ML20244D440
| ML20244D440 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/11/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20244D429 | List: |
| References | |
| NUDOCS 8904210341 | |
| Download: ML20244D440 (7) | |
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k UNITED STATES j
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NUCLEAR REGULATORY COMMISSION y,
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. NPF-10 SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA 1
THE CITY OF ANAHEIM, CALIFORNIA 1
SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 DOCKET NO. 50-361
1.0 INTRODUCTION
By letter dated November 7,1988 Southern California Edison Company (SCE),
et al. (the licensees) requested a change to the Technical Specifications l
for Facility Operating License No. NFF-10 that authorizes operation of San Onofre Nuclear Generating Station, Unit 2 in San Diego County, California. The amendment would revise Technical Specifications 3/4.4.8.1,
" Pressure / Temperature Limits;" 3.4.1.4.1, " Cold Shutdown-Loops Filled;"
3.4.1.3, " Hot Shutdown;" 3.4.8.3.1, " Overpressure Protection System, RCS Temperature less than or equal to 235"F;" and 3.4.8.3.2, " Overpressure Protection System, RCS Temperature greater than 235"F."
These changes would revise the pressure / temperature (P/T) and low temper-ature overpressure protection (LTOP) limits for operation through 10 effective full power years (EFPY). At the time of the submittal, the material analysis report of the first removed surveillance capsule had not been completed. Subsequently, the Battelle Columbus Laboratory, under contract to SCE, completed the report entitled " Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the San Onofre Nuclear Generating Station Unit 2."
This report found that the measured neutron fluence was 40% higher than the theoretical neutron fluence estimated in the November 7 submittal. By letter dated December 29, 1988, SCE submitted the report to the NRC and reduced the regeested effec-tive period of the P/T and LTOP limits from 10 EFPY to 8 EFPY. By letter dated February 23, 1989 SCE responded to the NRC's November 30, 1988 re-quest for additional information. The staff has reviewed the SCE submittals. Our evaluation and conclusions are described below.
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- 2 L 2.0 NEUTRON SURVEILLANCE Appendix H to 10 CFR Part 50 requires the' licensees to establish a surveillance program to periodically withdraw surveillance capsules from the reactor. vessel.. SCE removed the first surveillance capsule, which was located at 97-degree azimuth, from SONGS 2 on September 20, 1987 after 2.69 EFPY of operation. The 97-degree capsule was analyzed and tested by the Battelle Columbus Laboratory in cooperation with the Ohio State University..
2.1 Evaluation The objective of the pressure vessel neutron surveillance is to measure the fast neutron fluxi fluence) for energies greater than 1.0 MeV at the tensile specimen locations, to extrapolate these measurements to the pressure vessel, to identify the location-of_ the peak fluence and its spectrum,' and to verify all of the measurements with appropriate calcu-lations.-
Flux measurement is accomplished by the use of activation dosimeters and 'is covered by several ASTM procedures. The measurements and analyses in this report conform to the' requirements of all pertinent ASTM procedures.
The dosimeter monitors covered a wide range of energy and activity half
-lives. The monitors used included Al-Co, U, Ti, Fe, S, Ni, and Cu. Some were Cd shielded and encapsulated in stainless sted. The dosimeter activity was measured using a 4096-channel Ge(L1) detector with a full width, half-maximum resolution of 1.9 kev at 1,332.5 kev. Spectrum data were analyzed using the SPECTRAN-III computer code. For the calculation of the foil activation the daily reactor operating history was taken into account.
The energy distribution and the neutron flux in the reactor, in the surveillance capsule, and in the pressure vessel were calculated using the DOT 4.3 computer program, which is widely accepted for this type of calculation. The principal approximations' included third order P3 scat-l tering, S angular quadrature, 48 azimuthal segments and 47 energy groups.
8 The cross sections were based on the BUGLE-80 -library which has been de-rived from ENDF/B-V data. The DOT was used for R-and R-z runs which provided three dimensional neutron flux values in the locations of interest.
The power distributions (neutron source) were obtained for each fuel pin from the results'of PDQ depletion calculations.
In propagating the neutron field from the core to the pressure vessel, both the boron solution in the coolant (a cycle average value) and the coolant temperatures (densities) in the core and the downcomer were taken into account.
The' experimental flux determination, however, requires knowledge of the neutron spectra at the location of the measurement. Calculated neutron spectra were used, thus the flux values derived from the measured dosimeter activities are semi-experimental. However, this is the generally accepted
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Such check calculations have been performed in this. report and several adjustments related to the neutron spectra have been made. These adjust-ments are based on a spectrum which provides the closest possible agree-ment with'the. measured values of the flux from each of the dosimeters.
The methodologies, practices and computer programs outlined represent the generally accepted means for such calculations. They also represent what the staff has been requiring from licensees on neutron flux measurements i
'and calculations for licensing actions related to 10 CFR 50.61. They are therefore acceptable, j
i 2.2 Results
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Several dosimeters were found to be oxidized, and for their recovery special cutting and cleaning operations were applied. From a trtal of 27 monitors only three sulfur monitors were discarded. The standard deviation of the unadjusted spectra was about 12%. The calculated flux (with the adjusted spectra) is about 25% less than the measured flux at t
the.inside surface of the pressure vessel. The value accepted as the final value of the flux is the measured value, which is conservative.
In
-general the final values of the flux (and the corresponding fluence) fall-in the expected range. The results look reasonably conservative and are acceptable.
2.3 Conclusion 4
The measurement and analysis practices, the computer )rograms and the major approximations in the computations either conform wit 1 the relevant ASTM procedures, use the generally accepted methods and practices in nuclear reactor. dosimetry or comply with staff requirements for similar work. In adoitiontheresultsfallwithintheexpectedrange. Therefore, we find this report acceptable.
3.0- PRESSURE / TEMPERATURE LIMITS Appendix G to 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASMI Code and, in particular, to test the beltline materials in the surveillance capsules in accordance with Appendix H to 10 CFR Part 50. These tests define the condition of vessel embrittlement at the time of capsule with-drawal in terms of the increase in the reference temperature (RT
).
Appendix G also requires the licensee to predict the effects of E tron irradiation on vessel embrittlement by calculating the adjusted RT and upper shelf energy. A method of calculating RT thatisacceptMeto the NRC staff is described in Regulatory Guide 69, Revision 2.
An acceptable method for constructing P/T limits is described in Standard Review Plan (SRP) Soci. ion 5.3.2.
3.1 Evaluation The 97-degree capsule contained Charpy impact specimens and tensile speci-mens that were made from base metal, weld metal, and heat affected zone i
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- I metal. The specimens were designed and fabricated in accordance with ASTM E23-82 and ASTM E8-81. The Charpy impact, tensile, and hardness-tests were conducted in'accordance with appropriate sections of ASTM. In particular, the Charpy impact tests satisfied ASTM E23-82 and impact data 1
was prepared according to ASTM E185-82. The Charpy im i
that. the longitudinal-orientated base metal (C-6404-3) pact tests showed exhibited the l
largest increase in transition temperature (RTl9Il is'the limiting-
) of 51.1*F'at 30 ft-1b 1
impact energy. This indicates.that the base m (controlling) material. To verify this conclusion, the staff calculated the increase in RT o
Guide 1.99, RevisiOOT.f beltline material using methods in Regulatory 2
The staff found that the limiting material with th'e highest adjusted RT is the intermediate shell plate, Code No.
C-6404-3 (heat No. C-75N1). This plate has copper and nickel contents
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of 0.1% and 0.53%, respectively. The unirradiated RT
.I The staff's calculation confirmed that the C-6 N of the plate is 18*F.
3 plate is the-limiting material.
3.2 Results In the November 7,1988 submittal, SCE used the theoretical neutron fluence in the SONGS 2 FSAR to calculate the adjusted RT fluence was based on the neutron transport analyl9I. The FSAR neutron and was 40% lower than the measured neutron fluence in the surveillance capsule report.. The 40%
discrepancy between the measured and theoretical fluence values should not be a safety concern because uncertainties and random error in the transport analysis of this magnitude are expected.
In addition, the licensee has implemented a lower neutron leakage fuel management and a longer operating cycle from 18 to 24 months since the beginning of Cycle 4.
These measures will eventually reduce the actual neutron fluence to the level of the theoretical neutron fluence predicted by the transport analysis. To be
-conservative, the staff used the neutron fluence reported-in the capsule analysis to calculate a maximum adjusted RT of112.4*F(Table 1)for theC-6404-3plateat8EFPYand1/4T(T=Ngkicknessofreactorvesselat beltline). Substituting the RT of 112.4*F into equations in SRP 5.3.2, thestaffverifiedthattheprohledP/Tlimitsinheatup,cooldown,and inservice tests are within the acceptable values.
In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes P/T limits on the reactor vessel closure flanges.Section IV.2 of Appendix G states that when pressure exceeds 20 percent of the pre-service system hydrostatic test pressure, the temperature of the clo mre flange regions that are highly stressed by the bolt preload must exceed the RT of the material in those regions by at least 120'F for normal operate and by 90*F for hydrostatic pressure tests and leak tests.
Based on the flange RT of -10*F, the staff has determined that the proposed P/T limits saI9Ify Section IV.2 of Appendix G.
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Section IV.B of Appendix G requires that the reactor vessel must be thermally annealed if the predicted upper shelf energy (USE) at end of
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) life is below 50 ft-lb. At 2.69 EFPY, the measured limiting USE is 104.7 ft-lb for the transverse-orientated base metal. This is a 16.5% reduction from the unirradiated value of 125.5 ft-lb. Using the method in Regulatory Guide 1.99, Revision 2, the staff predicted that the limiting USE at end of life will still be above 50 ft-lb and thus the USE satisfies the Section IV.B requirement.
j 3.3 Conclusion The proposed SONGS 2 P/T limits on the Reactor Coolant System for heatup, cooldown, and inservice tests are valid through 8 EFPY because the limits conform to the requirements of Appendices G and H to 10 CFR Part 50. The SONGS 2 surveillance program also conforms to the requirements of Appendix H to 10 CFR Part 50. The P/T Limits may be incorporated into the SONGS 2 Technical Specifications.
TABLE 1 The staff's prediction of the adjusted RT for the limiting material.
NDT Code No. C-6404-3 Heat No. C-7595-1 Mat'erial:
Intermediate Shell Plate Copper 0.1%
Nickel 0.53%
Initial RT 84 NDT 19 FluenceatE0L(32EFPY)and1/4T 3.10 x 10 n/cm2 (capsuledata) 18 (capsuledata)
Fluence at 8 EFPY and 1/4T 7.75 x 10 n/ cme Adjusted RT at 8 EFPY and 1/4T 112.4*F NDT Adjusted RT at 32 EFPY and 1/4T 136.3*F NDT 4.0 LOW TEMPERATURE OVERPRESSURE PROTECTION LIMITS i
LTOP is provided by the shutdown cooling system (SDCS) relief valves, which must be aligned to the RCS when the RCS is below the specified temperature to provide assurance that the reactor vessel will be operated in the ductile region in accordance with 10 CFR Part 50, Appendix G, during both normal operation and overpressurization events due to equipment mal-function or operator error. Technical Specifications require alignment of the SDCS relief valves to the RCS whenever RCS temperature is below the temperature corresponding to the P/T curve pressurizer relief valve setpoint of 2500 psia,
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, The' updated P/T limits require the LTOP system to be' aligned whenevr RCS temperature is less than 287"F during RCS cooldown and less than 312'F during RCS heatup. The current Technical Specification.P/T limits only 1
require LTOP alignment when RCS temperature is less than 235'F during
'RCS cooldown and heatup. However, the licensee asserta that the LTOP analyses performed and documented in the FSAR. remain applicable to support the proposed Technical Specifications.
4.1. Evaluation
-The staff questioned whether or not the original LTOP analysis was still bounding for all transients. involving LTOP system design. This concern arose because the. initial RCS temperatures in postulated LTOP transients will be higher in accordance with the proposed LTOP alignment temperatures.
The licensee responded to the staff concern in its letter dated February 23, 1989. As documented in the FSAR, the two limiting pressure-transients for the LTOP system are (1) mass addition transient which assumes an inadvertent actuation of safety injection (startup of two HPSI pumps and three charging-pumps), and (2) energy addition transient which assumes RCP start with a
. temperature difference of 100*F between the steam generator and the reactor coolant system. For the mass addition transient, the SDCS relief valve was sized-to accommodate this LTOP transient for.SDCS temperatures from 120*F to 400*F and assuming a 417 psia relief valve setpoint. Therefore, the change in the LTOP alignment temperature from 235'F to 312'F is still bounded by the original relief valve sizing calculation. For the energy I
addition transient,.the original LTOP analysis assumed that one RCP was-started with a maximum cllowed differential temperature of.100'F between the primary and secondary systems. The most limiting energy addition trans-ient would be with the secondary system at 350*F and the primary system at 250"F. However, this condition is prevented from occurring through ad-ministrative controls. Changing the RCS temperature at which LTOP must be aligned from 235'F to 312*F would not change the results of the most limit-ing energy addition transient. The energy addition transient is driven by the differential temperature between the primary side and. secondary side rather than RCS initial energy. For a primary side temperature of 312*F and the maximum allowed secondary temperature of 350'F (SDCS maximum align-ment temperature), the differential temperature would be reduced from 100*F to 38'F thus reducing the severity of the pressure transient.
4.2 Results Based on the above, the staff agrees with the licensee's conclusion that
. changing the LTOP alignment temperature from 235'F to 312*F would not increase the severity of the most limiting LTOP transient. Therefore, the proposed Technical Specifications regarding LTOP system are bounded by the original FSAR analysis.
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, 4.3 Conclusion The staff finds that the proposed Technical Specification changes are reasonably conservative 6nd acceptable to support the updated P/T limits applicable for the period of 4-8 EFPY.
5.0 CONTACT WITH STATE OFFICIAL The NRC' staff has advised the Chief of the Radiological Health Branch, State Department of Health Services, State of California, of the proposed determination of no significant hazards consideration.. No comments were received.
6.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.37., and 51.35, an environmental assessment and finding of no significant impact have been prepared and published
.(54 FR 14303) in the Federal Register on April 10, 1989. Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.
7.0 CONCLUSION
4 We have concluded, based on the considerations discussed above, that:
'(1) there is reasonable assurance that the health and safety of the ublic will not be endangered by operation in the proposed manner, p(2) such activities will be conducted in compliance with the Commission's
. regulations and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
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Principal Contributors:
L. Lois J. Tsao C. Liang l
Dated: April 11,1989
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