ML20246F700
| ML20246F700 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/30/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246F698 | List: |
| References | |
| NUDOCS 8907130371 | |
| Download: ML20246F700 (7) | |
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1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 117 FACILITY OPERATING LICENSE NO. NPF-4 j
VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNIT NO. 1 DOCKET NO. 50-338
1.0 INTRODUCTION
In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement of Reuctor Vessel Materials and Its Impact on Plant Opera 1. ions," the Virginia Electric and Power Company (the licensee) requested changes to the pressure /
temperature (P/T) limits for the North Anna Power Station, Untt No.1 (NA-1)
Technical Specifications (TS). The request was documented in letters from the licensee dated November 30, 1988 and June 19, 1989. The purpose of the changes is to revise the P/T limits, which would be valid up to 10 effective full power years (EFPY). The proposed P/T limits were developed and based on the data from ectual surveillance capsules. The proposed revision provides up-to-date P/T lim!ts for operation of the reactor coolant system during heatup, cooldown, criticality, and itydrotest.
The licensee's letter dated June 19, 1989 provided supplemental information concerning the temperature difference between the water at the vessel sarface and the metal at the iT and 3/4T locations. This information did not change the staff's initial determination that the proposed amendment does not involve significant hazards considerations.
To evaluate the P/T limits, the staff used the following NRC regulations and guidance: Appendices G and H to 10 CFR Part 50; the ASTM Standards and ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
Regulatory Guide 1.99, Revision 2; Standard Review Plan (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 TS for the operation of the plant.
In particular, l
10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the TS. The P/T limits are among the limiting conditions of operation in the TS for all comercial nuclear plants in the U.S.
Appendices G and H to 10 CFR Part 50 oescribe specific requirements for fracture toughness and reactor vessel material surveillance which must be considered in setting P/T limits.
An acceptable method in constructing the P/T limits is described in SRP Section 5.3.2.
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, Appendix G to 10 CFR Part 50 specifies f racture toughness and testing requirements fer res;t:r vessel materials in accordance with the ASME Code and, in particular, to test the beltline materials in the surveillance capsules in accordance with Appendix H to 10 CFR Part 50. Appendix H, in turn, refers to the ASTM Standards.
These tests define the condition 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 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 Regulatory Guide (RG) 1.99, Revision 2 to predict the effect of neutron irradiation on reactor vessel materials. This RG defines the ART as the sum of the unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and margin to account for uncertainties in the prediction method.
Appendix H to 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 that are made f rom plate, weld, and h t at-affected-zone materials of the reactor beltline.
Based on the above changes, the licensee also proposed changes to several other TS to assure a proper low temperature overpressure protection (LTOP) at NA-1.
2.0 DISCUSSION 2.1 Pressure / Temperature Limits The staff has evaluated the effect of neutron irradiation embrittlement on each beltline material in the NA-1 reactor vessel. The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, Revision 2.
The staff has determined that the material with the highest ART (most erbrittled) at 10 EFPY was the circumferential weld between the intermediate and lower shell forgings.
The licensee has removed two surveillense capsules from the NA-1 reactor pressure vessel. The results from capsule V were published in a Babcock and Wilcox report BAW-1683. The results from capsule U were published in a Westinghouse report WCAP-11777. All surveillance capsules contained Charpy impact specimens and I
tensile specimens which were made from base netal and weld metal, and Charpy impact specimens from the heat-affected zone material.
For the limiting beltline material, the weld between the intermediate and lower shell forgings, the staff has calculated the amount of embrittlement per RG 1.99, 4
Revision 2.
The ART at 10 EFPY at AT was calculated to be 136*F.
The ART was calculated by applying the least squares extrapolation method in paragraph 2.1 j
of RG 1.99, Pevision 2 to hA-1 surveillance data.
l The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 136 F for the weld between the intermediate and lower forgings. The staff performed a similar calculation and verified the licensee's ART value to be acceptable I
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(see Table 1). Substituting the ART of 136*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, criticality, and hydrotest meet the beltline material requirements in Appendix G to 10 CFR Part 50.
TABLE 1 The NRC Staff Calculated Adjusted Reference Tempe'rature For The Limiting Reactor Beltline Material at horth Anna, Unit 1.
Limiting Beltline Material:
Intermediate-to-lower shell weld
. Code No.:
25531/1211/ SMIT 89 Copper Content:
0.086%
Nicke' Cc.?ar,t:
0.11%
Initial Reference Temperature:
19'F 2
Neutron Fluence n/cm at inside 1.39E19 surface at 10 EFPY:
136'F (Licensec Calculated 136'F)
In aedition to beltline materials, Appendix F to 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 pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions that are 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 -22*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section -IV.B of Appendix G requires the predicted Charpy upper shelf energy (USE) dt end-of-life to be above 50 ft-lb. At end-of-life, the 7.imitin (the At 5.9 EFPY, its measu$ and lowest Charpy USE materials with the highest adjusted RT is the lower shell forging.
Charpy USE is 93 ft-lb. Using the prediction method in RG 1.99, Revision 2, the Charpy USE at end-of-life will still be above 50 ft-lb.
2.2 Low Temperature Overpressure Frotection LTOP is provided by the. power operated relief valves (PORVs) on the pressurizer.
These PORVs are set at pressures low enough to prevent violation of the Appendix G heatup and cooldown curves should a reactor coolant system (RCS) pressure transient occur during low temperature operations. The licensee, in References 1 and 2, provides the results of analyses of the most limiting over-pressure transients in determining the PORV setpoints for LTOP.
A most limiting mass addition transient was analyzed assuming an inadvertent actuation of a charging pump. TS 3.1.2.4 and 3.5.3 allow only one charging
- pump to be operable when the RCS temperature is less than 324*F that is the
+ maximus RCS temperature during which the LTOP is required.
In response to the staff's request, the licensee provided in Reference 2 the results of a sensitiv-ity analysis.
It is indicated that an inadvertent initiation of an additional low head safety injection pump would not significantly increase the peak pres-sure during the mass addition transient. This is because the shutoff head of the low head safety injection pump will be reached in a short time and the contribution to mass addition of the low head pump will be negligible.
In this analysis, the licensee assumes initial RCS temperature of 100'F and 200'F each in combination with the PORV setpoints of 365 psia and 435 psia. The results indicate that the initial RCS temperature has a minor effect on the transient peak pressure. The heatup and cooldown P/T curves increase sharply 'at tempera-tures above 200*F. The slightly increased peak pressure due to the higher initial temperatures will be offset by the increased space between the PORV set-point, determined to prevent reaching tne Appendix G curve at a lower temperature and the Appendix G curve. Thus, the Appendix G curve allows higher pressure overshoot at temperatures above 200'F. Considering the above factors, the staff '
concludes that the assumptions applied to the licensee's analysis are reasonably conservative.
The licensee uses RETRAN 02/M0002 to perform its analysis in supporting the proposed TS changes.
Both RETRAN 01/M0003 and RETRAN 02/M0002 have been generically approved by the staff on Septenber 4,1984 (Reference 3). Also, the licensee's topical report on its plant-specific application of RETRAN 01/ MOD 03 was reviewed and approved by the staff on April 11,1985(Reference 4).
In addition, the licensee submitted comparisons between RETRAN 01/M0003 and RETRAN 02/ MOD 02 for a series of plant transients (Reference 5). This information demonstrated that the RETRAN 01/M0003 and RETRAN 02/M0002 code results are nearly identical for the NA-1&2 plant-specific models except for the changes caused by the nonequilibrium pressurizer model in RETRAN 02/M0002.
However, the LTOP transients deal with RCS in water solid conditions, thus, it is not affected by the nonequilibrium pressurizer model in RETRAN 02/M0002. While the licensee's Application on the use of RETRAN 02/M0002 has not been reviewed and approved by the staff, the staff finds reasonable assurance exists that the results of the licensee's analysis using RETRAN 02/ MOD 02 supports the proposed TS charges on the LTOP.
However, if any concerns should be raised during the staff't. review
.of the licensee's RETRAN 02/M0002 application, the staff will reevaluate the licensee's evaluation on the subject TS changes.
The modified Appendix G curve temperature corresponding to the pressurizer safety valve setpoint of 2485 psig is 324'F. This point is used to bound all of the low temperature transient analyses.
Below 324*F, the anticipated low tegera-ture overpressurization transients may be adequately mitigated by the automatic action of the pressurizer PORVs or by allowing sufficient time for operator response.
Automatic LTOP is required whenever any RCS cold leg temperature is less than 261'F. This LTOP temperature is determined in accordance with the requirements set forth in Section B.2 of the Branch Technical Position RSB 5-2.
Above 261.'F administrative control will provide adequate protection because of the Appendix G fracture criterion.
The analysis has an increased margin at higher temperatures.
In addition, operation of the RCS above 261*F decreases the effects of the two design basis transients.
Based on the results of the most limiting LTOP transient, the licensee's proposed TS PORV setpoints are less
.4, than or equal to 450 psig when the RCS temperature is less than or equal to 261*F and less than or equal to 390 psig when the RCS temperature is less than or equal to 150*F.
The licensee proposed changes to TS 3.1.2.2, 3.1.2.4, 3.4.1.3, 3.4.9.3, 3.5.2, and 3.5.3 and their associated bases sections reflect the above discussed LTOP alignment temperatures and the heatup and cooldown re.tes identified by the updated TS Figures 3.4-2 and 3.4-3.
The staff finds.that they are reasonably conservative and acceptable.
3.0 EVALUATION Based on the discussion above, the staff concludes that the proposed NA-1 P/T limits on the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 10 EFPY, because the limits conform to requirements of Appendices G and H to 10 CFR Part 50.
The licensee's submittal also satisfies Generic Letter 88-11, because it used the method in RG 1.99, Revision 2 to calculate the ART. Hence, the proposed P/T limits may be incorporated into the NA-TS.
In addition, the ;taff concludes that the licensee's proposed changes to TS 3.1.2.2, 3.1.2.4, 3.4.1.3, 3.4.9.3, 3.5.2, and 3.5.3 and their associated bases sections are acceptable to support the updated P/T limits applicable for a period up to 10 EFPY.
Therefore, the staff finds the proposed changes to be acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to surveillance requirements. The staff has determined that the amend-ments involve 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 significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that these amendments involve no significant hazards consideratiore and there has been no public coment on such finding. Accordingly, these amendments meet the eligibility criteria for categorical exclusion set fortn in 10 CFR 951.22(c)(9).
Pursuant to 10 CFR 551.22(b), no environmental fmpact statement or environmental dssessment need be prepared in connection with the issuance of these amendments.
5.0 CONCLUSION
We have 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, and (2) such activities will be conducted in compliance with the Comission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: June 30, 1989 Principal Contributors:
C. Liang B. Elliot
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6.0 REFERENCES
1.
Letter from W. Cartwright of Virginia Electric and Power Company to USNRC, " Proposed Technical Specifications Change," dated November 30, 1988.
2.
Letter from W. Stewart of Virginia Electric and Power Company to USNRC,
" Proposed Technical Sper*fications Change - Supplement," dated June 19, 1989.
3.
Letter from C. Thomas of USNRC to T Schnatz of UGRA, " Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5 and EPRI NP-1850-CCM "
cated September 1984.
4.
Letter from C. Thomas of USNRC to W. Stewart of. Virginia Electric and Power Company, " Acceptance for Referring of Licensing Topical Report VEP-FRD-41,~ ~
VEPC0 Reactor System Transient Analysis Using the RETRAN Computer Code,"
dated April 11, 1985.
5.
Letter from W. Stewart of VEPC0 to USNRC, "Surry and North Anna Power Stations Reactor System Transient Analyses," dated November 1985, e
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June 30. 1989 DATED:
June 30, 1989 AMEN 0 MENT N0.117 TO FACILITY OPERATING LICENSE NO. NPF-4-NORTH ANNA C{$$&^'LocalPDRs hW?
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