ML20049A218
| ML20049A218 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/11/1979 |
| From: | Moseley N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20033C447 | List: |
| References | |
| FOIA-82-176 NUDOCS 7911190277 | |
| Download: ML20049A218 (2) | |
Text
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jo UNITED STATES g
8 NUCLEAR REGULATORY COMMISS!ON
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SEP 111979
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f,7,j,,w[.u. /k SSINS 0410 Docket Nos. 50-361/362
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y MEMORANDUM FOR:
D. Vassallo, Acting Director, Division of Project Management, NRR D. Eisenhut, Acting Director, Division of Operating Reactors, NRR FROM:
Norman C. Moseley, Director, Division of Reactor o
Operations Inspection, IE
SUBJECT:
APPLICABILITY OF 10 CFR 50, APPENDIX A & B REQUIREMENTS TO ALL SAFEIY RELATED EQUIPMENT Enclosed for your information and for possible action is a memorandum from our Region V office which identifies several concerns regarding the applicability of 10 CFR 50, Appendix A and B to systems, subsystems and components that perform functions important to safety. The Appendix B concerns were identified during our inspector's review of the Southern California Edison Company's preoperational test program for San Onofre Unit 2.
However, the concerns appear to have broad applicability to the NRR approved QA program for construction as described in the last paragraph of the enclosed memorandum.
It appears that these concerns may also have generic applicability at other operating facilities and facilities under construction. This latter premise is based, in part, on discussions between B. Faulkenberry, RV, and H. Rood, NRR Project Manager for San -
Onofre 2 & 3, which indicate that the scope of the QA program reflected in the San Onofre 2 & 3 FSAR is consistent with NRR approved QA programs for similar facilities.
The specific concern resulted when a licensee representative informed our inspector that the pressurizer pressure and level control systems for San Onofre 2 and 3 are non-safety related systems and, therefore, not subject to the NRR approved QA program for operation. The licensee representative stated that the QA classification of structures, systems and components is identified in the Final Safety Analysis Report and that Quality Class I and II items are covered by 10 CFR 50, Appendix B.
We note that 10 CFR 50, Appendix B, was developed to assure that structures, systems and components that prevent or mitigate the consequences of 4
postulated accidents that could cause undue risk to the health and safety of the public are covered by the licensee's QA program. Table 3.2-1 of the licensee's FSAR identifies a number of systems as QA Class III or IV which appear to perform safety related functions as described above, and, based on the TMI accident, should be reclassified as QA Class I or II.
CONTACT:
F. J. Nolan, IE (x28019) 41/// 9 of 9 7
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NUCLEAR REGULATORY COMMISSION UWTED STATES M'
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REGION V 1990 N. CALIFO RNI A BOULEVARD o'
SulTE 202, WALNUT CREEK PLAZA WALNUT CREEK. CALIFO RNI A 94596 o
. August 29, 1979 l
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J. L. Crews, Chief, Reactor Operations and Nuclear Support MEMORANDUM FOR:
Branch B. Faulkenberry, Chief, Reactor Operations and Nuclear THRU:
Iq Support Section 2 t
i A. D. Johnson, Reactor Inspector l
FROM:
10 CFR 50, Appendix "A" AND."B" REQUIREMENTS
SUBJECT:
SOUTHERN CALIFORNIA EDISON COMPA!!Y, SAN ONOFRE UNIT RE:
N0. 2, DOCKET NO. 50-361 During a routine inspection of the preoperational test program at the referenced facility, I was informed by licensee representatives that the pressurizer pressure and level control systems for San Onofre Units 2 and 3 are non-safety related systems and, therefore, not subject to the li-censee's QA program established pursuant to the requirements of 10 CFR 50, Appendix "B."
The licensee representatives explained that structures, systems and, components have been classified as described in Section 3.2.3.1 of the Final Safety Further, Section 3.2.3.2 of Analysis Report (FSAR), pages 3.2-3 and 3.2-4.
the FSAR reads, in part, "For Quality Class I & II items, the applicable re-I quirements of 10 CFR 50, Appendix "8," Quality Assurance Criteria for Nuclear Power Plants...have been met to ensure the highest quality standards."
The licensee representatives stated and showed me in the project Quality Assurance Manual that the program only applies to items classified as quality class I and II and that, therefore, items classified as quality class III and IV were neither subject to 10 CFR 50, Appendix B requirements, nor the licensee's
-QA program submitted to and approved by NRC as a topical report.
For example, Table 3.2-1 of the FSAR provides equipment QA classifications.
on sheets 11 and 12, the following control systems are described as not required for safety and classified as indicated below.
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L J. L. Crews August 29, 1979 2
i QA Classification Systen.
III Reactor Regulating System Pressurizer Pressure Control III Pressurizer Level Control III IV Plant Computer System y
Core Operating Limit Supervisory System IV III Steam Bypass Control System III Baron Control System In-Core Instrumentation III l
(Startup and Control Channels) 2 III Ex-Core Instrumentation System Refueling Interlocks.
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III Feedwater. Control System i
Main Turbine Contols (Overspeed)
III f
l 10 CFR 50, Appendix B, Introduction, reads, in part, " Nuclear Power Plants
... include structures, systems and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. This appendix establishes quality assur-l ance requirements for the design construction and operation of those structures, systems End components. The pertinent requirements of this appendix apply to all activities affecting the safetyrelated functions of those structures, l
systems and components."
(Emphasis added.)
I Criterion 13,10 CFR 50, Appendix A, Instrumentation and Control, reads "Instrumentatier shall be provided to monitor variables and systems over l
l their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate
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safety, including those variables and systems that can affect the' fission process, the integrity of the reactor core, the reactor coolant pressure
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Appropriate con-boundary, and the containment and its associated systems.
trols shall be provided co maintain these variables and systems within prescribed coetating ranges."
Section A, paragraph 2 of the Regulatory Guide 1.68, reads, "Section XI, I
' Test Ccntrol,' of Appendix 9, ' Quality Assurance Criteria for Nuclear Pei""-
Plants and Fuel Reprocessing Plants,' to 10 CFR Part 50 requires that a test program be established to ensure that structures, systems, and components Since all functions designated in will perform satisfactorily in service.
the general design criteria (DGC) are important to safety, all structures, l
systems, and components requ. ired to perform these functions need to be tested to ensure that they will perform properly. These functions, as noted through-out the specific GDC, are those necessary to ensure that specified design conditions of the facility are not exceeded during any condition of normal i
operations, including anticipated operational occurrences, or as a result J
i f of postulated accident conditions." (Emphasis added.)
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Regulatory Guide-l.33 clearly indicates the systems classified above by the licensee as nonsafety related are incorrect and that the systems are safety related and subject to 10 CFR 50, Appendix B requirements during the operation of nuclear power plants, s
Regulatory Guide 1.26 fails to specifically address the issue of whether or not structures, systems and components important to safety or otherwise I
related to safety are subject to 10 CFR 50, Appendix B requirements, but rather relies on Criterion I of Appendix A as the basis to require that those items be designed, fabricated, erected, and tested to quality stan-d dards commensurate with the importance of the safety functions to be per-However, the licensee has not interpreted this to mean that the formed.
activities be subject to the requirements of 10 CFR 50, Appendix B, and so long as the items are designed, fabricated, erected and tested to prescribed codes and standards (even though not subject to a NRC required Quality Assurance program), NRC requirements are satisfied.
I am of the opinion that structures, systems and components subject to the requirements of 10 CFR 50, Appendix A, are important to safety or have other As such, all of those structures, systems and safety related significance.
components are safety related and subject to 10 CFR 50, Appendix B require-ments as specified in the introduction of the Appendix as quoted above.
Now therefore, I recommend that the Office of Inspection and Enforcement promulgate and enforce a policy which establishes that any and all structures, systems and components subject to 10 CFR 50, Appendix "A" be subject. to the 1
requirements of 10 CFR 50, Appendix B.
Should this be the policy of the NRC, as indicated in Regulatory. Guide 1.33, then IE:V should so inform the licensee that the activities to be conducted during preoperational testing of such structures, systems and components, of which the pressurizer pressure and level control systems are examples, must be conducted in accordance with their NRC approved Quality Assurance program.
Further, I recommend that the items classified by the licensee as III and IV be evaluated by NRC to determine whether or not the items have been constructed in such a mannr.r as to assure their safety during operatice of the nuclear power plant.
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