ML20129C716
| ML20129C716 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/17/1985 |
| From: | Snyder B Office of Nuclear Reactor Regulation |
| To: | Standerfer F GENERAL PUBLIC UTILITIES CORP. |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR GL-85-01, GL-85-04, GL-85-05, GL-85-08, GL-85-1, GL-85-11, GL-85-4, GL-85-5, GL-85-8, NUDOCS 8507290524 | |
| Download: ML20129C716 (2) | |
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Distribution: July 17,1985 docket-NoF50-3201 NRC PDR Local PDR DCS Docket No. 50-320 TMI HQ R/F TMI Site R/F
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BJSnyder WDTMavers MTMasnik
!!r. F. R. Standerfer RWeller Vice President / Director PGrant (Site)
Three 1111e Island Unit 2 RCook (Site) GPU Huclear Corporation CCowgill (Site) P.O. Box 480 LChandler, ELD fliddletown, PA 17057 IE (3) ACRS (16)
Dear Mr. Standerfer:
M-town Office
Subject:
Applicable Generic Letters - January 1,1985 through June 30, 1985 The Three flile Island Program Office has reviewed all generic letters issued from January 1,1985 through June 30, 1985. As a result of our review, we are forwarding to you five (5) generic letters that have been determined to be applicable to your facility. For those generic letters that require some actio9 by a certain date, you should lengthen the
" compliance by" date indicated on the letter by the number of days that has elapsed between the date of issuance of the generic letter and the date of this letter.
If there are any questions with regard to applicability or compliance, please contact Michael T. Itasnik of my staff. Sincerely,
/s/ R. A. Weller fgp Bernard J. Snyder, Program Director Three tille Island Program Office Office of Nuclear Reactor Regulation
Enclosures:
Generic Letters (85-1,85-4,85-5, 85-8, and 85-11) cc: T. F. Denmitt R. E. Rogan S. Levin 8507290524 h hPDR 20 W. H. Linton PDR ADOCK J. J. Byrne F A. W. fliller Service Distribution List [see etteched) "/e Ench.
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- U.S. GPO 1983-400-247
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TMI-2 SERVICE LIST Dr. Thomas Murley Willis Blaby, $tte Manager Regional Administrator. Region I U.S. Department of Energy U.5. Nuclear Regulatory Comission P.O. Bos 88 631 Park Avenue Middletow.. PA 17057 0311 King of Prussia. PA 19406 ( Oavid J. McGoff ! John F. Wol f e. Esq. , Chairman. Division of Three Mile Island Programs Administrative Judge hC-23 f 340g $hepnerd St. U.S. Department of [nergy Chevy Chase. MD. 20015 Washington. 0.C. 20545 Dr. Oscar M. Paris W11114m LeChstet Administrative Judge 104 Davey Laboratory Atomic Safety and Licensing Pennsylvania State university Board Panel Unt'ersity Park. PA 16802 - U.S. Nuclear Regulatory Comunission Washington D.C. 20555 Randy Myers. Editorial The Patriot Dr. Frederick M. Shon 812 Market St. l Administrative Judge Harrisburg. PA 17105 Atomic Safety and Licensing Board Panel Robert 8, Borsum U.S. haclear Regulatory Comission Babcock & W11 con Washington. 0.C. 20555 huclear Power Generation Division Suite 220 . Karin W. Carter 7910 Woodmount Ave. Assistant Attorney General Bethesda, M0. 20814 505 [necutive House P.O. Son 2357 Michael Churchh111. Esq. Marrisburg. PA 17120 PILCOP
- Dr. Judith M. Johnsrud 1315 Walnut St.. Suite 1632 Philadelphia. PA 19107 Envirormental Coalition on huclear Power Linda W. Little 433 Orlando Ave. 5000 Hermitage OR.
State College. PA 16801 Raleigh.NC 27612 George F. Trowbridge. Esq. Marvin I. Lewis Shaw. Pittman. Potts and 6504 Bradford Terrace Trowbridge Philadelphia PA 19149 1800 M. St.. NW. Washington. 0.C. 20036 Jane Lee 183 Valley Rd. Atomic Safety and Licensing Board Pane) Etters.PA 17319 U.S. huclear Regulatory Comission Washington, D.C. 20555
- J.B. Liberman. Escuire Atomic Safety and Licensing Appeal Panel Berlack.Israels. Liberman 26 droa hay U.S. Nuclear Regulatory Comissten new yorg, ny longa Washington. 0.C. 20555 Secretary Walter W. Cohen. Consumer Advocate Department of Justice U.S. huelear Regulatory Comission Strawberry Souare.14th Floor ATTN: Chief. Docketing & Service tranch Harrisburg. PA 17127 Washington. 0.C. 20555 Mr. Larry Hochendoner Edward 0. $wartz Dauphin County Ccognissioner Board of Supervisors P.O. Bos 1295 Londonderry Township Harrisburg, PA 17108-1295 RFD 81 Geyers Church Rd.
Middletown. PA 17057 John E. Minnich. Chairperson. Dauphin County Board of Comissioners , Robert L. Kr.upp. Esquire Dauphin County Courthouse Assistant solicitor Knupp and Andrews Front and Market Streets P.O. Bos P Marrisburg, PA 17101 a07 h. Front St. Dauphin County Office of Emergency Harrisburg. PA 17108 Preparedness Court House. Room 7 John Levin. Escuire
- Front & Market 5treets Pennsylvanta Public utilities Com.
Marrisburg, PA 17101 P.O. Box 3265 Marrisburg. PA 17120 U.S. (nvirofunental Protection Agency Region !!! Office ATTN: [IS Coordinator Curtis Building (Sisth Floor) 6th & Walnut streets Philadelphia, PA 19106 Mr. towin rintner taecutive vice President Thomas M. Gerusky. Director General Public Utilities nuclear Corp. Bureau of Radiation Protection 100 Interpace Parkway Department of [nvironmental Resources Parsippany, NJ 07054 P.O. Box 2063 Marrisburg PA 17120 Dan Kennedy Of fice of (nvironmental Planning Department of Environmental Resources
$Q 8 West Etng Street P.O. Son 2063 Lancaster. PA 17602 Marrisburg PA 17120
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# 'o UNITED STATES
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' \p NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
/ JAN 9 1985 TO ALL POWER REACTOR LICENSEES AND ALL APPLICANTS FOR POWER REACTOR LICENSES Gentlemen:
SUBJECT:
FIRE PROTECTION POLICY STEERING COMMITTEE REPORT (Generic Letter 85-01) Enclosed is a copy of the NRC Fire Protection Policy Steering Committee Report, dated October 26, 1984. The background and purpose of the Steering Committee is described in the report. A notice will be published in the Federal Reaister in the near future that will provide an opportunity for public comments on this report. No response to this letter is required.' el . Ei nhu , Division of LWcensing
Enclosure:
As stated R C ru .* T e a , N
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e e ENCLOSURE
#pa asog'o UNITED STATES
!' ,,q }p NUCLEAR REGULATORY COMMISSION
; .' WASHINGTON, O C. 20555
% . . . . . #' OCT 2 61984 MEMORANDUM FOR: William J. Dircks Executive Director for Operations FROM: Fire Protection Policy Steering Committee
SUBJECT:
RECOMMENDED FIRE PROTECTION POLICY AND PROGRAM ACTIONS Introduction At an August 27, 1984 meeting on fire protection, you directed that a review of current fire protection issues be conducted and that recomenda-tions for resolution of these issues be made within two months. This effort was to exaniine all current licensing, inspection, and technical issues to develop policy recommendations aimed at expediting Appendix R compliance for older plants and assuring consistent levels of fire protection safety at all plants. By your memorandum dated September 13, 1984 (Enclosure 1) to the NRR ana IE Office Directors and the Regional Administrators, you confirmed this direction and approved a Steering Committee to provide appropriate recom-mendations. You indicated that, among the issues to be considered, were: adequacy of current guidance to industry; interpretation of Appendix R requirements; treatment of technical and schedular exemptions; comparison of Appendix R and current NT0L plants for fire pro-tection safety; adequacy of current inspection practices; and identification and resolution plan for any outstanding technical issues. In response to this direction, the Fire Protection Policy Steering Committee (SC) has considered the broad range of fire protection issues necessary to arrive at policy recomendations. The SC has reviewed documents which provide the basis for current fire protection policies and which discuss many of the issues that could significantly delay Appendix R compliance and question consistency in fire protection safety at all plants. The SC held six meetings. These included meetings with the Senior NRC Managers, the NRR and IE Office Directors, and the fire protection engineers from NRR, IE, and the Regions. At the latter meeting, the candid views of the individuals intimately involved in the fire protection issues were solicited and received. A record of the SC meetings is included as Enclosure to GL 85-01, Re: Fire Protection Policy
s William J. Dircks 20 N Enclosure 7. Finally, the SC received significant input from Thomas Wambach, who acted as Secretary to the SC, and the Working Group, headed by Faust Rosa, NRR, and composed of V. Benaroya, C. Grimes, and V. Moore of NRR; S. Richardson, IE; W. Shields, OELD; and C. Anderson, T. Conlon, and W. Little of Regions I, II and III respectively. The recommendations of the SC are provided below. We believe that this responds to your direction, and when fully implemented, represent actions that will indeed expedite Appendix R compliance and assure consistent levels of fire protection safety at all plants. The SC is aware that not all parties will be fully satisfied with these actions. Nor have our recommendations been reviewed by the cognizant Offices or Regions. However, we believe that they represent sound judgments balanced with other competing safety priorities, and that with your approval the plan can and should be initiated promptly. Recommendations To expedite Appendix R compliance and assure consistent levels of fire protection safety at all plants, the Fire Protection Steering Comittee recommends that the following actions be taken:
- 1. Promptly issue the enclosed Generic Letter (Enclosure 2) informing all licensees that:
(a) Extensions to the 50.48(c) schedules will no longer be granted; (b) An expedited fire protection inspection program will be instituted; (c) Documentation of valid analyses supporting fire protection , features must be available for inspection; (d) Quality assurance applicable to fire protection systems is that required by GDC-1 of Appendix A to 10 CFR Part 50; and (e) The interpretations of Appendix R (Enclosure 3) which should facilitate industry implementation of Appendix R and the responses to industry questions (Enclosure 6) represent the official agency position on all issues covered. (It should be noted that The Comission requested these documents for their review prior to issuance to industry.) Enclosure to GL 85-01, Re: Fire Protection Policy
4 . OCT 2 61984 William J. Dircks 2. Conduct fire protection inspections within CY 1985 at ors and NT0Ls to include at least one site per licensee not subject to a previous Appendix R fire protection inspection. These inspections will assess the degree of fire safety, steer and promote licensee compliance, and take enforcement action where appropriate. A Temporary Instruction for this program will be issued by 11/15/84. To make this program of inspection most effective: (a) A workshop for the inspection teams will be conducted in mid December with SC, NRR, IE and Regional participation to assure common understanding of the objectives, scope and technical issues. Followup workshops will be held as needed; (b) The fire protection inspections will utilize new guidance for enforcement actions (Enclosure 4); (c) The processing of current fire protection enforcement actions will be expedited; and (d) A referee will be established to promptly resolve significant-differences between the inspection teams and licensees.
- 3. Upgrade regulatory documents and procedures to achieve an appropriate level of fire protection safety while maintaining consistency among plants. In particular:
(a) Impose a standard fire protection condition (Enclosure 5) in each operating license (already being implemented); (b) Reevaluate all fire protection guidance for consistency with the SC recommendations and compare fire protection requirements for ors and NT0Ls, both under the auspices of the Working Group; (c) Develop appropriate revisions to the Standard Review Plan and Standard Technical Specifications by March 31, 1985; and (d) Designate the Director, Division of Engineering, NRR as the central point of contact for interoffice / region fire protection issues.
- 4. To assure timely and on-track completion of these recomended actions, the SC will review progress at least quarterly, make mid-course corrections if appropriate, and report to the EDO.
' \
Enclosure to GL 85-01, Re: Fire Protection Policy
William J. Dircks OCT 2 61984 1 Discussion The recommended actions are grouped into three main areas dealing with (1) guidance to industry, (2) an expedited program of fire protection inspec-tions, and (3) a general upgrading of regulatory documents to reach and maintain consistent fire protection safety. This discussion section will l focus broadly on what the SC found during its deliberations to warrant the i focus of these recommendations and will indicate how this satisfies the agenda of issues cited in your memo of September 13. Details on these issues are provided in the record of the SC deliberations contained in Enclosure 7. With regard to guidance to industry, the SC concludes that adequate tech-nical guidance had been issued but that there were areas where confusion could arise. It was not clear where exemptions were needed, for example. However, a diligent reading of Appendix R and other staff documents did provide the basis for the satisfactory implementation of Appendix R at Calvert Cliffs. The SC concluded that it was neither needed nor appro-priate to develop new guidance, rather, bringing current technical and implementation guidange together in one Generic Letter and make the SRP, Tech Specs, and licenses consistent would suffice. The Generic Letter makes clear (1) that extensions to the 50.48(c) schedules will no longer be granted, (2) that an expedited inspection program will be instituted to see what fire protection fixes are in place and give licensees the inspection team judgements on the acceptability of future modifications, and (3) that the licensee judgements must be backed by documented and valid analyses. The SC believes that this will demonstrate to the licensee what action he must take and what our inspections will look for. The Generic Letter notes that, although the 50.48(c) schedules will not be extended, the relative safety priroities of fire protection modifications need to be considered in the development of "living schedules." One item of guidance in the Generic Letter that had not been uniformly disseminated is that the QA applicable to fire protection features is that required by GDC-1. This would not attempt to backfit any QA requirements. Rather it would assure that future design, procurement, installation, testing and maintenance of fire protection features would receive high industrial quality attention. The SC believes that this initiative fully responds to the first three issues in your memo of September 13. Turning now to the inspection program, the SC found that the current inspections are generally satisfactory but that steps must be taken to indicate NRC's view of the importance of expediting implementation of Appendix R. These steps are to (1) speed up the inspection process, (2) develop a sound policy for fire protection enforcement actions, and (3) issue enforcement actions currently pending. These steps, in our view, would help expedite licensee compliance because it would raise industry's Enclosure to GL 85-01, Re: Fire Protection Policy
6 . William J. Dircks OCT 2 61984 awareness to NRC's resolve in this area and, more importantly, would allow the teams to judge the current direction (for licensees still designing or installing fixes) and advise the licensee on its acceptability. This should save both industry and staff resources in the long run. In the short run, that is 1985, the SC believes that adequate resources exist for the inspection teams (one for R-I, R-II and R-III, and one for R-IV/V) to be taken from regional staff, augmented by contractor, NRR, and IE assistance. While this would take a modest amount of reprograming in the regional inspection program, we suggest that it's worth the effort to get Appendix R implementation behind us. Prior to the 1985 inspection, a several-day inspection team workshop would be held to discuss the inspection program, the technical issues, and reach a common understanding on acceptability of various configurations and required documentation. Since this workshop cannot solve all potential problems the inspection teams will encounter, a team at HQ would be set up to promptly resolve significant differences between the inspection per-sonnel and the licensee. This referee team would be headed by NRR (SES level) and would have an NRR, an IE and a regional technical member. Their decision would be issued in one week and would be sent to all teams for their information. The SC believes that this program of expedited inspections, aimed-at reaching ors and NT0Ls and to include at least one site from each licensee not previously subject to a fire protection inspection, coupled with denial of future schedular exemptions and a fire enforcement policy will result in a fair and uniform speed up of Appendix R compliance. Further, since the
~
resource cost is believed to be modestly above the already-programmed fire protection inspections, we believe the cost is well worth it and will even benefit industry by correcting false starts in Appendix R implementation where they are found. Although we found the current inspections adequate (fifth item in your memo of September 13), this program will continue to be focussed on safe shutdown, will be implemented more expeditiously, and will build on the resolution of other initiatives considered by the SC. A Temporary Instruction for this inspection program has been drafted and is undergoing final revisions. It will be in final form by November 15, 1984 and will include the elements discussed above, for example, the team set up to resolve inspection differences with the licensee. Finally, the SC considered means to assure and maintain consistent levels of fire protection safety at all plants. The Working Group researched the guidance documents currently available and how these are applied to old and 4 new plants. The SC discussed findings of the reviewers and inspectors who are close to the issues. As a result of this work, the SC found that the requirements for old and new plants were generally the same but that discrepancies do exist. The application of guidelines, both in the revies,.' process and the inspection process, leaves room for interpretation. The SC concluded that several steps needed to be taken in addition to those described above some of which were to assure and maintain consistency. O Enclosure to GL 85-01, Re: Fire Protection Policy
- O l
William J. Dircks OCT 2 61984 i These steps are to develop and implement standard fire protection license conditions, Standard Feview Plan, and Standard Tech Specs for all plants. The license condition developed for this is along the lines of the security plan and QA program (Enclosure 5). The Standard Review Plan should need minimum revision to assure that Appendix R is fully included. The Tech Specs, however, will require more research and development by the Working Group. We need to assure that the Tech Specs are soundly based to assure functioning of fire protection features but which require only those activities which are commensurate with other Tech Spec items in tenns of importance to safety. As part of the above tasks, all fire protection guidance needs review and culling to assure that only a consistent and appropriate set remains. This complete set of guidance will be referenced in the Standard Review Plan revision. A last step in achieving uniform technical requirements is the SC recommendation to designate an office as responsible for awareness and resolution of interoffice / region fire protection issues. This is felt to be needed since current fire protection review is conducted within three divisions within NRR and one in IE. Although there is a lead branch responsibility, it is not always kept informed and involved. Therefore, the SC believes that the Director, Division of Engineering in NRR should be designated as the central point of contact. Conclusion The Fire Protection Policy Steering Committee concludes that the actions described above will accomplish the goals set forth in your memo of
" expediting Appendix R compliance for older plants and assuring consistent levels of fire protection safety at all plants." We believe that these actions will facilitate industry implementation of Appendix R through the use of the " interpretations" and a consistent set of guidance, yet will provide the necessary regulatory tools to guide, monitor, and, where appropriate, enforce this implementation process. We feel strongly that the actions we propose are synergistic and therefore all need to be completed to be most effective.
The Fire Protection Policy Steering Comittee has found the assignment to be challenging and rewarding. We would be pleased to brief you on our efforts at your earliest convenience. I Fire Protection Policy Steering Comittee Enclosure to GL 85-01, Re: Fire Protection Policy
O O William J. Oircks OCT 2 6 ng4 FIRE PROTECTION POLICY STEERING COMMITTEE
< t
. Nelson Grace, Director Quality Assurance, Safeguards & Inspection Programs Office of Inspection and Enforcement
'ThomasT. Martin,Direcpr
[ Division of Engineering"and Technical Programs Region I 7-od ; William J. 61hstead, Director and Chief Counsel Regulations Division Office of the Executive Legal Director J n A. Olshinski, Director N ivision of Reactor Safety Region II
@ad M Richard L. Spiessardg Director Division of Reactor Safety Region III
/ Richard H. Vollmer, Directo A Division of Engineering Office of Nuclear Resctor Regulation (Chairman) s Enclosure to GL 85-01, Re: Fire Protection Policy
s l William J. Dircks 007 2 s us4 j
Enclosures:
- 1. Memo to H. Denton et al on Review of NRC Fire Protection Policy and Programs.
- 2. Generic Letter on Fire Protection
- 3. Interpretations of Appendix R
- 4. Guidance for Enforcement Actions Concerning Fire Protection Requirements
- 5. Fire Protection License Condition
- 6. Appendix R Questions and Answers
- 7. Steering Committee Memoranda i
j i i i ) i .i i l l ! Enc 1nsure to GL 85-01, Re: Fire Protection Policy
- e. .
UNITED STATES i NUCLEAR REGULATORY COMMISSION ENCLOSURE 1 msmuorow. o. c. nosss
- SEP 131984 MEMORANDUM FOR
- Harold R. Denton, Director, NRR Richard C. DeYoung, Director, IE Thomas E. Murley, Regional Administrator, R-I James P. O'Reilly, Regional Administrator, R-Il James G. Keppler, Regional Administrator, R-III i John T. Collins, Regional Administrator, R-IV John B. Martin, Regional Administrator, R-V FROM: William J. Dircks Executive Director for Operations
SUBJECT:
REVIEW OF NRC FIRE PROTECTION POLICY AND PROGRAMS At our Aue_g.st 27th meeting on fire protection, I directed NRR to review l and make recommenda.tions for resolving current fire protection issues within two months. The objective of such an ef. fort is to examine all current licensing, inspection, and technical issues and, based on such a review, develop policy recomendations aimid at expediting Appe.ndix R compliance for older slants and assuring consistent levels of fire protection safety at all plants. Some of the current issues that should be considered are:
. adequacy of current guidance to inoustry;
. interpretation of the Appendix R requirements vice staff guidance;
. treatment of expected future technical and schedular exemptions into late 1980s and early 1990s;
. comparison of Appendix R and current NTOL plants for fire protection safety;
. adequacy of current inspection practices; and
. identification and resolution plan for any outstanding technical issues.
i To implement the above, ! understand that you have set up a Steering ! Committee composed of Messrs. Grace, IE; Martin, R-1; Olmstead, ELD; Enclosure to GL 85-01, Re: Fire Protection Policy l i l t _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
e a , 2 Olshinski, R-11; Spessard, R-III; and Vollmer, NRR, Chairman. This ! Steering Committee is to decide the scope of issues to be considered, ! meet with HQ and regional personnel as necessary to consider these issues, make assignments as appropriate to a working group headed by Faust Rosa, NRR, for detailed consideration of certain issues, and make recommendations for actions along with supporting bases to me by October 26, 1984 I concur in these assignments and the general charter of the Steering Cannittee. All ongoing regulatory actions in your Offices regarding fire protection should be continued and should not be delayed or deferred pending the outcome of this review.
$!ps$VGliam1.D'fa$ .
William J. Dircks Executive Director for Operations cc: V; Stello G. Cunningham Steering Committee , i Enclosure to GL 85-01, Re: Fire Protection Policy
ENCLOSURE 2 GENERIC LETTER ON FIRE PROTECTION TO ALL LICENSEES AND APPLICANTS FOR OPERATING LICENSES Gentlemen:
SUBJECT:
IMPLEMENTATION OF FIRE PROTECTION REQUIREMENTS In the Spring of 1984, the Comission held a series of Regional Workshops on the implementation of NRC fire protection requirements at nuclear power plants. At those workshops, a package of recently-developed NRC guidance was distributed to each attendee which included NRC staff responses to industry questions and a document titled " Interpretations of Appendix R." The cover memo for the package explained that it was a draft package which would be issued in final form via Generic Letter following the workshops. The guidance approved by the Comission is appended to this letter, and is in the same format as the draft package, i.e., " Interpretations of Appendix R" and responses have been modified from the draft package, and a number of industry questions raised at or subsequent to the workshops have been added and answered. This package represents the official agency position on all issues covered, and where this guidance differs from previously issued guidance (including Generic Letter 83-33) on this subject, this guidance takes precedence. In the lettered sections below, some additional topics are covered which also bear on the interpretation and implementation of NRC fire protection requirements. The topics are: (A) schedular exemptions, (B) revised inspection program, (C) documentation required to demonstrate compliance. (0)applicabilityofGDC-1tofireprotectionsystems,and(E) notification of the NRC when deficiencies are discovered. A. Schedular Exemptions The Appendix R implementation schedule was established by the Commission in 10 CFR 50.48(c), promulgated together with Appendix R in November,of 1980. Allowing time to evaluate the need for alternative or dedicated shutdown systems, which require prior NRC approval before installation, and time for design of and NRC review of such systems, the Commission envisioned that implementation of Appendix R would be complete in four to five years, or approximately by the end of 1985. Many schedule ex-tensions were granted by the staff under the " tolling provision," 50.48(c)(6), and under 10 CFR 50.12, the longest of which now extends into 1987. Some licensees have proceeded expeditiously to implement Enclosure to GL 85-01, Re: Fire Protection Policy
3*% Appendix R and are now finished or nearly finished with that effort. Others have engaged in lengthy negotiations with the staff while continuing to file requests for schedule extensions, and thereby have barely begun Appendix R modifications needed to comply with Sections III.G and III.L. Schedule extension requests have been received seeking implementation dates of 1990 or beyond. As the 50.48(c) schedule was intended to be a one-time schedule commencing in the 1980-1982 in the 1985 time frame time frame, extensions well beyond and ending(particularly where this schedule major modifications remain to be completed) undermine the purpose of the schedule, which was to achieve expeditious compliance with NRC fire protection requirements. The NRC will therefore grant no further extensions to the 50.48(c) schedules. When a licensee's schedule expires, compliance is expected and appropriate enforcement action will be taken. If compliance cannot be achieved by that date, the licensee will be required to submit and justify a minimum schedule for completion of fire protection modifications, and to supply interim measures to compensate for the lack of compliance. In submitting a schedule which goes beyond the current 50.48 deadline *, the licensee will be required to demonstrate that it has endeavored in good faith to complete modifications on schedule. A showing of good faith attempt to complete implementation on schedule may mitigate - enforcement action for noncompliance with NRC requirements. The NRC is currently reviewing all dockets of plants covered by the 50.48 schedule to determine schedule deadlines. When this review is completed, each licensee will be informed of the deadlines.
- 8. Revised Inspection Program In 1982, the NRC developed an inspection program to verify compliance with the requirements of 10 CFR 50, Appendix R. This program was
! primarily oriented towards reviewing safe shutdown features of those ! pre-1979 licensees that had completed Appendix R modifications and a selected NTOL plants. From 1982 to the present, approximately seven ! Appendix R compliance inspections have been performed. In a number of cases, these inspections have discovered that licensees had made signi-l ficant errors in implementing a number of Appendix R requirements. l In order to expedite compliance verification and to provide the NRC l staff with earlier indication of problems associated with , implementation of fire protection features, the NRC will conduct fire l protection inspections of operating plants and plants currently under-going operating license review during 1985 to include at least one i l l
- Licensees submitting "living schedules" for NRC approval should be l aware that existing 50.48 schedules continue to apply. Licensees i
intending to include fire protection modifications within a "living schedule" are expected to assign within such schedules the relative safety priorities of remaining fire protection modifications. 4 Enclosure to GL 85-01, Re: Fire Protection Policy tw.
o . 3-site from each licensee who has not been subject to a previous NRC fire protection inspection. This inspection will review completed modifications and, in the case of incomplete modifications, review licensee plans and schedules for completing such modifications. C. Documentation Required to Demonstrate Compliance The " Interpretations" document attached to this letter states that, where the licensee chooses not to seek prior NRC review and approval of, for example, a fire area boundary, an evaluation must be performed by a fire protection engineer (assisted by others as needed) and retained for future NRC oudit. Evaluations of this type must be written and organized to facilitate review by a person not involved in the evaluation. Guidelines for what such an evaluation should contain may be found in: (1) Section B of Appendix R and (2) Section C.1.b of Branch Technical Position (BTP) CMEB 9.5-1 Rev. 2 dated July 1981. All calculations supporting the evaluation should be available and all assumptions clearly stated at the outset. Failure to have such an evaluation available for an area where compliance with Appendix R is not readily demonstrated will be taken as prima facie evidence that the area does not comply with NRC requirements, and may result in enforce-ment action. D. Quality Assurance Requirements Applicable Fire protection systems must meet the requirements of General Design Criterion 1 of Appendix A to 10 CFR Part 50. For such systems the licensee is therefore required to have and maintain a quality assurance program adequate to assure that these systems will perform their functions when called upon. Fire protection systems are not " safety-related" and are therefore not within the scope of Appendix B to 10 CFR Part 50, unless the licensee has consnitted to include these systems under the Appendix B program for the plant. NRC guidance for an acceptable quality assurance program for fire protection systems, given in Section C.4 of Branch Technical Position CMEB 9.5-1 Rev.2 dated July 1981, has generally been used in the review and acceptance of approved fire protection programs. E. Notification of the NRC When Deficiencies are Discovered Licensees are reminded of their obligation to notify the NRC of fire protection deficiencies which meet the criteria of 10 CFR 50.72 or 10 CFR 50.73 as applicable. Enclosure to GL 85-01, Re: Fire Protection Policy
Enclosure to Gl. 85-01, Re: Fire Protection Policy
a . ENCLOSURE 3 INTERPRETATIONS OF APPENDIX R
- 1. Process Monitoring Instrumentation Section III.L.2.d of Appendix R to 10 CFR Part 50 states that "the process monitoring function shall be capable of providing direct readings of the process variables necessary to perform and control" the reactivity control function. In I&E Information Notice 84-09, the staff provides a listing of instrumentation acceptable to and preferred by the staff to demonstrate compliance with this provision. While this guidance provides an acceptable method for compliance with the regulation, it does not exclude other alter-native methods of compliance. Accordingly, a licensee may propose to the staff alternative instrumentation to comply with the regulation. While such a submittal is not an exemption request, it must be justified based on a technical evaluation. The licensee may also propose alternatives to actual compliance with the regulation (e.g., instrumentation which does not provide a direct reading of the process variable) by filing an exemption request with adequate justification.
- 2. Repair of Cold Shutdown Equipment Section III.L.5 of Appendix R states that when in the alternative or dedicated shutdown mode, " equipment and systems comprising the means to achieve and maintain cold shutdown conditions shall not be damaged by fire; or the fire damage to such equipment and systems shall be limited so that the systems can be made operable and cold shutdown can be achieved within 72 hours."
This is not to be confused with the requirements in Section !!!.G.1.b of Appendix R. Section !!!.G.I.b contains the requirements for normal shutdown modes utilizing the control room or emergency control station (s) capabilities. The fire areas falling under the requirements of III.G.I.b are those for which an alternative or dedicated shutdown capability is not being provided. For these fire areas. Section III.G.1.b requires only the capability to repair the systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station (s) within 72 hours, not the capability to repair and achieve cold shutdown within 72 hours as required for the alternative or dedicated shutdown modes by Section !!!.L (noted above). With regard to areas involving normal shutdown, however, Section ! of Appendix R states that repairs must be made using only onsite capabilities. After repairs are made, cold shutdown can be achieved on a reasonable schedule using any available power source. 1 Enclosure to GL 85-01, Re: Fire Protection Policy
s
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- 3. Fire Damage Appendix R to 10 CFR part 50 utilizes the term " free of fire damage."
In promulgating Appendix R, the Commission has provided methods acceptable for assuring that necessary structures, systems and components are free of fire damage (see Section !!!.G.2a, b and c), that is, the structure, system or component under consideration is capable of performing its intended function during and after the postulated fire, as needed. Licensees seeking exemptions from Section !!!.G.2 must show that the alternative proposed provides reasonable assurance that this criterion is met. (Notealso that Section !!!.G.2 applies only to equipment needed for hot shutdown. Therefore, an exemption from !!!.G.2 for cold shutdown equipment is not needed.)
- 4. Fire Area Boundaries The term " fire area" as used in Appendix R means an area sufficiently bounded to withstand the hazards associated with the area and, as necessary, to protect important equipment within the area from a fire outside the area. In order to meet the regulation, fire area boundaries need not be completely sealed floor-to-ceiling, wall-to-wall boundaries. However, all unsealed openings should be identified and evaluated. Where fire area boundaries were not approved under the BTP process, or where such boundaries _')
are not wall-to-wall, floor-to-ceiling boundaries with all penetrations / sealed to the fire rating required of the boundaries, licensees must perform ' an evaluation to assess the adequacy of fire boundaries in their plants to determine if the boundaries will withstand the hazards associated with the area and protect important equipment within the area from a fire outside the area. This analysis must be performed by at least a fire protection engineer and, if required, a systems engineer. Although not required, licensees may submit their evaluations for staff review and concurrence. In any event, these analyses must be retained by the licensees for sub-sequent NRC audits.
- 5. Automatic Detection and Suppression Sections !!!.G.2.b and !!!.G.2.c of Appendix R state that "In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area..." Other provisions of Appendix R also use the phrase
" fire detectors and an automatic fire suppression system in the fire area..."
(see e.g., Section !!!.G.2.e). In order to comply with these provisions, suppression and detection suf-ficient to protect against the hazards of the area must be installed. In this regard, detection and suppression providing less than full area coverage may be adequate to comply with the regulation. Where full area suppression and detection is not installed, licensees must perform an evaluation to assess the adequacy of partial suppression and detection to protect against the hazards in the area. The evaluation must be performed Enclosure to GL 85 01, Ret Fire Protection Policy
3 by a fire protection engineer and, if required, a systems engineer. Although not required, licensees may submit their evaluations to the staff for review and concurrence. In ally event, the evaluations must be retained for subsequent NRC audits. Where a licensee is providing no suppression or detection, an exemption must be requested. 6.' Alternative or Dedicated Shutdown Section !!!.G.3 of Appendix R provides for " alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration." While "in-dependence" is clearly achieved where alternative shutdown equipment is out-side the fire area under consideration, this is not intended to imply that alternative shutdown equipment in the same fire area but independent of the room or the zone cannot result in compliance with the regulation. The
" room" concept must be justified by submission of a detailed fire hazards analysis that demonstrates a single fire will not disable both normal shutdown equipment and the alternative shutdown capability.
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l. a Enclosure to GL 85 01, Ret Fire Protection Policy
o . ENCLOSURE 4 GUIDANCE FOR ENFORCEMENT ACTIONS CONCERNING FIRE PROTECTION REQUIREMENTS
- 1. General Guidance A. Fire protection requirements are delineated by 10 CFR 50 Appendix A General Design Criterion 3, 10 CFR 50.48, 10 CFR 50 Appendix R. Facility License Conditions, facility technical specifications and other legally binding requirements, as applicable. A Notice of Violation will be issued for violation of requirements. However, failure to meet fire protection com-mitments will be designated as deviations.
B. Failures to meet regulatory requirements for protecting trains of equipment required for achieving and maintaining safe hot or cold shutdown are serious violations. The specific violations should be reviewed individually and as a group to determine their root cause(s). This guidance gives examples of violations at various severity levels and should be used to determine the appropriate enforcement action. For purposes of this guidance, required structures, systems, and components are those which are necessary to achieve and maintain hot and/or cold safe shutdown and which require the application of fire protection features as described in the licensee's fire hazards analysis report and safety evalua-tion, report. C. Fire protection violations may involve inoperable or inadequate: fire barriers, separation, suppression or detection systems, repair parts, procedures or other conditions or items required to protect safe shutdown equipment from fire and/or pennit the operation of safe shutdown equipment during a fire or to restore safe shutdown equipment to service following an actual fire. D. Numerous violations of fire protection requirements which in-dividually may be classified at lower severity levels may cumulatively be classified at a higher severity level due to inadequate implementation of the fire protection program. Enclosure to GL 85-01, Rei Fire Protection Policy
2 4 l
- 2. Severity Categories A. Severity !. Violations of fire protection requirements estab.
lished to protect or enable operation of safe hot shutdown equip- ' ment concurrent with an actual fire which damages that equipment , such that safe hot shutdown could not be achieved or maintained
- using the equipment dedicated for the purpose.
B. Severity !!. Violations of fire protection requirements established , to protect or enable operation of safe cold shutdown equipment concurrent with an actual fire which damages that equipment such that safe cold shutdown could not have been achieved and maintained using the equipment dedicated for this purpose in accordance with the applicable requirements, t C. Severity !!!. Violations of fire protection requirements established to protect or enable operation of safe shutdown equipment such that a fire in the area could damage that equipment to the extent that safe hot or cold shutdown could not have been achieved and l maintained using the equipment dedicated for this purpose in 3 accordance with applicable requirements. Failure to have a written
, evaluation available for an area where compliance with Appendix R .
is not readily demonstrated will be taken as prima facie evidence ; l that the area does not comply with NRC requirements and may result ; in enforcement action at the severity level. I D. Severity IV. Violations of one or more fire protection requirements i that do not result in a Severity Level ! i whichhavemorethanminorsafetyorenvIro!!or!!!violationand nmental significance. E. Severity V. Violations of one or more fire protection requirements > that have minor safety or environmental significance. t i Enclosure to GL 85 01, Re: Fire Protection Policy
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d ,8 k ' t i ENCLOSURE 5 4 I i FIRE PROTECTION LICENSE CONDITION l 1 ) 1. The licensee shall implement and maintain in effect all provisions ! of the approved fire protection program as described in the Final i Safety Analysis Report for the faci ity and as approved in the SER j subject to provisions 2 and 3 below. ( ) 2. The licensee may make no change to the approved fire protection program I which would decrease the level of fire protection in the plant without [ 3 prior approval of the Comission. To make such a change the licensee j must submit an application for license amendment pursuant to 10 CFR 50.90. j ]
- 3. The licensee may make changes to features of the approved fire pro- !
1 tection program which do not decrease the level of fire protection j , without prior Comission approval provided (a) such changes do not , i otherwise involve a change in a license condition or technical spect- i fication or result in an unreviewed safety question (see 10 CFR 50.59), j ] and (b) such changes do not result in failure to carry out the fire ; protection program approved by the Comission prior to licence issuance. d r i The licensee shall maintain, in an auditable form, a current record of i all such changes, including an analysis of the effects of the change on l l the fire protection program, and shall make such records available to NRC inspectors upon request. All changes to the approved program made without j prior Comission approval shall be reported annually to the Director of i the Office of Nuclear Reactor Regulation, together with supporting j analyses. l 4 l l l l i i j ; 1 l ! i j ! r i ! r Enclosurw to GL 85 01, Ret Fire Protection Policy i
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I i i 4 i Enclosure to GL 85 01, Ret fire Protection Policy [ t
O e APPEN0!X R l QUESTIONS AND ANSWERS
- 1. INTRODUCTION
- 2. OVERVIEW
- 3. !!! G FIRE PROTECTION OF SAFE SHUTDOWN CAPA81LITY 3.1 Fire Area Boundaries 3.1.1 Fire Area Definition 3.1.2 Previously Accepted Fire Area Boundaries 1 3.1. 3 Exterior Walls i 3.1.4 Exterior Yards 3.1.5 Fire Zones 3.1.6 Documentation 3.2 Fire Barrier Qualifications 3.2.1 Acceptance Criteria 3.2.2 Deviations from Tested Configurations 3.2.3 Fire Door Modifications 3.3 Structural Steel 3.3.1 NFPA Approaches 3.3.2 Previously Accepted Structural Steel 3.3.3 Seismic Supports 3.3.4 Cable Tray Support Protection 3.4 Automatic Suppression System 3.4.1 Water Density 3.4.2 NRC Consultation 3.4.3 Sprinkler Location 3.4.4 Fixed Suppression System In Fire Area 3.4.5 Sprinkler Head Location 3.4.6 Previously Approved Suppression Systems
- 3. 5 Separation of Redundant Circuits 3.5.1 Twenty Foot Separation Criteria 3.5.2 Floor to Floor Separation
Appendix R 2 -- 3.6 Intervening Combustibles 3.6.1 Negligible Quantities of Intervening Combustibles 3.6.2 In Situ Exposed Combustibles 3.6.3 Unexposed Combustibles
; 3.7 Radiant Energy Shields 3.7.1 Fire Rating 3.8 Design Bases ; 3.8.1 Fire Protection Features NFPA Conformance 3.8.2 Design Basis Fire l 3.8.3 Redundant Trains / Alternate Shutdown 3.8.4 Control Room Fire Considerations
- 4. !!! J. EMERGENCY LIGHTING 4.1 Illumination Levels
- 5. !!! L ALTERNATIVE AND DEDICATED SHUTDOWN CAPA81LITY I'
5.1 Safe and Alternative Shutdown > 5.1.1 Previously Accepted Alternative Shutdown Capability 5.1. 2 Pre-Existing Alternative Shutdown capability 5.1. 3 !!! L Backfit 5.2 Procedures
- 5. 2.1 Shutdown and Repair Basis 5.2.2 Post Fire Operating Procedures 5.2.3 Alternative Shutdown Capability 5.2.4 Post Fire Procedures Guidance Documents
- 5.3 Safe Shutdown and Fire Damage 5.3.1 Circuit Failure Modes 5.3.2 " Hot Short" Ouration 5.3.3 Hot Shutdown Duration 5.3.4 Cooldown Equipment 5.3.5 Pressurizer Heaters 5.3.6 On-Site Power !
5.3.7 Torus Level Indication 5.3.8 Short Circuit Coordination Studies 1 5.3.9 Diagnostic Instrumentation 5.3.10 Design Basis Plant Transients i 5.3.11 Alternate / Dedicated Shutdown vs. Remote Shutdown Systems Enclosure to GL 85 01. Ret Fire Protection Policy
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. ENCLOSURE 6 r
r, f I i APPENDIX R QUESTIONS AND ANSWERS
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i i I i 1 i l l* I I l Enclosure to GL 85-01, Re: Fire Protection Policy
Appendix R 3
- 6. III 0, OIL COLLECTION SYSTEMS FOR REACTOR COOLANT PUMP 6.1 Lube Oil System Seismic Design 6.2 Container
- 7. BRANCH TECHNICAL POSITION CMEB 9.5-1 7.1 Fire Protection and Seismic Events 7.2 Random Fire and Seismic Events
- 8. LICENSING POLICY 8.1 Fire Hazards Analysis / Fire Protection Plan Updating 8.2 Fire Protection License Condition 8.3 III G. J and 0 Exemptions for Future Modifications 8.4 Future Changes 8.5 Schedular and Blanket Exemptions 8.6 Trivial Deviations 8.7 Revised Modifications 8.8 Smallest Opening in a Fire Barrier 8.9 NFPA Code Deviations 8.10 " ASTM E-119" Design Basis Fire 8.11 Plants Licensed After January 1, 1979 8.12 Cold Shutdown Equipment Availability 8.13 Guidance Documents 8.14 Deviations from Guidance Documents 8.15 Staff Interpretations of Appendix R 8.16 Dissemination of New Staff Positions 8.17 Equivalent Alternatives 8.18 Coordination Study Updates 8.19 Exemption Request Threshold 8.19.1 Penetration Designs Not Laboratory Approved l 8.19.2 Individual vs. Package Exemptions 8.19.3 Exemption Request Supporting Detail 8.19.4 50.12 vs. 50.48 Exemption Requests 8.21 Post January 1,1979 Plants and Exemption Requests 8.22 NRC Approval for BTP CMEB 9.5-1 Deviations
- 9. INSPECTION POLICY 9.1 Safety Implications 9.2 Uniform Enforcement 9.3 NTOL Inspections Enclosure to GL 85-01, Re: Fire Protection Policy
s Appendix R 4 9.4 Future TI 2515/62 Revisions 9.5 Documentation Supplied by Licensee 9.6 Subsequent Inspections 9.7 NRC List of Conforming Items 9.8 Inspection Re-review
- 9. 9 List of Shutdown Equipment Enclosure to GL 85-01, Re: Fire Protection Policy
1 INTRODUCTION A major fire damaging safe shutdown equipment occurred at the Browns Ferry Nuclear Station in March 1975. The fire damaged over 1600 electrical cables and caused the temporary unavailability of some core cooling systems. Be-cause this fire did substantial damage, the NRC established a Special Review Group which initiated an evaluation of the need for improving the fire protection programs at all nuclear power plants. The group found serious design inadequacies regarding fire protection at Browns Ferry, and its report, " Recommendations Related to Browns Ferry Fire" (NUREG-0050, February 1976), contained over fifty recommendations regarding improve-ments in fire prevention and control in existing facilities. The report also called for the development of specific guidance for implementing fire protection regulations, and for a comparison of that guidance with the fire pratection program at each operating plant. NRC developed technical guidance from the technical recommendations in the Special Group's report, and issued those guidelines as Branch Technical Position Auxiliary Power Conversion Systems Branch 9.5-1 (BTP APCSB 9.5-1), 1/ " Guidelines for Fire Protection for Nuclear Power Plants." This juidance did not apply to plants operating at that time. Guidance to operating plants was provided later in Appendix A 2/ to BTP APCSB 9.5-1 which, to the extent practicable, relies on BTP APCSB 9.5-1. The guidance in these documen.ts was also published for public comment as Regulatory Guide 1.120,
" Fire Protection for Nuclear Power Plants" (June 1976). In response to public comment, the NRC issued an extensively revised version of Regulatory Guide 1.120 for further public comment.
-1/ Rather than serving as inflexible, legal requirements that must be followed by licensees, issuances such as regulatory guides and branch technical positions are meant to give guidance to licensees concerning those methods the staff finds acceptable for implementing the general criteria embodied in the NRC's rules. See, le. ., Petition for Emergency
& Remedial Action, CLI-78-6, 7 NRC 400, 406 (1978); Gulf States Utilities Company (River Bend Station, Units 1 and 2) ALAB-444, 6 NRC 760, 772 (1977).
-2/ Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976.
Enclosure to GL 85-01, Re: Fire Protection Policy
2 In May 1976, the NRC asked licensees to compare operating reactors with BTP APCS 8 9.5-1, and in September 1976, those licensees were informed that the guidelines in Appendix A would be used to analyze the consequences of a fire in each plant area. In September 1976 the licensees, were also requested to provide a fire hazards analysis that divided the plant into distinct fire areas and show that redundant systems required to achieve and maintain cold shutdown are adequately protected against damage by a fire. Early in 1977 each licensee responded v.ith a Fire Protection Program Evaluation which included a Fire Hazard Analysis. These evaluations and analyses identified aspects of licensees' fire protection programs that did not conform to the NRC guidelines. Thereafter, the staff initiated discussions with all licensees aimed at achieving implementation of fire protection guidelines by October 1980. The staff held many meetings with licensees, conducted extensive correspondence with them, and visited every operating reactor. As a result, many fire protection items were resolved, and agreements were included in Fire Protection Safety Evaluation Reports issued by the NRC. Several fire protection issues remained unresolved with a number of licensees. By early 1980, most operating plants had implemented most of the guidelines in Appendix A. However, as the Commission noted in its Order of May 23, 1980, the fire protection program has had some significant problems with implementation. Despite the staff's efforts, several licensees had expressed continuing disagreement with, and refused to adopt recommendations relating to several generic issues, including the requirements for fire brigade size and training, water supplies for fire suppression systems, alternate and dedicated shutdown capability, emergency lighting, qualifications of seals used to enclose places where cables penetrated fire barriers, and the prevention of reactor coolant pump lubrication system fires. To estab-lish a definitive resolution of these contested subjects in a manner con-sistent with the general guidelines in Appendix A to the BTP and to assure timely compliance by licensees, the Commission issued a proposed fire protection rule and its Appendix R, which was described as setting out minimum fire protection requirements for the unresolved issues (45 Fed. Reg. 36082 May 29, 1980). The fire protection features addressed included protection of safe shutdown capability, emergency lighting, fire barriers, associated circuits, reactor coolant pump lubrication system, and alternate shutdown systems. The Commission stated that it expected all modifications (except for alternate and dedicated shutdown capability) to be implemented by November 1, 1980. 4/ 3/ 11 NRC 707, 718 (1980) 4/ Id. at 719 Enclosure to GL 85-01, Re: Fire Protection Policy
r 3 1 As originally proposed (Federal Register Vol. 45. No. 1&5, May 22, 1980), Appendix R would have applied to all plants including those for which the staff had previously accepted other fire protection modifications. After analyzing comments on the rule, the Commission determined that only three of the fifteen items in Appendix R were of such safety significance that they should apply to all plants, including those for which alternative fire protection actions had been approved previously by the staff. These items are protection of safe shutdown capability (including alternate shut-down systems), emergency lighting, and the reactor coolant pump lubrication system. Accordingly, the final rule required all reactors licensed to operate before January 1,1979, to comply with these three items even if the NRC had previously approved alternative fire protection features in these areas (45 Fed. Reg. 76602 Nov. 19, 1980). However, the final rule is more flexible than the proposed rule because Item III. G now provides three alternative fire protection features which do not require analysis to demonstrate the protection of redundant safe shutdown equipment, and reduces the acceptable distance in the physical separation alternative from fifty feet to twenty feet. In addition, the rule now also provides an exemption procedure which can be initiated by a licensee's assertion that any required fire protection feature will not enhance fire protection safety in the facility or that such modifications may be detrimental to overall safety (10 CFR 50.48(c)(6)). If the Director, Nuclear Reactor Regulation determines that a licensee has made a prima facie showing of a sound technical basis for such an assertion, then the implementation dates of the rule are tolled until final Commission action on the exemption request. Most licensees requested and were granted additional time to perform their reanalysis, propose modifications to improve post fire shutdown capability and to identify exemptions for certain fire protection configurations. In reviewing some exemption requests, the staff noted that some licensees had made significantly different interpretations of certain requirements. These differences were identified in the staff's draft SER's. These differences were also discussed on several occasions with the cognizant licensee as well as the Nuclear Utility Fire Protection Group. These discussions culminated in the issuance of generic letter 83-33. ( Enclosure to GL 85-01, Re: Fire Protection Policy
4 ,
- 2. OVERVIEW Section 50.48 Fire Protection of 10 CFR Part 50 requires that each operating nuclear power plant have a fire protection plan that satisfies General Design Criterion 3 of Appendix A to 10 CFR 50. It specifies what should be contained in such a plan and lists the basic fire protection guidelines for this plan. It requires that the Fire Protection Safety Evaluation Report which has been issued for each operating plant state how these guidelines were applied to each facility.
Section 50.48 also requires that all plants with operating licenses prior to January 1, 1979 satisfy the requirements of Section III.G, III.J and III.0, and other Sections of Appendix R where approval of similar features had not been obtained prior to the effective date of Appendix R. By a separate action, the Commission approved the staff's requirement that all plants to receive their operating license after January 1,1979 also sat-isfy the requirements of Sections III.G, III.J and III.0 and that a fire protection license condition be established. Deviations from Appendix R requirements for pre-1979 plants are processed under the exemption process. Deviation from other guidelines are identified and evaluated in the Safety Evaluation Report. i A standard fire protection license condition has been developed and will be included in each new operating license. Present operating licenses will be amended to include the standard license condition. The Regions initiated inspections of operating plants and identified several significant items of non-compliance. The Nuclear Utility Fire Protection Group requested interpretations of certain Appendix R requirements and provided a list of questions that they thought should be discussed with the industry. The NRC held wo.rkshops in each Region to assist the industry in understanding the NRC's requirements and to improve the Staff's under-standing of the industry's concerns. This document presents the NRC's response to the questions posed by the l industry and supplemented with additional questions identified at the workshops as being of interest to the industry or the staff. These re-sponses may be used as guidance for design, review and inspection activi-ties. The questons have been reformatted according to their applicability l to Sections of Appendix R, BTP CMEB 9.5-1, licensing policy or inspection policy. 1 t Enclosure to GL 85-01, Re: Fire Protection Policy
5
- 3. SECTION III G, FIRE PROTECTION OF SAFE SHUTDOWN CAPABILITY 3.1 Fire Area Boundaries 3.1.1 Fire Area Definition Question Section III.G states the fire protection features required for cables and equipment or redundant trains of systems required to achieve and maintain hot shutdown that are located within the same fire area. Is the fire area of '
Section III.G, the same fire area referred to in BTP APCSB 9.5-1, Appendix A; and the supplementary guidance of September 1976?
Response
Yes. Prior to the issuance of Appendix R. fire area boundaries should have been established using the BTP guidelines. The concept of fire areas was described in BTP APCSB 9.5-1. Also, definitions were given for fire areas, fire barriers and fire ratings. The same fire areas were referred to in our supplementary guidance of 1977 and Appendix A to BTP APCSB 9.5-1. The same fire areas concepts definitions are carried over to BTP CMEB-9.5-1. During the " Appendix A" reviews, some licensees performed their fire hazards analysis using these definitions, some did not. Licensees sometimes called " fire zones" " fire areas." Section III.G sets forth fire protection alterna-tives within a fire area. If new fire areas are identified they should be established using the BTP guidelines. The concept of fire areas was described in BTP APCSB 9.5-1:
"C. Establishment and Use of Fire Areas The concept of separate fire areas for each division of safety equipment which requires redundancy will facilitate the installation of automatic water extinguishing systems since it will reduce the possibility of water damaging redundant safety-related equipment.
Fire areas should be established based upon the amount of combustible material present and considering suitably chosen design basis fires so that adequate protection can be provided for safety-related systems and equipment. Design basis L fires are those fires that result in the most severe exposure Enclosure to GL 85-01, Re: Fire Protection Policy
6 to the area or systems being considered. For this condition, it is assumed that no manual or automatic fire suppression action has been started and the fire has reached its peak burning rate and involves all combustibles present. Within each area special attention should be given to limiting the amount of combustible material and to providing effective barriers and fire resistive coatings to reduce the spreading of a fire in these areas. A design basis fire should be assumed and provisions should be made to limit the consequence of such a fire by providing fire barriers with suitable separation between redundant systems and components which are provided to carry out required safety functions. This separation is enhanced if the plant is divided into suitable fire areas since redundant safety equipment can then be placed in separate fire areas. Particular design attention should be given to the use of separate isolated fire areas for redundant cables to avoid loss of redundant safety-related cables. Provisions should also be made to limit the consequences of a fire by suitable design of the ventilation systems so that the spread of the products of combustion to other areas of the plant is prevented. Means should be provided to ventilate or isolate the area as required. The power
, supply and controls for the area ventilation system should be from outside the area, and the power and control cables should not pass through the area.
The fire detection systems should be designed using detectors of the right types at locations suitable to detect the particular type of fire expected in each area. In the design, consideration should be given to provide personnel access to and escape routes froa cach fire area. The emergency plans for all plants should lay out access and escape routes to cover the event of a fire in critical areas of the plant." In addition, definitions of the basic fire area components were given:
" Fire Area - that portion of a building or plant that is separated from other areas by boundary fire barriers (walls, floors or roofs),with any openings or penetratione protected with seals or closures having a fire resistance rating equal to that of the barrier.
Fire Barrier - those components of construction (walls, ' floors and roofs) that are rated by approving laboratorias , in hours for resistance to fire to prevent the spread of fire. Enclosure to GL 85-01, Re: Fire Protection Policy
7 Fire Rating - refers to the endurance period of a fire barrier or structure and defines the period of resistance to a standard fire exposure elapsing before the first critical point in behavior is observed. (Refer to NFPA 251). Fire Zones - subdivisions of fire areas in which the fire suppression systems are designed to combat particular types of fires. The concept of fire zone aids in defining to the fire fighter the fire parameters and the actions which would be necessary." The supplementary guidance, stated information to be provided in the fire hazards analysis for each fire area or fire zone established.
"In order to perform a proper fire hazards analysis, the services of a qualified fire protection engineer should be utilized. To demonstrate the results of the fire hazards analysis the following information must be provided:
- 1. Provide plan and elevation views of the plant that show the plant as divided into distinct fire areas.
Provide a description of the various systems, both safety-related and non safety-related, which occupy the fire area and could provide cooling to the core to safely shutdown the reactor, including decay heat removal. Provide a description of areas of the plant that contain radioactive material that may be released to the exclusion area or beyond should a fire occur in-those areas. For each fire area, provide the following: a) Describe the fire barrier that defines the fire area; the consequences of the design basis fire for that area; the consequences of the fire if the fire protection system functions as designed, b) Identify the safety related equipment and associated cabling. Provide the design criteria for the fire protection related to such equipment. Provide . the design criteria for protection of such equipment ! against inadvertent operation, careless operation or rupture of extinguishing systems. c) Provide a list of the type, quantity, and other pertinent characteristics of combustible materials associated with each fire area. e d) Provide a list of the fire loading which represent the combustibles identified in (c) above for each fire area. Enclosure to GL 85-01, Re: Fire Protection Policy
t 9 3.1.2 Previously Accepted Fire Area Boundaries QUESTION If a fire area boundary was described as a rated barrier in the 1977 fire hazards analysis, no open items existed in this area in the Appendix A SER, and the barriers have not been altered, then need those barriers be reviewed by licensees or the Staff under Appendix R?
RESPONSE
If a fire area boundary was described as a rated barrier in the 1977 fire hazards analysis, and was evaluated and accepted in a published SER, the fire area boundary need not be reviewed as part of the re-analysis for compliance with Section III.G of Appendix R. Openings in the fire barriers, if any, should have been specifically identified and justified in the fire hazards analysis performed in the Appendix A process. If openings in the fire area boundaries were not previously evaluated, such an evaluation should be performed as a basis for assessing compliance with Appendix R. See Item #4 of the " Interpretations of Appendix R." In BTP CMEB 9.5-1, Fire Barrier is defined as:
" Fire Barrier - those components of construction (walls, floors, and the supports), including beams, joists, columns, penetration seals or closures, fire doors, and fire dampers that are rated by approving laboratories in hours of resist-ance to fire and are used to prevent the spread of fire."
The term " fire area" as used in Apnendix R means an area sufficiently bounded to withstand the hazards associated with the fire area and, as necessary, to protect important equipment within the fire area from a fire outside the area. ' In order to meet the regulation, fire area boundaries need not be completely sealed with floor-to-ceiling and/or wall-to-wall boundaries. Where fire area boundaries were not approved under the Appendix A process, or where such bound-aries are not wall-to-wall or floor-to-ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, licensees must perform an evaluation to ' assess the adequacy of fire area boundaries in their plants to determine if the boundaries will withstand the hazards associated with the area and protect important equipment within the area from a fire outside the area. This analysis must be performed by at least a fire protection engineer ! and, if required, a systems engineer. Although not required, l licensees may submit their evaluations for Staff review and ' concurrence. In any event, these analyses must be retained
\ by the licensees for subsequent NRC audits.
Enclosure to GL 85-01, Re: Fire Protection Policy
w William J. Dircks 0E Enclosure 7. Finally, the SC received significant input from Thomas Wambach, who acted as Secretary to the SC, and the Working Group, headed by Faust Rosa, NRR, and composed of V. Benaroya, C. Grimes, and V. Moore of NRR; S. Richardson, IE; W. Shields, OELD; and C. Anderson, T. Conlon, and W. Little of Regions I, II and III respectively. The recomendations of the SC are provided below. We believe that this responds to your direction, and when fully implemented, represent actions that will indeed expedite Appendix R compliance and assure consistent levels of fire protection safety at all plants. The SC is aware that not all parties will be fully satisfied with these actions. Nor have our recomendations been reviewed by the cognizant Offices or Regions. However, we believe that they represent sound judgments balanced with other competing safety priorities, and that with your approval the plan can and should be initiated promptly, i Recommendations To expedite Appendix R compliance and assure consistent levels of fire protection safety at all plants, the Fire Protection Steering Comittee recomends that the following actions be taken:
- 1. Promptly issue the enclosed Generic Letter (Enclosure 2) informing all licensees that:
(a) Extensions to the 50.48(c) schedules will no longer be granted; (b) An expedited fire protection inspection program will be instituted; (c) Documentation of valid analyses supporting fire protection features must be available for inspection; (d) Quality assurance applicable to fire protection systems is that required by GDC-1 of Appendix A to 10 CFR Part 50; and (e) The interpretations of Appendix R. (Enclosure 3) which should facilitate industry implementation of Appendix R and the responses to industry questions (Enclosure 6) represent the official agency position on all issues covered. (It should be noted that The Comission requested these documents for their review prior to issuance to industry.) Enclosure to GL 85-01, Re: Fire Protection Policy
11 3.1.5 Fire Zones QUESTION
, Appendix R, Section III.G.3 states " alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area room or zone under consideration...." What is the implied utilization of a room or zone concept under Section III.G of Appendix R? The use of the phraseology " area, room or zone under consideration" is used again at the end of the Section III.G.3. Does the requirement for detection and fixed suppression indicate that the requirement can be limited to a fire zone rather than throughout a fire area? Under what conditions and with what caveats can the fire zone concept be utilized in demonstrating conformance to Appendix R?
I
RESPONSE
Section III.G was written after NRC's multi-discipline review teams had visited all operating power plants. From these audits, the NRC recognized that it is not practical and may be impossible to subdivide some portions of an operating plant into fire areas. In addition, the NRC recognized that in some cases where fire areas are designated, it may not be possible to provide alternate shutdown capability independent of the fire area and, therefore, would have to be evaluated on the basis of fire zones within the fire area. The NRC also recognized that because some licensees had not yet performed a safe shutdown analysis, these analyses may identify new unique configurations. To cover the large variation of possible configurations, the requirements of Section III.G 'were presented in three parts: Section III.G.1 requires one train of hot shutdown systems be free of fire damage and damage to cold shutdown systems be limited. Section III.G.2 provides certain separation, suppression and detection requirements within fire areas; where such requirements are met, analysis is not necessary. Section III.G.3 requires alternative dedicated shutdown capability for configurations that do not safisfy the requirements of III.G.2 or where fire suppressants
; released as a result of fire fighting, rupture of the system or inadvertent operation of the system may damage redundant equipment.
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I Enclosure to GL 85-01, Re: Fire Protection Policy
m William J. Dircks OCT 2 61984 Discussion The recommended actions are grouped into three main areas dealing with (1) guidance to industry, (2) an expedited program of fire protection inspec-tions, and (3) a general upgrading of regulatory documents to reach and maintain consistent fire protection safety. This discussion section will focus broadly on what the SC found during its deliberations to warrant the focus of these recommendations and will indicate how this satisfies the agenda of issues cited in your memo of September 13. Details on these issues are provided in the record of the SC deliberations contained in Enclosure 7. With regard to guidance to industry, the SC concludes that adequate tech-nical guidance had been issued but that there were areas where confusion could arise. It was not clear where exemptions were needed, for example. However, a diligent reading of Appendix R and other staff documents did provide the basis for the satisfactory implementation of Appendix R at Calvert Cliffs. The SC concluded that it was neither needed nor appro-priate to develop new guidance, rather, bringing current technical and implementation guidance together in one Generic Letter and make the SRP, Tech Specs, and licenses consistent would suffice. The Generic Letter makes clear (1) that extensions to the 50.48(c) schedules will no longer be granted, (2) that an expedited inspection program will be instituted to see what fire protection fixes are in place and give licensees the inspection team judgements on the acceptability of future modifications, and (3) that the licensee judgements must be backed by documented and valid analyses. The SC believes that this will demonstrate to the licensee what action he must take and what our inspections will look for. The Generic Letter notes that, although the 50.48(c) schedules will not be extended, the relative safety priroities of fire protection modifications need to be considered in the development of "living schedules." One item of guidance in the Generic Letter that had not been uniformly disseminated is that the i QA applicable to fire protection features is that required by GDC-1. This would not attempt to backfit any QA requirements. Rather it would assure ( that future design, procurement, installation, testing and maintenance of fire protection features would receive high industrial quality attention. The SC believes that this initiative fully responds to the first three issues in your memo of September 13. Turning now to the inspection program, the SC found that the current inspections are generally satisfactory but that steps must be taken to indicate NRC's view of the importance of expediting implementation of Appendix R. These steps are to (1) speed up the inspection process, (2) develop a sound policy for fire protection enforcement actions, and (3) issue enforcement actions currently pending. These steps, in our view, would help expedite licensee compliance because it would raise industry's Enclosure to GL 85-01, Re: Fire Protection Policy
t 13 3.2 Fire Barrier Qualification 3.2.1 Acceptance Criteria QUESTION Recently the Staff has applied a 325*F cold side temperature criterion to its evaluation of the acceptability of one-hour and three-hour fire barrier cable tray wraps. This criterion is not in Branch Technical Position (BTP) APCSB 9.5-1, Appendix A as an accceptance criterion for fire barrier cable tray wraps and is not contained in Appendix R. It appears to represent post-Appendix R guidance. What is the origin of this criterion and why is it applicable to elec-trical cables where insulation degradation does not begin until jacket temperatures reach 450 F to 650*F?
RESPONSE
Fire barriers relied upon to protect shutdown related systems to meet the requirements of III.G.2 need to have a fire rating of either one or three hours. S50.48 references BTP APCSB 9.5-1, where the fire protection definitions are found. Fire rating is defined:
\ " Fire Rating - the endurance period of a fire barrier or structure; it defines the period of resistance to a standard fire exposure before the first critical point in behavior is observed (see NFPA 251)."
The acceptance criteria contained in Chapter 7 of NFPA 251,
" Standard Methods of Fire Tests of Building Construction and Materials," pertains to non-bearing fire barriers.
Thes~e criteria stipulate that transmission of heat through the barrier "shall not have been such as to raise the tem-perature on its unexposed surface more than 250*F above its initial temperature." The ambient air temperature at the beginning of a fire test usually is between 50*F and 90*F. It is generally recognized that 75*F represents an accept-1 able norm. The resulting 325*F cold side temperature criterion is used for cable tray wraps because they perform
! the fire barrier function to preserve the cables free of fire damage. It is clear that cable that begins to degrade at 450*F is free of fire damage at 325*F.
Enclosure to GL 85-01, Re: Fire Protection Policy
William J. Dircks OCT 2 61984 These steps are to develop and implement standard fire protection license conditions, Standard Review Plan, and Standard Tech Specs for all plants. The license condition developed for this is along the lines of the security plan and QA program (Enclosure 5). The Standard Review Plan should need minimum revision to assure that Appendix R is fully included. The Tech Specs, however, will require more research and development by the Working Group. We need to assure that the Tech Specs are soundly based to assure functioning of fire protection features but which require only those activities which are commensurate with other Tech Spec items in terms of importance to safety. As part of the above tasks, all fire protection guidance needs review and culling to assure that only a consistent and appropriate set remains. This complete set of guidance will be referenced in the Standard Review Plan revision. A last step in achieving uniform technical requirements is the SC recommendation to designate an office as responsible for awareness and resolution of interoffice / region fire protection issues. This is felt to be needed since current fire protection review is conducted within three divisions within NRR and one in IE. Although there is a lead branch responsibility, it is not always kept informed and involved. Therefore, the SC believes that the Director, Division of Engineering in NRR should be designated as the central point of contact. Conclusion The Fire Protection Policy Steering Canmittee concludes that the actions described above will accomplish the goals set forth in your memo of "expedidng Appendix R compliance for older plants and assuring consistent levels of fire protection safety at all plants." We believe that these actions will facilitate industry implementation 'of Appendix R through the use of the " interpretations" and a consistent set of guidance, yet will provide the necessary regulatory tools to guide, monitor, and, where appropriate, enforce this implementation process. We feel strongly that the actions we propose are synergistic and therefore all need to be completed to be most effective. The Fire Protection Policy Steering Committee has found the assignment to be challenging and rewarding. We would be pleased to brief you on our efforts at your earliest convenience. l l l Fire Protection Policy Steering Committee i ! Enclosure to GL 85-01, Re: Fire Protection Policy l l .__ - _ -_
o . J 15 1
- 4. The application or "end use" of the fire barrier is unchanged from the tested configuration. For example, the use of a cable tray barrier to protect a cable tray which differs in configuration from those that were tested would be acceptable. However, the use of structural steel fire proofing to protect a cable tray assembly would not be acceptable.
- 5. The configuration has been reviewed by a qualified fire protection engineer and found to provide an equivalent level of protection.
3.2.3 Fire Door Modifications , QUESTION Where labeled and rated fire doors have been modified to incorporate security hardware or for flooding protection, is an exemption from Appendix R required?
RESPONSE
Where a door is part of a fire area boundary, and the modr-fication does not effect the fire rating (for example, installation of security " contacts"), no further analysis need be performed. If the modifications could reduce the fire rating (for example, installation of a vision panel),
. the fire rating of the door should be reassessed to ensure that it continues to provide adequate margin considering the fire loading on both sides.
An exemption is required if fire doors installed in a fire barrier used to satisfy Section III.G.2 are modified such that the labeled rating no longer applies. 3.3 Structural Steel 3.3.1 NFPA Approaches QUESTION Ooes the NRC's definition of structural steel supporting fire barriers completely accomodate approaches described in NFPA guidance documents and standards? ( Enclosure to GL 85-0I, Re: Fire Protection Policy
'N William J. Dircks 007 2 6 564 J
Enclosures:
- 1. Memo to H. Denton et al on Review of NRC Fire Protection Policy and Programs.
- 2. Generic Letter on Fire Protection
- 3. Interpretations of Appendix R
- 4. Guidance for Enforcement Actions Concerning Fire Protection Requirements ,
- 5. Fire Protection License Condition
- 6. Appendix R Questions and Answers
- 7. Steering Committee Memoranda
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Enclosure to GL 85-01, Re: Fire Protection Policy
17
RESPONSE
In general, cable tray supports should be protected, regard-less of whether there is a sprinkler system. If (1) the qualification tests were performed on wrapped cable trays with unprotected supports, and the supports are shown to be adequate, or (2) a structural analysis is performed which demonstrates failure of the unprotected support (s) will not cause a loss of the cable tray fire barrier, then they need not be protected. An exemption is not required; however, the qualification tests and applicability or the structural evaluation should be documented and available for audit. 3.4 Automatic Suppression System 3.4.1 Water Oensity QUESTION Staff guidance provided in Generic Letter 83-33 concerning automatic suppression coverage of fire areas interprets the phrase "in the fire area" in Section III.G as meaning "throughout the fire area." What delivered water density or occupancy standard as specified in NFPA-STD-13 must be achieved to meet this guidance?
RESPONSE
Individual plant areas are diverse in nature. The designer should determine the particular water density or occupancy classification. Those areas which contain a limited quan-tity of in-situ and anticipated transient combustibles and which feature contents such as tanks and piping, may be considered as " Ordinary Hazard (Group 1)", as defined by NFPA Standard No. 13. For those areas containing large amounts of cables or flammable liquids, an occupancy class-ification of " Extra Hazard" may be warranted. The decision as to which classification should be applied should be made by a qualified fire protection engineer. Once the occupancy classification is determined, the minimum water density should be based on the Density Curves in table 2.2.l(B) of NFPA 13. Any density equal to or in excess of the curves would be in conformance with our guidelines as delineated in Section C.6.c of BTP CMEB 9.5-1.
-o Enclosure to GL 85-01, Re: Fire Protection Policy
2 7 < Olshinski, R-II; Spessard, R-III; and Vollmer, NRR, Chairman. This Steering Committee is to decide the scope of issues to be considered, meet with HQ and regional personnel as necessary to consider these issues, make assignments as appropriate to a working group headed by Faust Rosa, NRR, for detailed consideration of certain issues, and make ' recommendations for actions along with supporting bases to me by October 26, 1984 I concur in these assignments and the general charter of the Steering Cannittee. All ongoing regulatory actions in your Offices regarding fire protection should be continued and should not be delayed or deferred pending the outcome of this review. pps$ %3!iari).Dh)J8 i William J. Dircks Executive Director for Operations ) cc: V; Stello G. Cunningham Steering Committee .. i 1 I 1 Enclosure to GL 85-01, Re: Fire Protection Policy
F , a . 19 3.4.4 Fixed Suppression System In Fire Area QUESTION Are fixed suppression systems required by Section III G.3 to be throughout the fire area, room or zone under consideration?
RESPONSE
No, but partial coverage must be properly justified and documented. See Item #5 of the " Interpretations of Appendix R."
... suppression less than full area coverage may be adequate to comply with the regulation. Where full area suppression and detection is not installed, licensees must perform an evaluation to assess the adequacy and necessity of partial l suppression and detection in an area. The evaluation must be performed by a fire protection engineer and, if required, a systems engineer. Although not required, licensees may submit their evaluations to the staff for review and concurrence. In any event, the evaluations must be retained for subsequent NRC audits..."
3.4.5 Sprinkler Head Location QUESTION I l If stacks of horizontal or vertical cable trays extend from l ceiling to floor, are sprinkler heads required (1) under i the lowest horizontal trays, near the floor for vertical trays; (2) at some intermediate level between the floor and ceiling, and (3) at the ceiling?
RESPONSE
I l Sprinkler heads should be located at the ceiling. Sprinkler heads at other locations may be necessary depending upon the hazard and the cumulative effect of the obstructions to the j discharge of water from the sprinkler head. The sprinkler system design should meet NFPA 13. Enclosure to GL 85-01, Re: Fire Protection Policy
Appendix R and are now finished or nearly finished with that effort. Others have engaged in lengthy negotiations with the staff while continuing to file requests for schedule extensions, and thereby have barely begun Appendix R modifications needed to comply with Sections III.G and III.L. Schedule extension requests have been received seeking implementation dates of 1990 or beyond. As the 50.48(c) . schedule was intended to be a one-time schedule commencing in the 1980-1982 time frame and ending in the 1985 time frame, extensions well beyond this schedule (particularly where major modifications remain to be completed) undermine the purpose of the schedule, which was to achieve expeditious compliance with NRC fire protection requirements. The NRC will therefore grant no further extensions to the 50.48(c) schedules. When a licensee's schedule expires, compliance is expected and appropriate enforcement action will be taken. If compliance cannot be achieved by that date, the licensee will be required to submit and justify a minimum schedule for corrpletion of fire protection modifications, and to supply interim measures to compensate for the lack of compliance. In submitting a schedule which goes beyond the current 50.48 deadline *, the licensee will be required to demonstrate that it has endeavored in good faith to complete modifications on schedule. A showing of good faith attempt to complete implementation on schedule may mitigate 1 enforcement action for noncompliance with NRC requirements. j The NRC is currently reviewing all dockets of plants covered by the 50.48 schedule to determine schedule deadlines. When this review is completed, each licensee will be informed of the deadlines. B. Revised Inspection Program , In 1982, the NRC developed an inspection program to verify compliance with the requirements of 10 CFR 50, Appendix R. This program was primarily oriented towards reviewing safe shutdown features of those pre-1979 licensees that had completed Appendix R modifications and selected NT0L plants. From 1982 to the present, approximately seven , Appendix R compliance inspections have been performed. In a number of l cases, these inspections have discovered that licensees had made signi-ficant errors in implementing a number of Appendix R requirements. In order to expedite compliance verification and to provide the NRC staff with earlier indication of problems associated with implementation of fire protection features, the NRC will conduct fire protection inspections of operating plants and plants currently under-going operating license review during 1985 to include at least one Licensees submitting "living schedules" for NRC approval should be aware that existing 50.48 schedules continue to apply. Licensees intending to include fire protection modifications within a "living schedule" are expected to assign within such schedules the relative safety priorities of remaining fire protection modifications. Enclosure tn GL 85-01, Re: Fire Protection Policy
21 3.5.2 Floor-to-Floor Separation QUESTION Where redundant circuits are separted by floor elevation but are within the same fire area due to open hatchways, stairs, etc., what is the NRC',s position with regard to separation criteria? If train A is located twenty feet from an open hatchway on the lower elevation and train B is located ten feet from the same opening on the next elevation, would this be considered adequate separation?
RESPONSE
If a wall or floor / ceiling assembly contains major unprotected openings such as hatchways and stairways, then plant loca-tions on either side of such a barrier must be considered as part of a single fire area. Refer to the staff position on Fire Areas in Generic Letter 83-33. As to the example provided, if train A was separated by a cumulative horizontal distance of 20 feet from train 8, with no intervening combustible materials or fire hazards, and both elevations were provided with fire detection and suppression, the area would be in compliance with Section III.G.2.b. 3.6 Intervening Combustibles 3.6.1 Negligible Quantities of Intervening Combustibles QUESTION Twenty feet of separation with absolutely no intervening combustibles is a rare case in most nuclear plants. What is the most acceptable method of addressing intervening combustibles? How are various utilities addressing this subject, and what would be sufficient justification to support an exemption request?
RESPONSE
If more than negligible quantities of combustible materials (such as isolated cable runs) exist between redundant shut-down divisions, an exemption request should be filed. [" Negligible quantity" is an admittedly judgmental criterion, and this judgment should be made by a qualified fire protec-tion engineer and documented for later NRC audit.] Justif-ications for such exemptions have been based on the following factors: Enclosure to GL 85-01, Re: Fire Protection Policy
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Enclosure to GL 85-01. Re. Fire Protection policy
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- b. Material having a structural base of noncombustible material, as defined in a., above, with a surfacing not over 1/8-inch thick that has a flame spread rating not higher than 50 when measured using ASTM E-84 Test
" Surface Burning Characteristics of Building Materials."
In Generic Letter 83-33, we state:
" Staff Position: Section III.G.2.b requires the " separation
... with no intervening combustibles ..." To meet this requirement, plastic jackets and insulation of grouped electrical cables, including those which are coated, should be considered as intervening combustibles."
For fire protection, "no intervening combustibles" means that there is no significant quantities of in-situ materials which will ignite and burn located between redundant shutdown systems. The amount of such combustibles that has signifi-cance is a judgmental decision. As with other issues, if the licensees fire protection engineer is concerned that the quantity of combustibles between shutdown divisions may not be considered insignificant by an independent reviewer, an exemption could be requested, or the staff consulted. Transient materials are not considered as an intervening combustible; however, they must be considered as part of the overall fire hazard within an area." Cables that are in covered cable tray should also be considered as intervening combustibles. Coated cables with a fire retardant material are also considered as intervening combustibles. 3.6.3 Unexposed Combustibles QUESTION Are unexposed combustibles, such as oil in sumps, closed cans, or sealed drums, or electrical cable in conduits, considered as " intervening combustibles"?
RESPONSE
Only oil in closed containers which are in accordance with NFPA 30 or electrical cables in conduits are not considered as intervening combustibles. In situ oil in sumps is con-sidered to be an intervening combustible. Enclosure to GL 85-01, Re: Fire Protection Policy
- 3. Fire Damage Appendix R to 10 CFR Part 50 utilizes the term " free of fire damage."
In promulgating Appendix R, the Connission has provided methods acceptable for assuring that necessary structures, systems and components are free of fire damage (see Section !!!.G.2a, b and c), that is, the structure, system or component under consideration is capable of performing its intended function during and after the postulated fire, as needed. Licensees seeking exemptions from Section III.G.2 must show that the alternative proposed provides reasonable assurance that this criterion is met. (Notealso that Section III.G.2 applies only to equipment needed for hot shutdown. Therefore, an exemption from III.G.2 for cold shutdown equipment is not needed.)
- 4. Fire Area Boundaries The term " fire area" as used in Appendix R means an area sufficiently bounded to withstand the hazards associated with the area and, as necessary, to protect important equipment within the area from a fire outside the area. In order to meet the regulation, fire area boundaries need not be completely sealed floor-to-ceiling, wall-to-wall boundaries. However, all unsealed openings should be identified and evaluated. Where fire area boundaries were not approved under the BTP process, or where such boundaries ;
are not wall-to-wall, floor-to-ceiling boundaries with all penetrations ' sealed to the fire rating required of the boundaries, licensees must perform an evaluation to assess the adequacy of fire boundaries in their plants to determine if the boundaries will withstand the hazards associated with the area and protect important equipment within the area from a fire outside the area. This analysis must be performed by at least a fire protection engineer and, if required, a systems engineer. Although not required, licensees may submit their evaluations for staff review and concurrence. In any event, these analyses must be retained by the licensees for sub-sequent NRC audits.
- 5. Automatic Detection and Suppression Sections III.G.2.b and III.G.2.c of Appendix R state that "In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area..." Other provisions of Appendix R also use the phrase
" fire detectors and an automatic fire suppression system in the fire area..."
(see e.g., Section !!I.G.2.e). In order to comply with these provisions, suppression and detection suf-ficient to protect against the hazards of the area must be installed. In this regard, detection and suppression providing less than full area coverage may be adequate to comply with the regulation. Where full area suppression and detection is not installed, licensees must perform an evaluation to essess the adequacy of partial suppression and detection to protect against the hazards in the area. The evaluation must be performed Enclosure to GL 85-01, Re: Fire Protection Policy
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3.8 Desian Bases 3.8.1 Fire Protection Features NFPA Conformance QUESTION Should the fire protection features required by Section III.G conform to the NFPA Codes?
RESPONSE
Yes. For example, Section III G.2 requires an automatic suppression system. Our guidelines would recommend that the system be in accordance with an NFPA Code. If deviations are made from the Code, they should be identified in the FSAR or FHA. 3.8.2 Design Basis Fire QUESTION Why isn't the industry allowed to design to protect against a design basis fire?
RESPONSE
Neither the industry nor the Staff has been able to develop criteria for establishing design basis fire conditions because the in-situ and potential transient combustibles vary widely in different areas of the plant. 3.8.3 Redundant Trains / Alternate Shutdown QUESTION Confusion exists as to what will be classified as an alternate shutdown system and thus what systems might be required to be protected by suppression and detection under Section III.G.3.b. For example, while we are relying upon the turbine-building condensate system for a reactor building fire and the RHR system for a turbine building fire, would one system be considered the alternative to the other. If so, would suppression and detection be required for either or both systems under III.G.3.b? An explanation of alternative shutdown needs to be advanced for all licensees. 9 Enclnsure tn GL 85-01, Re: Fire Protection Policy
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Enclosure to CL 85-01, Re: Fire Protection Policy
. 27
- 1. The reactor is tripped in the control room.
- 2. Offsite power is lost as well as automatic starting of the onsite a.c. generators and the automatic function of valves and pumps whose control circuits could be affected by a control room fire.
The analysis should demonstrate that capability exists to manually achieve safe shutdown conditions from outside the control room by restoring a.c. power to designated pumps, assuring that valve lineups are correct, and assuming that any malfunctions of valves that permit the loss of reactor coolant can be corrected before unrestoreable conditions occur. Note that the only manual action usually credited in the control room by this analysis is the reactor trip. Any additional control room actions deemed necessary would have to be justified under the exemption process. After the fire, the operators could return to the control room when the following conditic7s have been met:
- 1. The fire has been extinguished and so verified by appropriate fire protectic.n per:annel.
- 2. The control room has been deemed habitable by appropriate fire protection personnel and the shift supervisor.
- 3. Damage has been assessed and, if necessary, corrective action has been taken to assure necessary safety, control and information systems are functional (some operators may assist with these tasks) and the shift supervisor has authorized return of plant control to the control room.
- 4. Turnover procedures which assure an orderly transfer of control from the alternate shutdown panel to the control room has been completed.
After returning to the control room, the operators can take any actions compatible with the condition of the control room. Controls in any area (cabinet) where the fire occurred would not be available. Smoke and fire suppressant damage in other areas (cabinets) must also be assessed and correc-tive action taken before controls in such cabinets are deemed functional. Controls in undamaged area (cabinets) could be operated as required. Minor modifications inside the control room may be performed to reach cold shutdown. i Enclosure to GL 85-01, Re: Fire Protection Policy
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- 2. Severity Cateaories A. Severity I. Violations of fire protection requirements estab-lished to protect or enable operation of safe hot shutdown equip-ment concurrent with an actual fire which damages that equipment such that safe hot shutdown could not be achieved or maintained using the equipment dedicated for the purpose.
B. Severity II. Violations of fire protection requirements established to protect or enable operation of safe cold shutdown equipment concurrent with an actual fire which damages that equipment such that safe cold shutdown could not have been achieved and maintained using the equipment dedicated for this purpose in accordance with the applicable requirements. C. Severity III. Violations of fire protection requirements established to protect or enable operation of safe shutdown equipment such that a fire in the area could damage that equipment to the extent that safe hot or cold shutdown could not have been achieved and maintained using the equipment dedicated for this purpose in accordance with applicable requirements. Failure to have a written evaluation available for an area where compliance with Appendix R is not readily demonstrated will be taken as prima facie evidence that the area does not comply with NRC requirements and may result in enforcement action at the severity level. D. Severity IV. Violations of one or more fire protection requirements that do not result in a Severity Level I, II or III violation and which have more than minor safety or environmental significance. E. . Severity V. Violations of one or more fire protection requirements that have minor safety or environmental significance. Enclosure to GL 85-01, Re: Fire Protection Policy
_ 29
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- 5. ALTERNATIVE AND DEDICATED SHUTOOWN CAPABILITY 5.1 Safe and Alternative Shutdown 5.1.1 Previously Accepted Alternative Shutdown Capability QUESTION As part of the Appendix A review process, some plants had committed to an alternative shutdown system in the form of a remote shutdown panel or remote shutdown system. Footnote 2 to Appendix R describes alternative shutdown capability as being associated with " Rerouting, relocating, or modifying of existing systems." To the extent that an existing remote shutdown system previously reviewed and approved under Appendix A to BTP 9.5-1 does not require modifications, rerouting, or relocating of existing, systems, are the requirements of Section III.L of Appendix R backfit?
RESPONSE
Yes. Existing remote shutdown capabilities pre.viously re-viewed and approved under Appendix A to BTP APCSB 9.5-1 do not categorically comply with Section III.G.3 of Appendix R. Licensees were requested to re-analyze their plants to deter-mine compliance with Section III.G. If the licensee chooses to use the option of III.G.3 for provision of safe shutdown capability for certain areas, the criteria of Section III.L'. are applicable to that capability for that area. See also the response to 5.1.3. 5.1. 2 Pre-Existing Alternative Shutdown Capability QUESTION Some licensees defined safe shutdown capability for purposes of analysis to Section III.G criteria as being composed of both the normal safe shutdown capability and the pre-existing redundant or remote safe shutdown capability which was previously installed as part of the Appendix A process. This definition often took the form of two " safe shutdown trains" comprising (1) one of the two normal safe shutdown trains, and (2) a second safe shutdown train capability which was being provided by the pre-existing remote shutdown capability. This definitional process, which was undertaken by a number of licensees, makes a significant difference in the implementation of Appendix R. Under such a definition, does Section III.L criteria apply when the Commission did ? not call out Section III.L as a backfit? Enclosure to GL 85-01, Re: Fire Protection Policy
4 Enclosure to GL 85-01, Re: Fire Protection Policy
o . 31 7
RESPONSE
Safe shutdown capabilities including alternative shutdown capabilities are all designed for some maximum level of fire-damage (system unavailabilities, spurious actuations). Since the extent of the fire can not be predicted, it seems prudent to have the post-fire shutdown procedures guide the operator from full system availability to the minimum shutdown capability. As for repair procedure, similar conditions exist. A repair procedure can be written based on the maxi-mum level of damage that is expected. This procedure would then provide shutdown capability without accurately predic-ting likely fire damage. 5.2.2 Post Fire Operating Procedures QUESTION Does the NRC have any requirements regarding whether post-fire operating procedures should be based upon fire areas, systems, or by symptom-based?
RESPONSE
The NRC does not have requirements, nor do we propose any requirements regarding whether post-fire operating procedures should be based upon fire areas, systems or be symptom-based. We suggest that the post-fire shutdown capabilities designs be reviewed with the plant operation staff and procedures written with their input'. See also responses to 5.2.1 and 5.2.3. . 5.2.3 Alternative Shutdown Capability QUESTION Is it acceptable to develop post-fire operating procedures only for those areas where alternative shutdown is required? (For other areas standard, emergency operating procedures would be utilized in the presence of potential fire damage to a single train.) l l l l l - l ! Enclosure to GL 85-01, Re: Fire Protection Policy l l
. o Appendix R 2 3.6 Intervening Combustibles 3.6.1 Negligible Quantities of Intervening Combustibles 3.6.2 In Situ Exposed Combustibles 3.6.3 Unexposed Combustibles 3.7 Radiant Energy Shields 3.7.1 Fire Rating 3.8 Design Bases 3.8.1 Fire Protection Features NFPA Conformance 3.8.2 Design Basis Fire 3.8.3 Redundant Trains / Alternate Shutdown 3.8.4 Control Room Fire Considerations
- 4. III J, EMERGENCY LIGHTING 4.1 Illumination Levels
- 5. III L, ALTERNATIVE AND DEDICATED SHUTDOWN CAPABILITY 5.1 Safe and Alternative Shutdown 5.1.] Previously Accepted Alternative Shutdown Capability
- 5.1.2 Pre-Existing Alternative Shutdown Capability 5.1.3 III L Backfit 5.2 Procedures 5.2.1 Shutdown and Repair Basis 5.2.2 Post Fire Operating Procedures 5.2.3 Alternative Shutdown Capability 5.2.4 Post Fire Procedures Guidance Documents
- 5.3 Safe Shutdown and Fire Damage 5.3.1 Circuit Failure Modes l 5.3.2 " Hot Short" Duration j 5.3.3 Hot Shutdown Duration 5.3.4 Cooldown Equipment 5.3.5 Pressurizer Heaters 5.3.6 On-Site Power 5.3.7 Torus Level Indication 5.3.8 Short Circuit Coordination Studies 5.3.9 Diagnostic Instrumentation 5.3.10 Design Basis Plant Transients 5.3.11 Alternate / Dedicated Shutdown vs. Remote Shutdown Systems ll i
Enclosure to GL 85-01, Re: Fire Protection Policy
33 5.3.2 " Hot Short" Duration QUESTION If one mode of fire damage involves a " hot short" how long does that condition exist as a result of fire damage prior to terminating in a ground or open circuit and stopping the spurious actuation?
RESPONSE
We would postulate that a " hot short" condition exists until action has been taken to isolate the given circuit from the fire area, or other actions as appropriate have been taken to negate the effects of the spurious actuation. We do not postulate that the fire would eventually clear the " hot short." 5.3.3 Hot Shutdown Duration QUESTION Since hot shutdown cannot be maintained indefinitely, hot shutdown equipment needs to be protected for only a limited period of time. How long must a plant remain in that condi-tion in order to meet the requirement for achieving hot shutdown with a single train of equipment?
RESPONSE
Section III.G.1 requires that the one train of systems needed to achieve and maintain hot shutdown be free of fire damage. Thus, the systems needed are to be completely protected from the fire regardless of time. If the intent of the question concerns how long these systems must operate, these systems must be capable of operating until the systems needed to achieve and maintain cold shutdown are available. 5.3.4 Cooldown Equipment QUESTION Certain equipment is necessary only in the cooldown phase when the plant is neither in hot nor cold shutdown condition as defined by technical specifications. Is this equipment considered hot or cold shutdown in nature? I
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Enclosure to GL 85-01, Re: Fire Protection Policy
I 1 .1 1 l r Enclosure to GL 85-01, Re: Fire Protection Policy
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RESPONSE
These statements are meant to indicate that the alternativo shutdown capability should be powered from an onsite power system independent (both electrically and physically) from the area under consideration. Further, if the normal e=argency onsite power supplies (diesel generators) are not available because of fire damage, then a separate and independent onsite power system shall be provided. As an excr.ple, some plants are utilizing a dedicated onsite diesel generator or gas turbine to power instrumentation and control panels v!hich are a part of the alternative shutdown capability. 5.3.7 Torus Level Indication OUESTION For BWRs, I&E Information Notice 84-09 suggests that licen-sees need to have torus level indication post-fire. If an analysis shows that a level does not change significantly during any operational modes or worse case conditions, is level indication still required? Is an analysis in file adequate or is an exemption request required?
RESPONSE
It continues to be our position that torus (suppression pool) level indication is the preferred post-fire monitoring instrumentation in order to confirm the availability of the torus (suppression pool) as a heat sink. We recognize that existing analyses indicate that suppression pool level is not significantly changed during emergency shutdown conditions. However, we believe the operator should be able to confirm that spurious operations or other unantici-pated occurrences have not affected the torus function. An analysis of torus level change by itself is not considered an acceptable basis. 5.3.8 Short Circuit Coordination Studies QUESTION Should circuit coordination studios consider high impendance faults? Enclosure to GL 85-01, Re: Fire Protection Policy
- O Appendix R 4 9.4 Future TI 2515/62 Revisions 9.5 Documentation Supplied by Licensee 9.6 Subsequent Inspections 9.7 NRC List of Conforming Items 9.8 Inspection Re-review 9.9 List of Shutdown Equipment l
i i i i i 4 i Enclosure to GL 85-01, Re: Fire Protection Policy I
I
,. 37
- b. The safe shutdown capability should not be adversely affected by a fire in any plant area which results in the loss of all automatic function (signals, logic) from the circuits located in the area in conjunction with one worst case spurious actuation or signal resulting from the fire; and
- c. The safe shutdown capability should not be adversely affected by a fire in any plant area which results in
-simultaneous spurious actuation of all valves in high-low pressure interface lines.
5.3.11 Alternative / Dedicated Shutdown vs. Remote Shutdown Systems QUESTION What is the difference between the alternate / dedicated shutdown systems required for fire protection and the remote shutdown systems recommended under Chapter 7 of the SRP?
RESPONSE
The remote shutdown systems recommended under Chapter 7 of the SRP are needed to meet GDC 19. These remote shutdown systems need to be redundant and physically independent of the control room in order to meet GDC 19. For GDC 19, damage to the control room is not considered. Alternate shutdown systems for Appendix R need not be redundant but must be both physically and electrically independent of the control room. l l i ( .. l l Enclosure to GL 85-01, Pe: Fire Protection Policy
2 In May 1976, the NRC asked licensees to compare operating reactors with BTP APCS 8 9.5-1, and in September 1976, those licensees were informed that the guidelines in Appendix A would be used to analyze the consequences of a fire in each plant area. In September 1976 the licensees, were also requested to provide a fire hazards analysis that divided the plant into distinct fire areas and show that redundant systems required to achieve and maintain cold shutdown are adequately protected against damage by a fire. Early in 1977 each licensee responded with a Fire Protection Program Evaluation which included a Fire Hazard Analysis. These evaluations and analyses identified aspects of licensees' fire protection programs that did not conform to the NRC guidelines. Thereafter, the staff initiated discussions with all licensees aimed at achieving implementation of fire protection guidelines by October 1980. The staff held many meetings with licensees, conducted extensive correspondence with them, and visited every operating reactor. As a result, many fire protection items were resolved, and agreements were included in Fire Protection Safety Evaluation Reports issued by the NRC. Several fire protection issues remained unresolved with a number of licensees. By early 1980, most operating plants had implemented most of the guidelines in Appendix A. However, as the Commission noted in its Order of May 23, 1980, the fire protection program has had some significant problems with implementation. Despite the staff's efforts, several licensees had expressed continuing disagreement with, and refused to adopt recommendations relating to several generic issues, including the requirements for fire brigade size and training, water supplies for fire suppression systems, alternate and dedicated shutdown capability, emergency lighting, qualifications of seals used to enclose places where cables penetrated fire barriers, and the prevention of reactor coolant pump lubrication system fires. To estab-lish a definitive resolution of these contested subjects in a manner con-sistent with the general guidelines in Appendix A to the BTP and to assure timely compliance by licensees, the Commission issued a proposed fire protection rule and its Appendix R, which was described as setting out minimum fire protection requirements for the unresolved issues (45 Fed. Reg. 36082 May 29, 1980). The fire protection features addressed included protection of safe shutdown capability, emergency lighting, fire barriers, associated circuits, reactor coolant pump lubrication system, and alternate shutdown systems. The Commission stated that it expected all modifications (except for alternate and dedicated shutdown capability) to be implemented by November 1, 1980. 4/ 3/ 11 NRC 707, 718 (1980) 4/ Id. at 719 Enclosure to GL 85-01, Re: Fire Protection Policy
39 points would be safely collected and drained to the sump. The sump should be shown capable of safely containing all of the anticipated oil leakage. The analysis should verify that there are no electric sources of ignition. l (-~ Enclosure to GL 85-01. Re: Fire Protection Policy
4
- 2. OVERVIEW Section 50.48 Fire Protecticn of 10 CFR Part 50 requires that each operating nuclear power plant have a fire protection plan that satisfies General Design Criterion 3 of Appendix A to 10 CFR 50. It specifies what should be contained in such a plan and lists the basic fire protection guidelines for this plan. It requires that the Fire Protection Safety Evaluation Report which has been issued for each operating plant state how these guidelines were applied to each facility.
Section 50.48 also requires that all plants with operating licenses prior to January 1, 1979 satisfy the requirements of Section III.G, III.J and III.0, and other Sections of Appendix R where approval of similar features had not been obtained prior to the effective date of Appendix R. By a separate action, the Commission approved the staff's requirement that all plants to receive their operating license af ter January 1,1979 also sat-isfy the requirements of Sections III.G, III.J and III.0 and that a fire protection license condition be established. Deviations from Appendix R requirements for pre-1979 plants are processed under the exemption process. Deviation from other guidelines are identified and evaluated in the Safety Evaluation Report. A standard fire protection license condition has been developed and will be included in each new operating license. Present operating licenses will be amended to include the standard license condition. Tne Regions initiated inspections of operating plants and identified several significant items of non-compliance. The Nuclear Utility Fire Protection Group requested interpretations of certain Appendix R requirements and provided a list of questions that they thought should be discussed with the industry. The NRC held workshops in each Region to assist the industry in understanding the NRC's requirements and to improve the Staff's under-standing of the industry's concerns. This document presents the NRC's response to the questions posed by the industry and supplemented with additional questions identified at the workshops as being of interest to the industry or the staff. These re-sponses may be used as guidance for design, review and inspection activi-ties. The questons have been reformatted according to their applicability to Sections of Appendix R, BTP CMEB 9.5-1, licensing policy or inspection policy. Enclosure to GL 85-01, Re: Fire Protection Policy _= _ _
41 f, Our guidelines on the seismic design of hydrogen lines is delineated in BTP CMEB 9.5-1 C.S.d(5): (5) Hydrogen lines in safety-related areas should be either designed to seismic Class I requirements, or sleeved such that the outer pipe is directly vented to the outside, or should be equipped with excess flow valves so that in case of a line break, the hydrogen concentration in the affected areas will not exceed 2%. All PWR's have a hydrogen line going to the Volume Control Tank (Make-up Tank) that needs to be protected. To identify plant specific situations in which seismic events could initiate a fire in a specific plant area, the fire protection engineer and systems engineer performing the fire hazards analysis should be concerned with in-situ combustible materials which can be released in a manner such that they could contact in-situ ignition sources by a seismic event. An example of this would be the rupture of the RCP lube oil line directly above the hot reactor coolant piping. The fire protection engineer should also be concerned with seismic induced ignition sources, electrical or mechanical, which could contact nearby in-situ combustible materials. 7.2 Random Fire and Seismic Events QUESTION Is a random fire to be postulated concurrent with a seismic event?
RESPONSE
Our position, as stated in Section C.l.6 of BTP CMEB 9.5-1, is " Worst case fire need not be postulated to be simultaneous with nonfire-related failures in safety systems, plant accidents, or the most severe natural phenomena." Where plant systems are designed to prevent the release of combustible materials caused by a seismic event, such as a dike around a fuel oil tank transformer, or seismic supports for hydrogen lines, then no
- fire need to be arbitrarily assumed to take place in the fire hazards analysis.
Because it is impossible to completely preclude the occurrence of a seismically induced fire, Section C.6.c(4) of CMEB 9.5-1 states:
" Provisions should be made to supply water at least to standpipes and hose connections for manual firefighting in areas containing equipment required for safe plant shutdown in the event of a safe shutdown earthquake. The piping system serving such hose stations should be analyzed for SSE loading and.should be provided r with supports to ensure system pressure integrity. The piping
(.. Enclosure to GL 85-01, Re: Fire Protection Policy
r-6 to the area or systems being considered. For this condition, it is assumed that no manual or automatic fire suppression action has been started and the fire has reached its peak burning rate and involves all combustibles present. Within each area special attention should be given to limiting the amount of combustible material and to providing effective barriers and fire resistive coatings to reduce the spreading of a fire in these areas. A design basis fire should be assumed and provisions should be made to limit the consequence of such a fire by providing fire barriers with suitable separation between redundant systems and components which are provided to carry out required safety functions. This separation is enhanced if the plant is divided into suitable fire areas since redundant safety equipment can then be placed in separate fire areas. Particular design attention should be given to the use of separate isolated fire areas for redundant cables to avoid loss of redundant safety-related cables. Provisions should also be made to limit the consequences of a fire by suitable design of the ventilation systems so that the spread of the products of combustion to other areas of the plant is prevented. Means should be provided to ventilate or isolate the area as required. The power supply and controls for the area ventilation system should be from outside the area, and the power and control cables should not pass through the area. l The fire detection systems should be designed using detectors of the right types at locations suitable to detect the particular type of fire expected in each area. In the design, consideration should be given to provide personnel access to and escape routes from each fire area. The emergency plans for all plants should lay out access
- and escape routes to cover the event of a fire in critical
! areas of the plant." i In addition, definitions of the basic fire area components were given:
" Fire Area - that portion of a building or plant that is separated from other areas by boundary fire barriers (walls, floors or roofs) with any openings or penetrations protected with seals or closures having a fire resistance rating equal to that of the barrier.
I Fire Barrier - those components of construction (walls, floors and roofs) that are rated by approving laboratories in hours for resistance to fire to prevent the spread of fire. Enclosure to GL 85-01, Re: Fire Protection Policy i
r-43 i
- 8. LICENSING POLICY 8.1 Fire Hazard Analysis / Fire Protection Plan Updating QUESTION What constitutes the fire protection plan required by 50.48(a)?
Should licensees have programs to maintain the fire hazards analysis and the fire protection plan current or updated periodically? How often should the plan be updated? Must revisions be provided to the NRC?
RESPONSE
The basic elements required in the fire protection plan are described in 10 CFR 50.48(a). The fire protection program that implements that plan should include the details of the fire hazards analysis. The plan and program may be separate or combined documents and must be kept current with the fire hazards analysis updated prior to making modifications. We would expect that for most plants licensed after January 1, 1979, the fire protection plan and program would be part of the FSAR and therefore, would be updated and submitted to the NRC in conformance with the requirements of 10 CFR 50.71(e). For plants whose fire protection plans and programs are not part of the FSAR, we would expect that they would be updated prior to making modifications and kept at the site in an auditable form for NRC inspection. 8.2 Fire Protection License Condition QUESTION What is the significance of the fire protection license condition?
RESPONSE
For those plants licensed prior to January 1,1979 (Appendix R plants), the license condition is the legally enforceable requirement for the fire protection features other than those required by III.G, III.J. and III.0 that were accepted by the NRC staff as satisfying the provisions of Appendix A to Branch Technical Position BTP APCSB 9.5-1. For those plants licensed af ter January 1,1979, the license condition is the legally enforceable requirement for all fire protection features i at the facility. Appendix R is only enforceable on Post 1979 plants through the license condition. 10 CFR 50.48 makes Appendix R applicable only to plants licensed prior to January 1,1979. Refer to 10 CFR 50.48(e), t
!( ; ss_
l Enclosure to GL 85-01, Re: Fire Protection Policy
. o 8
e) Describe all the extinguishing and detection capabilities within each fire area. Discuss all means for containing and inhibiting the progress of a fire, e.g., the use of fire stops, coatings, curbs, walls, etc. Describe the extinguishing equipment outside an area which has access to the area. NOTE: If large fire areas are divided into fire zones for the purpose of fire protection, the above information should be provided for each zone.
- 2. Where redundant safety related equipment or cabling is located in a given fire area, describe the design features which prevent the loss of both redundant trains in a common fire, e.g., the separation provided by distance, physical barriers, and electrical isolation.
Where control, power or instrument cables of redundant systems used for bringing the reactor to safe, cold shutdown are located in the same cable trays, either provide a bounding analysis demonstrating that the worst consequences as a result of a fire in the cable trays are acceptable or show that redundant systems required to achieve and maintain a cold shutdown are adequately protected against damage by the fire." The guidelines for the fire rating of fire area bound-aries and their penetrations were given in Appendix A to BTP 9.5-1.
" APPLICATION 00CKETED BUT CONSTRUCTION PLANTS UNDER CONSTRUCTION PERMIT NOT RECEIVED AS OF 7/1/76 AND OPERATING PLANTS (j) Flonrs, walls and ceilings enclosing (j) SAME. The fire hazard separate fire areas should have minimum in each area should be fire rating of three hours. Penetrations evaluated to determine in these fire barriers, including conduits barrier requirements.
and piping, should be sealed or closed to If barrier fire resis-provide a fire resistance rating at least tance cannot be made equal to that of the fire barrier itself. adequate, fire detection Door openings should be protected with and suppression should be equivalent rated doors, frames and hardware provided, such as: that have been tested and approved by a nationally recognized laboratory. Such (1) water curtain in doors should be normally closed and case of fire, locked or alarmed with alarm and annuncia- (ii) flame retardant tion in the control room. Penetrations for coatings, ventilation system should be protected by a (iii) additional fire standard " fire door damper" where required, barrierf." (Refer to NFPA 80, " Fire Doors and Windows") Enclosure to GL 85-01, Re: Fire Protection Policy
-.___4 -
n -,_--4 ---. + - - -. s -. 45 l 8.4 Future Changes QUESTION Will future changes (no matter how minor) to approved configurations be required to be reviewed by the Staff in an exemption request? At what point may the process of 10 CFR 50.59 be invoked?
RESPONSE
If a future modification involves a change to a license condition or technical specification, a license amendment request must be submitted. When a modification not involving a technical specifica-tion or license condition is planned, the evaluation cade in conform-ance with 10 CFR 50.59 to determine whether an unreviewed safety question is involved must include an assessment of the modification's f impact on the existing fire hazards analysis for the area. This part of the evaluation must be performed by the person responsible for the fire safety program for the plant. The assessment must include the effect on combustible loading and distribution and the consideration of whether circuits or components, including associated circuits, for a train of equipment needed for safe shutdown are being affected or a new element introduced in the area. If this evaluation concludes
, that there is no significant impact, this conclusion and its basis must be documented as part of the 50.59 evaluation and be available for future inspection and reference. If the evaluation finds that there is an impact that could result in the area either not being in conformance with Appendix R, or some other aspect of the approved fire protection program, or being outside the basis for an exemption that was granted for the area involved, the licensee must either make modifications to achieve conformance or justify and request exemption (or, for the post 1979 plants, approval) from the NRC. See also responses to Questions 8.1 and 8.2.
8.5 Schedular and Blanket Exemptions QUESTION If an exemption is warranted and at the same time the provisions of the rule indicate that the appropriate schedular deadlines have passed, should a scheduler exemption be filed at the same time as the technical exemption request? If as part of the exemption request the utility is proposing to make modifications to achieve a reasonable level of conformance with t l Appendix R, and if the associated " clock" has run out for that type of modification, should the technical exemption request and the l description of the modification be filed with a schedular exemption? l Enclosure to GL 85-01, Re: Fire Protection Policy
10 , 3.1. 3 Exterior Walls QUESTION Must exterior walls to buildings and their penetrations be qualified as rated barriers?
RESPONSE
Exterior walls and their penetrations should be qualified as rated barriers when (1) they are required to separate a shutdown-related division (s) inside the plant from its redundant (alternate) counterpart outside the plant in the immediate vicinity of the exterior wall, (2) they separate safety related areas from non-safety related areas that present a significant fire threat to the safety related areas, or (3) they are designated as a fire barrier in the FSAR or FHA. Usually exterior walls are designated as a fire area boundary; therefore, they are evaluated by the guidelines of Appendix A. A FHA should be performed to determine the rating of exterior walls, if req'uired by the above criteria. 3.1.4 Exterior Yards QUESTION How should a utility define the boundaries of fire areas comprising exterior yards?
RESPONSE
An exterior yard area without fire barriers should be con-sidered es one fire area. The area may consist of several fire zones. The boundaries of the fire zones should be determined by a FHA. The protection for redundant / alternate shutdown systems within a yard area would be determined on the bases of the largest credible fire that is likely to occur and the resulting damage. The boundaries of such damage would have to be justified with a fire hazards analysis. The analysis should consider the degree of spatial separation between divisions; the presence of in-situ and transient combustibles, including vehicular traffic; grading; available fire protec-tion; sources of ignition; and the vulnerability and criti-cality of the shutdown related systems. Enclosure to GL 85-01, Re: Fire Protection Policy
- 47 8.8 Smallest Opening in a Fire Barrier QUESTION What is the smallest opening allowed in a fire area barrier for which an exemption request is not needed?
RESPONSE ' Unsealed openings in the configuration for which approval was obtained by an approved laboratory or the NRC staff would be acceptable. Our position on openings is given in Section 5.a(3) of BTP CMEB 9.5-1:
"(3) Openings through fire barriers for pipe, conduit, and cable trays which separate fire areas should be sealed or closed to provide a fire resistance rating at least equal to that required of the barrier itself. Openings inside conduit larger than 4 inches in diameter should be sealed at the fire barrier penetration. Open-ings inside conduit 4 inches or less in diameter should be sealed at the fire barrier unless the conduit extends at least S feet on each side of the fire barrier and is sealed either at both ends or at the fire barrier with non-combustible material to prevent the passage of smoke and hot gases. Fire barrier pene-trations that must maintain environmental isolation or pressure differentials should be qualified by test to maintain the barrier integrity under such conditions."
The unsealed opening (s) allowed in a fire area boundary or a barrier which separates redundant shutdown divisions should not permit flame, radiant energy, smoke and hot gases to pass through the barrier and cause damage to redundant shutdown divisions on the other side. The licensee should assess the adequacy of existing protection and should determine the minimum size based on a fire hazards analysis and con-servative fire protection engineering judgment. If the significance of openings in fire barriers is marginal, a formal exemption request could be submitted or the staff consulted. The basis for the lack of significance should be available for review by NRC Inspectors. Our acceptance of unprotected openings in fire barriers would depend
- upon the quantity and nature of combustible materials on either side I of the barrier; the location of the opening (s) in relation to the l ceiling (for openings in walls); the location, vulnerability and importance of shutdown systems on either side of the barrier; and compensating fire protection.
t l ! i. l Enclosure to GL 85-01, Re: Fire Protection Policy L
l . . 12 Section III.G recognizes that the need for alternate or dedicated shutdown capability may have to be considered on the basis of a fire area, a room or a fire zone. The alter-native or dedicated capability should be independent of the fire area where it is possible to do so (See Supplementary Information for the final rule Section III.G). When fire areas are not designated or where it is not possible to have the alternative or dedicated capability independent of the fire area, careful consideration must be given to the selection and location of the alternative or dedicated shut-down capability to assure that the performance requirement set forth in Section III.G.1 is met. Where alternate or dedicated shutdown is provided for a room or zone, the capability must be physically and electrically independent of that room or zone. The vulnerability of the equipment and personnel required at the location of the alternative or dedicated shutdown capability to the environments pro-duced at that location as a result of the fire or fire suppressant's must be evaluated. These environments' may be due to the hot layer, smoke, drifting suppressants, common ventilation systems, common drain systems or flooding. In addition, other interactions between the locations may be possible in unique configurations. If alternate shutdown is provided on the basis of rooms or zones, the provision of fire detection and fixed suppression is only required in the room or zone under consideration. Compliance with Section III.G.2 cannot be based on rooms or zones. 3.1. 6 Documentation QUESTION In Generic Letter 83-33 at pg. 2, the NRC Staff referred to the guidance in Appendix A to BTP 9.5-1 to establish the rating of the barrier. What level of documentation must be provided to verify that the fire area meets the requirements of Appendix R?
RESPONSE
The documentation required to verify the rating of a fire barrier should include the design description of the barrier and the test reports that verify its fire rating. Reference can be made to UL listed designs. Enclosure to GL 85-01, Re: Fire Protection Policy i
49 ( The guidelines identified in the footnotes to 50.48(a) Guidelines documents issued af ter January 1,1979. Commitments made to meet the requirements of Appendix R or specific sections such as III.G, III.J and 111.0. BTP CMEB 9.5-1, which includes the requirements of Appendix R and the previous guidance documents incorporated into the Branch Technical Position. The license for each plant licensed after January 1,1979 contains a license condition which identifies by reference the approved fire protection program for that plant. 8.12 Cold Shutdown Equipment Availability QUESTION A. Can a licensee achieve compliance with III.G.1(b) by demonstrating that one train of cold shutdown equipment will remain free of fire damage? [ B. In demonstrating that one train of cold shutdown equipment will remain free of fire damege, is a licensee limited to the three alternatives in III.G.2
RESPONSE
A. Yes. , B. No. 8.13 Guidance Documents QUESTION Please list all NRR guidance documents and position papers issued since Appendix R was promulgated.
RESPONSE
Fire Protection Guidance Issued Since January 1, 1975: IE Information Notices No. 83-41: Actuation of fire suppression systems causing inoperability of safety related equipment.
/ No. 83-69: Improperly installed fire dampers at nuclear power plants.
Enclosure to GL 85-01, Re: Fire Protection Policy
t J 14 , During the Appendix A review, licensees began to propose fire barriers to enclose cable trays, conduit, fuel lines, coolant lines, etc. Industry did not have standard rating tests for such components or for electrical, piping or bus duct penetrations. The NRC issued a staff position giving acceptance criteria for electrical penetration tests. These
, criteria require an analysis of any temperature on the unexposed side of the barrier in excess of 325'F. In the past, manufacturers designed their own qualification tests.
Nuclear Insurers, and the Institute of Electrical and Electronic Engineers have issued tests for some of these components. These tests usually exposed the component to the ASTM E-119 time temperature curve, but all had different acceptance criteria. Conduit and cable tray enclosure materials accepted by the NRC as 1 hour barrier prior to Appendix R (e.g. some Kaowool and 3M materials) and already
- installed by the licensee need not be replaced even though 1
they may not have met the 325'F criteria. However, new material should meet the 325'F criterion. Justification of
- temperatures which exceed 325'F is required.
I 3.2.2 Deviations from Tested Configurations 4 QUESTION Due to obstructions and supports, it is often impossible to
; achieve exact duplication of the specific tested configura-tion of the one-hour fire barriers which are to be placed around either conduits or cable trays. For each specific instance where exact replication of a previously tested ; configuration is not and cannot be achieved, is an exemption necessary in order to avoid a citation for a violation?
! RESPONSE No. Where exact replication of a tested configuration cannot be achieved, the field installation should meet all of the following criteria:
- 1. The continuity of the fire barrier material is maintained.
I i 2. The thickness of the barrier is maintained.
- 3. The nature of the support assembly is unchanged from the tested configuration.
! Enclosure to GL 85-01, Re: Fire Protection Policy
i 51 (
- 7. Letter dated 10/31/80 - Enclosing new 10 CFR 50.48 regarding
- fire protection schedules for operating nuclear power plants.
- 8. Letter dated 11/24/80 - Enclosing a copy of revised 10 CFR 50.48 and new App. R to 10 CFR 50, and a summary of open items from the SER for the BTP APCSB 9.5-1 review.
(
- 9. Letter dated 2/20/81 - Generic Letter 81-12 identifying informa-tion needed for NRC review of modifications for alternative shutdown capability.
l
- 10. Letter dated 4/7/82 - Provided clarification to Generic Letter 81-12 and guidance on information needed for NRC review of exemption requests.
. 11. Letter dated 10/6/82 - Generic Letter 82-21; provided criteria for annual, biennial, and triennial audits required by Technical Specifications.
*12. Letter dated 10/19/83 - Generic Letter 83-33; NRC Positions on Certain Requirements of Appendix R to 10 CFR 50.
Staff Generic Positions
- 1. Letter, Denton to Bernsen, dated 4/20/82 - Control room fires.
*2. SECY 83-269, dated July 5, 1983 - Attachments B and C.
- 3. Memo, Eisenhut to 01shinski, dated 12/20/83 - Physical independ-ence of electrical systems.
- 4. Memo, Eisenhut to Jordan, dated 10/24/83 - Bullet resistant fire doors.
" Staff positions regarding the need for certain exemptions delineated 4
in this guidance document have been revised per the " Interpretations
; of Appendix R".
8.14 Deviation From Guidance Documents QUESTION If a utility determines that a deviation from a guidance document exists, does an exemption request need to be filed? If so, what is the legal basis for this requirement?
RESPONSE
No. ( ( Enclosure to GL 85-01, Re: Fire Protection Policy ,
e
- I I I
! l f , 16 ,
{ i l RESPONSE l l-The NRC does not define the structural steel supporting i fire barriers. This steel is identified by the licensee.
] Qur position regarding the need to protect the structural < steel, which forms a part of or supports fire barriers, is ;
l consistent with sound fire protection engineering principles i i as delineated in both NFPA codes and standards, and The ! Fire Protection Handbook. !
- 3.3.2 Previously Accepted Structural Steel l 1
[. QUESTION Is it necessary to protect structural steel in existing i fire barriers where those barriers were approved in an ,
- Appendix A SER7
- RESPONSE i No. l i
! 3.3.3 Seismic Supports : i : QUESTION [ ! Does structural steel whose sole purpose is to carry dynamic i Ioads from a seismic event require protection in accordance l with Section !!!.G.2a of Appendix R7 I RESPONSE I No, unless the failure of any structural steel member due j { to a fire could result in significant degradation of the - i fire barrier. Then it must be protected. 1 3.3.4 Cable Tray Support Protection - i l
- QUESTION ,
i I i Should cable tray supports be protected if there is a : 1 sprinkler system in the fire area? Under what conditions may cable tray supports be unprotected? Do unprotected j i supports require an exemption? ' I i l l { i l l Enclosure to GL 85-01, Re: Fire Protection Policy r i
53 8.18 Coordination Study Update QUESTION Circuit modifications are an ongoing process. How recent must a coordination study be in order to be valid in protecting circuits associated by common power source?
RESPONSE
We would expect that as circuit modifications are made, the design j package would address the electrical protection required and the effects of this protection on the coordination of the protection for the power distribution system. This type of consideration should be included in the evaluation required by 10 CFR 50.59 Changes, Tests and Experiments. The design package and modification evaluation could not be complete without consideration of the coordination study. Therefore, we would expect that the coordination studies would be current with the last circuit modification made. 8.19 Exemption Request Threshold QUESTION (a) What is the threshold for exemption requests? (b) Is it necessary to file a request for each and every possible deviation from Appendix R7 RESPONSE i Typical examples are discussed in the response to Questions 8.21.1 through 8.21.6. ; (a) The licensee must develop its criteria for an exemption request threshold. ; (b) No. 8.19.1 Penetration Designs Not Laboratory Approved QUESTION Where penetration designs have been reviewed and approved by NRC but have not been classified by an approval laboratory, will it be necessary to submit an exemption request?
RESPONSE
No. 4 Enclosure to GL 85-01. Re: Fire Protection Policy
e ,
, t ,' ,
e i
\ ! 't b]
- i
'18 '
t 3.4.2 NE Consultation QUESTION
'f i Section 4.1.2 of NFPA-STDc13 allows for " partial installa-tions" or partial conr %e. The standard states that s "the authority having ;,urirdiction shall be consultedlin nach case." li'th the NRC ds .sutharity in this insta.1ce, must consulation occur only through the exemption prciess? (
RESPONSE
No. The staff is always available to consult with utility ' representatives and provide guidance as to the acceptability of a particular fire protection configuration in individual pl6nt areas. 3.4.3 Sprinkler Location QUESTION , s How does a supprossion system designer know whethic the~ term "throughout the area" means that sprinkler heads must be above or below cable trays when, in nis judgment, the i hazard of concern is a floor based f sre? ,
RESPONSE
( ' ' Section C.6.c(3) of BTP CMEB 9.5-1 states:
"(3) Fixed water extinguishing systems should conform to requirements of appropriate standards such as NFPA-13,
" Standard for the Installation of Sprinkler Systems,"
ard NFPA-15. " Standard for Water,Speny Fixed System?"." x i s, This question pertains to those sprinkler systems covered ' by NFPA-13. Chapter 4 of NFPA-13 provides guidance as to %'
- the location of sprinkler heads in relation to co7; mon obstructions. In general, to achieve complete area-wide ,
coverage, sprinklers should be located at the ceiling, with additional sprinklers provided below significant obstructions such as wide HVAC ducts and " shielded" or solid bottom stacked cable trays, To the extent that an existing or proposed sprinkler system desigrs deviates from this concept, the g design woulo have to be justified by a fire hazards Analysis. 3
\
Fire P'otection Policy Enclosure to GL 85-01, Re: ,
t 55 8.21 NRC Approval for BTP CMEB 9.5-1 Deviations QUESTION Do future deviations from BTP CME 8 9.5-1 guidelines require approval by the NRC? Do such deviations constitute a violation of license conditions?
RESPONSE
Compliance with guidelines in the BTP is only required to the extent that they were icorporated in the approved Fire Protection Program as identified in the license condition. When the new license condition is in place (See Response 8.2), future deviations may be made in accordance with the procedure stated therein. With present nonuniform license conditions, such deviations may or may not require a license amendment. In the absence of a license amendment, a violation may exist. l l l t l l \ Enclosure to GL 85-01, Re: Fire Protection Pnlicy L
I 20 3.4.6 Previously Approved Suppression Systems QUESTION Must suppression systems approved and installed under BTP APCSB 9.5-1, Appendix A be extended or altered to meet the total area requirements of Section III.G (as interpreted by tne Staff) or does this " requirement" only apply to new installations?
RESPONSE
Suppression systems installed in connection with Appendix A may or may not have to be extended as a result of III.G. { The licensee must analyze each area where suppression is required by III.G, and where only partial suppression has been provided, determine if the coverage is adequate for the fire hazard in the area. The licensee may consult with the staff during this review. In any event, the Appendix R analysis showing that the suppression provided .is adequate must be retained and available for NRC audit. z 3.5 Separation of Redundant Circuits ,s 3.5.1 Twenty-Foot Separation Criteria .. QUESTION Assuming that a licensee is utilizing the 20-foot separation for circuit protection, could an exemption request be granted for a portion of the circuit that did not maintain the 20-foot minimum separation if that portion was protected by one-hour barrier until 20-foot was achieved? This barrier would not be firewall-to-firewall, and the circuit protection would not be claimed under the one-hour barrier rule.
RESPONSE
With the erection of a partial' qualified one-hour rated barrier for portions of the circuits with less than 20 ft. separation, if 20 feet of horizontal separation existed between the redundant unprotected portions of the circuits without intervening combustibles or fire hazards, and if the fire area was protected by automatic fire detection and suppression, compliance with Section III.G.2.b would be achieved. These types of configuration have to be evaluated on a case-by-case basis. Enclosure tn GL 85-01, Re: Fire Protectinn Policy _ _ _ _ _ _ _ _ _ _ _ _ . - . . - )
}
I 57 l
RESPONSE
Yes, NT0Ls will be subject to the Appendix R audit; the TI 2515/62 is being revised to reflect the appropriate requirements for NT0Ls' and it is our intent to conduct such inspections prior to issuing the operating license. 10 CFR 50.48 requires each such plant to have a fire protection plan. Their operating license will contain a specific license condition to implement their approved fire protection program which must identify deviations from Appendix R. The fire protection inspections will be against the particular license conditions.
- 9. 4 Future TI 2515/62 Revisions
~ QUESTION Does the NRC plan to issue a new or revised version of Temporary Instruction 2515/62 for future Appendix R audits?
RESPONSE
Yes. 9.5 Documentation Supplied by Licensee QUESTION Temporary Instruction 2515/62 provided a list of documentation that the NRC needs to review as part of the audit process. In past audits, the NRC has requested additional information other than that contained on the list. Will a new list of documentation be developed?
RESPONSE
The documentation listing provided in TI-2515/62 does not restrict the inspection team from enhancing inspection efficiency by requesting a licensee to provide additional relevant documentation. A new listing of documentation for TI-2515/62 is not being developed. 9.6 Subsequent Inspections QUESTION To what extent will Appendix R issues be raised at future Regional I&E Fire Protection audits after a successful Appendix R audit? For example, if an area has already been reviewed and no noncompliance found, will it be subject to later review and reinterpretation by the Staff?
\
Enclosure to GL 85-01, Re: Fire Protection Policy
22
- 1. A relatively large horizontal spatial separation between redundant divisions; all cables qualified to IEEE-383.
- 2. The presence of an automatic fire suppression system over the intervening combustible (such as a cable tray fire suppression system);
- 3. The presence of fire stops to inhibit fire propagation in intervening cable trays;
- 4. The likely fire propagation direction of burning inter-vening combustibles in relation to the location of the vulnerable shutdown division;
- 5. The availability of compensating active and passive fire protection.
Any future changes in the cable configuration due to modif-ications could be handled under 50.59. See the provisions of the license condition in the response to question 8.2. 3.6.2 In-Situ Exposed Combustibles QUESTION } Within Appendix R, Section III.G.2.b, the phrase " twenty feet with no intervening combustible or fire hazards" is utilized. What is the definition of "no intervening combus-tible"? Is the regulation focused predominantly on the absence of fixed combustibles?
RESPONSE
There is no specific definition of "no intervening combus-tible." The regulation is focused on the absence of in-situ exposed combustibles. Non combustible materials would not be considered as intervening combustibles. In BTP CMEB 9.5-1, noncombustible material is defined as:
" Noncombustible Material
- a. A material which in the form in which it is used and under the conditions anticipated, will not ignite, burn, support combustion, or release flammable vapors when subjected to fire or heat.
Enclosure to GL 85-01, Re: Fire Protection Policy
t 59
RESPONSE
To the extent that a licensee's submittal to NRR is comprehensive and sufficiently detailed, the basis for the OI&E Appendix R inspection will be the assumptions, shutdown paths and equipment selections approved by NRR. If the inspection results in new information that casts doubt upon the approved configuration, the Regional inspectors have the responsibility to resolve such doubts. 9.9 List of Shutdown Equipment QUESTION What lists of shutdown equipment will be used by the Regional inspectors, if the shutdown analysis has not been reviewed and approved by NRR?
RESPONSE
Regional Inspectors will use the lists of shutdown equipment the licensee has identified in his fire protection plan. Generic Letter 81-12 and its clarification documents expect licensees to show how they will shutdown if a fire area is not provided with redundant train separation. Inherent within this expectation is *he assumption that the licensee will identify the equipment to be usee It is because the licensees have not had fire hazard analyses at ali for non-alternative shutdown fire areas that the inspectors to date have resorted to using the only lists available (the alternative shutdown equipment list used by NRR in their reviews). It is unlikely there would not be a list of at least those systems to be used for alternate shutdown, since 10 CFR 50.48 requires NRR review and approval of the means of alternate shutdown. i Enclosure to GL 85-01, Re: Fire Protection Policy
24
- 3. 7 Radiant Energy Shield 3.7.1 Fire Rating QUESTION Recently, the NRC Staff indicated that non-combustible radiant energy shields should be tested against ASTM-TD-E-Il9 based, apparently, on the requirements of BTP CMEB 9.5-1, Rev. 3, a document issued after Appendix R was promulgated.
This new requirement would not appear to be required by Appendix R or BTP APCSB 9.5-1 Appendix A. Could the Staff clarify the requirements in this area? ANSWER During the Appendix A reviews, we observed that inside some containments, there were large concentrations of cables converging at electrical penetration areas. In some cases, where the penetrations were grouped by division, shields were placed between the divisions so that radiant energy from a fire involving the cables of one division would not degrade or ignite cables of the other divisions. s These shields also directed the convective energy from the fire away from the surviving division. These shields were usually constructed of 1/2-inch marinite board in a metal . frame. Appendix R, Section III.G.f refers to these shields as "a noncombustible radiant energy shield." The guidelines in BTP CMEB 9.5-1, Section C.7.a(1)b. indicate that these
~ shields should have a fire rating of 1/2 hour. In our opinion any material with a 1/2 hour fire rating should be capable of performing the required function.
The guidelines of BTP CMEB 9.5-1 relating to a firerated radiant energy shield are being considered in our current reviews of NTOL plants. However, to the extent that an l applicant can justify that a proposed radiant energy shield l can achieve an equivalent level of safety, we have been accepting shields that have not been tested against the j acceptance criteria of ASTM E-119. In our Appendix R reviews, we have accepted non-fire-rated radiant energy shields that have been demonstrated by fire hazards analysis to provide an acceptable level of protection against the anticipated hazard of a localized fire within the containment. We have also accepted fire-rated metal-sheathed mineral insulated cables, as a radiant energy shield in specific con'igurations. I I Enclosure to GL 85-01, Re: Fire Protection Policy l
E p* *Ecoq
# jo, UNITED STATES E s.( j NUCLEAR REGULATORY COMMISSION ENCLOSURE 7
, .; j WASHINGTON, D. C. 20555 h[****'
cox / SEP 211984 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, IE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
FIRST MEETING OF FIRE PROTECTION POLICY STEERING COMMITTEE Sumary At its first meeting, the Fire Protection Policy Steering Committee (SC) discussed the issues identified in the ED0 memo of September 13, 1984 The SC made the following recommenda' ions and assignments:
. the SC recommended that Appendix R implementation policy follow the " Interpretations" discussed at the Regional Workshops, as modified by details discussed below, rather than requiring prior i
i staff review or exemptions for deviations from Generic Letter s 83-33;
. the SC recommended that the issuance of the fire protection enforcement policy and enforcement actions be expedited;
. the SC recommended that QA for fire protection be clearly defined as that required by GDC-1; and
. the SC assigned to the Working Group tasks dealing with the adequacy of available guidance, comparison of fire protection requirements for ors and NT0Ls, the adequacy of current inspection practices, and identification of outstanding technical issues.
Introduction As a result of the ED0 memo of September 13, 1984, regarding a review of NRC fire protection policy and programs, the SC held its first meeting
- on September 13 and 14 The objectives of this meeting were to:
- Attended by: I. Martin, J. Olshinski, L. Spessard, N. Grace, F. Rosa, ~
S. Richardson, W. Shields, W. Little, T. Wambach, W. Olmstead, K. Cyr and R. Vollmer.
's Enclosure to GL 85-01, Re: Fire Protection Policy
e 26
RESPONSE
If the system is being used to provide its design function, it generally is considered redundant. If the system is being used in lieu of the preferred system because the redundant components of the preferred system does not meet the separation criteria of Section III.G.2, the system is considered an alternative shutdown capability. Thus, for the example above, it appears that the condensate system is providing alternative shutdown capability in lieu of separa-ting redundant components of the RHR System. Fire detection and a fixed fire suppression system would be required in the area where separation of redundant components of the RHR system is not provided. However, in the event of a turbine building fire, the RHR system would be used for safe shutdown and is not considered an alternative capability. However, one train of the RHR system must be separated from the turbine building. 3.8.4 Control Room Fire Considerations QUESTION What considerations should be taken into account in a control room fire? What is the damage that is considered? What actions can the operators take before evacuating the CR? When can the control room be considered safe after a fire for the operator to return?
RESPONSE
The control room fire area contains the controls and instru-mental redundant shutdown systems in close proximity (i.e. ! usually separation is a few inches). Because it is possible i to provide shutdown capability that is physically and electrically independent of the fire area, it is our opinion that alternative or dedicated shutdown capability and its associated circuits for the control room be indepen-dent of the cables system and components in the control room fire area. The damage to the system in the control room for a fire that causes evacuation of the control room cannot be pre-dicted. A bounding analysis should be made to assure that ! safe conditions can be maintained from outside the control room. This analysis is dependent to the specific design. The usual assumptions are: Enclosure to GL 85-01, Re: Fire Protection Policy l
O 8 The SC will take advantage of the Working Group chaired by Faust Rosa and make assignments to that group and to certain individuals. It was our aelief that use of currently available resources, including the Working Group, would be sufficient to achieve our objectives and schedule. Issues
- 1. Adequacy of current guidance to industry.
The SC felt that enough guidance has been generated but that it needed to be made consistent with our recommendations and cleared up technically in some areas. The mechanism for this would be a generic letter, super-ceding previous guidance, to be sent to all licensees as promised at the , rtjional workshops. This generic letter would be sent to the Commission, as requested at the Commission's May 30 meeting, before'being issued to industry. The SC agreed that it would be best to utilize guidance already available, to the extent possible, to minimize possible confusion both within ind.stry and the NRC. It was also agreed that we should utilize either Generic Letter 83-33 or the " Interpretations" rather than a third option for dealing with the exemption issue. This is discussed below. The Working Group was tasked with reviewing all current guidance and outstanding technical questions and to revise the Regional Workshop package to incorporate in one place a comprehensive set of guidance that is consistent with the SC's policy recommendations and the approved technical recommendations of the Working Group.
- 2. Interpretation of the Appendix R requirements vice staff guidance.
The basic issue is whether industry can deviate from the contents of Generic Letter 83-33 without prior staff review and approval. This issue is fully developed by the current Staff DFO. The SC had the benefit of ELO's views and Faust Rosa's recommendations resulting from his assign-ment by Mr. Denton to make an independent assessment of this OPO. The SC decided that it could only support the contents of 83-33 as guidance to industry, consistent with the " Interpretations" drafted for the Regional Workshops since ELD advised that Generic Letter 83-33 stated requirements which went beyond the terms of Appendix R itself. However, the SC also felt that some of the clarifying language contained in the DP0 should be utilized and that specific guidance should be supplied to clearly indicate the level of fire protection to be achieved and the documentation necessary to demonstrate it. In addition, the SC felt that, at an appropriate time prior to the Appendix R inspection, the licensee should be requested to provide information necessary to ( Enclosure to GL 85-01, Re: Fire Protection Policy
t 28 .~, t
- 4. EMERENCY LIGHTING 4.1 Illumination Levels QUESTION
~
What is the requisite intensity level for emergency lighting for egress roJtes and areas where shutdown functions must be performed? What are the bases for determining these levels of lighting? ANSWER The level of illumination provided by emergency lighting in access routes to and in areas where shutdown functions must be performed is a level that is sufficient to enable an operator to reach that area and perform the shutdown functions. At the remote shutdown panels the illumination levels should be sufficient for control panel operators. The bases for estimating these levels of lighting are the guidelines contained in Section 9.5.3 of the Standard Review Plan, which are based on industry standards (i.e., Illuminating Engineering Society Handbook). -my
)
Where a licensee has provided emergency lighting per Section III.J of ' Appendix R, we would expect that the licensee verify by field testing that this lighting is adequate to perform the intended tasks. Enclosure to GL 85-01, Re: Fire Protection Policy
- 6. Identification and resolution plan for outstanding technical issues.
The Working Group was requested to identify any outstanding technical issues that impede Appendix R compliance. These will be discussed with the SC and a plan and schedule for resolution will be developed. One issue that was discussed is the QA standard to be applied to fire protection compliance. The SC felt strongly that the requirements of GDC-1 apply to fire protection features, recognizing that the activities are underway to resolve the GDC-1/ Appendix B and safety-related/ important-to-safety issues. Without trying to muddy the water in these areas, the SC felt that, at minimum, licensees be made aware that the fire protection program falls under GDC-l. The Working Group should look into thic also and review the QA commitments made by some licensees in their pre-Appendix R SERs. The next meeting of the Fire Protection Steering Committee will be held on September 27 at 9:00 a.m. in P-202A. Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello R. Minogue T. Murley J. O'Reilly J. Martin E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa (10) SC Members t Enclosure to GL 85-01, Re: Fire Protection Policy I
l 30 ..
RESPONSE
The definitional process mentioned considers an alternative shutdown capability provided under the Appendix A review as a redundant shutdown capability under the Appendix R review. This definitional process is incorrect. For the purpose of analysis to Section III.G.2 criteria, the safe shutdown capability is defined as on.e of the two normal safe shut-down trains. If the criteria of Section III.G.2 are not met, an alternative shutdown capability is required. The alternative shutdown capability may utilize existing remote shutdown capabilities and must meet the criteria of Sections III.G.3 and III.L of Appendix R. See also the response to 5.1.3. 5.1.3 III.L Backfit QUESTION Why do the Staff interpretive memoranda regarding the criteria for satisfaction of Section III.L form the audit-able basis for determining compliance to Appendix R when the Commission failed to backfit this section to all plants?
RESPONSE
Although 10 CFR 50.48(b) does not specifically include Section III.L. with Sections III.G., J., and O. of Appendix R as a requirement applicable to all power reactors licensed prior to January 1, 1979, the Appendix, read as a whole, and the Court of Appeals decision on the Appendix, Connecticut Light and Power, et al. v. NRC, 673 F2d. 525 (D.C. Cir., 1982), demonstrate that Section III.L. applies to the alter-native safe shutdown option under Section III.G. if and where that option is chosen by the licensee. 5.2 Procedures 5.2.1 Shutdown and Repair Basis QUESTION With regard to the term " post-fire procedures" the Commission states that it is impossible to predict the course and extent of a fire. Given this, how does one write post-fire shutdown and repair procedures that are both symptomatic and usable to an operator? Enclosure to GL 85-01, Pe: Fire Protection Policy
#s o arco f 'o UNITED STATES l' n NUCLEAR REGULATORY COMMISSION
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....* SEP 2 61984 MEMORANDUM FOR: Fire Protection Policy Steering Committee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
AGENDA FOR SECOND MEETING The second meeting of the Fire Protection Policy Steering Committee will be held at 9:00 a.m. on September 27, 1984 in Room P-202 A. At this meeting we will be briefed on the Working Group activities and on current inspection content and schedules and have the benefit of the IE and NRR Office Director's views on our activities. Our tentative schedule is as follows: 9:00 a.m. Executive Session 9:30 Harold Denton and Ed Case 10:00 Dick DeYoung and Jim Taylor 10:30 Faust Rosa will discuss Working Group activities 1:00 p.m. Steve Richardson will discuss Appendix R inspections 2:00 Committee work In addition to our consideration of issues raised, by the above agenda items, we need to consider policy approaches to deal with technical and schedular exemptions which put implementation of Appendix R into the distant future. Any approaches you have to deal with this issue will be welcomed. Y r- _ Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello H. Denton R. DeYoung T. Murley R. Minogue J. O'Reilly J. Keppler J. Collins J. Martin G. Cunningham E. Case J. Taylor D. Eisenhut 1 R. Bernero G. Arlotto F. Rosa Enclosure tn GL 85-01, Re: Fire Protection Policy
32 s
RESPONSE
Yes. The only requirement for post-fire operating procedures is for those areas where alternative shutdown is required. For other areas of the plant, shutdown would be achieved utilizing one of the two normal trains of shutdown system. Shutdown in degraded modes (one train unavailable) should be covered by present operator training and abnormal and emergency operating procedures. If the degraded modes of operation are not presently covered, we would suggest that the operation staff of the plant determine whether additional training or procedures are needed.
- 5. 2. 4 Post Fire Procedures Guidance Documents QUESTION 00 any NRC Staff guidance documents exist relative to the extent, form, nature, etc. of Appendix R post-fire operating procedures?
RESPONSE
No. Other than the criteria of Section III.L, no specific post-fire shutdown procedure guidance has been developed. '
)'
See also responses to 5.2.1, 5.2.2 and 5.2.3. 5.3 Safe Shutdown and Fire Damage 5.3.1 Circuit Failure Modes QUESTION What circuit failure modes must be considered in identifying circuits associated by spurious actuation?
RESPONSE
Sections III.G.2 and III.L.7 of Appendix R define the circuit failure modes as hot shorts, open circuits, and shorts to ground. If the concern is spurious actuation of equipment, actual circuit failure modes could be bypassed by assuming all possible failure states for the equipment (valves could fail either open or closed). Enclosure to GL 85-01, Re: Fire Protection Policy
/ 'o
~,, UNITED STATES
!" o NUCLEAR REGULATORY COMMISSION j, .I wAssimoTON. D. C. 20555 0.,
% . . . . . #g SEP 2 71984 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, 0IE Thomas E. Murley, Regional Administrator, R-I James P. O'Reilly, Regional Administrator, R-II James G. Keppler, Regional Administrator, R-III John T. Collins, Regional Administrator, R-IV John B. Martin, Regional Administrator, R-V FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
NRC FIRE PROTECTION MEETING The Fire Protection Policy Steering Comittee requests that the fire protection engineers in your organization attend a meeting in Bethesda on October 2, 1984, to give their views on licensing and inspection fire protection issues. This is an information gathering session to help the Comittee formulate its policy recommendations. A candid discussion of how these issues are viewed by the fire protection reviewers and inspectors, and recomendations they may have, would be welcome. A meeting agenda is enclosed.
~
ichard H. Vollmer, Chairman ~ Fire Protection Policy Steering Comittee
Enclosure:
Agenda cc: W. Dircks V. Stello R. Minogue G. Cunningham E. Case J. Taylor, IE D. Eisenhut R. Bernero G. Arlotto, RES F. Rosa N. Grace, IE L. Spessard, R-III W. Olmstead, ELD J. Olshinski, R-II T. Wambach Enclosure to GL 85-01, Re: Fire Protection Policy
34 s
RESPONSE
As stated in Section III.G.1, one train of systems needed to achieve and maintain hot shutdown conditions must be free of fire damage. Systems necessary to achieve and maintain cold shut-down can be repaired within 72 hours. Thus, if this certain equipment necessary only in the cooldown phase, is used to achieve cold shutdown, it can be repaired within 72 hours. If the certain equipment is maintaining hot shutdown while repairs are being made, one train must be free of fire damage. 5.3.5 Pressurizer Heaters QUESTION Most PWRs do not require pressurizer heaters to maintain stable conditions. In fact, the Commission does not con-sider heaters to be important to safety and they are not required to meet Class IE requirements. Are they required for hot shutdown under Appendix R? If yes, then how does a plant meet the separation requirements of Section III.G.2.d,
- e. or f without major structural alterations to the pressurizer? s
\
RESPONSE / One train of systems necessary to achieve and maintain hot shutdown conditions must be free of fire damage. PWR licen-sees have demonstrated the capability to achieve and maintain stable hot shutdown conditions without the use of pressurizer heaters by utilizing the charging pump and a water solid pressurizer for reactor coolant pressure control. 5.3.6 On-Site Power QUESTION Appendix R, Section III.L.4 states in part, "If such equipment and systems will not be capable of being powered by both on-site and off-site electrical power systems because of fire damage, an independent on-site power system shall be provided." Again, in Appendix R, Section III.L.5, the statement is made "If such equipment and systems used prior to 72 hours after the fire will not be capable of being powered by both onsite and offsite electrical power systems because of fire damage, an independent onsite power system shall be provided." An interpretation is needed of the meaning and the applicability of these two quotes relative to alternative shutdown capabilities. Enclosure to GL 85-01, Re: Fire Protection Policy
NRC Fire Protection Meeting Date: October 2, 1984 at Location: P-422 Phillips Building 9:00 a.m. Opening Remarks - R. Vollmer 9:10 Region I Remarks 9:40 Region II Remarks 10:10 Region III Remarks 10:40 Region IV Remarks 11:40 Brookhaven Remarks 1:00 p.m. NRR Remarks 2:00 I&E Remarks 2:30 to 4:00 Group Discussion 4:00 to 5:00 Each group given. opportunity to provide any additiona,1 comments. 5:00 Adjourn Enclosure to GL 85-01, Re: Fire Protection Policy
36 -
RESPONSE
Yes. To meet the separation criteria of Section III.G.2 and II.G.7 of Appendix R, high impedance faults should be considered for all associated circuits located in the fire area of concern. Thus, simultaneous high impedance faults (below the trip point for the breaker on each individual ciruit) for all associated circuits located in the fire area should be considered in the evaluation of the safe shutdown capability. Clearing such faults on non essential circuits may be accomplished by manual breaker trips governed by written procedures. 5.3.9 Diagnostic Instrumentation QUESTION What is diagnostic instrumentation?
RESPONSE
Diagnostic instrumentation is instrumentation, beyond that previously identified in Attachment 1 to I&E Information Notice 84-09, needed to assure proper actuation and func- ' tioning of safe shutdown equipment and support equipment (e.g., flow rate, pump discharge pressure). The diagnostic instrumentation needed depends on the design of the alter-native shutdown capability. Diagnostic instrumentation, if needed, will be evaluated during the staff's review of the licensee's proposal for the alternative shutdown capability. 5.3.10 Design Basis Plant Transients QUESTION What plant transients should be considered in the design of the alternative or dedicated shutdown systems?
RESPONSE
Per the criteria of Section III.L of Appendix R. a loss of offsite power shall be assumed for a fire in any fire area concurrent with the following assumptions:
- a. The safe shutdown capability should not be adversely affected by any one spurious actuation or signal resulting from a fire in any plant area; and Enclosure to GL 85-01, Re: Fire Protection Policy
e *
>R Rf Cp o UNITED STATES
~
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{ $ WASHINGTON, D. C. 20555 k+..../ October 3,1984 MEMORANDUM FOR: Faust Rosa, Chairman Fire Protection Working Group FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
ASSIGNMENTS FOR FIRE PROTECTION WORKING GROUP At the first meeting of the Fire Protection Policy Steering Comittee, the Committee assigned action tasks to the Working Group. The purpose of this memo is to provide better focus on the Working Group assignments as a result of the second and third Comittee meetings. The Working Group's first priority should be the revision of the enclosure to the Generic Letter discussed at the fire protection workshops. This package should include answers to the ques' ions posed by industry and revision of the
" interpretations" section to utilize clarifying aspects gf the DP0 and to emphasize the level of fire protection expected by staff guidance, e.g.,
Generic Letter 83-33, and the documentation necessary to demonstrate equiva-lence to staff guidance. This package should not contain any new guidance and should be internally consistent and consistent with the Comittee's proposed policy on the " interpretations." To the extent possible criteria acceptable to the staff should be clearly identified. This package should be available for Comittee review by October 12 and whatever efforts are necessary to meet this date should be expended. Bill Shields will prepare a draft of the Generic Letter in which he will lay out the elements for expediting Appendix R implementation. It will discuss the proposed policy on schedular exemptions, the proposed augmentation of plant inspections, an~d indicate generally what is expected of licensees and what enforcement action may occur. In addition, it was agreed that the 50.48 schedule expiration for each p.lant would be indicated. The item of next priority is the development of an inspection plan which would accomplish the objectives of a " helpful" inspection. That is, in a one week, four-man inspection, we need to find out how the licensee is approaching his Appendix R implementation, the status of the plant with respect to compliance, and plans and schedules for complete implementation. This inspection should identify improper paths being taken by the licensee, and allow the staff to assess current fire protection safety and need for any plant specific enforcement action. Finally, the Working Group should compare fire protection requirements for ors and NT0Ls and work on any outstanding technical issues. The outline Enclosure to GL 85-01, Re: Fire Protection Pclicy
I . e 38
- 6. OIL COLLECTION SYSTEMS FOR REACTOR COOLANT PUMP 6.1 Lube Oil System Seismic Design QUESTION If the reactor coolant pump lube oil system and associated appurtenances are seismically designed, does the lube oil collection system also require seismic design? Is an exemption required?
RESPONSE
Where the RCP lube oil system is capable of withstanding the safe shutdown earthquake (SSE), the analysis should assume that only random oil leaks from the joints could occur during the lifetime of the plant. The oil collection system, therefore, should be designed to safely channel the quantity of oil from one pump to a vented closed container. Under this set of circumstances, the oil collection system would not have to be seismically designed. An exemption is required for a non-seismically designed oil collection system. The basis for this exemption would be that random leaks are not assumed to occur simultaneously with the seismic event, since the lube oil system is designed to withstand the seismic event. However, the Rule, as written, does not make this allowance. 6.2 Container QUESTION It would appear that a literal reading of Section III.0 regarding the oil collection system for'the reactor coolant pump could be met by a combination of seismically designed splash shields and a sump with sufficient capacity to contain the entire lube oil system inventory. If the reactor coolant pump is seismically designed and the nearby piping hot surfaces are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, does this design concept conform to the requirements of the rule?
RESPONSE
If the reactor coolant pump, including the oil system, is seismically designed and the nearby hot surfaces of piping are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, and it could be demonstrated by engineering analysis that sump and splash shields would be capable of preventing a fire during normal and design basis accident conditions, the safety objective of Section III.0 would be achieved. Such a design concept would have to be evaluated under the exemption process. The justification for the exemption should provide reasonable assurance that oil from all potential pressurized and unpressurized leakage Enclosure tn GL 85-01, Re: Fire Prntection Policy
1
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! n NUCLEAR REGULATORY COMMISSION
{ E WA$HINGTON, D. C. 20655
/ 41984 OCT MEMORANDUM FOR: Fire Protection Policy Steering Comittee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
AGENDA FOR THE FOURTH MEETING The fourth meeting of the Fire Protection Policy Steering comittee will be held at 9:00 a.m. on October 10, 1984 in the Region II offices in Atlanta. At this meeting we need to reach agreement on the enforcement policy guidance and develop our decision, reached at the second meeting and fortified in the third meeting, on a FY 85 inspection for most plant types. I request that John prepare a rewrite of the enforcement policy and Nelson a general plan for inspections for Comittee consideration. Other items we need to discuss are:
. Focus of inspections relative to fire protection and safe shutdown;
. Working group assignments;
. What should be included in technical specifications;
. Resolution of licensee / inspection team disputes; and
. Disputes of issues raised at third meeting.
I hope at this meeting we can agree on our overall approach to make our recomendations to Dircks coherent and consistent so that we can begin writing to have a draft report ready at our fifth meeting. Richard H. Vollmer, Chair an Fire Protection Policy teering Committee cc: See next page Enclosure to GL 85-01, Re: Fire Protection Policy
o . 40 l l
- 7. BRANCH TECHNICAL POSITION CMEB 9.5-1 7.1 Fire Protection and Seismic Events QUESTION For which situations other than the reactor coolant pump lube oil system are seismic events assumed to be initiators of a fire?
RESPONSE
The guidelines for the seismic design of fire protection systems which cover other general situations is delineated in BTP CMEB 9.5-1 C.1.C(3) and (4):
"(3) As a minimum, the fire suppression system should be capable of delivering water to manual hose stations located within hose reach of areas containing equipment required for safe plant shut-down following the safe shutdown earthquake (SSE). In areas of high seismic activity, the staff will consider on a case-by-case basis the need to design the fire detection and suppression systems to be functional following the SSE.
(4) The fire protection systems should retain their original design capability for (a) natural phenomena of less severity and greater frequency than the most severe natural phenomena (approximately once in 10 years) such as tornadoes, hurricanes, floods, ice i storms, or small-intensity earthquakes that are characteristic l of the geographic region, and (b) potential manmade site-related events such as oil barge collisions or aircraft crashes that have a reasonable probability of occurring at a specific plant site. The effects of lightning strikes should be included in the overall plant fire protection program." We have considered California as being a high seismic activity area. For those plants reviewed under Appendix A, our position is (A.4):
" Postulated fires or fire protection system failures 'need not be considered concurrent with other plant accidents or the most severe natural phenomena" l Our guidelines on the seismic design of fire protection systems l
installed in safety related areas are delineated in Regulatory Guide l 1.29 " Seismic Design Classification", paragraph C.2. The failure of l any system should not affect a system from performing its safety l function. l l Enclosure to GL 85-01, Re: Fire Protection Policy
/ 'o UNITED STATES !" % NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555
{g=.caj,,E October 12, 1984 MEMORANDUM FOR: Fire Protection Policy Steering Comittee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
AGENDA FOR THE FIFTH MEETING The fifth meeting of the Fire Protection Policy Steering Comittee will be held at 9:00 a.m. on October 17, 1984 in Room P-202A. At this meeting you should be prepared to discuss the generic letter and its attachments relative to the ability of these documents to satisfy Steering Comittee decisions. We should also be prepared to settle on the enforcement policy guidance and the inspection module revision if these are available. Finally we will need to discuss format, content, and writing assignments for our October 26 report to the EDO. b . . _ _ ichard H. Vollmer, Director Fire Protection Policy Steering Comittee cc: W. Dircks V. Stello H. Denton R. DeYoung R. Minogue , T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-III J. Collins, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa Enclosure to GL 85-01, Re: Fire Protection Policy
42 and valves for the portion of hose standpipe system affected by this functional requirement should, as a minimum, satisfy ANSI B31.1, ' Power Piping.' The water supply for this condition may be obtained by manual operator actuation of valves in a connection to the hose standpipe header from a normal seismic Category I water system such as the essential service water system. The cross connection should be (a) capable of providing flow to at least two hose stations (approximately 75 gpm per hose station), and (b) designed to the same standards as the seismic Category I water system; it should not degrade the performance of the seismic Category I water system." The post-seismic procedures should include a damage survey, and a determination of whether any fires were initiated as a result of the seismic event. See also the response to Question 7.1. l l l i i i l l l i l 1 Enclosure to GL 85-01, Re: Fire Protection Policy l [
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/gapafCp 'o UNITED STATES
! *' s., ^,j NUCLEAR REGULATORY COMMISSION
- a WASHINGTON, D. C. 20555 a -
AV E qcK.f.... October 12, 1984 MEMORANDUM FOR: Harold R. Denton, Director, NRR Richard C. DeYoung, Director, OIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
SECOND MEETING OF THE FIRE PROTECTION POLICY STEERING COMMITTEE HELD ON SEPTEMBER 27, 1984 Summary At its second meeting, the Fire Protection Steering Committee (SC) con-sidered policy approaches to deal with technical and schedular exemptions which threaten to put Appendix R implementation into the distant future, discussed SC activities with the IE and NRR Office Directors and Deputies, and was briefed on Working Group activities. As a result of this meeting, the SC made the following decisions:
. the SC decided to hold a meeting with all HQ and Regional fire protection engineers, as a body, to candidly discuss their views on fire protection issues, problems, and possible future actions;
. the SC decided that the most promising way to expedite Appendix R compliance is to initiate an aggressive inspection program which would steer and promote licensee compliance, assess the degree of fire safety, and exercise enforcement policy where appropriate; and
. the SC decided that no further schedular exemptions should be granted.
Discussion
- 1. Meeting with Office Directors and Deputies.
The SC met with Messrs. Denton, DeYoung, Case and Taylor to discuss their views on the fire protection problems, potential solutions, and SC activities. They, viewed the problem as a lack of staff cohesiveness in a highly judgemental area exacerbated by industry reluctance to meet Appendix R requirements. They felt that a meeting between the HQ and Regional fire protection staff to air any problems, issues, and possible solutions, would benefit the SC's work. We agreed and set up such a meeting for October 2nd. ( Enclosure to GL 85-01, Re: Fire Protection Policy
a . 44 ,
.The NRC has drafted new language for this license condition which delineates the circumstances under which the fire protection plan may be revised. We are now including this language in all new licenses and are considering amending present licenses. The revised language is as follows.
"9.5 Fire Protection Program (Section 9.5, SER)
- a. The licensee shall maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment _ and as approved in the SER through Supplement _ , subject to provisions b & c below.
- b. The licensee may make no change to the approved fire protection program which would decrease the level of fire protection in the plant without prior approval of the Commission. To make such a change the licensee must submit an application for license amend-ment pursuant to 10 CFR 50.90.
- c. The licensee may make changes to features of the approved fire protection program which do not decrease the level of fire pro-
- tection without prior Commission approval af ter such features s have been installed as approved, provided such changes do not otherwise involve a change in a license condition or technical /
specification or result in an unreviewed safety question (see 10 CFR 50.59). However, the licensee shall maintain, in an auditable form, a current record of all such changes including an analysis of the effects of the change on the fire protection program and shall make such records available to NRC inspectors upon request. All changes to the approved program made without prior Commission approval shall be reported annually to the Director of the Office of Nuclear Reactor Regulation, together with supporting analyses." 8.3 III G, J and 0 Exemptions for Future Modifications QUESTION Is an exemption required from Appendix R Sections other than III.G, III.J and III.0 for future modifications that do not comply with such sections?
RESPONSE
Yes. The exclusion of the applicability of Sections of Appendix R other than III.G., III.J., and III.0 is limited to those features
" accepted by the NRC staff as satisfying the provisions of Appendix A to Branch Technical Position BTP APCSB 9.5-1 reflected in staff fire protection safety evaluation reports issued prior to the effective date of the rule."
\
Reference:
10 CFR 50.48(b). N 3 1 Enclosure to GL 85-01, Pe: Fire Protection Policy ' ____ _ - - ._ - _ __--_ _- _ - - _ A
0 0
~~
October 12, 1984
- 3. Technical issues.
Two technical issues were identified to be pursued by the Working Group: (1) the Dow Corning ascertion that penetration material was not being properly installed; and (2) whether our criteria on control room fires are appropriate and consistent.
- 4. Working Group discussion.
Faust Rosa discussed the Working Group activities, in particular, the out-line of the Working Group report. All items of interest to the SC were covered in the outline of this report. However, it appeared that com-pletion of the work as outlined was too ambitious for the time and resources provided. The SC indicated that priority attention be given to the completion of the generic letter package. The SC also indicated in its first meeting that the Working Group should compare fire protection requirements for ors and NT0Ls, review the adequacy of current inspection practices, and identify outstanding technical issues. These tasks should proceed except for the review of inspection practices which is superceded by the development of the program described in item 2 above.
, hadbh_ _.
ichard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello H. Denton R. DeYoung R. Minogue T. Murl ey, R-I J. O'Reilly, R-II J. Keppler, R-III J. Collins, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa t i Enclosure to GL 85-01, Re: Fire Protection Policy
' < o .
46 i When filing a schedular exemption under 550.12, it is not always clear fromwhatspecificparagraphsof$50.48anefemptionshouldbesought. , (. Is it acceptable to request a blanket exemption from the schedular l provisions of 10 CFR 550.48 without a spacification by paragraph? s If an exemption request is submitted to meet newly published interpre- , tations of Appendix P., when does the licensee need to be in compliance? . Is the schedule prefented in Appendix R still the guideline or must a ' new schedule be d?veloped under a different criteria? ' RESPONSE > We do not intend to issue any further extensions of the 50.48(c) schedules. When a licensee determines that a 50.48(c) schedule cannot be met, the appropriate NRC Region must be notified. This policy is further explained in the generic letter transmitting this package. 8.6 Trivial Deviations , QUESTION What guidance can the NRC Staff give the industry regarding when a deviation from the literal interpretation of Appendix R is.sufficiently trivial as to not require a specific exemption? RESPONSE s The significance of a deviation must be judged as part of a fire hazards analysis. The cor.clusion of this analysis is always subject to review by the HRC inspe. tor. 8.7 Reviaed Modifications QUESTION , What is the process for' altering configurations not yet implemented for plants with Appendix R SERs? - RESPONSE , If licensees propose changes to their NRC approved modifications, they must submit their new proposal and revised schedule for imple- , mentation for NRC approval. This exchange must be justified as to.(1) the reason for the change, ' (2) the basis for the revised schedule, and-(3) the interim measures , that will be provided to assure post fire shutdown capability untii the final modifications are implemented. Whether or not enforcement action will be taken based upon continued noncompliance with Appendix R will be decideo by the NRC Regional Administrator in consultation , with NRC Headquarters. ., Enclosure to GL 85-01,'Re: Fire Protection Policy,
/ #o UNITED STATES !* o NUCLEAR REGULATORY COMMISSION
{ $ WASHINGTON, D. C 2045
% ..... / OCT 191984 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, OIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
THIRD MEETING 0F THE FIRE PROTECTION POLICY STEERING COMMITTEE, HELD ON OCTOBER 2, 1984 Sumary At its third meeting the Fire Protection Policy Steering Comittee (SC) met with the fire protection engineers from HQ and the Regional Offices to obtain their candid views on licensing and inspection fire protection inues. Coments were presented by fire protection engineers from NRR, IE, and Regions I, II, III, and IV. In addition, Jane Axelrad discussed the proposed fire protection enforcement policy with the SC. Highlights of these coments and ensuing discussions are as follows:
. The responsibility & fire protection was viewed as fragmented since CMEB, ASB, LQL. 2nd QUAB in HQ are involved in addition to Regions. Comentors believed a central point of contact is needed in HQ;
. The fire protection guidelines and scope of Tech Specs were considered by some to be inconsistent and inadequate, and a list of " minimum requirements" for fire protection was requested;
. Regional inspectors indicated need for an enforcement policy and for policy on QA for fire protection;
. Need for control room electrical review policy was indicated; and
. Coments were voiced for and against use of the " interpretations."
This meeting was very helpful to the SC in better defining the issues and clarifying where action was most needed. However, the SC indicated it would not be able to resolve or even address all of the issues raised. Of particular note was the attitude expressed by the fire protection engineers of a desire to resolve the issues promptly and of a willingness to support the recomendations of the SC. A list of attendees is provided in the enclosure. Enclosure to GL 85-01, Re: Fire Protection Policy
48 8.9 NFPA Code Deviation QUESTION Is an exemption / deviation required for deviations from NFPA Codes?
RESPONSE
Deviations from the codes should be identified and justified in the { FSAR or FHA. An exemption is not required for NFPA codes. NRC guidelines reference certain NFPA codes as guidelines to the systems acceptable to the staff, and therefore such codes may be accorded the same status as Regulatory Guides, j When the applicant / licensee states that-its design " meets the NFPA codes" or, " meets the intent of the NFPA Codes" and does not identify any deviations from such codes, NRR and the Regions expect that the design conforms to the code and the design is subject to inspection against the NFPA codes. 8.10 " ASTM E-119" Design Basis Fire QUESTION Is an exemption / deviation required, if component.s are designed to withstand an " ASTM E-119" fire? 9
RESPONSE
Some cables are being develooed for high temperature (e.g., 1700*F) applications. An exemption would be required if such cable is used in lieu of the alternatives of III.G.2 or III.G.3 in a pre-1979 plant. A deviation from the guidelines would be required for similar applications in a post 1979 plant. 8.11 Plants =1.icensed After January 1, 1979 QUESTION What fire protection guidelines and requirements apply to the plants licensed after January 1,1979?
RESPONSE
Post-1979 plants are subject to: GDC 3 10 CFR 50.48(a) and (e) Enclosure to GL 85-01, Re: Fire Protection Policy
- . OCT 191984 j Region IT also indicated that guidance for inspectors needed revising and expansion and that an inspection module was needed for NT0Ls. It was also pointed out that the fire protection inspection must be done early in the inspection phase when the licensee has the opportunity to make changes. It was suggested that Regional inspectors accompany NRR reviewers in their site visits and that a general improvement in com-munication and understanding of SER commitments was needed. Finally, Region II voiced the view that the definition of fire areas in 83-33 must be retained, that guidance is needed for suppression systems and intervening combustibles, reiterated that the inspection module needs improvement by supplying minimum acceptance criteria, and stressed the need for an enforcement policy in this area.
Region III generally endorsed the comments of Region I and II. In addition, they pointed out the need for QA guidance in the area of fire protection. They stated that deficiencies in Tech Specs resulted from omission of fire damper surveillance, and inconsistancies of Code require-ments. It was suggested that the present inspection modules be cod ined into one for all plants. Region III indicated that they felt the need to explain the rule requirements to industry; for example, 20 feet separa-tion. They requested that RES be tasked to supply the technical basis. They felt that inspectors needed such information to guide them in making judgments and evaluations. The SC pointed out that the items in the rule were based on the best information at the time and that inspectors needed not feel obligated to explain rule obligations to 1Jcencees. If there are areas where the inspector feels safety is not well served by meeting rule provisions, such concerns should be elevated to management but that the rule, including its defense indepth provisions, seemed i adequate. Finally, Region III indicated that the three things most needed were: (1) enforcement policy, (2) minimum requirements, and (3) consistent levels of inspection. To take care of (3), a training program would be needed. When asked, Region III cited the following as their three biggest frustrations: (1) the adequacy of licensee analyses, (2) the adequacy of regulatory requirements, and (3) the inconsistent reviews l and inspection criteria. l Region IV has inspected Fort Calhoun, Fort St. Vrain and some NT0Ls. They endorsed most of the conenents of the previous Regions. In particular, they felt the need for acceptance criteria, enforcement policy, and-up-to-date Tech Specs. Brookhaven National Laboratory (BNL) comented on problems with specific I compliance vice meeting the " intent" of Appendix R. In particular, NTOLs allege that they meet the intent of Appendix R through a nu d er of ( Enclosure to GL 85-01, Re: Fire Protection Policy
e . 50 No. 83-83: Use of portable radio transmitters inside nuclear power plants.
*No. 84-09: Lessons Learned From NRC Inspections of Fire Protection Safe Shutdown Systems (10 CFR 50, Appendix R)
Standard Review Plan 9.5.1, Rev. 1 Fire Protection System, dated 5/1/76 9.5.1, Rev. 2 Fire Protection Program, dated 03/78 9.5.1, Rev. 3 Fire Protection Program, July 1981. Regulations 10 CFR Part 50: Proposed fire protection program for nuclear power plants operating prior to January 1, 1979, dated May 29, 1980. Federal Register Vol. 45, No. 105, 36082. 10 CFR Part 50: Fire protection program for operating nuclear power plants, dated November 19, 1980. Federal Register Vo. 45, No. 225, 76602. 10 CFR Part 50: Fire protection rule corrections, dated September 8, 1981. Federal Register Vo. 46, No. 173, 44734. Generic Letters Note: The following documents were obtained from the Palisades file Docket No. 50-255. Similar documents should be in the fiTe for other operating facilities. The dates may vary slightly.
- 1. Letter dated 9/28/76 - Enclosing App. A to BTP APCSB 9.5-1 and supplementary guidance on information needed for fire protection program evaluation.
- 2. Letter dated 12/1/76 - Enclosing sample Technical Specifications and an errata sheet.
- 3. Letter dated 8/19/77 - Enclosing " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance."
- 4. Letter dated 6/8/78 - Re: Manpower requirements for operating reactors.
- 5. Letter dated 9/7/79 - Re: Minimum fire brigade shift size.
- 6. Letter dated 9/14/79 - Enclosing staff positions - safe shutdown capability.
Enclosure to GL 85-01, Re: Fire Protection Policy
g DCT 191984 meet Appendix R as well as deviations or exemptions, there would be fewer inspection and enforcement problems, fewer citations and a better overall fire protection image. They pointed out that, if the licensee's evaluation is kept in house and not docketed under oath, it could be inaccurate. The NRR representatives stated that the practical effect of the interpre-tations would be to relax requirements because an additional burden is placed on reviewers and inspectors that changes to licensee fixes are needed. They also asked for the agency to characterize the priority of fire protection in plant safety.
- 2. Views of IE enforcement staff.
Jane Axelrad discussed the current policy and indicated that it has not yet been issued because of a lack of general policy on what constitutes compliance with the rule. She gave background on the enforcement policy and comments on efforts to apply policy consistently across the regions. The SC indicated that it would provide its revision of the enforcement policy guidance for EDO approval. 7$4I Richard H. Vollmer, Chairman Fire Protection Poli ~cy Steering Comittee
Enclosure:
As stated cc: W. Dircks V. Stello R. Minogue T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-III R. Martin, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa SC Conrnittee
\
Enclosure to GL 85-01, Re: Fire Protection Policy
52 8.15 Staff Interpretation of Appendix R QUESTION How does the Staff initiate interpretations of Appendix R in a manner which ensures their technical adequacy and consistency with the rule's objectives (e.g., presentation to ACRS, issue for comment as in draft regulatory guides, etc.)?
RESPONSE
Staff positions are initiated when our experience shows that generic issues are identified that require clarification. These positions are reviewed for accuracy and consistency by the cognizant Division Directors. Usually, they are not issued for comment. However, Generic Letter 83-33 was commented on by the NUFPG since it was initiated, in part, at their request. 8.16 Dissemination of New Staff Positions QUESTION Will licensees be automatically sent a copy of new Staff position papers as they are developed?
RESPONSE
- The Staff positions on generic subjects are considered for issuance in Generic Letters from ONRR and Information Notices or Bulletins from OI&E. Staff positions issued for specific questions on specific plants are not given generic promulgation because they normally involve plant specific deisgn considerations. 8.17 Equivalent Alternatives QUESTION How does a licensee demonstrate that alternative measures are equiva-lent to the measures of Section III.G.2 in order to obtain an exemp-tion lacking a formal definition of the term " free of fire damage"?
RESPONSE
See Item #3 of " Interpretations of Appendix R." Enclosure to GL 85-01, Re: Fire Protection Policy
/,pa asc oq4 UNITED STATES l '
- y. ~ ,j NUCLEAR REGULATORY COMMISSION
=
- t WASHINGTON. D. C. 20$55 o
- hs . . . . . ,e' October 19, 1984 MEMORANDUM FOR: Harold R. Denten, Director, NRR Richard C. DeYoung, Director, OIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
FOURTH MEETING OF THE FIRE PROTECTION POLICY STEERING COMMITTEE Sunmary At its fourth meeting, the Fire Protection Policy Steering Committee (SC) again considered the generic letter to licensees, the enforcement policy guidance and the scope, timing and resources for the expedited fire protection inspections. The SC focussed on some of the peripheral issues which would be important to the success of the expedited inspections and other SC initiatives. At this meeting, the SC decided that:
. Prior to the fire protection inspections, a workshop would be held with IE, NRR, and Regional partici' pants in the inspections to assure understanding in the objectives, scope, and technical issues and to help provide consistency between inspection' teams;
~
. A team would be established to promptly handle disputes between licensees and the inspection teams; and
. That a standard condition should be incorporated into all licenses, requiring maintenance of the fire protection commitments but allowing change under 50.59 which do not decrease the level of fire protection with annual reporting to the Commission of such changes.
Discussion
- 1. Expedited inspections.
The SC discussed the scope, timing, and resources for the expedited inspections. There was a discussion of whether this should be a review or an inspection. The SC felt that the concept was one of an inspection rather than a review and that the availability of enforcement was im-portant to the process. The SC discussed resources and concluded that an adequate pool existed; however it was not clear if or how the expedited s Enclosure tn GL 85-01, Re: Fire Protection Policy
54 8.19.2 Individual vs. Package Exemptions QUESTION How do we submit future modification exemption requests, etc.? Would NRC prefer them individually, or developed and submitted in packages for review and approval?
RESPONSE
Future exemptions should be submitted individually, if they are independent of each other. 8.19.3 Exemption Request Supporting Detail QUESTION When an exemption request is filed, what criteria are used to determine the level of detail needed to support the request?
RESPONSE
See Enclosure 2 of NRC's letter to all licensees dated April-May 1982. 8.19.4 50.12 vs. 50.48 Exemption Requests QUESTION With regard to exemption requests for future modifications, will they be submitted under 50.12 or 50.48?
RESPONSE
10 CFR 50.12. 8.20 Post January 1, 1979 Plants and Exemption Requests QUESTION Do plants licensed after January 1,1979 which have committed to meet the requirements of Section III.G, III.J and III.0 and are required to do so as a license condition, need to request exemptions for alternative configurations?
RESPONSE
No; however, deviations from the requirements of Section III.G, III.J and III.0 should be identified and justified in the FSAR or FHA and the deviation would probably require a license amendment to change the license condition. See responses 8.1 and 8.2. Enclosure to GL 85-01, Re: Fire Protection Policy
o -
- e Harold R. Denton October 19, 1984
[ Richard C. DeYoung t cc: W.'Dircks V. Stello R. Minogue T. Murley; R-I
~J. OReilly, R-II J. Keppler, R-III R. Martin, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa SC Committee i
Q. i Enclosure to GL 85-01, Re: Fire Protection Policy
56
- 9. INSPECTION POLICY 9.1 Safety Implications QUESTION Since the Commission states that fire damage cannot be defined and fire spread cannot be predicted, how does the Commission determine which Appendix R violations have "important safety implications?"
RESPONSE
III.G.2 provides alternatives to ensure that one of the redundant trains is free of fire damage. Fire spread within one area cannot be predicted, but damage is limited to ona fire area. Determination of the Appendix R violations that have "important safety implications" are based on the equipment, components, and systems that are located in the same fire area that are needed for safe shut-down or can adversely affect safe shutdown, and are not protected by the features of III.G.2, III.G.3 or an approved alternative. . 9.2 Uniform Enforcement QUESTION How does the Commission ensure that violations of the rule are uniformly treated between regions? ,
RESPONSE
Each Region evaluates violations in accordance with the NRC Enforcement Policy, 10 CFR 2, Appendix C. The Policy provides guidance for the determination of appropriate enforcement sanctions for violations. The Office of Inspection and Enforcement provides guidance for and monitors Regional implementation of the Policy to ensure a uniform application. In addition, the policy requires that all escalated
- enforcement actions be approved by the Director of the Office of Inspection and Enforcement.
! 9.3 NTOL Inspections QUESTION Will NTOLs be subject to an Appendix R audit now being performed on plants licensed to operate prior to January 1,1979? Or, will the current review and analysis being performed by the Staff be satisfactory? Enclosure to GL 85-01, Re: Fire Protection Policy
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(, **"* / OCT 191984 MEMORANDUM FOR: Fire Protection Policy Steering Committee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee SUBJECT : AGENDA FOR SIXTH MEETING The sixth meeting of the Fire Protection Policy Steering Comittee will be held at 9:00 a.m. on October 22, 1984 in the Region III offices. At this meeting you should be prepared to discuss and finalize versions of:
. Enforcement Policy
. Generic Letter and Interpretations
. Technical Issues Package
. Standard License Condition
. Inspection Module In addition, we need to prepare our final report to the E00. I will FAX an outline to you for your consideration today. In addition, there are a number of issues and suggestions still left hanging. For example:
the central point of contact for fire protection issue; the status of NFPA codes; what to do about Tech Specs; and the format of the workshop in advance of the expedited inspections. hdh - ichard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: See next page. l l 1 (, ' Enclosure to GL 85-01, Re: Fire Protection Policy
58 _1
RESPONSE
The Appendix R inspections are conducted on a sample basis. These inspections do not certify that all possible items of noncompliance with Appendix R have been identified. The inspection results do provided a basis for a determination of the adequacy of a licensee's Appendix R reanalysis, modification and preparation. When a noncom-pliance with Appendix R requirements is identified, a notice of viola-tion will be issued to ensure adequate corrective action. In ' those cases in which the licensee believes that the staff has invoked a reinterpretation of adequacy in areas which had previously been reviewed, NRC's procedures for appeal would be applicable. 9.7 NRC List of Conforming Items QUESTION At the end of the audit, will the NRC provide a list of items that had been reviewed and found in conformance with Appendix R? To date, only areas of nonconformance have been specifically identified in exit interviews.
RESPONSE
Subsequent to an Appendix R inspection, the NRC will not provide a . list of items reviewed and found to be in conformance with Appendix R. We do list the areas inspected and where non-compliances were not found. 9.8 Inspection Re-review QUESTION Where assumptions are made and clearly stated within the analysis submitted to NRR for review, will such assumptions be subject to a second review by OI&E during the inspection process? Where assumptions are made in conjunction with the analysis, should i exemption requests be filed just to provide protection for the licensee? If NRR accepts a licensee's selection of equipment and shutdown paths as being sufficient to meet the Appendix R shutdown criteria, will Ol&E review and have the right to challenge the approved shutdown paths and approved equipment selection? Or will they only check the shutdown paths and equipment in question to see that they meet the Appendix R requirements, i.e., separation? I Enclosure to GL 85-01, Re: Fire Protection Policy I
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' UNITED STATES
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% C WASHINGTON D. C. 20555 s J g .. f
..... OCT 2 6 084 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, OIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
FIFTH AND SIXTH MEETINGS OF THE FIRE PROTECTION POLICY STEERING COP 911TTEE At the fifth and sixth meetings of the Fire Protection Policy Steering Consnittee (SC), held in Bethesda on October 17 and the Region III offices on October 22, respectively, the final version of (1) enforcement policy guid-ance, (2) Generic Letter, (3) standard license condition, (4) temporary instruction for fire protection inspections, and (5) technical issues package of questions and answers were discussed, edited, and put into final form. No new initiatives were discussed but the impact and consistency of all initiatives developed by the SC were reviewe h The SC also assured that all issues included in the ED0 memo of September 13 had been fully addressed and that all issues raised to the SC's attention by other parties had been fully considered. gr-tbt w
' Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: See next page.
1 Enclosure to GL 85-01, Re: Fire Protection Policy
Enclosure to GL 85-01, Re: Fire Protection Policy
m a,,
#c UNITED STATES
{[o 7, ( NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE 7 ("CRf SEP 211384 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, IE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
FIRST MEETING OF FIRE PROTECTION POLICY STEERING COMMITTEE Sumary At its first meeting, the Fire Protection Policy Steering Committee (SC) discussed the issues identified in the EDO memo of September 13, 1984. The SC made the following recommenda' ions and assignments:
. the SC recommended that Appendix R implementation policy follow Me " Interpretations" discussed at the Regional Workshops, as modified by details discussed below, rather than requiring prior i .
staff review or exemptions for deviations from Generic Letter 83-33;
. the SC recommended that the issuance of the fire protection enforcement policy and enforcement actions be expedited;
. the SC recommended that QA for fire protection be clearly defined as that required by GDC-1; and
. the SC assigned to the Working Group' tasks dealing with the adequacy of available guidance, comparison of fire protection requirements for ors and NT0Ls, the adequacy of current inspection practices, and identification of outstanding technical issues.
Introduction As a result of the E00 memo of September 13, 1984, regarding a review of NRC fire protection policy and programs, the SC held its first meeting
- on September 13 and 14. The objectives of t.is meeting were to:
- Attended by: I. Martin, J. Olshinski, L. Spessard, N. Grace, F. Rosa, S. Richardson, W. Shields, W. Little, T. Wambach, W. Olmstead, K. Cyr l and R. Vollmer.
( Enclosure to GL 85-01, Re: Fire Protection Policy ! l
. discuss the background leading to the formation of the SC;
. establish a general charter, objectives, schedule, and working arrangements; and
. discuss issues currently in hand and make decisions for resolution if appropriate.
The discussion of background included the events leading up to and includ-ing the fire protection regional workshops, the May 30, 1984 meeting with the Commission, the August 27 meeting with the EDO, and the events surrounding the OP0 signeJ by three fire protection reviewers and two inspectors. As part of tn's background, ELD representatives discussed the Rule and the distinction between its legally enforceable requirements and staff guidance issued subsequent to the Rule. The background dis-cussion also focussed on the issues identified in the EDO memo of Sept mber 13. - Charter and Schedule The chartc of 'le SC is the review of NRC fire protection policy and prograrn leading to policy recommendations which would expedite compliance with Appendix R at older plants and assure consistent levels of fire
}
protection safety at all plants. To implement this Charter the SC agreed / that current licensing, inspection, legal, and technical issues needed to be examined. The SC's cbjective would be to make specific recomenda-tions to the E00 which could be carried out (1) within the existing framework of 50.48 and Accendix R, and (2) without making a disruption in the effort already unterway to imolement fire protection requirements. The SC would attempt to make recommendations th?- could be immediately effective. However, we recognize that there may be some instances in which further study was needed. In such cases, we agreed to recommend a specific assignme t and end date for such study. Finally, the SC agreed that its work wou~.d be completed through issuance of its report to the EDC by October 26, 1984. Working Arrangements The SC discussed how it could use the available resources most effectively. It was decided that the SC would not need to meet at this time with HQ and Regional seople since their views have been expressed extensively in written and transcribed material. The SC did not feel the need to meet with industry for the same reason. However, the SC does wish to meet with Vic Stello and hereby offers the opportunity of a meeting with the recipients of this memo. Enclosure to GL 85-01, Re: Fire Protection Policy
o
- The SC will take advantage of the Working Group chaired by Faust Rosa and make assignments to that group and to certain individuals. It was our Jelief that use of currently available resources, including the Working Group, would be sufficient to achieve our objectives and schedule.
Issues
- 1. Adequacy of current guidance to industry.
The SC felt that enough guidance has been generated but that it needed to be made consistent with our recommendations and cleared up technically in some areas. The mechanism for this would be a generic letter, super-ceding previous guidance, to be sent to all licensees as promised at the rti onal i workshaps. This generic letter would be sent to the Comission, as requested at the Comission's May 30 meeting, before being issued to industry. The SC agreed that it would be best to utilize guidance already available, to the extent possible, to minimize possible confusion both within ind,stry and the NRC. It was also agreed that we should utilize either Generic Letter 83-33 or the " Interpretations" rather than a third option for dealing with the exemption issue. This is discussed below. The Working Group was tasked with reviewing all current guidance and outstancing technical questions and to revise the Regional Workshop package to incorporate in one place a comprehensive set of guidance that is consistent with the SC's policy recommendations and the approved technical recommendations of the Working Group.
- 2. Interpretation of the Appendix R requirements vice staff guidance.
The basic issue is whether industry can deviate from the contents of Generic Letter 83-33 without prior staff review and approval. This issue is fully developed by the current Staff DPO. The SC had the benefit of ELD's views and Faust Rosa's recommendations resulting from his assign-ment by Mr. Denton to make an independent assessment of this OPO. The SC decided that it could only support the contents of 83-33 as guidance to industry, consistent with the " Interpretations" drafted for the Regional Workshops since ELO advised that Generic Letter 83-33 stated requirements which went beyond the terms of Appendix R itself. However, the SC also felt that some of the clarifying language contained in the DP0 should be utilized and that specific guidance should be supplied to clearly indicate the level of fire protection to be achieved and the documentation necessary to demonstrate it. In addition, the SC felt that, at an appropriate time prior to the Appendix R inspection, the licensee should be requested to provide information necessary to ( Enclosure to GL 85-01, Re: Fire Protection Policy
support the inspection for HQ and regional review. In addition, the licensees should be encouraged to meet with the staff to discuss plans before extensive hardware modifications are initiated. The Working Group was tasked to work on language to support this recom-mendation as part of its work on item 1.
- 3. Treatment of expected future technical and schedular exemptions into late 1980s and early 1990s. ,
This item was discussed extensively and tabled when no clear direction was apparent. The concern is that some licensees may unduly request staff reconsideration of technical findings and/or implementation schedules which would in effect defer compliance. This item will receive priority consideration at the next meeting and no assignments were made. However, the SC felt strongly that the fire protection enforcement policy and current enfcrcement packages consistent with that policy should be issued promptly to demonstrate NRC resolve in this area and that the backlog in NRR be processed expeditiously. 4 Comparison of Appendix R and current NT0L plants for fire protection safety. Based on statements made at the May 30 Commission meeting and correspond- 1' ence from the Regions, there may be differences in the licensing evaluation . and inspection practices for Appendix R plants and NT0Ls. The Working ~ Group was tasked with investigating if such differences do exist and how the goal of consistent levels of fire protection safety at all plants might be achieved. in pursuing this, it was suggested that the Working Group meet with representatives from HQ and the Regions active in fire i protection reviews and inspecticas. It was also acknowledged that the existing Temporary Instruction for Appe dix R safe shutdown inspections must be revised to be consistent with the new interpretations. The i Workino Group must also verify that the guidance documents referenced in j the Temporary Instruction for use by the inspectors ha"e been officially i sent to all licensees and applicants. l l 5. Adequacy of current inspection practices. The SC was informed that consistent inspection practices and schedules have been set up for Appendix R plants but that the NTOLs have been , hand'ed on a somewhat ad hoc basis. Some regions have instituted a l fairly extensive review of the capability to achieve a safe snutdown i following a fire and others look more at SER-specific hardware items. ! Steve Richardson of IE was tasked with reporting to the SC, at the next neetina on current inspection content and schedules for all classes of plants, suggestions for modifying current practices, and, specifically, when the NT0L's thould be inspected to assure effective compliance with NRC's fire prctection requirements. 1 Enclosure to GL 85-01, Re: Fire Protection Policy l
'6. Identification and resolution plan for outstanding technical issues.
The Working Group was requested to identify any outstanding technical issues that impede Appendix R compliance. These will be discussed with the SC and a plan and schedule for resolution will be developed. One issue that was discussed is the QA standard to be applied to fire protection compliance. The SC felt strongly that the requirements of GDC-1 apply to fire protection features, recognizing that the activities are underway to resolve the GDC-1/ Appendix B and safety-related/ important-to-safety issues. Without trying to muddy the water in these areas, the SC felt that, at minimum, licensees be made aware that the fire protection program f alls under GDC-1. The Working Group should look into thic.also and review the QA commitments made by some licensees in their pre-Appendix R SERs. The next meeting of the Fire Protection Steering Committee will be held on September 27 at 9:00 a.m. in P-202A. d Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello R. Minogue T. Murley J. O'Reilly J. Martin E. Case J. Taylor D. Eisenhut R. Bernero G.-Arlotto F. Rosa (10) SC Members 4 v Enclosure to GL 85-01, Re: Fire Protection Policy
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/ SEP 2 61984 MEMORANDUM FOR: Fire Protection Policy Steering Committee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
AGENDA FOR SECOND MEETING The second meeting of the Fire Protection Policy Steering Committee will be held at 9:00 a.m. on September 27, 1984 in Room P-202 A. At this meeting we will be briefed on the Working Group activities and on current inspection content and schedules and have the benefit of the IE and NRR Office Director's views on our activities. Our tentative schedule is as follows: 9:00 a.m. Executive Session 9:30 Harold Denton and Ed Case 10:00 Dick DeYoung and Jim Taylor 10:30 Faust Rosa will discuss Working Group activities 1:00 p.m. Steve Richardson will discuss Appendix R inspections 2:00 Committee work In addition to our consideration of issues raised by the above agenda items, we need to consider policy approaches to deal with technical and schedular exemptions which put implementation of Appendix R into the distant future. Any approaches you have to deal with this issue will be welcomed. kr _ _ Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello H. Denton R. DeYoung T. Murley R. Minogue J. O'Reilly J. Keppler J. Collins J. Martin G. Cunningham E. Case J. Taylor
- 0. Eisenhut i R. Bernero G. Arlotto .
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%.....* SEP 2 71984 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, OIE Thomas E. Murley, Regional Administrator, R-I James P. O'Reilly, Regional Administrator, R-II James G. Keppler, Regional Administrator, R-III John T. Collins, Regional Administrator, R-IV John B. Martin, Regional Administrator, R-V FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
NRC FIRE PROTECTION MEETING The Fire Protection Policy Steering Comittee requests that the fire protection engineers in your organization attend a meeting in Bethesda on October 2, 1984, to give their views on licensing and inspection fire protection issues. This is an information gathering session to help the Comittee formulate its policy recomendations. A candid discussion of how these issues are viewed by the fire protection reviewers and inspectors, and reconrnendations they may have, would be welcome. A meeting agenda is enclosed. d$rv-ichard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
Enclosure:
Agenda cc: W. Dircks V. Stello R. Minogue G. Cunningham E. Case J. Taylor, IE D. Eisenhut R. Bernero G. Arlotto, RES F. Rosa N. Grace, IE L. Spessard, R-III W. Olmstead, ELD J. Olshinski, R-II T. Wambach Enclosure to GL 85-01, Re: Fire Protection Policy
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i Enclosure to GL 85-01, Re: Fire Protection Policy
t l NRC Fire Protection Meeting Date: October 2,1984 at Location: P-422 Phillips Building
, 9:00 a.m. Opening Remarks - R. Vollmer 9:10 Region I Remarks 9:40 Region II Remarks 10:10 Region III Remarks 10:40 Region IV Remarks 11:40 Brookhaven Remarks 1:00 p.m. NRR Remarks 2:00 I&E Remarks 2:30 to 4:00 Group Discussion
- 4:00 to 5:00 Each group given opportunity :o provide any additiona,1 comments.
5:00 Adjourn l' i Enclosure to GL 85-01. Re: Fire Protection Policy
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%, . . . . . st MEMORANOUM FOR: Faust Rosa, Chairman Fire Protection Working Group FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
ASSIGNMENTS FOR FIRE PROTECTION WORKING GROUP At the first meeting of the Fire Protection Policy Steering Comittee, the Committee assigned action tasks to the Working Group. The purpose of this memo is to provide better focus on the Working Group assignments as a result of the second and third Comittee meetings. The Working Group's first priority should be the revision of the enclosure to the Generic Letter discussed at the fire protection workshops. This package should include answers to the questions posed by industry and revision of the
" interpretations" section to utilize clarifying aspects emphasize the level of fire protection expected by stafi'gf the OP0e.g.,
guidance, and to Generic Letter 83-33, and the documentation necessary to demonstrate equiva-lence to staff guidance. This package should not contain any new guidance and should be internal-ly consistent and consistent with the Comittee's proposed policy on the " interpretations." To the extent possible criteria acceptable to the staff should be clearly identified. This package should be available for Comittee review by October 12 and whatever efforts are necessary to meet this date should be expended. Bill Shields will prepare a draf t of the Generic Letter in which he will lay out the elements for expediting Appendix R implementation. It will discuss the proposed policy on schedular exemptions, the proposed augnentation of plant inspections, and indicate generally what is expected of licensees and what enforcement action may occur. In addition, it was agreed that the 50.48 schedule expiration for each p.lant would be indicated. The item of next priority is the development of an inspection plan which would accomplish the objectives of a " helpful" inspection. That is, in a one week, four-man inspection, we need to find out how the licensee is approaching his Appendix R implementation, the status of the plant with respect to compliance, and plans and schedules for complete implementation. This inspection should identify improper paths being taken by the licensee, and allow the staff to assess current fire protection safety and need for any plant specific enforcement action. Finally, the Working Group should compare fire protection requirements for ors and NTOLs and work on any outstanding technical issues. The outline Enclosure to GL 85-01, Re: Fire Protection Policy
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you provided of the Working Group's program review appears far too ambitious ' in view of current time and resource constraints. We should discuss this further at our October 10 meeting in Atlanta. - x s. . S Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee i cc: Steering Comittee ' W. Dircks; , 1 H. Denton ' R. DeYoung ' E. Case J. Taylor
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OCT 4 1984 MEMORANDUM FOR: Fire Protection Policy Steering Comittee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
AGENDA FOR THE FOURTH MEETING The fourth meeting of the Fire Protection Policy Steering comittee will be held at 9:00 a.m. on October 10, 1984 in the Region II offices in Atlanta. At this meeting we need to reach agreement on the enforcement policy guidance and develop our decision, reached at the second meeting and fortified in the third meeting, on a FY 85 inspection for most plant types. I request that John prepare a rewrite of the enforcement policy and Nelson a general plan for inspections for Comittee consideration. Other items we need to discuss are:
. Focus of inspections relative to fire protection and safe shutdown;
. Working group assignments;
. What should be included in technical specifications;
. Resolution of licensee / inspection team disputes; and
. Disputes of issues raised at third meeting.
I hope at this meeting we can agree on our overall approach to make our recomendations to Dircks coherent and consistent so that we can begin writing to have a draft report ready at our fifth meeting. Richard H. Vollmer, Chair an Fire Protection Policy teering Comittee cc: See next page Enclosure to GL 85-01, Re: Fire Protection Policy
l . . O e s 2 , e cc: W. Dircks
. V. Stello
- H. Denton R. DeYoung R. Minogue T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-!!!
J. Collins, R-IV J. Martin, R-V $ G. Cunningham E. Case J. Taylor
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October 12, 1984 MEMORANDUM FOR: Fire Protection Policy Steering Comittee FROM: Richard H. Vollmer, Chainnan Fire Protection Policy Steering Comittee
SUBJECT:
AGENDA FOR THE FIFTH MEETING The fifth meeting of the Fire Protection Policy Steering Comittee will be held at 9:00 a.m. on October 17, 1984 in Room P-202A, At this meeting you should be prepared to discuss the generic letter and its attachments relative to the ability of these documents to satisfy Steering Comittee decisions. We should also be prepared to settle on the enforcement policy guidance and the inspection module revision if these are available. Finally we will need to discuss format, content, and writing assignments for our October 26. report to the EDO.
& bc ichard H. Vollmer, Director Fire Protection Policy Steering Comittee cc: W. Dircks
- v. Stello H. Denton R. DeYoung R. Minogue ,
T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-III J. Collins, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa i Enclosure to GL 85-01, Re: Fire Protection Policy
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October 12, 1984 MEMORANDUM FOR: Harold R. Denton, Director, NRR Richard C. DeYoung, Director, OIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
SECOND MEETING OF THE FIRE PROTECTION POLICY STEERING COMMITTEE HELD ON SEPTEMBER 27, 1984 Sumary At its second meeting, the Fire Protection Steering Committee (SC) con-sidered policy approaches to deal with technical and schedular exemptions which threaten to put Appendix R implementation into the distant future, discussed SC activities with the IE and NRR Office Directors and Deputies, and was briefed on Working Group activities. As a result of this meeting, the SC made the following decisions:
. the SC decided to bold a meeting with all HQ and Regional fire protection engineers, as a body, to candidly discuss their views on fire protection issues, problems, and possible future actions;
. the SC decided that the most promising way to expedite Appendix R compliance is to initiate an aggressive inspection program which would steer and promote licensee compliance, assess the degree of fire safety, and exercise enforcement policy where appropriate; and
. the SC decided that no further schedular exemptions should be granted.
Discussion
- 1. Meeting with Office Directors and Deputies.
The SC met with Messrs. Denton, DeYoung, Case and Taylor to discuss their views on the fire protection problems, potential solutions, and SC ; activities. They viewed the problem as a lack of staff cohesiveness in a ' highly judgemental area exacerbated by industry reluctance to meet Appendix R requirements. They felt that a meeting between the HQ and Regional fire protection staff to air any problems, issues, and possible solutions, would benefit the SC's work. We agreed and set up such a meeting for October 2nd. Enclosure to GL 85-01, Re: Fire Protection Policy
October 12, 1984 ~ It was also noted that, although industry was less than enthusiastic about fire protection, many problems have been solved and there have been signif-icant improvements in fire protection of plants. To assist in defining fire protection problems and assessing safety significance, it was also suggested that we sort out fire protection and safe shutdown issues. The inclusion of fire protection features in Tech Specs was also discussed; namely, that they are not currently consistent, that augmenting of Tech Specs in this area has been proposed, and that such activities must be considered relative to general goals of simplifying Tech Specs. Finally, it was suggested that we need to better integrate the disciplines involved in this licensing and inspection area and identify a central point of responsibility.
- 2. Exemptions.
The SC discussed what, if anything, could be done to keep schedular and technical exemptions from dragging Appendix R implementation into the next decade. ELD stated that there was little we could do on technical exemptions since, if the utility has a valid reason for requesting one, then NRC must review it. On schedular exemptions, however, we can make a policy decision not to grant any more. Such a decision would have a legitimate basis since the Commission's Appendix R record viewed implementation in four or five years. Further, many 50.48 schedules have or are near running out. The SC decided that no further schedular exemptions should be granted. The discussion turned to means of assuring that licensees recognized and could implement their responsibility for Appendix R implementation. The SC felt that the staff was taking on too much of the burden and that a well defined set of technical criteria, coupled with a program of inspection, and an enforcement policy would provide the best incentives. The SC felt that about five teams each consisting of a team leader and a fire protection, electrical, and systems engineer, should be set up. Each team, beginning in February 85, should go to one plant per month for a one-week inspection. These inspections would target plants of each type, each utility, and each A/E including NT0Ls. Recognizing that plants in varying degrees of com-pliance would be inspected, the inspections should focus on safe shutdown. Where Appendix R implementation is still being engineered, the team should steer and promote licensee compliance in a technically supportable way. These inspections would also establish where each plant stands vis a vis Appendix R, and would use enforcement action where appropriate in a prompt fashion. It was suggested that there should be central and prompt reso-lution of any licensee / team disputes. This could be handled by a team consisting of a member of NRR, IE and the appropriate Regional management. The decision of this team would state the NRC position followed by a confir-matory letter or order as appropriate. Following such an inspection program, decisions on the long term fire protection inspections, both in extent and timing, would follow. Enclosure to GL 85-01, Re: Fire Protection Policy
a O October 12, 1984
- 3. Technical issues.
Two technical issues were identified to be pursued by the Working Group: (1) the Dow Corning ascertion that penetration material was not being properly installed; and (2) whether our criteria on control room fires are appropriate and consistent.
- 4. Working Group discussion.
Faust Rosa discussed the Working Group activities, in particular, the out-line of the Working Group report. All items of interest to the SC were covered in the outline of this report. However, it appeared that com-pletion of the work as outlined was too ambitious for the time and resources provided. The SC indicated that priority attention be given to the completion of the generic letter package. The SC also indicated in its first meeting that the Working Group should compare fire protection requirements for ors and NTOLs, review the adequacy of current inspection practices, and identify outstanding technical issues. These tasks should proceed except for the review of inspection practices which is superceded by the development of the program described in item 2 above. v t/b ^ - , ichard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: W. Dircks V. Stello H. Denton R. DeYoung R. Minogue T. Murl ey, R-I J. O'Reilly, R-II J. Keppler, R-III J. Collins, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa
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**.....* OCT 191984 l MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, 0IE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
THIRD MEETING OF THE FIRE PROTECTION POLICY STEERING COMMITTEE, HELD ON OCTOBER 2, 1984 Sumary At its third meeting the Fire Protection Policy Steering Committee (SC) met with the fire protection engineers from HQ and the Regional Offices to obtain their candid views on licensing and inspection fire protection issues. Comments were presented by fire protection engineers from NRR, IE, and Regions I, II, III, and IV. In addition, Jane Axelrad discussed the proposed fire protection enforcement policy with the SC. Highlights of these coments and ensuing discussions are as follows:
. The responsibility for fire protection was viewed as fragmented since CMEB, ASB, LQB, and QUAB in HQ are involved in addition to Regions. Comentors believed a central point of contact is needed in HQ;
. The fire protection guidelines and scope of Tech Specs were considered by some to be inconsistent and inadequate, and a list of " minimum requirements" for fire protection was requested;
. Regional inspectors indicated need for an enforcement policy and for policy on QA for fire protection;
. Need for control room electrical review policy was indicated; and
. Coments were voiced for and against use of the " interpretations."
This meeting was very helpful to the SC in better defining the issues and clarifying where action was most needed. However, the SC indicated it would not be able to resolve or even address all of the issues raised. Of particular note was the attitude expressed by the fire protection engineers of a desire to resolve the issues promptly and of a willingness to support the recomendations of the SC. A list of attendees is provided in the enclosure. Enclosure to GL 85-01, Re: Fire Protection Policy
OCT 191984 The SC believes that its previous decision on expediting plant inspections was reinforced by the comments in this meeting. The SC indicated its view that inspections should go forward rapidly to get on with the identi-fication and resolution of problems. To require more documentation in areas not specifically required by the Rule would slow compliance down. Discussion
- 1. Views expressed by fire protection engineers.
Region I discussed the inspections of Vermont Yankee, Salem and Calvert Cliffs. It was indicated that confusion generated was caused by differ-ences between Generic Letter 83-33 and " interpretations" but that the Region endorsed the interpretations because they would expedite the process and not create inspection problems. The implementation of Appendix R at Calvert Cliffs was successful because the licensee did a very thorough evaluation of his alternate shutdown needs, had substantial communication with the licensing staff, had some unit-specific features which benefitted shutdown, and had support of licensee management. Other comments made by Region I were that: (a) Vermont Yankee was con-fused by the Appendix R implementation letter, (b) at Salem the inspection was complicated because many exemptions were needed just prior to the inspection as a result of 83-33, and (c) all Region I licensees appear to be taking Appendix R seriously and making good faith efforts. Region II discussed their experience with Appendix A and Appendix R. They saw Appendix R and fire protection as a moving target in particular since different utilities take different approaches and when these are accepted in licensing, confusion in the inspection process results. They also noted that utilities were concerned that NRC was going beyond reactor safety and getting into loss prevention. Region Il raised an l issue, generally endorsed by others in the meeting, that the responsi-bility for fire protection is fragmented because of all the disciplines
- responsible. In particular, in NRR responsibility lies in engineering, systems interaction and human factors safety. Licensing work also resides in the Quality Assurance Branch in IE. This, along with differ-ent Regional views and inspector approaches, results in confusion and inconsistency. It was suggested that a central contact was needed at HQ to provide central authority for fire protection. It was also indicated that reviewers and inspectors need additional guidelines, in particular, minimum acceptance criteria. Region II also pointed out differences in license requirements and Tech Specs dealing with fire protection. A discussion evolved concerning the need for augmenting Tech Specs in relation to other safety significant items. The general concensus of the fire protection engineers was that Tech Specs needed to be expanded in this area.
Enclosure to GL 85-01, Re: Fire Protection Policy
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Region II' also indicated that guidance for inspectors needed revising and expansion and that an inspection module was needed for NT0Ls. It was also pointed out that the fire protection inspection must be done early in the inspection phase when the licensee has the opportunity to make changes. It was suggested that Regional inspectors accompany NRR reviewers in their site visits and that a general improvement in com-munication and understanding of SER commitments was needed. Finally, Region II voiced the view that the definition of fire areas in 83-33 must be retained, that guidance is needed for suppression systems and intervening combustibles, reiterated that the inspection module needs improvement by supplying minimum acceptance criteria, and stressed the need for an enforcement policy in this area. Region III generally endorsed the coments of Region I and II. In addition, they pointed out the need for QA guidance in the area of fire protection. They stated that deficiencies in Tech Specs resulted from omission of fire damper surveillance, and inconsistancies of Code require-ments. It u s suggested that the present inspection modules be codined into one for all plants. Region III indicated that they felt the need to explain the rule requirements to industry; for example, 20 feet separa-tion. They requested that RES be tasked to supply the technical basis. They felt that inspectors needed such information to guide them in making judgments and evaluations. The SC pointed out that the items in the rule were based on the best information at the time and that inspectors needed not feel obligated to explain rule obligations to 1jcencees. If there are areas where the inspector feels safety is not well served by meeting rule provisions, such concerns should be elevated to management but that the rule, including its defense indepth provisions, seemed adequate. Finally, Region III indicated that the three things most needed were: (1) enforcement policy, (2) minimum requirements, and (3) consistent levels of inspection. To take care of (3), a training program would be needed. When asked, Region III cited the following as their three biggest frustrations: (1) the adequacy of licensee analyses, (2) the adequacy of regulatory requirements, and (3) the inconsistent reviews and inspection criteria. Region IV has inspected Fort Calhoun, Fort St. Vrain and some NT0Ls. They endorsed most of the coments of the previous Regions. In particular, they felt the need for acceptance criteria, enforcement policy, and-up-to-date Tech Specs. Brookhaven National Laboratory (BNL) comented on problems with specific compliance vice meeting the " intent" of Appendix R. In particular, NT0Ls allege that they meet the intent of Appendix R through a nuder of Enclosure to GL 85-01, Re: Fire Protection Policy
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OCT 191984 different ways. BNL indicated that there were serious problems with consistency and interpretation of control room fires and that we lack the rationale or basis for these views. They question in particular how long is the control room habitable, what action can be taken, where two units share a control room are both units affected, and must both units shut down outside the contrcl room. BNL also stated that we needed specific guidelines for associated circuit analysis and indicated that the SER was not always a reliable indicator of licensing commitments for inspection. A representative of ASB indicated the scope of review for alternate shutdowns and that the criteria used were consistent and had been in use for most plant reviews. The criteria were not well documented however. He expanded on the systems used for safe shutdown, the requirements for physical separation and electrical separation for safe shutdown. With respect to the associated circuits analysis it was indicated that the evaluation assured, assuming offsite power loss, that safety could be demonstrated assuming one spurious signal, a loss of all automatic signals, and spurious operation of motor-operated valves in the high/ low pressure interface. It was indicated that this included review of licensee's sununary of operator actions and that, during inspection, the actual procedures are walked down. IE's discussion focused on item 3 of the " interpretations" which states that licensees must show equipment must be " free of fire damage" before, during and after a fire. He was concerned that although Section III.G.2 specifies free of fire damage, the interpretation would allow less than this, in particular, scorched and severely heated equipment which are still barely sufficient to perform their intended functions. He said that the rule language would not allow this and that it is not appro-priate and conservative. In many of the above comments from Region and HQ representatives the SC 2 detected a belief of bad faith by the licensees and practices which would subvert the spirit and the technical intent of the Commission's requirements. The SC pursued this to some extent but noted that there seemed to be a lack of specifics. Since the importance and safety significance of each requirement was somewhat judgmental the SC felt that the NRC needed to shoulder some responsibility for lack of com-pliance because of the evolution of Appendix R. NRR representatives indicated their belief that the NRC should stick with the Generic Letter 83-33 approach, which in their view has been working, and issue enforcement policy. They felt that if the licensees were required to submit for review their entire program, both how they Enclosure to GL 85-01, Re: Fire Protection Pnlicy
e 007 1 9 1984 meet Appendix R as well as deviations or exemptions, there would be fewer inspection and enforcement problems, fewer citations and a better overall fire protection image. They pointed out that, if the licensee's evaluation is kept in house and not docketed under oath, it could be inaccurate. ' The NRR representatives stated that the practical effect of the interpre-tations would be to relax requirements because an additional burden is placed on reviewers and inspectors that changes to licensee fixes are needed. They also asked for the agency to characterize the priority of fire protection in plant safety.
- 2. Views of IE enforcement staff.
Jane Axelrad discussed the current policy and indicated that it has not yet been issued because of a lack of general policy on what constitutes compliance with the rule. She gave background on the enforcement policy and comments on efforts to apply policy consistently across the regions. The SC indicated that it would provide its revision of the enforcement policy guidance for EDO approval. Richard H. Vollmer, Chairman Fire Protection Policy Steering Cannittee
Enclosure:
As stated cc: W. Dircks V. Stello R. Minogue T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-III R. Martin, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa SC Committee
\w-Enclosure to GL 85-01, Re: Fire Protection Policy
l 4 i l l l l l Enclosure to GL 85-01, Re: Fire Protection Po11cy
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October 19, 1984 MEMORANDUM FOR: Harold R. Denton, Director, NRR Richard C. DeYoung, Director, 0IE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Comittee
SUBJECT:
FOURTH MEETING OF THE FIRE PROTECTION POLICY STEERING COMMITTEE Sumary At its fourth meeting, the Fire Protection Policy Steering Committee (SC) again considered the generic letter to licensees, the enforcement policy guidance and the scope, timing and resources for the expedited fire protection inspections. The SC focussed on some of the peripheral issues which would be important to the success of the expedited inspections and other SC initiatives. At this meeting, the SC decided that:
. Prior to the fire protection inspections, a workshop would be held with IE, NRR, and Regional particibants in the inspections to assure understanding in the objectives, scope, and technical issues and to help provide consistency between inspection teams;
. A team would be established to promptl'y handle disputes between licensees and the inspection teams; and
. That a standard condition should be incorporated into all licenses,
, requiring maintenance of the fire protection commitments but allowing change under 50.59 which do not decrease the level of fire protection with annual reporting to the Commission of such changes. Discussion
- 1. Expedited inspections.
The SC discussed the scope, timing, and resources for the expedited inspections. There was a discussion of whether this should be a review or an inspection. The SC felt that the concept was one of an inspection rather than a review and that the availability of enforcement was im-portant to the process. The SC discussed resources and concluded that an adequate pool existed; however it was not clear if or how the expedited m i N .d Enclosure to GL 85-01, Re: Fire Protection Policy
October 19, 1984 Harold R. Denton Richard C. DeYoung inspections would affect other programs. The Working Group /IE was asked to complete a module to be used in these inspections and to plan out which licensees would be inspected to best achieve the objectives. A one-year schedule of plants and resources was requested by the SC. The SC also discussed how to assure a common understanding of the objectives, scope, and technical issues by all participants in these inspections. To be effective, such inspections need to be uniform. The SC decided that a workshop would be held prior to initiation of the inspections wherein the SC would lay out the philosophy and intent, and technical and procedural issues would be discussed. The SC also decided that it would be important to have a central authority available to promptly and uniformly resolve disputes between the licensees and inspection teams. This was felt to be necessary so that major issues would not be left unsettled and 50 that 'icensees would not pursue fire protection solutions that would be unlikely to be acceptable to NRC. The Chairman of the SC was assigned to draft a charter and membership for this authority for SC's consideration. It was also agreed that the resolution of disputes by this authority would be promptly published to all inspection teams.
- 2. Generic Letter.
The need for a standard fire protection license tandition was discussed. Problems resulting from inconsistent license conditions have been raised during the past year. It was agreed that a standard condition, similar to that drafted by ELD and NRR for current NT0Ls, should be prepared. The SC discussed the tim ing for reporting of fire protection chan'ges conducted under 50.59 and it was decided that annual reporting was appro-priate based on other important 50.59 issues that utilize annual reporting. A ciscussion ensued on how to get such a license condition applied across the board to all licensees. The use of a Generic Letter or a 50.54(f) letter was discussed. ELD was tasked with developing a recommendation.
- 3. Enforcement Policy.
The SC discussed in detail a proposed fire protection enforcement policy. Some issues were settled and it was decided that the SC would complete this activity at its next meeting. Finally, the SC decided to hold its next meeting on October 17 in Bethesda and to hold its final session in Chicago, for two days or more if needed, to complete activities and prepare a report to the EDO.
,- N Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: See next page Enclosure to GL 85-01, Re: Fire Protection Policy
4 o Harold R. Denton October 19, 1984 Richard C. DeYoung cc: W. Dircks V. Stello R. Minogue T. Murley', R-I J. 0Reilly, R-II J. Keppler, R-III R. Martin, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor D. Eisenhut R. Bernero G. Arlotto F. Rosa
-SC Comittee
\
Enclosure to GL 85-01, Re: Fire Protection Policy
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i Enclosure to GL 85-01, Re: Fire Protection Policy
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; WASHINGTON. D. C. 20S55 j
- s., j OCT 191984 MEMORANDUM FOR: Fire Protection Policy Steering Committee FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Consnittee
SUBJECT:
AGENDA FOR SIXTH MEETING The sixth meeting of the Fire Protection Policy Steering Committee will be held at 9:00 a.m. on October 22, 1984 in the Region III offices. At this meeting you should be prepared to discuss and finalize versions of:
. Enforcement Policy
. Generic Letter and Interpretations
. Technical Issues Package
. Standard License Condition
. Inspection Module In addition, we need to prepare our final report to the EDO. I will FAX an outline to you for your consideration today. In addition, there are a number of issues and suggestions still left hanging. For example:
the central point of contact for fire protection issue; the status of NFPA codes; what to do about Tech Specs; and the format of the workshop in advance of the expedited inspections. hdh - ichard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: See next page. (( Enclosure to GL 85-01, Re: Fire Protection Policy
i 4 i Enclosure to GL 85-01, Re: Fire Protection Policy
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' . U. . #' OCT 2 61984 MEMORANDUM FOR: Harold R. Denton, Director, ONRR Richard C. DeYoung, Director, DIE FROM: Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee
SUBJECT:
FIFTH AND SIXTH MEETINGS OF THE FIRE PROTECTION POLICY STEERING COMMITTEE At the fifth and sixth meetings of the Fire Protection Policy Steering Comittee (SC), held in Bethesda on October 17 and the Region III offices on October 22, respectively, the final version of (1) enforcement policy guid-ance, (2) Generic Letter, (3) standard license condition, (4) temporary instruction for fire protection inspections, and (5) technical issues package of questions and answers were discussed, edited, and put into final form. No new initiatives were discussed but the impact and consistency of all initiatives developed by the SC were reviewed.<- The SC also assured that all issues included in the EDO memo of September 13 had been fully addressed and that all issues raised to the SC's attention by other parties had been fully considered. pct [t v
' Richard H. Vollmer, Chairman Fire Protection Policy Steering Committee cc: See next page.
s Enclosure to GL 85-01, Pe: Fire Protection Policy
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cc: W. Dircks V.-Stello R. Minogue T. Murley, R-I J. O'Reilly, R-II J. Keppler, R-III R. Martin, R-IV J. Martin, R-V G. Cunningham E. Case J. Taylor ! D. Eisenhut i R. Bernero G. Arlotto F. Rosa SC Comittee i Enclosure to GL 85-01, Re: Fire Protection Policy 1
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3 j WASHINGTON, D. C. 20555 1 o, ,f 1 January 29, 1985 , l TO ALL POWER REACTOR LICENSEES AND APPLICANTS FOR AN OPERATING LICENSING Gentlemen:
SUBJECT:
OPERATOR LICENSING EXAMINATIONS (Generic Letter 85-04) This letter is to request your best estimate of your need for operator licensing examinations for the remainder of this fiscal year (February 1,1985 to ~ September 30,1985) and fiscal years FY 1986, FY 1987, and FY 1988 (October 1 to September 30 of each yearl. This information is needed to update the schedules you provided in response to Generic Letter 83-40. We are also raquesting that you provide requalification examination schedules for this same time period. Please identify the dates you have scheduled your requalification examination and your anticipated requests for licensing examination sito visits and the number of examinations for each visit.
. Your best estimates are needed to plan for NRC resources to meet your operator licensina needs. Please be aware that in response to budget reductions the NRC has resources for only two visits to each site per year for administering licensing examinations. To meet this goal in FY 19R6 and beyond, the regional offices may be required to radistribute the reauested facility operator examinations visits across the entire fiscal year to even out the examination workload and alininete hinh demand periode. Therefore, your submittal of this schedule does not guarantee the number or date of examinations reauested.
However, an accurate estimate of the need for examinations will allow us to propose budget modifications, if necessary. You should also keep us infonned of significant chances in your estimates as they occur, so that we can keep our data base current. Your schedules, in the enclosed suggested fonnat, should be returned to - Mr. Don Beckham, Chief, Operator Licensing Branch, AR-5221, Washington, D.C., 20555, with a courtesy copy to the appropriate Regional Administrator 30 days from the receipt of this letter. We appreciate your assistance. If you .', have any questions, please call Mr. Don Backham, Chief, Operator Licensing Branch, at (301) 492-4868. { y.
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t : i i j 7_ / This request was approved by the Office of Management and Rudoet (Oun) under
-., clearance OMB-3150-0018, which expired December 31, 1987. .
incerely, i q . g k r,. Sise h Q rec br si rr. ,
., .j Division of(Licensing .1
Enclosures:
} 1. Operatino Licensing Examination
; Schedule 1 2. Requalification Examination Schedule
! 3. List of Recently Issued i Generic Letters I
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Enclosure:
~' List of Generic Letters 1 .1 -i a -i ,
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w 1 m. LIST OF RECENTLY ISSUED GENERIC LETTERS GENERIC LETTER NO. SUBJECT DATE 84-15 Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability 7/2/84 84-16 Adequacy of On-Shift Operating Exper-ience for Applicants '6/27/84 84-17 Annual Meeting to Discuss Recent Develop-ments Regarding Operator Training, Qualifications and Examinations 7/3/84 84-18 Filing of Applications for Licenses and Amendments 7/6/84 84-19 Availability of Supplement 1 to NUREG-0933 "A Prioritization of Generic Safety Issues" 8/6/84 84-20 Scheduling Guidance for Licensee Submittals
,- of Reloads that involve Unreviewed Safety
( Questions 8/20/84 s 84-21 Long Term Low Power Operation in PWR's 10/16/84 84-22 Not used 84-23 Reactor Vessel Water Level Instrumentation in BK9s 10/26/84 84-24 Clarification of Compliance to 10 CFR 50.49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 12/27/84
, 85-01 Fire Protection Policy Steering Committee Report 1/9/85 85-03 Clarification of Equivalent Control Capacity 1/28/85 For Standby Liquid Control Systems 85-04 Operator Licensing Examinations 1/29/85 85-05 Inadvertent Boron Dilution Events 1/31/85 i
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\*...+/ May 23, 1985 TO ALL HOLDERS OF CONSTRUCTION PERMITS AND OPERATING LICENSES
SUBJECT:
10 CFR 20.408 TERMINATION REPORTS - FORMAT GENERIC LETTER NO. 85-08 Pursuant to 10 CFR 20.408, licensees are required to submit to the NRC a report of each individual's exposure to radiation and radioactive material when the individual terminates employment or work assignment at their facility. These exposure reports are commonly referred to as $20.408 termination reports. Previously, we have not specified a preferred reporting format for compliance with this regulation. However, the NRC is currently receiving approximately 100,000 termination reports each year, and this number is steadily increasing. For purposes of efficient automatic data processing, it is important to use a standard format. Processing in a more timely fashion will make the data more useful to the NRC and others in their performance of various duties (see Enclosure 1). For future S20.408 termination reports we request that you voluntarily use the attached Standard NRC Form-439. Instructions for completing the form are attached to the form. Questions regarding these instructions should be directed to Barbara G. Brooks, Office of Nuclear Regulatory Research, Washington, D. C. 20555, (301) 427-4577. The NRC is also conducting a pilot program for the electronic transmission of the termination data to the NRC via computer tapes or by direct linkup to the NRC's computing facility, and we would like to encourage you to consider participating. Should you desire more information about this program please contact Ms. Brooks. The form is intended for use in connection with the information collection requirement established in Section 20.408, 10 CFR Part 20 and approved under 0MB Clearance Number 3150-0014.
/
N1 0 ugh L. Thompson, J'., Dir or D "v ion of Licens ng 0 ce of Nuclear Reacto egulation
Enclosure:
- 1. Uses of Radiation Exposure Data
- 2. Five copies of NRC Form 439 With instruqtions
- 3. L(ist of Generic Let)ters L
u I?y.
L . 4 T i- Uses of Radiation Exposure Data f A number of NRC licensees have inquired as to how occupational radiation exposure data (from reports required by the NRC) are used by the NRC staff. This is a very appropriate inquiry that may be of importance to many affected 1.icensees. In combination with other sources of information, the principal
- uses of the data are to provide facts regarding routine occupational exposures to radiation and radioactive material that occur in connection with NRC-licensed activities, including individual and collective radiation doses from external sources as well as pertinent information on the inhalation of radioactive material (nuclides involved, bioassay results, exposure magnitude, etc.) These facts are used by the NRC staff as indicated below
- 1. The external-dose data permit evaluation of the radiological risk associated with NRC-licensed activities, including the size of the workforce and the collective dose.
e i 2. The data permit evaluation, from the viewpoint of trends, of the
; effectiveness of the overall NRC/ licensee radiation protection and ALARA efforts. They also provide for the identification (and subsequent correction) of unfavorable trends.
- 3. The data provide for governmental monitoring of the potential transient-worker problem.
1
- 4. The data are used in the establishment of priorities for the utilization of NRC health physics resources: research, standards development, regulatory program development.
- 5. The data are considered in reviews of inspection frequencies that are programmed for various categories of licensees.
- 6. Licensing action decisions are often influenced by the data.
- 7. The data are used for comparative analyses of radiation protection performance: US/ foreign, BWR's/PWR's, civilian / military, plant by plant,
. nuclear industry with other industries, etc. i l
- 8. The data permit analysis of annual dose distribution changes which can trigger investigations as to the cause.
- 9. The data are used for purposes of justification in the annual budget process.
! 10. The data provide facts for evaluating the adequacy of the current i risk-limitation system (e.g., are individual lifetime dose limits, worker population collective dose limits, requirements for optimization, etc., l needed). l f 1
- 11. The effectiveness of dose-reduction measures is evaluated using the data (e.g., methods for reducing individuals doses that may increase the collective dose).
- 12. The data provide facts for answering Congressional and Administration inquiries and for responding to questions raised by public interest groups, special interest group's, labor unions, etc.
- 13. The data permit comparisuns of occupational radiation risks with potential public risks when action for additional protection of the public involves worker exposures.
- 14. The data provide information which can be used in the planning of epidemiological studies.
With regard to routine work-place conditions, the annual statistical summary reports required by 20.407, the termination reports required by 920.408, and the annual dose data reported by work function in accordance with Subsection 6.9.1.5 of the standard technical specifications for nuclear power plants provide the only centralized data base available to assist the staff in the performance of its duties as listed above. It is to everyone's advantage if these duties are performed by a well-informed staff in the light of factual informa tion. l i 2
1 DATE 0' kip 0af Nec Foau 438 U.S. NUCLEAR REaOLATsRY COMMISSION is ene to CFA $20 400 2 "" C ''"5' "u" "' S ' REPORT OF TERMINATING INDIVIDUAL'S OCCUPATIONAL EXPOSURE SIE THE ATTACMf D INSTRUCTIONS PART 1. LICENSEE AND INDIVIDUAL 60ENTIFICATION DATA 3 NAME AND ADDRESS Op mEPOaTING LICENSEE 4 NAME OF INDivlDUAL (frat. snakse ewel est) AND ADDAES$ dopnons#7 6 NAME AND ADDaESS 06 EMPLoven IF Disf taENT anOM Asovt sOptenser 6 SOC 4At $tCL,AiTT NUM8tR I MONTM 5 OAy I vgAa SiRTM t I t PART ll. EXTERNAL DOSE DATA l4 PERSONNEL MONeTOniNG FOR ENTERNAL ERPOSURE TO RADsAfaON WAS NOT PaOVIDED 11 ExTatMITY DOSE #<wres 9 PEmeOD'Se OF EXPO,SURE DEEP SM ALLOW esses
, ,,,g, SMALLDW tsee#
a TOTAL to NEUTRON e TOTAL e gETA e PART 111. INTERNAL EXPOSURES TO RADIOACTIVE MATERIAL l 12 PERSONNEL MONIToniNG FOn tuPO5URE TO RADiOACTivi MATER 6AL WAS NOT PROviDED 16 SiOA55Av RESULYS 17 DOSE ESTIMATES #semse I f, , , 14 NUCLIOt 15 FORM a eN viv0 inces b URINALY$15 a COM b ANNUAL c ORGAN ', y , 15 in NESULTSI La MiTTE D DOSE c11 BUADEN 126 ORGAN DOSE
' 01Mtm SiOA$$ Av At$utf 5 20 iF TMis REPOAT 15 SEiNG U$to 70 $AYisFv TMt NOYiFiCAfiON mEOUsmEMENTS OF to CFa to i3 CHECa TME FOLLOWiNG Boa The ,eport a turnened so you un e the swoveari af she Nucmar mesMetor, Cc- a roguasten to CF# Parf ff % shouM grenerve the esport for furene, ,,#prence l YES
thSTRUCTIONS FC COMPLETING NRC FCM 439, Report of Teemmassar Individuars Occuparsonal Esposure if you are Irensed by the U.S Nuclear Regulatory Commnuon 4NRC) as specified m $20 408 st.10 CFR Pan 20. you are required to submn sermination radiauon esposure repons on cenma mdaviduals no the Director. Offwe of Nuclear Regulasory Research. U $ Nuclear Regulasary Commauson. Washington. DC 20555 Thn aformeren u ao be taken from door records thee neust be maananined under $20 a05 for endsveduals likely so receive esponere to radiauon that exceeds a cena a percentage of the NRC done standards for the ehule body, skin or entremstws-2SS for workers of age 18 years or more. $% for workers yoiinger than 18 The term " individual" as ined beloe to represent the worker for whom this report is submmesd The serm "doie" as used in Form 439 and m these masructsons refers to the done m rems as dermed a 520 deal and subesquendy designaaed "done equivalent" in ICRU Repon U (1960s The tune to be covered by than repon as that penod of employment. or work ass gnmes in your facdnyts). ohrh ended enh the most recent termansuon and was not innernspeed by any previous wrmmataan durmg which personnel monnorms was required by 620 202(as and'or biosuays were required by your hcense "Termmatann" a defmad in $20 3saul9) Pans il and III of ihn fann reflect regulatory requirements as well am requests miended to standardize reponeg methods. requests are clearly identsised as such PARTI IJCEN5EE AND INDIVIDUAL IDENTIFICATION DATA This pan idenures the trenner submemns the report and the termmarms mdividual la must be completed even af only one of the remaanmg Parts of this form a appleable Emer the follooms desa ITEM NUMBER I Dear that the repon oss prepared 2 Currene NRC hcenne naamber ausgned to the facdnytsi m which the indavidual received the reponed done, if more than one Irense n savolved, enter the license number for the facday or activay under whxh rnosa of the done was incurred as the farsa esmber If thas is nos practsal. enser the lacenae insmbers en the order of ongmal assuance 3 Peame and address of your facday as at appears on your NRC Irense. 4 The undsvedual's foru nanw. sniddle inanal. and surname a Addrew of the adsvedual may be acluded. Ins a is not emered smo the NRC records synem ) 5 The name and addreu of the undsvidual's employer. if a as differem frorn the reponing luennee (Optusial. nos emered inso the NRC records system ) 6 The nadavidual's social securuy number if noe avadable. enser the word " unknown " 7 The undividual's dase of birth PART II EXTERNAL DOSE DATA For the purpone of the form, the deep done a drfmed as the done assesand as a tusue depth of 3E) or 1.000 mg'cm8 (or leup. the shallom duee as defmed as the dose so the slun of any part of the luuly and the estremnes are defined as hands and forearms, feet and ankles. Isem Numter8 If the mdividual was not monnored for essernal esposure to radianon. you are requested to check the bon to the leh and go to Part 111 COLUMN NUMBER 9 Specify the reponmg surrvals spenods of esposures that the undsvidual was monnored at your facdwyts) pursuant to $20.202 You are requeued so une an-nual uncrements up to the year of wrminanon and uncremens not so encmed one quaner for die year m ehrh the endavishaal errminated ANNUAL: Indraer the month and year of the tugmmng dose of esposure when shoemg annual meremens se g . May 1979) and mdacase the year only for subesesem annual increments.
- QUARTER. For each completed quaner of the year of terminanon, undzase the quaner and year by dese.
CURRENT QUARTER Specify the begmning and ending deses of the acasal esposure penod (month, day, year)-. EJuer the following data 10m Unless the eyes are shweded. enser the deep done anaessed at a tissue depth of 300 mg'em8 tiens depth) or less. If the eyes are pronected by shielding ohnch has a tassue egmvalem thmkness of 700 mg>cm8 or more, the deep done may be annessed at 1000 mg'cm3 (goned depths or less Enger the total done of record, i c.. the highess done received as the selected depth, from all types of enwrnal radiation sources. at any locanon on the body encept the skan and die entremitaes thands and forearms. feet and ankles). 10c For all skan areas. escepi ther of the entremmes, enser a coisma 10c the shallow done of record Racced die total done to the skm 6.e. the lughest done dehvered by all radiauon incudent on the skan. including non4nvini doses from ska consammation. which penetrates to die depth at ehach the shallow done a is desermand. The done at a dapsh of 7 mg?cm or less, everaged over I cma , as acceptable. If Column IOc is len Idank, n udl be assumed that the entry in 10s as appluable also for the shnilow done Therefore, an eury for shallom done is required only if n escends the deep done. 10b & d You are regnseed to emer in column 10b the connbunon made by neutron radeauon so the done reponed m Column 10s and so emer in Column 10d the counbuuan made by hem radianon to the done reponed in Column 10c. Enser XXX if n is known that there was no ewe to redaanon of the type specified en the cohann heading Enwr UNK sf a detectable esposure as reponed in 10s or IOc whrh could have uncluded a been or neutron comenbution of unknown maganude IO & 11 You are requested to enser m or aero ten each column of 10. or in ll)if the done was andreactable. 4 e . ehe redaanon so whsch die worker's dosunseer was espmme produced a response that you consadered to be sasustrally . ' ' - from the response caused by mhetent vanabdars of the donameter system. Nose h is sometunes required to add m ser as equivalents to a real number, although NRC regulanons do nos specify a summahon procedure. the NRC saaff arbitrardy assigns 10 mrems to be a value of m tassuming 0 $ L 410 4 L. ehere L is the desaction tunn) for the purposes of seaustacal analyses. Il Reporting of the entremry does is required. You are regnsand so comply a the following snanner. Eneer the done of record, i s. the highest done, averaged over any area of I cm'. determaned for the skan of the hands and forearms or feet and ankles dunns the reponed penod. It is unneessnary to specify she entremary thes received Jhr done. doses to different estremstws should not be added together. The done is to unclude that delivered by all radia-non acidens on the skm. mcludang non-envial dones from skin comanunstpon, which penetrases to the depth at which die shallom done a determmed. The done at a depth of 7 mg'cma or less es acceptable If Column 11 as leh blank, a mill 12 assumed that the entry in 10c es appixable also for the entremny done. an entry m Coismn 11 is regmrod only of the shallow dose escended the deep done PART til INTERN AL EXPOSURE 5 TO RADIOACTIVE M ATERIAL If you are hcensed by die NRC as specified in 120 400tal 10 CFR Pan 20. and if your hcense requires bionssay services for workers at your farday. you are required to submn ter, nunsuon repons on personnel esposures to r=t-=1 ave enaunal, conssimag mformauen thea you have obsamed in comphance enh the Irense and reconied a compinance enh 620 401 You are requeseed to unclude m each eernunation report information that you have chemined en canyhence with $20103 aW3) and 20103dcN2) and recorded m compliance enh $20 401. This pan proemdes for the reponing of inerrnal monnonns proceduren in irrms of bionsasy results. done estimates. or smake Any one tot moest of these reporung methods may tw used The term "indivulual" is used be60w to represent the worker for whom this report as submnied The errm "esposures to redsoective masenal" es used a con-necten enh these termanstoon reports so represent the eury of redsoecove masenal amo the body. e leern Numher 12 If she endsvidual was not enannored for esposure to radmosctive maernal, you are requesand to check the bon to the left. odwrwene, emer the folloemg data. Column Number 13 If bmassay results are reponed tColumn 165. you are reewued to use du followmg format Sunenanze by year. seperseely hsung the number of measurernents which andscased quantities or concentrations that were andreactable, e e . an the dreactson system used, the redsonuclide present (if any) pro-duced a respanne that you considered no be sasusucally indistinguishable from ns background In Column 13 emer each year biomany was performed. an-ciudmg the year of sernunesen la Columns 14 through 16. use reo lues for each year, as en the eaample shown below, the upper ime for dreettable results and the lower hne for those undetectable In 16e and/or 16b upper har, enter the numhet f actudmg terop of desectable measurements follooed an parenthesis by the highem venfund renutt, of any, lower har. enter the number sincludmg erros of measurements mdcateg undetectable amounts
1%5TRUCT10N5 FC3 COMPLETING NRC FCOM 439 (Conesnueds COLUMN NUMBFR 13 iConunueds Column 13 Column 14 Column 13 Column 16 less Ie g 16b fpCi/Lp 1932 U.nat "I" 0 lung 2fil 2 lung 10 1933 U-nassTh 24: "I* 11 7: lung esos I by 3 19n4 U-nat Th 234 "I" 2:14: lung 12:13> 0 lung 0 Unm. for the numbers m parenthews shown m Column-16b are en be specified a the heading for Column 16b lf Columns 87 or il are comp 6esed nosa-tions en Column 16 are unnecesary if the done commament ISCLycar ensegrased domes en reponed. endraie m Column 83 by begmmng and endmg daies amomh. day, year) the penod dunng whwh the assouaied radeoective matenal man taken into the Imdy if annual doses are reponed. emer m Column 13 the calendar year over whuh each done man ensegrened incleading the farne and any succeeding years of this employment or merk assignmem and the year following the termmanon date For emnes m Column is semaket specify the reponmg nuervals apenada of emposures dunns =hgh the undnadual man esponed to concentrmons of radioactive masenal. uung annual meremerits up 10 the year of termmahon and mcrements mm to enceed one quaner for the year m whsch the endividual termmated The periods of espusure for imakes should appear an follows ANNUAL Indume the month and year of the beginning due of esposeire when showing annual uncrements se 3. June 1933) and indacale the year only for subsequent annual mcrements QUARTER For each comp 6eted quaner of the year of terminaten. Indicate the quaner and year by date CURRENT QUARTER 5pecify the begmmns and endmg deses of the actual caposure penod ernanth. day. year) Reponed entakes whuh mclude only the quammies required so be assessed m accordance with 620103 ax3) are acceptable 84 Bdemify the symbol used m 10 CFR Pan 20. Appendia 8. for the radionuclade or maature of radionuchden for whrh an vivo and/or unnelysm meawremems were performed se 3. Co 60. U 23$3 If the meawred quantny of activry for one radionucluse u also used to emunner other radenuclade quamuses. edemify the redsonucluse actually measured m paremhews emmediately aher the radionuclade luned a Column 14. See der enamp6e given en the directions for Co6umn 13 where U nattTh 234 a emered m Column 14 mdzatmg that the uransum lung burden was drwnmned from n of Th 234 photons 13 Enter the form. 5 for soluble or I for insolub6e. of the radenuclide to ehrh the worker man esposed if unknown. une quaan around the lener. thus m-dKatang whKb concentrasson value in Pan 20. Appendia 8. Table 1. Column 8. =an assumed so apply
- 16. 17 & 13 These columns allow for the reportmg of the resula of the sneernal manneonng procedures an terms of biomassy results. or done asuneses, or maake You may uw one or more of thew methods leadIe & a(2s For each year dunng which m esto measuremems were performed. as shown m Column 13. enser m Column 16mit the number of desactable maswremems followed by the higheu venfeed remit (in nanocurws) m parenthews On the neat ime an the column. emer the number of - ihot endicased undetectable amounts Specify m Column 16st2p the organ in whach the widzaled redsonucinde was found See the saample given a the dwections for Column 13 16b First. emer the gravunetnc or reduwnetrw unit en which the unnalysm resultb are reponed se g . macrograms per leer, nanocenes per Inere en the blank space of the headmg for Column 16b in Column 16b. for each year dunng thrh unnalyses were performed enser the number of detectable results followed by the highest numerical value of the concentranon m unne of the radumuclide lised a Column 14 for the year specified a Column 13 On the nest lme in the column. enser the number of measuremems mdicatmg undetectable amounts See the taample given m the dwectens for Column 13 57a. b. A c Specify in Column 17c the organ or tusue recesving dosn estimated in Chlumn 87a or 17b tNone that a es noe necessary to provide both the commmed and annual doers 1 For Columns 57a and 17b you are requessed to follow the procedures below. af any miserneuve procedures are used. desenhe them on the back of this form In 17e emer the dose imegreed from t so SO years, where t is the taginning deer shown a Column 13 la 17b emer the done m-segrated over each calendar year shown for thm purpose m Column 13 include the Erst and any succeadmg years of this employment or work assignment and the year following the errmmenon date Base done esemanes on the quantsy ses a minamums of the redenuclate. Column 14. sehen meo de body at your facdwytsi dunng thu employmem or work-assignmem penod is Repomag of radionucVr gniakn. an determmed try air sampling. in nos requwed by 10 CFR 20 408 However. should thss opuan he chosen. andcaw the tune weighted contem.. mm of radioactive matersal 8 e . MPC-houru to ehsch de mdividual was espnand dunng the time penods endsaand an Column 13 Refer to the last peregraph of the msanactens for Column 13 for the twne puervals en be used Comptree Columns 13.14. and l$ for each entry a Column 18 larm number 69 Any bonessay results thal cannot be reponed an desenhed above should be emered here tiem number 20 if you enh to send a sgy of thn reppn en the termmatmg mdandual to natafy the nuufacanon sequwements of 10 CFR 1913. check the "Yes" boa PRIVACY ACT STATEMENT Purwant to 5 U 5 C. 5$2asew3L enacted imo las by sectum ) of the Pnvecy Act of 1974 #Pubin Lam 93 579s. the followmg staerment en hannahad so indiviensis and persons wieo supply informanon to the Nuclear Regulaeory Commission on NRC Form 4W Than aformation is maintamed m a sysum of records designesed as NRC.27 and desenbed at 40 Federal Regineer 45344 aOctober 1.1975:
- 1. AUTHoluTY. Sectons $3. 63. 65. 81.103.104.161sbe. and 161 son of ihr Aeonus Energy Act of 1954. an amended 142 U.5 C 2073. 2093 2095. 2111. 2133. 2134.
220 lfts. and 220lton The autherwy for musicame the social securwy number am 10 CFR Part 20 2 PRINCIPAL PURPOSEtsi The aformaium en used by the NRC in an evaluatmn of the ruk of radiaten esposure anacciaeed enh the lEensed activsy and m esercising as sumsory responsibdwy so enomsor and regulate the safety and health practres of its licenneen The data perms a meanaghd compenson of both current and long. term es-posure esperwnce amarg types of lacensees and among Irensees enhen each type Data on your esposure to radiation are swadable so you upon reepsena 1 ROUTINE U5ES The informesma may lw weed so provide data ao other Federal and State agencies evolved in monnonng and/or evakeeung raammam esposure received by mdividuals employed as radiation workers on a permanene or ermporary basis and esposure received by monnored visnors The ufonneuan sesy also be ducioned so an ap-propnaw Federal. Sasse, or local agency in the evem the aformaton indicanes a volanon or posemma violaren of law and m de courns of an adnuniurenve or judsel Proceedmg 4 WHETHER DISCLOSURE IS M ANDATORY OR VOLUNTARY AND EFFECT OF NOT PROVIDlNG INFORMATION ON INDfVIDUAL OR PERSON It is voluntary that you furmsh the requessed informaten. encludmg name. date of binh. and wcnal securry number The social secunry number u used to aannte that NRC has an accurue idenufer not subject en the co acidence of memdar names sw berth denen among the large number of penons on whom date in memtamed Please note. however, that the licennee must file a wrmmsium repen santaient senaan required mformatum. auch an ancial wcurwy number. for nach sadevadual whose employmem or wort assignmets han termensied and for whom permanet rmanorms ma= required under 10 CFR 20 202. Fa41ere of the luenner se provide the afonnataan under 40 CFR 620 202 and 20 408 may subject the trenner to enforcemeni acteun under to CFR 20 e01. 5 SYSTEM MANAGER #5s AND ADDRESS Director. Offwe of Nuclear Regulasor) Rewarth U S Nuclear Regulamry Counmewsm hhinghe. DC 20939
1 DATE OF *iPOa1 - U.S. NUCLEAR LEAULATO .Y COMMISSION lngeOaM ase 10 CF A (20 aos 2 =RC uCENSE NuM.ER,5, REPORT OF TERMINATING INDIVIDUAL *S OCCUPATIONAL EXPOSURE SEE THE ATT ACMf D INSTRUCTIONS PART 1. LICINSEE AND INDIVIDUAL IDENTIFICATION DATA 3 NAME AND ADDo*SS Os mEPOAflNG LeCENSEE a luaME Of INDtviDualifest. snasse reef asser AND ADOnt$$ #epterWf S Naut AND ADDRESS OF EMPLOYER 8 DeFFERENT FROM ABOVE tOpterW8 6 SOOAL 5tCua Tv NUM8Ea I MONTn i oA, e vtAn tiATM l t l PART 11. EXTERNAL DOSE DATA l8 PERSONNEL MONITORING fOR INTERNAL EXPOSURE TO RADeATiON WAS NOT PROVIDED 9 Pinf00tSt OF ERPOSURE n EstaEm DOSE #w UEEP 5 MALLOW seems e TOTAL t> NEUTRON c TOTAL e geTA SMALLOW resel e PART lit. INTERNAL EXPOSURES TO RADIOACTIVE MATERIAL l 12 Pf a50NNEL MONitoniNG FOn EXPO $unt TO R ADOACTivE MATERIAL WAS NOT PmOvtDED IS SCA55AvRESutTS 17 DOSE ESTIMATES sesmos
,', , , ' , 14 NuCLIDE 15 FORM e IN VIVO lac 8 D URINALY$ss a COM D ANNUAL c OAGAN ', y ',
t$ is REsutTS i Li M TTED DOSE 118 SURDEN E2'OAGAN DOSI le 1 Otaf m 0iOA*5Av aE5utTS 20 af TMi$ REPORT IS SEtNG ustD TO SAfl5f v THE NOfif tC ATION ilEOuenEMENTS 08 to C5R 1913 CMECK YME FOLLOWiNG sos l vt $ # The report e furnehed so you umser the provere of rae Nurem Aegusorm Con;wesea a reguaeren 80 C84 Parf F9 tou enouM prensaw the report b furtnse refame i
l l INSTRUCTIONS FC3 COMPLETING NRC FCOM 439, Report of Termmaung Indeculuars Occupoisonal Enemure If you are lsensed try the U.S. Nuclear Regunasory Commauson (NitC) as speedied m $20 408sa).10 CFR Part 20. you are required to sutuna normannison redneten capaaure reporta e coronen endividuals to the Dwector. Office of Nuclear Reguissory Research. U 5 Nuclear Regulaeory Comminion. Washingson. DC 20555 This aformanon is no tu taken from ee records shot annet be ' under $20 401 for mesveduals takely to receive caposure to redashan that enceeds a certam percemage of the NRC does seandards for the whole ady. ska or en . 25% for workers of age 18 years or more. 5% for workers younger than 18 The Ierm "endnulual" es used trios to represent the worker for whom this report es nutunened The term "done" as used a Form 439 and en these matructaons refers to the done in rems as defund a $20 esas and subnaipsently designeed "done equivalent" in ICRU Report 11 (1968). The inme so be covered try than report is than permd of employment, or work ausgnment an your facdayfs). which ended eith the mosa recent termanesson and was nos marrmpsed try any prevanns eernunation dunag which personnel monnonng man required try $20 202(a) and/or twaansys were respaired try your liceme "Termmauon" is defend a $20 3asil!9: Parts 11 and 111 of this form reflect reguisiory reqmrements as well as rapsests onwnded so asandardize reporting methods. requests are clearly identified as such PART1. I.JCENSEE AND INDIVIDUAL IDENTIFICATION DATA Thas part edenties the Irensse submsneg the report and the termmating endsvidual it must be completed even af only one of the rennamag Parts of than form is apphcable Enter the following dass ITEM NUMBER I Daar that the report een prepared 2 Current NRC Irenne number suagned to the facdayist m whnh the endividual received the reponed done. If more than one trense as moolved, enter the licenne number for the facilny or acuvay under =hech mass of the dune was ancurred 6. the first mester II clus is not premacal esser the hcense numbers en the order of onginal issuance 3 Name and addreu of your facddy as e appears on your NRC license d The adavuhaars first name. ensedie samal. and surname (Address of the indivuhaal may be secluded. tua a es not ensered uso the NRC records syneem i i 5 The name and addren of the uuhvadual's employer, if n is differem from the reportmg luenase (Opensial, noe eneered uno sh NRC records sysum i 6 The umhviduars social securwy saemter; if not avadable. enser the word "uniumpen " 7 The endtvedual's dase of turth PART II EXTERNAL DOSE DATA For the purpose of this form, the deep done is defmad as time done anaenasd at a tasaut depth of 300 or 1.000 mg'cm8 for less). the shallom done es defused as the done to the ska of any pan of the body. and the entrenutes are defined as hands and forearna. feet and ankles. lesen Number 8 If the uuhvidual was not monnored for esseraal esposure to radiation. you are rupsessed to check the bon to the ten and go to Part III. COLUMN NUMBER 9 Specify the reporting meervals trenods of esposure) that the endsvadual was mondored at your incday(s) pursuant so $20.202. You are requesand to one an-emal increments up so the year of .. and acrements not so encmed one quaner for the year m ehuh the sadivahial eernunnsed ANNUAL: Indecaer she momh and year of the begenug dose of esposure when showmg annual ancremens te g May 1979) and d- the year only for suh=a9 ent annual uncrements; QUARTER For each compleemd genner of she yet of wrmuemuom. andmane the spearwr and year by dem. CURRENT QUARTER. Specify the begansag ami endang alenes of the arenal esposure pened (mosh. day, year). Enter ehr folloeng data s 10m Unless the eyes are sheidad. enter the deep done annessed at a tissue depth of 300 mg'em (lena dupshi or less if the eyes are proacted by shelding whach has a tueur agmalem thschness of 700 mg'em8 or more, the deep done egy be --' at 1000 mg'cm3 (goned dersh) or less Eamer the casal done of record, a e.. the highese door received at the notected daysh. from all types of enesenal redamuon sources, as any 8~- en the body except the ska and the entresuuss (hands and forearms. feet and anhaes). i 10c For all skan areas. except thes of the causenmes. enser a column 10c the shallow done of record. Record due total does to the ska. i e ithe highest done dehvered by all radiation ancident on she eksn. including non<nvial doses from ska comensunssion, ohnch penetreams no the depeh at which the shallow done is denermined The done as a depdn of 7 mg'cm8 or less, evermond over I cm8 is f==ha. If Column 10c as len blank. a od! be assumed ahme she ansry en iqs is applicable also for the shallow done Therefore. an entry for shallow done as rapsired only of a ev-en the demy done. 10b A d You are rW to amme in column 10b the caserrbuuan made by neutron redensson to the done sported a Column los. and so esser m Column lod the comentazion made by been redantson to the does reponed me Column 10c. Eamer XXX sf a es known ihm there was no esponsre no redestaan of the rype specified a she column heading Emser UNK if a deesciable et==e as reponed an los or 10c wisch could have anchsdad a been or neutron cournbunon of unknown magasande 10 & 81 You are to essor m or aero (in each column of 10. or en 11)if she done was undessetsNr. 6 e.. ihe ed=== to chsch the worker's desameer was esposed a reapanne shot you connadered to be ==*a-ally und . from the reapase caused by enheroes vensbilmes of she donameser syneem Near h as sommemes roeputed to add m for as agnvaienti to a real numtwr although NRC regulaun=a eks nas speedy e - procedure. the NRC saaff arbitrerdy assigns 10 mrems so tw a value of m (anaunung 0 $ L 4 80 4 L. where L as the doesctson lund) for the purposes of sentisencal analy ses. Il Raperung of the entrenuty done es required You are ressessed so casuply en the following misaner Eaner de does of secord, ie , du highest dose, averaged over any area of I em8. determmed for the skan of the hands and forearms or feet and ankles dunng the reparend pened. It is unnecessary to specify the entremry that received Air done, doers to diNerent entremstes should not tw added together Tlue does as so unlude that dehvered try all redes-tion encadres on the skm. uncludeg non-envial doors from skan comenesnation. which penetransa no der depth as which due shallow done is driermmed The 4 does at a depeh of 7 mg'cmaor less en acceptable if Coeven 11 es len lulank. W eill tw assumed thes the entry in 10c as applicable also for the entrempy does, an entry a Column 11 m regmrod only of du shallow does escended the dorp done PART III INTERNAL EXPO 5URES TO RADIOACTIVE MATERIAL If you are iscensed by the NRC as specified in $20 408tal.10 CFR part 20. and af your luenne regures beansesy servnces for workers as your ferdwy. you are regered to sistuna ser-mammaion repna on personnel esposures to rad ==ruve amenal. comasung infonemaan ihas you howe chenmed a complance en* the hcense and recorded in compliance enh
$20 408 You are p to anclude en each sernuestson report aformasson thes you have chenmed in compliance enh 820 Wstast3) and 20.103cc)(2f and recorded en compliance enh 620 401 This pan provides for the reparung of meeraal nionnonng procedures in terms of bioasesy results. done estunsees. or maakt. Any one for niere) of three reportmg snethods may be used The tenn "andsvidual" is used beteo to represem die worker for whnm this repri is netwnmed The term
- esposures ao rad a etive meernal" is used in con-noctuon enh three eernumanon reports to represes the entry of radioactive meernal sneo the body Jeem Number 12 If the undzwadual eas me mosusored for esposure to redsomcuve enasenal you are ragsenend to check the bon to the len. oderwise, enter the follooms desa Column Number 13 If biensaay results are reponed (Column 16s. you are reesessed to see lhe following format Sunenersee try year esymreely lisung the numtwr of
... . s ehwh andrased epsantaars or concentranosis shot were undreactable, e e . in the drisction system unad. the redsonuclide pretene (if any) pro-4
' duced a response ihme you consadered so be massastrally undisunguinhable from as background In Column 13 enter each yet biensaay was performed. m-cluding the year of eermanssion in Columns 14 through 16. une two lines for each year, as en the enample shown below, the upper line for doesctable results and the lower har for those undetectable in les endor 46b iqper har. enser the number (unludang aerot of desciable --- _ followed an paremhrsas by the highest venfied result. if any. lower hne. einer the numtier haciuding serop of ---- _ endicasag undnectable amounts.
- - - . - .,- _ - , . , - _ _ . -- _ - . - - ~ . , -. ..-_ , - - . - _ ,-_ - - _ ._ .. - _ .
INSTRUCTIONS FCJ COMPLETING NRC FORM 439 (Conemund COLUMN NUMBER 13 (Camannede Column 13 Column to Column 15 Cohamn 16 lem Il leef2p leb (pCs/Lt i932 U.a. r- O hug zile 2 lung 10 f 1933 U.nnuTh 234t "1" If7s lung 446s I sung s 1984 U-nastTh 2341 "I" 2:14 lung 12113) 0 jung 0
, Unus for the numbers in parenehetes shoen en Column 166 are en be specifend a che headmg for Column leb. If Columns 17 or le are compleasd. nons-tsom en Column 16 are unnecessary if the done - IS4 year umegrened doses a reponed, andscase m Column 13 by beguuung and ending deers ementh. day, years the persed dunng which Ihr associesed radscactive maneraal som saken uno the body
- if annual doers are reponed. sener a Column 13 the calendar year over which auch done een anergrened. encludang the Gras and any ===eng years of the employners or work assagesnam and the year follooms the arrnunnsson daae For eurws a Column le (menkes. ipacify the reportmg enservals trenods of esposures dunas etach the m&vulual was espuesd to cr=r===ssions of redsoncirve masenal. unang annual incremenn up to the year of eermannon and incremenes nor to saceed one apsarier for ihr year in which the indsvedual termmesed The penada of esposure for umakes should appear am follo==.
ANNUAL Indicane the month and year of the hugenang dose of esposure whes nemeng annual secrements se g . June 19:33 and endscaer ihm year only for j subespre annual encremenas i QUARTEA For each compieend apsareer of the year of isnnensioon. moscaer the apsarert and year by dame l CURRENT QUARTER 5pecify the begnuung asal ending deaes of the actual esposure pened (muneh. day. yeart Raponed emakes which seclude only the gmanmeses reywed so be amassmed e acwdmace endi 620.103(aW3s are assageshie 14 Bdemity she synihol used in 10 CFR Part 20. Appendia B. for ihr redsomuciade or aunesse of radionucladas for whsch a vivo anfor unnalysas
. were performed se 3. Co 60. U 235s if the nasamured esanswy of activry for one redimesclude en aho used to eenmase asher radiomustede apaammass. edennfy the redsonucinde arenally measured en parendernes usunadeserly ahrt ehr redsomucinde tuned en Column 84. $se lhe taample gives a the derections for Column 13 where U-asetTh 234s as enerred m Coimnn 14 undscateg thes the uranum lung banden was desernumed freen _ of Th 234 phana=
IS Esure ihr form. $ for solutet or I for umoluble. of ihr rhlsde to thsh the earter man empened if unknome, une apunes arumed the louer. Wass m-decatmg which -. _ value en Pan 20. Appends: B. Tahoe 1. Column 1. een anaemus en apply 16.17 A is Thane =h=== allow for the reportang of the results of ihr marrnal momennas procedures e ornas of huommesy resehs. or dass amamens. or ummhe ,Y,3 mesy mer one or more of the e nenhads 14 41) & a(23 For each year dunas which in vrvo --- ..._. mere perfonned. am shown en Column 13. esser en Colune limillihr ameer of duencimble
. follooed by the highre venfed resuh (in menacuruss a penmuhrers On the arms har en this column. asser er number of - shot
=aaaad undnerisNe - 5pecify a Colman less2s the orgen m ensch the indicated redsonuctede man found. $se the samuple given a che dwortsons for Coeusna 13.
16b Ferit, esser ihr gravuneens or redsometnc una m sluch the unmalysis resehs are reponed te g macrograms per test. samsonnes per leert en Ihr blank space of she hendung for Cainme let la Column 186. for each year dunas etuch unnelyses were perfonned. esser es sumuter of doesnable foeman foieomed by the highnet numersal value of she conmatranon in unne of the rahamelade head an Colume 14 for te year spactned a Coimne 13 On the ment luar en ihm column. enter the aanbre of andscanag 2 amounts $st the eaample given a de dwertaans for Column 13 17s. b. A c Specify in Column 17c she organ or sensue receiving dones assunmend en C'oeumn 17s or 17b iNoer aims e as mas amosasary to provsde hash the ====ad and annual danos For Columm 17s and 17b you are repasand so follow the procedmes below; si any ahornahve praememos are mand, daarnte sham on me back of shes fann la 17a. eaur she door umegemed from e ao 30 years. where e a she begamag ther shone o Colume 13. In 17b seer the deer m-ergreed over each caisader year shoes for ehm purpose en CcIusnn 13. laclude ihr Erst and any ==adeg years of des amployenus or work assagense and me year foltooms ihr senmannom dmse Base done euensees on the apsenewy (as a nununuma of ihr antamacher Column 14. enhen esto the lady at your factinytst dunag then employamse or work-assagnmem pened le Reporting of rudeanuclade uushan. en esarrunned IP) aer senplung. m nas repared Iry 10 CFR 20 40s Homever, should aus epsen te chneen. endesear the tune eeiglund concemrataans of radamenve meerrial de e.. MPC.hourts to which the m&vedusa was esposed Asnag me tune penuds sedussed a Cahans 13 Refer to the less peregraph of the enuructsons for Column 13 for the tune uusrvals en te used. Ca=pam Columns 13.14. and l$ for each eeuy a Cahuun is. Inun ausnher 89 Any beassesy seensits ther cannat he reported an desenhed above ehnold be eaarred here. leem munkrr20 if you enh to send a copy of dem reynri su she arraumanag sadevadual so natafy the asenfecessna separemsees of 10 CFR 1913. check te "Yes" boa. PfUVACY ACT STATEMENT Pterienne so S U $ C. $52ssen3s. enacted uno law by necison 3 of the Pnvacy Act of 1974 (Pishis be93-5796. Ihr falloming samassent u Asnushed to edivehiels and persons who supply efonnesson to the Nuclear Reguisanry C on NRC Form 4W Then aformanon a maassened e a sysum of records designmed as NRC 27 and dancnted at 40 Federal Register 45344 (Oceater I.1975s 1 AUTHORITY. Sections 53. 61. 63. gl.103 104.161(bs. and 161 sos of der Aaosme Emergy Act of 1954. en assaded q42 U.S C. 2073. 2003. 2095, till. 2133. 2134 2201 b1. and 220lton The aushorwy for noisemens she social necurwy number en IO CFR Part 20
- 2. PRINCIPAL PURPOSE 45: The ofermssene m used by the NRC en an evehessman of the nsk of reaa=== espusere assocised wish the luensed acewwy and en enerrung as sumssary . . Nwy so neonnor and reguiser the safety and hoekh practacen of at licenners The does pernus a mammaghd . . of bush cuness and longarnn en-pneurs esperience asname types of lueenses and amnag licenames enhen each type Dess on your empneure ao ransmen are svealsNe to you upon repost.
- 3. ROUTINE USES The unionusnan may be used to provede does en emer Federsi and Somer agences savolved in ainassanns and/or evolumens ressman espamme received by eneweduals employed as redessaan worters on a pannement or semporary buses and espesure receeved by mannored vousers. The unienneman may aho be discissed to an ap-prepnaar Federal. Same, or local agency en er evens ihr enformsessa endecessi a visionen or poemmel ventanon of les and en the course of an - _, or Judscial Pressuems 4 WHETHER DISCLOSURE 15 MANDATORY OR VOLUNTARY AND EFFECT OF NOT PROVIDING INPORM ATION ON INDIVIDUAL OR PERSON k a voluntary that you fennah the roguessed afonnesson. unludung name dame of terth, and nucial escurvy number The social escurwy number a used to anmere lhet NRC has an accureer adonnfier nue subsect to the coincedence of susmier names or terth deem amung the large nanbre of personn on ohnen esas in annanened Please neer. homever. shes the hceasse misu Hee a ernnusseenn repnet conseemng sertsen repwed sedermante. nach am nacial escunty numhet, for each enhvedual wheer asuptsyment or work assignmem han erruunmed and for ohnm perunusel nuunsering man regu red under 60 CFR 20 202. Feelure of she lueener to provede Ihr enfonneteen under 40 CFR 620 202 and 20 dog may eutgest she lusasse hs enforcemens acten enader 10 CFR 2U 601
$ SYSTEM MANAGERi$e AND ADDRESS Duector. Ofnce of Nuclear Regulaeary Re=earch U $ Nuclear Reguesenry Cesnme==nm Wanlungne. DC 20W
e senc"70au 43B U.S. CUCLEAR LE;;ULATORY COMMISSION ' 0*T E C'
- EPoa t saan to Csa $20 aos 2 ""C '"5' "u"' *"' S' REPORT OF TERMINATING INDIVIDUAL'S OCCUPATIONAL EXPOSURE Sit THE ATTACMED see87 RUCTIONS PART 1. LICENSEE AND INDIVIDUAL IDENTIFICATION DATA 3 haut AND ADDRESS Os mEPOaTING L8CtNSEE 4 NAME OF INDsviouAL trase masse eur w asses AND ADDAES$ deprensu 6 Naut AND ADDatSS 08 tMPLOTER W De5FintNY paOM AsOvt rOprenso 6 SOCIAL SECUAiTV NvusER 5 Mv4TM 3 Day B vtAn 3lATM I a s PART ll. EXTERNAL DOSE DATA l8 PEm$0NNEL MONiTOa:NG pom EXTERNAL EXPO $URE 70 AADIATiON WAS NOT PROvtOED 10 wwOLf 000v Dost emme' 11 INTREMITY D05E arense
, g, g, p g, ,,
, , g,, DEEP SMALLow essas e 70 fat e NEvYaoN e forat e ogva SMALLOW < sacs a
PART 111. INTERNAL EXPOSURES TO R ADIOACTIVE MATERIAL l 12 Pt#10NNEL MONITOR 64G SOR EXPOSURE TO RAQsOACTsyt Maf amiAL wAS NOT PROVIDED 16 SCAS$4v atSULYS 17 005E ESTIMATES e,wnst
' ,$,[,,[,$" 14 NVCLIO4 15 FOAM e psvlvOsnCd e UnNALysis a COM b ANNUAL e OAGAN yp ',
t$ H AESULTS i Li MiTTE D DOSE sti SumDIN (2: ORGAN 0054 b OTMEa seOA$54, mE Sutts 30 15 TMil A(Pont is sting U$to 10 satisf y fMt NOfeseCatsON af 0uenEMENts Os to Cf n it i3 CMtCa TMs FOLLOwiNG aOn l vtS e The <srurt e ruraenso as pov wresse rae are e et sne Nurnur moguesear, Commesem a reevasten to CF# Part 19 tow enover sweeerw #ne report No furmsr reNesare e i I
INSTRUCTIONS FC COMPLETING NRC FCOM 40, Report of Termenstang lascadual n Occupmenni Esposure if you are hcensed by the U $ Nucient Regedeenry Commessen (Nad as syncified in 520 400aar. 80 CFR Part 20. you are required to nubes sermanasson redessen esposure reports on cenne mdaveduals to the Derector. Offwe of Nuclear Regunasary Reeseech. U.S Nucteer Regulmary Commauson. Washington. DC 20555. Thu sformaten a to tw inten from done records shoe must tw ' under $20 401 for und>=eduals hkely a scesse esposure en radiation that escends a censen percemage of the NRC done esmulards for she whole huly. skan or entremnas-2$% for workers of age 18 years or neorr. $1 for workers younger than is The term "andevedual" as used below to represent the worker for whom slus report is subnueled The term "done" as used a Form 439 aM in shene matructens refers en the done am rems as defined m $20 eaa) and =aah==ta===iy deseguesd "done egmtalent" en ICRU ltepon il 41968 The tune to be covered by this repon a thss pened of employnerm. or work assignment a your facdwyds) whech ended wedt Ihr most recent termanmuon and eas not emnerrupeed by any prevenus errmanation durms which personnel monnorms was regured by (20 202tal and or beenmanys were vuonrad by your hceme "Ternunason" m defined en $20 3 san 19s Parts il and Ul of the form reAact regulatory requirements as well am requenas mornded to standard 6te reportag usehods, rogueses are clearly edenufend as such PARTl LJCENSEE AND INDIVIDUAL IDENTIFICATION DATA This pan adsnures the Isenese subaustag the report and the ternunatmg undstadual it must be completed even af only one of she resunenmg Parts of this form a appinable. Emer the fo6 lowing does ITEM NUMBER I Deer thee the repon was perpared 2 Currens NRC license number assigned to the facdwytst en ohech the undstuheal receased the reponed dose If more than one trense is envolves. enser the license number for the imodery or acavny under which most of the done was securred as the forst munber. If Ihas as nos premacal, enter the luense numbers a the order of onginal issuance 3 Name and address of your facdwy as e appears on your NRC license 4 The mes=edual i first name. meddie inesial. and surumme e Address of the in/.avulmal may be erluded. Ins a is not enerred uno the NRC records symem 3
$ The nome and address of the endevalual s empecyer, if n is different from the reportes Irenese (Opsemial, nne eseerad esso the NRC records syurm )
6 The endevedual s saceal secunty number, of not avadahir. enwr the word " unknown " 7 The andevedual's deu of birth PARTD EXTERNAL DOSE DATA For the purpoor of shes form, the deep door a drfined as she done assesand at a tuaue depth of 300 or 1.000 mg'cm8for lesas, du shallow done a defined as the done so der sksn of any pan of ihr body. and she estremetes are defined as hamts and fermarms. feet and ankles. less Number 8 If she uuhvuheal was not monnored for enerrmal esposure so radassme. you are requessed to check the hos so the lea and go so Pan lij. COLUMN NUMBER 9 Specify the reponsa meervals (pereuds of espoures that the mdevedual was monnored at your incdwytsi pursuene to $20 202. You are reeersted to see an. nual servantems up to the year of wrmmasson and screments nos to esmed one quener for du year a ehech the andsvedsel termassed ANNUAL. Indeceu ihr month and year of the begenug deu of esposure when showing annual incremsens (e g . May 1979) and unicaer the year only for subesguent annual morements. QUARTER. For each comp 6 seed gunner of ihr year of errmamuon. andscar the sNarier and year by dem. CURRENT QUARTER Specify the begenang and endeng deses of the seneal esposum pened amonh. day, year). Ener ihr failemme deia los Unless the eyes are shw6ded. emer she deep done aseenamd at a hsaut daysh of 300 mg'em8 (lens daysh) or less If the eyes are prosorted by abwiding whach has a tensus ogmvalene aheckness of 700 mg>cm8 or more, the deep done may be asassend at 1000 mg'cm8 (gened dupshi or less. Emer the eanal done of record. 6 e . the hightse done receeved as the ariscted dersh. froen all types of essernal redention naurces, se any locasion on the body excepe the skan and the entrenutess thands and forearms. feet and ankles) 10c For all enan areas. escept ther of the estremmies, esser a column 10c the saisino= done of record. Recoed du assal done to the skan, i e . the highest duee dehvered by all redession necedent on the skan uncludeng emn-tnvial doses from ska - _ which penserness so the daysh as whech the shallom done es deerrmaned. The does as a depeh of 7 mg'em8or less. sveraged over I cm8 Is arraf=8da If Column IOc a lea blank, a wdt be assenned ther der eury a los as apphcab6e a6eo for ihr shallom done Threefore. an eury for eheilow done en roguered only if n eaceses the deep done lab A d You are recessend so esser en column 10b sier counbution made by asutron redesioon to the done suponed an Columen los. and to eneer m Colwnn 60d the comentution made by beta radealmn to the done reponed m Colwns loc. Enner XXX if n es knows shot ihree was no esponsre to rahetson of the type spacefend in Ihr column heading Enser UNK of a doesclahie esposure es reponed a los or 10c elech could have nucluded a been or neutron - - of unknown maganude 10 A 11 You are ao esser m or aero ha each coeumn of 10. or in ll) if the done oss undreactober. I e . Ihr red as m to which the worker's enunsert was e mp==d a response thma you consedered so be ananian-ally -dine =pmakside from the roepmese cemend by enherent venehdnass of the dosameert synerm Note le a sommemes regered to add m (or as opstaient) to a real number. although NRC regulations do amt speedy a sununation procedure the NRC naaff ortnerardy assigns 10 mrems to be a value of m dessunung 0 5 L 4 40 4 L. ehere L es the doesttaan tuna) for Ihr purposes of ensamtwal analyses Il Reportag of the entremory done es required You are regassesd to comply a the follooms meiner Eneer the does of esconi. I e . the highest done, averaged over any area of I cm 8. determmed for the ihm of the hands am' forearms or fre and enkles dunng the reponed pened It sa un-ocesaary to apscify etw entremory thes received Jhe done. dunes to defferene entrenuters shoeind not be added eagether The does as es include that dehvered by all redes-tonn uncedem on ehr ik n. uncluding ann-trnial doses from ska comamenstson. ohrh prnerreens to she depsh at whech de shallow doet en determmed The does at a espa of 7 mg'cm8 or less es accepeah6e if Column il as left blank. a edi be assumed ther the eury en 10c as applwable also for the entremmy done, an omry a Coeven il a repared only of the shallow dune eacemend du deep done PART 111. INTERNAL EXPOSURES TO RADIOACTIVE M ATEALAL Bf you are luemed by er NRC as specifud en $20 400 sal.10 CFR part 20. and of yonr lecense re, seres banansey servues for workers at your farday. you are required to eastens art-menneson repons on personnel espanwes to redmactive masenal, commeneng mformaison shst you have obenened m complance enh the lacesse and recorded a complean6e enh l $20 401. You are g to enclude en each arrmenstson repon saformseson shot you have chenmed a congdeence enh $20103 san 3) and 20103ccN21 and recorded in comparence i enh $20 401 This pan provides fnt ihr repnetmg of enerneal monmanns precedures m terms of bmasesy renwits. dose estunnus or anneke Any one for sempel of three reportmg nerehnds may be used The erre "endividual" as maad beto= no repreerne the worter for whom this repen en siebmneed The erre "espneures en red ==rtive snownal" e used in con-nurtson enh dwee termenstion repons to reprenem the omry of redescerve masenal enso the body. i j lerm Number 52 If the mdevedual was not neonesored for espneure to redeosclave ensernal, you are requested to check the lies to du let, odwrwise, enter afie following data Column Number 1) If bonesesy reseelts are reponed EColumn 66p. you are regeriend so use du fuisommg fennst Sunenense by year. esparessly leenng the number of l measuremens which mdwaned speanenws or concentreasons ehet were undreactable. 6 e en ihr duesction syneem used. the redenuctede present of anyp pro-disced a response shst you considered to tu nashasually sadistanguishetde from as backgrossad in Coimna Il enser each year bensseay was performed, m-chedmg the year of eermanssion la Columns 14 thronagh 16. une two ines for each year, as en the esemple afmma below. Ihr upper Imr for deesctobte retuha and the lower lane for those endreactable la les ensor leb upper Imr. eneer the number tuaciudeng eerot of duenable measurements followed m i parentheus by the highret venfeed resielt. si any, lower ler enser the number includeg aero)of maaeurements mdwasmg undeesciable amounas 1 1
lhSTRUCTIONS PCO COMPLETING NRC FCOM 439 (Contmunds * , COLUMN NUMBER 13 sCrossmundi Column 13 Celueen 14 Column IS Coluspa 5% leal ti lear 21 le escal 1982 U-nal "I" O lung 2(l) 2 lung 10 ite) U-asuTh 234e ~l" is76 lung de6 1 lung g 19es U-usrTh 234 -l" 2:14e lung 12:83 0 lung 0 Unst for ihr numbers in paremheten shown in Coh'unn le are to be gecified a the heading for Colume leb If Columns 17 or 18 are composed. ama. teonn m Column 16 are unneceuary if the done commenent q$4 year unsgemed domes a reponed. merase m Column 13 by beganing and ending dates sniomh. day. years the pened dunng ohrh Ihr asseressed redsoortive material een inhen suo the body
- If annual dunes are reported. emer a Columen 13 ihr calendar year over weerk each does was uurgrund. martuding the Grue and any succendag years of ihis employairs or work assspunes and ihr year felicong ihr r done For earnes en Cohann le tuushet. epecify the reponeg uservals openada of esposures durung eluch she sadevadual was espuesd to roarentranons of redsoortive masenal. unang annual arresuenes up to er year af . and untrements not so enceed one spanner for er year m whsch the endsvedual errmmesad The periods of esosaure for innahes sMisid appear as follons ANNUAL Indicaer ihr snamh and year of ihr begiamag esse of esposure when shomung samsal urrennents (e 3. June 198h and adscme ihr year only for subanesem annual intronmass QUARTER For each counplssed genreer of ihr year of errmmasson. nadicene she ipsarter and year by deer CURRENT QUARTER Specify she beginneng and ending dears of the ocesal esposure pened imanh. day. yeart Rapaned manhos stuch seclude only the apuumten requand to be amassess e accordmace ese $20.10314N31 age assepsable le idennfy she symbol used m 80 CPR Pan 20. Appendia R. for ihr redsmurtade or numane of redesmurtedes for sench a ervo antor unamiysis
. een performed te s . Co 40. U 2356 If es meenured amassay of arnvuy for one redesmustade is also used no amanuar osher resanuelade apsarmies sdennfy the redesnurlade eranelly measured in paresuheers unmediasety aher ihr redenuchde Ismed a Celune 84 See lhe esemple given an the directions for Coeuen 13 where U ass:Th 234e n enarred in Column 14 undicases shot ihr urassum lung bewden een doesnmand from - of Th 234 yh==a l$ Enter Ihr funn. $ for soluble or I for smaalvbie. of she redsomuchde to stuch er oorber een esponed if unknown. mar spesass afaund Ihr lamer. dan en-draung stuch cessenwereon sehr an Pere 20. Appendes R. TsNe 1. Column I. een assumed to apply 16.17 & IS These cohanns ellee for the sapensag of ihr results of ihr esseruel memesnag psusedures en eenna of humassey resules. or done seemessa. or sambe g may use ear or mese of ihreem lenill & arts For each year eenne south en vivo --- . _ _ mere perfanned. an shown m Column 43, esser a Column lemiliihr number of deserteMe followed by the lughter venfed seemit im amanrunest en pensuheses On ihr neat ime an ens cohann. emner er munher of thss ademed undsomessWe asummann Specify a Column lems2)ihe organ m otuch the ,Am.d ym.m a.4 oss fuund Ser she enauple given e ihr duestion-for Column 13 I le Fwst, esser ihr gravuusent or . una e shech ihr unamiyte reentes are repened te g musesgrams per hast, naamsunes per iner) e ihr blank space of ihr handag for Ceeman leb in Colume leb. for earte year dunne stuch annelyiss mere perienned. esser es museer of dansceaWe reauks follomed be ihr highnes mumencal value of ihr conosauranon an unne of ihr redesnuriale haamd a Celune le for ihr year spanned a Column 13 On the nest ime en alus columm, esser the member of : a edumme endsectaMe ammenes See er esemple given a che duerisons for Column 13 17a. b. & e Spanfy in Column 17c she organ or taase receivsng denes assunseed a Cohenn 87a or 17b (None est a en est asemesary to provsde bush ihr comummed and annual duess i For Caiumns 17a and 17b you me requemed to folloe the peeredures below. of mey missemenve proseduses ese used. deernha them on ihr heck of ihis forme in 17a. enser er done unegreed from i to 30 years. obem e as Ihr hugunug esse shame a Cahmen 13 la 176 emer ihr dame so-espresad over each caender year shown for ilus pawyner a CoIumn 13 larlude er 8rst and any succeshng years of dus engtoyamuu or work assagasnam and me yest folloemt me sonenamen due Rene dame sammenen en ihr apentwy em a mununum) of er omsmauchde. Column 84. shon esso er body a your facilayant ewing ilus employneens or eart-assagense pened 18 Rapenmg of radiasnerinde ensches. as dearrauned by air aangsing. m nas regewed by 10 CPR 20 408 However. shouhl this ephen he chassa. sadscaer es tune.eeigheed conrenerenon. of rednernve meirnal se e . hePC houru so south er adivedual was espumed eines she same penads edussed a Column 13-Refer so she less paragraph of siw meansessasm for Column 13 for me eune unerweis en he used Congisme Columns 13.14. and 15 for each omry a Colman i II lam musher 19 Any bismaaey reensits thei cannot he reponed en dearrihed above themed be entered here
{ lese number 20 if you ==h to send a styy of en reynn su the termaanseg udsvedual to asanfy te assificaessa segureensaas of 10 CPR 1913 chark the Yes" hos PIUVACY ACT STATEMENT Purussar to S U $ C. SS2asenh. ensried uno lee by tertson ) of the Pnvary Act of 1974 dPuMir lae 93-579e. ihr feito=mg sensement a fenushed to andevaduals and persons who supply oformonen so me Nurtser Reguisenry C an NRC Form 4W The enfonnatum n mamesmed m a synaam of records dungamed as NRC 27 anj desenhed at 40 Pederal Regineer 453a4 lOresher 1.1975r.
- 1. AUTHORITY. $ssteens !). 63. 65. Bl.103.104.1614b3. and 16 lies of ihr Asomer Energy Act of 1954. as amended 842 U $ C. 207). 2003. 2005. 28tl. 2133. 2134.
2201(bl. and 2201eops The ausherwy for isierseng ihr morial escurwy number a 10 CFR Port 20 2 PRINCIPAL PURPOSE 4Si The infonnmum in used by the NRC in an evaimenom of the nah of reeseson empanure annarased esdilhe lesensed arnvey and en enerriseg as ! saaneery reopensatday no memsor and reguisse ihr nacery and hashh pretteces of as lerenners The duas perme a sneensagful cessganaan of hash curmes and long.aerm es-pneure espreisere among types of lerenners and anung hcenesen wishe eerb fype Dans on your empueure to redesteen am seedshir to you upon sognet 3 ROUTINE USES The enfunnaman may be used so provide does to aler Pederal and Sener agenriss eveeved a mannenne ankar evolusang retenen enpseur, received by endevinsais employed as sneansa merhers en a permansa or emnparerv hans and espueswe resswed by mismoored vismars The entennsmen misy sens he diertamed so an ap-pseynner Pederal. sese, or Imrel agenry a er event ihr unfunnenen indsessen a visiessen or pumanal visimman of les and an er coures of en eenmuserehve or jederial pretending 4 WHETHER DISCLOSURE 15 MANDATORY OR VOLUNTARY AND EPPECT OF NOT PflOVIDiNG INFORMATION ON INDIVIDUAL OR PERSON lt a volonear) shot you fornesh ihr regsseemd utfennamon, earludeng amme. doet of benh. and inzial escuruy saunber The morial isruary munber as mand so asawe as NItC has an arcereer ideneefert ans nuhyset to che remrulence of musuler names sw hweh denen among she large number of pencen on etusen dass es - Please amar. Immever. Ihas te Isenmar muss nie a ernsummann repen samuseamp sertsen requwed sneennssue. nuch an intel terurwy number. for each mdsvatusi einet emptsyment or mort senegenernt aermessed and for seman perwunst nuanorms oe. reguwed under 10 CFR 20 202 Fedure of er trenese so provuor er endormonen under 10 CFR 620 202 and 20 408 a subyers the lesenser to enforcemens arean under 10 CFR 2U 006- , 5 SYSTEM MANAGERa$p AND ADDRESS Dworint. OfRre of Norteer Reguimor) Remeersh U S Nuclear Reguiamry Ceenmemum waJungne DC 20999
- NAC Foau 435 U.S. NUCLEAR 4.EAULATs Y COMMISSliN 1 DAf t O'
- tP0a' id e>
c CFa62oem 2 ""C t"St *u"'t "' 5' REPORT OF TERMINATING INDIVIDUAL'S OCCUPATIONAL EXPOSURE set TNE AnAcNeo NsfnuctiO=s PART 1. LICENSEE AND INDIVIDUAL IDENTIFICATION DATA 3 NAME AND A00ntS5 Of atPOmfiNG LeCENSEE 4 NAME Of INDIVIDUAL trrss svucedie swer isers AND ADDat55 taprena## 6 NAMt AND ADDatSS Of eMPLOvia se DettentNT paOM AsOvi top #enses 6 SOCIAL SECua#T v NuMeta , I MONf M I Day 3 ytAm einfM i I e PART ll. EXTERNAL DOSE DATA l0 Pla50NNEL MOMTOa NG FOa tufgaNAL ERPO$unt 70 nADeatiON WAS NOT PaOviDED
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PART lit. INTERNAL EXPOSURES TO RADIOACTIVE MATERIAL l 12 Pea 50NNet MoNifon+NG son anPO5uas 70 m AD OACfive MAYsa+AL
- As NOf PaOviotD 16 0#0A55Av atSULTS IF DOSE ISYlM AYES desmet
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is h at$utf$ e L, MifftD D058 its sumDEN #2 OaGAN D054 Of Mt a SiO A $5 A v at Sul f 5 20 69 TMil al#0af f5 sting ultD TO lafessy fMt NOfif sCafiON asOuintMENTS Of to Cf a 1913 CHECa f Me FOLLOwiNG eOm l Vil iFNs repars e furnened se you ir dse a ene poveen e ene s Nur apar seguesvor, Commescri e repuderem f0 Cf a Pere ff toit samm pess ve the report for 4,rener reve eye t
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INSTRUCTION 5 FC] COMPLETING NRC FC3M 439, Report of Terminesms insivulual's Occupational Esposure if you are trensed by the U S Nuclear Regulasory Comnuumn (NRC) as specined a $20 406sas.10 CFR Pan 20. you are required to submm errmmanon radianon esposure reports on certain mdaviduals so the Director. Office of Nuclear Regulatory Research. U $ Nuclear Regulatory Comminion. Washmgton. DC 20555 Thu sformauan a no be taken from done records that must be mesmaaned under l20 401 for indsvaduals likely to receive esposure so radiation that enceeds a cenam percemage of the NRC dune standards for the whose body skan or entremmes-25% for workers of age 18 years or more. 5% for workers younger than is The term " individual" m used belon so reprnem the worker for whom this report es submmed The wrm "done" as used m Form 439 and in these insaructions refers to the done en rems as def ned a 520 dias and subeespernaly desigmaed "done espuvalem" in ICRU Repon II (1963) The time to be covered try thm repon a that penod of ernploynwm. or work awgnmem m your facdaytu. =hrh ended enh the most recem termmauan and was am enwrnspeed by any prevema semunanon dunng whuh personnel monnorms was required by $20 202tal and or bionssays were required by your license "Termmanon" es dermed an (20 3 tait 19 Parts II and III of thu form reflect regulatory requirements as well an requeus miended to saandardue repones metimds. requests are clearly identified as such PARTI LJCENSEE AND INDIVIDUAL IDENTIFICATION DATA Tlus part edenufes the (Kenaar submaning the report and the sermmsung endevulual 11 must be completed even if only one of the remaming Parts of this form m applicahie Enter the following data ITEM NUMBER I Dese that the report was prepared 2 Current NRC trenne amnber auigned to the facdaylis m whah the endividual received the reponed done If mnre than one Ixense is involved. enier the license number for the facdwy or actawny under = huh most of the done was ancurred as the Grit mamber. If this a not praasal, emer the Irenas nwnbers a the order of ongmal usuance 3 Name anal address of your facddy as a appears on your NRC 'scense 4 The moniasual's first name. meddle innial. and surname (Addren of stu mdivusual may be ancluded. tna a a nos ensered saio the NRC records synaem p S The name and address of the nadividual's employer. if a as different from the reporung licenase. dOpuonal. not entered suo the NRC records sysaem 3 6 The endividual's sacsal securwy number. of not avadable. emer the word "unknome " 7 The endavulual's deu of Infth PART11 EXTERNAL DOSE DATA For the purpose of this form the dorp done as de medr as the done assesand at a tanur depth of 300 or 1.000 mg'em8 tor lesse. the shallow done a defined as the done lo the en of any pan of the body. and the entrempus are defmed as hands and forearms. feet and ankJes tem Number S If the mdivutual mas nos morutored for enternal esposure to radianon. you are requeuad to check the bon so the leh and go to Pan ill COLUMN NUMBER 9 Specify the reporung enerrvals trenods of esposure) that the mdividual was monnored at your incdwyts pursuant to 820 202. You are requesied to une an-nua' ancrements up so the year of errmmanon and increments nos to escend one <paarter for the year a whsh the andivuhaal errmanneed. ANNUAL Indnase the momh ans year of the begmning daae of esposure when shommg anniaal incremens de 3 . May 1979) and indicate the year only for ahaent annual increments.
^
QUARTER. For each . _ quaner of the year of ernmnahon. unfuase the quaner and year by date. CURRENT QUARTER 5pecify the tegmnas and ending daies of the acesal esposure pened (momh. day, yearp. Enner the follooms data los Unless the eyes are shelded. esser the dorp done assessed at a tissue depth of 300 mg/cm8tiens depths or less. If the eyes are preencted by ahmiding ohuh has a tusue agavalem thukness of 700 mg'cm or more. the deep done may tw annessed as 1000 mg'em8 (soned depths or less Enter the total done of record. s e., the highesi dune recesved at the selected depih. from all types of essernal redeuon sources, at any locanon on the body entept the ska and the entrenutaea (bands and forearms. feet and ankles) 10e For all skan areas. encept than of tem entremans. enwr in column 10c the shallom done of record Record er ensal does so the skin. 6 e . the highese done dehvered try all radiation incident on the en. mcluding non-anvial doses from en consanunanon. ehsch penetreses to the depth at ehrh the shallow done es deerrnuned The done as a dersh of 7 ms<cm8 or less, everaged over 1 cm8 is accepiable If Column lac as leR Idank, n edt be esaumed thus the eury a 10m is applKable also for the shallom done Therefore. an eury for ihmilow done na required esdy if a enceeds ihr deep done. 10b & d You are regneasd to emer m column 10b the comnbuuan made by neutron radiaten so the done eponed en Column I0n. and to eneer a Column lod the comentymann made by been radiation to the done reponed m Column 10c. Emser XXX if n is known thee there was no espueste to redeuon of the type specifend en the column handag Enser UNK sf a duenciable esposure as reponed m los or loc eluch could have included a tisia or neutron comenbuuan of unknown magnesde IO & il You are requessed so eneer a or aero to each column of 10. or in Ill if ihr done was undresciaber. 6 e , the redaten to ehrh the worker's dosuneter mas esposed pr=6 cad a response that you considered to be seauancelly indesunguishoble froen the resecame caused by wherent venabdans of the desuneter syseem None h as sommemes regorod to add m (or na eqmvalem) to a real number. although PiltC regulatoons do not specdy a summenon procedure, the NRC siaff arbitrardy assigns 10 mrems to be a value of m tassuming 0 $ L 410 4 L where L is the deesction luruel for er piirposes of assustacal analyses Il Reporting of the entremay done is reqmred You are regneasd to conyty en the follooms manner Emeer the done of socord. I e , the highest done. 8 averaged over any area of I cm dreennened for the en of the hands and forearms or fees and ankles dunng she reponed pened it es unnecessary to specify the entremwy thes receivedJhr done, doers to diNerent entrenutes should nos he added together The done as to unclude thss dehvered by all radia-saan nacutem on the skm. encludes non-envial doses from en conearnensaion ehach prartraess to the depth at shach die shallow done es determined. The dune at a depth of 7 mg'ema or less as accepisbie if Column il as leR blank, a edt be assumed that the entry a 10c as applicable aino for alw entremwy done. en eury na Column il as regered only if the shallee done escended the deep done PART III INTERN AL EXPOSURES TO RADIOACTIVE M ATERIAL If you are lacensed by the NitC en specifsed a $20 400f a).10 CFR Pan 20. and if your license reqmten benessey nervices for markers as your facday. you are regered so submd err-mmation reports on personnel esposures to radaometsve masenal. comamans mformaten that you have chiamed a compliance enh the ascenas and recorded en conspliance enh
$20 405 You are regssened to eclude a each errnunanon reppn enformaten that you have etwamed a canyl.ance enk $2010 ham 3) and 2010hcW2p and recorded a compliance enh 120 401 1has part prowsies for the reponing of marrnal monotonns procedures in serms of biossaay results. dnee esaimates or maake. Any one for more) of these reponmg methmis may be used The term "andevidual" as used below to represent the worker fnt ohnm this report is submaned The serm "esposures to radioactive maserial" in used in con-necten enh these erresuieuon reports to represent the entry of radaometeve annernal meo the body.
liens Number 12 If the nidivedual was ma momsored for esposure to redacective masenal, you are requesand so check the lion to the len. otheresse, emer the follomsg data Column Nurnber 13 If tunassey results are reponed EColumn 161. you are requeued to mar du follooms format Surrenanze by year. separessly lesung the number of measuremems shach mdacased quantaters or concemratsuna that mere andreactable. e e , in the detectson synaem used, the radenucher pretene Of anyI pro-duced a response thes you consedered so be saatinascally undanunguishoble from as background la Column 1) enwr each past benanaey mas performed. in-cludeg the year of errnunatum la Columns le through 16. une reo hnes for each year. as en Ihr eaample shown belon, the upper luw for enactable results and the lower line for those undreactable in 16e end or leb apper har. emer the numbre (including aerol of denectable measurements followed m parenthesas by the highest venfied result. if any. looer line, emer the number (includmg eero) of measurements mdecateg undeten-table amounts
Ih5TRUCTIONS FC3 COMPLETING NRC FORh8 439 (Coneinuedi COLUMN NUMBER 83 aConunued Column Il Column 14 Column 15 Column lei g 16sd 21 16b tyCi/L) 1982 U-nat ~l" 0 tung 2(1) 2 lung to 1943 U-nauTh 214: ~1" f(7s lung 4#6s I bg 8 1964 U-nauTh 234: "1" 2s 14e lung 12:13e 0 kg 0 Units for the numbers en parenthews shown m Column 166 are to be spece6ed an the headmg for Column leb !! Columns 17 or 18 are completed. nasa-tion = m Column 16 are unnecessary if the done commament ESO year insegrened domes in reputed indica e m Column 1) by begening asal emhng dawn amonth. day. years the permal dunng which the asnociased radioacteve enetenal mas taken uno the body If annual doses are reponed. emer a Column 13 the calendar year over whuh each done man ensegraied. encluding the Erse and any t-v==d ng years of i tha employmem or =crt assignment and the year folioming the termanosum date For earn m Column 18 tuusket specify the reponmg unerwalm apenods of esponeres dunng whuh the endivulual een espoemd to concemrations of radioactive matenal. uung annual mcrements up to the year of errmmation and incrementa not to escend one quaner for the year m whrh the endivulual errmmated The penuds of esponere for satakes should appear as follo== ANNU AL Indwate the month arnJ year of the beginnmg daar of esponere when showing annual incremenin se 3. June 19831 and indsais the year only for subsequem annual incremems QUARTER For each completed quaner of the year of terminatum. mdgaw the quaner and year by date CURRENT QUARTER 5pecify the beginning and endeng dalen of the octual esposure peruns (month. day. years Reponed imakes which include only the ip.antasm requwed so be anaessed en accordame wie $20103vew3# are --- 14 Idrmify the symbol used sn 10 CFR Part 20. Appendia B. lor the radionuclube or musture of radaanucindes for which an vivo amt/ar unnalysis measurements were performed se 3. Co 60. U 233) If ihr measured quanswy of actnwy for one radanniscinde en also used so enummer other reduziuclube quantnies. edemify the radionuclide actually measured in parenthews immediately aher the redsonuclade lened in Column 14 See the taample given a the directiosa for Column 13 =here U natiTh 234 en emered on Coisma le andmatmg that the uran um fung burden man deernnmed from --- . _ of T) 234 pivions IS Emer the form. 5 for soluble or i for maoluble. of the radionuclule so which the worker man esponed if unknown. une quuses around the lener. thus m-dicasmg ohuh contentratum value a Pan 20. Appendia B. Table 1. Column 1. =an annumed so apply
- 16. 17 & 18 These coisman allow for the reponing of the resulm of the marrnal monmanns procedures in errms of biossasy results, or done eenmanes, or sesake b may one one or more of thrw methods 16asil & ad2e For each year dunng abch m tivo measuremems were performed. an shown en Column 13. enser o Column 16stla the number of dresciable measurements followed by the highese venfeed result fin nanocursrst in parentheses On the near hne en this column. enter the naamber of --- _ thai endwaned undetectable amoums Specify in Column 16ed2) the organ m eluch the endrased radeanuclide was found See the esample given en er directions for Column 1) ,
i leb Fau. enter the gravunetru or raduwnetru une m which the unnalyus results are reponed te g . nucrograms per here annocunes per laer) as the blank space of the headmg for Column 16b la Column leb. for each year dunng which unnalyws were performed. eneer the nasedwr of de: actable results followed by the higterst numerwal tatue of the canamrasson a unne of the redsonuchde haasd a Column 14 for the year specified in Column 13 On the nest har in ihn sciumn. emer the number of measurements indrating undeactable amounts See ihr eaample given a the derectuum for Column 13 8 7a. b. & 6 Specify in Column 87c the organ or tusue receiving dotes enimated m Column 17e or 17b (Nose thas a en me secessary so protute boek the commmad and annual dotes i For Columnn 17a and 17b you are requessed to follow thr' pnxadures below. af any alarrneuve procedures are used dernbe them on the back of thes form le I?a. enter the done energreed from e to 50 years. =here t u the tagsnneng dear shown an Column 13 in 17b enser the done en-ergrated over each calendar year shown for than purpose m Column l) Instude the Erst and any succeeding years of dus employment or work assegnmem and the year follomang the errmination date Bane door esamaws on the quantwy tan a meannump of du th-lute. Column 14. taken suo the body at your fanisyist dunng tNs employnwas or wort assignmrm period 18 Itepnrtmg of radionuchde intakes. an determined by as namphag. es nos required by 10 CFit 20 408 However, should this optum be chosen. andraw the time.weigheed contemranons of radeonctive mairnal es e . MPC 4tourse to which the endsvedual was espused dunng ihr tame penods undrased m Column l} Refer to the lass paragraph of the instructums for Column 13 for the owne meervals en be used Compeste Columns 13.14. and 15 for each entry a Column 18 lum numher 19 Any becamsay results that cannot be reponed as dracnhed above should be enetred here hem number 20 Il you wish en send a 69s ni this repnrt eu the errmmatmg mdividual to matmfy the maifwation resperremens of 10 CFR 1913. check the "Yes" bon PRJb ACY ACT STATEMENT Purwant to S U S C 552asen)e. enacted ima lam by ectum ) of the Pnvecy Act of 1974 Pubin I.a. 91579i. the followmg staiemem is funnahed so umhviduals and persons who supply mformatum to the Nuclear Regulatnes Commnainn on NRC Form 419 Thm mlormatuin es snamiained in a syssevn of recordt desigviesed as NRC 27 ased descritied as 40 Federal Regisert 451a4 aOctober I.1975: 1 AUTHORITY 5ections 53.63 65.II.103.104 161rts. and 161sor of the Asomw Energy Act of 1954. an amended e42 U $ C. 2073, 209). 2005. 2ill. 21)L 2134. 220 lfts. and 220lton The authorwy for iulumns the social iecurwy number en 10 CFR Pan 20 2 PitJNCIPAL PURPOSEal) The aformatum n used by the NRC m ett evaluange of the n6k of radiataan espneere atacciand enh the Irensed activwy and m enerttung es siatusory respnnubilwy en monitor and regulaw the safety and health practKen of at igenneen The dass permit a meanmsful companion of boeh cunent and long term es. pimure esperwnce among types of Iremees and amnng luenwen enNn each type Data on your esposure so radiaeum are swaalable to you upon rapiena
) ROltilNE U$ES The alarmauan may tw used so provide data en other Federal and State agentses evolved 2 monitonag ander evalustang rodaaton esposure recetted by endneduals employed sa rahation workers on a permanene or ermporary bases and esposure racerved by monieured vaport The enformaison may aho he disclosed to an ap-propnaw Federal Senee or knal agency a the event the enformation mduanes a violatum or poorniaal vioisium of law and m the course of an admmistrative or Judwial Proceedmg 4 WHETHEst DISCLO5URE 15 M ANDATORY OR VOLUNT ARY AND EFFECT OF NOT PROVIDING INFORMATION ON INDIVIDU AL OR PERSON le n voluntary that you furenh the requeued aformaton. entluding name dew of twth and unial necweg number The social nacunty number se used to assare that NRC has an accuraw edenhfwr nie nubpect en the emacidente of sumlar names tw thnh down among the large number of peru n= on ohnm dass a mamasined Please nose. however that the twennee enugt file a errminciann repnet somamans tertain required informaimm nuch a. usial teevnry numhet, for each mdivedual ehune employment or work assagnment has errmmaiad and lor ohnm perummet nuanormt es* required under 10 CFR 20 202 Failure of the licenise so prowuer the entormaten under 10 CFR 620 202 and 20 a08 may subpect the irennee es enfortemem action under 60 CFR 2U 601 $ SYSTEM M AN AGERi$i AND ADDRES5 Dwrtior. Offre of Nutlear Regulamry Researsh U $ Nuclear Regulainry Counmmam nasahnghe IEC 20999
LIST OF RECENTLY ISSUED GENERIC LETTERS GENERIC LETTER NO. SUBJECT DATE 84-20 Scheduling Guidance for Licensee Submittals of Reloads that Involve Unreviewed Safety Questions 8/20/84 84-21 Long Term Low Power Operation in PWR's 10/16/84 84-22 Not used 84-23 Reactor Vessel Water Level Instrumentation in BWRs 10/26/84 84-24 Clarification of Compliance to 10 CFR 50.49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 12/27/84 85-01 Fire Protection Policy Steering Committee Report 1/9/85 85-02 Staff Recommended Actions Stemming From NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity 4/15/85 85-03 Clarification of Equivalent Control Capacity 1/28/85 For Standby Liquid Control Systems 85-04 Operator Licensing Examinations 1/29/85 85-05 Inadvertent Boron Dilution Events 1/31/85 85-06 Quality Assurance Guidance for ATWS Equipment that is not Safety-Related 4/16/85 85-07 Implementation of Integrated Schedules 5/02/85 for Plant Modifications 85-08 10 CFR 20.408 Termination Reports - Format 5/23/85 85-09 Technical Specifications for Generic Letter 83-28 Item 4.3 5/23/85 85-10 Technical Specifications for Generic Letter 83-28, Items 4.3 and 4.4 - 5/23/85
UNITED STATES .Asvcassman NUCLEAR REGULATORY COMMISSION '087"*j, * '"5 '** WASHINGTON, D.C. 20566 wasu o e PtAMIT foo G47 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. 8300 O e e
.s 3IIIg 8(gk kg UNITED STATES NUCLEAR REGULATORY COMMISSION '
rj WASHINGTON, D. C. 20555 s.,
/
RIN 2 6 1985 TO ALL LICENSEES FOR OPERATING REACTORS Gentlemen:
SUBJECT:
COMPLETION OF PHASE II 0F " CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS" NUREG-0612. (GENERIC LETTER 85- 11 ) _ On December 22, 1980, NRC issued a generic letter (unnumbered) which was supplemented February 3, 1981 (Generic Letter 81-07) regarding NUREG-0612,
" Control of Heavy Loads at Nuclear Power Plants". This letter requested that you implement certain interim actions and provide the NRC information related to heavy loads at your facilities. Your submittals were requested in two parts; a six month response (Phase I) and a nine month response (Phase II).
All licensees have completed the requirement to perform a review and submit a Phase I and a Phase II report. Based on the improvements in heavy loads handling obtained from implementation of NUREG-0612 (Phase I), further action is not required to reduce the risks associated with the handling of heavy loads (See enclosed NUREG-0612 Phase II). Therefore, a detailed Phase II review of heavy loads is not necessary and Phase II is considered completed. e However, while not a requirement, we encourage the implementation of any
~, actions you identified in Phase II regarding the handling of heavy loads .
that you consider appropriate. For each plant which has a license condition requiring commitments acceptable to the NRC regarding Phase II, an application for license amendment may be submitted to the NRC to delete the license condition citing this letter as the basis. If you have any questions, contact your Project Manager or Don Neighbors (301) 492-4837. Sincerely, u . Thompson , ector Di ion of Licensing
Enclosure:
As Stated f' 1
Enclosure 1 1 NUREG-0612. " CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS" RESOLUTION OF PHASE II Generic Technical Activity A-36 was established to systematically examine the staff's licensing criteria, adequacy of measures in effect at operating plants and reconsnend necessary changes to assure the safe handling of heavy loads. , The task involved review of licensee information, evaluation of historical
~
data, performance of accident analyses and criticality calculations, development of guidelines for operating plants, and review of licensing criteria. The review indicated that the major causes of load handling j accidents include operator errors, rigging failures, lack of adequate inspec-l tion and inadequate procedures. The results of the review culminated in the i issuance of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants" in July 1980 NUREG-0612 described a resolution of Task A-36. l
~
NUREG-0612 presents an overall philosophy that provides a defense-in-depth approach for controlling'the handling of heavy loads. The approach is directed I ~~
- to preventing load drops. The following suunnarizes this defense-in-depth approach:
l 1. Assure that there is a well designed handling system, i 2. Provide sufficient operator training, load handling instructions, and equipment inspection to assure reliable operation of the handling system.
- 3. Define safe load travel paths and procedures and operator training to j assure to the extent practical tt.at heavy loads are not carried over or
. near irradiated fuel or safe shutdown equipment.
- 4. Provide mechanical stops ore 'lectrical interlocks to prevent movement of j heavy loads o,ver irradiated fuel or in proximity to equipment associated l with redundant shutdown paths.
l
'o 2 l l j
- 5. Where mechanical stops or electrical interlocks cannot be provided, i provide a single-failure-proof crane or perform load drop analyses ,
to demonstrate that unacceptable consequences will not result. ' i T I l By Generic Letters dated December 22, 1980, andFebruary3,1981.(Generic i Letter 81-07), all utilities were requested to evaluate their plants against l the guidance of NUREG-0612 and to provide their submittals in two parts; Phases I I (six month response) and Phase II (nine month response). Phase I responses f were to address Section 5.1.1 of NUREG-0612 which covers the following areas: I i _ 1. Definition of safe load paths l 2. Development of load handling procedures
- 3. Periodic inspection and testing of cranes
- 4. Qualifications, training and specified conduct of operators-l l_ 5.
6. Special lifting devices should satisfy the guidelines of ANS! N14.6 6. Lifting devices that are not specially designed should be installed and used in accordance eith the guidelines of ANSI 830.9 i 7. Design of cranes to Ah51 830.2 or CMAA-70 l
-Phase II responses were.to address Sections 5.1.2 thru 5.1.6 of NUREG-0612
! - which cover the need for electrical interlocks / mechanical stops, or alternatively, single-failure-proof cranes or load drop analyses in the spent' fuel pool area (PWR), containment building (PWR) reactor building (BWR), other areas and the specific guidelines for single-failure-proof' handling systems. We have completed our review of the utilities' submittals for Phase I for nearly all operating reactors. Only one plant still remains to be reviewed. During our review we verified that the seven guidelines listed above were l providing the desired level of safety indicated in NUREG-0612. By way of the utilities' responses to the criteria of NUREG-0612, Section 5.1.1 and through discussions with our consultants based on their experiences ! from the reviews, we have concluded that the Phase I guidelines have provided ! an increased awareness by the utilities of the importance of heavy load handling. - Our review has indicated that satisfaction of the Phase I guidelines assures that the potential for a load drop is extremely small. We have noted l t
3 improvements in heavy load handling procedures and training and crane and handling tool inspection and testing. These changes have been geared to limiting the handling of heavy loads over safety-related equipment and spent i fuel to the extent practical, but where this can not be avoided r,to accomplishing it with the operational and other features of the-tirocram imolemented in Phase I. We therefore conclude that the guidelines of Phase I are adequately providing the inten'ded level of protection against load drop accidents. To date we have received Phase II submittals from all licensees. We interpret Phase II of NUREG-0612 as an enhancement to Phase I. Thus, prior ~ to undertaking a review of the utilities' Phase II response for all of the operating reactors, and as a test of the adequacy of the Phase I program, we decided to undertake a pilot program with a limited number of plants. The fir. dings from the pilot program would then provide a basis for a decision on whether to proceed with the review of the Phase II submittals for all operating reactors, to reduce the scope of the review, or to totally eliminate the review. The pilot program involved the review of operating reactors at 12 sites, a total of 20 reactors (eight SWRs and 12 PWRs). Of the 20 reactors, 5 BWRs ,, (Browns Ferry 1, 2 and 3 and Peach Bottom 2 and 3) have single-failure-proof, cranes for all heavy load lifts. " Single-failurg-proof" is used to mean a crane which meets the guidelines of NUREG-0554, " Single-Failure-Proof Cranes for Nuclear Power Plants." Three BWR units (Dresden 2 and 3 and Big Rock Point) have taken credit for a combination of single-failure-proof cranes in some plant areas and load drop analyses in others. Five PWR reactors (Nillstone 2. Prairie Island 1 and 2 and Surry 1 and 2) have utilized the load drop analysis approach. One plant (Kewaunee) has taken credit for a combination of electrical interlocks in some plant areas and load drop analyses in others. The remaining six reactors (Davis Besse. Indian Point 2. Arkansas 1 - and 2 and C.sivert Cliffs 1 and 2) chose to take credit for a combination of administrative controls, procedures and Technical Specification restrictions in conjunction with some type of load drop analysis. This approach does not meet the criteria of Sections 5.1.2 to 5.1.6 of NUREG-0612. Rather, it is an amplification of the quidelines of the Phase I effort, reflecting Section 5.1.1 of NUREG-0612.
. 4 I .
!' It should also be noted that we have completed our review of Phase II for five operating license applicants. Ofthese,two(WNP-2ar.dFermi-2)have, single-failure-proof cranes. The remaining three (Callaway Wolf Creek and Catawba 1 and 2) employ a combination of electrical interlocks, aghanical i stops, limit switches and load drop analyses. % l In addition to the detailed reviews of the Phase II reports in the pilot program and in connection with the five operating licesse applications, we have performed a sufficient review of all other Phase II reports to flag any outstanding plant-spedific concerns reported. I I o-j From our pilot program and OL Phase II reviews, together with the above-I mentioned reviews of the other Phase II reports,we have concluded that the risks associated with damage to safe shutdown systems are relatively small because:
- 1. nearly all load paths avoid this equipment
- 2. most equipment is protected by an intervening floor
- 3. of the general independence between crane failure probability and safety-relar.ed systems which has been observed
~~ 4 . redundaner of components We did r.ot identify any outstanding plant specific safety concern associated with heavy loads handling.
Therefore, most of the risk appears to be associated with carrying heavy loads over or in a location where spent fuel could be damaged. The single most important example of this concerns loads handled over the open reactor vessel during refueling (such as the reactor vessel head). However, as previously mentioned, this is limited to the extent practical and where necessary, is performed with a specifically implemented program in confomance with the Phase I guidelines. I From the pilot program and OL reviews, we noted that nine of the twenty reactors, all PWRs do not have single-failure-proof cranes. To date, we have , not identified any PWRs with single. failure-proof cranes. Further, since f electrical interlocks and mechanical stops are not possible for PWR polar I
l ! 5 cranes, these reactors would be required to perfonn costly detailed load *
- drop analyses. If satisfactory results could not be demonstrated from these analyses NUREG-0612 would call for installation of a single-failure-proof crane.
Based on the above, since a single failure proof crane becomes me l only solution for satisfying the NUREG-0612 criteria, the cost / benefit should be examined. Because we are dealing'primarily with PWRs the cost for modification of a polar crane to meet single failure criteria j (NUREG-0554) guidelines)isapproximately$30million. This includes,
- as the dominant cost element, the cost of the extended shutdown which is required in order to gain access to containment. On the benefit side, ;
given the improvements obtained from the Phase I implementation and the : infonnation obtained in the course of the pilot program and OL Phase II l reviews, we cannot perceive a significant enough benefit in conversion to single-failure-proof polar cranes to warrant the high costs. (See Attachment I for a cost-benefit analysis.) We believe that the cost / l benefit analysis in NtlREG-0612 is no Innger valid because of the benefits realized by Phase I implementation. t
& - ~~ ' We believe the above assessment is further borne out by the industry 4
experience with handling of heavy loads over the years. Precautions *
) have been and are being taken such that no heavy load drop accidents affecting any features of the defense-in-depth against severe core-l damage accidents have occurred.* This determination is also supported by the recomendation of our contractor for the pilot program reviews j (Franklin Research Center) and our benefit-cost analysis suggesting
) that we accept other, less strigent but less costly means for Phase II l compliance as an alternative to the criteria of NUREG-0612 with respect
- to conversion to single-failure-proof cranes.
( Conclusion and Recomendation
- Based on the above, we believe the Phase I implementation has provided sufficient protection such that the risk associated with potential heavy load i
. *There have, however, been recent occurrences of lesser severity. (Seefor i examMe, IE information Notice No. 85-12: Recent Fuel Handling Events LER 84-015, Fort Calhoun 1. Load Over the RCS; and LER 84-006 SanonofreI nothing in this determination should be, Polar Crane Malfunction). regarded as a basis Accordinglyimphasis for any as- of continued attention to the l _ _ safe handlim Gf heavn Imdan
6 i , drops is acceptably small. We further conclude that the objective identified in Section 5.1 of NUREG-0612 for providing " maximum practical defense.in depth" is satisfied by the Phase I compliance, and that the Phase II analyses ~ did not indicate the need to require further generic action at this time. t Thisconclusionhasbeenconfirmedbytheresultsobtained'fromtNePhaseII oilet program and additional Phase II reviews, which identified no residual heavy loads handling concerns of sufficient significance to demand further generic action. All plants have examined their load handling practices against the recomendations of Phase II and submitted the Phase II report. In this way, the utilities were required to identify any unexpected problems to the staff. t
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ATTACHMENT I
SUMMARY
OF COST-BENEFIT ANALYSIS OF PWR POLAR CRANE CONVERSION TO SINGLE-FAILURE-PROOF FEATURES SCOPE The safety benefit of converting the polar crane in the containment of an
"~' operating or completed or nearly completed PWR to single-failure proof features and the cost of the conversion were estimated and compared.
The safety benefit was estimated in terms of the resulting reduction in the risk of a severe accident, involving major radioactive material release, during the remaining plant life. The risk was expressed as the product of the accident probability and the population radiation dose from the release, should the accident occur.
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The cost estimate included the cost of shutdown (or extension of a non-operating period) needed to accomplish the conversion. . ACCIDENT FREQUENCY ESTIMATES Crane Failure Frequency There were 32 crane LER events in the approximately 400 reactor years of U.S. power-reactor operation in the 10 year period July 1969 to July 1979 (NUREG-0612, p. 4-6). None resulted in radioactive release. Of the 32 events, 17 (i.e., just over half) were apparently due to hardware design or fabrication causes, the other 15 to human factors. (Navy crane statistics, cited in NUREG-0612, for 40 load-drop or potential load-drop events in 1974-77 show 80% of the events to be due to human factors.) i 4 4 i
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It may be assumed, as a rough approximation, that Phase I of NRC's heavy-loads generic program is addressed to all the human factors causes and one-half of the hardware causes and succeeds in reducing the affected part vT the failure frequency to a quite small fraction of the frequency - originally present. Since human factors and hardware each contribute about one-half of the failures, approximately 3/4 of the total crane failures can be expected to be eliminated by the Phase I program. Single-failure proof (SFP) cranes should substantially reduce the remaining 1/4 of the failure
-- frequency, though those failures would not be eliminated altogether, since the SFP feature (as defined in NUREG-0554) does not protect against all types of possible failure (e.g. , the bridge is not SFP and the SFP feature itself is subject to defeat by some types of human error). On the other hand, the SFP feature would make the cranes more " forgiving" of imperfections in the Phase I implementation. Accordingly, one may reasonably assume that the SFP !
feature would have a' net effect of eliminating 1/4 of the pre-Phase I
. failure frequency.
enu ----Frequency of Accidents Involving Radioactive Release Not all LER events involve radioactive release. In over 600 reactor years of U.S. power-reactor operation to date [1982] there have, to our knowledge, been no radioactive releases due to load drops. The 10 year period covered by the survey in NUREG-0612, which included 32 crane LER events, all without release, represents about 60% of all U.S. power-reactor operating time to date. An assumption of a pre-fix frequency of some radioactive release once in 1,000 reactor years appears' consistent with the LER-reflected failure experience, taken together with the absence of releases to date. With 1/4 of these releases averted by an SFP crane feature, the pertinent release frequency reduction would be 1 in 4,000 RY. For the most part, these can be assumed to be minor releases due to limited fuel damage in the spent-fuel storage pool or in the reactor.
Frequency of Accidents Involving Major Releases For a load-drop event to cause a major accident, with major radioactive release, special circumstances need to be present -- circumstances that Phase I is intended to make much less likely to occur. A highly damaging heavy load drop, such as one that could destroy a core cooling feature through violation of -- or imperfections in -- Phase I provisions combined with other failures, should be unlikely, and very unlikely to lead to major release, because of back-up safety provisions (e.g. , independent additional core cooling provisions). Review of typical load paths and associated crane-operation frequencies suggests that of all load drops in a typical PWR plant that could have radiological consequences, some 1/4 could involve equipment with a role in i safe reactor shutdown, including primary-system piping. If one assumes that there is typically a 1% probability'that back up revisions would also fail, then the pertinent major release frequency is 1 in 1,600,000 reactor years. W -
= = - Frequency Reduction
- Single-failure proof cranes may reasonably be expected to eliminate most, perhaps 90%, of the residual load-drop probability after the Phase I improvements. Thus, the frequency reduction for major release is approximately 1 in 1,800,000 RY (90% of 1/1,600,000).
l l It should be noted that these estimates are sensitive to plant layout. Plant-specific evaluations could, depending on case specifics, point to a much higher or lower major-release frequency estimate for a specific case. For example, should layout of a specific plant be such that a particularly { unfortunate load drop could destroy all means of core cooling or incapacitate the control room (possibilities suggested by the situations at Montecello and Arkansas Nuclear 1, respectively, before remedial actions were taken at those plants), the above generic analysis could be wide of the mark
. for such a plant. The major-release accident frequency could well be an order of magnitude higher for such a plant (i.e., of the order of 1 in 100,000 reactor years) -- or even higher, depending on plant and crane features, load paths, and operating practices.
CONSEQUENCES ESTIMATE
, Potential radiological consequences of load-drop accidents encompass a wide range of possibilities, depending on specific features of plant design,
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operating practices, and the nature and location of the specific load-drop event. We assume that some -- though very rough -- indication of the severity of the load-drop accident risks may be gained by using in these simplified calculations certain selected release categories described in WASH-1400, Appendix VI, pp. 2-1 to 2-4. Category PWR 4 was selected for the major-release estimates for pressurized water reactors. In PWR 4 core cooling and containment both fail. Core melt occurs. This release category is used to explore consequences of a load drop that incapacitates core cooling (during or promptly after reactor operation), with containment open. ! The release estimates, stated as resulting public dose, based on l l representative generic estimates, for a hypothetical site with a projected Year 2000 mean U.S. power-reactor-site population density, developed by 1 Battelle Pacific Northwest Laboratories (NUREG/CR-2800) is 2,700,000 person-rem. COST ESTIMATE l Costs of change-over to single-failure proof cranes are subject to wide plant-specific variation, depending on the number of features of the specific cranes involved and other aspects of plant design and status. l l l l t
1 Based on advice from the Auxiliary Systems Branch, DSI, and limited vendor and utility contacts, we take the following estimates as representative (as of 1982, when the estimates were made). For future plants, the cost differential for original inclusion of SFP features is estimated at about $250,000 for PWRs (based on information from Ederer Crane Co.). -- At the pre-operating-license stage, with no startup delay, the costs -- including planning, engineering, hardware, installation, and testing -- are estimated at $2 million per plant. This is based on the Monticello experience (1 M in 1976, adjusted for inflation). (The Monticello information was obtained from the licensee through the NRC resident inspector.) For operating PWRs the estimated costs are dominated by plant shutdown during modifications of the polar crane located inside the containment - - ---- building. (The shutdown may be an extension of a shutdown for refueling or other purposes.) The cost effect of a startup delay for a completed or nearJy completed plant would be similar. With a 3-month shutdown and with shutdown costs taken as determined by the cost of replacement power at $300,000 per day, representative total change-over costs for operating PWRs are estimated at about $30 million. RISK REDUCTION Based on the foregoing frequency and consequences estimates, the " expected value" of the risk subject to being affected by the possible Phase II SFP feature, i.e., the magnitude of release times the frequency of its
6-occurrence, integrated for the remaining plant life taken as 20 years, is as follows: Major release risk = 20 x 2,700,000 = 30 person-rem / reactor 2,800,000 COST-BENEFIT RATIO The cost-benefit ratio indicated by the foregoing estimates is approximately
$1,000,000/ person rem. This estimate is subject to wide plant-to plant variation as well as large uncertainties in the underlying estimates of accident frequency and consequences. Nevertheless, it is possible to conclude with reasonable confidence that the benefit-cost ratio for the crane conversion would fail to meet a $1,000/ person rem worthwhileness criterion by a large margin.
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