ML20133N719
| ML20133N719 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 12/30/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20133N699 | List: |
| References | |
| NUDOCS 9701230384 | |
| Download: ML20133N719 (5) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20666-0001
.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED AMENDMENT NO. 80 TO FACILITY OPERATING LICENSE NO. NPF-85 PHILADELPHIA ELECTRIC COMPANY, LIMERICK GENERATING STATION. UNIT 2 DOCKET NO. 50-353 9
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1.0 INTRODUCTION
By letter dated August 1, 1996, the Philadelphia Electric Company (the licensee) submitted a request for changes to the Limerick Generating Station, Unit 2, Technical Specifications (TSs). The requested changes would revise TS Section 3/4.4.6 (i.e., Figure 3.4.6.1-1) to reflect the addition of two hydrotest curves, effective for 6.5 and 8.5 Effective Full Power Years (EFPY),
to the existing Pressure-Temperature Operating Limit (PTOL) curves for LGS Unit 2.
The P/T limits are used to operate the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
The staff evaluates the P-T Limits of pressurized water reactors (PWRs) based on the following NRC regulations and guidance:
10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11; GL 02-01, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP) Section 5.3.2.
GL 88-11 advised licensees that the staff would use RG 1.99, Revision 2 to review P/T Limit Curves.
RG 1.99, Revision 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.
GL 92-01, Revision 1, requested that licensees submit their reactor pressure vessel (RPV) data for their plants to the staff for review.
GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.
These data are used by the staff as the basis for the staff's review of P-T Limit submittals, and as the basis for the staff's review of pressurized thermal shock assessments (10 CFR 50.61 assessments). Appendix G to 10 CFR Part 50 requires that P-T Limits for the RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code.
SRP Section 5.3.2 provides an acceptable method of calculating the P-T Limits for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code.
The basic parameter of this methodology is the stress intensity factor K, which is a function of the stress state and flaw configuration.
The 3methods of Appendix G postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress. The flaw in the RPV is postulated to have a depth that is equal to one-fourth of the RPV 9701230384 961230 PDR ADOCK 05000353 P
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l beltline thickness and a length equal to 1.5 times the RPV beltline thickness.
The critical locations in the RPV beltline region for calculating heatup and cooldown P/T Limit Curves are the 1/4 thickness (1/4t) and 3/4 thickness i
(3/4t) locations, which correspond to the depth of the maximum postulated flaw, if initiated and grown from the inside and outside surfaces of the RPV, 4
respectively.
The Appendix G, ASME Code methodology requires that licensees determine the adjusted reference temperature (ART or RT at the maximum postulated flaw The* ART is defined as the sum of 6y) initial (unirradiated) reference depth.
e temperature [RT,b]y, irradiation (ART the mean value of the adjustment in reference temperature cause m ), and a margin (M) term.
The ART,
is a product of a chemistry factor and a fluence factor.
ThechemistryfaItor is dependent upon the amount of copper and nickel in the material and may be determined from tables in the RG or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the RT is a plant-specificoragenericvalueandwhetherthechemistryfaIt'or>wasdetermined cu i
using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of RT copper and nickel contents, fluence and calculational procedures.
RG 1.NcuNevision 2, describes the methodology to be used in calculating the margin term.
The extent of the P/T limits revision in this submittal was limited to the adding of two hydrotest curves to the current P/T limits.
Hence,'the evaluation was focused on this revised part of the P/T limits.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on the limiting beltline material in the Limerick 2 reactor vessel. The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Revision 2.
The twt, hydrotest curves of 6.5 and 8.5 EFPYs, which are to be added to the P/T limits of Limerick 2, were derived using the same material parameters and loading information as those of the existing hydrotest P/T limits of 10 EFPYs. The only difference between the added curves and the existing one is the different fluence values as reflected by their different EFPY values. The limiting material with the highest adjusted reference temperature (ART) at 6.5 and 8.5 EFPYs for Limerick 2 is plate 14-2 (heat number B3416-1), which has 0.14% copper (Cu), 0.65% nickel (Ni), and an initial RT, of 40*F.
Presently, no surveillance capsules have been withdrawn from the reactor pressure vessel.
Hence, the chemistry table in RG 1.99, Rev. 2 was used to determine the chemistry factor in calculating the ART The staff calculated the highest ART for both EFPYs.
For plate 14-2 at 6.5 EFPY, the staff calculated the ARTS to by 79.0 F at 1/4t, and 64.4 F at 3/4t.
The staff used a fluence of 2.42E17 n/cm at 1/4t and 1.15E17 n/cm at 3/4t.
2 At 8.5 EFPY, the ARTS were calculated to be 86.0 F and 69.2"F using fluence
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3 values of 3.17E17 n/cm and 1.51E17 n/cm at 1/4t and 3/4t, respectively. The i
fluence values were interpolated from the internal-diameter (ID) valut of l
1.73E18 at the end of license (EOL); the ARTS were determined using Section 1 of RG 1.99, Revision 2, because no surveillance capsules have been withdrawn from the reactor pressure vessel.
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The licensis calculated that the RT, initial RT, of 40*F.
would shift upwards by 41.1*F for 6.5 EFPY and 48.0*F for 8.5 EFPY from an Therefore, the i
respective ARTS would be 81.1*F and 88.0*F.
These agree well with the staff's calculated. values of 79.0*F and 86.0*F, considering that the staff estimated i
j the fluenc'e' values for 6.5 and 8.5 EFPYs based on a linear interpolation.
Substituting these ARTS into equations in SRP 5.3.2, the staff verified that the added P/T limits for hydrotest meet the beltline material requirements in i
j Appendix G of 10 CFR Part 50.
The staff concludes that the two added P/T limits for the reactor coolant system hydrotest are valid through 6.5 and 8.5 EFPY because the limits conform l
to the requirements of Appendix G of 10 CFR Part 50. The P/T limits also e
satisfy Generic Letter 88-11 because the licensee used the method in RG 1.99, Revision 2 to calculate the ART. Hence, the proposed P/T limits and their j
corresponding Bases may be incorporated into the Limerick 2 Technical i
j Specifications.
3.0 STATE CONSULTATION
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In wcordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment.
The State official had no commentis.
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4.0 ENVIRONMENTAL CONSIDERATION
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i The amendment changes a requirement with respect to installation or use of a 2
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, j
of any effluents that may be released offsite, and that there is no i
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 (61 FR 57490). 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 i
the amendment.
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5.0 CONCLUSION
T The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the
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public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common i
defense and security or to the health and safety of the public.
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Principal ' ontributor:
S. Sheng C
Date: December 30, 1996 s
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6.0 REFERENCES
1.
Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988 i
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NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature Limits, dated July 1981 3.
Letter from G. A. Hunger, Jr., PEC0, to U.S. NRC Document Control Desk,
Subject:
Limerick Generating Station, Unit 2 - Technical Specifications l
Change' Request No. 96-15-2, dated August 1, 1996
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