ML20039A182

Response to NRC Request for Admissions.Affidavit & Certificate of Svc Encl.Related Correspondence

ML20039A182
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	11/28/1981
From:	Stillman I CITIZENS FOR NUCLEAR REACTOR SAFETY, INC.
To:	NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
References
NUDOCS 8112160328
Download: ML20039A182 (16)

Text

-

. ,I __,_ y -

?,$U 'O' ';' w, ,,

-ll

(( t } , .

//

9' DEC 1 5 198 Wse i

SA# UNITED STATES OF AMERICA DEC 111981, -

/ NUCLEAR REGULATORY COMMISSION. y, g. . . . ,

. e. ., .,$

4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ro 1 l'""

^

W b In the Matter of #co j3 ARMED FORCES RADI0 BIOLOGY ) Docket No. 50-170 RESEARCH INSTITUTE ) (Renewal -of Facility

) License No. R-84)

(TRIGA-Type Reactor) )

INTERVEN0R CNRS's RESPONSE TO NRC STAFF REQUEST FOR ADMISSIONS

' Now comes the Intervenor in the above-captioned proceedings and, pursuant to 10 C.F.R. 52.742, responds to the NRC Staff request for admis-sions as follows:

1. Omit words "the only" then we admit it as truth.

2. Substitute the. word "Primarily" instead of "Only."

3. No, the statement says "1.4% of the fuet inventory resulting in release of 100% of the noble gases and 0.2% of the iodines" - this is quoted from the clad failure DEA.

4. We agree that the licensee may use a fuel temperature of approxi-mately 600 C in their calculations, however, we believe that temperatures of 800 C or more are likely under conditions that are apt to produce cladding failures.

5. We carnot testify to the truth of this statement since we have no direct information or observation of the way in which the high flux safety channels are checked.

6. 'According to specifications, the high flux safety channels are

^

supposed to scram the reactor at 1.1 MW. However, we cannot admit to the -

811216'O 338F7 hp

s truth of this statement since we have no direct information or observation

-of the way in which the high flux safety channels scram the reactor at elevated power icvels.

7. Whether or not the temperature safety channels will scram the reactor (at approximately 600 C) depends largely on the competent function'-

ing of .the fuel temperature safety channels themselves. There is at least one documented reported failure of the fuel temperature safety channel of' the AFRRI-Reactor. [See " Fuel Temperature-Automatic Scram System" Malfunction Report Jan. 29,1974.]

8. Whereas we believe that the fuel material in the AFRRI TRIGA Re-actor is zirconium hydride alloyed with uranium, we surmise that the exact ratio of hydrogen to zirconium within the alloy is not fixed but may vary within a narrow but significant range.

9. Tests of chemical reactivity (such as those described in your assertion) were performed with TRIGA fuel elements at the General Atomic Corp. Laboratories. To have complete confidence in their results would be comparable to asking the tobacco industry to determine the effect of cigaret-te smoking on the incidence of lung cancer. In other words, General Atomic is the manufacturer of both the TRIGA-Reactor and its fuel elements, hardly an impartial scientific study.is to be expected.

10. We have no reason to doubt the va"idity of this statement, never-theless, we cannot truthfully verify it either (without more direct evidence).

Respe tfully submi,t d, j /

.y ,

W/~ ^

n Irving M. Stillman Dated'at Bethesda, Maryland this 28th day of November,1981

r ..

e UNITED STATES'0F AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSI'NG'B0ARD In the Matter of ARMED FORCES RADI0 BIOLOGY ) Docket No. 50-170 RESEARCH INSTITUTE ) (Renewal of Facility

). License No. R-84)

(TRIGA-Type Reactor) )

INTERVENOR CNRS's RESPONSE TO NRC STAFF's FIRST SET OF INTERROGATORIES Now comes the Intervenor in the above-captioned proceedings and, pursuant to 10 C.F.R. 52.740b, responds to the NRC Staff's first set of Interrogatories as follows:

General Matters

1. (a) Irving.M..Stillman, M.D., Ph.D.

5480 Wisconsin Avenue Chevy Chase, Maryland 20815 B.S. 1955 Queens College, C.U.N.Y.

M.D. 1959 Washington University School of Medicine Ph.D. 1968 Polytechnic Institute of Brooklyn Earl A. Gulbransen, Ph.D. , M.E.

Research Professor Department of Metallurgical & Materials Engineering University of Pittsburgh Michio Kaku, Ph'.D.

Associate Professor of Theoretical Physics -

Physics Department City College, City University of New York ~

4

+

^

_2-Ernest J. Sternglass, Ph.D.

Professor of Radiological Physics University of Pittsburgh

^

Daniel Pissello, Ph.D.

Associate Professor of Physical Sciences Fordham University, New York City.

Leonard R Solin, Ph.D.

Division of Radiation Safety

. Department of Hea l th, New York City Mr. Leo Goodman

.2948 Macomb Street, N.W.

Washington, D.C.

4 e .

' h 9

0

o .

t 3-

~

(b) See. bracketed language in the body of the following answers

' for substantiation of Dr. Stillman's views. As to the other persons listed in 1(a) this information is not available to the Intervenor at this time.

(c) The answer to this question would be very lengthy and

. - technical. It is, therefore, a matter reserved for direct scientific testimony by the Intervenor's experts during the actual hearing.

(d) These summaries are unavailable to the Intervenor at this time.

(e) This information is unavailable to the Intervenor at this time.

e 4

.. _ _ . _ _ _ . _ _ _ _ _ ._-_-___._-.-____m. -

. \

2. (a). No.

(b) _See responses given to " Request for Admissions."

3. See responses given to " Request for Admissions."

STIPULATED CONTENTIONS -

Contention 1

1. HSR does not specifically state that the cladding failure would

" occur at a peak fuel temperature of less than 100 C," however, in describing this accident they estimate that only 0.2% of the iodine will be released.-

Since iodins v'olatizes at about 183 C one must assume that they are using rather icw temperatures (less_than 100 C) to account for only a 0.2% release.

2. Because surface cracks have been observed in fuel element clad-dings primarily when they were overheated (see H. H. Hausner and J. F. Schumar in " Nuclear Fuel Elements," p. 84).

3. At fuel element temperatures of 900 C or more the quantity of fission product sap activity could easily reach 5% or more of-the radioactive inventory [see Fig. 5-1 in "The U-ZrH x Alloy: Its Properties and Use in TRIGA Fuel," by M. T. Simnad].

C_ontention 2

4. Since experiments performed at AFRRI are authorized on an in-dividual basis, it would be virtually impossible for us to predict (at this time) which future experiment is likely to fail.

5. See answer 4. .

6. See answer 4.

7. On February 1,1973 the Reactor Core Position Safety Interlock System that coordinates the lead door / core movement malfunctioned due to a faul ty 'de-energizing' relay. This-system, when functioning properly, brings the lead, door.into near ccntact with the core shroud. This provides important 6

'O

I v l

- shielding.from any excessive radiation developed within the core during reactor -

operation.

8. The Applicant describes several built-in safeguards that would alert the reactor operator (i.e., safety system channels, safety system set-tings, and radiation monitoring systems) to effect the necessary corrective measures (e.g., initiate a reactor scram, engage the appropriate interlock sys-tem, add water coolant, close the ventilation system, etc.) should an accident even threaten. However, the Petitioner contends that human errors coupled with equipment fail'u'res can render these safeguards ineffective, as has occurred

. repeatedly in nuclear reactor accidents world-wide. To demonstrate that'such failures can also occur in this particular reactor (AFRRI-TRIGA) we shall cite several instances of relevant malfunctions involving these safeguards reported by AFRRI to the appropriate federal regulatory agency, including:

(a) On February 1, 1973 the Reactor Core Position Safety Inter-lock System that coordinates the lead door / core movement malfunc-tioned 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 mechanical contacts of the TZ output meter that initiated the automatic scram through a relay."

(c) On August 26, 1975 the Radiation Monitoring System mal-

> functioned, i.e., the reactor room ventilation dampers failed to close when'the Continuous Air Monitor Sensing Device was manually

~

triggered. " Inspection revealed that two wires in the control' box were loose" and that this was the apparent cause of the malfunction.

l (d) On July 10, 1979 there was a malfunction of the Pool Water j

Level Sensing Float Switch that monitors the reactor pool water level.

in case of an impending LOCA. "The malfunction was caused by wear

• ~ '

- _g . - y gy ,+-- ,, g __ y. . _A- y

I on the jacketing around the wires leading to the switch which p o-vided a path to ground, thereby circumventing 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 Applicant admits that "since this system monitors the principal safety parameter of the reactor, it was felt that a more secure ground was required."

(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.

(g) On March 15, 1980 there was a malfunction of Safety Channel One such that most of the scram indicators on the reactor central console were illuminated. Furthermore, an inspection on March 17, 1980 " revealed that Safety Channel One would nc+, initiate a scram in accordance with the Technical Specifications of Reactor License R-84." The cause of the malfunction was attributed to a damaged operational amplifier on a Safety Channel One circuit board "when electrical power had been reapplied to the console after a power outage."

The Applicant alleges that even if there is a power excursion in the AFRRI-TRIGA and the built-in safeguards malfunction, the reactor will automati-cally shut down due to the negative temperature coefficient (i.e., -0.017%

reactivity decrease per 1 C rise in fuel temperature). This automatic shut-down is entirely dependent on the relative amount and energetic state of the hydrogen nuclei within the uranium-zirconium hydride alloy. Thus, any signifi-cant deviation of th'e hydrogen parameters from their expected values or cur.ves r

will' cause a drastic change in the negative temperature coefficient. .The 4

- ~ . , ,we.

l 1

exact accident scenarios and their resulting releases of radiation are de-scribed in detail in the following sections of these . interrogatories.

9. The hydrogen nuclei contained within the TRIGA fuel alloy (U-ZrH x

) can theoretically exert their " moderating effect" in two general ways (based on relatively elastic collisions between neutrons and hydrogen nuclei).

(1) The transfer of energy from " fast neutrons" colliding with

" cool hydrogen nuclei" to slow down or "thermalize the neutrons."

(2) The transfer of energy from " warm hydrogen nuclei" to speed up neutrons already thermalized by the cool tank water (via elastic collisions).

Since fission requires neutron capture by the uranium nuclei which, in turn, is a function of the neutron's velocity, we see why the " hydrogen nuclei" can exert the crucial moderating effect with respect to the rate of fission.

10. Experiments performed at Brookhaven National Laboratory were ex-plained by assuming that in the TRIGA fuel alloy the hydrogen-atom lattice vibrations can be described by an Einstein model with a characteristic energy hz = 0.130 electron volts. This descriptiorcis consistent with the theory that the hydrogen atom occupies a lattice site at the center of a regular tetrahedron of zirconium atoms. For the most part, zirconium hydride is not very effective in thermalizing neutrons (because hz > KT), but it can speed up neutrons already thermalized (i.e., mechanism (2) in the answer to Question 9 above) by transferring to each of them a quantum of energy hz, via a relatively elastic collision with a hydrogen atom. However, unusual, rapid changes in temperature and/or pressure are apt to produce serious changes in the structure of the alloy affecting its lattice vibration frequency, z (e.g.,' phase transi-tions of the zirconium hydride, such as the alpha phase transition that occurs at 530 C). Mathematically, the negative temperature coefficient is primarily

e 8-a function of e-hz/kT so that any significant change in z (which is possible during accident conditions when the fuel elements are damaged) will necessarily modify the coefficient. (See A. W. McReynolds, M. Nelkin, M. N. Rosenbluth, and W. Whittemore, " Neutron Thermalization by Chemically Bound Hydrogen and Carbon," published in the Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy," Geneva, 1958.]

11. Yes.

12. All 85 fuel elements could fail under certain conditions explained under what we contend could be a maximum credible accident.

13. The answer to this question would be very lengthy and technical. It is, therefore, a matter reserved for direct scientific testimony by the Intervenor's experts during the actual hearing.

Contention 4

14. Since the statistical uncertainty in the annual perimeter dose per monitoring station (at the 95% confidence level) is t 20 mrad, it is likely that the annual population exposure at several unrestricted area stations has exceeded .the EPA limit (i.e., 25 mrem) for just about every year during the pasc 19 years of the AFRRI Facility ' operation. Furthermore, since the environ-i l mental film dosimetry method employed at the monitoring stations detects only external gamma radiation, the population radiation exposure dose due to the.

l inhalation or ingestion of radio nuclides is entirely neglected. The absence

of data due to omission of internal radiation doses makes it virtually impos-1: rible to evaluate the true population exposure to radiation let alone determine whether or not federal regulatory limits have actually been exceeded. There have been no corrective actions taken by the AFRRI to prevent or compensate for these deficiencies.

15. Argon-41, Nitrogen-13, Oxygen-15, Xenon-133, Krypton Sodium-24,

' Tritium.

16. and 17. We would need an accurate and complete list of radio

' ~

nuclides emitted (i.e., gaseous, liquid, and solid effluents) tefore we could knowledgeably answer these questions.

18. Neither the Applicant nor the Petitioner have any way of answer-ing that question since apparently no one has done a good quantitative analysis of the particulates actually released after filtration by the_AFRRI Reactor.

The Applicant admits that the particulate radio activity 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. Furtnermore, the only other AFRRI stack radioactive effluent monitoring system, the radio-active gas monitor, is certainly not reliable for particulate radioactive

~

I sampling. Hence, one can only obtain crude estimates of the airborne radio-active particulates that have been dispersed into the environment. *l,e true values may, in fact, have exceeded public safety limits.

19. Because its being " isokinetic" would enable it to give somewhat more quantitative evaluations.

20. Possibly. Jha have not received any recent data obtained by Gross Beta Effluent Analyses. Even more important, we are not privy to observ-ing or monitoring the Applicant's environmental sampling and analysis pro- .

l cedures.

l

!' +

~

21. Effluent analyses by the Washington Sanitary Sewage Commission during the past two years strongly implicate the AFRRI as having released unhealthy quantities of radioactive liquid effluents (see their recent evaluations].

22. Samples should be taken more frequently and undergo more stringent, quantitative analyses. Furthermore, analyses should be done by an independent scientific laboratory employing some of the newer analytic techniques.

23. The answer to this question would be very lengthy and quite technical. It is, therefore, a matter reserved for direct scientific testimony by the Petitioner's experts during the actual hearings.

24. 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 mqnths. Unless one con-cludes that the required elevation for discharge of effluent is wholly unneces-sary and to that extent superflucus, one must assume a high probability that at one time or another (over a period of several months) the 10 C.F.R. part 20 limits were exceeded.

25. We are certain that prior to establishment of this reasonable regulation, the appropriate federal agencr did the requisite calculations demon-strating a potential public danger if the reactor ventilation were to be ex-hausted at ground level. Otherwise, there would be no such regulation; please refer to those predictive calcu'lations.

26. and 27. Without any special exchange of new information concern-ing the operation of the AFRRI-TRIGA, we cannot be expected to know the more recent change.s that might reduce the concentration of the Argon-41 released, to the atmosphere.

UNSTIPULATED CONTENTIONS Contention 1

28. About 250 gallons of cooling water per minute.

29. The actual pulse mode interval is irrelevant because the pulsing can be repetitive (i.e., the pulse mode interval can be repeated over and over

-again) especially if the safeguard systems malfunction.

30. The Applicant claims that "in the event that the only mechanism for heat removal after the loss of coolant (LOCA) is natural convection of air through the core, the fuel element cladding would not fail and the fission products would be retained within the tuel element." We contend that.with a rapid loss of coolant (e.g., mere than 250 gallons of water per minute) in an actively operating reactor core (i.e., in a recurrent pulsing mode) there could be a sudden temperature elevation (> 900 C) sufficient to effect multiple cladding failures attended by the release of their gaseous fission products into the reactor room atmosphere. Surface cracks in overheated (> 900 C) fuel ele-ment claddings have been demonstrated elsewhere on several occasions (see H. H. Hausner and J. F. Schumar in " Nuclear Fuel Elements," p. 84). At fuel temperatures of 900 C or more the release of radioactive gas into the fuel-ele-ment gap would be ba'out 5% of the element's radioactive inventory which, in turn, would seep out of the cracks or degradations in the fuel element cladding and into the reactor room. Any breach of containment within the reactor room would permit radiation leakage throughout the entire AFRRI Facility (e.g.,

l through the ventilation system) and possibly extend beyond the building itself into the outside air as effluent or by direct radiation to the M'Je (t'hrough

~

i the roof and walls of the building). The Applicant has failed to address or l

even anticipate the effect this would have on the public health and safety. -

31. and 32. We do not know of any documented cases in which a TRIGA

[ Reactor-suffered a LOCA in the pulse mode causing multiple cladding failures,

'~

9

-o if there are'any such cases you should bring them to our attention. However, there are many documented examples of other nuclear reactors that suffered multiple cladding failures resulting from _a LOCA, Three-Mile Island being the T

most famous. .

33. -Once again, we must insist that such calculations form an in-tegral part of our expert testimony which we would rather reserve for the actual hearings.

Contention 2

34. _There are several ways in which a Power Excursion Accident (PEA) could be initiated 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 excessive reactivity (i.e., radioactivity associated with the experiment itself) to produce a dangerous transient (i.e.,

excess reactivity above cold' critical).

(3) A Stuck Transient Rod - if the most reactive control rod (i.e., the transient rod) is stuck out 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 withdrawn to achieve a steady state power greater than 1 MW).

The common factor in all 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

n

-A C, s ,

- excursion,or transient) capable of causing multiple cladding failures at ele-

'I n Vated fuel element temperatures.

s, ,

35. Multiple cladding failures could reduce the thermalizing effect of hydrogen in at least two ways:

(1) The hydrogen collected within the fuel element gaps could leak out through degradations in damaged fuel element claddings.

This would disrupt the hydrogen (pressure) equilibrium within the

s fuel elements such that more and more hydrogen would be released fron$ the fuel alloy itself into their respective gaps and then out

.through the cracks in the claddings, thereby resulting in a serious depletion of the hydrogen in the zirconium hydride. Without this hydrogen there can be no "thermalizing effect cf hydrogen."

(2) The hydrogen pressure change (via the mechanism just de-scribed) could easily effect a phase change of 'het zirconium hydride lattice (poisoned with uranium atoms) such-that the vibration fre-quency of the remaining hydrogen, z, is significantly modified. A change in z produces a modified quantum of vibrational energy, hz, which is necessarily exchanged with the neutrons during the process of neutron thermalization. This could certainly modify the " therma-lizing effect of hydrogen" and the all-important " negative tempera-ture coefficient" which is an exponential function of hz.

36. Calculations already exist that demonstrate 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 Re-actor located at Columbia University in New York City). These calculations show that a LOCA (in this tank-type TRIGA Reactor) can result in fuellelement tem- .

peratures up to 1,200 C. At these elevated temperatures (when they occur rapidly as is the case with a LOCA) cracks in the fuel element claddings are-

}

quite likely (see H. H. Hausner and J. F. Schumar in " Nuclear Fuel Elements,"

p. 84).

I

37. Large amounts of heat are released both in the air and steam re-actions. For the reaction of zirconium with oxygen, 262 kcal/g mole of zirconium is released at 1,000C; for the reaction of Ir U H with air even more heat 0.034 x is released. Once started, there is no easy way to stop the explosive reaction.

For the reaction of zirconium with steam, 141 kcal/g mole of zirconium is re-leased at 1,000 C. In addition, 20% more hydrogen is released, which could cause more problems if the hydrogen is confined or if mixed with air to give an explosion. Again, there is no easy way to stop the zirconium-steam reaction once it gets started. Temperatures for initiating these catastrophic reactions are between 800' and 1,000 C (see letter to Professor Jay Marcus, June 1, 1979, Columbia University from Professor Earl A. Gulbransen). Similarly, an NRC Re-

- port indicates that the rate of a violent zirconium-water (or steam) reaction becomes significant at about 900 C (see NRC memorandtyn to Roger Mattson from R. O. Meyer, dated April 13,1979, " Core Damage Assessment for TMI-2," p. '25).

Unless designated otherwise, all responses to Request for Admissions and all answers to the Interrogatories herein were answered by me.

Respectfully submitted,

. /'" '

/

Irving M Stillman Dated at Bethesda, Maryland this 28th day,of November, 1981

, ::c::- .-- m r:

$g~ *

//  :

~ .-

' O Og ?"= DEO 11198f

• %':n: q AFFIDAVIT OF IRVING STILLMAN 6^ 'U I '

NI I, Irving Stillman, being duly sworn, do state:

1. That the responses of Intervenor Citizens for Nuclear Reactor Safety, Inc. to the NRC Staff Request for Admissions, to the~ NRC- Staf f's First Set of Interrogatories, and to the Licensee AFRRI's First Set of. Interrogatories, that are desig-nated "Stillman," were asnwered by me.

2. That said responses are true to the best of my knowl-edge, informatior and belief.

. N '

W Irving S illman

/

SUBSCRIBED AND SWORN to before me this 2f day of / V O Vf 7M 6,f/-

1981.

g4 .-

Notary Publ AN ':Sn*:.CMTE NOTARY r.'i UC STA72 CF MA.i."AND. * .

- My Commissien Er.!res Ju!y 1,1982 no a

r i-7 ,

UNITED STATES OF AMERICA . . ;. _

~

NUCLEAR REGULATORY COMMISSION A

~~

~C,

. .d '

C/

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD - ,61 Ly/g,.

g ,7., y In the Matter of f$

ARMED FORCES RADI0 BIOLOGY ) Docket.No. 50-170 RESEARCH INSTITUTE ) -

) (Renewal of Facility .

(TRIGA-TYPEReactor) ) License No. R-84 i

l CERTIFICATE OF SERVICE i I hereby certify that copiesrof "Intervenor CNRS's Response to NRC Staff _ Request for Admissions," "Intervenor CNRS's Response to NRC. Staff's First Set of Interrogatories," and Affidavit of Irving Stillman have been served on the following by depositing in the United States Mail, first class, this 3rd day of December, 1981:

Louis J.' Carter Esq., Chairman e

Mr. Richard G. Bachmann, Esq.

Administrative Judge Counsel for NRC Staff

' Atomic Safety and Licensing Board Office of the Executive Legal Director

-23 Wiltshire Road U.S. Nuclear Regulatory Comission 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 txecutive Legal Director P.O. Box 808, L-123 U.S. Nuclear Regulatory Comission ,

Livermore, CA 94550 Washington, D.C. 20555 -

Mr. David R. Schink Atomic Safety and Licensing '

Administrative Judge Board Panel .

Departnent of Oceanography U.S. Nuclear Regulatory Comission Texas A & 11 University Washington, D.C. 20555 College Station, TX 77840 Atomic Safety and Licensing Docketing nd Service Section Appeal Panel Office of the Secretary U.S. Nuclear' Regulatory Comission U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Washington, D.C. 20555 Mr. Robert L. Britt'igan, Esa. '

General Counsel .

Defense Nuclear Agency

. Washington, D.C. ~20305

, i r ,. 1 j ' 4.l ? Y

'~ ,' ; . d 1 [ ? Q Elizabeth B. Entwisle-Counsel for Intervenor u .

, , , J- ,

,