ML20211G906

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Safety Evaluation Supporting Amends 7 to CPPR-135 & CPPR-136
ML20211G906
Person / Time
Site: Seabrook  
Issue date: 06/10/1986
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20211G905 List:
References
NUDOCS 8606200263
Download: ML20211G906 (7)
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SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 7 TO CONSTRUCTION PERMITS CPPR-135 AND CPPR-136 PUBLIC SERVICE COMPANY OF NEW PAMPSPIRE SEABROOK STATION, UNITS 1 AND 2 DOCKET NOS. 50-443 AND 50-444 INTRODUCTION By letters dated August 9, 1984, February 1, 1985, and July 10, 1985, Public Service Company of New Fampshire (applicant) on behalf of itself and its joint owners

The limited schedular exemption granted by the Commission permits the applicants to eliminate the installation of protective devices and the consideration of the dynamic effects and loading conditions associated with postulated pipe breaks in the four primary loops in the Seabrook Station, Units 1 and 2 primary coolant system for a period ending at the completion of the second refueling outage, pending the outcome of rulemaking on this subject. The February 1,1985 letter also included an analysis of the occupational radiation dose reductior which constituted a value-impact analysis associated with the exemption request. The value-impact analysis together with the technical information contained in Westinghouse Report WCAP-10567, provide a comprehensive

,iustification in support of applicant's request for a partial exemption from the requirements of GDC-4.

EVALUATION The staff's detailed evaluation and basis for granting the partial exemption to the requirements 'of GDC-4 are delineated in the Exemption enclosed with the staff's November 22, 1985, letter. A summary of the staff's evaluation findings and conclusions are immediately below.

SUMMARY

OF EVALUATION FINDINGS From its evaluation of the analysis contained in Westinghouse Report WCAP-10567 for the Seabrook Station, Units 1 and 2, the staff found that the applicants presented an acceptable technical justification which adequately addressed the staff's evaluation criteria, to: (1) eliminate the need to postulate longitudinal and circumferential pipe breaks in the reactor coolant loop (RCf.) piping (hot leg, cold leg, and cross over leg pipingl; (2) eliminate the

  • The current construction permit holders for Seabrook Station are: Bangor Pydro-Electric Company, Canal Electric Company, Central Maine Power Company, Central Vermont Public Service Corporation, Connecticut light & Power Company, Fitchburg Gas & Electric light Company, Pudson light and Power Department, i

Maine Public Service Company, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Public Service Company of New Hampshire Taunton Municipal I.ighting Plant, the United Illuminating Company, Vermont Electric Generation and Transmission Cooperative, Inc., and Washington Electric Cooperative, Inc.

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L need to install associated pipe whip restraints in the RCl primary piping and (3) eliminate the requirement to analyze and design for the dynamic effects of these breaks including jet impingement, reactor cavity pressurization and load combination assumptions.

This finding does not 19 any way affect the design bases for the cor.tainment, the emergency core cooling system, environmental cualification, engineered safety feature systems response, or the design of the Reactor Coolant System (RCSI heavy components support. This finding is predicated on the fact that each of the parameters evaluated for Seabrook Station, Units 1 and 2 is enveloped by the generic analysis performed by Westinghouse, contained in Westinghouse Report WCAP-9558, Revision 2, and accepted by the staff in Enclosure (1) to NRC Generic l.etter 84-04 (February 1, 1984). Specifically, the NRC determined that:

(1) The loads associated with the highest stressed location in the main loop primary coolant system piping are 2332 kips (axial), 37,045 in-kips (bending moment) and result in maximum stresses of about 97% of the bounding stress used by Westinghouse in Reference 1.

Further, these loads are approximately 88% of those established by the staff as limits.

(2) For Westinghouse plants, there is no history of cracking failure in reactor primary coolant system loop piping. The Westinghouse reactor coolant system primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g. intergranular stress corrosion cracking), water hammer, or fatigue (low and high cycle). This operating history totals over 400 reactor-years, including five (Si plants each having 15 years of operation and 15 other plants with over 10 years of operation.

(3) The leak rate calculations performed for the Seabrook plants using an initial through-wall crack of 7.5 inches are identical to those of Reference 2.

The Seabrook Station has an RCS pressure boundary leak detection system which is. consistent with the guidelines of Regulatory Guide 1.45, and it can detect leakage of one (1) gpm in one hour. The calculated leak rate through the postulated flaw results in a factor of at least 10 relative to the sensitivity of the Seabrook plants' leak detection system.

(4) The margin in terms of load based on fracture mechanics analyses for the leakage-sized crack under normal plus SSE loads is within the bounds calculated by the staff in Section 4.2.3 of Enclosure 1 to Reference 2.

Based on a load-limit analysis, the load margin is about 2.0 and based on the J-limit, the margin is at least 1.1.

. (5) The margin between the leakage-size crack and the critical-size crack was calculated by a limit-load analysis. Again, the results demonstrated that a margin of at least 3 on crack size exists and is within the bounds of Section 4.2.3 of Enclosure I to Reference 2.

(6) In addition to the wrought stainless steel pipes, the Seabrook plants have cast stainless steel fittings and associated welds in the RCS.

As an integral part of its review, the staff's evaluation of the material properties data of Reference 3 is enclosed as Appendix I in the Exemption enclosed with the staff's November 22, 1985 letter.

In Reference 3 data for ten (10) plants are presented and lower bound or " worst case" material properties were identified and used in the analysis performed in the Reference 4 report by Westinghouse. The applied J for the Seabrook plants in Reference 4 for cast stainless steel fittings was less than 3000 in-lb/in2 Pence the staff's upper bound 3000 in-lb/in2 on the applied J (refer to Appendix I, page 6) was not exceeded.

ENVIRONMENTAL. ASSESSMENT In advance of issuing the Exemption, the Commission published in the Federal Register on November 19, 1985 (50FR 47468) an " Environmental Assessment and Finding of No Significant Impact." It was stated in that assessment that the planned Exemption action would not have a significant effect on the quality of the human environment. The Exemption granted involves. design features located entirely within the plant restricted area as defined in 10 CFR Part 20; does not affect plant radioactive and non-radioactive effluents; has no other environmental impact and does not involve the use of resources not previously considered in the Final Environmental Statement (construction permit) for Seabrook Station, Units 1 and 2.

The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has determined that the amendments involve no significant hazards considerations. Accordingly, the amendments meet 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 this amendment.

CONCLUSION In granting the limited schedular Exemption, the staff found that the advanced fracture mechanics techniques used by the applicants provided assurance that flaws in primary system piping will be detected before they reach a size that could lead to unstable crack growth.

For this reason, further protection t

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. provided by protective devices against the dynamic effects resulting from the discharge from postulated breaks in the primary piping is unnecessary.

Additionally, consideration of such dynamic effects associated with previously postulated pipe breaks is unnecessary. With full protection against dynamic effects provided by adv=.n %d analysis techniques, and based on the considerations discusste above, we conclude that:

(1) the proposed amendments to Construct sn Permits CPPR-135 and CPPR-136 permitting the use of the Exemption in construction of Seabrook Station, Units 1 and 2 do not involve a significant increase in the probability or consequences of accidents previously considered, do not create the possibility of an accident of a type different from any evaluated previously, do not involve a significant decrease in a safety margin, and thus do no involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be in compliance with the Commission's regulations, and the issuance of the amendments will r.ot be inimical to the common defense and security, or to the health and safety of the public.

Date of Issuance: JUN 101986 4

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provided by protective devices against the dynamic effects resulting from the discharge from postulated breaks in the primary piping is unnecessary.

Additionally, consideration of such dynamic effects associated with previously postulated pipe breaks is unnecessary. With full protection against dynamic effects provided by advanced analysis techniques, and based on the considerations discussed'above, we conclude that:

(1) the proposed amendments to Construction Permits permits CPPR_135 and CPPR-136 permitting the use of the Exemption in construction of Seabrook Station, Units 1 and 2 does not involve a significant increase in the probability or consequences of accidents previously considered, do not create the possibility of an accident of a type different from any evaluated previously, do not involve a significant decrease in a safety margin, and thus do no involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense and security, or to the health and safety of the public.

Date of Issuance: JUN 101986 e

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amendments to Construction Permits permits CPPR-135 and CPPR-136 permitting the use of the Exemption in construction of Seabrook Station, Units I and 2 does not involve a significant increase in the probability or consequences of accidents previously considered,li>xnot create the possibility of an accident of a type different from any evaluatedsp're'viously, do not involve a significant decrease in a safety margin, apd thus do no involve a significant hazards consideration; (2) there is-reasonable assurance that the health and safety of the public will not be-e'ndangered bf' operation in the proposed manner; and (3) such activities'will be in compliance with the Commission's regulations, and the issuance of the amendments will~not be inimical to the common defense and security, or to the health and safety of the public.

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T l.IST.0F REFERENCES (1) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WCAP-9558, Rev. 2, May 1981, Westinghouse Class 2 proprietary.

(2) NRC Generic f.etter 84-04, " Safety Evaluation of West'inghouse _ Topical Reports Dealing with Elimination of Postulated Breaks in PWR. Primary Main I. oops," February 1,1984. -

(3) Westinghouse Report WCAP-10456, "The Effects of Thermal Aging on _the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983, Westinghouse Class 2 proprietary.

(4) Westinghouse Report WCAD-10567, " Technical Bases for Elininating 1.arge Primary I.oop Pipe Ruptures as the Structural Design Basis for Seabrook Station, Units 1 and 2," June 1984, Westinghouse Class 2 proprietary.

I.IST 0F REFERENCES (1) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WCAP-9558, Rev. 2, May 1981, kestinghouse Class 2 proprietary.

(2) NRC Generic letter 84-04, " Safety Evaluation of Westinghouse Tcpical Reports Dealing with Elimination of Postulated Breaks in PWR Primary Main't. cops," February 1, 1984.

(31 Westinghouse Report WCAP-10456, "The Effects of Thermal Aging on the Structural -Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983, Westinghouse Class 2 proprietary.

(41 Westinghouse Report WCAD-10567, " Technical Bases for Eliminating large Primary 1.oop Pipe Ruptures as the Structural Design Basis for Seabrook Station, Units'I and 2," June 1984, Westinghouse Class 2 proprietary.

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