ML20059L280
| ML20059L280 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/27/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059L278 | List: |
| References | |
| NUDOCS 9402030342 | |
| Download: ML20059L280 (5) | |
Text
, [ p " %,[ j UNITED STATES
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NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D.C. 20555-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.170 TO FACILITY OPERATING LICENSE NO. DDR-6]
NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY MJLLSTONE NUCLEAR POWER STATION. UNIT NO. 2 DOCKET NO. 50-336 1.0 IflTRODUCTION By letter dated June 11, 1993, supplemented by letter dated November 15, 1993, Northeast Nuclear Energy Company (NNEC0/the licensee) submitted proposed Technical Specification (TS) changes to Facility Operating License No. DPR-65 for the Millstone Nuclear Power Station, Unit 2 (Millstone 2).
The proposed amendment revises the pressure / temperature (P/T) limits for the reactor vessel.
Specifically, Figure 3.4-2, " Millstone Unit 2 Reactor Coolant System Presure-Temperature Limitations for 12 F ll Power Years," on page 3/4 4-19, is revised to reflect the change in the curves and the title change to " Millstone Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20 EFPY"
[ effective full power years). The P-T limit curves are revised to reflect the increase in the nil-ductility reference temperatre of reactor vessel beltline materials.
The November 15, 1993, submittal provided information that did not change the initial no significant hazards consideration determination.
2.0 BACKGROUND
The licensee's arguments were based in part on the results of surveillance capsule W-104, the report was submitted to the staff on November 27, 1991 (Ref. 2).
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 American Society for Testing and Materials (ASTM) Standards and the American Society of Mechanical Engineers (ASME) Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2); Regulatory Guide (RG) 1.99, Rev. 2; Standard Review Plan (SRP)
Section 5.3.2; and Generic Letter 88-11.
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Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide TS 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 for operation in the TS for all commercial nuclear plants in the United States. Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness 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 withdrawal in terms of the increase in reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vesse1 ~ embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees use the methods in RG 1.99, Rev.
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 i
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 monitor embrittlement of reactor vessel materials. Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and be removed from the reactor vessel periodically for testing. The capsules should contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
3.0 EVALUATION FOR FAST NEUTRON FLUENCE FOR 20 EFPY Millstone Unit 2 has shifted to a low leakage core loading pattern in Cycle
- 10. Thenewloadingpatterncontainsonenewassemblyjntheperimeter,which 0
shifted the azimuthal peak fluence location from the 27 'to 1 on the quadrant. The azimuthal shift resulted in a lower estimated peak value at the new peak location. Surveillance capsule W-104 verified the calculated estimate.
The analytical method was based on the two dimensional DOT 4.3 finite differencing code. The (r,6) solution used a 47 group cross section set based on ENDF/B-IV. The cross section scattering approximation was P and the 3
geometrical quadrature was S. The source and the source spectrum were based-on a combination of U-235 and Pu-239 spectra to account for the use of 4
depleted assemblies in the periphery. A pin-by-pin. power distribution was used for the outer assemblies.
The pin power distribution was derived from the assembly average power at Millstone and the pin power shape distribution
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from St. Lucie Unit 2, which is a sister plant to Millstone and has had identical loading patterns to Cycle 10. The analytical method as described above is acce; table.
The licensee stated that up to Cycle 15 the Cycle 10 loading will be used.
However, beyond Cycle 15, the licensee stated that it will either continue.the Cycle 10 loading pattern or chang'e to an all burned assembly scheme in the periphery, which will further reduce the fluence to the vessel and it will be conservative.
From the preceding discussion we conclude that:
(1) the low leakage pattern introduced in Cycle 10 will reduce the peak vessel irradiation, (2) the analytical method used for the estimation of the fluence to 20 effective full power years of operation is acceptable because it complies with staff requirements and because it is conservative, and (3) the Cycle 10 loading pattern will be used to.15 effective full power years of operation and possibly to 20 effective full power years.
If, however, it is changed after 15 effective full power years it will become more conservative.
From the above we conclude that the Millstone 2 fluence estimate to the pressure vessel l
for 20 effective full power years of operation is acceptable.
3.1 Evaluatior for P/T limits l
The staff calculated the ART for each beltline material in the Millstone 2 reactor vessel in accordance with RG 1.99, Rev. 2.
The staff determined that the material with the highest (limiting) ART at 20 EFPY was the intermediate shell plate C-505-2 (heat number C5843-2) having 0.13% copper (Cu), 0.64%
I nickel (Ni), and an initial RT, of 25'F.
The staff calculated the ART to be 143*F at the 1/4T location (T - reactor l
vessel beltline thickness) and J13*F for the 3/4T location. The staff used a 2
neutron fluence of 9.32E18 n/cm at 1/4T and 2.79E18 n/cm at 3/4T. The staff j
used the method in Regulatory Position C.2, Surveillance Data Available, in RG 1.99, because the surveillance data are credible.
For the same limiting material, the licensee calculated an ART of 145'F and i
ll8'F at the 1/4T and 3/4 locations, respectively, using Regulatory Position l
C.1, Surveillance Data Not Available, in RG 1.99.
RG 1.99 allows use of either methods to calculate the ARTS.
RG 1.99 states that if Position C.2 gives a higher value of ARTS than that given by using Position C.1, the ARTS from Position-C.2 should be used.
If Position C.2 gives lower ARTS, either may be used. The licensee's ARTS are more a
conservative than the staff's ARTS and, therefore, acceptable to use.
j Substituting the ARTS of 145'F and ll8'F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and nydrotest meet the requirements in Paragraphs IV. A.2 and IV. A.3 of Appendix G of 10 CFR Part 50.
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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.A.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 exceed the reference temperature of the material in those regions by at 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 proposed P/T limits satisfy Section IV.A.2 Of Appendix G.
Paragraph IV.A.1 of Appendix G requires that reactor vessel beltline materials maintain a Charpy upper shelf energy (USE) throughout the life of the vessel of no less than 50 ft-lbs unless it can be demonstrated that lower values of USE will provide margins of safety against fracture equivalent to those I
required by Appendix G of the ASME Code. The conformance of upper shelf energy to paragraph IV.A.1 will be determined pending the staff evaluation of the licensee's response to Generic Letter 92-01.
The licensee has removed surveillance capsules W-97 and W-104 from Millstone 2 and has performed required tests. The test results from capsules W-97 and W-104 were published in reports by Combustion %gineering and Babcock &
Wilcox, respectively (Ref. I and 2). The co' ormance of the surveillance program to Appendix H to 10 CFR Part 50 also will be determined under the staff evaluation of the licensee's response to Generic Letter 92-01.
The staff has performed an independent analysis of the P/T limits to verify the licensee's proposed limits. The staff concludes that the proposed P/T limits for heatup, cooldown, hydrostatic tests, and criticality are valid through 20 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50 and Generic Letter 88-11. Hence, the proposed P/T limits may be incorporated in the. Millstone 2 Technical Specifications.
The conformance of the upper shelf energy and reactor vessel material surveillance program to Appendices G and H will be determined pending the staff resolution of Generic Letter 92-01 in 1994.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Connecticut. State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATIO_B t
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no i
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. significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (58 FR 39054). Accordingly, the amendment meets the eligibility criteria 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 need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the. health and safety of the public.
7.0 REFERENCES
1.
S. T. Byrne, " Post-Irradiation Evaluation of Reactor Vessel Surveillance Capsule W-97," TR-N-MCM-008, Combustion-Engineering, April 1982 2.
Letter from J. F. Opeka of Northeast Utilities to USNRC, dated November 27, 1991
" Analysis of Capsule W-104 Northeast Nuclear Energy Company Millstone Nuclear Power Station, Unit No.
2," B&W Nuclear Service Company, BAW-2142, November 1991 l
Principal Contributors:
J. Tsao L. Lois Date:
January 27, 1994
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