ML17262A764
| ML17262A764 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/06/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17262A763 | List: |
| References | |
| NUDOCS 9203120294 | |
| Download: ML17262A764 (7) | |
Text
~gq RE00 (6
P0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 SAFETY EVALUATIOtl BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED At)ENDMENT N0.48 TO FACILITY OPERATING LICENSE NO. DPR-18 ROCIIESTER GAS AND ELECTRIC CORPORATION R. E.
GIHHA NUCLEAP, POWER PLANT DOCKET NO. 50-244 INTRODUCTION
1.0 INTRODUCTION
By letter dated February 15, 1991, as supplemented on May 14, 1991, the Rochester Gas and Electr'fc Corporation (the licensee) requested an amendment to Facility Operating I icense No. DPR-18 for the R. E. Ginna Nuclear Power Plant.
The proposed amendment would revise the pressure/temperature (P/T) limits in the Ginna Technical Specifications (TS), Section 3.1.
The revised P/T liIIIits have then re-evaluated the low temperature overpressurization protection system (LTGPS) set point and its associated basis in the Ginna Technical Specifications, Section 3.1.2, 3.3.1, and 3.15.1.
1.1 TS P
T Limits The proposed P/T limits are valid for 21 effective full power years (EFPY) and were developed using Regulatory Guide (RG) 1.99, Revision 2, "Radiation
,Embrittlement of Reactor Vessel Material."
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations,"
recommends RG 1.99, Rev. 2, be used in calculating 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; and 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 (SRP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications (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 TS.
The P/T limits are among the limiting conditions of operation in the TS for all commercial nuclear plants in the U.S.
Appendices 6 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.
9203120294 920306 PDR ADOCK 05000244 P
Appendix 0 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 emhrittlement at the time of capsule withdrawal in terms of the increase in reference temperature (RT
).
Appendix 6 also requires the licensee to predict the effects of neutron irr36ation 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, Revision 2, to predict the effect of neutron irradiation or reactor vessel materials.
RG 1.99, Revision 2 defines the ART as 1) the sum of unirradiated reference temperature;
- 2) the increase in reference temperature resulting from neutron irradiation, and 3) a margin to account for uncer tainties 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, fn 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 beltlinc.
1.2
~ETD The Reactor Coolant System (RCS) Pressure-Temperature (P/T) limits during plant heatup and cooldown for R. E. Ginna Nuclear Power Plant are specified in Technical Specification (TS) Figures 3.1-1 and 3.1-2.
The requirements for low temperature overpressure protection (LTOP) system are specified in TS 3.15.
The restriction of the safety injection pump operability is specified in TS 3.3.1 in consistence with the assumption used in the analysis supporting LTOP setpoint.
By letter, dated February 15, 1991 (Ref. 1), Rochester Gas and Electric Corporation, in response to GL 88-11, proposed amendments to the P/T curves in TS Figures 3.1-1 (for heatup) and 3.1-2 (for cooldown),
changes in the LTOP system setpoint and restriction of the safety injection pump operabi lity specified in TS 3.3.1'and 3.15.1.
The licensee also provided the results of a safety analysis to support its proposed TS changes.
2.0 EVALUATION P.l TS P
T Limits The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Ginna reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Revision 2.
The staff has determined that the material with the highest ART at 21 EFPY was the circumferential weld between the intermediate and lower shells (SA-847/MR-19) with 0.25% copper (Cu), 0.55% nickel (Ni), and an initial RTndt of O'.
The licensee has removed three surveillance capsules from Ginna.
The results from capsules V, R, and T were published in Westinghouse reports FR-RA-1, WCAP-8421, and WCAP-10086, respectively.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld octal, and HAZ metal.
For the limiting beltline material, weld SA-847/WR-19, the staff calculated the ART to be 210'F at 1/4T (T = reactor vessel beltline thickness) and 178.)'F for 3/4T at 21 EFPY.
The staff used a neutron fluence of 1.57E19 n/cm at 1/4T and 7.20E18 n/cm at 3/4T.
The ART was determined by the least squares extrapolation method using the Ginna surveillance data.
The least squares method is described in Section 2.1 of RG 1.99, Revision 2.
The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 210'F at 21 EFPY at 1/4T for the same l,imiting weld metal.
Substituting the ART of 210'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 RT for the reactor vessel closure flange materials.
Section IV.A.2 of Appeal)x G states that when the pressure exceeds
'20K of the preserv'.ce system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the RT << of the material in those regions by at least 120'F for normal operation and 5$'0'F for hydrostatic pressure tests and leak tests.
Based on the flange RT of 60'F, the staff has determined that the proposed P/T limits satisfy SecttQ IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb.
The staff has identified weld SA-847/WR-19, a Linde 80 weld, as the limiting material which has the lowest USE of all reactor vessel beltline materials.
Based on data from surveillance capsule T and on RG 1.99, Revision 2, the staff has calculated USE at current EFPY (about 15.2) to be 54.2 ft-lb.
The USE at 21 EFPY is calculated to be 52.2 ft-lb, and the End-of-Life (EOL)
USE is expected to be 49.2 ft-lb.
The licensee has also predicted that the USE of the limiting material at EOL to be below 50 ft-lb as shown in the Babcock 5 Wilcox (BSW) report, BAW-1803.
The licensee is a
member of the B
8 W Owners Group (BSWOG) established to study and resolve the low USE issue.
The staff is following the progress of the B&WOG's study.
2.2
~ET 0 I In the current TS, when the RCS cold leg temperature is less than or equal to 330'F; LTOP protection is provided by either PORVs with a lift setting of less than or equal to 435 psig, or a RCS vent of greater than or equal to I.l square inches.
The operability of two PORVs or an RCS vent of greater than or equal to 1.1 square inches ensures that the RCS will be protected from pressure
4 transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS cold leg temperature is less than or equal to 330'F.
Each PORV has adequate relieving capability to protect-the RCS against P/T limits when the transient is limited to either (1) the start of an idle RCS with the secondary water temperature of the steam generator less than or equal to 50'F above the RCS cold leg temperature or (2) the inadvertent actuation of a safety injection pump and its injection into a water solid RCS.
Consistent with the above assumption, the current TS 3.3.1 permits one operable safety injection pump when the RCS cold leg temperature is less than or equal to 330'F.
The licensee indicated in its letter, dated February 15, 1991, there is no safety relief valve at the suction side of the RHR system to protect the RHR system from potential overpressurization, thus the LTOP system also protects the RHR system from overpressurization when the RHR system is connected to the RCS.
The licensee also determined that the allowable peak RCS pressure is more limiting for the RHR system protection than that for the protection against the limits of Appendix G to 10 CFR Part 50.
The licensee, in its letter, dated February 15, 1991 (Reference 3), proposed a
change of the PORV setpoint from 435 psig to 424 psig.
Mith this reduction of PORV setpoint, the licensee stated that the required LTOP system specified in Technical Specification 3.15.1 is adequate for protection against the limits of Appendix G to 10 CFR Part 50.
The design basis transients of either energy addition or mass addition as described above apply.
However, the licensee has determined that in order to protect the RHR system from overpressurization by the PORV the design basis mass addition transient needs to be changed to a charging/letdown mismatch with three charging pumps in operation.
Consistent with this change, the licensee has proposed Technical Specification 3.3.1.8 to require that all three safety injection pumps shall be inoperable and safety injection discharge flow paths to the RCS isolated whenever overpressure protection is provided by the pressurizer PORV.
- Also, the licensee proposed that Technical Specification 3.3.1 allows no more than one safety injection pump to be operable when the overpressure protection is provided by a RCS vent of greater than or equal to 1.1 square inch.
This is because the results of the licensee's recent analysis indicated that the mass addition from the inadvertent operation of a safety injection pump will not result in RHR system pressure exceeding allowable limits when overpressure protection fs being provided by a RCS vent of greater than or equal to 1.1 square inch.
The staff has evaluated the licensee's proposed Technical Specification changes regarding LTOP setpoint and the analysis supporting these proposed changes.
lIe have concluded that the proposed Technical Specifications are consistent with the assumptions used in their supporting analysis and therefore are acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to installation cr 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 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 (56 FR 33960).
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.
5.0 CONCLUSION
Based on the staff evaluation in Section 2.0 above, the staff concludes that the 1'.censee's proposed Technical Specification
- changes, Sections 3.1.2, 3.3.1.7, 3.3.1.8, and 3.15.1, are acceptable.
The staff concludes that the proposed P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 21 EFPY because the hmits conform to the requirements of Appendices G and H of 10 CFR Part 50.
The proposed P/T limits also satisfy Generic Letter 88-11 because the method in RG 1.99, Rev.
2 was used to calculate the ART.
- Hence, the proposed P/T limits may be incorporated into the Ginna Technical Specifications.
The staff 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, and (2) such activities will be conducted in compliance with'he Commission's regulations, and issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.C REFERENCE~
1.
Regulatory Guide 1.99, Radiation Embrfttlement of Reactor Vessel Naterials, Revision 2, tray 1988.
2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressur e-Temperature.
3.
Letter from R.
C. Necredy, Rochester Gas and Electric Corporation, to USNRC, "R. E. Ginna Nuclear Power Plant, Modifications to Low Temperature Overpressure Protection Systems,"
February 15, 1991.
4.
Niay 14 1991, Letter from G. J. Wrobel (RGIIE) to A. R. Johnson (USNPCI,
Subject:
Addftfonal Information on Gfnna.
5.
T. R.
Y<ager et al., "Analysis of Capsule V from the Rochester Gas and Electric R. E. Ginna Unit No.
1 Reactor Vessel Radiation Surveillance Program,"
Westinghouse Nuclear Energy Systems, FP-RA-1, April 1, 1973.
6.
S.
E. Yanichko et al., "Analysis of Capsule R from the Rochester Gas and Electric R. E. Ginna Unit Ho.
1 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corporation, MCAP-8421, November 1974.
7.
S.
E. Yanichko et al., "Analysis of Capsule T from the Rochester Gas and Electric R. E. Ginna Unit No.
1 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corporation, WCAP-10086, April 1982.
Principal Contributors:
John Tsao C. Liang Dated:
t1arch 6, 1992
tx 4
'I