ML20244B841

Forwards Matls Engineering Branch Input to Ser.Addl Applicant Info Required for Section 5.3.1 Re Reactor Vessel Matls

ML20244B841
Person / Time
Site:	Seabrook, 05000000
Issue date:	01/25/1982
From:	Johnston W Office of Nuclear Reactor Regulation
To:	Tedesco R Office of Nuclear Reactor Regulation
Shared Package
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References
FOIA-87-51 NUDOCS 8202100084
Download: ML20244B841 (9)

Text

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V xy-Docket No. 50-443/444 MEMORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing Division of Licensing FROM: William V. Johnston, Assistant Director Materials &

Qualification Engineering Division of Engineering

SUBJECT:

SEABROOK NUCLEAR STATION UNITS 1 AND 2, PUBLIC SERVICE CO. OF NEW HAMPSHIRE PLANT NAME: Seabrook Nuclear Station Units 1 and 2 SUPPLIES: Westinghouse; United Engineering Construction DOCKET NUMBERS: 50-443/444 RESPONSIBLE BRANCH & PROJECT MANAGER: LB 3; L. Wheeler REVIEWER: B. J. Elliot, INEL l DESCRIPTION OF TASK: Safety Evaluation Report Input for Sections 5.3.1, 5.3.2 and 5.3.3 REVIEW STATUS: Section 5.3.1 - Additional Confirmatory Information Required Section 5.3.2 - Complete i Section 5.3.3 - Complete The Component Integrity Section, Materials Engineering Branch, Division of Engineering, has reviewed the Final Safety Analysis Report (FSAR) for Seabrook Nuclear Station Units 1 and 2. Based on our review of the information in FSAR amendments through no. 43 and Enclosure 1 to PSNH letter dated November 27, 1981, we have prepared our input to Safety Evaluation Report (SER) Sect-ions 5.3.1, 5.3.2, and 5.3.3 which is attached. In this SER we have granted

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an exemption to one paragraph in Appendix G,10 CFR Part 50. t i We have reviewed the Seabrook Nuclear Facilities in accordance with the Standard Review Plan (SRP) in NUREG-0800, Rev.1 as indicated in this SER.

1 The review status of each section of the Seabrook SER are identified above.

The additional confirmatory information required to complete our review of Seabrook Units 1 and 2 were identified in our previous draft SER and questions.

The applicant's response to those questions was that they will respond at a l later date. When the applicant provides the information requested, we will confirm the applicant has complied with the regulatory requirements. -

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' William V. Johnston, Assistant Director

' # Materials & Qualification Engineering

}()[{()(,C[ Division of Engineering

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Robert L. Tedesco 2 JAN 2 5 1982 1

Enclosure:

As stated

Contact:

B. J. Elliot X-28083 cc: D. G. Eisenhut L. Wheeler '

R. H. Vollmer W. S. Hazelton W. V. Johnston R. W. Klecker F. J. Miraglia B. J. Elliot S. S. Pawlicki P. K. Nagata (INEL)

E. Sullivan Distribution:

Docket File MTEB Reading File MTEB RE 1-1 Seabrook i

i Concurrences: [ s y DE:MTEB898 DE: E DE: TB DE:AD/MQE BElliot; jar RKl ker SP licki WVJohnston 1//r/82 1/g/82 1/' /82 1/pt/82

l ATTACHMENT 1 SEABROOK STATION UNITS 1 AND 2 SAFETY EVALUATION REPORT MATERIALS ENGINEERING BRANCH COMPONENT INTEGRITY SECTION l

5. 3.1 Reactor Vessel Materials and 5.2.3 Reactor Coolant Pressure Boundary Material The staff of EG&G, Idaho National Engineering Laboratory has reviewed the fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, and the materials surveillance program for the reactor vessel beltline. The acceptance criteria and references which are the basis for this evaluation are set forth in paragraph II.3.a of Standard Review Plan l (SRP) Section 5.2.3 and paragraph 11.5, II.6 and II.7 (Appendices G and H, i 10 CFR Part 50) of SRP Section 5.3.1 in NUREG 0800 Rev.1 dated July 1981.

A discussion of this rev'.,ew follows.

General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," Appendix A,10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, maintenance, and test conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized. General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary," Appendix A,10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed to permit an appropriate saterial surveillance program for the reactor pressure vessel.

The fracture toughness requirements for the ferritic materials of the reactor coolant pressure boundary are defined in Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Requirements" of 10 CFR Part 50.

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We have reviewed the materials selection, toughness requirements, and extent of materials testing conducted by the applicant to provide assurance that the ferritic materials used for pressure retaining components of the reactor coolant pressure boundary possess adequate toughness under operating, maintenance, and testing conditions.

The Seabrook Station Units 1 & 2 (hereinafter SB 1 & 2) reactor vessel, piping, pumps, valves, and steam generator were designed to the 1971 Edition of the ASME Boiler and Pressure Vessel Code (hereinafter "the Code") Section III:

" Rules for Construction of Nuclear Power Plant Components" including Addenda through Winter 1972. Based on the July 7,1976, construction permit date, Section 50.55a, " Codes and Standards," 10 CFR Part 50 requires that the ASME Code Edition and Addenda applied to the above reactor coolant system components be no earlier than those of the 1971 Edition, Summer 1972 Addenda. The design and construction of the RCPB (reactor coolant pressure boundary) components of SB 1 & 2 are, therefore, in compliance with the requirements of Section 50.55a, 10 CFR Part 50.

Compliance with Appendix G,10 CFR Part 50 Dur evaluation of the SB 1 & 2 FSAR to determine the degree of compliance with the fracture toughness requirements of Appendix G,10 CFR Part 50, indicates that the applicant meets al? the requirements of this appendix.

Compliance with Appendix H,10 CFR Part 50 The materials surveillance program at SB 1 & 2 will be used to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region, resulting from exposure to neutron irradiation and the thermal environment as required by General Design Criterion 32, " Inspection of l Reactor Coolant Pressure Boundary." Under the SB 1 & 2 surveillance program, fracture toughness data must be obtained from material specimens that are representative of the limiting base, weld, and heat-affected zone materials in the beltline region. These data will permit the determination of the con- t ditions under which the vessel can be operated with adequate margins of safety l

against fracture throughout its service life.

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l The fracture toughness properties of reactor vessel beltline materials must be monitored throughout the service life of SB 1 & 2 by a materials surveillance program that meets the requirements of Appendix H of 10 CFR Part 50.

We have evaluated the a,71icant's information for degree of compliance to ,

these requirements. Bases on our evaluation we conclude that the applicant has met all the requirements of Appendix H,10 CFR Part 50 with the exception of Paragraph II.B.

Paragraph II.B, Appendix H, to CFR, Part 50, requires that reactor vessels constructed of ferritic materials, with a peak neutron fluence (E > 1MeV) at 2

the end of the design life of the vessel exceeding 10 17 n/ cm , shall have their beltline regions monitored by a surveillance program complying with the specifications of ASTM Standard E185-73 except as modified by Appendix H.

ASTM Standard E 185 requires that material placed in the surveillance capsules represent the material that may limit operation of the reactor during its lifetime. The selection of the base metal, heat affected zone (HAZ), and the weld metal surveillance specimens is based on consideration of the fracture toughness properties of all of the beltline materials (RTNDT and upper-shelf.

CVN energy) in the unirradiated condition, the chemical composition and the r,eutron fluence to establish the limiting materials. The applicant has indi-cated that SB 1&2 will comply with all the requirements of Appendix H,10 FR Part 50, but has not submitted sufficient information for us to confirm that the reactor vessel materials surveillance program complies with the requirements of Appendix H,10 CFR Part 50. The applicant has indicated that they will provide the following confirmatory information:

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a. Identification of all materials in each surveillance capsule.

b. The lead factors for each capsule.

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c. The azimuthal locations of each surveillance capsule.  !

d. The withdrawal schedule for the capsule.

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When the applicant provides the above data, we will confirm that the applicant has complied with the requirements of Appendix H,10 CFR Part 50.

1 Conclusions for Compliance with Appendices G and H,10 CFR Part 50 Based on our evaluation of compliance with Appendices G and H,10 CFR Part 50, we conclude that the applicant has met all the fracture toughness requirements of these Appendices except for Paragraph 11.8 of Appendix H. The applicant has indicated that in 1983 and 1984 he will provide the data for units 1 and 2 respectively to confirm that these units comply with the requirements of Paragraphs ll.B of Appendix H,10 CFR Part 50.

Appendix G, " Protection Against Non-Ductile Failure,"Section III of the ASME Code, will be used, together with the fracture toughness test results required by Appendices G and H,10 CFR Part 50, to calculate the pressure-temperature limitations for the SB 1 & 2 reactor vessel.

The fracture toughness tests required by the ASME Code and by Appendix G of 10 CFR Part 50 provide reasonable assurance that adequate safety margins  ;

against the possibility of non-ductile behavior or rapidly propagating fracture can be established for all pressure ret,aining components of the  !

reactor coolant boundary. The use of Appendix G,Section III of the ASME I

Code, as a guide in establishing safe operating procedures, and use of the results of the fracture toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adequate safety margins during operating, testing, maintenance, and anticipated transient conditions.

Compliance with these Code provisions and NRC regulations constitutes an acceptable basis for satisfying the requirements of General Design Criterion 31.

The materials surveillance program, required by Appendix H,10 CFR Part 50, will provide information on material properties and the effects of irradiation of material properties so that changes in the fracture toughness of the material in the SB 1 & 2 reactor vessel beltline caused by exposure to neutron radiation can be properly assessed, and adequate safety margins against the possibility of vessel failure can be provided.

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Compliance with Appendix H,10 CFR Part 50 assures that the surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness of the reactor vessel material and satisfies the requirements of General Design Criteria 32.

5.3.2 Pressure Temperature Limits The staff of EG&G, Idaho National Engineering Laboratory has reviewed the applicant's pressure temperature limits for operation of their reactor vessels.

The acceptance criteria and list of references which are the basis for this evaluation are set forth in the Standard Review Plan (SRP) Section 5.3.2 of NUREG 0800 Rev. I dated July 1981. A discussion of this review follows.

Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR Part 50, describe the conditions that require pressure-temperature limits and provide the general bases for these limits. These appendices specifically require that pressure-temperature limits must provide safety margins at least as great as I those recommended in the ASME Code,Section III, Appendix G, " Protection Against Non-Ductile Failure." Appendix G,10 CFR Part 50, requires additional i safety margins whenever the reactor core is critical, except for low-level I physics tests.

The following pressure-temperature limits imposed on the reactor coolant l pressure boundary during operation and tests are reviewed to ensure that they provide adequate safety margins against non-ductile behavior or rapidly pro-pagating failure of ferritic components, as required by General Design l

Criterion 31:

a) Preservice hydrostatic tests, b) Inservice leak and hydrostatic tests, l c) Heatup and cooldown operations, and 1

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d) Core operation. l

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The applicant has presented (Figures 3.4-2 and 3.4-3 of the Technical Specifications) heat up and cooldown pressure-temerature curves for 3.9 )

effective full power years (EFPY). We have evaluated the curves and have )

found them acceptable for at least 3.9 EFPY. The applicant has indicated that I new curves will be presented in 1982. We will evaluate them at that time to determine whether they conform to the requirements of Appendix G,10 CFR Part 50.

I The pressure-temperature limits to be imposed on the reactor coolant system j i

for all operating and testing conditions, to ensure adequate safety margins against non-ductile or rapidly propagating failure, must be in conformance with established criteria, codes, and standards to be acceptable to the staff.

The use of operating limits based on these criteria, as defined by applicable regulations, codes, and standards, will provide reasonable assurance that non-ductile or rapidly propagating failure will not occur, and will constitute an acceptable basis for satisfying the applicable requirements of General Design Criterion 31.

5.3.3 Reactor Vessel Integrity We have reviewed the FSAR sections related to the reactor v tssel integrity of I SB 1 & 2. Although most areas are reviewed separately in accordance with i

other review plans, reactor vessel integrity is of such importance that a special summary review of all factors relating to reactor vessel integrity is warranted.

The staff of EG&G, Idaho National Engineering Laboratory has reviewed the fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, the pressure temperature limits for operation of the reactor vessels, and the materials ., surveillance program for the reactor vessel beltline. The acceptance criteria and references which are the basis for the evaluation are set forth in paragraphs 11.2, II.6 and II.7 (Appendices G and H,10 CFR Part t0) of Standard Review Plan (SRP) Section 5.3.3 in NUREG 0800 Rev. I dated July 1981. A discussion of this review follows.

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l We have reviewed the above factors contributing to the structural integrity of the SB 1 & 2 reactor vessel and conclude that the gplicant has fully complied with the required regulations, codes and standards except for Paragraph II.B of Appendix H.

The applicant has indicated that SB 1&2 will comply with all the requirements i of Appendix H, 10 CFR Part 50, but has not submitted sufficient information for us to confirm that the reactor vessel materials surveillance program complies with the requirements of Appendix H, 10 CFR Part 50.

We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude there are no special considerations that make it necessary to consider potential reactor vessel failure for Seabrook Units 1 ,

and 2.

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