ML20070R225
| ML20070R225 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/24/1983 |
| From: | Potter T, Rodger W, Stratton W CONSOLIDATED EDISON CO. OF NEW YORK, INC., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20070R218 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8301270347 | |
| Download: ML20070R225 (107) | |
Text
{{#Wiki_filter:.. .. 1 UNITED STATES OF AMERICA
' NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD I
Before. Administrative Judges: James P. Gleason, Chairman Frederick J. Shon Dr. Oscar H. Paris l
)
In the Matter of )
)
CONSOLIDATED EDISON COMPANY OF ) Docket Nos. NEi' YORK, INC. ). 50-247 SP l (ladian Point, Unit No. 2) ) .50-286 SP ! ) POWER AUTHORITY OF THE STATE OF ) NEW YORK ) January 24, 1983 (Indian Point, Unit No. 3) )
)
l LICENSEES' TESTIMONY OF WILLIAM-R. LTRATTON, WALTON A. RODGER, AND THOMAS E. POTTER ON QUESTION ONE ATTORNEYS FILING THIS DOCUMENT: Brent L. Brandenburg Charles Morgan, Jr. Paul-F. Colarulli CONSOLIDATED EDISON COMPANY Joseph J. Levin, Jr. OF NFN YORK, INC. 4 T.rving Place MORGAN ASSOCIATES, CHARTERED New York, New York 10003 1899 L Street, N.W. (212) 460-4600 Washington, D.C. 20036 (202) 466-7000 v,.s-T
TABLE OF CONTENTS I. ' Presentation of Panel, Qualifications.................... 1 II. Initial Review of Indian Point Probabilistic Safety Study Source Terms................................ 5 Table 1a.................................... 7 Table 1b............................................. ......... 8 III. Indications _which Support Reduction in Source Term...... 11 A. History............................................. 11 Table 2............................................. 16-B. Post-TMI Developments............................... 20 IV. Chemistry of Important Elements......................... 27 V. Physical and Chemical Behavior of Key Elements in Post-LOCA Containment................................ 30 VI. Review of Experimental De ta. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 A.- Early Work.......................................... 37 B. More Recent Small Scale Experiments. . . . . . . . . . . . . . . . . 39 C. Containment System Experiments................. D. G e rm a n S t u d i e s . . . . . . . . . . . . . . . . . . . . . . . . .....
. . . . . 44
.........4 VII. Review of Accident Data................................. 46
, A. Windscale No. l..................................... 46 B. SL-1............................................. 47 Table 3............................................. ... 49 C. THI-2............................................... 50 Figure 1............................................ 51 Table 4............................................ 53 Table 5..................................-.......... . 55 VIII. Suggested Sourca Terms.................................. 57 Table 6........ .................................... 59 IX. Impacts of Suggested source Terms....................... 61 Figure 2........................................... 64a Figure 3........................................... 64b Figure 4........................................... 64c Figure 5........................................... 64d Figure 6........................................... 64e Figure 7........................................... 64f Figure 8........................................... 64g References.............................................. 65 Appendix................................................ 68 Table A-1............................................G9 Table A-2........................................... 71 Table A-3........................................... 73 Table A-4........................................... 75 Table A-5....................... Table A-6....................... ................... 77
................... 79
J I. PRESENTATION OF PANEL, QUALIFICATIONS My name is William R. Stratton. I received.a Ph.D. in Physics at the University of Minnesota in 1952 and was employe6 at the Los Alamos National' Laboratory from mid-1952' until February 5, 1982. Prior to. undergraduate and graduate studies at Minnesota, I served as a Naval aviator during World War II. During my career at Los Alamos, I have-worked in the-following areas related to nuclear energy: theoretical nuclear weapons design, criticality experiments and analysis, criticality safety, and fast reactor safety and analyses. In 1965-1966 I was the United States Atomic Energy Commission representative in the Cadarache Nuclear Laboratory in Southeastern France. I was a member of the United States Atomic Energy Commission Advisory Committee on Reactor Safeguards-for nir.e years (1967-1975), and I was Chairman during 1974. In 1978, I was a member of the team that assisted the Canadian Government in the recovery and analysis or the Soviet satellite and nuclear reactor that entered the atmo-sphere above Northern Canada in late January of that year. In 1979, I was a member of the staff reporting to the commiasion appointed by President Jimmy Carter to investi-gate the accident at Three Mile Island. This commission l l _ . _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _J
L
-:2-was1 chaired by John Kemeny,:then President of Dartmouth
~
l University. , In-the Spring of~1981,'I was presented the American 2 Nuclear Society's'"Special Award" and was recognized by the University of R 'consin (River Falls) as a " Distinguished i alumnus." lam a member.of the American Physical' Society, Lthe Society of Sigma Xi, the American Association for the Advancement of Science, and a fellow of the American Nuclear
-Society.
I have been= invited to speak at national meetings of the American Physical Society,-the American Nuclear Society,
-the Health Physics Society, the: American Chemical Society, and. international meetings sponsored _by the International Atomic Energy Agency (IAEA) and the United Nations (Second International Conference on the Peaceful Uses of Atomic
-Energy).
1 My.name is Walton A. Rodger. I received a Ph.D. in , . Chemical Engineering at Illinois Institute of Technology in , 1956, a M.S. in Chemical Engineering, and undergraduate
. degrees in Chemical and Metallurgical Engineering from the University of Michigan. My working career covers 42 years, 40.of them in the nuclear field.
I spent 17 years at three of the Atomic Energy Commis-sion National Laboratories, seven years in private industry 4
- , , - - . . -~..-.,.,-n -.-n...--,- , , .,.w-,., , , , , , . _ . , , , , - , , , . , - .,-g, _,,,_,,,.n,. n.., - , . ,,_,_-,,--e.,- -. - , , . - - . ,
- r. -
and 18 years in private consulting. My major fields of endeavor have been reprocessing of spent nuclear fuel, management of nuclear wastes in all aspects of the fuel cycle, isolation of specific fission products, radwaste systems for reactors, cost benefit studies, and source term analysis. la 1959, I served as Consultant to the Congress at hearings-held on. nuclear waste disposal. In the early 1970's, I servedLas a principal witness for the nuclear industry in the-RM 50-2 Hearings which led to establishment of Appendix I to 10 C. F.R Part 50. In 1979, I served as one of the "gegen-Kriticker" at the Gorleben Hearinge held in Lower Saxony to review the plans of the Federal Republic of Germany for reprocessing spent fuel and management of the wastes therefrom. Immediately thereafter, I served with the Three Mile Island Recovery Team with the special respon- '
.sibility of trying to minimize releases of iodine from the site.
In 1981, I was presented the American Institute of Chemical Engineer's Robert E. Wilson Award. I am a fellow of the Institute and a charter member of the American Nuclear Society. My name is Thomas E. Potter. I am a consultant at Pickard, Lowe and Garrick, Inc., in public health conse-quence analysis of radioactive releases. I was a principal
4-
. - investigator on the: Indian Point Probabilistic Safety
' Study. A statement of.my professional qualificacions-i.*
-- attached.
~
My. responsibility-in this testimony was to prepare risk estimates based upon source-terms developed by Drs..Stratton.and Dodger. More detailed statements of our professional qualifi-cations are attached. i - J p --, - -.- - . - - - - - - - ,c, - - -
II. INITIAL REVIEW OF INDIAN POINT PROBABILISTIC SAFETY STUDY SOURCE TERMS
~ This testimony sets forth our conclusions as to more realistic but still conservative assumptions about the
. amount and composition of radioactive materials that could reasonably be expected.to be released to the environment in the event of a serious accident at Indian-Point. This is commonly referred to as the " source term." We emphasize that some of the-retention factors we have used have been substantiated by experiments and calculations, while others are based on our judgment of the chemical and physical factors involved. Although the entire Indian Point Probabilistic Safety Study (IPPSS) has been-available to us, our review has.
.concent ra et d on-Section 5 and a portion of Section 6,
- relating to descriptions of accident sequences and the selection of. source terms for those-sequences.
As this testimony will detail, we find that the source terms which were used in the IPPSS, published in March 1982, are very conservative. Overly conservative estimates of the l magnitude of the source term released to the environment can I lead to inaccurate and misleading overestimates of the consequences of various accident sequences and to inappro-priate responses to these hypothetical accidents. 1
'our1 estimates-of the upper' limits of releases to be expected.from the two sequences which appear.to dominate the risk,.as described,in the IPPSS, are shown in Tables.la and lb,;for source term. categories 2 and 2RW, respectively, together.with the IPPSS estimates-for these same sequences >
. and the resultant reduction factor are also'shown for each
; isotope. .If our estimates of-the releases (which we believe.
- are;still conservative) were used in the IPPSS' consequences analyses, we would expect the consequence' analysis to show:
There would be no early fatalities from any accident scenarios that might occur at Indian-Point;
.The risk :of . latent f atalities is
-extremely small, and the consequences of even a " worst case"~ accident would tua 1 similar.to other,-large-scale industrial accidents.
' ~
~
There-would be a reduction-in property damage estimates by an order of magni tude ; .- Thefbenefit of additional mitigating devices would be reduced by a factor of 10; and A smaller Emergency _ Planning Zone may be justified, and the selection of protec-tive measures ~would warrant review. The IPPSS was started after the accident at Three. Mile Island (TMI), but before the current-state of understanding . relative to fission product chemistry and aerosol behavior became widely available and more generally accepted. As a 1
r_. - TABLE la Radioactive Release Fractions For Source Term Category 2 Reduction
~ Isotope' R/S IPPSS Factor
. Organic Iodide Neg1igible E-03 --
Iodine 1.5E-02 7E-01 47 Cs-Rb 1.0E-02 5E-01 50 Te-Sb 1.0E-03 3E-01 300 Ba-Sr 1.0E-03 6E-02 60 Ru- 1.0E-03. 2E-02 20 La 1.0E-03 4E-03 4 Noble Gases. 1.0E-00 9E-01 0.9 R/S -'Rodger/Stratton _
.IPPSS - Indian Point Probabilistic Safety Study Reduction Factor - IPPSS R/S
- s
TABLE lb
^
Radioactive -Release Fractions For Source Term Category 2RW Reduction Isotope R/S IPPSS Factor Organic Iodide Negligible 7E-03 -- Iodine 1.5E-02 lE-01 6.6 Cs-Rb' l.0E-02 3E-01 30 Te-Sb' l.0E-03 4E-01 400 Ba-Sr 1.0E-03 3E-02 30 Ru 1.0E-03 3E-02 30 La 1.0E-03 5E-03 5 Noble Gases 1.0E-00 9E-01 0.9 R/S - Rodger/Stratton IPPSS - Indian Point Probabilistic Safety Study Reduction Factor - IPPSS R/S f 4
9-i result, 'the source' term assumptions used in the IPPSS are closely related-to those used in WASH-1400, the Reactor Safety Study (n.SS75). . WASH-1400 assumed that the transport behavior of-iodine and cesium is such that iodine is present as molecular I2, that cesium is atomic, and that both, as wsil' as other fission products, behave essentially as noble gases. As a consequence, the IPPSS allows for practically
~
f no attenuation in the reactor vessel and primary system. Thus, the results in the IPPSS for some of.the potentially very severe accident sequences show little change from WASH-
.1400.
However, the assumptions in WASH-1400 and earlier docu-ments as to the chemistry of fission products are now believed to be incorrect. A major point is that in June 1981; the Commission agreed that the predominant form of iodine escaping from fuel is CsI (NUR81); concurrent studies have'shown that the remainder of the cesium will be dominated by the compound cesium hydroxide, CsOH. The fact that these two are the dominant forms will make a profound
-change in the attenuation-of fission products as they leave the core, pass through the upper plenum structure of the reactor vessel, through the primary system, then to the con-I tainment, and subsequently to the environment.
The cechnical knowledge developing since TMI and at an l accelerated pace during the writing of the IPPSS, and the l l new understanding now becoming more wiCely accepted, illus-l l
10 -
.trate an interesting and very important change in our knowledge of fission product behavior. There is little question but that the retention.of fission products as they move"from fuel to primary system to containment and perhaps -
to the environment will be very large. -The resulting reduc-Ltion in-consequences will also be-very large, as we have already illustrated. The uncertainties relate to the' magni-tude of the reduction, not to the direction of change. i i t 1
'III. INDICATIONS WHICH SUPPORT REDUCTION IN SOURCE TERM A. History Prior to World War II, the radioactive material posses-sed and used by man was almost entirely radium and its associated decay products and residues _of purification. The total amounts world-wide must have been measured in ounces. With the development of a nuclear industry during World War II, the amount of radioactivity created and handled must have been equivalent to hundreds of thousands of tons of radium (in terms of the hazards involved). This task was accomplished by creation and enforcement of strict and stringent safety rules resulting in, to our knowledge and with the exception of some criticality accidents shortly after World War II, no fatalities during this remarkable development.
Part of the tradition started during this development was the custo.a of overestiraating hazards by making bounding assumptions if complete data were not available. Some of this practice carried over into the Atomic Energy Commission (AEC) when it struggled with organizing both regulation and development of this new industry. At this time, serious consideration was given to the problem of establishing
. - rsquircm:nta-and rsgulations which would cdcquetely protect the health a.1d safety of the public.
WASH-740 (TPO57). In 1957, the first major attempt to assess the potential consequences of a major reactor accident to a commercial reactor was' completed at Brookhaven National Laboratory. .The objective of that study wae to provide an. estimate of the upper limit to the consequences (without regard to probabilities) that might be involved in such accidents in order to help the Congress ensure that legislation then being censidered to provide government indemnification of the public would be adequate, i Based on ultra-conservative and extraordinary assump-tions, very large consequences were postulated. For years, the response to this was not to question the validity of the assumptions, but rather to say the accident simply could not happen and to eliminate the so-called Class 9 accident from consideration. TID-14844 (TID 62). This document, issued in 1962, was the AEC's resolution of the regulation / promotion conflict And was adopted to specify the design, safety, and licensing requirements for reactors. A paraphrase of the assumptions proposed-therein follows: Given a severe accident to a nuclear plant, the reactor shall be assumed to be damaged to the extent that,
- a. 100 percent of the noble gases shall escape into containment, 4
i e -
- b. 50 psrcant of the-hologans shall escape likewise, but. half of this quantity will-be assumed to plate out, and
- c. 1 percent of the solid metallic fission products shall be airborne in the-containment.
It was generally agreed that.the presence of these quantities of fission products in the containment atmosphere would imply an accident very much worse than could be rea-
.sonably assumed, and if protection for this case were provided, a plant could be considered for construction and operation licenses.
These assumptions, like Oliver Wendell Holmes' one-hoss shay, were near-perfect in that both regulation and promo-tion could be accommodated with small penalty to either , effort. Unfortunately, the technical aspects of TID-14844 were not backed by adequate knowledge of fission product chemistry and physics, thermal-hydraulics, and aerosols. ! Further, the apparent near-perfection of the assumptions (for regulatory purposes) did little to stimulate investi-gation and consideration of the actual behavior of reacotrs during accident conditions. In fact, two decades of l application of this and related documents have created an air of realism and belief about the postulated escape of fission. products from core to containment. Even now, nearly four years after TMI, a residium of belief exists that I2 l i
h will occEpe from fual and will act during eccidents as do the noble-gases. WASH-1400 (RSS75). _This apparent belief regarding I 2 may have been reinforced by-the " Reactor Safety Study," . WASH-1400, published.in 1975, but available in draft-form for comment in.1974. This study evaluated risk for nuclear electric power stations. Risk required evaluation of both the consequences of the-accidents and probability of the accident. Generally, this massive and precedent-setting study showed_that unless fuel and' cladding both were damaged, primary system were damaged, and containment failed, the public health and safety consequences were very small and even negligible. Only if certain of the engineered safety features were to fail would the consequences to~the public be important. However, the
- probability of'such an accident was found to be very small l
and, hence, risk to the public was acceptably low -- i_.e_., significantly less than comparable accidents in other industries. In our opinion, because the probabilities of these very severe accidents were predicted to be so low, it was com-fortable to continue the tradition of using bounding assump-tions of little or no attenuation of fission products as they move from core to primary system to containnent and l possibly to the environment. L i
l An extremely important step in determining the public consequences of a reactor accident is to establish the
" source term," the amount and chemical form of the inventory of radioactive materials in the core which may be released >
1 to the containment atmosphere and then to the environment ' Table 2, and thus to members of the surrounding population. - abstracted from WASH-1400, gives the environmental source It terms used in that study for major accident sequences. the assumption, in is impoz/ tant to note that WASH-1400 made ' the case of an accident which involved breach of the containment, that materials which were once released from the core passed through the primary system into containment and thence through containment into the environment without It is this extra-any significant subsequent retention. ordinarily conservative assumption, an assumption which the IPPSS essentially repeated, which is now being seriously -' questioned by the body of knowledge being discussed in this testimony. TMI. At 4:00 a.m. on March 28, 1979, there began an event at the Three Mile Island Unit 2 which has profoundly affected the nuclear industry: the way the public perceives it, the way the Commission approaches their responsi-itself. The bilities, and the way the industry looks at loss of coolant accident (LOCA) which occurred was not the massive pipe break which in the past had occupied so much design and regulatory attention, but rather a much more
~ TABLE 2 WASH-1400 Release Categories Release Fraction of Core Inventory Released Ca t.egory Xe-Kr I Cs-Rb Te-Sb Ba-Sr g PWR . 0.9 0.7 0.4 0.4 0.05~ 0.4 PWR-2 '09
. 0.7 0.5 0.3 0.06 0, < 2 l-l t
4 . _ _ . , . .c._y.. . ..y-- , . - . - . _ ,c , - - .
--17. '
likbly l (in: rotrocpect) cmnller lone-of coolant through
-failure of a valvelin the primary = pressure boundary (with a concomitant but nonassociated' procedural failure in the auxiliary feedwater 9ystem). In any-event,-there was massive core damage (but, to our1 knowledge, rur fuel melting) 1and significant release of radioactive materials from the
. core. In' fact,.the core. releases materially substantiate the WASH-1400 core releases shown in Table 2 for the major accident sequences.- What was.not substantiated, in' fact what was clearly contraindicated, was the lack of attenu-ation of radionuclides other than. noble: gases in passing through the. primary system into containment and from con-tainment to the environment. Although.possibly as much as 13 million curies of noble gases were vented'from the system,-10. percent'of.the initial core inventory of-Xe-133, on.ly about 80 curies of I-131 were released to the environ-ment over.the entire course of the event and most of that came from the Auxiliary Building. Since about 64 million curies of I-lll were initially in the core, the retention-factor for iodine 64E+06/20 = 3E+06 is in striking contrast to the factor of about 1.5 used in-WASH-1400 and IPPSS calculations for risk-limiting sequences.
-Despite the very limited release of fission products other'than noble gases at.TMI, the:e was a reluctance at that time to consider seriously the concept of a limited accident.-- which'is exactly what happened. The emphasis 4
e
n ' invaricbly wac'"what-if this" or "what if thot" hopponed or
- did . not happen, with 'the implication that a very serious accident,- even a disaster, was close at hand and very r.early occurred. In fact, the accident was not close to. a more -
serious release of fission-productsLto the environment. KEMENY COMMISSION (KEM79, 79a). The two issues, the behavior of iodine and the "what if" matter were addressed by' the Kemeny Commission (appointed by President Jimmy Carter) in the staff document, " Alternative Event Sequences." This document derived from two closely related initiatives:
- 1. An early objective of the Kemeny Commission and its staff was to explain and understand the very low escape of iodine in view of the'large escape of noble gases, primarily' xenon. A subset of this study was an explanation of why none of the relatively volatile metals escaped. It was generally believed by the staff that if the behavior of iodine could be explained, understanding of cesium, tellurium, rubidium, etc., would follow without difficulty.
-2. A direct request came from the Kemeny Commission to'its staff to investigate the "what if" matter; the directive'was to examine'the sequence of events for
.the first few hours, to consider the operator actions or non-actions, and equipment working or failing, and for
- the significant cases, to assume the ~
reverse, and follow the now modified sequence of events. In this way, the staff was to determine if the plant was ever close to a more serious accident, and, if so, how serious.
Questions relative to the behavior of iodine under accident conditions were directed by the staff of the Kemeny Commission to chemists throughout the country. An implicit assumption was that the explanation would be related to molecular (gaseous) iodine, but the Oak Ridge National J Laboratory group that replied quantitatively developed, instead, the bases for the chemical state of the core and, hence, the chemical form (valence state) of the iodine (KEM79). This chemical state was reducing, as opposed to oxidizing, and hence the iodine would be driven to iodide (I-) and would most likely be associated with hydrogen (H+), cesium (Cs+), rubidium (Rb+) or some other metallic ion. Because the control rods contained several thousands of pounds of silver, a strong candidate might be silver iodide. Most of the metallic iodides (except AgI) ar^ highly soluble in water; once in solution they would stay in this state and would not become airborne with the possi-bility of a release to the environment. This description of the chemistry involved was so powerful it provided adequate explanation of the behavior of iodine observed at Three Mile Island. This same basic ' chemistry would more or less be involved in all light water
} reactor accidents. It appea s that molecular iodine or organic iodine molecules play little role in water moderated and cooled reactors. It developed later that the most
.20 -
probrblo compound of iodina, parhcps even in the fuol element, is CsI. This peint will be discussed later. The second request from the Kemeny Commission was addressed tar a concerted study. The result is best stated Izt quoting one of the findings from the Kemeny Commission's report to the President, one primarily concerned with the integrity of containment: If the high pressure injection system (HFIS) had not been turned on and if no heat sink were allowed, large scale fuel melting could occur throughout the core. This hypothetical situation was examined and bounded ,1n7 postulating a fuel-melting accident under total ab-sence of heat removal from the reactor vessel. . This study found that contain-ment would not be violated and opened to the environment by a steam explosion, by over-pressure, or by penetration of the reinforced concrete base of the contain-ment (basemat) by the action of molten fuel. Because the containment integrity would not have been violated, the re-lease of fission products would not be
~
changed by a large factor over what actually occurred at TMI-2. (KEM79). B. Post-TMI bevelopments At the conclusion of the study by the Kemeny_ Commission in October;1979, and the following Nuclear Regulatory Commission study led by Mitchell Rogovin (ROG80), it was apparent that the importance of chemistry in the study of
' reactor accidents had not been appreciated cad that the 'large, dry containment buildings were considerably stronger than was generally acknowledged.
I
However, additional generalizations to other reactor ( and plant designs and to other postulated accidents were In a real sense, a period of some months in ceveloping. l time to assimilate, understand, and appreciate these re s-il tively new ideas was necessary and this process is st l continuing. More than 20-years acceptance of the assumption 2 (and near-noble) that iodine should be treated as a gas (I ) is not easily instead of as a compound (CsI or other iodide) rejected and replaced by the new and chemically correct state of affairs. Before mid-1980, though, ideas were beginning to crystallize among various individuals and working groups f that the iodine matter was generalizeable, the chemistry o and for a large class fission products was_very important, some of the assumptions buried in of accidents (perhaps all) WASH-740, TID-14844, and WASH-1400 were not only highly Incorrect assumptions, as conservative but also wrong. opposed to merely conservative assumptions, could lead to the wrong design of engineered safeguards, wrong response to an accident, and a wrong plan for emergency procedures. This was an it portant matter and had to be brought to the attention of the Nuclear Regulatory Commissioners. This likelihood that serious errors were or had been built into existing dose calculations was called to the attention of the Nuclear Regulatory Commission on August 14, 1980, by Stratton et al. (STR80) and, on September 2, 1980,
- 22"-
indspsndantly by~Starr, Lsvencon, and Wall (STA80), and by
.Kouts (KOU80). ?The subject was discussed during a-meeting of the Commission on November 18, 1980 and in detail at an-AUS session on November 20,.1980, chaired I(M Professor Norman Rasmussen (ANS80). The papers were published in the
.-journal " Nuclear Technology" in May 1981. An additional review was completed by the Nuclear Safety Oversight
. Committee in their report to President Carter on December
':21, 1980. The response of the Commission was to order a review.of the matter.by their staff. This review led to the document " Technical Bases for Estimating Fission Product Behavior during LWR Accidents," NUREG-0772, published-in June 1981. This initial reponse by the NRC staff and.its contractors acknowledged.that the predominant chemical form 4 of iodine was cesium iodide, but continued to emphasize i uncertainties and extreme conservatisms associated with
' highly improbable early' failure accident sequences. Peer review of the document was generally critical and it will be re-issued in 1983.1 The NRC has a number of continuing studies, the results of which will become.available in the I .next year or so. . t
- 1. There has been a limited distribution of the first revisions of some sections of NUREG-0772 (designated NUREG-
- .. 0956) for review'and comment. At this time, we have been E
unable to review these documents. I r l l , , - . , ,
The. Department of Energy responded-to these develop-ments by creating a large. study group of experts from all facets of the nuclear industry. This group prepared the review document, " Source Terms: An Investigation of Uncer-tainties, Magnitudes, and Recommendations for Research," NUS-3808, October 1981. .This document notes the importance of CsI and CsOH and is an important contribution to the existing body of knowledge in this important area. Additionally, an industry sponsored activity known as the Industry Degraded Core Rule Making Program (IDCOR) is preparing its own-investigation of the source term issue. The Atomic Industrial Forum is the lead organization and various electric utilities, reactor vendors, architect l engineers, and the Electric Power Research Institute are , members and contributors. A serious in-depth study is being-conducted. Dr.-Rodger is responsible for a portion of.the l fission product transport section of this study and Dr. Stratton serves on the Technical Review Panel for this study. The Electric Power Research Institute has conducted its own studies and investigations during and since the accident ! at Three Mile Island. Its studies contributed significantly to the November 1980 meeting and to the momentum of research and study now under way. The American Nuclear Society started, in July 1982, its l L own review of the source term matter. An interim report is
. - ~ .. . . . . - . . - - - . . - . . _ _ - _ - ~ _ . . -
' Cch0dulcd fog May 1983. Dr. Stratton is chairman of. thio ;
committee and Dr. Rodger is a member. The International Atomic' Energy Agency has created a broad-based Technical Committee to study the source term . 4 matter and make recommendations to the Agency. The Commit-tee has . met1twice '(October 1981 and : August 1982) . Some of its conclusions'regarding fission product chemistry are l* I(IAEA81): Equ'ilibrium thermodynamic calcu-
,1ations for-reducing steam conditions in the primary system indicate.that CsI is the dominant iodine chemical form.
Similar calculations for cesium show that CsOH'and CsI are the most stable ' species. Tellurium is predicted to appear as elemental.Te at temperatures below 500C and as Te Te, and H2Te at - higher' temperatures.2,Since reaction rates are fast.at high temperatures, the equilibrium calculations may give a , reasonably accurate description of primary system chemistry. . Below about
^
500C, the vapor pressures of the I, Cs, and Te-species will be low and very-little'will.be in the gaseous state.
- 'If water is contacted in the primary
- , system, iodine and cesium will.be con--
verted-to non-volatile I- and-Cs+. Cs+ does not form volatile aqueous species 4 and will not be transported to the gaseous state. Iodine, in dilute solu-tions, exists primarily as non-volatile I and IO over a broad range of condi-
- . tions and3 little partitioning into the gas phase is expectedLif proper water
~
t chemistry conditions are established.
-In this regard, it was noted that dis-solved impurities were unlikely to lead to oxidizing conditions.
More general conclusions relative to very serious accidents are: There is general agreement that there will be reductions in the source terms developed in earlier US and German Risk Studies but there is also uncertainty in the amount of these reductions. For sequences in which water or wet steam is present in the flow path, or in which the containment remains intact, or in which failure is delayed for some hours or more, the reductica in source term could well be several orders of magni-tude. These reductions could be important in consideration of emergency planning and siting, but possibly would not significantly affect overall risks as currently evaluated. For sequences involving early containment failure, it is less clear that such significant reductions in source term will result from new assessments. However, even a breached containment may delay the release of aerosols for many tens of minutes and so give an order of mag-nitude attenuation factor; this is especially so if condensing conditions are predominant. Thus, it can be seen that there is a growing awareness , in the technical community that accident source terms used in the past (and in the IPPSS) are much too high. Clearly it has been demonctrated that if a core is allowed to get hot anough, whether or not melting occurs, nearly 100 per-cent of the noble gases and 50 percent of the cesiums and lodines can become volatile. There is little or no quarrel ! with the initial core release source term. However, the l proposition that these materials once vaporized are chem-l l ically inert and remain volatile as they encounter lower l l l
temperatures, cooler surfaces, water and water vapor, and particulates in the air as they make their tortuous way through the primary system and containment and possibly into the external atmosphere, defies not only the laws of physics and chemistry but flies in the face of evidence available from experiment and actual accident experience. That the presently used source terms are overstated is certain. The only uncertainty is the degree to which they are overstated. i: l l l t
}
IV. CHEMISTRY OF IMPORTANT ELEMENTS An examination of the precursors to the critical iso-topes-shows that within 0.5 hours after the accident, the critical elements have indeed achieved essential equilibrium and it is these elements with which we must deal. There-fore, the elements of most importance are: o Tellurium o Iodine o Cesium o Ruthenium o Cerium o Strontium (Yttrium) o Bariur (Lanthanum) o Antimony o Noble Gases Intuitively, one might feel that the time at which the release to the environment begins would be highly critical to the immediate dose to individuals. The data presented in Table 3 show that, over the time period from 0.5 to 24 hours after the accident, this is not no. Compared to the 3.5-hour release dose calculated in WASH-1400, a release begin-ning at 0.5 hour would produce an individual does at 0.5 mileo from the reactor -- only 6 percent higher than the 3.5-hour release. Conversely, a release beginning at 24 hours after initiation of the incident would produce a dose
cnly 17 percOnt loco then the 3.5-hour rolocos. Thic, of course, is based on the WASH-1400 assumption of essentially no attentiation of the core release source term. As dis-cussed herein, the provision of additional time for
' operation of the various gas phase removal processes to take place would be expected to even further reduce the 24-hour dose.' However, based only upon the processes of radioactive decay, the time period 0.5 hour to 24 hours produces less effect than one might intuitively expect.
Dr. Rodger has studied the chemistry of.these elements with particular emphasis on compounds which might be expec-ted to remain volatile. The basic conclusion is that at high enough temperatures some of the elements themselves and some of their compounds may be expected to vaporize. How-ever, tne lower temperature which will be encountered in any pathway which could reach the environment will result in the condensation of these materials. In particular, there appears to be ample evidence that the expected form of iodine will be as cesium iodide and not as elemental iodine. It may be noted that there are about 10 atoms of cesium in the core of a reactor for every atom l of icaine, and cesium iodide is thermodynamica11y the more stable chemical form of the two elements. Therefore, nearly all of the iodine may be expected in the form of cesium or i I L - l
othcr iodidac.1 Tho r;mmindar of the very activa cacium will form compounds with the hydroxyl ion or with oxygen. Cesium hydroxide is the most stable chemical form, second. only-to cesium iodide. The importance of the conclusion that the iodine is in
-the form of iodides, rather than elemental iodine, cannot be overemphasized. There are many more mechanisms.which will be operative in removing iodides.than would be-operative on elemental iodine. As a result, much larger removal factors for iodine can be expected than were assumed in the IPPSS.
t
- 1. Other iodides may also be expected, particularly AgI because there is a large amnount of silver in the control rods at Indian Point.
{ l l
V. PHYSICAL AND CHEMICAL BEHAVIOR OF KEY ELEMENTS IN POST-LOCA CONTAINMENT In hypothetical severe core damage accidents for light water-cooled reactors, it is to be expected that fission products "eleased from the fuel will undergo chemical and physical changes and will deposit on various surfaces after they are released from the fuel and cladding and as they are transported through the primary system to the containment. The escape of fission product activity under LOCA conditions requires successive releases from four physical barriers: (1) the fuel matrix, (2) the fuel rod cladding, (3) the primary system envelope, and (4) the containment system boundary. The release from the primary system has traditionally been assumed to be instantaneous and complete for all species released from the fuel. This conservative assump-i tion is derived from previous limited analyses which indicated that iodine should almost completely escape the j primary system under conditions of pure steam flow and high 1 surface temperatures (WASH-1400). To try to improve upon L
.this assumption and to include consideration of various other chemical species and possible physical states for l
these species, a study was set up at Battelle (GIE77) by the Commission to develop a more realistic primary system j fission transport and deposition model which is consistent l'
".with the' state of the released fission products and the thermal-hydraulic f actors due to ECCS operation. While.this study concentrated on~the primary system, Lit nonetheless covered 'in ogood detail many of the operative processes. We
-thereforeLborrowed liberally from that study.
.The.following. elementary mechanisms-can conceivably play'a role in fission product transport and-deposition I during a- light-water reactor . accident.
- 1. Phase changes of the-various. fission
, product compounds and of'the coolant.
Compounds whose melting or boiling point - is near the temperature of the transport fluid can experience a phase change as it heats up in the core or steam gener-3 ator and is cooled by the ECC water or in transport through the primary system into containment. Fission products in supersaturated concentration can con-dense on-solid surfaces,-liquid sur-faces,-particles, and water-droplets. Given sufficiently high. saturation, a vapor species may nucleate on ions that will be present in the steam phase due-toi the ambient radiation.
- 2. Chemical reactions among the fission products, chemical reactions between-fission products.and the system, and i chemical reactions solely among com- '
r ponents of the' system may occur.
- 3. Adsorption and Desorption. Fission products can be adsorbed chemically and physically on surfaces from the vapor and particle (liquid-or solid) phases.
They-can also desorb.- The rate at and degree to which they do so depends on the physical and chemical natare of the-surface and the amount and species of-fission product already adsorbed. 4, Solution in water. Solubility in water
- plays a role in the transfer of fission
product vcporo to water droploto.
~
It plays a significant role in the ' degree . to which~ particles grow as a result of water vapor condensation. In addition, particles that dissolve in the bulk water-may. experience a different trans-port.from the water to the steam phase during boiling than do those that remain intact..
- 5. Particle diffusion through a gas boundary layer. Because the primary system flow generally is turbulent, rapid bulk mixing will occur and dif-fusion of particles to surfaces will be rate controlled by Brownian diffusion through a laminar boundary layer. This effect is dependent on particle size.
- 6. Particle impaction on surfaces.
Inertial impaction is possible on obstructions to flow in the upper plenum, the upper annulus, on excessively rough surfaces, and in regions of sharp directional. changes in flow, such as the pump and steam gener-ator U tubes. This effect depends on the particle size and is critically dependent on assumptions or model calcu-lations of particle size and size dis-tribution.
- 7. Diffusiophoresis. Diffusiophoresis refers to the. transport of particles as a result of molecular dif fusional trans-port in the suspending medium. It may occur, for instance, in the: presence of a condensing fission product species.
- It . is expected to be insignificant except in connection with steam conden-sation.
- 8. Thermophoresis. Thermophoresis refers to the transport of particles in the
. presei.ce of a temperature gradient of the suspending medium. .The transport velocity is proportional, but opposite in direction, .to the~ temperature gradient. The principal effect due to thermophoresis is expected to be an enhancement of particle wall plate out.
. . _ . . _ ~ . .- - . - . . - .. - - - .
. 9._ . Sedimentation. . Particles can drop'out of fluid suspension and deposit on hor-izontal surfaces. iThe degree to which they do so depends on the particle size.
and_is therefore critically dependent on particle size assumptions and/or cal-culations.
- 10. Particle-particle agglomeration. The rate at w;11ch particles agglomerate' .
depends on the-particle concentration in :
.the gaseous phase. fit is anticipated that particle concentrations will.be too-small for.significant' agglomeration-to
- occur durin, the time required to flow lthrough a loop. Note,that this conclu-
~
! sion is disputed by work at-Karlsruhe i_
-(BUN 80)_which concludes that giant agglomerates will form-in milliseconds (see-Section VI). Note also that all of the mechanisms which depend on particle concentration will be greatly enhanced by the presence of particulate matter foriaad byJthe vaporization of structural materials. Such materials can be expected to exceed the actual fission product material several fold.
4
- 11. Reentrainment froml dry walls. Sticking probabilities will. depend on the nature of'the adhesion bond,.whether chemical or physical. Given the flow . velocities
- throughout the primary system, reen-trainment must-be considered for ,
physically bonded materials, o- Thermodynamic considerations and f- literature review'suggest that most ll > of L the fission products can be l expected to form compounds whose l boiling ~ points (bp) are
.sufficiently high that they are not likely to remain volatile for
- long. In particular the expected form of iodine is cesium iodide (bp 1280' C), not I2 (bp 184' C).
(AgI has.a bp of 1504' C). _
.o Deposition of vapor species directly onto solid surf aces occurs l'
-. ., . s -- . . , - . . -. .. - , , _ _ ,- . , , , . , . . . - - . - - - - -
r
~
and can be enaracterized by deposi-tion velocities. Genco et al. (GEM 69) showed that the deposition velocity for HI was one.to two orders of magnitude greater than
.for.I2+ The value for-Te was anotherJfactor of: ten higher.. CsI mayfbe expected to behave more like
.HI orHTe than I2*
o Gieseks (GIE77) has developed a code-called TRAPI which calculates the fraction of fission products escaping from the' core which will be held up in the. primary' system. A sensitivity study suggested that:
- The primary-system (water and dry interior surfaces) may be expected to retain.20 to 65 percent of the iodine (which they assuma to be elemental) after it is seleased from the fuel rods.
- About 99 percent of the re-leased cesium is predicted to be retained by condensation on interior primary system sur-faces of a PWR.
It should be noted that this study assumed cesium would be present as CsOH and would be perfectly soluble. Iodine, on the other hand, was assumed to be present as I2 and to With be only moderately iodine present. as. soluble. soluble CsI, it would be expected to behave like cesium hydroxide. Therefore this study can be interpreted to suggest that about.
- 1. The TRAP code has been modified since 1977. It is one of the computer programs used in the studies leading to the revision of NUREG-0772, referred to earlier. Indepen-dent . groups in industry have and are continuing to develop computer programs to' serve the same purpose as the TRAP code.
t
~
.99 percent of both cesium and i iodine would be rretained on 'the interior surf aces -of -the primary system. This would be consistent with the observation at TMI~that both cesium and iodine continued to i come out of the core.(or off-pri-mary surfaces) for,15 days-after the event.
.o Ins'ide containment fission. products can be. expected to be further removed by:
Agglomeration with water drop- < lets and condensation of steam
.on aerosol particles, and
- Gravitational deposition and
. ' deposition on interior contain-ment surfaces. o Particulates also are removed by sprays and by deposition on sur-faces. An important parameter :is the size of the particle to be expected. . For concentrated aero-sols agglomeration may be expected to be substantial and rapid.-
VI. REVIEW OF EXPERIMENTAL DATA Over the past quarter century or morc, a large amount of research'has been completed which aimed at understanding the subsequent behavior of volatilized fission product ele-ments. It is difficult to understand why more weight has not been'given to these efforts in establishing release source terms for radiological assessments of severe acci
'd en ts . A reasonable explanation is that until recently the use of TID-14844 assumptions has been so easy and obviously so very conservative that new, original -ideas for licensing requirements were not encouraged or needed.
A review of much of this literature has been carried out, and is briefly sammarized below, with special attention being paid to: aerosol behavior effect of steam or air as the carrier environment
- condensation and plateout of the released species.
In general, the results reported herein on both small and large scale experiments, as well as destructive tests on reactors, show that while fairly large fractions of many fission products can volatize from the core should cooling be lost (fractions similar to those assumed in reactor safety studies such as WASH-1400 and the IPPSS), these materials, with the exception of noble gases, do not long
remain in the gaseous phnsa and reloccas to the environmsnt as large as those postulated, even for a breached contain-mant, are certainly overstated in present safety studies. A.- Early Work The concept that Reactor Safety Study assumptions for release of isotopes to the environment are too high.is far from a new thought. Almost twenty years ago, Parker et al. , (PAR 63a) noted that the release fractions used in the licensing of light water reactors were rather conserva-tive. They argued even then that there are numerous influential factors which control the chemical state and amount of the release products and lead to their removal from the gas phace. Even earlier, Hilliard et al. (HIL61) studied the release of' fission products from irradiated normal uranium heateo at 1440' C. Experiments were conducted in air, i steam, and helium. Eight key elements were observed: iodine, tellurium, xenon, cesium, ruthenium, strontium, zirconium, and barium. At temperatures up tc 1440* C, after four hours of heating, the fission product release in steam was a factor of 2 to 10 lower than in air, while the release in helium was low. Iodine, tellurium, and xenon were . released in proportion to the amount of uranium oxidized. The release of strontium, barium, and zirconium was <0.12 percent under all the conditions tested.
Although the fuel form in the Hilliard experimer.ts was not U02 , the behavior of the fission products once released is of great interest. The following is quoted from the original work. Fission products that are released from a fuel will not leave the reactor quantitatively but will be held within the reacter core, biological shield and associated ductwork by -condensation and deposition. . . . A significant portion of the released material tends to deposit as soon as it leaves the heated area. Emphasis added. Browning et al. (BR063) melted ss-clad UO 2 in the atmo-sphere. Almost all of the I, Te, Cs and almost half of the Sr, 2r, Ru, Ba, Ce, and U were released from the fuel sample. However, escape of the latter isotopes from the high temperature zone (^-1000* C) was less than 3 percent. This important early observation shows that the volatilized materials condense rather quickly as soon as even modestly reduced temperatures are experienced, or conversely, that very high temperatures are required to maintain most of these species in the gas phase. l Still another early study at ORl1L (PAR 63b) showed that in. partially melted multi-pin experiments, only very minor j . amounts of particulate activity escaped the immediate furnace liner surrounding the experiment. This was in striking contrast to more commonly performed single-pin experiments which released relatively more then 100 times as l
- -139 -
much of the fission product invantory ec was obesrvcd in the
~
These results show that the unmelted multi-pin experiments. parts of the fuel and surrounding structure offer plate-out surfaces for the release of fission products. B. More Recent Small Scale Studies Recently, interest in this.20-year-old field is being revived. For example, cesium and iodine were observed to condense on surfaces in a 1980 ORNL study (LOR 80). This same study has confirmed che formation of CsI in the fuel, prior _to the release from the matrix. The formation of Cs2T e (FOR76, CUB 78) is also expected. In 1979, experiments at Rockwell International (MOR79) on high temperature, high concentration U02 aerosols demon-strated that the fallout of microparticles is generally characterized by two relaxation constants. The first is of the order of a few seconds, during which time more than 90 percent of the' mass of the airborne particles is removed from the air, while the second one is of the order of tens of minutes during which the fine particles settle out. It is observed that at extraordinarily high concentrations (70-1000 g/m3) the agglomeration is-so rapid (milliseconds) and the resulting particles are so large that giant agglo-merates (100-400 microns) are formed which will rapidly fall out under gravity. The effect is even_ greater in the presence of steam.
1 Experiments related to aerosol leakage (NEL75) have shown that .valls, irregular surf aces, structures, and cracks trap' aerosols. Any moisture present in the cracks would further collect the aerosols. Thus the process of removal in containment by sedimentation would be further' enhanced by these processes during escape _through containment. C. Containment System' Experiments During'a'six-year period, from 1964-1970,_ an important series of experiments called the. Containment Systems Experi-ment Program (CSE) was carried out at Hanford (HIL70). The CSE' facility was sized to represent a 1/5 scale model of a 1000-Mwe BWR. Many. runs wer'e made in the CSE and information perti-nent to predictionlof release source terms was developed in four important areas: '
- Leakage from containment vessels;
- Fission product transport in containment;
- Removal of fission product by sprays; and Operstion of filter-adsorber system.
Results of each of the above are summarized below. Leakage -- Air . Several experiments were performed during which the containment vessel atmosphere was heated, permitting the
~
effect of temperature on-leak rate to be measured. These tests showed that the leak rate decreased markedly as the l-
temperature was raised from room temperature to 107' C. It was deduced that expansion of the vessel with temperature closed off the leakage. paths. This suggests that the increased temperature of containment which would be expected to accompany a LOCA would work to reduce containment leakage rates. Leakage -- Air Steam Mixtures The presence of a condensable vapor may lead to forma-tion of liquid in the leakage path. Because the liquid viscosity is much higher than that for the gas, a marked reduction in the leak rate compared to the noncondensable gas results. Leakage ratcs measured at a known leak point fell drastically when steam was introduced into the vessel; in fact, they fell to immeasurably low values. It was concluded that under accident conditions leak rates will be less than those predicted on the basis of tests at ambient temperature. Leakage -- Fission Products The leakage of fission products from artificial leak points was measured during several of the natural transport experiments. Two important conclusions from these measure-ments are (1) there is a holdup of fission product activity in the leak path, aad (2) most of the leaked material is in water solution and thus unavailable for airborne transport. These factors introduce even more conservatism into dose calculations which neglect leakage path deposition. a
)
o l
. . . . I Fission Product Transport -- Natural' Processes
- - .Two natural process mechanisms play important roles in 1,
' depleting fission. products from post-accident containment a tmosphere s. One is surface deposition of. gases and particles. . Studies of natural transport processes. have
-shown than~the bulk'of the~ gas phase in a given compartment is.well mixed.. A boundary layer, driven by thermal and
. concentration buoyancy.at the wall, is the. locus of the mass
' transfer resistance.
The second mechanism is gravity settling of fission products present'as particles. While gravity settling is relativ'ely unimportant for submicron particles which may be produced initially by fuel overheating, agglomeration and
.w ater absorption cause rapid growth to much larger sizes. il For the larger: particles,fgravity settling can give appre- >
_ ciable removal rates. Because rapid agglomeration is expected.for concentrated aerosols, this.effect is most
-impo rtant.
Fission Product Removal by Sprays
- Sprays are used in most PWR containment buildings, ,
including Indian Point. Their most important function is to condense steam and thereby limit pressure buildup. This 1 t protects thefintegrity of.the containment vessel. A second important-function,is to remove fission products from the t
~
gas space. A great deal of literature attests to the
I efficacy of sprays in removing fission products from con-tainment air. The following conclusions are supported by the CSE spray experiments in which iodine was introduced as ele-mental iodine and cesium as the hydroxide: o Good agreement was obtained between experimental values of i..itial removal rate for elemental iodine and pre-dictions based on gas phase limited transfer and on.mean drop size. The measured concentration half lives of 0.6 and 2.0 min are in the range expected for large power reactor systems. o Iodine was not re-evolved from spray liquid which was recirculated for 20 hours at 250' F. Continuing chemical reactions in this liquid phase bind iodine in a nonvolatile form. It may be noted that CsI would be even less likely to be re-evolved and may well have been created during the experiment. o The 2-hour time-integrated dose reduc-tion factor (DRF) for a large PWR con-tainment is estimated to be about 50 for elemental iodine. The DRF continues to increase with time. o Aerosol particles are eftectively removed by sprays. Particle agglomer- ! ation and water adsorption ensure that l- particles generated under the maximum hypothetical accident case will be larger than those generated in the CSE vessel. Conservative use of CSE data leads to a 2-hour DRF of 12 for a large PWR containment vessel. l l L P l l I
German Studies D. Recent studies at Karlsruhe (BUN 80) have also ' investiga'ted'the validity of the presently used environmental source' terms. . They make the interesting ' ~ observation that'to be consistent with these source term assumptions, it would be necessary to vaporize from 1 to 2.3 tonnes of material.. This quantity would be made up of mostly structural materials (steel),_ fission products, and
~
.perhaps a little fuel. It would first have to be distrib-uted in the reactor vessel 'and then in the containment shell'
~
before release to the environment. Even if the entire primary system volume (v'350m.) 3 were available for the !' initial distribution of the aerosols,_the resulting concentration would be 7000 g/m3 Evenly distributed in the _ entire containment volume (e-57,000m3), the concentration
-would be 40 g/m3 The Karlsruhe studies, as well as work quoted earlier (MOR79), show that at these high concentrations, even in l air, agglomeration occurs in milliseconds-to form giant
! _ agglomerates . -(100-4 00' microns) which fall out almost immediately due to gravitational settling. Thus, if the quantities of fission products which are assumed to escape conta'.nment were actually vaporized in the first place, the res':lting concentrated aerosols would be promptly removed by [ natural processes. This will occur even in the absence of l-l
the moisture. When moisture is present, the aerosol depletion would be even faster. Baumgaertner and Heil (BAU79) arc melted simulated high-level waste solids and passed the resulting aerosols through a 1.5-meter long quartz tube which had a large thermal gradient ranging from greater than 1200* C at the furnace end to room temperature at about the 1-meter distance. Some deposition of the solids occurred within that portion of the tube which was in the furnace and essentially all had depositied by the 1-meter point. The final half meter of the tube was almost clear of any deposition. These experiments were done with no air flow through the tube. Subsequently, Dr. Baumgaertner repeated this experiment with air introduced into the bottom of the tube to produce an air sweep of 5 cm/sec and 10 cm/sec. The results were essentially identical (BAU82). Note also that in all the experiments and studies discussed in this testimony the highest concentrations used experimentally were 0.1 g/m3 Compare this to the lowest concentration predicted above (40 gm/3) which assumes distribution in the entire containment volume. It can be seen that the removals which can be expected from such cotuentrations should be much higher than these experimental results suggest.
VII. REVIEW OF ACCIDENT DATA - There have been some accidents at reactors involving significant' core damage where no appreciable amounts of radioactive material were released to the environment (MOR81). The following reactor accidents reaulted in radioactive releases to the environment. 1957 Windscale No. 1, England 1961 SL-1, Idaho 1979 TMI-2, Pennsylvania Both the Windscale No. 1 and SL-1 accidents occurred in noncommercial reactors. Neither of these two reactors had . containment buildings. Nevertheless, the escape of. fission products was still quite limited. In all these accidents, 'the. point of interest is the fractional inventory release, jl .;e . , the amount of radioactivity escaping relative to the radioactivity in the core. Details about these accidents are given below. A. Windscale No. 1 In October 1957, at Windscale No. 1 -- an aircooled reactor -- burning of the graphite and uranium core and lack of a containment system allowed the escape of fission products from the core and stack.- A part of the reactor continued to burn for more than two days. Almost all of the
iodine which existed in this part of the core was released from the burning fuel; however, only a fraction exited the stack. About 12 percent of the gaseous iodine was released . from the stack. Almost all the particulates were trapped by the filters (CLA74). The highest radiation level, at a single location 1 mile from the reactor, was 4 r.r/hr, The chemical environment (burning. uranium and graphite) is believed to have been oxidizing in nature instead of reduc-ing, as during the accident at TMI. B. SL-1. On January 3, 1961, the SL-1, a small BWR at the Idaho National Reactor Testing. Station, experienced a reactivity insertion accident. This was caused by the sudden with-ocawal of a control rod following a shutdown for mainten-aace. The subsequent. increase in reactivity resulted in a ! power. excursion and extensive core melting. Three employees l died due to physical injuries sustained in this accident.
- l. The-fuel that melted contained about 19 percent of the total core fission product inventory. However, despite the l
f l fact that the sheet metal building which housed the reactor l ! was "draf ty" and vented to the atmosphere, less than 0.1 percent of the non-gaseous inventory actually reached the l l atmosphere during the first two days. About 10 curies of i I-131 (0.04 percent) were released during the first 16 hours
- and an additional 70 curies (vs0.3 percent) were released r
i
during the~next 30 days (HOR 63). The release of Cs-137 was estimated to be 0.5 curie (0.02 percent) and that of Sr-90 to'be 0,1 (0.003 percent). These very limited environmental
-releases occurred in spite of the fact that the reactor i building was not designed as a containment building. In addition, most water was lost from the pressure vessel during the accident.- The' reactor building was filled with steam which leaked to the atmosphere through open doors in the control room and from the exhaust on the fan floor. The water expelled from the core inside the building was highly contaminated >ith fission products, including the radio-iodines. Nevertheless, only the limited releases described above occurred.
A particularly interesting comparison of the measured releases at SL-1 with those which would be predicted by WASH-1400 CORRAL (RSS75) . calculations has been made by Mendoza et al. (MEN 81). Their results, given in Table 3, show that' the actual releases were more than two orders of , l magnitude lower than the calculated releases. i-L
=.. - .. . _ . . . . _ _ - ..
49 - TABLE 3 SL-1: CORRAL Results Pbr Early Release Fission Curies Released to Atmosphere Ratio Product Calculated Measureda Calculated / Measured I-131 2150 10b 215 Cs-137 J17 0.5 234 Sr 29 0.1 290
=.
a Estimated fran post-accident envirorsnent surveys. b Released within about 1/2 day after accident
Reference:
MEN 81 l l
C. TMI-2 By far the most familiar reactor accident is that which occurred ct Three Mile Island Unit 2 on March 28, 1979. Extensive core damage occurred to that reactor and large fractions of noble gases, iodines, cesiums, and, to a lesser extent, other fission products were released from the core. Although about 5 percent of the Xe-133 core inventory was released to the environment over the first few days, less than 2 x 10-5 of the I-131 was released over a period of 30 days-(MOR81). No Cs-137 or Sr-90 was found in the environment beyond that due to weapons testing in China -- a tribute to the sensitivity of modern measuring techniques. There was no breach of containment at TMI during the accident but there was a limited bypass of containment. The releases to the environment which did occur were from materials which transferred to the Auxiliary Building prior to containment building isolation. Because there was no breach of containment, instructive information from TMI is obtained from reactor coolant samples and from containment concentrations of I-131, Cs-137, anu Cs-134 which was kept on a running basis by Dr. Rodger during the course of the post-accident recovery operations. Figure 1 shows that both aesium and iodine continued to come out.of the core into the reactor coolant for 15 to 18 days after the event, and that they ceased coming out at i
. . _ _ _ . . _ _ , _ . - , _ ,,___.m __ _, _ , . - ,
L Ficure 1 Reactor Coolant Concentration Follo' wine the Dien: at ?.ree .v.ile Island Unit 2 tss - j t. ..r c..ta. ':...., s. :4 .ser. s... .. :1
- hree . Mile Island "Jais :
1
,,. > h
- Pressurtser Samples 9 s
N'q.- .- s .~ 's,
*4 .
. , , 8 .
~~.
'% % :D-Day % 1f-Lifa (Feet & $1e.4) s
'\'-131 3
j N
$ N'%N- \
2t
;ts p_. s N.s\,
!-.\,N
-{ . N* N~t20-0av \ I-!ay 41!41!e
- 04-131 __ . ' italf-Lif e 9.
3- - N. K NN - l ;1 -- - s4x 5 4 g .
~
, 23-Day italf-Life i
...,,. N.
~
= -
\.
\ .
N.
\.
t':
. . . N l ,. to :. 3. i , : ,o to io .o ==
u:,, - .. w , .. a n. ., w,, a s::a, u w:, =;:, uws .~u - u=> ~, Dave A! er :Me Ivet t I l
about_the same time, which even at the time suggested CsI as the probable chemical form. While there is still some uncertainty concerning the exact aucunt of the various isotopes which were released from the core,1 these core releases clearly were of the order of those predicted by WASH-1400. As indicated above, however, releases. to the environment were very much lower than would be predicted by WASH-1400 methodology. This is illustrated by comparing'the ratios of various isotopic releases to those of noble gases for the range of accidents considered to the same observed ratlos at TMI. The com-parison of the ratios predicted to those observed are shown in Table 4. It can be seen that for-iodine, WASH-1400 assumptions would produce doses a minimum of 1000 times too hig'h and more likely 100,000 times too high. We have not found data for other isotopes which permit quantitative ca]- culations. It is reported (KUM79) that neither cesium nor strontium were detected in the environment above weapon fallout levels, strongly suggesting that these-would also be i overpredicted by even larger factors. For Te, Ru, and La, there seem to be no data at all. However, had they I
- 1. The uncertainty lies in how much water went through the primary system and onto the containment floor between the time significant ccre damage began (probably about 3 i
hours after the event began) and when the first reactor t coolant sample was taken the afternoon of the next day. l l
~ _ ._ _ _,
TABIE 4 (bmparison of Predicted (WASH-1400) and Observed ('IMI) Release Ratios of Isotope / Noble Gases Isotope WASH-1400 Pred Ratio Ratio Ratio Cbserved: Predicted Range of Average Iowest Es7v Based on Based on Over-for IHR for PWR at '1NI Average PWR Iowest IHR_ Prediction I 0.34 3E-03 4E-06 1.2E-05 1.3E-03 lE+03 to lE+05 Cs-ab 0.22 3E-03 Te-Sb 0.20 3E-03 tm - - - Ba-Sr 0.027 3E-04 ' Rua 0.085 2E-04 tm - - - Lab 2E-03 2E-05 im - - - a Includes Rh, Co, Mo, 'It b Includes Y, Zr, Ib, Ce, Pr, Nd, Pu, Am, On NR No record i 7 M'
been present in a detectable quantity, it is reasonable to assume.that they would:have been observed and reported. Clearly, release ratios much lower than predicted with present source terms were experienced at 2MI for all isotopes.- Another way to approach this same point .is through containment air samples.taken during the course of the event. We have only becn able to find data on Kr-P5, Xe-133, I-131, and a very little bit on I-133 and Cs-137 (DAN 80).- The ratios of these isotopes to noble gases are shown in Table 5. -Again, the extreme level of overpredic-t tion is clearly evident. These data strongly suggest that t-the initial airborne concentrations in containment for , Jodine, certainly, and for cesium, probably, were low enough that even had containment been breached the releases of these.ieutopes would have been lower than calculated by the use of WASH-1400 source terms by at least a factor of 1000 and more likely by a factor of 10,000. l Very generally, the history of the accidents, inci-dents, and experiments supports the theses developed during the THI-2 investigations, namely, that the chemistry is very important, and that the significant chemical forms are CsI, I Cs0H, and tellurium metallic vapor. The first two of these t
.are very soluble in water and, once in solution (condensing l steam, for example), they stay in solution. Oxidation to a l
[ gaseous I l i t
55 -
- ' TABLE 5 Inferred to 'IMI Containmtent Air Concentrations Extrapolated _t o Total Quantity Core Fraction Ratio
. Containmenc Air in Cont Air Inventory (bre Inventory Isotope to
' Isotope Conc uCi/cc curies curies in Cont Air Noble Gas Kr-85 1 .5.7E+04 9.6E+04 0.58 1 Xe-133 1500 8.SE+07 1.5E+08 0.59 1 I-131 0.1 5.7E+03 6.4E+07 8E-05 1E-04 I-133 0.25a ~1.4E+04 1.5E+08 8E-05 1E-04 Cs-137 0.05b 2.8E+03 8.4E+05 3E-03 SE-03 Cs-137 SE-05c 3 8.4E+05 3E-06 SE-06 . .a Based on a single measurement taken on 3/31/79 at 0700.
b Based on a single sample taken on 4/2/79 c Based on the. highest of four cubsequent-samples. )
form is very slow. Tellurium vapor has an extraordinary affinity for steel, iron, and zirconium.(PAR 82). It it should be vaporized, little would be expected to escape even the core region, much less the reactor vessel. Cesium hydroxide is stated to be one of the strongest bases known and vigorously attccks other materials, even glass. It is expected to plate out in the reactor vessel or be dissolved in water.
.Thus, the general conclusion of this section is that
'the history of licensing and regulation shows that con-servative, but some fundamentally wrong assumptions led to designs that adequately protected the health of the public, but which were so easy to implement that they led to incor-rect understanding of the nature of severe core damage reactor accidents. The accident at Three Mile Island has triggered a veritable avalanche of studies and experimental prog rams . The results of all studies of which we are aware conclude that the very large escape fractions of the " risk dominant accidents" in WASH-1400 are in error on the high side (except for noble gases, of course, which are chemi-cally inert). The only debatable point is the magnitude of the reduction in the escape fractions, i,.e., 1/10, 1/100., 1/1000, etc. More recent evidence has been in the direction of verification and support, rather than of any contradic-tion.
VIII. SUGGESTED SOURCE TERMS
' ThelIPPSS considers 13 release categories. The source terms for these' release categories are given in Table 5.4-2 of the IPPSS.- Two of the release categories -- 2 and 2RW --
represent the largest contributors to risk. For these two categories we have developed, based upon the principles and 2 information included in the previous sections of this testi-mony, revised source terms which we believe to be much more representative of what could reasonably be expected if such sequences were to occur. In developing these revised source terms, attenuation for each element through various barriers is considered and a decontamination factor (DF) for each mechanism is assigned; the net attenuation is the product of attenuation by each physical or chemical barrier. The available barriers between the core and the environment include:
- 1. Retention in the core material itself;
- 2. The primary system envelope:
- a. retention on primary surfaces,
- b. transfer into primary coolant; l-
- 3. Removal within the containment system:
- a. by natural processes,
- b. by action of Emergency Core Cooling System; and i
- 4. Retention on passing through the l-
- containment barrier into the environment.'
.Tne detailed development of these revised source terms is given in 12 tables included as Appendix A of this testi-mony. In-these tables, for each of the three categories and for six isotope classes, the following are given:
- 1. Time elapsed since incident began;
- 2. Step-by-step description of an accident sequence-(as described in the IPPSS) which leads to this release category;
- 3. Expected behavior of.the particular isotope' class during this portion of the sequence;
- 4. Fraction of the core inventory which is airborne in containment air; and
- 5. Fraction of the core inventory which escapes to the environment during each time interval.
The total release for.each sequence is the sum of the releases identified at (5). In developing the revised source terms, we started wf.th the releases shown in IPPSS. Table 4.3 and used them as releases from'the core. The only modification was that the
" vaporization" releases assigned to I2, Cs, and Te were included with the melt releases because all of these more volatile elemer.ts will be released well before the " vapor-ization" release can occur.
Our revised source term estimates are summarized in Table 6 -- which is a revision of part of Table 5.4-2 of the IPPSS. Throughout this testimony, we have discussed the
Table 6' Radioactivity Release Fractions for Source Term Categories (Fraction of Core Inventory Release fran Contairunent) Fission Product Class Designations Organic For This Study Iodide .Iodinea Cs-Rb .Te-Sb Ba-3r g g Ibbles 2 neg 1.5E-02 lE-02 lE-03 lE-03 lE-03 lE-03 lE-00 21W neg 1.5E-02 lE-02 lE-03 lE-03 lE-03 lE-03 lE-00 a Iodine is taken to behave the same as Cs. 4
- < - - - , m - _ . . . . . , -
2 m A _ 60 - evidence supporting the appearance of iodine in the. form of CsI under. accident conditions. The. behavior.of iodine will be the:same as that of.CsI, and we have used this compound
- to characterize the behavior of both-iodine and cesium.
4 4 i t i l 6
-'-~. .- -, , , , , . . ., ,
r .. L e I X .- IMPACTS'OF SUGGESTED SOURCE TERMS
As stated in testimony under Contentions _2.1(a) and 2.l(d) and elsewhere under Commission Question 1, smaller
. releases _of radionuclides than those used in the IPPSS would have a significant impact on risk and on emergency protec-tion measures. .These other pieces of testimony presented sensitivity studies on risk and on emergency protection measuresL where the IPPSS fission product release fractions were reduced by factors of 10 and 20, with.the exception of the noble gases which were not changed. The approach here is similar, except that the suggested source terms in Table
- 6 are.used.
The sensitivity analysis results in this section are intended to illustrate relative-risk. Except as noted otherwise in'this testimony, these results are based on input assumptions used to produce the point estimate site matrix and are comparable to results in the IPPSS, Figures 6.4-1 through 5 and Tables 8.5.2-4a through e. Mean release frequencies were used to convert conditional risk estimates
. to absolute risk format. Expressions of uncertainty are not included in these results. These uncertainties would not affect the conclusions drawn.
Four major areas are examined with these suggested source. terms: early fatalities, latent fatalities, man-rems, and property damage. 1 ( - , , ---- .
- 1. Early Fatalities Figure 2 is a graph of frequency versus early fatal-ities~for the IPPSS release category:2 source term and for the IPPSS source term with todines and particulates reduced by a factor of 10 (IPPSS/20), using the same emergency response assumptions as in the IPPSS. Analyses with the IPPSS iodine and particulate source term reduced by a factor of 20 (IPPSS/20) resulted in no early fatalities.
Consequently no curve exists for IPPSS/20 on this figure. By interpolation, about a factor of 15 reduction-in the IPPSS source term is sufficient to effectively eliminate early fatalities when there is cvacuation. The (R/S)
~
suggested source term also resulted in no early fatalities with evacuation as the emergency response. The R/S source term is approximately equal to an IPPSS/50 value for parly fatalities. Additional analyses were made with the R/S source term assuming no evacuation and no sheltering for 24 hours and even under these conditions there were no early fatalities calculated. When the IPPSS source term is used with this response scenario, early fatali;tes are calcu-lated, but over 95 percent of the early f atelity risk occurs within four miles of the Ir'ian Point plants. With the R/S source term, because no early fatalities were calculated, this measure is no 'onger relevant.
- 2. Latent Fatalities Figures 3 and 4 display frequency versus number of
latent fatalities for Indian Point 2 and 3, respectively. Latent' fatality risk is almost completely dominated by accidents that lead to slow overpressurization of the con-tainment (2RW releases). The R/S 2RW source-term is approximately equal to a reduction in'the IPPSS 2RW source term of'about 20. With the IPPSS source term, the mean value of the number of latent fatalities per year for Indian Point 2 is approximately 0.18. With the R/S source term, the Indian Point 2 mean value for latent fatalities is 0.026 per reactor year. For Indian Point 3, the corresponding IPPSS and R/S values are approximately .034 and .005 latent fatalities per year, respectively. In Tables III-1 and III-2 of the Licensee's Testimony in response to Commission Question one, latent fatalities were expressed as a percent increase over the background cancer rate. For example, using best estimate analyses, the probability of increasing the cancer rate by one percent is about once in 10,000,000 years for Indian Point 2 and about ! once in 100,000,000 years for Indian Point 3. These values are based on the IPPSS source term. If the R/S source term
- were used, there would be less than one percent increase for the same frequencies.
- 3. Man-Rem Figures 5 and 6 display frequency versus man-rem for Indian Point 2 and 3, respectively. For Indian Point 2 the mean value of the man-rem for the IPPSS source term is 3600, i
1
while the corresponding mean value of the man-rem using the R/S source-term'is 360.. For Indian Point 3, the correspon-ding values are 700 und 70 man-rems, respectively. This is a 10-fold reduction for Indian Point 2 and 3, respective-ly. As stated ~in the Licensees' Testimony on Contentions 2.1(a) and 2.l(d), the worth of mitigative (and also pre-ventive) devices are measured in terms of the man-rems averted. With the R/S source terms, the potential worth of adding a mitigative or preventive device woald be correspondingly reduced by a factor of 10 at each plant.
- 4. Property Damage Using the same approach as before, the impact of the R/S source term on property damage was. calculated. Figures 7 and 8 display this information for Indian Point 2 and 3, respectively. For Indian Point 2, the mean value of the property damage (in 1982 dollars) is reoaced from
$100,000/ year to SS,800/ year when going from the IPPSS source term to the R/S source term; corresponding values at Indian Point 3 are approximately one-fifth of these.
L l - -
Figure 2 Sensitivity of Early Fatality Risk to Source Term IP2 and IP3 l o
'-1 : l
~
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, a5 .IPPSS Release Categor r2 c -
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No early Ea'talities 8 [ calculate 3 for Rodger-St catton { 4 Release C ategory 2 N e or for IP ?SS RC2 ' e l Iodirns a la a _ ~ Particula tes 20
-t
- E:
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O , , , ,,,,, , , , ,,,,, , , ,.....
~ _ . . . . . . .
10' 10' 10' 10' 10' 10" 10* Early Fatalities _7 _y Mean Frequency 5.4 x 10 yr IPF3S Emergency Response Assumptions I
Figure 3 Sensitivity of Latent Fatality Risk to Source Term IP-2 a I o M -: I l T y , i l I 4 o x - : - Rodger-Str atton E Release Cc tegory 2RW IPPSS Re] ease
$ - Category 2RW g ,
Tables 6.2 - 16,17 i o M u
- 6. -
e u o
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Figure 4 Sensitivity of Latent Fatality Risk to Source Term IP-3 a e o M : 1 I 4 o_ ,
>. r -
@ IPI'SS Release Category 2RW a Tables 6.2 - L6, 17 e _
O e - o x . m (4 g a 0 w o O2 - 1 i
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- Tables 6. 2-16,17 u .a >i 1 O
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t Figure 8 l Sensitivity of Property Damage to Source Term IP-- 3 l l o i M: ~ i \ - i - 3
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_ p, _ lp; _ x _. _ . . 65 - REFERENCES ANS80 Winter meeting of the American Nuclear Society, Washington, D.C., November 1980. BAU79 Baumgaertner, F. and F. Heil, "Das Eindampfverhalten und die Verfluechtigung der Radioaktiven Komponenten einer Hochaktiven Abfallloesung (HAW) aus der Wiederaufarbeitung," report presented at Karlsruhe, Germany, November 23, 1979. BR063 Browning, W.E. et al., " Release of Fission Products during In-Pile Melting of UOg," Transactions of the American Nuclear dociety 6, p. 125, 1963. BUN 80 Bunz, H. and W. Schoeck, "The Natural Removal of Particulate Radioactivity in an LWR Containment During Core Meltdown Accidents," Thermal Reactor Safety Meeting, Knoxville, TN, April 7-11, 1980. CLA74 Clarke, R.H., "An Analysis of the 1957 Windscale Accident Using the Weerie Code," Ann Nuc Sci & Eng 1,, p. 73, 1974. CUB 78 Oubicciotti, D. and J.E. Sanecki, J Nuclear Mat 78,, p. 96, 1978. DAN 80 Personal communication, Jack Daniels to W.A. Rodger, June 1980 . FOR76 Forsyth, R.S., et al., " Volatile Fission Product Behavior in Reactor Fuel Rods under Accident Conditions," Proc Specialist Meeting on the Behavior of Water Reactor Fuel Elements under Accident Conditions, OECD Nuclete Energy Agency (Norway), 1976. GEN 69 Genco, J.M. et al., " Fission Product Deposition and Its Enhancement under Reactor Accident Conditions: Deposition on Primary-System Surfaces," BMI-1863, March 1969. GIE77 Gieseke, J.A. et al., " Analysis of Fission Product Transport under Terminated LOCA Conditions," BMI-NUREG-1990, Battelle Columbus Laboratories, December 1977.
, -~66 -
.HIL61 Hilliard, R.M. : et al. , " Fission Product Release ia Efrom Overheated Uranium," Health Physics 7, pp. l- '
10,.1961. HIL70L ~Hilliard,.R.K.'.et al., " Removal of Iodine and Particlcr.from Containment Atmospheres.by' Sprays
-- Containment Systems Experiment Interim Report,"
'BNWL-1244,'.Battelle Pacific Northwest Institute, 1970.-
HOR 63 _Horan,tJ.R..and W.P. Gammill, "The' Health Physics p Aspects of the SL-1 Accident," Health Physics 9, - p.-177, 1963. IAEA81 Ir.sernational Atomic Energy Agency,. Division of 4 Nuclear Safety, Technical Committee Meeting on-Airborne Fissi0n Product Release Following ExtensiveLCore Damage Accidents, October 12-16, 1981, Vienna, Austria.
' ICR77 Recommendations of the International Commission on Radiological Protection, Annals of the ICRP, Vol.
1, No. 3,-ICRP Publication 26, 1977.
' KEM79 Kemeny,1 John G., Chairman, " Report of the
. President's Commission of the Accident at Three tile Island," October 1979. K0 '80 - .Kouts, H. , " Iodine Release From . a Nuclear Power-
~
Accident," Brookhaven National' Laboratory, June 18, 19E0. (Sent to Joseph M. Hendrie, September , 12,-1980). LEV 81 Levine, S. et al.,'" Source Terms: ,An-Investigation of Uncertainties, Magnitudes, and . Recommendations for Rer,earch," ALO-1008, NUS-3808, March 1982. l i LOR 80 Lorenz,'R.A. et al., NUREG/CR-0722, February 1980. I MEN 81 Mendoza, Z.T.fet al., " Radiation Release from the-SL-1. Accident," Nuclear' Technology 53, p. 155, - 1981. MOR 79 Morewitz, H.A. et al., Nuclear' Technology --- 46, p. 33 2, December =1979. MOR 811 Morewitz, H.A., " Fission Product and Aerosol Behavior Following Degraded Core Accidents," ' Nuclear Technology 53, p. 120, 1981. e .,,,,,,;-,, _ . , , - - - . - . , . - .,,,m ,_, , ,.__7. . _ , ,
._,_,,_g. .
+ NEL75 Nelson, C.T. und R.P. Johnson, " Aerosol Leakage Tests," ERDA-56, 1975. NUR81 NUREG-0772, Technical Bases for Estimating Fission Product-Behavior During LWR Accidents, United States Nuclear Regulatory Commission, June 1981. PAR 63a Parker,-G.W., " Fission Product Release'Research," Transactions of the American Nuclear Society 6, p. 120, 1963. PAR 63b Parker,_G.W. and R.A.'Lorenz, " Fission Product Release'from Meltdown of a Cluster of Center-heated UO 2 Fuel Pins," Transaction of the American Nuclear' Society 6, p. 124, 1963. IU)G80 _Three Mile Island a Report to the Commissioners and to the Public, Mr. Ragovin, Director, United States Nuclear Commission Special Inquiry Group, April 5, 1979. RSS,5 WASH-1400, " Reactor Safety Study -- An Assessment of Accident Risks in U.S. Commercial Wuclear Power Plants," NUREG-75/014, United States Nuclear Regu-latory Commission, October 1975. STA80 Starr, C., M. Levenson and I.B. Wal.1, " Realistic Estimates of the-Consequences of nuclear Accidents," NRC briefing, November 18, 1980. STR80 Letter from W.R. Stratton, C.P. Malinauskas and D.O. Campbell to Chairman John Ahearne, August 14, 1980.
- TPO57 WASH-740, " Theoretical Possibilities and l Consequences of Major Accidents in Large Nuclear Power Plants," USAEC, 1957.
TID 62 Dinunno, J.J. et al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, 1962. l l
APPENDIX Detailed Development of Revised Source Terms This appendix includes 12 tables giving the details of the development of source terms for two IPPSS release categories (2 and 2RW) and six isotope groups. The specific table identifications are as follows. Sequence-Isotope Group 2RW 2 Cs-Rb .A-1 A-7 Kr-Xe A-2 A-8 Te A-3 A-9 Ba-Sr A-4 A-10 Ru A-5 A-ll La A-6 .A-12 l' I
. Table A-1 Evaluation of Envirormental Releases for Irullan Point Seoence 7_1W Fraction of Core Inventory Tire Event Segjerce Expected Betavlor of Cs-Hb(1) in Contairment Air Release:d to hr Envirormenta Explanation of Reasoning 0 Break in small lire on hot leg- Only the RC FP Inventory is involved at regligible nose Ho core damage and pressure 600 to 1200 ps!. Release this point mest of which will remain contairment intact at to contalement as fir hlrq steam- in the llgild. this point, water mixture. Flow path:
upper pleram hot leg break contalrment Residerce times 10 to 30 talrutes. 2 to Safety injection systems inoper- Cap selease ( 2) s.lus en additional 2f5 of 0.25x0.1 = 2.%-02 IE-06 Core release from 10PSS 2.5 allve, core tegins to test (p. Cs will be releLed during teatup Nisse, and EIMI/flGR studles.
. Expect if of at least 10 by plateout on ,
Major removal by gnavita-toper plenum and interconnecting pipirg. tional sedimentation. Correlation used: egn E-19, t4JtEG-0772. Gieseke i results [GIE77] suggest even higter (f. $ 2.5 Core meltlig proceeds to about Remainder of Cs released via same flow 2.%-02 4E-36 Dominant process through- - I to 4 3r*, corium tw;;lrs to run into path as above. Expect [f of at least 10 out is gravitatlonal sed-19ttom of RPV. System pressure as befrre in plenum and piping. Process of Amentation. See previous rises. Irpit to contairment plus depletion in con- entry. Hemoval in corw talrment by natural settllrq should lead talrment based on ESE to airborne corcentrations about as before. studies [fDS 71] which Host of tte unretained isotope [7.35 to 10E] suggest removal telf-times will be in the RPV. of about 30 mirotes, a Based on 15/ day leakage from unruptured contalrment and an air turnover time in contalrment of 2 hours after contairment tes failed. i W
t Table A-1 ' (ct;otirmed) Evaluation of Environmental Ibeleases for Indian Point Sequerce 2HW
~
Fraction of Core Inventory Time Event Sewerce Expected Behavior of Cs-It(1) In Containment Air Released to . hr Environmenta Explanetton of Reason!rg 4 RPV fails and the awrox SOE of Although all Cs would have been releasci lE-01 neg dJring tie Allows for transfer of tie ' core which is molten falls into from tie core by this time, the depressuri- approx 100 secs airborne material in RPV reactor cavity and into water zation of the RPV would transfer the approxi- regilred for de- to cortalrument during : which flestes to steam and re- mately 7.315 to 10E which remained altborne pressurization. deptersurizatinn. , solldlfles core. In the RPV loto contstrument where it would add to that already L..ere. 4-5 Hemainder of core melts,' falls . Cs long gone--or any residial release 5E-02 IE-05 CSE studies (see supra) into bottom of failed RPV and would have been in RPV and path wuuld teve ' indicate removal by thence into reactor cavity where been as described above. Ite fraction in factor of 4; used 2. It boils off residual water. containment air would teve had time to reduce furtter by a factor of at least 2. , 1 i 5-12 Core-concrete interaction, non- Cs all released previously. Material in IE-02 SE-05 Used removal half time I condensible gases generated contalnnent air would teve time to reduce of 2 1/2 hrs to allow q 1rcreasing system pres ure, further by a factor of at least 5. for earlier removal of o' larger particles arul re-duced tiet almost by two. I 12 Contalr' ment falls due to over- Wien conLinment falls about 905 of the IE-02 SE-03 Allows for removal i pressurization--about 140 psla. contained gas would be released carrying during pressure surge, with it most of tie airborne activity although some removal of particulates by impaction could be expected. 12-10 Contaltunent open. Assume core Most of remaining airborne materl.a1 reg IE-U2 Allows remainity lut to querched Ly 20 hr by introduc- would be released. te discharged. .Some tion of water from outside source. would be retained, a Based on It'1ey leakage from unruptured containment and an air. turnover time in containment of 2 hours af ter contairveent has failed. t t
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Table A-2 (continued) , Evetuation of Envirornental Releases for Indlan Fbint Segjente 2RW Fraction of Core Inventory - Time Event Sequerce Ewted Behavior of Kr-Xe in Contairment Air Heleased to-br Envirormenta Explanetion of Reasoning 4-5 Hemalruler of core melts, falls No rediction of NG concentration in con- 9.8E-01 2E-04" . Containment leakage. into bottom of failed RPV and tainment other then tr/ small contairment trence into stry reactor cavity leakage. wtere it boils off rasidual water. 5-12 Core-corcrete interaction, non- See previous comment. 9.8E-01 1E-03 Contalraent leakage. condensible gases generated increasing system pressure. 12 Contalsment falls due to over- When contalrment falls about 905 of the 8E-02 9E-01 Allows for removal' pressurization--about 140 psia. the contaired gas would te released carry- dJrirg pressure surge. Ing with it a comarable fraction of the noble gases. 12-20 Contairment open. Assume core Nost of remaining airborne material neg IE00 Allows remaining 1(E to be I gjenched by 20 hr by introduction would be released. discierged. Some would tu y of water from outside source. retained. w I a Based on 1%/ day leakage from unruptured contairment.at:1 en air turnover time in containment of 2 hours after containment tes failed. i I
L Table A-3 . Evaluation of Envirormental Releases for Irelan Point Semence 2RW Fraction of Core Inventory Time Event Sequerce Expected Behavior of Te in Contairment Air Released to - hr Envirormenta Explanation of Reasoning 0 Break in small lire on hot leg- Only tre RC FP Inventory is Irwolved at negligible none No core d. mage ard ctus-pressure 600 to 1200 psi. Release thlr. point most of which will remain taltment. intact at this to contairment as flashirg steam- in tte ligild.
- point.
water mixture. Flow path: upper plerum hot leg break contairment Residerce times 10 to 30 minutes. 2 to Safety injection systems inoper- Negilgible gap release but ini'lal wla- IE-03 neg Core release from IFT5S 2.5 ative, core begins to heat tp. tility about egJa! to Cs (255) released and EfRI/IOQJR studies, from core but reduced by facter at least Parker experleents (FRH82) ten because of interaction with Zr. Tien stapjest zero release of expect additional if at least 10 by plate- Te. Used only one factor out on teper plenta and interconnecting of ten reduction. Major I pipirg. removal by sedimentation, y Cleseke (CIE77) sujgests w even higter tr. I 2.5 Core meltirg proceeds to about Remainder of Te released daring this IE-03 neg Ouminant removal process to 4 505, corita begins to run into period beteving as abuve. Expect DF=10 throughout is gravita-bottom gf RfY. System pressure as before in plerum and piping. Processes tional sedimentation. rises. of irput to contairment plus depletion See previous entry. He - trerein by natural settling should lea.d to naval in contairment based airborne concentrations about as before. on CSE studies (IUS71) Most of the unretained isotope (1%) will which stapjest removal be in ite RPV. telf-times of about 30- ,
. mirvates.
a Based on 12/ day leakage from unruptureo containment and an air turnover time in containment of 2 teurs after contairment tvis failed. l l i
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Toble A-4 Evaluation of Environmental Heleases for Irdian Point Se<3Jence AW Fraction of Core Inventory ilme Event Sequerce Expected Behovlor of Be-Sr in Contairment Air Released to hr Envirormenta Explanetton of Reasoning. O Hreak in small lire on hot leg- - Only the RC FP Inventory is involved at neg!!gible- ruse h core damage and corw pressure GD to 1203 ps1. Releasa this point most of which will remain telnment irtact at this . to contairment as flashis.g steam- in tre licpaid. point. water mixture. Flow path: ageer plenum lot log break contalement Hesiderce times 10 to 30 strotes. - 2 to Safety injection systems inuper- Lip release (reg) plus en additional 15 of ' O.01x0.1 = IE-03. nog ' Core release from IPf55 2.5 alive, core begins to lest tp. Sr will be released disring heetw phase. and EfMI/IDCLM studies. Expect UF cf et least 10 by plateout on . Major removal by gravi-teperpienta ard interconnecting piping. tettosol sedjeerdotimi.
. See previous comments on Cs ard Te. I 2.5 Core melting proceeds to about Another 45 of Sr released wie same path ~ IE-03 nog Dominent process 'is gra- us to 4 NE, corium begins to run 3rdo as above. Expect DF of at leest 10 as vitational settling. See bottom of RPV. System pressure diove. Processes of Irput to contalruent estries la Cs table. I i rises. plus depletion in etsitairment should leed
, to airborne concentrations about as before. t Most of the unretained isotope (0.5t) will be in RPV. > a Based on 15/ day leakage from unnetured conteltment and en air turiever time in contairment of 2 hours after conteirment les f*' led. I t k 1 .
s Table A-4 (cordinued) Evaluation of Envirormental Releases for Indian Pbint Seeserce au Fraction of Core Inventory Time Event Seg;erce Expected Behowlor of Be-Sr in Contalrument Air Released to 4
'hr Envirorumenta Explanetton of Reasonity 4 RPV falls ar.J the aprox 505 of No aMitional Sr comes off at this potrt but ' 6E-03 neg - Allows for transfer of core which is molten falls into the depressurization of the RPV will trans- aifoorte meterial in RFV reactor cavity and into water "er tiu aitterne Sr (approm 0.55) into to contattument daring 4
which totally flashes to steam contelnnent. depressurization. and resolidifies core. 4-5 Remainder sf core melts, falls No additional release of Sr during this 3E-03 IE-06 CSE sttalles irmlicate re-Into tmttom of failed RPV and period. Tte fraction in containment air movel by factor of 4; therce into reactor cavity where would have had time to redre by a factor ural 2. It bolls off residjel water. of at least 2. 5-12 Core-corcrete .ateraction, enm- Additional SE of Be-$r come out into cork IE-03 2E-05 A mrom egillibrium would 4 condensible gases generated in- talrveent with concrete interaction. be set w between intro-creasing system pressure, duction of meterial into , contairement and depletion by natural deposition. w m ' 12 Containment falls due to over- when contr.1rument falls about 9GE of the IE-03 SE-04 Allows for removal darir- - ' pressurization--about 140 psia, contained gas woutri be released. Some pressure surge. additloral remova; of perticulates would te expected due to lapaction on the many obst:uctions. 12-20 Contairement open. Assume core Nost of remaining airborne meterial 'would neg_ IE-03 Allows remaining IGE to gjerched at 20 hrs by introduc. be released. be disctergest. Some would tion of water from outside sourco. be retained. a Based on 11E/doy leakage from unrupturh1 contalruent and an air turnover time in , contalrveent of 2 hours after containmert has failed. i 1
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Table A-6 Evaluation of Environmental Releases for trullan Fbint Se@ence 2HW Fraction of Core Inventory Time Event Segerce Expected Behavior of La In Contaltr2nt Air eteleased to hr - Envirormenta Explanation of Reasonitg 0 Break in small lire on tot leg- Only tie RC FP inventory is involvevi at negligible none- No core damage erut cow pressure ao to 1200 ps1. Release this point most of which will remain talement intact at this
- to contalruent as flashing i, team- In tie 11cpild. point.
water mixture. Flow path: upper plenum . , itt leg break contalrment Residence times 10 to 30 minutes. 2 to Safety injection systems inoper- Neg!!gible release at trils point, regligible none 2.5 ative, corr Wins to test to. 2.5 Core meltirg proceeds to about Small amount (0.2%) La released via IE-04 neg Core release frta IF15S to 4 505, corlue begins to run into upper plenue arut tot leg. Expect if at and EFHl/IDCUR studies. Dominant process is gra- I tottom af RPV. System pressure least 10 on plenum and piping. Most of rises 2 tte unretained La (0.025) will be in tie RPV. vltational sett Mng. See y entries in Cs table. e neg Allows for transfer of I 4 RPV falls eruj Lie approm SOE of N) additional La at this point but de- 2E-04 core which is molten falls into pressurization will transport the airborre alttorm material in RPv . reactor cavity and into water La (agrrom 0.025) into contalrment. Into-Malrment djring which totally flestes to steam depressur! ration, arul resolidifies core. a Based on 1%/ day leakage from unrupturec* contairment and an air turnover time in containment of 2 hours after containment has failed. o %p
A lable A-6 (contirused) Evaluation of Envirormental lleleases for Indian Point Sewre 3tW Fraction of Core inventory Time Eveat Sequerce Expected Othawkr of La - Jn containment Air reeleasea to nr Envirormenta - Explanation of Reasantry 0-5 Hemainder of cote melts, falls No arkj1tional release of I a durirg this IE-04 neg ' CSE sttrjies indicate re-Into bottom of failed RPV aruj period. Fraction in containment air would moval by factor of 4; thence into reactor cavity wtere teve had time to rethce by a factor of at used 2. It bolas off residal water. least 2. 5-12 Core-corerete interaction, non. Additional 15 of La come out into con- IE-03 2E-05 /4 pros equ111brium would condensible gase:6 generated in- tairment with concrete Interaction. te set q) between intro-creastrq system pressure, duction of material into containment azul depletion by natural deposition. About Contairmer$t falls due to owwr- When contairment falls about 90E of the IE-03 9E-04 Allows for removal durirg 13 pressurization--obout 140 psia. contained gas would be released. Some pressure surge. additional removal of particulates would be expected (kse to' impaction on tie many 8 obstructions. m o 12-20 Contairment open. Assume core Most of remainirg airborne material would neg IE-03 Allows for release of most (peerted at 20 hrs with water be released. of remaining airbotte I from outside source. material. a Based on 15/ day leakage from untuptured containment and an air turt' aver time in contabuent of 2 hours after contairment tes ?isited. =
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i' ~ 11WE OF. ElRTH: 15 l!ny 1927 m [
~ ;larried i '
iGRlTAL STATUS I ROFESSION OR TRADE: Physi cist . U' r
.":111 CAT 10S * .
H1 CHEST DEGREE
' LOCATION AND DATE FIELD OF STUDY Scil 00L River Fa11s High River Fa115, WI 1940 !1 " " " ncne und ergradunt e f
U.~ of Wisconsin i ilinneapolis, I.!N A.B.,.1947 Physi cs, !!ath U. of !4innesota Physies, !4:sth
" " Ph.D., 1952 U. of *:innr . otre 5 v -
li!EFIS ADY1SOR: - .ichn H.1 il'15 ams
'iiii 515 71TLE: The. Experimental Determination of Cross Section Per Uni: Solid j Angle for the Elastic' Scattering of Deut erons .by Trit ons and, f the Elastic Scattering of Protons by Dentrons as a Function o A
1 'S:attering Angle 'and Inciden- Deuteron Energy. I E;4DJ.OD17.NT AND EXPERIENCE:
- d. '
. From: IJ i;ovember 1977 To: Present i;se- :ind Address of Employer: LASL, Grcup TD-7
' Supervisor: Alfree T. Feeslee, Jr.
I Business: Intelligence
- : i'our Position: Staf f 14e ber L i-lork performed' and equipnent used: Project 142 nager for foreign technology
- snalysis.
k l I November 1977
- Frc=: .1 !4 arch 1977 To.
h: ice and Address of Employer: LASL, Group Q-7 h Supervi sor: Flichael Stevensen Reactor Safety E Eusiness: David Hall
. Your Position: Sin fi tie =her T o::ci c:
i ::eri, perfor .cd - end equij. ent used: Adei>ory work in the field of fast s:. ie t y . <
\
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- From: l'ay 1957 To: ;'.a rch 1977 - -
Iace and' Address of Employer: .LASL, Greap N-2 , , _
. Business: Supervisor: Hugh Paxton-Dad d Hall .
Your Position: Staff Member p ifork performed and equipment used: Theoretical studies connected with cri- > ticality safety, parameters associated eith critical and near critical l. f
-systems, and studies of the dynamic h:hsvior of supercriticn1 reacting
( ' syr ems; occasional efforts connected with critical assemblies. Safety e .
~
of Rover reactors and test operations. Design of and predictien and analysis' of the I;iwi-TNT experiment. 9 From: August 1952 To: 1957 Name and Address of Employer: LASL, Group If-4 Business: r: capon Design Supervisor: Art Sayer Your Position: Staff i; ember 1lork performed and equipment used: Theoretical weapon design. Fro =: April 1952 To: July 1952 Name and Address of Employer: Univ of Minnesota Business: Nuclear Research Supervisor: John H.11111inas Your Position: Research Associatte tiork performed and equipment used: Research with the 11niv of Minnesot: Van de Graaff accelerator, assisting graduate students. ~ 11 PROFESSIONAL
REFERENCES:
David Hall H. C. Paxton R. E. Schreiber 1 Norris Bradbury 3 OTHEP.- RELEVANT INFOR".ATION: 1 Member, Advisory Committee on Reactor Safeguards, 12/01/66 - 12/31/75 l L Member, Los Alamos Criticality Safety Ccamittee, 1960 - 1
. U. S. Representative, Cadarnche 1.aboratory, France, - 1965-1966; special interests: fast reactor safety, criticality safety, critical experir.:en t s .
1 h c 1 'J 4-W ~
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- ik;:Ler, Tne '_'Li11ihr." C.: sittee, _ h*a ti.ircion, l<, 9/5c- 2/L.; . Ths 121;irin k Cummittec war. a tpecini ad hoc ,,dvitory :;roup in the President's S:ience -
l.
.idvir.ory Ccen.itice.
m ~ il::::.ber, American Physical Society a .. .
- p. -l1:=ber, Socicry of the Sigma Xil r .
V . 14:mber, American Association for the Advancement of Science . i ; Fe13cw;, American 'uc3' car Society , 4 3 Board of Governors,J Amerienn Nt; clear Society: 1975-197S Censultant to Advisory:Cc:=5ttee cn Rescior Safc;;uards. F Consultant to the . Division _of Reactor Development- and Demonstration. Past activities at the LASL have included theoretical nucIcar weapon ;
' design, Rover reheter design and ' safety, wide experience in criticality safety r.nd- studies in parameters appropriate- to critienlity. A continuous effort since 1957 has been in the general arca of fast reactor safety.
~
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- 1. Late January-February,197E, I wss atomember -
north,rnofCanada the special Nucir ar to assist Emergency Scarch Tea:: that was sent r.a tel li t e in. the search for and recovery of frapents' of theofSovietThis satell
" Cosmos 954".
created a special ~ ht:ard when it returned toI the wascarth an advisor cast Yellowknife Northwest Territorics, Canada. for criticality and safer:/ prob 1ces. . asked to work with the staff supporting and
- 2. On May 20, 1979, 1 S.25 T~nree reporting to the Tresident's Cc=ission on the Accident at Mile Island. ' The chairman of this Cc: mission was Dr. John Tc=en president of Dart 7outh College. I was the principal author of time entil the end of October,1979. ibuted the staff document " Alternative Event Sequences" and contrl i Assessment to other documents from the group known is the Techn ca Task Force.
- 3. Publications:
An assessment "US USSR Cocperation in Fast Ereeder Ecactors:
.(a) of the Agreement; by Paul M. Giles, John L. Rand Sponsored by the US Department of Energy.
of the Destructive Energy Created (b) " Estimates of the Upper Limit Reactor Power Excursions, John Subsequent to Postulated Fast T. Larkins, Thc=as P. .t:cLaughlin and ~ William LA-UR 7c-534, f. Strat 1979.ten, Los Alamos Scientific Lsboratory document, 26, 1979, "Are Fortions (c) An article in Science cagazine, Ocicber Stration, Danny of the Urals Really Contaminated?" by W. R. y Stillman, Sumner Barr and Harold Agnew. 4 i 1 8 b s 5 1
,)
l
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. W. R. STRATTON-1 .I 'I Accidents Involving Postulated Large Changes in Rcactivity 'l (EN). Str . t. ton , W. R.
I,ulletin of the American .chysi-cal Society, Vol. 21, 14, pp. 604-605, Les A3ames
- Scientific Lab. (1976).
Analysis of the F.iwi-TNT Experinent. Stratton, M. R., and King, L. D. P. Trans. Amer. Hucl. Soc., Vol. E, il, { pp. 126-127 (June 196c). g Analysis of the Kiwi-TNT E'E:cursionStratten,
. and itr Relaticn M. R.; Adler, to .cver F
Reactor Accident Predictions. D. M. AIAA F. T.; Altomare, P. M.; and Pcterson,
,i
~
Propulsion Specialists Conference, Colorado Springs, CO (June 14-18, 1965).
.l j,
Analysis of the Kiwi-TNT E.x periment R.; and Phoabus Conceptral Accident Study. Stratton, M. Altcmarc, P.; and f. Peterson, D. M. Chapter IV; Sections S r.nd D, LA-335S ".S. 4 e.nalysis of Prompt Excursions in Simple Systems. Stratten, q W. R., and Colvin, T. M. Tast Recctor Informa. ion Meeting, Chicano, Su:: aries of Papers, p. 14, DTIE 3 4 (TID-7548) (1957).
.l An . lysis of Prompt ExcursionsR.; in Simple Syste:ts nd Idea 3 iced Stratton, U. Colvin, T. M.: and Lazarus,
.i Reactors.
}
R. E. International Conference on the Peaceful Uses of Atcmic Energy, 2nd, Geneva (1945), Proceedincs. Vol. 12, pp. 196-206 (1958).
'l ..ngular Distribution of the Reaction 1:e3 t (d,p)He4 3etween 240 H.; and KV and 3.56 MeV. Yarnell, Y. L.; Lovberg, R.
4 Stratton, M. R. Phys. Rev. 90, pp. 292-297 (1953). Studies. Applications of the PAD Code to LMFER R.; Power and Transient Peterson, D. M. Engle, L. B.; Stratton, W. f Transactions of the ANS, 15, (2, p. 820 (Ucvember 32-17, 1972). Conceptual Prompt Power Excursions in Propulsion Reactors. 4, il, Pp. Stratton, H. R. Trans. Amer. Nucl. Soc., L 1949-50 (June 1961) .
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;Correlkt'icns of'Experin.cntal'end Theoretice). Critical Dr.ta.
- Comparativef Rel'iabilit?, Safety . Factors ' fer Criticali:y A., Jr.;
2 ; Control. Pa>: ton,f H. C.t Carlson,-E. G.;.Goodwin, Hansen,1G.'E.; Mills, Cc B.; Roach, W. H.; Sa f onc:, G. ; - _. 4 LAMS-2537.-(March 1961) .
'and Stra tton', W.. R.-
1Correlatilons.of E;;perimentsTand Calculations. Stratten, W. R.- : Nuclear -Criticality Safety ~ Mational Topical
' ' ~
' Nceting, Trinity and Southern Nevada T.ccticns, . J. erican 91-104,- e Nuclear ' Society, ~ Las ; Vegas, MV,' : Procee6ines , : pp.
- Sandia Cor pora tion, --.SC-DC-67-1305 - (Dc .:c=ber 13-15, 1966).
h1 ,< 3., and 9: ' Coujd ed Neutronic-Dynamic jroblems. ; La zarus , ;T.. M.R.. computer methods-in Reactor-Physics, a*l X- / e Stratton, " H. Greanspan Ed., Chapter 7, pp. 509-533, Gordon-and. c
~
-- Bresch f(1968) .- .
Critical Dimensions of Tiomogeneous Spheres Containing 2350,~
"2380, and Carbon for various and C/2350 Stratton, W. R.
Rctics and 2350 I '
"nrichments. Engle,.L. B.,
LA-3SS!,-MS 1(May .15, 1968). n . Critical Dimensions of ~ nranium- (93.5)-Graphite-Mater L.N:S-2 955. Spheres, (Hay Stratton, M. R. y
' Cylinders,:and Elaus. -
' a -
1962).
~
l Criticality.-Data and Factors Affecting Criticality of Single Stratton, W. R. LA-3612 (Septem-t a liomogeneous US.its.
'ber 22, 1967).
CriticalityjResearch at the Los'Alamcs Scientific Laboratory R. - ANS, and ^at ' the Rocky Flats Plant.- .Stratton, H.
- c Transactions,.Vol. 11, # 2 ~, p. 690 (1968).
Cross Section of the G12(p,y)N13 Reaction Phys. at LcwRev. ' Energies. 77, ) Bailey, C. L. , and Stratton, W. R. f C pp._194-196 (1950).. Differential Cross-Section Measurements fcr the Scattering R. J.; Freier, G. D.; H of Protons by Deuterons. Erown,
- L.
f H..D.; Str'a tton , N. R.; and Yarnell, J. 1 Holmgren,
! Phys. Rev. 88_, pp._253-256 (1952).
s Stratton,
)
.The Elastic. Scattering of. DeuteronsG.byR.; Tritons. Rankin, D.; and
? i i
'W.~
R.; Freier,-G. D.; 1:cepin,
*"1 Stratton, T.:F. Phys. Rev. 88, p. 25_7 (1952).
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. Energy fielear.c f rcm Ihltde;:n J.ccidents. c.tri;icn, M. A.:
t Engle, L'. D.; cnd Peterson, D. 11. f.or. Al ar.:s Srd :nti- ~ r fic Lah., HM, Trans. A:per. 1;u cl . Soc., 17 0;cycr.he r 1973T. -- N Energy-nelt:ase f rom IM1 ten-Fuel Recriticality accidents
~ $ P.; Peterson, D. M.; aad Strattan, d' . (Eli) . McLaughl, T.
4 H. R. Trans. Amer. Nucl. Soc., Vol. 18, pp. 198-199, Los Alamos Scientific Lab., PO3.1663, Les Alamos,-FM (June . 23, 1974). Stratton, Explosive Energy -Release fron lieltdown Accidents. ANS Meeting,- W. R.; Engle, L. B.; and Peterson, D. M.
.I San Francisco, CA (1;ovember 12-16, 1973).
Feasibility Study of the liuc1ccr Destruct Ce.ncept for King, Pcst Operational Disposal of the Rover Type Focket. Joint f. E L. D. P., and Stratton, W. _ R. AIAA Propulsion Specialist Conference,. Colorado Springs, CO-(June 14-15, ' f 1965). Gcdiva II. An Unmoderated Pulse-Irradiation Reactor. R.; and Wood, Mimett, T. F.; White, n'. H.; Stratton, M. D. P. Muclear Sci. and Enc. 8, pp. 691-70S (1900).
. Kiwi-TNT Experiment. Stratten, W. R. Air Force /Public Eealth Service Of f-site E::perimental Study Symposium, s- Cocoa Beach, FL (Decemhcr 1-3, 1964).
Fenstermacher, C. A.; I'ing, , b Kivi Transcient 13uclea:- Test. M. P. . AIAA Bulletin, Vol. 2, L. D. P.; ead Stretton,
- o. 226 (1965).
Fensternacher, C. A.;
, Kiwi Transient Nuclear Test (U).
P.; and Stratton, M. R. Confidential ED, King, L. D. m. LA-3325-MS (June 17, 1965). Engle and Peterson. ANS Trans. 13, -3 2 , Large Accidents. . [ pp. 721-7 22 (November 1970) . e LMFDR Disassembly Analyses. McLaughlin, T. P.; Jackson,_ P J. ; and Stratton, J. F.; Forehand, H. M.; Koelling, J. Los Alamos Scientific Lab., Report LA-UR-76-2208, "t W. R. Ai:S Meeting on Fast Ecactor Safety (October 1976). i!ECKLACE--A Computer Program Concerning the Transient Tem-peratures of Fissioning Spheres Embe6ded W. R. LA-3115-MS in Graphite. (March Chezem, C. G., and Scratton,
-' 1964).
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"1 Ucutronics !!andbook f or Ei.ti-A' and Kivi-A3 Cperation at
. !:TS . Darten, D. M.; G? aves, G. A.; Orndoff, J. D.; -
P a >: t o n , H. C.; and Stratton, M. R. Muclear Safety Aspects of the Rover Program (U). Graves, G. A.; Harris; P. S.; Langham, M. H.; Anderson, C. G.;
- Peider, R.; Stratton, W. R.; Van Dilla, M. A.; Ander-d son, U. C.; Haddad, E.; Pansen, G. E.; Kerr, T. C.; and i Walton, R. B. Secret PD, L7.- 2 4 0 9 (March 1960).
PAD: A One-Dimensional, Coupled I!eutronic-Thercod>namic-Hydrodynamic Computer Code. Pcterson, D. M.; Stratton, W. R.; cnd McLaug?lin, T. P. LA-6540-MS (December 1976).
)
The Pajari.o Dynamics Code ~ with Applications to P.cactor Experiments. Strctton, W. R.; Peterson,-D. M.; and h jl Engle, L. 3. Transactions of the AMS, Vol. 15, 62,
- p. 819 (November 12-17, 1972).
t Proyosal for a Pulse 6 Reactor. Stra tton, M. R. N-2-Sll7 (August 31, 1966).
- }
i PAC--A Cc=puter Program for P.eactor Accident Calculations. ,d Chezem, C. G., and Stratten, W. R. LA-2920 (January L8 k 1963). I i Beacter Power Excursion Studies. Stratten, S D'. ; Fngle, L. B.; and Peterson, D. II. International Conference on the Engineering of Fast Rcactors for Safe and Reliable Ope'ra tion , Karlsruhe, Germany (October 9-13, 1972). Reactor Power Transient Ctudies. Stratton, W. R.; Engle, L. B.; and Peterson, D. M. Los Alamos Scientific Lab._, NM. LA-DC-72-660 (June 2, 1972). ka[ , i} Report on Integrity of Reactor Vessels for Light, Cater Power
- Reactors (EN). Stratton, W. R. Nuclear Engineerinc e and Desion, Vol. 28, 92, pp. 147-195, Atom. Energy e Comm., nashington, D.C. (1974)
Review of Criticality Accidents. Stratton, W. R. LA-3611 (January 1967) . Review of Criticality Accidents. Stratton, W. R. Progress in Uuclear Enercy, Series V. Vol. 3, pp. 163-205 (1960). Y ' \ Jt - f: - - : - m w7,7c;e w; 3 ~ 7, ,, . m. yp,.y.7- . _
i
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A Review of Criticoiity Inci6ents. . Criticality ContrcI in Chemical 1and lietallurgical P1cnt. Stratton,.W. R.
+ .
Paris, European nuclear Energy Agency, pp. 491-533
- (1961).
Stratton, Rover Reactor Powr Transient Calculations. Proceedincs, Muclear.Propul-W. R., and Chezem, C. G. 1, sion Conference,.Uaval Post Graduate School, Monterey,
)
CA (1962). TISE (1963), pp. IS4-191 (TID-7653, Pt. 2). Rover Re,'ctor Transient Analyses. Stratton, W. n., and Che:e:n, C.--G. Proceedines. Acrospace Nuclear Safety National Topical Meeting, LT.erican Nuclear Society,
.Albucuercue,.EM (October 1-4, 1963).
l the Scattering .of Protons-by Tritons. Classen, R.'S.; f 3rown, R. J.'S.; Freier, G. D.; and Stratton, M. R. Phys. Rev. 82, pp. 589-596 (1951). Scre 1:nstable Reactors. Stratton, W. R. M-2-8535 (Jan-uary 14, 1972). Su-sary of Co r.ents Of fered at the Meetings on Recriticality r Ene rgetics. .Stratten, W. R. Los Alamos Scientific Lab.,'!W. LA-1:n-76-887 (1976). 1.NS Trans. 12,.#1, p. 11 (June h Thermal Reactor Safety. 1969). i 1 v. g . a w i t< . Q-y , Y f l o
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@ nucicer-safety associates
- WAU
- CN A m uor.4 EDUCATION Dr Rodger holds BS degrees in both Chemical and Metallurgical Engineering from the University of Michigan (1939) and an MS degree in Chemical- ,
Engineering, also from the University of Michigan (1940). His doctorate in Chemical Encineering was awarded in 1956 by the Illinois Instiv.ute of Technology. , Dr.Bodger has spent seven years ir. private industrf PRCFESSIONAL EXPERIDCE -(pharmaceutical and nuclear) in production and plant construction, 17 years in research and development at AEC National Laboratories, and 17 years in private consulting. Much of his working lifetime has been spent in the field of waste control. He is recogn'ized i- internationally as an expert in the control of.
; - radioactive waste and has given ' papers on the subject at numerous. national and international meetings in the US'and half a dozen foreign countries.- He has written ;
extensively, is the author of about three dozen papers, and is the author of sections of several nuclear engineering handbooks. - In the course of his career Dr Podger was responsible for the design of the liquid waste handling
. facilities, both nuclear and industrial, for Argenne
. National' Laboratory, Argenne, Illinois, and for the total waste handling facilities for the West Valley, NY, plant of Nuclear Fuel Services, Inc--the weld's first ccmnercial nuclear fuel reprocessing plant.
Dr Fedger has done extensive work throughout the entire nuclear fuel cycle, including research and develogreent, design, construction, operation, and consulting. He has also hal a large amount of experience in dealing with the licensing of nuclear facilities.
()nuclocr safety associates WALTON A ROCGER page 2 PRCFESSIONAL For several years Dr Rodger was involved in preparing EXPERIE!CE and presenting testimony for the utility industry at (continued) the NBC's Appendix I Rulemaking Hearings aimed at determining what releases are "as low as practicable" for light water reactors. This involved evaluating all possible release paths from the reactor ccrnplex, the effective doses therefrom, the costs associated with reducing emissions, and a cost-benefit analysis of such reductions. This work was done for 4 group of about thirty utilities. Iater that croup was recenstituted to ccmuent on EPA's efrorts to further limit effluents frem the nuclear fuel cycle. Dr Rodger has also appeared as a technical witness in several individual licensing actions testifying on the ALARA implications of radwaste systems. Following his post-24I service at CMI-2, which consisted largely of following the behavior of iodine in the system and assuring that no action w s taken which would significantly increase the environmental release of iodine, Dr Rodger became interested in the calculation of accident source terms, since the 24I data strongly suggested that the then current methods of calculation over-predicted the actual source term. Since then he has prepared for Long Island Lighting Cc9ny an extensive review of nearly 30 years of literature which pertains to release of fission pro 6 cts from overherted er moltan fuel. Pertinent - accident data were also reviewed and compared to calculation. He is curre.ctly serving on the Peer Review Cemitte for the Shoreham PPA study and is in charge of Section ll.?--Fission Product Transport--of the industry's degraded core study--IDQR. He is aise doing studies for the Electric Power Research Institute of the release oi. fission products from both degraded and miten ceres.
.a.
(.)riuclGCr safety asscciates WALEN A PCOGER page 3 17tILITY . In addition to the Consolidated Utility Group for whom CLIENTS the Appendix I work was done, utilities with whom Dr. Podger has consulted during the past decade ixlude: General Public Utilities Iowa Electric Light & Pcwer Bochester Gas & Electric Co Pacific Gas & Electric Co
. Boston Edison Co long Island Lighting Co Florida Power & Light Co Houston Lighting & Power Co Carolina Power & Light Co San Diego Gas & Electric Co Vermont Yankee Metropolitan Edison Co Jersey Central Power & Light Co PPCFESSICNAL Dr Podger is a Fellow of AIChE 'and the 1981 recipient AFFILIATIOt3 of AIChE's Pobert E Wilson Award. In 1960 he was Chairman of the Nuclear Engineering Division of the Institute. He is Liso a member of ANS ard AIF. In 1959 he served as Technir:al Consultant to the Joint Comittee on Atomic Enegy of 6e 86th Congress at the Hearings on Industrial Fadioactive Waste Dis;csal, and in 1972 prepared testimony before the Joint Comittee on Atcmic Energy's Hearings on Ultimate Disposal of Nuclear Waste. Until mid-1975, be was Chairman of the AISI N-48 Comittee on Ultimate Waste Disposal. He is also a nember of the ANS Ad hoc Cc:rmittee on Faactor Accident Source Terms.
NAME THOMAS E. POTTER _ECUCATION M.S., Environmental Science, University of Michigan,1972. B.S. , Chemistry, University of Pittsburgh,1963. PROFESSIONAL EXPERIENCE General Summary Consultant on health and safety aspects of nuclear power. ' Performing ' environmental dose assessments for nuclear power plant safety analysis, environmental reports and operating reports. Assisting clients in design and implementation of radiological or environmental monitoring programs and interpretation of results. Providing independent review of in-plant radiological protection programs and effluent analysis programs. Consultant in radiological health aspects of nuclear power. Prepared radiological health section of safety analysis reports and environmental monitoring programs and evaluated data from those programs. Developed a mathenatical model to predict radiation doses from nuclear power plant effluents. . License administrator, plutonium fuel facility health and safety supervi sor. Provided radiological safety review of major facility modi fic1tions. Used these analyses and nuclear criticality analyses perfomed by others to prepare AEC special nuclear materials and bypr: duct license applications. Served as corporate contact with AEC in matters related to licensing. Organized and supervised a radiological protection pecgram for a plutonium fuels fabrication facility and hot cell facility. Instituted personnel monitoring programs using thermoluminescent dosimetry and breathing-:ene areosol sampling in 1967. - Served as secretary of-a plant safety committee which inspected all ' i operations and reviewnd detailed. written procedures for operators. l Served as member of a corporate safety committee which determined l 1 corporate policy regarding health and safety matters. Chronological Sutmary 1973-Present Consul tant, Pickar 2, Lowe and Garrick, Inc. 1972-1973 Consultant to Dr. G. Hoyt Whipple, Universit.y of Michigan. 1963-1970 Nuclear Materials and Equipment Corporation (NUMEC). i License administrator, plutonium fuel facility health and safety supervisor.
POTTER - 2 MEMBERSHIPS
' American Chemical Society.
American Nuclear Society. . Health Physics Society. , Certified by American Board of Health Physics. . REPORTS AND PUBLICATIONS Woodard, K. , and T. E. Potter, " Consideration of Source Term in Relation to Emergency Planning Requirements," presented to the Workshop of Technical Factors Relating Impacts from Reactor Releases to Emergency Planning, Bethesda, Maryland, January 12-13, 1982. Ga rrick, B. J. , S. Xaplan, G. Apostolak'is, D. C. . Iden, X. Woodard, and T. E. Potter, " Seminar: Probabilistic Risk Assessment of Nuclear Power Pl ants," PLG-0141, July 1980. 3 Garrick, B. J. , S. Kaplan, G. E. Apostolakis, D. C. Bley, and T. E. Potter, " Seminar: Probabilistic Risk Assessment as Applied to . Neclear Power Plants," PLG-0124, March 1980. > Woodard, K., and T. E. Potter, " Modification of the Reactor Safety Study Consequences Computer Program (CRAC) to Include Plume Trajecturies," presented to the 1973 ANS 25th Winter Meeting, San F ancisco, California, November, 11-15, 1979. Woodard, X., and T. E. Potter, " Assessment of Noble Gas Releases from the Three Mile, Island Unit 2 Act s' dent," prasented to the 1979 ANS 25th Winter Meeting, San Francisco, California, November, 11-15, 1979. Garrick, B. J. , S. Kaplan, P. P. Bientar=, K. Woodard, D. C. Iden, H. F. Perla, W. Of eter, C. L. Cate, T. E. Potter, R. J. Duphily, T. R. Rcbbins, D. C. Bley, and S. Ahmed, "0E _1A, Oyster Creek Probabilistic Safety Analysis," (Executive Summary, Main Report, Appendixes), PLG-0100 DRAFT, August 1979.- Woodard, K.,'and T. E. Potter, "Prcbabilistic Prediction of X/Q for Routine Intennittant Gaseous Releases," Transactions of the American Nuclear Society, Vol. 26, June 1977. 1 f
- e. m i J
T 19 . UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
' ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:
James P. Gleason, Chairman F;ederick J. Shon Dr. Oscar H. Paris
.. )
In the Matter of )
)
CONSOLIDATED EDISON COMPANY OF ) Docke t ' No s. NEW YORK , INC. ) 50-247 SP (Indian Point, > Unit No. 2) ) 50-286 SP
)
POWER AUTHORITY OF THE STATE OF ) Jan. 24, 1983 NEW YORK ) (Indian Point, Unit No. 3)- )
)
CERTIFICATE OF SERVICE I hereby certify that on the 24th day of January,1983, I caused a copy of Licensees' Testimony of William R. Stratten, Walton A. Rodger, and Thomas E. Potter on Question One to be served by first class mail, postage prepaid on the 1 following: l-l
-6 ,
t James' P. Gleason, Chairman Charles M. Pract, Esq. Administrative Judge Stephen L. Baum, Esq. Atomic Safety and Licensing Board Power Authority of the 513 Gilmoure Drive State of New York Silver Spring, Maryland 20901 10 Columbus Circle New York, New York 10019 Mr. Frederick J. Shon Administrative Judge Janice Moore, Esq. Atomic Safety and Licensing Board Counsel for NRC Staf f U.S. Nuclear Regulatory Office of the Executive Commission Legal Director Wachington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Oscar H. Paris Administrative Judge Brent L. Brandenburg, Esq. Atomic Safety and Licensing Board Assistant General Counsel U.S. Nuclear Regulatory Consolidated Edison Company Commission of New York, Inc. Washington, D.C. 20555 4 Irving Place New York, New York 10003 Mr. Ernest E. Hill Administrative Judge Ellyn R. Weiss, Esq. Lawrence Livermore National William S. Jordan, III, Esq. Laborato_y Harmon and Weiss University of California 1725 I Street, N.W. , Suite 506 P._O. Box 808, L-123 Washington, D.C. 20006 Livermore, CA 94550 Charles A. Scheiner, Co-Chairperson Docketing and Service Branch Westchester People's Action Office of the Secretary Coalition, Inc. U.S. Nuclear Regulatory Commission P.O. Box 488 Washington, D.C. 20555 White Plains, New York 10602 Joan Holt, Project Director Alan L?.tman, Esq. Indian Point Project 44 Sunset Drive New York Public Interest Research Croton-On-Hudson, New York 10520 Group 9 Murray fitreet Ezra I. Bialik, Esq. New York, New York 10007 Steve Leipzig, Esq. Environmental Protection Bureau Jef frey M. Blum, Esq. New York State Attorney New York University Law School General's Office 423 Vanderbilt Hall Two World Trade Center 40 Washington Square South New York, New York 10047 New York, New York 10012 Alfred B. Del Bello Charles J. Maikish, Esq. Westchester County Executive Litigation Division Westchester County The Port Authority of New York 148 Martine Avenue and New Jersey White Plains, New York 10601 One World Trade Center
..ew York, New York 10048 Andrew S. Roffe, Esq.
New York State Assembly Albany, New York 12248
i Marc L. Parris, Esq. Atomic Safety and Licensing Eric Thorsen, Esq. Board Panel County Attorney- U.S. Nuclear Regulatory Commission County of Rockland Washington, D.C. 20555 11 New Hempstead Road New City, New York 10956 Atomic Safety and Licensing Appeal Board Panel Phyllis Rodriguez, Spokesperson U.S. Nuclear Regulatory Commission Parents Concerned About Indian Washington, D.C. 20555 Point' P.O. Box 125 Honorable Richard L. Brodsky Croton-on-Hudson, New York 10520 Member of the County Legislature Westchester County Renee Schwartz, Esq. County Office Building Paul Chessin, Esq. White Plains, New York 10601 Laurens R. Schwartz, Esq. Margaret Oppel, Esq. Zipporah S. Fleisher Botein, Hays, Sklar and Hertzber9 West Branch Conservation 200 Park Avenue Association New York, New York 10166 443 Buena Vista Road Honorable Ruth W. Messinger Member of the Council of the Mayor George V. Begany City of New York Village of Buchanan District #4 236 Tate Avenue City Hall Buchanan, New York 10511 New York, New York 10007 Judith Kessler, Coordinator Graster New York Council Rockland Citizens for Safe Energy on Energy 300 New Hemstead Road c/o Dean R. Corren, Director New City, Neti York 10956 New York University 26 Stuyvesant Street David H. Pikus, Esq. New York, New York 10003 Richard F. Czaja, Esq. i Shea & Gould L Joan Miles 330 Madison Avenue Indian Point Coordinator New York, New York 10017 New York City Audubon Society 71 West '23rd Street, Suite 1828 Amanda Potterfield, Esc.
- New York, New York 10010 Johnson & George l
528 Iowa Avenue L Richard M. Hartzman, Esq. Iowa City, Iowa 52240 l Lorna Salzman l Mid-Atlantic Representative Ruthanne G. Miller, Esq. Friends of the Earth, Inc. Atomic Safety and 208 West 13th Street Licensing Board Panel
-New York, New York 10011 U.S. Nuclear Regulatory i Commission l Stanley B. Klimberg, Esq. Washington, D.C. 20555 Generel Counsel New Yr'rk State Energy Office 2 Rockefeller State Plaza L Albany, New York 12223 i
~
7 Mr. Donald.Davidoff Director, Radiological Emergency Preparedness Group-Empire State Plaza
. Tower Building , Rm.' 1750 Albany, New York 12237 Craig Kaplan, Esq.
National Emergency Civil Liberties Committee 175 Fifth Avenue, Suite 712 New York, New York 10010 Michael D, Diederich, Jr. . Esq. Fitgerald, Lynch & Diederich 24 Central Crive Stony Point,- New York 10980 Steven C. Sholly Union.of Concerned Scientists 1346. Connecticut Avenue, N.W. Suite 1101 Washington, D.C. 20036 Spence W. Perry Of fice of General Counsel Federal Emergency Management Agency 500 C Street, S.W. Washington, D.C.- 20472 Stewart M. Glass Regional Counsel Room 1349
-Federal Emergency Management Agency 26 Federal Plaza New York, New York 10278 Melvin Goldberg Staff Attorney New York Public Interest Research Group l
9.Murray Street l New York, New York 10007 Jonathan L. Levine, Erg. P. O. Box 280 New City, NEv York 10958 l
- 1 l l
}
C N Paul F. Cola'rulli s
. - _ . -. - . --}}