ML20136F647
| ML20136F647 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 09/20/1995 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20136C539 | List:
|
| References | |
| FOIA-96-485 NUDOCS 9703170019 | |
| Download: ML20136F647 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
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ON THE FLORIDA POWER AND LIGHT COMPANY'S SUBMITTAL DATED SD 7/1995 t
i "PRDPOSED LICENSE AMENDMENT - RCS PRESSURE /TENPERATURE LIMITS" I
i MATERIAL INTEGRITY SECTION-M TERIALS AND CHEMICAL ENGINEERING BRANCH l
l QH.I_SION OF ENGINEERING i
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1.0 INTRODUCTION
f By letter dated May 17, 1995, the Florida Power and Light Company (the l
l licensee) requested permission to revise the pressure / temperature (P/T) limits i
in the St. Lucie 1 Technical-Specifications, Section 3.4.
This request stems from new information on beltline materials data, which becomes available as a j
result of Generic Letter (GL) 92-01 review. The proposed P/T limits were requested for 23.6 effective full power years.(EFPY), and were developed using l
Regulatory Guide (RG) 1.99, Revision 2.
GL 88-11, "NRC Position on Radiation l
Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations,"
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recommends AG 1.99, Rev. 2, be used in calculating P/T limits, unless the use i
j of different methods can be justified. The P/T limits provide for the operation of the reactor coolant system during heatup, cooldown. criticality, j
and hydrotest.
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To evaluate the P/T limits, the staff uses the following NRC regulations and i
guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the l
' ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
i RG 1.99, Rev. 2; Standard Review Plan (SRP) Section~5.3.2; and GL 88-11.
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.Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the l
plant.
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications. The P/T limits are 4
among the limiting conditions of operation in the Technical Specifications for i
all consercial nuclear plants in the U.S.
Appendices G and H of 10 CFR l ]i Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits.
4 j-An acceptable method for constructing the P/T limits is described in SRP l_
Section 5.3.2.
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Appendix G ef 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 cf 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards. These tests define the extent 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 ATTACMENT 1 9703170019 970301 PDR-FOIA BINDER 96-485 PDR
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g effects of neutron irradiation on vessel embrittlement by calculating the j~
adjusted reference temperature (ART) and Charpy upper-shelf energy (USE). GL
'~ 11 requested that licensees and permittees use the methods in RG 1.99, Rev.
j 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.
l Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor i
vessel. Appendix H refers to the ASTM Standards which, in turn, require that i
the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zona (HAZ) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the St. Lucie I 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 23.6 EFPY is the lower longitudinal weld (wire Heat No. 305424) with 0.28% copper (Cu),
0.635 nickel (N1), and an initial RT of -60*F.
The ART calculated by the staff for this limiting material is N1.5'F at 1/4T (T - reactor vessel beltline thickness) and 137.1*F at 3 values calculated by the licensee, u/4T at 23.6 EFPY. The corresponding sing Chemistry Factor Table in Section 1.1 of RG 1.99, Rev. 2, are 191*F at 1/4T and 136*F for 3/4T. Both calculations used a neutron fluence of 1.063E19 nfca' at 1/4T and 3.78E18 n/ca' at 3/4T.
The staff judges that a difference of 0.5'F between the licensee's ART of 191*F and the staff's ART of 191.5'F is negligible. Further, since the licensee back-calculated the fluence in this application based on the ART of 191*F, which is the same as that appeared in the 1989 P/T limits submittal, the proposed P/T limits for heatup, cooldown, and hydrotest are identical to the current P/T limit curves and meet the beltline material requirements in Appendix G of 10 CFR Part 50. The only changes made in Figure 3.4-2a and Figure 3.4-2b by the licensee are EFPY values in the legends associated with these P/T limit curves.
In addities to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based en the reference temperature for the reactor vessel closure flange materials.- Section IV.A.2 of Appendix G states that when the pressure exceeds 205 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. Since the flange reference temperature of 30*F remains the same as that in the 1989 submittal and the proposed higher fluence value does not enter the equation for determining P/T limits for flange materials, the staff has determined that the proposed P/T limits satisfy Section IV.A.2 of Appendix G.
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Section IV.A.1 of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb. The limiting material is the intermediate shell with 76 ft-lb. Using RG-1.99, Rev. 2, the staff calculated that the end of life USE will be 58 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable.
3.0 CONCLUSION
The staff concludes that the proposed P/T limits for the reactor coolant i
system for heatup, cooldown, leak test.. and criticality are valid through 23.6 EFPY because the proposed limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The proposed P/T limits also satisfy GL 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 St. Lucie 1 Technical i
Specifications. The staff also accepts the editorial changes in the St. Lucie 1
4.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2: Pressure-Temperature Limits 3.
May 17, 1995, letter from D. A. Sager (FPL) to USNRC Document Control Desk, subject: Proposed License Amendment, RCS Pressure / Temperature Limits 4.
December 5, 1989, letter from J.H. Goldberg (FPL) to USNRC Document Control Desk, subject: Proposed License Amendment, P-T Limits and LTOP Analysis
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i SALP INPUT 4
h FACILITY NAME:
St. Lucie 1 SUBetARY OF REVIEW ACTIVITIES The staff has reviewed the licensee's pressure / temperature (P/T) limits in the St. Lucie 1 Technical Specifications as a part of Generic Letter 88-11 review.
Generic Letter 88-11 requires the licensee to use Regulatory Guide 1.99, Rev.
2, to calculate the nil-ductility reference temperature, RT which is a parameterusedinestablishingtheP/Tlimits.ThestaffcaIEu,latestheRT of the limiting beltline material, and compares it to the licensee's RT,.,,
Based on the limiting RT usingStandardReviewPla$,5.3.2.the staff verifies the licensee's P/T limits NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE-FUNCTIONAL AREA i
ENGINEERING / TECHNICAL SUPPORT The licensee's analysis for calculating P/T limits resulted in conclusions in close agreement with the staff's.
This suggests that appropriate attention has been given by the licensee to the engineering adequacy of the submittal.
The licensee is also very responsive to staff's request for an additional document.
SAFETY ASSESSMENT /00ALITY VERIFICATION The licensee's calculation of RT P/T limits are within the limits,n follows the method in RG 1.99, Rev. 2. The of SRP 5.3.2 results. Quality control in preparing the P/T limits is evident.
Author: Simon Sheng, DE/ENCB 415-2708
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