ML20040B078

Memorandum Supporting Motion to Compel Answers to Licensee First Interrogatories to Citizens for Nuclear Reactor Safety,Inc.Proposed Order & Certificate of Svc Encl

ML20040B078
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	01/20/1982
From:	Brittigan R ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE, DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20040B075	List: ML20040B074 ML20040B078
References
NUDOCS 8201250112
Download: ML20040B078 (25)

Text

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I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ARMED FORCES RADIOBIOLOGY Docket No. 50-170 -

p RESEARCH INSTITUTE (Renewal of Facility (TRIGA-Type Research Reactor) License No. R-84)-

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MEMORANDUM IN SUPPORT OF MOTION TO COM?EL ANSWERS TO LICENSEE'S FIRST INTERROGATORIES TO THE CITIZENS FOR NUCLEAR REACTOR SAFETY, INC.

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INTRODUCTION On August 31, 1981 this Board entered an Order which provided, inter alia, that the parties would file their first sets of interrogatories on September 30, 1981 and the answers to the first sets of interrogatories on October 30, 1981. All parties filed their first sets of interrogato-rica on September 30, 1981. The Licensee responded to the Intervenor's first set of interrogatories on October 30, 1981. The Intervenor , however , on October 31, 1981, requested a 30 day enlargement in which to respond to Licensee's first set of interrogatories. A purported response to Licensee's first set of interrogatories was ultimately filed by the Intervenor on December 29, 1981 and received by the Licensee 8201250112 820115 0 PDR ADOCK 05000 o j

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J t on January 4, 1982. Comparison of the Licensee's First Interrogatories to the Citizens for Nuclear Reactor Safety, Inc. (Exhibit A) with Intervenor CNRS's Response to Licensee ~'s First Set of Interrogatories (Exhibit B) discloses that many of the purported answers are evasive, incomplete, or both. Accordingly, Licensee has submitted herewith its Motion to Compel Answers to Licensee's First Interrogatories to the Citizens for Nuclear Reactor Safety, Inc. .

QUESTION PRESENTED Although 10 C.F.R. g 2.740(f) provides that " failure to answer or respond shall not be excused on the ground that the discovery sought is objectionable unless the person or party failing to answer or respond has applied dor a protective order," a condition not here present, it may be helpful to discuss the Licensee's purpose in submitting its first set of interrogatories before addressing the respon-siveness of the Intervenor's answers to specific interrogato-ries. The August 31, 1981 Order issued by this Board clearly contemplated submission of motions for summary disposition upon the conclusion of the second round of discovery and it is the intent of the Licensee to submit such a motion.

Many of the contentions involved in these proceedings are, in the opinion of the Licensee, either irrelevant to the safety of reactor operations or totally without scientific 2

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b d t merit. Because the contentions are broadly stated and set forth only the ultimate conclusions of the Intervenor, their

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relevancy and scientific merit, if any, can only be determined by examining the underlying hypotheses, calculations, and assumptions. The intent of the Licensee, throughout the first set of interrogatories, was to obtain additional informa-tion and more clearly focus the issues. As will be demonstrated below, the non-responsive answers submitted by the Intervenor completely frustrate this effort.

Answers which are incomplete or evasive must, under 10 C. F. R. s 2.740 (f) , be treated as failures to answer or respond. In determining whether an answer is incomplete or evasive, the Board should consider the fact that these proceedings have now been pending for more than a year and are rapidly approaching the originally established summary disposition phase. If the Intervenor has any bases for its contentions, such bases must now be disclosed. If, on the other hand, the Intervenor's contentions have no foundation, both the Licensee and the Board are entitled

, to be informed of that fact.

For the convenience of the Board and the parties, the following discussion is keyed to the number of the interroga-l tory and purported answer concerned.

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. SPECIFIC FAILURES TO RESPOND WHICH ARE THE SUBJECT OF THE ATTACHED MOTION INTERROGATORY 6. The Intervenor refused to respond to this interrogatory, asserting that it did not yet know what experts will be called. At this stage of the proceedings at least some information regarding anticipated expert testimony must be available to the Intervenor. Moreover, the Intervenor, in refusing to respond to interrogatories numbers 27, 28, 29, 30, and 34, specifically identified " experts" by name and stated that the named individuals would testify as to ,

those matters at the hearing. Licensee submits that the Intervenor's responses to the substantive interrogatories are utterly inconsistent with the position taken by the Intervenor with regard to interrogatory number 6.

INTERROGATORY 7b. The Intervenor failed to answer the inter-rogatory as stated. Although the Licensee specifically requested the basis and supporting documentation for the Intervenor's inference that the Licensee's "HSR erroneously assumes that during an inadvertent transient a peak fuel temperature would be below 100 C," the Intervenor in its response commented upon 600 C rather than 100 C. Moreover, the Intervenor refused to provide any supporting documentation.

The Intervenor's response is irrelevant to the interrogatory and must be considered evasive. If the Licensee and this 4

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4 i Board are to be expected to assess and evaluate'the Intervenor's inference, the underlying basis and references must be provided.

INTERROGATORY 7c. The interrogatory was not answered as stated. The Licensee specifically requested " references that state the cladding temperature accompanying the ' elevated fuel temperature' postulated by the petitioners." The Inter-venors responded by stating an assumption and referring to any elementary thermodynamics text. Although the Licensee does not expect or require specific references defin'ing the Adiabatic Model for Heat Transfer, the interrogatory clearly requires the Intervenor to cite the references, if any, which support their assumption that the " cladding temperature will essentially mirror the fuel temperature" during a pulse operation or inadvertent transient. Absent such information, neither the Licensee nor the Board can '

test the validity of the Intervenor's assumption.

INTERROGATORY 7d. The Intervenor failed to answer this interrogatory as stated. The Licensee requested a specific temperature associated with the Intervenor's alleged " elevated fuel temperature." The Intervenor's response merely cited a rise in fuel temperature of "several hundred degrees" without even specifying the units of measurements (types of degrees). Although the Intervenor provided a list of '

variables in an effort to justify its vague, non-specific response, the variables cited are irrelevant to both the 5  !

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i e I interrogatory and the Licensee's reactor. The " elevated

-fuel temperature" was postulated by the Intervenor in its contention. The Licensee submits that the Intervenor can and should be required to provide specific information regard-ing its contention. Otherwise, neither the Licensee nor the Board can assess the likelihood of occurrence or probable

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INTERROGATORY 7el. The Licensee agrees with the statement made by the Intervenor in response to this interrogatory.

Unfortunately, however, the response does not answer the l

i interrogatory as stated because the Licensee asked the Inter-l venor to provide a specific reactivity insertion which the l

Intervenor considers to be associated with its definition l of inadvertent transient. The Intervenor's response is evasive and fails to provide either the Licensee or the Board with any specific information regarding its contention.

I INTERROGATORY 7e2. The Licensee agrees with the statement l made by the Intervenor in response to this interroaatory.

Unfortunately, however, the response does not answer the interrogatory as stated because the Licensee asked the Inter-venor to provide a specific maximum power level which the Intervenor considers to be associated with its definition of inadvertent transient. The Intervenor's response is evasive and fails to provide either the Licensee or the Board with any specific information regarding its contention.

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a t INTERROGATORY 7e4. Although the Intervenor attempted to answer the interrogatory by asserting that multiple human and mechanical errors can, taken collectively, render some system safeguards ineffective, the response is totally irrele-vant because the Intervenor never even attempted to explain why the behavior of the reactor would differ.

INTERROGATORY 7i. Although the Licensee specifically requested that the Intervenor provide references, the Intervenor failed to do so. Accordingly, the response is incomplete and provides ,

no documentary basis upon which to evaluate the Intervenor's contention. .

INTERROGATORY 7n. Although the Licensee specifically requested that the Intervenor provide references, the Intervenor failed i to do so. Accordingly, the response is incomplete and provides no documentary basis upon which to evaluate the Intervenor's j contention. [

l INTERROGATORY 8b. This interrogatory was not answered.

The Intervenor failed to define in its words, " greater severity,"

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and its basis for the same. It is apparent that the response  !

i i to this interrogatory was evasive and does not provide the Licensee with any basis upon which to take additional action. l INTERROGATORY 8c. The Intervenor failed to answer this f l interrogatory as stated. Licensee requested a quantitative 7 i t

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, i likelihood with justification, for accidents proposed by Intervenor. The Intervenor did not even attempt to characterize and justify the relative likelihood of its proposed accidents that are of a different kind.

INTERROGATORY 9a. Intervenor failed to answer this interroga,-

tory as stated. Licensee requested bases and justification for " assuming that the failure of a storage rack would result in an alleged accident of a ' greater severity' than those in the HSR." Intervenor cites information which is irrelevant to the question of " greater severity," cites information dealing with more than one fuel element storage rack failure, ,

and fails to cite any supporting documentation or justification pertaining to the interrogatory.

INTERROGATORY 10a. Intervenor failed to provide a complete answer to this interrogatory. Intervenor provided an adequate definition of an " experiment failure," but failed to respond to "what specifically occurs during an experiment failure and what results are to be expected?"

INTERROGATORY 10h. Intervenor failed to answer this interrog-atory as stated. Licensee requested, "What spccific MPC levels (radionuclide and concentration) art the petitioners contending would be exceeded by the alleged failure of an experiment?". Intervenor stated that the answer, " depends on the specific experiment sanctioned by the AFRRI." The 8

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Intervenors' response provides no clarification whatsoever.

The Licensee is interected in specific consequences associated with experiment failures that the Intervenor considers possible and important.

INTERROGATORY 12a. The Intervenor failed to answer this interrogatory as stated. The Licensee requested consequence data for the alleged malfunction and presumed accident.

1 The Intervenor stated that it needs more detailed information about the physical layout and operational history in order to answer this interrogatory. Licensee contends that this information is available in the public recorJ or has been directly provided to the Intervenor in documents prepared and submitted in conjunction with this relicensing effort.

Moreover, the Intervenor cited the alleged malfunction as part of its contention. Unless the Intervenor has the minimal information required to answer this interrogatory, there would appear to be no basis for inclusion of the alleged malfunction as part of the contention. In any event, the Intervenor's response is clearly evasive and must be treated as a failure to respond.

INTERROGATORY 12b. The Intervenor failed to answer or even attempt to answer this interrogatory. The Intervenor's response in no way addressed the question, "How was this failure cited above detected?". The Licensee contends the l

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. i answer to this interrogatory is readily available and a matter of public record. The Intervenor should be compelled to answer this interrogatory which the Licensee contends is of significant importance to the associated contention.

INTERROGATORY 12c. The Intervenor failed to answer this ,

interrogatory as stated. The Licensee requested Intervenor's

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feelings which does not require more detailed information on the physical layout or operational history at AFRRI as Intervenor states. The Intervenor's response therefore is irrelevant, incomplete, and evasive.

INTERROGATORY 12d. The Intervenor failed to answer this ,

interrogatory as stated. The Intervenor stated that it needs more detailed infarmation about the physical layout and operational hiotory in order to answer this interrogatory.

Licensee contends that this information is available in the public record or has been directly provided to the Inter-venor in documents prepared and submitted in conjunction with this relicensing effort. Moreover, the Intervenor cited the alleged malfunction as part of its contention.

Unless the Intervenor has the minimal information required to answer this interrogatory, there would appear to be no basis for inclusion of the alleged malfunction as part of the contention. In any event, the Intervenor's response is clearly evasive and must be treated as a failure to respond.

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4 i INTERROCATORIES 13, 14, 15, 16, 17, 18, 19, 20, 21, and

22. The Intervenor has clearly failed to answer or even l

l attempt to answer any of these interrogatories. The Intervenor has incorporated in its Failure of Experiment contention a series of alleged " malfunctions of confinement safeguards at AFRRI." The relationship of the alleged malfunctions to the Intervenor's postulated " release of radiation-in excess of occupational and offsite limits" is not clear.

Interrogatories 13 through 22 inclusive were designed to elicit clarification of this relationship, if any. The Intervenor totally refused to respond to these interrogatories and, indeed, appears to have attempted to disassociate itself from its own contention. The Licensee and the Board are clearly entitled to know whether the Intervenor contends that any of the alleged malfunctions could give rise to the effect postulated by the Intervenor.

INTERROGATORY 23a. The Intervenor failed to answer this interrogatory as stated. The Licensee requested the Intervenor to cite the mechanism by which a damaged TRIGA fuel element's moderating characteristics would be changed and also specifi-cally requested documentation and references to support such a claimed change in the fuel's moderating effect.

Intervenor provided a treatise, with references, on neutron moderation for uranium-zirconium hydride fuel and cited 11

, i a claimed potential mechanism for altering the moderation characteristics of damaged TRIGA fuel but never cited any substantiation or supporting documentation to verify that its claimed mechanism is valid or is possible. That is, Intervenor cited a mechanism as requested but failed to cite specific references supporting and verifying that its proposed mechanism can indeed alter the moderating effect of damaged uranium-zirconium hydride TRIGA fuel as specifically requested. Therefore, the Intervenor has failed to provide a complete answer to this interrogatory.

INTERROGATORY 23b. The Intervenor has failed to answer this interrogatory due to its failure to provide specific references to support its claims as requested.

INTERROGATORY 23c. Intervenor fe . led to answer this interrog-atory. Licensee requested Intervenor's understanding of the differences between a thermal and fast reactor in order to gain an understanding of the Intervenor's claims and l contentions which address criticality and inherent shutdown mechanisms. Intervenor states that this interrogatory is

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irrelevant. Licensee contends that it is relevant because it has a direct bearing on the Licensee gaining an insight into the Intervenor's contentions and claims concerning the AFRRI reactor, as well as the Intervenor's understanding '

of the mechanisms controlling thermal reactors in general.

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INTERROGATORY 23d. The Intervenor has failed to answer this interrogatory due to its failure to provide specific references to support its claims as requested.

INTERROGATORY 23g. The Intervenor failed to answer this interrogatory as stated due to its failure to answer the .

question, "Would this be a positive or negative effect?",

and its further failure to cite supporting documentation for any claims.

INTERROGATORY 23h. The Intervenor failed to answer this interrogatory due to its failure to cite references or show calculations to support its claims as requested. ,

INTERROGATORY 23i. The Intervenor totally refused to answer this interrogatory. Rather, it stated that this interrogatory will be addressed by direct testimony at the hearing. The Intervenor's response is clearly evasive and in direct contra-diction to the purpose of the discovery phase of these proceed-ings. Absent a response to this interrogatory, the Licensee can neither proceed with the second round of discovery nor prepare its motion for summary disposition.

INTERROGATORY 24a. Intervenor failed to answer this interrog-atory as stated. Licensee specifically asked "What is the basis that the HSR or SAR does not consider multiple cladding failure accidents?" Intervenor's response stated that "They do not believe multiple cladding failure accidents are credible."

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} Licensee is at a loss as to whom the term "they" refers.

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l More significantly, the Intervenor never cites to any specific  ;

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l INTERROGATORY 24c. Intervenor failed to answer this interrog,-

l atory as stated. Licensee specifically requested Intervenor f' to identify and cite references to support claimed mhchanisms i

causing " cladding failure." The Intervenor responded by ,

j addressing predominantly mechanisms for causing fuel material i l

failure--not necessarily to include the cladding. '

INTERROGATORIES 24g, 24h, AND 241. Intervenor simply and . [

I f clearly failed to answer or attempt to answer these interroga- I t

i tories. [

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INTERROGATORIES 25a, 25b, AND 25c. Intervenor has failed I to answer these interrogatories. Although it states that, f I

"Liz Entwistle will respond," to date she has not done so. [

I INTERROGATORY 26a. Intervenor failed to answer this interrog- r atory. Licensee requested, "What specific waterborne radioac- '

tive emissions are generated by rou'ine operations?". Inter-i

} venor responded by quoting the Licensee's statement that ,

"no waterborne radioactive emissions are generated by routine .

operations," and then proceeded to ask a question. If the [

Intervenor does not accept the Licensee's statement quoted j l

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interrogatory, to disclose the basis for its contention.

INTERROGATORY 26q. Intervenor failed to answer this interrog-  ;

L atory. Licensee requested the Intervenor to discuss, "how corrective actions are inadequate to detect and prevent ,

l violations of regulatory limits?". Intervenor made,a philo-  ;

sophical or political statement but never addressed'the interrogatory as stated. If the Intervenor's contention ,

l t is based solely on the response provided to this interrogatory, it clearly should be seeking rulemaking action rather than  ;

intervention in license renewal proceedings. ,  !

INTERROGATORY 261. The Intervenor failed to answer this interrogatory. In response to this interrogatory the Intervenor  ;

i merely alluded to its responses to Interrogatory 26, parts f, g, and h. Review of the cited responses discloses that i the Intervenor only cited what it considers inadequacies, i

but did not provide an answer to Interrogatory 261, specifi-j cally; "What safeguards (equipment or procedure) would satisfy }

, the Petitioners that proper monitoring methods are in use  !

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1 at AFRRI for checking emissions from routine operations."  ;

, i j INTERROGATORIES 27, 28, 29, AND 30. Intervenor clearly failed to answer these interrogatories. In its response, ,

the Intervenor directs the Licensee to, "see testimony of Professor Ernest J. Sternglass to be presented at the hearings." f i

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o s specific information and insight into the Intervenor's concerns, contentions, and the bases for its concerns and contentions.

Through evading the Licensee's interrogatories, by attempting to defer response until the hearing, the Intervenor clearly violates the purpose and intent of the discovery phase and ,

provides no opportunity for the Licensee to gather information and evidence for summary disposition. Indeed, the I'ntervenor's refusal to respond leaves the Licensee totally at a loss as to what it must defend against.

INTERROGATORIES 31b AND 31c. Intervenor failed to answer these interrogatories as stated. Intervenor's response .

to Interrogatory 31 is accepted as an adequate response to Licensee's interrogatory 31a. However, Interrogatories 31b, and 31c, are never addressed or answered. Furthermore, e

Interrogatory 31 was intended to elicit information which would disclose whether this extremely serious allegation has any foundation. If the allegation of the Intervenor is, as Licensee believes, utterly without foundation, that fact should now be disclosed and the inflammatory contention summarily dismissed. If the Intervenor has a foundation for its contention, it must disclose that information in response to the Licensee's interrogatory.

INTERROGATORY 32b. Intervenor failed to answer this interrog-atory as stated. Intervenor postulates, without citing 16

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, i supporting documentation as requested, that " internal gap pressures could rise to 1800 PSI or more depanding on the l temperature elevation." The Intervenor's response to this very specific interrogatory is vague and evasive. Porthermore, Intervenor failed to " prove," through calculations and refer-ences, that its claimed peak internal gap pressures could t

indeed be achieved for a LOCA at the AFRRI reactor facility, as specifically requested by the Licensee.

INTERROGATORY 32c. Intervenor has failed to completely i

answer this interrogatory since it refused to provide a  :

i response to a significant portion of the interrogatory. ,

Specifically, "Have any resulted in fission product release  !

to the unrestricted area environment? If the answer is yes, please provide documentation or reference it." The Intervenor's limited response does not afford Licensee suffi-cient information to proceed with the second round of discovery or prepare for summary disposition.

INTERROGATORIES 34a , 34b, 34c, 34d, 34e, AND 34f. The Inter-venor totally refused to answer these interrogatories. <

i In its response, the Intervenor states that, "We expect this topic to be discussed by several of our witnesses at the time of the full hearings, including Dr. Irving Stillman, Dr. Ernest Sternglass, and Dr. Irwin Brass." The purpore of discovery is to provide the Licensee with specific informa-tion and insight into the Intervenor's concerns, contentions, 17

o e and the bases for its concerns and contentions. By evading the Licensee's interrogatories, through deferment of informa-tion in response to interrogatories until the hearing, the Intervenor clearly violates the purpose and intent of the discovery phase and provides no opportunity for the Licensee to gather information and evidence for summary disposition. 1 Moreover, the Intervenor 's refusal to respond leaves- the  ;

1 Licensee totally at a loss as to what it must defend against.

This matter has now been pending for more than a year. ,

At some point the Intervenor must be required to disclose the bases, if any, for its inflammatory and prejudicial allegations. The Licensee submits that this time has come.

INTERROGATORY 35e. The Intervenor failed to answer this Interrogatory as stated due to its failure to " justify and support any statements" made in its response as specifically requested.

INTERROGATORIES 35f. AND 35h. The Intervenor failed to answer these Interrogatories as stated due to its failure to " cite references to support the elements given" as specifi-cally requested by the Licensee.

INTERROGATORIES 35g. AND 351. Intervenor clearly failed to answer or even attempt to answer these interrogatories as stated. Licensee requested that Intervenor "show and justify" with documentation that all elements and conditions necessary for an alleged explosive zirconium-air or zirconium-steam interaction are indeed present in the AFRRI TRIGA l

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l core. In response to this request, Intervenor provided evasive and incomplete responses replete with conjecture.

Absent supporting references, the Licensee cannot be expected to understand the basis for Intervenor's conjecture. If the Intervenor truly has a documented basis for its contention concerning explosive chemical interactions in the AFRRI TRIGA, then it should disclose this information to the Licensee and Board at this time, i

INTERROGATORY 36a(2). Intervenor failed to answer this interrogatory as stated due to its failure to "specify which monitor location" and "which year" as specifically requested .

by the Licensee.

INTERROGATORIES 36a(3) AND 36a (4) . The Intervenor failed to answer these interrogatories as stated. The Intervenor

! states that these interrogatories, "have already been discussed" but does not indicate where in the text the Intervenor feel's

, it has discussed the cited interrogatories. Review of the responses to previous interrogatories fails to disclose answers to the questions asked in Interrogatories 36a(3) and 36a(4). Accordingly, the Intervenor has failed to respond l to these interrogatories.

i INTERROGATORY 36c. The Intervenor failed to answer this interrogatory as stated. The Intervenor stated that data is contained in Tables 1 and 2 given above. Licensee agrees

! that data is indeed presented in Tables 1 and 2 but this i

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data has no relevance to Interrogatory 36c. Specifically, Licensee requested Intervenor to specify the residential area in which the population would ' highly probably' have received a dose in excess of 10 C.F.R. 20. This information is not contained within Tables 1 and 2 or elsewhere in the Intervenor's response.

INTERROGATORY 36d. The Intervenor failed to completely answer this interrogatory since it failed to answer or address the following parts: (1) "Please give any informati~on you possess as to why and under what conditions these measurements were made?", (2) "Were they due to environmental releases?",

(3) " Cite all references to support any statements made."

INTERROGATORY 369 Intervenor clearly failed to answer this interrogatory. Licensee here is requesting the Intervenor to explain and show how the AFRRI reactor core rad.3 active '

inventory has anything to do with routine emissions as the Intervenor contends. Intervenor does not answer this question and is evasive. Intervenor's response is totally irrelevant to the interrogatory.

REQUEST FOR STAY As indicated in the attached Declaration, counsel for the Intervenor and the Licensee have discussed the matters '

raised above and agreed to attempt to resolve the question of the adequacy of the Intervenor's response informally.

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i Although the motion to compel is being filed at this time  !

to comply with the time requirements set forth in 10 C.F.R.

2. 740 (f) and protect the rights of the Licensee, we request l

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j that the Board stay consideration of the motion for a period l 1

of 45 days. The Licensee will submit a supplemental memorandbm f; i to advise the Board of the progress made, if any, in the l

attempt to amicably resolve this matter. -

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Respectfully submitted l > I i

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f ROBERT L. BRITTIGAN j General Counsel l i Defense Nuclear Agency -

l Counsel for Licensee ,

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UNITED STATES OF AMERICA  !

NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of '

4 Docket No. 50-170 ARMED FORCES RADIOBIOLOGY -

RESEARCH INSTITUTE *

(Renewal of Facility (TRIGA-Type Research Reactor) l LICENSEE'S FIRST INTERROGATORIES TO THE t

CITIZENS FOR NUCLEAR REACTOR SAFETY, INC.

PLEASE TAKE NOTICE, that pursuant to 10 C.F.R. 2.740b and the Special Prehearing Conference Memorandum and Order i

. filed herein on September 1, 1981 you are hereby directed to answer in writing and under oath or affirmation the j following interrogatories within thirty (30) days. '

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INSTRUCTIONS A. These interrogatories are directed to the citizens

. for Nuclear Reactor Safety, Inc. (CNRS) and the answers hereto are to be completed to the best knowledge of CNRS, its officers, agents, attorneys, investigators, employees, technical advisors, and other representatives.

B. These interrogatories are continuing in nature i and CNRS is requested to provide by way of supplementary responses hereto in accordance with 10 C.F.R. 2.740 (e) such additional information as may hereafter be obtained EXHIBIT A Myfts .

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. a by CNRS or any person on CNRS's behalf, which will augment

' or otherwise modify any answers now given to these interrog-atories. Such supplementary responses are to be filed and served upon the Licensee within thirty (30) days after

receipt of such information, but no later than fourteen .

(14) days before the final prehearing conference.

C. Where an individual interrogatory calls'for an answer which involves more than one part, each part of i

the answer should be fully set forth and numbered or lettered to correspond with the appropriate sub-part of the interrog-atory.

D. For the convenience of the Atomic Safety and

' Licensing Board, the Nuclear Regulatory Commission Staff, l

and the parties hereto, the following interrogatories are keyed to specific contentions raised by CNRS. In ,

1 each case the relevant contention or applicable portion thereof is identified in capital letters immediately preced-ing the interrogatories.

! GENERAL QUESTIONS INTERROGATORY l Please list the name and address of all members of

! the petitioning group known as Citizens for Nuclear Reactor ,

Safety, Inc. (CNRS) .

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! INTERROGATORY 2 ,

Please list the nan.e and address of all persons who l are members of CNRS and participated in preparing the l contentions submitted to the Atomic Safety and Licensing 1

Board.  !

INTERROGATORY 3 e i 4

i Please list the name and address of any person or [

t firm used by the Citizens for Nuclear Reactor Safety, [

4 Inc. to prepare any contention before the Atomic Safety  ;

i and Licensing Board. Indicate which contention they were i

f involved with, the approximate time spent, how much, if any compensation was paid, l

I j INTERROGATORY 4 '

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] Please list the qualifications and experience of j l

each and every person involved with formulating the conten-  :

tious, basis for the contentions and any research required l to submit the contentions. If there were no technically i

oriented individuals preparing the contentions, please give the source of technical information required to formu-

, late the contentions.

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INTERROGATORY 5  !

Please list all persons answering these questions.

Indicate which specific question was answered.

If these persons are different than those listed in questions 3 '

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t i or 4, please give name, address, and qualifications to

{

! answer questions. .

{

I I I INTERROGATORY 6 I

Please list the names, addresses, and employers, f

c if any, of any expert or other technical witnesses which  !

i l CNRS intends to call in support of its contentions. With  !

l  !

respect to each such witness state the professional qualifi-  !

l cations, the contention with respect to which the witness  !

l  !

is expected to testify, and the substance of the expected .

l testimony.

t I

i ACCIDENTS I - THE ANALYSIS OF THE " FUEL ELEMENT CLAD FAILURE ACCIDENT," ONE OF THE TWO DESIGN BASIS ACCIDENTS  !

(DBAs) WITHIN APPLICANT'S HAZARD

SUMMARY

REPORT (HSR)  !

IS FAULTY IN THAT: THE ANALYSIS OF THE " FUEL ELEMENT  !

CLAD FAILURE ACCIDENT" ERRONEOUSLY ASSUMES THAT CLADDING l FAILURE DURING A PULSE OPERATION OR INADVERTENT TRANSIENT i WOULD OCQUR AT A PEAK FUEL ELEMENTTHAT TEMPERATURE OF LESS THAN 100 C. PETITIONER CONTENDS SUCH CLADDING FAILURE l WOULD BE MUCH MORE LIKEgY TO OCCUR AT ELEVATED FUEL TEMPERA- f TURES (IN EXCESS OF 400 C), RESULTING IN FAR GREATER GAP ACTIVITY AND FISSION PRODUCT RELEASES THAN THE HSR POSTULATES. , [

4 I

l l

J i

4 i I

1 s

- _ . ~

s INTERROGATORY 7

a. Where specifically in the AFRRI Hazard Summary Report or SAR does it assume that a cladding failure would occur at a peak fuel element temperature of less than 100 C7

b. What is the basis for t7e inference-that the HSR erroneously assumes that during an inadvertent. transient a peak fuel temperature would be below 100 C? State the references to support your basis,

c. Give references that state the cladding temperature accompanying the " elevated fuel temperature" postulated by the petitioners,

d. Specifically what temperature is the elevated fuel temperature supposed to reach?

e. What is an inadvertent transient? '

l. What reactivity insertion is to have occurred?

2. What is the maximum power level this "inadver-l tent transient" is supposed to reach?

l i 3. How is this " inadvertent transient" initiated?

4. How does this " inadvertent transient" behave differently such that the same temperature shutdown mechanism  ;

! that controls a planned transient would not control the i

inadvertent one? -

l 5

J I

l.

/

f. How is the elevated fuel temperature (in excess j i

of 400 C) determined, by the petitioners, during a pulse [

operation that assumes a cladding failure? Please provide  !

I a reference to any documents relied upon and a copy of l

t any calculations made. l l

g. Give the references that show the "much more i likely" occurrence of a cladding failure at fuel temperatures in excess of 400 C. I i

l

h. Quantitatively, how "much more likely" is a cladding  !

failure at a fuel temperature at 400 C? Above 400 C? ,

Please state references and provide a scenario for achieving i i l your cited temperature above 400 C. r i

1. How many degrees above 400 C is necessary to [

t cause the "more likely occurrence" of a cladding failure? i Please state references.

,j

j. What mechanism results in "the greater gap activity" (

L following a pulse rather than a steady state operation? i State conditions necessary for the mechanism and cite ,

references.

k. How does a greater gap activity by itself result l in a greater fission product release?

I

1. The statement " Petitioner contends tham .ach

• f cladding failure would be much more likely to occur at -

i i

elevated fuel temperatures (in excess of 400 C)," indicates {

I l

i t

i 6 f

, i

i

L i

that the petitioners feel that the probability of a cladding }

failure is the same or essentially the same below 400 C

! l without respect to temperature. Is this statement correct?  :

i

m. At what fuel temperature would the petitioners l 1 I recommend AFRRI calculate fission product releases resulting f

- t from a cladding failure and shat core history (considering  !

I maximum operations) should be used? e

n. What conditions (specifically temperature, pressure, etc.) are necessary to cause a cladding failure? State j references.

t

'n ACCIDENTS II - ACCIDENTS CAN DE EXPECTED TO OCCUR  !

I AT THE AFRRI REACTOR OF A DIFFERENT KIND AND GREATER SEVERITY

i THAN THOSE DESCRIBED IN THE HSR. SUCH ACCIDENTS SHOULD i BE MORE PROPERLY DESIGNATED DBA's TO ENSURE TiiAT SUCH  !

J , ACCIDENTS WOULD NOT RESULT IN RELEASES IN E.(CESS OF REGULATORY LIMITS. l i

> INTERROGATORY 8  !

a. What specific accidents other than described 1 i in the AFRRI Hazard Summary Report or Safety Analysis Report could be " expected to occur" and what is the basis f

I j for the statement that these accidents can be expected [

i to occur? i I

i

b. What is " greater severity"? Please state a specific I i

magnitude or level, and show the reason for selection l

. of this level.  !

t i

1 5 i  !

) i t

)

c. Give and justify the quantitative likelihood (probability) that these accidents of a different kind will occur.

d. What qualifications are necessary for a person to be considered qualified to " properly designate" an accident as a DBA for the AFRRI TRIGA. reactor?

e. What specific regulatory limits are AFRRI supposed to ensure would not be exceeded as a result of these alleged accidents?

f. Since one of the purposes of a DBA is to evaluate the hazard of one or two possible " worst case" accidents, what specific accident and what specific magnitude should

'be considered as the AFRRI DBA. Please site references, data, and any other information to substantiate that any proposed " worst case" accident is in fact" valid and more ,

relevant that those considered in the AFRRI Safety Analysis Report or Hazard Summary Report.

1) FUEL ELEMENT STORAGE RACK FAILURE. THE HSR DOES NOT PROVIDE REASONABLE ASSURANCE THAT SUCH AN ACCIDENT CANNOT OCCUR IN THAT: a) IT FAILS TO PUBLISH THE CALCULATIONS FROM WHICH IT CONCLUDES THAT A CONTACT CONFIGURATION OF THE TWELVE ELEMENTS STORED IN APPLICANT'S POOL WOULD NOT RESULT IN A CRITICAL MASS; b) IT DOES NOT CITE THE SOURCE FOR ITS STATEMENT THAT EXPERIENCE SHOWS IT TAKES APPROXIMATELY 67 CLOSELY PACKED FUEL ELEMENTS TO ACHIEVE CRITICALITY.

INTERROGATORY 9

a. What are the petitioners bases for assuming that the failure of a storage rack would result in an alleged 8

/

'i ,

j accident of a " greater severity" than those in the HSR?

j Please cite specific references from which the bases are [

t derived, l f

b. What is reasonable assurance? Please give a [

specific statement that defines reasonable assurance. ,

c. What specific regul~ation requires that the referenced .

j criticality calculation be pu,11shed? e i

d. What is the definition of contact configuration?

Explain why this configuration is the most impor' tant to analyze. Please cite specific reference or show calculation.  !

, e. What would provide the " reasonable assurance" l

L i

that a 12 element configuration would remain sub-critical?

i 1

f. What documents would petitioners consider an  !

i adequate " source" for a statement of experience?

1 9 Precisely what calculations and data would be '

f required to satisfy the petitioners that the failure of i i

, a storage rack, fully loaded with twelve TRIGA fuel elements 1

i would present no safety hazard to either operational person-nel or the general public?  !

t i

2) FAILURE OF AN EXPERIMENT. APPLICANT IIAS FAILED TO  !

S!!OW TIIAT SEVERAL INSTANCES OF MALFUNCTIONS OF CONFINEMENT  !

SAFEGUARDS AT AFRRI COULD NOT RECUR DURING AN EXPERIMENT l FAILURE, RESULTING IN TiiE RELEASE OF RADIATION IN EXCESS  !

OF OCCUPATIONAL AND OFFSITE LIMITS. -

l t

t l

9 i

i i

~ . _ _ _ . - . _ _ . . _ _ , . _ _ _ - -. .

--_.-.. - _ _ . , _ . . - - ~ . _ . , . , . ~ _ , . _ , . - _ . - . . . - .

INTERROGATORY 10

a. What is an " experiment failure"? Please define specifically what occurs during an experiment failure, and what results are to be expected,

b. What type of experiment failure could occur that-would release " radiation" to the environment?

c. What type of experiment failure could o'c, cur that would release " radiation" in excess of occupational limits?

d. At what points within the facility would these alleged releases in excess of occupational limits occur?

e. What activity levels would be required to exceed occupational limits within the facility?

f. What experiment failures at AFRRI would result in exceeding regulatory limits should there be a malfunction of confinement safeguards?

g. What activity levels are required at the location of an alleged experiment failure to exceed MPC for offsite releases?

h. What specific MPC levels (radionuclide and concen-tration) are the petitioners contending would be exceeded by the alleged failure of an experiment?

i. What specific safeguards do the petitioners feel could prevent the occurrence of an experiment failure ,

coincident with a malfunction of a confinement safeguard?

10

f SUCH MALFUNCTIONS INCLUDE:

i a) A BREACH OF CONTAINMENT CAUSED BY MISSION RUBBER GASKET  !

SEALING MATERIAL ON THE DOUBLE DOORS TO THE CORRIDOR BEHIND i THE REACTOR CONTROL ROOM, IN VIOLATION OF APPLICANT'S l TECHNICAL SPECIFICATION, I.A.4. (SEE, NOTICE OF VIOLATION, l APP. A, NRC INSPECTION REPORT DOCKET NO. 50-170, 10/13/78); j I

INTERROGATORY ll -

a. By what mechanism would radiation be rel, eased to the environment even if the rubber gaskets were totally f

] -

i removed from the double doors?

t i

! T

! i b) FAILURE OF THE REACTOR ROOM VENTILATION DAMPERS  ;

[

TO CLOSE ON AUGUST 26, 1975 WHEN THE CONTINUOUS AIR MONITOR ;

) WAS ALARMED (SEE, DNA ABNORMAL OCCURRENCE REPORT TO DIREC- I I TORATE OF REACTOR LICENSING, DATED SEPTEMBER 3, 1975, I a DOCKET NO. 50-170, 9/10/75.); j 4

INTERROGATORY 12 .

a. In the event of a damper failure during an experi- [

ment, what activity levels would be required, and in what ,

l areas, to cause a release to the environment above MPC?  !

b. How was the failure cited above detected? j f v

c. What additional safeguards do the petitioners  ;

feel are necessary to attenuate releases in the event J  :

! of a failure of the automatic dampers to close after a f

CAM alarm? .  !

l j

f t

v l

f 11  !

i 1

, - ~ , - , , - - - . _ - . , , -e-, r,- ,- -n-. , , - , . - . . . , - - - - - , , .

, , - e , ---w . .n..-, ----- , ,,w-,- - , , 9 e

d. How does the postulated concurrent failure of an experiment and damper closure lead to an environmental release?

c) FAILURE OF THE LEAD SHIELDING DOORS TO STOP OPENING AT THE FULLY OPENED POSITION (SEE DNA ABNORMAL OCCURRENCE REPORT, DATED JULY 27, 1976,* DOCKET NO. 50-170, 8/16/76); ,

l e

I INTERROGATORY 13

a. How does the failure of the lead shield doors to stop at the fully open position result in the release of isotopes above MPC?

b. How does the failure of the lead shield doors to stop at the fully open position result in an experiment malfunction?

c. How doec the failure of the lead shield doors '

to stop at the fully open position have any effect on anything related to personnel radiation exposure, offsite  ;

releases or confinement safeguards at AFRRI?

d. Please state how the lead shield door stop is part of a confinement safeguard or experiment failure,

e. However, even in light of Question c, what safe-guards would satisfy petitioner that in the event of a lead shield door failure to stop at the full open position there would be no release of isotopes to the environment above MPC?

12 l

i d) REACTOR CORE FOSITION SAFETi INTERLOCK MALFUNCTION ON FEBRUARY 1, 1973 (NOT RECORDED IN DOCKET NO. 50-170).

P INTERROGATORY 14

a. How does the failure of the core position interlock result in release of activity above MPC?

b. How does the core position safety inter'ock l relate to a confinement safeguard or experiment failure? ,

c. What safeguards would satisfy the petitioners that should the reactor core position safety interlock malfunction there would be no release to the environment i I above MPC? I PETITIONER CONTENDS THAT HUMAN ERROR COUPLED WITH FAILURE l OF BUILT-IN SAFEGUARDS COULD LEAD TO A SERIES OF EVENTS  !

RESULTING IN RELEASES OF RADIOACTIVITY IN EXCESS OF REGULA- .

TORY LIMITS AND CITES THE FOLLOWING PAST MALFUNCTIONS AT AFRRI AS EVIDENCE THAT SUCH FAILURES COULD OCCUR THERE IN THE FUTURE:

INTERROGATORY 15

a. Have any of the cited failures resulted in releases to the environment? If so, please document when, where, and by what means these releases occurred. Also list the amounts of specific isotopes released.

b. Specifically, what humsn errors coupled with what series of safeguard failures could lead to a release of radioactivity in excess of regulatory limits?

13

I

c. Which specific regulatory limits could be exceeded by the alleged series of events in b. above?

d. Please give any design (engineered) safeguards not provided at AFRRI that the petitioners feel would further enhance the safety of the AFRRI-TRIGA reactor '

by decreasing the likelihood.of an environmental release above regulatory limits. ', -

a) MALFUNCTION OF SAFETY CHANNEL ONE ON MARCH 15, 1980.

AN NRC INSPECTION ON MARCH 17, 1980 " REVEALED THAT SAFETY CHANNEL ONE WOULD NOT INITIATE A SCRAM IN ACCORDANCE WITH (APPLICANT'S) TECHNICAL SPECIFICATIONS",

INTERROGATORY 16

a. How would the failure of Safety Channel One result in or cause the release of levels above MPC?

b. What would cause the simultaneous failuro of redundant safety channel systems such that no safety channel would be operable and a release of activity would be possi-ble?

c. Are such failures, multiple or single, at any TRIGA reactor documented to have caused releases to the environment in excess of 10 C.F.R. 20? If your answer is yes please cite specific references.

l

d. Cite the source for the statement "An NRC inspec-tion on March 17, 1980, revealed that Safety Channel One would not initiate a scram."

14

e. Name the NRC inspector that discovered the alleged malfunction.

I

f. What additional safeguards would satisfy the petitioner that should Safety Channel One completely fail

a release to the environment above the limits in 10 C.F.R.' )

20 could not occur? -

e b) REACTOR EXHAUST SYSTEM MALFUNCTION ON AUGUST 9, 1979 '

CAUSED BY AN ELECTRICAL FIRE IN THE EF-1 CUBICLE OF THE l l MOTOR CONTROL CENTER, IN TURN CAUSED BY A POWER SURGE

j. DUE TO A FAULTY TRANSFORMER; INTERROGATORY 17 k

a. List references for documents that show any releases of any radiation from any TRIGA reactor in excess of 10 C.F.R. 20 caused by an electrical fire in a motor control center.

i

b. Which faulty transformer caused the alleged power surge? ,

c. What could AFRRI do to prevent the power surge?

1 i

d. What additional safeguards do the petitioners 1 feel could be placed in service such that failure of the

reactor exhaust system (EF-1) would not. result in the -

release of isotopes above MPC in the environment?

4 1

15 l

}

c) MALFUNCTION OF THE FUEL ELEMENT TEMPERATURE SENSING CIRCUIT CAUSED BY A " FLOATING SIGNAL GROUND," REPORTED BY DNA ON AUGUST 1, 1979; INTERROGATORY 18

a. How would the malfu'nction of a fuel element temper-ature sensing circuit cause a radioactive release.to'the environment above MPC?

b. How was the malfunction detected?

c. What additional safeguards could AFRRI install to prevent a malfunction of this nature?

d. What additional safeguards could AFRRI implement to satisfy the petitioners that in the event of such failures, radioactivity above MPC would not be released to the environ-ment?

d) MALFUNCTION OF THE POOL WATER LEVEL SENSING FLOAT SWITCH CAUSED BY WEAR,0N THE JACKETING AROUND THE WIRES LEADING TO THE SWITCH, REPORTED BY DNA ON JULY 31, 1979; INTERROGATORY 19

a. If the pool level float switch fails during an

~

experiment, how much activity do the petitioners postulate would be released to the environment? -

b. How was the malfunction discovered?

I 16 i

l

c. What can be done to prevent such a malfunction?

Ilow would such actions or changes reduce the likelihood of releases in general or above MPC in particular?

d. Were there any reported radioactivity releases associated with this malfunction? -

e. Please list and document any releases from any TRIGA reactor Jesulting from malfunctions of a water level sensing float switches.

f. What additional safeguards could AFRRI place in service to satisfy the petitioner that in the event of this type of malfunction radioactivity above MPC would not be released to the environment.

e) MALFUNCTION OF RADIATION MONITORING SYSTEM CAUSED BY TWO LOOSE WIRES IN THE CONTROL BOX AND RESULTING IN A FAILURE OF THE REACTOR ROOM VENTILATION DAMPERS TO CLOSE ,

(ON AUGUST 26, 1975 (REFERRED TO IN CONTENTION 2b), ACCIDENTS II, SUPRA) ;

INTERROGATORY 20 -

a. Please describe an experiment failure which, when coupled with the stated malfunction, would result l

in releases above regulatory limits. Show that such experi-ments are performed at AFRRI.

b. Should the experiment described by the response '

for Interrogatory 20a and the stated malfunction simultane-ously occur, what prevents an operator on duty from manually closing the ventilation dampers?

17 i

7

c. Please list and document any releases from any TRIGA reactor that have resulted from a malfunction of this type. .

d. What additional safeguards do the petitioners feel would alleviate any potential releases in the event of a malfunction of this type?

e f) MALFUNCTION OF THE FUEL TEMPERATURE - AUTOMATIC SCRAM SYSTEM ON JANUARY 29, 1974, CAUSED BY A BUILD-UP OF HIGH RESISTANCE MATERIAL ON THE MECHANICAL CONTACTS OF THE TZ OUTPUT METER; INTERROGATORY 21 .

a. How does the malfunction of a fuel temperature automatic scram system result in the release of isotopes in excess of MPC to the environment?

b. Were any releases reported in the same malfunction report? In any report in Docket 50-170 associated with this type function, were any releases reported?

c. Please list and document any releases from any TRIGA reactor associated with this type of malfunction.

d. How was this malfunction detected?

e. What additional safety features could be incorpor-ated into the automatic protection circuit to satisfy the petitioners that, should this malfunction occur again, no isotopes would be released to the environment in excess of 10 C.F.R. 20?

18 i

l I

g) MALFUNCTION OF THE REACTOR CORE POSITION SAFETY INTERLOCK SYSTEM ON FEBRUARY 1, 1973, CAUSED BY A FAULTY DE-ENERGIZING RELAY (REFERRED TO IN CONTENTION 2d), ACCIDENTS II, SUPRA).

INTERROGATORY 22

a. How does the failure of the core position interlock result in the release of radioactivity above MPC?

b. Did this malfunction result in any relea'se to the environment or were any releases reported in the same ,

malfunction report? ,

c. Has a malfunction of this nature at any TRIGA reactor resulted in or contributed to a release to the <

environment above MPC. If your answer is yes please cite ,

! specific references.

d. What safeguards would satisfy the petitioners that, should the core position safety interlock system ,

fail, there would be no core damage?

e. What safeguards would satisfy the petitioners that should the core position safety interlock system fail there would be no release to the environment above MPC?

APPLICANT HAS NOT SHOWN THAT THE TRIGA REACTOR'S NEGATIVE TEMPERATURE COEFFICIENT WILL AUTOMATICALLY SHUT DOWN THE REACTOR IN ACCIDENT SITUATIONS WITH DAMAGED FUEL ELEMENTS, '

WHERE THE MODERATING EFFECT OF THE HYDROGEN NUCLEI IN THE U-Zr-Hx ALLOY MAY BE SIGNIFICANTLY REDUCED AND THE  :

VALUE OF THE NEGATIVE TEMPERATURE COEFFICIENT IS CHANGED.  !

l 19

INTERROGATORY 23

a. By what mechanism does a " damaged" fuel element l change the moderating effect of UZrli fuel? Please cite references.

1

b. What happens to the reactivity characteristics -

of a thermal reactor core if.the moderating effect in l the fuel is significantly reduced? Please cite specific references.

c. What constitutes the difference between a thermal reactor and a fast reactor?

d. By what mechanism is the negative temperature coefficient changed, and by how much is it changed? Cite supporting documentation.

e. What constitutes a damaged fuel element in the context of this contention?

f. Ilow does the petitioner wish AFRRI to show that the reactor's negative temperature coefficient will automati- l l

cally shut down the reactor during accident situations (i.e. which model, calculation, or experiment) ?

g. Ilow much of a reduction in the moderating effect would the petitioners expect from a single damaged fuel element? Would this be a positive or negative effect?

Provide supporting documentation. ,

20

h. How many fuel elements would have to be damaged to reduce the moderating effect as postulated by the peti-tioners? Cite references or show calculations.

i. Please give specific references or calculations to show how a reduction in the moderating effect of hydrogen nuclei in a TRIGA fuel matrix will fail to decrease the .

average energy loss per average logarithmic energy decrement necessary to sustain power in a TRIGA reactor cord.

j. What safeguards or data would satisfy the petitioners that the reactor's prompt negative tempcrature coefficient will always shut down (insert large quantities of negative reactivity) the reactor automatically? .

4) MULTIPLE FUEL ELEMENT CLADDING FAILURE ACCIDENTS HAVE NOT BEEN CONSIDERED IN THE HSR. SUCH ACCIDENTS COULD RESULT FROM: a) DEFECTS IN THE MATERIAL INTEGRITY OF i

THE FUEL ELEMENTS THEMSELVES; b) AN UNCONTROLLED POWER EXCURSION IN THE REACTOR CORE; c) LOCA: d) SABOTAGE, AIRCRAFT '

COLLISION OR NATURAL i"ACT OF GOD") ACCIDENT.

INTERROGATORY 24

a. What is the basis that the HSR or SAR does not consider multiple cladding failure accidents?

b. Have there been any prior multiple fuel element cladding failure accidents in any TRIGA reactor? Cite supporting documents.

c. How is an " uncontrolled power excursion" defined in a TRIGA reactor and what mechanism causes cladding failure for such a hypothetical uncontrolled power excursion in TRIGA fuel? Cite references.

21

1

d. How would a loss of coolant accident result in the AFRRI TRIGA reactor result in multiple cladding failures?

i

' Cite references or provide any calculations performed.

e. In the event of a LOCA what would be the power l

history required to breech the integrity of the fuel's .

cladding? Does the AFRRI TRIGA reactor have this hypotheti,-

cal power history required to achieve these temperatures?

f. What postulated historical basis (oper ati'onal) for a 1MW TRIGA reactor would meet the requirement for l

a cladding failure specified in e. above? Cite references. ,

g. How many elements should be considered for a multiple cladding failure? Provide the basis for your  ;

. answer.

h. How many years of operation or how many kilowatt-hours of energy production are necessary to prove the integrity of the fuel elements contained in the TRIGA core?

I ' i. To what speci,fic " natural ("Act of God") accident" i does petitioner refer in this contention?

EMERGENCY PLAN - THE EMERGENCY PLAN PuBPARED BY APPLICANT IN CONJUNCTION WITH ITS LICENSE RENEWAL APPLICATION DOES NOT COMPLY WITH THE STANDARDS SET FORTH AT 10 C.F.R. PART 50, APPENDIX E.

i INTERROGATORY 25 ,

$ a. Specifically which section of 10 C.F.R. 50 requires 1 that the information set forth in 10 C.F.R. PART 50, APPENDIX E for a research reactor be included in its emergency j plan.

1 22 i

1 i

. . - , ,_. -- . ,-.._,,-.._nn, _. . . , ., . - - , _ , , . _ , , - - - , , - - . , , , , _ . . . , . , . , , . . , .

_n...

b. Please document why, in light of USNRC Regulation Guide 2.6, the AFRRI emergency plan is not adequate.

c. What are the qualifications of the petitioners to assess that the emergency plan fails to provide " reason-able assurance" that appropriate measures will be taken ,

to protect the public health *and safety in the event of

^

offsite releases?

d. Please substantiate the statement made in Contention A.3, first paragraph, that the accidents described in ACCIDENTS I and II, SUPRA would cause a major offsite 1

release and necessitate the implementation of the emergency  ;

l plan. What constitutes a major offsite release in this context?

ROUTINE EMISSIONS I - APPLICANT HAS NOT DEMONSTRATED THAT AIRBORNE AND WATERBORNE RADIOACTIVE EMISSIONS FROM ROUTINE OPERATIONS AND DISPOSAL OF SOLID WASTES WILL BE MAINTAINED WITHIN THE LIMITS OF 10 C.F.R. PART 20 IN THAT ACTUAL AND PROBABLE VIOLATIONS OF THESE REGULATORY LIMITS HAVE TAKEN PLACE ON THE OCCASIONS LISTED BELOW AND APPLICANT'S RADIATION MONITORING METHODS AND CORRECTIVE ACTIONS ARE INADEQUATE TO DETECT AND PREVENT THEIR RECURRENCE.

INTERROGATORY 26

a. What specific waterborne radioactive emissions are generated by routine operations? ,

b. Which specific part of 10 C.F.R. 20 does the AFRRI reactor exceed with respect to the disposal of solid waste generated by routine operations?

23

c. What types of solid waste are generated by the reactor during routine operations? That is, which specific isotopes are produced and disposed of that would be beyond the limits of 10 C.F.R. 20?

d. What is a " probable violation?"

e. Which specific regu'latory limits allegedly have been violated by AFRRI? \

f. Which specific radiation monitoring methods are allegedly inadequate? Justify your answer.

g. How are " corrective actions" inadequate to detect and prevent violations of regulatory limits?

h. 'Which specific corrective action is inadequate to prevent which specific recurrence of which specific regulatory limit?

1. What are the petitioner's basis for stating that -

)

" probable violations" have occurred? I l

j. List all sources of data, information, or reasons I they believe that the ' alleged " probable violations" have occurred, include names of persons or list specific documents and the dates that show these " probable violations" have occurred.

k. The contention above states'that actual violations of 10 C.F.R. 20 have taken place at AFRRI. Please list .

specific violations including the isotope involved, the amount released, the date, the NRC inspector filing the 24

Notice of Violation, and the NRC report number. Do not i include " probable violations" requested above.

t

1. Khat safeguards (equipment or procedures) would satisfy the petitioners that proper monitoring methods [

i are in use at AFRRI for checking emissions from routine  !

operations. ,

- r e

1) APP.'sICANT'S EQUIPMENT, METHODS, AND REPORTING SYSTEM FOR MEASURING RELEASES INTO THE MONTGOMERY COUNTY SANITARY SEWERAGE SYSTEM AND AT ITS PERIMETER AND OFFSITE MONITORING STATIONS DO NOT PROVIDE REASONABLE ASSURANCE THAT VIOLATIONS OF REGULATORY LIMITS HAVE IN ALL INSTANCES BEEN OR WILL  ;

BE DETECTED. L i

INTERROGATORY 27

a. It is contended that AFRRI's equipment, methods  ;

and reporting systems do not provide " reasonable assurance" f

that violations of regulatory limits will be detected. ,

Which specific equipment, methods, and reporting systems do not provide "reasona,ble assurance" that alleged violations ,

have been or will be detected?

r

b. What specifically constitutes " reasonable assurance"? l

c. From the petitioners' point of view, what type of equipment and methods for any offsite monitoring would i be necessary to provide " reasonable assurance" that regula-tory limits if exceeded would be detected? Please provide detailed characteristics of the proposed equipment and methods.

25 ,

l

d. Which specific regulatory limits are the petitioners contending that AFRRI did or will exceed?

e. What is the basis for the contention that violations of regulatory limits have occurred and have not been detected?

Justify any statements made.

f. List specific sourc'es for the contention that regulatory limits have been violated and that such. violation was not detected. Include names or specific documents and dates.

9 What equipment, methods, and reporting systems and safeguards would provide reasonable assurance to the petitioners that violations of any regulatory limits would be detected?

h. What isotopes, subsequently released into the Montgomery County Sanitary Sewerage System, are generated by reactor operations?

ENVIRONMENTAL MONITORING IS INADEQUATE TO DETERMINE RADIATION DOSES TO THE PUBLIC DUE TO INHALATION OR INGESTION BECAUSE:

a) FILM DOSIMETRY DETECTS ONLY EXTERNAL GAMMA RADIATION.

b) THE PARTICULATE RADIOACTIVITY MONITOR FOR AIRBORNE

' EFFLUENTS (i.e. A PANCAKE-PROBE C-M COUNTER) IS NOT ISOKINE-TIC, AND THEREFORE CANNOT BE USED FOR MEANINGFUL EVALUATIONS.

APPLICANT'S ONLY OTHER STACK EFFLUENT MONITORING SYSTEM, THE RADIOACTIVE GAS MONITOR, IS LIKEWISE NOT RELIABLE l FOR PARTICULATE SAMPLING. (SEE, ENVIRONMENTAL RELEASE REPORT ISSUED 12/14/71, COVERING PERIOD 1/1/70 - 9/30/71, AND INSPECTION REPORT NO. 50-170/77-01-03.) '

INTERROGATORY 28

a. What specific isotopes are to be detected by the environmental monitors?

26

b. What type of dosimetry would be adequate to detect radiation doses due to inhalation or ingestion? Please elaborate.

c. What Federal regulation requires a particulate monitor at AFRRI?

d. What references state that a gross activity detector [

is required to be isokinetic? ,

e. How does detector type determine isokinetic charac-teristics?

f. What evaluations are required of a gross activity i

detector?

g. What is the petitioners concept of the purpose and

'use of a particulate monitor for airborne effluents as used at AFRRI.

h. Why is the stack gas monitor not reliable for ,

particulate sampling? I

1. What is " meaningful evaluations?"

j. Cite specific sources to support the statement that the environment monitoring is inadequate at AFRRI.

k. What type of environmental monitoring and stack effluent monitoring equipment would be adequate, from the petitoners' point of view, to determine radiation doses to the public due to inhalation or ingestion? Please ,

r be specific and provide the minimum characteristics of such equipment.

27 t'

r C) APPLICANT WAS' CITED BY THE NRC FOR A VIOLATION OF ENVIRONMENTAL S AMPLING AND ANALYSIS PROCEDURES. THE VIOLA-TION NOTICE OF GROSS BETA EFFLUENT ANALYSIS, BASED ON AN NRC INSPECTION CONDUCTED JANUARY 12-14, 1977, CITED APPLICANT FOR CALCULATIONAL OMISSIONS, METHODS FOR PREPARING

'AND ANALYZING SAMPLES, AND INSTRUMENTATION USED. THE GROSS BETA MEASUREMENTS WERE MADE WITHOUT THE USE OF A BETA SELF-ABSORPTION CORRECTION IN THE PRESENCE OF SIGNIFI-CANT AMOUNTS OF SUSPENDED SOLID MATERIAL. (SEE NRC INSPEC-TION REPORTS NO. 50-170/77-01-02 AND 50-170/77-01-03.)

MOREOVER, APPLICANT'S " ENVIRONMENTAL SAMPLING AND ANALYSIS

• PROGRAM DOES NOT PROVIDE ADEQUATE INFORMATION ON,HOW QUARTERLY L ENVIRONMENTAL SAMPLES OF WATER, SOIL, AND VEGETAT_ ION ARE PREPARED AND ANALYZED, NOR DOES IT. PROVIDE THE RAW DATA COLLECTED OVER THE PAST TEN YEARS.

INTERROGATORY 29

a. What Federal regulation requires AFRRI to publish the methods of sample preparation and analysis?

b. What type of " raw data" should be published by AFRRI?

c. What Federal regulation requires the publication of this raw data?

d. What information would be necessary to be considered adequate?

e. What was the net change in the activity released to the environment as a result of changing the beta self absorption correction factor in the effluent analysis?

f. Specifically how could AFRRI change the currently in use methods and equipment such that the environmental ,

monitoring would be deemed adequate by the petitioners?

D) THE " CONCENTRIC CYLINDER SET MODEL" USED BY APPLICANT TO DERIVE ITS DOSE ASSESSMENTS TO THE ENVIRONMENT, AND FROM WHICH IT CONCLUDES ITS EFFLUENTS ARE WITHIN REGULATORY LIMITS, IS AN UNREALISTIC MODEL.

28

INTERROGATORY 30

a. What specifically is unrealistic about the concen-tric cylinder set model? Cite supportive references for any statements made.

b. What dose assessment method would the petitioners' consider realistic? Explain" fully and provide characteris-tics and justification for proposed method. .

c. What model was used by the petitioners to determine that the effluents as projected by AFRRI were not within regulatory limits and that therefore the model used by AFRRI was unrealistic? Please cite the reference that e

insures that the model used by the petitioners is a realistic model?

d. Specifically which regulatory limits are of concern to the petitioners? -

2) AN NRC INSPECTION CONDUCTED JANUARY 10-12, 1979 REVEALED THAT, CONTRARY TO APPL,ICANT'S TECHNICAL SPECIFICATIONS GOVERNING DISCHARGE OF AIRBORNE RADIONUCLIDES, ARGON-41 AND OTHER RADIONUCLIDES WERE DISCHARGED AT GROUND LEVEL OUTSIDE THE REACTOR BUILDING FOR SEVERAL MONTHS THROUGH A LEAK IN THE VENTILATION EXHAUST STACK DRAIN LINE (SEE NRC INSPECTION REPORT NO. 50-170/79/-01). IT IS HIGHLY PROBABLE THAT THIS RESULTED IN RELEASES IN EXCESS OF THE MAXIMUM PERMISSIBLE CONCENTRATIONS SET FORTH AT 10 C.F.R.

PART 20, APPENDIX B.

INTERROGATORY 31

a. What is "higbly probable"?

b. Which specific maximum permissible concentrations set forth at 10 C.F.R. 20, Appendix B is AFRRI supposed to have " highly probable" exceeded?

29 i

i

c. What method was used to determine that it was l

" highly probable" that AFRRI exceeded MPC in 10 C.F.R. 20, Appendix B? ,

i I .

I APPENDIX "B" - CONTENTION 1 ACCIDENTS 1  ;

l THE ANALYSIS OF THE LOSS OF COOLANT ACCIDENT-(LdCA)

) AND THE TWO DESIGN BASIS ACCIDENTS (DBAs) WITHIN APPLICANT'S HAZARD

SUMMARY

REPORT (HSR) IS FAULTY IN THAT: .

1) IT ERRONEOUSLY CONCLUDES THAT IN EVENT OF ,

AN ACCIDENT DESCRIBED THEREIN AS " LOSS OF SHIELDING AND COOLING WATER," AIR CONVECTION COOLING WOULD BE SUFFICIENT TO PREVENT CLADDING FAILURE AND SIGNIFICANT PRODUCT RELEASE.

j INTERROGATORY 32 I

a. What is the basis for the conclusion on the part i

) of the petitioners that air convection cooling would not ,

be sufficient to prevent a cladding failure in the event of a Loss of Shielding and Cooling Water (LOCA) accident?

Please give references,to support any statements made.

b. AFRRI studies indicate that even in the event of such accident conditions the integrity of the cladding will remain. Since the petitioners (CNRS) indicate this is incorrect, please give (considerin.g the power history of the AFRRI core) the maximum internal pressure a single j fuel element would experience for a loss of coolant event.

i i

l t

i i 30 >

i I

Is this pressure sufficient to rupture the cladding?

Please give references or calculations to prove the above.

c. Do the petitioners have any documented evidence that any TRIGA reactor has had a cladding f ailure, caused by a LOCA? Have any resulted in any fission product release to the unrestricted area environment? If the answer is yes, please provide that document or reference it'.

d. How significant a fission product release does CNRS Inc. expect from their postulated LOCA induced cladding failure? Please be specific as to amounts and f.sotopes released to both the reactor room and the environment.

Please justify the answer with calculations and references.

PETITIONER CONTENDS THAT, IN THE EVENT OF A RAPID LOSS OF COOLANT WHILE THE REACTOR CORE IS IN THE PULSE MODE, THERE COULD BE A SUDDEN TEMPERATURE ELEVATION SUFFICIENT ,

TO CAUSE MULTIPLE CLADDING FAILURES AND FISSION PRODUCT RELEASES IN EXCESS OF THE LIMITS PROVIDED IN 10 C.F.R.

PART 20.

INTERROGATORY 33

a. Please describe in detail the events that could lead to a pulse causing a temperature elevation sufficient to cause multiple cladding failures during a LOCA. Show how these events occur with such timing to allow the alleged accident to occur. Please provide references to support any statements made. 1 1

31

4

b. Would the temperature reached during this alleged pulse be of such magnitude to cause cladding failure?

Please show how this is possible to occur and justify t with supporting documentation and references.  ;

c. Which specific limits of 10 C.F.R. 20 are proposed ' ',

l would be exceeded by the all'eged fission product release? -

)

i

d. What is the temperature history during the.511eged j transient and accompanying LOCA? Provide supporting documen-tation.

l BOTH OF THE DBA ANALYSES IN THE HSR (" FUEL ELEMENT DROP k ACCIDENT" AND " FUEL ELEMENT CLAD FAILURE ACCIDENT") ERRONE- -

OUSLY CONSIDER ONLY THOSE RADIATION DOSES TO HUMANS THAT WOULD RESULT FROM SUBMERSION EXPOSURE TO THE NOBLE GASES ,

l RELEASED. ,

PETITIONER CONTENDS THAT IF SUCH ACCIDENTS WERE TO I 1 OCCUR, INDIVIDUALS WOULD RECEIVE ADDITIONAL EXPOSURE DUE *

' TO INTERNAL EMISSIONS OF THE NOBLE GASES, SUSTAINING INJURIES FAR GREATER THAN THOSE PREDICTED IN THE HSR.

i INTERROGATORY 34 .

a. Please show how the dose to humans would be increased l

by considering internal emissions from the noble gases.  ;

i Substantiate with references. }

) b. Specifically, how much additional exposure would j r

an individual receive due to internal emissions resulting j from the noble gases?

c. Please be specific in the statement that " individuals would receive additional exposure due to internal emission of the noble gases, sustaining injuries far greater than r

I 32

those predicted in the HSR." Exactly how much is "far greater"? Please show and justify.

d. Please explain why in light of the given thyroid dose (SAR Table 5) resulting from a described accident (Fuel Element Drop) the petitioners contend that only radiation doses from the noble gases were considered as possible accident consequences. e

e. Please clarify why the petitioners consider the thyroid dose (Iodine) given in the SAR for acciden"t conditions as submersion exposure to the noble gases.

f. What is the petitioners' basis for contradicting the Health Physics standards for exposure to noble gases?

(For example, NCRP and ICRP guidelines)

TWO MAXIMUM CREDIBLE ACCIDENTS (MCAs) BEYOND THE DESIGN BASIS OF THE REACTOR (CLASS 9 ACCIDENTS) : A) POWER EXCURSTON ACCIDENT (PEA) RESULTING IN MULTIPLE CLADDING FAILURES AT AN ELEVATED TEMPERATURE WITH REDUCTION IN THE THERMALIZING EFFECT OF HYDROGEN, FOLLOWED BY AN EXPLOSIVE ZIRCONIUM-STEAM INTERACTION; AND,B) LOCA RESULTING IN MULTIPLE CLADDING FAILURES AT AN ELEVATED TEMPERATURE, FOLLOWED BY AN EXPLOSIVE ZIRCONIUM-AIR INTERACTION.

INTERROGATORY 35

a. What constitutes a "significant offsite release"?

b. Please be specific and show how significant the offsite releases would be. For example, which isotopes and how many micro-Curies would be releasca to an unrestricted area?

33

o .

L

c. Please demonstrate how "such accidents" would result in "significant offsite" releases. Justify and cite any supporting documentation.

d. Please show specifically how the petitioners' alleged power excursion accident (PEA) can occur at an -  !

elevated temperature with a reduction in the thermalizing i f

effect of hydrogen in the AFRRI TRIGA reactor. ( l

e. What specifically does CNRS propose happens if the thermalizing effect of hydrogen in the AFRRI.TRIGA reactor is reduced in a powar excursion accident? Please  :

justify and support any statements. f

f. What are the necessary elements and conditions [

t for an explosive Zirconium steam reaction? Please cite  ;

i references to support the elements given.

g. Please show and justify that all elements and (

conditions necessary for an explosive Zirconium-steam f t

reaction are present in the AFRRI TRIGA core. Please  ;

give references that support the elements and conditions {

i stated as being present are in fact present.

h. Please give the necessary elements and conditions [

for an explosive Zirconium-air interaction. Please be specific and give references to support the elements given,

i. Please show and justify that all elements and ,

conditions necessary for an explosive Zirconium-air interac-tion are present in the AFRRI TRIGA core. Please give i

I 34 i

b

references that support the elements and conditions stated as being present are in fact present. (This may include the loss of cooling water such that air is present.)

j. If the two maximum credible accidents given in

2) are, as stated, beyond the design basis of the reactor (class 9 accidents), how can~the petitioners state that ,

u they can be expected to occur? Please be specific in justifying this apparent discrepancy.

ROUTINE EMISSIONS I INTERROGATORY 36  ;

a. Please explain section 3 of Routine Emissions I contention #5. Please be specific as to:

1 (1) Specifically, where in the Environmental Report or EIA data does it state that the highest average unrestricted area exposure rate from airborne releases extends to residential areas? Historically, what was the highest reported dose in a residential unrestricted area?

(2) Has any environmental monitor since 1970 given a reported dose level that even ' approaches 0.5 rem / year?

If so, please specify which monitor location, which year? -

35  !

h (3) What bases were used to state that "it is j highly probable that such exposures have resulted and continue to result in doses to the public in excess of

.5 rem"? Please site specifics.

(4) What bases were used to questicn AFRRI's use and commitment to the ALhRA principle concerning emissions  ;

from operations? Please site specifics.

b. Please cite specific monitoring locations that i show all doses in excess of 10 C.F.R. 20 limits in any residential area resulting from operations under the reactor license.

c. Please specify the location of exposure rates in unrestricted areas discusscd in #4 of contention 5.

Please be more specific as to the dase rate of concern. ,

Additionally, please specify the residential area eluded to that population of which would " highly probable" have received a dose in excess of 10 C.F.R. 20.

d. Since it is implied in section 4 of contention 5 that measurements were made only "a few times a year" of the dose rates quoted, it is assumed that the dose i

rates mentioned were arrived at from measurements other than environmental station monitors; is this correct?

Please give any information you possess as to why and -

under what conditions these measurements were made. What is the source or cause of these donc rates? Were they 1

, 36 n

due to environmental releases? Cite all references to support any statements made.

e. It is assumed that the .5 rem /hr is a typographical error; if not, please give the references that support the use of a number of this magnitude. .

Were the dose rates cited in contention 5.B.

f.

section #4 attributed to reactor operations in tNe reports mentioned? Were the rates from environmental monitor locations?

g. Please explain how the question and answer (#11 Autumn 1979) referred to in 5.B.3 which contains a statement of by-product material (fission products and decay products) contained in the AFRRI core has any bearing on section 3 of Contention 5.

h. Please list names and addresses of CNRS members ,

who live within 600 feet of the AFRRI stack. Please give the address of any residence within 600 feet of the AFRRI stack. ,

CONTENTION 7 - SECURITY NEITHER THE PHYSICAL SECURITY PLAN FOR THE FACILITY NOR APPLICANT'S HISTORY OF SECURITY VIOLATIONS AND SUBSTANDARD MANAGEMENT AND OPERATION PROCEDURES DEMONSTRATE THAT THE CONTROLLED ACCESS AREAS CAN BE PROTECTED FROM SABOTAGE OR CIVERSION OF SPECIAL NUCLEAR MATERIAL ACCORDING TO '

THE STANDARDS SET FORTH AT 10 C.F.R. PART 73.

l l

37

Ti!E DRAFT AUDIT REPORT OF TIIE AFRRI FACILITY PREPARED BY Tile DEFENSE AUDIT SERVICE IN 1979 CITES FREQUENT INSTANCES OF SECURITY AND MANAGEMENT VIOLATIONS, INCLUDING:

1) EIGilTEEN ACTIVATIONS OF TiiE FACILITY ALARM SYSTEM DURING A 34-DAY PERIOD, CAUSED BY PERSONNEL LEAVING WORK AFTER NORMAL DUTY HOURS PROM UNAUTIiORIZED EXITS. '

AUDITORS WERE TOLD BY AFRRI SECURITY PERSONNEL AND OTIIER AFRRI OFFICI ALS TIIAT INVESTIGATIONS WERE NOT MADE OF Tile .

ACTIVATIONS AND Tl!AT NOT ENOUGli SECURITY PEOPLE WERE ON DUTY TO INVESTIGATE EACII TIME TIIE ALARM WENT OFF;.

2) UNAUTIIORIZED PEOP.LE ENTERING TIIE PACILkTY BY FOLLOWING EMPLOYEES IN WilO USED TIIEIR MAGNETIC CARDS TO UNLOCK TIIE DOOR;

3) FAILURE TO ESCORT VISITORS ATTENDING WEEKLY SEMINARS AND PROVIDE Ti!EM WIT 11 DOSIMETERS;

4) FAILURE OF EMPLOYEES ENTERING AND EXITING Tile BUILDING AFTER IIOURS TO SIGN A LOG SIIOWING TIIEIR TIME =

OF ARRIVAL AND DEPARTURE.

5) VIOLATIONS OF APPLICANT'S ACCOUNTING AND DISPEN-SING PROCEDURES FOR CONTROLLED SUBSTANCES SUCII AS NARCOTICS.

INTERROGATORY 37

a. Please document the statement that standard operat-ing procedures fail to, protect the CAA trom sabotage or diversion of special nuclear materials.

b. Of the eighteen cited activations of the facility alarm system, how many involved the reactor controlled access areas?

c. Please provide documentation that any unauthorized people entering AFRRI also entered the reactor controlled access areas.

38

d. Please explain how any of the cited instances of security and management violations reflected on the integrity of the controlled access areas.

Submitted by ROBERT L._BRITTIGAN Counsel for Licensee a

39 i

i

UNITED S*.'ATES OF AMERICA NUCLEAR REGULATORY COMMISSION l

BEFORE Tile ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No. 50-170 ARMED FORCES RADIOBIOLOGY (Renewal of Facility. [

RESEARCII INSTITUTE License No. R-84)  ;

(TRIGA-Type Research Reactor) ,

CERTIFICATE OF SERVICE OF DUPLICATE SIGNED COPIES OF 30 September 1981 FILING i

I hereby certify that true and correct copies of the foregoing

" LICENSEE'S FIRST INTERROGATORIES TO TiiE CITIZENS FOR NUCLEAR REACTOR SAFETY, INC." were mailed this 30th day of September, ,

1981, by United States Mail, First Class, to the following:

Louis J. Carter, Esq., Chairman Administrative Judge ,

Atomic Safety and Licensing Board l 23 Wiltshire Road Philadelphia, PA 19151 Mr. Er nes t E. 11111 Administrative Judge Lawrence Livermore La bor a tory University of California P.O. Box 808, L-123 Livermore, CA 94550 Dr. David R. Schink Administrative Judge Department of Oceanography Texas A&M University College Station, TX 77840 Mr. Richard G. Bachmann, Esq.

Counsel for NRC Staff U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ,

r

Elizabeth B. Entwisle, Esq.

8118 Hartford Avenue Silver Spring, MD 20910 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Panel (5)

U.S. Nuclear Regulat.ory Commission '

Washington, D.C. 20555 -

Secretary (21) -

U.S. Nuclear Regulatory Commission ATTN: Chief, Docketing and Service Section Washington, D.C. 20555 SMK ROBERT L. BRITTIGAN Counsel for Licensee ,

4 6

2

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of  :

Docket No. 50-170  ;

ARMED FORCES RADI0 BIOLOGY  : L RESEARCH INSTITUTE  : (Renewal of Facility

License No. R-84) .

(TRIGA-Type Reactor)  :

r

- I INTERVENOR CNRS's RESPONSE TO LICENSEE's FIRST SET OF INTERROGATORIES NOW COMES the Intervenor in the above-captioned case and pursuant to 10 C.F.R. 62.740b, responds to the Licensee's l first set of Interrogatories as follows:  ;

INTERROGATORY 1 ' l i

Answered by Entwisle.

The Intervenor objects to this question. The only rele-vance it has to this proceeding is whether the Intervenor has legal standing. This issue has already been resolved, and the names and addresses of members were given in the affidavits that were submitted to establish the Intervenor's standing.

INTERROGATORY 2 Answered by Entwisle.

Elizabeth B. Entwisle, Esq.

8118 Hartford Avenue Silver Spring, MD 20910 ,

Irving M. Stillman, M.D., Ph.D.  !

5480 Wisconsin Avenue Chevy Chase, MD 20815 l INTERROGATORY 3 ,

Answered by Entwisle.

None. l INTERROGATORY 4 Answered by Entwisle.

Entwisle is an attorney and co-author of a study prepared for the President's Council on Environmental Quality, "NRC's Analysis of Nuclear Accidents: Is it Adequate?" March 1980.

EXHIBIT B 9U 't

,6##'~7, 1

. o Stillman is a physician and physicist who has participated in the Three Mile Island proceedings before the NRC and whose advanced interdisciplinary training in medicine and physics quali-fies him to speak about the biological impact of radiation asso-ciated with the operation of the Licensee's reactor.

l

' The contentions are based on examination of the Licensee's documents, such as the Draft and Final Audit Reports, and of documents in the Licensee's docket in the Public Documents Room of the NRC, 1717 H Street, N.W., Washington, D.C. '

( INTERROGATORY 5

'l Answered by Entwisle. -

Same as Interrogatory 4.

INTERROGATORY 6 Answered by Entwisle.

As of the present, the Intervenor has not determined which, if any, expert witnesses will be called to testify on any contsn-tion.

INTERROGATORY 7 Answered by Stillman. Unless designated otherwise, every Interrogatory hereinafter is answered by Stillman. ,

a. The Licensee describes two DBAs involving clad fail-ures in their Safety Analysis Report (SAR), namely: a " Fuel Element Drop Accident" and " Fuel Element Cladding Failure Ac-cident." In the Drop Accident the fuel element "has been allowed to decay after being taken out of the operating core and placed in storage. The fission products released from the gap will depend upon the temperature of the fuel following two weeks delay. This temaerature is expected to be less than 50 C."

(See quote in sal, pp. 6-16.)

For the Cladding Failure Accident they postulate a gag activity of only 1.4 percent (of the total radioactive inventory in the fuel element) and a maximal release of only 0.2 percent of the iodines. Since the temperature needed to volatilize iodine is 183*C (see SAR, pp. 6-16), it follows that this DBA is, presumed to occur at a temperature far below 180 C in order to meet their own critaria for maximal iodine release and total gap activity.

b. Throughout the Hazard Summary Report (HSR) , peak fuel temperatures above 600 C are never acknowledged. Furthermore, the selected (fuel element) gap activities and potential radio-active gaseous releases are only realistically compatible with much lower temperatures. For example, the very low values for 1

2

l

! radiciodines that would be released in a cladding failure accident.

i For the record, the Intervenor believes that during an inadvertent transient the peak. fuel temperature could rise several hundred I degrees.

I

c. The Intervenor makes the conservative assumption that the cladding temperature will essentially mirror the fuel tempera-ture, i.e., an adiabatic transfer of heat between them. Since the l Adiabatic Model for heat trans fer is given in any elementary thermo-4 dynamics text it is u,nnecessary to cite specific references. '

d. During an inadvertent transient, the fuel temperature can rise several hundred degrees depending on the exact core con-ditions (e.g., specific fuel element configuration, position of the control rods, mechanical malfunctions, operator errors, pre-

ceding power history, experiments in progress, loss of coolant, placement of the core within the reactor pool).

l e. An " inadvertent transient" or power excursion occurs l when there is a sudden core insertion of excess reactivity above

, cold critical to produce a large rapid increase in neutron flux.

1. According to the Licensee's " technical specifica-tions" the maximum step insertion above critical can be as much as 2.8 percent sk/k reactivity in the pulse mode without any potential danger.

2. If the AFRRI-TRIGA reactor is functioning within the permitted specifications, the maximum reactivity ,

transient that could possibly occur (according to the Licensee) would be that produced by the rapid insertion of the entire available amount of reactivity, namely, 3.5 percent ak/k ($5.00) excess reactivity above cold critical (with or without all experiments in place).

The maximum power level associated with such a transient is < 10,000 MW. The Licensee maintains that based on the

operating experience of the Advanced TRIGA Prototype Reactor (ATPR) in the General Atomic laboratories and

~

calculations using the Fuehs-Nordheim mathematical model, "it can be concluded that the rapid insertion of the total excess reactivity of 3.5 percent ak/k would not represent an undue risk."

3. There are several ways in which an " inadvertent transient" could be initiated and trigger a Power Excur- ,

sion Accident (PEA), such as:

(1) Improper fuel loading - a reactor operator inadvertently inserts a fuel element into the reactor core when it is already critical.

(2) Failure of an experiment - resulting in an instantaneous insertion of excess reactivity (i.e.,

the radioactivity associated with the experiment it-i self) to produce a dangerous transient.

(3) A stuck transient rod - if the most reactive control rod (i.e., the transient rod) is stuck out 3

of the reactor when the core is already loaded to its total excess reactivity.

(4) Pulsing with the transient rod greater than

$3.00 (2.1% Ak/k) reactivity - after withdrawal of the three standard control rods (previously with-drawn to achieve a steady state power greater than 1 MW).

4. An " inadvertent transient" requires, by defini-tion, an unplanned error or malfunction in the operation of the TRIGA reactor. Such human errors or equipment .

failures are very often multiplied during the course of any reactor accident. The history of nuclear reactor accidents, in general, is literally replete with examples of a single malfunction or human error compounded by a series of errors and additional malfunctions. Such a set of circumstances could prevent the safeguards that normally control a " planned transient" from~ functioning properly, i.e., "within the permitted specifications."

f. During a pulse operation that results in a PEA with cladding failures, the Intervenor postulates that both the fuel-moderator matrix and the claddings will have reached temperatures of 900 C or more. In spite of repeated assurances by the Licensee that the built-in and natural safeguards of the AFRRI-TRIGA would prevent fuel temperatures from rising to and above the safety limit, 1000 C, we contend that such safeguards are not foolproof (see our Interrogatory Answers to question 8a, part 3) and further that there must be circumstances under which such temperature elevations are possible. To document this contention the Licens-ing Board is referred to

1. The AFRRI Hazards Analysis reviewed by the Test and Power Reactor Safety Bianch Division of Licensing and Regulation,' Docket No. 50-163, p. 3, 1963, which states that if the three standard control rods are with-drawn to obtain a steady state power of 2 MW, then puls-ing with the transient rod of $3.00 (2.1% Ak/k) reactivi-ty could raise the peak fuel temperature "to about 900 C due to the temperature at the steady state compounded with the temperature increase from pulsing." Clearly, then, pulsing with a transient rod of more than $3.00 reactivity could raise the peak fuel temperature to well above 1000 C.

2. Calculations have been made to determine the temperature rise in a central TRIGA fuel element if the cooling water is lost instantaneously (see the 1963 GA-2025 Hazards Summary Report for the 250 KW Mark II TRIGA Reactor [ Pulsing] located at the Columbia University in New York City). These calculations cicarly demonstrate that a LOCA (in this tank-type TRIGA reactor) can result in fuel element temperatures up to 1200 C.

3. The many experiments routinely performed during the last twenty years at the General Atomic Laboratories 4

I in California with TRIGA fuel elements in which tempera- t tures of 1000 C or more were rather easily attained (re-1 gardless of negative temperature coefficients). ,

g. Experiments have been performed on hydrided 10 wt%

U-Zr fuel elements that were rapidly heated by induction. "Results '

indicated that within about 75 see the surface temperature reached ,

930* to 970 C with only minor hydrogen evolution. Abruptly '

thereafter, the surface was observed to crack parallel with the (

cylindrical axis, with strong outgassing rates" (see "The U-ZrH x l Alloy: Its Properties and Use in TRIGA Fuel," M. R. Simnad,, .

pp. 2-18). Another reference is H. H. Hausner and J. T. Schumar

(" Nuclear Fuel Elements," p. 84) where surface cracks appear-d in e l fuel element claddings when they were overheated to 900 C or more.

', In addition to these specific references, a clear general mechanism 1 is present for concluding that cladding failures are "much more l j likely" at fuel temperatures greater than 400 C, n&mely, at ele- I vated temperatures there would be a corresponding increase in the total gap pressure (produced by the rise in fission gas pressure, i  ;

i residual air pressure, and the peak equilibrium hydrogen pressure) -

l 1

that would put the cladding under much greater stress. ,

h. It is not possible to quantitatively assess the risk' of a cladding failure at any fuel temperature. If the Licensee  ;

knows some exact way of determining such risks (without knowing -

the actual probability forleach component event) then they should  ;

share that knowledge with the rest of the world. Temperatures -

of operating fuel elements well above 400 C may easily be achieved through pulse heating (see R. E. Taylor's " Pulse Heating of Modi .  ;

fied Zr-H," U.S. AEC Report NAA-SR-7736, North American Aviation, i 1962). Another scenario for fuel temperature elevation is de- '

i scribed under the LOCA-induced multiple cladding accident scenario  ;

(see Answer to Interrogatory 24, parts d, e, f) .

i. Cladding failures are more likely the higher the fuel ,

, element temperature. They are less likely at temperatures below  !

800*C and become much more possible at temperatures of 900 C or more.

j For references, see Interrogatory answers to both 7.g. l*

j (given above) and 24.c. (given below). ,

). Repeated activity in the pulse mode may result in  ;

pulse beating. If the peak fuel temperature stays below 550 C the emission of radioactive gases is largely controlled by recoil

, effects which are not very sensitive to the fuel temperature. 3 However, should the pulse beating result in temperatures above .

3 600*C, the process of gas emission into the gap becomes mostly

diffusion controlled and results in " greater gap activity." By

i contrast with the recoil mechanism, diffusional gas emission is  !

extremely temperature sensitive so that gap activity rises rapidly [

as a function of increasing fuel temperature.  :

! k. " Greater gap activity" implies larger partial pressures '

of the radioactive gases contained within the gap. If there is  !

no cladding failure and the cladding maintains perfect structural  !

integrity, then " greater gap activity" will not,of itself, result l i 5 i t

---,,-__-_ _ _ _ _ , _ . . - . ~ . , - , . . . , - - - . - . -

in a greater fission product release. However, any structural disruption resulting in cladding degradation would result in

" greater fission product release" if there were " greater gap ac-tivity." Thus cracks or penetrations in the cladding would per-mit the radioactive gases in the gap to stream out under pres-sure and if the total gap pressures become great enough to ap-proach 1,800 psi, then these excessive pressures would cause additional breaks in the cladding permitting more rapid release of the gap activity. Total gap pressures in excess of 1,800 psi could even cause complete rupture of the cladding without any prior deterioration. -

1. Below 400'C the possibility of a cladding failure is relatively independent of the fuel temperature (e.g. t the tempera-ture-independent recoil mechanism is operating at temperatures of 400 C or less).

m. 1,200 C. A core history of at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of 1 MW steady state operation.

n. A cladding failure is more apt to occur at fuel tem-peratures above 1,000 C and total gap pressures in excess of 1,800 psi. The greater the temperature and gap pressure, the more likely the cladding failure.

INTERROGATORY 8

a. Accidents that might occur other than those described in the AFRRI-HSR include

1. Fuel element storage rack failures. Because of their very high radiation levels, spent fuel elements are stored under water for shielding purposes. They are there-fore stored in aluminum racks under the pool of water in the reactor tank at the AFRRI facility. The Licensee states that " experience shows it requires approximately 67 fuel elements, of the design used at AFRRI, in a close packed array to achieve criticality." The Intervenor would like to know the exact nature of the " experience" that demonstrates the requirement for "approximately 67 fuel elements" to achieve criticality, given the fact that unlike ordinary fuel elements (in most power reac-tors) that contain only 3% enriched Uranium - 235, the TRIGA fuel elements contain nearly 20% enriched Uranium-235. Furthermore, even by this overly conservative esti- ,

mate, the AFRRI could conceivably accumulate this number or tuel elements in their reactor pool over the next 20 years if on-site storage (versus Away From Reactor) re-mains the guiding principle for the handling of high level radioactive waste. There is also the possibility that un-foreseen conditions may require the rapid discharge of the full core load of fuel (i.e., 85 fuel elements). What plans has the AFRRI made to handle such an emergency?

Would it really be safe to store this number of fuel ele-ments in their aluminum racks inside the reactor tank?

6 l

2. Failure of an Experiment resulting in a signifi-cant release of radioactive material. This can result from a runaway experiment undergoing activation either within the reactor core (such as the CET) or in the ex-posure rooms. There have been at least two documented malfunctions that could effect the release of signifi-cant amounts of radiation (from a runaway experiment) into the reactor room, namely:

(i) a safety interlock malfunction that occur-red on February 1, 1973; (ii) malfunction of the lead door rotation on' July 27, 1976.

Technical Specifications, section I.A.4 sta'tes "The re-actor room shall be designed to restrict air' leakage when the positive sealing dampers are closed." To accomplish such containment "the door to the corridor.behind the reactor control room . . . is a double door that is sealed with compressible rubber gaskets and latched. The double doors at the opposite end of the corridor . . . is also sealed with a compressible gasket." Contrary to this specification, as of October 13, 1978, the above doors were not maintained as designed, in that gasket material was missing on both doors preventing fulfillment of the design function (see Notice of Violation, Appendix A, NRC Inspection Report Docket No. 50-170, 10/13/78). Hence any significant radiation release into the reactor room resulting from an experiment failure could have leaked out of the reactor room into the rest of the AFRRI facility.

Another example of this breach of reacter room containment occurred on August 26, 1975, due to failure of the Reactor Room ventilation dampers to close when the Continuous Air Monitor (CAM) was alarmed. Both of these examples plainly demonstrate that any radiation released into the reactor room has a distinct possibility of leaking out into the I

entire AFRRI building and even outside the building into the environment (given a large enough radiation release) thereby endangering the public health and safety (see the AFRRI Abnormal Occurrence Report to the Directorate of Reactor Licensing, dated 9/3/75) .

3. Failure of one or more of the " built-in safeguards,"

such as the:

(i) safety system channels (ii) safety system settings -

(iii) radiation monitoring systems (iv) " negative temperature coefficient of re-activity" mechanism for automatic shutdown.

The Licensee describes several built-in safeguards (listed above) that would either alert the reactor operator or some automatic mechanism to effect the necessary correc-tive measures (e.g., initiate a reactor SCRAM, engage the appropriate interlock system, add water coolant, close 7

the ventilation system, etc.) should an accident even threaten. However, the Intervenor contends that human

errors coupled with equipment failures can render these safeguards ineffective, as has occurred repeatedly in i

nuclear reactor accidents nationwide, such as

1 (i) the jumpered safety interlocks of Vermont l Yankee; (ii) the Dresden 2 blowdown in 1971; (iii) the Millstone seawater intrusion in 1972; (iv) the Brown's Ferry fire in 1975; '

(v) the inversion of control rods and the .

Rancho Seco control rod drive failure.in 1975; i

(vi) the relief valve that malfunctioned and stuck open during the accident sequence at Thr_ee-Mile-Island in 1979.

To demonstrate that such failures (both human and mechani-cal) can also occur with the TRIGA reactor at the AFRRI, we shall cite several instances of relevant malfunctions i involving these safeguards (reported by the AFRRI to the i appropriate federal regulatory agency), including:

(a) On February 1, 1973 the Reactor Core Position Safety Interlock System that coordinates the lead door / core movement (to bring the door into near con a tact with the core shroud) malfunctioned due to a faulty de-energizing relay.

(b) On January 29, 1974 the Fuel Temperature-Automatic Scram System malfunctioned "due to the build-up of high resistance material on the mechani-cal contacts of the T2 output meter that initiated the automatic scram through a relay." '

(c) On August 26, 1975 the Radiation Monitoring System malfunctioned, i.e., the reactor room ventila-tion dampers failed to close when the Continuous Air Monitor sensing device was manually triggered. "In-spection revealed that two wires in the control box were loose" and that this was the apparent cause of the malfunction.

(.d) On July 10, 1979 there was a malfunction of the Pool Water Level Sensing Float Switch that moni-tors the reactor pool water level in case of an im-pending LOCA. "The malfunction was caused by wear on the jacketing around the wires leading to the switch which provided a path to ground, thereby cir-cumventing the switch function."

(e) On July 30, 1979 there was a malfunction of ,

the fuel temperature indicators (i.e., fuel element temperature sensing circuit) ostensibly caused by a

" floating signal ground with respect to the system ground." The Licensee admits that "since this system l monitors the principal safety parameter of the re-actor, it was felt that a more secure ground was j required."

1 (f) On August 9, 1979 the reactor exhaust system malfunctioned due to an electrical fire (in the EF-1 cubicle of the motor control center) caused by a power surge due to a faulty transformer.

8 l

. _ _ - _ . , _ . . . . ~ . - . . , . . _ _ _ _ ~ ,_

k (g) On March 15, 1980 there was a malfunction of [

Safety Channel One such that most of the scram indi-

! cators on the reactor control console were illuminated.

Further, an inspection on March 17, 1980 " revealed that  :

Safety Channel One would not initiate a scram in ac-  !

cordance with the Technical Specifications of Reactor  !

License R-84. "The cause of the malfunctiod' was attrib-  !

uted to a damaged operational amplifier on a Safety Channel One circuit board "when electrical power had i been reapplied to the console after a power outage." .

The Licensee alleges that even if there is a power  ;

excursion in the AFRRI-TRIGA and the built-in safeguards malfunction, the reactor will automatically shut down due ,

t to the prompt negative temperature coefficient.(i.e., -0.126%  ;

i ak/k decrease per 1 C rise in fuel temperature). This ,

automatic shutdown is entirely dependent on,the relative ,

amount and energetic state of the hydrogen nuclei within [

the U-ZrH alloy. Thus, any significant deviation of the  ;

i hydrogen farameters from their expected values or curves [

will cause a drastic change in either the prompt and steady- l l state negative temperature coefficients. Such deviation of i the hydrogen parameters are likely under accident condit(,ons where large internal pressures and elevated temperatures may produce phase changes within the U-ZrH alloy (see a dis-cussion of the " hydride phases" in "Th6 U-ZrH Alloy" by .

M. T. Simnad, February 1980). Such phase changes will af-  :

fect the vibration frequency, , of the hydrogen nuclei  !

i and thereby seriously alter the negative temperature coef-i ficients of reactivity which depend on the transfer of -

energy quanta (of magnitude h ) from warm or excited hydrogen  ;

nuclei to slow or thermal neutrons (via elastic collisions). l j 4. Multiple cladding failure accidents. [

l  !

j Such accidents'may result from any one or more of the i

! following.  ;

1 (i) Defects in the material integrity of the I'

=

fuel elements themselves.

(ii) Uncontrolled power excursion (or inadvertent .

transient) in the operating reactor core (PEA).

(iii) Sabotage or a natural accident (e.g., "act of [

God") involving the AFRRI-TRIGA Reactor.  !

J  !

! The Intervenor contends that cladding failures may occur -

l

! during operation of the TRIGA reactor secondary to inherent  !

defects or weaknesses in the material integrity of the fuel  !

elements themselves. These may go unnoticed in the re-quired annual fuel element inspections by the Licensee.

Note, the frequency of these inspections (for the AFRRI) j

. was decreased from six to every twelve months (in May 1972).  !

Whereas three cladding failures have already been reported l by General Atomic in their Torrey Pines TRIGA reactor (see  !

t 9

I i

l I

l

the AFRRI Safety Analysis Report), we contend that there is little reason to believe that they cannot happen in Betnesda, Maryland. In fact, there apparently have been at least two reported cladding failures at the AFRRI it-self. The first was on August 17, 1964 whereupon a tele-gram was dispatched to the AEC in which "the institute has determined that at least one TRIGA type pulsing fuel ele-ment exhibits cladding failure." The second occurred on October 19, 1967 in which "small amounts of gas bubbles were observed to be released from a C-ring fuel element."

The report (to the AEC on October 24, 1967) goes on to say that "the leaking fuel element was taken out . . . and transferred to the fuel element storage rack-in the pool."

Clearly there have been and will continue to be cladding failures within the core of the operating AFRRI-TRIGA re-actor, o The several ways in which PEAS can be initiated was described above (see answer to Interrogatory 7.e., part 3). The common factor in all of these initiating incidents is that there is a sudden insertion of excess reactivity (within an already critical reactor core) to produce a rapid, large increase in neutron flux (i.e., a prompt power excursion or transient) capable of causing cladding failures at elevated fuel element temperatures. As noted above, if the AFRRI-TRIGA reactor is functioning "within the permit-ted specifications," the maximum reactivity transient that could occur would be that produced by the rapid insertion of the entire available amount of reactivity, namely 3.5%

ak/k. Based on the operating experience of General Atomic',s ATPR and calculations rooted in the Fuchs-Nordheim mathe-matical model, the Licensee concluded that "the rapid in-sertion of the total excess reactivity of 3.5% ak/k would not represent on undue risk." This confidence, however, may be misplaced since the ATPR is certainly not identi-cal with the AFRRI-TRIGA and is under considerably more expert scrutiny and experimental control by the General Atomic scientists. Consider the past history of mechanical malfunctions self, including:

within the core of the AFRRI-TRIGA reactor it-(i) cladding-damaged fuel elements on August 17, 1964 and October 19, 1967; (ii) separation of the transient rod from its con-necting rod discovered on July 17, 1973; (iii) crack detected in the top weld of the transient, control rod on May 1, 1974; (iv) detection of a tilted fuel element within the reactor core, reported January 31, 1975; (v) misalignment of two fuel assemblies occurred on August 22, 1978.

Consequently, one cannot presume that the AFRRI-TRIGA will always function "within the permitted specifications" and therefore is subject to a serious Power Excursion Accident (PEA) involving one or more cladding failures.

10

. . t

)

The potential for sabotage or terrorist activity was dramatically pointed out in an April 3, 1979 Draft Audit Report by the Defense Audit Service (DAS), charging that frequent security and safety violations were being commit-ted at the AFRRI. Specifically, the draft audit states that although "NRC's inspections have generally shown that AFRRI's security and safety operations have been satisfac-tory . . . our review showed that frequent safety and security violations were being committed." Even Admiral Robert Monroe, former Director of the Defense Nuclear Agency (DNA), admitted (see The Washington Star, August 14, 1979) that the possibility for sabotage is real when he said: "If a group of heavily armed, desperate men stormed into the building, there'd be nothing out there to stop them." Clearly, any serious explosion within the reactor room that permitted release of the radioactive inventory into the Facility and beyond, would seriously threaten pub- ,

lic health and safety.

As for the possibility of an accident, consider an air-plane crash into the AFRRI Facility. There are two major airports (National and Dulles) within 15 miles of the National Naval Medical Center producing extremely heavy '

air traffic above Bethesda, which is more than five times what is considered safe (from plane crashes) for any nuclear site according to the American National Standards Institute.

In addition, a helicopter pad, located on-site at the medi-cal center, is less than one-third of a mile from the nuclear reactor. Other types of accidents are also quite possible. .

For example, less than a thousand yards from the AFRRI Facility a new Metro subway station and tunnel are being constructed. The Intervenor warns that not only the drill-ing and dynamite explosions during construction, but future train accidents, might result in conditions that predispose the AFRRI-TRIGA to a LOCA resulting from either rupture of the reactor tank itself, damage to the AFRRI cooling tower, or damage to any part of the pumping system. Such a con-struction or train accident could also affect one or more of the safeguard systems (listed above) thereby potentiat-ing a dangerous PEA.

5. Maximum Credible Accident (class 9 accident) re-sulting from an (i) explosive zirconium-steam (water) interaction (at fuel temperature >1,000 C) following a PEA- .

induced multiple cladding failure (without a LOCA), or (ii) explosive zirconium-oxygen (air) interaction--

(at fuel temperature >1,000 C) following a LOCA-induced multiple cladding failure.

The several ways in which a PEA could be initiated and lead to multiple cladding failures were outlined above (see answers to Interrogatory 7.e.3. and 24.c.). As indicated, in those multiple cladding failure accidents due to uncon-trolled, prompt, power excursions there is likely to be an 11

.~ - - _ _ _ _ _ . _ _

associated elevation of the fuel-moderator temperatures (within the damaged fuel elements) to 900"C or more. In particular, those fuel elements reaching temperatures above 1,000 C might produce total gap pressures (>1,800 psi) capable of rupturing their already damaged stainless steel claddings. Rupture of a fuel element cladding would ex-pose hot Zr H x (the major component of the fuel element) to the tank water. An NRC report indicates that the rate of a violent irconium-water (or steam) reaction becomes significant at about 900 C (see NRC memorandum to Roger Mattson from R. O. Meyer, dated April 14, 1979, " Core Damage Assessment for the TMI-2," p. 25). _For the strong-ly exothermic reaction of zirconium with steam approxi-mately 140 k cal per g-mole of zirconium is released at 1,000 C. Each fuel element contains nearly 2 kg of zirconium-hydride, hence a pressure explosion within the ruptured fuel elements would essentially strip these ele-

! ments and release their entire radioactive inventory into the reactor room. This would also lead to a series of chain-like explosions from additional zirconium-steam inter-actions as well as other chemical explosions (e.g., from ignition of the hot hydrogen chemically reacting with th,e oxygen in the reactor room) that would ultimately release hundreds of thousands of curies of mixed fission and acti-vation products into the unprotected atmosphere. Un-protected because the AFRRI-TRIGA reactor (unlike a power nuclear reactor) is not enclosed in any reinforced contain-

. ment dome. Since there are a few hundred pounds of zirconium-hydride within the core of the TRIGA reactor, .

, which is explosively equivalent to almost half a ton of gunpowder when it reacts with water or steam (at tempera-tures >1,000 C), an explosion within.the reactor room would disperse radioactive material over a very densely populated area of many square miles.

Perhaps the most serious credible accident that might i befall the AFRRI-TRIGA Reactor would begin with a LOCA.

The water coolant could swiftly leave through an open water line, rupture the reactor tank and aluminum tank liner, or be pumped out of the reactor pool. Any of these could be initiated by sabotage, inadvertent accident, mechanical i malfunction, or human error either individually or in com-bination (as was noted in authoritative reports of the in-famous Three Mile Island Accident). If the water leaves rapidly (approximately 250 gallons per minute) then the -

fuel element temperature would rise suddenly (see the Hazards Summary Report, GA-?025, 1963, for the 250 kW Mark III TRIGA Reactor at Columbia University). As noted above (see H. H.

Hausner and J. F. Schumar in " Nuclear Fuel Elements,"

p. 84) a sharp temperature fluctuation of this nature is apt to induce multiple cladding failures. Calculations appearing in the Hazards Summary Report prepared by General Atomic scientists for the Columbia-TRIGA Reactor (a smaller but otherwise similar tank-type of nuclear reactor), show 12

)

the maximum fuel element temperature resulting from a LOCA might reach 1,200 C. Such peak fuel temperatures (i.e.,

>1,000 C) could produce excessive total gap pressures

> (i . e . , >1,8 00 p s i) sufficient to rupture several of the already damaged stainless steel claddings. Under these conditions, rupture of the claddings'would expose the Zr-Hx to air (or oxygen) at temperatures of or above .the 1,000 C safety limit. According to Professor Earl A.

i Gulbransen the chemical reaction between ZrUO .034 Hx with air is even more violently exothemic than the zirconium-water reaction, releasing more than 260 k cal per g-mole of zirconium at 1,000*C. Furthermore, once started, he claims there is no easy way to stop the explosive reaction. 'In a certain sense, the explosive mechanism becomes auto-catalytic in that a single explosion would rupture more_ fuel elements releasing additional zirconium and hydrogen which is then available for further explosive chemical interactions with the oxygen in the air. A series of such core explosions would result in dispersing the radiation inventory of the entire AFRRI nuclear facility over a very large area in and around Montgomery County. The public living within the 5-mile ingestion :one (more than 100,000 people) would, in effect, be showered with such radionuclides as Uranium-235, Strontium-89, Iodine-131, Cerium-144, Cesium-134, Yttrium-91, Krypton-85, Strontium-90, and Cesium-137; all with considerable activities and prolonged half-lives.

b. The Intervenor takes great exception to the Licensee's

. broad allegation that " accidents ranging from failure of experi-ments to the largest core damage and fission product release con-sidered possible, result in doses of only a small fraction of 10 CFR part 100 guidelines and are considered negligible with respect to the environment." In fact, all of the accidents described above (in Section 8.a. of these Interrogatories) could violate those guide-lines and especially the last two scenarios (involving circonium explosions) would absolutely result in radioactive releases far in excess of the 10 CFR 1Q0 gu'idelines. Since the HSR and SAR admit to only minimal population exposure (i.e., " doses of only a small fraction of 10 CFR part 100 guidelines") all of the acciden s de-scribed by the Intervenor should be considered of " greater co erity."

l

c. In order to quantitatively evaluate the risk of any

- reactor accident you must know the specific probability for each ite'm in the postulated event-tree as well.as the reliability of the subsystems involved. Without an adequate data base, calcula-tion of the probability for each component event is virtually im- '

possible. Similarly, there is a lack of reliability data on tany '

of the essential subsystems. Unfortunately, this type of data is not yet available and even if it were, there are associated theo-retical controversies that still plague interested scientists and mathematicians, the infamous Rasmussen Report being a case in i point. If only we could quantitatively evaluate the risk of ac-

] cidents this intervention would probably be unnecessary, for then '

. no one would dare put a nuclear reactor in Bethesda.

i 13 i

I a l

d. No cne person is really qualified to " properly desig-  !

nate" an accident as a DBA for the AFRRI-TRIGA reactor. It would I take a team of qualified experts representing several scientific (

disciplines including nuclear physics, nuclear engineering, mate- >

rials science, chemical physics, radiation medicine and health physics.. These experts should~all be Ph.D.s or M.D.s or both.

When dealing with a public danger of this magnitude, it behoovest us to use the very best talent we can muster to evaluate the true *

",W MA DBAsyt @ ether.itae for the AFRRI-TRIGA or any other nuclear r.reappor;,' fQ 2 TPo Ad I T d.M W9 % .

i i

,,r mTInce i Ib5 El O IP9 W47 Ekeysprespntl~ne_

I'C D c h.st. It

.,.,e. 'The" Fede rai!!Citfideping y '

. y'h, im twould q a'dWisable .to<incluA9cbpth 'ePEPAwas#ellf a's the NRC i

'E"WdkfThe twofyI$31

,t .,. ~ f '.

pkc[

m ]%{

bi$i $

c,r hikfil:cid en.tygde, scribed in4 Inter-

~

m o trogstory Answer 8.a. (5)aabo'vg yh'c'u'1dTF de' signa.ted- as DBAs 'for the i

AFRRETRIGifbecausest!' ey arg, thyts'6; po'ssib1'ei" wor:st case" acci-

dents; 4 .. ~

' ' ~ . siblelbiitTheaaccidgnTs"c 3

theyc a"90 al" ohoescribedtu}86mpared t rivial with the magnitude oftiie'H 1.mthe;thh36t~enttiafexposiveydyp iny accidpats ', For documentation,

" refeY, to th'e %es,t.imoh pYespptg PF6fbssor,, Dini' e P M. Pisello at j the 'EivTronmenta( Yrgtectibgdiomia, tee %ff 'the".j w' York City Council  !

• 1979.

Ecarj;nj

, ,. f* 459 '6dcchexHystds!6f

% GQi)?Qs,97ym y. 'Noglfr] W O u c ,z 1 Q , L l'

1. 6wer Plants,y]Jiine 15, INTERROGATORY-9 y .

Q* .M l

l o'

.y . a. JIt biscommon'knowfgTd#fEahs~ pent'fueltelementsfrom t

[

L powerradibactive.

highly $,9sctoistare!storgd[ip@f44ks,pnderwaterCbe'c~auseth ilf the W usbdnto n store the elements should

• failf then en60'gh7 fuel elements.may"c6me togetheraat the bottom of  ;

th~e.pbol'toErbachmcriticalit l cursion) . Itmise ar prob.1.ent> wh[c(h^(currently: concerns pany nuclej

, scientists. At the AFRRI the s)bnt*fuell el'ements are stored 12

'to a rack within the reactor pool itself and if the elements must  !

be k 9 pt at.the AFRRI Facility (because of no federally designated <

or avdilibic 0AFRJdisposair,gQc)i f gn the total number of spent c

fuel'eloments stoiedJineth ateactor 9 Joolcmdy become a serious hazard. [

Unlike the'fueltelementstincpd$prNeaEtorshwhich,are only 3% U-235 i enriched; thM TRIGA fuel elements W e"nominallyL20%tU-235 enriched

'and arej onsequ6ntly,aagregty'}'th.reat9 r "Itotherefore becomes neces- [

sary" to, determine the-mamignum 3 pu'mhe"r 'of spentr fuel elements that 2  !

a r'e, safe .to"storeon'ssumingL Ah,e ~ opt'imtiin"reactivei geometrical array i if they 'shotild"co'me :togeithem at? 'the'bott'om 'of the reactor pool . ,

The Licensed eassuresnus Ahat, iT feifer'lthan67 fuel elements come f togeth'er',' . nothing Lcan thappen.1i Vefiniply?wantmto . know the "experi '

ence" and calculationscon dhic [tl{istassurance is based. It also i becomes very important should" he' heeds saddenlyaaris9 to dismantle and temporarily close the entire core of-fuel elements, about 85.

What provisions have been made for such an emergency?

b. Reasonable assurance could come from two sources, namely: '

(i) experiments in which TRIGA fue'l elements are placed  :

in the contact configuration under water and the actual reactivity measured with local power range monitors (to >

measure the local power distribution in the fuel element i array); i 14 l

. , _ , _ . . . . . _ . . . . _ . ~ - _ ,_ --- _ _ _ . - . . _ _ _ . - _ _ _._......m _-

(ii) criticality excursion calculations for the worst possible geometrical array - that can be evaluated by non-government and non-industry scientists.

c. No specific regulation. However, this is a shortcoming that must be corrected immediately since inadequate fuel element storage now looms large as a terribly significant problem for the entire nuclear industry.

d. The contact configuration represents the optimum reac-tive geometry, that is, the geometric array most likely to achieve a critical power excursion (i.e., criticality).

e. The values obtained for keff <0.746 and m/mcrit. <0.415, are a " reasonable assurance" that a 12 element configuration would remain subcritical.

f. The Intervenor accepts the data represented in Figure 2 as adequately representing the experience of the Los Alamos Scientific Laboratory and the Oak Ridge National Laboratory in their experi-ments U-235 enriched fuel elements Memorandum for Record (January 19, 1981) submitted by the Licensee.

g. The Intervenor is satisfied by the data presented (in' their Memorandum for Record) that failure of a storage rack, fully loaded with twelve TRIGA fuel elements would present no safety hazard to either operational personnel or the general public. However, if additional storage racks are contained in the reactor pool, they should be limited to two or three.

INTERROGATORY 10

a. An experiment fails when it either results in an in-stantaneous insertion of reactivity into the reactor core (type I),

or there is a release of radioactive material from an experiment undergoing activation in the reactor (type II).

b. Either type I or II, but by an entirely different mechanism for each. If the type I failure resulted in a PEA (de-pending on the level of reactivity already operative within the core) and cladding failures (from overheating), then these sets of circumstances could lead to an escape of the radioactive gases from the damaged fuel elements into the reactor room and pool water.

A type II failure results directly in the' release of radioactive material that could also escape into the reactor room.

c. t he same as described in Part 10.b. above.

d. Initially in the reactor room. However, if there is a breach of containment (as described in answers to Interrogatory 1

8.a.) then the radioactive gases could reach other areas of the AFRRI Facility depending on the nature of the containment breach.

e. Please refer to the Federal Regulations for the occupa-tional limits on each radionuclide.

15

1

f. The same as described in part 10.b. above.

g. See Federal Regulations.

h. Depends on the specific experiment sanctioned by the AFRRI.

i. Make certain that the confinem.mt safeguards are intact and functioning properly by more frequent, competent, and independent '

third-party inspections.

INTERROGATORY 11

a. If the rubbergasketsaretotallyremovebthengaseous radionuclides can enter the ventilation system through adjacent rooms or even penetrate through these rooms to the entire facility.

INTERROGATORY 12 a.*

b. To answer these questions accurately, we would have c.' to have more detailed information concerning the physi-

d. cal layout and operational history of the AFRRI Facility.

INTERROGATORIES 13, 14, 15, 16, 17, 18, 19, 20, 21, 22 These questions all refer to specific malfunctions and violations incurred by the AFRRI during their operation of the TRIGA reactor. It makes no sense for us, as outsiders, to second-guess information which is more readily available to them through their own documents or by their direct observation and measure-ment. The charges we have made are a matter of public record and are in full agreement with the designated regulations and techni-cal specifications necessary to operate the TRIGA reactor safely.

If the Licensee is serious.about trying to remedy these situations by including our technical input, we recommend that they consult with us and the Union of Concerned Scientists on some formal basis.

INTERROGATORY 23

a. The moderating effect of the Zr Hx is largely mediated by the hydrogen nuclei. Experiments performed at the Brookhaven National Laboratory (on neutron thermalization by chemically bound hydrogen) gave results for Zr Hx compatible with a solid lattice of regular tetra hedra of zirconium atoms with the hydrogen atoms oc ,

cupying sites at the center of each tetrahedron. The hydrogen lat-tice vibrations could be described by an Einstein model with a char-acteristic energy hv =0.130 electron volts, where he is Planck's constant and v is the hydrogen lattice vibration frequency (see A. W. McReynolds, M. Nelkin, M. N. Rosenbluth, and W. Whittemore,

" Neutron Thermalization by Chemically Bound Hydrogen and Carbon,"

Proceedings of the Second U.N. International Conference on the Peaceful Uses of Atomic Energy, Geneva, September 1-13, 1958, Paper UN/P/1540). The moderating effect of the hydrogen nuclei may be achieved by clastic collisions with fast or slow neutrons; 16

l . ,

that is, prompt or fast neutrons can be slowed down or thermalized by giving up a quantum of their energy, hv, to the sluggish (or cool) hydrogen nuclei, or slow neutrons may be speeded up by re-  !

ceiving the quantum of energy, hv, from the energetic (or warm) hydrogen nuclei. For the most part Zr Hx is not effective in thermalizing neutrons (because hv> kT) , but it can speed up neutrons already thermalized (by the hydrogen nuclei in the tank water).

l Clearly, anything that changes the hydrogen lattice vibration

) frequency, v, will alter the " moderating effect of the UZr Hx fuel."

In turn, the vibration frequency depends on the fuel temperature, the equilibrium hydrogen pressure (between the hydrogen in the fuel-moderator and the gap hydrogen pressure), and the zirconium-hydrogen phase relationships. Damage to a fuel element is likely to affect one or more of these parameters and thereby change v, I

which controls the moderating effect of the hydrogen: nuclei in the l

U-Zr Hx.

b. Under normal operating conditions a reduction in the moderating effect of the Zr Hx would not appreciably affect the reactivity characteristics of a thermal reactor. However, if the fuel temperature goes above 600 C the hydrogen nuclei in the Zr Hx ordinarily reduce the reactivity (i.e., reduce the number of uranium fissions) by warming up the neutrons so they are no longer easily captured by the U-235 nuclei. If the moderating effect of the '

Zr H x is reduced (e.g., by loss of the hydrogen through cracks in the fuel element claddings) then the reactivity characteristics will show a positive increase (i.e., increase the number of uranium fissions). In other words,the protective " warm neutron effect" which would ordinarily decrease the positive reactivity (or equiva-lently, increase the negative reactivity) is no longer available .

because of the reduction of the moderating effects usually mediated by the hydrogen nuclei in the zirconium-hydride,

c. The question is irrelevant since the AFRRI-TRIGA is strictly a thermal reactor.

d. Mathematica11y,' the negative temperature coefficient is primarily a function of exp-hv/kT, so that any change in the hydrogen lattice vibration frequency, v, will necessarily modify this coefficient of reactivity. As discussed above, in part a, v is a function of the circonium-hydrogen phase relationships which, in turn, depends on the fuel moderator temperature and the equi-librium hydrogen pressure. Thus, any core condition that signifi-cantly changes these parameters within the fuel elements will af-fect the negative temperature coefficient of reactivity. Fuel element cladding failures that permit the escape of hydrogen, will .

undoubtedly affect the equilibrium hydrogen pressure which ulti-mately reduces the availability of hydrogen nuclei directly and may induce a phase transition indirectly (due to the reduced concentra-tion or density of hydrogen). Thus , cladding damaged fuel elements can profoundly change the effectiveness of the ordinarily protec-tive negative temperature coefficient by removing a substantial number of hydrogen nuclei (the direct effect) and by modifying v through a phase transition (the indirect effect). This is why the Intervenor contends that whereas the mechanism for the negative 17 i

temperature coefficient may operate well under ideal conditions, it may not work very well in a real accident situation (e.g., when the fuel elements may be bent, scratched, corroded, and inadequate-ly cooled) in which case the moderating effect of the hydrogen nuclei (within the U-Zr Hx) could be seriously impaired. There-fore, we argue that this automatic shutdown mechanism is, like other so-called " failsafe" mechanisms, not absolutely foolproof.

e. In the context of this contention, a damaged fuel ele-ment may be functionally defined as a fuel element that eithe,r leaks hydrogen or undergoes an unusual change in the magnitude of its hydrogen lattice vibration frequency, v, as the peak fuel tempera-ture goes beyond 600 C.

f. Design an experiment in which cJudding damaged fuel elements (that leak hydrogen) are rapidly warmed up (to temperatures of nearly 1,200 ) by pulse heating. Keep all the other variables in their usual condition. Be sure to do this experiment in a safe place, not Bethesda.

g. About 1% reduction in the total core moderating effect.

h. This depends on one's criteria for significance. If 10% reduction is significant, then about ten damaged fuel elements would be required.

i. These calculations will be presented by direct testimony at the hearing itself.

j. Data from experiments such as those described above in.

part f of this interrogatory.

INTERROGATORY 24

a. They do not believe multiple cladding failure accidents are credible. -

b. No* that we know of.

c. An " uncontrolled power excursion" may be defined as a large, rapid increase in neutron flux resulting from an unscheduled insertion of excess reactivity above cold critical. Associated with any power excursion is the abrupt rise in fuel temperature.

The combined effects of the sudden temperature elevation and the large rapid increase in neutron flux, stress the fuel elements (in-volved in the power excursion) in several important ways, including:

(i) thermal migration stresses - which arise when hydrogen migrates from the higher to the lower temperature regions of the fuel element, causing the colder regions to expand and the hotter regions to contract. This results in a " migration stress" which is in the opposite sign (or direction) of the thermal stress. The brittle nature of zirconium hydride makes it susceptible to " thermal stress cracking." [See Meyer, R. D., and J. G. LeBlanc, " Negative Thermal Expansion in UZr H Reactor Fuel," Trans. Am. Nucl. l Soc. 13, p. 2, 1970.] Furthermore, if the radial tempera- I ture gradient on the fuel element is asymmetric, bowing of j the rod will occur.

18 l

l

(ii) anomoulous oscillations (secondary to sustained cycling) of the fuel elements if during this redistribu-tion (of hydrogen) and bowing change the thermal gradients are altered, as would be the case in an operating reactor, the conditions for sustained cycling are obtained (i.e.,

anomalous oscillations). The oscillatory behavior.was probably due to the clustering of fuel elements under a thermal gradient, followed by an abrupt declustering (caused by the rehydriding of the fuel elements) which ap-plies a force in the opposite direction. [See Simnad, M. T., "The U-Zr H x Alloy: Its Properties and use in TRIGA Fuel," GA-4314, 1980, pp. 2-17.]

(iii) surface cracks in the fuel elements effected by the escape of hydrogen - measurements and calcu_lations have been reported of hydrogen loss from hydrided U-Zr fuel elements which were rapidly heated by induction to tempera-tures ranging from 900 - 1,000 C. After.75 sec at those temperatures, there was an abrupt crack in-the surface of the fuel elements associated with major hydrogen evolution (i.e., strong outgassing rates). [See Leadon, B. M. et al., " Aerospace Nuclear Safety-Measurements and Calcula-tions of Hydrogen Loss from Hydrided U-ZrH Fuel Elements During Transient Heating to Temperatures Near the Melting Point," Trans. Am. Nucl. Soc., 8, 1965, p. 8.]

(iv) excessive gap pressures - at elevated temperatures there is an increase in all three component gap pressures (i.e., residual air pressure, fission gas pressure, and the peak equilibrium hydrogen pressure) to produce a total gap pressure that way approach or even exceed 2,000 psi.

Note, at temperatures above 800 C the equilibrium hydrogen pressure is, by far, the major contributor. Gap pressures of such magnitude, must put considerable stress on the fuel elements.

(v) irradiation stress resulting from the high neutron fluence caused by, the power excursion (e.g., a peak power level of g,000 y MW will create a neutron fluence of about 1.0 x 10 nyt) - large neutron fluxes induce structural flaws in the substance and claddings of the fuel elements.

That is, bombardment by energetic neutrons will produce solid state defects which can migrate and coalesce to establish significant weaknesses and degradations in the  :

fuel elements.

d. As described above in Interrogatory Answer 8.a., part 5) ,

a LOCA would result in a sudden, large elevation of fuel temperature which is apt to produce multiple cladding failures by all of the '

stress mechanisms outlined in part c of this interrogatory (except that the neutron flux is generated by pulsing rather than an uncon-trolled transient).

e. The conditions necessary "to breech the integrity of the fuel's cladding" are entirely divorced from the past power history of the AFRRI-TRIGA reactor. However, the immediate power history of the reactor is important in that "a recurrent pulsing mode" of operation is a vital contributing factor.

19

f. If the IMW TRIGA reactor was not capable of the pulsing '

operation, it is unlikely that cladding failures would result from a LOCA involving that reactor. It is noteworthy, however, that the long operational history (about 20 years) of the AFRRI-pulsing I'RIGA reactor is more likely to have a multiple cladding failure accident simply because of its longevity (i.e., greater accumulation of radiation-induced solid state defects within the exposed fuel ele-ments themselves).

INTERROGATORY 25 -

a, b, c Liz Entwistle will respond.

D. The Licensee mus t concede that there are'at least three essential conditions to prevent serious fuel element" cladding failures, namely:

(i) fuel element temperature should never exceed the safety limit (1,000 C);

(ii) there must be no sudden core insertion of excess reactivity (i.e. , >3.5% Ak/k excess reactivity above cold critical) to produce a very large, rapid increase in neutron flux; (iii) the fuel elements must contain the proper ratio 5 of hydrogen to zirconium and the appropriate mixture of phases of the U-Zr H x, so that the hydrogen can remain an effective moderator and gap pressures do not become excessive (>1,800 psi).

Violation of one or more of these conditions can result in (multiple) cladding failure accidents (see answers to Ir.terrogatory 8). At fuel temperatures of 1,000 C or more, almost 10% of the radioactive gases contained in the cladding-damaged fuel elements (that have redistributed into the gaps) will then diffuse out through the degradations or breaks in the claddings (see In fact, the equiv,alent of one pound of radium per defective fuel element (in gaseous form) could leak out of the cladding-damaged TRIGA reactor. Another danger to public health and safety is contamination of the tank or cooling water by the (20% enriched) uranium and stored fission products in the cladding-damaged fuel elements. For example, in the event of even a single claddiag failure the water activity could easily reach a level of luci/cc and that level would remain elevated (because of the small decay constant of iodine-131 and iodine-133, among other reasons), which greatly exceeds the MPCs outlined in 10 CFR Appendix b. This would not only be a hazard to persons in the reactor room (occupational exposure) but if, by human error, the contaminated water leaked '

out into the sewage system of Montgomery County and the District of Columbia, it could cause considerable damage to the people living in these communities. Now contemplate a multiple cladding failure and the reason for our concern becomes quite understandable. The two worst accidents (i.e., maximum credible accidents) involving zirconium explosions would, of course, release the entire radio-active inventory of the reactor into the environment. Given the population density, clustering of hospitals, schools, churches, etc., in the Bethesda area, such an accident would be sheer catastrophe.

20

INTERROGATORY 26

a. The Licensee claims that "no waterborne radioactive emissions are generated by routine operations." Where then did the radionuclides found by the Washington Suburban Sanitary Com-mission reports (see WSSC reports for 1980-81) come from?

b. S c. The Licensee is legally responsible for disposing of its generated solid radioactive waste which included:

(1) contaminated animal carcasses, tissues and wastes, (2) mixed laboratory wastes, ,

(3) scintillation vials with scintillant, (4) filters which collect reactor by-products, (5) spent radioisotopic targets, ,

(6) worn-out or spent fuel elements.

Solid radioactive waste is disposed of in two different ways, namely:

(i) transferred to waste disposal contractor in steel barrels for shipment to radioactive waste burial grounds, (ii) transferred to the NNMC Radiological Safety De-partment in boxes for incineration.

The Intervenor contends that incineration of these solid wastes on an unrestricted area of the NNMC grounds results in the airborne release of radioactive gases and particulates that endanger the public health. For example, the AFRRI on one occasion reported the transfer of 160 boxes of contaminated waste to the NNMC for incinera-tion, noting that the principal isotope was Sodium -24. The esti-i mated total activity of this isotope (i.e., Na-24) was approximatel l ly 1.6 mci. Now the annual licensed quantity equivalent (per 10 CFR Appendix C) is no more than 10 pCi, so that this represents (more than a one-hundredfold excess) a clear violation of that safeguard. Another example of dangerous accumulated quantities of radioactive solid waste. occurred in 1973 when shipment records l indicate that 80 Ci of tritium (targets) were transferred offsite.

Since the annual licensed quantity equivalent (per 10 CFR 20 Appendix C) is only one m Ci, this again represents an infraction of Federal regulation and a potential public hazard (see NRC Inspection Re-ports covering 1975-76, specifically Report No. 50-170/ March 1977).

l

d. We define a " probable violation" as one in which a violation is apt to have been committed except that one or more of the following applies:

(i) the appropriate data were not recorded which would, have established the violation, (ii) the appropriate data were recorded but the reports have been doctored or kept secret, (iii) the instrumentation or method of measurement was inadequate to actually obtain the pertinent data demonstrat-ing the violation,

c. Specific regulatory limits that have been violated by the AFRRI include:

1 (1) The concentration of the gaseous radionuclide Argon-41 (Ar-41) released from the AFRRI stack in 21

1962, 1963, and 1964 exceeded the MPCs listed in 10 CFR 20 Appendix B.

(2) The yearly environmental monitoring data (obtained from the AFRRI Perimeter Monitoring System) demonstrate that the AFRRI has exceeded the well-known Federal regula-tion that the average yearly ambient radiation levels be less than the 0. 5 Rem per year for unrestricted areas.

This was the case for both 1962 and 1963.

(3) The data (see Table 2 below) clearly demonstrate that the AFRRI has consistently exceeded the annual ex-posure EPA limit of 25 mrem for unrestricted areas s'urround-ing a nuclear reactor (see EPA Resolution No. 40 CFR 190) .

(4) Technical specifications require the AFRRI reactor building ventilation to exhaust to a stack.having a minimum elevation of 18 feet above the roof level of the highest building in the AFRRI complex. Contrary to the above, a leak through a stack drain line discharged part of the exhaust at ground level outside the building for a period of several months.

(5) During the period January 1, 1970 to July 1, 1971 the " normal exposure rate" was 0.5 mrad /hr, however, there were several unrestricted areas where the exposure rate rose to 1 mrad /hr or more and at least one specific un-restricted area where the dose rate approximated 5 mrad /hr.

The maximal permissible annual exposure, by NRC regula-tions, is 500 mrem per year. Hence, any person who lived or worked in these unrestricted areas where the dose rate was 1-5 mrad /hr, would have received excessive radiation if they had been exposed for merely 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> during the entire year, or about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per week. Since the Licensee has failed to convincingly demonstrate that this, could not have occurred, it represents a clear violation of the ALARA principle (the goal embodied in 10 CFR part 50) as well as the Federal regulation (requiring that no one receive more than 500 mrem per year). The locations of maximum ionization chamber readings were partly in residen-tial areas. Note approximately 50-60% of the area within a one-mile radius of the AFRRI stack is, in fact, resi-dential.

(6) According to 10 CFR 20.201(b) the " Licensee shall make or cause to be made such surveys as may be necessary for him to comply with the regulations in this part."

Contrary to this directive, gross beta measurements of liquid effluents to assure compliance with 10 CFR 20.303,

" Disposal by release into sanitary sewage systems," made for the period January 1976 to January 1977 were inade- ,

quate in that the gross beta measurements were made with-out the use of beta self-absorption correction in the presence of significant amounts of suspended solid material.

(7) The AFRRI on one occasion reported the transfer of ,

160 boxes of contaminated waste to the NNMC for incinera- l tion, noting that the principle isotope was Sodium-24. '

The estimated total activity of this isotope (i.e., Na-24) was approximately 1.6 mci. Now the annual licensed quanti-ty equivalent (per 10 CFR 20 Appendix C) is no more than 22

10pCi, so that this represents (more than a one hundred-fold excess) a clear violation of that safeguard.

(8) Another example of dangerous accumulated quantities of radioactive solid waste occurred in 1973 when shipment records indicate that 80 Ci of tritium (targets) were trans-ferred offsite. Since the annual licensed quantity equiva-lent (per 10 CFR 20 Appendix C) is only one mci, this again represents an infraction of Federal regulation and a poten-tial public hazard.

f. Specific radiation monitoring methods that the Inter-venor considers inadequate,inc:ude:

(i) The statistical uncertainty in the annual perimeter dose per monitoring station is + 20 mrad at -the 95% con-fidence level. This is totally inadequate and could easily be remedied.

(ii) The environmental film dosimetry method employed at the monitoring stations detects only external gamma radiation. Thus, the population radiation exposure dose due to the inhalation or ingestion of radionuclides is entirely neglected.

(iii) The particulate radioactivity monitor for airborne radioactive effluents (i.e., a pancake-probe G-M counter) is not isokinetic and therefore cannot be used for any quantitative evaluations.

(iv) The dose rates (using ionization chamber type in-strumentation or an alternative) are not determined with sufficient frequency either for restricted or unrestricted areas (both on and offsite).

(v) The " Environmental Sampling and Analysis" program has been criticized for calculational omissions, the man-ner in which the samples were prepared for analysis, and the type of instrumentation used to perform the analyses.

For these and other reasons, this program should be adminis-tered by private and public scientific agencies outside of the Department of Defense (e.g., Washington Sanitary Sewage Commission, the EPA, the Sierra Club, etc.).

g. The principle of self-regulation when it comes to radia-tion monitoring is highly suspect. The public would be better served if such monitoring were left to private scientific laboratories under the authority and inspection of local government agencies. This puts responsibility for public safety and protection precisely,in the hands of the local people who need that protection. Federal agencies could advise, fund, and help implement the appropriate .

radiation monitoring methods when " corrective actions" are truly indicated to prevent violations of regulatory limits. As long as the Licensee itself has the primary monitoring responsibility they are likely to be inadequate.

h. We are not presently aware of detailed " specific correc-tive actions" being undertaken by the AFRRI. If the Licensee wishes to share such information, we would be ready to comment on its

" adequacy or inadequacy" to prevent a recurrence.

23

l

i. We believe " probable violations" have occurred in the following instances:

(1) Since the statistical uncertainty in the annual l

perimeter dose per monitoring station (at the 95% confidence level) is + 20 m Rad, it is likely that the annual popula-tion exposure at several unrestricted area stations has exceeded the EPA limit (i.e., 25 m Rem) for just about every year during the past 20 years of the AFRRI Facility l operation.

(2) The absence of data due to omission of internal radiation makes it virtually impossible to evaluate ,the true population exposure to radiation, let alone determine whether the Federal regulatory limits have.actually been .

exceeded.

(3) The only two AFRRI particulate radi$nctivity monitor-ing systems (i.e., the pancake-probe G-M counter and the radioactive gas monitor) are not reliable for quantitative particulate radioactive sampling. Hence, one can only ob-tain crude estimates of the airborne radioactive particu-lates that have been dispersed into the environment. The true values may, in fact, have exceeded public safety limits.

(4) The maximum permissible annual exposure, by NRC regulations, is 500 mrem per year. It is very likely that any person who lived or worked in an unrestricted area where the dose rate was 1-5 m Rad /hr, would have received radiation in excess of this " maximum permissible annual exposure."

l

j. The following is a list of sources used to document the contention raised above, that the past and present operation of the AFRRI reactor has resulted in probable violations of 10 CFR part 20:

(1) See the letter from the AEC to the AFRRI dated October 6, 1961 which predicts that, according to their calculations for Argon-41 concentrations in unrestricted areas, AFRRI will probably not be able to meet the MPC re-lease standards for unrestricted areas as stated in 10 CFR 20, Appendix B.

(2) See Environmental Release Report (AFRRI-TRIGA Re-actor) covering the period 1 Jan. 1970 to 30 Sep. 1971, .

issued on December 14, 1971.

(3) See Inspection Report No. 50-170/77-01-03 that dis-cusses (gaseous effluent) airborne particulate evaluation.

(4) See the AFRRI-TRIGA Reactor Environmental Release Report issued on December 14, 1971 by AFRRI and the DNA.

(5) In its Environmental Impact Appraisal the Licensee notes that the highest average unrestricted area exposure rates corresponding to given airborne releases are 4.1 m '

Rem /hr for Argon-41, 4.3 m Rem /hr for the combinatien of both Nitrogen-13 and Oxygen-1 , and 0.5 m Rem /hr for Xenon-133. These are high dose rates and since they admit-tedly extend to residential areas, it is quite possible that people may have received in excess of 500 m Rem (the regu-latory limit) in any given year. It is also another illus-tration of violation of the ALARA principle originally de-signed to protect the public'from excessive exposure.

(6) See the AFRRI Environmental Release Data and Perim-eter Monitoring Reports Docket No. 50-170 (e.g., May 27, 24

?*

1966 report, September 20, 1966 report, and December 14, 1977 report).

(7) See the AFRRI's written response to Mr. Joe Miller's (from Citizens for Nuclear Reactor Safety) question #11.

(8)

Table 2 Year Average Annual Perimeter Dose (per monitoring station) 1962 242 mrad 1963 231 mrad .

1964 89 mrad 1965 55 mrad *

k. The following is a list of sources used to document the contention raised above, that the past operatiom of the AFRRI Reactor has resulted in actual violations of 10 CFR part 20 (also see answer to part e. where the actual violations,are listed).

(1) With respect to the excessive release of Argon-41 please see the AFRRI Airborne Release Reports for 1962, 1963, and 1964; and the AEC Inspection Reports for 1962, 1963, and 1964, in Docket No. 50-170.

(2) The yearly environmental monitoring data (obtained from the AFRRI Perimeter Monitoring System reports) dem9n-strate that the AFRRI has exceeded the annual federal limit, 0.5 Rem, for the average yearly ambient radiation.

Table 1 Year Maximum Annual Exposure Specific Perimeter Station T707 > 500 mrad 2c 1963 > 500 mrad 16A .

1964 116 mrad 2A 1965 112 mrad 16A '

1970 76 mrad 16A 1978 30 mrad llA (3) The data (see Table 2 above) also clearly demonstrate that the AFRRI has consistently exceeded the annual exposure EPA limit of 25 mrem for unrestricted areas surrounding a ,

nu.: lear reactor (see EPA Resolution No. 40 CRF 190).

(4) Regarding the ground level leak of gaseous effluent, see NRC Inspection Report conducted on January 11, 1979 (i.e., Inspection Report No. 50-170/79-01).

(5) See Violation Notice of Gross Beta Effluent Analysis based on the NRC inspection of January 12-14, 1977. Also see Inspection Report No. 50-170/77-01-02.

(6) With regard to violations concerning solid radio .

active waste see NRC Inspection Reports covering 1975-76, specifically Report No. 50-170/ March 1977.

1. The answer to this question is essentially contained in the answer provided to Interrogatory 26, parts f, g, and h, given above.

INTERROGATORIES 27, 28, 29, 30 See Testimony of Professor Ernest J. Sternglass to be

, presented at the Hearings.

l 25

1 INTERROGATORY 31 All of the component questions center on the meaning given to the phrase " highly probable." In order to assess the probabili-ty that the MPCs (set forth in 10 CFR 20, Appendix B) have been exceeded, we would need considerable information (not made avail-able) regarding the individual radionuclide concentrations, air flow parameters and meteorological data occurring during those several months.

INTERROGATORY 32 .

a. Our contention that air convection cooling alone would not be sufficient to cool an operating TRIGA reactorecore during and immediately following a LOCA, is based on calculat' ions contained in the 1963 GA-2025 Hazards Summary Report for the 250 kW Mark II TRIGA Reactor (Pulsing) located at Columbia University in New York City.

b. Internal gap pressures could rise to 1,800 psi or more depending on the temperature elevation. This pressure is capable of producing breaks or penetrations of the fuel element cladding (see Figures 2-9, " Equilibrium Hydrogen Pressure over Zr Hl .65 '

versus Temperature" in the 1980 GA-4314 Report by M. T. Simnad).

c. TRIGA reactors have definitely had cladding failures.

However, whether any such failure has ever been the result of a LOCA is unclear.

d. The Licensee asserts that the maximum amount of fission products that could be released in the event of a cladding failure of a single average fuel element in the AFRRI-TRIGA core is less than 7 curies during steady state operation and also 7 curies dur- '

ing pulse operation (following steady state operation). These cal-culations, in turn, are based on the assumption that the fraction of gaseous fission products (i.e., radioisotopes of Iodine, Kryp-ton, and Xenon) released from U-Zr H x fuel into the gap between the fuel material and the fuel element cladding is only 0.1%. That assumption is valid, however, only if the fuel temperature is be-low 550 C where emission of radioactive gases is largely controlled by recoil effects. However, in a LOCA-induced cladding failure the temperature will rise way above 550 C, so that the process of gas emission into the gap becomes mostly diffusion controlled and radio-active gases begin to stream out of the cladding-damaged fuel ele-ment. In fact, a LOCA is apt to produce temperatures in excess of, 1,000 C, which means that nearly 10% of the radioactive gases in the fuel element escape into the gaa region between the fuel and the cladding. Practically all of taese gaseous radioisotopes will find their way out of the fuel element gap into the reactor room atmosphere through the penetrations in the damaged cladding. Thus, if 0.1% fraction of gaseous fission products results in a release of above 7 curies, then a fraction of nearly 10% would result in the release of at least 500 curies from a LOCA-induced cladding failure of a single average fuel element. The presence or absence of any breach of containment within the reactor room, will determine 26

the amount of gaseous radioisotopes (e.g., Iodine, Krypton, and

, Xenon) that ultimately leak into the outside environment (see j Figure 5-1, " Fractional Release of Gaseous Fission Products from TRIGA Fuel," p. 5-3 (in the 1980 GA-4314 Report by M. T. Simnad).

INTERROGATORY 33

a. Pulse heating leading to sudden elevations of tempera-ture sufficient to cause multiple cladding failures, has been described several times throughout the body of these Interrogatory Answers (e.g., see Answers to Interrogatories In the specific case of a LOCA-induced multiple cladding failure, the timing aspect is crucial. It is crucial because the loss of the water is not only a loss of coolant, for a TRIGA reactor it is also the loss of the primary moderator since it is the. hydrogen nuclei within the water which actually thermalize the prompt neutrons (i.e., the fast neutrons leave the fuel elements and enter the tank water where they give up their excess energy to bedome slow neutrons capable of initiating more U-235 nuclear fission). Thus, if the loss of tank water occurs too rapidly there can be no pulsing or inadvertent power excursions (transients). Realistically, we cal-culate that water could leave the tank at a maximum rate of about 250 gallons per minute (in the several possible ways described above). At this rate of moderator (water) loss, pulsing would nht be prohibited and yet cooling would be seriously impaired, produc-ing exactly the necessary conditions for cladding failures to occur.

b. Temperatures above 900 C are rarely, if ever, reached by a single pulse. Indeed, it would take many pulses to establish fuel temperatures of this magnitude rapidly. Temperature fluctua .

tions required to produce cladding failures in a reactor core could result by repeated activation of the pulsing mode following a period of steady state operation.

c. In the answer to Interrogatory 32, part d, we established a fission product release of about 500 curies from a single cladding failure accident if the fuel temperature goes above 1,000*C. Thus, even a single cladding failure would release gaseous radionuclides (e.g., Argon, Krypton, Xenon, Iodine) beyond the limits imposed by 10 CFR 20. One should also allow for radionuclide leaks into the water coolant, given a serious cladding failure (see Interrogatory Answer for details). A multiple cladding failure accident would result in a horrendous release of fission and activation products.

d. The temperature history during a LOCA associated with one or more transients depends entirely on a number of conditions,.

including the rate of loss of coolant (water), the number and magnitude of the pulses or transients, the previous events immediate-ly preceding the accident (e.g. , duration of steady state operation),

condition of the fuel elements, etc. Given these and other boundary conditions, we might be able to calculate the temperature history very approximately!

1 INTERROGATORY 34 i i

The medical concern over radiation exposure secondary to inhalation of gaseous radioisotopes (such as the noble gases) 27

- y -- -+----------w- ----*----y+ y -- v -- v - w g---

continues to engender controversy among physicians and health physicists. Internal emissions have been a topic of medical sym-posia such as the NIH Conference about one year ago. In order to do this subject technical justice will require considerable scien-tific explanation and justification. We expect this topic to be discussed by several of our witnesses at the time of the full hearings, including Dr. Irving' Stillman, Dr. Ernest Sternglass, and Dr. Irwin Brass.

INTERROGATORY 35 .

a. The release of any radionuclide (or'combinction thereof) into the environment that would expose the public to a dose rate

of 100 mrem /hr or more for at least one hour (i.e.,'a total ex-posure of at least 100 mrem), constitutes, in our op' inion, a "significant offsite release."

b. Since the AFRRI nuclear reactor has be'en operational for the past 18 years , one can estimate that most of the 87 fuel elements have a radioactive inventory equivalent to about ten pounds of radium per element. This inventory specifically includes, among others, such radionuclides as Uranium -235, Strontium-89, Iodine-131, Cerium-144, Cesium-134, Yttrium-91, Krypton-85,  ;

. Strontium-90, and Cesium-137; all with considerable activities and

prolonged half-lives. Thus, if the radioactive contents of even one fuel element were dispersed into the environment by a chemical explosion (described above) the fission and activation products j released to an unrestricted area would violate the 10 CFR (part 100) guidelines many times over.

S

c. Any of the chemical explosions described above in the two " worst-case" scenarios (e.g., hydrogen-oxygen, zirconium-steam and zirconium-oxygen) could trigger a series of such explosions.

, Since the AFRRI-TRIGA has no containment dome, these types of ac-cidents would widely disperse the radioactive inventory originally contained within the reactor core. These offsite releases would l not merely be "significant," they would be catastrophic.

d. This has been detailed in the answer to Interrogatory given above. However, the sequence may again be briefly outlined as follows: the power excursions produce fuel-moderator heating i in an abrupt or rapid manner. The acute temperature fluctuations i and massive, sudden increases in neutron flux, effects one or more

! cladding failures, causing a loss of the' hydrogen (that migrates

! to and accumulates in the gaps) through breaks in the element clad-i dings. Thus, the thermalizing effect of the hydrogen within the -

fuel elements is severely compromised.

4

e. The loss of hydrogen nuclei from the U-Zr H x necessarily reduces its moderating effects and via changes in hydrogen den.sity can also induce phase changes, both of which modify the prompt negative temperature coefficient. A reduction in the effectiveness of the negative temperature coefficient to protect the TRIGA re-actor during an accident, would have serious consequences.

28

4 1

f. The explosive zirconium-steam reaction requires zir-

conium, water or water vapor, and high temperatures (see

i

g. To the best of our knowledge, all of these necessary conditions can exist within an overheated AFRRI-TRIGA nuclear core (see i

t

h. The explosive zirconium-air interaction requires zir-j conium, air or oxygen, and high temperatures (see .

i. To the best of our knowledge, all of these necessary

, conditions can exist within an overheated TRIGA core following a I LOCA (see c

j. This is merely a matter of semantics. Obviously we  !

believe that the two maximum credible accidents (dbscribed above) l should be designated as the design basis accidents for the AFRRI- i TRIGA, rather than the two relatively trivial accidents presently '

4 designated as such. I INTERROGATORY 36

.I

! a. (2) During the period January 1, 1970 to July 1, 1971 the " normal exposure rate" was 0.5 mrad /hr, however, there were several unrestricted areas where the exposure rate rose to 1 mrad /hr or more and at one specific unrestricted l area where the dose rate approximated 5 mrad /hr. The maxi-mum permissible annual exposure, by NRC regulations, is .

. 500 mrem per year. Hence, any person who lived or worked ,

j in these unrestricted areas where the dose rate was 1-5 mrad /hr would have received excessive radiation if they

! had been exposed for, at most, 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> during the entire l year, or about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per week (see the AFRRI-TRIGA Reac-i tor Environmental' Release Report issued on December 14,

] 1971 by AFRRI and the DNA).

I (1) In its EIA, the Licensee notes that the highest average unrestricted area exposure rates corresponding to '

given airborne releases are 4.1 mrem /hr for Argon-41, 4.3 mrem /hr for the combination of both Nitrogen-13 and Oxygen-15, and 0.5 mrem /hr for Xenon-133. These are high dose rates and since they admittedly extend to residential areas, it is quite possible that people may have received in excess of 500 mrem (the regulatory limit) in any given year. .

i (3) and (4) have already been discussed.  :

l b. and c. Data contained in Table 1 and Table 2 given above. l l l

d. The dose rates alluded to were determined by ionization l chamber instruments, not the film badges at the environmental station i i monitors. The large dose rates were attributed to excessive X- l i

Radiation coming from a large X-Ray machine (called the Maxitron) .  !

! However, this interpretation was never verified. i 1 '

29

! l l >

. . , _ , , . - ,. -- ,~--, _ . . , _ _ _ _ _ _ _ , - - , , . _ , , _

e. If it is a typographical error then it was made by the AFRRI or DNA in the Environmental Release Report issued on Decem-ber 14, 1972.

f. We believe so. No.

g. This information describes the radioactive inventory contained within the AFRRI-Reactor Core. It was used whenever such information was needed to make a specific point.

h. Answer by Li Entwisle. No CNRS members live within 600 feet of the AFRRI stack. The Intervenor is without knowledge of the address of any residence within 600 feet of said stack.

INTERROGATORY 37

a. Admiral Robert Monroe, The Washington Star, Tuesday, August 14, 1979 ; " Colonel MacIndoe," Montgomery Jour.nal, by Sandy Golden, Wednesday, June 27, 1979.

b. The Intervenor is unable to answer this question with-out access to information not in the public record,

c. Same as b. i

d. Each of the cited instances demonstrates a break in the first and last layer of security between the controlled access areas and the public.

, _ Respectfully submitted, Elizabeth B. Entwisle Counsel for Intervenor l 30

r AFFIDAVIT OF ELIZABETH B. ENTWISLE 1

I, Elizabeth B. Entwisle, being duly sworn, do state:

1. That the Response of Intervenor Citizens for Nuclear Reactor Safety, Inc. to the Licensee AFRRI's Firs't Set of Interrogatories were preoared under my direction and l supervision. -

2. That the responses therein designated " Answered by Entwisle" were answered by me.

3. That the responses designated " Answered by Stillman" were answered by Dr. Irving Stillman.

i

4. That the responses are true to the best of my knowledge, information, and belief, t

Elizabeth B. Entwisle q fN SUBSCRIBED AND SMORN TO before me this cv 7 day of December, 1981.

I f ,

/Aut u.4 4 Notary Public (j' [) 7-/-8 4

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of  :

ARMED FORCES RADI0 BIOLOGY Docket No. 50-170 RESEARCH INSTITUTE -

(Renewal of Facility (TRIGA-Type Reactor)  : License No. R-84) .

CERTIFICATE OF SERVICE ,

I hereby certify that the foregoing Responic of CNRS to Licensee's First Set of Interrogatories and Affidavit was served on the following by depositing in the United States Mail, first class, this 29th day cf December 1981.

Louis J. Carter, Esq., Chairman Mr. Richard G. Bachmann, Esq.

Administrative Judge Counsel for NRC Staff Atomic Safety and Licensing Office of the Executive Legal Board Director ,

23 Wiltshire Road U.S. Nuclear Regulatory Commission Philadelphia, PA 19151 Washington, D.C. 20555 Mr. Ernest E. Hill Mr. Stuart A. Treby, Esq.

Administrative Judge Assistant Chief Hearing Counsel Lawrence Livermore Laboratory for NRC Staff University of California Office of the Executive Legal P.O. Box 808, L-123 Director Livermore, CA 94550 U.S. Nuclear Regulatory Commission l Mr. David R. Schink Washington, D.C. 20555 Administrative Judge Atomic Safety and Licensing Department of Oceanography '

Board Panel Texas A 6 M University U.S. Nuclear Regulatory Commission College Station, TX 77840 Washington, D.C. 20555 Docketing and Service Section Atomic Safety and Licensing Office of the Secretary Appeal Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 Mr. Robert L. Brittigan, Esq.

General Counsel Defense Nuclear Agency Washington, D.C. 20305 -

Elizabeth B. Entwisle Counsel for Intervenor

I

. 1 4

}

i i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 4

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of ARMED FORCES RADIOBIOLOGY Docket No. 50-170 RESEARCH INSTITUTE (Renewal of Facility (TRIGA-Type Research Reactor) License No. R-84) l l

l DECLARATION OF ROBERT L. BRITTIGAN The Intervenor's response to Licensee's First Interroga-1 tories to the Citizens for Nuclear Reactor Safety, Inc.

(Exhibit B), was received in my office on January 4, 1982.

The attached motion to compel, under the provisions of 10 C.F.R. 2.740(f), should have been filed on January 14, 1982.

On January 14, 1982, however, a severe snowstorm had rendered travel in the Washington, D.C. area hazardous and non-essential Federal employees were asked to remain at home. Accordingly, typing and reproduction could not be completed on that date.

It is the position of the Licensee that January 14, 1982 should be treated as a legal holiday for the purposes of I computing time under 10 C.F.R. 2.710 and 10 C.F.R. 2.740 (f) .

I have discussed the matters raised in the attached

- motion to compel with Ms. Elizabeth B4 Entwisle, Esq., Counsel for the Intervenor, and we have agreed to attempt to resolve i

n

the question of the adequacy of the Intervenor's response informally during the next 45 days. I have been authorized to inform the Board that Ms. Entwisle joins in the Licensee's request that the Board defer ruling on the attached motion to compel for a period of 45 days to permit informal resolution of the matter.

I hereby declare under penalty of perjury that the foregoing is true and correct.

SYE W C ROBERT L. BRITTIGAN '

Counsel for Licensee Signed at Headquarters, Defense Nuclear Agency, 6801 Telegraph Road, Alexandria, Virginia this 15th day of January 1982.

2

_ _ _ . _ _ _ . _ _ _ _ _ ._.__ =_ . _ . . __ . _ .

l l

i l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of l l ARMED FORCES RADIOBIOLOGY Docket No. 50-170 l j RESEARCH INSTITUTE -

t 4

(Renewal of Facility

{

(TRIGA-Type Research Reactor) License No. R-84)  !

1 l

ORDER  !

l Upon consideration of the Licensee's Motion to Compel i

Answers to Licensee's First Interrogatories to the Citizens l I for Nuclear Reactor Safety, Inc., the opposition thereto,  !

1 '

l and the entire record herein, the Board finds, pursuant [

to 10 C.F.R. $ 2.740 (f) , that Intervenor CNRS's Responses i

, to Licensee's First Set of Interrogatories are incomplete, l l

evasive, or both, with respect to interrogatories numbers l

6, 7b, 7c, 7d, 7el, 7e2, 7e4, 7j, 7n, 8b, 8c, 9a, 10a, 10h, (

l 12a, 12b, 12c, 12d, 13, 14, 15, 16, 17, 18, 19, 20, 21, (

22, 23a, 23b, 23c, 23d, 23g, 23h, 231, 24a, 24c, 249, 24h, j 241, 25a, 25b, 25c, 26a, 269 , 261, 27, 28, 29, 30, 31b, 31c, 32b, 32c, 34a, 34b, 34c, 34d, 34e, 34f, 35e, 35f, 35g, 35h, 351, 36a(2), 36a(3), 36a(4), 36c, 36d, and 369 and i i

! that the Citizens for Nuclear Reactor Safety, Inc. has there- l l f 4 fore failed to respond to the cited interrogatories.

l l

t I

h I

_ _ _ _ - - . - - , . _ _ _ _ _ _ _ , , . - _ - , . _ , . , , , - . , , y s____ , , . , . _ _ _ _

. m - e ,r-- --- --- "

/

Accordingly, it is hereby ORDERED that Intervenor Citizens for Nuclear Reactor Safety, Inc. provide to the Licensee full and complete answers to the interrogatories listed ,

above not later than 20 days from the datt 7f this ORDER.

1 i

FOR TIIE ATOMIC SAFETY AND -

LICENSING BOARD Louis J. Carter,' Chairman ADMINISTRATIVE JUDGE l

Dated at Bethesda, Maryland I this day of 1982.

i l 5 t

i L

6 F

2

UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION h'

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of .

ARMED FORCES RADIOBIOLOGY Docket No. 50-170 RESEARCH INSTITUTE (Renewal of Facility (TRIGA-Type Research Reactor) License No. R-84)

+

CERTIFICATE OF SERVICE OF DUPLICATE SIGNED COPIES OF 15 JANUARY 1982 FILING I hereby certify that true and correct copies of the foregoing j

" MOTION TO COMPEL ANSWERS TO LICENSEE'S FIRST INTERROGATORIES '

TO THE CITIZENS FOR NUCLEAR REACTOR SAFETY, INC.," " MEMORANDUM IN SUPPORT OF MOTION TO COMPEL ANSWERS TO LICENSEE'S FIRST INTERROGATORIES TO THE CITIZENS FOR NUCLEAR REACTOR SAFETY, INC," " DECLARATION OF ROBERT L. BRITTIGAN," and proposed form of " ORDER" were mailed this 15th day of January, 1982, by United States Mail, First Class, to the following:

Louis J. Carter, Esq., Chairman e Administrative Judge Atomic Safety and Licensing Board 23 Wiltshire Road Philadelphia, PA 19151 Mr. Ernest E. Hill Administrative Judge Lawrence Livermore Laboratory ,

University of California j

P.O. Box 808, L-123 Livermore, CA 94550 Dr. David R. Schink Administrative Judge Department of Oceanography '

Texas A&M University College Station, TX 77840 ,

l r

l

- _ =

l l

l l

l Mr. Richard G. Bachmann, Esq.

Counsel for NRC Staff U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 Elizabeth B. Entwisle, Esq.

8118 Hartford Avenue Silver Spring, MD 20910 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Panel (5)

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Secretary (21)

U.S. Nuclear Regulatory Commission ATTN: Chief, Docketing and Service Section Washington, D.C. 20555 -

RUAB&p ROBERT L. BRITTIGAN Counsel for Licensee 2