ML20198B299
| ML20198B299 | |
| Person / Time | |
|---|---|
| Site: | University of Lowell |
| Issue date: | 12/10/1998 |
| From: | Alexander Adams NRC (Affiliation Not Assigned) |
| To: | Bobek L MASSACHUSETTS, UNIV. OF, LOWELL, MA (FORMERLY LOWELL |
| References | |
| NUDOCS 9812180134 | |
| Download: ML20198B299 (10) | |
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f l Dr. Leo M. Bobek l
Reactor Supervisor Nuclear Radiation Laboratory University of Massachusetts Lowell One University Avenue Lowell, MA 01854
SUBJECT:
COMMENTS ON ANSl/ANS-15.17," FIRE PROTECTION CRITERIA FOR RESEARCH REACTORS"
Dear Dr. Bobek:
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1 l Please find enclosed comments on draft ANSI /ANS-15.17," Fire Protection Program Criteria for !
Research Reactors." The comments were prepared by a NRC fire protection engineer. The comments refer to appendices 1 and 2, which are summaries of two site visits conducted by the j fire protection engineer to gain familiarity with non-power reactors and fire protection practices )
at non-power reactors. Because these appendices do not contain any comments related to the standard, I have not included a copy of the appendices as part of the enclosure.
If you have any questions, please contact me at 301-415-1127.
Sincerely, )
Original Signed By:
Alexander Adams, Jr., Sr. Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-223 i
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Dr. Leo M. Bobek l Reactor Supervisor l Nuclear Radiation Laboratory l University of Massachusetts Lowell
! One University Avenue Lowell, MA 01854 l
SUBJECT:
COMMENTS ON ANSI /ANS-15.17," FIRE PROTECTION CRITERIA FOR RESEARCH REACTORS"
Dear Dr. Bobek:
Please find enclosed comments on draft ANSl/ANS-15.17, " Fire Protection Program Criteria for Research Reactors." The comments were prepared by a NRC fire protection engineer The comments refer to appendices 1 and 2, which are summaries of two site visits conducted by the fire protection engineer to gain familiarity with non-power reactors and fire protection practices at non-power reactors. Because these appendices do not contain any comments related to the standard, I have not included a copy of the appendices as part of the enclosure.
If you have any questions, please contact me at 301-415-1127.
Sincerely, Original Signed By:
Alexander Adams, Jr., Sr. Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-223
Enclosure:
See next page cc: See next page DISTRIBUTION:
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%, . m. /g December 10 1998 Dr. Leo M. Bobek Reactor Supervisor l Nuclear Radiation Laboratory University of Massachusetts Lowell One University Avenue Lowell, MA 01854 1
SUBJECT:
COMMENTS ON ANSI /ANS 15.17," FIRE PROTECTION CRITERIA FOR RESEARCH REACTORS" l
Dear Dr. Bobek:
Please find enclosed comments on draft ANSI /ANS-15.1.7," Fire Protection Program Criteria for i Research Reactors." The comments were prepared by a NRC fire protection engineer. The l comments refer to appendices 1 and 2, which are summaries of two site visits conducted by the fire protection engineer to gain familiarity with non-power reactors and fire protection pract(ces at non-power reactors. Because these appendices do not contain any comments related to the standard, I have not included a copy of the appendices as part of the enclosure, if you have any questions, please contact me at 301-415-1127.
Sincerely, Alexander Adams, Jr., Sr. ro'ect Manager Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-223
Enclosure:
See next page cc: See next page l
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' University of Massachusetts Lowell Docket No. 50-223 l
cc:
Mayor of Lowell City Hall. :
Lowell, Massachusetts 08152 {
Dr. David C. Medich.
j Acting Reactor Supervisor University of Massachusetts Lowell One University nvenue Lowell, Massachusetts 01854 j i
Office of the Attorney General I Environmental Protection Division 19th Floor
- . One Ashburton Place Boston, Massachusetts 02108 l
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1 PLANT SYSTEMS BRANCH COMMENTS ON DRAFT AMERICAN NUCLEAR SOCIETY STANDARD 15.17,
" FIRE PROTECTION CRITERIA FOR RESEARCH REACTORS"
1.0 INTRODUCTION
, Title 10 of the Code of Federal Regulations, Section 50.34, required each applicant for a l license to submit a Safety Analysis Report (SAR)in their application. NUREG 1537,
" Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors," specified that although regulations do not require the licensee for a non-power reactor to periodically update the SAR (as is required on 10 CFR 50.71(e) for power reactors), the NRC staff encourages non-power reactorlicensees to maintain current SARs on file at NRC after initial licensing or license renewal. This can be accomplished by submitting replacement pages along with applications for license amendment along with the annual report that summarizes changes made without prior NRC approval under 10 CFR 50.59.
Section 9.3 of NUREG 1537, " Fire Protection Systems and Programs," provides general information concerning the fire protection program required in the SAR which would be acceptable to the NRC. In NUREG 1537, the staff stated that " ANSI /ANS 15.17 1987 l contains generalinformation on fire protection."
By letter dated November 19,1997, Wade Richards, American Nuclear Society Standards Committee Chairman for ANS 15, provided the NRC with a draft of " ANSI /ANS 15.17, Fire Protection Program for Research Reactors," for review and comment.
2.0 REQUEST FOR REVIEW OF DRAFT ANSI /ANS 15,17, " FIRE PROTECTION PROGRAM CRITERIA FOR RESEARCH REACTORS" By work request dated December 3,1997, (TAC M40005), the Non Power Reactors Decommissioning Project Directorate requested that Plant Systems Branch (SPLB) review and comment on the draft ANSI /ANS 15.17, " Fire Protection Program for Research Reactors." As part of its review, the staff visited the NationalInstitute of Standards and Technology (NIST) on Apnl 21,1998 (Appendix 1), and the University of Maryland (UMD) on April 28,1998 (Appendix 2).
3.0 DISCUSSION There are several types of research reactors that produce rated power ranging from 250 kW (TRIGA Types) to 20 MW steady state. Some research reactors are also designed to pulse and may be found to pulse at approximately 1200 MW up to 6500 MW (University of Illinois). Research reactors are normally designed with a core immersed in a large pool of water, in which heat energy is transferred to this water by natural convective flow through the fuel region.
The majority of the research reactors are small(250 kW) and designed to remove decay heat by natural convection. In the unlikely event coolant is lost from smaller type reactors, natural air circulation is sufficient to remove decay heat and prevent fuel damage. In the few larger type research reactors, the core is immersed in a large pool of water sufficient l
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enough to remove decay heat from the reactor. A skirt surrounds the core and prevents loss of water protecting the core in the event of a pipe break. The skirt is designed to maintain sufficient water around the core to provide sufficient core and decay heat cooling.
l Natural air circulation can provide sufficient decay heat removal after cooling water is boiled off.
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An extremely large fire in the reactor area may impede natural convection cooling, l
however, it is not anticipated that sufficient combustibles and large heat release rate will be available to severely restrict the ability of the reactor to cool down.
The release of radioactive material to the environment due to fuel damage as a result of a fire in a research reactor facility is less likely to occur then in a commercial nuclear power reactor. The research reactors are significantly smaller than commercial power reactors.
Many of these small reactors do not produce sufficient thermal heat to damage fuel or control rods. These small reactors locate their fuel rods and control rod assemblies inside pools of water. In the highly unlikely event all of the water was removed from the pool, the reactor is designed to air cool without damage to fuel or control rods. Unlike large power commercial nuclear power plants, there is no need to protect many safe shutdown cables and equipment. The few "large" type of research reactors are also provided with intemal water tanks to provide makeup and decay heat removal. In the event of a control room fire, the operator may scram the reactor. No further operator action is required to maintain the reactor in a safe shutdown condition. Loss of power will also result in the reactor scram. No short circuits to valves and pumps were identified that would lead to a loss of coolant.
l The draft ANSI /ANS 15.17, " Fire Protection Program for Research Reactors," provioed general fire protection criteria for research reactors. This draft standard specified the need to preserve the capability to achieve and maintain safe shutdown of the reactor, and included consideration of both direct fire hazards and indirect or consequential hazards.
Draft ANSl/ANS 15.17, further specified, "It was prepared for those charged with the design, construction and operation, and protection of the research reactor facility.
Application of this standard requires irmut from knowledgeable personnel who are capable l of identifying the safety related systems that should be protected against fire, and from i
knowledgeable personnel who are capable of determining and applying fire protection principles to meet the objectives of this standard."
- a. Fire Protection Program Objective Draft ANSI /ANS 15.17, Section 3, specified that a fire protection program shall be l established for the research reactor facility which provides for fire prevention and reasonable assurance that safety related systems can perform their required functions
- and that the defined Loss Criteria is met. Loss criteria as described by draft l
ANSI /ANS 15.17 is determined through consideration of the credible consequence of a fire upon personnel safety, reactor safety system integrity, other safety-related system requirements, systems, the prevention of radioactive releases, property damage, and program continuity.
Draft ANSI /ANS 15.17. also specified that the fire protection program necessary to achieve the overall objective can be described in terms of its three program components: passive fire protection, active fire protection and fire prevention.
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ANSI /ANS 15.17 defines each of the these program components.
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Draft ANSI /ANS 1517, Section 4, specified that a Fire Protection Program shall be established for the research reactor facility which achieves the " Fire Protection Objective"
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(as described in "a" above). The Program shall be designed with input and periodic review ;
l by the reactor staff and knowledgeable fire protection staff of consultants, shall be adequately funded and shall be documented. An effective program requires support of 1 management. A written description of the ' ire protection program shall be prepared. The l l program description sha!! include, but not be limited to, the following:
l (1) A statement of management commitment to fire protection
! (2) The organizational structure which implements the program and the responsibilities and authorities of the organizational elements.
l (3) Identification of the safety-related systems in the facility (4) The Loss Criteria for the facility i (5) Identification of potential fire situations '
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(6) The combination of program elements provided to meet the program objectives I
(7) Controls established over fire protection activities and a description of such controls j c. Program Components l Draft ANSI / ANSI 15.17 specified that various methods, or elements, may be used to satisfy l
the program objective. It is not required that all of these elements be implemented, but a combination of them shall be selected ."hich will ensure that the program objective is met.
(1) Passive Fire Protection Elements (2) Active Fire Protection Elements (3) Fire Prevention Elements
- d. Fire Hazard Analysis Draft ANSI /ANS 15.17 specified that fire protection measures can be effective only if they are based on proper analysis and evaluation of the risk of fire. A complete evaluation is
, important to provide an optimum level of fire protection. The evaluation process to l determine the risk of a fire included reccgnizing existing and potential hazards, and l
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4 evaluating the adequacy of preventive measures. For research reactors, the evaluation process includes an analysis of radiological consequences.
- e. Fire Safety Assurance Draft ANSI /ANS 15.17 specified that area inspections shall be conducted at least once each calender month, with the interval between inspections not to exceed six weeks.
These inspections shall consists of visual reviews of facility areas in order to detect the !
existence of new potential fire situations and may be coordinated with the inspections and test.
Audits of the total Fire Protection Program shall be conducted on an annual basis, with the intervals between at.dits not to exceed 15 mer.'.hs. The audits shall be conducted not to l exceed 15 months. The audits shall be conducted in accordance with prepared plans.
- f. Appendix Draft ANSI /ANS 15.17 specified that this Appendix is not part of American National Standard Fire Protection Program Criteria for Research Reactors, ANSI /ANS 15.17 1981, however; it is included for information purposes only.
4.0 NRC COMMENTS / RECOMMENDATIONS Draft ANSl/ANS 15.17 provided general framework for development implementation and maintenance of a fire protection program at research reactors. The ANSI committee specified in the draft ANSI /ANS 15.17 that an experienced fire protection engineer and reactor engineer will be necessary to provide an effective fire protection program. The NRC staff agrees that this is needed to provide assurance that an effective fire protection program will be developed, implemented and maintained. However, additionalinformation should be considered in the draft ANSI /ANS 15.17 that further describes certain concepts as described below (Recommendations /
Comments are typed in " italics"):
A written oescription of the fire protection program shall be prepared. The program description shallinclude, but not be limited to, the items given in the following paragraphs.
4.1 A statement of management commitment to fire protection. This shculd also include a descnption of the upperlevel management position that retains ultimate responsibility for the fire protection program. This upper management position should be assigned to a person who has management control over all organizations involved in fire protection activities. This person should retain t?is responsibility even though the formulation and the assurance to the program is delegated.
4.2 The organizational structure which implements the program and the responsibilities and authonties of the organizational elements.
I The wntten descnption of the organizational elements responsible for the following should be clearly identified.'
- a. Fire Protection Program Requirements
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- b. Fire Hazards Analysis [ incudes safe shutdown analysis, Mrs circuit analysis including loss of power to reactor safety circuits)
- c. Design, surveillance and maintenance of all Mrs protection features (such as detection systems, suppression systems, fire pumps, Mre barriers, Sie dampers, fire doors, penetration seals and fire brigade equipment)
- d. fire prevention activities (administrative cor" mis and training)
- e. organizational responsibilities andlines of communication pertaining to Sie pmtection should be defined between the vanous positions through the use of organizational charts and functional descnptions of each position's responsibilities.
The following positions should be designated:
Upperlevel of management responsible for the formulation, implementation and assessment of the effectiveness of the fire protection pmgram.
Management position responsible for the overall administration of the plant operations and emergency plans which include the fire protection and prevention program and which provide a single point of control and contact for all contingencies.
Onsite positions which conduct periodic inspections to minimize combustibles,
' determine the effectiveness of housekeeping practices; inspect and test fire protection equipment; ensures prompt and effective corrective actions are taken to ,
correct conditions adverse to fire protection and preclude their recurrence.
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Fire Bdgade Positions indicating authority and responsibility should be clearly defined.
4.3 Identification of the safety-related systems in the facility The equipment that has been identified to achieve and maintain safe shutdown during a fire should be listed. A safe shutdown analysis describing the safe shutdown method to achieve and maintain stable shutdown condition should be pmvided. It is recognized that in many cases, an insertion of negative reactivity in combination with core cooling provided by the reactor poolis sufficient to achieve and maintain safe shutdown. These cases should be clearly descnbed and reference to acceptable analysis or safety reviews providing the technical bases. The analyses or safety reviews should be available for review.
4.4 The Loss Cnteria for the facihty should be clearly defined. In section 2.2, " Glossary of Terms," Loss Cnteria is defined as cnteria established by facility management in accortlance with all applicable regulations, as limits for risk to personnel, radioactive or toxic contaminant release, property damage, and programmatic intstruptions which might occur from a fire of maximum credible proportions or effect. The NRC recommends that further explanation and examples concemirg the methodology conceming terms " Loss Cntena" and " fire of maximum credible proportions be provided for review and discussion.
4.5 Identification of potential fire situations, and an appropriate assessment of the risk associated with each. Additional information on fire hazard analysis is provided in Section 6," Fire Hazards Analysis."
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6 In section 2.2, " Glossary of Terms,"' potential fire situations"is de&ned as a situation where a Em may occur and result in harm to life, property, or the environment. The NRC l
recommends that further explanation and examples concoming the methodology i conceming potential fire situations be provided for review and discussion.
l (6) The combination of program elements provided to meet the program objectives. The choice of elements shall be based on an assessment which considers the effects on safety related systems of such environmental transients as temperature, pressure, smoke and this assessment shall be part of the program description. The fire prevention elements included in 4.7 have priority in this choice of elements. The NRC requests that further explanation and examples concoming the methodology conceming combination of program elements be provided for review and discussio'1.
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