ML14318A906

Enclosure 2: Safety Evaluation Report

ML14318A906
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Site:	05000131
Issue date:	11/12/2014
From:	Beverly Smith Reactor Decommissioning Branch
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ML14318A624	List: ML14318A806 ML14318A872 ML14318A906
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Enclosure 2 SAFETY EVALUATION REPORT BY OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS RELATED TO THE DECOMMISSIONING PLAN AND AMENDMENT NO. 12 FOR THE OMAHA VETERANS ADMINISTRATION MEDICAL CENTER ALAN J. BLOTCKY REACTOR FACILITY FACILITY LICENSE R-57 DOCKET NO. 50-131 1.0 Introduction By letter dated May 21, 2014 (Agencywide Documents Access and Management System (ADAMS) Package No. ML14150A404), as supplemented by letter dated November 12, 2014 (ADAMS Accession No. ML14335A597), the Department of Veterans Affairs (VA or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or the Commission) to approve a license amendment and a decommissioning plan (DP) completed by its design and oversight contractor, AECOM Technical Services, Inc. (AECOM), for the Alan J.

Blotcky (AJB) Reactor located in the Omaha Veterans Administration Medical Center (VAMC).

Licensing activities related to decommissioning activities at the VMAC are limited to the Alan J.

Blotcky Reactor Facility (AJBRF), NRC Docket No. 50-131, Facility License No. R-57.

The DP proposed decontamination of the facility, dismantlement of the reactor, and termination of the facility license with no restrictions on future site use. The May 2014 DP submittal was a major revision of the DP submitted by the VA in a letter dated September 21, 2004 (ADAMS Accession No. ML042740512) and incorporated changes in response to the NRCs request for additional information dated May 13, 2008 (ADAMS Accession No. ML081080402) and commitments made in letters dated November 9, 2010 (ADAMS Accession No. ML103210235),

August 15, 2011 (ADAMS Accession No. ML11250A056), and March 8, 2012 (ADAMS Accession No. ML12075A202). The licensees March 8, 2012, letter requested that the NRC approve the licensees DP and terminate License R-57. The NRC noticed the DP in the Federal Register on June 20, 2012, and provided opportunity for the public to provide comments, request a hearing, and intervene (77 FR 37074). The May 2014 DP revised the release criteria to correctly reference Inspection and Enforcement Circular 81-07. Since this is a minor change from the DP noticed in 2012, a new notice was not required.

The VA also submitted a site characterization by letter dated April 6, 2006 (ADAMS Accession No. ML061140054), as supplemented by letters dated May 2, 2006 (ADAMS Accession No. ML061380584), and August 15, 2011 (ADAMS Accession No. ML11255A334).

Its former contractor, Duratek, Inc., performed the initial DP and site characterization for the VA.

This characterization included rooms, ventilation systems, drainage systems, cooling systems, storage areas, the reactor structures, and outside areas, as well as a historical site assessment.

NRC requested that the VA perform additional characterization to support conclusions and objectives presented in the initial DP. Additionally, some of the radiological conditions and disposal options changed in the time since the original DP was submitted, which required the VA to reevaluate low-level radioactive waste (LLRW) disposal options. The VA performed additional site characterization in response to the NRCs request, which provided a better

2 assessment of the isotopes of concern and examined other areas or systems they were not fully characterized in the original submissions. The revised DP provided more details of the proposed Final Status Survey (FSS) and followed guidance in NUREG 1757, Consolidated Decommissioning Guidance.

The staff has reviewed and approves the proposed license amendment and DP. This Safety Evaluation Report (SER) summarizes the staffs safety review of the licensees proposed license amendment and DP.

2.0 Regulatory Criteria 10 CFR Section 50.82(b)(4) contains the regulatory requirements for the contents of DPs for research and test reactors. This regulation requires that the proposed DP include the following items:

• The choice of the alternative for decommissioning with a description of activities involved (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 funds set aside for decommissioning, and a plan for assuring the availability of adequate funds to complete decommissioning (see section 3.14 below); and

• 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 the VA 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. 10 CFR 50.82(b)(5) states that, if the NRC finds that these criteria are met, after notice to interested persons it will approve, by amendment, the DP, subject to such conditions and limitations as the Commission deems appropriate and necessary. The DP will be included as a supplement to the safety analysis report or equivalent.

Section 50.82(b)(6) provides the requirement that follows the approval of a DP. This regulation states that the NRC will terminate the license if it determines that the decommissioning has been performed 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, Standards for Protection Against Radiation, Subpart E, Radiological Criteria for License Termination.

3.0 Evaluation 3.1 Decommissioning Alternatives

3 The objective of the AJBRF decommissioning activities is to remove licensed radioactive materials from the facility and adjacent areas in order to obtain NRC approval for release of the property for unrestricted use and to allow termination of the NRC license. The AJBRF intends the decommissioning pathway described in the DP to meet the necessary requirements to achieve this objective.

The DP Section 5.1 considered three alternatives available to the licensee: (1) the No-Action alternative (SAFSTOR); (2) the entombment option (ENTOMB); and (3) complete decontamination and release of the structures (DECON) for reuse as storage, laboratory space, or staff offices in the Omaha VAMC hospital.

The DP Section 5.2 states that the licensee considered SAFSTOR and ENTOMB to satisfy requirements for public protection while curtailing initial costs associated with time, finances, radiation exposure, and waste disposal capacity. Conversely, the licensee states that both SAFSTOR and ENTOMB alternatives require lengthy decommissioning schedules that would result in the AJBRF areas being unavailable for use extended periods. The extended period would require Omaha VAMC to expend resources to maintain operations for maintenance, security, and surveillance during decommissioning, with the increased probability of losing personnel with knowledge and experience in reactor operations. Therefore, the licensee rejected SAFSTOR and ENTOMB as decommissioning alternatives.

The DP Section 5.3 states that DECON is the licensees preferred option for decommissioning because it is the most environmentally suitable alternative and provides the maximum flexibility in future re-use of the site. The DECON alternative requires the site to be released and restored for unrestricted use. The licensee describes the decontamination, dismantling, and demolition of the reactor facility in Section 7. The licensee will demolish, remove, and dispose of the reactor and associated equipment as radioactive waste. The licensee is not sure if the building in which the reactor area is housed will be re-constructed or used at all. The licensee states that the area where the AJBRF is located will continue to function as a VA hospital post decommissioning and will use these areas for storage, laboratory space, or staff offices.

Therefore, the licensee will conform to the As Low as Reasonably Achievable (ALARA) principles and will minimize radioactive exposures to the Omaha VAMC staff, patients, and contractors performing the DECON process. The licensee will follow the DECON process with a comprehensive final contamination survey demonstrating that the AJBRF site meets the NRC criteria for release to unrestricted use.

3.1.1 Conclusion The NRC staff reviewed the licensees chosen method for decommissioning, DECON, and concludes that this choice satisfies the requirement of 10 CFR 50.82(b)(4)(i). This choice is acceptable because it provides for completion of decommissioning without significant delay (i.e., 12 weeks from DP approval to completion of the FSS).

3.2 Facility Radiological Status 3.2.1 Facility Description and Operating History

4 The Atomic Energy Commission issued Facility Operating License No. R-57 to the VA to operate the Alan J. Blotcky (AJB) reactor on June 26, 1959.1 The AJB reactor was a low-power Mark I Training, Research, Isotopes, General Atomics (TRIGA) nuclear reactor, which was authorized to operate at steady-state power levels up to 10 kW(t) and first reached criticality on June 30, 1959.2 Amendment No. 2 increased the steady-state thermal power level of the reactor to 18 kW(t) in September 1963 and Amendment No. 9 issued in April 1991 increased the reactor licensed power level from 18 kW(t) to 20 kW(t). The VA primarily used the AJB reactor in research and radioisotope production related to the diagnosis and treatment of disease, such as a source for neutron activation analysis of biological samples and for hot atom chemistry research. The VA also used the reactor for training Fort Calhoun Station nuclear power reactor operators from 1989 to 2001.3 The AJB reactor operated until November 5, 2001.

The AJBRF is located in the basement of the eleven-story wing of the main building of the Omaha VAMC. DP Figure 2.3 illustrates the AJBRF floor plan. The AJBRF is constructed of brick exterior with a concrete floor and concrete block exterior walls. Utilities such as municipal water and nonradioactive sewage, natural gas, and electricity are provided for joint use in the entire building. The AJBRF has a separate ventilation system exhausting through filters to the outside environment and a dedicated chiller system that removed heat from the reactor water.

Research and preparation laboratories are part of the AJBRF, and the chemical hoods in these laboratory rooms exhaust separately on the hospital building roof.4 The licensee states in DP Section 2.1 that the AJBRF fuel was removed and shipped to the US Geological Survey (USGS) TRIGA reactor in Denver, Colorado in June 2002. The applicant states in DP Section 2.2 that the remaining TRIGA reactor assembly is located in a 6.1 meters (m) (20 feet (ft)) deep steel reactor pool structure below-ground with only vertical access to the core. Gunite lines the reactor pool and concrete surrounds the outside by 25.4 cm (10 inches (in)) and on the bottom by 27.9 cm (11 in), which is between the steel and a corrugated steel casing.

DP Figure 3.2 illustrates the TRIGA reactor assembly. The TRIGA reactor assembly contains a graphite reflector enclosed in aluminum casing resting on a platform that raises the lower edge of the reactor assembly about 0.6 m (2 ft) above the reactor pool floor. Neutron activation during reactor operations caused the reactor assembly and parts of the reactor pool and surrounding concrete to be radioactive. The demineralized water within the reactor pool provides enough shielding so that radiation levels in and around the reactor room are background levels. The AJBRF entrance is through the door to the main reactor room, B526 (DP Figure 2.3).

The licensee prepared samples before irradiation typically in rooms B540 and B537 and stored isotopes in the isotope storage area in room B540A. The licensee conducted gamma counting analysis of irradiated samples in a large shielded cabinet in room B533, which is marked "hot cell" on DP Figure 2.3. Note that room numbers were changed in the period between when the 1 NRC, 2002. Safety Evaluation Report Related to the Renewal of the Operating License for the Alan J.

Blotcky Reactor Facility at the Department of Veterans Affairs, Nebraska - Western Iowa Health Care System, Omaha Division, August 2002, ADAMS Accession No. ML020630192, p 1.6.

2 Ibid.

3 VA, 2004. Decommissioning and Decontamination Plan for the Alan J. Blotcky Reactor Facility, Omaha, NE, March 2004, ADAMS Accession No. ML042740512, p 12.

4 NRC, 2002. Safety Evaluation Report Related to the Renewal of the Operating License for the Alan J.

Blotcky Reactor Facility at the Department of Veterans Affairs, Nebraska - Western Iowa Health Care System, Omaha Division, August 2002, ADAMS Accession No. ML020630192, p. 1.6.

5 first DP was submitted and site characterization was conducted and when this updated DP was submitted in May 2013.

Within the AJBRF are the following additional facilities (old room numbers in inner parentheses):

Locker Room (Room B522 (SW 1))

Restroom and Shower (Room B522A (SW 1A))

Radioisotope Reactor Research Laboratory (Room B526 (SW 2))

Nuclear Research Lab and Office (Room B533A (SW 2A))

Office and Darkroom (Room B533AA (SW 2B))

Nuclear Research Lab and Office (Room B537 (SW 2C))

Nuclear Research Lab (Room B535)

Walk-In Cooler (B535A (SW 2D)

Nuclear Research Lab and Office (Room B540 (SW 2E))

Isotope and General Storage (Room B540A (SW 2F)

The AJBRF used various radioactive sources for instrument calibration and reactor start-up and monitoring. All sources were in solid form. DP Table 2.1 describes the sources and the transfer/deposition dates, which Table 1 presents below:

Table 1. AJBRF Radioactive Sources Source Manufacturer Initial Activity Serial Number Date Received Transfer/

Disposal Date Po21O/Be Isotope Specialties 5 curies (Ci)

None 05/30/1959 06/01/1963 Po21O/Be U.S. Nuclear 7 Ci B673 07/13/1961 05/23/1966 Po21O/Be U.S. Nuclear 7 Ci J-102 05/03/1963 05/23/1966 Po21O/Be U.S. Nuclear 7 Ci L-202 07/02/1964 05/23/1966 Am241/Be U.S. Nuclear 2 Ci 618AM372 01/29/1968 07/24/2003 Am241 Siemens 12 mCi 2824LA 02/27/1987 07/24/2003 Am241 Siemens 12 mCi 5839LV 06/24/1987 07/24/2003 Po21O/Be U.S. Nuclear 7 Ci 0-178 01/12/1966 In addition to the reactor fuel, the licensee shipped all fission chambers to the USGS in Denver, Colorado. The licensee states in DP Section 1.4 that all licensed radioactive sources have been removed and properly disposed of except one Polonium-Beryllium (Po-Be) source, which is stored in a 18.9 liter (L)(5 gallon (gal)) pail of paraffin within a B-25 shipping container in the AJBRF room B535A. The DP Section 2.2.2 states that the source has decayed to negligible activity, but the stainless steel capsule containing the source is slightly activated.

3.2.2 Current Radiological Status of the AJBRF The AJBRF Section 4 describes the current radiological status of the facility and states that the VA performed an initial site characterization of the AJBRF in December 2002 to assess the radiological status of the facility and to support development of the initial DP (Duratek, 2003).5 5Duratek, Inc. 2003. Alan J. Blotcky Reactor Facility Omaha Veterans Administration Medical Center, Omaha, Nebraska, Characterization Survey Report, February (ADAMS Accession No. ML061140054).

6 The licensee states that the initial characterization study found no contamination outside the AJBRF, but identified several areas of contamination within the AJBRF laboratory rooms. The initial characterization found the highest concentrations of contamination in the laboratory hood in room B540 and the laboratory hood drain in room B535A. The licensee states that a later site characterization conducted in May 2011 confirmed these findings and provided additional information on the isotopes of concern.6 The DP states that the second site characterization also included sampling of the wastewater drain line piping and laboratory hood exhaust ventilation ducts, but did not identify detectable contamination above background concentrations in these additional areas. The licensee states that the second characterization identified a list of radionuclides of concern that are typical for TRIGA research reactor operations and confirmed within the AJBRF. These isotopes are in DP Table 4.1 and listed in Table 2 below. The licensee states that it only lists longer-lived isotopes with half-lives greater than 2 years because the reactor operations stopped in 2001 and all fuel removed in 2002. The only isotope listed, but not confirmed is Eu154. The DP states that one can characteristically find Eu154 associated with Eu152 in lower concentrations. The DP Section 4 states that there are no isotopes associated with reactor fuel or leakages. The licensee supports the absence of fuel-associated isotopes from the list of radionuclides of concern on information obtained from the site historical assessment (e.g., operating history) that there were no known releases and the site characterizations confirm the absence of the fuel-associated isotopes.

Table 2. AJBRF Radionuclides of Concern Isotope Half-Life Location Confirmed Present Screening Level (dpm/100 cm2)*

H3 12.3 yr Surface, Soil, Concrete, Dummy Fuel Elements Yes 1.2 x 108 C14 5730 yr Surface, Soil, Concrete, Dummy Fuel Elements Yes 3.7 x 106 Fe55 2.7 yr Surface, Soil, Concrete Yes 4.5 x 106 Co60 5.3 yr Surface, Soil, Concrete, Metal, Dummy Fuel Elements Yes 7.1 x 103 Ni63 100 yr Surface, Soil, Concrete Yes 1.8 x 106 Cs137 30.2 yr Surface, Soil, Concrete, Metal Yes 2.8 x 104 Eu152 13.6 yr Soil, Concrete, Dummy Fuel Elements Yes None Eu154 8.8 yr Soil, Concrete, Dummy Fuel Elements No None

• disintegrations per minute (dpm) per 100 centimeters squared (cm2)

The DP Section 4.1 states that the two site characterizations performed did not identify any significant areas of surface contamination (total or removable) above the NRC surface 6 AECOM, 2011. Alan J. Blotcky Reactor Facility Additional Characterization Report, AECOM Technical Services, Inc., July.

7 contamination screening levels (see Table 2 above). The licensee states in DP Section 1.5 that the contractor secured all loose radioactive materials during pre-decommissioning characterization activities performed in May 2011. The licensee states in DP Section 2.2.2 that it placed materials that could not be free-released, i.e., radioactive materials, in one B-25 shipping container and one 208-L (55-gal) drum, which the licensee stores in the AJBRF room B535A. The licensee stores the single remaining Po-Be source in the B-25 container with the other radioactive materials. The licensee states it attached an inventory of the items contained within the B-25 box to the outer surface of the box.

3.2.2.1 Systems and Structures The licensee states in DP Section 4.1 that removable H-3, C-14, total beta, and alpha activities were largely less than minimum detectable concentration (MDC) for removable contamination measurements (swipes) on floor and wall (structure) and reactor system surfaces. All detectable measurements of H-3 were below 10 percent of the H-3 surface screening criteria.

The licensee states that the site characterization identified C-14 as the isotope of concern in the drain from the lab vent hood in room B535. The licensee states that liquid scintillation analysis of the total beta activity was largely equal to the sum of the H-3 and C-14 activity in the samples, thus the licensee concludes that these samples did not indicate the presence of other hard-to-detect (HTD) isotopes such as Ni-63 or Fe-55.

The licensee states that the site characterizations measured total surface beta-gamma (/)

contamination greater than NRC screening criteria for Co-60 in the lab vent hoods in rooms B535 and B540, the reactor pool cover, and the floor and wall penetrations in room B540A.

Co-60 is the isotope with the lowest screening level of the isotopes of concern in SER Table 2.

The licensee states that most of the surfaces that the first site characterization detected removable contamination were cleaned with Maslin cloths during the second characterization in 2011. The licensee states that no detectable activity was found during Maslin surveys.

3.2.2.2 Reactor Pool and Components The licensee states in DP Section 4.2 that the first site characterizations surveyed the reactor assembly with an underwater dose rate meter and estimates that most of the dose is from Co-60 present in the Lazy Susan due to activation of the rotary rack bearings, which contain a high concentration of cobalt metal. The licensee used MicroShield (Version 8.03) to model the dose, estimate the activity of the Lazy Susan, and apply it to the reactor assembly as a whole.

The licensee states that the model estimates Co-60 activity 0.030 and 0.045 Curies (Ci). The licensee states that it assumes that Co-60 is the only significant gamma-emitting isotope in the reactor assembly. The licensee based the 0.030 Ci estimate modeling an external dose point on an annular cylinder that was equivalent to the maximum dose measured, 1.5 Roentgen per hour (R/hr)), on contact with top of the reactor assembly above the inner edge of the Lazy Susan. The licensee based the 0.045 Ci estimate on an internal dose point in an annular cylinder filled with water with 0.6 R/hr measured in the center of the top grid plate. The licensee estimates Co-60 activity is now between 0.009 Ci and 0.014 Ci after nine years of decay.

The licensee expects the minimal removable contamination within the tank, on the reactor assembly and other tank internal components because the monthly reactor pool water samples collected since 2006 indicate non-detectable activity. The VA has continuously circulated the reactor tank water. The licensee states the Decommissioning Operations Contractor (DOC) will filter the tank water prior to the proposed discharge to the sanitary sewer system because it has measured H-3, C-14, Fe-55, Co-60, Ni-63, Cs-137, and Eu-152 in the reactor water

8 demineralizer filter resins. The DOC will manage the demineralizer tank, resins, piping, and pumps that contacted reactor tank water. The licensee expects the chiller system components are free of internal radioactive contamination because the chiller system, part of the heat exchanger, was not in contact with the circulated reactor tank water. The licensee shut down the chiller system in 2003 and disposed of the refrigerant for re-use elsewhere in accordance with air pollution control regulations. The licensee assumes that the steel reactor tank, inner gunite layer, tank floor, and surrounding concrete are radioactive from neutron activation, but has not determined the extent of the activity.

3.2.2.3 Surface and Subsurface Soil The licensee states in DP Sections 4.3 and 4.4 that neither site characterization detected isotopes of concern or isotopes of potential concern above background concentrations in the surface and subsurface soil samples collected and analyzed. The licensee states that although the VA did not identify radioactive contamination in any samples, the VA did not perform analysis for HTD beta-emitting isotopes, such as H-3 and Ni-63. Therefore, the DOC will collect additional soil samples near the reactor tank as part of the FSS.

3.2.2.4 Surface Water and Groundwater The licensee states in DP Section 4.5 that there is no surface water on the Omaha VAMC campus or in the immediate vicinity. DP Section 4.6 states that the licensee installed shallow groundwater wells around the AJBRF in 2001, but the VA has not analyzed any samples for radionuclides. The licensee states that no soil samples collected during well installation indicated subsurface contamination, and thus the licensee assumes groundwater is not contaminated.

3.2.3 Release Criteria This section provides the specific radiological criteria that will be applicable for unrestricted release of the site and termination of NRC license R-57.

3.2.3.1 Unrestricted Release of Structures, In-place Systems, Soil, and Concrete The licensee states in DP Section 6.1 that the VA will use the radionuclide screening criteria in NUREG 1757, Volume 1, Rev 2, Appendix B, Tables B.1 and B.2, which has been calculated to correspond to 25 mrem/yr dose limit. The VA will use the criteria in these tables as site-specific release criteria for facility surfaces, in-place systems, and volumes of concrete and soil. The licensee commits to reduce contamination activity to ALARA and using the sum-of-fractions rule when multiple contaminants are involved. The licensee states that it does not need to do site-specific dose modeling because the VA will use the NUREG-1757 screening criteria.

DP Section 6.1.1 states that the VA will apply the Co-60 surface activity limit to all detectable /

surface activity (7,100 dpm/100 cm2). The VA will use an instrument such as a Ludlum Model 2929 or equivalent and apply 710 dpm/100 cm2 as the removable / criterion of for detectable removable surface contamination because the NRC developed total surface activity screening criterion assuming a 10 percent removable fraction. The licensee states it will use swipes analyzed in a liquid scintillation counter (LSC) to evaluate total surface contamination of HTD isotopes. The licensee states that the last site characterization identified HTD isotopes are present in the form of low-level surface contamination or in contaminated/activated materials.

9 The VA will assume a removable fraction of 10 percent when applying removable contamination measurements to total surface contamination screening levels.

The licensee will use the sum-of-fractions rule to evaluate all measured activity. The VA will use the total net beta activity measured against the Ni-63 screening criterion in DP Table 6.1 because it is the most conservative screening criterion for the HTD isotopes listed. The licensee states that some of the total net beta activity may be due to detectable isotopes; the licensee will use other analytical instruments (i.e., non-LSC) to determine the activity of these isotopes. The licensee will use the sum-of-fractions rule to evaluate the impact of all measured activity, such as the following equation:

7100

1.27 1.855 1

The licensee states that site characterization did not detect alpha contamination; therefore, the licensee does not plan to survey for alpha contamination. However, should either total or removable alpha contamination levels be assessed, the licensee will not release items with detectable alpha contamination based on detection limits listed in Table 6.3 of the DP, which are values from Inspection and Enforcement (IE) Circular 1-87. The alpha detection limits are a surface contamination criteria of 1,000 dpm/100 cm2 (average), 3,000 dpm/100cm2 (maximum),

and 100 dpm/100 cm2 (removable).

The licensee states in DP Section 6.1.2 that it will use the sum-of-fractions rule to evaluate soil and concrete measurement data against the screening criteria in DP Table 6.2. The licensee will remove activated concrete to ALARA and use release criteria in DP Table 6.3 to determine if concrete will remain in place. The VA will consider remediation complete when a structural engineer determines that decontamination and decommissioning (D&D) activities have reached the structural limits of the concrete.

The licensee commits to collecting soil samples during D&D activities to validate that surrounding soils are not activated or contaminated. The DOC will describe the sampling approach in the Final Status Survey Plan (FSSP).

3.2.3.2 Criteria for the Free-Release of Items and Articles from the AJBRF The licensee states in DP Section 6.2 that it will not release items from the AJBRF with detectable residual activity; and that detection limits will not be greater than the values in DP Table 6.3. The licensee states in DP Section 6.3 the VA may wish to apply alternative release criteria, but that the VA will submit any proposed criteria to the NRC for review and approval prior to deviating from the criteria in the DP.

3.2.4 Conclusion The NRC staff has reviewed the site characterization performed by the licensee. The NRC staff, based on its experience and engineering judgment, concludes that the licensee ensured that its contractors performed the characterization 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 also concludes that the release criteria proposed by the licensee based on the referenced generic screening thresholds are sufficient to demonstrate compliance with

10 10 CFR 20.1402 and are, therefore, acceptable. DP Section 16 specifies DP change criteria.

The licensee asserts that the licensee may make minor changes to the DP that do not change the original intent of the plan and that do not involve a situation that may impact radiation safety or worker/public dose. Further, the VA states that these changes may be approved by the Associate Chief of Staff for Research. The licensee states that if a significant change to the DP is required, the Reactor Safeguards Committee (RSC) will apply the test identified in 10 CFR 50.59 as it applies to non-power reactors in decommissioning. The licensee states that should the RSC determine that the change is significant and could pose a significant increase in potential worker, public, or environmental impacts, the licensee will obtain NRC approval prior to implementing the change.

The NRC staff has compared these criteria to those listed in Appendix 2 to NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, Rev.

1, and Appendix 17.1 to NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Part 2, Standard Review Plan and Acceptance Criteria, and has determined that the DP change criteria are consistent with NRC guidance for application of 10 CFR 50.59, and would not significantly increase potential worker, public or environmental impacts. However, other change criteria related to FSS in NUREG-1700 Appendix 2 are not included in chapter 16 of the DP. NRC staff has incorporated these criteria into section 2.D of the AJBRF license. Therefore, the NRC staff concludes that the change criteria in DP Section 16 taken with the changes to the AJBRF license are acceptable to ensure that NRC approval is required under the appropriate circumstances, such as the development of alternate release criteria.

Additionally, the licensees proposed screening values were compared to the trigger levels in the Memorandum of Understanding between the U.S. Environmental Protection Agency (EPA) and the NRC, reproduced in NUREG 1757, Appendix H, Consultation and Finality on Decommissioning and Decontamination of Contaminated Sites. The licensees proposed screening values are at or below the relevant consultation trigger values for industrial/commercial land use and therefore are considered appropriate.

3.3 Decommissioning Tasks 3.3.1 Scope of the Decommissioning Project DP Section 7 states that the VA plans to decontaminate and dismantle the AJBRF and associated systems in a safe manner and follow ALARA principles, the AJBRF Radiation Protection Program, and written procedures provided by the DOC.

The licensee states that contractors will decontaminate, dismantle, and demolish the facility, and remove both contaminated and non-contaminated materials. The licensee and independent consultants will supervise contractors. The licensee states that the DP is limited to those tasks required to decommission the reactor and associated support equipment. The licensee states that the DP does not address any re-construction efforts because the VAs goal of the DP is unrestricted use and therefore, any re-construction is immaterial to the evaluation of the DP.

The licensee states in DP Section 7 that decommissioning activities will start with the most contaminated components and structures, and progress to the least contaminated components and structures. The licensee states that this approach will minimize the potential spread of contamination to less contaminated or contamination free areas and will reduce overall

11 exposure to the decommissioning work employees by removing sources of exposure from the job site.

3.3.2 Schedule DP Section 1.8 states that D&D and the completion of the FSS will take 8 to 12 weeks. DP Figure 7.1 illustrates the proposed schedule upon NRCs approval of the plan and shows that from DP approval to completion of the FSS is approximately 13 weeks.

3.3.3 Individual Tasks DP Table 7.1 lists the following tasks and the estimated duration of each task:

Project planning 4 weeks Site Mobilization and Training 3 days Facility Preparation 2 days Reactor D&D 4 weeks o

Remove Control Rod Drives and Bridge o

Remove Fuel Storage Rack and Interference o

Remove Reactor Core Components o

Drain Tank and Process Water o

Remove Gunite, Liner, Tank, Soil Remove Ancillary Systems 2 weeks o Remove Water Filter System o Remove Water Cooling System o Remove pneumatic transfer system o Remove lab vent hoods and drain lines Remove Gunite, Liner, Tank, Soil Decontaminate Surfaces 1 week o Decontaminate Fuel Storage Pits o Decontaminate Room 540A (Source Storage) o Decontaminate Fuel Storage Pits o Decontaminate Miscellaneous Surfaces Perform FSS 1 week NRC Confirmation Surveys Not Applicable Secure Reactor Pit 3.3.4 Conclusion NRC staff observed the following:

Section 7 of the DP describes tasks and activities for site preparation and site D&D.

12 Section 4 of the DP describes locations and types of radioactive material, and Section 7 of the DP describes how components and systems will be removed.

Section 7 of the DP describes procedures for major decommissioning activities, including additional health and safety considerations where appropriate.

Section 7 of the DP provides a sequential schedule of D&D activities and interrelated tasks, including FSS.

Section 11 of the DP describes methods for decontamination of contaminated areas and waste packaging The NRC staff reviewed the DP with respect to decommissioning tasks including its FSSPs as noted above and concludes that the manner in which the licensee proposes to complete each of the decommissioning tasks is consistent with the guidance provided in NUREG-1537, Part 1, Appendix 17.1, Section 2.3, and is therefore acceptable. As described in Section 13 of the DP, the licensee will develop FSSPs in accordance with the standards in NUREG 1575; and NUREG-1757, Volume 1, Revision 2, Appendix B.

3.4 Decommissioning Organization and Responsibilities According to DP Section 8, the VA is committed to, and retains ultimate responsibility for compliance with the NRC license and all applicable regulatory requirements during decommissioning. The licensee states that it designed the decommissioning project management and organization to ensure that the licensee realizes the appropriate management control and oversight during decommissioning activities. DP Figure 8.1 illustrates the Omaha VAMC oversight of the decommissioning activities and the reporting relationships in a diagram.

DP Section 8 describes the following key licensee organizations and positions:

Omaha VAMC Director Has ultimate responsibility for VAMC Omaha operations including all licensed activity at the AJBRF.

Shall ensure that adequate funds are available to complete the decommissioning and final radiation survey activities in an efficient and timely fashion.

Has delegated authority for the overall management and oversight of the decommissioning activities at the AJBRF to the Omaha VAMC Associate Chief of Staff for Research (ACOS/R).

Omaha VAMC Industrial Safety Manager (ISM)

The Omaha VAMC ISM reports directly to the Omaha VAMC Director on all decommissioning matters.

Ensures safety of non-project personnel, patients, and hospital visitors.

Responsible for the development, communication and implementation of a safety culture at the Omaha VAMC and among the decommissioning contractor.

Has the authority and responsibility to stop or suspend any activity deemed unsafe or lacking in adequate safety controls and measures.

Reviews the D&D project Health and Safety Plan (HASP) to ensure proper controls are in place.

13 Investigates, documents, and evaluates any injuries that may occur during the decommissioning, and develops and monitors implementation of corrective measures.

Assists in the identification of hazardous materials within the AJBRF and maintains an inventory of such materials; and Inspects and monitors work sites and worker performance to identify any hazards and ensure safety program compliance.

Omaha VAMC Associate Chief of Staff for Research (ACOS/R)

The Omaha VAMC ACOS/R ensures that all personnel conduct decommissioning activities safely and the control of radioactive materials and radiation exposure is in accordance with regulatory requirements and ALARA.

Is responsible for the Quality Assurance (QA) program and for the effectiveness of the Radiological and Industrial Safety programs.

Has final approval authority of minor changes to the DP and to procedures that do not involve an un-reviewed safety issue.

Serves as the Chairman of the Reactor Safeguards Committee (RSC).

Ensures decommissioning activities comply with applicable federal, State, and local regulations.

Ensures the VA and contractor staffs perform decommissioning activities in a manner to protect the public, hospital staff, patients, and the environment.

Ensures suitable resources and qualified personnel are assigned sufficiently to safely achieve the decommissioning of the AJBRF.

Omaha VAMC RSC The Omaha VAMC RSC has the overall responsibility for radiation safety and adherence to the reactor license requirements and reports to the Omaha VAMC Director through the ACOS/R.

The RSC has ex-officio members to provide specific expertise in the areas of decommissioning, radiological engineering, environmental controls, and/or industrial safety. The following are standing RSC members in accordance with the NRC License and Technical Specifications requirements:

ACOS/R (Chairman)

Omaha VAMC Radiation Safety Officer (RSO)

Omaha VAMC Reactor Facility Manager The RSC reviews decommissioning procedures, decommissioning activities dealing with radioactive material, and radiological controls. The RSC evaluates all proposed DP changes to determine if un-reviewed safety issues exist.

Reviews proposed DP or procedure changes, including changes in monitoring or control equipment, systems, or testing to determine if there are safety issues as defined in 10 CFR 50.59;

14 Oversees and reviews any operations that could potentially release radioactivity to the environment and planned releases of radioactive material to the environment; Reviews all reports of violations of technical specifications, license, and procedures that involve personnel and/or public safety.

Reviews NRC inspection reports and the licensee responses and corrective actions.

Performs audits of plans and programs, such as ALARA, QA, and industrial safety to ensure compliance with procedures.

Reviews the dosimetry program and decommissioning workers radiation exposures.

The AJBRF technical specifications and the RSC charter govern these functions.

Reviews and approves all Radiological and Hazardous Work Permits (RHWP) and RWPs that have an estimated collective total exposure of 2.5 person-rem or greater VA Central Office (VACO) Project Manager (PM)

The VACO PM provides overall guidance with respect to project planning, scheduling and execution, and interactions with VA funding source, VA contracting, and the NRC PM. The VACO PM serves as the Contraction Officer's Representative (COR).7 As the COR, the VACO PM is responsible for monitoring contractor schedule adherence, schedule changes due to unforeseen issues, work scope expansion, or contractor performance. The VACO PM is an ex-officio member of the RSC.

Provides technical leadership for the overall D&D project.

Acts as primary overall project point of contact for the Omaha VAMC and VA staffs and the DOC.

Provides organizational leadership to decommissioning project.

Incorporates the contractor work schedules into a master schedule for the decommissioning project; Coordinates with the NRC PM.

Omaha VAMC Reactor Facility Manager The Omaha VAMC Reactor Facility Manager is responsible for ensuring compliance with the NRC license and technical specifications, including performance of necessary surveillance, testing, and inspections. The Reactor Facility Manager provides day-to-day interactions with VA's technical representatives, contractor, and the DOC. The Reactor Facility Manager reviews all D&D plans and provides technical oversight during decontamination and removal of reactor components and systems.

Provides technical representation for the Omaha VAMC RSC.

Interacts daily with VA's Technical Representative Contractor, the VACO PM, the DOC, and various elements within the VA and VAMC operations.

7 DP Section 8.1.5 states that the VACO PM serves as the Contraction Officer's Technical Representative (COTR). The Federal Acquisition Regulations no longer use this term and has replaced it with COR.

15 Omaha VAMC Radiation Safety Officer (RSO)

The Omaha VAMC RSO is responsible for ensuring the radiological safety of the Omaha VAMC patients and staff, primarily concerning radiation treatment and diagnosis procedures. The Omaha VAMC RSO plays no direct role in the decommissioning of the AJBRF. The Omaha VAMC RSO is a standing member of the RSC.

VA Resident Engineer / Senior Resident Engineer (RE/SRE)

The RE/SRE is responsible for monitoring contractor schedule adherence; monitoring overall safety concerns to the hospital at large, and is a source of information to the RSC and the DOC.

Serves as the construction site RE/SRE and COR for the DOC Has stop/start work authority Coordinates vendor work plans and logistics to minimize work delays and conflicts with Omaha VAMC Works with VA Technical Representative Contractor to ensure that decommissioning activities do not endanger VA personnel, patients, and visitors to the VAMC.

Follows budget, schedule progress, and prepares reports on progress, variance, and trends to the VACO PM and RSC.

VA Technical Representative Contractor The VAs design and oversight contractor, AECOM, will serve as the VA Technical Representative Contractor. The VA Technical Representative Contractor will provide daily radiological safety oversight of the DOC to ensure that the DOC is operating within the requirements of the DP, Radiation Safety Program (RSP), and other applicable operation and safety plans and procedures. AECOM will provide a certified health physicist with similar D&D experience as the VA Technical Representative Contractor to support on-site personnel.

Provides the D&D Project RSO Ensures the DOC and subcontractors comply with the VA-approved Quality Assurance Program Plan.

Initiates nonconformance reports (NCRs)/corrective action reports (CARs) as required and ensures that the VA technical contractor, DOC, and subcontractor staffs identify and perform any corrective actions.

Reviews D&D plans and procedures to ensure contractor and subcontractor personnel use the appropriate extent of controls during the execution of the decommissioning activities and that contractor and subcontractor personnel comply with the AJBRF license requirements and applicable regulations.

Reviews dismantling plans to ensure contractor and subcontractor personnel maintain appropriate configuration control for safety.

Reports to the RSC on the status of QA, compliance with plans, procedures, and safety.

Maintains records of quality and safety audits or inspections.

16 Provides the VA with technical input for communications with the NRC PM and inspectors.

Minimum qualifications for the VA Technical Representative Contractor are:

Health physicist and/or radiological engineer with more than ten years of D&D experience, which includes experience with decommissioning research reactors.

D&D Project RSO DP Section 8.2.1 and DP Figure 8.1 refer to the D&D Project RSO and the DOC RSO interchangeably. The D&D Project RSO will be responsible for the daily execution of the policies and practices pertaining to the safe decontamination, demolition, and disposition of radioactive material at the AJBRF. The D&D Project RSO serves as a non-voting consultant to the RSC and has the authority and responsibility to suspend any unsafe activity involving radioactive material and/or work performed in radiation areas. Such activity would include actions that conflict with ALARA principles, violate regulations, and the unmonitored or unplanned release of radiation to the environment.

Verifies contractor-provided radiation detection instruments used for measurement of radiation and contamination quantities are properly maintained and calibrated.

Maintains and implements the ALARA and personnel dosimetry program for the VA and DOC staffs.

Evaluates and distributes the results of dosimeter measurements.

Maintains an inventory of the radioactive material possessed under the authority of the AJBRF license.

Reviews radioactive material/waste shipping documents.

Oversees the environmental monitoring programs.

Provides or determines the adequacy of radiation safety training for all VA and contractor D&D personnel working at the AJBRF in accordance with the Omaha VA Radiation Protection (RP) Program.

Ensures consistency of DOC radiological protection procedures with the Omaha VA RP Program.

Ensures that personnel involved in D&D activities with potential radiological exposure comply with the AJBRF license, approved procedures, the DP, and applicable federal, State, and local regulations.

Reviews and approves all DOC Radiation Work Permits (RWP).

Evaluates procedural and RWP compliance by workers and supervision.

Prepares and maintains the documentation required by 10 CFR 20 and the AJBRF RP Program.

Advises the RSC about all matters regarding radiation monitoring and radiation safety during decommissioning activities.

17 DOC The DOC is responsible for the execution of the D&D in accordance with federal and State regulations, the AJBRF Technical Specifications, and the contractual requirements of the decommissioning contract. The DOC shall provide personnel and any subcontractors to provide all the services and skills required for the decontamination, dismantlement, disposal, and reconstruction activities. The DOC will develop and execute detailed work plans, policies, and procedures for the D&D, which includes the FSSP and final status survey report (FSSR).

Responsibilities include, but are not limited to the following:

Directs and supervises D&D activities and resolves work problems.

Coordinates, verifies, and validates the proper implementation of work control and oversight programs.

Manages D&D and waste management services.

Prepares radioactive waste manifests.

Prepares and implements the FSSP.

Supports the analysis and reporting requirements of the RSC.

Provides a Site Safety Officer Site Superintendent The DOC site superintendent directly oversees the D&D, reports directly to the VACO PM, and has the authority and responsibility to stop or suspend work of any personnel performing D&D work.

Executes the DP and FSSP Prepares the RHWPs and RWPs and submits them to D&D RSO for approval.

The Project RSO will review RHWPs/RWPs for compliance with Omaha VAMC and DOC safety programs and procedures.

Provides daily updates to D&D management team 3.4.1 Conclusion The NRC staff has reviewed the licensees DP with respect to the proposed decommissioning organization and responsibilities. The NRC staff concludes that this information satisfies the requirement of 10 CFR 50.82(b)(4)(v) regarding quality assurance provisions because the licensee has provided reasonable assurance that organizational structures needed to safely decommission the AJBRF are in place. In addition, the licensee has committed to ensuring that the Omaha VAMC RSC will properly oversee all decommissioning activities conducted by the DOC, 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 AJBRF is consistent with the guidance on the role and composition of the facility safety committee provided in Section 2.4 of Appendix 17.1 to NUREG-1537 and is, therefore, acceptable.

3.5 Training Program

18 According to DP Section 7.2, Site Mobilization and Training, initial mobilization of DOC personnel will include the travel of workers to the site, training of workers, and attainment of required equipment for the DP tasks. All personnel will receive safety and securities briefing provided by Omaha VAMC personnel. DP Section 8.5 states that unescorted individuals including employees, contractors, and visitors that require entry to the AJBRF Controlled Access Areas will receive radiation and industrial safety training in accordance with the AJBRF Radiation Protection and the Omaha VAMC Safety Manual. DOC will perform formal training sessions upon initial employment or on-site arrival for contractors.

General Site Training DP Section 8.5.3, states that the DOC will provide workers and regular visitors training to ensure compliance with the requirements of the NRC (Chapter 1 of 10 CFR), the EPA (40 CFR 24.311), and the Occupational Safety and Health Administration (OSHA) (29 CFR) regulations and the Omaha VAMC safety requirements. The training will provide personnel with the means to recognize and understand the potential risks involved with personnel health and safety issues associated with the work at the facility.

The Omaha VAMC ISM will review the DOC's safety plans and may conduct periodic safety inspections. The VA Technical Representative Contractor will verify training and documentation. The DP states that the DOC will conduct daily tailgate sessions with workers that will include deficiencies discovered during audits and inspections, training on procedural changes, and applicable industry events or occurrences. The licensee commits to conducting tailgate sessions in response to lost-time injuries and industrial and/or radiation safety violations. The training will include the following:

Training on the DP General safe work practices Radiological, chemical, and demolition-induced hazards Worker rights and responsibilities Use of personnel protection equipment Hearing conservation training Fall protection Radiation Worker Training According to DP Section 8.51, the DOC will provide general Radiation Worker Training (RWT) to D&D personnel working in restricted areas that are commensurate with the duties and responsibilities being performed. The D&D Project RSO will document and approve the initial or annual refresher training. The licensee states that training program objective is to ensure that personnel understand the responsibilities and methods for minimizing exposure to radiation and practices for safe handling of radioactive materials. The licensee states the training will ensure that D&D personnel have appropriate knowledge to work in controlled areas in accordance with the requirements of 10 CFR 19.12 and the AJBRF Radiation Protection program. The licensee states that it may exempt workers from participating in training if they have documented equivalent RWT from another site; however, no workers will be exempted site-specific training

19 on administrative limits and emergency response. Any personnel exempted from training will be required to pass a written examination and attend demonstration exercises. The licensee states that the Reactor Facility Manager will maintain training records. The radiation safety training provided to any worker will include the following:

Fundamentals of radiation Radiation monitoring techniques Radiation monitoring instrumentation Emergency procedures Radiation hazards and controls Contamination limits and controls Biological effects of radiation External and internal radiation exposure limits and controls Principles and techniques of D&D activities; ALARA philosophy and program Respiratory protection requirements, philosophy, and program Management and control of radioactive waste, including waste minimization practices The licensee states the workers will be required to participate in the following demonstrations:

Proper procedures for putting on and removing protective clothing Ability to read and interpret self-reading and/or electronic dosimeters; Proper procedures for entering and exiting a contaminated area Frisking techniques An understanding of the use of RHWP by working within the requirements of a given RHWP Respiratory Protection Training DP Section 8.5.2 states that the DOC will provide workers that require the use of respiratory protection devices with respiratory protection training in the devices and techniques that will be required for use. The licensee states that the respiratory protection training program will follow 10 CFR Part 20 requirements.

3.5.1 Conclusion NRC staff notes that Section 8.5 of the DP, as described above, contains the following elements identified in NUREG-1537, Appendix 17.1, Section 2.5:

Special training for D&D activities Contractor training Training in the principles of D&D

20 Health Physics Training Training in the use of monitoring and safety equipment Compliance with 10 CFR Part 19 Section 8.5 of the DP also demonstrates that the licensee is aware of the differences between normal operation and decommissioning Based on its review of the licensees training program as described in the DP, the staff has determined that the proposed licensee-training program is consistent with the NRC guidance for training for decommissioning research reactors provided in Section 2.5 of Appendix 17.1 to NUREG-1537, Part 2. Therefore, the NRC staff concludes that the licensees training program is acceptable. The licensee also recognizes 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 by 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.

3.6.1 Decommissioning Contractors DP Section 7.1 states that the VA will develop a scope of work and bid specifications to contract a DOC to decontaminate, demolish, package and dispose radioactive waste, and conduct FSS activities. After the VA has selected a DOC, the VA will coordinate with the DOC to schedule upcoming D&D activities to ensure that the licensee clearly identifies and documents the roles and responsibilities between the Omaha VAMC and the contractor. The selected DOC will manage the physical aspects of its portion of the decommissioning work including QA, health physics, safety, waste processing, and waste packaging and shipping. However, as described in DP Section 8, the VA will continue to maintain overall responsibility for health and safety, compliance with regulations, and applicable license conditions. The licensee states the Omaha VAMC ACOS/R will ensure the licensee selects the proper resources and qualified personnel to safely achieve the decommissioning of the AJBRF.

DP Section 12 states that the VA commits to a comprehensive and effective QA program as an integral part of the D&D effort. The licensee states that the QA program will provide a systematic approach to ensure that D&D efforts comply with established policies, procedures, and recognized good practices. The licensee states that to implement QA, VA will exclusively use contractors with demonstrated experience in D&D projects of similar size and scope and require the DOC to have a Quality Assurance Project Plan (QAPP). The licensee states that the RSC will review and approve the QAPP. The licensee states the QAPP will include the following:

Written definitions of authority, duties, and responsibilities of contractor managerial, operation, and safety personnel.

Defined organizational structure Assigned responsibility for review and approval of plans, designs, procedures, data, and reports.

Personnel training

21 Written procedures for D&D activities Documentation and data management Corrective action process 3.6.1. Conclusion NRC staff observes that Section 8.5 of the DP, as described above, contains the following elements identified in NUREG-1537, Appendix 17.1, Section 2.6:

Delegation of authority for decommissioning activities to contractors, while maintaining overall licensee responsibility for regulatory compliance Discussion of administrative control system Discussion of quality assurance program Description of contractor type and scope of work to be completed The DP does not discuss detailed performance history of the contractors, as they were not selected in advance of preparation of the DP. However, the NRC staff has reviewed the criteria the licensee used to select the VA Technical Contractor Representative, AECOM, (see Section 8 of the DP) and will use to select the DOC. The selection criteria cover all skill areas necessary for successful decommissioning project management and performance.

Consequently, the NRC staff concludes there is reasonable assurance that the licensee has selected a Technical Contractor Representative and will select a DOC contractor with adequate qualifications.

Therefore, based on its review of decommissioning contractor management as described in the DP, the staff has determined that the proposed licensee-training program is consistent with the NRC guidance for training for decommissioning research reactors provided in Section 2.6 of Appendix 17.1 to NUREG-1537, Part 2. Therefore, the NRC staff concludes that the licensees decommissioning contractor management program is acceptable.

3.7 Radiation Safety and Health Program 3.7.1 ALARA Program DP Section 9.2.2 states that the licensee will conduct the AJBRF D&D activities in such a way to minimize exposures to workers and the public ALARA. The licensee states that it expects all decommissioning personnel to be knowledgeable of D&D activities and to follow all ALARA requirements documented in policies and procedures, and be responsible for minimizing dose to themselves and others.

3.7.2 Radiation Safety Program (RSP)

According to DP Section 9.2.1, the Omaha VAMC commits to a RSP that controls radiation exposure to workers and members of the public that avoids unnecessary and accidental doses, and maintains effluents and doses to workers below regulatory limits. The DOC will prepare a Radiation Protection Plan (RPP) and a HASP for VA review and approval. The D&D RSO will implement the D&D projects RSP and the approved DOC documents will implement safety procedures for the D&D project.

22 The RSP will include the following:

Respiratory Protection Program Internal and External Exposure Controls Contamination Control Radiation Monitoring Instrumentation Program Control Environmental Releases Health Physics Audits, Inspections, and Record Keeping Respiratory Protection Program DP Section 9.2.3 states that the licensee will use respirators to control personnel exposure to airborne radioactive materials when administrative and engineering controls are not effective and the use of respirators result in the Total Effective Dose Equivalent (TEDE) being ALARA.

The DOC shall implement the respiratory protection program in compliance with the requirements of 10 CFR Part 20, Subpart H and under the supervision of the D&D RSO. The licensee states that the respiratory protection program shall include requirements in 10 CFR 20.1103 for radiation protection and 29 CFR 1910.134 for non-radiological hazards and comply with the guidelines in the NRC Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" and NUREG 0041, "Manual of Respiratory Protection Against Airborne Radioactive Material."

DP Section 9.2.3.2 states that the DOC will collect localized air samples including breathing zone air samples where the licensee expects airborne radioactivity concentrations are likely to exceed the criteria of 10 CFR 20.1502(b). The DOC will set alarm points for continuous air monitors (CAM) at 10 percent of the derived air concentration (DAC). The licensee states that if monitors or grab samples show airborne contamination greater than 10 percent of the DAC, the DOC will suspend work and notify radiation safety personnel.

DP Section 9.2.3.3 states the DOC will normally select respiratory protection equipment (RPE) that has a protection factor greater than the anticipated peak airborne concentration expressed as a multiple of total DAC. However, the DOC may select RPE that has a protection factor less than the anticipated peak airborne concentration expressed as a multiple of total DAC if the licensee expects use of that equipment will result in a lower TEDE. The licensee states that it will document the evaluation for determining the respiratory equipment in the RPP or the RWP.

The licensee will assign protection factors for respiratory protection equipment in accordance with Appendix A of 10 CFR Part 20 and 10 CFR 20.1703.8 DP Section 9.2.3.4 states that it will ensure that personnel that require RPE shall receive medical evaluations and have a physician certify that the individual is certified to wear RPE and that the individual is physically and psychologically fit prior to wearing any respiratory protection device. DP Section 9.2.3.5 states the licensee will ensure that personnel responsible for the cleaning and maintenance of RPE shall receive documented training necessary for fulfilling their responsibilities.

8 The DP has an apparent typo. The text reads that protection factors will be assigned in accordance with Appendix A of 10 CFR 20.1103. There is no 20.1103. Appendix A, Assigned Protection Factors For Respirators, is an appendix to 10 CFR Part 20 and 10 CFR 20.1703 addresses protection factors.

23 Internal and External Exposure Controls The DP Section 9.2.4 states the DOC RPP shall include air sampling and in vitro bioassay and/or in vivo bioassay monitoring in order to comply with requirements in 10 CFR 20.1204 and

§ 20.1502. The licensee plans to follow recommendations of the National Council on Radiation Protection and Measurements and on regulatory requirements to protect the embryo/fetus during a female worker's pregnancy. The licensee acknowledges that a declaration of pregnancy is entirely voluntary and that a female worker does not require medical proof. The licensee states that a female worker who declares pregnancy will need to inform the D&D RSO in writing with an estimated date of conception, so that the D&D RSO can estimate the dose to the embryo/fetus prior to the declaration. The licensee states that it will not permit a declared pregnant worker to enter airborne radioactivity areas nor be assigned to tasks that could lead to internal radionuclide intakes.

The licensee states in DP Section 9.2.6 the DOC shall provide appropriate dosimetry to determine D&D workers external radiation exposure that enter and/or work in the restricted area. The licensee states that the VA will continue to monitor its personnel. The DOC will use primary and secondary dosimetry to monitor external exposures. Optically Stimulated Luminescence Dosimeters or other primary dosimetry devices shall be capable of measuring the deep dose equivalent at a tissue depth of one centimeter (cm), measuring the lens dose equivalent at a tissue depth of 0.3 cm, and measuring the skin dose equivalent at a tissue depth of 0.007 cm. The licensee states that it may use secondary dosimetry (e.g., self-reading pocket dosimeters and alarming dosimeters) to monitor dose between dosimetry processing periods and as a backup to the primary dosimetry.

The licensee will establish administrative dose control limits that are less than the 10 CFR Part 20 limits to ensure personnel exposure do not exceed regulatory limits. The D&D RSO will review and approve dose assessments prior to assigning a dose other than that measured by a primary dosimetry device. The licensee will sum internal and external doses whenever positive doses are measured, excluding doses to the lens of the eye, skin, and extremities in accordance 10 CFR Part 20.

Contamination Control The DP Section 9.2.8 states it will control the spread of contamination to clean areas, minimize the need for respiratory protection devices, and maintain personnel internal and external exposures ALARA. The licensee states that it will achieve this goal by establishing good work practices. The DOC will conduct routine contamination surveys at a minimum on a weekly basis to monitor the spread of contamination and identify areas to be decontaminated. The licensee commits to performing non-routine surveys to detect contamination.

The licensee will conduct personnel contamination surveys (e.g., nasal and face smears, sampling of hair) when the work performed results in unexpectedly high levels of loose or airborne contamination. The licensee states that management policy requires personnel contamination is ALARA. The DOC RPP will specify contamination limits, but states that areas containing removable contamination that exceed 1,000 dpm/100 cm2 of beta/gamma-emitting radionuclides or 20 dpm/100 cm2 of alpha emitting radionuclides will be considered contaminated. Equipment, materials, and tools shall be controlled as radioactive material when total contamination exceeds 100 counts per minute (cpm) above background using a detector at least as sensitive as a pancake GM detector or removable contamination exceeds 1000

24 dpm/100 cm2 of beta/gamma-emitting radionuclides or 20 dpm/100 cm2 of alpha-emitting radionuclides.

Radiation Monitoring Instrumentation Program DP Section 9.2.9 states that the DOC's RPP will include procedures that comply with federal regulations and the VA license and procedures and address inventory, issuance and control, calibration, operation, response testing, maintenance, repair, and quality control of radiation protection and hazardous instrumentation and equipment. The AJBRF D&D project will use various types of radiological and hazardous measurement instrumentation for radiation protection and monitoring hazards.

Control Environmental Releases The DP Section 10 describes the Environmental Monitoring and Control Program. The licensee commits to conducting D&D operations in a controlled manner to minimize both public and occupational radiation exposures. The licensee states that the two site characterization surveys show that most structures and surface areas within the AJBRF are not contaminated and areas outside the AJBRF are not contaminated above natural background levels. The licensee states the DOC will decontaminate or remove contaminated portions of equipment and material in such a manner to minimize the spread of contamination and thus, reduce the prospects of elevated releases to the environment.

The licensee states that it will routinely monitor the environment by measuring the quantity of direct radiation and radioactive material releases outside of the AJBRF. The licensee states that environmental monitoring will begin prior to starting D&D operations, which will serve as the baseline data benchmark for comparison with in-progress and final survey results following completion of the D&D project.

The licensee states the VA is committed to maintaining occupational and public exposures, as well as effluent discharges ALARA. The licensee will use engineering controls and ALARA principles to accomplish this goal. The licensee has a zero release policy for liquids and particulate airborne effluents associated with the AJBRF. The licensee forbids discharge of hazardous or radioactive liquids by way of sinks and drains or other means of discharge to the environment during the decommissioning. The licensee states that the DOC will treat the contaminated reactor tank water through the existing demineralizer system and sample the treated water prior to discharge to verify that the tank water is free of radioactive contaminants of concern greater than 50 percent of the applicable isotope-specific criteria contained in Table 3 of 10 CFR Part 20, Appendix B.

The licensee states that the DOC will disable laboratory hoods in the AJBRF, which the site characterization showed were free of contamination, to prevent contaminating the exhaust system during D&D operations. The licensee will install and use a temporary High Efficiency Particulate Air (HEPA) filtration system, with the exhaust capacity required by AJBRF Technical Specifications, in place of the building ventilation system. The licensee will discharge to the ground level outside the AJBRF. The licensee will collect routine and special air samples from the exhaust point using portable air monitors. The licensee states monitoring instrumentation will have the capabilities of a CAM for real-time monitoring as well as capabilities to collect samples for laboratory analyses.

25 The licensee will use local engineering controls, such as containments and HEPA filters, to minimize the release of airborne radioactivity to general areas within the AJBRF. The licensee will establish airborne radioactivity action levels an administrative control concentration of 50 percent of the applicable criteria for Eu-152 in Table 2 of 10 CFR Part 20, Appendix B or 1.5E-11 pCi/mL. The licensee uses the Eu-152 value because it and Co-60 are the most probable contaminates in the activated concrete; however, the release limit for Eu-152 is lower than the Co-60 limit.

The licensee states that it will establish administrative controls to ensure that samples collected from the discharge areas are representative. The licensee will prepare duplicate and blank samples for tank water analysis. The licensee will periodically recount air samples analyzed on-site at a frequency of 20 percent.

Health Physics Audits, Inspections, and Record Keeping In DP Section 9.4, the licensee states that the licensees RSO9 will maintain all records related to health and safety of radiation workers and individual members of the public in accordance with the AJB reactor license technical specifications. The licensee states that it will prepare and maintain all records in accordance with the AJBRF Radiation Protection Program.

3.7.3 Occupational Dose Estimates In DP Section 9.1, the licensee estimates the occupational exposure to the DOC staff to complete the D&D Project to be 1 person-rem for each task listed in DP Table 9.1. This dose estimate for decommissioning of the reactor was prepared using the individual work activity durations and work crew sizes, based upon the results of the characterization results. The licensee states that the characterization studies show that the reactor facility and associated outside areas of the facility do not contain widespread nor significant levels of residual radioactivity. The licensee states that it considers the site to have low residual concentrations of radioactivity present in small areas of the facility and not widespread throughout the facility. The licensee based the dose estimates using its contractors professional experience involving similar work, task durations, projected man-loading, and schedule and are considered conservative for many tasks where dose rates are assumed to be 0.1 and 0.05 mrem/hr.

3.7.4 Nuclear Criticality Safety The licensee states in DP Section 9.3 that there is not fissile material at the AJBRF and consequently, no nuclear criticality safety issues of concern.

3.7.5 Industrial Safety Program In DP Section 9.5, the licensee states that the DOC prepares the HASP, which complies with OSHA and the licensees occupational health and safety requirements during D&D operations.

The RSC and Omaha VAMC ISM will review and approve the HASP. DP Section 11.3.1 describes the licensees D&D hazard communication program and states that site specific 9 DP Section 9.4 states the AJBRF RSO will maintain records in accordance with the Technical Specifications. The DP never makes any other reference to the AJBRF RSO nor describes this position.

The DP Section 8 and Figure 8.1 describe the Omaha VAMC RSO in the D&D project management oversight. Technical Specification 6.3(2) assigns the hospital RSO (e.g., Omaha VAMC RSO) with the responsibility of implementing the AJBRF radiation protection program.

26 training will include the types and identification of hazardous materials used at the AJBRF to comply with 29 CFR 1910.1200 and the Omaha VAMC's Right to Know Program. DP Sections 8.1.3 and 8.5.3 further describe the Omaha VAMC ISMs industrial safety responsibilities. The Omaha VAMC ISM inspects and monitors work sites and worker performance to identify any hazards and ensure safety program compliance; and has the authority and responsibility to stop or suspend any activity deemed unsafe or lacking in adequate safety controls and measures.

3.7.6 Conclusion The NRC staff has reviewed the licensees commitments to decommissioning ALARA and radiation safety programs and the licensees estimate of the potential dose from the AJBRF D&D activities. The NRC staff concludes that, because the radiation protection controls committed to by the licensee will maintain doses within the 10 CFR Part 20 limits and be ALARA, these commitments are adequate to protect occupational and public health and safety.

Therefore, these commitments satisfy the requirement of 10 CFR 50.82(b)(4)(ii), which states that the DP must include a description of the controls and limits on procedures and equipment to protect occupational and public health and safety. Additionally, the NRC staff review of the Industrial Safety Program indicates that it will ensure appropriate personnel protection consistent with OSHA requirements and is acceptable.

3.8 Radioactive Waste Management The DP Section 11.2 describes the licensees proposed program to manage and control the management and disposition of solid and liquid radioactive waste for the D&D project. The licensee states that AJBRF D&D operations will require the handling and processing of volumes of radioactive waste typical of a research reactor of this type to reduce the residual levels of radioactivity to levels that allow for license termination and the release of the site for unrestricted use. The licensee will manage the materials that the DOC does not decontaminate and/or release as radioactive waste. According to DP Table 11.1, the licensee estimates the D&D project will produce approximately 31.3 cubic meters (m3) (41 cubic yards (yd3)) of radioactive waste.

The DP states that solid radioactive waste consists primarily of neutron-activated aluminum/graphite from the reactor internals, concrete and structural materials from walls, exhaust hoods, steel and concrete/epoxy/gunite from the reactor tank, steel/rust from laboratory drains, and a small volume of mixed waste consisting of contaminated lead paint and contaminated asbestos-containing floor tiles. The NRC staff notes that waste disposal costs directly relate to the activity, volume, and weight of the materials requiring disposal. DP Section 11.2 states the licensee will use strategies to minimize waste throughout the D&D operation, including source reduction, reuse, decontamination, volume reduction, and waste stream segregation.

The DP Section 11.1.2 states that the DOC will be required to develop and implement a Waste Management Plan for the AJBRF D&D project. The Waste Management Plan will include detailed guidance for the characterization, sampling, classification, segregation, handling, packaging, manifesting, transporting, and disposal of waste generated by the decommissioning.

The licensee states the RSC will review and approve the Plan and it will define the waste management and pollution prevention programs prior to beginning D&D operations. The licensee will implement the program will by written administrative and technical procedures to meet the requirements of the AJBRF license Technical Specifications and applicable federal and State regulations.

27 The licensee states in DP Section 11.2 that that the DOC will perform appropriate processing, packaging, and monitoring of solid and liquid wastes generated during the decommissioning process in accordance with formally approved administrative and technical procedures. The licensee states that based on recent experience at the University of Arizona, the licensee expects all radioactive waste generated during D&D to meet the definition of Class A waste as defined in 10 CFR 61. The licensee states that disposal of Class A waste may occur at Energy Solutions in Clive, Utah. However, the licensee states that Waste Control Specialist in Andrews, TX can accept Class A waste and under certain circumstances can import these wastes from out-of-compact states. It is the DOCs responsibility to determine cost effectiveness of the waste disposal options while maintaining compliance with all applicable regulatory requirements.

3.8.1 Fuel Removal The DP Section 11.5 states that there is no longer any reactor fuel. The licensee states that it transferred fuel including fission chambers, to the USGS research reactor facility in Denver, Colorado in June of 2002.

3.8.2 Radioactive Waste Processing The DP Section 11.2.2 states that the DOC will process all radioactive waste in controlled areas to minimize radiation exposure to personnel and the spread of contamination. The licensee expects all waste to meet the definition of Class A waste as defined in 10 CFR 61. The licensee commits to packaging waste to meet the applicable requirements of the US Department of Transportation (DOT) regulations (49 CFR Part 173), 10 CFR Part 20, 10 CFR Part 71, as well as the disposal facility's criteria for transportation and disposal for each decommissioning waste stream. The licensee states that the Waste Management Plan will include instructions for determining the 10 CFR Part 61 waste classification, as well as methods to determine the radionuclide content of a container through a combination of direct measurements, radiation-shielding calculations, and the use of appropriate scaling factors.

DP Section 11.2.6 states that most of the liquid on-site is within the reactor tank. As discussed in SER Section 3.7.2, Control Environmental Release, the DOC will treat the contaminated reactor tank water through a demineralizer system and sample the treated water prior to discharge for compliance with Table 3 of 10 CFR Part 20, Appendix B. As stated in DP Section 10.2.3, the licensee must verify that the tank water is free of radioactive contaminants of concern greater than 50 percent of the applicable isotope-specific criteria contained in Table 3 of 10 CFR Part 20, Appendix B. The licensee states in DP Section 11.2.6 that it is highly probable that it cannot discharge residual water in the water filter/cooling systems because the residual water does not meet the discharge limits. The licensee states that the DOC will stabilize, solidify, and ship this residual liquid waste as solid waste. The licensee commits to including the management of this residual liquid waste in the Waste Management Plan.

In DP Section 11.2.3 the licensee states that solid radioactive waste awaiting shipment will be stored in accordance with 10 CFR Part 20 in posted controlled areas and away from personnel traffic or work areas. The DOC will perform periodic inspections of these areas and those containers stored within the controlled area boundaries to ensure that the licensee maintains package integrity. The licensee states that the DOC will implement applicable security and radiological controls, as necessary, to prevent unauthorized access to the area and the radioactive materials stored inside.

28 3.8.3 Radioactive Waste Disposal The licensee states in DP Section 11.2.4 that a certified radioactive waste broker will ship solid radioactive waste in accordance with applicable NRC and U.S. Department of Transportation (DOT) regulations and the Waste Management Plan. The licensee states that the licensee will apply the quality control requirements of 49 CFR 173.475 prior to each shipment. The licensee states that the licensee may use vendors to process waste at various stages, such as using a vendor to process the disposal of the Po-Be source. Table 11.1 provides estimated waste volumes, by area of the AJBRF, totally an estimated 31.3 cubic meters of low level radiological waste.

3.8.4 Mixed Low-Level Radioactive/Hazardous (Mixed Waste)

DP Section 11.2.7 states that the licensee will process and dispose of hazardous materials and hazardous wastes following the Omaha VAMC hazardous waste management program and commits to including this commitment in the Waste Management Plan. The licensee states that the initial site characterization identified potential lead-based paint and asbestos as a mixed waste. The licensee states that a certified waste broker will be classify and dispose of these materials at an EPA/NRC authorized facility if necessary. The licensee states that it does not foresee using hazardous chemicals or other hazardous substances during D&D operations that could generate a mixed waste.

3.8.5 Conclusion NRC staff reviewed the licensees proposed decommissioning waste management program and determined that it is consistent with the NRC guidance on research reactor decommissioning waste management provided in Section 3.2 of Appendix 17.1 to NUREG-1537, Part 2, and observes that the DP contains the following:

A discussion of the disposition of all fuel A description of waste management systems and procedures A description of planned radioactive waste disposal, including estimates of type and quantify of waste, transportation modes, and disposal locations The DP also discusses the general industrial safety program in Section 11.3.1, Hazard Communication, which includes a discussion of non-radiological accident prevention and mitigation Additionally, the NRC staff determined that the licensee has demonstrated experience in safely managing radioactive waste generated during facility activities performed in preparation for decommissioning. Based on these reviews, the NRC staff concludes that the licensees proposed decommissioning radioactive waste management program demonstrates a program sufficient to control, process, package and transport radiological waste, and is therefore acceptable.

3.9 Radiological Accident Analyses ABJRF DP Section 14 describes the potential radiological accidents during reactor decommissioning that the licensee evaluated using consequence levels discussed in US Department of Energy (USDOE) Standard DOE-STD-1120-2005, "Integration of Environment,

29 Safety, and Health into Facility Disposition Activities" and guidance in NUREG 1537, Part I, Appendix 17.1, Section 3.3. The licensee states that the most probable potential for radiological accidents during the ABJRF D&D project relate to handling contaminated liquids in the reactor tank and the ancillary water-cooling and filter systems. The licensee also acknowledges that uncontrolled releases of airborne contamination could also occur during the demolition and segmentation of activated and/or contaminated reactor components or during a transportation accident.

The licensee explains why it did not address accidents involving dropping an activated reactor assembly, fires, or fissile material. There are no fissile materials located on-site that could result in a criticality incident. The licensee states that the surface contamination on activated reactor assembly parts would not be sufficiently high to release significant quantities of radioactive materials if dropped. The licensee states that the most probable result would be an increase in external exposures in handling the reactor assembly and preparing it for waste shipment. The licensee states that fire releasing uncontrolled airborne contamination is unlikely because the majority of the combustibles on-site will be non-radioactive contaminated materials. The radioactivity on potentially contaminated combustibles, such as personal protective clothing, and clean-up rags and towels, would not be enough to result in a significant release during a fire.

The licensee considered the following radiological accidents in the DP to present the highest potential consequences:

Release of contaminated liquid Release of airborne contamination Hot Particles Transportation accident 3.9.1 Release of Contaminated Liquid ABJRF DP Section 14.1 states that an uncontrolled release of radioactively contaminated liquids could result in the contamination of workers, the ABJRF, or the environment. The spilling of contaminated water could occur during pool water pumping or liquid removal operations from the waste systems. These liquids may contain Co-60, Cs-137, and hard-to-detect (HTD) isotopes, such as H-3, Fe-55, and Ni-63. Hoses could leak or break, resulting in an uncontrolled release. To mitigate the extent of such releases, the DOC will only perform processes involving contaminated liquids with personnel present. Personnel will check for leaks and spills and respond by shutting down the activity, which will prevent additional loss of water from the system. Additionally, the pumps and filters will be in the vault, with limited space, and, a spill kit will be readily available to respond to any incidents. Thus, the licensee does not expect much impact from a spill within the vault.

The licensee estimates that an uncontrolled release of the contaminated water may result in minor ingestion, short-term dermal contact, and external exposures. The licensee used Co-60, which has the lowest annual limit on intake (ALl) of the contaminants of concern, to calculate worse case exposures from oral ingestion. The licensee calculates that the ALI for Co-60, 200

µCi, coincides with a committed effective dose equivalent (CEDE) of 5 rem. The licensee estimates that more than more than 6.4 L (1.7 gal) of contaminated water containing a Co-60 concentration of 0.06 µCi/ml would need to be ingested to reach the oral ingestion ALI; and 0.13 L (x gal) ingested to exceed the public dose limit of 100 mrem.

30 The licensee states in ABJRF DP Section 14.1 that external exposures from a spill would be far less than the current dose measured dose-rate in an accident scenario because the radioactivity concentration in the spill would be diluted over a large area. The NRC staff notes that the activity concentration in the liquid would remain the same (e.g., 0.06 µCi/ml) over the area, but the dose rate would decrease as the activity spread over the area of the spill. The licensee states in DP Section 14.1 that a more reasonable accident scenario may result in the ingestion of several milliliters of contaminated water and exposure to the material for an 8-hour period, which the licensee calculates to be 40.2 mrem. The licensee concludes that the resulting dose in an accident involving the release of contaminated liquids would be much less than one rem to off-site receptors and 25 mrem to on-site workers. The licensee concludes that safety management during operations, such as standard engineering and administrative controls, are sufficient for protection against spills.

3.9.2 Release of Airborne Contamination The licensee states in ABJRF DP Section 14.2 that an uncontrolled release of airborne radioactivity could occur during cutting and demolition activities involving contaminated or activated materials. The licensee states that these activities may take place inside temporary containment structures equipped with local HEPA filter ventilation systems or using localized ventilation ducts. The licensee states that failure of the containment structure or ventilation system could result in the release of airborne radioactive materials into the ABJRF. The licensee states that the HEPA air filter system would prevent release of airborne contaminants to the environment if D&D operations maintain negative pressure in the AJBRF at the time of such an incident. The licensee states that D&D activities could release airborne radioactive material to other areas of the Omaha VAMC or to the environment if the ventilation system in the ABJRF is not operating at the time of such an incident. The licensee commits to using alarming CAMs in the work areas to warn against the release of airborne radioactivity.

The licensee states in ABJRF DP Section 14.2 that Eu-152 has the most-limiting inhalation ALl of the contaminants of concern and the DAC for Eu-152 is 1 E-8 µCi/mL that results in 1 ALl or 5 rem to the exposed individual in a 2,000-hour work year. The licensee states that it does not know the Eu-152 concentration in the reactor tank Gunite or concrete at this time, but assumes a concentration of 10 pCi/g to calculate exposure based on the DAC. The licensee states it estimates that breathing air would be limited to a respirable particulate loading of 1.0 milligram (mg) of Gunite or concrete/ml of air before the Eu-152 concentration would exceed the DAC if the licensee assumes that concrete dust at its worst-case concentration becomes airborne after an uncontrolled release. The licensee calculates that approximately 28 kilograms ((kg) 61.7 lb) of contaminated material would have to become airborne to reach the DAC amount. The licensee states that it based its calculation assuming that the interior free volume of the reactor room is approximately 28,000 L (1,000 cubic ft (ft3)).10 Further, the licensee concludes that an accidental release as described above for a short duration would have minimal consequence on public dose because the effluent release to air control public dose is limited to 3 E-11 µCi/mL based on the publics exposure for one year.

10 The licensees calculation is correct for a room with a volume of 28,000 L, but incorrect for a room with a volume of 10,000 ft3. The licensee incorrectly converted 10,000 ft3 to 28,000 L (instead of 283,000 L) and used 28,000 L for the volume in the calculation. The licensee used the conversion factor of 2.8 L/ft3 rather than 28.3 L/ft3. As a consequence, the amount of contaminated material needed in a 10,000 ft3 room to exceed the DAC calculated by the NRC staff using the correct conversion factor of 28.3 L/ft3 is 283 kg, which is an order of magnitude greater than the licensees estimate.

31 3.9.3 Hot Particles The licensee states in ABJRF DP Section 14.3 that it does not expect to the find hot particles during D&D operations because the VA has never had any failed fuel within its facility. The licensee states that previous site characterizations did not identify hot particles at the AJBRF.

However, the licensee states that the DOC will conduct daily, weekly, and monthly surveillance surveys and provide continuous coverage by a health physics technician on all ongoing job activities. In addition, the licensee commits to having engineering controls such as, the constant recirculation of the reactor water, HEPA vacuum cleaners, and general radiation monitors, to avoid and identify any incidents involving hot particles during the D&D project.

3.9.4 Transportation Accidents The licensee states in ABJRF DP Section 14.4 that the DOC will ship various forms and quantities of radioactive waste from the ABJRF during the D&D project. The licensee states that the dose consequence from transportation accidents could be higher than the contamination accident scenarios described previously in SER Section 3.9 because high-activity reactor components could be involved. As such, the licensee states that there is a potential for a moderate dose consequence of between 1 and 25 mrem for the public following a transportation accident. However, the licensee concludes that adherence to NRC and DOT radioactive material packaging and transportation requirements is a sufficient control measure for mitigating transportation-related incidents.

3.9.5 Conclusion The NRC staff reviewed the postulated accidents identified in the ABJRF DP and determined that they are reasonable and reflect the accidents that could be expected during decommissioning activities. Because the licensee has removed and shipped off-site the reactor fuel, the radiological consequences of these postulated accidents are significantly less than those accidents occurring during reactor operations. The licensee considered accidental exposures from fire, radiological spill of pool water or waste systems liquids, release of airborne radioactivity during cutting and demolition, failure of air filtration systems, encountering hot particles, and transportation accidents. The licensee concluded that the potential dose from transportation accidents could be higher than for the other postulated accidents, but that the licensee will reduce this potential by following NRC and DOT radioactive material packaging and transportation requirements. The NRC staff agrees with this conclusion. The NRC staff reviewed the postulated radiological accident scenarios and the DPs risk mitigation plans, and concludes that the DPs radiological accident analysis is complete and that the mitigation plans are sufficient to protect the public health and safety and are, therefore, acceptable.

3.10 Proposed Final Status Survey Plan The licensee states in ABJRF DP Section 13 that the purpose of the FSS will be to document that the licensee has decommissioned the AJBRF to the extent necessary to meet the unrestricted use release requirements as specified in 10 CFR 20.1402 "Radiological Criteria For Unrestricted Use." The licensee will develop the FSS based on applicable guidance contained in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM).

The licensee states that the DOC will develop and implement the FSSP to ensure the licensee follows the requirements delineated in the DP to allow the NRC to terminate the license. The licensee states that the DOC will design the FSSP based on the assumption that the remaining

32 activity after decommissioning is essentially normally distributed relative to both the interior and exterior areas of the AJBRF. The licensee states that the DOC will revise the final survey implementation procedures to accommodate the survey findings if the results of the survey demonstrate that this assumption is not true. The licensee states that the VA will submit the FSSP for review and approval separately from the DP. The licensee states that the VA will finalize the FSSP after it submits the DP and after the DOC has started D&D activities because the licensee does not know the physical condition of the site at the time of DP submittal; the VA may want to propose alternative release criteria that would affect the implementation of the FSS.

3.10.1 Identification and Classification of Survey Units In DP Section 13.2, the licensee states that the FSSP will identify survey units that are classified based on contamination potential according to the methods described in the MARSSIM. The licensee states that the VA will subdivide the final survey classification for the AJBRF into three distinct classes of impacted areas: Class 1, Class 2, or Class 3 designations. Class 1 areas have the greatest potential for radioactive contamination in excess of the screening criteria.

Class 2 areas have a potential for radioactive contamination, but the VA does not expect the contamination to exceed the screening criteria. The VA does not expect any residual radioactivity in Class 3 areas, or expects any contamination present to a small fraction of the screening criteria.

The licensee states that Class 1 areas will receive the highest degree of survey efforts during the final survey receiving a 100 percent surface scan along with direct surface measurements or material samples at locations the DOC determines using the MARSSIM methods. Class 2 areas receive scan surveys between 10 and 100 percent of the surface and direct measurements. The licensee and DOC will use judgmental scans and direct measurements on Class 3 areas. The licensee described each class areas it anticipates in DP Table 13.1, AJBRF MARSSIM Classifications for FSS. The licensee has described the reactor tank wall and pit, the reactor water-cooling system vault, and the floors in Rooms B526, B540, and B540A as Class 1.

The walls less than 2 meters in Rooms B526, B540, B540A, B533A, B533AA, B537, and B535A and the floors in hall outside Room B526 and the stairs on the south side of Room B526 are Class 2. The walls higher than 2 meters and the ceilings in Rooms B526, B540, B540A, B533A, B533AA, B537, and B535A and the floors in Rooms B522 and B522A are Class 3 and the licensee expects all outside areas to be non-impacted.

3.10.2 Survey and Sampling Approach The licensee states in DP Section 13.3.2 that the DOC will establish instrument background for the specific types of surfaces that the DOC will survey. The DOC will take background measurements in non-impacted areas of the Omaha VAMC.

DP Section 13.3.1 states that the DOC will collect and analyze swipe samples and other material samples in accordance with a written procedure prepared by the DOC. The DOC will use rigorous chain-of-custody practices to ensure control and reproducibility of samples. The licensee states in DP Section 13.3.3 that the DOC will estimate the sample size for each survey area using the guidance in NUREG-1575 Table 5.3. However, the licensee states that based on the judgment of the DOC, the VA RSO, and the ACOS/R, the DOC may add a larger number of samples to survey units on a case-by-case basis.

3.10.3 Radiation Survey Instrumentation

33 DP Section 13.3.1 states that the radiation monitoring instrumentation that the DOC will use in the FSS will have the necessary sensitivities and capabilities to detect the radiation of interest and at levels below the screening criteria. The licensee states that the DOC will calibrate, operate, and maintain the instrumentation in accordance with written procedures.

3.10.4 Data Assessment and Evaluation The licensee states in DP Section 13.3.4 that quality control methods ensure the quality and accuracy of survey data and that these methods apply to field and laboratory instrumentation, sample collection, sample analysis, use of radioactive reference sources, and data processing.

The licensee states that it will implement quality control through the QAPP or FSSP. The licensee states that the DOC will include the scope of the FSS in the QAPP or include a strong quality assurance/quality control (QA/Quality Control) section in the FSSP. The licensee commits to having the DOC establish controls for activities affecting quality to ensure that DOC meets the prerequisites for the given activity, such as the appropriate equipment or the required environmental conditions.

3.10.5 Data Quality Objectives (DQO)

DP Section 13.1 describes the DQO process, which will consist of a series of planning stages that use a graded approach to safeguard the integrity of the radiological data used in decision making. The licensee states that this procedure ensures that the types, quantity, and quality of radiological data used in decision-making are appropriate for the intended application. The licensee states that the DQOs are qualitative and quantitative statements that clarify the process objective, determine the most appropriate locations and conditions for collecting data, and specify acceptable levels of decision errors that the licensee will use to determine the quantity and quality of data needed to support its decision. The licensee states it plans to use DQOs for the following:

Selection and independent verification of survey unit classification, Collection of sufficient high quality data to validate the release criteria for each survey unit to determine if residual radioactivity in each unit has been reduced to a level below the release criteria, Ensuring that the potential radiological risk from the AJBRF is below that represented by the dose limit release criteria.

3.10.6 The Final Status Survey Report The licensee states in DP Section 13.4 that the VA will develop a FSS Report when it completes the FSS and submit the report to the NRC. The licensee commits that the report will contain information needed by the NRC to make a decision on termination of the AJBRF license and authorization of the site for unrestricted use. The licensee commits to use applicable guidance from NUREG 1575 Section 8.6 to prepare the report.

3.10.7 Conclusion The NRC staff has reviewed the licensee DP and proposed FSSP and concludes that the licensee satisfies 10 CFR 50.82(b)(4)(iii), which requires a description of the planned final radiation survey. The licensees description of the FSS is adequate because the licensee commits to follow the NRC guidance in MARSSIM in the planning and conduct of FSS.

34 3.11 Technical Specifications The proposed amendment contains no changes to the Technical Specifications since Amendment 11, approved by letter dated August 5, 2002, (ADAMS Package No.

ML020630216) modified the Technical Specifications to identify when they are no longer applicable once the facility is shutdown. Thus, the operational Technical Specifications are also sufficient for the permanently shut down facility, and no changes are needed.

3.12 Physical Security Plan The VA states in DP Sections 2.2 and 11.5 that the AJBRF was permanently shut down on November 5, 2001 (ADAMS Accession No. ML020150541), defueled in June 2002, and the reactor fuel and fission chambers were transferred and removed to the United States Geological Survey TRIGA reactor facility in Denver, Colorado. As there is no longer any special nuclear material at the facility, a 10 CFR Part 73 physical security plan is no longer required.

The VA states in DP Section 1.5 that all loose radioactive materials were secured during pre-characterization activities performed in May 2011 (ADAMS Accession No. ML11255A334).

Section 7 of the DP states that during active decommissioning, an area inside the AJBRF will be utilized to package and store radiological waste and that VA hospital police will provide security.

Additionally, by letter dated November 12. 2014, as a revision to the DP, (ADAMS Accession No. ML14335A597) the VA indicates that storage and control of radiological material will be in accordance with 10 CFR Part 20, subpart I.

3.12.1 Conclusion The NRC staff reviewed the DP with respect to the licensees proposed physical security plan provisions to be in place during decommissioning. The staff concludes that these provisions satisfy the requirement of 10 CFR 50.82(b)(4)(v) regarding physical security plan provisions since the licensees commitment to facility security is consistent with regulations and is adequate for protection of facility radiological material.

3.13 Estimated Cost The VA states in DP Section 15 that the decommissioning cost estimate is $1,364,328 and summarizes the costs in DP Table 15.1, Decommissioning Project Cost Estimate. The estimate includes planning and management ($163,688), shipping and disposing of waste ($569,384),

and site decommissioning and FSS ($358,390). The VA includes a contingency of 25 percent

($272,866) in the decommissioning cost estimate as recommended in NUREG-1757, Vol. 3, Rev. 1, Section A.3.1, Preparing the Site-Specific Cost Estimate, to ensure that sufficient funds are available to cover costs that may result from unanticipated conditions or unforeseeable elements in the project scope. The NRC staff finds that unanticipated conditions or unforeseeable elements may include factors, such as waste disposal rates or waste volume increases from undiscovered contamination. In addition, the time interval between the development of the DP and the start of decommissioning activities can influence the costs associated with changes in the economy and regulatory requirements.

In accordance with 10 CFR 50.75(e)(1)(iv), the VA is a federal institution and as such, DP Section 15 states the VA has requested and received funds consistent with a more than

$2,000,000 project. The VA states that the licensee will develop and submit all budget and

35 legislative requests necessary to ensure that required funds for decommissioning of the AJBRF will occur should additional funds above and beyond those designated for the decommissioning be required.

3.13.1 Conclusion The NRC staff has reviewed the licensees decommissioning cost estimate and funding availability statement in its DP. The NRC staff concludes that the provision of this information satisfies the requirements of 10 CFR 50.82(b)(4)(iv) because the licensee has provided an updated cost estimate for the chosen alternative for decommissioning; and the licensee has stated that it has received full funding for this estimate.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment associated with this DP involves changes in the installation within the restricted area as defined in 10 CFR Part 20. Based on the staff review of the DP discussed above, (see sections 3.7 Radiation Safety and Health Program, 3.8 Radioactive Waste Management and 3.9 Radiological Accident Analysis), 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 NRC staff has reviewed the licensees proposed actions to decontaminate, dismantle, and dispose of component parts of the AJBRF, and to perform an 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 these levels, then the NRC will terminate the Department of Veterans Affairs Alan J. Blotcky Reactor Facility License No. R-57.

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 VA will dismantle the reactor and dispose of all significant reactor-related radioactive materials 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 AJBRF 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 that the personnel training program that the VA proposes is acceptable because its scope covers all aspects of the decommissioning activities that need to

36 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 AJBRF 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 AJB reactor. 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.

The NRC staff concludes that the licensee has provided, in accordance with 10 CFR 50.82(b)(4)(v), an acceptable description of the technical specifications, quality assurance provisions, and physical security plan provisions to be in place during decommissioning. Therefore, the NRC staff concludes that the licensees DP meets the requirements of 10 CFR 50.82(b)(4).

Thus, the NRC staff concludes that the DP demonstrates that the decommissioning will be performed in accordance with NRC regulations and will not be inimical to the common defense and security or the health and safety of the public. Additionally, as previously stated, the NRC has previously published notice of the DP to interested persons and solicited comments.

Therefore, in accordance with 10 CFR 50.82(b)(5), the NRC approves the DP by amendment, subject to the criteria in section 2.D of the AJBRF license.

Principal Contributors: Theodore B. Smith, NMSS Stephen Giebel, NMSS Tanya Palmateer Oxenberg, RES (formerly FSME)

Date: November 12, 2014