ML20029A506
| ML20029A506 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/07/1991 |
| From: | Silver H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029A502 | List: |
| References | |
| NUDOCS 9102250121 | |
| Download: ML20029A506 (3) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION f
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO._133 T0_ FACILITY OPERATING LICENSE NO. OPR-72 FLORIDA POWER _ CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50 302 INTRODUCTION By letter dated October 31,1989 as supplemented March 30, 1990 and August 10, 1990, Florida Power Corporation FPC or the licensee) requested an amendment to the Technical Specifications TS), appended to facility Operating License Nr <- DRP-72 for the Crystal River Unit No. 3 Nuclear Gent'ating Plant (CR-3).
The proposed amendment revises the reactor coolant system (RCS) pressure /
temperature P/T) limits. The current P/T limits are for 8 effective full power years EFPY) of operation. The proposed anendment provides P/T limits assed upon predicted reactor vessel neutron esbrittlement for up to 15 EFPY, using the mthods of Regulatory Guide (RG) 1.99, Rev. 2.
Generic Letter 88-11 requested that licensees use RG 1.99, Rev. 2 to predict neutron irradiation effects. The proposed amendment also changes limits on plant
-heatup and cooldown. ate to be more representative of actual plant capability.
FPC also requested TS revisions for low temperature overpressure )rotection (LTOP).
This Safety Evaluation does not address the proposed LTO) changes, which will be addressed in a separate amendant.
This Safety Evaluation applies only to P/T limits for the current format TSs.
-Revisions proposed in the format of the Technical Specification Improvement Program will be reviewed as part of that effort.
EVALUATION 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 the ASMECode,whichare~referencedin-AppendicesGandH;10CFR50.36(c)(2)
RG 1.99, Rev. 2;-Standard Review Plan (SRP) Section 5.3.2; and Generic Letter 88-11. Appendix G 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 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 a margin to account for uncertainties in the prediction method.
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l Appendix H to 10 CFR Part 50 requires licensees to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel. Appendix H refers to the ASTM Standard which, in turn, requires that the capsules be installed in the vessel before startup and that they contain test specimens that are made from plate, weld, and heat-affected zone (HAZ) materials of the reactor beltline.
- The staff. evaluated the effect of neutron irradiation embrittlement on each beltline material in the CR 3 reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2. The staff has determined that the material with the highest ART at 15 EFPY was the upper-to-lower shell girth weld, WF-70, with 0.35% copper (Cu), 0.59% nickel (Ni),
and an initial RT of -6 4.
ndt The ifcensee has removed four surveillance capsules from CR-3.
The i
results from capsules B, C, D, and F were published in Babcock & Wilcox.
I reports BAW-1679, BAW-1898, BAW-1899, and BAW-2049, respectively. All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal, i
for the limiting-beltline material, weld WF-70 the staff calculated the ART to be 191.1'F at 1/4T (T =-reactor vessel beltline thickness) and j41.4'F for i'
3/4T at 15 EFPY 2 The staff.used a neutron fluence of 2.99E18 n/cm at 1/4T and 1.08E18 n/cm at 3/4T.
The ART was determined by Section 1 of RG 1.99, Rev. 2.
The licensee calculated a higher ART (203'F) than the staff at 1/4T for the limiting material WF-70 because the licensee used a conservative safety margin.
At the 3/4 location, the licensee selected the atypical-weld as the' limiting material.
In 1978 Babcock & Wilcox performed chemical analyses of samples of beltline welds in.CR-3.
The results indicated that one of the welds had atypical concentrations of nickel and silicone for the 1.inde 80 submerged-arc welds in the reactor vessel.-- Based on the analysis, the atypiw1 weld showed a higher than normal.value of RT Thus, the licensee calculated a more conservativeARTof171'FattYt.3/4 location. The staff judges that the
-i licensee's ARTS are conservative and, therefore, acceptable.
Substituting the
- ARTS of 203*F and 171'F into equations in SRP 5.3.2, the staff verified that the.prooosed 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 the reference temperature for the reactor vessel closure flange ~ materials.
Section-IV.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 60'F the staff has-determined that the proposed.P/T limits satisfy Section IV.2 of Appendix G.
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. Finally, during the staff's review of the proposed TS changes, it was noted on page B3/4 4-11 of the application that a line was inadvertently dropped from tie first paragraph. This error has been corrected by the staff.
Based on the above evaluation, the staff concludes that the proposed changes to the P/T limits for up to 15 EFPY are acceptable.
ENVIR0fEENTAL CONSIDERATI0H This amendment involves a change to a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types of any ef fluents thatmaybereleasedoffsite,andthatthereisnosignificantIncreasein individual or cumulative occupational radiation exposure. The Commission has areviously issued a propcsed finding that this amendment involves no significant lazards consideration and there has been no public comment on such finding.
Accordingly, this 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 impect statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLUS10!!
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 Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: February 7, 1991 Principal Contributor:
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