ML20082B649

Decommissioning Plan for L-77 Research Reactor

ML20082B649
Person / Time
Site:	05000262
Issue date:	07/02/1991
From:	Brigham Young University, PROVO, UT
To:
Shared Package
ML20082B644	List: ML20082B642 ML20082B649
References
NUDOCS 9107160192
Download: ML20082B649 (22)

Text

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O DECOMMISSIONING PLAN FOR Tile L-77 RESEARCH REACTOR BRIGilAM YOUNG UNIVERSITY PROVO, UTAH PREPARED BY BRIGilAM YOUNG UNIVERSITY

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1.

SUMMARY

OF PLAN The BYU L-77 Research Reactor is a small, solution-type nuclear reactor designed for laboratory use. It was manufactured by the Atomics International division of North American Aviation, a predecessor of the Rocketdyne Division of Rockwell International. It consists of a reactor core tank, which contained the fuel solution, within an inner shield tank. The reactor is installed in the Nuclear Laboratory on the south edge of the Brigham Young University Campus.

1.1. DECOMMISSIONING METilOD Due to the need for the land upon which the reactor is presently located, DECON is the only acceptable decommissioning method.

1.2. ESTIMATED COST The cost for decommissioning the reactor facility is estimated to be

$35,000. The University has approved $35,000 for this purpose.

1.3. Maior Tasks The major tasks include: removal and transfer of the Pu/Be source, preliminary radiological surveys, dismantling the shield tank, dismantling the inner shield assembly, packaging and shipping radioactive waste, and completion of the final survey and report.

Target date for completion of the process is June 1, 1991.

1.4. Items Subject to Quality Control

' Radiation survey equipment shall be quality controlled in accordance with established protocols (see Section 7).

1.5. Items to Be Performed By Contractor All work shall be performed by Brigham Young University personnel.

1.6. Final Survey The final survey shall be performed using a format that shall give 95%

certainty of discovering contamination foci that average as low as 100 dpm/ square centimeter above background, assuming a population of one contaminated area in every 100.

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2. CHOICE OF DECOMMISSIONING ALTERNATIVE AND DESCRIPTION OF ACTIVITIES INVOLVED 2.1. Decommissioning Alternative Because of the need for the land for other purposes, no alternative to dismantling and decommissioning is acceptable to the licensee.

2.2. Decommissioning Activities; Tasks, and Schedules 2.2.1. Preliminary Survey: Preliminary radiological survey of the facility prior to initiation of the dismantling process.

2.2.1.1. Time to complete: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Ten 1 meter square sections on the floor and walls will be marked off and counted using ,

gamma, beta and alpha instruments as por the protocol for the final survey. In addition the entire outer surface of the shield tank will be surveyed f or contamination. This will be accomplished prior to the .

start of decommissioning activities. Five core samples will be taken  !

from the top and side of the tank and analyzed for alpha,-beta and gamma radiation. Four hours have been allocated for the core samples and analysio.

2.2.2. Preliminary Survey Review: The preliminary survey will be submitted to the full Decommissioning Committee for review prior to initiating dismantling procedures. The Committee will review the data obtained from the preliminary survey. No further action will be initiated until the Decommissioning Committee gives unanimous i approval. l l

2.2.3. Dismantle Shield Tanks Dismantle shield tant, and cut and crimp control and fuel lines. The shield tank will be drained and cut apart (see figure 1).  ;

2.2.3.1. Radiological status and exposure. Three five minute counts made at the surface of the tank using a thin window GM [

tube yioided results of 43, 50 and 47 cpm. This is compared to 46 cpm i as a normal background count. The surface reading using a NaI detector was 1000 cpm with background also measured at 1000 cpm.

There is no evidence to indicate any contamination of the outer shield tank. Thus preliminary estimates support the conclusion that there -

will be no occupational exposure due to-cutting apart the outer shield tank. The Decommissioning Committee will review all surveys including ,

core samples of the tank before proceeding with this phase of - the r project. 7 l

2.2.3.2. Time to complete: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This time is based ,

on an estimate of two days made by our Mechanical Department with one day added in order to cover contingencies.

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'. 2-2 2.2.4. Determino Radiological Status of Remaining Structure Surveys and core samples of the three inner shielding layers shall be obtained and submitted to the Decommissioning Committee. This shall include determining the surface activity of each layer prior to ,

initiating any dismantling of that layer. Time for these surface surveys will be 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the outer stainless steel structure, I hour for the first aluminum structure and 30 minutes for the inner aluminum cylinder. In addition core samples of each of these layers shall be obtained from five locations. One core shall be taken from the top of the structure and the other four shall be taken centered on the reactor vessel and at 90 degree increments around the surface of the cylinder. One hour has been allocated to obtain each core sample.

Four hours have been allocated for analysis of the core samples. An additional four hours have been provided for the Decommissioning committee to review the data obtained and make suggestions on appropriate safeguards to be taken in dismantling the remaining shielding structures. An additional day has been provided to insure sufficient time for this phase of the project.

2.2.4.1. Radiological status and occupational exposure.

A NaI scintillator was introduced into the beam tube. The observed count was 200 cpm eight inches into the tube. The reading was also at 200 cpm in direct contact with the outer surface of the reactor vessel. Background counts taken from other locations on campus and directly outsida the reactor were 1000 cpm. From these measurements

( it !s dif ficult to tell if there is any lonizing radiation due to contamination inside the reactor but radiation from inside the reactor cannot be greater than one fifth of the background ionizing radiation levels at our altitude. Exposures of this magnitude for the eight hours required in proximity to the reactor shell to complete this phase are essentially zero, l

2.2.4.2. Time to complete: 3 days.

2.2.5. Dismantle Remaining Shielding: The inner three layers of shielding will be dismantled. This involves cutting apart the I

stainless steel shall around the outer solid shield region. This is a 0.109 inch thick cylinder 43 inches in diameter and 57 inches high.

This cylinder shall be cut into four sections. The time estimated to complete this task is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Next the borated paraffin and lead will be removed with time to complete esti'.aated at I hour. The second solid shield structure is bounded by an cylinder 54 inches high and 35 inches in diameter made of 3/16 inch aluminum. This cylinder will be cut in half. The estimated time to complete - this operation is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The second layer of borated paraffin will then be removed with a time to complete the project of one hour. The final solid shielding structure is enclosed within a-cylinder 42 inches high and 28 inches in diameter consisting of 3/16 aluminum. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> have been allocated to cut this structure in half. Finally the lead and diphenyl will be removed from the reactor vessel. Estimated time to complete this task is I hour. (note time necessary for surveys has been allocated above)

2-3 2.2.5.1. Radiological status and occupational exposure.

Preliminary surveys (see above) indicate that in the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of potential exposure there will bo no meaningful occuWtional exposure.

Additional surveys and samples will be evaluated by the Radiation Safety Officer and the Decommissioning Committee as they are available.

2.2.5.2. Total Time to complete: 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.2.6. Packaqo Hadioactive Waster Radioactive Waste will be

( packaged labeled and shipped.

Time to complete: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2.2.7. Final Exit Survey.

Time to complete 10 days.

2.2.8. Preparation of Final Reports:

Time to complete: 5 days. 1 2.2.9. Total Time to Completion of Decommissioning: Total time for the completion of the project is 24 working days. Allocating an additional three days for review yields a total of 27 working days or five weeks for the project.

Time to complete 27 days.

2.3. Decommissioning Organization and Responsibilities.

All dismantling operations shall be performed by personnel from Brigham Young University.

2.3.1. Ultimate Responsible Party: The President of Brigham Young University maintains ultimate authority and responsibility for the reactor and the decommissioning project. The President ' has assigned the Associate Administrative Vice President to manage the reactor and decommissioning project.

2.3.2. filgh Level Administrative Control: The Administrative Vice President of Brigham Young University maintains administrative control over all activities relating to the reactor. He reports directly to the President in matters concerning the decommissioning of the reactor.

2.3.3. Project Supervisor 2.3.3.1. Duties: The Project Supervisor directs the day to 'ay activities involved in decommissioning the reactor, lie has the authority and the responsibility to stop any work or activity that ho deems unsafe. The Project Supervisor _ is responsible for the detailed planning and implementation of the Decommissioning Plan.

2-4 2.3.3.2. Line of Authority: The Project Supervisor reports directly to the Associate Administrative Vice President. His activities are reviewed by the Decommissioning Committee, the Radiation Safety Officor and the Project Safety Officer.

2.3.3.3. Training: The Project Supervisor shall have at least one year of experience in general construction as well as one year of experience in radiation safety and one year of experience in general industrial safety. In addition the Project Supervisor shall have a four year degree or it's equivalent in scienco with a minimum of 20 semester hours in physics and mathematics.

2.3.4. . Radiation Safety Officer (RSO):

2.3.4.1. Duties: The RSO is responsible to insure that no operation is performed or continued if it is unsafe. The RSO is also responsible to oversee the performance of radiation surveys. The RSO s' 11 be present in the Reactor Facility while the reactor is being semoved and while radioactive waste is being packaged.

2.3.4.2. Line of Authority: The RSO reports directly to the Associate Administrative Vice President on all decommissioning activities.

2.3.4.3. Training: The RSO shall have a minimum of one year of experience in radiation safety. In addition the RSO shall have obtained a four col' ego degree in physical science or an equivalent of experience and education. This position does not require a board certification.

2.3.5. Project safety Officer:

2.3.5.1. Duties: The Project Safety Officer shall have responsibility to insure that good industrial hygiene and general safety practices are observed during the decommissioning project.

2.3.5.2. Line of Authority: The Project Safety Officer shall report to the Associate Administrative Vice President.

2.3.5.3. Training: The Project Safety Of ficer shall possess a minimum of one year of experience in Industrial Hygiene and Safety.

2.3.6. Decommissioning Committee:

2.3.6.1. Composition: The Decommissioning Committee shall be composed of one Faculty member with a Ph.D in Physics, one Faculty Member with a Ph.D in Biological Science, One Faculty Member with a Ph.D in Mechanical Engineering, a representative from the Campus Safety Office, the Radiation Safety Officer, and an inspector from Physical Facilities.

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- 2-5 2.3.6.2. Duties: The Decommissioning Committoo shall ros iew and audit major decommissioning operations dealing with special nuclear material, radioactivo material, radiological controls, review proceduros, records, reportable occurrences under Parts 20 and 50, and changes made in accordance with 50.59. The committoo shall receive wookly reports f rom the RSO wh11.0 the project is in process. This committoo shall review all surveys and reports from the RSO prior to initiation of activities designated !n section 2.2.4 2.3.6.3. Line of Authority The Decommissioning Committoo shall report directly to the Associate Administrative Vice President.

2.4. Training Program j Eight hours of training shall be given to each individual directly involved in the decommissioning project. The training shall include radiation safety, hazard communication, and general industrial safety.

This training shall be given by the Radiation Safety Officer and the University Safety Officer. (soo enclosure # 1 for outlino and pertinent course materials) This training will be provided to any l Physical Plant personnel involved in the project. Our physical plant l personnel have no background in radiation protection or physics. l 2.5. Contractor Assistanco The only contractor involved in this project will be the contracted waste disposal company.

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3. PROTECTION OF OCCUPATIONAL AND PUBLIC llEALTil AND SAFETY 3.1. Facility Radiological Status 3.1.1. Facility Operating flistory_, The BYU L-77 reactor was delivered to the University by the Atomics International Division of North American Aviation, Inc. In August of 1967. Prior to that, this reactor had been used by Atomics International at several expositions in foreign count. ries where it had been loaded and opera t.ed for demonstration purposes.

The BYU reactor was loaded with urapyl-sulf ate solution on tho days of 12-15 September 1967 with a final

• U mass of 1447 grams. The reactor was mainly used for the next fif teen years in conjunction with reactor physics classes. All operations ceased on 12 May 1982. There were 355 separate operations undertaken prior to the end of operation in May 1982, and the cumulative total for these operations was 1779 watt hours. The reactor was defueled on May 5, 1987 and the fuel was shipped to EG&G Idaho.

There is no reason to believe that any contamination exists outside the confines of the Shield Tank. Neither radiological surveys nor records indicate the presence of such contamination.

3.1.2. Current Radiological Status of Facility: There cre no areas in the facility, outside the reactor shield, where radiation or radioactivity exceeds normal background levels. After draining the shielding water and removing the outer shield tank, radiation levels up to about 0.1 mrom/h may be found. No radioactive contemination outside of the core vessel is expected. Measurements of the radiation levels inside the reactor assembly were accomplished by insertirg a 2" NaI detector into the Beam tube and taking measurements at the surface, 8 inches, 12 inches, 16 inches 20 inches and at the surface of the core vessel and a background from outside the reactor f acility.

The observed counts obtained were respectively 1000 cpm (surf ace), 200 cpm, 200 cpm, 200 cpm, 200 cpm, 200 cpm, and 1000 cpm (background).

The shielding water has been analyzed and no contamination detected.

A sample of the water was also taken by Dean Chaney of NRC region IV for analysis in their laboratories. Pending confirmation of our survey results, along with chemical surveys the water shall be drained into the local sanitary sewer. This course of action has already been cleared with the Publicly Owned Treatment Works and is consistent with present technical specifications.

No direct measurements of the internal contamination levels of the reactor core have been made. However when the fuel was remover the vessel 2nand drip lines were rinsed with distilled water. In addition since U produces x-radiation or gamma radiation at energy levels above 80 key, the NaI detector used to measure residual levels would be sensitive to that type of radiation. Since very low levels of ionizing radiation were detected the implication follows that contamination by residual fuel is small.

l 3-2 Activation of the cobalt impurity in stainless stool components of the coro vessel is estimated to bo loss than 20 uCi at the present timo. No other significant activities are expected to exist.

Because of the sealed and contained nature of the L-77 reactor, no release of radioactive material to the environment is expected. The exposure rate to the public during dismantling and transport of the reactor core vossol to a disposal site is negligible. Thoro are no significant exposure pathways to the public.

The 0.5 Plutonium Bory111um source has been moved to the Physics Underground Laboratory under the direct suporvision of Dr. Gary Jonson who is the Facility Chief and a former licensed operator.

NRC region IV has boon notified of this chango in the facilities.

A. Radiation Protection

1. Decommissioning ALARA Programt It is the basic operating principle of the University that exposures to ionizing radiation shall be kept as low as reasonably achievable. In kooping with this policy the following operating principles shall be observed in decommissioning the reactors

a. Surveys. Throughout the decommissioning process the RSO shall make appropriate surveys and radiological measuromonts to allow the propor selection of protectivo equipment and work practicos in order to maintain occupational exposures as low as reasonably achievable. In addition the Decommissioning Committoo shall review disassembly plans which could involve occupational exposure to ionizing radiation to ensure that appropriate protectivo measures are followed.

b. Dosimotry. Each member of the decommissioning team shall wear a TLD badge. The TLD badges shall be road at the conclusion of the project. Due to the fact that we are unable to identify any lonizing radiation sourcos likely to cause an occupational exposure approaching 10% of the annual limit of 5 rem this type of-monitoring sooms sufficient.

c. Equipment. A Victoreen model 470A ionization chamber shall be on the decommissioning sito at all times while the building is occupied. This device shall be used to measure exposure rates in mrem /hr under the direction of the RSO. In k addition, an Eberline ESP-1 with a thin window Geiger Muller probe shall be used for surveys of bota contamination. An Eberlino ESP-2 with an alpha probe shall also be used under the direction of the RSO to survey for alpha contamination.

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d. Administrative Controls and Structure.

The ALARA program is the direct responsibility of the President of The President has designated the Brigham Young University.

Associate Administrative Vice President to serve as his administrative officer in charge of this program with regard to decommissioning. The RSO serves at the discretion and under the direction of the Associate Administrative Vice Presjdent and is the primary officer responsible for the implementation of the ALARA program. In addition the Decommissioning Committee is charged with reviewing and auditing surveys and procedures to assure that occupational exposures are kept ALARA. This committee also reports directly to the Associate Administrative Vice President.

The radiation protection

2. !jealth Physics Programs program to be used for this decommissioning project is an extension of the program used by the University under Utah Bureau of Radiation Control licenses URC 2500081 and URC 2500091.

Radiation monitoring equipment used during the completion of the work shall include:

1. An Eberline model ESP-2 (with dual calibration) equipped with a thin window pancake probe for beta detection and an NaI low energy gamma dotector.

2. An Eber11ne model ESP-1 with an alpha probe.

3. A Victoreen model 470 ionization chamber for dose rate measurements.

In addition swipes for removable surface contamination as well as various material samples shall be placed in liquid scintillation fluid and counted using a Packard model 1500 Liquid Scintillation Counter.

The quality control of termination survey instrumentation is described in Section 7.0.

In addition to radiation monitoring, personnel dosimetry is performed using TLD dosimeters provided by Eber11ne Dosimetry Servico. All radiation workers have an NRC Form-4 on file, and all occupational radiation exposure is i..cluded in their monitoring record.

Work involving radiation exposure or potential for release of radioactive material requires the presence or availability of the RSO. It is his responsibility to measure and assess-conditions and to assure that radiation exposures are kept ALARA and do not exceed regulatory limits.

3-4 The RSO, the Project Supervisor, and the Project Safhalt et.y Ofunsafe ficer authority and the responsibility to have the operations. It is the responsibility of the project supervisor to I

It is the specific avoid and prevent any unsafe operations.

responsibility of the RSO to prevent excessive exposures to ionizing radiation.

for compliance with applicable The ultimate responsibility regulations resides with the President of Brigham Young University.

radiation survey instruments for radiation protection Use of is purposes (including transportation and termination surveys)These limited to the RSO or his designated representative.

instruments are stored in a manner to minimize the risk of damage 1 to them. Instruments are calibrated by the instrument vendor.

Calibration records as well as quality assurance checks are l maintained by the R$0.

Instruments used for general surveillance are qualitatively tested prior to use by the RSO by checking the battery test, background, and at least one standard. Faulty or out-of-date instruments are withdrawn from service and sont for repair and calibration.

Prior to leaving an area subject to radioactive contamination, or an area with easily transpotted sources of radiation, each person, piece of equipment, lot of material, or container of waste, is subject to an appropriate survey. The work site 19 surveyed for removable surface contamination by smearing 100 cm with a small polystyrene swipe, placing the swipe in a vial containing 11guld l scintillation fluid and counting for radioactivity in a liquid l scintillation counter. Large areas may be surveyed by use of a mop or towel measured by use of a portable survey instrument. During dismantling, these surveys are performed as the work proceeds.

Protective clothing (such as lab coats or overalls), rubber or canvas gloves, and shoe covers are provided as needed.

B. Radioactive Waste Manaqoment

1. Fuel Disposal: The fuel has been shipped to the DOE.

2. Radioactive Waste Processing: The only anticipated radioactive waste generated in the decommissioning process is the core vessel. This structure is a stainless steel container approximately 18 inches in diameter. This material shall be packaged and sent to the-low level waste repository located in Hanford, Washington. Hanford has agreed to accept the waste material. Our Broker over the _ past 4 years has - terminated all contracts. We will select a broker to insure packaging and transportation meet federal as well as state requirements.

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3. Radioactive Waste Disposal _

Without compaction the  :

i total volume of radioactive waste should be less than 27 cubic feet

and should contain total activity of less than 20 microcuries of cobalt 60. In addition there may be traces of radioactive material inside the vessel. In addition to the core vessel the piping .

running to and from the vessel will be submitted as radioactive  !

j waste. Of course any material not meeting the criteria for disposal as non-radioactive waste discovered in the course of the decommissioning project shall be disposed of as radioactive waste.

4. Criteria for disposal as non-radioactive..

Material  ;

shall be considered radioactive if it f alls within three sigma units of the background mean for normal samples of that particular material. Normal samples of material shall be obtained . f rom-vendors and areas which are not associated with radioactive operations.

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5. Minimization of Airborne Hazards. Respirators and SCBA's shall be available to the personnel involved' .in .

decommissioning process. All regulations dealing with personal protective equipment as lister ir. 29 CFR shall be followod. The Radiation Safety Officer and the Project Safety of ficerproject shall determine the necd- for the above equiptaent as the ,

progresses and more detailed information is available. At present there is no anticipated- rel ease of radioactive material or hazardous material as described in Table Z of 29 CFR which would present a hazard to either the worters involveo or the public in general.

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6. Accident Analysis Once the Pu/Bo sourco is removed and transferred prior to the initiation of decommissioning I

I activities, there is no reasonably foreseeable accident which could cause excess ionizing radiation exposure to either the public or the decommissioning team. There are no industrial accitents which could endanger the public at *. rge.

Potential industrial accidents which could affect occupational safety include the standard industrial accidents due to falls, fires, lifting, falling objects, _ and power tool incidents. The Project Safety Of ficer shall review plans and monitor activities to reduce the likelihood-of such occupational injuries.-

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No significant amounts of any non-radiological hazardous materials-are anticipated in this project. No_ blasting shall be performed-and no solvents are needed. All members of the project crew have been trained in routine industrial hygiene and safety practices as applied to dismantling and decontamination operations.- Emphasis is '

placed on portable ' electric tools, eye protection, and mobile equipment. The Director'of Campus Safety is directly involved-in supervising the entire project.

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3-6 Both the Project Supervisor and the Project Safety Of ficer have the authority to halt unsafe operations. It is the responsibility of the Project Supervisor to avoid and prevent any unsafe operations.

BYU select.s UL or FM approved standard-grado commercial equipment available for industrial hygiene and safety purposes. No further selection criteria have been established.

Accident prevention is practiced by reviewing operations and equipment prior to starting, by providing both management and health physics supervision, and by emphasizing the need for safe work practicos. Accident response during the dismantling project shall be provided by on site and municipal fire protection and emergency medical services.

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1. PROPOSED FINAL HADIATION SURVEY PLAN After dismantling has boon completed, and all other equipment has from the facility, termination radiation survey been removed A one meter grid shall locations shall be selected and identiflod.

be established on the floor, walls and colling insido the ret.ctor room (see figure 2). The entire surface area of the reactor room shall then be surveyed for alpha, bota In and thegamma radiation remaining part ofusing the the instrumentation described below.

f acillty the survey shall be accomplished by establishing From each a virtual 3-m-3-m-square grid on the floor,* walls, a and colling.

single 1-m square shall be square (approximately 100 f t ),

selected that, as judged by the RSO, is most likely to represent the highest amount of residual radioactivity within the 3-m square.

The selected square meter is surveyed for alpha, beta, gamma, and removable surf ace contamination. This method results in 11% of the remaining surface area within the f acility being uniformly samplod.

This sampling shall be increased on the floor, as necessary, to assure that at least 30 locations on the floor are tested.

In addition, each drain and floor penetration shall be surveyed.

The survey measuromonts to be performed consist of determination of the average alpha surfaco radioactivity, average beta surface radioactivity, removable surfaco radioactivity, and the ambient If (gamma) radiation exposure rate at 1 m from the surface.

measurements of the average alpha and/or beta surface radioactivity indicate the presence of residual " hot-spots", a measurement of the radioactivity in a 100 cm' area at the hot-spot location shall be made.

Measurements of the averago alpha and beta surface radioactivity in a 1-m square are made by using portable scalors (Eberline ESP-1 and

2) with audible output, connected to an Eberline model Eber11ne 260alphathin window GM probe for beta activity, and an scintillator for alpha activity. Those probos are uniformly scanned over the surface (in close proximity) and the counts recorded for 5 minutes. Correction for detector efficiency, (estimated via Th-230 source for alpha and a Tc-99 source for bota) local background, detector area, and count timo results inI m*

a, measure of the residual radioactivity averaged over Measurement of the maximum surface radioactivity averaged over 100 cm* within that 1 if warranted, by similarly scanning an area of 100 cm,m' is dono,for a period of one minute and performing the necessary corrections as described.

Removable surface radioactivity is estimated for each 1-m square by smearing a 100 cm'. area, using a polystyrene swipe, dissolving the swjpe in 11guld scintillation fluid and counting the same in a Packard model 1500 11guld scintillation countor. This counter provides counting efficiencies of about 95% for betas with average energy of 49 key, approximately 100% for botas between 50 and 1000 l -

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4-2 key and an officiency of approximately 100% for alpha radiation.

In addition, gamma emitters are detected due to the presence of internal conversion electrons, auger electrons (efficiencies of about 50% at 6 kev) and compton scattering. Total efficiencies for gamma embitters vary by the nuclide involved but generally range from 10% to 70%. A minimum of 500 swipe tests shall be made.

Under the assumption that removable surface contamination is described by a Poisson distribution, there is a 99% probability of detecting contamination even when it is present at levels as low as one contaminated square por one hundred squares.

Gamma radiation is measured by means of an ESP-2 portable scaler with audible output connected to an Eberline PG-2 low energy gamma probe (Sodium Iodido scintillator). The probe is situated in the conter of the square meter being surveyed and an exposure is integrated over a five minute period.

Floor drains and similar features (utility boxes or trays, etc.)

that do not permit such a methodical approach shall be inspected by use of portable survey instruments for alpha, beta, and gamma activity, and by smears for removable activity.

Calibration shall be performed by the instrument vendor (Eberline) within three months of the termination of the survey.

The data shall be analyzed by use of a computer program that takes the fundamental instrument reading and performs the necessary calculations to produce the derived result, either dpm/100 cm* or uR/h, thus minimizing transcription and arithmetic errors. The irequency distribution shall be plotted and compared, via goodness of fit test, to the normal distribution with the mean equal to the average background count and the standard deviation equal to the square root of the background count divided by 5. The sample distribution shall also be compared to a background sample distribution, obt61ned from other locations on campus.

Instrument quality control charts with two sigma intervals are prepared for each instrument. The channel check and background are These, and the plotted daily whenever the are instrument is in reviewed use.

and compared with the fundamental data sheets, analytical results to confirm the survey quality.

A summary report of the survey results is prepared for submission to the NRC to support a requ st for termination of the license.

All survey documents are retained until after concurrence by the NRC that the facility is acceptable for release for unrestricted use.

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4-3 A. Acceptance Criteria The facility shall be. deemed suitabic r. -

for unrestricted use if the survey reveals for release .o measurements which a r e - m o r'a than three sigma units above i u established bar.kground measurements for a facility composed >

similar materials which has no history of use witti .acaoacti .

materials.

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11. COST ESTIMATE FOR THE DECOMMISSIONING PROJECT AND PLAN FOR ASSURING AVAILABILITY OF FUNDS FOR THE COMPLETION OF THE PROJECT.

Cost estimates have been Funds basedhaveon the beenworst casebased approved which on the the University deems possible.

following estimates, although there is a high likelihood that this estimate will significantly exceed actual costs.

A. Cost Estimates

1. Waste disposal $15,000 Equipment (rental / purchase) $10,000 Disposable gear $ 5,000 Miscellaneous $ 5,000 TOTAL $35,000 B. Funding Assurance.

The University shall insure that funding is available by appropriating $35,000 for fiscal year 1991 to a decommissioning budget account.

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9 6-1 IN PLACE III. TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS DURING DECOMMISSIONING Current technical specifications shall remain in place during the decommissioning process.

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l 7-1 IV. QUALITY ASSURANCE PROVISIONS IN PLACE DURING DECOMMISSIONING A. Radiation Survey Equipment All radiation survey equipment shall be calibrated by the equipment vendor within tow weeks of the completion of the final rurvey. In addition, a quality control chart with two sigma interval shall be prepared for each detector / probe combination. Quality control checks shall be made on a daily basis when the instrument is in use for the decommissioning project. If an instrument yields a reading which is out of control the channel check shall be repeated. If the instrument is still outside the control limits, the instrument shall be taken out of service and returned to the manuf acturer for repairs and calibration.

1. Beta Survey Instrument Channel Check.

Beta survey instrument channel check shall be accomplished by placing the probe over a Tc-99 (traceable to NBS) source and accumulating a one minute count. The same procedure shall be repeated for background. The above quantities shall be logged on a quality control chart.

2. Alpha Survey Instrument The same procedure as was listed for the Beta instrument using a Th 230 source.

3. Gamma Survey Instrument As outlined above using a Cs 137 source.

4. Victoreen Ionization Chamber.

The instrument shall be zeroed and placed in a field produced by a Cs 137 source. The resulting reading shall be plotted on a two sigma quality control chart.

5. Packard model 1500 Liquid Scintillation Counter (LSC).

The calibration of the LSC is checked using calibrated standard liquid scintillation vials purchased from the New England Nuclear subsidiary of Dupont. Three standards containing tritium, carbon-14 and background are placed in the counter at the end of each lot of material counted. The resulting counts are plotted on a quality control = chart with two sigma interval.

Any other quality control measures deemed necessary by the Decommissioning Committee shall be placed into effect by the order of that Committee.

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8-1 l SECURITY PLAN PROVISIONS IN PLACE DURING V. PHYSICAL DECOMMISSIONING The Physical Security Plan is no longer required. (see Docket No. J 50-262 dated November 1, 1989.)

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9-1 VI. ESTIMATED COLLECTIVE DOSE EQUIVALENT As is indicated above, f rom the measurements made to date, there is no identifiable source of exposure at levels approaching the background levels in this area.

Q 10-1 VII. REGULATIONS REGULATORY GUIDES AND STANDARDS. The University is committed to following the stanlards set by the NRC in 10 CFR part 20 as relates to the potential hazards associated with ionizing radiation. In addition ;he University is committed to following guidelines established in 29 CFR by the Occupational Safety and Health Administration with regard to general industrial safety.

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