Text

Argonaut Reactor Facility Decommissioning Plan
	ML20203G016
Person / Time
Site:	05000124
Issue date:	07/31/1986
From:	; VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSB
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ML20203G013	List: ML20203G007; ML20203G016;
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	NUDOCS 8607310231
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{{#Wiki_filter:. - _ _ _. _. T CHEM-NUCLEAR SYSTEMS,INC. I I VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY ARGONAUT REACTOR FACILITY DEC0!94ISSIONING PLAN 'I NRC LICENSE NO. R-62 DOCKET NO. 50-124 I i JULY,1986 1 VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY I DEPARTMENT OF HEALTH AND SAFETY BLACKSBURG, VIRGINIA 24061 I I I I I nraa888a n8Mr 1l'

I TABLE OF CONTENTS NUMBER _SECTION PAGE

1.0 BACKGROUND

AND MANAGEMENT 1 1.1 SUMARY DESCRIPTION 1 1.1.1 Facility 1 1.1.2 Reactor 1 1.1. 3 Support Facilities 3 1.1.4 Decommissioning Option 4 1.2 FACILITY OPERATING HISTORY 4

1. 2.1 Reactor Operations 4

I 1.2.2 Solid Waste Management 4

1. 2. 3 Liquid Waste Management 5

1.2.4 Gaseous Waste Management 5 I 1.2.5 Incidents 5

1. 2. 6 Conclusions 6

1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY 6 1.4 DECOMMISSIONING ALTERNATIVE 7 1.5 DECOMISSIONING ORGANIZATION AND RESPONSIBILITIES 7 1.5.1 Virginia Tech Administrative Controls 8 1.5.2 Contractor's Organization 10 1.6 REGULATIONS, REGULATORY GUIDES AND STANDARDS 15 1.6.1 Quality Assurance 15 1.6.2 Q.A. Program 16 1.7 TRAINING AND QUALIFICATIONS 16 2.0 OCCUPATIONAL & RADIATION PRCTECTION PROGRAMS 16 2.1 RADIOLOGICAL PROTECTION PROGRAM 16 2.2 INDUSTRIAL SAFETY AND HYGIENE PROGRAM 17 2.3 CONTRACTOR ASSISTANCE 17 2.4 COST ESTIMATE AND FUNDING 17 3.0 DISMANTLING AND DECOM ISSIONING TASKS AND SCHEDULES 18

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I TABLE OF CONTENTS (CONTINUED) NUMBER SECTION PAGE i 3.1 TASKS 18 3.1.1 Phase I 18 I 3.1. 2 Phase II 18 3.1. 3 Phase III 19 3.2 SCHEDULE 20

3. 3 TASK ANALYSES 20 g

3. 3.1 Baseline Radiological Survey 20 g

3.3.2 Core Dismantlement 21 3.3.3 Experimental Facilities Dismantlement 22 3.3.4 Decontamination and/or Dismantlement of I Fuel Storage Wells 23 3.3.5 Dismantlement of Water Process System 23 3.3.6 Activated Concrete Removal 24 3.4 SAFE STORAGE 25 4.0 SAFEGUARDS AND PHYSICAL SECURITY 25 5.0 RADIOLOGICAL ACCIDENT ANALYSES 25 6.0 RADI0 ACTIVE MATERIALS AND WASTE MANAGEMENT 25 6.1 FUEL DISPOSAL 25 6.2 RADI0 ACTIVE WASTE PROCESSING 25 7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS 26 8.0 PROPOSED TERMINATION RADIATION SURVEY PLAN 26 I I lI lrI r I 11 lI I ,I o

I FIGURES (Page 29) NUMBER TITLE PAGE lA VIRG!NIA STATE COUNTY MA? F-1 1B ROANOKE/BLACKSBURG AREA F-2 1C MONTG0MERY COUNTY MAP F-3 1D CAMPUS LAYOUT F-4 2A ROBESON HALL - FIRST FLOOR PLAN F-5 2B ROBESON HALL - SECOND & THIRD FLOOR PLANS F-6 2C VTAR - BASENENT PLAN F-7 3A VTAR BIO-SHIELD PLAN VIEW DRAWING F-8 3B VTAR EXPERIMENTAL FACILITIES DRAWING F-9 4A VTAR SIDE VIEW SECTIONAL DRAWING F-10 4B VTAR PLW/ ELEVATION SECTIONAL DRAWING F-ll 5 VTAR VENTILATION SYSTEM DIAGRAM F-12 6 VTAR CIRCULATING SYSTEM DIAGRAM F-13 7 VTAR RADIOLOGICAL CONTROL SURVEYS F-14 8 DEC0lHISSIONING PROJECT ORGANIZATION F-20 I 9 MAJOR TASK & MILESTONE SCHEDULE F-21 I I I I I I l 4

i TABLES (Page 30) NUMBER TITLE PAGE I VTAR DECOMISSIONING PROJECT MAN-REM ESTIMATE T-1 II VTAR SPECIFICATIONS /0PERATING HISTOP.Y T-2 III VTAR RADI0 ACTIVE MATERIAL DISCHARGE / DISPOSAL HISTORY T-5 IV REGULATIONS AND STANDARDS T-6 Y ACCEPTABLE RADIOLOGICAL CONTROL RELEASE LIMITS T-7 VI CRITERIA FOR RADIOLOGICAL CONTROL OF EQUIPMENT & MATERIAL T-8 i VII INSTRUMENTATION TO BE USED DURING THE VTAR DECOMISSIONING T-9 VIII VTAR DECOMMISSIONING RADI0 ACTIVE WASTE / SHIPPING SCHEDULE T-10 I I I lI

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I APPENDICES (Page 31) ITEM TITLE PAGE A YTAR TECHNICAL SPECIFICATIONS Al-21 B ENVIRONMENTAL IMPACT REPORT B1-6 C PHYSICAL SECURITY PLAN Cl-4 I I I I I l I 'I I I l I I I V I I I

1.0 BACKGROUND

AND MANAGEMENT This plan addresses the dismantling and decomissioning (D&D) of the Virginia Tech Argonaut Reacter (VTAR) Facility and the return of the l facility to unrestricted use. A contractor, Chem-Nuclear Systems, Inc., who is experienced in the dismantling of research reactors will perform the major decommissioning operations. As the licensee, I Virginia Polytechnic Institute will monitor the overall project to ensure compliance with all applicable regulations. The format of this plan follows the outline proposed by the Standardization and Special I Projects Branch of the Nuclear Regulatory Commission. 1.1

SUMMARY

DESCRIPTION 1.1.1 Facility The Virginia Tech Argonaut Reactor (VTAR) Facility is I located in Robeson Hall on the northwest corner of the main campus of Ytrginia Polytechnic Institute and State University (Virginia Tech), between the Appalachian and Blue Ridge Mountains in southwest Virginia, I approximately 35 miles west of Roanoke, Virginia. The reactor was used as part of the Nuclear Engineering curriculum for basic research in neutron physics, I neutron radiography, neutron activation analysis, technical training and Reactor Operator training, as well as experiments associated with health physics and I nuclear engineering. Operation of the VTAR was generally limited to eight hours per day, five days per week, or approximately 2000 hours per year. Within Robeson Hall are Physics faculty, staff and graduate student offices, other research laboratories, shops and classrooms which are primarily used for Physi cs. The building itself has free access; however, entrance to the reactor facility is strictly controlled in conformance with the Physical Security Plan. See Figure 1 for an overview of the Reactor Facility, its I location within the state, county, campus, and major access routes. Figure 2 shows the floor plan layout of Robeson Hall and the VTAR Facilities. 1.1. 2 Reactor The VTAR is an Argonaut type research and training reactor originally designed and installed by American Standard Nuclear Division. The reactor began operation in June,1959. Originally licensed for a maximum power I level of 10 kW(th), the reactor was modified and the license (R-62) amended to allow a maximum power of 100 (1 ) 0758J lI llI

I I kW(th) in late 1966. The reactor was shut down on July 14, 1983 and a possession only license was issued in April, 1985. The reactor fuel was shipped to DOE in I late 1985 and early 1986. Figure 3 shows an overall view of the experimental facilities available around the reactor, as well as a general layout of the reactor cell. Figure 4 shows cutaway views of the shielding I and the core area. The VTAR core is heterogeneous in design, using 93 I percent enriched MTR type uranium-aluminum matrix fuel el ements. Thermal power output of the VTAR was limited to 100 kW(th) with water used as a coolant and as part of the moderator. The remainder of the moderator I consists of graphite blocks which surround the boxes containing the fuel and the water moderator. The reactor core has a two-compartment fuel assenbly geometry which was composed of two separate water-filled aluminum boxes. Each box contained six fuel assemblies surrounded by reactor grade graphite. The primary coolant (demineralized light water) was pumped upward over the fuel elements, then fed by gravity through the overflow orifices to the heat I exchanger, where reactor heat was transferred from the primary coolant to the secondary coolant. Experimental facilities located in the VTAR are: (1 ) Shield tank on east face of core ( 2, iwo (2) six-inch beam ports on the north and south faces of the reactor (3) One central penetration into the center of the core area. This penetration also contains a graphite sample block for up to 49 3/5-dram vials. (4) Three top-penetrations into the core area (1 in. x 2 in. each) (5) 4 in. x 4 in. removable blocks in the thermal column l (6) Pneumatic " rabbit" receiver in central region of thermal columns I (7) Pneumatic " rabbit" receiver in central region of core Shield plugs were normally inserted into these facilities unless I the facilities were being used (see Figure 3). 0758J (2) I

I 1.1. 3 Support Facilities a) Located within the reactor facility is a chemistry laboratory with a fume hood and associated I chemical laboratory equipment. This area is now used for counting of samples and is radiologically clean. b) A 10 ton overhead crane services transfers between the reactor core and fuel pit. The crane is also { I used to transport various heavy loads. The crane i is a Robin and Myer F5 double girder, three motor, / overhead crane. Support for the crane is supported by two "I" beams, supported at 4 points 5 on each beam by two center posts and by the wall on either end. c) The ventilation system for the VTAR facility provides negative pressure in the reactor room. If small amounts of loose contamination should occur, this feature helps to keep it inside the reactor I room and allow for the monitoring of the exhaust stack for any radiation release to the environment. The negative pressure in the I ventilation system is provided by a centrifugal fan mounted on the roof. The output from the fan is directed into the input of an axial fan, which I dilutes the stack exhaust and provides additional fl ow. Intake from the ventilation fan is provided by an 18 inch diameter galvanized duct which runs from the roof down to the intake duct in the I reactor room. Monitoring within the duct is accomplished by a I G-M detector and two TLD's mounted in the duct. A j connection is provided to exhaust the chemical fume hood in room 6. A damper is provided to close off the fume hood if necessary. See Figure r 5 for a diagram of the ventilation system. l l g d) The reactor cooling system served the dual ! g functions of removing heat generated by fission and of moderating the fast fission neutrons in the VTAR. The cooling system consists of a primary l l system and a secondary system. The primary system transfers the heat from the reactor to the heat I ,I 0758J (3) I

I exchanger; this heat is then removed by the I secondary system with no mixing of water between the two systems. A functional diagram of the cooling system is shown in Figure 6. 1.1. 4 Decomissioning Option The VTAR site is currently licensed by the USMIC for possession only under License R-62, Amendment No. 6 issued April 16, 1985. The areas covered by the license, shown in Figures 3 and 5, include rooms 108, I 6, 8, and 10 (the " Project Site"). The only decomissioning option planned is the release I of the facility for unrestricted use. The total cost of the decomissioning operation is estimated to be $600,000. The date for the completion of the I decomissioning is estimated to be January,1987. The collective dose equivalent for the decomissioning i operation is estimated to be less than 2 man-rem (see Table I). 1.2 FACILITY OPERATING HISTORY 1.2.1 Reactor Operations A sumary of the VTAR operating history is presented in Table II. There was no fracture of or leakage from the I fuel elements during reactor operations. Solid, liquid, and gaseous radioactive effluents were produced by the VTAR. Table III provides the details on the I types of waste generated, the nuclides present, their activities, and how each type of waste is treated at the VTAR facility. f 1.2.2 Solid Waste Management l Low-level radioactive solid wastes were generated during routine reactor operations and maintenance. These wastes consisted primarily of contaminated gloves, paper, plastics, tools, clothes, samples from I activation analysis and ion-exchange resins. Isotopic analysis was performed on unknown solid waste using a germanium-lithium (GeLi) semiconductor detector. I Solid wastes were then packaged in 55-gallon drums and shipped to an approved disposal site in accordance with applicable NRC and D0T regulations. The tota'.*ctivity I of solid waste disposed of between November, I975 and June,1983 consisted of 67.45 millicuries (listed in Table III). Cobalt-60, Zinc-65, and Silver-110 I comprised 98.2% of this total. 0758J (4) I

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1. 2. 3 Liquid Waste Management Liquid wastes generated from reactor operations consisted primarily of contaminated water from the I

shield tank, dump tank, primary coolant, and rinse water from the decontamination of tools. Samples of the water to be disposed of were either analyzed on a I GeLi detector or evaporated, with the residual solid analyzed for gross alpha and beta activity. The water was discharged to the sanitary sewerage system if the activity levels were within the acceptable levels for release into an unrestricted area as cited in 10 CFR 20, Appendix B. The release was further diluted by an average flow of approximately 900,000 gallons of I sanitary sewerage per day from the university out fall. The liquid waste released from the VTAR facility into the sanitary sewerage system from 1977 to 1983 is presented in Table III. Liquid wastes that could not be disposed of in the sanitary sewerage system were placed in suitable I containers and stored on site until arrangements for off-site disposal had been made. The liquid was then absorbed into vermiculite in 55 gallon drums and I shipped to an approved disposal site in accordance with applicable NRC and DOT regulations.

1. 2. 4 Gaseous Waste Management The primary gaseous effluents of concern was Argon-41.

This radionuclide was produced in the VTAR by the I neutron activation of Argon-40. The calculated concentrations of Ar-41 in unrestricted I areas around the VTAR represents 0.24% and 0.46% of 10 CFR 20 limits. Argon-41 emissions were limited by VTAR Technical Specifications to 1 x 10-4 Ci/sec and 315 C1/ year. Table III shows the Ar-41 stack discharge rate in Cf/sec from 1978 to 1983. It also shows the total amount of Ar-41 released per year for the period 1978 to 1982. 1.2.5 Incidents In 1971, an incident occurred during reactor operations which resulted in the release of irradiated U-235 products into areas near Room 10 and Room 106. I Virginia Tech performed decontamination efforts under AEC cognizance and received AEC approval of the final site survey following decontamination. 0758J (5) I

I In 1975, the core region was flooded with reactor coolant which drained through the core cavity drain line into the process pit sump. There were no I environmental releases during this episode. All liquids were treated and analyzed prior to release. The decontamination of this drain line is anticipated to be part of the D&D Project. 1.2.6 Conclusions Wastes produced at the VTAR consisted of only low-level radioactive wastes. These wastes were packaged and shipped to an approved disposal site in accordance with I applicable NRC and D0T requirements. Liquid wastes were discharged to the sanitary sewerage system when the activity was less than those specified in 10 CFR 20, Appendix B - Table II for unrestricted I areas. Liquid wastes not discharged to the sanitary sewerage system were packaged and shipped to an approved disposal site in accordance with applicable I NRC and D0T requirements. Gaseous effluents released to the environment consisted mainly of Ar-41. Releases to unrestricted areas were well below the maximum allowable limits cited in 10 CFR 20, Apendix B. I Release rates in Ci/sec and Ci/vear were less than the limits specified in the Technical Specifications for the VTAR. Based on the above analysis the radioactive waste management program at the VTAR facility has operated in a safe manner consistent with NRC and DOT regulations. 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY I l 5 The facility has been operated in an extremely clean f radiological fashion. Since fuel has been removed, Virginia Tech has undertaken significant efforts to remove 'E non-radioactive and salvage waste and has generally made site !3 conditions ideal for turnover to the decomissioning contractor. 1 ( The radioactive portions of the facility are confined almost l exclusively to the Reactor Assembly / Bio-Shield shown in Figure 6, with the possible exceptions of the fuel storage pit and pump pit which have low, almost releasable levels of contamination. I Figure 7 shows the Experimental Facilities which, with the activated concrete and steel surrounding these components, l contain almost all of the total inventory of by-product material on-si te. I 0758J (6) I I

I The contractor will perform a detailed baseline survey prior to on-site mobilization to fully characterize radioactive materials inventories on-si te. The primary sources of radiation in the VTAR Facility consist of the following: (a_) Neutron activated materials: Reactor fuel channels and coolant / moderator piping Reactor instrumentation components (detectors) I Reactor core, thermal column, and tank duct graphite blocks / stringers Beam port plugs and steel liners Biological shield structure (reinforced concrete) I Reactor core closure plugs (steel encased high density concrete blocks) (b) Contaminated /potentially contaminated materials: Reactor coolant / moderator / process system (internals) I Below grade fuel storage wells Ventilation system (internals) Radiological Surveys of the facility are currently being I conducted on a quarterly basis in conformance with the facility Possession Only License requirements. Select copies of these and past surveys are presented as Figure 7. 1.4 DECOMISSIONING ALTERNATIVE In order to restore the areas now occupied by the reactor to I productive use, Virginia Tech is dismantling and decommissioning the reactor facility for unrestricted use. No other alternative is under consideration. Future activities at the reactor l facility will not require an NRC license. 1.5 DECOMISSIONING ORGANIZATION AND RESPONSIBILITIES I Figure 8 shows the Organization Chart for the dismantling and decomissioning of the Virginia Tech heactor Facility. E Personnel experienced in reactor operation and radioactive 5 material handling will be responsible for all dismantling operations. 1.5.1 Virginia Tech Administrative Controls The overall responsibility for the Virginia Tech I reactor has been assigned to the Vice President for Administration and Operations. He has assigned the responsibility for administering the facility to the I 0758J (7) I

Head of Department of Safety and Health Programs. The Reactor Safety Comittee has the responsibility for monitoring reactor activities to assure compliance with NRC regulations, provisions of the terms of the reactor license and the Technical Specifications. Virginia Tech has established a decomissioning organization which includes personnel with experience in reactor operation and radioactive material handl ing. To the largest extent possible the personnel in the organization are drawn from existing university I personnel epxerienced in reactor operations. A contractor will be employed to perform specific tasks of the decomissioning. The ultimate responsibility I for decomissioning activities rests with the university administration. I The decomissioning staff line organization is illustrated in Figure 8. The minimum requirements for key personnel, their education and training and experience requirements are discussed below. a) University Administration and Related Services - Academic administration, business administration, and services such as planning, university police and utilities are provided by persons within the university organization that are either appointed I or hired to fulfill the requirements of the many university functions. Education, training and experience are determined by university policy for each of the various personnel positions. I Preparation of contract documents and dispersal of funds are two of the responsibilities delegated to specific personnel within the University I Administration and Related Services. b) Reactor Safety Comittee - This committee will advise university administration on matters of radiological safety. It will also provide review of overall planning and direction to the Safety 3 Office on matters of radiation safety. A member ' 3 of the safety office will inform the comittee of the overall planning and dismantling activities. The University plans to assign these responsibilities to a special six-person Reactor Safety Subcomittee of the Radioisotope Safety Comittee that reviews activities of radiation lE5 safety of various University programs. The l members are appointed by the Vice-President for Administration and Operations. 0758J (8) I

I c) Health & Safety Department - Responsibilities for industrial and radiation safety programs are delegated to the staff of the Health & Safety Department. Persons employed by the office consist of a Health & Safety Department Head, Radiation Safety Officer, Fire Marshal, Environmental Safety Specialist, and Occupational I & Health Specialist. Decomissioning planning and tasks will be reviewed by the designated persons of the safety office to determine the safety I specifications applied to decomissioning activities. Industrial safety rcquirements will be implemented by actions of safety office I personnel or incorporated into the specifications for specific dismantling tasks. Review of radiological control plans and performance of the decomissioning activities will be done by the I Radiation Safety Officer and the results provided to the Radiation Safety Comittee, d) Crafts and Labor - Craft supervisors and crew leaders from the Virginia Tech physical plant will provide assistance to some of the various decomissioning tasks. Persons are to be I designated contracted by the university that are knowledgeable of the proposed activities and that have the appropriate experience for each task. I Selection assistance of available contractors for each task may be provided by the same contractor that provides the Decomissioning Project Team. e) Health Physics - A health physicist will be assigned to the decomissioning of the reactor. This person will be responsible for recomending ( and enforcing radiological safety policy. Responsibilities include maintenance of radioactive exposure records, implementation of i the environmental survey program, ensuring compliance with work procedures, training and, if necessary, assigning additional health physics technicians to specific work tasks. In addition, I the health physicist will be responsible for oversight of the development and implementation of i the facility radiation protection program, the I survey instrumentation program including j calibration, bioassay of personnel, airborne radioactive monitoring, supervision and I documentation of radioactive waste packaging and ALARA planning. 0758J (9) I

f) Quality Assurance - The Quality Assurance I Supervisor will be responsible for performing and implementing the quality assurance plan for decomissioning, working with all branches of the I organization. In order to insure independence of the quality assurance program this person will report directly to the University administration. I This person maintains audit and job performance records and verifies that established safety review procedures are followed. A person j experienced in Quality Assurance activities or j I trained for the specified dismantling activities will be designated to perform and implement quality assurance procedures. 1.5.2 Contractor's Organization I The contractor project team will have an on-site component and a Corporate Office component. The on-site component will consist of the Project Manager, the Site Superintendent, the Radiological Control I Supervisor, the Decontamination Supervisor and up to eight trained radiation workers. The Corporate component will consist of the Senior Project Engineer, I the Technical Advisory and Review Committee, and corporate support personnel. Contractor personnel directly responsible for the project and their general duties are listed below, a) Project Manager: The Project Manager will have overall authority and responsibility for the contractor's conduct of the Virginia Tech Research Reactor project, reporting to Virginia Tech on all project activities and deliverables and to his Corporate I l Office for oversight and management control. He j will provide technical direction to and be responsible for the professional conduct of staff activities in the delivery of completed tasks in I ( accordance with the terms and conditions of the contract. In carrying out these basic responsibilities, the Project Manager will, with the assistance of subordinate staff, accomplish the following: Preride Virginia Tech with a liaison for reporting i and quality assurance, and for the adequacy and l technical accuracy of all reports submitted; I 0758J (10)

I Prepare monthly schedule status reports for transmittal to Virginia Tech; Continuously measure and evaluate overall performance, initiating necessary corrective measures; Comunicate with the Virginia Tech contact person at least weekly to maintain continuity of information on the status of the Project; Conduct monthly progress meetings with the Superintendent and Radiological Control Supervisor; i Meet with, report to, and receive guidance from the Technical Advisory and Review Comittee, initially and at least monthly thereafter; and Coordinate the preparation of the final report and data summary to Virginia Tech, b) Site Superintendent: The Site Superintendent will closely and continuously monitor the contractor's project activities, for which he will have total site res ponsibility. He will provide technical I direction to, and be responsible for, the technical conduct of the staff. He will be responsible for assuring the completion of project i E tasks in accordance with applicable regulations as

5 well as with the terms and conditions of the contract with Virginia Tech. The Site Superintendent will report to the Project Manager l

for technical oversight and financial control, and to Virginia Tech for project activities, schedule, l and milestones accomplishment. In carrying out j these basic responsibilities, the Site Superintendent will accomplish the following tasks f on a routine basis: Coordinate, administrate, and monitor overall project operations for which he will have total responsibility at the Virginia Tech; I Evaluate project performance and intitiate corrective actions when necessary; Administer the required Safety, Quality Assurance, and other applicable programs; I 0758J (11) I

Ensure compliance with work permit and procedures; Requisition the procurement of materials and services; Maintain the project log and record progress daily. Prepare weekly schedule and cost status reports and alert the Project Manager to potential problems with the project; and { Support and work cooperatively with the Radiological Control Supervisor to assure effective leadership, safety of operation, and 1 I regulatory compliance. j l c) Radiological Control Supervisor: The Radiological Control Supervisor will report directly to the Site Superintendent for project administration and coordination. In order to maintain appropriate " checks and balances," however, the Radiological Control Supervisor will also report directly to the contractor's Corporate I Radiological Safety Officer for Health and Safety standards application and for assistance in regulatory compliance. He will supervise the health physics program and will be responsible for regulatory compliance. In carrying out these responsibilities, the Radiological Control Supervisor will, with the assistance of his I Radiological Control Technicians, accomplish the following: Devise and conduct the general health physics and Virginia Tech site-specific training program for all project personnel, and test and certify I locally hired personnel as " radiation workers"; l Assist the Site Superintendent in the preparation of work procedures; Specify radiological and industrial safety requirements on Radiological Work Permits and l Radiological Work Procedures; Maintain copies of all reference doccments, licenses, procedures, and training and qualification records; 1 i 0758J (12)

I Monitor radiological work performance; Review and maintain records of all radiological surveys; Prepare and submit weekly safety and regulatory compliance reports to his Corporate RSO for review and for review by the Virginia Tech Reactor Safety Committee; i Comunicate freely with the Corporate RSO and with I members of the Technical Advisory and Review { Comittee for technical assistance and guidance; Comunicate at least weekly with the Virginia Tech RSO to discuss current Project status and any radiological control problems encountered; Manage the inventory of radiological equipment and supplies; Prevent performance of work by any project personnel when, in his opinion, such work is being performed in an unsafe manner. Inform the Site Superintendent imediately of stop-work decisions and expeditiously seek resolution; and I Inspect and assist with radioactive materials shipmen ts, d) Radiological Control Technicians: Radiological Control Technicians report to the Radiological Control Supervisor and are primarily I responsible for meeting radiological requirements and for safe performance cf all work in radiologically controlled areas. In carrying out I these responsibilities, they will perform the } following functions: Conduct surveys to ensure that all radiological I conditions are met prior to work performance in the area; I I I

I t Monitor work in progress to ensure compliance with standard radiological practices, work permits, and procedures, and stop work when radiologically adverse conditions are indicated; Perform radiological surveys and prepare records of these surveys as directed by the Radiological Control Supervisor; Advise Decontamination Technicians on techniques to minimize exposure during radiological work; j Supervise the packaging of radioactive material; Inform the Radiological Control Supervisor of any unusual situations, requirements for consumable i supplies, or suspected malfunctions of instruments; and Maintain radiological, auf pment free of defect to facilitate immediate c:: cess and use. l e) Decontamination Supervisor: Schedule, coordinate, supervise, and directly participate in daily work activities; Promote safe work practices; Review and assure accuracy of time sheets; Advise the Project flanager of any abnormal situations as they occur; and Inform the Site Superintendent of all requirements for equipment, tools, and consumable supplies in a timely manner. f) Decontamination Technicians: ( Perform work as assigned in a safe and efficient manner and in verbatim compliance with prescribed procedures. l 1.6 REGULATIONS, REGULATORY GUIDES AND STANDARDS The Dismantling and Decommissioning Operations will be governed by the relevent Federal and State regulations, regulatory guides and standards. These are listed in Table IV. l 0758J (14)

I t The Virginia Tech staff will act in an oversight and advisory capacity to the subcontractor for the dismantling, decontamination, demolition and shipping operations. Virginia Tech and the subcontractor will ensure full compliance with all I Federal and State regulations. All operations will be planned and executed in such a manner as I to minimize radiation exposure of the workers, in keeping with j the ALARA principle. The target values of Table VI will be employed for radiation control during the D8D Project. Items or areas may be released to disposal or salvage above Table V limits, after detailed technical surveys are performed. In no circumstance will limits above Regulatory Guide 1.86 or the NRC staff position of 5 microR/hr at 1 meter be exceeded for releasable equipment or areas. The decontamination operation will ensure that the final condition of the facility complies with the limits specified in Table V. 1.6.1 Quality Assurance For the duration of the decommissioning operations a quality assurance (QA) program will be carried out to ensure conformance with the decomissioning plan's procedures and for procurement of equipment involving personnel and public safety. The quality assurance program will cover the following areas: I a) Review of new decomissioning procedures to ensure i that adequate consideration is given to radiation, j safety, security, QA/ quality control aspects, I reliability and the choice of processes and i material s. l b) Procurement document control to verify that any QA requirements for vital equipment and services are I accurately identified in procurement documents or j purchase orders. I c) Formal documentation control of work instructions and procedures, drawings and information management, including changes as they occur. iI 0758J (15) j I

1.6.2 Q.A. Program The Contractor's Quality Assurance Program meets or exceeds the requirements of the following: o 10 CFR 71 Subpart H o 10 CFR 50 Appendix B o ANSI N45.2 1.7 TRAINING AND QUALIFICATIONS The safety and quality of the Project Plan are paramount to the overall success of the project team at Virginia Tech. The Contractor has developed radiological, industrial, and occupational safety policies and programs which are dedicated to i making all personnel knowledgeable of the hazards associated with the job; this knowledge makes each person responsible for his own safety and the safety of those under his supervision. The Contractor will use its own Field Service Radiological Protection Procedure to ensure that all on-site personnel are properly trained and qualified. (See Section 2.0. ) Contractor I personnel have completed training and qualification programs under this or similar programs. Prior to on-site operations, the Contractor will submit to Virginia Tech records of personnel training. The Contractor's on-site radiation safety organization will supervise all work activities to assure compliance with NRC and Virginia Tech policies. The Contractor's NRC approved Quality Assurance Program will be implemented by project management to meet the requirements of its Corporate Quality Assurance Policy and NRC requirements. 2.0 OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS 2.1 RADIOLOGICAL PROTECTION PROGR#1 Virginia Tech understands that all work must be conducted in conformance with pertinent regulations (NRC, Virginia Tech I requirements) and safe Radiological practices to ensure worker health and safety. To that end, a radiological protection plan j which places worker health and safety as the top priority for I the entire decommissioning project has been developed by the contractor. The radiological control program will meet all the regulatory requirements of the NRC and Virginia Tech, as well as site-specific conditions. 0758J (16) I

During decomissioning, surface contamination will be monitored with suitable alpha, beta and gama sensitive instruments (see Table VII). These instruments detect individual events so direct measurement of detected counts per minute is possible. I From there, a consideration of the subtended solid angle and the detector efficiency will convert counts per minute to disintegrations per minute. These detectors will be calibrated l with sources traceable to NBS. For measurements of dose rate at 1 meter from surfaces, a Pressurized Ion Chamber (PIC) will be used to establish a correlation factor to event counters. This factor will be used to modify the response (readings) of hand-held "microR" meters (NaI scintillators - Eberline PRM-7 or Ludlum Model 19) to the I actual gama spectrum of the facility. 2.2 INDUSTRIAL SAFETY AND HYGIENE PROGRAM The Industrial Safety and @giene Program is equally as important as the Radiation Protection Program. Virginia Tech maintains a strong industrial safety and hygiene program and will take all necessary precautions to assure the safe completion of the dismantling and decomissioning operation. The Project Supervisor is responsible for assuring that safe I working condition are maintained and that the personnel in the working area are properly protected. He will review and approve all operations that employ hazardous operations or materials as defined by OSHA regulations, and will ensure that the plans for dismantling and demolition are adequate to control these hazards, such as airborne debris and hazardous material. The appropriate supervisory personnel will be responsible for carrying out the operations in a safe manner in conformity with the approved plan.

2. 3 CONTRACTOR ASSISTANCE The dismantling of the Reactor, decontamination of the Facility, j l shipping of the radioactive material, and demolition of the i E reactor shield structure will te done by the Contractor. The schedule and specific task to be done by the Contractor will be 3

planned in advance, and detailed work procedures written for the 3 significant operations. In addition to the detailed instructions for performing the work, prerequisites will be i defined in writing before the work can be started, such as health and safety precautions, and the protective clothing requirements. Due to the specific expertise of the Contractor in these areas, he will supply the personnel and the equipment to accomplish these tasks in a safe and expeditious manner. I 0758J (17) I

I 2.4 COST ESTIMATE AND FUNDING The total program cost of the decomissioning operation is estimated to be $600,000. The cost of the contracted work is estimated to be less than $400,000. 3.0 DISMAKTLING AND DECOMISSIONING TASKS AND SCHEDULES The Dismantling and Decomissioning Tasks and Schedules will include the following operations. 3.1 TASKS The Contractor, Chem-Nuclear Systems, Inc., is experienced in I dismantling and decomissioning of research reactors. The following tasks are to be performed by the Contractor, with Virginia Tech personnel overseeing the operations. 3.1.1 Phase I Virginia Tech's management and the Contractor will review all contractual documents and plans to ensure they comply with the requirements of the Virginia Tech Facility License and with all Federal and State I regulations. A predismantling (baseline) radiological survey of the Facility and the environs will be compl ete d. The essential equipment, tools, and I materials needed to accomplish the dismantling and decomissioning tasks will be scheduled for the appropriate time periods. The Contractor will assemble the trained personnel required for the dismantling I tasks. By the end of Phase I, Contractor personnel move to the I Facility. The site will be prepared for the dismantling operation. Equipment, supplies, and approved shipping containers will be assembled at the Facil i ty. 3.1. 2 Phase II Phase II will encompass the removal and disposal of all contaminated and neutron activated material followed by the facility final release survey described in Section l 8.0. All activated and contaminated material of concern is currently confined, almost exclusively, to the reactor shielding structure. I I 075fk1 (18) I

I I Phase II constitutes the major portion of the decontamination and decomissioning activity. Major subtasks include: o Removal and disposal of in-core hardware and graphite shielding blocks, including activated core support structures; I o Removal of thermal column and tank duct graphite, beam port plugs and items embedded in the reactor shielding structure; o Removal of the aluminum tank window and gama curtain sections at both ends of the reactor cavity, including thermal column and core support plates; o Removal of all non-radioactive coraponents and materials from the reactor facility. o Removal of the pump pit and secondary piping and I tank followed by decontamination of the walls as necessary; and o Segmenting, chipping, and removal of steel rebar and activated concrete adjacent to the reactor cavities to a depth of roughly twenty four (24) l inches; o Decontamination and dismantlement of the contaminated fuel storage pit (as required). All work performed in the reactor shield cavities will be closely monitored and controlled to minimize personnel exposure and to preclude the dispersal of contamination to uncontaminated areas of the facility. The Contractor will, with oversight by Virginia Tech, I implement proven techniques and procedures to meet these concerns. Of special concern is the control of personnel exposures. By employing shielding and custom 3 tooling, total crew exposures will be minimized during 3 the most important dose related task: removal of the i core components. Environmental monitoring of controlled areas will be done by continued placement of TLD's at locations surrounding the project site (see Appendix B), and will be in compliance with the Virginia Tech technical I specifications for the site. 3.1. 3 Phase III Upon completion of the decontamination activities,l the final release survey will be conducted, and a fina I report with a request for license termination submitted to NRC. i 0758J (19)

1 Following verification by the NRC that the release limits are satisfied, the NRC releases the Facility for unrestricted use. 3.2 SCHEDULE The tentative Major Task & Milestone Schedule is shown in Figure 9. 3.3 TASK ANALYSES This section describes the approach which has been scheduled to accomplish the most important tasks of the Virginia Tech Research Reactor Project. It has been divided into the I following main headings: 3.3.1 Baseline Radiological Survey 3.3.2 Core Dismantlement 3.3.3 Experimental Facilities Dismantlement 3.3.4 Decontamination and/or Dismantlement of Fuel 3 Storage Wells D 3.3.5 Dismantlement of Process Water System 3.3.6 Activated Concrete Removal

3. 3.1 Baseline Radiological Survey During Phase I, waste characterization data will be collected (in compliance with 10CFR61).

This will include collecting concrete samples from the shield. The Contractor will use several methods to collect the baseline radiological data. Swipe analysis of the I overall facility and specific suspect areas and components will be conducted. Samples will be taken from the concrete shield to confirm activation analysis I calculations and to identify the radionuclide content of the concrete waste. Other radionuclide identification samples will include: o Graphite, aluminum, and steel from the core area o Sludge and rust from the fuel storage pits o Activated lead Additional analyses of ventilation ducting and support area samples will be performed as necessary to locate l and identify contamination in these areas. I 0758J (20) I

Also, based on the isotopes handled at VTAR and previous contamination surveys, there appears to be essentially no potential for alpha contamination. The Contractor will make a cursory alpha survey (direct readings and smears) of random points to confirm the absence of alpha contamination. In the event that alpha contamination is found, further surveys will be performed to define the extent of alpha contamination. All samples will be analyzed off site. Sample analysis will identify weak beta emitters and will establish scaling factors for waste characterization in accordance with 10CFR61. Additional data gathered for all samples will be done by gamma emitter identification to establish the scaling factors for "hard-to-measure" nuclide concentrations from the measured ganna emitter concentrations. The scaling factors developed from the baseline data base will be used to complete the waste shipment manifests and address radiological controls necessary I during operations. Background dose rates will be measured in buildings and I outdoor areas within an approximate quarter-mile radius from the reactor using a Pressurized Ion Chamber (PIC) and microR meter. To determine the background dose rate for Room 10, a number of measurements will be made I in rooms of similar 4 pi concrete geometry. 3.3.2 Core Dismantlement The dismantlement of core components is expected to be executed in a straightforward manner with no attendant mechanical difficulties or unreasonable personnel exposure. To minimize personnel exposures, the dismantlement of core components has been scheduled l early in the overall dismantlement program in order to l u remove the largest sources of radiation before dismantling the experimental facilities and setting up the equipnent to remove the activated concrete. Fo. the removal of the core components, the ALARA principle will be strictly enforced to assure minimum personnel exposure and full regulatory compliance. Crew training in tooling, handling, and task requirements will minimize the time required. Use of local lead blankets will reduce direct radiation exposure and for some tasks long handled cutters and tools will be used to maximize the distance to the 0758J (21 ) i

sources. The anticipated maximum dose rate (1-2 R/hr at one foot) should not be so severe as to restrict the removal of the components. The most radioactive items will.be removed first by employing remote removal I techniques and temporary shielding to reduce working area background radiation as personnel enter the core cavity. After the removal of all non-structural core components and upon local shielding of the steel items enbedded in the concrete, the working background in the core cavity should be in the 25-50 mR/hr range. The total radiation exposure for the core components removal is expected to be on the order of 0.75 Man-Rem. 3.3.3 Experimental Facilities Dismantlement I Removal of the thermal colum and tank duct graphite and support plates can preceed concurrently with the core dismantlement, since the gamma curtain will significantly reduce any interfering radiation from I either source. The two beam tube enbedments will be subsequently removed with the activated concrete. Graphite from both the Thermal Column and the Shield Tank Duct will be removed by hand and then placed directly in approved shipping containers. Experience I has shown that careful handling of the graphite logs can control airborne particulate contamination. After the first graphite logs are removed, in front-to-back sequence, the intake nozzle of a 2000 CFM, I HEPA-filtered exhauster will be inserted in the cavity. The exhauster will establish an inward air flow at the Thermal Column face. Additionally, each I graphite log will be damp wiped as it is pulled from the stack. The operation, including shipment container loading, will be restricted to a surface-protected, controlled surface contamination area. Air sampling I will be conducted continuously. Removal of the Thermal Column support plate will I imediately follow the graphite removal. Unbol ting at the embedded studs and the breaking of the tack welds will accomplish this task. At this point, the enbedded channel iron and studs will be removed by use of a jackhammer. During this operation a 2000 CFM HEPA l filtered blower will provide exhaust for the Thermal Column cavity. I 0758J (22) I

I I After vacuum cleaning the Thermal Column cavity, the ganna curtain will be removed, making one continuous cavity. 3.3.4 Decontamination and/or Dismantlement of Fuel Storage Wells l The fuel storage wells will be manually and chemically (if necessary) decontaminated following removal of shield plugs and fuel element guides. The plugs will be removed, decontaminated, and disposed of as clean waste; guides will be removed by metal-cutting rhisels and disposed of as contaminated waste. Decontamination of the tubes will consist of vacuum cleaning, wire brushing, and cloth swabbing. Additional ef fort, if required, will make use of chemicals for paint and rust removal for getting trace levels of mixed fission products into solution. Should all decontamination efforts fail, it may be necessary to remove the entire storage well assenbly. In such event, the proposed solution would be to extract the tube and concrete matrix as a discrete block by saw cutting the facil.'ty floor at the I perimeter of the storage tube assembly. The block would then be segmented by saw cutting or stitch-drilling into two manageable pieces, each weighing less than 10 tons. Each piece would then be lifted by the overhead crane. 3.3.5 Dismantlement of Prrgess Water System The new secondary water cooling system inside room 10 (which was not used for reactor operations) will be disassenbled and surveyed to verify the absence of contamination. At the point where the piping exits room 10, smears of the interior piping will be taken to I confirm that no contamination is present, and to release the balance of the system. The old cooling system, including all piping in the I pump pit, will be removed, disassembled or cut-up. While probably releasable, all pit piping and coolant return piping will be surveyed prior to disposition. The pump pit, which will be used as a water collection sump during concrete sawing and breaking, will be desludged and decontaminated after concrete removal is complete, to meet release criteria. ) I 0758J (23) I

I All exposed primary piping will be removed. The embedded supply piping is not expected to be contamina ted. 3.3.6 Activated Concrete Removal Activated Concrete will be removed, sectioned, etc., by means of jackhammers, or other techniques, to a depth at which the radiological survey indicates compliance with the criteria of Regulatory Guide 1.86 and 5 microR/hr above background radiation level at 1 meter. I The activated rubble is removed and packaged for dis posal. Removal of the concrete will take place under a secure containment structure to greatly reduce I the chance of airborne contamination spreading to unrestricted areas. Based on the baseline survey, coring of the shield, and a surface dose rate and sample survey of the interior surfaces of the shield, a depth profile and contour map of activated concrete which is above release limits I will be produced. The exterior portions of the shield which are composed of releasable materials (below the Table y limits) will i be removed, surveyed, sampled, and shipped to the Virginia Tech-owned solid waste disposal site. Sampl es I will be analyzed on-site and retained with the survey da ta. This releasable concrete will be buried in the same manner as other construction debris by the University. Concurrent with the work on the reactor biological shield, all remaining areas of the Facility are surveyed for radioactivity and the appropriate action taken to decontaminate these areas. Only low levels of radioactivity are expected for areas outside the reactor structure, with the exception of the west fuel storage pi t. Any contaminated material or waste that is found will be packaged and shipped to an approved reci pi ent. Concurrent with and following the removal of radioactive material from the Facility, radiological surveys will be conducted to determine the levels of radioactivity remaining in the Facility. Based on these surveys, the appropriate actions will be taken to j remove, package, and ship any remaining radioactive M material or waste, until surveys indicate compliance wi th Tabl e V. 0758J (24) I

3.4 SAFE STORAGE There will be no requirement for Safe Storage since all radioactivity will be removed from the VTAR Facility and the I Facility released for unrestricted use. 4.0 SAFEGUARDS AND PHYSICAL SECURITY I The Virginia Tech Security Organization will be responsible for the safeguard and physical security of the Facility. The existing Virginia Tech Physical Security Plan for the Reactor Facility has been modified, to the extent necessary, for the Dismantling Operation. All Special Nuclear Material (SMi) listed under NRC License R-62 has been transferred from the reactor facility. See Appendix C for the Physical Security Plan and Appendix D for the Emergency Plan. 5.0 RADIOLOGICAL ACCIDENT ANALYSIS A discussion of radiological accidents related to fuel handling is not presented here since the Reactor fuel has already been safely transported off-site to approved recipients (Section 6.1). 6.0 RADI0 ACTIVE MATERIALS AND WASTE MANAGEMENT The following sections describe the procedures to be used in handling the radioactive waste generated at VTAR. 6.1 FUEL DISPOSAL Irradiated fuel has been transferred to the Savannah River Pl ant. New (non-irradiated) fuel was transferred to Oak Ridge I National Laboratory. 6.2 RADI0 ACTIVE WASTE PROCESSING I This section discusses waste handling methodologies and estimates of waste volume for the disposal of all contaminated materials associated with the decontamination activities. All I practicable volume conservation techniques will be utilized to minimize waste volumes. The dismantling and decontamination operation is not expected to generate liquid or gaseous radioactive waste. The solid waste is expected to be low level waste and will be packaged in D0T approved metal packaging boxes and wooden box overpacks for packaging the low-level radioactive waste. All packaging and shipping of the waste will be carried out by the contractor with Virginia Tech Health Physics personnel monitoring the operation. I 0758J (25) I

3 Each waste container will be packaged, sealed, surveyed, E labeled, and loaded in accordance with Federal regulations. No radioactive waste other than Class A is expected to be generated in the Virginia Tech VTRR decontamination activity. No hazardous wastes are expected to be generated during the D&D Project. Non-radioactive wastes, including pipe, wood, plastic sheeting, concrete rubble and blocks, steel and other materials which meet Table V requirements will be disposed of at the University-owned landfill, located approximately 3 miles from the reactor. Table VIII lists the expected inventory of radioactive wastes to be shipped from the VTAR facility. 7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS I The Technical Specifications ( Appendix A), the project organization, adherence to 10 CFR 20 guideliness, and compliance with all State and Federal requirements for health physics, industrial safety, and shipping assure that the dismantling and decomissioning tasks will be accomplished with no significant impact on the environment or the health and safety of the public. 8.0 PROPOSED TERMINATION RADIATION SURVEY PLAN The final release survey will conform to NUREG-2081, " Monitoring for I Compliance with Decomissioning Survey Criteria." The survey will use a formalized procedure based on a sampling inspection plan to assure adequate coverage and statistical analysis of the results to provide efficient use of the data. 8.1 The Facility will be divided into three types of areas for data collection and analysis: o High Potential Contamination Areas o Medium Potencial Contamination Areas o Low Potential Contamination Arear 8.1.1 Areas of low contamination potential are those areas where previous Virginia Tech survey results indicate contamination below release limits and the function of that area was not conducive to creating a contamination I problem. Areas included in this description are the control room, offices, and all ceilings. These areas will be sectioned into 10' x 10' (3m x 3m) grids on the I floor, walls, and ceiling. These grids will be surveyed (as described below) on a statistically random 0758J (26)

l basis. A total of 30 grids will be surveyed in these u areas to meet statistically satisfactory release cri teria. 8.1. 2 Areas of medium contamination potential are those areas where either Virginia Tech survey results indicate contamination near or above release limits and/or the l function of that area constitutes a potential for contamination. Areas included in this description are room 10 floors and walls and the Chemistry Lab and adjacent rooms. Subsequent to decontamination, these areas will be sectioned into 6' x 6' (2m x 2m) grids on the floor and up the walls. Approximately 30 grids will be surveyed on a statistically random basis. In addition to the 30 grids surveyed on a random basis for each of the low and medium potential contamination areas, approximately 10 specific grids will be surveyed on a stratified basis. These grids will be in (1) doorways and room corners, (2) near penetrations or obstructions, and (3) specific areas which have I previously indicated contaminated materials or equipment. 8.1. 3 Areas of high potential contamination will be the reactor shield floor area, remaining structures, and pump-process pit. Subsequent to decontamination these areas will be sectioned into 3' x 3' (im x 1m) grids; all of these grids will be surveyed. 8.2 The survey procedure will be in accordance with NUREG-2082. I Each designated grid will be surveyed in the following manner, A gamma (micro-R/hr) point reading will be taken at 1 meter o above the center of the grid using a gama scintillatien detector. Beta-gama and gama contact readings will be taken at five I equally spaced points within the grid. o A G-M beta-gama surface scan survey of the grid will be o I conducted and the maximum beta-gamma point identified and surveyed. o A smear survey and gama dose rate survey will be taken at the maximum beta-gamma point. The smear will be counted for gross beta-gama activity, The five point surveys will be recorded and the average of o the five point recorded as an unbiased measurement. The l measurements at the beta-gama maximum report will be recorded as a biased measurement. l l l 0758J (27) 1

8.3 The procedure for surveying pipes, drain lines, and ductwork will be basically the same. The nearest accessible locations will be surveyed with suitable instruments and wiped for removable contamination as far inside as can reasonably be reached. These locations will include the ends of accessible pipes, the interior surfaces of inlet and outlet vents, and the water traps and exit points of drain lines. If no I significant radioactivity is found at the entry and exit locations, a " wipe" will be pulled through representative pipes, drain lines, and duct work (if possible) to determine internal contamination levels. If, on the other hand, contamination exceeding limits established by Regulatory Guide 1.86 is found, cach such pipe, drain line, and duct work will be " wiped" through its entire length, or removed. 8.4 All instruments used for these surveys will be calibrated in a manner acceptable to the NRC. All calibration sources are traceable to the National Bureau of Standards. Use of a scintillation (event) counter for dose rate measurements will be correlated with a Pressurized Ion Chanber. l 075&) (28) i

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FIGURE 5 -I Booster Fan i Centrifugal Blower I Roof

1. Stack Monitor

2. Sample Point

3. TLD

4. APFPM 14

5. TAP

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6. Rabbit System Exhaust Room 308

7. Reactor Room Ventilation Plenu

8. Dumo Tank Booster Fan Duct

9. Cor'e Ventilation Fan Duct

10. Fume Hood i

I Room 208 2 3-Room 108 Room 10 (Reactor Room) 4 5 \\ (3 [ 9 l ,7 I i 6 U [ i0 8 Reactor Room 8 Damper Valve I VTAR VENTILATION CHARACTERISTICS COMPONENT DESCRIPTION Blower Fan 855 rpm Centrifugal Exhaust Unit Blower Fan 855 rpm Vain-Axial Booster Fan Blower Fan 1725 rpm, Dump Tank Centrifugal Booster Fan 1/8 hp (Process Pit) l. Total Flow Rate 12,700 cfm at Stack Discharge l I

M M"M M M M M M (upper) Operating Closure l I I I I I I

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( Dump Tank Booster Fan 1 l Core Overflow Drain 1 i l Schematic Diagram of Purge Points

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8 FIGURE 8 O Virginia Polytechnic Institute Argonaut Reactor (VTAR) Deconmissioning Project Organization UNIVERSITY ADMINISTRATION I I Health & Safety Dept. (Quality Assurance) Health Physics PlanningServices[ I (RS0) l 1 I 1 e Crafts & Labor I Reactor Safety g I ,g Corsni ttee 1 -..____...........____......d...____....______..______......__....d_____..___...___..__.__.__.II I i t s Contractor Personnel j Decomissioning Project Ma nager (Contractor) s I I I i sEMIOR fHOJCCI (MasettR

M OsreC[ F W ir N

+ o Caerecal l o F inance & ace n. str at eoa l I o Legal Counses I s e IL SUP(H emet (edt NT o Conteact Aame n & s t r et ioa i o Oue 8 * 't A= sur e N e 8 o Licenseng g I o ee-e s t a en. s e c s l f l I I#5 R AD IOL OGI CAL CO* T HOL St# v Equipment Hentets o I DLCONIM*INAllON $UPLHy8$UH a (g,, o Meter t al s/Suppe ter o Transpor tet ton Ofsposea o H. P. I( Cted e C e am I I F-20

M M l M AJOR Mll.ESt%ES ( 3

1. CONTR ACT AW ARO 1

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2. D & D PLAN SUBMITTED l

FIGURE 9

3. PREPARATION COMPLETE MAJOR TASK AND MILESTONE SCHEDULE

"^""'"' **" VTRR DECOMMISSIONING PROJECT "^" ' " " **""'

e. PIN AL REPORT SUSMITTED w

J DATE JUNE 16 AUG.18 O CT.2 7 JAN.S FE8.11 12345678910,1112131415161718192,021222324252627282930313233343530 WEEK NO. n i I I l l 1 l l l PHASEI l PHASE 11 PHASE Ill l N SEPT.so PRGJECT PREPARATION g l a h'""' ** NRC REVIEW & APPROVAL g l -l 0& DPLAN l l in WEEx3 i i ON-SITE MOttLIZATION l g i i l l l I l oCT. 7 Nov.17 l PACILITIES OfSMAfffLEMENT l g { g l l @oEC.io l CONCRETE oEMouTioN l I g l I I o E e,,7 l DECONTAMINATION l il l l I l I RELE.S E S-ETS i l I i l i I i l RADWASTE SMIPPING l l l g j DEMOttL12AT80N l l JAN.14 I PIN AL REPORT l ~

E< d NUMBER TITLE PAGE I VTAR DEC0tNISSIONING PROJECT MAN-REM ESTIMATE T-1 II VTAR SPECIFICATIONS /0PERATING HISTORY T-2 III VTAR RADI0 ACTIVE MATERIAL DISCHARGE / DISPOSAL HISTORY T-5 IV REGULATIONS AND STANDARDS T-6 V ACCEPT /0LE RADIOLOGICAL CONTROL RELEASE LIMITS T-7 VI CRITERIA FOR RADIOLOGICAL CONTROL 0F EQUIPMENT & MATERIAL T-8 VII INSTRlNENTATION TO BE USED DURING THE VTAR DEC0bt11SSIONING T-9 l VIII VTAR DECOMMISSIONING RADI0 ACTIVE WASTE / SHIPPING SCHED'JLE T-10 I I I I I I I 0758J (30) I I<

I TABLE I Virginia Tech Decontamination / Decommissioning Estimated Personnel Dose Equivalent I TASK COLLECTIVE MAN-REM Fuel Removal

0.25 Miscellaneous Equipment Removal
0.10 Baseline Survey 0.07 Core Cavity Material Removal & Packaging 0.75 Concrete Removal 0.35 Routine General Area Exposure (s) 0.10 TOTAL 1.62
Indicates that task has already been completed.

l 1 'I l 0771J T-1 l l

Table I I A Nominal VTAR Characteristics General Features Reacto r Type............................. He te rogeneous, Thermal Maximum Power Level...................... 100 kw Maximum Thermal Neutron Flux in Central I 12 2 Irradiation Stringer..................... 1.3 x 10 N/Cm see Excess Reactivity (@ 750F )..............

0. 60% &/k maximum clean, Cold Critical Mass................

3031.45Ggams I Prompt Neutron Lifetime.................. 1.3 x 10- see Vo id Co e f f i ci e n t......................... -0.184% &/k/1% Void Temperature Coefficient.................. -0.004% &/k/ F I Re f le c to r................................ Graphi t e Mod e ra t o r................................ Pu re Wa te r, Gra phi t e Startup Source........................... 1 Curie Pu-Be Fuel Fue1.................................... 93% Enriched, U-Al Fuel Loading............................. Approximate ly 3. 00 Kilograms Pl a t e Thi ckn e s s.........................

0. 8 0 i n Plate Width.............................

3.0 in Pla t e Le ng th............................ 2 6. 0 i n Plate Separation......................... 0.40 in Cladding Thicknes s......................

0. 020 in 235 Fuel Composition.........................

14.0 w/o U Primary Coolant Typ e..................................... De mi ne ra li z e d Li g h t Wa t e r I Flow @ 100 kw............................ 42 GPM US (Nominal) Equilibrium Core Inlet Temp. (@l00 kw).. 840F Equilibrium Core Outlet Temp.(@l00 kw)... 100 F Secondary Coolant Ty pe..................................... H 0, Po t a b le, Tr e a t e d 2w/ Potassium Chromate Flcw @ 100 kw............................ 400 GPM US I Equilibrium Heat Exchanger Inlet Temp.... 70.0 F Equilibrium Heat Exchanger Outlet Temp.. 71.7 F Control Blades i i Type..................................... Boral, Windowshade l Number.................................. 3 Safety, 1 Regulatfag Material................................. Boral Shim & Safety Rod Dimensions............. 7" x 7" x 0.125" Regulating Rod Dimensions................ 2" x 2" x 0.025" Shim & Safety Rods Worths................ Approximately 0.65% ak/k Regulating Rod Wo rth..................... App roximate ly 0.12% &/k Maximum Reactivity Insertion Rate........ 9.02% &/k/sec reactivity for the VTAR is limited to 0.6% ak/k. Calculations by Tuley and SPERT data indicate that an uncontrolled reactor period of 3.2 msec is required to produce the heat necessary to melt the fuel. A step insertion of 0.6% ak/k will result in a period of 200.0 msec at the T-2

TABLE IIB VIRGINIA TECH REACTOR OPERATING HISTORY (Reactor went from 10 kw to 100 kw in late 1966) YEAR KILOWATT-HOURS I

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

1959 0 1960 63 1961 1,699 1962 1,511 1963 1,521 1964 2,266 I 1965 4,573 1966 4,781 1967 60,911 1968 74,219 1969 53,653 1970 73,525 1971 44,934 I 1972 92,170 1973 74,971 1974 94,893 I 1975 127,736 1976 124,220 1977 113,837 1978 90,174 1979 105,846 1980 70,293 1981 45,422 1982 4,636 1983 9,097 1984 0 1985 0 1986 O TOTAL KW HOURS 1,276,951 I T-3 I I

I TABLE IIC Materials Incorporated in VTAR Component Materials Used A. Beam Port and Plugs Carbon Steel I Mild Steel Ordinary Concrete B. Concrete Shield Aluminum Alloy G2 42-A-I Aluminum 6061-T6 Concrete-Class C 4211 Rubber Coated Packing I Grout-Embeco Premixed Mild Steel Wood C. Core Support Plate Aluminum 6061-T6 Reactor Grade Graphite D. Core Support' Plate Grout-Embeco Premixed Hardened Steel Mild Steel E. Core Tank Aluminum Alloy G-5 11A-T6 Hild Steel F. Gamma Curtain Aluminum 1100-F Aluminum 6061-T6 Lead Mild Steel G. Graphite Shield Tank Duct Aluminum 6061-T6 Grout-Embeco Premixed Mild Steel Nuclear Grade Graphite l H. Graphite Stringer Aluminum 6061 3 Aluminum 6061-T6 Reactor Grade Graphite I. Thermal Column Door Boral Grease High Density Concrete I Lead t Mild Steel ~ J. Thermal Column. Support Plate Aluminum 6061-T6 Grout-Embeco Fremixed Mild Steel l l T-4

M M M M M M c= TABLE III. Liquid Waste Released from VTAR to Sanitary Sewerage System 1977 to 1983 Solid Waste Disposal, VTAR (Nov. 1975 - June 1983) Date Volume of Nuclide Activitv Water ( pC1/m 1) Radionuclide Quant f.ty '(mci) (liter) Co-60 39.654 3-25-77 9117.4 s-35 1.7x10 Zn-65 13.592 3-21-78 113.4 Ho-166 3.6x10-6 Ag-110 13.050 3-9-79 37.8 Co-60 2.4x10-6 Th-232 0.0902 9-5-79 75.6 Unknown 5.9x10-11 Cr-51 0.0001 9-12-79 37.8 Unknown 2.5x10-11 Eu-154 0.0010 9-26-79 75.6 Unknown 1.2x10-9 Cd-109 0.0042 10-11-79 73.6 Unknown 8.7x10-10 Cd-113 0.0460 5-2-80 34.0 Na-22 5 x 10-7 Ba-133 0.0562 2-25-81 109.6 Unknown 1.3x10-8 Fe-59 0.750 3-4-81 94.5 Unknown

1. 0x 10-8 Ra-226 0.00003 4-9-81 680.4 Na-24 2.5x10-7 Po-209 0.0000012 5-21-81 7098.8 Na-24 1.8x10-7 Lo-176 0.00006 3r-82 5.0x10-8 Cs-137 0.1002 12-1-82 8316 Unknown 2.7x10-3 Ac-227 0.00003 y

7-22-83 8316 Unknown ND+ Pb-210 0.000056 7-26-83 831.6 Unknown ND Co-57 0.000132 m Na-22 0.0001 non-detectable Cs-134 0.1 Eu-152 0.050 Liquid wastes that cannot be disposed of in the sanitary sewerage TOTAL 67.494 system are placed in suitable containers and stored on site until arrangements for off site disposal have been uude. The liquid is then Ar-41 Stack Discha rge Rate * (Ci/sec), VTAR 1978 to L983 %,,,,. s Ar-41 Stack Discharge Yearly Totals, VTAR Date 1978 - 1982 Discharge Rate (Ci/sec) Year Ar-41 (Ci/yr) kWh 10-25-78 3.00xto-5 4-27-79 3.28x10-5 10-23-79 4.86x10-5 1978 144.7 90,174 11-29-80 1979 169.9 105,846

4. 7 7x 10-5 11-3-80 1980 113 70,278 1.63x10~5 5-15-81 1981 73.6 45,845 5.48x10-5 8-9-32 4.79x10-5 1982 7.4 4,636 1-26-83 3.29x10-5

flow rate of 1x106 ml/sec

TABLE IV Regulations & Standards Related To Decommissioning Of The Virginia Tech Research Reactor Agency Regulation / Standard Description Federal: Radiation Protection NRC 10CFR20 Radiation Protection Standards 10CFR50 Domestic Licensing of Production & Utilization Facilities 10CFR71 Shipping of Radioactive Material Reg. Guide 1.86 Decontamination Limits D0T 49CFR Shipping of Radioactive Material EPA NEPA Preparation of an EIS OSHA 29CFR Worker Health & Safety Guidelines NRC NUREG-1756 Decommissioning of Research & Test Reactors NUREG-2082 Termination Survey Criteria NUREG-0586 Generic EIS for Decommissioning Nuclear Facilities Staff Position " Guidance and Discussion of " Guidelines" Requirements for an Application to Terminate a Non-Power Reactor Facility Operating License", Revision 1,1984 ANSI ANSI /ANS-15.10 Decommissioning of Research Reactors ANSI /ANS-N13.12 Decontamination Limits I 0771J T6

Table V ACCEPTABLE SURFACE CONTAMINATION LEVE13 b b NUCLIDEa AVERAGE c MAXIMUMbd REMOVABLE e U-nat, U-235, U 238, and 5,000 dpm a/100 cm2 2 2 15.000 dpm a/100 cm 1,000 dpm a/100 cm associated decay products Transuranics, Ra 226, Ra-228, 100 dpm/100 cm2 2 300 dpm/100 cm 20 dpm/100 cm2 Th-230, Th-228, Pa-231, Ac 227, I 125,1-129 2 2 Th-nat, Th-232, Sr-90, 1000 dpm/100 cm 3000 dpm/100 cm 200 dpm/100 cm2 Ra-223, Ra-224, U-232, I126,I131,1133 2 2 Beta-gamma emitters (nuclido. 5000 dpm #9/100 cm 15,000 dpm #9/100 cm 1000 dpm Sq/100 cm2 with decay modes other than alpha emission or spontaneous fission) except Sr 90 and others noted above-aWhere surface contamination by both alpha-and beta-gamma-emitting nuchdes exists, the hmits estabhshed for alpha-and beta-gamma-emitting nuclides should apply independently. DAs used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. CMeasurements of average contaminant should not be sveraged over more than 1 square meter. For objects of less surface area, the average should be derived for each such object. 2 dThe maximum contamination level spplies to an area of not more than 100 cm, 2 'The amoant of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount cf radioactive material on the wipe with an appropriate instrument of known efnciency. When removable contamination on objects of less surface area is determined, the pertinent leveh should be reduced proportionally and the entire surface should be wiped. Maximum Radiation Level (at 1 meter from surface) 5 JR/hr (above background) l I T-7 l l

l I TABLE VI CRITERIA FOR RADIOLOGICAL CONTROL OF EQUIPMENT AND MATERIAL UNRESTRICTED USE I TYPE ACTIVITY TOTAL REMOVABLE 2 2 Alphas 100 dpm/100 cm 22 dpm/100 cm I 2 Beta-Gammas 0.1 mrad /hr at 220 dpm/100 cm 2 0.1 cm thru 7 mg/cm I absorber LINITED TO POSTED CONTAMINATED AREAS 2 500 dpm/100 cm2 Alphas 2500 dpm/100 cm 2 Beta-Gammas 1 mrad /hr at 2500 dpm/100 cm 2 1 cm thru 7 mg/cm absorber Equipment and material with activity between these limits shall be evaluated I and reviewed in each specific case. I I I 0771J T-8

TABLE VII RADIATION DETECTION INSTRUMENTS WINDOW TYPE OF RADIATION SENSITIVITY THICKNgSS INSTRUMENT NtNBER DETECTED RANGE (mg/cm ) USE Eberline ES20 2 Beta, Gamma 0-2K mR/hr 30 Survey Eberline F120 2 Beta, Gamma 0-50K cpm 1.4-2.0 Survey Eberline ESP-1 1 Alpha, Beta, Gamma Various N/A Survey Eberline RM 14 2 Beta, Gamma 0-50K cpm 1.4-2.0 Survey Eberline RAS-1 2 Alpha, Beta, Gama N/A Air Sampler Eberline AMS-3 1 Beta, Gamma 0-100K cpm 1.4-2.0 Air Sampler Eberline PAC 4G 2 Alpha 0-50K cpm 1.4-2.0 Survey Eberline PRM-7 1 Gamma 0-5 mR/hr N/A Survey Eberline PIC6A 1 Beta, Gampa 0-lK R/hr N/A Dose Rate Eberline SPA-3 1 Gamma Detector N/A Detector Ludlum 19 1 Gamma 0-5 mR/hr 30 Survey Staplex 2 Alpha, Beta, Gama N/A fl/A High Volume Air Sampler Teletector 1 Peta, Gamma 0-lK R/br N/A Survey Tennelec LB 5100 1 Al pha, Beta, Gamma 0-999999 dpm 0.5 Survey Victoreen 495 2 Beta, Gamma 0-50K cpm 1.4-2.0 Survey (Frisker) Keithley 36100 1 Beta, Gamma 0-20 R/hr Survey Victoreen 856-20 3 Gamma 0.1-104 mR/hr N/A Area Monitors Reuter Stokes PIC 1 Gamma 1-999 uR N/A Low Level l Gamma Surveys 0771J T-9 I

i TABLE VIII RADI0 ACTIVE WASTE SHIPPING SCHEDULE ESTIMATED RAD LEVEL ESTIMATED ESTIMATED AT CONTACT CUBIC FT. WEIGHT OF DISPOSAL INCLUDING INCLUDING SHIPMENT / UNIT CONTENTS CONTAINER PACKAGING PACKAGING 1 / Box Pb Bricks, Vertical Beam 20 mR/hr 96 8978 Tube Plugs (2 each), Drum of lead Instrument Assay, Tank Window, Thermal Column Graphite Pieces 1 / Item 2 Thermal Column Door 0.5 mR/hr 63 7820 1 / Item 3,4 Inner Core Plugs (2 400 mR/hr 67 12,160 each) 1 / Item 4,5 Outer Core Plugs (2 40 mR/hr 70 12,780 each) Shipment 1/ Totals 296 41,737 2 / Box 2 Remaining Thermal 20 mP/hr 96 6333 Column Graphite Pieces E 2 / Box 3 Beam Tubes, 2 each, 96 8679 !E Concrete Slabs, 3 each, Concrete Feinforced Slabs, 4 each, Pb Brick, Small I Concrete Block, Pit Piping, TC Support Plate 2 / Box 4 Tank Duct Graphite, 200 mR/hr 96 7680 Gamma Window, Core Vetal, CRD Plugs, Concrete 2 / Box 5 Core Graphite 100 mP/hr 96 6000 2 / Box 6 Ccncrete 15 mR/hr 96 8100 2 / Item 6 Tank 0.5 mR/hr 45 300 Shipment 2/ Totals 525 37,09? 0771J T-10

TABLE VIII Continued RADI0 ACTIVE WASTE SHIPPING SCHEDULE ESTIMATED I RAD LEVEL ESTIMATED ESTIMATED AT CONTACT CUBIC FT. WEIGHT OF DISPOSAL INCLUDING INCLUDING SHIPMENT / UNIT CONTENTS CONTAINER PACKAGING, PACKAGING Shipment 3/ Box 7-11 Concrete 15 mR/hr 480 42,500 Shipment 4/ Box 12-16 Concrete 15 mR/hr 480 42,500 I 1781 163,829 Total Project Radioactive Waste I l l 1 I 0771J T-11

l [ h lI "^"' ^ ^'-"' I B ENVIRONMEMTAL RADIOLOGICAL SURVEY B1-6 C PHYSICAL SECURITY PLAN Cl-4 l I I I I I I I I 'I I 0758J (31 ) l< J

lA I APPENDIX A TECHNICAL SPECIFICATIONS FOR THE VIRGINIA POLYTECHNIC INSTITUTE AND I STATE UNIVERSITY RESEARCH REACTOR DECONTAMINATION AND DECOMMISSIONING FACILITY LICENSE R-62 DOCKET N0. 50-124 I JULY,1986 I I I

I I

I p I

iEN0 MENT NO. 7 A-1 0803J I

i

I I m m 0,c0 m N1S g ouMeER Sect 10N ,AeE 1.0 DEFINITIONS A-3 2.0 LIMITING CONDITIONS FOR DISMANTLENENT & A-4 I DISPOSITION LICENSE 3.0 RADIATION MONITORING A,-5 4.0 BUILDING EVACUATION PLAN A-5 5.0 SURVEILLANCE REQUIREMENTS A-6 6.0 DESIGN FEATURES A-6 7.0 AD41NISTRATIVE CONTROLS A-7 FIGURE A DEC0l411SSIONING PROJECT ORGANIZATION A-8 I I I I I

iEND4ENT NO. 7 A-2 0803J

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

1.0 DEFINITIONS 1.1 ABNORMAL OCCURRENCE I An abnormal occurrence is any one of the following: A. An observed inadequacy in the implementation of I administrative or procedural controls, such that the inadequacy causes or could cause an unsafe condition with regard to reactor radiological practices. B. A malfunction of the instrumentation such that a release of radioactive material could occur or a release could occur without being detected. C. An accidental or unanticipated release of radioactivity to the environment in excess of the limits specified in I 10 CFR, Part 20. 1.2 CONFINEMENT I A closure of the reactor room cell (room 10) which controls the movement of air into and out of the cell through a controlled pa th. 1.3 OPERABLE A system or component which is capable of performing its intended function in its normal manner. 1.4 OPERATING A systen or component which is performing its intended function in i ts normal manner. 1.5 REACTOR DISABLED The reactor is considered disabled whenever: A. The means of placing and maintaining moderator in the core area is removed and, B. The control rod drive mechanisms are physically disconnected from the control rods in the core and, C. All control rods are fully inserted and, D. The fuel is not in the reactor core. I MENOMENT N0. 7 A-3 0803J l l I

I 1.6 CONTROL R0D A device fabricated from a neutron absorbing material which is used to control the rate at which fissionable material in I reactor fuel reacts. In the present case, the control rods will remain fully inserted in order to assure that the reactor fuel would remain subcritical, even if the fuel were loaded into the I Core. 1.7 SHALL, SHOULD, OR MAY The word "shall" is used to denote a mandatory or required action or condition; the word "should" a recommended or desirable option; and the word ";aay" a permissible or discretionary course, without bias. 2.0 LIMITING CONDITIONS FOR THE DISMANTLEMENT AND DISPOSITION 2.1 REACTOR FUEL There is no fuel at the reactor facility and at no time will any I fuel be placed in the reactor core assently. 2.2 CONFINEMENT I A. Normal Confinement Operations that require normal confinement are operations in which irradiated or contaminated equipment or materials are being handled. B. Conditions to Achieve Normal Confinement The existing conditions necessary to achieve normal I confinement are all reactor room accesses are maintained closed and lccked except as necessary for controlled personnel movements. C. Emergency Confinement Emergency confinement shall require as a minimum: 1. The reactor room shall be evacuated and all means of access closed and locked. 2. Robeson Hall in which the reactor is located shall be evacuated and reentry prohibited, except for emergency I response personnel, until such time as the emergency condition is cleared. 3. All ventilation systems used in the reactor area will be turned off. /NENDMENT NO. 7 A-4 0803J I

3.0 RADIATION MONITORING Sufficient portable radiation survey instruments shall be maintained at the facility to permit monitoring of personnel and the facility to assure safety of reactor personnel and the public. Demonstration of the absence of escaping radioactive material will be made by swipe and area surveys at the points of egress from the reactor room. 3.1 PORTABLE RADIATION MONITORS Portable radiation monitoring equipment shall be maintained at the reactor facility during the period of " Dismantlement and Disposition Status" and thereafter at the Radiation Safety Of fice, a devision of Safety and Health Programs. This, instrumentation shall consist of: A. A beta-gamma instrument capable of measuring dose rates of 0.1 to 20,000 mrem per hour. B. A neutron instrument capable of monitoring dose rates of 0.1 to 10,000 mrem per hour. C. A portable air sampler for collecting airborne activity. D. A G-M type "frisker" or equivalent for personnel and equipment surveys and for determining airborne particulate activity level s. E. An Eberline Model PAC-4G low-level alpha counter. 4.0 BUILDING EVACUATION PLAN The building shall be evacuated whenever the following condition exists: A. A condition exists which, in the opinion of a responsible person, e.g. a member of the reactor staff, a radiation safety person, contractor personnel, or a member of the Reactor Safety I Connittee, would pose a danger to personnel within the reactor facility or members of the general public, would result in evacuation of personnel under directions of those persons listed above. 4.1 ACTIONS When a building evacuation occurs, the following actions will take place: I AMENDMENT NO. 7 A-5 0803J I

A. Facility operations will imediately cease; all personnel within the reactor laboratory area will imediately evacuate, and assist in the evacuation of personnel from the remainder of Robeson Hall. The doors to the facility I l shall be locked. The VPI&SU Police will be notified and will respond bringing an emergency kit and will prevent anyone from reentering the building. If the emergency occurs when no one is present within the reactor facility, the Police will notify Safety and Health Department Head and radiation safety personnel from an emergency call list I maintained by the police. Reentry of the facility will be only by a decision of the radiation safety person present. Due care will be taken to insure the protection of all individual s. Only after it is proven that it is safe to do I so will occupancy of the portion of Robeson Hall be' permi tted. 5.0 SURVEILLANCE REQUIREMENTS The general conditions within the reactor room will be inspected quarterly with a walkthrough type inspection for obvious damage. 6.0 DESIGN FEATURES 6.1 SITE The Virginia Tech Argonaut Reactor is located in Robeson Hall of I VPI&SU. Robeson Hall is located in the northwest corner of the main campus. Buildings located within a 500 foot radius include Davidson, Williams, Derring, Cowgill, Burruss and Pamplin. Additions are planned to Robeson and Pamplin within the next year. 6.2 REACTOR FACILITY The area to be maintained secure under the " Dismantling and Disposition Status" will be Room 10, and portions of Room 108. Other areas currently within the r estricted area may be made available for other use by constructing walls or by providing alternate access. Rooms 108, 6, 8, and 10 will be included within the area defined as the reactor facility. I

iENDMENT NO. 7 A-6 0803J I

7.0 ADMINISTRATIVE CONTROLS The overall responsibility for the reactor while in " Dismantling and Disposition Status" has been assigned to the Vice President for I Administration and Operations. He has assigned the responsibility for administering the facility to the Head, Department of Safety and Heal th. The Reactor Safety Comittee has the responsibility for monitoring reactor activities tc assure compliance with NRC regulations, provisions of the terms of the reactor license, and these Technical Specifications. A Radiation Safety specialist is assigned to the reactor. The organizational chart (Figure A) reflects the organization which will exist during dismantlement and disposition. 7.1 REACTOR SAFETY COMMITTEE The Reactor Safety Comittee will remain a separate subcommittee of the University Radiation Safety Committee. The Reactor Safety Comittee will be chaired by an administrator familiar with radiation and appointed by the Vice President for Administration and Operations. There will be at least 5 other menbers of whom one will be the Head, Department of Safety and Health Programs, and the Reactor Safety Of ficer. The others will be chosen according to their background in radiation applications, engineering or other relevant disciplines. In each case, a quorum is defined as half the total membership plus one. Decisions of the comittees require a majority of the menbers present and voting. 7.1.1 MEETINGS The Reactor Safety Comittee shall meet quarterly or as needed, or by request of any Committee member. 7.1.2 REPORTS The reactor staff, when one exists, and the Reactor I Safety Of ficer shall make a report to the Reactor Safety Comittee or its successor, at each meeting, of reactor activities, radiation exposures, and transfers and discharges of radioactive materials during the I preceeding quarter. 7.1. 3 RECORDS Records shall be maintained of all information pertinent to the facility under " Dismantlement and I Disposition Status" to allow others at a later time to determine the function and condition of the facility at such time as the facility is finally decomissioned. i l

iENDMENT NO. 7 A-7 0803J I

e h Virginia Polytechnic Institute Argonaut Reactor (VTAR) Deccnmissioning Project Organization UNIVERSITY ADMINISTRATION Health & Safety Dept. I _.e (Quality Assurance) lPlanningServices! Health Physics (RS0) e I t i Crafts & Labor I Reactor Safety I 4 Comittee i 1 L----.--.----..-------.--.. I I Contractor Personnel * { Decomissioning Project I I Manager (Contractor) i t i staim twoxcr to =tte 1 r- \\I o c,e,.< e f .< w cn act ru~crio~s o r.a.oce a ao.i..ser.eio. l l o t.9.i c~.s i i siit sueta. :r.u ur o c,,,,,,,, u....s i i o w a s... as s.,.y. e t.<cas.ng i o o .....e.s.,s e I i I surominac tms arinai oc.cm cc,.swot sun. " tw p-..e n e.es a o U UMIN' *tAIGON sUPtRWGsUN 'I g,,, o M.e., s.s ses ,e i., o t<..sewe.,no. o os sen. s . e. ri o sc e a. I n-8 lI

lA I I I I caxuets iso otterious 10 18e rec 8 mica < sgec1ric 11ous lI I L1 CENSE R-62 15 JULY 86 DOCKET 50-124

4ENDMENT 7 I

I I I I I I I I

License R-62 15 July 86 Amendnent 7 I CHANGES AND DELETIONS TO THE TECHNICAL SPECIFICATIONS I PARAGRAPH REASON 1.1.C. Delete since no fuel is on-site

1. 2, 1. 3, 1. 4, 1.5, 1.7 These definitions are not necessary because no devices of this type are necessary during dismantl ement, so delete.

2.2 Control Rods - Control rods will be disposed as radioactive waste. All fuel has been removed - dele te. 2.3 Dump Valve - The dump valve will be removed during dismantlement - delete.

2. 4. A.1 Containment - No irradiated fuel will be handled I

- del e te. 2.4.B.2 This condition is deleted since during I dismantlement, the contractor will provide localized liEPA filtered exhausters and/or localized containment for those operations which could generate airborne contamination. 2.4.C.3 Delete since this system will not be operating.

3., 3.1, 3. 2 Emergency Power - No systems which require emergency power are necessary.

Radiation monitoring equipment usod during dismantlement are battery powered - delete this section. 4.1.A,B,C Deleted sections which apply to fixed monitors due to authorization to deactivate system per I Paragraph 4. 4.2 Change " Possession Only Status" to " Dismantlement and Disposition Status".

5. A,B Change: Automatic System will be dismantled and replaced with administrative controls.

(1 ) 0802J

License R-62 15 July 86 Amendnent 7 I CHANGES AND DELETIONS TO THE TECHNICAL SPECIFICATIONS PARAGRAPH REASON

5. C. (Para.1 )

Add " contractor personnel", delete "a manual I initiation of the evacuation alarm" and replace with " evacuation of personnel under direction of those personnel listed above."

5. C. (Para. 2)

Delete due to dismantlement of the systerii. 5.1 Change " building alarm" to " building evacuation." 5.1. A Delete " automatically" and replace with " man ual ly". Delete " alarm" and replace wi th " emergency". bbdify organizational titles to reflect current personnel titles.

6. A,B,C Delete in entirety since all fuel has been I

removed. [ 7. Delete since al1 fixed instrumentation wil1 be l dismantled. 7.1 Delete "to maintain integrity of reactor systems."

7. 2. A,B Delete in entirety since the system will be disman tl ed.

Ventilation control during dismantlement will be provided by contractor

I personnel and portable HEPA filtered exhausters used with local containment.

I 7.3 Emergency Power - Delete since no electrically l powered systems are needed to issue containment l during power outages, f 7.4 Building Evacuation System - Delete because system will be dismantled. 7.5 Padiation Monitoring System - Delete since these j systems will be dismantl ed. 8. Delete " Design features" and the following l l sentence, since changes to the system will be made during removal. l l l (2) l 0802J l

License R-62 15 July 86

.iendment 7 CHANGES AND DELETIONS TO THE TECHNICAL SPECIFICATIONS PARAIRAPH REASON 8.2 Reactor Facility - Change " Possession Only" to "Dismantl ement and Disposition", and delete references to fuel.

Delete reference to roof since it has already been elimated from vertical area. 8.3 Reactor Coolant - Delete in entirety since all coolant has been removed. 8.4 Shield Tank - Delete since the tank is empty and will be removed. 8.5 Fuel Storage Pits - Delete since the fuel has been removed. 9. Administrative Controls - Change " Possession Only" to " Dismantlement and Disposition", and change personnel position titles to reflect corrent organization, and absence of fuel. 9.1 Reactor Safety Committee - Revise entire paragraph to reflect safety organization during I dismantlement and disposition. 9.1.1 Delete "or its successor" since current organization is successor 9.3 Change to " Dismantlement and Disposition" and revise paragraph to reflect decomissioning the facility as only option under consideration. 9.4 Revise entire organization chart to reflect current 0 & D staffing plan. l FWG:sl (3) l 0802J l

r APPENDIX A I TECHNICAL SPECIFICATIONS FOR THE VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY RESEARCH REACTOR FACILITY LICENSE R-62 i DOCKET NO. 50-124 MARCH 1985 I I I I I I AMENDMENT NO. 6

..l. '~ ~ TABLE OF CONTENTS 1.0 DEFINITIONS 2.0 LIMITING CONDITIONS FOR POSSESSION ONLY LICENS E 3.0 EMERGENCY POWER 4.0 RADIATION MONITORING _.5. 0 BUILDING EVACUATION SYSTEM I. 6.0 FUEL HANDLING 7.0 SURVEILLANCE REQUIREMENTS 8.0 DESIGN FEATURES l 9.0 ADMINISTRAflVE CONTROLS I I AMENDMENT NO. 6 I

1 Definitions . 1.1 Ghngtmal Qggutangg An abnormal occurence is any one of the f ollowing: A. An - observed inadequacy in the implementation of admirn strative or procedural

controls, such that the inadequacy causes or could cause an unsafe condition with regard to reactor radiological practices.

B. A malfunction of the instrumentation such that a release of radioactive material could occur or a release could occur without being detected. f C. A release of fission fragments from one or more of the reactor fuel elements of a magnitude to indicate a failure of the fuel element cladding. I D. An accidental or unanticipated release of radioactivity to the environment

in excess of the limits specified in 10-CFR, Part 20.

1.2 Gbt00gl The combination of sensor,

lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter.

l 1.3 GbEODEl GBliktBliGD I The adjustment of the channel such that it's output responds within acceptable range and accuracy to known values of the parameter which the reactor measures. 1.4 Gbannel Qantabillix Gbagh Oualitative verification of acceptable performance ,'by observation of channel behavior. 1.5 Gbencel Inst The int.roduction of a signal into the channel to verify that it is operable. 1.6 Ggafingmget A closure of the reactor room cell (room 10) which controls I the movement of air into and out of the cell through a controlled path. I 1 AMENDMENT NO. 6 l l

'47 Uggsuccd Yalua The value of a parameter as it appears on the output of the channel. 1.8 Oggtahlg A system or component which is capable of performing its inten~ded function in its normal manner. 1.9 QEEttiiDD A system or component which is perf orming its intended function in its normal manner. 1.10 Bggetgt Dingklad The reactor is considered disabled whenever A. The means of placing and maintaining moderator in the core area is removed and, B. The control ' rod drive mechanisms are physically disconnected f rom the control rods in the core and, C. All control rods are fully inserted and, D. The fuel is not in the reactor core. 1.11 G90ttel BQd A device fabricated from a neutron absorbing material which is used to control the rate at which fissionable I material in reactor fuel reacts. In the present case, the l control rods will remain fully inserted in order to assure l that the reactor fuel would remain subtritical, even if the fuel were loaded into the core. 1.12 Shellt Shgyld, gt day The word "shall" is used to denote a mandatory or required action or condition; the word "should" a recommended or desirable options and the word "may" a permissible or discretionary course, without bias. 2 Limition Geodittoos fet the Eennessima Qaly Lignosa 2.1 Beastot Euni At no time will the fuel be placed in the reactor core assembly. AMENDMENT NO. 6 LI

' 2.2 GeDttet!Beda The control rods will be fully inserted and uncoupled from the control rod drive mechanisms except at such times when tests are being performed on the control systems. 2.3 pymg yglyg The ~ dump valve shall be disabled in the full open condition except at such times when tests are being performed on the control systems. i 2.4 GQDfiDRERQt I A. Ngtmal GgDilDtmgDi Operations that require normal confinement ares I i 1. Irradiated fuel is being handled. 2. Other operations in which irradiated or contaminated equipment or materi,als are being handled. B. GgDditiGDE 19 6EbitXR NQCERl GQQfiDROrDi The existing conditions necessary to achieve normal confinement are: 1. All reactor room accesses are maintained closed and locked except as necessary for controlled personnel movements. 2. The reactor room ventilation is operating and I maintaining a negative pressure within room 10 with respect to the surrounding areas. The need for an operating ventilation system will be evaluated, after the fuel has* been removed according to provisions of Section 16.7.2.B. C. EmgtggDgy GgDilDemtDt Emergency confinement shall require as a minimum: 1. The reactor room shall be evacuated and all means of access closed and locked. 2. Robeson Hall in which the reactor is located shall be evacuated and reentry prohibited, except for emergency I response personnel, until such time as the emergency condition is cleared. I 3. The ventilation system for the reactor cell will be turned off. I AMENOMENT NO. 6

'l l 3 ESCte20EY EQWZC A backup emergency power supply shall be available to provide uninterrupted power to such instruments as will be maintained in an operating condition during the possession only status. 3.1 GgmQentat1 I The emergency power supply consists of a converter

unit, a

battery

bank, an

inverter, and connections to the instrumentation.

3.2 gancifiggligen The battery bank shall be capable of delivering a nominal 48 VDC to the inverter unit for at least 15 minutes under a full load and the inverter shall be cabable of delivering a single phase, nominal 220 VAC output. 4 Badiatino deoitation While the reactor fuel remains at the facility, the fixed radiation monitoring system shall remain functional and sufficient portable radiation survey instruments shall be maintained at the facility to permit monitoring of personnel and the facility to assure safety of reactor personnel and the public. After the fuel is

removed, sufficient I

components of the fixed system will remain functional for a period of six months to permit demonstration that no radioactive materials or effluents are escaping from the

facility, or that radiation conditions within the facility I

do not cause the requirements of10 CFR Part 20 to be exceeded. After six months if the data support the premise that no radioactive material has escaped the facility during the period nor is likely to do so in the future, a formal request will be made to turn off the fixed system. Continued

demonstration of the absence of escaping radioactive material will be made by quarterly swipe ahd area surveys at the points of egress from the reactor room.

4.1 Eingd Badiatien deoitetton Erstam The fixed system shall consist of the following: A. A G-M type detector with proportional compensation on the West wall of the reactor room. This detector has a range of 0.1 to 10,000 mrem per hour. The normal setpoint will be 15 mrem per hour or less, except during transfer of I irradiated fuel when it may be increased to avoid i ni ti ati on of a building alarm due to momentary (less than one minute) excursions at that time. An individual Will be at the I console observing the meter to insure that the levels do not exceed safe levels. When tripped, the monitor initiates the l 4 ANENDMENT NO. 6 I

building Gvccuotion alarm and daansrgizos the reactor r o cin ventilation. B. A G-M type detector with proportional compensation on the East wall of the reactor room. This detector has a I range of .1 to 10,000 mrem per hour. The normal setpoint will be 5 mrem or less. When tripped, the monitor initiates an over tolerance area alarm. C. A G-M type detector with proportional compensation is located within the reactor room vent exhaust stack as the stack passes through the library on the 3rd floor of Robeson Hall. The detector has a range of 0.1 to 10,000 mrem per hour. The normal alarm setpoint is 15 mrem or less. When tripped, the monitor initiates the building evacuation alarm and deenergizes the reactor room ventilation. 4.2 Enttehlt Badiation deoitets Portable radiation monitoring equipnent shall be maintained at the reactor f aci.11ty during the period of preparation f or " Possession Only Status" and thereafter at the Radiation I Safety

Office, a

division of Safety and Health Programs. This instrumentation shall consist of: A. A beta-gamma instrument capable of measuring dose rates of 0.1 to 20,000 mrem per hour. B. A neutron instrument capable of monitoring dose rates I of 0.1 to 10,000 mrem per hour. C. A portable air sampler for collecting airborne activity. I D. A G-M type "frisker" or equivalent for personnel and equipment surveys and for determining air-borne particulate activity levels. E. An Eberline Model PAC-4G low level alpha counter. I 5 E911diD9 EYAEMallen SYElRO A building evacuation system for Robeson Hall shall be maintained in an operable condition while the reactor fuel remains at the facility and for a period of at least si:: months thereafter. At the end of the interval a formal request will be made to deactivate the system based on submission of supporting data that no accidental release of radiation due to the presence of the defuelled facility is likely to occur. The building shall be evacuated whenever any one of the following three conditions exist: A. The building monitor on the West wall of the reactor room exceeds it's set point (normally 15 mrem per hour except as noted in section 16.4.1.A). j 5 AMEN!) TENT NO. 6 I

11 ~ l B. The ventilation stack monitor exceeds itt setpoint of s 15 mrem per hour. C. A condition exists

which, in the opinion of a

I responsible person, e.g. a member of the reactor staff, a radi a't i on safety person, or a member of the Reactor Safety Committee, would pose a danger to personnel within the reactor facility or members of the general

public, would result in a manual initiation of the evacuation alarm.

If maintenance is required on either of the two devices I which would initiate an automatic building evacuation

alarm, the facility would be secured and all operations other than those involved in the maintenance would cease.

During normal working

hours, surveys using portable instrumentation would be perf ormed at the beginninn end of the work day.

\\ 5.1 8E11901 When a building alarm occurs, the following actions will I take place: A. Facility operations will immediately cease; all personnel within the reactor laboratory area will immediately

evacuate, and assist in the evacuation of personnel from the remainder of Robeson Hall.

The doors to the facility shall be locked, and the ventilation system will automatically shut down. The VPI&SU Police will be i notified and will respond bringing an emergency kit and will ( prevent anyone f rom reentering the building. If the alarm occurs when no one is present within the reactor

facility, the Police will notify reactor staff (until the fuel is,

removed from the facility and the

Director, Safety and I

Health Programs thereafter) and radiation safety personnel from an emergency call list maintained by the police. Reentry of the f acility will be only by a joint decision of [ the senior reactor staff member and the radiation safety person present. Due care will be taken to insure the protection of all individuals. Only after it i*s proven that it is safe to do so will occupancy of any portion of Robeson Hall be per.v.itted. 6 Ewel Bandlion It is not intended that the fuel be moved from the fuel storage pits, where it is currently located, until such time that removal of the fuel from the f acility shall tal.e place.

However, if it is necessary to move one or more elements for inspections, auditing or safety

reasons, the I

fuel shall be handled under the following conditions: 6 AMENDMENT NO. 6 I

L c,. A. All =cvamcnt of fuel wil1' be made by a team of at least . E three persons: a reactor staff member, a health physicist, l and an assistant. When the fuel is being transferred ~to a commercial firm for removal from the site, the commercial firm shall have the responsibility for-handling the fuel. ( B. Radiation monitoring will be done using the portable and

fixed radiation monitoring equipment described in Section 16.4.

C. All fuel element movements will be logged. i D. A fuel transfer cask will be used for movement of all " hot." fuel. Acceptably " cold" fuel as determined by radiation surveys and specifically permitted by the Radiation Saf ety Of ficer may be moved without using the fuel transfer cask. E. A maximum of one " hot" element shall be allowed in transition at any one time. \\ 7 SytvtillEDER BR9Mitgmgats Under the possession only statuw, only a limited amount of checking of the instrumentation will be required. The f intervals will range f rom annual, not to exceed 15

months, semiannual, not to exceed 7 months, and quarterly, not to exceed 4

months. Six months after the fuel has been

removed, tests of specific systems will be discontinued

( (after approval has been obtained from the Radioisotope Committee). The following systemc will be inspected and/or functionally checked as noted at the prescribed intervals. s I l I 7.1 Bragtgt,Bggs The general conditions within the reactor room will 'b e I inspected quarterly with a walkthrough type inspection for obvious damage to maintain the integrity of the reactor systems. g 4 7.2 Bragtgt Bggs 2g0111311g0 5 Until such time as the fuel is disposed of, the ventilation ( system will be checked for the following items: A. The ventilation system shall be checked quarterly to ( assure that a negative pressure is being maintained in the reactor cell. B. The capability of the systeba to automatically turn off the ventilation system will be performed semi-annually. 7 AMENlMENT NO. 6 I

.S.' .e Aftcr tho fuol hon bacn renovsd frea tha facility, tha ducto 1eading to the ventilation stcck and the upper end of the stack shall be covered and the ventilation fans and motors protected from the environment. If found to be necessary to prevent radioactive material leaving the facility, according to the swipe and area surveys specified in Section 16.4, the ventilation system will be temporarily reactivated until the dispersion problem is resolved. 7.3 EmttgenEy Egwnt A functional transfer and load test shall be done quarterly or following any repairs or modification. A commercial power outage during-normal working

hours, with proper operation for 15 minutes or

longer, will count as a

successful test. 7.4 EWildiDE EYBEWAtiGQ EYEtts Building evacuation drills shall be performed semi-annually until such time as the fuel has been removed, at which time the f requency will be changed to annually. 7.5 BAdiatiGO M90119C109 EYEttmE The reactor

Area, Building and Stack monitors shall be I

calibrated semi-annually until such time that permission is secured to turn these systems off. 8 Dggigo igatytgs The f ollowing f eatures are described in order that changes shall not be made while under the " Possession Only Status". 8.1 Site E l E The Virginia Tech Argonaut Reactor is located in Robeson Hall of QPI&SU. Robeson Hall is located in the nor thwest corner of the main campus. Buildings located within a $00 l foot radius include Davidson,

Williams, Derring,

Cowgill, j

Burruss and Pamplin. Additions are planned to Robeson and j Pamplin within the next year. 8.2 BRActQt EAsility The area to be maintained secure under the " Possession Only Status" will be room 10, and portions of room 106. Other areas currently within the restricted area may be made available for other use by constructing walls or by providing alternate access after the fuel has been removed. Until that time, rooms 106, 6, 8 and 10 will be included within the area defined as the reactor facility. The building roof has been part of the restricted area. Since the reactor will not be operating during the " Possession 8 AMENDMENT NO. 6

..'.o l Only Ctatua" tha rcof crco will no lcngar bo pcrt cf tha I restricted area. . l s.3 BeacteC Coelant The water in the dump tank will be removed and the water I supply to the reactor coolant system shall be disconnected so t' hat i t wi l l be impossible to fill the reactor coolant system. I S.4 Ebigld Inch The shield tank shall be emptied and the tank will remain I empty. I B.5 Ewel Stetant Eitz While under " Possession Only Status", the only places where the fuel may be stored, pending it's removal from the

facility, are the East and West Fuel storage pits.

Only a single element or the equivalent number of plates may be stored in each storage hole in these pits. 9 Bdministtatire Genttelz 1 The overall responsibility for the reactor while in L " Possession Only Status" has been assigned to the Vice President for Administration and Operations. He has assigned the rsponsibility for administering the facility to l the Director of Safety and Health Programs. The Reactor Safety Committee has the responsibility for monitoring reactor activities to assure compliance with NRC I regulations, provisions of the terms of the reactor license, I and these Technical Specifications. Operations of the facility while fuel is still present is under a Reactor Supervisor. A half-time senior operator assists the Reactor Sup er vi s or,. A Radiation Safety specialist is assigned to the reactor When the fuel is removed from the f acility, the reactor supervisor and senior operator positions will be transferred to other areas within the University. The organizational chart in section 16.9.4 reflects the organization which will exist as long as the fuel is present. When the fuel is removed, the Reactor Safety Committee will be replaced by the Radioisotope Committee on this chart and the Reactor Supervisor and Reactor staff will be deleted. 9.1 SteEtet ggfgty Ggmmittgg During the period while the fuel remains at the University, the Reactor Safety Committee will remain a separate subcommittee of the University Radiation Safety Committee. The Reactor Safety Committee will be chaired by an administrator familiar with radiation and appointed by the I 9 AMENIMENT NO. 6

o. Vice President for Administration and Operations. There i g will be at least 5 other members of whom one will be the g Reactor Supervisor, another the Director of Safety and Health

Programs, and a third the Reactor Safety Officer.

The others will be chosen according to their background in radiation applications, engineering or other relevant disciplines. Once the fuel has been removed from the

facility, the Reactor Safety Committee will cease to exist as a separate committee and the oversight function will be assumed by the Radioisotope Committee, the other subcommittee of the Radiation Safety Committee.

The membership of this committee is defined under the terms of the University's Broad License f or Use of Radioisotopes. In each

case, a

quorum is defined as half the total membership plus one. Decisions of the committees require a I majority of.the members present and voting. 9.1.1 bggtings The Reactor Saf ety Committee or its successor shall meet quar terl y or as needed, or by request of any Committee member. 9.2 Benetin The reactor staff, when one exists, and the Reactor Safety Officer shall make a report to the Reactor Safety Committee or it successor, at each quarterly meeting, of reactor activities, radiation exposures, and transfers and I discharges of radioactive materials during the preceeding quarter. 9.3 BREQCdn Records Whall be maintained of all information pertineht to I the facility under " Possession Only Status" to allow othe'rs at a later time to determine the function and condition of all the reactor systems at such time as the facility is finally decommissioned or a decision is made to seek oper at i onal status. 5 f94 QCQEDLIBliGDBl Chttt (On next page) i 10 AMENDMENT NO. 6

Q808NII8I1QNBL Gd8BI I I I President i I I I _______1_______ IVice President i I for i IAdministration i I & Operations I l__________- __I I ______________________l_____________________ l I ________1_________ ________1_________ l Reactor Safety i I Director, Safety I Committee l_________________________I & Health Programst i I I I I I l l__I Reactor I i I i Supervisort 1 I l_____________I l__________________________________,__I _1____________ l i Reactor I I I Staff I l l_____________I I l I Reactor I l__t Radiation i I Safety i I Officer I l_____________I 11 ANENMENT NO. 6

' A I I I APPENDIX B VIRGINIA TECH ARGONAUT REACTOR (VTAR) ENVIRO WENTAL IMPACT REPORT FOR DECOM ISSIONING 14 JULY 86 I NRC LICENSE NO. R-62 DOCKET NO. 50-124 VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY DEPARTMENT OF HEALTH & SAFETY I BLACKSDURG, VIRGINIA 24061 1 I I I I I

m-- NUMBE_R SECTION PAGE B-3 1.0

SUMMARY

2.0 FACILITY DESCRIPTION B-3 3.0 WORKER EXPOSURE DURING DISMANTLING B-4 4.0 PUBLIC EXPOSURE DURING DISMANTLING B-5 5.0 PUBLIC EXPOSURE AFTER DISMANTLING B-6 I I l l l (B-2) 0797J

1.0 SUPMARY This report addresses the environmental considerations of the decomissioning of the Virginia Tech Argonaut Reactor (VTAR) Facility and the return of the facility to unrestricted use. Chem-Nuclear Systems, Inc.. who is experienced in the dismantling of research reactors will perform the decomissioning operations. As the licensee, Virginia Polytechnic Institute will monitor the overall project to ensure compliance with all applicable regulations. With the exception of an incident that occurred in 1971 (described in Secticn 1.2 5), there has been no uncontrolled release of radiation to the environment during the operating history of the reactor; no fracture of, or leakage from, the fuel elements; no 1ong-term contamination of the facility from radiochemical spills; and no uncontrolled emissions to the environment. The reactor fuel is no longer at the site; it was shipped to DOE in late 1985 and early 1986. The intent of this Environmental Report is to provide the supporting information needed by the Nuclear Regulatory Commission to determine that the dismantling and decomissioning operations will not have a significant effect on the quality of the human environment, and the decomissioning of the VTAR will have no environmental impact. 2.0 FACILITY DESCRIPTION The Facility is located in Robeson Hall on the northwest corner of the main campus of Virginia Polytechnic Institute and State University (Virginia Tech), between the Appalachian and Blue Ridge Mountains in southwest Virginia, approximately 35 miles west of Roanoke, Virginia. The reactor was used as part of the Nuc1 car Engineering curriculum for basic research in neutron physics, neutron radiography, neutron activation analysis, technical training and Reactor Operator training, as well as experiments associated with health physics and nuclear engineerin g. 0797J I

The VTAR is an Argonaut type research and training reactor originally designed and installed by American Standard Nuclear Division. The reactor has been in operation since June,1959. Originally licensed for a maximum power level of 10 kW(th), the reactor was modified and the license (R-62) amended to allow a maximum power of 100 kW(th) in late 1966. During licensed operating periods, all radioactive wastes were disposed of properly by shipment to a licensed burial site or discharged to the sanitary sewer within license limits after monitoring. During - dismantling, all radioactive waste will be disposed of at a licensed disposal site. 3.0 WORKER EXPOSURE DURING DISMANTLING The dismantling operations will be performed in such a manner as to maintain the radiation exposure of the dismantling team workers as low as reasonably achievable (ALARA). The following factors are in line with the ALARA principle: 1) The dismantling operation has been delayed for three years after shutdown in order to lower the level of the radioactivity from the short-lived comportent of the radioactive material. 2) Experienced and well trained personnel will be used in the dismantling operation. 3) A competent and well trained health physics team will monitor all dismantling operations. 4) Approved protective clothing will be worn by the workers in the dismantling operation. (0-4) 0797J

5) The dismantling plan will be reviewed and approved by the Reactor Safety Connittee. 6) The fuel elements have been transferred to the DOE. I The movement of radioactive dust will be minimized by the use of 7) enclosures, controlled atmosphere, and filtration systems. 8) The radioactivity in area outside the controlled area in the VTAR has been maintained at levels lower than the acceptable limits, and these same controls will be maintained during the dismantling operation. The collective radiation dose to all the workers for the entire dismantling operation is estimated to be less than 1.7 man-rem. 4.0 PUBLIC EXPOSURE DtfRING DISMANTLING The exposure of the general public to radioactive material released from the Facility during the dismantling operation will be near or at zero detectable. Effluent discharge shall be in accordance with, and within the limits specified in,10 CFR 20.303. All radioactive waste material will be packaged and shipped from the Facility by approved methods, and they will be transferred to a licensed disposal facility. The general public may be exposed to very low radiation IcVels for very short periods of time during the transportation of the radioactive material. There will be no long-term exposure of the general public in excess of the limits pennitted by 10 CFR 20. No hazardous wastes will be generated during dismantling. l (B-5) 0797J l i l l l

5.0 PUBLIC EXPOSURE AFTER DISMANTLING There will be no long-term exposure of the general public in the environs of the VTAR following the license termination. The reactor fuel has been packaged, transported and received by 00E. Similarly, the sources have been transferred to approved recipients for further use or storage. The radioactive material from the reactor structure and the concrete shield will be packaged and shipped to an approved disposal facility. These radioactive materials will be in controlled areas and, therefore, will not present an exposure hazard to the general public. In sunmary, there will be no significant exposure of the general public I during the dismantling operation, or following the license terminati on. There will be no environmental impact as a result of the VTAR deconnissioning. l 'I (B-6) 0797J

lA I I APPENDIX C PHYSICAL SECURITY PLAN FOR VIRGINIA TECH ARGONAUT REACTOR I (VTAR) DURING DISMANTLEMENT & DECO MISSIONING (D&D) I 14 JULY 86 NRC LICENSE NO. R-62 DOCKET NO. 50-124 VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY DEPARTMENT OF HEALTH & SAFETY BLACKSBURG, VIRGINIA 24061 I

I i

I

TABLE OF CONTENTS i NUMBER SECTION PAGE 1.0

SUMMARY

C-3 2.0 PHYSICAL SEC'JRITY PLAN C-3 2.1 LOCKED 000RS C-3 2.2 KEY ISSUANCE C-3 2.3 KEY LIST C-3 2.4 SECURITY SURVEILLANCE C-4 f l i l (C-2) 07961 j

I

. ~..

1.0 SUNiARY Virginia Tech has developed a physical security plan which addresses the security measures that will be implemented during the dismantling and decommissioning (D&D) project. All Special Nuclear Material (SM4) sources and all irradiated and new fuel have been transferred to other licensees or to the 00E; thus, there is no SNM associated with t'le licensed areas of the VTAR. Activated and contaminated materials generated from normal reactor operations are the only radioactive materials currently remaining within the facility. I The physical security plan for the D&D of the VTAR has been modified with consideration to only the control of byproduct material, I consisting of soli'd, fixed activated reactor components associated with the biological shield. 2.0 PHYSICAL SECURITY PLAN - I 2.1 LOCKED 000RS I All doors providing access to the VTAR will be kept locked at all times unless personnel are present in the facility. 2.2 KEY ISSUANCE Keys to the doors of the VTAR will be issued by the Head of the Department of Heal th & Safety, Dr. Kei th Furr. 2.3 KEY LIST Persons who will be issued keys shall be limited to only those persons who need access and will include: (C-3) 0796J

A) Chem-Nuclear Systems, Inc. D&D Contractor B) Virginia Tech Health & Safety Department Staff C) Master Key Held by Virginia Tech Key Shop The issuance of keys will be logged by the Head of the Department of ilealth & Safety. 2.4 SECURITY SURVEILLANCE The Virginia Tech Police Department will maintain periodic surveillance of the controlled areas of the VTAR 24 hours per day and will have a phone list for all keyed individuals as well as the Virginia Tech Emergency Plan phone list. I (C-4) 0796J _}}

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