ML20052G687
| ML20052G687 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/05/1982 |
| From: | James Shea Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| Shared Package | |
| ML20052G688 | List: |
| References | |
| TASK-03-01, TASK-3-1, TASK-RR LSO5-82-05-002, LSO5-82-5-2, NUDOCS 8205180572 | |
| Download: ML20052G687 (6) | |
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MAy 5,1982 4
A Docket No. 50-245 LS05-82-05-002 Mr. W. G. Counsil Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 061 01
Dear Mr. Counsil:
SU8 JECT: SEP TOPIC III-1, QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS - MILLSTONE NUCLEAR POWER STATION UNIT 1 Enclosed is the staff's draft safety evaluation of SEP Topic III-l for the Millstone Nuclear Power Station Unit 1.
Ourevaluation(Enclosure
- 1) is based upon our contractor's final evaluation (Enclosure 2) of this topic. This assessment compares 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 confirming that the facts are correct, or by identi-4 fying errors and supplying the correct information.'
i The staff was unable to complete this topic due to the lack of informa-tion of original design requirements for various components. We have concluded, for those components where a comparison of codes was possi-ble that the changes in the codes since the original design, do not
[Of significantly affect the safety of the plant. 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.
/
Your response to requested within 30 days of receipt of this evaluation. D"^ M I
If no response is received in this time we will assume the evaluation is correc:t.
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- c. sthy Sincerely, g
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7-D. Pggi.Wric Original aigneW.
m James J. Shea, Project Manager 8205180 f'.
D Operating 1.eactors Branch No. 5
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Division of Licensing i
Enclosures:
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b Millstone' Unit 1 Docket No. 50-245 Revised 3/30/82 Mr. W. G. Counsil CC William H. Cuddy, Esquire State of Connecticut Day, Berry & Howard Office of Policy & Management Counselors at Law ATTN:
Under Secretary Energy One Constitution Plaza Division Hartford, Connecticut 06103 80 Washington Street Hartford, Connecticut 06115 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission Region I Office 631 Park Avenue King' of Prussia, Pennsylvania. 19406 Northeast Nuclear Energy Company ATTN:
Superintendent 1
Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC P. O. Box Drawer KK Niantic, Connecticut 06357 First-Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385 John F. Opeka Systems Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203
o SYSTEMATIC EVALUATION PROGRAM TOPIC III-l MILLSTONE MUCLEAR POWER STATION UNIT 1 TOPIC:
III-1, Classification of Structures, Systems and Components (Seismic and Quality) 1.
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 gener-ally recognized codes, standards and criteria in effect at that time; however, the codes, stand 3rds and criteria have been periodically re-vised. 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 6re presented in Appendix A of Technical-Evaluation Report C5257-432, " Quality Group Classification of Components and Systems - Millstone Nuclear Power Station Unit 1," pre-pared 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 as-pects deleted in Topic III-l are identified below.
III-6 Seismic Design Considerations III-7.B Design Codes, Design Crtieria, Load Combinations and Reactor Cavity Design Criteria V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity XVII Operational Quality Assurance Program The resolution of Topic V-8 is part of Unresolved 3afety Issues A-3, A-4 and A-5.
-2.
IV.
REVIEW GUIDELINES The review guidelines are presented in Section 3 of Report C5257-432 (attached).
V.
EVALUATION The basic input for this report is Table 4.1 in Section 4 of Report C5257-432. 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 iden-tified 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 C5257-432 with a detailed discussion 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:
j 1.
Fracture Toughness - The current code requires that pressure retain-ing materials be impact tested.
For 4 of 66 components reviewed, sufficient information was available to exempt them from this require-ment.
2.
Radiography Requirements - We have determined that:
a.
In general, past full radiograpy requirenents for pressure vessels were more conservative than current requirements, with the excep-tion of Category C welds of currently classified ASME Code Class 2 vessels designed to Class C or Section VIII requirements.
b.
Full radiography requirements for Class 1 or Class 2 welded joints in piping valves and pumps were not required in the past codes. However, if provisions 2 and 3 of Code Case N-7 were applied in the construction of these components, then the current radiography requirements were met.
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 62 of 66 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 C5257-432.
Table 5-1 of the Report identifies those components which this infor-mation is necessary.
2.
Radiography Requirements - The licensee should provide the following information:
a.
Radiography requirements imposed on Class 2 pressure vessels.
b.
Radiography requirements imposed on all Category welds of Class 1 and 2 piping and valves.
c.
Radiography requirenents imposed on Class 1 and 2 pumps.
d.
Piping and valves for which ASME Code Section I (1965) was invoked, confirm that the full radiography requirements of t at code were impl emented.
3.
Valves - Provide, on a sample basis for Class 1, 2 and 3 valves, in-formation regarding the design of the valve in order to evaluate if they meet current body shape.and pressure-tenperature rating require-ments.
4.
Pumps - Pumps designed to standards other than ASME Code Sections III or VIII (1965), should be demonstrated to meet current fatigue analysis requirements.
5.
Storage Tanks - Provide the following:
a.
Confirm that the atmospheric storage tanks meet current compressive stress requirements.
I b.
Confirm that the 0 to 15 psig storage tanks meet current tensile l
allowables for biaxial stress field conditions.
c.
Specifications for tanks built to codes other than ASME Section III l
or VIII (1965).
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- 6.
Missing Information - The following information, which is incomplete or missing from Table 4-1 or Tables 4-2(a), (b) and (c) of Report C5257-432 (attached), should be provided:
a.
Information for 19 components (out of 76) regarding the code, code class (if available), and code edition used in their de-sign (Table 4-2),
b.
Response to questions raised by notes in Table 4-1 c.
Any specifications or calculations used in designing pumps, valves, and tanks that may assist in conducting this evaluation d.
Information on the code, code class (if applicable), and code edition used in the design of the feedwater heat exchanger (shell side) c.
The assumption of 100*F temperature drop from 100% to 0% power that was used in the usage factor calculation for fatigue load-ing should be confirmed.
A more detailed esplanation of the information to be provided may be found in Report C5257-432 (attached).
VI.
CONCLUSION We have determined that for the following, changes. between current and original code requirements for the' Millstone Nuclear Power Station Unit I will not significantly affect the safety functions of the systems and components 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 information is fiscussed 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.