ML20058F500
| ML20058F500 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/18/1989 |
| From: | Fischer D, Holahan G, Wigginton D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19311B205 | List: |
| References | |
| NUDOCS 9312080153 | |
| Download: ML20058F500 (7) | |
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4 generally initiated in a timely manner."
NRC Region I Inspection Report No. 50-423/91-15 (Millstone Unit 3) (August 8, 1991) at 14.
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Again, in 1992, the NRC noted that NU's reporting procedures were a "[p]erformance strength based on the low threshold for l
initiation, prompt performance of detailed, multidisciplinary
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management reviews, an effective tracking and trending system, and aggressive management of the program overall."
NRC Inspection Report No. 50-213/91-81 (Haddam Neck) (January 31, 1992)-(Executive
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Summary).
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Significantly, the NRC has had this to say specifically I
about Millstone Unit 1 in August 1992:
" Safe and conservative operational decisions and strong self assessment were demonstrated in [four enumerated instances]
In each of the above cases, 3
the licensee demonstrated conservative actions that resulted in voluntary reactor shutdowns or the extension of an ongoing outage."
i NRC Systematic Assessment of Licensee Performance (SALP) Final t
Report Nos. 50-245/90-99, 50-336/90-99 and 50-423/90-99 (August 4, 1992) at 28-29.
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The conservatism in NU's reportability process was i
acknowledged by the NRC also in an NRC inspection report for NU's l
t Haddam Neck
- plant, in which the NRC praised NU for its
" conservative approach to reporting and [NU's] identification'of design deficiencies, which reflects a strong safety ethic and the j
I effectiveness of the programs to identify, evaluate and report potential safety problems." NRC Inspection Report No. 50-213/92-26 (February 26, 1993) at 14.
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i 9312080153 931130 1-
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NU Actions Related to Personnel Considerations Have Also Adequately Addressed Any Underlying concerns That May Have Existed In early 1991, as a result of an in-depth review by the Company of the overall performance of the individual who made the questionable comment during the November 14, 1989 meeting, and, in particular, the individual's personal style in communicating and dealing with people, he was transferred to a pos'ition involving i
cost and scheduling at the corporate offices in
- Berlin, f
Thus, the Company has already taken appropriate personnel action in this case and no further purpose would be served by enforcement action.
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Millstone Nuclear Power Station, Unit No.1 l
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EQUIPMENT OPERABILITY PROMPTLY DETERMINING OPERABILITY AND ESTABLISHING CORRECTIVE ACTION PLANS FOR DEGRADED OR NONCONFORMING SAFETY EQUIPMENT BY GARY M. HOLAHAN, ACTING DIRECTOR DIVISION OF REACTOR PROJECTS - III, IV, Y AND SPECIAL PROJECTS 0FFICE OF NUCLEAR REACTOR REGULATION AND DAVID C. FISCHER AND DAVID L. WIGGINTON OFFICE OF NUCLEAR REACTOR REGULATION FOR PRESENTATION AT THE NRC REGULATORY INFORMATION CONFERENCE THE MAYFLOWER HOTEL i
WASHINTON, D.C.
APRIL 18-20, 1989 i
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6 EOUIPPENT OPERABILITY PROMPTLY DETERMINING OPERABILITY AND ESTABLISHING CORRECTIVE ACTIONS PLAk5 FOR DEGRACED OR NONC0hf0RMING SAFETY EQUIENENT In recent years, the nuclear industry and the NRC have increasingly discovered various forms of physical degradation of equipment and systems and noncon-formances of materials and components.
Detection and analysis methods have improved which have uncovered heretofore undetected deficiencies and close scrutiny of past purchasing and marketing practices has resulted in the identi-fication of ncnconformances of materials and components.
While these deficiencies and past practices are of deep concern to us, and must be addressed at their source, the immediate problem for licensees is to assure that the uncertainty raised by the discovery does not threaten the capability of the affected system to perform its specified function.
As such, equipment operability determinations have been called upon more often as degraded or nonconforming ccoditions are revealed.
The nonconformances incluce such examples as:
Codes and Standards specified in the FSAR are not met, As built equipment, or as modified, does not meet FSAR design requirements, Physical evidence of degradation exists such as when heat exchanger fculing has reduced heat removal capability below the FSAR or design
- value, Design documentation is not available or operating experience or engineering reviews have raised questions of design adequacy, and Documentation required by rules, such as 10 CFR 50.49 is not available.
Many of these potentially nonconforming conditions manifest themselves as conditions adverse to quality.
For structures, systems, and ecmponents that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public,Section XVI " Corrective Action" of 10 CFR Part 50, Appendix B " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants", requires prompt identification, correction, and documentation.
The most notable equipment problems in the recent past have been associated with questionable manufacturing, marketing, and quality assurance practices of the suppliers of equipment and components.
In addition to identification provided by licensees and the industry, the NRC has issued bulletins and generic letters which provide, in most instances, acceptable guidance for equipment operebility cecisions and elements of corrective action programs specific te the deficiency.
Most recent examples of these include the NRC Eulletin 88-05 on Ncnconforming Materials (etc) which dealt primarily with flanges and NRC Eulletin 88-10 on Nonconforming Molded Case Circuit Breakers.
In making the eculpment operability determinations, licensees are first guided by the definition of OPERABLE in their Technical Specifications. This standard definiticn was provided in an April 10, 1980 letter to all power plant licensees.
That letter was issued to clarify the meaning of the term OPERABLE
as it applies to the single failure criterion for safety systems in power reactors.
This clarification assures that all conditions of systems and necessary support systems are to be censidered to further assure that no set 7 equipment outages would be alloweo that would result in the facility being in an unprotected condition.
Except for minor word changes, this definition continues today in the Standard Technical Specifications.
As a point of explanation, the systems in the Technical Specifications and any other structure, component, or system that provides a direct capability to prevent or mitigate the consequences of postulated accidents should use the concept of equipment operability determinations.
Equipment that provides support for accident prevention and mitigation must assure the availability.
of tt:e support fcnction in their equipment operability determinations to the extent that tney are needed for the prevention and mitigation structures, systems, and components to perform their functions.
Simply meeting the Technical Specification surveillance requirements for a system with existing nonconforming conditions may not be sufficient for these determinations.
The following are principal points to be considered in dealing with the issue of degraded or nonconforming equipment.
If an immediate threat to public health and safety is identified, action to place the plant in a safe condition should begin as soon as this circumstance is known and should be completed as soon as is reasonable. Shutdown is not always the safest condition or the most appropriate response.
Prompt Determination of Operability Determining equipment operability and plant safety is a continuous decision making process that cannot be avoided.
If, at any time, equipment is found to be inoperable or the plant is found to be in an unsafe condition, action is required.
Upon the identification of the potential nonconformance and specific structures, components, or systems involved, a determination of the operability should be made as soon as possible consistent with the safety importance.
In most cases, it is expected that the decision can be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery even though complete information may not be available.
The decision should be based on the best information available within this time period.
A licensee should examine the full scope of the design basis including the Technical Specifications and FSAR commitments, to establish the conditions and performance requirements to be met for determining operability.
The operability decision may be made based on analysis, a test or partial test, experience with operating events, engineering judgment or a combination of these factors considering equipment functional requirements.
An initial determination regarding operability should be revised, as appropriate, as new or additional information becomes available.
There cust te reasonable assurance that tne equipment will perform its specified safety function.
There must also be a logical and defensible basis for that conclusion.
?elevant past operating experience or identified margins of safety in analyses, including conservatisms used as margin in lieu of detailed analyses at the time of licensing, may be censidereo in making the operability determination.
A 10 CFR 50.59 evaluation should be performed where acorocriate.
If the operability issue can De resolveo Dy a 10 CFR 50.59 cnange anc/or evaluation, no additional action other than that required by 10 CFR 50.59 is necessary.
Followuo Licensee Actions After making the decision on operability, the actions taken by a licensee will depend on whether or not the equipment is covered by the Technical Specifica-tions. Corrective actions to restore or requalify the nonconforming structures, systems, or components to the FSAR or design basis should begin as soon as is reasonably possible.
For equipment covered by the Technical Specifications and determined to be operable, continued plant operation is authorized by the license, however, a corrective action plan to resolve the issue is required.
For equipment covered by the Technical Specifications which is determined to be not operable, the licensee should normally follow the Technical Specification Action statement and commence repair or resolution. Where safety consideration would allow or dictate another course of action, an emergency license amendment or other regulatory action may be requested by the licensee.
For potentially nonconformino ecutoment which is not covered by Technical Specifications and is ceterminea to be operable, centinued operation of the plant is acceptable.
The operability determination should include the basis for determining that the equipment is operable and among other things, the corrective action plan to justify the time needed to complete the corrective action.
For this same equipment determined to be not operable, the licensee must decide i
if reasonable assurance of continued safe operation can be justified, i
If reasonable assurance of continued safe operation cannot be justified, the reactor should be placed in a safe condition.
If continued safe operation can be justified, plant operation can continue while the licensee proceeds with corrective actions as soon as reasonably possible considering the safety significance of the issue.
Where the ' licensee proposes to allow degraded or nonconforming conditions to remain unchanged, a 50.59 evaluation is required.
In addition, prolonged operation with a degraded or nonconforming condition could call for a 50.59 evaluation.
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Scope of Ocerability Determinatiens The scope of operability determinations needs to be sufficient to address the functicnal capability of the equipment to perform its safety function.
Operability determinations should therefore include the following actions:
determine the equipment that is potentially nonconforming, determine the safety function (s) performed by the equipment, determine the circumstances of the potential ncnconformance; including the possible failure mechanism, determine the requirement or commitment established for the equipment, and why the requirerrent or comitment may not be met, determine by what means and when the potentially nonconforming equipment was first discovered, and determine safest plant configuration including the effect of transitional action.
As mentioned earlier, operability determination should be based on:
- analysis, test or partial test, operating experience, and engineering judgement i
Licensees are reminded of 10 CFR Part 21 requirements and are encouraged to maite use of the nuclear industry reporting mechanisms for sharing important information with other utilities.
Cerrective Action Plans The corrective action plan must address the means and schedule for corrective action and should include or discuss the following to achieve final resolution:
repairs or.edifications,
- analyses, tests, documentation activities, license amendments, and 50.59 evaluations cr other actions.
The initial corrective acticn plan for managcment review is expected to be updated as cher:ges in the ccrrective Ettions are enccuntered.
For nest situaticns, the above assessment would be acequate.
However, for equipment not coverec by the Technical Specifications, which is determined to be inoperable, continued safe creration must be justified based on additional i
ccr.sicerations.
The basis for this justification should include consideration i
of the following:
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the availability of redundant or backup eouipment, compensetory measures taken.
events for which the ecuipment must function for prevention cr mitigation.
existing values in relation tc values assumed or implicit in original design or analysis, conservatisms or margins in criginal design or analysis which are used to ameliorate the circunstances, and probability that t.5e event or postulatec accident will occur while the nonccnformance crists, that the nonconformance will result in a failure that initiates the event or accident, and that the equipment will f ail to prevent or mitigate an accident or event.
The written documentatier cf the operability determination process should be kept cn file at the plant site.
Equipment cperability ceterminations and Appenaix B ccrrective actions are the subject of a proposed Generic Letter which should be issued in the near future.
The principal elements of the Generic Letter have been presented here.
The proposed Generic Letter, while general' in nature, is not meant to replace component specific guidance already provided in bulletins or generic letters.
Licensees should continue those ongoing efforts as established.
Examples of guidance which may not follow the propcsed Generic Letter but which should continue to be followed are Generic Letter 87-02 on " Verification of Seismic Adequacy of Mechanical and Electrical Equipment..." and Generic Letter 88-07 on
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...Envi ronmenta l Qualifications of Electrical Equipment Important to Safety..."
For very specific reasons, these and other bulletins and generic letters have been prepared and are directed by special requirements of licensees.
The Generic Letter will provide a definition cf the equipment involved and it is meant to be all inclusive for system, structures, and components where prevention cr mitigation of postulated accidents and unaue risk of public health and safety has been addressed.
The generic letter will not establish new reporting requirements but is intended to help reinforce the safety basis for decisions and to clarify the relationship among Technical Specification Operability, 10 CFR Part 50 Appendix B, OA requirements, corrective action plans and modifications performed under 10 CFR 50.59.
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l Docket No. 50-245 B14692-i i,
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1 - Exhibit 2 Millstone Nuclear Power Station, Unit No. 1 i
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NUCLE AR REGULATORY COMMIS$10N i
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wEwRuruw FOR:
William T. Russell Regional Amninistrator. Region !
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Stewart D. Ebneter Regional Aaministrator Region 11 j
A. Bert Davis kegional Administrat&r, Region !!!
I Robert D. Martin Regional Aabninistrator Region !Y John B. Martin Regional Administrator. Region V FROP:
Thomas E. Murley. Director Offict of Nuclear Reactor Regulation GUIDANCE ON ACTION TO BE TAKEN FOLLOWING DISCOVERY CF POTENTIALLY NONCONFORMING EQUIPMENT
SUBJECT:
i As a result of the EDC Senior Kanagement Meeting on June 25-29, 1988. NRR was
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askee to develop guidance concerning the length of time a licensee should be allcwee to cecice the cuestien of operability when an equipment deficiency After careful consideration. NRR has concluded that it is is ciscovereo.
unnecessary to issue such guidance. There is no generally apprcpriate timeframe in whten operability seteminations shoule be made.
Ecuipment which is clearly inoperable (such as' a valve that will not. open or pump that will not run) should inneetately be~ declared inoperable ard the appropriate tecer.ical specification followee if the equipment is inclueen in the techt.ical specifications. Ah aconformances which call into cuestion the capacility
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of a syster or congonent tc perfoWTtTspecified functions l'nclude lbut are not limiteTto) situations wnere lic'essees ao not meet a Tinarlafety~ Analysis Riptrt70saittJneWUTTtoWor' standard to which they have cossnit'ted. Tver.
TPorimch honconYcTerTries 'may ~not be AccompanGd by physical evidence cf ceTrTYaTion, a_guesTi~9a_Ts raisedfrejana.1rtg Ihe_eperability of the, equipment _.
In 1.e., toe capability of tne enuiprent to perforw. its specified function.
thne cases.ToerebTilty' deter 31 nations theulo be mace by licensees as'soon I
as practicat,le, and in a tisefrant commensurate with the applicable eculpment's importance to safety, using the best informtion available (e.g., analyses. A test er partial test. experience with operating twents, engineering judgment 4
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Multiple Accressees
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cf a coccination of these f actors). We believe that this is the current practice, anc issuarce of prescriptive quicance in this area could isipose an unnecessary tcreen j
cn licensees with no improvement in safet;..
igecific cuestiens on operattlity can, where necessary, be discussed with WRR personrel ty contacting the appropriate Projects Division.
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Thomas E. Murley, Director Office of Nuclear Reactor Regulatier.
cc:
V. Stelle. Ett J. P. Taylor, ECO i
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E Docket No. 50-245 i
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l t - Exhibit 3 Millstone Nuclear Power Station, Unit No.1 l
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December 1993 1
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