ML20203F240
| ML20203F240 | |
| Person / Time | |
|---|---|
| Issue date: | 10/25/1985 |
| From: | Ramsey C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Davis A, James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20203F188 | List: |
| References | |
| GL-85-01, GL-85-1, NUDOCS 8604250087 | |
| Download: ML20203F240 (12) | |
Text
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,7 NUCLEAR REGULATORY COMMisslON E
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\\.~.w... p OCT 2 5 1985 MEMORANDUM FOR:
James G. Keppler, Regional Administrator A. Bert Davis, Deputy Regional Administrator THRU:
b Carl J. Paperiello, Director, Division of Reactor Safety, A g Luis A. Reyes, Chief, Operations Branch, Division of Reactor Safety William G. Guldemond, Chief, Operational Programs Section.
Operations Branch, Division of Reactor Safety FROM:
Charles Ramsey, Reactor Inspector, Operational Programs Section, Operations Branch, Division of Reactor Safety
SUBJECT:
REQUE5T BY THE COMMISSIONERS FOR THE DP0 ORIGINATORS TO WRITE A
SUMMARY
STATEMENT REGARDING THEIR POSITION ON THE FIRE PROTECTION POLICY STEERING COMMITTEE'S RECOMMENDATIONS CONTAINED IN GENERIC LETTER 85-01 During the Comission Meeting of October 3,1985, held to discuss the " Staff Recomendations Covering the Implementation of Appendix R to 10 CFR 50,"
- i the Comissioner's requested the DP0 originators to write a sumary statement r.egarding their position on the Fire Protection Policy Steering Comittee's recomendations.
The staff DP0's were filed with the intent of alerting management attention and achieving formal resolution tc issues related to fire protection and safe shutdown capability of nuclear power plants. Many of these issues appear to be technically or administratively incorrect. Certain issues appear to be inconsistent with the requirements of Appendix R to 10 CFR 50 with the connotation of undemining the basis for originally promulgating the regulation. The significance of the DP0's primarily concerns the protective features required for safe shutdown capability in the event of fire. However, policy and minimum acceptable conditions are also significant issues.
The inspector has reviewed the Staff Recomendations covering the implementation of Appendix R to 10 CFR 50 (Generic Letter 85 01). As a result of this review, the following coments are offered to the comission.
(These views do not necessarily represent the views of Region III management.
Region III management is informed of the inspector's views via this memorandum):
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James G. Keppler 2
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1.
Adequacy of Current Guidance to the Industry ISSUE:
Whether current regulatory and guidance documents are sufficient to achieve an appropriate level of compliance while maintaining consistency in interpretations among plants.
INSPECTOR,C0tNENTS:
The inspector feels that there is a need to review, consolidate and, where necessary, upgrade this lengthy list of documents in order to achieve the appropriate level of regulatory consistency and technical merit in utility compliance with the regulation. Confusion dees exist within the industry and the NRC.
The inspector views the following steering comittee recomendations as appropriate actions that could be taken:
Extensions to the 50.48(c) schedules sh uld no longer 9
be granted.
Institute an expedited fire protection inspection program.
Quality Assurance applicable to fire protection systems is that required by GDC-1 of Appendix A to 10 CFR 50.
(See supporting coments Enclosure
.No. 2)
Conduct fire protection inspections at operating and NTOL plants focusing on at least one site per licensee not subject to a previous Appendix R inspection. These inspections should assess the degree of licensee compliance, promote licensee compliance, and take enforcement action where appropriate.
Issuance of a Temporary Instruction to make this program of inspection effective.
Conduct workshops for inspection teams to assure comon understanding of the objectives, scope and technical issues.
Expedite processing of fire protection enforcement actions and issue guidance for enforcement actions.
g Impose standard fire protection license conditions in each operating license.
5
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4 James G. Keppler 3
OCT 2519o5 Reevaluate all fire protection guidance for consistency and compare these requirements for operating and NTOL plants under the auspices of one working group utilizing fire protection and systems expertise.
Develop appropriate revisions to Standard Review Plan 9.5-1 and Standard Technical Specifications.
Designate a central point of contact for interoffice /
regional fire protection issues utilizing fire protection and systems expertise.
To assure timely and on-track completion of the S.C.
recomendations, review progress at least quarterly, make mid-course corrections as appropriate and report progress results to the EDO.
2.
Documentation Required to Demonstrate Compliance ISSUE:
Whether licensees should be pennitted to perform analyses of fire hazard potential which result in plant modifications or hardware installations without prior NRC review ano approval.
The previous approach required licensees to submit such analyses for prior NRC review and approval so that agreement on modifications or new hardware installations could be reached at the earliest possible stage and the schedule for completion correnced well in advance of a NRC audit.
INSPECTOR COMMENTS:
The inspector disagrees with this S.C. recommendation because the extent that the industry can deviate from the guidance of Generic Letter 83-33 and comitment made to established fire protection code and standards needs to be supported by superior engineering evaluation and empirical data that equal or improve the level of fire safety that is provided by adherence to traditional fire protection methods.
It is the inspector's view that the purpose of the requirements containedinAppendixRistoprov[ideassurancethroughavailable proven methods that the release o radioactivity to the environ-ment will be minimized during and following any nuclear plant fire. This goal is accomplished by maintaining containment integrity, prompt detection and suppression of fire, and main-taining reactor vessel and fuel cladding integrity. To do this, it is necessary to maximize the availability of systems and equipment required for safe reactor' shutdown through maintenance /
surveillance activities and Quality Assurance (Technical Specifications and Quality Assurance requirements).
(See Enclosure No. 1)
2 I I355 James G. Keppler 4
00T For a given plant fire, a safe reactor shutdown will require a minimum of the following: reactor pressure control; circulation of reactor coolant; removal of decay heat and establishment of sub-criticality conditions through the reactor scram, reactivity control, sub-cooling and equipment support functions. Assurance that at least one redundant train of systems and components needed to perform these functions is provided through fire hazard analysis and the resulting diverse methods (permitted by the regulation) to minimize fire damage to this equiprent.
(See EnclosureNo.1)
Clearly, the fire hazard analysis, all supporting evaluations, assumptions and predictions made about the occurrence of fire in a particular plant configuration is an important part of compliance with the regulation. Existing guidelines (Section B) of Appendix R and Section C.1.b of Branch Technical Position (BTP) CMEB 9.5-1, Revision 2, dated July 1981 do not provide sufficient guidance to assure that the required level of conservatism and technical merit for either content or technique to be utilized in such analyses (NUREG CR/2607).
(See Enclosure No. 3)
From staff reviews and audits that have been conducted.
weaknesses that have been identified in previous fire hazard analyses have included the following:
(1)
Improper analysis of fire growth and fire spread which results in the installation of untested and non-rated hardware (i.e., partial height-partial width walls acting as fire barriers).
(2)
Improper assessment of fire area boundaries.
(3)
Inadequate assessment of component fragility to fire and combustion product damage (i.e., fire damage threshold dut to exposure to corrosive species).
(4)
Inadequate overall systems analysis that proceeds through the containment.
(5) Failure to perforn physical or analytical fire models that support predictions about fire occurrences.
(6)
Improper analysis of fire detection / suppression system effectiveness.
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James G. Keppler 5
g :I r.
(7)
Inadequate analysis of fire detection / suppression (manual and automatic) in conjunctior. with fire spread and growth as competing processes.
,(8) Failure to incorporate fire detection / suppression capability into computer based models.
(9) Failure to develop an analytical model for automatic sprinkler system effectiveness which provides the time of actuation and extinguishment for a given fire.
The inspector is concerned about such analyses being performed and retained by licensees for the following reasons:
(a) With the previous approach, Regions II and III inspection teams have averaged approximately 1200 manhours per site performing Appendix R audits.
Inspections were scheduled so that there was a two week preparation period prior to the one week site audit and one week report writing time.
NRR Fire Protection and system reviewers spend
.i approximately 160 manhours reviewing these analyses
/
and other licensee submittals. Unless the site audit is extended at leart another month, it does not appear to be reasonable to expect inspection teams to perfonn this level of review during a site audit.
(b)
If an appropriate level of review is performed at the site during an audit and the inspecticn team takes issue with the adequacy of a licensee's analysis, attempts at resolution would have to be taken.
If resolution could not be reached during the audit, the issues would have to be deferred as unresolved items for subsequent NRR action.
If NRR concurs with the inspection tean and concludes that the analysis is inadequate, this could result in a reduced level of safety at the site and delayed compliance with the regulation until resolution is achieved at.d appropriate modifications are made. This was stated in the staff DPO, but in addition, the inspector is concerned about how the backfit rule applies to this situation.
(c) Whether the analysis is perfonned and retained by licensee's for future NRC audit or the previous practice is continued, the inspector feels that the
James G. Keppler 6
g; ; r industry should be provided with additional guidance which identifies acceptable methods or techniques to be utilized in perfoming such analyses so that the margin of inadequacies and uncertainties can be minimized. This additional guidance includes new guidance to the industry, as well as guidance to the staff for conducting reviews or audits.
3.
Interpretations of Appendix R ISSUE:
The interpretations deviate from previous staff positions with the connotation of undemining the basis for originally promulgating Appendix R.
INSPECTOR C0994ENTS:
Refer to staff DP0 of May 2, 1984.
4.
Appendix 2 Questions and Answers ISSUE:
Potentially confusing and may convey inconsistent staff positions to the industry.
INSPECTOR COMMENTS:
The inspector disagrees with the S.C. recomendation because answers to some of the industry questions are confusing in that they do not respond in a clear and concise manner to the issue raised in the question.
An example of this observation is the response to Question 7.1 concerning fire protection and seismic events. The answer, as in several others, is primarily a reprint of previously published staff guidance that is basically irrelevant to the issue addressed. Similarly, the answer tc Questions 3.1.5 and 3.6.2 are not clear, or appear to be inconsistent with the requirements of the regulation.
From a broader perspective the inspector feels that the questions and answers should not be published. This would formalize the position taken in the responses and impose an added burden on the industry to ascertain the degree that a utility's fire protection program conforms to these positions.
This may result, as it did subsequent to the issuance of Generic Letter 83-33, in additional -
submittals to the staff requesting femal approval of deviations from these positions. Also, based on recent regulatory experience, these positions are not needed to
James G. Keppler 7
OCT 2 5 1985 facilitate continued regulatory efforts in the field of fire protection.
It should be remembered that the questions were originally submitted enly to help the staff develop an agenda for the regional fire protection workshops and, as such, were to constitute nothing more than " talking points" for the workshops.
G.d b --
C. B. Ramsey,
Reactor Inspector i
Enclosures:
As stated I
cc w/ enclosures:
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W. J. Dirks, EDO W. J. Olmstead, ELD D, Kubicki, NRR of. Rose, NRR R. H. Volmer, IE R. L. Spessard IE
./
T. T. Martin, RI J. N. Grace, RII P. Madden, RII J. A. Olshinski, RII J. M. Ulie, RIII J. Holmes, RIII e
e
o ENCLOSURE NO. 1 I
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ENCLOSURE NO. 2 Supporting Connents for QA Requirements l
10 CFR 50.48(a) states, in part, "Each operating nuclear power plant shall have a fire protection plan that satisfies Criterion 3 of Appendix A to this part. The fire protection plan shall describe the overall fire protection program for the facility....
The plan shall describe specific features necessary to implement the program described above such as i
administrative controls...and the means to limit fire damage to structures, systems or components important to safety so 4
that the capability to safety shutdown the plant is ensured."
Since licensee's commit to follow the fire protection guidance contained in BTP 9.5-1, Appendix A to BTP 9.5-1 and supplemental guidance document, " Nuclear Plant Fire Protection Functional Responsibilities Administrative Controls and Quality Assurance," dated June 14, 1977, the inspector feels that these Quality Assurance requirements are adequate for this purpose if enforceable and followed by the industry. The inspector's concern in this area is that the same level of enforceability that can be applied to the installation of fire protection hardware cannot be i
l applied to the quality of these installations because of a licensee's failure to comply with a " commitment" versus
,1 the failure to comply with a " regulation".
While the inspector recognizes GDC-1 and the current "important to safety"/" safety-related" controversy is embraceo by another proceeding, QA requirements f or F.P. are already specified in the B.T.P.
Tnerefore, the inspector concurs with the steering committee that GDC-1 is the appropriate regulatory requirement that is applicable to ensure that fire protection features perform their intended safety function.
1 J
ENCLOSURE 3 SUPPORTING COMMENTS FIRE HAZARD ANALYSIS There is a whole arsenal of fire hazard analysis methods and techniques that contribute to the problem. All of them share the connon framework that includes identification of critical areas of vulnerability and calculation of calorific heat value (BTU /sq.ft.)ofcombustibles. They may identify all critical areas where important safety equipment could be damaged by fire, but this concept is further narrowed to a criterion that emphasizes areas where only redundant equipment could be com-promised by fire. Attention is generally given to questions such as the potential for the cross-zone spread of fire and the likelihood that transient combustible fuels might supplement insitu combustible fuels. While this part of such analyses remains more of an art than a science, the general consensus is that it is reasonably mature in the sense that the uncertainties introduced are thought to be smaller than uncertainties from other aspects of the analyses.
There are limitations in each fire hazard analysis method or technique that inhibit the ability to precisely model and quantify the interaction of the discrete contributors to fire risk in exact tems.
Further research is necessary in this area i
in order to narrow the margin of large uncertainties that exist so that the usefulness of bottom line fire hazard analysis values can be verified to ensure accuracy and completeness.
Past studies have shown that credible fires in nuclear power plants initiate and grow from a finite number of combustible fuels (cable insulation, lubricants, housekeeping and maintenance activities, etc.), some of which may be in predictable config-urations (vertical and horizontal cable trays). The energy and mass release rates of fire from these configurations, including release rates of certain corrosive products of combustion must be known in order to better determine affects of environments resulting from fire. The specific correlation between combustible fuel load (BTV/sq. ft.) in an enclosure and the fire severity it could produce in tems of a fire exposure represented by minutes on the ASTM E-119 Standard Time Temperature Curve is not a complete data base that provides sufficient detail to evaluate fire exposure to equipment important to safety in nuclear power plants. Other aspects of the fire environment such as ventilation characteristics, rocm or enclosure configuration, themal conductivity, fire detection and suppression effects, etc., need to be known in order to better characterize equipment performance in such environments.
In most cases, analytical fire models can be utilized to obtain better engineering insights that enable more accurate predictions about fire environments.
The more difficult aspects of predicting the likelihood of disabling equipment due to damage resulting from fire environments can be obtained through evolutions of physical modeling (full scale fire testing). Research in this area is continuing.
However, it is nnt possible to physically test every nuclear plant enclosure configuration. Some empirical data exist from actual physical model testing of specific plant configurations which can be useful in fire hazard analysis predictions.
Several important analytical fire models have been developed and used in recent years that can assist in calcclating the likely progression of the fire phenomena. Although, even in the best cases, with the use of these analytical fire models, uncertainties remain large in a numerical sense (NUREG/CR 3239-1983). The compounded problems of modeling fire detection / suppression systems, actual combustible fuel availability (amount and character of transient fuels), the stochastic nature of fire growth over time, the size of the affected secondary zone where hot and corrosive gases can cause equipment failures and issues of access for fire fighting, are difficult to determine through extrapolations.
Furthermore, the level of degradation of systems and equipment due to fire environments and consequential spurious actuation of fire suppression systems where the continued operability of safe shutdown systems and equipment can no longer be ensured (analysis of system effects - fire damage threshold of equipment) needs to be included in such analyses. For example, although it is common to find operating temperature ratings on various pieces of electrical equipment, fire damage threshold ratings are not specified and there is no current guidance for performing this type of qualification testing.
The damage threshold of systems and equipment is related to parameters of the environment in the irmediate vicinity of the equipment. These parameters will include temperature, beat flux and exposure to corrosive species f rom fire environments which may result in reaching a fire damage threshold for equipment through the combination of the environmental parameter and the duration of exposure. The loss of operability may be due to loss of function. For example, a motor fails to start because of the disconnection of a cable; or, because of spurious actuation (a relay spuriously makes or breaks a circuit).
The existing data base on fire occurrences in nuclear plants indicates that threat of fire damage to systems and components important to safety is significant. Since the 1975 Erowns Ferry fire, over 100 fires have occurred in nuclear f ants. Many of these fires affected systems and components important to safety.
The examination of the fire risk sequence as a precurson to potential radioactive release to the environment should require quantitative estimates based on further research of fire burning and spreading characteristics, as well as fire occurrence analysis which result in Probabilistic Risk Assessments (PRAs) that more precisely predict the potential for the fire occurrence.
2
Several different PRA methods exist in available literature (NUREG CR/2300), but only recently has the PRA of internal i
nuclear plant fires become an accepted part of a full scale nuclear plant PRA (i.e., Big Rock Point, Zion and Indian Point).
The methodology used in these studies is by no means mature; however, the few recent applications have provided new engineer-ing insights about nuclear plant vulnerabili'ies to fire damage.
It appears that these methods permit a more realistic simulation of the affects of nuclear plant fires and they are fully capable of identifying many important types of vulnerabilities in a qualitative sense which includes the ranking of relative importance.
Although previous staff poritions have rejected the use of PRAs, these new engineering insights may merit reevaluation of previous positions.
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