ML20211N448
| ML20211N448 | |
| Person / Time | |
|---|---|
| Issue date: | 09/02/1999 |
| From: | Spector A NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9909100196 | |
| Download: ML20211N448 (25) | |
Text
I September 2, 1999 MEMORANDUM TO: File FROM: August K. Spector, Communication Task Leader inspection Program Branch (Original signed by:)
Division of Inspection Program Management Office of Nuclear Reactor Regulation
SUBJECT:
PUBLIC MEETING ON REACTOR OVERSIGHT PROGRAM ISSUES MAY 26,1999 On May 26,1999, a public meeting was held between the NRC and the NEl to continue exchanging information on the reactor oversight program. The meeting agenda, a meeting summary, a list of attendees and a copy of written information exchanged at the meeting are attached.
Attachments: As stated
Contact:
August K. Spector 301-415-2140 i
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/ September 2, 1999 MEMORANDUM TO: File FROM: August K. Spector, Communication Task Leader inspection Program Branch i Division of Inspection Program Management Office of Nuclear Reactor Regulation
SUBJECT:
PUBLIC MEETING ON REACTOR OVERSIGHT PROGRAM ISSUES l MAY 26,1999 On May 26,1999, a public meeting was held between the NRC and the NEl to continue exchanging information on the reactor oversight program. The meeting agenda, a meeting summary, a list of attendees and a copy of written information exchanged at the meeting are attached.
Attachments: As stated
Contact:
August K. Spector 301-415-2140 l
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' P BLIC MEETING
SUMMARY
, 1 ISSUES DISCUSSED
- 1. Discussed Frequently asked questions. See attachment for Updated Q&A submitted by NEl and reviewed by NRC.
~
- 2. Reviewed by Region i public meeting training session. Generally agreed that the
' training was well received session. A lot of information was presented by a wide variety of presenters during the meeting.
- 3. Discussed pilot project 4; Fire Protection. See attachment information which was submitted related to Fire
. Protection issues discussed in the meeting.
' 5. NRC and the industry discussed Information System Support and Logistics for the overisght process performance dada such as performance indicator plant issue motion
. and actions taken by the licensee /NRC
- 6. Participants asked questions regarding SDP process and methods of coloring finding and performance indicators. The NRC staff explained in detail how the assessment and SDP process work. Detailed draft guidance will be issued to pilot plants before the
. . implementation of the pilot program.
~ 7. NEl/NRC discussed their plans to survey public to assess their perception about the i new process and to verify success criteria of the new process.
- 8. Responses for survey conducted to the licensee's plant manager level (see attached) were discussed briefly discuss the meeting.
- 9. Industry questioned the licensing fees charged during the pilot program implementation.
The NRC staff stated that the guidelines will be discussed during the future meetings.
- 10. NEl stated that Mr. Lockbum will issue a report on the new program. The staff stated that it should be directed to OPA.
- 11. Maintenance Rule definition and other maintenance rule related issues ;
were discussed (see attached).
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! ' ATTENDEES Public Meeting MAY 26,1999 Mil John Butler Tom Houghton NRL August Spector j Roy Mathew Don Hickman l Tim Frye { '
Alan Madison Ron Frahm, Jr.
Tom Boyce OTHER f Dennis Zarmoni Gabe Salamon ,
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l AGENDA FOR MAY 26,1999, NRC/NEl MEETING TO DISCUSS THE CONTINUED DEVELOPMENT OF RISK-INFORMED PERFORMANCE ASSESSMENT PROCESS AND INSPECTION PROGRAM IMPROVEMENTS I
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- Introduction '
. Purpose of Meeting l
4 Review / Discuss the Risk-Informed Inspection Program and Assessment Process
- Planning for Future Interaction
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- g 9
4 Comment. Three commentors expressed similar views related to high-risk activities.
One noted that, under suitable controls, a shorter time in a more risk-significant configuration may be safer than a longer time in a less risk-significant configuration. Another noted that high risk-significant activities should be recognized and avoided, where practical, and limited in duration when they are necessary. The third noted that the proposed rule does not address situations in which failure to perform a maintenance activity may have a greater impact on risk than performing the high safety-significant activity.
Response The NRC agrees that the proposed rule precluded entering risk-significant configurations, no matter the duration, when, in fact, situations may exist that would yield a net '
safety benefit by performikmaintenance in a risk-significant configuration for a short time. The rule has been revised to require licensees to understand their options with respect to risk and to manage their maintenance activities according to their best judgment, considering insights from operating experience and deterministic and probabilistic analyses.
- 8. Emeroent maintenance reauirements.
Comment. Two commentors stated that the proposed rule does not address expectations for revising assessments upon the discovery of a previously unknown condition requiring maintenance (emergent maintenance). They also expressed concems that if certain emergent maintenance activities are not completed immediately, the plant could be at greater risk.-
Response Under the revised rule, an assessment is required to be initiated following the discovery of emergent failures or changes in plant conditions to determine the safety impact
' of the failure or the change in piant conditions. For additional information on this subject, please see the discussion in item 4 of Section Ill, "The Final Rule," below. -
- 9. Qggumentation of the assessment.
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' Comment. Three utility commentors stated that the proposed rule is not explicit enough
. regarding assessment documentation expectations.
Response The rule has no explicit documentation requirements. Instead, the rule emphasizes performance. A licensee's assessnient process is expected to identify the impact on safety that is ' caused by the performance of maintenance. Licensees should use documentation to the extent necessary to assure themselves that the requirement for an assessment has been acknowledged and performed adequately. NRC expectations are that a
- licensee will have a requirement for the assessments and an explanation of the process to be followed in its maintenance rule program, along with a description of assessment tool (s) to be used and their limitations, implementing procedures, and explicit direction covering instances when the plant configuration is or is proposed to be outside the span of the assessment tool.
Further, the assessment process is expected to be incorporated into the maintenance planning and scheduling process and into work package requirements. Moreover, control room operators, who are expected to understand, use, and know the limitations of the assessment
- tools, generally use and maintain a variety of documents, such as logs and checklists, that
' contain information relating to out-of-service SSCs.
- 10. Definition of availability.
Comment. Three commentors stated that the definition of availability will be key to this rulemaking.. They also stated that the availability definition should take into account the time required to restore the functionality of an SSC and should also be risk informed.
Response. A definition of availability forlicensee maintenance rule programs is set forth in NUMARC 93-01, Revision 2, which was endorsed by the NRC in Regulatory Guide 1.160, !
Revision 2, of March 1997. According to that document, availability is "(t)he time that a(n) SSC
-is capable of performing its intended function (expressed) as a fraction (usually as percent) of
F-11 the total time that the function may be demanded." Also according to that document, under the definition of " unavailability," is the following statement: "An SSC that is required to be available for automatic operation must be available and respond without human action." Additionally, in the instance where an SSC is taken out of service for testing but could be manually activated, the NRC has accepted that, as long as the dedicated operator's written procedure specifies a single action that would permit an automatic initiation of the out-of-service SSC in the event of an accident or transient during the test, the SSC could be considered available. (Meeting
]
Summary - November 19,1991 NRC/NUMARC Public Meeting on the Development of Guidance Documents for the implementation of the Maintenance Rule (10 CFR 50.65), R.P.
Correia, Office of Nuclear Reactor Regulation, memorandum to E.W. Brach, Office of Nuclear Reactor regulation, dated November 23,1991.) The NRC's expectation is that, by procedure, the dedicated operator is stationed at the equipment and is ready and qualified to perform that single action in a moment. An acceptable single action could be the rapid repositioning of a switch or a lever; an unacceptable action would be racking in a breaker or, in some instances,
- opening a manual gate valve.
With respect to risk-informing the maintenance rule definition of availability, the reliance of initial availability performance measures on probabilistic risk assessment (PRA) data provided such a basis. However, in quality maintenance programs, availability is monitored to identify and trend the performance of equipment, thereby permitting certain conclusions to be drawn about the effectiveness of the equipment's maintenance program. Paragraph (a)(3) of the rule requires that the prevention of SSC failures (reliability) through maintenance is appropriately balanced against the objective of minimizing unavailability. Omitting unavailability time from the maintenance effectiveness determination analysis is flawed logic. Omitting unavailability time because, in an accident scenario, the equipment may not be needed for the time it may take to
j .
e 12 restore its safety function recognizes the role of the equipment but masks the actual requirement for maintenance. The maintenance rule requires licensees to monitor the effectiveness of their maintenance programs. Omitting significant details, such as how much maintenance time an SSC requires in order to attain the objective of preventing failures, is contrary to the purpose of the rule.
Note also that maintenance rule " availability" is not technical specification " operability."
- 11. Backfit and reaulatory analyses.
Comment. One commentor stated that the regulatory analysis does not justify the expansion of the maintenance rule to " normal shutdown operations" and that a revision of the analysis to better consider such expansion would show through backfit considerations that the
. expansion is not justified. Another commentor also presented a concern that the overall .
implications of the rule were not supported by the backfit analysis.
Resoonse. The new preamble to the rule is an introductory sentence clarifying that the
\
rule applies under all operating conditions, including normal shutdown. The Commission intended the rule to apply to all operating conditions, and it has been implemented by the NRC staff consistent with such an interpretation. Moreover,' Section 11.2.3 of NUMARC 93-01 i
spe_cifically states that "ascessment applies during all modes of plant operation." The overall implications of the rule were assessed in the backfit analysis for the original maintenance rule, which was issued July 10,1991.
I
- 12. Reaulatory analysis cost estimatesc Comment. One commentor rais* ' the concem that if facilities are required to develop numerical models for every combin; .,on of low safety-significant SSCs, the cost of implementing the program would be significantly higher than estimated in the regulatory analysis.
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l August Spector-clirify.wpd___ ~
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'Pageh NRC Clarification Questions: " Stakeholder Comments on Draft FP Inspection Module" Prepared by: Leon Whitney, FPES/SPLB/NRR/NRC Prepared on: May 25,1999 First Commenter "The inspection basis (last sentence) is inconsistent with the objective."
NRC Clarification Comment: Please explicitly specify the sentences which are inconsistent and state what the perceived inconsistencies are.
"Section 02.02.b appears to be demanding that we have an unannounced fire drill in a high risk area at a time of NRC's choosing."
NRC Clarification Comment:
The NRC notes that the subject sentence begins with the phrase "may also include" rather than "
will include." Nevertheless, NRC inspection teams do expect the licensee to conduct one fire drill during the onsite inspection period. The location of the drill may be unannounced to the fire brigade. However, the inspectors will have discussed with the licensee the timing and character of the fire drill during the information gathering visit weeks in advance of the inspection. A mutually agreeable scenario would be agreed upon at that time.
Is the commenter's issue the phrase "high risk area?" Are there other phrases which the commenter could suggest (e.g. " plant area in which a fire could require conduct of a post-fire safe shutdown")? Should the passage have some other form of rewrite?
Is the commenter's issue that the fire drill is unannounced? This is not equivalent to unplanned or unsupervised.
Please expand the comment to illustrate exactly what the commenter's fire drill issue is.
"The triennial general guidance should pick a different high risk area than the previous inspection, to be most comprehensive. It should also include a sampling check of lower risk areas. Otherwise the utilities might be tempted to " teach the test" and rigorously keep the high risk areas at 100% and let other medium and low areas degrade."
NRC Clarification Comment: Only in two places does the phrase "high risk" appear,02.01b and 02.02b, both of which are related to observation of fire brigade drills or actual responses.
Everywhere else the inspection planning is described as looking at selected risk significant areas, in some cases up to five such areas during each inspection, the selection of which is driven by a high number of determinants. Does the commenter believe that the selection processes are flawed such that the same plant locations will always be selected? If so, please elaborate on how this is so.
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[Kugust Spector-c!frify.wpd .
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"Section 03.01/2: Include the referenced Information from the FPFI module to allow this to be sufficient."
NRC Clarification Comment: The basic premise of the development of the baseline procedure was that the FPFI team inspection procedure will be published as part of the inspection program (that is, as a final rather than draft document), that the FPFI team inspection 1 procedure would be the fundamental' source of inspection lines of inquiry for the inspectors who conduct baseline / triennial activities, and that the inspectors who conduct baseline activities would be well qualified to extract information from the FPFI team inspection procedure. This obviates the necessity to replicate FPFI team inspection procedure information within the baseline procedure. Considering this information, does the commenter still believe that extracts !
should be put in the baseline procedure?
"Section 03.02a should also request supporting calculations."
i NRC Clarification Comment: There was a typographical error whereby 03.02b was labeled 03.02a. The first 03.02a does ask for supporting calculations. Given that the "SRA's report will not focus on the validity fo the modeling assumptions of the IPEEEs," it is not clear why (in the true 03.02b on the bottom of the fourth page) the SRA would necessarily, need to ask for )
supporting calculations and analyses. Please elaborate or explain.
]
Section 03.02b: The inspection results and non plant specific fire event information have little to do with ranking fire areas according to risk; this should not be the responsibility of the SRA to acquire / address."
NRC Clarification Comment: Consider that inspection results (hypothetically for example repetitive findings of large quantities of transient combustibles in a given plant location) may affect SRA fire risk calculations. Also, the term "non-plant specific fire event information" refers to generic data which the SRA would use as input to his plant area fire risk calculations. It is )
not understood why the team's SRA should be prohibited from either of these information sources. Please elaborate.
Second Commenter "The main comment I have on the proposed guidance deals with the source of their criteria. The source of the data for the sprinkler systems, for example, comes from i recent versions of NFPA 13. Those of us with older plants will not meet this criteria, I even though we will be in full compliance with our committed code of record. This could j result in being classified with a high degradation category deficiency, while being i completely in compliance with our commitments.
One specific example is as follows: Older sprinkler codes only considered an obstruction below the head to be a problem if it was greater than 48" wide. The guidance provided states that an obstruction below the head of 24" or greater requires a i head below the obstruction. A high degradation condition would be identified if two or i more heads were obstructed in this manner. This would not be that uncommon with the presence of 24" wide cable trays and ducts.
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[ ugust Spector- clarify.;wpd_
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It seems like we will be expected to backfit our systems to meet the current code if we j are to have a satisfactory inspection." l l
l NRC Clarification Comment: This stakeholder comment appears to be directed toward the criteria used in the FPRSSM ti sh significance determination process. The staff considers situations in which FPRSSM risk values would result in Commission backfit actions to be unlikely. Nevertheless, risk assessments should consider all relevant information. Within the l FPRSSM process, equipment and systems which were installed under commitments to older I codes may be characterized as degraded (hypothetically, judged to be less than fully effective relative to newer, state-of-the-art designs). It is possible that such a degradation could be considered as an input into a risk significance determination calculation. However, the stringent criteria of 10 CFR 50.109 (Backfitting) would need to be met before a change to licensee code
' commitments would be directed by the Commission.
Third Commenter if parts of the " Draft FPFl" module are going to be the basis / criteria applied in the triennial (or any other inspections) associated with this module, the specific review criteria should be incorporated into this module to make it a " stand alone" document.
Referencing a " draft" docurnent is not consistent with accepted practice in the nuclear regulations.
NRC Clarification Comment: As stated above, the basic premise of the development of the baseline procedure was that the FPFI team inspection procedure will be published as part of the inspection program, that is, as a final rather than draft document. Furthermore, as stated above, the FPFI team inspection procedure will be the fundamental source of inspection lines of inauiry for the inspectors who conduct baseline / triennial activities. As such, the line items in the FPFI procedure are not requirements upon the licensee.
" Risk significant"should be more stringently defined as " risk significant fire safe shutdown areas." Due to the IPEEE analysis methodology, an area may be " risk significant in IPEEE terminology, but not necessarily in safe shutdown terminology."
NRC Clarification Comment: The point of this stakeholder comment is not clear. It would be helpfulif the commenter could elaborate on the specific baseline procedure passages in which the term " risk significant" causes confusion, and then elaborate on the confusion which is perceived to result.
"The module does not Indicate if the corrective action program will be included in the triennial review scope."
The NRC Reactor Assessment and Oversight Task Force plans to pilot and issue an inspection module specifically directed at the assessment of licensee corrective action programs. There is no intent to duplicate that effort within the fire protection inspectable area.
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( August Spector - NRC FP Biselinil_nspection ProceduriCifrificatiorIQu stions/Comm:nts Pag 3'1] ,
From: Leon Whitney To: Alan Madison, August Spector, INTERNet:fae@nei.o..
J Date: Wed, May 26,1999 4:08 PM
Subject:
NRC FP Baseline inspection Procedure Clarification Questions / Comments Fred:
Attached are NRR's Fire Protection Baseline inspection Procedure Clarification questions and comments for industry consideration.
Augie Spector will put the attached document in the meeting minutes for either the 5/24 meeting or the 6/28 meeting.
Separate FPRSSM clarification questions will be sent next week by J.S. Hyslop and Pat Madden.
Leon Whitney
t .,
Surrested Anoroach for Assessing Fire Protection Inspection Findings Fire protection deficiencies are but one way that safe shutdown equipment may not be able to perform their intended function. Other reasons may be due to environmental qualification, seismic, maintenance or quality assurance issues, etc.
The Significant Determination Process.(SDP) for the reactor safety cornerstones (initiating events, mitigation and barriers)is capable of assessing the significance of a finding that impacts plant safety equipment regardless of the cause. Fire can be
' both an initiating event and can affect mitigation capability. But, this is not unique
- the same can be said for events, such as Loss of Offsite Power.
The attached SDP should be able to be use'd to assess fire protection deficiencies discovered during inspection activities. Fire protection deficiencies can be translated into an estimated frequency of having a meaningful fire (one that could reasonably disable safe shutdown equipment). This would be done as follows:
Meaningful fire initiating frequency = : Fire ignition frequency times DID factor Where DID factor = (automatic suppression) x(automatic detection)x (fire barriers)x (fire brigade effectiveness).
The fire ignition frequency for an area would be multiplied by the DID factor to arrive at a meaningful fire initiation frequency. This frequency would then be used as an entry point into Table 1 of the SDP. A robust DID factor would result in 1 lower meaningful fire frequencies and thus any associated inspection findings l would be assessed as having less significance. Conversely, a weak DID factor would result in higher meaningful fire frequencies and associated findings would be assessed as having higher significance.
With this approach, there would be no need to have a significantly different SDP
. process for fires. The only difference would be an adjustment for the entry .
frequency in Table 1 based on the DID factor, which is the direct expertise of the i fire protection evaluators.
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