ML110470594

University of Arizona - Decommissioning Plan Approval Safety Evaluation Report

ML110470594
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Site:	05000113
Issue date:	04/15/2011
From:	Tran A Research and Test Reactors Licensing Branch
To:	Williams J Univ of Arizona
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ml110470589	List: ML110470594
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Download: ML110470594 (22)

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SAFETY EVALUATION BY OFFICE OF FEDERAL AND STATE MATERIALS AND ENVIRONMENTAL MANAGEMENT PROGRAMS RELATED TO THE DECOMMISSIONING PLAN FOR THE UNIVERSITY OF ARIZONA RESEARCH REACTOR FACILITY LICENSE R-52 DOCKET NO. 50-113

1.0 INTRODUCTION

By letter dated May 21, 20091, The University of Arizona (the University or the licensee) submitted a request to the U. S. Nuclear Regulatory Commission (NRC) to approve its decommissioning plan (DP), dated May 21, 20092,and supplemented on March 26, 20103, for the University of Arizona Research Reactor (UARR) located in Tucson, Arizona.

Licensing activities related to decommissioning activities at the UARR are limited to the training reactor and isotopes production, General Atomics (TRIGA) reactor, NRC Docket No. 50-113, Facility License No. R-52.

As described in the plan, the option chosen for decommissioning is the decontamination (DECON) option and will consist of a transfer of licensed radioactive equipment and material from the site and decontamination of the facility to meet the unrestricted release criteria provided in Title 10, of the Code of Federal Regulations (10 CFR) Section 20.1402, Radiological Criteria for Unrestricted Use. In its DP the licensee described how the final status survey (FSS) plan will be developed and submitted to the NRC for review. The licensee will perform an FSS to verify and document that the decommissioned areas and structures meet the requirements for release for unrestricted use. The licensee will then submit documentation of the satisfactory completion of its FSS to the NRC for review and acceptance.

A notice, titled Notice and Solicitation of Comments; Pursuant to 10 CFR 20.1405 and 10 CFR 50.82(b)(5) Concerning Proposed Action To Decommission University of Arizona Research Reactor, was published in the Federal Register on December 14, 2009, (74 FR 66173).

1 ADAMS Accession No. ML100620926 2

ADAMS Accession No. ML091490076 3

ADAMS Accession No. ML100920089

As part of approving the DP, the facility license is amended to reflect license conditions deemed appropriate and necessary for the approval of the DP.

2.0 REGULATORY BASIS The regulatory requirements for the contents of DPs for research and test reactors are contained in 10 CFR 50.82(b)(4). This regulation requires that the proposed DP include the following items:

• the choice of the alternative for decommissioning with a description of related activities (see section 3.1 below)

• a description of the controls and limits on procedures and equipment to protect occupational workers and public health and safety from ionizing radiation (see section 3.7 below)

• a description of the planned FSS (see section 3.10 below)

• an updated cost estimate for the chosen alternative for decommissioning, a comparison of that estimate with present decommissioning funds set aside, and a plan for assuring the availability of adequate funds to complete decommissioning (see section 3.14 below)

• a description of quality assurance (QA) provisions, physical security plan provisions, and technical specifications (TS) in place during decommissioning (see sections 3.4, 3.12, and 3.11 below)

The NRC conducted its review of the DP submitted by University of Arizona in accordance with 10 CFR 50.82(b)(5) to determine whether the preferred decommissioning alternative would be performed in accordance with applicable regulations and would not be inimical to the common defense and security or to the health and safety of the public. Furthermore, should the NRC find that these criteria are met, and after notice to interested persons, it may approve the DP as an amendment to the referenced license, subject to such conditions and limitations as deemed appropriate and necessary.

License conditions for this amendment are based on Appendix 2 to NUREG-1700, Standard Review Plan for Evaluating Nuclear Power License Termination Plans, Revision 1, issued April 2003. Furthermore, the NRC staff established a license condition, in accordance with the requirement of 10 CFR 50.82(b)(5), that the approved DP supplement the safety analysis report or equivalent.

The requirements following the approval of the DPs are provided in 10 CFR 50.82(b)(6). This regulation states that the NRC will terminate the license if it determines that the decommissioning was in accordance with the approved DP and that the FSS and associated documentation demonstrate that the facility and site are suitable for release in accordance with the criteria for decommissioning in 10 CFR Part 20, Subpart E Standards for Protection Against Radiation.

3.0 EVALUATION 3.1 Decommissioning Alternative The licensees stated objective for decommissioning the UARR is the regulatory release of the facility for unrestricted use. The DECON option is the decommissioning alternative chosen by the licensee to permit termination of the reactor licenses and provide beneficial reuse of the property. Decontamination of facility equipment and structural components will be conducted to minimize radioactive waste. Structural portions of the building and materials found to be radiologically contaminated and/or activated will be decontaminated, sectioned and removed, and/or processed, as necessary. The licensee will perform a FSS to demonstrate that the UARR meets the NRC criteria for unrestricted use. The FSS results will be documented in a report to be submitted to the NRC in support of a request that the site be released for unrestricted use and the reactor license terminated.

3.1.1 Conclusion The NRC staff concluded that the choice of DECON and the associated proposed plans meets the provisions of 10 CFR 50.82(b)(4)(I) for decommissioning without significant delay and is, therefore, acceptable.

3.2 Facility Radiological Status 3.2.1. Facility Description and Operating History The UARR is a TRIGA pool-type reactor designed and constructed by the General Atomics Division of General Dynamics Corporation (now GA Technologies of San Diego, California).

TRIGA stands for Test, Research, Isotope production, General Atomics. The UARR is located within the University of Arizona Nuclear Reactor Laboratory (NRL) on the 325 acre campus of the University of Arizona in Pima County Arizona in the city of Tucson. The University is about 65 miles north of the Mexican border at Nogales, Arizona, 110 miles southeast of Phoenix, Arizona, and 120 miles from the western border of New Mexico.

The UARR is located on the first floor of the north wing of the Engineering Building. The Engineering Building is made of brick and reinforced-concrete construction, including most floors and ceilings. Four adjacent rooms in the Engineering Building are permanently established as the NRL and are designated a controlled access area. These are the Control Room (Room 122); the Reactor Room (Room 124); the Equipment Storage and Experiment Setup Room (Room 124A) and the room directly above the reactor room, which was originally designed to receive a beam of neutrons from the reactor (Room 216).

The reactor is located near the bottom of a circular pit approximately 14 feet below ground level.

The pit contains a steel tank resting on a 1-foot-thick concrete slab. Eight inches of poured concrete surrounds the outside of the tank, except for a window 4 foot wide by 1 foot 10 inches high, which was left in the concrete to allow for the insertion of a thermal column at a later date, and a 3-inch-diameter circular opening, which was intended to accept a van de Graff generator beam tube. The inside of the steel tank is covered on the sides by a layer of Gunite approximately 2 inches thick and on the bottom by a layer 4 inch thick. The entire inner surface of the Gunite is coated with Amercoat (epoxy-base paint).

The reactor was constructed at the University of Arizona in 1958 and went into operation in December of that year. The licensed power was 10 kW thermal with operation at 30 kW possible for short periods of time. Subsequently facility was licensed to allow operations at 100kW. Based on a revised Safety Analysis, a license amendment was approved in June 1972 allowing for receipt and possession of additional fuel for a complete change over from aluminum-clad to stainless-steel-clad fuel. In December of that year, 87 partially used stainless-steel-clad TRIGA fuel elements were obtained, permitting operation in the pulsed mode. When operations ended the reactor was licensed for a steady state power of 110 kW (thermal), with a pulsing capability up to peak powers of approximately 650 MW.

There is no history of major accidents or spills. In 1974, a crack was discovered in the cladding of a fuel element, however, there were no releases of long-lived fission products or nuclear fuel from the fuel element.

3.2.2 Current Radiological Status of the Facility The licensee submitted a Radiological Characterization Report of the University of Arizona Nuclear Reactor Laboratory dated May 20, 20094. The purpose of the characterization was to assess the current radiological status of the facility for the development of the DP. Due to ongoing reactor operations, the scope of the characterization was limited to facility areas outside the reactor pit, its associated components and operations systems.

The survey used for the characterization was designed to determine whether or not elevated levels of radioactive material contamination were present in the radiologically impacted rooms of the NRL. In each radiologically impacted room, professional judgment was used to select locations that may have increased potential to be radiologically contaminated. These locations were then surveyed for total alpha, total beta, removable alpha, and removable beta. Additional sample locations were selected to investigate the presence of tritium. General area gamma radiation dose rates were also collected.

The radiological characterization surveys showed that there were no radiologically contaminated areas outside of the reactor pool that are greater than the potential release criteria.

The reactor components and portions of the tank are neutron activated. An activation analysis was performed on the core internals and reactor tank to determine the activated product radionuclide concentrations of reactor core components. Results from this analysis indicate that all radioactive waste associated with the NRL is considered Class A LLRW (low-level radioactive waste) and that no Class B&C LLRW is expected in the NRL. This activation analysis report was submitted by the licensee5.

Water samples are collected quarterly from the UARR pool. Gamma spectroscopy data indicates no activity significantly above background. The water does contain tritium at approximately 3x10-6 Ci/mL.

4 ADAMS Accession Number ML091490105 5

ADAMS Accession Number ML091490104

On October 13, 2010, the licensee submitted a report of characterization surveys of areas that were not accessible during operation6. The characterization report included information required in NUREG-1575, Sections 2.4 and 5.3; and NUREG-1757, Volume 2, Section 4.2 to allow NRC staff to verify that the University has adequately characterized the radiological condition of the site.

3.2.3 Release Criteria The decommissioning alternative proposed in this DP includes the removal of all activated and contaminated materials, equipment and components. The remaining equipment and surfaces will be released to the NRC required 25 millirem annual (mrem/yr) Total Effective Dose Equivalent (TEDE) following guidance contained in NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)7 (NRC 2000). The release criterion will be determined to have been met by demonstrating surface or volumetric activities, as appropriate, meet NRC screening values8, or to an alternative site-specific release criteria as discussed below.

Alternative site-specific release criterion may be developed using a dose modeling software code such as the RESRAD family of codes. Argonne National Laboratory (ANL) developed the RESRAD family of computer codes under the sponsorship of the U.S. Department of Energy (DOE). The codes have been used widely by DOE and its contractors, the NRC, U.S.

Environmental Protection Agency (EPA), U.S. Army Corps of Engineers, industrial firms, universities, and foreign government agencies and institutions. The codes are pathway analysis models designed to evaluate potential radiological doses to an average member of the specific critical group based on a defined occupancy or site reuse scenario. The licensee committed to obtain NRC approval if alternative site-specific release criteria are developed.

3.2.4 Conclusion The NRC staff has reviewed the characterization performed by the licensee for the principal radioactive components. The NRC staff, based on its experience and engineering judgment, concludes that the characterization was performed in accordance with NRC guidance in NUREG-1757 and that the licensees estimates of the radiological conditions and radiation measurements are acceptable.

The NRC staff concludes that the release criteria proposed by the licensee based on the referenced generic screening thresholds are sufficient to demonstrate compliance with 10 CFR 20.1402 and are, therefore, acceptable. A license condition will be added for review and approval by the NRC if alternative release criteria are developed for release of the NRL.

6 ADAMS Accession Number ML102930034 7

ADAMS Accession Number ML003761476 8

Acceptable license termination screening values of common radionuclides for building surface contamination were published in the Federal Register on November 18, 1998 (64 FR 64132). Screening values of common radionuclides for surface soil contamination were published in the Federal Register on December 7, 1999 (64 FR 68395).

3.3 Decommissioning Tasks 3.3.1 Scope of the Decommissioning Project The licensee proposes to remove all radioactive materials from the UARR facility, dismantle the reactor and its peripheral support systems, release the licensed area for unrestricted use and seek termination of Facility License R-52.

Many of the reactor components and systems that are either activated or contaminated and will need to be segregated from non-radiological components and surfaces so that they may be disposed of as LLRW. Building materials such as the reactor tank and dry storage pits will need to be evaluated for radiological activity, removed, and disposed of according to their radiological status, as necessary.

The following are the decommissioning tasks, which are necessary for site release:

• Further characterization

• Remove loose equipment

• Remove hazardous materials (lead, cadmium) and asbestos

• Remove the control rod drives, rotary rack drive, and bridge

• Remove fuel storage rack, holsters, and cooling coils

• Remove the reactor structure, reflector, and irradiation components

• Remove and disposition pool water

• Segregate and package materials according to radioactivity levels and classification

• Remove auxiliary systems (rabbit system, water purification, ventilation)

• Remove Gunite, activated portions of the tank liner, concrete, and affected soils

• Decontaminate or remove the dry storage pits

• Decontaminate building surface

• Ship radioactive waste for disposal

• Perform the Final Status Survey (FSS)

• Submit required reports that demonstrates to the NRC that the facility meets the release requirements

• Request license R-52 termination

• Restore the facility for future use by the University.

The FSS will be developed by the licensees Decommissioning Contractor (DC) using the criteria provided in NUREG-1575. Since it is anticipated that no subsurface foundations or soils were impacted by the operation of UARR, the FSS will only cover the exposed concrete and soil surfaces remaining within the UARR facility after the reactor and activated components have been demolished and removed. This is contingent on the results of the characterization of subsurface soils under the reactor. If additional radionuclides are identified in the characterization of the subsurface soils beneath the reactor, they will be added to the list of radionuclides of concern in the DP and may impact the FSS.

3.3.2 Schedule The UARR shut down operations in May 2010. The on-site decommissioning tasks are expected to start after fuel removal in 2011 and are anticipated to last about 4-6 months.

3.3.3 Conclusion The NRC staff concludes that the manner in which the licensee proposed to complete each of the decommissioning tasks is acceptable. A license condition will be added to require the University to provide NRC with a FSS prior to conduct of license termination surveys. The University committed to develop the plan in accordance requirements in NUREG 1575 (MARSSIM); NUREG-1537, Part 1, Appendix 17.1, Section A; and NUREG- 1757, Volume 2, Chapter 4.

3.4 Decommissioning Organization and Responsibilities The University of Arizona is committed to, and retains ultimate responsibility for, full compliance with the existing NRC reactor license and the applicable regulatory requirements during decommissioning. The Decontamination and Decommissioning (D&D) Project and the UARR is under the supervision of the Nuclear Reactor Lab Director. The University will appoint a University Project Manager (UPM) to oversee the decommissioning process.

The University of Arizona plans to select a contractor to perform all or parts of the UARR Decommissioning Project. The team will consist of UARR personnel and the selected contractor. The selected DC will manage the physical aspects of their portions of the decommissioning work including QA, health physics, safety, waste processing, and waste packaging and shipping. However, the University will continue to maintain overall responsibility for health and safety, compliance with regulations, and applicable license conditions.

The following key licensee organizations / positions are described in the DP:

Nuclear Reactor Lab Director Nuclear Reactor Lab Director is responsible for assuring that all D&D project activities are conducted in a safe manner and within the requirements of the UARR NRC License, the DP, the UARR Radiation Protection Program, and the provisions of the Reactor Committee.

The minimum qualifications and experience for this position are as follows:

• Advanced degree (MS or PhD) in Nuclear Engineering or related discipline and five (5) years experience in nuclear reactor operations and/or decommissioning.

• Familiarity with the UARR NRC License, the DP, the UARR Radiation Protection Program and with applicable federal and state regulations.

• Trained at the level required by the UA Radiation Protection Program for a permit holder to be in the possession of radioactive materials of the types known to be present at the licensed reactor site.

Reactor Committee The Reactor Committee reviews reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license. The structure, member qualifications, and responsibilities of the committee are specified in the UARR TSs.

With respect to the DP, the responsibilities of the committee include:

• Approval of all plans and procedures required for decommissioning

• Review and approval of all proposed changes to the facility, procedures and TSs and the DP

• Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in the Technical Specifications as required by 10 CFR 50.59, and review and approval of required safety analysis University Project Manager (UPM)

The UPM is responsible to University administration and management for the successful completion of decommissioning work. The role of UPM includes many activities from prequalifying and selecting contractors to reviewing and recommending changes to the contract, as well as ensuring communication between members of the University staff during the decommissioning process.

Specifically, the role of UPM includes the following duties:

• Selecting a DC in accordance with the University of Arizona procurement guidelines

• Overseeing the DCs performance relative to the terms of their contract

• Overseeing the DCs performance relative to the Decommissioning Plan

• Overseeing the DCs performance relative to all subsequent plans and procedures The minimum qualifications for the UPM are as follows:

• Bachelor's degree in Architecture, Civil, Electrical, Mechanical or Structural Engineering or related field AND five years of construction experience which included one year of construction supervisory experience; OR, nine years of progressively responsible construction experience which included one year of construction supervisory experience; OR, any equivalent combination of experience, training and/or education approved by the University Human Resources department.

Reactor Supervisor (RS)

The RS is responsible for the following duties:

• Overseeing all activities of the DC personnel on the licensed site

• Ensuring that all decommissioning activities are performed in compliance with applicable regulations and license conditions

• Review of all plans and procedures required for decommissioning

• Reviewing and submitting to the Reactor Committee all needed changes and subsequent plans and procedures that do not change the original intent or result in an unreviewed safety question

• Communicating with all appropriate regulatory agencies

• Communicating with the Nuclear Regulatory Commission, the University Administration, and the DC and sub-contractors The minimum qualifications for the RS are as follows:

• Bachelor's degree in Engineering or science field AND completion of an approved training program in nuclear reactor maintenance AND three years of nuclear reactor operation and maintenance experience; OR, completion of an approved training program in nuclear reactor maintenance AND seven years of nuclear reactor operation and maintenance experience; OR, any equivalent combination of experience, training and/or education approved by Human Resources

• Familiarity with all physical plant and equipment at the UA reactor at a level equivalent to that required for an SRO license at an operating reactor of the same type

• Knowledge of the UARR license conditions, Safety Analysis Report and DP and procedures as approved by the Reactor Committee

• Knowledge of security requirements at the site and ability to meet the requirements for unescorted access to it

• Trained at the level required by the UA Radiation Protection Program for worker with radioactive materials of the types known to be present at the reactor licensed site

• Knowledge of Federal regulations applicable to the possession, transport and use of radioactive materials University Radiation Control Office (RCO)

The University RCO is responsible for monitoring and overseeing radiological safety at the UARR and NRL. The RCO has the responsibility and authority to stop any plan or activity that has the potential to result in an unacceptable radiological hazard. This function ensures that the activities involving potential radiological exposure are conducted in compliance with the applicable licenses, Federal and State regulations, and NRL standard operating procedures.

The minimum qualifications for RCO staff are as follows:

• Radiation Control Office directorate. Two, or more, individuals with advanced degrees (MS or PhD) certified in Health Physics and Diagnostic Radiological Physics, each with 10 years experience in radiation safety and health physics.

• Radiation Control Office staff. Associate's degree or equivalent certification in Radiation or Physical Sciences AND two years of radiation monitoring experience; OR, four years of radiation monitoring experience; OR, any equivalent combination of experience, training and/or education approved by the University Human Resources department.

3.4.1 Conclusion The NRC staff concludes that the licensee has provided reasonable assurance that organizational structures needed to safely decommission the UARR are in place. In addition, the licensee has committed to ensuring the Reactor Committee will properly oversee all

decommissioning activities conducted by the DC, including the review and approval of changes to the facility and of decommissioning-related plans and procedures. The NRC staff finds that the project management structure for the decommissioning of the UARR is consistent with the guidance on the role and composition of the facility safety committee provided in Appendix 17.1 to NUREG-1537 and is, therefore, acceptable.

3.5 Training Program The licensee will have a general training program designed and implemented by the DC and approved by the RCO to provide orientation to project personnel and meet the requirements of 10 CFR Part 19, Notices, Instructions, and Reports to Workers: Inspection and Investigations. General site training will be required for all personnel assigned on a regular basis to the D&D project. General site training will include but is not limited to: project orientation, security, and access control; introduction to radiation protection; quality assurance; industrial safety; emergency procedures; and packaging and transport of radioactive materials.

additionally, for specific areas of work supplemental training may be required in: radiation worker training; hazardous waste operations and emergency response training; respirator training and fit testing; hazardous material training; hearing conservation training; permit-required confined space entry training; lockout/tagout hazardous energy control training; and trenching and excavation training.

The DC will be responsible for the radiation worker training of its employees and verifying that subcontractors are also adequately trained in radiation safety commensurate with their work activities in accordance with the requirements of 10 CFR Part 19. The DC Site Radiation safety Officer (RSO) will be responsible for on-site radiation safety training of workers and verifying pervious training and qualification. The DC's radiation safety training program will be administered by a Site RSO who will approve all training materials and qualification of workers.

The University RCO may provide additional training or verification of support staff training prior to providing dose monitoring badges such as thermoluminescent dosimeters (TLD). A written exam will be required to demonstrate proficiency with the radiation worker training topics.

Radiation worker training will also include a practical factors demonstration and evaluation.

3.5.1 Conclusion Based on the review of the licensees training program as described in the DP, the NRC staff concludes that the licensees training program is acceptable. The licensee also recognized that specific training would be required to reflect the unique hazards associated with decommissioning operations. While the NRC does not regulate non-radiological hazards as specified in the Atomic Energy Act, the licensee is aware that personnel involved with decommissioning project activities are subject to training requirements administered by other Federal, State, and local government agencies and has committed to provide training commensurate with the potential hazards to which individuals may be exposed.

3.6 Decommissioning Contractor (DC)

The NRC staff reviewed the criteria specified by the licensee for selecting a contractor to manage and supervise all or part of the UARR Decommissioning Project. The selected DC will manage the physical aspects of their portions of the decommissioning work including QA, health physics, safety, waste processing, and waste packaging and shipping. However, the University

will continue to maintain overall responsibility for health and safety, compliance with regulations, and applicable license conditions.

In selecting the DC, the University will prepare a request for proposal, which will define the qualifications and experience necessary for prospective DCs and subcontractors. Prior history and performance of the prospective contractor on non-power reactor or similar decommissioning projects will be used to help the University select a qualified DC to perform the facility D&D.

The contractor qualifications and experience required include the following:

• Task experience: the University will require the selected DC to have at least 5 years prior experience in radiological site decommissioning. Specific experience in decommissioning test reactors, power reactors, and/or materials licensed sites will be required. DC submittals of project descriptions, references, and other supporting information will be required prior to contract award. Specific DC project management documentation will be required in the areas of work plan development, training, QA, work management, reactor dismantlement and decontamination, waste packaging, waste shipping, work documentation, radiation protection, final status surveys, regulatory interface, and supporting the preparation of the final decommissioning project report.

The minimum expectation of the University for a DC is verification of company experience in these tasks plus proof of financial viability.

• QA program: the University will require the selected DC to have an existing QA program that was used to support decommissioning projects and final status surveys. The minimum expectation of the University for a DC is verification of company QA experience in these tasks.

• Personnel experience: the University will require the selected DC to support the project with an experienced work force with at least 5 years prior experience in radiological site decommissioning. Specific individual experience will be required in the areas of work plan development, training, QA, work management, reactor dismantlement and decontamination, waste packaging, waste shipping, work documentation, radiation protection, final status surveys, regulatory interface, and supporting the preparation of the final decommissioning project report. The minimum expectation of the University for a DC contractor is verification of personnel experience in these tasks plus a commitment to provide experienced personnel for the duration of the project.

3.6.1. Conclusion The NRC staff has reviewed the criteria the licensee will use to select the DC. The selection criteria cover all skill areas necessary for successful Decommissioning Project management and performance. Therefore, the NRC staff concludes there is reasonable assurance that the licensee will select a prime contractor with adequate qualifications.

3.7 Radiation Protection 3.7.1 ALARA Program The licensee committed to conducting decommissioning activities in a manner that will ensure that radiation exposures will be maintained as low as reasonably achievable (ALARA). The University RCO, the DC Site RSO and DC health physics staff will be responsible for:

implementing ALARA principles; providing radiation worker training; establishing administrative-level occupational and public dose limits; monitoring personnel for occupational exposures; controlling exposures; providing and maintaining radiation monitoring equipment; performing radiation surveys and monitoring; and, maintaining records and generating reports as necessary to comply with regulatory and licensing requirements.

The DC will prepare a Radiation Protection and ALARA Plan that will incorporate provisions for minimizing occupational and public radiation exposures. The ALARA Plan will describe specific administrative and engineering controls that will be put in place during specific D&D project activities. Examples of administrative and engineering controls include limiting access to certain areas, mock-up training, use of remote-handling devices, temporary shielding, containment structures, portable HEPA (high efficiency particulate air) filtered ventilation, and specialized protective equipment and respiratory protection.

3.7.2 Health Physics Program The project Health Physics Program will be implemented under the authority of the University RCO with the assistance of the DC Site RSO. The Health Physics Program will be designed to satisfy the following commitments that should be established by the Radiation Protection Program:

• Implement the procedures defined in the Radiation Protection and ALARA Plan.

• Ensure radiological safety of the public, occupationally-exposed personnel, and the environment.

• Monitor radiation levels and radioactive materials.

• Control the distribution and release of radioactive materials.

• Maintain potential exposures to the public and occupational radiation exposure to individual within administrative limits and the regulatory limits of 10 CFR Part 20 and ALARA.

• Monitor personnel internal and external exposure in accordance with 10 CFR Part 20 requirements. The University RCO will provide and manage TLDs; however, full compliance and costs (other than TLDs), will be the responsibility of the DC.

3.7.3 Dose Estimates The total projected occupational exposure to complete the decommissioning of the UARR is estimated to be 2.39 person-rem. This estimate was prepared using the individual work activity durations and work crew sizes, based upon the results of the characterization results to date and recent experience in performing similar activities at the University of Washington. The licensee provided this estimate for planning purposes. Detailed exposure estimates and exposure controls will be developed during detailed planning of the decommissioning activities.

The dose estimate to members of the public as a result of decommissioning activities is estimated to be negligible. This is because the area immediately surrounding the facility is under the control of the Reactor Supervisor and because the area where decommissioning activities are taking place are fully contained within the facility (with the exception of loading and unloading of shipments of equipment and radioactive materials). This is consistent with the estimate given for the reference research reactor in NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, issued 1988. The dose to the public during DECON and truck transport of radioactive waste from the reference research reactor referred to in the Final Generic Impact Statement is estimated to be negligible (less than 0.1 man-rem).

Activated pieces and any contaminated debris will be removed and shielded, if required, to meet U.S. Department of Transportation (DOT) shipping requirements and the disposal site waste acceptance criteria.

3.7.4 Conclusion The NRC staff reviewed the licensees ALARA program, HP program, and the estimate of potential radiological doses attributable to decommissioning operations. The licensee has committed to implementing its existing radiation protection and contamination control programs in support of decommissioning the UARR. The NRC has reasonable assurance that these mature programs will achieve their intended purposes because they have been successfully implemented by the licensee in support of reactor operations. Furthermore, based on review of the ALARA, personnel monitoring, radiation protection/instrumentation, and respiratory protection programs, the NRC concludes that these programs are adequate to safely support decommissioning operations at the UARR and are therefore acceptable. Based on its review of the UARR DP and the reference research reactor in NUREG/CR-1756, the NRC staff concludes that the licensees estimates for occupational and public doses during decommissioning activities are reasonable. The NRC staff also finds that the estimates of occupational dose may be revised as additional characterization data are developed. As such, the licensee has provided reasonable assurance that decommissioning operations will be conducted safely and within the allowable radiation dose limits specified in 10 CFR Part 20.

3.8 Radioactive Waste Management 3.8.1 Fuel Removal As stated in the DP, the fuel will be removed from the UARR reactor after the shut-down date of May 22, 2010 and transferred to DOE for reuse.

3.8.2 Radioactive Waste Processing The licensees decommissioning of the UARR will result in the generation of solid and LLRW, mixed waste, and hazardous waste. This waste will be handled (processed and packaged),

stored, and disposed of in accordance with Federal, State, and local requirements.

The licensees waste management program includes provisions for waste minimization or volume reduction, radioactive and hazardous waste segregation, waste characterization, neutralization, stabilization, solidification, and packaging.

3.8.3 Radioactive Waste Disposal The licensee stated that prior to disposal, all waste streams will be properly characterized according to the requirements of the disposal facility. This characterization will include qualification of primary radionuclides of concern as well as hard-to-detect radionuclides.

Additionally, those radionuclides that have specific limits for Class A waste will be directly quantified or estimated based on ratios to concentrations of other radionuclides. The licensee stated that surfaces and materials destined for reuse, recycling, or disposal as clean waste will be shown to be free of detectable surface contamination in accordance with the guidelines provided by the NRC in IE Circular 81-079. The volume of LLRW is estimated at 753 cubic feet.

The DP notes that all waste will be shipped to an acceptable waste disposal site in accordance with applicable NRC and DOT regulations regarding waste packaging, labeling, and placarding.

The licensee stated that each LLRW shipment will be accompanied by a shipping manifest as required by Section I of Appendix F to 10 CFR Part 20, "Requirements for Low-Level Waste Transfer for Disposal at Land Facilities and Manifests." The waste will be manifested consistent with its classification. Only licensed transporters will be used to transport wastes from the UARR.

Mixed wastes may be shipped to a licensed processing facility or directly to a licensed land disposal facility depending on the nature of the waste and the treatment options available.

3.8.4 Industrial Safety Program DC industrial safety and hygiene personnel, such as Certified Safety Professionals or Certified Industrial Hygienists, along with project management personnel, will be responsible for ensuring that the D&D project complies with all applicable federal safety requirements and general safe work practices. The DC will prepare a site specific Health and Safety Plan (HASP) to document safety requirements and accident response procedures.

At a minimum, the HASP will include: hazards assessment; general site safety procedures; a requirement for a daily site safety meeting; site inspection procedures; emergency response procedures; emergency contact telephone numbers; material safety data sheets for hazardous materials present on-site; training requirements for specific activities such as permit-required confined space entry or hot work; and local emergency medical information.

The HASP will be reviewed and approved by the Universitys Risk Management Department.

The HASP will direct site activities necessary for ensuring that the UARR D&D project meets occupational safety and health requirements for protection of project personnel. The functional purpose of the HASP will be to ensure compliance with the Occupational Safety and Health Act (OSHA) of 1973.

9 ADAMS Accession Number ML082490470

All DC personnel working on the D&D project will receive health and safety training in order to recognize and understand potential hazards and risks. Training requirements for DC subcontractors will be determined by the DC Site Health and Safety Officer (SHSO) based on the specific task the subcontractor is performing.

3.8.5 Conclusion The NRC staff reviewed the licensees program that will be implemented to support waste management operations. Based on the review of the licensees program and the licensees experience in safely managing radioactive waste generated during normal operations, the NRC staff concludes that the licensees proposed radioactive waste management program that will be implemented to support decommissioning operations is adequate and, therefore, acceptable.

The NRC staffs review of the Industrial Safety Program indicates that it will ensure appropriate personnel protection consistent with OSHA requirements and is acceptable.

3.9 Radiological Accident Analyses The licensee evaluated potential radiological accidents during decommissioning of the UARR by determining UARR components and areas that contain the highest radioactive material inventory. The proposed decommissioning activities and methods in which radioactive material could be released to the work area or environment were considered. Since all special nuclear material will have been removed prior to decommissioning, the majority of the accidents discussed in the current license are not applicable. The accident identification process was supplemented by reviewing experiences at other non-power reactor decommissioning projects.

The following radiological accidents were considered to present the highest potential consequences:

• Fire

• Spill tritium loaded water into the environment

• Release airborne contamination to the environment

• Transportation accident The consequences of a fire during decommissioning of the UARR were considered and are not significantly different than the consequences of a fire during reactor operations. The likelihood is low that a fire would start or that a fire could become intense enough to release radioactive material. Any fire in dry radioactive waste would be limited to a few microcuries of radioactivity.

The spilling of tritium-loaded water could occur during pool water pumping or removal operations the total tritium inventory in the pool is less than 100 Ci. All other radionuclides in the pool water are at negligible concentrations. If the entire pool inventory were released to the environment and conservatively assumed to evaporate in one day, releasing 100 Ci, the maximum exposure to an individual would be 6.4 mrem.

An uncontrolled release of airborne radioactivity could occur during cutting and demolition activities involving contaminated or activated materials, such as removal and segmentation of reactor components, or removal of tank steel and concrete. The licensee will utilize safety management operations (standard engineering and administrative controls) for protecting against such accidents. While the actual concentrations of airborne radioactive materials are unknown at this time, the dose consequence of an uncontrolled release is expected to be

low (< 1 mrem off-site impact and < 25 mrem to on-site workers). Air sampling will be performed to identify the actual airborne radioactive material concentrations and confirm that the concentrations are below the 10 CFR 20, Appendix B values.

Various forms and quantities of radioactive waste will be shipped from the UARR during the D&D project. The dose consequence from transportation accidents could be higher than the contamination accident scenarios described above because high-activity reactor components could be involved. As such, there is a potential for a moderate dose consequence of between 1 and 25 mrem for the public following a transportation accident. However, licensee adherence to NRC and DOT radioactive material packaging and transportation requirements is considered a sufficient control measure for mitigating transportation-related incidents.

3.9.1 Conclusion Based upon information contained in the DP, the NRC staff finds that the postulated accidents as reported by the licensee adequately reflect the types of potential accidents that could reasonably be expected to occur during decommissioning operations. In addition, since the reactor fuel will have been removed from the reactor and shipped off site, the radiological consequences that could result during decommissioning are far less than those types of accidents that formed the licensing basis for the reactor during operations. Furthermore, the licensee is required by regulation to perform reviews in accordance with 10 CFR 50.59 before removal of significant safety controls (e.g., fire suppression systems) during decommissioning operations. The NRC staff finds that it is reasonable to conclude that offsite radiological consequences, as reported in the DP, would be well below the NRC dose limits for individual members of the public (100 rem/yr). As such, the NRC staff finds that, given the likelihood and low radiological consequences to workers and members of the public attributable to potential accidents that could reasonably occur, there is reasonable assurance that decommissioning operations can be performed while maintaining the health and safety of the public and protection of the environment.

3.10 Proposed Final Status Survey Plan The licensee provided a plan for the development, review, and approval of the FSS plan in Section 4.0 of the DP. The licensees stated objective for the FSS is to ensure that the facility meets the unrestricted release criteria of 25 millirem annual (mrem/yr) Total Effective Dose Equivalent (TEDE).

3.10.1 General Survey Approach The outline for the proposed FSS plan prepared by the licensee is intended to provide information to the NRC for determining the adequacy of the licensees understanding of the proposed FSS approach, commitments, and objectives needed to ultimately demonstrate compliance with the radiological criteria for license termination.

The licensee proposes that the FSS plan will be developed following the guidance provided in MARSSIM to demonstrate compliance with the release criteria. The MARSSIM process emphasizes the use of data quality objectives (DQO), proper classification of survey areas (survey units), a statistically-based survey and sampling plan, and an adequate quality assurance/quality control (QA/QC) program.

The FSS will be performed in accordance with an FSS Plan by trained DC technicians experienced in performing FSS. The technicians will follow written procedures regarding surveys and sampling, sample collection and handling, chain-of-custody, and recordkeeping.

The FSS Plan will define sampling locations, required analysis, and survey types.

3.10.2 Identification and Classification of Survey Units The licensee proposes to classify survey units based on contamination potential according to the methods described in MARSSIM. In general, there are two overall classifications, non-impacted and impacted. Non-Impacted areas have no reasonable potential for residual contamination because there was no known impact from facility operations.

Impacted areas may contain residual radioactivity from facility operations. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2 or Class 3 designations. Class 1 areas have the greatest potential for residual activity while Class 3 areas have the least potential for impacted areas.

The licensee has classified the areas of the NRL based on the operating history of the NRL, and from the radiological characterization performed in February 2009 as described below.

Class 1 Areas - The reactor pit is the only Class 1 area in the NRL. Areas in the bottom of the reactor pit will have contamination greater than the current proposed Derived Concentration Guideline Levels (DCGL).

Class 2 Areas - The Class 2 areas of the NRL are limited to the floors and walls up to 2 meters of the Reactor Room and the Equipment Storage Room, Room 124 and Room 124A, respectively. Radioactive material and sources were handled in these areas, however, if residual contamination is found, it should be a small fraction of the DCGL.

Class 3 Areas - The Class 3 areas of the NRL are the walls above 2-meters in Rooms 124 and 124A; all surfaces in Room 122, the Reactor Control Room; and all surfaces in Room 216, the Second Floor Storage Room. The walls above 2-meters in Rooms 124 and 124A have very little potential to have any level of contamination, but there are no barriers to prevent possible contamination from being spread to these areas. There is also very little potential for any contamination in Rooms 122 and 216, but these rooms are a part of the NRL licensed area and are to be included in the FSS.

3.10.3 Instrumentation The FSS will include surface gamma surveys using appropriate instrumentation. Surface and subsurface soil samples will be collected using either a random-start grid pattern or randomly generated locations as appropriate commensurate to the classification of the survey area. Soil samples will be analyzed for contaminants of concern using standard analytical methods including liquid scintillation counting for hard-to-detect beta-emitting radionuclides (i.e., Carbon-14 and tritium) and gamma spectroscopy for gamma-emitting radionuclides.

3.10.4 Data Collection Buildings, equipment, and components that are destined to remain after license termination require the following surveys to demonstrate they meet the appropriate release criteria.

Buildings, equipment, and components require two-stage scan measurements as part of the FSS process at appropriate coverage rates and speeds. Gross beta and/or gross alpha measurements are utilized as appropriate to the potential contamination. The measurements typically are performed at a distance of 1 cm or less from the surface. Adjustments to scan speed and distance may be made in accordance with approved procedures. Direct measurements are required for buildings, equipment, and components as part of the FSS process. The required quantity of direct measurements is a calculated value. The calculation is described in MARSSIM. Direct measurement data for buildings, equipment and components is collected with an appropriate detector. As much as practical, the detector is of an appropriate size to maintain the surface to detector distance of no greater than the calibrated distance +0.5 cm.

Soil areas will require gamma scan measurements as part of the FSS process at appropriate coverage rates and speeds. Volumetric samples are required to demonstrate a soil area meets the appropriate release criteria. In lieu of volumetric samples, soil areas may receive direct measurements using in-situ gamma spectroscopy, as equipment and trained personnel are available. Volumetric sampling differs slightly depending on the situation for which the sample is desired. The required quantity of volumetric samples for an open land survey unit is a calculated value that is discussed in MARSSIM.

3.10.5 Data Evaluation Data evaluation is performed on FSS results for individual survey units to determine the whether the survey unit meets release criterion. Appropriate tests will be used for the statistical evaluation of survey data. Tests such as the Sign test and Wilcoxon Rank Sum (WRS) test will be implemented using unity rules, surrogate methodologies, or combinations of unity rules and surrogate methodologies, as described in the MARSSIM and NUREG-1505 Chapters 11 and 12.

If the contaminant is not in the background or constitutes a small fraction of the DCGL, the Sign test will be used. If background is a significant fraction of the DCGL the Wilcoxon Rank Sum (WRS) test will be used. It is anticipated that the sign test will be the only statistical test applied to the collected data because of the small fraction of the DCGL that background radionuclides will contribute.

3.10.6 Data Quality Objectives The licensees stated data quality objectives are to permit demonstration at the 95 percent confidence level that the criteria are met. Decision errors selected by the licensee are limited to 5 percent for both Type I () and Type II () errors.

3.10.7 Conclusion The NRC staff has reviewed and finds the licensees DP concerning the planning of the FSS acceptable. However, the NRC staff notes that the information provided does not constitute a Final Status Survey Plan and the submittal of a complete FSS plan will be required prior to

conduct of the FSSs. Therefore, a license condition has been included with requires the submittal and NRC approval of the FSS plan prior to performance of the surveys.

3.11 Technical Specifications By letter dated May 20, 2010, as supplemented on August 13, 2010, September 20, 2010, October 8, 2010, January 7, 2011, and January 25, 2011, the University submitted a request to revise the license and technical specifications (TS) to support decommissioning activities. The requested amendment consisted of changes to the Facility License from possess, use, and operate to a possession only license, and changes to the facility TS to eliminate those specifications which are no longer needed and to revise other specifications to reflect the reactors permanently shutdown status. The staff approved the proposed amendment on February 15, 201110.

3.11.1 Conclusion This item has been satisfactorily addressed by the amendment.

3.12 Physical Security Plan The licensee stated that they will maintain a physical security plan in accordance with the regulations in Section 73.67(c)(1) of 10 CFR Part 73 which require facilities to maintain a physical security plan when they possess special nuclear materials of moderate strategic significance or 10 kg or more of special nuclear material of low strategic significance. The licensee intends for the nuclear fuel to be shipped off-site prior to the start of decommissioning activities. Once the license has been amended for no possession of nuclear fuel, a 10 CFR Part 73 physical security plan is no longer required.

The license acknowledges that the regulations in Sub Part I, Storage and Control of Licensed Material of 10 CFR Part 20 are applicable to the remaining byproduct and special nuclear materials that would be possessed by the UARR. All UARR licensed materials that are in storage will be secured from unauthorized access or removal; and licensed materials that are not in storage will be maintained under the control and constant surveillance of authorized UARR personnel as required by 10 CFR Part 20.

3.12.1 Conclusion The licensees commitments for facility security are consistent with regulations and adequate for protection of the material.

3.13 Emergency Plan The University of Arizona has a reactor facility emergency plan for responding to emergencies at the reactor facility. The purpose of this plan is to minimize any emergencys effect on the public, personnel, the reactor facility, and the environment surrounding the facility. The removal of spent fuel from the site will considerably reduced the potential for significant release of radioactive material off site. Any airborne or liquid releases resulting from decommissioning 10 ADAMS Accession Number ML102980413

activities would have a negligible impact off site. The most likely accident scenario is a contaminated and/or injured individual. The plan covers events involving the potential or actual release of radioactivity and provides measures for facility evacuation, reentry and recovery. The plan also covers medical support for afflicted personnel. The UARR D&D project will adopt the Emergency Plan as written.

3.13.1 Conclusion The NRC staff finds that the current reactor facility emergency plan is acceptable for responding to emergencies that may arise while decommissioning the UARR.

3.14 Estimated Cost The licensee stated that dismantlement and decommissioning of the NRL will be accomplished without dismantlement of the building. The detailed estimated cost to decommission the NRL licensed areas is presented in Table 1-1 of the DP. The estimate includes a breakdown by major project activities. The licensee estimated that the project will cost $1,990,686 with a 25 percent contingency included for cost escalation. The licensee stated that in accordance with 10 CFR 50.75 (e)(1)(iv), the University of Arizona is a state institution and as such will provide financial assurance with a statement of intent containing a cost estimate for decommissioning, indicating that funds will be obtained when necessary.

3.14.1 Conclusion The NRC staff has reviewed the licensees decommissioning cost estimate. The NRC staff finds that the cost estimate provided in Table 1-1 of the DP is consistent with the scope of work covering dismantlement and decommissioning of the WCNS. The NRC staff concludes that the licensee is committed to providing adequate funding for decommissioning the Ward Center for Nuclear Studies (WCNS.)

4.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes in inspection and surveillance requirements. The NRC staff has determined that this amendment involves no significant hazards consideration, no significant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

5.0 CONCLUSION

The licensee has safely conducted licensed operations involving the use of radioactive materials since the UARR became operational in 1958. As such, the operational, security, QA, waste management, and radiological programs required under the referenced license have been effectively carried out. Because these programs will continue to remain in effect in accordance with regulatory and license requirements, the NRC staff focused its review on the manner in which these required programs would be maintained and subsequently transitioned to support the mission of safely decommissioning the UARR.

The NRC staff has reviewed the licensees proposed actions to decontaminate, dismantle, and dispose of component parts of the UARR, and to perform a FSS. After decommissioning activities are completed, the NRC will review the licensees FSS report to determine if the facility has been adequately remediated to levels commensurate with unrestricted use in accordance with 10 CFR 20.1402. If the NRC concludes that the facility has been successfully decommissioned to permissible levels, then the University of Arizonas Facility License No. R-52 will be terminated.

Based on the NRC staffs review of the licensees application for approval of decommissioning, the NRC staff finds that the licensee is adequately cognizant of its continuing responsibilities to protect the health and safety of both workers and the public from undue radiological risk. The licensee provided reasonable assurance that the dismantlement of the reactor and disposal of all significant reactor-related radioactive materials would be conducted safely and in accordance with applicable regulations and NRC guidance.

The NRC staff concludes that the choice of the DECON decommissioning option is acceptable and meets the requirements of 10 CFR 50.82(b)(4)(i) for decommissioning without significant delay. The NRC staff concludes that the licensee provided acceptable organizational structure and control to decommission the NRL while maintaining due regard for protecting the public, the environment, and workers from significant radiological risk. Furthermore, the NRC staff concludes that the licensees plan for radiation protection and radioactive material and waste management is acceptable based on the use of standard guidance and practices for such programs. The NRC staff finds the personnel training program that UARR proposes to be acceptable because its scope covers all aspects of decommissioning activities that need to be performed safely. The industrial safety program and procedural and equipment controls are consistent with such programs at decommissioning reactors, and they are therefore acceptable.

The NRC staff concludes that the accident analyses show potential radiological consequences to be well within acceptable limits.

The NRC staff concludes that the licensees DP contains a description of the controls and limits on procedures and equipment to protect occupational and public health and safety as required by 10 CFR 50.82(b)(4)(ii).

The NRC staff concludes that the licensee has adequately described the radiological status of the NRL reactor facility and has proposed acceptable release criteria for the facility. The licensee has acceptably described the tasks, the sequence of activities, and the schedule needed to decommission the NRL. The NRC staff also concludes that the licensee has provided an acceptable description of its planned final radiation survey as required by 10 CFR 50.82(b)(4)(iii).

The NRC staff concludes that the licensee has provided, in accordance with 10 CFR 50.82(b)(4)(iv), an acceptable updated cost estimate for the DECON decommissioning option and has an acceptable plan for assuring the availability of adequate funds for the completion of decommissioning.

Therefore, based on the discussion above, the NRC staff concludes that the licensees DP meets the requirements of 10 CFR 50.82(b)(4).

The NRC staff has added licensee conditions to the facility license that are deemed appropriate and necessary for the approval of the DP.

Principal Contributors: John B. Hickman, FSME Stephen Giebel, FSME Date: