATION AMENDMENT NO. In TO FACILITY OPERATING LICENSE NO. DPR-44 PHILADELPHIA ELECTRIC COMPANY PUBlit' SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. ?
00rVET NO. 50 ??7 1.0 INTPODUCTION By letter dated May 15, 1989, Philadelphia Elec+ric Company requested an amendment to Facility Operating License No. DPR-44 for the Peach Bottom Atomic Power Sta+ ion, Unit No, 7.
The proposed amendment modifies the pressure-temperature limits for the reactor vessel.
In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Ya'erials and I+s Effect on Plant Operations," the Philadelphia Electric Comnany(P/T) limits in +he Peach (the licensee) requested permission to revise the pressure / temperature Bottom atomic Power Station, l' nit 2 (hereinafter, Peach Bottom 2)
Technical Specifications, Section 3/4.6.
The purpose of the revision is to change the e#fectiveness o' the P/T limits for 32 effective 'ull power years (EFPY).
The proposed P/T limits were based on Regulatory Guide (RG) 1.99, Pevision 2.
The proposed revision provides up-to-da+e P/T limits for +he operation of the reactor coolant system during hea+up, cool-down, 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 Revision 2; Standard Review Plan (SDP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is reovired 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 lirriting 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 o' 10 CFR Part 50 describe specific requirements for fracture toughness and reac+0r vessel material surveil 16nce that must be considered t
l in settino P/T limits. An acceptable me+ hod for constructing the P/T limits is described in SRP Section 5.3.2.
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2 Appendix C of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASM{ Code and, in particular, that the beltline materiais in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50.
Apoendix H, in turn, refers to ASTM Standards.
These tests define the extent of vessel entrittlement at the time of capsule withdrawal in terms of the increase in reference temperature. Appendi) G also requires the licensee to predict the effects of neutron irradiation on vessel Charpy upper shelf energy (g the adjusted reference temperature (ART) and embrittlement by calculatin USE). Generic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation on reactor vessel materials.
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 licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel. Appendix H refers to the ASTN Standards which, in turn, require that the capsules be installed in the vessel before startup anc that they contain test specimens made from plate weld, and heat-affected-2one(HAZ)materialsofthereactorbeltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each belt-line material in the Peach Bottom 2 reactor vessel. The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, Rey, 2.
The material with the highest ART at 32 EFPY was the lower-intermediate shell plate C2873 1 with 0.12% copper (Cu) and 0.57%
3 nickel (Ni), and an initial RT of -6 F.
The licensee has removed one surveillance capsule from Peach Bottom 2.
1he results from that surveillance capsule were published in General Electric Report SASR 88-24, DRF B13-01440, which is an attachment to a letter from J. W. Gallagher to T. E. Murley dated Nay 23, 1988. The surveillance capsule contained Charpy impact specimens and tensile specimens which were made from base metal, weld metal, and HAZ metal.
For the limiting beltline material, C2873-1 the staff calculated the ART to be 51*F at 1/4T (T = reactor vessel beltline thickness) for 32 EFPY.
The staff calculated the ART by the method described in Section 1 of RG 1.99, Rev. 2 because only one surveillance capsule had been withdrawn from the Peach Bottom 2 reactor vessel.
Thejicenseecalculatedthesame 51*F for the ART. Substituting the ART of 51 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 requirenents in Appendix G of 10 CFR Part 50.
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In addition to beltline materials, Appendir G of 10 CFR Part 50 also j
imposes P/T limits based on +be re'erence temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when +he pressure exceeds 70% of the preservice system hydrostatic test pressure, the +emperature of the closure flance regions highly stressed by the bolt preload must exceed ghe reference temperature of the gaterial in those regions by at leas + !?O F for normal operation and by 90 F for hydrostatic pressure tests and leak tests. Fioure 3.6.2 of the proposed Technical Specifica+1ons for Peach Bottom 2 shows that the tempera +ure 0
for hea+up or cooldown following nuclear shutdown is approximately 165 F at 300 psig.
Figure 3.6.1 of the proposed Technical Specifications for Pearh Rottom 2 shows that the minimum temperatgre for pressure tests required by Section XI c' the ASME Codg is 100 F at 312 psig.
Based on the flange reference temperature of 10 F, the stef# has determined that the proposed P/T limits satisfy Section IV.? of Appendix G for normal creration, hydrostatic pressure and leak tests.
Section ]V.B of Appendix G requires that the predicted Charpy USE at end of life (EOL) be above 50 ft-Ib.
The initial USE for the limiting beltlire material, the lower-intermediate shell plate metal (C2873-1), was not supplied.
However, the calculated USE 'or the surveillance base metal (C2761-2) at EOL is 04.5 ft-lb, which is higher than the Appendix G EOL USE requirement.
The strveillance base metal (C2761.2) was produced by the same manufacturer to the same ASTM specification as the limiting beltline material, and has copper and nickel contents that are very close to those of the limiting beltline material (0 11% Cu and 0.54% Ni for CP761-2 vs 0.1?T Cu ane 0.57% Ni for CP873-1).
Pased on this comparisor, the staff believes that the E0L USE of the limiting beltline material (C?873-1) meets the Appendix G 50 ft-lb requirement.
The s+aff concludes + bat the proposed p/T limits for the reactor coolent system 'or heatup, cooldown, leak test, and criticality are valid through 32 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
The licensee's submittal also satisfies Eeneric Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2, to calcula+e the APT.
Hence, the orooosed P/T limits may be incorporated into the Peach Bottom ? Technical Specifications.
The licensee also proposed certain edministrative chances to the Technical Specification pages involved with +he changes discussed ebove.
These changes include the deletion of Figure 3.6.4 which provides in'ormation on estimated transition temperature shi't relative to #1uence; rewording T.S. 3.6.A.3 to more accurately describe the vessel materials and appur'enances involved; revision of the " neutron flux specimen" terminolony in T.S. 4.6.A.2 to " surveillance specimen"; revisions to T.S.
page 144 te reflect removal and testing o# a surveillar.ce capsule and l
deletion of Figure 3.6.4; related changes in the List of Figures; and l
minor forma + end typographical chances on page 143 and 144.
Pelated changes to the T.S. Pases are also proposed.
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' The staff finds that these proposed changes reflect the results of material analyses conducted as part of the reactor coolant pressure boundary material surveillance program.
These changes are consistent with the proposed changes to the reactor vessel pressure-tempers +ure limits and are thus acceptable.
The staff also finds the proposed addition of Figure 3.6.5 to the Lis+ of Figures properly re'1ects its addition in a previously approved license amendment, and is thus acceptable.
.0 ENVIRONMENTAL CONS 1pEPATIONS This amendment involves a change *o a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part P0 and changes to the turveillance requiremen+s.
The staff has determirtd that the amendment irvolves no signi'icant increase in the amounts, and no significant change in the types, of ery effluents that may be released of fsite and that there is no significart increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that this amendment involves no sienificant hazards considera+1on and +here has been no public coment on such finding. Accordingly, this amendmen+ r'eets the eligibility criteria for categorica' exclusion set forth in 10 CFR St.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact s+atement nor environmental assessment need be prepared in connec+1on with the issuarce of this amendment.
4.0 CONCLUSION
The Commission nede a proposed determination that +be amendment involves no significant hazards consideration which was published in the Federal Pegister (54 FR 31116) on July 26, 1989 and consulted with the Comonwealth of Pennsylvania.
No public c3mments were received and the Comenwealth of Pennsylvania did not have any corments.
The staf# has concluded, based on +he considerations discussed above, tha+:
(1) there is reasonable assurance that the heal +h and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be ccnducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety n' +he public.
Principal Contributor:
J. Tsao Dated:
October 25, 1989
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TABLE 1 The NRC Staff Calculated Adjusted Reference Temperature for the Liiniting Reactor Beltlir.e Paterial at Peach Bottom Atomic Power Station, Unit 2.
Limiting Beltline Material:
Lower-intermediate shell plate Code No.:
C2873-1 Copper Content:
0.12%
Nickel Content:
0.57%
0 Initial Reference Temperature:
-6 F Reactor Vessel Celtline Thickt ess (in.)
6.31 Reactor Vessel Beltline Inside Padius (in.)
125.5 i
Chemistry Factor (CF) Used in Calculation 82.4 l
2 Neutron Fluence n/cm at 32 EFPY:
At I.D.
1.0E18 At 1/4T 0.69E18 AT 3/4T 0.33E18 Fluence factor At I.D.
0.417 i
At 1/4T 0.347 At 3/4T 0.231 i
0 I
Margin 28.5 F 0
0 ART at 1/4T at 32 EFPY:
51 F (Licensee calculated G1 F) 0 ART et 3/4T at 32 EFPY 32 F (Licensee did not provide an ART for 3/4T in the GE report, SA 88-24) i 5
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