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Supplemental Contentions to 800522 Petition to Intervene. Alleges Omission of Essential Info by Applicant Re Severe Reactor Damage & Lack of Spare Parts.Supporting Documentation & Certificate of Svc Encl
	ML19331D906
Person / Time
Site:	05000142
Issue date:	08/25/1980
From:	Hirsch D; COMMITTEE TO BRIDGE THE GAP
To:	; Atomic Safety and Licensing Board Panel
References
	NUDOCS 8009040203
	Download: ML19331D906 (140)
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NUCLEAR REUULATORY COM'4ISSION v

BEFORE THE ATOMIC SAFETY AND LICHISEIG 20ARD In ,the Matter of THE REGBITS OF THE UNIVERSITY Docket No. 50-142 0F CALIFORNIA (Proposed Renewal of (UCLA Research Reactor) i COMMITTEE TO 3RIIGE THE GAF'S SUPPL 34 ENTAL CONTSITIONS TO PETITION FOR LEAVE TO INTERVHIE Daniel Hirsch 1637 Butler Street Ios Angeles, CA 90025 President: Campus Connittee To Brid 6e The Gap e

8009040203 G w ,w -

O 6 UNITED STATES OF AMERICA NUCLEAR REUIATORY COMMISSIQi BEFORE THE ATOMIC SAFETI AND LICalSING .30ARD In the Matter of THE REDITS OF THE UNIVERSITY Docket No. 50-142 0F CALIFORNIA (Proposed Renewal of Facility License)

(UCIA Research Reactor)

COMMITTEE TO BRIIGE THE GAP'S SUPPIlMHITAL CCITINTIONS TO PETITION FOR LEAVE TO DiTERValE INTRODUCTIQi On April 25, 1980 a notice was published by the Commission offering a hearing on the renewal of the operating license held by the Regents of the University of California (Applicant) for the UCIA Research Reactor.

& May 22, 1980 the Conmittee to Brid 6 e the Gap (Petitioner) filed a petition for Leave to Intervene. @ July 22, 1980 the Atomic Safet/ and Licensing Board scheduled a pre-hearing conference for September 18, 1980, which was subsequently rescheduled for September 25, 1980. In its orier !

scheduling the pre-hearing conference the Boari invited Petitioner to submit supplemental contentions to its original petition for Leave to Intervene. These supplemental contentions are hereby being submitted by Petitioner in response to the Board's invitation and in confor=ance with the requirements of 10 CFR 2 714(b).

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10 CFR 2.714(b) allows an intervention Petitioner to supplement his petition with the " contentions which petitioner seeks to have liti-gated in the matter, and the bases for each contention set forth with reasonable specificity " prior to the Special Pre-hearing Conference.

At the Conference the Boari is to determine whether petitioner's contentions meet the requirements of 10 CFR 2.714(b), and thereby deter-mine the scope of discovery.

These supplemental contentions submitted by petitioner hereby incoporate by reference, and are supplemental to, although not limited by, the petition for Leave to Intervene. The supplemental contentions cover a broad range of issues and problems, reflecting the wide variety of deficiencies in the reactor operation and the fact that the license request may be jud 6 ed against a twenty year operatinJ!; history. The supplemental contentions 'dll focus en three major areas of argument:

1. That the application filed by UCLA is incomplete, misleading, contains material errors of fact, and is generally inadequate to support the issuance of the requested license 2. That the history of deficiencies in the reactor opo stion over the previous twenty years makes it impossible for the Applicant to reasonably assure that, in the future, they will comply with the regulations applicable to them, and that they will not endanger the public health and safety; 3 That inherent problems of the reactor, such as, age, seismic vulnerability, location in a densely i populated area, etc.) indicate that the reacter cannot be operated in a manner that will not be inimical to the public health and safety.

Each contention set forth below must be considered in the context of the fact that PetPioner has had no opportunity for discovery and has

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3 had no access to a public reading room, and is merely setting forth the areas of contention to be litigated in the licensing process. The burden in this licensing process, under the provisions of 10 CFR 2.732, is on the Applicant. . They must meet the common standard set forth in

! 10 CFR 50.40(a) and in 10 CFR 50 57(3) that the whole effect of the license application and the past operation is to " collectively provide reasonable assurance that the applicant will conply with the regalations in this chapter, including the regulations in Part 20, and that the 4

health and safety of the public will not be endangered." Each of Petitioner's contentions ultimately goes to the fact that either the information included in the application, the past operating practices j and history, or current characteristics of the reactor, indicate that the Applicant has not met the burden necessary to support the issuance of an operating license.

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1. APPLICATI0:I G?OSSLY LADECUATE The Application--including the supporting appendices--is so substantially flawed that it fails to meet even mini al standards for -

such applications and fails to adequately demonstrate that the health and safety of the public will be properly protected if the license is Cranted, thus r.aking approval of the Application impossible on the basis of its inadequacies alone, j 3pecifically,

1. Cmisdon of essential infomation.

a. Applicant stated on page II/3-1 of Application:

The UCLA Reactor has been stfojected to experinental vibration.

The results were reported by C.B. Smith at the 'dinter Meeting of the American :iuclear Society, ::ovember,1968, in a paper titled " Vibration Testing and Earthquake Response of !!uclear Reactors."

I However, Applicant failed' to state that the UCLA reactor failed the test in question and that damage to the reactor was so severe that a control blade eventually stuck in the outfosition, requiring dismantling of the reactor core, necessitating substantial radiation exposure to repair workers and raising serious questions about the safety of the

! UCLA reactor in case of an earthquake. The report the Applicant cites above without describing its contents states the follow' ig:

About six nonths after the vibration experiment routine tests indicated that ene of the control blade insertion timen had increased. A few months later safety blade !!o.1 stuck in the "out" position during a routine prestart checkout of the reactor control system.

'Jhen the reactor was disnantled, we discovered that lead shielding bricks had been displaced upward, causing the shaft to bind.

" Vibration Testing", p. 24

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- b. In identifying agents responsible for the reactor construction and design, Applicant failed to mention that the company that built the reactor, AMF, is no longer in the reactor business and that this has led to difficulties in getting spare parts:

Some of the reactor instmnentation is still workable, but sometimes unreliable, and is very difficult to i

repair due to its age and the resultant problem of

' obtaining parts (e.g. , vacuum tubes, specialized switches, indicators, and meters).

HEL 1976 Annual Report, p. 35

c. Despite certification (notarized) by Dean O'lieill, Dr. ;fegst, i

and Vice Chancellor Hobson on page 10 of the Application that they certify that these applications are prepared in conformity with Title 10, Code of Federal Regulations, Parts 50 and 70, I and so solemnly swear (or affim) that all infomation contained herein, including any supplements attached hereto, is true and correct to the best of our knowledge and belief s

there is virtually no other mention of Part 70 in the entire document.

The document has not be 'n prepared in confomity with 10 CFR 70 (the requirements for applicatiens for Special nuclear Mater _.als licenses) and virtually none of the infomation required by 10 CFS 70 is included.

- 2. Submission of an original acclication, i An elementary requirement for an adequate application for facility license is that it be written about the facility for which the license is being applied. And yet much of the Application submitted by the j

University of California is not original but was lifted verbatim from a 20-year-old Hazards Analysis and fron a 6-year-old AEC neto on l i research reactors as a group. Dr. ;tegst, in the cover letter to the Application, cdnits that "this application contains only ninor changes 3

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2a (listed in the forward to appendix 5) from tho orginal Bic] application."

The failure.to write an original application about this specific reactor as it exists today and as it ely to exist duling the 20-year period for which it has requested a license is so fundamental as to call into question the good faith of the Applicant's effort and mandates the su:n.ary- rejection of the Application.

a. Safety Analysis Largely a Retroed Version of 1960 Hazards Analysis The SAR submitted with this application is a virtual verbatim copy, with some pages added cnd a few passages removed, of the Hazards Analysis for this reactor written twenty years ago. The reactor's characteristics have changed significantly since then, as has the state of nuclear safety knowledge, but the 1980 SAR and the 1960 Hazards Analysis are substantially identical (as can be seen in the sa=ples included herein on pages/4/O In 1960 the reactor was in its own two-sto:y building; today classrooms and officos have been attached on almost all sides. The reactor now operates at 10 times the power, four times the excess reactivity, with a pneumatic rube " rabbit" system not in existence in 1960. Since the Hazards Analysis was written new characteristics such as a positive graphite temperature coefficient and a different void coefficient have been identified. Further:. ore, new information is now available about seisnology, meteorology, hydrology, and reactar i

perfomance. 'fet with all this new information available the Applicant mereb retyped much of the old 1960 Hazards Analysis with minird updating and no new analysis. What changes have been made, besides some changes updating such things as population figures, are thtt the mention of some safety features such as a deflector plate (to prevent repeated excursions)

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and limits on excess reactivity to less than that needed for prompt criticality have been recoved. All in all, there is no new analysis of the safety. of this reactor, merely repition of an outdated analysis.

In short, the Applicant has analyzed the hazards of a reactor far different from the one it seeks to have licensed,

b. Lack of orieinal environmental imoact acuraisal for-this reactor.

Applicant has ostensibly filed an IIA for this particular reactor, but much of the language has been lifted, without attribution and virtually verbatin, fion Daniel Huller's AEC meno of January 23, 1974, on "Envimnnental Considerations Regarding the Licensing of Research Reactors and Critical Facilities." There is virtually nothing on pages II/3-1 thmugh 7-1 that was written by the Applicant nor can it be said that the contents of those pages represent a review of the environnen'4 aspects of Applicant's specific facility. Applicant made no showing that Muller's general conclusions fit the specific circumstances of UCLA; nor for that matter did they identify the language as anything but theirs.

3. Misleading and inaccurate, statements.

a. Page 5 of the Application states that the use to which the facility will be put is:

the education of *enior undergraduate and graduate studen'.s ;

in nuclear engineering and related sciences. In addition to fomal course 9 and demonstrations, the reactor vill be used to cupport research at the Z.S. and Ph.D.1cvels.

This statement is at best misleading, becauce a chart on p. III/1-5 of the Application indicates that last year only 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> l

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s of reactor operation were spent on instruction. Further, a substantial portion of the reactor's "research" usage is rental and sale of services to comercial concerns, prbndly activation analysis for ore assaying.

(for details, see contention on "'Irong Class of License.")

b. Applicant states on page 7 of the Application that "no structural weaknesses (earthquake vulnerability) have ever been identified."

Yet Professor Catton's 1976 Annual Report states: "The Februar/1971 earthquake gave rise to minor problems that worsened wi*h time and ultimately required a najor maintenance effort in 1972}" a seven-month shutdown. (p.3.1976 report). Applicant's statenent is further contradicted by the earthquake sinulation test mentioned in 1.a. above,

c. The Foraani to the Technical Specifications (p. V/i) states:

The Technical Specificatione c.; stained in this appendix, enbody the earlier Technical . 3cifications (of 1971 as amended in 1976), in revised 'rnat and expanded content.

iith four exceptions noted be' ;r, no attempt has been nade to alter the content and provr.&ns of the earlier Technical Specifications, and an/' ot.her caserepancies should be interpreted as typographical errors or editorial deficiencies.

Notwithstanding the above t.ssertion by Applicant, significant c' r.nges besides the four noted in the Forward have been nade in the T .:h Specs included in the Application. '!hether their cause be typogr .phical errors, editorial deficiencies, or attempts to alter the content, the following changes appear to have been nade without so stating:

1. the requirement for an annual heat balance-inst: enent calibration has been renoved.

ii. the definition of " annual" for the remaining ca ibrations has been changed from once everf 12 nonths to once ever 14 months.

iii. the excess reactivity limitation has been changed fron 2.3shk/k to 53.54, apparently utilising ad of .0065. If the /3 of the 1960 hasards Analysis, upon which the change to 2 3s was based, uere used, that would mean an actual increase of excess reactivity to 2.62% k/k, without any approval or request for approval.

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l iv. removal of language requiring that ALARA be met.

' v. removal of specifications regarding height of exhaust stack and flow rate of emissions out of exhaust stack, i l

The above items are discussed in more detail in the contention about

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the Technie.a1 Specifications. !

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! d. ' Applicant on page III/3-1 lifts verbatim the language of the l ic60 Hasards Analysis that 4

Uo deep wells have been drilled on the campus or in the vicinity of the campus. The water table is estimated to lie 200 feet below the surface of this area.

However, page /f of these contentions shows a hydrology map for the area indicating that there are a number of wells near the campus,

e. Applicant on p. II/3-1 lifts verbatim the language of p. 3 j of the Huller AEC memo that Accidents ranging from failure of experiments to the largest core damage and fission product release considered possible result in doses of only a s=all fraction of 10 CFR 100 guidelines and are considered negligible with respect to the environnent.

However, p. III/3-6 of the Application indicates that in event of an accident Applicant's own estinate includes i thyroid dose to members of the public of ].800 ren to the thyroid, considerably ja excess of the 10 CFR 100 guideline of 300 rem thyroid dose. (See contention regarding Failure to Heet 10 CFR 100 Guidelines.)

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4. U_nsupuorted and/or unsunnortable _ statements,

a. Applicant on page II/5-1 essential'y repeats the Muller language, There are no suitable or more eco tonical alternatives which can accomplish both the educatior.a1 and the research i objectives of this facility.

Applicant fails to support this statensnt with any serious consideration of the alternatives. (see contentien on F.nvironnental Impact Appraisal).

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b. Page V/3-6 of the Application states SPERT and SORAX tests showed that plate type fuel elenents survived step reactivity insertions of J3.54 As a careful reading of Appendix A of the SAR indicates, SPERT and SORE tests showed no such thing. It was merely extrapolated from SORE data, which was for a water-moderated reactor, not a water and graphite moderated one like the UCLA reactor, that .6J5 6 k/k was relatively safe because it wasn't until one got in the range of 2.7) A k/k that a cladding nelt could occur. As ic seen in the Appendix narked Figure D-5 in the Hazards Analysis, there is no 203AI data available in that range. Furthernore, the translation of 2.3,5 4 k/k into :J3 54 is highly questionable because of the use of a S tant differs from the one used in the intial analysis.

(For a detailed discussion, see contention on excess reactivity.)

CC:CLUSI0l The above citation of omissions, =isleading statements, inaccuracies and inadequacies contained in the Application represent but a snall number of such items that have been found by Petitioner. However, further deb %g of these itens would be burdensone on all parties at this stage because of their great nunbor. Furthermore, many of the itens not included here are included in other contentions. It has been our intent nerely to show sufficient basis for our centention regarding the inadequacy of the Application itself.

The above citations are sufficient to present a picture of an n

Appli cation uhich is gmssly inadequate to meet the requirments for the issuance of a reactor license. There is clearly insufficient accurate infomation about the facility as it now exists and is likely to exist .

during_the preposed license period to serve as a basis for the 3oard to 4

conclude that the reactor uill be operated in a nanner which complies j with the regulations and uhich will not endanger the public health and safety. One of the principal burdens upon an applicant for license 1

] is the preparation of an adequate application. 'lithout such an adequate

! application it is h:possible for the requested license to be issued.

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Furthemore, the onissions and misleading and inaccurate statenents are so serious as to call into question Applicant's compliance with Section 186 of the Atenic ;i.nerg Act of 1954 (42 U.S.c. 2 2236) as interpreted in ViW_.ia Z1ectric G Fower Co. (Uorth Anna Power Station, Units 1 & 2) ALA3-324, 3 !RC 347, (1976).

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3.0 ENVIRONMENTAL EFFECTS OF ACCICENTS .

Accidents ranging from failure of experiments to the largest core damage and fission product reTease'~ considered possible result in. doses of only a small fraction of 10 CFR Part 100 guidelines and are considered neg1Tgible with respect to the environment. The UCLA Reactor has been_Suhted to .N

/ experimental vibra 1jon. The results were reported by C. B. Smith at the '

[ W1nter Meeting of the American fiuclear Society, November,1968, in a paper titled ""ibration Testing and Earthquake Response of Nuclear Reactors".

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VIBRATION TESTING AND EARTHQUAKE . :%KTML..s@g w.asi m~ ~~.ga RESPONSE OF NUCLEAR REACTORS

                           .                                                                          K EYWORDS: reactors, reactor broteon         ng
   ,'}           CRAIG B. SMITI! and R. B. MATTHIESEN Nuctsar Energy Laboratory,                      s
        .'       Earthquake Engineersng and Structures LaboratorSe                                    mathematics, sensitivity, sta.

1 'os Angeles, California 90028 t bility, analysis, motion, earth. quakes, UCLA, EGCR, CYTR

      'i             Received November 25, 1968
      ,;             Revised February 24, 1969 s

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O., jD reactors in strong motion earthquakes, it will be necessary to predict their performance with Vibration testing of nuclear reactors is dis- studies based on simulations and analysis. cussed as a part of the determination of the To discuss vibration testing of nuclear reactor response of such systems to earthquakes. The systems, one needs to consider the use that will

           ;      basic theory of vibration testing is presented be made of the tests. The obvious use is to deter-along inith a comparison ofimpulse, ambient, and mine the dynamic response of key reactor sys-
   *;             steady-state testing. Steady-state tests provide a tems. We believe that this is important, but it is
         ;        method of obtaining the complete dynamic charac- also important to use the test results to check the teristics of a system and of selectively studying validity of mathematical models of structures.
       ,          each of the components of the system; e.g., con- There is considerable need for analytical models
       ';         tainment, steam generator, pressure vessel, in- that will accurately predict the response of large
          '       simmentation, etc. Generally, both impulse and nuclear power plants to the vibration effects of.

i ambient studies do not provide as much detailed earthquakes. information ichile being less time consuming and Much work has been done in the fields of sels-

       )l ;s      creating less interfere.ce tvith other operations. mology and earthquake engineering,and we believe
 .Tl                    A series of tests performed on the UCLA re- that it is possible today to construct a "first ap-
   .?,;           search reactor, the Carolinas-Virginia Tube Re- proximation" to a complete analytical modeL We actor, and the Experimental Gas-cooled Reactor are surveying this work and are attempting tc

.4 y; :' at Oak Ridge are used to illustrate results ob- draw it together to construct an overall model.

1 tained icith steady-state tests. These illustrate Where possible, we plan to use our own experi-the effect of the vibrations on instmmentation as mental work or the wcrk of others to verify the Q~, icell as the response of the reactor cores, fuel model.

l. elements, biological shielding, steam generators, in addition, we expect that the experimenta:

       ^l          exhcust stacks, and the containment stmetures. work we have done will indicate areas, if any I    The tests of the UCLA reactor included tests zcith where nuclear power plant design requires furthet
               >    the reactor atfull porcer. The examples illustrate r,esearch and development. Once a complett the complexity of the soil-stmeture-reactor sys- analytical representation of the earthquake-soil-tem ichile also iruficating the nature of the results structure nuclear reactor system is available, i which may be obtained with vibration tests.           will be possible to study the sensitivity of the model to variations of its parameters. Sensitivil',
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6 mo , w M analyses can pinpoint areas in the system where additional research is required or where addi-tional research would lead to significant improve-INTRCOUCTION ments in the stability or safety of the system. Knowledge of the effects cf earthquakes on nu- In this paper we discuss some analytica clear reactor safety will be increasingly important models and several experimental techniques fo: as more nuclear power plants are constructed in testing reactor structures. We compare the ad seismic regions. Until the time when we have ex- vantages and disadvantages of the several testin-perienced the actual behavior of large power technique, based on our experience in the field 6 NUCLEAR AFF1dCAT!ONs is TECHNOLOGY VOL.7 JULY 196

l l Smith and Statthiesen EAllTilQUAKE RESPONSE OF NUC Li..s'A llEACTORS ,f h I I) l.

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3 . None were observed. Ab_out 6 months _ :dter. the graphite columns and the top grid plate show vibration experiment routtgests_lndi.cated that peaks, one at ~3.9 cps and another at 4.2 cps. 'T one et themntroillade insertion ~ times had forced vibration tests reveal that the peak

             'increaseCA~fe_w 'mgnths later safety blade' Nofi                                                4.2 cps is the primary response of the but                                                                                       $

sEick'in the "out" position during a routine the 3.9-eps response is due to the core i ? prestart checkout of the reactoi control-system. When the shaking direction is switched to E%p l 7 hen the reactor was dismantled. we dis- the grid plate has a sharp peak at 4.6 cps, wy Ehp j covered that lead shielding bricks under the is equal to the natural frequency of the buildi:p *[* control blade drive shaft had been displaced the east-west direction. Q e l-upward. causine the shaf t to bind. The lead shield Figure 25 shows another interesting aspect blocks were stacked on lead shot which had been EGCR core response. The grid plate res, - d-F poured in the void spaces between the crachite and indicates the unstable jump phenomenon ass N biolocical shield. Subsequently the lead shot has ated with a nonlinear softening spring. As gl i been canned in steel containers, and a steel frequency of the forced vibrations increases. ,,, shroud has been welded in place to protect the amplit'.'de of accelerations increases unifor .

   }                                                                                                           8.87 cps. At 8.90 :ps, the amplitude nearly                                                                                      l drive shaft from interference.

f The response of the EGCR core (Figs. 22 bles, and then falls off at higher frequency. . through 24)is interesting. The acceleration curves the forced vibration frequency is lowered. I (e (north-south shaking) for both the center of the acceleration amplitude : retra . I Nt: CLEAR APPLICAT oNS & TECllNOLoGY VOL.7 JULTI M 24 -d Y:

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, Section 50.34

a. Not applicable

b. Final Safety A_nalysis R_eport, (FSAR) .

An FSAR, the Argonaut Safety Analysis Report (ASAR) is included with this appTication.- (1) Environmental monitoring results are discussed in the Technical Specifications, Appendix V, Section 7.0. (2) See the ASAR, Appendix III. (3) Se the Environmental ,I_mpact A_opraisal (EIA), Accendix II. (4) No structural weaknesses (earthquake vulnerability) have ever been identified. The bioloaical shield wn_a_ugmged in 1963 M-- L caccor power was increased to 100 kwt. Aluminum primary coolant lines, embedded in concrete beneath the ' reactor core and shield, were replaced (by-passed) by new lines in 1971 because of (external) corrosion problems. The originally planned PuSe start-up was replaced by a RaBe source prior to initial operation. The Rase source was replaced in 1976. Ventilation stack monitoring problems (type of monitor and calibration) were prevalent until 1975. The present monitor, a 4 liter, flow-through ion chamber, is believed to be quite satisfactory.

          .(5) Safety questions raised during the Construction Permit stage are unknown at UCLA today.

(6) (i) An organization chart is provided in the Technical Specificaticns, Figure V/6-1. Principal respcnsibilities are designated. " Demonstrated Ability" is the most comon personnel qualification at intermediate and higher administrative levels. (ii) Not applicable. (iii) Not applicable. (iv) Plans have been replaced by Tecnnical Soecifications, ; Appendix V, and (imolicitly) Procedures. l l (v) An Emergency Resconse Plan is included in tnis application, Aapendix IV. (vi) Technical Specifications are included in this application, Appendix V. (vii) Not applicable.

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                                                                                                             /2, ANNUAL REPORT NUCLEAR ENERGY LABORATORY January 1, 1976 to December 31, 1976 Ivan Catton, Director

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il subsidiary repairs. It is not possible to visually inspect the core other than by entry, and it is not practical to provide routine preven-tive maintenance within-the roreT The Fe$ruary 1971 earthquake gave rise to minor problems that N worsened with time and ultimately required a major maintenance effort in i , 1972. A coolant leak late in 1974 required reactor down-time from y j

               -August to earlv nac-W      The cmh wp ouwn rrom April 15 to July                  )

3, 1975, pending resolution of a Nuclear Regulatory Commission concern ov+r argon-41 emissions, and again from November 23, 1975 to March 18, 1976 to replace a leaking encapsulated neutron source. At the present l writing, May 1977, the reactor has continued to be operational since March 1976, and no major maintenance has been required since December of 1974. There are no current symptoms indicative of a significant main-tenance requirement.

      .           Technological changes influence reactor demand, and adaptability to change through finding new markets for reactor services continues to                !

influence reactor productivity. The reactor is no longer new, and reactor physics research projects with the UCLA reactor have become non-existent. The advent of the Medical Cyclotron on the UCLA campus has i displaced the reactor in the field of medical radioisotope production. But, new interests in activation analysis by geophysicists, geologists, and meteorologists have replaced these vanishing acti/ities. Of course, the reactor continues to be a valuable instructional teel both for the academic and the non-academic consnunity. The current reactor activities are discussed in Chapters IV and V of this report. Non-reactor activi-ties of tb3 laboratory are becoming a financially significant factor for the NEL snd are described in Chapter VI. Chapter IX describes new l 3 l

Q p {l C e 'A * - N 4.0 UNAVOIDABLE EFFECTS OF FACILITY CONSTRUCTION AND OPERATION The unavoidable effects of construction and operation involves the materials used in construction that cannot be recovered and the fissionable material used in the reactor. No adverse impact on the env,ironment is expected from either cf the unavoidable effects. 5.0 ALTERNATIVES TO CONSTRUCTION AND OPERATION OF THE FACILITY There are no suitable or more economical alterna.tives which can accomplish Is

    -        both the educational and the research objectives. of this facility.

These objectives include. the training of students in the operation of nuclear reactors, the proddction of radioisotopes, its use as a source of neutrons. for neutron activation analysis, and also its ut.e as a demonstration tool to familiarize the general publi_c with nuclear reactor operations. 6.0 LONG-TERM EFFECTS OF FACILITY CONSTRUCTION AND OPERATION s The long-term effects of a research facility such as t.se UCLA Nuclear Energy Laboratory are considered to be beneficial as a result of the contribu-tion to scientific knowledge and training. This is especially true in view of the relatively low capital costs involved and the minimal impact on the environment associated with a facility such as the UCLA Nuclear ; r 7.0 COSTS AND BENEFITS OF FACILITY AND ALTERNATIVES s. The cost for a facility such as the UCLA Nuclear Energy Laboratory is on the order of Si million with very little environ.9 ental impact. The benefits include, but are not limited to: , (a) education of students and public: , (b) research (activation analysis and production of short-lived isotopes); and (c) training. Some of these activities could be conducted using particle accelerators or radioactive sources, but these alternatives are at once more costly and less efficient. There is no reasonable alternative to a nuclear research reactor of the type presently used at the Nuclear Energy Laboratory for conducting the broad spectrum of activities indicated above. I

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UNITED STATCS

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gM[ 25 BN 5$$ r. w... 2 ff) D. Skovholt, Assistant Director for Operating Reactors, L 3 ENVIRORIENTAL CONSIDERATIONS RECARDING THE LICENSING OF RESEARCH REACTORS EP=:- AND CRITICAL FACILITIES .

Introduction

                                                                                                                                          .# '- Q....      p.g n                                   f5.

Thf8 df****sion deaz8 Vith re... and co m tio whic't' Me dga 'r .. . 'i citon eration involves the [.) ynavoidabfe Effects ofotFacility be recovered Const and uof theadverse const {- effects No ef fects. i.

                   'The unavoidable                        constructionreactor             that .cann   un.svoidable l

i materials used incaterial d from u:ed in theeither of these . f fissionable environ =ent is expecte ion and0$cration of fM h reactors, there ,are . tra Alternatives to objectives

                                                             .         Construct      associated Some of these with researccbjectivesof rdd in I'

To acconplish thealternatives.of reactors, product oconduct expericents. are no suitable operation ray beams to students in theof neutron and ga=na and use .- tion and Oneration of Facility Construc considered to beand) j are scidntific knobledge l Lone-Ter= Effects long-tcrn effects of research f aciliticscontrib'ution to a result of the TM ' urces invo1Ed and and beneficial as . amouni of' capital resoittle irreversible training. relatively low facilities. Beca use of the environment- very lassociated with such the small i= pact on theco=it=ent is irretrievable with very ility and Alternatives and Benefits of Fac . lions of dellarsare notanalyses, limited

                                                  .-                              order ofThe    several benefits    milinclude, but                         .

The on the costs areenviron=entaf ir. pact.of the follouing: conduct of activationf op c: ore could be to, some combinationradiogrcphy, little unich could training oactivities Le a Sc=c of thess cources alternative toi

            %.                        conduct of neutron                                     or radioactivereaconablespectru:a no                                              of a wis~

education of students. accelerators There is i using particle efficient. conducting this l

                                   ' costly and lessresearch reactor for 2 nuclear l

l

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                                                                                        /
                                                                                       /o Table III/1-2                                       l REACTOR ANNUAL USE Year Number of Runs Megawatt Hours 1973          76            13.8 1974          76            14.8 1975          91            11.9 1976          82            13.1 1977        106             15.9 1978        132             20.3 1979        149             29.0 Table III/1-3
                                                                                        ~~

REACTOR ACTIVITY Activity Hours per Year 1973 1974 1975 1976 1977 1978 1979 Research 145 177 146 158 188 244 411 Class - Instruction 46 28 39 27 88 60 (' _ s Maintenance 12 52 31 23 14 36 1 TOTAL 203 257 216 208 290 340 446 f 9A]'y (4 *

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f 7g C * * ; ' SECTION I b *' - A. REACTOR SITE De reactor is located in a newly-constnicted, permanent, reinforced concrete build-ing on the campus of the University of California at bs Angeles. He location of the building and its relationship to its surroundings is indicated in figure I-1. Le normal building populations during a school day are given in Figure I-2. At times other than school days, campus biailding populations are reduced to small fractions

           ;             of the figures shown.

l %e 400-acre campus is located on coastal plain approximately five miles east of the i Pacific Ocean and 15 miles west of the bs Angeles civic center. To the south of the campus is a business and shopping district, and to the north, west and east are residen-l tial areas. A map of this general area is given in Figure I-3. Geoloey De UCLA campus is situated on a coastal plain, and is approximately 400 feet above aca level. %e coastal plain consists of a terraced alluvial fill,. 200 feet deep at the reactor site, overlying sedimentary rock of rather recent origin. L e coastal plain lies I at the base of the Santa Monica hntains which are 2000 feet high.* Le most important I I formation in these ciountains is Santa Monica slate, an old sedimentary layer 2000 feet thick. Overlying this slate stratum are several more recent sedimentary layers. A cross section of the coastal plain near 'he campus is given in Figure I-4. E is section is at - right angles to t!ie anticlinal folding of the Santa Monica Mountains.

                       !!1JL_a.L.AI No deep wells have been drilled on the campus of UCLA or in the vicinity of the campus. 'Ihe water table is estimated to lie 200 feet below the surface in this area.

A log of a typical test well made by a foundation engineer near the site of the reactor l building is shown in Figure I-5. Surface runoff water is collected in concrete-lined storm drains which empty into the ocean. %is drainage system has been adequate to prevent any flooding of the campus by heavy winter rains. Le maxinum rainfall in any 24-her period during the last 75 years was ten inches, as indicated in Figure I-6. It is aarely conceivable that runoff I

    '                  from the watershed area north of the campus could flood 3cstwood blevard and the area                                                          -

s ., to the west of the reactor site. Ifowever, the reactor core lies about ten feet above

this level, and a rainfall equal to the largest ever recorded would not flood the re- ,%

ac tor. In the unlikely event that such flooding should occur, it *ould pose an extreme e' operational inconvenience, but would not create any radiation hazard. '

    ,                 sciantnev                                             -

j Southern California is seismically active. De locations of known active faults g are indicated in Figure I-7. %e nearest of these to the reactor site is the Inglewood j fault running in a north-westerly direction about two miles cast of the campus. In Southern Cali fornia, the region from the Mojave Desert to beyond the off-shore islands

is traversed by a series of active faults. Ecsc faults extend from 20 to 50 to many hundreds of miles in length, and the trend is generally Letween north and west. !!ow-ever, they are only roughly parallel, and in certain instances a major fault zone divided into two or more well delined faults. In general, these faults are from five to twenty miles ali art and apparently extend to depths of 15 or more miles below the surface. D

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s M ) 3.0 GENERAL ptANT DESCRIPTION O s 3.1 SITE LOCATION The reactor is located on the 400 acre campus of the University of California at Los Angeles. It is housed within the Nuclear Energy Laboratory (NEL) in a specifically designed and constructed reinforced concrete building. 3.2 SITE GEOLOGY The UCLA campus is situated on a coastal plain, and is approximately 400 feet above sea level. The coastal plain consists of a terrac'ed alluvial fill, 200 feet deep at the reactor site, overlying sedimen-tary rock of rather recent origin. The coastal plain lies at the base of the Santa Monica Mountains which are approximately 2000 feet high. The most important formation in these mountains is Santa Monica slate, an old sedimentary layer 2000 feet thick. Overlying this slate stratum are several more recent sedimentary layers. 3.3 SITE HYDROLOGY No deep wells have been drilled on the campus of UCLA or in the vicinity of the campus. The water table is estinated to lie 200 feet below the surface of this area. Surface runoff water is collected in concrete-lined storm drains which O . ematy sato the oceea. This ereiaese system hes eeen edeauete to pre-vent any flooding of the campus by heavy winter rains. The maximum , rainfall in any 24-hour period during the last 75 years was ten inches. It is barely conceivable that runoff from the watershed area north of the campus could flood Westwood Boulevard and the area to the west of the reactor site. However, the reactor core lies about ten feet above tnisilevel, and a rainfall equal to the largest ever recorded would not.' flood the reactor. 3.4 SITE SEISMOLOGY Southern California is seismically active. The nearest major fault to the reactor site is the Inglewood fault running in a north-westerly direction about two miles east of the campus. In Southern California, , the region from the Mohave Desert to beyond the off-shore islands is traversed by a series of active faults. These faults extend from 20 to 50 to many hundreds of miles in length, and the trend is gene-rally between north and west. However, they are only roughly paral-lei, and in certain instances a major fault zone is divided into two , or more well defined faul ts. In general, these faults are from five i to twenty miles apart and apparently extend to depths of 15 or more i miles below the surface. Earthquakes have occurred in California for a long time in the geo-logic past, and it is extremely probable that they will recur from i time-to-time in the future. In the southern coastal section, shocks e l III/3-1 i

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g II. WRONG CLASS OF LICM SE The Applicant has applied for the wrong class of license. Applicant has applied for a Class 104 license despite the fact that in the past, more than fifty percent of reactor funding and more than fifty percent of the hours of reactor usage have been devoted to the sale of services, rather than research or education. Given this history, and without any indication that Applicant intends to chan6e reactor usage, Applicant under 10 CFR 50.21(b) and 10 CFR 50.22, should have applied for a Clasc 103 license.

a. Applicant's financial statements indicate that more than half of the reactor funding comes from sources other than the UCLA School of Engineering and Applied Sciences (SEAS).

1. Applicant states on pa6e I/1-1 of the application that 52 5%

of the NEL income in 1979 came from " reactor earnings" and "other income", and that these two categories are projected to increase to 58% of the 1980 earnings.

2. Applicant also indicates on page I/1-1 that in 1979 and 1930 less than half of NEL's funding came from SEAS "in pursuit of the University's teaching and research mission."

b. Infor::iation provided by the application, though lacking sufficient dete.11 to support definitive assertions, indicates t&5 more than half of the reactor operating time is spent on commercial, non-educational projects.

1. Instruction accounts for only a small portion of the reactor t

l operating hours. An examination of the table provided on page III/1-5 of the application shows that in 1979 less than 8% of the

g reactor operating time was devoted to instruction, and that over the last seven years, instruction has averaged only 16.4%.

2. Research, the only other category of non-maintenance reactor operating time listed in the application, combines both commercial and non'-

commercial projects. Statements by the reactor staff indicate that the reactor hours for non-scholarly commercial use account for a majority of the 411 hours listed in the table on page III/1-5 of the app]ication as "research" hours. The application indicates that the " greatest number of researchhoursareforactivationanalysis(bothprayspecEscopyand delayedneutroncounting),andfissiontrackdatingprojects."(Application at page III/1-3), That the hours spent on activation analysis are likely to have been devoted to commercial projects, rather than bona fide research, can be inferred from the following staff statements: One business fim, paying $65/ hour to use the reactor, changes the color of gems such as rubi9s, garnets and topazes to make them more valuable. After the gens are removed from the reactor, they are allowed to lose their radioactive energy and are then shipped to customers.

                        -infomation attributed to Charles Ashbaugh of the reactor staff in UCIA Daily Bruin, May 31, 1979 Dr.Kalil,[whorrnsauraniumoreassa~ingbusiness uses the reactor to determine the economic value of the uranium ore samples through a technique he calls ' neutron activation'.
                        -same article, UCIA Ihily Bruin, May 31, 1979 Anyone can have a sample analyzed. Itcosts$75/ hour.
                        -same article quoting Mr. Ashbaugh, UCIA Ihily Bruin, May 31, 1979 Reactor manager Neil C. Ostrander said a substantial portion of a                 the 411 hours classified as research i- done for private firms.

Valley News, July 11, 1980, page 14 Trends in the amount of instructional time for which the reactor is used and the intended amount of time the reactor management desires

o , y to devote to non-university, commercial activities also indicates that the Class 104 license is inappropriate for the reactor. The 1976 Annual states: In order to attract more outside business and to eliminate our reactor user's from shopping elsewhere for a higher neutron flux, the reactor may be slightly altered to go to higher power levels, i.e. 500kw or 1Mw, the current licensed porer is only 100kw. The best possible use for this higher power level would be in activation analysis. If the money is found, our antiquated activation analysis laboratory must be modernized. It is currently about 10 years behind the state of the art.

                       -1976 Annual Report at page 35-36
       +r 4

l i l 1

g M. I:!AIITILT2 ,OAG22IAL .CD AHII:!IST?dTI72 CC::T.TLS . Applicant has failed to denonstrate adequate =anagerial and adninistrativo controls in the Application, as required by 10 CFR 50.34(6)(11), and, further, has de.onstrated throughout its operating histor/ grossly inadequate nanagerial and adninistrative controls. i l Those inadequacies nahe it inpossible to fhd that Applicant's nanagerial and m M nistrativo controls are adequate to responsibly l t

protect the public health and safety.

Specifically,

1. Anelicant has failed to crovide the infornation required in i

10 CF2 50.34(6)(11) retarding "nanaterial and ad-inistrative contro,1,s! ; l to be used to ascure safe onoration": this failuro na'tes a firdiM. adequacy innossiblo. On page 7 of Application, in responso to information required by 10 CFR 50 34(6)(11), Applicant has nerely responded "not applicable." Uo explanation is given, nor any hfor.ation provided. An e=nination of that section of the regulations p:xrrides no clue as to why Applicant vieus itself as exenpt from that section. Indeed, the histor/ of grossly inadoquate reactoi nanagenent and the gravity of the matter for reactor safety vould seen to require a substantial docunentation of adrf nistrative changes to rebut the presunption of inadequacy raised by the reactor's I record.

2. Zailure to ;ot nrior anorov_al fron 2ee.ctor Use Comittee or 2eactor Director for changes in reactor systens and for non-standard exeerinents.

Motice of '.!onconpH ar.ce, dated April 24, 1973:

2. r I Technical Specification VIII H.1 requires that the Radiation Use Co:r.ittee review proposed changes to the facility, when such changes have safety significance, and shall determine whethor they involve an acendnent to the license, a change

in the Technical Specifications or an unreviewed safety

. question. , l l { Contrary to this requirenent the Reactor Use Co=mittee did l j not review and nake the required findings with respect to

;                                the change of the reactor logic system from 110 V a.c. to l

28 Y d.c. ! i In a memo from AEC Reactor Inspector U.S. 7etter, dated August 21, i 1969, the following areaof noncompliance nas cited: ! Contrar/ to Condition 1 of the license, which incorporates i the UCLA Training Reactor Easards Analysis Report dated March 1960, the reactor was operated for the purpose of , i conducting a bean tube experiment during periods of non-

!                                standard reactor oporating conditions (byoassed scram c.i rcui_ty,z) . The periods of nonstandard reactor operation were not approved by the laborator/ Director nor had they j                                 been reviewed by the Reactor Hazards Co=mittee. The requirements for review and approval for nonstandard reactor

operation ars described on page 49 of the UCLA Training

    +                            Reactor Hazards Analysis inw/ Report.                                                                       ;

(emphasis added) Iote that the failure of ad=histrative and managerial oversight resulted in what could have been an extraordinar/ safety hasard-- the b/passing of scram circuitry. j 3. Failure to get trior aonroval from the Co:3;31 ssion for facility changes. 4

eno, Spencer to O'Reilly, covering inspection report 50-142/69-01,ne=o dated July 15,1%9, btates: -

                                 ...one of the items of nonconpliance involves a significant facility nodification (which was not authorized by the Comission as required by the facility license) and represents,

! for all intents and purposes, a repeat itee of noncompliance. That is, following the previous inspection, the licensee agreed, categorically, that no f acility changas of any significance would be carried out without prior approval ! by the Co==ission. ag- , , - - , - - - - - - , - - , - . + = . y -+ w , -- -- p--- >m. - g- -, yeti-W-- '-

3 l Senior Roactor Inspector Spencer conthues i As you know, we had previously been extending " tacit approval" for type 50.59 changes for this facility b the absence of technical specifications (and contrary to the requirements of the license) so that the licensee could continue a vigorous program requiring reasonable flexibility in the area of expeditious facility modifications. dowever, during the previous inspection we concluded that the licensee had overstenced " reasonable bounds" (see cover memo, CO Report Mo. 50-142/66-2) and consequently, we infomed both ::hs Reactor Director and the Assistant Reactor Director that any

                '    future modifications to the facility would require Commission approval. 3oth of the aforementioned individuals agreed, without reservation, to the now lbitations on the basis of the facility license requirenants. Obviously, the licensee has failed to adhore to the recuirenents of the license...

(emphasis added) Thus, there is considerable basis for the concern that the facility has I in the past been modified without prior Commission approval, the.'. the

 ,      problem has been a repeat pmblem, and that lack of administrative and nanagerial control (absence of Reactor Director) is in part the cause.

A review of the Application and recent inspection reports gives no j indication that these problems have been substantially resolved. The concem that the facility could be modified without prior Commission appmval raises many serious safety concerns, among them that excess reactivity limits, already dangerously high, could be altered by the licensee by changing fuel loading without prior approval. I 4. Lack of involven_e_nt of Lab Director and/or Assistant 9. rector for extensivo neriodsLother indications of inadecuate sucerricion. A history of accidents, radiation spills, worker exposures, inaccurate calibrations, violations of regulations and high number of unintentional scrams, among other indications of an inexcusably sloppy operation described in other contentions indicate grossly inadequate superiision. i l P 1

Y Commission inspecticn reports have froc tine to time come to the same conclusion, and at thes cited as cause the lack of involvement of the i Director and/or Assistant Director in the daily running of the facility. The Spencer memo cited above concluded by saying: Obviously the licensee has failed to adhere to the i requirements of the license and although the reasons appears l to be lack of involvement on the part of the Reactor Director,rather than deliberate disregard, we have concluded that anything short of C0 Headquarters enforcement action would be inadequate. le feel that the events which led to the items of noncompliance, I as well as the poor housekeeping observed during the current inspection, all reflect overly heavy workloads and a lack of 4 clear definition of responsibility insofar as the reactor supervisor is concerned. 4 In Inspection Report 69-01 we find the following In addition to the apparent noncompliance items listed above, it aceears that neither the Director or the Assistant Dire'c' tor of the facility has been actively participating , i in the day-by-day operation of the facility. For this l reason, and because one of the items of apparent nonco=pliance i (unauthorised f acility modifications) is essentially a repeat coeurrence, Headquarters action was requested.

p. 2 i

and continuing: } ... Professor Snith made a brief, unofficial visit to the reactor facility (not concerned with the inspection). At this time, the inspector iterated previous statements relative to the noncompliance aspects of the inspection. The acuarent degenera_ti_on of housekeeoing oractices was also discussed. Snith could offer little or no rebuttal, exclaining that he had been away from the facility for six nontj]s, and expressed i essential agreement with the inspector's co=nents. In restonse to a stated conclusion on the cart of the insoector

j' that Mr. Hornor annea: ed to be plely r'esconsible for 'the f,apqph a oneration without benefit __o_f designaEed (documented) _responsibilit, Professor Smith said that this situation was covered by the facility operating procedures and or license 4 correspondence. However, a search of the_ license records by both Smith and Hornor to verify Smith's statement was fruitless, j p.3 (emphasis added) i l

   . - -   ~,  .--v               ,,a  w,                                                                            .m--              .w...

y ., .pg-- 4--,--..,.-. . - .. ,m - _ , . ,.an, .,,rg--.---, , , , r,- - - - , -

5

5. Unlicensed visitors to the reactor facility have been invited to_

operate the reactor _contr_o1s,_in violatio_n of 10 CFR 502,_jg. and 1, and 10 CFR_M 2. 55.3a _& b, 55.4d_& f. and 55.9 a & b. Despite clear requirenents in the above cited regulations that only licensed operators are pernitted to operate the controls of a nuclear reactor, for obvious safety reasons, the Applicant has nonetheless invited unlicensed operators to operate the controls.

a. Attenots to have "Open House" visitors ocerate the controls,

_ including brbgine the_ reactor to scran coi_nts "in_, order to cublig demonstrate reactor _ safety." The bibliography for this reactor's docket is replete trith references to UCLA's desire to pemit "Open House" visitors to nanipulate the reactar." For example, a telephone call made on Hovember 10, 1965, by UCLA to the Comission requesting exenption of requirements of 10 CFR 50.541... so that visitors to their facility can operate their... as part of an open house denonstration on 11-14 (elipses in the original docket bibliography) A few nonths later UCLA sent a letter (dated 2-9-66) which the docket record describes as: subnitting a rough draft of a pamphlet they propose to distribute next fall when they hold open house, and citing precautionary steps they plan to take beforo allowing visitors to manipulate the reactor. Also St ating that_ visitors wg1 also be eergtted _to reach the, scran trio yoin_tL n_o_rier i to rublicly demonstrat_e reactor safe;t;r. (emphasis added) The Connissien responded in a 3-22-66 letter advising that ... DRL does not agree to an exemption from the regulatier.s to pernit operation of their reactor, in ref to their ltr of 2-9-66 which discusses "Open House." i This restonse from the Comission, declining to waive the regulations l 1 _ l which prohibit unlicensed personnel from operating the reactor controls, , i I I

tl l l l did not end the natter. UCLA, the docket bibliography indicates, sent a letter to the Connission on 6-13-66 requesting additional censideration to their request ! that visitors be allowed to particioate in the operation of their research reactor, in ref th our [AE_C7 3-22-66 ltr re their plans for an "Open Houre". And once again the Commission responded: we do not consider it advisable to waive the requirements of AT regulations 10 CFR Part 50 and 55 for this purpose.

b. Recent invitations to novie actors to ocerate the reactor.

In a length letter to Michael Douglas, who produced and starred in the film The China Svndrene, lIII., staff criticized the film and saids i If you wish further elaboration, which is a must due to the conplexity of the subject natter, or would like to talk or come by and visit our facility, and even oeerste our d Eth nuclear reactor, please give us a call. , EL letter to Hichael Douglas, April 9,1979 (emphasis added) Copies of this invitation were sent as well to the other principals in The China Syndrone--Jane Fonda, Jack Len, and James Bridges. The requirenent that only trained and licensed individuals j operate nuclear reactors is a very sensible one. It appears from the April 9,1979, letter to Michael Douglas, Jane Fonda, Jack Lemon, and James Bridges that the Comission's desire that the sections of 10 CFR 50 and 55 cited above be obeyed still has not sunk in with the reactor staff. With the a=ount of excess reactivity that is potentially available, and the lack of a containment structure should a radiation release occur, the operation by non-licensed personnel would be a serious threat to public health and safety and indicates ! very poor adninistrative and managerial contmls.

7l ! l l

6. Inadequate reco:d keeoing,_ including loss of key, records.

The Applicant has been consistently criticized by the Connission's Inspection and Enforcenent division for inadequate record keeping.

a. ].o,ss of naintenance log. Notice of Violation dated October 15, 1974:

Section VIII K.3 of the technical specifications requires that a reccrd be maintained of the principal maintenance activities and reasons therefore. Contrar/ to this requirement, the record of maintenance activities prior to May 1974 was nissing. (Severity Category III) As Inspection Report 74-01 stated about the loss of the record (p.4): The loss of this log was of particular concern since records such as instrument calibrations were not otherwise available, and two key laboratory personnel uith knowledge of previous maintenance activity had left. The discussion in the contention regarding inadequate instrument calibration and equipment naintenance indicates how disastrous the loss of this log was. In a November 4,1974, response to the Notice of Violation, 'darold V. 3:own, Environmental Health and Safety Officer for the University, replied: Section VIII k.3 of the Technical Specifications requires a maintenance log to be kept on the reactor and supporting equipnent. This record was lost in April,1974 before the > previous reactor supertisor, Mr. J. Brewer, left this facility. It appears that it will never reappear. Interestingly, inspector G.S. Spencer had said only a few years before (in a May 3,1971 neno te J.P. O'Reilly) that the problems in record keeping UCLA had had prior to that time were hoped to be resolved by the addition of 3 rower: The recent addition to the reactor staff, Mr. 3 rower, an ex-nuclear Havy =en, appears to be an individual oriented towards

9 oceratiens in accordance with written procedures and by the " book." This influence appears to have resulted in more detailed and organized documentatien to chew that operations have been perfomed in accordance with licensed requirements. Unfortunately, the addition of Brower did not end the record-keeping problems,

b. Other inadeauate record-keeoing. The above mentioned neco from Spencer continued:

You will note that our inspector discovered an error in the licensee's technical specifications governing the absolute reactivity worth of experiments. This error tras a nisplacement of a decimal point, an error of an o:tler of =agnitude that tras--this time at least--in the safe direction. Should such an error occur in the other direction, serious consequences could result, as is shown in the contentien on excess reactivity. The inspection reports available to Petitioner at this ti=e are filled with citations for inadequate record-keeping. One example is j an April 24, 1973 letter from iGC's Spencer to Thornburg

!                 The inspection disclosed three items of noncenpliance. ...

The instection disclosed a weakness in the keeping of i operating and maintenance records. The nost glaring deficiency was associated with the void coefficient experiment. . . . The licensee had no naintenance log record, however,this l type of information was found in various locations (i.e., operating log sheets, ninutes of the Reactor Use Committee and facility change order safety analyses).

7. Failure to hold ad-%istrative meetings and conduct reviews recuired by the Tech Snees. :!otice of lioncompliance, dated April 24, 1973:

1. Technical Specification 71II h requires that the Radiation Use Cordttee meet at least semiannual 17 Contra:/ to this requirenent, the ?.adiation Use Cordttee had not met since March 23, 1972,

q

3. Technical Specification VIII H.3 requires that the Radiation Use Co=mittee make an in-depth review of facility operations at least annually.

Contrary to this requirement only one review of facility operations has been made (in December 1972 or Januar/1973)

;                   since the Technical Specifications became effective on March 1,1971.

Similar problens are reported in other inspection reports. CONCLUSION The problems with inadequate administrative and managerial controls are long-standing and there is no indication they have been or will be resolved. The facility has a long history of questionable management. As Spencer wmte to Thornburg, meno, October 15, 1974, trans=itting inspection report of September 30, 1974: The itens of noncompliance appear to be oversights which indicate a need for nore disciclined nanagenent. This conclusion is reinforced by crevious excerience with this licensee. Consequently, we intend to broaden the inspection effort at this facility until inproved perfo: :ance is evident. (e=phasis added) And as R.H. Engleken wrote to UCLA the same' day as the S encer J memo was sent, in a cover letter transmitting another notice of violation: In addition to the need for corrective action regarding these specific violations, we_are concerned about Qee j=ulenentation of yot e manarement contml system _that zesulted in these violations. Consequently, in your reply, you should describe in particular, those actions taken or planned to improve the effectiveness of yotr nanagenent control system.

p. 2 (e=phasis added)

The history of gressly inadequate reactor nanagement, and the importance of responsible management to the operation of a safe reactor, venld seen to require a substantial demonstration of administrative i and managerial changes to rebut the presunption of inadequacy raised by the reactor's record. Applicant has failed to make such a demonstration l l

lo - I6 and therefore the present application cannot cupport issuance of the requested licenso . 1 l

4 9 ". O

                                                                                                                                    ",,')

UNIVERSITY OF CALIFORNIA, LOS ANGELES N

                                                                                                            ,l  SANTA BARBARA * $ 4NTA CB1/Z SERKELEY

Davis
InvTNE
LOS ANCELES
RIVER $fDE * $AN DIECO
SAN FRANCISCO D'

SCitOOL OF ENCINEER!NC AND APPLIED SCIENCE LOS ANCELE.5, CALIFORNIA 90024 Boelter Hall 2567 April 9, 1979 Mr. Michael Douglas Producer, I.P.C. Films

                                                                                                                         .i O

Post Office Box 900 Beverly Hills, California 90213

Dear Mr. Douglas:

Last week I went to see "The China Syndrome". I enjoyed it and I, therefore would like to inform you of some of the technical inaccuracies.

l. The Ventana power plant was rather unusual in that it appeared to be a cross between a Pressurized Water Reactor (PWR) and a Boiling Water Reactor (BWR).

a. Physical setting was that of a PWR.

b. The utility's (CGaE) PR description including the flow diagram on how the Ventana reactor worked, is descriptive of a PWR.

. c. The plant terminology or equipment description was that of a BWR.

d. The accidents were descriptive of a BWR.

2. Assuming that the power plant was a BWR, the plant must have, while Ventana did not, the following: n. ore than one feedwater line, more than one high pressure injection system, a couple of residual heat removal systems, a stand by liquid control system, an automatic depressurization system, and a suppression pool. A licensed commer-cial auclear power plant must have these and hence would prevent a meltdown from taking place, assuming though, that they all worked.

3. fluclear power plants cannot work with the entire high pressure injec-tion system out of service for repair or whatever. The built in electronic safety circuits would prevent the reactor from going to power.

4. All chart pen recorders are preceded by a digital, meter, or CRT readout, so a' stuck recorder pen really can only fool fools. Chart recorders are for record keeping or long term time analysis only.

5. The decay heat residual in a reactor core is very dependent on both the power level and the operational ~ time interval a : the power level .

It takes weeks at full power to build up enough decay heat to provide enough heat to ~ melt uncovered reactor fuel, just after a reactor shutdown.

6. One U02 fuel pellet has the equivalent amount of energy of about one ton of coal or 1/100 of a train car of coal, not six car loads. I

1 i 1 NUCLEAR ENERGY LABOR ATORY IVAN CATTON. Dinctor

O O April 9, 1979 Page Two t OV

7. MiiB engineer's quote of,"You are lucky to be alive and that is the same for Southern California," and the " molten core hits ground water and causes a blast of radioactive steam ..." is very exaggerated and blown clear out of proportion. It probably could be termed literary license.

8. The radiation instrument used to find the water leak under the pump was I

a rad gun which is a current reading device and therefore does not have a speaker which goes click, click, click. You should have used a G-M survey meter for that operation.

9. All nuclear ~ power plants have one area called a two man entry zone, outside the control room, where the turbine can be made to trip off line and cause the reactor to scram; so, you do not need a crew of electricians to cross a million wires to eventually cause a reactor j

scram..

10. Not every power plant has been given the o.k. by the AEC or later the NRC to be licensed. There were many which have and will be turned down.

r Just look at the history of our own Los Angeles Department of Water and

 '             Power; i.e. Malibu, Bolsa Chica, and San Joaquin Nuclear Projects.

These are examples of just one utility's fight to obtain a nuclear power plant. Due to the above, I would seriously consider laying off your three

          " engineers"???? from the MiiB Technical Associates. All in all, though, a.

very good accurate action packed factually correct nuclear thriller could and maybe should be made. The acting and the emotional ties to the audiences of "The China Syndrom" made it, but the errors and some of the plot did not quite make it. In a way though, maybe that was good. For example, while we were leaving the theater, we were jumped by a group called the " Alliance for Survival" who were taking advantage of the subject matter of the film for political purposes and therefore were there handing out poorly documented anti-nuclear literature. This was when I became upset. Why can't a good nuclear thriller be made without having all of the hidden " heavy man" stuff be generated af terwards. I know you said you tried in your (especially Fonda's

  !       remarks) during publicity campaigns for the movie. I hope that the audiences i       do not turn against nuclear power but just enjoy the movie, or we may really be in trouble some day. Look at OPEC price in:reases, gasoline prices, our

> poor trying to make it into the middle class, and of course, then use more energy. With a few years of increasing demand far energy, and a possible curtailed nuclear power industry, we then may have to rush back into nuclear and increase the possibility of a china syndrome, or go to war for oil. Whatever you do, you do have more sway with the general public than you probably should have. But that's my own value judgment. You do have the sway, and I hope you do use it, and not abuse it, for the good of your audiences. _ One last comment; I'd like to redefine the tera. China Syndrome as "A psychological euphoric malady common among anti-nukes, characterized by i a paranoid fear of a reactor core melting through the center of the earth and popping out in China, where it presumably irradiates the natives on

        . both sides". The China Syndrome so to speak has really been blown out of proportion. If a mel tdown would occur, the most likely effects would be

o . M e f ', ; % \ g I j April 9,1979 Page Three i

                                                                                      !Q                                     :

l a large loss of money and equipment to the utility involved, a lot of workers would catch a few extra rays cleaning up the mess, but I seriously doubt that the general public would feel any real effects. We have had several partial meltdowns in the past, i.e. EBRI, Fermi, and Three Mile 1 Island. Only the last one involved a release of radioactive gas and this occurred either due to operator error or to premature transfer of radio-active water out of the containment building. A lot was learned and then implemented after the Browns Ferry fire and the same will occur after the Three Mile Island incident is fully investigated. So, I'd say, keep the faith and remember that risk versus benefit, and apply this to all other-forms of generating electrical energy. .I have, and that's why I'm j a pro-nuke. h Well, thank you for taking the time to read this letter, if indeed it got that far. If you wish further elaboration, which is a must due to the complexity of the subject matter, or would like to talk or come by and visit our facility, and even operate our .1 MWth nuclear reactor, please give us a call. Sincerely yours, CAA s alL-/ Charles E. Ashbaugh, M.S., P.E. ] Associate Development Engineer / Lecturer UCLA Nuclear Energy Laboratory 825-2040, 825-2825 l l CEA/li

  !         cc: Jane Fonda            '

Jack Lemmon - James Bridges

                                                          -,,           -,w .-  ,      , , ,   , . , . , - - , , _ . - - , ,

( . . l IV. VIOMTIONS OF NRC REDULATIONS Applicant has been consistently been cited for violations of NRC i

regulations as well as violations of the provisions of its own Technical
Specifications. This consisten pattern of regulatory non-compliance and

! the lack of assurances that the pattern will not continue in the future, a indicates that the Applicant cannot adequately demonstrate that future operation of the facility will comply satisfactorily with the regulations to protect the public health end safety. , From the inspection reports available to 1%titioner prior in the establishment of a pubite reading room for this docket and the granting of discovery rights a persistent pattern of violation of regulations, averaging roughly one per inspection, is evident. Ebme of thesF-Citsdr0 ri'61ations are:

1. Operation on two occaisions with the secondary coolant fission productmonitorscramcircuitrybypassed.8/69

2. Operation on two occaisions with core excess reactivit greater than permitted during certain kinds of experiments. 69 3 Conducting a beam tube experiment during periods of non-standari operating conditions (by-passed scram circuitry); failurc to get prior approval from Iab Director and Reactor Fazards Committee. 8/69 4 Failure of the Radiation Use Committee to make an in-depth review of facility operations at least annually. 4/73

5. Failure of Use Committee to meet at least semi-annually. 4/73

6. hilure of Use Committee to give prior review to proposed facility changes. 4/73

7. Record of maintenance activities prior to May 1974 missir4 11/74

8. Contrary to technical specifications, acceleration nozzle had been removed from the reactor exhaust stack. 11/74 l

  .       _      .                          .   .                      _ _ .    .      .                 -                 _ =

i 9 Failure to calibrate the reactor mon area radiation monitors and i the radioactive gaseous effluent monitor at the required frequeeny. 1/75

10. Ventilation exhaust air from the reactor room not being diluted to 14,000 C m and not being released at 125 feet above ground level as required by the Technical Specifications. 1/75

11. Changes to the operating proceduros not approved in writing l by the Reactor Supervisor.

The history of violations of NRC regulatiens and Applicant's Technical i j Specifications set forth above are basee. on the limited number of inspec-l tion reports now available to Petitioner. However, the pattern of persistent J ] violations is clear. Given this history of regulatory violations, i Applicant must demonstrate changes in its operation that will reduce such violations in the future. Applicant has failed to make such a demonstra-l tion and therefore has not provided reasonable assurances that the operation of the reactor will be conducted in compliance with the re6ulations. A l i finding of such assurances is required by 10 CFR 50.57(3)(11) before i an operatin6 license may be issued. l l l l i J

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

l VTeo :mca messs ancTrgg The 2 cur.t of excess reactivity which is perzitted by the Technical Specifications to be installed in this reactor is too gret in that it is potenticily sufficient to cause a serious power , excursion uhich eculd bring about nelting of the fuel cladding and significant release of fission products, seriously endangering the public health and safety. Specifically,

1. The enount of excess reactivity pernitted in a facility such as this uhich is used at tines for the instruction and training of students nust be quite linited in order to leave a large margin of safety.

The 19c0 Hasa:ds Analysis for this reactor nade quite clear the I requirenents that a training facility such as this one nust neet in order i to be a safe facility at which training could take place: A reactor which is to be used for student instruction must be designed so that safety is insured without exercising greater restraint on the activities of students than is no@* advisable in a university laborator. This necessitates: (1) that the total available excess reactivity be limited to something less than that needed for prompt criticality; (2) that the reactor have a high degree of demonstrated inherent safety; and 3) that it be limited to low-power operation, p.19, Hazards Analysis Petitioner vill outline in what follows considerable basis for the concern that the three factors detemined by the writer of the 2 aforo-nentioned Hazards Analysis as essential for the safe operation of this facility have each in the subsequent years been considerably nitigated.

2. The reactor has lost several significant self-liniting features

7 since it uas originally licensed.

a. The excess reactivity has been substantially increased over the level necessa17 for prompt criticality. (see iten 3 below).

b. The deflector plate to prevent repeated excursions appear to f

have been removed. (see contention on lack of adequate safety features).

c. The assumption that the reactor has a large negative temperature coefficient appears to be erroneous because of the positive graphite temperature coefficient. (see contentien on safoty features),

i d. the power of the reactor has increased fmn 10 kw to 100 kw. The absenca or nitigation of each of these self-liniting features reduces the nargin of safety that was odginally assumed to exist at this facility.

3. The excess reactivity pemitted in the reactor is no longer less than that needed for prompt criticality.

The reactor was originally designed to have a 14 itation on excess reactivity of .6$ 0 k/k. "'his limitation has been changed to 2.7jhk/k(Applicationp.III/6-5). Pronpt criticality for this reactor is somewhere between .65:j6k/k and .7184k/k depending upon which of the four figures for Ogiven by the Applicant variously in the Application and original Hazards Analysis is the correct one (sr.e discussion for point 6 below). Thus it is clear that the current li-4 tation puts the level of excess reactivity far beyond that a=ount neceseary for prompt criticality. In fact, those parts of the original Hasards Analysis that have been put into the SAR in the current Application no longer nention as a safety feature the restriction on excess reactivity as being inport. ant to below that needed for prompt criticality. (compare

p. III/5-1 of the application with p.19 of the Hasards Analysis).

As the Hazards Analysis original.ly said, it is possible to operate the reactor with an amount of excess reactivity uhich is well below that required for pronpt criticality,i Under these conditions, the reactor neets the safety requiraients of a training reactor and can tolerate considerable operational error without danage, p.19, Haza:ds Analysis The above-mentioned safety feature no longer exists, neaning one less feature to help the reactor " tolerate considerable operational error." As can be seen in the contentions on inadequate management controls and on the facility's histor/ of accidents and unintential scrans, a good deal of operational error must be aniicipated. The loss of any i protec -ion to help tolerate such error is worrisone. 4 4. The licensed anount of excess reactivity (2.33 A k/k) currently cernitted is that amount uhich the 1960 Hazards Analysis found could cause melti_n.!; of the fuel cladding. The sections of the 1960 Hazards Analysis reproduced in the SAR indicate that at .(dD k/k,the level of excess reactivity to which the reactor was initially limited, it was unlikely any substantial damage could be caused if all that excess reactivity was o inserted at one time, while at 2.$ A h/k one would have a power excursion 1 of 28.4 ini-see uith a corresponding period of 9.1 milliseconds, if the extrapolations from the Sorax I tests were correct. That period and energy release are, according to the Hazards Analysis reprinted in the SAR, precisely that anount which is estinated necessarf to raise the fuel tenperature to the nelting point of the aluminun cladding. (Application p. III/A-3,5). The language in the analysis is somewhat 4 contradictorf, in that they first say that an energy release of 41 LLi see or a period of 6.7 r4114 seconds for the 3orax--and 28.4 EJ-see and 9.1 milliseconds for the UCLA reactor--would be enough to raise the maxinun tenperature of the fuel plate to the nelting point of aluminun (said in

             -,    s           ,   -        -              ,,    n           -  -,--  , ,

the Hazards Analysis to be 1000 ? over the boiling point of water). Later they imply that both reactors could survive reactivity insertion ug tp, those levels. It seems clear that the Hazards Analysis, which was written to demonstrate that the .65 limitation then in effect was a prudent one, merely was attempting to demonstrate that there was a significant safety nargin between the .6% lirAtation and the 2.$ 0 k/k range uhere nelting could occur. That safety margin no longer exists.

5. The void coefficient _for the reactor nas ch_anced since the iniXal c_alculations were done, cutting the reactor ae uresently licensed over the level necessagr for a nouer eccursion _tyg, could _ result J.n clad _djglg gelti_ne_and fission uroduct release.

The void coefficient used in the analysis included in the Application (reproduced fron the 1960 Hazards Analysis) en pages III/A-1 through A-7 is given therein as .18 ,$k/jcoolant void; while the current void coefficient, listed on page III/6-5, is =.164k/) coolant void. Replacing the old void coefficient with the new one in the calculations ! on page III/A-4--which one nust do if one is analysing the reactor as it l now exists rather than as it was in 1960 :inen the analysis was first done-- one finds that the ratio of coolant voids of the Sorax reactor to the UC reactor has also changed, altering the entire set of calculations. (One assumes that the tem UF is a typographical error derived from attenpting to utilice calculations from the University of Florida Argonaut reactor's Hazards Analysis-hopefully, as we shall mention later in this contention, there was no trcnsposition of data as well). The calculations thus would have to be altered tire- ir. oifer to - bring then into conformity with the 1?co void coefficients l l 1 i

g 126), 1989 CSorax 0.24 C3 erax ,, 0.24 ,1,g4

                            .33                  C UCLA     0.164 C

UCLA. 0.lo, =

     *(see above note about UF error) hl :r.; sec
                    = 31lG see                         El IM sec     =  28 :G sec 1 33                                                1.46 31 :".i sec x .82 x 1.12 = 28.4151 see          28 !?1 see x .82 x 1.12 = 25.7 :G see 28.4 :N seconds corresponds to                 25.7:LI seconds corresponds a period of 9.1 nillisec &                     to a period greater than 10 2 32 6 k/k excess reactivity                   nilliseconds and 2.%2 4k/k to bring cladding to the nelting pt.           reactivity to bring to nelting pt.

(the conversion fron :U seconds to /26 k/k and exponential period is made

11. both cases fron graphs D-5 and D-7 in the old Hazards Analysis).

Thus, if the equations presented in the Hazards Analysis are correct--and the Applicant relies heavily upon then--the 2 3'2d k/k excess reactivity limitation which, with 1960 characteristics, would have brought then just to the melting point of the cladding, woul_d_, _t_oday be _ considerably over that_ nelting no_inA 2.1;s would be the corresponding 1 biting figure today, because of the changed void coefficient. Thus, being licensed for 2.3 5A.k/k puts then considerably over a level that was questionably safe to begin with and creates the clear possibility of a catastrophic reactivity insertion causing cladding failure and a serious release of fission products.

6. W convMinz 2 T's A k/k e;; cess mantivity NWt4em in +'4a current tech stees into _Q3.54 in the,procosed tech specs _ included in the Application--and using a different h than the one used in the "a ards Analysis to make that conversion tit is possible that the Arelicant, has, uithout so stating, shifted its tech see.: limit fron _2,35 to 2,_6_2'2 d k/k, taking it further into the range, indicated by its own_ Hazards Anahc_in capable of causing a serious power;,sxegsion and cladding nelting.

o a g The Applicant has, in its Application, given three different figures, ! vithout clear explanation, for /3, uhile the Hazards Analysis upon which all the calculations are based gives a fourth figure. In converting fron 2.s A w/k in the present technical specifications to the $3 54 limitation cited in the ones in the Application, applicant has used a 6 of .0065. If any of the other three @'s they cite are actually the correct one, this potential ermr could mean a non-Sproved increase in excess reactivity from the 2.$ limit up to 2.62,6 0 k/k if the Hazards Analysis h of .0074 were used. (The .0074 figure is given on pages D-12,13,14 and 15 of the Hazards Analysis; the two other h's cited are .0068 and +.0070 on pages III/6-5 and 6-4 of the Application respectively: the h of .0065 apparently used by the Applicant for converting 9Ok/kintodollarsconesfromApplicationp.III/6-4and5.) Since Ge calculations on safe levels of excress reactivity are based entirely on the 1960 Hasa:ds Analysis, use of otherh figures uithout explanation seens questionable. (For that natter, having four different figures for the sane supposed constant floating around without explanation is very

uorrisone.) If the 1960 Hazards Analysis 0 is not correct, then it may

!           have throtm off the calculations upon which UCLA relies, because the graphs used to determine exponential period versus excess reactivity inserted use .0074 as [7           Either way, explanation is in order and a potentially

] serious increase in excess reactivity nay have been included.

7. lhe central assunction _of the Hazards _ Analysis, that the Borax I

                       .ts can be extracolated to the UCLA reactor operationu acpear to be seriously _ incorrect.
    .y         . - - -
                                                                                , - , , ,       r. n    ., v-_ -

7 Bor.iax uns unter cooled and moderated. UCLA reactor is unter-cooled and.rartially water-moderated, but also uses graphite as a noderator and reflector. i.later has a negative temperature coefficient but graphite appears to have a positive tenperature coefficient, throwing off nuch of the extrapolation of the Sorax I data to the UCLA situation. The Hazards Analysis "es_t_dnation of effects of assumed large reactivity additions" begins with the following statements i It has been demonstrated repeated in the 3orax and SPERT reactors that water-cooled, uater-coderated reactors of suitable design may have a ver/ substantial self-protection against the effects of reactivity accidents, even in the absence of corrective action by the roactor control system. This self-protection is providad by the negative steam-void coefficient of reactivity and the negative tecuerature coefficient of inactivity, both of which can result in important reactivity reductions as the reactor power

ises. The UCT2 has been designed with a high degree of self-protection of this type. In this tppendix estinates 1 are :nde of the behavior of the reactor under various hypothetical conditions of excess reactivity addition with no corrective action by the control system, i

Appli:ationp.III/A-1,emphasisadded Although there is indeed evidence, as indicated above, that water-cooled, uater-coderated reactors have substantial self-protection nochanism, prinarily negative reactivity coefficients, the UCLA reactor is not the sane as the Sorax and S?E2T reactors in that they are exclusively water-

            .oderated and the UCLA reactor is moderated by both water and graphite.

8. P_ositive tencent_ure coefficient of grachite. Eight years after the Hazards Analysis was uric. ten, UCLA was notified by the /J:I that the University of 'iashington Argonaut reactor, shnar in design to UCLA's, had discovered a cositive temperature coefficient for graphite of I

l l

g

     +0.01$ / F.      The 100 inspector at the time of infoming them of the University of '.!ashington finding inquired of U':LA if they had noted the sane thing. UCLA infomed the inspector that they had tried an experiment to test it, but the experinent had failed. The inspector made a rough calculation on the spot based on log entries and confimed a positive graphite temperature coefficient-By reference to the console logbook data concerned with core reactivity changes as a function of time and the tenperature of the uater noderator, it acoears that a cositive tencerature
                                                                          ~

graphite temperriture of 0.00$ delta k/k/0F exists.

p. 6, inspection report 50-142/68-1 UCLA connitted itself to experi:wntally detemining the graphite temperature coefficient "as soon as pronising test equipnent could be developed."

But not long aftor that Jr. Smith, the lab director who had nade the connitment to the ICC, vent on leave, and not long thereafter a new Director was apointed. There is no evidence in any of the documents presently aved .lable to the Petitioner that would indicate whether that experimental detemination uas ever nade. A positive temperature coefficient for the graphite--particularly one that is larger than the negative coefficient fo, the water (+.006% conpared to .00M/l bk/k/ F)-raises extremely serious questions about whether the principal self-liniting design safety feature, that of a negativo tenperature coefficient, exists forthis reactor. It also nake? .::y questionable uhether any of the data from 3orax can be used because it uns an entirely water-moderated reactor with a clearly negativo tenperature coefficients. Questions also are raised about the conparability of the s oid coefficients when the graphite is taken into account. Even the Applicant recognized the analytical probleas l

o . q introduced by the presence of a graphite nederator, although it didn't address the p:ublems directly, apparently thinking that the graphite like the water had a negative coefficient and could thus be ignored. But the uncertainties are clear: The relative importance of the two noderators, graphite and water, in detemining the effective neutron tenperature introduce uncertainties in the theoretical connutation of this

                   ~ coefficient.

Application,III/A-2,emphasisadded

9. Aeolicant's analysis relies on twenty-five-year-old data and fails

     .to_ include any__naw data, ad,d_ing nore uncerta_inties.

A najor part of the Applicant's SAR is a copy of a section of the Hazards Analysis written twenty years ago for the initial reactor license. The Hazards Analysis lists as its only references twenty-five year old reports done on the Sorax I experinents in the early fifties. (The Borax I was a reactor designed to test the effects of rapid reactivity insertion; it self-destructed in the late su=ner of 1954 with a 135 :rd-soc release of energy and a steam explosion accompan4.ed by release of fission products. For details, see Thonpson and 3eckerley, The Technolo.w ofIuclear Reactor Safe +gg, p. 622). The SA?. and current Technical Specifications included in the Application all nention the SP22T tests as providing evidence of the self-liniting features of the UCLA reactor. However no data fron the SPZP.T tests are cited or utilized in the .11R, presunably because the bulk of the tests were conpleted after the Hazards Analysis upon which the SAR relies so heavily : ras conpleted.

     *For confirnation of a positive coefficient for graphite, see Thonpson and Jeckerley's description of the pouer excursion and related partial neltdown of the graphite-moderated, sodium-cooled 32Z (p. 643); euctectic nelting--

unalyzed in Applicant's SAR--was a significant factor for the SE nelt; see photo of rods attached to back of this contention.

                                                                                                                                         /o i

I I A great deal of knouledge has been gained in the last twenty-five years concerning the behavior of reactors and fuel during power i excursions and cladding failures. The SL-1, S2E, S;iAP-8, SPERT, S8Ed, - t 0133, and T.CGA reactors have all valuable experience to aid in analysis. (For example, the T.EGA reactor, another type of training reactor, has been shotm to be capable of a cladding failure trith a $2.00 i insertion of excess reactivity). But the Applicant has considered no new data and continues to rely exclusively on the 1953 and 1954 3orax I tests. Attached to the end of this contention are two graphs, one from the Applicant's Hasards Analysis, relying on twenty-five-year-old data, and another one, fron Thompson and 3eckerley, relying on data that is only fifteen to twenty years old. One can easily see that the feu extra years of experience with reactor safety has pmvided nuch more data; certainly a conprehensive review of even nore current data is called: for in re. king I en analysis of this gravity.

!                     37 failing to update their Hasards Analysis, even to the point of shply xeroxing the section on excess reactivity and including it 4

untouched in their current Application, Applicant has fe.iled to nake 1 even a ninimal effort to demonstrate that the reactor trill not endanger the public health and safety. I 10. Aeplicant's Easa:ds Analysis regarding excess reactivity is based on nunerous unverified assunctions,n_any of_ynich are not identified t by Apylicant as such, and cannot be _u_ sed as the basi _s_for anything nore than estinating a range of excess reactivity _a_ddi_tions at (n_ich hasard f

              -j.ght o,xist._lurther.gadoolicant _does not give _the error bars for its conoutations and analyses, errog bars _that must be assuned to be very large.

L i

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                                                                                                                                         - -,-e- -
       ,w. .         . . . - , . - ,      - - - ---. ,         . - . . , , , - . ~ , . , -. .

11 a,The entire section of the Hazards Analysis reproduced in the SAR is filled trith terns describing the calculations clearly as estimates and extrapolations: On the assunot'.on that this nininun value is the true v_a_lue, a rise of water tennerature from near 0 C to SOC would reduce - reactivity by 0.6) ice rf. III/A-2emphasisadded The characteristics ,3f the UCTh which determine its behavior during power transients resulting fron large reactivity additions are quite similar _to, but not identical with, those of the Borax 1 reactor. III/A-1emphasisadded ggery.nents of the Sorax and SPET types have n_ot been nade uith reactors havinE uidely different neutron lifetines. The generalfriden_e_e of the experbents, however, sunoor_ts

                -the suunosition that ...
                               ~

III/A-3 emphasis added In comparing the behavior of different fuel plates, it must be recognized that the total energy release of the power excursion can no longer be considered _ as_a_definitiv_e_ variable... III/A-4 emphasis added l The text is replete with phrases about esti=ation, assunption, uncertainties, suppositions, and so on. The Hazards Analysis was .esi;;ned to merely I roughly estimate the upper 1%4t of the safoty margin--not to say i exactly where danger will lie. That it cannot do. 4,The fig 2re of 2 34Ol:/k being the louer bounds of the danger cone was extrapolated fron Bora I data and entailed a considerable number of assunptions, the accuracy of which are open to significant question. The Hazards Analysis itself nakes clear that the figure 2 3. is nerely a rough estination in order to deternine how large a safety nargin the reactor would have at the then-licensed lhit of .6,4k/k. This is I nade clear when in the Hazards Analysis, af%r analysing what an insertion ' of the then-lhit of .6% would do, it goes on to say 1

i 12 1 It is useful to aweto the value of excess reactivity which, Af suddenly inserted and not renoved by the control systen, uould raise the maxinun temperature in the hottest fuel plate

;                                     to the meltine point...

The first step in the procedure is the estination of the exponential period corresponding to the excess reactivity uhich would have characterized a power excursion of sinilar i effect in Jorax I. i III/A-3 emphasis added . The Hazards Analysis cannot now be used to declare that 2.?$ 6 k/k is safe. It merely shous that .6% 6 k/k is safe and has a reasonable safety nargin to compensate far the potential errors in extrapolating from the l 3orax data. ' l c. The SAR uses as a " constant" an interpolation off a line ' 4 drawn inposed on 21 data points from the Borax I exceriment. (see ! chart marked Figure 3-5 at the end of this contention). The line itself is a rough approxination on a log-log chart with the line only capturing 5-7 points plotted, with no data points located anfwhere i in the range of the point of interpolation. There is obviously a significant probability of error involved here, particularly on a log-log chart like this, but no error bars are given, in the absence of which, i a significant extra safety nargin must be made. The SAR nakes a similar assumption of a " constant" on p. III/A-3 by assuming that the maxt:un fuel-plate temperature rise is, _to_ within

eenerimentg error, proportional to the maximum energy release ,

of the pouer excursgon. The proportionality was determined to be constant 24.4 F per .74-sec." (enphasis added).

                      'lithout defining the bounds of that possible experinental error, that "c.pnstant" is used to determine at what excess reactivity the UCLA reactor
       . .      . . -                     . . .    .       .        --    --      . - - -     - - ~ . - -

                                                                                                                 \3 4

would suffer a cladding nelt, and then to justify licensing it at exactly th2t level. Obviously, such interpolation off rough data points on a log-log chart, and such attempts at creating a " constant" from a few data points in a 25-year-old experinent, are subject to a high degree of probably error without even taking into account the other assumptions involved in applying the data to the UCLA reactor, yet the Applicant does not acknouledge these error factors or give statistical error bars. At i least Applicant should have built in very large nargins of safety,

d. There are indications that the Hasard Analysis sections included in the SAR dealing with excess reactivity additions uere prepared by using Hasards Analyses for other Argonaut facilities. At one point, where the initials UCT3 (University of California Training 2eactor) should have been given in an equation, the initials UF are inserted. (p.IH/A4).

And the confasion about the differenth may be in part due to the use of graphs in UCLA's Hazards Analysis from another facility where there was that h, The UF mentioned above can likely mean the University of Flordia, where another Argonaut-type reactor is located; it is thus at least possible that the author of this Hazards Analysis was working from the University of Flordia analysis, which is cited elsewhere, and following computation done for the University of Florida reactor, transposes their initials into the equation. The question is thus raised whether any figures or non-transferrable graphs were also transposed. (This question is further underlined by the fact that the charts cited in the Ha have different page d's than the citations would indicate.)

- ~

N

11. I!ew nechanism for ranid reactivity insertion. Since the analysis included in the Application was written, 20 years ago, a significant new nechanism for the rapid insertion of excess reactivity has been added to the facility-a pneunatic " rabbit' system. This systen provides a new nechanism for the rapid insertion of a sa=ple of positive north as well as for the rapid rmoval of a sanple of negative worth, either of which could initiate or contribute to a serious excursion. In addition, core characteristics key to the calculations included in the Ea:ards Analysis nay have been altered through the core nodifications necessary for the installation of the rabbit systen.

A portion of the ~5asards Analysis not included in the current Application recognizes sanple renoval as a potential cause of a power excursion: One procedure to achieve nxcbun excess reactivity in the reactor would be to insert into the reactor a sanple with sufficient absorption to prevent start-up. ;ihen the controls were fully withdrawn and criticality was not achieved, the naxinun reactivity would be added if the canple were removed without reinserting the control blades.

p. 60

12. The licensed li.it _en conbined exceriments--03.54,_ negative or positive--creates an unacceptable nethod Ng which such a catastrochic accident could occur. Excerinents worth +43.54, inproperly inserted, could I

double an already excessive excess reactivity installed in the reactor. Rapid insertion of a negative worth experhent could also have a 4 catastrophic result.

13. The Ha:aMs Analysis indicates that the renoval of one of the bean __ tubes could insert excess reactivity into the reactor by renov_i_n._-s neutron _absorntion and increasinue_flector savings. ; lith the reactor i

apparently being right at or over the point at which nelting could occur

                                                                                                           /.5 with the reactivity presently permitted, and with the uncertain accuracy of the calculations by which that point was detor=ined, beam tube renovals--

singly or severally--could contribute to an unacceptable excursion.

14. The reactor oower has been increased since the ori4n_al Hazad palysis was writteng there is no_ adequ_ ate revieu contcined _in the analvsis as to how tile increased cower nay alter the excestr_e_ activity cal _c_ulations frcn 20 - ears _ago_and one-tenth _the couer.

15 . On at least two occasions Applicant has violated excess reactivity linits, suggesting th_at even if the licens_ed linit were safe, it would not yrevent_;ossible excursions at this facili_t,z.

!otice of Violation, dated August 15, 1969:

Contrary to Amendnent 2 of the license, which incorporates the application dated March 21, 1961, the reactor was operated with a core excess reactivity greater than 0.6% delta i' /k during the' conduct of experiments designed to measure the effect of water level on reactivity. We uncerstand that on February 5,1969, and March h,1969, the core excess reactivity was approximately 1.5% delta k/k. In addition, the Applicatin lists the 1960 limit as 0.6% (at 32 F) and the amount ^ installed then as 1.5% (at room temperature); even taking into account the negative temperature coefficient of water--ana not ecmpensating for. the apparent positive coefficient of graphite--that 1.5% amcunt installed would appear to be in excess of the then-licensed limit. (see III/6-5). It certainly at least merits some explaining in the Application. Furthermore, the current technical specifications and the proposed enes in'the application make no reference to the tenperature at which the reactor is currently limited to 2.3% delta k/k. And;to make matters worse, inspector A.D. Johnsen during inspection No. 50-lh2/71-1 found an error in the technical specifications concerning

o

/(

the excess reactf.vity limit. The limit was off by a decimal point-- an order of magnitude. Fortunately, the error was on the safe side, but next time fortune may not be so kind. And inexplicably, once corrected, the error reappeared in the Technical Specifications. An error of a decimal point in the other direction and the failure of someone to catch it could contribute to a disaster.

16. The Borax data itself needs error bars. Thompson and Beckerly p. 623 indicate that "the somewhat unexpected destructive-ness of the test resulted in losing some information which might have been gained. They continue: " Instrumentation and other means for obtaining information from transient tests should be planned on the basis of overestimates of the possible destructiveness of the tests if maximum information is to be gained." They report the period as .0026 seconds, describing it as "the minimum period measured," so the few data points that are charted for Borax in the Hazards Analysis may not be as firm as one would like. It is clear from the H.A. chart that there are no data points in the region of interest, that all that is done is extrapolation.

17. Failure to analyze potential for euctectic melting. Other serious reactor accidents (e.g. SRE) have had as partial causation a melting caused by creation of a fuel-cladding alloy at a cercain elevated temperature, that new alloy having a lower melting point than either the fuel or cladding alone. Analysis of the potential for a euctectic melt is lacking from the SAR.

4 l I m- . - _ . . _ . , , _ ,. - _ _ ._ , _ . .

l}7 18.. Excess reactivity hazard exacerbated by numerous related safety and operation problems.

a. Applicant has a history of inaccurately calibrating its instruments (see contention on inadequate maintenance and sinstrument 1

calibration), raising sbrious questions about their ability to know the actual worth of an axperiment, set their trip points, make sure 4 that reactor instability is properly corrected for, and avoid hazards that could be caused by inaccurate readings for neutron flux, power, temperature, etc.

b. Applicant's history of occurences causing unscheduled shutdowns raises questions about the ability to guarantee that incorrect procedures will never be involved in something like unsafe withdrawal of beam tubes,

c. Applicant's problems with the pneumatic tube system (leaks, need to replace tube itself, " archaic counting room") could result in a large' reactivity insertion.

d. Applicant's history of lax administrative and managerial controls such as inviting non-licensed operators to run the controls mandates that the reactor have nowhere near the amount of excess reactivity permitted to cause a damaging power excursion,

e. Applicant's history of by-passing interlocks and safety systems j raises an unacceptable probability that such an accident could occur through such a by-passing.

f. Applicant's troubles with control blades (pinning, failure of drive logic, lack of replacement motors) creates an unacceptable likeli-hood that the 10.CFR 50 Appendix A III guidelines cannot be met by this facility with regards redundancy and capability of reactivity control i l l system.

i I _.

,I

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g. Positive temperature coefficient for graphite moderator raises serious concerns that should an excursion take place, normal self-limiting features once thought inherent in this reactor will not be available.

h. Lack of air-tight containment structure and siting in a crowded building on a crowded ca= pus make the reliance on the cladding as the "' principal barrier"to fission product release (Application, V/1-4) worrisome and call into question Applicant's ability to demonstrate protection by multiple fission product barriers.

1. There are only four control blades and there are no back-up blades. The Commission has expressed concern about blade drive systems and lack of back-up motors. The drive logic in the past has malfunctioned, blades have repeatedly become stuck. The worth of the blades is $1.80 for one and $2.40 to $2.70 for three. Anthony.Nero, Jr., says in A Guide-book to Nuclear Reactors (UC Press, 1979, p. 263):

Individual control rods are often " worth" slightly less than one dollar, indicating that a single control rod cannot be responsible for a catastrophic reactivity insertion."

j. The problem with the heat balance and long-term stabilization of neutron channels and apparent lack of heat balance-nuclear instrumentation calibration as required by the Technical Specifications (see contention on calibration) raise serious questions that the automatic scram systems may not be able to shut down when power exceeded the trip point because of inadequate calibration of those trip points. Facility has had trouble with this in the past to a small degree; a larger calibration failure on the trips could result in failure of a back-up system when there are few others. (see 19 annual report, p.2)

t

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4 4

k. Graphite swelling and cracking--a problem with numerous other i

reactors that use graphite (Hallam and Pfqua as examples) has not been f examined for this reactor, as well as the possible effects of soaking the graphite by steam or water. Graphite swelling could make 1::t3rtion of control blades difficult; graphite soaking could change reactivity coefficients.

1. Fuel blade warping has not been examined in the Application, despite repeated problems with tie bolt failures in the past (see inspection reports for 1968).

m. The apparent lack of a deflector changes many of the assertions made in the Hazards Analysis about protection against repeated i

excursions, if indeed the deflector has been removed as it appears. CONCLUSION The plethora of safety problems and practices which impinge on . and interrelate with the excess reactivity characteristics of the reactor again emphasize the need at this facility for a very large margin of safety to be built into the excess reactivity limitations, as in fact the designers of this facility had intended. The gradual remeval, one by one, of those safety features raise serious safety Concerns. The twenty-year-old calculations, the positive temperature coefficient of the graphite moderator, the minimization of other 1 self-limiting features, and the number of ways in which the excess reactivity of the reactor could be affected lead to only one possible conclusion, that the Applicant cannot give reasonable i

20 t 4 l assurances that the reactor will not endanger the public health I and safety. 1 ) There are few areas of the reactor operation _ that are more l critical for safety considerations than reactivity control and cautious and prudent limitations on the amount of excess reactivity i that can ever be present. Given the gravity of this matter, Applicant has a heavy burden of proof to demonstrate chat the current limitations l on excess reactivity are sufficient to adequately protect public ; health and safety. Virginia Electric and Power Co.; ALAB-256, 1 NRC 10, 17 at n. 18 (1975). It should go without saying that t Applicant cannot meet such a heavy burden by relying on an analysis based on dated references, questionable assumptions and comparisons, i and failure to take into account changed characteristics. Consequently, the Application as it stands cannot support the issuance of the l 3 requested license. 1

It appears from all the material in the above contention that l this facility may very well have the potential of a catastrophic excursion caused by rapid insertion of too much excess reactivity, 4

causing a' cladding melt and significant fission product release. The facility appears to be considerably over the safe level of j excess reactivity; it has the means to insert it; and it has an operating history that indicates an extraordinarily unacceptable i likelihood that such an error could occar. Page 60 of the old Hazards Analysis, not included in the SAR, indicates that of the two methods by which they thought a reactivity accident could be l ' initiated, both would " require the . . . violation of the operating ' i . l I f

        , , - - - - ,                  ,~r      - - - -              -      -,

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rules by the operator." If protection from a catastrophic reactivity accident. rests on the assumption of reactor operator compliance with procedures, with this facility's history of problems, a threat-to public health and safety of considerable proportions exists. While it is unlikely that a reactivity insertion could bring the fuel to the temperature of the melting point of the uranium, Applicant's own analysis indicates an unacceptable likelihood that such a temperature for the melting point of cladding can b, reached with the excess reactivi2y presently permitted. The result would be the same, in essence, from a public safety point of view; the primary reason to worry about fuel melting is fear of breaching of the containment. Since this facility has no containment structure, a breach of the cladding would be sufficient to create unacceptable exposures. (for details about likely exposures in case of such an accident, see contention about 10 t-" 100). i E~ l 5

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1. Estimated safe limit for Borax I.

2. Estima.ted safe limit for Borax I with. Educator Reactor plate spacing.

K, 3, Estimated safe limit for Educator Reactor. ENERGY OF EXCURSION FIGURE D-5 Total Energy of Excursion Minus Energy Required (4 mw-sec) to Raise Temperature of Center of Average Plate to Boiling Point. (f rom Fig. 41 AECD-3668, Appendix F. Ref. F-3) l 1 f'ro,n l$40 2AfDC l l ,_ 4AJA4yQ5 - I _. . _ _ _ . r.

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I 5 10 50 I00 40 0 i SL-1 accident cust withdrawal INITIAL REACTOR PERIOD ( msec ) G r: actor. This i. translint whose FIC. 3-30 Preemed and measured nuclear nerTy release vs. pertad for EORAX-1.SPERT !. and SL-1 j vt in Table 3-9 Total energy unless otherwise s;ecined. Circles are BCRAX data trom reference (19), squares SPERT-1 DU 12/25 data from (65a] and triangles SPERT-l des:ru::s* test aata from [65L is withdrawal of pressure region. The pressure wave front which rIactor critical a single slug. The water level in the tank was , about 2.5 ft (76.2cm)belowthetcpof the vessel and developed no doubt spread out in all directions, m calculate t%t striking the vessel side walls next to the core a contLnuing the the slug, therefore, had this distance to acquire a (tha position in kinetic energy. This slug hit the bottom of the first and btuging them, then stri'dng the bottom idwiracollapsed of the plate area in thecentral16 elements reached head and giving a net downward force on the the vaporization temperature and this caused vessel," and finally acceleratingupwards the entire t!!:vs that power mass of water above the core. It appears I!kely d:nat:ly 4 maec more steam prodetion and violent destruct 10n of this region. About 20% of the entire core shows that the wate- moved upwards more er less as tation ttrminated 3 0.4 x 10 4 3tw, melting proceeding to the clad surfaces. General ep rature in the Electric estimates that the total nuclear transient had just reached energy was 133,10 Alw-sec and that no more than *Apparently no one has looked into this down-an additional 33 3 w-see of energy ibest estimate ward fo rce and one can only cenjecture as to whether C60'C (3767'F). this dowiward force was sufficient to sever the 11:. or 0.8 M mm) 24 +10 alw-sec) was releasediachemicaireactions ttr surfaces had between the molten or vaporized metal and water. pipe connections to the tank. It is difficult to From the start The formation of the steam void terminated judge the resistance to such a shock provided h3 excursion,3~i the nuclear transient, but it also created a high by the vessel supports. c;I */ J gt e.~ f /fdI <r

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4 E EXCESSIVE RADIATION: VIOLATION OF RADIATION STANDARDS; INADEOUATE MONITORING Applicant has in the past and is at present emitting excessive radiation and violating radiation standards and conducting inadequate radiation monitoring. Applicant has failed to demonstrate in Application l or recent performance any evidence that these conditions can reasonably be expected to improve in the future, in the absence of which demonstration a granting of an operating and SNM license cannot be made without undue threat to public health and safety. Specifically,

1. Past and present emissions excessive.

The detailed basis for this contention is contained in document submitted as part of initial Petition for Leave to Intervene. That document, "The UCLA Nuclear Reactor: Is It Safe?" gives in considerable specificity bases for contention that both past and present emissions have been and are excessive. A brief su= mary follows:

a. In a January 1975 Inspection Report, No. 50-142/75-01, page marked " summary", the inspector noted several violations of the technical specifications and 10 CFR 20:

1. The exhaust stack was found to be 17 feet short.

2. The ventilation exhaust from the reactor room was not being diluted to the required 14,000cfm rate.

3. The radioactive gaseous effluent monitor had not been calibrated with the required frequency.

4. An error of a factor of ten in the calibration curve for the gaseous effluent monitor had been identified.

i

1? 4 1 I l S. Extensive construction of new facilities around the reactor had resulted in a condition which could conceivably have personnel immersed in the discharge plume from the ventilation exhaust stack.

b. The error factor in the effluent monitor was later found to be a factor of 300 instead of 10.

(see letter from Charles Ashbaugh III, Reactor Supervisor, to David Jaffee, USNRC, dated April 23, 1975, correcting past calculations of Argon-41 releases).

c. In a 1974 Inspection report (no. 50-142/74-01, UCLA was cited for not having an accelerator nozzle on the exhaust stack as required.

d. In April of 1975 the NRC responded to UCLA's efforts to rectify some of the problems above. The NRC found their response "an unacceptable response" in part because

, the revised calibration figure on the gaseous effluent monitor had revealed that annual average discharge concentrations were above limits permitted 3

by the Technical Specifications (essentially 10 CFR 20 limits).

NRC Memo: Enforcement Conference

and Subsequent Actions, UCLA,

) Docket 50-142, April 22, 1975. 1 j Thus, UCLA had been determined by the Commission's Inspection and l l Enforcement division to be "above limits permitted by the Technical Specifications (essentially 10 CFR 20 limits" in part because of all the other inadequacies--stack too short, no accelerator nozzle, not l enough flow rate, placement of new structures around reactor stack, and so on. That finding of noncompliance with both tech specs and 10CFR20 limits, due in large measure to significant underestimation of radioactivity releases, makes clear that the violation had gone on for many years previous because it was only in 1975 that the error in calibration was detected, despite the fact that correspondence of concern between the Commission and the Applicant goes back into ! the early sixties on the question of calibration of the Argon monitors. l l I

                                                                                          - jf   1 1

2. Future emissions

, Several of the underlying conditions that caused the emission to exceed the Technical Specificacions before they were + amended have not been changed and therefore there is reason to believe that persons will continue to be exposed to such excessive emissions,

a. The reactor stack nozzle has been removed from the stack.

b. The height of the stack has never been increased.

c. The-overall emissions of the reactor have been increasing over the last few years and will continue to increase if reactor use is increased.

i Releases of radioactive Argon gas have increased from 33 curies in 1976 to doubtle that amount, 65.6 curies, in 1979. (p. II/2-5). Thus, the only demonstration made by the Applicant about future intentions is the increase in emissions. Fromises of installation of decay tanks are more than offset by Applicant's linking of said tanks to increased reactor use factor and/or power. The only evidence presented about future emissions is that they are likely to be more endangering to public health and safety than they are now. Especially if the placement of Math Science airvent remains where it is, directly downwind of the reactor stack.

3. Failure to Adequately Monitor-Applicant has not in the past, nor has it in tne present Application, been able to reasonably demonstrate that exposure in I unrestricted areas are not excessive. The radiation monitoring system,

devices and programs are wholly' inadequate. !

             -a. Applicant has never directly measured the Argon anywhere except 4
                  -          -,e    -                    ,-  p                      y ---- - - -

Lj

at the stack, where its concentration at 100 kw is varicusly reported as between 1 ano 2 x 10 -5 C1/ml of air, 50 times MPC even when the

, reactor utilization factor (18.8%) is figured in. No measurements of Argor ' ave every been taken elsewhere.

b. Applicant has had repeated difficulties calibrating their aeriitoring equipment, leading to past errors in emissions calc'21ations oi 300 fold.

c. In preparing his study on "The Atmospheric Dispersion of Argon-41 from the UCLA Nuclear Reactor," Mark Rubin found that the required systems for measuring Argon directly "were not available at UCLA...

and finances seemed to preclude the development of the required radioae-ive decay system." (p.4) Rubin then chose to simulate Argon dispersion with a tracer gas, sampled with 35cc syringes. (p.9) The sampling technique produced almost 10-fold variations in single location samples. These wide variations were caused by the lack of adequate sampling equipment (Rubin, p.24). However, Rubin utilized these results obtained through faulty sampling methods in reaching his conclusions that exposures with within the limits of 10 CFR 20, although the greatest likely public exposures were seen to be within the Math Sciences building, where only a few of the SF6 samples were taken.

d. Contradictory data. A TLD study done by the Applicant produced i

results that they admitted were "not free of ambiguous interpretation," (1978 annual report, p.24) even af ter rejecting the high readings, something they did without testing the supposition that led them to do it, that the TLDs were picking up radiation from the concrete. They l

       -lost a few TLDs to birds and " curious individuals," threw out the remaining high readings, and were left primarily with TLDs upwind of l

l 1 1

jT the stack. Readings on the stack were similar to readings 100 feet away, despite the assumption that dilution would be at work. One TLD was moved from near the stack to directly on top of it; if anything, its reading went down'

e. The film badge data reported in the application contradict even further the TLD data. Page II/2-la of the Application indicates that badga xi located in the reactor stack (at the same position :=

I the TLD NEL reports as averaging 43.2 mrem / year gamma + estimated 10 mrem / year beta, according to Tom Collins, Assistant Dean of Engineering, i

in the Daily Bruin 11/16/79) reports a reading of 350 millirem / year leta only. When one adds in the gamma Jadiation at a rate of four times beta, based on Dean Collins' estimate cited above, the resultant 1

exposure is 1750 mrem / year. The figure is over 25 times the level the TLD was reading at the same location. i The above basis indicates that the studies conducted by the Appl 1 cant have either suffered from such serious methodological 4 flaws or demonstrated such inconsistent results with each other that the reliability of all of the studies arc c.'lled into question. Given the lack of reliable empirical data it is impossible for Applicant to assure that the concentrations of Argon-41 or other radioactive

     ' substances being released from the reactor are not reaching the public in quantities inimical to their health. One last example makes that very clear. Applicant (p. II/2-1) states that measured levels of direct radiation in uncontrolled areas near the reactor "are not detectable above background ('J 0.0410.03 mrem / hour)."                    That sounds like a very reassuring statement until one realizes that 0.04 mrem / hour is 350 mrem / year, and that background in Los Angeles is 80-100 mrem / year.                  So Applicant 1
         . - . _   , , _ - . _ .,   _ __         ___   . , . _ , _ - -               ___ _ _ _           . _ ~ ,     --

d A 4 l can't detect radiation above background because they define background as about four times what it is, can't measure below that level, and t even at it have a probab1$11ty of error of 75%. Thus, it is possible that radiation levels in unrestricted public areas could be as high as 612 mrem / year and the detector could read zero (350+75%).

It is clear that Applicant is incapable of adequately demonstrating emissions are not excessive and comply with federal radiatic.. standards and their own technical specifications. The film badges that Applicant has placed in various locations in Boelter and elsewhere have threshholds such that the levels of radiation likely to be found withir t

,                  those buildings, unnecessarily hazardous though they may be, would

.1 not approach those threshholds, and the readings would be zero, nothing over background, even though there might be a sizeable exposure 4 taking place. (see sensitivitias of film badges in memo by Jack Hornor, November , 1979, on film badge locations).a s -

4. Failure to meet radiation standards a.10 CFR 20 (appendix B & 20.IO6.b and 20.106b(1) and (2)
1. NRC Memo: Enforcement Conference and Subsequent Actions, Docket 50-142, April 22, 1975 stated:

the revised calibration figure on the gaseous effluent monitor had revealed that annual average discharge concentrations were above limits permitted by the Technical Specifications (essentially 10 CFR 20 limits). It would appear that this finding of violation of 10 CFR 20 limits would indicate that the Applicant had been in

,                                  violation of those regulations for years, because the 1

calibration error was a long-standing one. 1 3

                                                                                               , _ _. _,  w

77

2. 20.106(b) states that an application for a license or amendment may include proposed limits higher than those specified in Appendix B of Part 20 and that the Commission will approve the proposed limits if the applicant demonstrates (1) That the applicant has made c reasonable effort to minimize the radioactivity contained in effluents to unrestricted areas; and (2) That it is not likely that radioactive material discharged in the effluent would result in the exposure of an individual to concentrations of radioactive material in air or water exceeding the limits specified in Appendix "B" of this part.

(c) An application for higher limits pursuant to paragraph (b) of this section shall include information demonstrating that the applicant has made a reasonable effort to minimize the radioactivity discharged in effluents to unrestricted areas. . . . 1 It is therefore contended that UCLA's Application for a license includes proposed limits higher than those specified in Appendix B of Part 20, that Applicant has failed to include information " demonstrating that the applicant has made a reasonable effort to minimize the radioactivity discharged in effluents to unrestricted areas," and that Applicant has not adequately demonstrated, nor can it with the inadequate monitoring and simulation' tests done to date, that no individual in an unrestricted area will be exposed to excessive levels of radioactive material discharged in Applicant's effluent, and that therefore, Applicant's request for license must be turned down because concentrations of Argon-41 in Applicant's effluent are higher now than when the Commission cited them in1975 for violating 10 CIR 20 limits and Applicant has failed

[f to provide the information required and to make adequate showing (of reasonable efforts to minimize radioactive releases and of demonstrable unlikelihood trat a person in an unrestricted area could be exposed to excessive emissions from the facility) to qualify for I Commission ruling permitting discharges higher than 10 CFR 20 Appendix B concentration limits,

b. 10 CFR 50.36a--ALARA Requirement of Technical Specifications; 1) Applicant was cited in 1975 for violation of Technical Specifications limits on emissions, for a situation that
was, when discovered in 1975, one that had existed for a great many years previous. Thus there was long-term violation of the Technical Specifications emissions limits that paralleled the violation of 10 CFR 20 limits; since ALARA limiting values are to be small fractions of 10 CFR 20 limits, it is clear that in 1975 and for a great many years previous (while Applicant was underestimating Argon releases a

by a factor of 300) Applicant was in violation of ALARA as well. It is important to note that the concentration of. emissions in stack effluent has not gone down since 1975-- annual releases have tripled since then (II/2-5); thus there is reasonable basis to assume that ALARA remains unmet. The reasonable efforts to reduce concentrations of effluents , in unrestricted areas--raising the stack, increasing the flowrate, restricting the roof or adequately posting it, moving ventilator 4 ! intake or stack itself, putting accelerator nozzle back on--

                                                                                                                      <[  ,

are all undone. The principle of ALARA--quite literally taking every 4 action that can reasonably be taken to keep emissions as low as reasonably achievable--clearly is not followed by Applicant. There are a host of reasonable efforts Applicant could have made and still could make to keep emissions as low as reasonably achievable, yet Applicant has made none of them. The recommended way of showing compliance with 10 CFR 50 radiation standards (limiting conditions for operations and criteria for licenses) is through Appendix I to CFR 50. Reg. Guide 1.109 is here very useful in determining numerical guidelines for Appendix I. Applicant has made no effort to show that it meets the guidelines of Appendix I or 1 the numerical guides in 1.109 Reg Guide, nor has it proposed any other way in which to measure its performance with regards the ALARA standard.. If one takes the limited amount of data, flawed though it is, that is available from Applicant regarding possible radiation levels in unrestricted areas, it is clear that ALARA as normally defined is being violated. The SF6 study, for example, indicated likely exposures inside the Math Science Building to be roughly 12% of MPC, which would be several times ALARA. The TLD and film badge results similarly shev expcsures in excess of the 5 mrem whole body 10 CFR 50 Appendix I criteria. If a big power plant must adequately demonstrate that they will meet those design criteria

,      before being licensed, and in the absence of any proposal by Applicant for alternate criteria, it seems most reasonable to assess their license by the same criteria. And according to those criteria, Applicant has failed to meet them and the Application for license should be turned down ,
                                      -w.--,           m   _____.y_ _ . , _ .,. _ - , . , ,          ,,
                                                                                                          ,.,._,,,-,y

!                                                                                                                                   /d>

c. Failure to demonstrate meeting of the 40 CFR 190 standards that members of the public will.not be exposed to 25 mrems whole body or 75 mrems thyroid by.the planned discharge of the reactor. The Rubin data, the TLD and film badge data, and the concentration figures
of the Argon monitor all indicate the 40 CFR 190 standards are not being i

j met. .

                      . d. 10 CFR 100 criteria. This is discussed in a separate l              contention regarding 10 CFR 100. Suffice it to say here that Applicant's own Hazards Analysis' indicates that thyroid doses in case of an accident 4
;             could be 1800 rem to members of the public, considerably over the 10 CFR 100 limit of 300 rem thyroid dose.

CONCLUSION: Applicant has for long periods violated 10CFR20 requirements and the requirements of its own Technical Specifications. Applicant has clearly demonstrated that it cannot meet the 10CFR50.34a and .36a and Appendix I guidelines for a showing of adequate meeting of ALARA; likewise it has clearly demonstrated it does not meet the siting criteria of 10 CFR 100 to ensure that radiation exposures in case of serious 4 accident are kept below the 10 CFR 100 levels. Failura to meet the 10 LFR 50 and 100 licensing criteria mandate rejection of the Application; j history of violation of 10 CFR 20 and the Applicant's own Technical Specifications over long periods of time indicate a serious safety ! threat would be posed if the license were granted; and failure of Applicant to make showing of compliance in the future with the above 4 standards and to show significant improvement in its monitoring and i l emissions control systems make it a public health and safety threat I. I for such'a license to be granted absent serious showing of changed practices. l

o i

i 1 E LACK OF OPERATIONAL RELIABILITY PERSISTENT PATTERN OF UNSCHEDULED SHUTDOWNS, ABNORMAL OCCURRENCES, AND ACCIDENTS i The reactor has in the past experienced a persistent pattern of 4 numerous unscheduled shutdowns, abnormal occurrences, and accidents. 4 These occurrences are so pervasive that they evince a pattern of unreliability which makes it impossible for the Applicant to reasonably assure that the reactor can be operated in a manner which does not endanger the public health and safety.

a. A review of UCLA's annual reports for the years 1971-1978 indicate forty-five unscheduled shutdowns and four abnormal occurrences.

During this period the reactor was shutdown for several extended periods r totalling approximate ly in excess of one and one half years. The major shutdowns were due in part to the 1971 earthquake (the resulting maintenance required the exposure of three workers to greater than 5 rem / year each; pg. 6 1972 Annual Report), a coolant leak in 1974, 4 and a leaking Radium-Berylium source in 1975. Some examples of the causes of the scrams are: " Operator made incorrect range change. Ranged up rather than down. Operation was reviewed with operator." " Operator inadvertinently /Jiy pressed scram switch while lecturing on console instrumentation and circuitry. I Revised lecture procedures." (annual report 1971, p. 1). Another scram which raises questions about the general reliability of the 1 l i i l l

2L reactor operation occurred in 1977 when the reactor supervisor blocked the main exhaust vent to prevent a tritium build-up, causing the Argon-41 ronitor alarm to go off. "The supervisor was cautioned to exercise more diliger.ee in the future." (annual report 1977, p. 2-3). Only two of the inspection reports available to Petitioner at this time give a yearly average for the rate at which spurious scrams occur. The figures are .923 per month and 1.14 per month from Inspection Report 68-01. Those rates correspond to rates of roughly 240 per year for full time operation (facility cperates about 5% of the year), a rate that would clearly be unacceptable in a big power plant or any other facility that could pose a safety hazard.

b. The Applicant also has a history of leaks and spills.

Leak in reactor shield tank found in 1968--C0 Report No. 50-142/68-2. Coolant leak late in 1974 " required reactor down-time from mid-August to early December." and down again"from November 23, 1975 to March 18, 1976 to replace a leaking encapsulated neutron source." Catton, 1976 Annual Report, p.3. In 1979 demineralizer tank on the floor directly overhead of the console leaked--leak continued all weekend as the staff of NEL did not know how to turn it off and didn't check to see whether someone from physical plant had taken care of the problem--short-circuiting console instrumentation, necessitating a week's shutdown for dry-out and repair. (Daily Bruin, November 21, 1979). Su=mer 1979 spill of radioactive liquid during clean-up from previous week's spill due to failure of sample and pneumatic tube, necessitating clean-up and replacement of tube. (KNBC report October 1, 1979). The reactor also sprung a leak in a reactor gasket in 1974 (R0 Inspection Report 50-142/74-01. The above-mentioned shutdown, leaks and abnor=al occurrences do

b not by any means represent all of the instances of reactor events substantially outside normal operating procedure. They are intended merely to show that there is a basis for the concern about an apparent history of operational unreliability that can represent a pattern of sloppiness on the part of personnel and unreliability on the part of reactor equipment that can pose significant safety hazards in the future, particularly when coupled with the safety inadequacies in other portions of this series of contentions (see particularly inadequate calibration, managerial controls, lack of safety features, and violations of regulations contentions.) There is considerable basis from the brief operational history cited above to base a contention that the Applicant has not reasonably demonstrated that the reactor operation is reliable, that a good safety record exists, and that the reactor operation will not endanger the public health and safety. l l

e mee #f N-gom- e o % Reactorishut down , ; f* $5.,:,; ; the leak on Friday afternoon by water damagel.".' By Mary Astadourign ,

                                                                                                - -- - ~sieni lw Office and told them to do it, ,

shut off the water valve, because siert wrii" he did not know how. Ilut he - A water leak developed m the control room of the Boeller' ' ; forgoi to check and see if they

     .Ilall nuclear reactor over the weekend, causing exten,sive                                did, the caller said.
    .> damage to the main control panel, and rendermg the reactor                                   Ashbaugh denics this. "Two',

inoperable for a week or more. .

                                                                              ,                  people from Physical Plant
         'An anonymous' phone caller told the,liruin about the leak..                    .

which started Friday in the de-ionized water tank, and rapidiY , ' [ came Friday that they would takemorning and said care of the sincapacitated the control room but pid not release.any

leak - it was just a few drops at .

radioactive water. Y' the time," he said. Ashbaugh l Reactor supervisor Chuck Ashbaugh confirmed the report' t , believes the major leak began. Tuesday. Ashbaugh explained that two weckt ago th'c main j % over the weekend when the water pipe in the School of Engineering broke. A temporary $ reactor was not in operation , pipe was installed but not depressurized; , :>' and when none of the reactor "This increase in pressure caus'ed the machine that mak'es'I personnel was present. "~l he ' I purified water for the reactor to leak," Ashbaugh said. lie also - custodian found it and Iold the explained that this water is used for experiments and"doesn't custodian supervisor, who belong to th: reactor." - . N,. called Physical Plant," Ash-

T he damage, according to Ashbaugh, is not serious."Some ! baugh said. of the instruments got wet and we're drying them out," he 1 i" Ashbaugh said he would said, adding,"We don't think Jhat anything got burned out, .; % allow the Bruin to take pictures just wet. ,

                                                     * .                              :N of the damage in the Nuclear i,N Ashbaugh beh, evesi ,t will cos,t about,5500, primarily fo,r                            Energy Laboratory. ' llowever, labor, to get the reactor operatmg again.                                      .

I)can Tom Collins of the School The ansnymous caller blamed the damage on Ashbaugh's . i ,of Engineering refused to allow carelessness. Ihc caller reported that Ashbaugh discovered a Bruin photographer to do so. (Coptinued on Page 11) -You can clardy the incident." said Collins,"but I don't think you need any pictures." ,

               ~
                                                                                                                                           --C

I g, FAIII."?2 TO 70.W 10 CFR 100 SITC:3 CRITERIA REGARDII!G RADIATION BUASE I:! A I AOCID'.!C 2.e Applicant has failed to meet the siting criteria of 10 CFR 100 regarding radiation exposure in case of an accident. The calculations contained in the SAR are based on numerous assu=ptions that unrealistically

        =ini=1::e the extent of exposure in case of a major release of fission products, and that even with these unrealistic assu=ptions, Applicant's own SAR indicates that it is in violatien of 10 CFR 100 standards for th/roid dose to the public in case of a =ajor accident. The standards set a li.it of 300 re for thyroid exposure (10 CFR 100,11(a)(1) & (2))

and the Application finds that the possible dose fro = a raactor accident is 1300 re=s for thyroid dose. (Application at III/B-6) Specifically,

1. Even with nu.erous unrealistic assu=ptions, outlined below, and despite claims to the contrar/, UCLA's own SAR indicates that an accident would cause public exposures in excess of the li=its in 10 CFR 100.11(a)(1) and (2). The regulations require that applicants for licenses de=onstrate that =e=bers of the public would not receive thyroid doses .in excess of 300 re=s in event of a major accident involving significant fuel =elting and fission product release. The Application, 7 despite many unrealistic assumption which limit the estimate of release, l

indicates a thyroid does of 1300 re=s, considerably in excess of the 4 i li :it. This is in direct contradiction to the statement on page II/3-1 of the i= Applicatien which states: Accidents ranging fro = failure of exper:.nents to the largest core damage and fission products release considered possible

o . 2 result in doese of only a small fraction of the 10 CFR 100 guidelines and are considered negligible with respect to the environ =ent. This statement - - which appears to have been taken word .for word frc= page 3 of a January 1974 meco about all research reactors written by Daniel ::uller, Assistant Director for Environ = ental Projects, Directorate of Licensing, A3X:, called " Environmental Considerationa Regardirq the Licensing of Research Reactors and Critical Facilities." Applyire this assertion to the UCLA reactor is clearly inappropriate in light of the i Applicant's own figures on page III/3-6. 1800 re= to the thyroid is certainly greater than the 10 CFR guideline of 300, not "a small fraction of the 10 CFR 100 guidelines."

2. "he analysis is considerably flawed, in part because of its age, in part becuase of its reliance on dated references, in part becuase of changea in reacter characteristics and the site at uhich the reactor is located and largely because of highly unrealistic assumptiens.

a. Applicant assumes a release that is limited to only 10% of the 4

volatile fissien products and none of the non-volatile products. Tnis j is a c0=pletely unrealistic assumption. To =cet 10 CFR 100, one is

l. required to use the following assu=ptions:

l l The fission product release assu ed for these calculations should be based upcn a major accident, hypothesized for purposes of site analysis er postulated frc= censiderations of possible accidental events, that would result in potential hazards not exceeded by those frc= an/ accident considered credible. Such accidents have generally been assumed to result in a substantial =elting of the core with subsequent release of appreciable quantities of fiscion products.

                                          - footnote to *0 CFR 100.11(a).

Tnis particular reactor need not consider core celting, cerely i

cladding =elt, because it has no containment structure and admits that the cladding is the " principal barrier " against radiation release. (Application 7/1 h)

M

 .. .                                                                                        l
                                                  <                                       ~$ l Applicant assumes that "none of the nonvolatile fission products are transferred to the building air..."

The foregoing set of circumstances is consistent with the reasonable assu=ptions made here that the incident is not , violent enough to blow off the top and side biological shields so as to cause an intense spary of water-stea=-radioactivity mixture into the building. Application, page III/3-1 This is a most unreasonable assu=ption. A steam explosion is a verJ-clear possibility. 10 CFR 100 provisions are to be met by considering the maxi um credible accident, not a cediu=-sized accident. S e biological shield for this reactor is not designed to prevent such a steam explosion

and is not bery strong. There is absolutely no reason to rule out the

possibility of such an occurance.

           'uSH-740, in its discussion of the possible consequences of major accidents at nuclear plants, used two scenarious for rolcase: 100 percent of the volatiles plus one percent of the Strontium; and 50 percent of the volatiles plus the non-volatiles. It was these assumption that should have been used by the Applicant. Assu=ing no volatiles are release and only ten percent of the volatiles is an asst' .ption that could result in a considarable underestimatien of fission product release, at least by an order of magnitude.

l b. Applicant assu=es reactor has been operated at 10 kw "long enough to have attained equilibrium concentrations of relatively short-lived fission products, i.e. the iodine, bromine, and krypten isotopes." (p. III 4 -1). The reactor is new at ten times that power level. Applicant argues on page IIIM-1 4

The calculation of fission product. inventory is based upon a steady state equilibrium inventory at 10 kat, and certain assu=,tions

                    ;      concerning leak rate frem the building.

The consequential dose calculations were appa ently unreviewed in the approval of A=endment 3 (1963) that increased the nazi =um licensed pcwer level to 100 kwt. S ey were reviewed by the

Division of Licensing and Regulati:n in procession the application for Amendnent 7 (referred to above) / increasing limit on excess reactivity /. In view of the current restriction of the UCLA Reactor operating hours to 55 of the year, t. e maxi =u . average po.ter is now 5 kwt, a factor of two less than the 10 kwt used in th original calculations. The above argument by the Applicant cannot be supported. The Hazards Analysis from 1960 based its calculations on 10 kwt, but did not assume that it ran 100 percent of tha year, only "long enough to have attained equilibriu= concentrations of the relatively short-lived fission products."

   *dhile UCLA is limited to operating five percent of the year at 100 kwt, there is no evidence presented to show that its maxi =um fission product inventory is half that of a reactor vhich runs at 10 kwt "long enough to have attained equilibrium concer     ations" of the short-lived fission products. Neither the very long-lived nor the ver; short-lived isotopes would be affected by UCLA's current five percent restriction; UCLA could run continuously for 13 days at 100 kut, or 36 days at 50 kwt, or three
    .onths continuously at 20 kwt and still be within its licensed limit.

The Application does not state the " equilibrium period." It is possible that the equilibrium period and UCLA's current restriction could catch. Purthermore, the only part of the inventory affected by five percent restriction, if an-/ is at all, would be the "relatively short-lived fission products" -- the oth:rs would be unaffected. No attempt has been made to accurately detemine the maximum fission product inventory of the reactor at 100kwt and five percent of the year operating limit. It is clear that it is very unlikely that the reactor at 100 kwt would have a may.i=u= '.nvertor-/ of one-half of the reactor inventory at 10 kwt, even with the former restricted to 433 full-power hours per year and the latter measured at the end of the equilibriu= period for relatively short lived products. It should also be noted that the calculatiens at TCkwt assu=ed thct the reactor operated only 20 percent of the time. See, 1960 Hazards Analysis at page 62. It is clearly possible that

5 the failure to recalculate possible release levels at 100 kwt could result in an underestimation of the release by nearly an order of i . magnitude.

c. Applicant assumes the reactor is in a two-story building with a

place possible exposure to the public occuring outside the building. l That analysis was useful in 1960 when those were the conditions at ! l 4 l hand; however, now the reactor is housed in a massive building complex i

;                   used by perhaps several thousand students, staff, and facul'ty. The i

calculations based en plu=e travel time and air dispersion are obviously 1J useless when people now can be exposed not berely on the outside of l

the building as the radioactive gas passes but by being 1:ntersed in i radioactive gas which passes throughout a huge building aided by the ventilation systems. Exposure codels for inside the building clearly l l tust be used. The Applicant did net use such models. '"his could result 4 in significa.t underestication of exposure.

d. Page III/Y-2 assumed a building leakage rate of 20$ of the reactor room volume per hour for a 30 mile per hour wind, ass =ed to be directly proportional to wind velocity. '"his assumption is erreneous in part because

<                                                                                                                                                                                                          4 I

it is based on leakage Out of the building into the open air instead of the current situation in which a primar/- area of exposure wo"'d be leakage out of the reactor room into other parts of the Boelter complex and in I part because the ass =ptien is predicated on the assertion that "all I access doors will be weather-stripped and emergency doors leading directly to the cutside, caulked and sealed for =ind=um leakage." Any such atte=pt to prevent air leakage has 1cng cince been renoved, as evidenced by the nassive airflow under the doors because of the negative 4 pressure kept incide the reactor room under nomal operations due to venting up the stack. If the stack is shut dcwn, as it is supposed to be 6

     --    -, -                       ,w-,.e--7w-,. ,    y      ~.            a , . + , , . .n-, ,- ,,, , - . . , - - - - , - , , , - . , - , , - - -              .,,,,,,.e,            p - - - .   , - ,

[ in an accident, that negative air pressure will cease and the radioactive material will flow out the same way the air previcusly flowed in. Any accident denerating encugh heat - - and perhaps a steam explosion - - to =elt the cladding would create an oce. pressure inside the reactor roc = that would force the air out. Once out of the reactor room area, the building's nor=al air circulation system will very effectively transport it throughout the buliding. The assumption of 20 percent release per hour in a 30 mile per hour wind outside (two percent release per hour in a , 1 three =ile per hour wind) with the only exposure being to someone outside the building is thus a vast underesti=ation, caused no doubt by the difference in situations in 1960 when the analysis was done (before the buildings were added to the reactor structure) and tcday,

e. Finally, the use of references frc= 1953 /nen new dose and dispersion models are now available, the failure in twenty years to test 1.ny of the
assu ptions upon which the analysis is based (e.g. why estimate who much 4

4 of the reactor air volume would leak per hour - - why not =easure it?) t i =ake the analysis ucrthless, particularly given the chances that have taken 4 j place in the reactor since the analysis vis done. It is not the seco a reactor that the 1960 Hazards Analysis analyzed. 1.n analysis twenty years old should not be used to demonstrate the safety of a current reactor for the twent/ years to come. "he Hazards Analysis was so far off in its l esti=ation of Argon Enissiens (p. 62 of the Hazards Analysis estimates at a radioactive concentration of 3.2 4c per c=3 of air when air flow of 5000 cf= and reactor running twenty percent of the year. Current air flow is three times that and the reactor is restricted to five percent cf the year; despite the fact that the concentration should be twelve ti es less than the Ha:ards Analysis predicted, it is actually ten times more - - an i i

                                .                 -- - , . - ~ - .      ,    -~,-,,e ,,n ,. ,,      e -- -.  ,.v.:

7 error of tuo orders of cagnitude. This certainly casts doubt on the accuracy of other estimates in the Hazards Analysis. Although the reactor iJ de31cned to Operate at a maxi =u= steady pouer of 10 kw, it is not planned t; perate it at this cwer icyc1 continuously. Much of the operation for the training progrs= will be at considerably lever power icvels and ',dil be inter-nittant. It is esti=ated that the reactor will be in Operatien

              /about 1300 hours each year or about 20 percent of the total tice/.

With thic type of operational progra=, no very largo a=ount of new fission products will ever exist in the core. Hazards Analysis at page 59 Thus, the Hazards Analysis was not censidering the 1960 reactor to be running full time at 10 'ce as the CAR would suggest; thus the assu=ption that the present 100 krt reacter's max 1=u inventory would be less than it was at 10 '.re see=s highly questicnable. Finally, it cast be said that the fo11cwing statc=ent in the Hazards Analysi: secticn on radiation doeses is clearly challenged by what has been shown in the section on Ixcess Reactivity.

               ... cuch an event is not censidered even plausible because of the limitati:ns en available excess reactivity and because of the inherent self-li=iting characteristics of the reactor...

Application at page III/3-1 Petitioner has shown elsewhere that since the Hazards Analysis was written the excess reactivity pennitted has been increased to well over the a=ount necessary for prc=pt criticality and indeed perhaps the a=ount needed for cladding =elting, and that the inherent self-limiting characteristics of of the reactor thought to be existing in 1360 as the Mazards Analysis was written are either no longer in existence or never were (the deflector shield has been recoved, no longer protecting against repeated excursicns; there may well be a positive temperature ccefficient because of the graphite moderatcr; and so on.) Despite all of these changes and erroneous assu=pdons, the Application still shows that the potential dose to the thyroid excees per=issible levels according to the siting standard set out in 10 CFR 100.11(a).

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                                                       % INADEQUATE MAINTENANCE AND CALIBRATION i

The Applicant, in the past, has not adequately maintained its equipment , nor calibrated its instruments properly, thereby increasing the chances of , i ! equipment failures and erroneous instrument eadings. Applicant has

;               failed to demonstrate that its maintenance and calibration effort will i

l improve in the future. This failure precludes the conclusion that the i 1 issuance of the license will not be inimical to the public health 1 and safety. i

.                     1. Applicant has failed to calibrate instruments at the required intervals.                                                                                                                                -

?

a. Inspection Report 050-142/75-01 reports as an item requiring i

j enforcement action: The licensee had not calibrated the reactor room area radiation monitors and the radioactive gaseous effluent monitor at the j frequency required by the Technical Specifications.

;                                                                                                p.1

b. The same Inspection Report indicates that the Gaseous Effluent i Monitoring system was also required by Section V.C. of the technical i

i specifications to be calibrated semi-annually, and that i the maintenance log shows no record of this monitor being calibrated. The licensee representative against stated he was l unaware of. the semiannual requirement of the tech spec. . . p.7

2. Applicant's personnel are not familiar with the calibration requirements of their own technical specificatiens.

4 1

a. The Inspection Report mentioned above explains the calibration violation

!               in part on the present Applicant's ignorance of the requirements in their own technical specifications:

2 The licensee representative stated that the area monitors were last calibrated about one year ago. The representative was unaware of the requirement for semiannual calibration and was under the impression thatthe calibration frequency was the same as for nuclear instrumentation (annually).

p. 6 75-01

3. Applicant has failed to maintain, or has lost, calibration records, making accurate instrument calibratins and data interpretation impossible.

a. When questioned about the validity of the calibration curve and the detector response to Ar-41 versus C-14, the licensee stated that the calibration curve was experimentally generated years ago and that the documentation no longer exists which shows how the curve was developed or what error it may have.

inspection report 75-01 p. 6-7

b. The Applicant lost the facility's maintenance log for all years prior to 1974, thus making accurate maintenance and calibration far more difficult because previous results were unavailable for comparison, along with records of calibration and maintenance methods. (SEE NRC Notice of Violation dated October 15, 1974:

Section VIII k.e of the technical specifications requires that a record be kept of the principal maintenance activities and the reasons therefore. Contrary to this requirement, the record of maintenance activities prior to May 1974 was missing. (Severity Category III)

4. Applicant has significantly underestimated radioactive emissions for extensive periods of time due to errors in its calibration methods.

a. The licensee stated that a recent calculation performed to c.mpare the exicting response of Ar-41 to C-14 inoicates that the existing calibration curve is in error by a factor of ten.

The licensee representative further stated that he is convinced these calculations are correct....

p. 7 (emphasis added)

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b. The calibration error reported above, when corrected, was not, as the licensee insisted above, the only error. When Argon emissions records were finaly corrected, it turned out the error was actually a factor of 300. Charles E. Ashbaugh III, then-Reactor Supervisor, in a letter to David Jaffee, USNRC Directorate of Licensing, on April 23, 1975, corrected previous years' reports of Argon emissions. As is quite evident, the error was much larger than just a factor of 10:

1971 -124.9 Ci released instead o.* .303 Ci reported 1972 - 41.9 Ci released instead of .1046 Ci reported 1973 - 52.9 Ci released instead of 0.248 Ci reported 1974 - 56.2 released instead of 2.39 Ci reported (the last year the facility had detected the initial calibration error and compensated for it by multiplying their readings by 10)

5. Applicant has had continuing problems with heat balance calibrations,

a. Inspection Report 68-1 indicates a long-term problem with key nuclear instrumentation related to calibration discrepancies:

Entries in the censole logbook and reactor checkout forms indicated that the nuclear instrumentation had performed correctly during the period covered by the current visit. A possible exception would be several occasions wherein nuclear instrumentation power level indications were not consistent with heat balance calculations. However, the maximum speed was noted to be approximately five percent and detector positioning corrections were made without undue delay. Dr. Smth said that the nuclear instrumentation--heat balance power level discrepancies have been a long-term, but not increasing problem. He said that work was continuing to stabilize nuclear channel long-term operation so that the need for detector relocation can h kept to a minimue.

p. 5-6, emphasis added

b. There is no indication in subsequent inspection reports thatche heat balance problem has been resolved.

l I c. Applicant's report of only one hour of reactor operation for maintenJn e and instrument calibration means that the heat balance calibration was . ] , not done last year, because it takes considerably longer than one hour. ' to do said calibration.

d. Applicant's technical specifications included in the application j differs from those previously in effect in that the rqquirement for conducting heat balance calibrations is not any longer in the tech specs.

f

!          6. Applicant has, by making undeclared changes in their technical specifications, j

] relaxed or discontinued essential calibration standards and requirements. i A 7 a. Applicant, despite statements that no changes of substance 1 between the present Tech Specs and the ones included in the Application 1 exist besides those outlined in the forward, has relaxed some instrument i calibrations without so declaring. This relaxation, especially in view of laxity in periodic calibrations, seems quite unwarranted. Specifically, I the present tech specs say that the log N period channel, the power level safety channels, and the linear power level channel will be " calibrated at intervals not exceeding 12 months, or any time a change in channel performance is i note." (p.5) However, the new tech specs say: I The requirements listed below generally prescribe tests or inspections to verify periodically that the performance of reqired systems is in accordance with specifications given above in Sections 2 and 3. i In all instances where the specified frequency is annual, the interval between tests is not to exceed 14 months; and when semiannual, i the interval should not exceed 7 months. V/4-1 Application I l

a a _ _ 0

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b. Applicant has also removed, without so declaring, the requirement that "the neutron channels shall be calibrated against an independent measure of core power at intervals not to exceed 12 months."

7. Applicant has not cevoted adequate time to maintenance and calibration Page III/1-5 of the Application states that only one hour of reactor operating time was spent last year in maintenance tests or instrument calibrations reb,uired by the teactor's technical specifications.

The Technical Specifications requirements ca-not possibly be met in one hour of reactor operation. The " heat balance" alone takes longer than that. It is clear that the reactor simply did not

do most of the maintenance and calibration last year that is required by the tecnnical specifications.

1 A review of the correspondence bibliography between the Applicant and the Commission fromthe early sixties on indicates that the a calibration problems have been long-term and are continuing. There is no indication that the managerial and administrative problems which underly the calibration inadequacies will be resolved in tie future. Applicant must demonstrate that.these problems will be i resolved in order to meet t'.te burden of providing reasonable assurances l that regulations will be complied with and that the public health and safety will not be endangered as required for the issuance of a license. a I I

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1 T,INADBaUATE HIVIR0fiMBITAL IMPACT APPRAISAL The Divironmental Impact Appraisal submit'ted by Applicant is lacking in detail, largely copied from material not related to this particular facility, relies on unsupported assumptions and conclusions

and is generally so inadequate that it cannot possibly support the issuance of a license or support the finding by the Commission that the licensing is not an action that significantly effects the quality of the human environment.

1. Iack of original environmental impact appraisal for this reactor, f

Applicant has ostensibly filed an EIA for this particular reactor, but much of the language has been lifted, without attribution and virtually verbatin, from Daniel Muller's AEC memo of January 23,1974, on "Divironmental Considerations Regarding the Licensing of Research Reactors and Critical Facilities." There is virtually nothing on pases II/3-1 ' I through 7-1 that was written by the Applicant nor can it be said that the contents of those pa6es represent a review of the environtal aspects of Applicant's specific facility. Applicant has made no showing that Muller's general conclusior.s fit the specific circumstnces of UCLA, nor for that matter did they identify the language as anything but their own.

2. Analysis of envimnnental affects of facility operation inadeouate.

Applicant's description and analysis of the envinnmental effects of the normal operation of the reactor, is based on faulty assumptions and unreliable monitoring equipment and methods,

a. Applicant has failed to discuss the effect of gaseous emissions other than through the reactor stack,

b. Applicant's definition of background radiation cited on page II/2-1asapproximately0.04A 0.03 aren/ hour, or 350 mrem I 262 aren

2.

        / year means that radiation levels of 612 mren/ year would be ignored by Applicant as background and insignificant even though backgmund in Ios Angeles is considered to be 80 to 100mnn/ year.

c. Applicant in interpreting its film badge data fails to analyze and predict the total gamma plus beta raditition dose based on the beta readings of the badges. This failure leads Applicant to understate the impact of its emissions by a factor of four,

d. The significant discrepancies between the film badge readings and the TLD reading, and among the TLD readings themselves, indicates that the Applicant has not demonstrated accurately what the levels of radiation it emits are, and has ro basis for a conclusion that there are no si6nificant levels of radioactiva emissions reaching uncontrolled areas.

e. A more detailed discussion of the levels of radioactive effluent given off by Applicant in its normal operation and its inability to monitor and measure such emissions can be found in the contentions on excessive emissions, and inadequate maintenance and calibration.

3 Analysis of environmental effects of accidents inadeauate. Applicant's description and analysis of the environmental effects of accidents is cursory and conclusory, despite the fact that a major accident at this facility could endanger thousands of lives. a. Applicant uses verbatim the language from Muller's memo on research reactor impacts without any attempt to justify the application of his conclusions to its facility.

b. Applicant's conclusion that the releases from the greatest core i

damage possible are within the limits of 10 CFR 100 are not supported by any data whatsoever, In fact the conclusion is inconsistent with 6he l l

3 ( data included in other portions of the Application. (see contention on compliance with 10 CFR 100)

c. Applicant's statement that the reacter was subjected to experimental vibration and that the results were published in a paper by C.B. Smith, is inadequate to support any conclusion. Especially, given the fact that the vibration test caused some danage to the reactor, and that the facility was shut down for seven months for repairs following the 1971 earthquake. (see contentien on seismic vulnerability)

d. This reactor facility is vulnerable to a major accident with serious consequences for the public health and safety. The reactor operates at or over the limits for prompt criticality, in a facility plagued with managerial problems, and in a buildir4 surrounded on three sides and attached to large classroom facilities. Given the reactors lack of self-limiting features and its close proximity to thousands of people, the facility represents a major potential environmental effect in the case of an accident. (see contention on the maximum credible accident).

4 Applicant does not discuss, analyze or describe any altematives to the operation of the reactor facility,

a. Applicant's conclusion that there are no suitable and economic alternatives to the reactor is taken directly from Muller's memo and is not supported with any discussion or data regarding the UCIA facility,

b. Alternatives to the reactor do exist, both for research and for training and education. In the research area for example the Medical Center has its own cyclotron for makir4 isotopes. Furthermore, the Nuclear Ehergy Iab has experienced a major decrease in research activity 1 l

1

Y by University researcher in the past several years, due to low power and antiquated design. Nothing could indicata more clearly that there are alternatives to the research uses of this reactor. (see contentions on wrong class license)

c. The minimal use of the reactor for education anr1 training, 34 hours or 8%, in 1979, suggests that such education and training might well be conducted more efficiently at other facilities in the Ics Angeles area.

d. Applicant cites the use of the reactor as a public relations tool as one to which there is no alternative. Petitioner submits that since, under the regulations the public cannot run the reactor, any purpose could be accomplished with a reactor mock-up.

5 Awlicant's discussion of the lene ter t effects of the reactor is conclusorf and inadecuate. In fact, they fail to mention the impact of decommissioning the reactor, a $753,000 item in 1980.

6. Applicant's discussion of the costs and benfits of the reactor is inadequate and conclusorf,

a. The value of the education of students is minimal given the en11 amount of reactor time devoted to it,

b. Applicant fails to discuss the need for the facility in light of the decline in the use si the facility for research by Uni /srsity personnel.

c. Applicant fails to discuss the value of the facility as a training tool in light of its antiquated design, or the fact that such training could well be done at other facilities for much less cost than running the UCIA reactor.

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d. The applicant in concluding that alternatives to the reactor operation would be more costly for some of the activitiea concucted fails to consicer that the cost of doing certain experiments on different types of equipnent is irrelevant if the research is not being conducted at the reactor anyway.

e. The applicant fails to give any data en the cost of using alternative facilities or means to conduct the activities that now cn at the reactor.

The application subnitted by the Applicant represents a total failure to comply with the intent and spirit of NEPA. Applicant has failed to adequately explore and discuss alternatives, costs and benefits, and above all has not accurately assessed the effect of the issuance of the proposed license on the quality of the hintan environment. Given such failures this appraisal cannot support the issuance of a license or support the finding that the action will not have a significant effect on the quality of the human environment. Furthennore, it is contended that the iseuance of this license will indeed have a significant environmental impact and that therefore an Environmental Impact Statement should be prepared en this action. Finally, Petitioner would like to nete that it wishes to be apprized of any and all opportunities for public input into the environmental impact and assessment process consistent with Council on Environmental Quality regulations. I l l

1 E. LACK OF ADERUATE SAFEIT FEATURES Applicant lacks key intrinsic and engineered safety features and other safety features are substantially inadequate particularily lacking are features that are redundant and independent. Furthermore, a number of safety features built into the reactor initially are no longer existent. Tha intrinsic safety features of the reactor have been substantially mitigated er removed. The 1960 Hazards Analysis for this reactor begins a discussion of " General Safety Considerations" with the following statement: The inherent safety of the reactor is based on four points. First, the amount of excess reactivity in the reactor is limited 1 to about 0.6%. Second, the reactor has negative thermal and void codfficients. In addition, the reactor is pro.2ded with sufficient interlocks and safety trips to make a ha::ardous incident extremely improbable. Thiri, the amount of contained fission products will be relatively mull since the reactor is to be limited to a maximum power of 10 kw. Fourth, there is no credible way in which the fission products can be made to escape. P. 59 As we shall see in what follows, each of the above four bases for the supposed inherent safety of the reactor has since been substantially mitigated. First, the amount of excess reactivity in the reactor is no longer limiced to .6%. Second, the reactor has a positive thermal graphite coefficient. In addition, the reactor's staff has over the l years found ways to disconnect the interlocks and safety trips, and the value of the latter has been brought into serious question by the lack of accurate calibration, particularily heat balance instrument calibra-tiens. Third, the amount of contained fission products is no longer small relative to twenty years ago since the reactor power limit has increased to 100kw. Fourth, there ara a number of credible ways in which fit,sion products can be made to escape. (for a detailed discussion

0 0 2. of excess reactivity see contention T , release of fission products contention E ) In this section, we will show how some of the key safety features have been removed, c'hers c were discovered to not exist as previously thought, while still others have never existed or have been proven to be substantially inadequate.

1. positive Temperature Coefficient for Graphite. The reactor was apparently built with the assumption of a negative temperature coefficient based on the negative coefficient for water; but since the reactor is also moderated with graphite (graphite is also used as a reflector),

temperature effects on graphite must be considered. Eight years after the Hazards Analysis upon which most of Applicant's current SAR relies was written, an AEC inspector reported the following (Inspection Report 68-01): A report to the Commission by the University of Washington (letter to D.J. Skovholt from A.L. Babb, dated January 4,1976) deals with a positive graphite temperature coefficient which had been noted during operation of the University of Washington Argonaut reactor. As a result of the subject report, an effort was made during the current visit to identify possible sh W r effects relative to operation of the UCLA Argonaut reactor. Dr. Smith informed the inspector that he had received a copy of A.L. Babb's letter and that he had attempted, unsuccessfully, to measure the effect of graphite heating in the UCLA reactor. He said that preparations for the test had involved the fabrication of a graphite log, which was to be inserted adjacent to a fuel can and heated, incrementally, to determine possible reactivity effects. Smith said 'he experiment had never been perfomed because the heater wires around the graphite log persistently

            " burned out" during out-of core ?.ests. He said the problem was ornof inadequate heater wire insulation.

However, during the review of the console logbook, the inspector noted that several, three to four hour, reactor operating periods at 100kw had been performed. By reference to the console legbook data concemed with core reactivity changes as a function of time and the temperature of the water moderator, it appears that a positive graphite temperature of 0.006% k/k/oF exists. This is about c:.e-half of the coefficient measured during the University

3 of Washington experiment. Dr. Smith said that in spite of the foregoing, M intended to experimentally determine the graphite 4 temperature coefficient as soon as promising test equipment could beideveloped

p. 6 (emphasis added)

It is unclear from any of the documents availsble to us at this time whether further tests were ever conducted. The positive nature of the coefficient for graphite found at University of Washington and at UCLA is confirmed by the experience of the SRE, which was also graphite moderated. Thompson and Ieckerley, in l The Technoloay of Nuclear Reactor Safety (prepared under the auspices of the U.S. AEC, Vol. I, p. 643(1964)) indicate that during the power excursion that centributed to the partial meltdown of the SRE, the slow btt steady rise at a rate of +0.04% in a 3 minute rampt in spite of gradual control rod insertion and tha negative Doppler effect is attributed to an abnormal rise of the temperature of the moderator (graphite 7 which has a reactivity coefficient of +1.7 x 10 '/"F and perhaps also to some sodium vapor famation in partially plugged channels. (The selium void coefficient is positive...)

p. 643 (emphasis af.ded)

Note in particular that the temperature coefficient estimated for the UCLA graphite, +.006%, is greater thatn the water coefficient of

       .0048%. If the University of Washnigton coefficient is more accurate than the inspector's logbook calculations, then the difference is even more significant. In either case, the self-protection inherent in the negative temperature coefficient of the water appears to be more than offset by the positive graphite coefficient, therefore an important inherent safety feature designed into the system does not appear to exist. In addition, the void coefficients used by Applicant are calculated based on water voids and may not have taken into accoalt any factors arising from the fact that this reactor is both water and graphite moderated.

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2. Deflector Shield Renoved. The language on page 27 of the 1960 Hasards Analysis report when compared with the language of the SAR suggests that the deflector shields may have been removed.

The top of each box is closed by a plug (Figure II-8) which extends upward through the graphite which forms the base for the vertical thermal column. The upper part of the plug consists of8-3/8inchesofgraphiteontopoffourinchesoflead. The lower part of the plus consists of a flanged section which fits at the top of the fuel box, and a deflector plate. The diaphragm is used to keep water vapor from the boxes from getting into the graphite space where it might condense. In case of a power excu.75 ton of sufficient magnitude to expel water from the fuel boxes, this barrier would be easily broken by the force of the water. The deflector plate located above the diaphragm insures that water, once ejected from the box, will not find its way back. If one compares the above language with p. III/5-8, one finds that once again the SAR is a virtual copy of the 1960 HA, except in this case the mention of the deflector plate has been removed. There is lan6uage on page III/A-6, another section copied from the old HA, which discusses the " baffles" en the UCIA reactor. It would appear then that either the reference to the deflector should not have been removed from the language on page III/5-8 or that the deflector itself has been removed. B. The engineered safety features are inadecuate to trotect the tublic health and safety. The Technical Specifications included in the Application list only two Ehgineered Safety Features, a containment and a safety high level radiation monitor. Petitioner contends that there is no containment and that the safety high level radiation nonitor is inadequate. Sicne Applicant fails to list any other engineered safety features in the Technical Specifications, or anywhere else in the application, one must conclude that the facility lacks such other features. 1

1. No ai--ticht containment stracture exists for this reactor. On

p. V/1-4 of the applicatien, Applicant indicates that the " principal l

l fi physical barrier" against release of radioactivity is the fuel cladding As William Kastenberg, then UCIA reactor director, wrote to Karl Goller, Assistant Director of Operating Reactors for the NRC on November 5,1975: The UCIA reactor was desi 6 ned and built before the irradiation and diffusion of interstitial air (with Argon-41) became a recognized problem. The biological shield is stepped and reasonably well designed to eliminate streaming radiat; ion, it was not designed as a containment vessel. Kastenberg letter p. 4-5 (emphasis added) i The Applicant also described the lack of a containment vessel in the original Hazards Analysis: "Therefore, no containment vessel has been provided for the building and no airlock closures have been provided." , (HA,page18). The reactor is simply placed in a room in a building. In fact:, as i part of the reactor operation,the air pressure in the room is kept slightly negative, requiring air flow spaces under the doors and other places. '4 hen the reactor is operating these spaces must be sufficient to allow at least 14,000 cubic feet of air per minute to flow into the room as the exhaust fan pushes it out of the e::haust stack. In the event of an occurence whereby fission products were released and the room pressure went positive (likely when the exhaust stack is closed in emergencies) there would be no barrier to prevent the contemination of the surrounding areas.

2. The Hish Level Radiation Monitor system is inadecuate. Applicant has received notice of violations from the NRC for bypassing the scram circuitry and shielding the< monitor in other radiation systems, indicatir4 e.n insensitivity to the importance of these safety monitoring devices.

Further, the Applicant was cited in 1975 for failure to calibrate the reactor room radiation monitors at the required frequency. (Inspection i

(o Report 050-142/75-01,p.1) 3 Iack of Dnerrency control Systems. The Application and the other documents presently available to Petitioner do not mention any of the following emergency systems: an adequate boron-injection system, radioactivity removal cystem, emergency liquid and gaseous emissions holding tanks, EPA filters, emer6ency core cooling system, an emergency set of control blades, or spare control blade motors. Iacking such systems Applicant cannot reasonably assure that the public health and safety will not be endangered. 4 Iack of adecuate Shielding and Access Restriction in areas where the public' might be exposed to radiation. The reactor building does not include adequate shielding, particularily above the reactor, to protect persons outside of the reactor building.

a. The reactor building was constructed without a view toward shielding as it was a self-contained and separate building. Now the reacte building is surrounded by classroon buildings including directly overhead. In recognition of the lack of shielding in the reactor roof, the third floor void area is interlocked to prevent reactor operation when someone is working in that area. However, the probability of irradiation on the third floor raises a serious question about the lack of any shielding for the persons who work and attend classes above the void area.

b. The area on the third floor directly above the reactor ic fenced in as is the arw. directly above the adjacent Tokomak laboratory. However, there is a heavily used walk way between these two fenced in areas.

1 The question is raised as to how it can be dangerous enough to fence in l l

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                                                                                          'T the areas above the labs yet safe enough to allow people to walk between the fences.

c. The roof areas surrounding the reacter exhaust stack on Boelter Hall and the Math-Sciences building are readily accessible to the public.

The roof is accessible through two elevators and several unlocked doors. There are no fences restricting foot traffic, nor radiation hazard signs 't warning people on the roof. There is one locked door to this " restricted area" however, during at least one NRC inspection this door was found to be propped open: D2 ring a tour of the area by NRC inspectors it was observed that one door to the roof area had been propped open and access not adequately linited at that tine. Licensee representative indicated that this was unexplained and unusual and that the linited access control plan would be reviewed with physical Plant personnel to assure that the doors to the roof area remain locke and access will be centrolled and linited to reactor staff and physical plant naintenece personnel who are aware of the restrictions. Inspection Report 50-142/76-02p.2 5 Inadeauate or non-existent Interlock systems. Applicant has inadequate interloeks and has in the past by-passed such systens as it has. In 1968 Applicant was cited by the NRC for by-passing and shielding safety interlock systens, increasing the chance of excessive irradiation of personell or other accidents. (Inspection Report No. 50-142/69-1p. I 4) l

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6. Lack of missile shields, particularly for control blade drives. No information provided in Application indicates that any missile shields exist at all.

7. Hazard from swelling of graphite in reactor core.

As the Piqua and Hallam reactor experience shows, graphite used in reactors has a tendency to swell and crack. In this facility, control blades can stick or have trouble getting inserted, and other damage to fuel assemblies or otherparts of the core can ensue. A thorough analysis of this potential problem is lacking from the Application.

8. FUEL FAILURES. Applicant has had a history of fuel failures, particularly tie bolt failures, which raises questions about thermal stresses, warping, etc. In addition, in Application p. III/6-2 in describing who constructed fuel fon the facility, it is mentioned that A.I. did the fuel, then in parenthesis (2nd time) . Was there something wrong with the first fuel? This should be explained in the Application.

9. CONTROL BLADE PROBLEMS. Perhaps the most worrisome of these safety problems are the persistent problems the Applicant has with the control blades. They have oftsn become stuck: after earthquake simulation reported on in 1968 inspection report; inspection report 74-01 deals with a sticking control' rod drive in May 1974; and just last December an NRC inspectj7n report once against dealt with control blade problems. In addition. the control blade drive logic has malfunctioned, Under abnormal occurences in the 1975 Annual Report, the following incident is recorded:

9 1 4 7 l the reactor operator noted that the control rods were not

!                           functioning normally. This can be summarized as follows:

(1) Rod #1 would not drive out when the " rod drive up" switch , was depressed and rod #1 and, rod #2 would both drive down when the " rod drive down" switch was depressed. (2) Rod #2 i would not drive either way. (3) Rod #3 would not drive either way. -(4) Rod #4 qwuld not drive out when the " rod drive up" switch was depressed and rod #2 and rod #4 would both drive down when the " rod drive down" switch was depressed. As we have seen in the contention on reactivity, reactivity control mechanisms for this reactor in particular are essential to work without problems. The control blade problems sketched out above raise serious questions about the adequacy of the reactivity control system as a whole. CONCLUSION: Petitioner has shown that a significant number of the reactor's safety features are either inadequate or missing in entirety. The specific features mentioned are those that are known by Petitioner at this

.         stage of the proceeding, prior to discovery and prior to the establishment of a public reading room for this docket. Petitioner contends that sufficient basis has been established to support the contention that safety features generally are lacking and inadequate, but is willing to specify features not detailed above after a public reading room is set up and discovery completed.

i k t 1

 . a 10 The bases presented above demonstrate that Applicant's facility lacks basic safety features necessary to operate the reactor in a manner that will not endanger the public health andsafety. Given the gravity of this matter Applicant has a heavy burden of proof to demonstrate that the lack of adequate safety features will be overcome in the future sufficiently to allow the safe operation of the reactor. Applicant has failed to make such a demonstration, and therefore the application cannot support the issuance of a license.

I

o . 1 y , SPECIA1, !!UCLEAR I'A?DMS tlCEMSE The Applicant's equipcent, facilities, and procedures for handling and using special nuclear materials are inadequate to protect health and minimize danger to life and property and therefore cannot support the issuance of a license. Specifically,

1. Applicant does not include an application for a special materials license with their Application for an operating license. 10 CFR 70.22(b) requires that the application centain "a full description of applicant's progran for c0ntrol of an accounting for spec.a1 nuclear material which will be in Applicant's peccession under th' license, to show how conpliance with the requirements of 70 53 will be acce=plished." : either cection 22 Or section 53 have been cog lied with.

2. The little info mation in the Application that is relevant to the special nuclear materials license is inadequate and does not meet the re-quirenents of the regulations.

a. There is no W_dence in the Application that the stored spEcTal nuclear =aterials are monitored for criticality as required by 10 CFR 70 33

b. Applicant states en page five of the Application that they are applyin; for a lice.se to use h700 gr: s U-235 (irradiated), h730 grans U-235 (fresh), a.d Fu-239 and a 2 Curie, Fu-Ee neutr:n sou-ce. This des-cription does not meet the requirements of 10 CFR 70.22(h) that they cpe-cify "the name, a=ount, and specifications (including the chemical and physical fc = and, w.ere applicable, isotopic content) of tha special nuclear material..."

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1 3 The general safety, security and operating history of this Applicant suggests that the license for bomb-grade (93l$ enriched) uranium should not be granted. i

a. The quantity of U-235 requested represents 60 percent of that needed l

for an ato=ic bo=b assuming a sphere of 19 g/cm3, with 15 em natural uran- l ium reflector, according to Reviews of Modern Physics: Vol. 50,'Jo. 1 , 1

!            Part II, January 1973, page 323.
!                        The tragi-comic incident involving UCLA's first and only shipment of spent fuel en June 21 of this year further indicates tha* the Acplicant is qualified to handle neither fresh nor spent fuel. (Attachment, Valley
             'Tews, June 22,1980). S e failure to notify local officials, the inability i

to keep the ship =ent secret from =edia or public, and the failure of the i shipnent to take the appropriate route (and instead traveling and extra t 100 miles through highly populated areas), all indicate that there is a serious health threat to the public if the Applicant is licensed to handle irradiated fuel and a serious proliferation problem if they have fresh fuel. (Valley News, June 22, 1930). The above indicates that the Applicant has failed to sub=1t an application for a special mtterials license despite the certification page statement that: 1 S e applicant or any official executing this certificate on behalf of the applicant certify that these applications are prepared in confor ity wi-h Title 10, Code of Federal Regulations, Parts 50 and 70, and so sole =nly swear (or affi m) that all info mation contained herein, including any supplements attached hereto, is true and correct to the best of our knowledge a-d belief. i Given this failure it is i=poss E.e for the'3oard to issue such a i license or to conclude that the Applicant will handle special nuclear

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8 0 3 materials in such a manner as to protect the public health and to minimi::e danger to life and propert*/. i I i

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H E 1 1" M W U L4 Sf]t003 TWC00Er J i DT8Y@ifD80

                                                                                                                                      <hd not see it as a vrunus threat Dy ADAM DAWSON Background                                                                 to the puhnc. "but an> thmg deahng with a reactor is in the The seini-secret . shipment of                                                                                      'Y ""O i """id hk' '* h' U"

high level ranhoactive waste 11Cl Es nucle;r energy laboratory has been m operatwn since "*"*' from IICI.h's nuclear reactor IM Smce that Imn'. spent fuel has been stored on the sste. Foler- While Karbus said the.langrr carly Saturday annoyed I.os al appuural tras granted recently to Iransfer the material. to the puhhc was inmunal. he Angeles County health ofhrials who had not been told alunit the noint thece coi:hl be s.une lor.d move. on its way to an Idaho I' alls fa- the Nuclear llegulatory Com- ra hation contanunation if the mission and liCl.A Monday to truck carrying the spent fuel The enriched uranimn. far a ihty for cheimcal reprocessing. had an acude nt en route . nut When told of the si..pment by hnd out why hical officuns were less than needed to construct a kept in the d.u k.

the cask contanung the spent twanh. moved from the West- a Valley News relmrter, Joseph fuel lu oke tspen. wmul c.uninis about 10 a.in.. M. l'arhus, heaal of the county's lie s.nd the epiantity of en. down Wilslure lloulevard anil radiation management office, riched tiratump being shippeil **Therc is a rertain heahh h.it-onto brst the San lhega and said he should have been in-- was small enough anniut two then the Santa Monica freeways fmined anel nonnised i to conta< t gunnuts of spent biet. so that he flects the amount, not the and associated with the han- radiation levels, of the ship, ilhng of it," said Neill C. Ostran- ment. Karbus said hundreds of ship-der, manager of UC1,A's nuclear engmecring. lab. Ile added the ments of radioachve material cr sr. l os Angeles Coanty each amount of enriched uranium week although moe t of them are

                                                            ' contained 'in' the spent fuel        medicalisotopes with low levels
                                                            . plates was far less than the         of radiation. Such low radiauon amount needed to construct a          material can be huried safely, hcnnh.    ,

officials saul, in contrast to the "This reactor is under NRC high-level radioactive _ waste. shrection." Karbus said. "We " Enriched uraniinn certainly woiild look to them to advise has greater siginficance than

                                                              "*                                   normal medical isotopes,"
                                                         .       Ifnder NRC rules local offi.      Kas hus said.

coals are not required to be noti- lie noted Sou'thern California ficd of shipments of low strate. Ikhsena Co. hypasses les Ange-nic; significance. That rating re- les County and all heavily popu. Sec.1-Sun June 72,1900/VAliff NEWS

                                                                                                                                                                             < b.

W L a shios . ' Dig 'D-len el nuclear waste measured at two roentgens per tor becomes a public issue and lated areas when it trucks spent the other .vav around." liirsch hour c'. one meter. That is equal the only way to keep thmgs in fuel from its San Onofre Nucle- said. If UCLA is this careless to 17,520 roentgens a year and perspective is to be informed. ar Generatmg Station south of with security about a shipment the legal limit of exposure for San Clemente. of high. level radioactive waste Ostrander said he had been imagine how lax they might be the general public is .5 "I think the NitCshould have advised Ihe NilC wouhl contact with an incoming shipment of roentgens (or 500 millirems) per advised us about this." Karbus the necessary agencies. "I'm fuel." year. said. sorry to be told he (Karbus) Ostrander said there was fly the time the fuel plates, wasn t informed about this, weighing 35 pounds, were The spent fuel, weighing a to-plenty of campus security pres. tal of 35 pounds, consists of ura. flegion'.' NilC officials were ent during the actual loading of placed inside a cask weighmg 32.000 pounds on the back of a nium. aluminum plates 26 mehes unavev...e er comment. the material onto the truck. al. long. 2.5 mehes wide and 1 %G of Daniel : . Msch, of the anti. though once the loading had long flatbed trailer UCLA radi. ation safety officers measured an inch thick. said UCLA,s Os-nuclear mice to llridge been completed. and prior to 8.he trander. the Gap.. of he 'vas surprised vehicle's departure. no security less than .I of a millirem at one county ~ 4.4 . .i<ials wouhl was visible. meter. The plates in five bundles of not k%w nimut 4 #hipment of Iladiation from the spent fuel. Karbus said because of the 11 were the first shmment of spent me; and a ' comber of the which Ostrander sahl had been Three Mile Island nuclear acci- spent fuel from the UCLA reac-medta would. dent in Pennsylvania last year, for in the 20 years it has been in dry storage at UCLA for at operational.

      It seems to mo it should be least the past five years, was         anything that happens to a rcac-
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                        . IMFERE;T PROBLE'_S _Ei ARGC?!AUT TYPE PIACTORS Problems inherent in the design of Argonaut type reactors have been identified at other Argonaut reactor facilities, and until Applicant has I      de=cnstrated that these p.uble=s have been adequately resolved at UCLA, the Applicant cannot assume that the operation of the reactor will not be inimical to the public health and safety.

Specifically,

1. Graphite Te=rerature Coefficient: ~he UCLA reactor is graphite =oder-ated. The positive temperature coefficient of graphite increases the likelihood of a reactor runaway and decreases the effectiveness of the reactor's self-limiting features,

a. "A report to the Cc==ission by the T*niversity of 'Jashington (letter to D.J.Skovholt from A.L. Babb, dated January 4,1968)

                          &.otes] a positive graphite tesperature coefficient which Oras foundJ during operation of the University of 'dashington Argonaut reactor." ...
                        " hCLM stte=pted unsuccessfully to =sasure the affect of graphite heating in the UCLA reactor". . . .
                        "However, during the review of tne censole logbook, the inspector noted that several, three to four hour, reactor operating periods at 100 kw had been perforced. Ey reference to the console logbock and the temperature of the water moderator, it appears that a pos-itive graphite te=perature cf 0.0065 ok/t./*? exists."
                                         - '4.E.Vetter Report at inspection C0 Report
                                           ;0 50-1h2/68-1, page 6.

b. Sere d.s no in"or=ation included in the application or otherwise presently available to Petitioner which indicates that Applicent has ex-perimentally dete. ined the coefficient or made any changes in their op-eration to cc=pensate for the effect of heat en graphite.

2. Centrcl Fod Motcrs: Cther Argonaut facilitie s have experienced con-trol rod motor problems, and therefore **CLA shculd have replace =ent

1

     =otors on site.

1

a. "Further inquiry was =ade regardirg replacement regulating rod drive motors (P.eference: 11/20/72 ce=o Keppler to Spencer re-garding rod drive motor proble=s experienced at University of Florida). UCLA does not have any replacement motors."

                              - Me=o fro: G.S. Spencer to H.G. Sornburg reL M ng April 3-5, 1973 inspection at UCLA; dated: April 24, 1973                                 ,

4

b. Tnere is no infor=ation included in the Application or otherwise presently available to Petitioner which indicates that Applicant has pro-curred replace =ent motors or taken other steps to guard against motor failures.

3. Water pressure Problems: S e University of Florida has experienced water pressure proble=3 in the coolant syste= because the system is sup-plied by the city water =ain. UCLA reactor coolant system is also sup-plied from the city water supply and should have a pressure ec=pensation syste= installed.

a. "~he nuclear reactor at the University of Florida has a problem - -

the cooling system calfunctions when sc=eone flushes the toilet...

            /fowriskexperimentsrunonasecondarycoolingsyste=tiedinby l

a city water main to the toilet. Untimely flushes have caused the reactor to be shut detn five ti=es in the past three years sending students' experiments down the drain."

                              - Washington Post: August 19, 1977 The above indicates that Argonaut reactors have certain operating problems inherent in their design. However, Applicant has failed to identify these problems or their resolution in their application. Given this failure, it is impossible to conclude that the reactor will be operated in a manner which avoids these inherent proble=s and will not endanger the public health and safety.

1 l 1 I 4

1 g , SITING The site characteristics of this reactor such as population density i and soisnic activity are such, and have changed sufficiently over the perica of the first license, that the reactor cannot be operated in such 1 a manner as to assure that the public health and safety will be protected.

1. Increased toculation density. The population density inmediately surrounding the reactor and in the general vicinity have increased to such a degree it is no longer safe to operat the reactor in this facility at this site.

a. Since the reactor building was built as a self-contained structure the facility has been surrounded on all sides by new construction, including classroon facilities directly above the reactor and directly adjacent to it,

b. The main air-conditioning intake manifold for one of these classroom structures, the Math Science Building, is located less that 100 feet directly upwind of the aactor exhaust stack.

c. The student populatic.1 of the campus has increased to 30,000 during the licessed period with at least a proportionate increase in staff, employees and others.

d. More than one half million people live within a five

      - mile radius of the reactor with over two million people 1_ ting within a ten mile radius.      These population fi 6ures represent a 13% increase over the past decade, a level of increase that will probably continue for several years. (See the Application, page III/3--3)

I i

1

2. 10 CFR 100 Siting Criteria. This reactor without a containment structure and being surrounded by such a dense population can potentially expose the public to levels of radiation far in excess of the guidelines set forth'in 10 CFR 100 or by any standard necessary to protect the public health and safety,

a. The reactor has no air tight centainment structure.

b. In the Applicant's Safety Analysis report they state that the exposure to the tliyroid of a person within fifteen meters downwidd frca a reactor building leah would be 1800 rens in en eight hour period.

Application, page III/2-6. This level of enposure is sin tinos the enposure alloued by the guidelines cet forth in 10 C73100.11 (a) (1), (2), for exclucion areac and loa population cones. Given the fact a reactor leak could cccur into the adjacent buildings and their environmental centrol cystens, hundreds of people could be expoced to enceccive doces of radiation. 3 5eicnic 7ulnerability. The reactor is cited in a seicnically activo area and in the last twenty years has bec dancged by ceienia activity. Thereforo, given the other characterictica of the site, in the event of a major carthquake the reactor uculd cndanger the public healtu and safety,

a. The reactor 10 cited in an area uhere earthquake intensity can be c::pected to reach levels of intensity of 7III and higher on the ::odified

     !!ercalli Intoncity Scale. :Tero: A Guidebook to Nuclear 3eactors. 3;;. 57 . (1979).

b. The earthquake of Februar/1971 cauced encuch dana.:;e to the reactor to require chutdor. for a najor naintenance effort. 1976 Annual Report page 3.

I 1

3

c. In 1960 the reacter staff conducted an earthquake cinQlation tent. The ratults of the test as raported in .hcicar Arnlications and Tecimolo~y, 'lol. 7. Jul'y 1969, entitled "71bration Testing crA Ir.rthquake 3esponce of :uclear 2: actors: Crias 3. Snith and 3. 3. lathiecen of :!ZL.

                 "Kr.cutsix ncnths after the vibration ex;crinent routine test: indicated that one of the control blade insertion tines had incresccd. A fe:r nonths later safety blada ::o. 6 stuck in tr.c cut ;ocition during a routing pre-start check-out of the reactor control cycten.      l hen the reactor was dienantled, we discovered that lead shielding bdcks had been displaced up1cri, caucing the shaft to bind.

Vibration Testing, Icse 23-24 Tne above bases indicate that the reactor ::ac built at i time ::hsn the po;ulation surroundir.g the reactor ::ac nuch less dcnce and nuch lecs proxinately located to the reactor. The denographic changes that have cccurred over the pact t::cnty years prevcnt this reactor fren being cafely operatel over the nont twenty years.

1 R , REACTOR IS TOO OLD The m . :or in question is so old that it poses an unacceptable hazard should 2.t be -elicensed, particularly for a twent/-year period. The key equipment is already so aged as to be unreliable and other equiptent is antiquated and outdated. Because of the age of the reactor it is ve:7 difficult to obtain spare parts and key safety features required of newer facilities are lacking in this facility. he reactor is thus too old to function safely and reliably now, let alone at the berinning of the next centu:"/. Specifically,

1. The reactor was built in the late 130's by a ec=pany that shortly thereafter went out of the business of building reactors. S.e company in question (X?, which merged with 'Toit Cc=pany) now pri~.arily makes basketballs, swi= fins, and other sports equiptent. The fact that the reactor vendor is no longer in the business cf building reactors has hampered proper maintenance and significantly reduced the safety of the reactor, in large part due to the difficulty in obtaining spare parts.

2. "'he age of the reactor makes instrumentation unreliable, difficult to repair and ha-d to find spare parts for. In his Annual Report about the UC'l Nuclear Energy Lab, Professor Ivan Catton, DEL Director, wrote:

Some of the reattor instrumentatien is still w:rkable, but scre-ti=es unreliable, and is very difficult to repair due to its age and the resultant problem of obtai.ing parts (e.g. vacm=1 tubes, specialized switches, indicators, a-d =eters). J. :hch equipment is outdated and, because the Applicant dces not have l l the financial means to update it, this equipment is only likely to get worse. frofessor Catton: If the I'lL receives extra funds, the orderly updating of sensole instrumentation will proceed...

   . e 1

If money is found, our antiquated activation analysis laboratory must be modernized. It is currently about 10 years behind the state of the art. 1976 Annual Report, page 35-36.

4. "'he reactor was build before features such as containment vessels and 2:::ergency Core Cooling Syste.ms were realized to be i=portant safety features.

In a November 5,1975 letter to the Nac's Karl coller, Willica Kasteru-rs, then lab director, wrote:

                "he biological shield is stepped and reasonably well designed to eli :inate stree-ing radiation, it was not designed as a containment vessel. (emphasis added) page 4

5. '"he reactor was built before the problem of Argon 41 producticn with such reactors was recognized. 3 e Kastenberg letter, cited above:

                "he UC M reactor was designed and built before the irradiation and diffusion of interstitial air (with argon 41) became a re-cognized proble=.

4 Kastenberg letter, 11/3/75, page h.

6. Age has severely reduced the usefulness of the facility. Professor Catton's 1976 report, page 3:

The reactor is no longer new, and reactor physics pro,Jects with the UCLA reactor have becoce non-existent.

            ~he UCLA reactor is so old that it car.not be operated safely and rel:. ably.   ~5e reactor was built prior to the development of =cdern safety features and lacks these features. The existing equipment is-dangerous due to age, antiquated design and the difficulty of obtaining replacement parts. S e reactor therefore cannot be operated safely and reliably over the twenty year license period.

. o 1

                      % , SETTIC VUll ERABILI""f 2 e facility for which the license has been requested is inadequately protected from seismic activity. Further= ore, there is a basis for concern that the facility is vulnerable to seismic activity and its location is one of the most seismically active regions of the country. The reactor therefore poses a serious threat to public health and safety.

Specifically,

1. Applicant has o=itted key infe:=ation from the Application regarding seis=ic vulnerability.

a. Applicant claims on page 7 of the Application:

No structural weakness (earthquake vulnerability) has ever been identified. However, page three of the 1976 Annual Report by Professor Catten states: 3 e February 1971 earthquake gave rise to minor problems that worsened with tire and ulti=ately required a. major maintenance effort in 1972.

b. Application on page II/J-1 states:

The IJCLA Reactor has been subjected to experimental vib-ration. The results were reported by C.B. S=ith at the

               'dinter meeting of the A=erican : uclear Society, I!ove=ber, 1963, in a paper titled " Vibration Testing and Earthquake Response of Nuclear Reactors."

However, Applicant neglects to include the results of that ex-peri = ental vibration - - significant shifting of the core causi.T sticking of a control blade in the out posit a n requiring dismantlird of the core. S e paper cited in tha Application states the following: About 6 =enths after the vibration enperi=ent routine tests indicated that one of th2 control blade insertion times had increased. A few =onths later safety blade

                !o. 1 stuch in the "out" position during a. routine prestart checkout of the reactor control system. ~!    < hen

2. the rcactor was dismantled, we discovered that lead shielding bricks under the c:ntrol blade drive shaft had been displaced upward, causing the shaft to bind. page 2h. This failure is confi red by AEC inspection report y% 142/63-2, which stated that the pinning of the centrol blade was caused by shifting of leed shot inside the reactor core: I < caused by, or at least aggravated by, an experi=ent durips the previous year to dete d e the effect of eaM hquakes on reactor operations. During this experiment, the reactor i superstructure and core were subjected to relatively severe shStins. - pr.ge 7-8. Furthermore, the Application did not =ention an October 2966 shake test corresponding to "an earthquake of light to =cderate intensit/", about "a magnitude of 4" that indicated a power oscill3 tion during the shake test and also a 1963 si=ulation of effects of vibration in a fuel bundle con-fiming that such pcwer oscillations can result from fuel bundle vibrations changing fuel spacing. ("Si=ulation of Earthquake-Induced 7ibrations in a UCLA Reactor Fuel Eundle" by R.L. Eud=an, 1963; and "Si=ulation of Ea-thquake Effects on the UCI.A Reactor Using Structural Vibrators" by R.3. 7.atthiesen and C.3. Z=ith). A review of the results of all three studies - - and any others that might enst - - seems in order.

2. Failu e to Consider !!ew Data

a. The Application simply repeats the language of tae 1960 Hazards Analysis when it states that the reactor was built "according to the accu =ulated ,

wisden of the Uniform Building Cede" in effect in the 1rta "i #"a s when it was built. *4e question whether the Uniform Euilding Code then in 4 effect is sufficient for a nuclear reactor today; certainly Applicant should have reviewed the facility against current standards. In acdition, new &

         ' data (for example, the results of the l=perial Valley earthqua.e of 1979 which damaged the services building of the County despite the fact that the earthquake had a peak =agnitude of 6.6 cn the Richter scale and the

3 building was supposedly engineered to withstand an earthquake of magnitude 3." See"'Carthqrske-Resistant Buildings" by R. Serger, Science,.Febraary 1, 1950) should have been considered. It is important to note that at least two of the earthquake sirralation tests to which the UCLA reactor or fuel were subjected were only representative of an earthquake of a

       =agnitude of 4 - - several orders of =agnitude below what could occur in the area.

b. In addition, the Applicant has failed to report that it has been dis-ccvered that four UCLA buildings don't =eet earthquake standards and thr.t 4

the University architects were -- as of April 1930 -- conducting a study of buildings on a UC ca:puses to rank the according to their need for seis=le renovation, according to the UCLA Daily Eruin, April 30, 1950. Surely the rer2 L of that study should be taken into account,

c. A review of the experience of other Argonaut or test reactors during a

4

      . earthquakes was not done and would be in order.

d. The Applicati:n =erely restates the language of the 1960 Hazards Analysis regarding site seis= ology. Surely there is new information available in the twenty years since the initial repcrt was written. As I

has been shown elsewhere in these contentions, the claim that there are no wells in the area is not trae; what other largaage from that 1960 report that wcs just retyped for this application is also untrae? A thorough review is in order. i J. possible Earthcuake Hazards "ot Analyzed i

a. Possibility that in an earthquake, the supports in the void area

)' l I between the third and fifth floor of Boelter, directly above the reactor, could be da= aged, causing floors five and above to ecce crashing dcwn onto the roof of the reactor. The acceleratien of so =uch mass through two stories, landing en the roof of the reactor, could easily crash through i 1

4 that roof, =assively damaging the reactor core below. This could result in broken fuel bundles and a release of significant amounts of fission products.

b. Earthquake causing vibration of the fuel bundles, changing spacing and creating potentially hazardous power oscillation.

c. Earthquake causing control blades to stick as the coderate shak:ng did previously,

d. Earthquake shifting a " secured" experiment of significant reactivity; cr breaking open the pneumatic tabe and dispersing its contents inside the reactor; or causing pipe breaks or other leaks involving possible 4

radiation enposure.

e. Initiation of criticality in stored fuel -- fresh or spent.

;               f. Earthquake initiating a stea= explosion, disasse=bling core.

I g.. Earthquake causing significant ccre damage without upper floors crashing through the roof of the reactor.

h. other potential hazards from earthquakes that would be apparent if and when a full disclosure of relevant seis=ic studies and data and calculations are ade.

The reactor is situated in an area of major seis=ic activity; in the absence of a containment structure and in the midst of a populated building and campus and co=ranity, a thorough review of potential seis=ic problems is essential. Applicant made no such review in the Application. In fact, 4 the few statements made about seismic issues in the application were misleading at best. In the absence of a thorough review of the potential I for damage from seis=ic activity, the Applicant cannot be said to have l 1 fu'. filled its burden of de=onstrating adequate protection against such dr= age. I l

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VIBRATION TESTING AND EARTHQUAKE w n.J M.mA Whm..u. encmu RESPONSE OF NUCLEAR REACTORS

                            .                                                                          KEYWORDS: reactors, reocta j            CRAIG B. SMITII and R. B. MATTHIESEN Nuclear Ensrgy Laboratory,                                   6ro on es
        .'         Earthquake Engineering and Structxcres Laboratory,
         ,         Los Angstes, California 9003                                                        mothematics,         sensitivity bilit Y, onelysis, motion,  earth

quakes, UCL.A, ECCR, CYTc

                   Receivd November 25, 1968 Revised February 24, 1969 3

3 r4 we i m1L wC . iJTZ gi 90 reactors in strong motion earthquakes, it will b@

    .I                                                                     necessary to predict their performance witt Vibration testing of nuclear reactors is dis- studies based on simulations and analysis.

cussed as a part of the determination of the To discuss vibration testing of nuclear reactow response of such systems to earthquakes. The systems, one needs to consider the use that will

      .;            basic theory of vibration testing is presented be made of the tests. The obvious use is to detere
     \              n!ang with a comparison ofimpulse, ambient, and mine the dynamic response of key reactor sys-j             steady-state testing. Steady-state tests provide a tems. We believe that this is important, but it 19 j             method of obtaining the complete dynamic charac- also important to use the test results to check thG teristics of a system and of selectively studying validity of mathematical models of structures.!
      ,             each of the components of the system; e.g., con- There is considerable need for analytical modelf j              tainment, steam generator, pressure vessel, in- that will accurately predict the response of large
        !           strumerlation, etc. Generally, both impulse and auclear power plants to the vtbration effects 00

1 ambier.! studies do not provide as much detailed earthquakes.

1 a infortnation while being less time consuming and Much work has been done in the fields of seise i creating less interference with other operations. mology and earthquake engineering,and we believG p A series of tests performed on the UCLA re- that it is possible today to construct a "first ap4

       , , .        search reactcr, the Carolinas-Virginia Tube Re- proximation" to a complete analytical modeL Wa

q. actor, and the Experimental Cas-cooled Reactor are surveying this work and are attempting tg l: ? at Oak Ridge are used to illustrate results ob- draw it together to construct an overall model)

M tained with steady-state tests. These il ustrate Where possible, we plan to use out own experi-y the effect of the vibrations on instmmentation as mental work or the work of others to verify th; well as the response of the reactor cores, fuel model. 3l , elements, biological shielding, steam generatos s, In addition, we expect that the experimental al exhaust stacks, and the containment structures. work we have done will indicate areas, if anyi a! The tests of the UCLA reactor included tests with where nuclear power plant design .equires furthes t the reactor atfidt power. The examples illustrate r,esearch and development. Otce a complets the complexity of the soil-structure-reactor sys- analytical representation of thc earthquake-sol 1= tem while also indicating the nature of the results structure nuclear reactor system is available, iB which may le obtained with vibration tests. will be possible to study the sensitivity of tha model to variations of its parameters. Sensitivity

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                              , i Md -           a a - 4:.4 _m c.d         analyses can pinpoint areas in the system whera additional research is required or where addi=

INTRODUCTICN tional research would lead to significant improvec ments in the stability or safety of the cystem. Knowledge of the effects cf earthquakes on nu- In this paper we discuss some analytica~ clear reactor safety will be increasingly important models and several experimental techniques fo3 as more nuclear power plants are constructed in testing reactor structures. We compare the ad-seismic regions. Until the time when we have ex- vantages and disadvantages of the several testint perienced the actual behavior of large power techniques, based cn our experience in the fleid 6 NUCLEAR APPLICATIONS & TECHNO!.OGY YoL. 7 JULY 196

d Smith and .\latthiesen t'AltTilQUAKE ltLS1' OSSE OE N1lCLEAlt itEACTOltS I g "j 7 7 ~~.3 n O

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Fig.19. UCLA reactor-core area. , 4

l None were observed. About 6 months _after the graphite cobers and the top grid plate show '~ vibration experiment routine t_ests_irgd.icated that peaks, one at ~3.9 cps and another at 4.2 cps. one7f the-- tDntrol blade . insertion _ times had forced 'tibration tests reveal that the pe ,

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     '                                                                                                                                                                                                                                                   ~l stuelIin~ - ~the "out" position during a~ routine the 3.9-eps response is dae to the core i
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prestart c'hecicou~t of the reacto r control' system. When the shaking direction is switched to Ed N.,- Y' "When the reactor was dismantled, we dis- the grid plate has a sharp peak at 4.6 cps, w':i covered that lead shielding bricks under TtHe is equal to the natural frequency of the building . y O control blade drive shaft hed been displaced the east-west direction. E ypward. causine tha shaf t to bina. The lead shield Figure 25 shows another interesting aspek .d J I blocks were stacked on lead shot which had been EGCR core response. The grid plate res hiY; ' poured in the void spaces between the crachite and indicates the unstable jump. phenomenon ass h biological shield. Subsequently the lead shot has ated with a nonlinear softer.ing spring. As i been canned in steel containers, and a steel frequency of the forced vibrations increases., & shroud has been welded in place to protect the amplitude of accelerations increases uniform! dd "l 8.8" cps. At 8.90 cps, the amplitude nearly l drive shaft from interference. The response of the EGCR core (Figs. 22 bles. and then falls off at higher frequency. the forced vibration frequency is lowered. $ k[ p through 24)is interesting. The acceleration curves I (north-south shaking) for both the center of the acceleration amplitude retraces the same CT x;;' M,

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M1 NUCLEAR APPLICAT!oNS & TECl!NOLOGY VOL 7 JULN 24 41 [ i

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M , INADEQUATE FINANCIAL CUALIFICATI0 tis The Applicant does not possess, and cannot give reasonable assurance of obtaining funds sufficient to cover the costs of operating the facility over the twenty year license period plus the cost of permanently shutting the facility down and maintaining it ti a safe condition. Given this lack of assurance Applicant fails to qualify financially for an operating license.

1. Deferred Maintenance. Applicant has, in the past, neglected or postponed the repair and improvement of safety instruments and systems, due to lack of funds. These financial difficulties indicate that Applicant must make a strong showing that in the future they will obtain sufficient funds to maintain and operate the reactor,

a. On March 13, 1975, Thomas Hicks, then Director of NEL, in response to vblations cited in NRC Inspection report no. 050-142/75-01, 1

proposed to replace an exhaust fan noter and add footage to the exhaust stack. In theltech 13th letter Hicks stated:

                     "The cost of bringing the ventilation system to conformance 1-with the Technical Specifications will be substantial and beyond the means of the Nuclear Ehergy Iaboratory or the School of Ehgineering and Applied Science. The School of Ihgineering and Applied Science is currently seeking University support for the revision and hopes to accomplish the work within 6 to 9 months." (emphasis added)

March 13, 1975 Letter by Thomas Hicks to NRC j i The NRC Inspection and Ehforcement Division found this an " unacceptable ' response" in part beacuase of the delay caused by the lack of funding end convened an 31forcement conference in Walnut Creek with the Applicant. (See Memo to File by: F.A. Wenslawski, on Ehforcement Conference and Subsequent Actions, UCLA, Dcoket no. 50-%2, April 30,1975)

i

b. Applicant, despite the knowledge that the highest likely radiation exposure to the public would occur within the adjacent Math Science building has never directly measured the Argon-41 concentrations within that bi W ing. A masters thesis by Mark P. Rubin in 1976 done in connection with NEL indicates that direct measurement was not done for financial reasons:

                      "... some method of radioactive decay analysis is the only way to achieve the sensitivity necessary (to measure the Argon)... However, the required systems were not available at UCIA and building them wculd run into thousands of dollars for the ion chambers, and tens of thousands of dollars for a scintillation system. Since virtually no funding existed for this research, finances seemed to preclude the development of the required radioactive decay detection system."

Rubin: Atmospheric Dispersion of Argon-41 from the UCLA Nuclear Reactor, pg.3-4 (1976)

c. In 1975 Applicant determined that a decay tank capable of reducing the Argen-41 emmissions could be built for about $1,000.00.

These tanks have not been installed. Letter from '41111am E. Kastenberg, Reactor Director to Karl Goller, NRC; November 5,1975.pg.5-6. .

d. Applicant has acimitted that needed updating of the reactor's 4

aging equipment has suffered because of a lack of funds:

                     "If the NEL receives extra funds, the orderly updating of console instrumentation will proceed. Some of the reactor instrunentation is still workable, but sometimes unreliable, and is very difficult to repair due to its age and the resultant problem of obtaining parts.

UCIA NEL Annual Report: 1776, pg 35

3

e. The Applicant has failed to carry out other relatively inexpensive safety improvements recommended over the past several years, such as extending the stack height and installing a fence around the stack area, (See Study for UCLA done by Applied Nucleonics: Atmospheric Dispersion Analysis of Argon-41 Discharges from the UCLA-NEL Nuclear Reactor, (February,1975))which recommended that raising the stack height would significantly reduce public exposure.

The littany of failures to repair, maintain, and calibrate equipment, and the failure to conduct reasonable amounts of monitoring and systems checking set forth . in other contentions all suggest that the Applicant does not have sufficient funds at its disposal to adequately and safely operate the reactor.

2. political Funding. Applicant, because it is part of a public institution and subject to funding on a yearly basis cannot assure that it will be able to obtain sufficient funding for operation or decommission over the license period.

a. Applicant states in its application that its funding levels are

          " Subject to the availability of funds from the State of California, continuing positive recommendation by the faculty, and continuirg programmatic need . . ." Application pageI/1-1

b. The continuing programmat13 ned for the antiquated reactor is questionable. In the 1976 Annual Report Professor Catten stated:

     "The reactor is no longer new, and reactor physics projects with the UCIA reactor have become non-existent". (Annual Report 1976 pg 3); see also other contentions dealing with the antiquation of the reactor and lack of research projects.

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c. If the reactor is antiquated and lacking in utility in 1980, and will become more expensive to maintain as its age increases on what basis will the University justify appropriating money for its operation in the year 2000?

3. Decommissioning Bcpeg. A. :licant has made no provisions to assure that they will be able to obtain the funds for the $754,000.00 (1980 dollars) cost of decommissioning the reactor. An expense equal to over five times the annual University appropriation for the NEL.

The reactor operation has had financial difficulty in the past. In the future it faces increased maintenance costs associated with age and the enormous cost of d9 commissioning. These factors when balanced against the political nature of the NEL funding suggest that Applicant cannot reasonably make the financial assurances necessary to support the issuance of a twenty year license. { t 4 4

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      % FAILUR*:. TO ADEQUATELY EXAMINE MAXIFUM CREDIBLE ACCIDENT FOR THIS REACTOR Because of the placement of this reactor in a crowded building on a highly populited campus in a highly populated urban area, and because of the lack of a containment structure to effectively isolate fission products that might be released in an accident, and because the reactor is operated at times in an .astructional and training situation, it is essential that a thorough analysis be conducted of various possible scenarios by which a major accident might occur.

The only attempt in the application to do this, that of the 20-year-old xeroxed section of the Hazards Analysis from 1960 dealing with excess reactivity insertions, is the subject of thorough review in the contention regarding excess reactivity. Its inadequacies are quite clear. l No other attempt has been made in the Application to deal with the j l- questijynofthemaximumcredibleaccidentforthisfacility. Among the scenarios that should be examined thoroughly if an adequate application were prepared would be the nature of maximum potential i damage that could be caused by ! 1. earthquake i

a. damaging core, breaking open fuel, releasing fission products

2. sabotage 1

a. from inside

b. from some person or group outside I 3. reactor runatray i.

2

4. plane crash into reactor building

5. multiple failure modes

6. maximum potential-operator error

7. Other failure modes that can best be determined by Applicant or after discovery.

This reactor need not have damage of the fuel meat itself to release significant fission products. Any damage to the cladding is a potentially serious release of fission products directly to a highly populate-d area. Scenarios by which that cladding damage could occur are essential for a thorough safety analysis and for a basis for the assertion by the Applicant that the license can be granted without undue risk to the public. Given the gravity of potential results if the Applicant is not worthy of licensing and nonetheless is able to continue operating this facility, a substantial burden rests with Applicant to consider ways in which serious accidents could occur in order to provide a basis for an accurate j assessment of their likelihood and consequences. In the absence of such information, license cannot be granted without an undue threat to public health and safety. 1

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

, 3 MflSICAL SECURITY PIAN Applicant's Physical Security Plan is inadequate and fails to meet the requirements set forth in 10 CFR 50.M(c). This contention is based on the fact that, while Petitioner has been unable to examine the security plan at this point in the proceeding, the general lack of attention shown to regulatory requirements in the rest of the application, and the history of lax security practices at the facility, strongly suggest the security plan is inadequate. Until Petitioner is adnitted as an intervenor and given the opportunity to examine the plan, .htitioner cannot possibly make this contention more specific. Therefore, Petitioner requests admission on this contention, with the understanding that specific cententions regarding the security plan will be submitted at the close of discovery.

a. The events surrounding the shipment of high level wastes on June 21, 1980 is one example of lax security practices. The security during i

this operation was virtually non-existent. No security was visible between the loading and the departure. The month of the shipment was published in the application to relicense. No security car followed the truck as it left UCLA. Furthermore, the Applicant failed to notify the local authorities that the shipment was to take place. (See Article, VglLtr 6 , June 22, 1980).

b. Applicant has a history of inviting and allowing unlicensed and unqualified persons to operate the reactor. (See contention on man 36erial and administrative controls for detailed discussion).

a

= . 1 g DfET,DICY RESPONSE PIAN Petitioner contends that Applicant's Daergency Response Plan is insufficient to demonstrate that the plan provides reasonable assurance that appropriate measures can and will be taken in the event of an emergency to protect public health and safety and prevent damage to pzuperty. Given this insufficiency Applicant's Daergency Response cannot support the issuance of an operating license.

1. The prohibition in the plan that no non-university individuals be contacted until instructions to do so come from Campus police entails an unnecessary and unreasonable delay in placing non-University emergency response personnel on alert. Plan a 1.2.12-14

2. The requirement in the Plan that evacuation of Boelter Hall and Math Sciences addition be cleared through the vice-chancellor's office entails an unreasonable and unnecessar/ delay in evacuating these facilities since both facilities are directly adjacent to or surrounding thereactorfacilityandhaveair-conditioning /heatingsystemsthatcould be contaninated inmediately. Plan 5 1.2.13 3 The Plan does not adequately provide for alternative personnel with the authority to make initial evacuation decisions in the event that the vice-chancellor's office is unable to respond.

4. The Plan does not provide for alternative personnel to carry out the role of the Health Ihysicist, as general director and supertisor of the emergency response, in the event that the Health Physicist is unavailable, i

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2. 5 Applicant does not have adequate measurirg devices to accurately determine the extent and seriousness of an accident .

a. (See contention on inadequate monitoring systems).

6. The Plan has no provisions indicatirs that there is an evacuation plan for the entire campus which could be successfully implemented in the svent of an emergency.

7. The Plan does not provide for any emergency centars other than the UCIA Medical Center, despite the fact that the Medical Center is no more than one quarter of a mile away from the reactor. In the event that there was a serious accident and the wind was blowing in the direction of the Medical Center, it would be unusable as an emergency center.

8. The Plan lists available equipment and the locations at which such equipment is available. However, the Plan fails to indicate which equipment and in what quantities it is available for each location.

Plan EE 2.1,2.2

9. The Plan will only be effective in the event of an acutal emergency if the annual trahing excercises and drills set forth in the Plan have actually been carried out. Petitioner wishes to reserve the d

r16ht to further specific contentions in this area until the close of discovery, because they have no way of gathering such information at this point in the proceedirgs.

o e g SAFEGUARDS CNTINGENCY PIRI Petitioner contends that the Applicant has not submitted a safe-Suards contingency plan as part of their applicatien and therefore the Plan is inadequate and cannot support the issuance of a license. The plan must at a mini- include, the four factors set forth in 10 CFR 50 34(e); Backgrcund, Generic Planning Base, License Planning Base, and Responsibility Matrix. None of these items appear in the application. If the Applicant contends that the safeguards contingency plan requirements have been met in the contents of the physical security plan the Betitioner must centend also that the plan is inadequate to protect the public health and safety. This contention is based on the fact that although Petititioner has not had an opportunity at this stage 6f the proceedings to examine the physical cecurity plan, the generally inadequate nature of the I application and the past history of lax security practices suggest that the plan will also be inadequate. If the plan is incidued in the Physical Security Plan, Petititoner contends that the Board must admit this centention for the purpose of discovery with the understanding that Petitioner must submit specific cententions regarding the safeguani.s con-tingency plan at the close of discovery.

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The Tech.ical Specifications included in the Application contain provisions which unacceptably reduce safety standards and pore a threat to public health and safety; numerous substantial changes have been

     =ade in the Technical Specifications without so describing in the intro-duction.

1. Change in excess reactivity limits he current Technic a1 : Specifications limit excess reactivity to 2 3 percent A%/.:; the new Technical Specifications describe the limit as 03 54, a.oparentlyusinga/of.0065 '. .e Hazards Analysisgivea8of.007h. Elsewhere in the Application 13 is described as .0063 and .0070. If any of these thrse other $'g are orrect, the trans-position of excess reactivity from 2 3 percent to 33 5h is an unapproved increase in reactivit/. Furthur, the Technical 3pecifications do not say at what temperatu e -.- 32*F Or room temperature -- the new 03 5h limit is defined.

2. ?*ev definiticn far "'_ rual" in celibratien ree;2 ire .ents: The old Tech.ical

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Removal of requirerent that ALJA be net. The old Technical Specifications require the facility to meet AL;Ja standards. "'he new Technical Spec-ifications have removed that larguage. Furthennore, the discussion about AL/OA 10 replest with er onecus conclusions and statenents (see contentions on emissions for details); in particular, Applicant cannot take credit for meetira ALARA now with plcns for somethir4 Applicant intends to do later, (i.e. the installation of decay tanks., particularly since this installation has been tied to an increase in operating time in the Applicant's statenents and thus to an increase in the production of Argon. The references to 500 cr are irrelevant to a discussion of ALOa; 500 =r is :GD, AL!Pa is cenerally defined as one percent of !?D.

5. Fenoval of srecifica+1onc regarding height of exhaust stack, flew rste cut of exhaust stack and no mention of restricted area on roof. Se cover letter to the Application by Dr. 'ialter 'iegst of UCLA states that "This application centains only minor changes (listed in the forward to appendix

5) from the original application." n at forward states (page v/i):

The Technical Specifications contained in this appendix, e= body the 1 earlier Technical Specifications (of 1971 as amended in 1976), in revised fonnat and expanded content. 'dith four exceptions noted below, no attempt has been made to alter the content and pro-visions of the earlier Technical Specifications, and any other discrepaa.cies should be interpreted as t/pographical errors or editorial deficiencies.

        *iithout attenpting to judge the cause of the alterations of the Technical Specifications -- t.hether by attempts or by typographical or editorial err 0r -- nonetheless significant changes have been .ade that re not repor':ed in that fonrard.

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M.PROCEEURAL CONTENTIONS

1. Censolidation of proposed changes to the reactor facility with the licensing process.

Applicant has made it clear in the Application, in newspaper articlas, and elsewhere that it wishes to increase operating hours and power. In fact, Applicant has stated that the proposal to increase power will be submitted to the NRC by Sept ember 1, 1980. Any significant change in the operating hours or power will have a major impact on the I;ssues and questions before the Board in the licensing proceeding. Consequently, exlusion of imminent changes from the process will undemine the legitimacy of the proceeding, unduly burden the Board with subsequent duplicative hearings and work against the interest of public participation in the licensing process. Therefore, Petitioner contends that the proposed changee in power and hours be consolidated with this relicensing process.

2. A twenty year license period is excessive considering the present age condition and design of this reactor. Petitioner, contends that the license period be shortened, j

kVysY ff86 ff TATLf 1 SIG:.F.3 l l M.icl Eirsch 1637 Outler 2trnet Los Angeln, CA cC025 hecident: Canpaa Cc =ittee To 2 ridge the Jap

e , , DECLARATIQi 0F SERVICE 3Y MAIL On the AI day of Au m;/' 1980, I have served copies of the foregoing COMMITTEE TO BRIDGE THE GAP'S SUFFLEMENTAL CNTENTIONS TO FLTITIm FOR LEAVE TO INTERVENE, by mailing them through the United States mails, first class posta6e prepaid, on each of the following: Elizabeth S. Bowers, Esq. Dr. Walter F. Wegst U.S. Nuclear Regulatory Commission Office of Research and Occupational Safety. Atomic Safety and Licensing Boari University of California, Ios Angeles Washington, D.C. 20555 405 Hilgard Avenue los Angeles, CA 90024 Dr. Emmeth A. Inebke U.S. Nuclear Regulatory Commission Mr. James W. Hobson Atomic Safety and Licensing Board Vice Chancellor Washington, D.C. 20555 University of California, Ios Angeles 405 Hilgard Avenue Dr. Oscar H. Paris los Angeles, CA 90024 U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Mr. Neil Ostrander Washington, D. C. 20555 Nuclear Ehergy Lab, Boelter Fall University of California, Ios Angeles Office of the Executive Le6a1 Director Ios Angeles, CA 90024 Attention: Mr. Joseph Gray U.S. Nuclear Regulatory Commission Ms. Majorie J. Woolman Washington, D.C. 20555 Secretary of the Regents 689 University Hall

        !!r. Jin 11111er                                      Berkeley,   CA   94720 U.S. Nuclear Regulatory Comnission Washington,      D.C. 20555                        Ms. Christine Helwick, Esq.

590 University Hall Mr. Hal Bernard Berkeley, CA 94720 U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. Roger Holt, Esq. Deputy City Attorney Docketing and Service Section 200 N. Main Street Office of the Secretary City Hall East U.S. Nuclear Regulatory Commission Ios Angeles, CA 90012 Washington, D. C. 20555 Dated: f/A-/ru sy: - v,r

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ML19331D906: Difference between revisions

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Dear Mr. Douglas:

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ML19331D906: Difference between revisions

Latest revision as of 22:21, 15 March 2020

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Dear Mr. Douglas:

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