ML13331A781
| ML13331A781 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/31/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13331A780 | List: |
| References | |
| TAC-55027 NUDOCS 8502010558 | |
| Download: ML13331A781 (6) | |
Text
UNITEDSTATES o0 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 0Z ENCLOSURE 1 EVALUATION OF PROPOSED TECHNICAL SPECIFICATION CHANGE FOR SAN ONOFRE NUCLEAR GENERATING STATION-UNIT 1 (PROPOSED CHANGE NO. 137)
OFFICE OF NUCLEAR REACTOR REGULATION 100 INTRODUCTION In a letter from K. P. Baskin to H. R. Denton dated May 17, 1984 the Southern California Edison'Company requested a revision to the San Onofre-1 Appendix A Technical Specification 3.1.3, "Combined Heatup, Cooldown and Pressure Limitation." To support this change the licen see referenced Westinghouse WCAP-9520, "Analysis of Capsule F from the Southern California Edison Company, San Onofre Reactor Vessel Radiation Surveillance Program."
This analysis was submitted for staff review in a letter from K. P. Basin to H. R. Denton dated April 18, 1980.
2.0 EVALUATION The purpose of the reactor vessel surveillance program is to monitor the effect that neutron irradiation and the thermal environment will have on the beltline materials' reference temperature (RT
). The method recommended by the staff for predicting the effect oD eutron irradiation and the thermal environment on beltline materials' RT is documented in Regulatory Guide 1.99, "Effects of Residual Elements o Predicted Radiation Damage to Reactor Vessel Materials." In revision 1 to this Regulatory Guide, the increase in RT TRT is dependent upon the amount of copper, phosphorus and neutron fluence.
Tdraft revision 2 dated July 23, 1984 (Ref. 1) the increase in RT Tis dependent upon the amount of copper, nickel and neutron fluence. DrafqDrevision 2 was prepared from analysis of commercial reactor vessel material surveillance data generated during the staff's review of the issue of "Pressurized Thermal Shock" and is being prepared for review by the Committee for Review of Generic Requirements (CRGR). Table I compares the actual increase in RT for the Capsule F material surveillance data to the amounts predicted T Regulatory Guide 1.99 Rev. 1 and 2. This.
comparison indicates that the actual increase in RT-T for the surveillance materials was less than that predicated by Regulato TGuide 1.99, Rev. 1 or Rev. 2. However, the increase predicted by revision 2 was much closer to that measured on the surveillance materials. Hence, the staff believes that revision 2 would be more appropriate for predicting the increase in RTNDT for the San Onofre-1 beltline materials.
9502010558 850131 PDR ADOCK 05000206 P
-2 Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G to 10 CFR Part 50, which became effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G, 10 CFR Part 50, are dependent upon.
the initial RTNeT for the limitinq materials in the beltline and closure flange regions N the reactor.vessel and the increase in RT resulting from neutron irradiation dAmage to the limiting beltline maYR ial.
The initial RT for the limiting materials in the San Onofre-1 reactor vessel beltlinND nd closure flange regions were 820 F and 60'F, respectively.
These values were estimated by the licensee using the method recommended by the staff in Branch Technical Position MTEB 5-2, Standard Review Plan 5.3.2 (NUREG-0800, Rev. 1 dated July 1981). The licensee indicates that the increase in RT would result in an adjusted RT at the 1/4 thickness location of 21PT after 16 effective full power 9 rs (EFPY) of operation.
This adjusted RTNDT was used to calculate the licensee's pressure-temperature curves.
As previously discussed, the increase in RT predicted using Draft Regulatory Guide 1.99, Rev. 2 is dependent Un the neutron fluence and the amount of copper and nickel in the limiting beltline material. Using the method recommended in this draft Regulatory Guide, the adjusted RT for the limiting beltline material in the San Onofre-1 beltline at theN9 4 thickness location after 16 EFPY of operation is 208 0F.
The limiting beltline material is plate W7601-5, which the licensee indicates has 0.14 percent copper. The staff has used the licensee's estimates of the amount of copper and neutron fluence and an assumed nickel of 0.20 percent to calculate the adjusted RT for the limiting beltline material.
Since the adjusted reference tN erature predicted using Draft Regulatory Guide 1.99, Rev. 2 is less than that used to calculate the San Onofre 1 pressure-temperature limits, the staff concludes that the pressure-temperature limits provide sufficient margin to account for neutron irradiation damage for 16 EFPY of operation.
Appendix G to 10 CFR Part 50, also requires that the temperature of the flange regions that are highly stressed by bolt preload must exceed the RT of the materials.in those regions by at least 120 0 F for normal operatiNT and 900F for hydrostatic pressure and leak tests whenever the pressure exceeds 20 percent of the preservice system hydrostatic test pressure unless.a lower temperature can be justified by showing that the margins of safety for those regions are equivalent to those required for the beltline, when it is controlling.
The licensee indicates in its May 17, 1984 letter that the curves were revised based on the requirements in Appendix G to 10 CFR Part 50.
The staff has not completed its evaluation of this statement and whether the beltline region in San Onofre 1 reactor vessel is more limiting than the closure flange region. However, since the surveillance capsule data indicate that the licensee's pressure-temperature curves have adequate safety margins, the staff concludes that it is acceptable for the licensee to utilize the
-3 pressure-temperature curves until the completion of Cycle 8. We request that the licensee clarify the method of determining that the beltline region is more limiting than the closure flange region. The requested ihformation should be sent for staff review and approval at-least six months prior to the start of Cycle 9.
3.0 REFERENCE Draft Regulatory.Guide 1.99, Revision 2, "Radiation Damage to Reactor Vessel Materials," -Working Paper C, July 23, 1984, Office of Nuclear Regulatory Research.
4.0 ACKNOWLEDGEMENT B. Elliot prepared this evaluation.
TABLE I COMPARISON OF ACTUAL INCREASE IN RFFERENCE TEMPERATURE FOR MATERIAL IN CAPSULE F TO THAT PREDICTED BY REGULATORY GUIDE 1.99, REV. 1 AND REV. 2 Surveillance Neutron Fluence Increase in Ref. Temp. (oF), R.G. 1.99 Materials F X 1019 n/cm2 Actual Rev. 1 Rev.2*
Corr. Monitor 1.20 120 192 139 2.36 150 269 159 5.14 130 385 174 Plate W7601-8 2.36 110 246 149 5.14 120 363 174 Weld Metal 1.20 80 213 161 5.14 145 385 196
- R.G. 1.99 Rev. 2 is mean increase in reference temperature plus 2 sigma uncertainty in prediction method.
ENCLOSURE 2 SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE NUCLEAR UNIT 1 DOCKET NO. 50-206 REQUEST FOR ADDITIONAL INFORMATION
- 1. Provide the nickel composition for all plate materials in the reactor beltline.
- 2. Provide pressure temperature limit curves that comply with explicit closure flange material temperature requirements of the revised (May 27, 1983), Appendix G, 10 CFR 50, or provide responses to Items 3 and 4 below:
- 3. Provide the analysis that shows that the closure flange region is less limiting than the beltline region. Include as minimum the following information:
- a. A description of the finite element analysis used to determine the stresses within the closure flange region.
- b. Indicate the peak bolt-up, pressure, and thermal stresses determined by the finite element analysis at the inside and outside surface locations of the flange to head and flange to shell junctions.
- c. Indicate how the bolt-up, pressure, and thermal stresses were combined to determine the maximum applied stress intensity factors.
- d. Indicate the flaw geometry used to calculate the maximum applied stress intensity factors.
- e. Indicate the maximum applied stress intensity factors for the flange-to-head and flange-to-shell junctions.
- f. Indicate the non-destructive examination methods that will be used during inservice examination to determine that.the critical flaw size, which was used in determining the maximum applied stress intensity factors, is not within the flange-to-head and flange-to-shell junctions.
- g. Indicate whether the non-destructive examination methods identified in (f) have been evaluated to demonstrate that the examination methods are capable of locating and sizing flaws of the geometry used for calculating the maximum applied stress intensity factors.
Indicate the results of the evaluation.
-2
- 4. In order to demonstrate that the beltline region is more limiting than the flange region, indicate the minimum metal temperature at the flange and beltline regi6ns which result from the fracture analysis. During a heatup and cooldown what are the required minimum water temperatures to ensure that the limiting flange locations will be equal to or greater than the required minimum metal temperatures?