ML20134G259
| ML20134G259 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/21/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20134G257 | List: |
| References | |
| NUDOCS 8508230197 | |
| Download: ML20134G259 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
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WASHINGTON, o. C. 20655 j
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4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
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SUPPORTING AMEN 0 MENT NO. 89 TO PROVISIONAL OPERATING LICENSE NO. DPR-20 i
CONSUMERS POWER COMPANY l
PALISADES PLANT (PNP)
DOCKET NO. 50-255
1.0 INTRODUCTION
i 14, 1985, the Consumers Power Company (CPC) submitted By letter dated June a request for changes to the Palisades Plant Technical Specification Sections i
3.1.2 and 3.1.3.
j The amendment provides new reactor vessel pressure-temperature limits for.
heat-up, cooldown and hydrostatic test. The last-surveillance capsule report submitted to the staff by the licensee was Westinghouse WCAP 10637, entitled l
" Analysis of Capsule T-330 and W-290 from the Consumers Power Company Palisades j
Reactor Vessel Radiation Surveillance Program." This report was submitted to the NRC by letter dated October 31, 1984.
A Notice of Consideration of Issuance of Amendment to License and Proposed
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No Significant Hazards Consideration Determination and Opportunity for 4
l Hearing related to the requested action was published in the Federal
'l Register on July 3, 1985 (50 FR 27504).
No comments or requests for 4
hearing were received.
2.0 DISCUSSION Pressure-temperature limits must be calculated in accordance with the require-ments of Appendix G, 10 CFR 50, which became effective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the require-for the ments of Appendix G, 10 CFR 50 are dependent upon the initial RTN limiting materials in the beltline and closure flange regions of N e reactor vessel and increase in RT resulting from neutron irradiation damage to NOT the beltline materials.
4 The PNP reactor vessel was procured to ASME Code requirements, which did not i
for each of the specify fracture toughness testing to determine the RTfoYDInterialsinthe reactor vessel materials.
Hence, the initial RT closure flange and beltline region of the PNP retNor v.lssei couid not be determined in accordance with the test requirements of the ASME Code.
for these materials must be estimated from Therefore, the initial RT test data from other simiN materials used for fabrication of reactor
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. The licensee indicates that the limiting closure flange region materials were forgings, which were fabricated to ASME Code SA 508 C12 requirements.
Tne licensee has estimated the RT for these materials in accordance with Branch Technical Position - NB 5-2, " Fracture Toughness Requirements,"
which are contained in NUREG-0800, "USNRC Standard Review Plan 5.3.2, Pressure-temperature Limits".
This branch technical position provides con-servative estimates of RT for reactor vessel materials.
This branch technical position resultbIn an RT for the closure flange forgings of NDT 60*F.
The limiting materials in the PNP reactor vessel beltline are weld metals, which were fabricated by Combustion Engineering using the submerged arc weld process with RACO 3 and MIL B-4 Mod (Mn Mo Ni) weld wires.
The RACO 3 sub-s.
merged arc welds were fabricated using a second wire of pure nickel, identified as Ni 200.
In all submerged arc welds the flux utilized was Linde 1092.
The initial RT for these weld materials was estimated by the licensee as -56*F withastabrddeviationof17*F.
These initial RT and standard deviation valueswererecommendedbythestaffinCommissionRhrtSECY84-465," Pres-surized Thermal Shock" for welds fabricated by Combustion Engineering using Linde 1092 flux.
The increase in RT resulting from neutron irradiation damage was estimated by the licensee us k the method documented in Draft Regulatory Guide 1.99, Revision 2, " Radiation Damage to Reactor Vessel Materials." Although this regulatory guide is only a draft, its methodology is considered by the staff to be the most up-to-date method for predicting neutron irradiation damage.
This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and nickel in the beltline material.
The licensee has conducted a detailed search of vessel and surveillance fabrication records at Combustion Engineering to determine the heats of wire used in their reactor vessel beltline and their surveillance welds.
As a result.of this search, the licensee indicates that the PNP surveillanc'e weld was fabricated using heats of wire, which were different from those used in fabrication of the PNP beltline welds.
The search confirmed that RACO 3 heat numbers W5214 and 348009 and MIL B-4 Mod (Mn Mo Ni) heat number 27204 were utilized to fabricate the PNP reactor vessel beltline.
During fabrication of the PNP reactor vessel, chemical analyses of the PNP beltline welds were not performed.
However, the licensee in Attachment III to its June 14, 1985 memorandum has established the amounts of copper and nickel in each of the beltline welds. The amounts of copper and nickel were estimated from chemical analyses of reactor vessel surveillance welds and other nuclear vessel welds, which were fabricated by Combustion Engineering using the same heats of weld wire as the PNP beltline material.
Since the amount of copper and nickel should be consistent within a heat of weld wire and the wire is the source of copper and nickel in a weld, the use of chemical analyses from surveillance welds and other nuclear vessel welds fabricated with the same heats of wire as the PNP beltline weld should provide reliable estimates for the amounts of copper and nickel in the PNP beltline welds.
t The licensee's proposed pressure-temggraturg limits have been calculated The am'unt of time using a neutron fluence of 1.30 x 10 n/cm (E > IPeV).
o corresponding to this neutron fluence incident on the reactor vessel is dependent upon a radiological evaluation of the core and the PNP vessel.
Report WCAP-10637 contairs a description of the radiological analyses per-formed by Westinghouse on the PNP core vessel. This analysis results in a lead factor of 1.P8 between the capsule and the vessel location receiving the highest neutron flux. The Westinghouse radiological calculation predicts the end of life (2530 Myg for 32 effective full-power years) peak neutron fluence to be 6.56 x 10 n/cm (E > IPev), when the axial peaking factor at the core midplane is 1.20.
The licensee has evaluated its previous core peeking factors and detemined that the axial peaking factor for the nidplene of the core was 1.15. This decrease in the axial peaking fggtor cguses the end of life peak neutron fluence to be reduced to 6.29 x 10 n/cm (E > 1MeV).
Report WCAP-10637 contains the Westinghouse analysis of the desimetry in Surveillance Capsule W-290. The calculated peak neutron fluence at the end of life using the results from tgg Capsyle W-290 dosimetry and the predicted lead factor of 1.28 is 5.38 x 10 n/cm (E > IMeV). Since the peak neutron fluence from the Capsule W-290 dosimetry is less than that calculated using the Westinghouse radiological analysis with 1.15 axial peaking factor, the Westinghouse calculated value will conservatively estimate the end of life neutron fluence for the PNP reactor vessel.
3.0 EVALUATION The staff has used the method of calculating pressure-temperature limits in '
USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev. 1 July 1981 to evaluate the proposed pressure-temperature limits. The amount of neutron irradiation damage to the beltline materials was estimated using the method documented in Draft Regulatory Guide 1.99, Working Paper F dated May 21, 1985. The inputs used were the amounts of copper and nickel reported in Attachment III to the licensee's letter dated June 14 1985 and the calculated end of 79 life peak neutron fluence of 6.29 x 10 n/cm (E > IMeV). The pressure temperature curves submitted by the licensee have been accepted by the staff; however, the licensee's submittal would permit use of these curves without reassessment for a period of 9 effective full power years. Upon review by the staff, a mathematical error was found which, when corrected, would require that these curves should be used only for 6.6 effective full power years before they are reassessed. The licensee hes accepted the staf# assessment. This change does not change the substance of the amendment requested by the licensee and there are no otber differences between the anendment requested by the licensee and the amer.dment authorized by the staff. Our conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G,10 CFR Part 50 for 6.6 effective full power years and may be incorporated into the plant's technical i
specifications.
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4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the.
installation or use of facility components located within the restricted The staff has determined that the area as defined in 10 CFR Part 20.
amendment involves no significant increase in the amounts, and no sianificant change in the types, of any effluents that may be released offsite and that
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there is no significant increase in individual or cumulative occupational The Commission has previously issued a proposed finding radiation exposure.
that this amendment involves no significant hazards consideration and there Accordingly, this amendment has been no public comment on such finding.
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
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statement or environmental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the haalth and safety of the public will not' be endangered by operation in the r.-posed 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.
6.0 ACKNOWLEDGEMENT This Safety Evaluation has been prepared by B. J. Elliot.
Dated:
August 21, 1985.
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