ML20138M905

Decommissioning Plan:Phase I:Dismantlement & Radiological Assessment of Ucla Argonaut Reactor Facility

ML20138M905
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Site:	05000142
Issue date:	10/31/1985
From:	CALIFORNIA, UNIV. OF, LOS ANGELES, CA
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ML20138M891	List: ML20138M886 ML20138M905
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NUDOCS 8511040475
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DECOMMISSIONING PLAN =

Phase I: Dismantlement and Radiological Assessr.:ent of the UCLA Argonaut Reactor Facility License R-71 Docket 50-142

[ October 1985 School of Engineering and Applied Science University of California, Los Angeles 8511040475 851029 PDR ADOCK 05000142 P PDR

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Table of Contents Section page

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1. 0 Plan Background and Management 1 1.1 Summary Description

1. 2 Facility Doerating History 2 i

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1. 4 3 Current Radiological Status of Facility 3 Decomissioning Alternative 4

1. 5 Decommissioning Organization and Responsibilities

1. 6 4 Regulations, Regulatory Guides, and Standards 5

1. 7 Training and Qualifications 6

2. 0 Occupational and Radiation Protection Programs 6 2.1 Radiation Protection Program 7

2. 2 Industrial Safety and Hygiene Program

2. 3 Contractor Assistance 8

2. 4 Cost Estimate and Funding 8 8

3. 0 Dismantlement Tasks and Schedules 3.1 Tasks G

3. 2 Schedule 8

3. 3 Task Analyses 9

3. 4 Safe Storage 9 12

4. 0 Safeguards and Physical Security 1 12

5. 0 Radiological Accident Analyses 12

6. 0 6.1 Radioactive Fuel DisposalMaterials and Waste Manacement 12

6. 2 12 Radioactive Waste Processing 13

7. 0 Technical and Environmental Soecifications 14

8. 0 Proposed Termination Radiation Survey Plan 14 Appendix A:

License and Technical Specifications (Requests incorporated by reference)

Appendix B:

Preliminary Radiation Survey (Radiological Status Assessment)

Appendix C: UCLA Dismantlement Experience I

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I List of Tables Table Page

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P-1: Radionuclide Inventory 3 P-2: Decommissioning Organi:ation 5 P-3: Regulations. Guidance, and Standards 6 P-4: Radiation Monitoring Equipment 7 P-5: Task Identification 8 P-6: Non-Essential Equipment 10 P-7: Estimated Radioactive Waste 13 P-8: Essential Eauipment 13 s

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DECOMMISSIONING PLAN Phase I: Dismantlement and Radiological Assessment of the UCLA Argonaut Reactor Facility

1. 8 Plan Background and Management phase I of the Decommis.sioning Plan (DP) pertains to disman-tiement of the nuclear research reactor located in Boelter Hall on the campus of the University of California at Los Angeles (UCLA). The objective of Phase I is to define the radiological status of the facility oy gaining access to the interior of the biological shield and assessing the distribution of neutron in-duced activity in the foundation, side walls, and removable shield blocks. The work is to include surveys of the fuel stor-age oits, the floor drains, and the decontamination facilities.

The format of the Plan follows the outline of the Standardization and Special Projects Branch of the U. S. Nuclear Regulatory Commission (October 15, 1984).

The scope of dismantlement is to include removal and dispo-sal of the reactor core-reflector, graphite thermal column, shield tank, and certain peripheral equipment. Phase I is to culminate with a report discussing final decommissioning alterna-tives, DECON and SAFSTOR. This report is to provide the basis for selecting the ultimate Decommissioning Plan to be implemented in Phase II.

1.1 Suggary_Descriptipp The UCLA Argonaut reactor wac water-cooled, water-moderated, and graphi te -reflect ed. It was operated from October 1960 until it was shut down in January 1984. Operations were by the Nuclear Energv Laboratory (the NEL) within UCLA's School of and Aoplied Science. Engineering It was operated to provide student instruc-tion and in support of research.

The first license period extended from 1960 to March 30, 1970. Technical Specifications were then added and the licenso was extended to March 30, 1980. A timely application for renewal was filed in February of 1980, extended pending NRC action.

and the license was automatically In June 1984, University informed the NRC that University intended to withdraw the renewal applica-tion and decommission the reactor.

The most recent licensing action (Amendment 14) deleted the license to oossess Special Nuclear Material and the attendant recuirement for a Physical Security Plan. It also eliminated the Operator Training requirements of 10 CFR Part 55. A request for related changes to the Technical Specifications has been submitted to the Nuclear Regulatory Commission.

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l Tne oreferreo Decommission 1na Plan (DP) is DECON. but SAF-STOR a

is regarced as a possible alternative mode to be elected if radiation ticnal circumstances.

survey at the close of Phase I discloses any excep-The Phase I project will retain the con-crete shield blocks to provide the crimary barrier for SAFSTOR until tion.

such time as that mode is rejected as a logical continua- ,

Dismantlement is to include removal of non-essential exter-nal eouipment, unstacking removable concrete blocks, core contents removal of (generalized to include the thermal column and at least part of the shielc tank). packaging of radioactive mate-rials for transport, and disposal of those materials.

the concrete The assayasof residual activity will include core sampling of well as wipe tests of the fuel storage pits, drain lines, and decontamination facilities. The work is to be done in sufficient detail to characterire the decontamination l

requirements and the costs attendant to final decommissioning.

The cost of completing the phase I dismantlement mated to be approximately $65.000 exclusive of internal is esti- i adminis-trative and supervisory costs.

UCLA expects that the unstacking and materialc packaging of core can be accomplished in two to four weebs unless com-clications arise in characteri:Ing the materials as thev are removed. The core sampling, analyses, report preparation, and precaration of the Final Decommissioning Plan are to be accom-l olished in the subsecuent six to eight weeks.

j The principal items subject to quality assurance are: rad-i

' lation worker dosimetry; worker environmental protection; control of air-borne radioactivity and soreadable contamination; and com-4 ollance with packaging, transport, and burial regulations.

i UCLA experience in major core maintenance (dismantlement and reconstruction) is described in Apoendix C. That experience f

which included fuel handling and core entries within three 3

after shut down, yielded exposures of 35 to 45 man-rem. The weeks fuel is cone and the reactor has been shut down since January, lo84.

i LCLA does not exoect total exposures to exceed 10 man-rem.

12 E2Cill%2 Opergt199_distgcy The reactor operated at a maximum power of 100 kw (thermal ) ,

with a 24 year time-average cower of about 2. 5 kw(t). The corresponding time-average thermal flux near core center is about 3.75 E10 per see per sq cm. The reactor last operated in January of 1984.

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If one knew the concentration of trace elements such as cobalt in steel and stainless steel, eurooium in graphite and concrete, and silver i r. lead. It might be*possible to theorett-cally calculate the discribution of activation products in those naterials. However, those concentrations are unknown, and the flux has never been mapped throughout the core-reflector region.

UCLA favors the use of a radiation survey to define the cecontamination requirements; i . e. the volume of concrete to be removed, the status of the fuel pits, the drain lines, and the decontamination facilities.

There are no known radioactive sollis that might aggravate the decontamination and deconmissioning. tiowever, the possibility of activation by neutrons streaming through the control blade shrouds, the horizontal pipeways of the primary water lines. and down the vertical crain line below the reactor core center is reccgnized.

1. 2 G9CCeDt_Bgdiglggicel_Stetyg_gf_Eacilliy Samoles of graphite, lead, and concrete have been taken from the reactor core-reflector region. The findings of this work are cescribed in Appendix B.

There are no known gaseous radionuclides at the site except for those which occur naturally.

There are several gallons of water in the sumo and some sludge resulting from the wet core sampling of graphite. Eu-152 may be present.

The principal racioisotopes in the solid materials are to aue trace elements not normally specified i r. describing the ccmoosition of materials. Mn-54, Fe-55, Co-60, and 2n-65 are expected in tne metallic core parts. The principal products

centified in the graphite are Eu-152, Eu-154, and Co-60. Tne lead is known to contain Ag-108 and Ag-lle. The major radionuc-

1. des found in the magnetite concrete are Eu-152, Co-60, Eu-154, Mn-54 and Cs-134. The first two of these account for about 94%

of the neutron induced activity in the concrete.

The inventory of radioactive material in the facility is estimated in Table P-1.

Table P-1: Radionuclide Inventory Metallic Components '

l. 5 Ci Grachite ( 4.2 Ci Lead ( 0.1 Ci Concrete

• 1.5 Ci 3

The quantity of metallic components near the core center is small relative to the graohite and lead.

exoected to be more intensely radioactive,Thus, although they are may the total inventory be no more than that in all of the otner materials combined.

The graphite and lead estimates are based upon samples taken near the core center, and the same concentration in those materials was conservatively assumed (high-side estimate) to prevail ipde-cendently of where they were located in the reactor.

The observed incuted activity in the concrete was estimated using the 2.2 inch attenuation length calculated in Appendix B at B.2 and taken perpendicular to an area of 240 sq ft at a surface concentration of 377 nano-Ci per gram. The assumed area is 2.4 times the area of the four faces cube that are adjacent to the concrete.

of the five-foot core-reflector from edge The factor of 2.4 arises and corner effects which engage a that which is projected perpindicularly to the faces of a cube.

larger volume than

1. 4 Dgegmmiggignigg_Altgenative The purpose of the Phase I program is to define the feast-Dility, environmental acceptability, and cost, of DECON versus SAFSTOR. The interest in SAFSTOR follows from the fact that a seven story structure and rests upon the reactor room foundations.

tne cecommissioning oeyond the reactor facility.

activities may have implications that go

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Deggmmissigning_Organizatign_and_Rggggnsibilitigg Many divisions of UCLA may be peripherally involved in the cismantlement work. The principal organizational lines that will be directly involved are shown in Table P-2.

UCLA will provide a manacer and a Health Physicist at the site whenever work is in progress.

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Table P-2: Dismantling Oigani:ation IChancellori I I

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i i l i 1 Executive i IAdministrative i IVice Chancellori IVice Chancellori I I I i IDean, School ofi 1 Assistant i i Engineering __I IVice Chancellori i l i I I Director I i Director i f NEL i i OROS *** i i

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• 1 1 Manager i I Radiation I i NEL i l _Sa fet y_Of f _i_ cer_1 I

I l_1 Assistant ! I Health l I l__danas9C-_1 l_Pbysisistisl- l i

I l_1 Temporary l l Techs I

• • • For the dismantling operations, the manager of the NEL will take direction from the Director of the Office of Research & Occupational Safety (OROS).

1. 6 Begylgtigns _Begglatgty_Gyldest_and_

t Standards Dismantlement, decontamination, and decommissioning will be governed by the applicable Federal and State regulations, regula-tcry guides, and Standards. Tnese include the following:

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Table p-3: Regulations. Guidance, and Standards Radiation 10 CFR 20 protection Standards Protection Reg Guide 1.86 Decon Standards And Surveys NUREG 2082 Termination Survey Worker Title 8. CAC (Cal OSHA) Worser Health & Safety Safety Environ- EPA-NEpA (40 CFR) Environ Impact Statement mental NUREG 0586 " " " "

Transport 10 CFR 71 Packagina & Transport of 49 CFR (DOT) packaging & Transpcrt Radioactive Material Most UCLA employees, including some employees working in areas adjacent to the NEL, are under 17 CAC and the requirements of 1335-70 (July 1984). This Plan pertains to the NEL site a rid to the federal regulations governing the packaging, transport, and burial of radioactive materials.

1. 7 Training _and_Gualifications UCLA will use experienced radiation worker-technicians, and will provide orientat ion and training for the UCLA facility.

This is to include a brief review of radiation properties, han-allng of radioactive materials, personnel dose and contamination control, applicable permissible doses, radiat ion wor's permits, and an introduction to the physical structure of the UCLA Argo- ,

naut reactor.

2. 0 Occupational and Radiation Protection Programs The Director of the Office of Research and Occupational Safety (OROS) is resoonsible for all matters relating to occupa-tional safety at UCLA, and he will be the Director of the disman-tlement work. The Radiation Safety Officer (RSO) who heads the campus Radiation Safety Office, is directly responsible to the OROS Director. A qualified Health physicist reporting to the Radiation Safety Officer will be present (full-time) to assure that appropriate cor.tamination control procedures are used; to monitor all materials packaged for disposal or released for other use: to prepare appropriate records and shipping papers; and to generally assure that all radiation exposures are maintained ALARA.

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2.1 Radiat ion protect ion prograrn The radiation protection program for dismantling and decom-missioning will be an extension of the supervisory role which the Radiation Safety Office normally exercises over the use of radio-isotopes at UCLA. The practices will conform with the procedures specified in the UCLA Radiation Protection Manual.

All workers will wear personnel dosimeter film badget and accket dosimeter ionization chambers. The ionization chambers will be read and doses recorded daily. All workers will be given pre- and post- dismantlement whole body counts. Based upon pres-ently available data, extremity dosameters will not be required, but should that Judgment prove wrong such dosimeters are avail-able.

The radiation monitoring eaulpment normally available at the 1 reactor site is shown in Table p-4. The Radiation Safety Office will provide back-up emergency instrumentation and services if necessary. The emergency instruments will not necessarily duplicate any of the instruments identified in the Table.

Table p-4: Radiation Monitoring Equipment Portable 1 Ludlum Model 3 with shielded pancake probe Survey Model 44-40 and unshielded probe 44-9 4 Meters 1 Ludlum Model 3 with pancake GM probe 1 Ludlum Model 14C with shielded pancake probe Model 44-40 l 1 Ludlum Model 14C with side window GM probe 1 Eber-line RO-2 ionization chamber 4

1 Eberline alpha survey meter 1 Teletector Model 6112B, 0-1000 R/hr (uust be operated in horizontal plane)

Wipe I Counters 4 Technical Associates Multiscalers with thin end-window GM tubes i

1 Nucleus Multiscaler with thin end-window GM tube

• Gamma ND-66 (4096 channel) with 110 cu em GeLi crystal i Spectroscoce (princeton Gamma-Tec)

, particulate Stack Effluent Monitor, fixed filters, changed Monitors and counted weekly I

High Volume Air Sampler personnel Film Badges by Landauer Monitors pocket Dosimeters plus charger and case 6 0- 200 mr I

2 0- 500 mr 2 0-1500 mr i Hand & Foot Counter, HMF2 7

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22 10d ug;nig1_ggfgly_gpo bygiggg_Prggdag The Office of Research and Occuoational Safety (OROS) in-cludes specialists in toxic materials. electrical safety, macnine safety, and fire protection. The Director of that Office has been a physicist and nuclear engineer and is a Certified Health Physicist. He will direct the dismantlement operations.

Some operations will involve the use of half-face dust respirators. These respirators will be fitted and leak tested by the Industrial Hygienist of the OROS staff. Only NIOSH/MEH4 approved respirators wil' be used and air sampling will be done to determine the extent of any air borne contamination.

2. 3 Qge.iractgr_Asg15t ance The dismantlement will be done by temporary technicians under the direction of the UCLA staff. Techn3 clans will receive training as described in paragraph 1. 7 and will be provided with scotective garments. safety shoes, safety glasses, anc respira-tces as necessary. UCLA olans to use a consultant for the con-crete core sampling.

2. 4 Qgst _Es t i ma t g_ a nd_Fu ngi ng UCLA estinates the cost of dismantlement to range from 150.000 to 165,000 decenalng upon the cost of disposing of the graphite rec 1pient).

and lead (burial versus transfer to a licensed

3. 0 Dismantlement Tasks and Schedules 3.1 Tagds The major tasks of Pnase I are identified in Taple p-5.

Table P-5: Task Identification

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1 Planning and Review 2 Mobilization 2 Removal of External Equipment 4 Disassemble Reactor Core 5 Package Materials for Transport E Transport Materials from Site 7 Assessment of Radiological Status

a. Biological shield (concrete coring)

b. Fuel storage pits

c. Floor drain lines

d. Decontamination facilities 8 Prepare Report and Final Decommissioning Plan 8

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3. 2 Schedule UCLA expects that the unstacking and packaging of core materials can be accomplished in two to four weeks unless com-plications arise in characterizing the materials as they are removed. The core sampling, analyses, report preparation, and preparation of the Final Decommissioning Plan are to be accom-plished in the subsequent six to eight weeks.

Schedules are subject to unforeseeable delays due to approv-als and/or other events beyond the control of UCLA.

3. 3 Task _ Analyses

3.3.1 planning

UCLA has been actively engaged in planning. 3rocurement of special shipping containers, and of garments for workers has been initiated, and a source of experienced radiation worker /techni-j clans has been identified. The general unstacking procedure is largely dictated by the geometry of an Argonaut reactor, i . e.

must work one from the top and thermal column face toward the core center.

3. 2. 2 Mobilization:

and UCLA A source of raciat ion worker technicians has been Ident i fied expects to use the first three scheduled work days in orientation, training, garment fitting, and bioassays. A reasonable number of shipping containers will be on the site. No scecial tools or equipment will be required other than a heavy

duty fork truck for moving containers.

Such fork trucks can be rented locally.

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3.3.3 Removal of External Eauipment:

l peripheral reactor equipment, external to the core and nor.-

1 essential to the prosecution of decommissioning, is to be sur-veyed and classified as either radioactive or non-radioactive.

i Non-radioactive material is to be retcoved from active material (if any! is to be either decontaminated or pack-the site. Radio-aged and chipped to a licensed disposal site. Non-essential peripheral equipment is listed in Table p-6.

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Table P-6: Non-Essential Equipment

1. The dump tank and internal heat exchanger.

2. Primary water circulation system (except for primary pump) including flow meters, control valves, and de-minerali:er circuit components.

3. Secondary water system.

4. Air lines and gas vent lines.

5. Shield tank water curification system.

6. The shield tank (non-essential, but physical removal is impractical prior to core disassembly).

Items 1 through 4 are located in the process pit. Air and water supply lines are to be removed to a convenient shut off valve location.

3.3.4 Disassembly of the Reactor Core:

Disassembly involves the unstacking of concrete blocks, lead, and graphite, the unbolting of flanges and control rod mechanisms, and the cutting of tubing and piping to convenient packaging lengths. The shield tank is to be cut flush to the top of the biological shield, and the well covered with temporary decking. This will provide more work space on the reactor top and will facilitate removal of the west lead wall which resides in the shield tank. The principal embedded component in the core (other than pipina and drain lines) is the control blade support structure. It is built of five-inch, 6.7 lb channel, and it will #

not be removed unless it interferes with the core sampling work to follow. ,

UCLA experience with core disassemoly is described in Appen-cix C. As the circumstances of those disassemblies were quite cifferent from the present circumstances, the radiation exposures of that experience are not applicable to the present conditions.

3.3.5 package Materials for Transport:

This task is to proceed concurrently with core disassembly.

i The materials are to be surveyed and catagorized in sufficient detail to provide e shipper's descript ion of the materials and, when applicable, to satisfy burial site requirements. It should be noted that large variations in specific activity are to be expected because of the spatial distribution of the reactor flux.

Where appropriate the self-shielding properties of these mate-rials will be used to reduce the surface radiation level of packages. The available information characterizing the radioiso-topes in these materials will be found in Appendix B.

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1 Removal of metallic parts (fuel boxes, control blade system components, piping, etc) is to include removal of the protruding portions of embedded pioing where such piping can be removed by non-abrasive cutting (shearing or sawing). Torch cutting is not contemplated in phase I.

3.3.6 Transoort Materials from Site Most, if not all, of the external equipment is expected to be non-radioactive. Possibly some of the graphite from the thermal column will also be non-radioactive. Such materials will De disposed of as non-radioactive material.

The metallic parts are expected to be the most radioactive and the most likely to interfere with the radiation survey attendant to defining the radiological status of the facility.

These materials are to be promptly shipped to burial at a licen-sed burial site.

The graphite and lead are expected to be of low specific activity and some may be non-radioactive. The cost of disposal of those materials has been estimated to be about $42.50 per cubic foot (burial cost included).

3.3.7 Radiological Assessment Survey The Survey is to assess the radioisotopic composition (cualitative and cuantitative) in the concrete biological shield and embedments, and to determine the extent and kind of radio-activity (if any) in the drains, the fuel storage oits, and other facilities as necessary to estimate the reauirements and cost of decontamination and ultimate decommissioning.

. 3.3.8 Report:

At completion of Task 7, a recort is to be prepared describ-ing the cost of decontamination and decommissioning the reactor.

Tne report is to describe soecific decontamination reautrements includino the extent (area, volume, and type) of material to be removed. The report is to consider both DECON and SAFSTOR modes in the context of satisfying the regulatory requirements of those alternatives.

The report is to be used by UCLA to provide the basis for selecting the mode of decommissioning, and for soliciting bids for the DECON mode of decommissioning.

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The report is to include an estimate of the cost of com-cletion of DECON and imolementation of SAFSTOR. Cost is to include removal, t rar.s oor t , and burial of radioactive material, and cost of the termination survey. The estimated cost of SAF-STOR is to be based upon the assumption of current regulations and dollars; and is to include initial cost, annual cost, and termination cost.

2. 4 Safe _Stgcaqq A crolongea perloc of safe storage would be invoked only under exceptional circumstances. The provisions here certain to eitheri (1) an interim SAFSTOR Deriod to adjust to new and unex-pected findings; or (2) a long term SAFSTOR if the DECON continu-ation would imply structural damage to the building or unacceot-able environmental effects.

4. 0 Safeguards and Physical Security There is no fuel on the site and there are no 10 CFR 73 Safeguards or physical Security requirements. Normal industrial security measures will be retained. The physical security system may be used in connection with the SAFSTOR mode if that becomes necessary.

5. 0 Radiological Accident Analyses With no fuel on the site, there is no significant potential for radiological accidents that might affect the public. The potential for worker over-exposures is always implicitly present WhEn handling radioactive materials, but this is a matter to De managed by a combinat.cn of worker training and radiation oro-tection practices. See sections 1.7 and 2.1.

6. 0 Radioactive Materials and Waste Management There are no gaseous radioisotopes present other than those which occur naturally. A small amount of licuid waste (water) will be generated in concrete curing and tool decontamination.

The vast bulk of radioactive material will be the solid compo-nents removed from the core. Those materials will be packaged and removed from the site in accordance with all applicable regu-lations.

6.1 Fugl_Disgggal Not apolicable, there is no fuel on the site.

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6. 2 Radioactive _ Waste _P t ocessing Wet core sampling will generate some contaminated waste water. If cross-contamination problems are not prohibitive, the volume of such water may be reduced by using a settling system to recover recycle water. Residual sludge will be solidified and treated as solid waste.

j Past UCLA experience indicates that water for tool decontam-ination is used in ouite small amounts, and the waste water is generally within the limits of 10 CFR 20 for discharge to the sanitary sewer. When outside those limits, a modest dilution will suffice to bring the water within prescribed limits.

The solid waste consists of three distinct types of mate-rial: metallic components (aluminum, magnesium, cadmium, struc-tural steel, and stainless steel); graphite; and lead. Esti-mates of the volumes and masses of these materials are summarized in Table P-7.

Table p-7: Estimated Radioactive Waste Cu Ft Lbs

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Metallic Comuonents 26.5 1325 Graphite (reflector) 120 12,000 Graphite (thermal column) 86 8,600 Lead bricks 22.4 15,900 Lead sheet 16.7 11,860 Lead shot 2. 7 1,910

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TOTAL 274 61,600 The total volume of radioactive material is not well known.

In of Table P-7 it is assumed that the metallic components consist the known core parts (780 lbs) and 545 lbs representing one-fourth of the shield tank mass. For the metallic parts, the volume per cu ft.

is based upon an assumed packaging density of about 50 lbs The graphite and lead masses are based upon volume estimates converted to mass using intrinsic (maximum) densities.

The following eauipment is considered to be essential:

Table P-8: Essential Equipment

1. Sump oump

2. Holding tanks

3. Decontamination facilities

4. Primary Dump

5. Ten ton bridge crane

6. Ladders, stairways, walkways, platforms

7. Ventilation system 13

Items 1 through 4 are potentially useful to the mana gerae nt of liquids generated in the Phase I work and are to be retained for similar service in the Phase II program. The retraining items are regarded as assets to decommissioning and to the subsequent use of the building.

7. 0 Technical and Environmental Specifications The total volume of radioactive material estimated in Table P-7 is less than 20% of the radioactive material annually gener-ated by UCLA and is well within UCLA's capacity to manage. The Technical Specifications; the project organization; adherence to the applicable federal and state regulations governing worker exposure, packing, transport, and burial of radioactive mater-lais; all contribute to assurance that dismantlement can be accomplished without significant impact upon the health and safe-ty of the public.

8. 0 Proposed Termination Radiation Survey Plan This is not a final Decommissio..Ina Plan and a Termination ,

Radiation Survey will not be undertaken.

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l APPENDIX A LICENSE AND TECHNICAL SPECIFICATIONS (Amendment Requests) for i

Docket 50-142 j License R-71 A request for License Amendment 14 was submitted to the NRC by letter of Wegst (UCLA) to Denton dated August 30, 1985. A request for corresponding changes to the Tech-nical Specifications was submitted by letter of Wegst to Denton dated September 25, 1985.

Those submittals are hereby incorporated in this Plan by reference and are not reproduced herein.

October 28, 1985 f

i School of Engineering and Applied Science University of California, Los Angeles

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APPENDIX B: PRELIMINARY RADIATION SURVEY B.1 ACTIVATION PRODUCTS IN GRAPHITE AND LEAD Samples of graphite and lead from the reactor core have been examined with a GeLi detector and multichannel analyzer to j determine speci fic activities and radioisotopic composition.

Graphite stringers taken from the central region of the reactor core in March of 1985 exhibited surface radiation levels

from 20 to 50 mr/hr near the center of those vertical four-foot

, stringers. At a perpendicular distance of one foot from the center of the stringers, the level fell to 5 to 8 mr/hr. Samples

of graphite taken from locations within the central region near core mid-height were found to contain 13.6 year Eu-152 as the principal radioisotooe. Observed specific activities ranged from

0. 3 to 0.45 micro-Ci per gram (3-20-85). Eu-154 and Co-60 were also observed, each at a specific activity about one order of magnitude less than that of the Eu-152.

The lead above the reactor core consists of two layers of lead bricks, each layer is two inches thick. Samples from the upper layer at a corner of the reactor were measurably radio-

, active (0. 5 mr/hr on the surface), and silver-110 was identified at a concentration of about 240 pico-Ci per gram. Bricks taken from the lower course in the vicinity of the fuel boxes were appreciably more radioactive (5 mr/hr on the surface). The pr' i nc i pa l isotopes are silver-108 (127 yr) and silver-110 (252 4

day). The results for six samples, three from each of two bricks, are shown in table B-1.

4 Table B-1: Activities in Lead (5-13-85)

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! Sample Ag-108 Ag-110

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1-1 11.3 gms 110 2510 1-2 30.3 gms 186 3410

1-3 17.4 gms 158 3100 Average 164 3100 l 2-1 38.0 gms 91 2300 i 2-2 20.5 gms 87 2600 2-3 20.3 gms 173 3500

Average 110 2670 t

i Activities are in pico-Ci per gram, the " averages" are the mass-weighted average for each brick.

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B. 2 ACTIVATION PRODUCTS IN CONCRETE The reactor beam port plugs are composed of concrete cast in aluminum sleeves. Samples taken from the south beam port plug and counted on May 15, 1985, showed the following activation products (in order of dominance):

Table B-2: Activation Proaucts in Concrete (5-15-85)

Isotope Activity *  %

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1. Eu-152, 13.6 y 183 52

2. Co-60, 5.3 y 157 42

3. Eu-154, 8.6 y 19 5

4. Mn-54, 312 d 9 2

5. Cs-134, 2.1 y 9 2

TOTAL 377 100

• Activities are in nano-curies per gram. The results in Table B-2 pertain to a sample taken one inch from the interior face of the plug (four inches into the core from the biological shield).

The total activity, as a function of distance from the inner face of the biological shield was found to be:

Table B-3: Specific Activity, Concrete (5-15-85)

Distance, inches Activity, n Ci/gm

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-4 377

-2 229 0 (Note 1) 203 2 126 4 36.6 7 11.5

• 13 1.46 (Note 2) 19 Background Note 1: Zero is taken as the inner face of the concrete bio-logical shield.

Note 2: At the 13 inch depth, Eu-154 and Cs-134 were not dis-cernible. The sum of the other isotopes identified in Tatle B-2 was divided by 0.93 to obtain the activation product concentration of 1.46 nCi/gm. In that same sample a number of short-lived radium-thorium daughter products appeared.

A least-squares best fit of an exponential to the five points from zero to nineteen inches indicates a relaxation length (e-folding distance) of about 2.2 inches for the activation products in this concrete.

2

l The concrete was black in color, evidently the plugs are filled with magnetite concrete. It is premature to assume that the samples are representative of all of the concrete and higher levels of activity may apcear due to streaming neutrons through the control blade shrouds or along pipeways. This question cannot be answered without unstacking the core.

B. 3 RADIATION SURVEY OF THE SOUTH BEAM PORT A radiation survey (by Teletector) of the south beam port in December of 1984 yielded the following readings as a function of distance from the internal end. The "zero" distance is 16 inches interior to the inner face of the biological shield; i . e. the 16 inch measurement is at the inner surface of the biological shield.

Table B-4: South Beam Port Observations Distance, inches Radiation, m rem /hr

========== =============

0 3000 8 2000 i

16 650 28 55 l 52 4. 5 The high readings corresoond to locations within a few inches of the steel blade drive support bearings and structure.

i B. 4 RADIATION MEASUREMENT IN CORE CENTER VOID The reactor core was uncovered to the top of the fuel boxes t in March 1985 for a radiation survey and to collect samples. ,

Fifteen vertical graphite stringers (4 inches by 4 inches by 4 feet) were removed thereby creating a void 12 inches by 20 inches in hori:ontal cross-section and four feet deep. The radiation field in the void was observed to have a nearly uniform value of one rem per hour.

J B. 5 ACTIVATION OF CORE METALLIC PARTS l Other than lead, aluminum is the predominant metallic core component. The fuel boxes, the shield tank, and various plumbing and tubing are composed of aluminum. The vertical port liners are composed of type 6061 aluminum which according to NUREG/CR-i 1756, vol 2, Table E.1-1, contains 0.25% =ide. A small sample of a vertical port liner examined with the GeLi detector showed both 1

2n-65 (244 day) and Co-60 (5.27 yr) in the atomic ratio of about 2:3. The calculated total activity based upon the reactor operating history followed by 20 months decay is about six micro-3

Curies per gram. The calculated result agrees well with a meas-ured radiation field of 10 mR/hr at a distance of three inches from the tube center line at a location far from either end of the tube.

The highest concentrations of radioisotopes are expected in the structural steel and the small amount of stainless steel in the core. Calculated values for these tctivities, based upon the spatial-maximum, time-averaged neutron flux, are shown in Table B-5.

Table B-5: Calculated Activities in Steels Structural Steel Fe-55 1.2 mci /gm l Stainless Steel Fe-55 0.8 mci /gm Co-60 0.3 mci /gm Ni-63 0.1 mci /gm l The majority of the steel parts are not in the highest flux region of the reactor, and the indicated values are htqh upper-bound estimates. '

t I

l l

i 1

I 3

i i

I l l

l 4 l

APPENDIX C: UCLA CORE DISMANTLEMENT EXPERIENCE The UCLA staff, aided by student volunteers, has dismantled the core for maintenance on a number of occasions. Typically, a three week interlude between shutdown and core entry was used for p l ann i r.g, preparation, and procurement. The general plan of entry is straightforward, the planning and preparation related to i finding volunteers, garment selection, procurement of special l materials, review of special tooling requirements, and instru-i ment checking.

1 During this phase, holding tank and sump waters were tested for release to the sanitary sewer to accommodate any new liquid waste that might arise. Coromon air conditioning filters were fixed to the reactor room exhaust grill with duct tape.

Controlled areas were defined within the reactor room, paper was laid, and entry points were established.

Core entry is initiated by removing the top concrete blocks, removing the fuel box cover plugs, and transferring the fuel to dry storage. A third layer of top blocks, the east face blocks, the graphite thermal column, and the lead wall are successive 1v removed. A concrete block (process pit cover block) is then rigged to the crane and used as an elevator to transport lead and graphite from the core top to the floor level. The work was done i with two people at the core top, two unloaders at the floor level, a crane operator, a " gopher," and a surveillant health j chysicist. The gopher remained in the clean area to fetch and deliver materials and tools to the controlled area boundary.

Because of exposure considerations, student help was limited to unstacking and restacking the upper courses of lead and grapn-ite and to floor operations outside the core.

, personnel were rotated on a three-to-four hour basis.

Typical radiation exposures were 35 to 45 man-rem distributed over a twenty to thirty man work force in a period of approxima-tely three months. Fuel handling and reactor reassembly were sionificant contributors to the total exposure. Neither are now recuired. All fuel has been removed from the site, and the ex-core packaging of raaterials for t ransport should be easier l

than the in-core reassembly of the same compnnents to dreanding tolerances.

It might be possible to improve upon the procedures used by i UCLA. oarticularly in man-power utilization. Reactor dismantle-

, ment normally commenced about three weeks after termination of normal operations. Now, nearly two years after the termination of operations, the radiation exposure conditions should be much more benign.

i 1

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