ML20054C495

Forwards Draft Safety Evaluation & Technical Evaluation Rept of SEP Topic III-1,quality Group Classification of Components & Sys.Requests Licensee to Examine Facts Upon Which Staff Based Evaluation

ML20054C495
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/16/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Vandewalle E CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
Shared Package
ML20054C496	List: ML20054C495
References
TASK-03-01, TASK-3-1, TASK-RR LSO5-82-04-050, LSO5-82-4-50, NUDOCS 8204210190
Download: ML20054C495 (8)

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Docket No. 50-155 Q

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Mr. David J. VandeWalle 8

Nuclear Licensing Administrator y

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Consumers Power Company 1945 W Parnall Road cv N

Jackson, Michigan 49201

SUBJECT:

SEP TOPIC III-1, QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS (BIG ROCK POINT NUCLEAR PLANT)

Enclosed is the staff's draft safety evaluation of SEP Topic III-l for the Big Rock Point Nuclear Plant. Our evaluation (Enclosure

1) is based upon our contractor's final evaluation (Enclosure 2) of this topic. This assessment com ares your facility with the criteria currently used for licensing new facilities.

You are requested to examine the facts upon which the staff has based its evaluation and respond either by confiming that the facts are i

correct, or by identifying errors and supplying the correct 5 go 5

infomation.

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The staff was unable to complete this topic due to the lack of

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3 infomation of or final design requirements for various components.

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We have concluded, for those components where a comparison of codes was possible that the changes in the codes since the original

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design do not significantly affect the safety of the plant.

Baseo on our sa@ ling of code comparisons to date we do not expect the remaining items to pose a significant hazard to safe plant operation.

Your response is requested within 30 days of receipt of this evalua-tion.

If no response is received in this time we will assume the evaluation is cori ect.

Sincerely, DC tc field 4//(/82 AD. A:DL Dennis M. Crutchfield, Chief GL inas Operating Reactors Branch No. 5 Division of Licensing 4/{/82 h6 m

l Enclosure's: As stated!

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.Mr. David J. VandeWalle cc Mr. Paul A. Pe'rry, Secretary U. S. Environmental Protection Consumers Power Company Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN:

Regional Radiation Representative Judd L. Bacon, Esquire 230 South Dearborn Street

- Consumers Power Company Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Peter B. Bloch, Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission Isham, Lincoln & Beale Washington, D. C.

20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.

20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.

20555 505 Peoples Building Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Commission Atomic Safety & Licensing Appeal Board Washington, D. C.

20555 U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Big Rock Point Nuclear Power Plant ATTN:

Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44' Charlevoix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Mr. Jim E. Mills Route 2, Box 108C Route 2, Box 108C Charlevoix, Michigan 49720 Charlevoix, Michigan 49720 William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard County Board of Supervisors Ann Arbor, Michigan 48103 Charlevoix County l

Charlevoix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC Room 1 - Capitol Building RR #3, Box 600 9

Lansing, Michigan 48913 Charlevoix, Michigan 49720 Herbert Semmel Counsel for Christa Maria, et al.

Urban Law Institute Antioch School of Law 263316th Street, NW Mcshington, D. C.

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s Mr. David J. VandeWalle cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ms. JoAnn Bier 204 Clinton Street Charlevoix, Michigan 49720 Thomas S. Moore Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission

. Washington, D. C.

20555 James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 9

L SYSTEMATIC EVALUATION PROGRAM TOPIC III-l BIG ROCK POINT NUCLEAR PLANT TOPIC:

III-1, CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS (SEISMIC AND QUALITY)

I.

INTRODUCTION SEP plants were generally designed and constructed during the time span from the 1950's to the late 1960's. The plants were designed to generally recognized codes, standards and criteria in effect at that time; however, the codes, standards and criteria have been periodically revised.

Therefore, the SEP plants may have been designed and constructed to codes, standards and criteria no longer in effect or acceptable to the NRC.

The purpose of Topic III-l is the review of the classification of structures, systems and components of as-built plants compared to-the current classifications required for seismic and quality groups in the codes, standards and criteria.

Since the review of seismic classifica-tions is addressed in other SEP topics (See Section III of this evalua-tion), this topic has been limited to the evaluation of quality group classifications.

II.

REVIEW CRITERIA The review criteria for this topic are presented in Appendix A of Technical Evaluation Report C5257-434, " Quality Group Classification of Components and Systems - Big Rock Point Nuclear Plant," prepared for the NRC by Franklin Research Center (attached).

III.

RELATED SAFETY TOPICS The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review are performed in related topics.

As stated previously, the seismic aspect of this topic has been deleted.

The quality aspect for the reactor vessel and steam generators (PWRs only) and the quality assurance have been deleted.

The related safety topics, and the subject matter covered in the topics, that cover the aspects deleted in Topic III-1 are identified below.

III-6 Seismic Design Considerations III-7.8 Design Codes, Design Criteria, Load Combinations and Reactor Cavity Design Criteria V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity XVII Optrational Quality Assurance Program The resolution of Topic V-8 is part of Unresolved Safety Issues A-3, A-4 and A-5.

IV.

REVIEW GUIDELINES The review guidelines are presented in Section 3 of Report C5257-434 (attached).

V.

EVALUATION The basic input for this report is Table 4.1 in Section 4 of Report C5257-434.

Table 4.1 is a compilation of all systems and components which are required to be classified by Regulatory Guide 1.26 and the original codes, standards and criteria used in the plant design.

After comparing the original codes, standards and criteria with those currently used for licensing facilities the following areas were identified where the requirements have changed:

1) Fracture Toughness

2) Quality Group Classification

3) Code Stress Limits

4) Radiography Requirements

5) Fatigue Analysis of Piping Systems An evaluation of each of these areas is presented in Section 5 of Report C5257-434 with a detailed discussion included in the Appendix of the report.

We have determined that changes in the following areas have not signi-ficantly affected the safety functions of the systems and components reviewed in this report:

1) Quality Group Classification 2)

Code Stress Limits

3) Fatigue Analysis of Piping Systems In the remaining two areas we have concluded the following:

1) Fracture Toughness - The current code requires that pressure retaining materials be impact tested.

For 4 of 71 components reviewed, sufficient I

information was available to exempt them from this requirement.

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2) Radiography Requirements - We have determined that:

a) Vessels built to ASME VIII (1959) and currently classified as l

Class 1 vessels meet the current radiography requirements if Code Cases 1270N and 1273N were invoked.

b) Vessels built to ASME VIII (1959) and currently classified as Class 2 or 3 vessels meet the current radiography requirements for weld joint thicknesses greater than 1-1/2 inches if Code Cases 1270N and 1273N were invoked.

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. c) Piping built to ASA B31.1 (1955) and Code Cases N-1, N-7, N-9 and N-10; piping built to ASME I (1959) or USAS B31.7 (1968);

or piping built to Code Case N-7, N-9 or N-10 meets the current full radiography requirements.

Our review has not identified any significant deviations from past codes.

However, we were unable to complete our evaluation due to insufficient information for the following:

1) Fracture Toughness _- For 67 of 71 components there is insufficient information on materials to complete our review.

The licensee should provide the necessary information using the format provided in Tables A4-4 through A4-6 in Appendix A of Report C5257-434.

Table 5-1 of the Report identifies those components for which this information is necessary.

2) Radiography Requirements - The licensee should provide the following information:

a) Radiography requirements imposed on Class 2 and 3 vessels for which Code Case 1273N was not invoked and with welded joint thicknesses less than 1-1/2 inches.

b) Code edition and Class for components designed to ASME Section III.

c) Radiography requirements for piping and valves designed only to ASA B31.1 (1955).

d) Radiography requirements for all Class 1 and 2 pumps.

e) Confirm that radiography requirements were inforced for safety relief valves designed to ASME Section I.

f) Radiography requirements for Class 1 vessels for which Code Case 1270N or 1273N were not invoked.

Valves - Provide, on a sample basis for Class 1, 2 and 3 valves, in-l 3) information regarding the design of the valve in order to evaluate if t

they meet current body shape and pressure -temperature rating require-ments.

Pumps _- Provide the codes of standards or manufacturers specifica-4) tions used for designing the Core Spray System, Fire Protection System, Reactor Shutdown Cooling System and Reactor Cooling Water System pumps.

. 5) Storage Tanks - Provide the following:

a) Confirm that the atmospheric storage tanks meet current compressive stress requirements.

b) Confirm that the 0 to 15 PSIG storage tanks meet current tensile allowables for biaxial stress field conditions, c)

Specifications for tanks built to codes other than ASME Section VIII (1959).

d)

Information on the standards used in the design of the reactor cooling water tank and nitrogen bottles.

6) Piping - Calculations similar to those presented in Section 4.2, Appendix A of Report C5257-434 should be provided in order to assess the impact on the usage factor of gross discontinuities in Class 1 piping systems for a medium and large number of cyclic loads.

7) Pressure Vessels - Demonstration of compliance with current fatigue analysis requirements for all Class 1 vessels.

8) Missing Information - The following information, which is incomplete or missing from Table 4-1 or Tables 4-2(a), (b) and (c) of this report, should be provided:

a)

Information on codec, class and code cases used in the design of 7 out of 86 compor.ents (Table 4-2).

In addition, provide this information for the core spray spargers and the suction strainers (these two components were not listed in Tables 4-1 or 4-2 and are currently classified as Class 2).

b) Any specifications or calculations used in designing pumps, valves, and tanks that may assist in conducting this evaluation should be provided.

c) Confirm assumptions made regarding code editions (See Table 4-1).

A more detailed explanation of the information to be provided may be found in Report C5257-434 (attached).

VI.

CONCLUSION We have determined that for the following, changes between current and original code requirements for Big Rock Point will not significantly affect the safety functions of the systems and cpmponents reviewed:

1) Quality Group;

2) Code Stress and

3) Fatigue Analysis for Piping Systems.

. We were unable to complete our review due to insufficient information regarding various other systems and components.

The required informa-tion is discussed in Section V of this evaluation.

Based on our sampling of code comparisons to date we do not expect the remaining items to pose a significant hazard to safe plant operation and, therefore, have determined that the schedule and need for providing the remaining information can be determined during the integrated plant safety assessment.

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