ML20235T509
| ML20235T509 | |
| Person / Time | |
|---|---|
| Issue date: | 09/19/1987 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20235T514 | List: |
| References | |
| RTR-NUREG-CR-4981, RULE-PRM-50-44 NUDOCS 8710130113 | |
| Download: ML20235T509 (5) | |
Text
y f(p uom[g UNITED STATES y
g NUCLEAR REGULATORY COMMISSION 5 l WA$HIN9 TON, D. C. 20555
% Y OEP 3. 9 1987 -
.a l MEMORANDUM FOR: Victor Stello, Jr.
Executive Director for Operations FROM: ,
Thomas E. Murley, Director .
Office of Nuclear Reactor Regulation i i
SUBJECT:
PETITION FOR RULEMAKING-50-44, " REDUCE FIRE HAZARD FROM NUCLEAR REACTOR GRAPHITE"
- D , ,
Register notice (Enclosure 1) denying , }h Enclosed the subjectfor your signature petition is a the received from Federal Ce _ idnittee To Br.idge the The Gap. a petitioner requested that the Commission amend its regulations to require '
licensees whose reactors, employ graphite as a neutron moderator or reflector l to forrou*, ate and submit for NRC approval fire response and evacuation plans for combatting a reacto'.* fire involving graphite and other constituent parts -l (e.g. fuel). Also, the petitioner requested that licensees perform '
measurements of the EWigser energy" stored in the graphite at their reactor and submit these measurements to the NRC for review together with a revised ,
safety analysis that addresses the risk and consequences of a reactor fire. !
The petition was published for comment in the Federal Register and there were 21 responding entities who opposed and 6 who favored the petition. Abstrac ts of all comments received are included in Enclosure 2. These abstracts were,ij prepared by Science and Engineering Associates of Albuquerque, New Mexico who ,/ i were on contract for technical support to the staff. Enclosure 3 is a * !
contractor NUREG report entitled "A Safety Asse,sment of the Use of Graphite inNuclearReactorsLicensedbytheU.S.NRC"'.(NUREG/CR-4981)preparedby Brookhaven National Laboratory. This report reviews existing literatdre e3d knowledge on graphite burning and on stored energy accumulation and release in i order to assess what role, if any, stored energy release can have in initieting i or contributing to hypothetical graphite burning scenarios in non-power reactors.
It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident.
The staff is recommending denial of the petition for reasons given throughout the respcnse to the petition and the summary on page 17 (Enclosure 1).
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50-44 PDR r '-
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Victor Stello, Jr. SEP i s 1987 Notice: A letter will be sent to the petitioner advising him of this action (Enclosure 4) and a copy will be placed in the Public Document Room. The appro-priate Congressional Committees will be advised of this action (Enclosure 5).
Coordination: The Office of Administration (Rules and Records) concurs in this matter. The Office of the General Counsel has no legal objection. The Office of Public Affairs recommends that a public announcement not be issued.
E Thomas E. urley, Director Office of Nuclear Reactor Regulation
Enclosures:
- 1. FR Notice of Denial of Petition for Rulemaking
- 2. Abstracts of Comments Received 1
- 4. Letter to the Petitioner
- 5. Congressional Letters CONTACT:
T. Michaels NRR/PDSNP Ext. 28251 DISTRIBUTION:
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SEP 1g my Victor Stello, Jr. i l
Notice: A letter will be sent to the petitioner advising him of this action (Enclosure 4) and a copy will be placed in the Public Document Room. The appro-priate Congressional Comittees will be advised of this action (Enclosure 5).
l Coordination: The Office of Administration (Rules and Records) concurs in this matter. The Office of the General Counsel has no legal objection. The Office of Public Affairs recommends that a public announcement not be issued.-
- /
.dri %
Thoma E. Murley, Director _
1 Office of Nuclear Reactor Regulation l
Enclosures:
l
- 1. FR Notice of Denial of i Petition for Rulemaking !
- 2. Abstracts of Coments Received 1
- 4. Letter to the Petitioner
- 5. Congressional Letters l
j CONTACT:
T. Michaels ,
NRR/PDSNP l Ext. 28251 !
Approval for Publication Pursuant to 10 CFR 1.40(o), the Executive Director for Operations is authorized I to deny petitions for rulemaking of a minor or nonpolicy nature where the grounds for denial do not substantially modify cxisting precedent.
The enclosed denial for a petition for rulemaking denies PRM-50-44, "To Reduce Fire Hazard from Nuclear Reactor Graphite," received from the Committee to Bridge the Gap. The petitioner requested that the Commission amend its regulations to require licensees whose reactors employ graphite as a neutron moderator or reflector to formulate and submit for NRC approval fire response and evacuation plans for combatting a reactor fire involving graphite and other constituent parts (e.g. fuel). Also, the petitioner requested that licensees perform measurements of the "Wigner energy" stored in graphite at their reactor and submit the measurements to the NRC for review together with a revised safety analysis that addresses the risk and consequences of a reactor fire.
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The petition does not raise policy issues and the grounds for the denial are in accordance with existing precedent. I, therefore, find that the denial is within the scope of my delegation of authority and am proceeding to issue it.
Date /
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V'ictor SteUo, Jrff !
Executive Directsf for Operations l
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, 3 NUCLEAR REGULATORY COMMISSION 4
10 CFR 50-Docket No. PRM-50-44 Committee To Bridge the GAP; Denial of Petition for Rulemaking AGENCY: Nuclear Regulatory Commission.
ACTION: Denial of Petition for Rulemaking.
SUMMARY
- The Nuclear Regulatory Comission (NRC) is denying a petition for -
rulemaking submitted by the Committee To Bridge the Gap. The petitioner requested that the Commission amend its regulations to require all-licensees l
whose reactors employ graphite as a neutron moderator or reflector and whose- ]I licensed power is greater than 100 W to: (1) fonnulate and submit for NRC approval fire response plans for combating a reactor fire involving graphite and other constituent reactor parts (e.g., fuel); (2) formulate and' submit for NRC approval evacuation plans in case of a reactor fire; and (3) perform {
measurements of the Wigner energy stored in the graphite _ of their reactors and ]
submit these measurements to the NRC for review, together with a revised safety analysis that shall address the risks and consequences of a reactor fire.
The petitioner believes these requirements are necessary because the previous NRC safety evaluations of these reactors allegedly were based on a belief that graphite fires were not credible and on an inability of the NRC and its contractors
- l. to properly calculate Wigner energy in the graphite. The Commission is denying the petition because Fort St. Vrain Nuclear Generating Station and all NRC-licensed researchandtest(non-power)reactorshaveapprovedplansfordealingwith
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emergencies in accordance with existing regulations. The protective actions are based on conservative dose calculations consistent with those proposed by-the petitioner.
Graphite burning is a very low-probability (i.e., noncredible) event and its potential is essentially independent of stored energy in graphite. Empirical .
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measurements of stored energy in graphite are not needed to perform an evaluation of the releasable stored energy. Furthermore, the requirement for,such measure-ments could result in personnel exposures that would be inconsistent with NRC's as low as is reasonably achievable (ALARA) principle. -
ADDRESSES: Copies of the petition, public comments and abstracts of the comments received on the petition, and the Brookhaven National Laboratory Report j NUREG/CR-4981 are available for inspection and copying under Docket No. PRM-50-44 l in the NRC Public Document Room, 1717 H Street NW, Washington, DC. Copies of I
NUREG/CR-4981 may be purchased through the U.S. Government Printing Office by calling (202)275-2060 or by writing to the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. Copies may also be purchased from the National Technical Information Service, U.S. Department of Comerce, 5285 Port Royal Road, Springfield, VA 22161.
1 FOR FURTHER INFORMATION CONTACT: Theodore S. Michaels, Standardization and Non-Power Reacter Project Directorate, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, Washington, DC 20555, Telephone (301) 492-8251.
SUPPLEMENTARY INF0PBATION:
The Petition A petition for rulemaking was filed by the Comittee To Bridge the GAP (CBG) on July 7, 1986. The petition was docketed by the Comission on July 7,1986 and
- was assigned Docket No. PRM-50-44. A notice requesting comments on the peti-tion was printed in the Federal Register on September 3, 1986 (51 FR 31341).
The petition requests that the Comission amend its regulations.
Basis for the Request The petitioner offered the following justification for the proposed ~ revision of the regulations:
The occurrence of a graphite fire at the Chernobyl plant in the Soviet Union demonstrates that such fires are credible events. The 2
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, NRC and its licensees have mistakenly dismissed graphite fires as j 4
noncredible events. .
1 New experimental data show that NRC's generic analysis of stored a 1 energy in research reactor graphite significantly underestimates Ne actual amount of stored energy, and thus underestimates the ass'oci-ated risk of graphite fire. ,
I The NRC failed to require basic safety measures that could help to !
reduce the threat of such a fire. Licensees whose reactors use i graphite, including dozens of non-power reactors and one commercial power reactor, have no fire response plans for combating graphite ,
fires in their reactors. Non-power reactor licensees do not have adequa'te emergency plans to evacuate members of the public in the event of a graphite fire or other severe accident. ,
1 For these reasons, the petitioner would require all licensees whose i reactors employ graphite as a neutron moderator or reflector and whose licensed power is greater than 100 W to: l 1
1 (a) Formulate and submit for NRC approval fire response plans for '
combating a reactor fire involving graphite and other constituent reactor parts (e.g., fuel) which might be involved in such a fire, taking into consideration the potential for explosive reactions.
Response plans shall identify precisely which materials will be used to suppress a fire without increasing the risk of explosion, and shall indicate where and in what quantities these materials will be stored.
(b) Formulate and submit for NRC approval evacuation plans for a reactor fire. Plans should include evacuation out to a sufficient -
distance from the reactor such that no member of the public l receives a dose to the thyroid greater than 5 rem, assuming a release to the environment of 25% of the equilibrium radioactive iodine inventory.
(c) Perform measurements of the "Wigner energy" stored in the graphite of their reactor, and submit these measurements to NRC for review-3
l together with a revised safety analysis, which shall address.the risks and consequences of a reactor fire. A sufficient number of q graphite samples shall be measured to identify the location of maximum stored energy, and to determine the maximum quantity of stored energy within 10%.
I Public Comments on the Petition On September 3,1986, the Commission published a notice in the Federal Register ]
(51 FR 31341) requesting comments on the petition. The NRC received nine }
requests for an extension of the comment period. An extension of the coment period was granted, changing the closing date for the coments from November 3, 1986, to February 2, 1987. A total of 27 coments were received, six of which supported the petition and 21 of which opposed the petition. Of the six comenters supporting the petition, two were individual citizens and four were from citizen's groups. Of the 21 comenters opposed to the petition,15 were universities or university-related organizations, four were companies involved l l
with the nuclear industry, one was a state government agency, and one was an individual citizen.
Of the comments in support of the petition, none offered any specific technical insights but rather simply endorsed the infonnation and basis of the petition.
These comments covered general concerns that include:
the potential for graphite fires, training of firefighters to manage graphite fires.
- evacuation of persons on-site and in nearby areas in the evt:nt l of an accident.
1 Highlights from the comments opposing the petition are as follows: .
- CBG's comparison of research reactors to the Chernobyl-4 (RBMK) reactor ignores the extreme differences in power level, core size, fission product inventory, operating temperature, reactor control systems, and inherent design characteristics.
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- CBG's inference that graphite fires were the initiating events in both the Chernobyl and Windscale accidents cannot be substantiated.
- The operating temperature of the-Chernobyl graphite (700*C) dismisses CBG's contention that stored' energy in the. irradiated graphite played ~
any role in the Chernobyl accident.
- CBG ignores the necessity for an initiating event to raise the graphite temperature 50C'-100C* above its nomal operating temperature before any Wigner(stored)energyingraphitecanbereleased. i
- CBG ignores the fact that only the releasable stored energy, not the total stored energy, in graphite, in accordance with the annealing temperature, can contribute to a graphite temperature increase.
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- The conditions necessary for graphite burning do not exist nor can they be created by random events in non-power reactors.
- The conditions necessary for graphite burning do not exist in the Fort St. Vrain reactor.
Operating temperatures of the graphite in the Fort St. Vrain reactor preclude the accumulation of any significant quantity of stored energy (i.e.,thegraphiteisself-annealing).
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- NRC-approved emergency plans (required by 10 CFR 50, Appendix E) are in place at all NRC-licensed reactors and are adequate and accepteble.
- Measurement of stored energy is not consistent with the ALARA philos-ophy, since it requires the unnecessar/ exposure of reactor personnel.
CBG fails to provide a technical basis for any of the petition',s proposed requirements.
The comments opposing the petition are too numerous to address individually.
However, each comment has been considerc' by the staff and its contractors in analyzing the petition and in developing the NRC position. Abstracts of all 5
coments received and the full. text are available.at the NRC Public Document
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Room in the Docket file PRM-50-44, as noted in the address'section above, e
Analysis of the Petition 6
(1) The petitioner asserts that "the occurrence of a graphite fire at the Chernobyl plant demonstrates that such fires are indeed credible events."
CBG filed its petition on July 7, 1986. Consequently, only fragmentary information, mostly conjecture, was available before the petition was filed. More detailed and definitive information was first made available, outside the Soviet Union, during a meeting held.by the International Atomic Energy Agency (IAEA) in Vienna, Austria, on August 25 to 29, 1986. Without the benefit of the detailed Soviet report, the basis of the petition is seriously flawed.
In response to the CBG assertion regarding the Chernobyl event, the NRC selected BrookhavenNationalLaboratory(BNL),operatoroftheBrookhavenGraphiteResearch Reactor, whose staff is recognized internationally for its research on reactor-grade graphite and its properties, to review the published information and detennine its relevancy to the use of graphite in NRC-licensed reactors. In addition, BNL personnel reviewed the Chernobyl and Windscale accidents and the role, if any, of the graphite moderator in these events. The results of this review are contained in NUREG/CR-4981, "A Safety Assessment of the Use of Graphite in Nuclear Reactors Licensed by the U.S. NRC," July 1987. .This report is available as noted in the address section above.
The staff has used the BNL report, comments received from the public, and its own understanding of and expertise relevant to the use of graphite in non-power reactors and Fort St. Vrain to evaluate and respond to the assertions and proposed requirementsoftheCBGpetition(PRM-50-44).
In their evaluations of the Chernobyl accident, both Soviet and international scientists agree that graphite burning did occur during this accide'nt. However, most of the experts, including the scientists at BNL, consider the graphite burning a secondary or corollary event resulting from the explosions that occurred as a result of a very rapid reactivity insertion that overheated the fuel and-cladding. The explosion created the conditions necessary to initiate and 6
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l sustain graphite burning (e.g., fragmentation of fuel and graphite, rupture of the moderator inert gas boundary, admission of air, a favorable ratio of graph-ite volume-to-surface area, sustained heat input from asphalt fire's, and decay heat). Although the petiti;n considers the Chernobyl accident a demonstration j of graphite fire credibility, the accident confirms that initiation and's~ustained ]
burning of graphite require the existence of a complex combination o'f ideal conditions, which are extremely difficult to achieve in any real situation and are virtually incredible in the reactors being considered under this petition.
The words " credible" and incredible" have been used in many AEC/NRC safety analyses. As used by the staff, these words have always been a qualitative j statement of the likelihood or probability of an event or condition occurring. I i Accordingly, the staff's conclusion that sustained or self-sustained graphite burning is not a credible event in NRC-licensed reactors is still valid (i.e.,
the random simultaneous occurrence of the several conditions necessary for j sustained graphite burning or self-sustained graphite burning is an event with a very small probability of occurring). The staff thus concurs in the conclu- j sion reached in the BNL report: "There is no new evidence associated with the l analyses of either the Windscale accident or the Chernobyl accident that indi-cates a credible potential for a graphite burning accident in any of the reactors considered in this review. Nor is there any new evidence that detailed l case-by-case safety analyses of the role of graphite in NRC-licensed reactors are warranted." Accordingly, there has been no change in the staff's assessment of graphite burning, the Chernobyl accident notwithstanding, in NRC-licensed reactors, and no changes are required in the staff's previous findings in the safety evaluation reports prepared for these reactors.
(2) The petitioner states that "the NRC has failed to require basic safety measures to reduce the threat of a graphite fire."
The petitioner did not identify the " basic measures" the NRC has failed to require and provided no basis for this statement. The staff considers that the elements of the NRC regulatory and licensing process represent the basic safety measures required of licensees to ensure the safe design and operation of their reactors as well as to provide specific plans and procedures for managing and responding to off-nonnal conditions and accidents. Some examples that are relevant to fire ldetection, protection, and mitigation are listed below:
Covers all types of fires, including graphite fires.
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- Safety reviews of non-power reactors include an assessment of the fire protection systems at each facility. Fire detection, fire extinguishers, fire alarms, fire prevention, fire fighting training of facility personnel, and onsite and offsite response _ to fire alarms are typical
-areas included in the safety review. Inadequacies identified duririg7h~e review must be corrected before a license is granted.
- Each non-power reactor licensee is required by conditions of the license (Technical Specifications) to provide a safety review for experiments to be inserted in their reactors and for changes in reactor operation.
Among many other safety considerations, an assessment of fire potential (e.g., flammable materials) is included.
- Each non-power reactor licensee has responded to the requirements of 10 CFR 50.54(q) and 10 CFR 50, Appendix E, in submitting an emergency plan for NRC review and approval. All licensed non-power reactors now have approved emergency plans and the necessary implementing procedures. These plans were reviewed against ANSI /ANS-15.16-1982 and Regulatory Guide 2.6, proposed Revision 1, as outlined in NUREG-0849,
" Standard Review Plan for the Review and Evaluation of Emergency Plans for Research and Test Reactors."
Examples of the evaluation items that are relevant to " basic safety measures to reduce the threat of... fire" are listed below:
e (a) The [ emergency) plan should also describe non-radiological monitorsorindicators....(2)Firedetectors....
(b) The emergency plan should describe an initial training and periodic retraining program designed to maintain the ability of emergency response personnel to perform assigned functions for the following: ,
...f. Police security, ambulance, and fire fighting personnel....
(NUREG-0849, Sections 8.0and10.0)
The licensee for Fort St. Vrain has satisfactorily met the requirements of 10 CFR 50,48 and 10 CFR 50, Appendix R. Appendix R " Fire Protection Program 8
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for Nuclear Power Facilities Operating Prior to January 1,1979," sets forth fire protection features required to satisfy Criterion 3 of Appendix A to 10 CFR 50. These NRC requirements include the " basic safety meastires to reduce the threat of a... fire." ;_
It is the staff's judgment that the NRC has required adequate basic safety measures to reduce the threat of fire as well as to mitigate the consequences of any fires that do occur. These measures have been reviewed, approved, and implemented for all licensed reactors. They generally apply to all fires and have been found to provide acceptable protection for the health and safety of the public.
(3) The petitioner alleges that " licensees have no fire response plans for graphite fires."
As discussed in item 2, above, all licensees have NRC-approved emergency plans in accordance with 10 CFR 50.54(q) and 10 CFR 50, Appendix E. These plans provide for response to fires, for training of fire fighting personnel, and for periodic drills to demonstrate proper operation of the plan in accordance with procedures developed for each facility. One commenter opposing the petition !
reported that the offsite fire fighters and their supervisors were regularly trained in fire fighting procedures for their facilities and that the fire fighters were confident that they were prepared to deal with the type of fires- they could encounter, including a fire involving graphite. This is consistent with BNL research,2 which recommends a basic fire fighting technique for graphite fires, j
that is, exclude air or oxygen and cool the graphite. Success in using this basic " cool-and-smother" technique was demonstrated during the Chernobyl accident.
Cold nitrogen gas was pumped into the bottom of the reactor to successfully cool the graphite and fuel debris while excluding oxygen to smother any burning.
Also at Chernobyl, graphite blocks were successfully quenched using water (NUREG-1250,pp.4-12,4-21,and7-23). Since this basic cool-and-smother 2
R.W. Powell, R. A. Meyer, and R. G. Bourdeau, " Control Radiation Effects in a Graphite Reactor Structure," Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Vol. 7, 1958,
- p. 293.
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technique is effective for'most fires, the staff has concluded-that the ifcensee' existing emergency plans provide an adequate response for graphite fires as well as any other type of fire. l (4) The petitioner asserts that "non-power reactors do not have adequ te-emergency plans to evacuate members of the public in the event 'of a graphite fire." l Neither the petitioner nor any of the citizens' groups or individuals support- !
ing the petition provided a basis in support of this assertion. The staff has reconsidered the need to provide a plan to evacuate members of the public located off site in the very unlikely event of a graphite fire and, in the !
course of evaluating this petition, has not identified any such need._ ,
As stated in Regulatory Guide 2.6, Revision 1:
In the judgement of the NRC staff, the potential radiological hazards to the public associated with the operation of research and test reactors are considerably less than those involved with nuclear power plants. In addition, because there are many different kinds of non-power reactors, the potential for emergency situations arising and the consequences thereof vary from facility to facility. These differences and variations are expected to be reflected realistically in the emergency plans and procedures developed for each research and test reactor facility.
Accordingly, each non-power reactor licensee has developed an emergency plan based on the identified characteristics of its reactor facility. To assist licensees in meeting the requirements of 10 CFR 50, Appendix E, Regulatory Guide 2.6 (ANSI /ANS .15.16-1982, Table 2) provides an " Alternate Method for j DeterminingtheSizeofanEmergencyPlanningZone(EPZ)." Table 2 is based on highly conservative dose calculations that are generically applicable to non-power reactors. These calculations include the very conservative' assumption !
for non-power reactors that 25% of the equilibrium radioactive iodine is gaseous and will escape from the reactor building into the environment. It is the current 4 and standard practice of the NRC staff to use the 25% iodine source term with regard to 10 CFR 20 recommended dose considerations in its safety evaluations of non-power reactors. Table 2, which is based on power level, recommends 10
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I that reactors with power levels less than or equal to 2 MW use their " operations I boundary" for their EPZs, which essentially recognizes that a reactor of this power level will only need to initiate protective actions for members of the general public on site and will not pose an unacceptable radiological hazard to members of the public off site. There are only five licensed non-power reactors l containing graphite that have power levels greater than 2 MW. Three of the j reactors have power levels less than 10 MW, one has a power level of 10 MW, and one has a power level of 20 MW. Table 2 recommends an EPZ of 100 meters for non-power reactors with power levels greater than 2 MW and equal to or less j than 10 MW, and 400 meters for those with power levels greater than 10 MW and equal to or less than 20 MW. The licensee for each of these reactors has an NRC-approved emergency plan that takes into consideration the specific charac-teristics of each reactor (e.g., fission product inventory and engineered safety ]
features) in the development of the action levels, procedures, and protective I actions necessary to protect all members of the public within its EPZ. Regula-tory Guides 1.3 and 1.4 recommend the use of the 25% radioactive iodine source l
i term in determining the compliance of power reactors with the siting, containment, and dose guidelines of 10 CFR 100. The staff believes the current regulatory practices are suitable to ensure that the basic statutory requirement, for adequate protection of public health and safety, is met. l 1
These emergency planning considerations are appropriate for reactors utilizing graphite components. Because the graphite contains no fission products 3nd very few activation products, even the remote possibility of the graphite burning would not contribute to the radiological source tenn. Therefore, a graphite fire in and of itself presents essentially no radiological hazard to the public.
l Because of the major differences in design, power level, core size, fission product inventory, reactor control systems, and inherent reactor neutronics, comparison of the Chernobyl accident and its consequences with accidents and the resulting consequences for non-power reactors is not appropriate, nor is it meaningful. Many of the comments received in opposition to the petition speak of the impropriety of comparing NRC-licensed non-power reactors with the Chernobyl RBMK-1000 reactor.
The petitioner has not provided any proof of inadequacy in the emergency plans for non-power reactors. On the basis of a review of the guidance for l 11
emergency planning contained in Regulatory Guide 2.6 and ANSI /ANS'15.16-1982 and the requirements fo 10 CFR 50, Appendix E, the staff has conc 1,uded that the emergency plans previously approved by NRC are still appropriate and adequate.
Neither the petitioner nor the commenters supporting the petition haveisupplied
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information_that demonstrates that, even in the remote case.of graphite burn-ing, there is a need to modify any existing emergency plans. j (5)' The petitioner states that "NRC's generic analysis of stored energy in j research reactor graphite significantly underestimates the actual amount of -
stored energy and thus underestimates the associated risk of graphite fire.
i The conditions necessary for stored energy. releases in graphite are described in Section 3 of the BNL report. The staff agrees with the methodology derived for estimating the stored energy that can be released from graphite and in the analysis applied to the estimation of stored energy releases in Section 6'of the BNL report.
i In Section 2 of the BNL report, the necessary conditions for graphite to burn
- are discussed in detail. A reassessment of the literature on the experiments-previously performed at BNL and the reported details of the Windscale and Chernobyl accidents are included in the BNL study. The conclusions reached as a result of these analyses are:
[T]he potential to initiate or maintain a graphite burning. incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for each research reactor and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, rea: tion product i
removal and a favorable heat balance must be maintained. There is no new evidence' associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for.a graphite burning accident in any of the reactors considered in this review.
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On the basis of its review of the BNL report, the literature on BNL experiments, 1 and the information on the Windscale and Chernobyl events, the staff finds that .
the conclusions reached by BNL are correct and adopts them as its own.
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(6) The petitioner asserts that " actual empirical measurements of Wigner energy will be required to assess the magnitude of the energy stored in
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research reactor graphite."
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Measurements of stored energy in its research reactor graphite were made' by the University of California, Los Angeles, in the course of decounissioning its Argonaut research reactor. Several things learned from its program of sampling and measuring stored energy were reported by a connenter who opposed the petition. This information was also reported in a paper by Ashbaugh, .
3 Ostrander, and Pearlman at the American Nuclear Society annual meeting in {
June 1986.
Stored energy decreases with increasing distance from the fuel region j (e.g.,5.61 cal /gmat18 inches,1.34 cal /gmat22 inches,andanunmeas- f urableamountat26 inches).
Within the graphite island, stored energy decreases from 33.3 cal /gm at the fuel box graphite interface to 19.2 cal /gm about 3 inches from the fuel box toward the center of the graphite island.
These results illustrate the principles associated with the proposed requirement to measure the Wigner energy stored in the research and test reactor graphite.
The significant changes in stored energy with relatively small differences in location demonstrate the difficulty in selecting the locations and the number of samples needed to characterize the " maximum stored energy and ?.o determine i the maximum quantity of stored energy to within 10%."
The bases for storage and release of Wigner energy in graphite are delineated in the BNL report, which shows that there is no unique connection between total stored energy and the releasable energy. Thus, estabhshing the magnitude of the stored energy in non-power reactor graphite by empirical measurements would not provide the information needed to evaluate this potential. Because the releasable stored energy saturates, an upper bound on the stored energy that 3C. E. Ashbaugh, N. C. Ostrander, and H. Perlman, " Graphite Stored Energy in the UCLA Research Reactor," Transactions of the ANS, Vol. 52, 1986, p. 372.
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can be released to 700 C can be determined from existing data. Therefore, no measurement of stored energy is required. ;
Also, because of the several conditions required to initiate graphite burning in addition to a graphite temperature of 650'C, the potential to initiate or maintain a graphite-burning incident is essentially independent of stored '
energy in the graphite. This further supports the conclusion that no measure-ment of stored energy is needed. l Many of the comenters who opposed the petition cited a violation of ALARA considerations because stored energy measurements would not provide needed infomation, but would incur radiological exposures. The impracticality of l taking the samples and making the measurements was also pointed out. For i example, sampling the graphite reflector pieces in the ends of a TRIGA fuel l j
pin would require breaching the fuel pin cladding as well as providing l l shielding against the fuel pin's radioactivity. Similar challenges would be !
associated in taking a sample from graphite reflector components clad with l metal. In addition, it was pointed out that numerous samples would be required to establish the true magnitude of stored energy in the various graphite components.
The staff has considered the relevant BNL findings and the connents received j l
l and has concluded that empirical measurement of stored energy in non-power reat tor graphite components is not practical nor is it necessary to ensure the health and safety of the public.
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(7) The petitioner refers to "one comercial power reactor," indicating that l it has no fire response plans for combating graphite fires. The petitioner j also states that " graphite is used as a moderator in the Fort St. Vrain nuclear power plant in Colorado." ;
i Other than the lack of graphite fire response plans, the petitioner does not {
identify specific concerns related to Fort St. Vrain. However, it is implied !
that all reactors using graphite components are subject to CBG's concerns and j assertions. In reality, the petition and requirements are really directed at !
NRC-licensed non-power reactors. ;
l 14 j
_ _ _ -_ - - - - - _ ]
1 Fort St. Vrain is a high-temperature gas-cooled reactor (HTGR) owned and oper--
ated by Public Service Company of Colorado. Its design capacity is'330 MWe.
]
It uses a ceramic fuel particle (uranium and thorium carbide) clad with j silicon carbide and multiple layers of pyrolytic carbon. The fuel panicles-are compacted into small rods and installed in fuel holes in the hexagonal graphite fuel blocks. Including the reflectors there are 500 tons of reactor' graphite in the core. The reactor coolant is helium with an average inlet temperature of 762'F (405*C) and an outlet temperature of 1445 F (785'C). The average graphite moderator temperature is 1380*F (749'C). These characteristics are far different than those of the non-power reactors. BNL has reviewed Fort St. Vrain parameters in relation to graphite stored energy and concludes in Section 7 of its report, " Fort St. Vrain operates at temperatures that preclude accumulation of stored energy. There are no known problems associated with stored energy in graphite for operating temperatures associated with HTGRs." ;
The staff agrees with BNL's conclusion and can find no reason to empirically l measure the stored energy in Fort St. Vrain's graphite components. !
In response to an NRC request, Public Service Company of Colorado addressed I
the implications of the Chernobyl accident for Fort St. Vrain. The licensee submitted a final report entitled " Design Differences, Air Ingress and Graphite Oxidation, and Steam Ingress and Water Gas Generation" (P-86641, l December 4,1986). The staff has reviewed the report and concludes that the I only significant similarity between Chernobyl and Fort St. Vrain reactors is that they both contain a large amount of graphite moderator.. There are I design differences between these reactors that preclude an accident similar to the Chernobyl accident at Fort St. Vrain.
l l Furthermore, on the basis of its review, the staff concluded that the structural integrity of the Fort St. Vrain prestressed concrete reactor vessel would be maintained during and after the assumed accident scenarios. Although the initiating events are beyond the plant's original design basis, the plant design appears to have an adequate margin of safety to withstand these events.
l The staff's comments and conclusions can be found in the NRC Public Document Room under Docket No. 50-267, in a letter dated April 1, 1987, Accession No.
8704090248.
15
The petitioner's assertion that graphite burning and oxidation were not included in the staff's evaluation for Fort St. Vrain is in errori LThis subject was thoroughly reviewed in both the construction permit and operating license safety evaluations. These staff evaluations may be found in the P0b11c
^
Document Room in the 50-267 docket file. The licensee's updated Fort St.
Vrain Final Safety Analysis Report, Section 14, contains much of the infonnation and analyses submitted for NRC review. The staff concluded that significant graphite oxidation at Fort St. Vrain was not credible. (Note: In
{
addition to the previously discussed conditions necessary for graphite burning, j Fort St. Vrain must suffer simultaneous independent structural failures resulting in the release of the inert helium and the subsequent supply of an adequateair/oxygenflow). The stati finds no basis for changing its previous conclusions. The licensee for Fort St. Wain has met the requirements of 10 CFR 50, Appendix R (which sets forth fire protection features required to i satisfy Criterion 3 of 10 CFR 50, Appendix A) and has an NRC-approved emergency plan that meets 10 CFR 50, Appendix E. The Fort St. Vrain fire protection program and emergency plan specify the necessary organization, plans, and j procedures to provide the necessary protection of the health and safety of the public even in the very unlikely event of a graphite fire.
r e
16 l
Ee BASIS FOR DENIAL
_n ,
The NRC denies the petitioner's request to amend 10 CFR 50 to require licensees whose reactors employ graphite as a neutron moderator or reflector and whose licensed power is greater than 100 W to:
(1) formulate and submit for NRC approval fire response plans for combating a reactor fire involving graphite and other constituent reactor parts (e.g., fuel);
i (2) formulate and submit for NRC approval evacuation plans in case of a reactor fire; and l l
(3) perform measurements of the Wigner energy stored in the graphite of their reactors, and submit these measurements to the NRC for review together 9 with a revised safety analysis that shall address the. risk and consequences l of a reactor fire.
This denial is based on the following:
l (1) Each licensee of a non-power reactor has submitted an emergency plan that l has been approved as meeting the requirements of 10 CFR 50, Appendix E. i The petitioner has not demonstrated that these plans do not provide an l appropriate level of protection of the health and safety of the public.
l (2) The licensee for Fort St. Vrain has an approved emergency plan that meets the requirements of 10 CFR 50, Appendix E, as well as an approved fire l protection program that meets the requirements of 10 CFR 50, Appendix R.
In addition, at the request of the NRC, the licensee has submitted a report addressing the implications of the Chernobyl accident for Fort St.
Vrain. The report has been reviewed and approved by the staff. The petitioner has not provided a technical basis that would show that an
- additional fire response plan would enhance the protection provided for the health and safety of the public by the existing emergency plan and fire protection program.
17
9 (3). Measurement of maximum stored energy in non power reactors are no.t.ne.ces- !
sary to ascertain the releasable stored energy in graphite components below 650 C. Existing knowledge provides-this-information-which is adequate for a safety evaluation of the effect of stored energy on the potential for graphite burning and the associated danger to the health and ,
'I safety of the public. Additionally, such measurements are contrary to the NRC's ALARA principle, since unneeded knowledge would be sought at the expense of unnecessary personnel exposure.
Accordingly, the Commission denies the petition. ]
l Dated at Bethesda, Maryland, this ed_b day o 1987.
l 1
For the Nuclear Regulatory Commission i i
Victet Stell M r.
l Executive Directo i for Operations ;
e 1 i 1 .
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4 Enclosure 3 1 l l l l 1 l 1 l l 1 I L______-_____.____ .j
NUREG/CR-4951' BNL-NUREG-52092' A SAFETY ASSESSMENT OF THE USE OF GRAPHITE IN NUCLEAR REACTORS LICENSED BY THE U. S. NRC l D. G. Schweitzer D. H. Curinsky l i E. Kaplan C. Sastre Manuscript Completed: May 1987 Date Published: July 1987 Prepared by Brookhaven National Laboratory Upton, New York 11973 Prepared for Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 FIN A-3855
E.- 4 CONTENTS' ABSTRACT . . . . . . .. . . . .:. . .-. . . ... . . ...... 'l
'l. INTRODUCTION . . . . . . . , . . .. . . . . . . . .-. . ...... 2
- 2. GRAPHITE BURNING . . . .. . . ... . ...... . . ...... '2L
- 3. STORED ENERGY . . . . . . . . .... . .... . ... .... 7? j 3.1 Sommary . . . '.. . . .. .. . . ...... . ...... 7 3.2 Wigner Energy -- Its Generation and Buildup. . . . . . . . .. 7 3.3 Stored Energy Releases . .- . . . ... .... . ...... '12 3.4 Calculational Approaches ... ... ... ... ....... . 16 4 THE CHERNOBYL ACCIDENT-. . . . .. . . ........
...... 17 5._ " ACCIDENT AT.'INDSCALE W NO. 1 PILE ON 10th 0F OCTOBER, 1957" ... 18
- 6. U. S. RESEARCH REACTORS . . . . ... ... .... . . ...... 19 6.1 Criteria for Stored Energy in Graphite . .... ...... 19 6.2 Stored Energy in Graphite. . . ... ..... . ...... 20- '
6.3 Graphite Burning . . . . . . . . ...... . . ...... 22
- 7. FORT ST. VRAIN - GRAPHITE STORED ENERGY . .... . ....... 24
- 8.
SUMMARY
. . . . . . . . . .. .... ... .... ... .... 24 8.1 Graphite Burning . . . . . . . . . . . . . . . .... ... 24 8.2 Stored Energy in Graphite .. .... ... .. ...... 25 8.3 Safety Assessment . . .. .. . ... ... . ....... 27 i l
1
- 9. CONCLUSIONS . . . . . . . .. ... . ....... . ...... 27 j i
- 10. GLOSSARY . . . . . . . . . . . .. . . ... . . .-. . . . .... 28
- 11. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 l
- 12. BIBLIOGRAPHY . . . . . . . . . ... . . ...... .. ..... 33
]
I
-l 1
l, j i i i : i j l l
.]
E______________________________.________ _ _ _ _ _ _ . _ _ _ _ __.._.m __j
[' 1 FIGURES Figure 1. Graphite burn configuration . . . . . .. . . . . .... . 5 Figure 2a. Total vs released stored energy .. . .. . . . . ... . . 10 l Irradiation = 30*C, Tanneal = 800*C l Figure 2b. Total vs released stored energy . . . .. ... . . ... . 11 Irradiation = 30'C, Tanneal = 400'C Figure 3. Stored energy released . . . . . . . . . .. ... . ... . 13 Figure 4. Cumulative energy release; .. . . .. .... . . ... .. 14 l exposure of 500 mwd /AT or les,s Figure 5. Cumulative energy release; .. . . . ... ... . ... .. 15 irradiations at 70*C and above i l l 1
)
i I J i 11 i _-___-___-_m
.I
'~ 1 A Safety Assessment of ~ the Use of. Graphite )
in Nuclear Reactors Lic'ensed by the U.S. NRO. I D. G. Schweitzer, D. H. Curinsky, E..Kaplan and C. Sa's'tre j ABSTRACT , This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases.in order _to assess what' role, if any, a stored energy release can.have in initiating or con hypotheticalgraphiteburningscenariosinresearchreactors.{ributingtn It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). . The conditions necessary to initiate and maintain graphite burn'ing are summarized and discussed. - From analyses of existing information it is con-cluded that only stored energy accumulations and releases below the burning temperature (650*C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that-stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the. energy produced during the first few minutes of a burning incident. The Windscale and Chernobyl accidents are summarized and reviewel. It is shown that there is no evidence from the Chernobyl event that stored energy releases played a role either initiating or contributing to this accident. An improperly controlled process of annealing the graphite at Windscale with nu-clear heat resulted in damage to the fuel elements that initiated fuel burning which resulted in a graphite fire. Stored energy releases did not initiate or contribute to this accident either. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is: essentially independent of the stored energy in the graphite, and depends on other factors that are unique for each 1 research reactor and for Fort St. Vrain. In order to have self-sustained l rapid graphite oxidation in any of these reactors, certain necessary condi- ! tions of geometry, temperature, oxygen supply, reaction product removal, and a .! favorable heat balance must be maintained. There is no new evidence associ- 1 ated with ei'.her the Windscale Accident or the Chernobyl Accident that indi-cates a credible potential for a graphite burning accident-in any of the reactors considered in this review. 1 I i
- 1. Research reactors as used herein means research, test, and training reactors.
I I
- _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ = _ _ _ _ _ .
- 1. INTRODUCTION on September 3, 1986 the NRL published in the Federal Register [51FR3134, 1986) a notice of receipt of a petition for rule making filed by The Committee to Bridge The Gap to consider the subject of graphite fires in U.S. research nuclear reactors. Under contract with the NRC staff, Brookhaven National Laboratory staff with past experience in safety evaluation of graphite burning and stored energy releases initiated a reevaluation of graphite burning and stored energy information. The objective of this evaluation was to develop an analysis of the potential role of stored energy releases in initiating or con-tributing to graphite burning scenarios, as well as an analyses of graphite ignition and self-sustained combustion in the event of a LOCA accident.
The 1986 accident at Chernobyl motivated studies describing the causes for the accident. As a result of this new information, BNL has undertaken a reevaluation of the Windscale Accident, graphite burning studies, and stored energy information that might be relevant te hypothetical graphite burning scenarios in nuclear reactors. Prior to a detailed analysis of the Windscale Accident, the British mis-takenly assumed that the accident might have been initiated by a stored energy release that took place during the anneal of the reactor. Subsequent work by both the team at Brookhaven National Laboratory and the British showed that this was not true, and that the accident was triggered by an uranium fire. In the Prime Minister's report to Parliament, [ Penney, 1957}, the following ! statement was made,
"...the most likely cause of the accident was the combined
. effect of the rapid (nuclear) heating and the high temperature l reached by the fuel elements in the lower front part of the pile. In all probability, one or more end caps of the cans of fuel elements were pushed off, and uranium exposed." As a result of the extensive full scale work carried out at BNL, a great deal of detailed information was developed on the factors affecting both the burning of graphite and the stored energy releases that occurred during anneals [Schweitzer, 1962c; Kosiba, 1953].
- 2. GRAPHITE BURNING For reasons that are well understood, graphite is considerably more dif-ficult to burn than is coal, coke, or charcoal. Graphite has a much higher thermal conductivity than have coals, cokes or charcoals, making it easier to dissipate the heat produced by the burning and consequently making it more i difficult to keep the graphite hot. Concomitantly, coals, cokes and charcoals develop a porous white ash on the burning surf aces which greatly reduces radi-ation heat losses while simultaneously allowing air to reach the carbon sur-f aces and maintain the burning. In addition, coals, cokes and charcoals are heavily loaded with impurities which catalyze the oxidation processes.
Nuclear graphite is one of the purest substances produced in massive quantities. 2
. j i
I The literature on the oxidation of graphite under a very wide range of' l conditions is extensive. Effects of temperature, radiation, impurities, por- ) osity, etc., have been studied in great detail for many different types of graphites and carbons [ Nightingale, 1962]. This information served as a foun- I dation for the full scale detailed studies on graphite burning accidents.in ) air-cooled reactors initiated and completed at Brookhaven National Laboratory j [Schweitzer, 1962a-f]. After British experimenters at Harwell confirmed the results obtained at BNL [ Lewis, 1963] there appecred to be no new cot lusions from additional work in this field. The aspects of the work pertiner to evaluating the potential for graphite burning accidents are described here in l some detail. i
)
Burning, as used here, is defined as self-sustained combustion cf graph-ite. Combustion is defined as rapid oxi4 tion of graphite at high tempera-tures. Self-sustained combustion produces enough heat to maintain the react-ing species at a fixed temperature or is sufficient to increase the tempera-ture under actual conditions where heat can be lost by conduction, convectian, and radiation. In the case where the temperature of the reaction increases, 4 the temperature will continue to rise until the rate of heat loss is just equal to the rate of heat production. Sustained combustion is disti'guished from self-sustained combustion when, in the first case, the combustion is sus-tained by a heat source other than the graphite oxygen reactions (e.g., decay heat from reactor fuel). Early attempts to model the events at Windscale [ Robinson, 1961; Nairn, 1961] were followed by the BNL work described here. Some 50 experiments on graphite burning and oxidation were carried out in I 10-fcot long graphite channels at temperatures from 600*C to above 800*C. To obtain a lower bound on the minimum temperature at which burning could occur, the experiments were specifically designed to minimize heat losses from radia-tion, conduction, and convection. The objectives of the full scale channel experiments were to determine j under what conditions burning might initiate in the Brookhaven Graphite : Research Reactor (BGRR) and how it could be controlled if it did start. Chan-nels 10-feet long were machined from the standard 4 in. x 4 in. blocks of AGOT2 graphite used in the original construction. The internal diameter of the BGRR channel was 2.63 inches. Experiments were also carried out on chan-nel diameters of one to three inches on 10-foot long test channels in order to obtain generic information. The full length of the channels was heated by a temperature controlled furnace and was insulated from conductive heat losses. At intervals along the length there were penetrations in the furnace through which thermocouple used to read the temperature of the graphite and air were introduced, and from which air and air combustion products were sampled. A preheater at the inlet of the graphite channel was used to adjust the air to the desired temperature. The volume of air was controlled and monitored by flow meters to allow flow measurements in both laminar and turbulent flow conditions.
- 2. Trade name for nuclear graphite used in the BGRR.
3
In a typical experimental run the graphite was first heated to a prese-lected temperature. The external heaters were kept on to minimize heat losses by conduction and radiation. The temperature changes along the ' graphite chan-nel were then measured for each flow rate as a function of time with the heaters kept on. It was observed that below 675*C it was not possible to , obtain temperature rises along the channel if the heat transfer coefficient ! (h) was greater than 10 " cal /cm-sec *C. Below 650*C it was not possible to get large temperature rises along the channel with 30*C inlet air temperatures at any flow rate. For h values lower than 10 " cal /cm-sec *C maximum tempera-ture rises were 0-50*C and remained essentially, constant,for long periods of time (five hours). For h values greater than 10 " cal /cm-sec *C the full length of the channel was cooled rapidly. There were two chemical reactions occurring along channels. At low tem-peratures the reaction C + O2 to form CO2 predominated. As the temperature increased along the channel C0 formed eitner directly at the surface of the channel or by the reaction CO2 + C. At temperatures above 700*C, CO reacts in the gaseous phase to form CO2 with accompaniment of a visible flame. It was observed that the unstable conditions which were accompanied by large and ) rapid increases in temperature involved the gas phase reaction CO + O2 and 1 occurred only for h values below 10 " cal /cm-sec *C below 750*C. Temperature ' rises associated with the formation of CO2 from C + O2 were smaller than those due to CO + O2 and decreased with time. They too occurred at h values below 10 " cal /cm-sec *C. In a channel which was held above 650*C there was an entr.mce region run-ning some distance down the channel which was always cooled. A position was reached where the heat lost to the flowing gas and the heat lost by radial conduction through the graphite was exactly equal to the heat generated by the ! oxidation of the graphite and of the CO. This position remained essentially 1 constant with time. Beyond this point rapid oxidation of graphite occurred with the accompaniment of a flame (due to the C0-0 gas phase reaction). Under conditions of burning, the phenomena were essentially independent of the bulk graphite chemical' reactivity. Rate controlling reactions during burning were 1 determined by surface mass transport of reactants and products. The experiments were used to develop an equation which expressed the length of channel that can be cooled as a function of temperature, flow rate (heat transfer coefficient), diameter and reactivity of the graphite. It was found that the maximum temperature at which thermal equilibrium (between heat generated by graphite oxidation and heat removed by the air stream) will occur in a channel can be predicted from the heat transfer coefficient, the energy of activation and a single value of the graphite reactivity at any tempera-ture. Above this maximum temperature the total length of channel is unstable and graphite will burn. The studies show that the bounding conditions needed to initiate burning are:
- 1. Graphite must be heated to at least 650*C.
- 2. This temperature must be maintained either by the heat of combustion or some outside energy source.
4
3
'~
- 3. There must be an adequate supply of oxidant (air or oxygen). j
- 4. The gaseous source of oxidant must .' flow at a. rate capable of removing-gaseous reaction products without excessive cooling.of the graphite surface.
- 5. In the case of a channel cooled by air these conditions can be met.
However, where such a configuration is not. built into the structure it is necessary for a geometry to develop to maintain an adequate flow of oxidant and removal of the. combustion. products from the , reacting surface. Otherwise, the reaction ceases. To illustrate how difficult it is to " burn" graphite thg following was excerpted f rom a report by Woodruff and Boger'. [ Reich,'1986] . These tests-were carried out in a search for methods'for extending the useful life of the-N-Reactor. (The following is quoted directly from text of the report.):
" Dry Burning Test: Three pieces of graphite were weighed and stacked together as indicated in Figure 1. Grafoil and carbon felt were placed under and around the blocks. This wrapping material was used as thermal insulation to hold heat in the blocks, and as a buf-fer to prevent catalysis by contact with the stainless steel tank used to contain the test. Thermocouple were placed at 5 locations in the blocks to monitor temperatures through the test. Two oxy-acetyleng torches delivering a combined heat output of approximately 2.7 x 10 BTU /Hr. through rosebud nozzles were positioned about 2 l inches above the graphite. Oxygen flow rates to the torches were Ierth lotetfonti i
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- 3. The receipt of this report f rom Mr. W. Quapp. of United Nuclear ;
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l I adjusted to produce nearly neutral flames. Still photographs and a ' ~ video tape were made to visually record the test. 1 "Five minutes after ignition, the surface of the top block in j regions directly below the torches was glowing yellow-white at an i estimated temperature of 1832*F (1000*C).
" Twenty-five minutes after ignition, the lower blocks were also j red over their entire surface. Block temperatures continued to rise ;
at rates of a few degrees centigrade per minute until fuel to the , torch over the thermocouple was shut off 57 minutes into the test. l The peak recorded temperature for thermocouple #1 was 2300*F i (1260*C). Other temperatures appear in Table 1. Using an optical i pyrometer, the blocks maximum surface temperature was estimated to I be approximately 3000*F (1650*C) directly under the torches. TABLE 1: FEAK TEMPERATURE DATA Thermocouple Dry Test TC #1 2300*F (1260*C) TC #2 2140*F (1170*C) l TC #3 1890*F (1030'C) l TC #4 1615'F ( 880*C) ' TC #5 1515*F ( E25'C) i FUEL AND BLOCK WEICHT DATA ) l 5 Acetylene Consumed: 13.0 lb (2.69 x 10 BTU) l 0xygen Consumed: 20.0 lb Total Block Weight Loss: 1.314 lb BTU /lb Weight Loss: 2.05 x 10 5 "With the acetylene to one torch shut off, oxygen was being blown onto the hot block at a rate of approximately 0.16 pounds per minute (1.9 cfm). The oxygen alone could not sustain a reaction with the graphite and the region below the nozzle cooled quickly. Sixty-five minutes after starting the test, both torches were removed, and the blocks were allowed to cool. When cool, the blocks were reweighed to determine weight loss.
"In the dry burn test, small craters were formed directly beneath each of the two torches. They are approximately 2 inches in diam-eter and their bottoms average 3/8 inch below the original graphite level. These craters account for only a small portion of the total weight loss. The remainder of the weight loss is the result of oxidation on the blocks surfaces that were exposed to air.
"In the interface areas where one block rested on top of or beside another, there are no visible signs of oxidation.
6
" DISCUSSION:
There is a common perception taken from our experiences with coal and charcoal that when a mass of these fuels achieves a glowing red condition a self-sustaining combustion is underway. Transferring this perception to graphite has led to repeated references to " burning" graphite when in fact a self-sustaining reaction was not in progress. The test sequences described in these tests demonstrate how difficult it can be to achieve conditions for self-sustained combustion of graphite."
- 3. STORED ENERGY 3.1 Summary i
A review was made of existing literature and knowledge on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650*C) are I pertinent. A review of existing information on stored energy has shown that stored energy releases do not occur spontaneously but are initiated by mecha-nisms that raise the graphite temperature above the irradiation temperature. Moreover, the maximum releasable graphite stored energy that could be produced by combustion from any reactor containing graphite is a very small fraction of the energy produced if graphite burning were to occur. Conclusions from these analyses are that the potential'to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite. I 3.2 Wiener Energy -- Its Generation and Buildup i From the earliest days of the Manhattan Project, E. P. Wigner [Wigner, 1946] recognized that if graphite was used as a moderator in nuclear reactors ! used to produce plutonium, "the collision of neutrons with the atoms of any substance placed into the pile (reactor) will cause displacement of these atoms. ... The matter has great scientific interest because pile irradia-tions should permit the artificial formation of displacements in definite num-bers and a study of the effect of these on thermal and electrical conductivity, tensile strength, ductility, etc. as demanded by theory." The theoretical prediction has been amplified by the work of F. Seitz [Seitz, 1958], the experimental work of Burton [ Burton, 1956) and many others. One of the many observed effects of neutron bombardment of graphite in slowing down the fast neutrons produced in fission to thermal energies is the production of large numbers of displaced carbon atoms and vacancies. Many of these displaced atoms of carbon come to rest in between the planes which constitute the structure of the graphite. The rest of the displaced atoms may 7 L_____-___________ _ _ _ _ _ _ _ _
. _ .~
k either wander back to their equivalent positions in the lattice, or to crystal boundaries. This introduction of new atoms between the planes increases the spacing between the original planes. This can be measured by the~ increase in the dimensions of the C-axis. This change in C-axis dimensions is reflected i by a change in the gross dimensions of the graphite specimen. Distortion of j the lattice reruits in an increased energy of the overall system. This j increase in la.tice energy is called the Wigner energy or stored energy. It was recognized that these two effects, dimensional changes and Wigner ; energy, might prove to be troublesome-in the operation of graphite moderated i reactors. The total stored energy of the graphite increases with neutron exposure and is a function of the temperature of the exposure, and the energy , distribution of the neutrons. The stored energy that can be released is I spread over a range of temperatures. It has been shown that when graphite j irradiated at moderate temperatures (less than 100*C) is heated above the i I irradiation temperature some of the stored energy.is released as heat when the temperature of the test specimen is rcised some 50-100*C above the irradiation temperature. Increases in exposure to fast neutrons increases the total energy stored. Eventually the stored energy which is releasable up to a tem-perature of 700*C saturates even though the total stored energy can continue to accumulate with increasing exposure. Total stored energy can be determined by combustion of the sample. Stored energy releases also can be measured by differential thermal analysis where the difference in behavior of an unirra-diated specimen and an irradiated specimen are compared in a calorimeter by increasing the temperature in a pre-determined manner. Broad experimental programs were undertaken during the Manhattan Project. This work was followed by basic and applied programs in the late forties and fifties. Much of this early work was presented at the first Geneva Conference on The Peaceful Uses of Atomic Energy held in Geneva in 1954 [ Woods, 1956]. By the early fifties it waa known that large dimensional expansions take place in reactor graphite structures and that stored energy accumulated. The British decided to control the stored energy of the Wind-scale reactor by heating up the graphite moderator (annealing). This process was carried out at regular intervals. The Brookhaven graphite gas cooled research reactor (BGRR) was annealed to reduce the dimensional changes ; (growth) caused by irradiation and to release the stored energy. Prior to i carrying out this work considerable experimental work was carried out to determine the rate of growth and the rate of buildup of stored energy as a function of irradiation exposure and temperature of exposure. A large body of complex literature exists on the accumulation of stored energy at different irradiation temperatures and fast neutron exposures. Much of this work is not pertinent to the problem of how tuch stored energy can be released below a given temperature. In this report we have analyzed existing , information in order to identify the factors needed to determine the quantity l of stored energy that can be released below the bounding temperature (650*C) needed to initiate graphite burning. l l l l l 8 l
The energy required to raise graphite from some initial temperature T' o to some higher temperature, T, is the enthalpy, which is calculated from the ; integral of the specific heat at constant pressure over the temperature inter-val of interest [ Schick, 1966). Consider a starting temperature of 30*C, and a final temperature of 650*C, the minimum temperature required for graphite to burn. The energy required to go from 30*C to 650*C is 202 calories per gram. Energies required to reach 650'C from various starting temperatures are shown below: Starting Final _ Temperature Temperature Enthalpy (C) (C) (cal /g) 30 650 202 50 650 105 150 650 175 200 650 160 Observed stored energy accumulation is non-linear, and depends upon irra-diation temperatures, levels of exposures to fast neutron fluxes, neutron energy spectra, spatial distribution of the flux, properties of specific graphites, geometries of individual reactors, etc. At low temperatures and at low exposures, the displaced carbon atoms move into interstitial positions [Kircher,1964; Schweitzer,1962a], and the re-sulting forces between these displaced atoms and planes in the lattice force the lattice apart, leading to expansions that are initially linear with fast neutron exposure. As neutron irradiati i continues, the number of simple ; i defects increases until they begin inte .cting and result in the formation of l l larger complexes (Schweitzer, 1964b]. Similarly, initial stored energy in-creases are linear with neutron irradiation, until a dose is eventually reached at which the stored energy tends to saturate. 4 Figure 2a shows that a sample exposed for 5000 mwd /AT at 30*C has a l total stored energy of 620 cal /g, but only 275 cal /g is released in annealing temperatures up to 800*C [Davidson, 1959, in Nightingale, 1962]. Similar results for other exposures and annealing temperatures up to 400*C are shown in Figure 2b [Kinchin, 1956]. l Results of calorimetric and heating experiments show that stored energy will not be released until the annealing temperature exceeds the irradiation temperature by some specific amount. This threshold temperature increase has been reported between 50*C to 100'C above irradiation temperatures [Kircher, 1964; Cottrell, 1958; Woods, 1956].
- 4. Units of neutron dosage are reported in different units by different authors. For this report we genera 1{y use the conversion one megawatt-day per adjacent ton [NWd/AT) = 3.9 x 10 thermal neutrons per square centi-meter [nyt(th)]. hi n, 1956] and Bridge
[ Bridge, 1962], we For use data from 1 mwd /ATKinchin
= 5.56 x[ Kin 10 19 nyt(th). For these data we were unable to obtain conversion factors for fast neutron flux.
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At irradiation temperatures above about 150*C the rate of accumulation of total stored energy is very low [ Bridge, 1962; Neubert, 1957; Nightingale, l 1958, 1962]. At about 30*C and at low total exposures, the total" stored energy increases with exposure at a near linear rate of about 40 10 cal /g per 100 mwd /AT. As the exposure continues, the rate of accumulation of total i stored energy decreases, and the stored energy that can be released below the ! minimum bounding temperature to initiate graphite burning (i.e. 650*C) satu-rates and then appears to decrease. An upper bound on the stored energy that can be released to 700*C can.be found from existing data. Figure 3 shows this as about 120 cal /g for an irradiation in the temperature 0 range of 35-70*C at an exposure of 930 mwd /AT (equivalent to about 3.6 x 10 nyt (thermal) [Neubert, 1957]. (This is about 1/60 the heat of combustion of graphite.) l 3.3 Stored Energy Releases A great deal of evidence exists demonstrating that stored energy is released through a series of complex and interactive thermally activated pro-cesses. Release of stored uergy is generically attributed to the recombina- , tion of various interstitial defects with vacancies, or the annealing of the interstitial to edge atoms or other voids in the graphite crystal. Removal l of interstitial species from between the graphite planes reduces the stored energy, lattice parameter increases, and other forms of radiation damage. I Existing views of irradiation changes in graphite support the claim that t irradiation produces different defects that thermally anneal with different activation energies (i.e. different energies are required to initiate the I releases). The type of defects and their respective quantities depend upon i the magnitude of the irradiation, the temperature of the irradiation, and whether or not the graphite was subjected to anneals, between irradiations. In the latter cases [Schweitzer, 1964a, 1964b] data show that defects interact with each other and that changes that occur during such anneals are very j different from the changes observed after a single irradiation. 1 At any given temperature the stored energy that can be released with time ] can result from several different processes whose rates decrease as the de- ! fects anneal. No evidence exists that stored energy releases are spontane- I ous. The observation that a 50-100*C increase above the irradiation temper- I ature is required to observe finite release rates is consistent with the exponential changes in release rates with reciprocal temperature associated with thermally activated processes. From our review of the literature on Wigner energy we have compiled data on releaseable stored energy at various combinations of exposures, and irradi-ation and annealing temperatures and have plotted this information in Figure 4 and Figure 5. In both figures a curve is shown of the amount of energy re-quired for a sample of carbon to go from 100*C to the particular temperature of interest (i.e., the enthalpy between 100*C and some temperature T). Also i shown are curves entitled " envelope of releases," which simply delineate an upper bound on stored energy releases found in the technical literature. Data l above the enthalpy curve indicate a region where a sample in an adiabatic environment would heat up to the upper intersection of the enthalpy curve and the envelope of releases. Figure 4 shows that the maximum releasable stored l l 1 12 l _ _ _ _ ___a
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energy in irradiations below 500 mwd /AT (irrespective of irradiation temper-atures) is sufficient to raise the carbon temperature from 100*C to about 450*C. Figure 5 illustrates the amount of releasable stored energy at exppI sures in the range 16-5700 mwd /AT [ equivalent to about 9 x 10 18
- 3.2 x 10 nyt (thermal)] and irradiation temperatures greater than 70*C. Figure 5 indi-cates that irradiations at 70*C or above (irrespective of exposures) have resulted in temperature rises from 100*C to no more than about 300*C.
3.4 Calculational Approaches
~
Buildup of stored energy in graphite is a result of~ the formation of a large number of ill-defined defects each of which can be associated with a stored energy release of unknown specific magnitude, unknown activation energy and unknown temperature range. Since the sum total of these defects deter-eines the accumulated stored energy and since this in turn depends upon the level of the irradiation, the temperature of the irradiation, and the history of irradiations and anneals, BNL does not.believe that any of the calcula-tional approaches. involved in the past UCLA license renewal hearings can be defended. Other calculational approaches such as'the bounding method used by Spinrad (Spinrad, 1986] rely heavily on a number of empirical correlations which involve appreciable uncertainties. These include determining the frac-tion of energy transferred to carbon atoms by neutron moderation that goes into atomic displacement energy. This must be combined with the fraction of stored energy that self-anneals at various irradiation temperatures. Aside from the direct dependence of this method on measurements showing a great deal
~
of uncertainty, these models cannot account for the non-linear buildup of stored energy, che saturation effects, the temperature dependence of releases, l the exposure dependence of releases and the complex consequences of I irradiations combined with several anneals. After review and analyses of existing information on estimating stored energy pertinent to graphite burning scenarios, we believe the approach pro-posed in this report is consistent with existing data and is acceptable for safety assessments. Total stored energy accumulation has no overall correla-tion with the stored energy that can be released at temperatures below 650*C. j The stored energy that can be released below this temperature saturates at a value that can be bounded from existing knowledge. The dependence of the sat-uration value of the stored energy released on irradiation temperature can also be bounded from existing data. This approach allows for safety analyses irrespective of the uncertainties in total exposure and total accumulated stored energy. We emphasize again, that the adiabatic assumption that all the released stored energy goes into heating the graphite is bounding but unrealistic. l Under adiabatic conditions where the decay heat is transfered from the nuclear I fuel to the graphite, steady increases in the graphite teraperature could occur that are much larger than those due to the hypothetical single spike f rom the release of stored energy. l I i 16 , J l
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Because heating graphite to at least 650*C is necessary but not suff'i-cient to initiate burning, the conclusion of these analyses is that the poten-tial to initiate or maintain a graphite burning incident is esse'ntia?.ly independent of the stored energy in the graphite.
- 4. THE CHERNOBYL ACCIDENT BNL has examined recent studies analyzing the Chernobyl accident to determine if any additional information on graphite burn _ing has been devel-oped. The accident summary described here has been taken from Kouts [Kouts, 1986]:
On April 25-26, 1986, "The accident took place during an experiment con-ducted at the start of a normal reactor shutdown scheduled for routine main-tenance. The operating staff had prepared to do what they considered to be a test of some electrical control equipment that was meant to serve a safety purpose." The objective of the experiment was to see whether the coastdown of the turbine of the nuclear reactor system would supply power long enough to allow for start-up of the standby diesels. The test required that the reactor power had to be reduced to a level (700 MW[th]) just above the value which was known to be low enough to become unstable. In approaching this level, a series of unfortunate operations were carried out in which many safety systems were intentionally by passed for unknown reasons. In one of these operations, the power level began to decrease rapidly, and fell to an estimated 30 MW(th) before the operator could halt the drop by control rod motion. After the operator had stopped the rapid drop, he managed to achieve some measure of control at 200 MW(th). At this point, the number of control rods in the reactor were far less than regulations permitted. Further manipulation of the cooling and feed-water systems resulted in other problems eventually leading to a rapid power surge estimated at 300,000 MW(th). Six violations of safety requirements, eventually resulted in a steam explosion that blew off the top of the reactor. The explosion disintegrated the fuel elements, fragmented the graphite, and exposed the graphite and fuel to air. The force of the steam explosion blew pieces of the core and fuel i through the roof of the reactor building. A second explosion lifted the cover l plate shearing the fuel channels releasing primary system steam pressure to the exterior. Falling hot projectiles ignited asphalt roofing materials causing exter.sive fires. , l Graphite burned for many days supported by aspha.1 t fires and decay heat from the buried fuel. Soviet teams tried to put out the fires by dropping massive amounts of materials from helicopters. The attempts were not success-ful presumably because the dropped material insulated the hot debris. Even-tually liquid nitrogen was used to cool and inert the burning debris. No evidence exists that stored energy in graphite played any role in this accident. 1 I l 17 l j _ _ _ _ _ _ _ _ _ _ _ _ _ _ b
- 5. " ACCIDENT AT WINDSCALE NO. 1 PILE ON 10th CF OCTOBER, 1957"5 Windscale Pile No. 1, was a graphite moderated, air cooled reactor, fueled by natural uranium metal encased in sealed aluminum cans to prevent the uranium from reacting with the components of the air and to contain the gaseous and solid fission products produced in fission. In 1952, the Wigner (stored) energy was found to be releasing on a shutdown of this reactor be-cause the graphite temperature rose above its normal operating temperature when the forced cooling was reduced on reactor shutdown.
To avoid a recurrence of such an incident the Windscale piles were there-fore regularly heated above their normal operating temperature to bring about a controlled release of the Wigner energy. The accident developed during the course of one of these controlled releases on October 7th, the day of the start of the Wigner release. Nuclear heating was used, but with cooling f essentially shut down to increase the temperature of the graphite above its normal operating temperature. In this instance the first nuclear heating was thought to have inadequately heated enough of the core graphite. To bring l about a more uniform temperature throughout the graphite structure the reactor was " pulsed again" but according to the investigators of the accident the rate of increase of nuclear energy input was too rapid, and caused the uranium cladding to break and expose uranium to air. Uranium is an extremely reactive metal. It reacts readily with oxygen, nitrogen, and hydrogen with the release of a large amount of heat. There is also the possibility that the initiating event in this accident may have been the failure of some aluminum clad magnesium lithium cartridges which were in the reactor at the time. The operator of the reactor was not aware of the cladding f ailure due to an inadequate number of thermocouple and inadequate radioactive sensing de-vices at the outlet of the cooling channels. Radioactivity sensing was done at a point some distance from the channel. Since the anneal procedure re- , quired allowing the heat to be conducted through the graphite structure by l maintaining the cooling shutdown for a day or longer the failed slugs heated adjacent ones and they too failed. Finally after a couple of days during I which the temperatures of portions of the reactor were noted to be rising, efforts were made to cool the reactor by admitting air. These efforts failed to cool the hot sections of the reactor. On October 10th a plug in the charg- ! ing wall of the reactor was removed. The uranium cartridges in the four chan- 4 nels which could be viewed were at red heat. Water was finally used to cool down the reactor after other efforts failed. l There is no evidence that stored energy releases initiated or played a q l significant role in the evolution of the Windscal? accident. ;
- 5. Title of a report presented to Parliament by the Prime Minister by command of Her Majesty, November 1957. Other sources on this accident " Final Report of the [ Alexander Fleck) Committee Appointed by the Prime Minister to Make a Technical Evaluation of Information Relating to the Design and Operation of the Windscale Piles and to Review the Factors Involved in the ,
j Controlled Release of Wigner Energy." Presented to Parliament by the ; Prime Minister by command of Her Majesty, July 1958. 1 l l 18 , I l
- 6. U.S. RESEARCH REACTORS 6.1 Criteria for Stored Enercy in Graphite Analyses of existing information indicate that the conditions associated with the initiation and maintenance of craphite burning scenarios are essen-tially independent of the stored enercy in the graphite, irrespective of its value.
As shown in Section 3, if the irradiation temperature of the graphite was ! 70'C or above, the maximum stored energy releasable below 650'C for any level ! of irradiation cannot raise the graphite temperature to the minimum value which would be required for initiating a self-sustained burning reaction. For graphiteirradiagontegperaturesbelow70'Ctotale>.posuresofabout500 mwd /AT (3.5 x 10 nyt) are required to continue to heat the graphite from about 100*C to 650*C if an external heat source can raise the graphite from its ambient temperature to 100*C. We have assumed that if the stored energy in the graphite cannot be bounded, any process that heats the graphite to 100'C should be treated as if it heats the graphite to at least 650'C. The analyses and conclusions on stored energy releases and graphite burn- f , ing conditions described above provide a meaningful method of categorizing ! nuclear reactors with respect to stored energy releases below 650*C (the threshold temperature for graphite burning) as follows: (1) Any reactor containing graphite in which the lowest irradiation is j 70'C or higher, can be excluded from stored energy safety concerns. l l (2) Any reactor in which the graphite is irradiated at te'.npe rature s below70'CbuthasreceivedatotalJastneutronexposurethat is much less than 500 mwd /AT (3.5 x 10 nyt) can be excluded from stored energy safety concerns. (3) Those reactors which have graphite that has received more than about 500 mwd /AT (3.5 x 10 1 nyt) of fast neutron irradiation below 70*C without thermal anneals or subsequent reirradiation at higher tem-peratures would require detailed heat transfer analyses to determine if the graphite were capable of reaching 650*C following an event that raised its ambient temperature to about 100*C. It is important 4 to recognize that even under conditions that allow the graphite to reach 650'C or above, this is not sufficient to initiate burning. In order to separate reactors into these categories, it is necessary to j determine only the total fast neutron exposure reached by graphites irradiated at temperatures below 70*C. 1 I
- 6. Estimated fast neutronfluegcewasconvertedtoMWd/ATusingthe )
conversion f actor: 7 x 10 A nyt = 1 mwd /AT. 19
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t 6.2 Stored Enercy in Graphite , The significance of stored energy for U.S. research reactors under NRC's licensing authority was assessed in light of criteria in Section 6.1. The information used in the assessment was obtained from Safety Analysis Reports (SAR's) and other readily available data representing the main types of these , reactors. The objective of the assessment was to determine if stored energy l releases can initiate or significantly contribute to the evolution of graphite l burning accidents, and if graphite wo'uld play a role in previously reviewed potential accident scenarios. For the purpose of overall screening of the research reactors, rough l estimates of the graphite exposure were made. Only operating research reac-cors containing graphite and licensed to operate at powers greater than 100 W were included in the survey. For TRIGA reactors GA Technologies publication GA-4361 [ West, 1963] was used to derive a maximum neutron fast flux (above 0.1 MeV) in the side reflec-tor. In addition, an analysis performed by GA Technologies [GA Technologies, i 1987] shows, for three out of the four locations where graphite is found in I the reactor (i.e., graphite reflectors in the top and bottou of the fuel ele- l ments and in the radial graphite reflector) that stored energy would not be l sufficient to raise the graphite temperature to 650*C. The reason for this is l that these locations satisfy, in essence, either criterion 1 or 2 in Section 6.1. The dummy elements, which are not in every TRIGA reactor, were found to have enough stored energy such that the graphite could reach 650*C if the tem-perature of the graphite is elevated to at least 120*C. However, no normal or abnormal operation would produce an initiation temperature of 120*C. Even if , this temperature were react'ed, water cooling of the aluminum clad surrounding i the graphite would preserve the integrity of the clad and prevent exposure of the graphite. Additional discussion on the significance of stored energy in TRIGA reactors is found in Section 6.3. The remaining research reactors were reviewed tc assess their stored energy accumulation. These reactors are listed in Table 2. Values of fast flux at the graphite were obtained from the licensees. Where licensee data were not available, peak fast neutron flux data for the reactor core compiled by the American Nuclear Society [ Burn,1983] were used, keeping in mind that the neutron flux that could be expected at a graphite reflector located close to the core would be about a factor of 2 to 10 lower. In the case of MTR reactors, the published data on power and fast flux in the ANS compilation were correlated, removing an outlier, to arrive at a flux-to power conversion factor. The total neutron exposure in some reactors was available from the licen-sees in terms of mwd of operation. In those few cases where these data were not directly available they were estimated based on data of first full power operation and reported equivalent days of full power operation for 1983. From the survey (see Table 2) it appears that four reactors (General Electric, North Carolina State University, University of Lowell, and University of Virginia) have stored energy greater than 500 mwd. However, the 20
.)
I presence of stored energy above the 500 mwd threshold in parts of the reac't6r graphite is not by itself taken as a safety concern, as discussed in greater i i detail in the preceding sections of this report and in Section 6.3. i i i Table 2. Stored energy calculations in graphite for non-TRIGA Research Reactors l
~
l - 1 Fast Irradiated l Power Duty Total Flux Dose Tenterature Reactor Identifier Type kW Year h/yr mwd n/cmsq/s nvt mwd /AT *C Ceneral Electric Co. Spec. 1.00E+02 100.0i 5.00E+11 4,3E+19 617 j Westinghouse Electric Spec. 1.00E+01 - 3.00E+11 - - i N. Carolina State U. Pulster 1.00E+03 403.0 1.30E+12 4.5E+19 647 ; Georgia It.st. Tech. MTR D2 0 5.00E+03 708.0 5.00E+10 6.lE+17 9 M.I.T. MTR D2 0 5.00E+03
- I M .no l National Bureau Stds.
MTR D2 0 2.00E+04 52013.0 2.00E+09 4.5E+17 6 Cintichem HTR 5.00E+03 1961 7800 42250.0 2.80E+08 2.0E+17 3 Ohio State U. HTR !.000+0! 1961 200 2.2 2.60E+11 4.9E+18 70 ! Purdue U. MTR I.00E+01 - - - Rhode Island NTR 2.00E+03
- 14H.00 U. Lowell MTR 1.00E+03 140.0 5.00E+12 6.0E+19 864 U. Missouri (Columbia) MTR 1.00E+04
- 100.00 U. Missouri (Rolla) HTR 2.00E+02 1962 62 12.9 4.86E+12 2.7E+19 387 U. Virginia HTR 2.00E+03 1702.0 3.50E+12 2.6E+20 3676 Worcester Poly. MTR I.00E+0! 1960 100 - - -
lowa State U. Argonaut 1.00E+01 - - - U. Florida Argonaut 1.00E+02 1959 213 24.9 1.30E+11 2.BE+18 40 U. Washington Argonaut 1.00E+02 1967 100 8.3 1.30E+11 9.4E+17 13 i NOTF.S t Year - Year of initial operation at (at least) one half of full power. Duty - Number of hours of operation per year, reported for 1983. Total - Total MW days of operation to date. Fast Flux - Peak fast neutron flux in the core or graphite reflector. Dose - Product of years of operation, duty, and fast flux. Represents maximum possible dose to any graphite. HWd/AT - Equivalent dose in mwd /AT. Factor 7e16 nvt = 1 mwd /AT. Irradiated Temperature - Normal maximum operating temperature of exposed graphite.
? - The graphite in the General Electric Co. reactor was annealed in 1976 when the reactor fuel container was i replaced for a leak in the weld area. Total HWd since that anneal is 44 mwd. l
- - Not significant because of low power.
* - Since irradiated temperature is above 70*C stored energy was not estimated.
l 21
l 6.3 Graphite Burning
)
J Research reactors which use graphite in or near their cores'and are licensed to operate at power levels greater than 100 watts (thermal) were categorized with respect to:
- 1. Quantity and location of graphite in and near the core,
- 2. Geometry, l
- 3. Accident conditions considered by the NRC staff ~in the licensing bases of the reactors, i
- 4. Fast neutron flux,
]
I
- 5. Normal operating sequence, and j i
- 6. Graphite irradiation temperatures. J Although present information indicates a great deal of variation in fast flux, operating sequences and graphite temperatures for reactors within a {
given type, our analyses of existing information shows that these factors are not significant to those factors related to graphite burning. In scenarios ], that postulate graphite burning, the quantity of graphite that can burn is an i important factor in determining the consequences of burning. However, the ) credibility associated with a postulated burning accident depends upon the existence of all of the conditions necessary for graphite burning, including the capability to heat the graphite to temperatures above 650*C and maintain-ing this temperature in the presence of much cooler flowing air. In any given reactor, this not only depends upon the original geometry, but also upon the 1 geometry resulting from the accident that allowed the graphite to heat up in I the presence of air. In assessing the potential for graphite burning in the research reactors licensed by h7C, consideration has been given to conditions during normal operation and conditions that may exist following a LOCA. The LOCA was j selected as having conditions most likely to result in high temperatures in I the fuel and graphite and, therefore, most likely to release the graphite stored energy and to result in conditions with the potential for graphite l burning.
]
All TRIGA reactors operate in water pools. Since graphite does not burn under water, all accidents in which the core and graphite reflector remain submerged will not be subject to graphite burning. GA Technologies [CA Tech-nologies, 1987] has estimated in a response submitted to the NRC on January ' 28, 1987 that aluminum clad graphite in dummy elements could, under loss of coolant conditions for some of the reactors, reach 770'C and result in melting of the cladding. GA Technologies claims that the hot graphite at 770'C cannot burn because the specific requirements for graphite burning cannot be met since the graphite radiates its energy rapidly and quickly cools to the ambi- ) ent air temperature. Our assessment of this claim is based on the experiments discussed in Section 2. That is, radiant heat losses to the cooler 22 1 L_____-__-____-__--___-__-__ _ _.
i surrounding structures coupled with convective cooling by the cooler air - ' surrounding the graphite could cool the graphite and preclude its. burning. Analysis of a LOCA in an Argonaut reactor predicts peak fuel temperatures of about 120'C [Chen, 1981). This, coupled with the insignificant stored energy of the graphite suggests no change in the conclusions already reached during the evaluation related to license renewal. The likelihood of graphite i fires was reviewed in NUREC/CR-2079 [Hawley,1981]. Reactors with MTR fuel and the PdLSTAR reactor have their fuel located in l a water pool. In accidents in which the water level in'the pool remains above i the core top the graphite could not burn. During a LOCA the maximum fuel plate surface temperature for any of these reactors is 500*C and for many it is much lower except for two cases where it has been calculated to reach 510*C and 582*C. In these two cases, however, emergency core cooling spray systems are activated during a LOCA and the actual fuel temperature would be much lower than the calculated fuel temperatures [NUREG 0928, Section 14.1.3, p. 14-3; NUREG 1059, Section 14.1, p. 14-2]. The stored energy is unlikely to raise the temperature to 650*C under non-adiabatic conditions that exist. Also, the graphite will not burn if the conditions to sustain burning are not present. If the fuel plate surface temperature is always less than 500*C, the heat losses from the graphite by radiation to the cooler structures of the pool coupled with convective cooling by the cooler air in contact with the graphite should preclude conditions necessary for graphite burning. The Safety Analysis Report [GE, 1981] for the General Electric Nuclear Test Reactor was reviewed for potential impacts of graphite stored energy on j the safety analysis of the reactor. The loss-of-coolant accident analysis in the report predicts maximum fuel temperatures of 300-320'c depending on assumptions about peaking factors. Such temperatures pose no danger to the aluminum clad fuel. However, there is no indication that the loss in thermal conductivity of irradiated graphite, or the releasable stored energy in the irradiated graphite, have been included in the thermal analysis. The reduced thermal conductivity could in principle lead to higher local graphite temper-atures which in turn could result in some stored energy release. Since in this postulated accident the graphite acts as an effective heat sink, the potentially higher graphite temperatures could have an impact on maximum fuel i temperatures. Without a numerical analysis accounting for the space depend- ) ence of the thermal conductivity, for the time dependence of the rate of i energy release, and for the concomitant changes in thermal conductivity of the graphite, it is not possible to estimate the impact of the irradiated graphite on the course of this postulated accident. However, in connection with Amend-ment No. 9 to the General Electric license, the NRC staff evaluated the conse- I quences of a postulated maximum hypothetical accident which assumed, nonmech-anistically, that all of the fuel in the core melted (NRC Safety Evaluation, ; Section 3.4, dated June 30, 1969). This scenario encompasses any potential ' impact of degraded thermal properties of irradiated graphite on the conse-quences of a loss-of-coolant accident. The resulting radiological doses to an i individual at the site boundary under the extremely conservative assumptions of the analysis were well below the allowable 10 CFR Part 100 guidelines. I 23
I I 1 The MTR-D 2 0 reactors have the graphite located away from the core, in a cavity with restricted air interchange. In the analysis of loss of-coolant scenarios of the SAR for the National Bureau of Standards reacto'r [NRC, 1983c], NRC staff agreed that a LOCA will not result in melting of the fuel. Under such conditions it appears implausible that the graphite could be ! subjected to temperatures compatible with burning. 4
- 7. FORT ST. VRAIN - GRAPHITE STORED ENERGY !
Fort St. Vrain operates at temperatures that preclude accumulation of j stored energy. There are no known problems associated with stored energy in- i graphite for operating temperatures associated with HTCR's.
- 8.
SUMMARY
8.1 Graphite Burning ) l The f actors needed to determine whether or not graphite can burn in air are the graphite temperature, the air temperature, the air flow rates, and the I ratio of heat lost by all possible mechanisms to the heat produced by the burning reactions [Schweitzer, 1962a-f]. In the absence of adequate air flow, ' graphite will not burn at any temperature. Rapid graphite oxidation in air removes oxygen and produces CO2 and CO which, along with the residual nitrogen, suffocate the reaction causing the graphite to cool through unavoidable heat loss mechanisms. Self-sustained rapid graphite oxidation cannot occur unless a geometry is maintained that allows the gaseous reaction l products to be removed from the surface of the graphite and be replaced by ' l fresh reactant. This necessary gas flow of incoming reactant and outgoing l products is intrinsically associated with a heat transfer mechanism. When the l incoming air is lower in temperature than the reacting graphite, the flow rate I is a deciding factor in determining whether the graphite cools or continues to heat. Experimental ctudies on graphite burning have shown that for all the geometries tested which involved the conditions of small radiation and conduct 1-on heat losses, it was not possible to develop self-sustained rapid oxidation for graphite temperatures below about 650*C when the air temperatures were below the graphite temperature. At both high and low flow rates, the graphite was cooled by heat losses to the gas stream even under conditions where other heat loss mechanisms such as radiation and conduction were negligible. < I At temperatures above about 650*C, in realistic geometries where radia-tion is a major heat loss mechanism, graphite will burn only in a limited range of flow rates of air and only when the air temperatures are high. At ' i low flow rates, inadequate ingress of air restricts burning. At high flow t rates, the rate of cooling by the flowing gas can exceed the rate of heat I produced by oxidation. l 24
l 1
. s l
Studies have shown that burning will not occur when there is no mech'anism to raise the graphite temperature to about 650*C [Schweitzer, 1962a-f). If the temperature is raised above 650*C, burning will not occur unless a flow pattern is maintained that provides enough air to sustain combustion but not j enough to cause cooling. Since the experiments were designed to minimize all heat losses other than those associated with the air flow, 650*C can be , considered a lower bound for burning.
]
8.2 Stored Enercy in Graphite
~
Fast neutron irradiation of graphite results in the development of stored (Wigner) energy. For a research reactor that has accumulated 30 cal /g of graphite after years of operation, this energy corresponds to about 1/250 of the energy released by combustion. Existing data show that for grapMte irradiated at temperatures of 30*C or above, the stored energy that on be released et 650*C saturates at a value that is less than 1/30 of the combustion energy. l ! . Analyses of the Windscale Accident and the Chernobyl Accident have shown 4 l that stored energy releases were not initiating events nor did they play any significant role in the evolution of the accidents. Although precise details of the buildup and release of stored energy vary with reactor geometry and factors relating to reactor operation, this review and analysis did not un-cover any substantiated evidence or credible scenario in which stored energy releases were responsible for an accident leading to graphite burning [ Fleck, 1958; Kouts, 1986]. In assessing the role of stored energy releases in graphite burning sce- , narios only the stored energy released below the burning temperature was con- I sidered pertinent. Stored energies released at or above the burning tempera-ture are a small fraction of the energy released by the burning process. A large volume of literature exists on the accumulation of stored energy at different irradiation temperatures and different fast neutron exposures. 1 l Total accumulation of stored energy is a complex phenomenon that depends upon I many factors related to reactor geometries, fast flux distributicas, graphite ; properties, reactor operating schedules and other conditions. At irradiation temperatures above about 150*C, the rate of accumulation of total stcred energy is very low with negligible releases occurring if the graphite tempera- ; ture remains below the graphite threshold burning temperature of 650*C. At ' about 30*C and at low total exposures, the total stored energy increases at a near linear rate of about 40 210 cal /g per 100 Mud /AT [ Nightingale,1962). As the exposure continues, the rate of accumulation of total stored energy decreases, and the stored energy that can be released below 650*C saturates and then appears to decrease (Nightingale, 1962; Neubert, 1957; Woods, 1956]. ' From existing data, an upper bound on the stored energy that can be released , below 800*C is 280 cal /g if the graphite was irradiated at 30*C. If the ' graphite was irradiated at 70*C, data indicate that the maximum stored energy releasable below 700*C is about 150 cal /g. The saturation value for an irradiation temperature of 135*C is about 50 cal /g released below 700*C, 25 l _ _ _ - __ _A
1 Although there appears to be significant differences in the estimates of total accumulated stored energy calculated in the past [Hawley, 1981; NRC, 1983a, 1983b}, these values have little relevance to graphite burning condi-tions. The total stored energy is always greater than, and is not directly 3 proportional to, the stored energy that can be released below the threshold j temperature associated with graphite burning. It requires about 200 cal /g of j stored energy to raise the graphite temperature from 30*C to 650*C if there are no heat losses. Similarly, it requires about 190 cal /g to raise the graphite temperature from 70'C to 650*C and 180 cal /g to raise it from 130*C to 650*C. The evidence on maximum stored energy releasable below 650*C shows that if graphite is irradiated at 70*C, or above, the maximum energy released below 650'C is not sufficient to raise the temperature to the burning tempera-ture even under the hypothetical conditions of a spontaneous release under totally adiabatic conditions. In an assumed adiabatic LOCA scenario, the decay heat in any nuclear reactor should be the major source for raising i graphite temperatures. The analyses and conclusions on stored energy releases and graphite burn-ing conditions described above provide a meaningful method of categorizing nuclear reactors with respect to stored energy releases below graphite burning temperatures: (1) Any reactor containing graphite in which the lowest irradiation tem-perature is 70*C or higher, can be excluded from stored energy j safety concerns. i 1 (2) Any reactor in which the graphite is irradiated at temperatures below 70*C but has received a gtal fast neutron exposure that is less than 500 mwd /AT (3.5 x 10 nyt) can be excluded from stored energy safety concerns. (3) Those reactors which have graphite that has received more than about 500 mwd /AT (3.5 x 10 1 nyt) of fast neutron irradiation below 70*C without thermal anneals or subsequent re-irradiation at higher tem-peratures require detailed heat transfer analyses to determine if the graphite is capable of reaching 650'C in an accident that heated it initially to about 100*C. We emphasize again that graphite tem-peratures exceeding 650*C are necessary but not sufficient conditions to initiate and support burning. In order to separate reactors into these categories, it is necessary to determine only the total fast neutron exposure reached by graphites irradiated at temperatures below 70'C. One pound of graphite releasing a stored energy of 200 cal /g is equiva- ~ ' lent to running a 100-watt light bulb for one hour. Recognizing that such releases cannot occur unless another energy source raises the graphite temper-ature above its operating temperature, spontaneous stored energy releases can- ; not be considered credible initiating events for graphite burning phenomena. ' Since the maximum energy that can be stored below 700*C is about 1/30 of the 26
j
* . .. 1 1
combustion energy, the single release of stored energy that might occur d'uring I a graphite burning accident is an insignificant portion of the total energy I released in the first few minutes of burning reactions. These. conclusions are consistent with analyses of both the Windscale and Chernobyl accidents. 8.3 Safety Assessment ! Consequences of graphite burning accidents depend upon the amount of graphite that can burn, and the inventory of radionuclides that can be re-leased. Both the amounts of graphite and the inventories of radionuclides in the Chernobyl and Windscale reactors were many orders of magnitude greater than in NRC-licensed research reactors operating in the U.S. Analyses of the actual reactor accidents in which graphite burning occur- ] red and analyses of hypothetical accidents show that some mechanism must lead ' to either fuel or graphite heatup under conditions where air is available. The review of a number of research reactors representing the various classes or types of research reactors currently licensed to operate in the U.S. (e.g. the TRIGAs, ARGONAUTS, PULSTAR, GE-NTR, MTR-D 2 0, and MTRs) found that under normal operating conditions their design features and/or environments should preclude graphite being heated to a temperature at which burning could be ini-tiated. In addition, under LOCA conditions it was judged to be plausible that the potential for cooling the graphite by passive means (e.g. radiation, conduction, natural convection) also should preclude graphite burning.
- 9. CONCLUSIONS i
After review and analyses of existing information on graphite burning, stored energy accumulations and releases, and causes of the Windscale and Chernobyl accidents, we have concluded that the above phenomena are suffi-ciently well understood to allow the following evaluations of U.S. research reactors and Fort St. Vrain. l The conclusions of these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored l energy in the graphite and depends on other factors that are unique for each l research reactor and for Fort St. Vrain. However, in order to have self-sus-tained rapid graphite oxidation in any of these reactors certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal and favorable heat balance must exist. The reactors considered in this review have all undergone safety evalua-tions and have been granted operating licenses by the NRC. There is no new evidence associated with the analyses of either the Windscale Accident or the
- Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review. Nor is there any new evidence that suggests that detailed case-by-case safety analyses of the role of graphite in NRC licensed reactors are warranted.
l 27 1
p a -.- , , j l
- 10. GLOSSARY BCRR Brookhaven Graphite Research Reactor .,
BNL Brookhaven National Laboratory BTU /Hr British Thermal Units per Hour cal /g Calories per gram i CBG Committee to Bridge the Gap CO Carbon monoxide CO2 Carbon dioxide FSAR Final Safety Analysis Report LOCA Loss-of-coolant accident mwd /AT Megawatt days per adjacent ton NRC Nuclear Regulatory Commission nyt(th) Exposure in terms of thermal neutrons per em2 02 0xygen SAR Safety Analysis Report 3 1 l l 1 2 28 i i
- 11. REFERENCES
[ Bridge, 1962] H. Bridge, R. T. Kelly and B. S. Gray, " Stored Energy and Dimensional Changes in Reactor Graphite," Proceedings of the Fif th Conference on Carbon, Volume 1, 1962. [ Burn, 1983] R. R. Burn, Editor, Research. Training, Test and Production Reactor Directory, American Nuclear Society, Second Edition, LaGrange Park, Illinois, 1983. [ Burton, 1956] R. Burton and T. J. Neubert, "Effect of Fast Neutron Bombardment on Physical Properties of Graphite: A Review of Early Work at the Metallurgical Laboratory," J. Appl. Phys, 27, 557-572 (1956). [Chen, 1981} W. Chen, " Final Safety Analysis Report Submitted to the U. S. Nuclear Regula-tory Commission in Partial Fulfillment of the Requirements for a Class 104 License," College of Engineering, University of Florida, Gainesville, Florida, January 1981. [Cottrell, 1958] A. H. Cottrell, J. C. Bell, G. B. Grenough, W. M. Lomer and J. H. W. Simmons,
" Theory of Annealing Kinetics Applied to the Release of Stored Energy From Irradiated Graphite in Air-cooled Reactors," Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Volume 7 (1958).
[Davidson, 1959] J. M. Davidson, " Stored Energy in Irradiated Graphite," US/UK Graphite Con-ference, Held at St. Giles Court, London, December 16-18, 1957. U. S. Atomic Energy Commission Report TID-7565 (Pt. 1), pp 11-20, March 16, 1959. [Emmons, 1965] A. H. Emmons, D. G. Fitzgerald and E. L. Cox, " University of Missouri Research Reactor Facility Hazards Summary Report, In Support of an Application to the United States Atomic Energy Commission for a Class 104 Utilization Facility License," University of Missouri, Columbia, Missouri, July 1, 1965. [ Federal Register, 1986] SIFR31341, 51:170, Wednesday, September 3, 1986. [ Fleck, 1958] Sir Alexander Fleck, J. Cockcroft, W. Penney, R. Spence, J. Diamond, J. M. Kay and H. W. B. Skinner, " Final Report of the Committee Appointed by the Prime FEnister to Make a Technical Evaluation of Information Relating to the Design and Operation of Windscale Piles and to Review the Factors Involved in the Controlled Release of Wigner Energy," presented to Parliament by the Prime Minister by command of Her Majesty, London, UK, 1958. 29 l l
[CA Technologies, 1987] GA Technologies, P. O. Box 85608, San Diego, California, 92138, Letter to U. S. Nuclear Regulatory Commission, GEN-1010, January 28, 1987. [GE, 1981] General Electric Company, " General Electric Nuclear Test Reactor Safety Analysis Report," NED0-12727, 80NED029, Class I, April 1981. [Hawley, 1981] S. C. Hawley, R. L. Kathren and M. A. Robkin, " Analysis of Credible Accidents for Argonaut Reactors," NUREG/CR-2079, Pacific Northwest Laboratory report (PNL-3691) prepared for D. S. Nuclear Regulatory Commission, April 1981. [Kinchin, 1956] G. H. Kinchin, "The Effects of Irradiation on Graphite," Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Volume 7, pp. 472-478 (1956). [Kircher, 1964] J. F. Kircher, and R. E. Bowman, Editors, Effects of Radiation on Materials and Components, Reinhold Publishing Corporation, New York, 1964 [Kosiba, 1953] W. L. Kosiba, G. J. Dienes, D. H. Curinsky, "Some Effects Produced in Graphite by Neutron Irradiation in BVL Reactor," Second Conference on Carbon, pp. 143-148, 1953. [Kouts, 1986] l H. J. C. Kouts, "The Chernobyl Accident," Brookhaven Lecture No. 227, BNL-52033, 1986. [ Lewis, 1963] l J. B. Lewis, "The Thermal Reactivity of Nuclear Grade Graphites to Oxygen," Progress in Nuclear Energy Series IV, Volume 5, p. 145, 1963. [Nairn, 1961] , J. S. Nairn, and V. J. Wilkinson, "The Prediction of Conditions for Self- I Sustaining Graphite Combustion in Air," Proc. of the US/UK Meeting on the l l Compatibility Problems of Gas-Cooled Reactors, TID-7597, 1961. 1 [Neubert, 1957] T. J. Neubert, and R. B. Lees, " Stored Energy in Neutron-Bombarded Graphite," Nuclear Science and Engineering, 2:748-767 (1957). 1 l [ Nightingale, 1958]
- l R. E. Nightingale, J. ?!. Davidson and W. A. Snyder, " Damage to Graphite Irra- '
diated up to 1000*C," Proceedings of Second United Nations International j Conference on the Peaceful Uses of Atomic Energy, Volume 7, 1958. ; [ Nightingale, 1962] 'i R. E. Nightingale, Nuclear Graphite, Academic Press, New York, 1962. 30
1 l1 l [NRC, 1983a] Nuclear Regulatory Commission, "In the Matter of UCLA Research Reactor (Proposed Renewal of Facility License)," Docket No. 50-1420L, page 1870, California, July 23, 1983. [NRC, 1983b] Nuclear Regulatory Commission, " Testimony of CBC Panel II - Chemical Reac-tions, In the Matter of UCLA Research Reactor (Proposed Renewal of Facility License)," Docket No. 50-142 OL, page 13, California, October 13, 1983. [NRC, 1983c] Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Renew-al of the Operating License for the National Bureau of Standards Reactor," Office of Nuclear Reactor Regulation, Washington, D. C., NUREG-1007, September 1983. [NRC, 1984] Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Renew-al of the Operating License for the General Electric-Nuclear Test Reactor (GE-NTR)," Office of Nuclear Reactor Regulation, Washington, D. C., NUREG-1069, September 1984. 1 [ Penney, 1957] Sir William Penney, B. F. J. Schonland, J. M. Kay and J. Diamond, " Accident at Windscale No. 1 Pile on 10th October 1957," Presented to Parliament by the Prime Minister by command of Her Majesty, London, November 1957. ) [ Reich, 1986] I F. R. Reich, and R. J. Nicklas, "An Evaluation of Methods for Reducing or I Delaying the Effects of Graphite Distortion to Extend the Production Life of the N-Reactor Core, Appendix E-2," United Nuclear Corporation, Inc., UNI 3680 l September 1986. [ Robinson, 1961] P. J. Robinson, and J. C. Taylor, " Thermal Instability Due to 0xidation of a Graphite Channel Carrying an Air Flow," Industrial Group Headquarters, Risley, Warrington, Lancashire, UK, IGR-R/W- 302; also in " Proc. of the US/UK Meeting on the Compatibility Problems of Gas-Cooled Reactors," TID-7597, p. 471, 1961. [ Schick, 1966] H. L. Schick, Editor, Thermodynamics of Certain Refractory Compounds, Volume II, Academic Press, New York, 1966. [Schweitzer, 1962a]
- D. G. Schweitzer, " Activation Energy for Annealing Single Interstitial in Neutron-Irradiated Graphite and the Absolute Rate of Formation of Displaced Atoms," Phys. Review, 128:2, pp.556-559, October 15, 1962.
[Schweitzer, 1962b] D. G. Schweitzer, "0xidation and Heat Transfer Studies in Graphite Channels IV," Nuclear Sci. and Eng., 12:59, 1962. 31 1
- a s
[Schweitzer, 1962c] D. G. Schweitzer, " Fundamental Studies of Radiation Damage in Graphite," BNL Lecture Series, No. 16, BNL-745, April 17, 1962. [Schweitzer, 1962d] D. G. Schweitzer, and D.H. Gurinsky, "0xidation and Heat Transfer Studies in Graphite Channels 11," Nuclear Sci, and Eng., 12:46, 1962. [Schweitzer, 1962e] D. G. Schweitzer, G. C. Hrabak and R. M. Singer., " Oxidation and Heat Transfer Studies in Graphite Channels I," Nuclear Sci. and Eng., 12:39, 1962. [Schweitzer, 1962f] D. G. Schweitzer, and R. M. Singer, " Oxidation and Heat Transfer Studies in Graphite Channels IV," Nuclear Sci. and Eng., 12:51, 1962. [Schweitzer, 1964a] D. G. Schweitzer, and R. M. Singer, " Removal of Radiation Damage from Graphite by Alternate Reirradiations and Low-Temperature Anneals," Nuclear Sci. and Eng. , 19, p. 385, 1964. [Schweitzer, 1964b] D. G. Schweitzer, R. M. Singer, S. Aronson, J. Sadofsky and D. H. Gurinsky,
" Decomposition of Defects by Neutrons in Reirradiated Graphite," Nuclear Sci.
and Eng., 18, p. 400, 1964. [Seitz, 1958] F. Seitz, and J. S. Koehler, "The Theory of Lattice Displacements Produced During Irradiation' Proceeding of the Second United Nations International Conference on Peaceful Uses of Atomic Energy 1958, Vol. 7, 615-633, USA, 1958. [Spinrad, 1986] B. I. Spinrad, "Storea Energy in Reflector Graphite of Research Reactors," correspondence to Nuclesr Regulatory Commission in reply to notice of rule making, Washington, D. L . 1986. [UCLA, 1981] University of California at Los Angeles, " Safety Evaluation Report Related to Renewal of the Operating License for the Research Reactor at the University of California at Los Angeles, Los Angeles, California, July 1981. [ West, 1963] G. B. West, and J. E.' Larsen, " Calculated Fluxes and Cross Sections for TRIGA Reactors," General Atomic, Division of General Dynamics, San Diego, California, CA-4361, August 14, 1963. ' [Wigner, 1946] E. P. Wigner, " Theoretical Physics in the Metallurgical Laboratory of Chicago," J. Applied Physics, Vol. 17, No. 1, November 1946, p. 857; also an address presented to the Am. Physical Soc. at the Chicago Meeting, June 22, 1946. 32
, :*s ,
)
[ Woods, 1956] W. K. Woods, L. B. Bupp and J. F. Fletcher, " Irradiation Damage to Artificial Graphite," Proceedings of the International Conference on the Peaceful Uses of l Atomic Energy, Volume 7, pp. 455-471, United Nations, 1956.
- 12. BIBLIOGRAPHY Beattie, J. R., J. B. Lewis and R. Lind, " Graphite Oxidation and Reactor i Safety," Proceedings of the Third United Nations Interna _tional Conference on the Peaceful Uses of Atomic Energy, Volume 13, Paper P/185, 1965. i j
Dalmasso, C. and G. F. Nardelli, "The Wigner Release in Graphite-Moderated I Reactors," Energia Nucleore, English translation in USAEC Report AEC-tr-4545, May 1961. Dickson, J. L. G. H. Kinchin, R. F. Jackson, W. M. Lomer and J. H. W. Simmons, "BEPG Wigner Energy Release," Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Volume 7, 1958. Fox, M. and R. W. Powell, "The Annealing of the Graphite Moderator Structure in the BNL Reactor, BNL-275, January 1954. Hawley, S. C., R. L. Kathren, " Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, NUREG/CR-2387, Pacific Northwest Laboratory report (PNL-4028) prepared for U.S. Nuclear Regulatory Commission, April 1982. Kosiba, W. L., D. H. Gurinsky and G. J. Dienes, " Evaluation of BNL Pile Graphite," BNL-255. October 5, 1953. Kosiba, W. L. and G. J. Dienes, "Effect of Displaced Atoms and Ionizing Radiation on the Oxidation of Graphite," Advances in Catalysis, Academic Press, New York, 1957. ' Lewis, J. B., P. Hawtin and R. Murdoch, " Thermal 0xidation of Nuclear Graphite," J. British Nuclear Energy Society, pp. 95-98, April 1964. Meyer, W. A., Jr., " Stored Energy in Irradiated Graphite," University of Missouri, Research Reactor Facility, December 10, 1986. Nightingale, R. E., " Record of Proceedings of Session E-21," Proceedings of Second United Nations' International Conference on the Peaceful Uses of Atomic Energy, Volume 7, 1958. Powell, R. W., R. A. Meyer, and R. G. Bourdeau, " Control of Radiation Effects in a Graphite Reactor Structure," Proceedings of the Second United Nations l International Conference on Peaceful Uses of Atomic Energy, Vol. 7, P/462, l USA, pp. 282-294, 1958. 33
1 y 's . l Rimer, D. E. and W. M. Lomer, " Calculations on the Release of Stored Energy in Graphite," Atomic Energy Research Establishment, liarwell, U. K. , A.E. R.E. M/R.2063, June 1958.
~
j Rimer, D. E., "The Validity of the Constant Activation Energy Model for the l Release of Stored Energy in Graphite," Atomic Energy Research Establishment, Harwell, U.K., A.E.R.E.-R.3061, August 1959. I Schweitzer, D. G. and R. M. Singer, "Effect of Irradiation Temperature and Annealing Temperature on Expansions and Contrac.tions in _ Alternatively Irradiated and Annealed Graphite," Trans. Amer. Nuclear Society, 6:2, pp. 383-384, November 1963. I Schweitzer, D. G. and R. M. Singer, " Anomalous Stored Energy and c-Axis Changes in Alternatively Irradiated and Anncaled Graphite," Carbon, 2:4, pp. 185-191, 1964. Schweitzer, D. G., " Determination of the Single Interstitial Migration Energy From Stored Energy and Thermal Resistivity Changes in irradiated Graphite," Carbon, 2:4, pp. 404-412, 1965. 1 l l 1
)
l l l i l l q 1 I 34
h Enclosure 4 1
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s _ e4 Mr. Steven Aftergood, Executive Director Committee To Bridge the Gap 1637 Butler Avenue #203 Los Angeles, California 90025
Dear Mr. Aftergood:
This letter responds to your petition for rulemaking, dated July 8,1986, to ^ reduce the fire hazard from nuclear reactor graphite. The petition requested that the Nuclear Regulatory Commission amend 10 CFR 50 to require licensees whose reactors employ graphite as a neutron moderator or reflector to formulate : and submit for NRC approval fire response and evacuation plans for combating a reactor fire involving graphite and other constituent parts (e.g., fuel). The petition also requested that licensees perform measurements of the "Wigner energy" stored in the graphite at their reactors, and submit these measurements to the NRC for review together with a revised safety analysis that addresses the risk and consequences of a reactor fire. You were notified of receipt of the petition and of the notice of the petition ; and request for comments published in the Federal Register. Copies of the j public comments were sent to you. After reviewing the petition, the comments i received from the public on the petition, our contractor's (Brookhaven National Laboratory) report on the subjects of stored energy in graphite and graphite burning, and the existing regulatory framework, the petition for rulemaking has been denied. The basis for this decision is set forth in the enclosed Federal Register notice of denial. We appreciate your concern about and interest in the health and safety of the
, public.
Sincerely, i Vc Executive Direc for Operations ,
Enclosure:
As stated I Sfjf Y
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