ML21012A274
| ML21012A274 | |
| Person / Time | |
|---|---|
| Site: | General Atomics |
| Issue date: | 12/31/2020 |
| From: | General Atomics |
| To: | Marlayna Vaaler Doell Reactor Decommissioning Branch |
| Doell M | |
| Shared Package | |
| ML21012A268 | List: |
| References | |
| 38/67-5040 | |
| Download: ML21012A274 (62) | |
Text
FINAL STATUS SURVEY REPORT FOR THE GENERAL ATOMICS TRIGA REACTOR FACILITY USNRC Facility Operating Licenses R-38 and R-67 Prepared by General Atomics Prepared for the U.S. Nuclear Regulatory Commission December 2020
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page iii of viii TABLE OF CONTENTS
1.0 INTRODUCTION
.................................................................................................................. 1 2.0 HISTORICAL SITE ASSESSMENT..................................................................................... 2 2.1 SITE DESCRIPTION............................................................................................................................. 2 2.2 SITE HISTORY.................................................................................................................................... 4 2.3 PREVIOUSLY RELEASED AREAS........................................................................................................ 7 2.4 MARK I DECOMMISSIONING HISTORY............................................................................................... 9 2.5 MARK F DECOMMISSIONING HISTORY.............................................................................................. 9 2.6 CHARACTERIZATION & REMEDIATION OF REACTOR PITS................................................................. 9 3.0 DESIGN OF THE FINAL STATUS SURVEY................................................................... 13
3.1 BACKGROUND
DETERMINATION FOR SURFACES AND STRUCTURES................................................ 14
3.2 BACKGROUND
DETERMINATION FOR LAND AREAS........................................................................ 14 3.3 DATA QUALITY OBJECTIVES (DQOS)............................................................................................. 14 3.4 AREA CLASSIFICATIONS.................................................................................................................. 14 3.5 NON-IMPACTED AREAS................................................................................................................... 14 3.6 IMPACTED AREAS............................................................................................................................ 15 3.7 CLASS 1 AREA................................................................................................................................. 15 3.8 CLASS 2 AREA................................................................................................................................. 15 3.9 CLASS 3 AREA................................................................................................................................. 15 3.10 SURVEY UNITS (SUS)...................................................................................................................... 15 3.11 SCANNING SURVEYS....................................................................................................................... 16 3.12 TOTAL SURFACE ACTIVITY MEASUREMENTS.................................................................................. 17 3.13 DETERMINING THE NUMBER OF SAMPLES....................................................................................... 17 3.14 DETERMINATION OF THE RELATIVE SHIFT...................................................................................... 17 3.15 DETERMINATION OF ACCEPTABLE DECISION ERRORS.................................................................... 17 3.16 DETERMINATION OF NUMBER OF DATA POINTS (SIGN TEST).......................................................... 17 3.17 DETERMINATION OF SAMPLE LOCATIONS....................................................................................... 18 3.18 REMOVABLE CONTAMINATION SURVEYS....................................................................................... 19 3.19 SURVEYS OF BUILDING MECHANICAL SYSTEM INTERNALS............................................................ 19 3.20 VENTILATION SYSTEMS.................................................................................................................. 19 3.21 TRENCHES AND DRAINS.................................................................................................................. 19 3.22 BURIED PIPE.................................................................................................................................... 20 3.23 SUB-CRITICAL PIT........................................................................................................................... 20 3.24 SURVEY ACTION LEVELS................................................................................................................ 20 4.0 INSTRUMENTATION........................................................................................................ 21 4.1 INSTRUMENT CALIBRATION............................................................................................................ 21 4.2 INSTRUMENT EFFICIENCY DETERMINATION.................................................................................... 21 4.3 FUNCTIONAL CHECKS..................................................................................................................... 21 4.4 DETERMINATION OF COUNTING TIMES AND MINIMUM DETECTABLE CONCENTRATIONS............... 22 4.5 COUNTING UNCERTAINTY, ERROR PROPAGATION, AND CONFIDENCE INTERVAL........................... 25 4.6 INSTRUMENTATION SPECIFICATIONS............................................................................................... 25 5.0 SURVEY UNIT DOCUMENTATION, VALIDATION, AND DESCRIPTIONS............. 27 5.1 SURVEY DOCUMENTATION............................................................................................................. 27 5.2 DATA VALIDATION......................................................................................................................... 27 5.3 SURVEY UNIT OVERVIEW............................................................................................................... 27 6.0 FINAL STATUS SURVEY RESULTS............................................................................... 35 6.1 MARK I AND MARK F REACTOR PITS.............................................................................................. 35 6.2 SURFACES AND STRUCTURES.......................................................................................................... 36 6.3 OPEN LAND/SOIL AREAS................................................................................................................ 37
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page iv of viii 7.0 BURIED PIPE DOSE ASSESSMENT................................................................................ 40
7.1 BACKGROUND
................................................................................................................................. 40 7.2 DETERMINATION OF SOURCE TERM................................................................................................ 42 7.3 DOSE ASSESSMENT METHODOLOGY............................................................................................... 44 7.4 RESULTS......................................................................................................................................... 46
7.5 CONCLUSION
................................................................................................................................... 46 8.0 DATA QUALITY ASSESSMENT & INTERPRETATION OF SURVEY RESULTS...... 48 8.1 PRELIMINARY DATA REVIEW.......................................................................................................... 48 8.2 DETERMINING COMPLIANCE FOR SURFACES AND STRUCTURES SURVEYS...................................... 48 8.3 VERIFICATION OF REQUIRED NUMBER OF SAMPLES FOR SURFACES AND STRUCTURES.................. 48 8.4 DETERMINING COMPLIANCE FOR OPEN LAND/SOIL AREA SURVEY UNITS..................................... 49 8.5 VERIFICATION OF REQUIRED NUMBER OF SAMPLES FOR OPEN LAND/SOIL AREAS........................ 49 9.0 WASTE MANAGEMENT................................................................................................... 50 9.1 WASTE TYPES, VOLUMES, AND ACTIVITY...................................................................................... 50 9.2 CLASSIFICATION.............................................................................................................................. 51 9.3 STORAGE & TRANSPORTATION....................................................................................................... 52 10.0 EXECUTIVE
SUMMARY
.................................................................................................. 53
11.0 REFERENCES
..................................................................................................................... 54 LIST OF ATTACHMENTS ATTACHMENT A - SURVEY UNIT MAP ATTACHMENT B - MAPS OF SURVEY UNITS ATTACHMENT C - INSTRUMENT CALIBRATION CERTIFICATES ATTACHMENT D - SITE SPECIFIC BACKGROUND MEASUREMENTS ATTACHMENT E - MDC CALCULATIONS ATTACHMENT F - EXAMPLE SURVEY UNIT PACKAGE INSTRUCTIONS ATTACHMENT G - GAMMA SPECTROSCOPY RESULTS ATTACHMENT H - STATICS, WIPES AND EXPOSURE RATE RESULTS ATTACHMENT I - TESTAMERICA RESULTS ATTACHMENT J - BURIED PIPE MICROSHIELD OUTPUT FILES ATTACHMENT K - BURIED PIPE RESRAD AND MICROSHIELD CASE SCENARIOS ATTACHMENT L - VOLUMETRIC SAMPLE RESULTS
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page v of viii LIST OF FIGURES FIGURE 1: GENERAL ATOMICS AERIAL VIEW 3
FIGURE 2: TRIGA REACTOR FACILITY 4
FIGURE 3: MARK I REACTOR CORE 6
FIGURE 4: NON-NRR AREAS RELEASED IN 2007 8
FIGURE 5: MICRO-PILE SYSTEM 11 FIGURE 6: SCREW JACKS 11 FIGURE 7: MARK F REACTOR 11 FIGURE 8: BARE STEEL LINER 12 FIGURE 9: REACTOR PIT WALL 12 FIGURE 10: MARK I MAPPING METHODOLOGY 28 FIGURE 11: MARK F MAPPING METHODOLOGY 29 FIGURE 12: BROKEN PIPE 41 FIGURE 13: PIPE OPENING 41 FIGURE 14: PIPE CONFIGURATION 42
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page vi of viii LIST OF TABLES TABLE 3.1: DEFAULT SURFACES & STRUCTURES SCREENING VALUES 13 TABLE 3.2: DEFAULT CONCRETE & SOIL SCREENING VALUES 13 TABLE 3.3: SURVEY UNIT DESCRIPTIONS 16 TABLE 3.4: SCAN SURVEY COVERAGE BY CLASSIFICATION 16 TABLE 3.5: SURVEY SAMPLE PLACEMENT OVERVIEW 18 TABLE 3.6: SURVEY INVESTIGATION LEVELS 20 TABLE 4.1: INSTRUMENTS USED AT THE TRIGA FACILITY 26 TABLE 4.2: TYPICAL INSTRUMENT OPERATING PARAMETERS AND SENSITIVITIES 26 TABLE 6.1: SURVEY UNIT 1 SAMPLE ANALYSIS RESULTS 35 TABLE 6.2: SURVEY UNIT 2 SAMPLE ANALYSIS RESULTS 36 TABLE 6.3:
SUMMARY
OF SURVEY RESULTS FOR SURFACES AND STRUCTURES 37 TABLE 6.4: SURVEY UNIT 5A SAMPLE ANALYSIS RESULTS 38 TABLE 6.5: SURVEY UNIT 12 SAMPLE ANALYSIS RESULTS 39 TABLE 7.1: BURIED PIPE SURVEY RESULTS 47 TABLE 9.1: NNSS WASTE SHIPMENTS 50
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page vii of viii ACRONYM LIST ALARA ANSI BOP As Low As Reasonably Achievable American National Standards Institute Balance of Plant Cs-137 CsI CFR Cesium-137 Cesium Iodide Code of Federal Regulations cm Co-60 cpm D&D centimeter Cobalt-60 counts per minute Decontamination and Decommissioning DCGLEMC Derived Concentration Guideline Level - Elevated Measurement Comparison DCGLW Derived Concentration Guideline Level - Wilcoxon Rank Sum DOD DOE dpm DQO Department of Defense Department of Energy disintegrations per minute Data Quality Objective DSV Eu-152 Eu-154 FLAIR Default Screening Value Europium-152 Europium-154 Flashing Advanced Irradiation Reactor FSS FSSP GA HEPA HSA Final Status Survey Final Status Survey Plan General Atomics High-Efficiency Particulate Air Historical Site Assessment HTGR HVAC IP-1 ISO LBGR High Temperature Gas-Cooled Reactor Heating, Ventilation, Air Conditioning Industrial Package Type 1 International Organization for Standardization Lower Bound of the Gray Region m
meter MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC MDCR Mark F Mark I Mark III MLLW Mrem NaI NIST Minimum Detectable Concentration Minimum Detectable Concentration Rate TRIGA Mark F Reactor TRIGA Mark I Reactor TRIGA Mark III Reactor Mixed Low-Level Waste millirem Sodium Iodide National Institute of Standards & Technology NNSS NMSS Nevada National Security Site Nuclear Materials Safety and Safeguards NRC U.S. Nuclear Regulatory Commission NRR NUREG ORAU pCi/g POL Nuclear Reactor Regulation Nuclear Regulatory Commission Guidance Document Oak Ridge Associated Universities picocuries per gram Possession Only License
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page viii of viii RESRAD R&D Computer Program Developed at Argonne National Laboratory Research and Development SR Sr-90 SU Scientific Research Strontium-90 Survey Unit TA TEDE TKF TRF TRIGA
µR USAEC UZrH XRF yr.
TestAmerica Total Effective Dose Equivalent TRIGA King Furnace TRIGA Reactor Facility Training, Research, Isotope Production, General Atomics Micro Roentgen United States Atomic Energy Commission Uranium Zirconium Hydride X-ray Fluorescence year
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 1 of 54 1.0 Introduction This report presents the results of the Final Status Survey (FSS) of the unreleased portion of the General Atomics (GA) TRIGA Reactor Facility (TRF) and associated land area located on the GA Torrey Pines campus in San Diego, California. The TRF housed three TRIGA reactors: the Mark I, Mark F, and Mark III which operated under Nuclear Regulatory Commission (NRC) licenses R-37, R-68, and R-100 respectively. This report only addresses the Mark I and Mark F Reactors. The Mark III Reactor license was terminated in 1975. Rooms 109-115, adjacent to and including the Mark III Reactor, and associated yard areas were previously decommissioned and released in 2007[1].
The final status surveys were designed in accordance with NUREG 1575[2], Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) and NUREG 1757[3], Consolidated Decommissioning Guidance and used to demonstrate compliance with the release criteria in 10CFR20 Subpart E. Surveys were performed from August 2019 to May 2020 and the impacted areas were divided into 16 survey units. A total of 742 static measurements and 126 volumetric samples were collected and analyzed. Wipe surveys to assess fixed and removable contamination levels as well as exposure rate measurements were performed in every survey unit. Volumetric samples were collected in the Mark I and Mark F reactor pits, the Mark F storage canal, Mark F Excavation Area and the surrounding land area. On-site sample analysis using gamma spectroscopy was performed to support decision making in the field.
Based on the results of the FSS, the remaining building structural surfaces, surface soils, and subsurface soils at the facility meet the requirements for unrestricted release specified in 10CFR20, Subpart E, Radiological Criteria for License Termination and are suitable to be released for unrestricted use.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 2 of 54 2.0 Historical Site Assessment The purpose of a historical site assessment (HSA) is to determine the current status of the site including potential, likely or known sources of radioactive contamination by gathering historical data from various documents and records. This data includes physical characteristics and location of the site as well as information found in site operating records, including radiological surveys. GA prepared the TRF Process Knowledge Report[4] in 1996 which presented all of the available historical data and process knowledge information pertaining to the TRF. This report identified the type, quantity, condition and location of radioactive and hazardous materials which are, or which may be, present as residual contamination. The information from the Process Knowledge Report was incorporated in the GA TRF Decommissioning Plan[5] which was approved by the NRC in January 2000.
Based on this HSA and operating characteristics of the Mark I and Mark F reactors, potential contaminates were identified as Eu-152, Eu-154 and Co-60 (activation products) and Cs-137 and Sr-90 (fission products). Activation products Eu-152 and Eu-154 have been found in the concrete biological shields of both reactors. The Mark F reactor was constructed with an extensive rebar matrix leading to the presence of Co-60 from iron activation. Fission products Cs-137 and Sr-90 were found in the east yard area adjoining the Mark F reactor room and in soil underneath the northeast portion of the Mark F reactor room floor. These radionuclides are thought to have originated in a resin bed ion exchange system located in the east yard. Accidental spillage during maintenance or resin replacement likely resulted in the deposition of Cs-137 and Sr-90 onto surfaces in the yard and into a trench extending from the Mark F pit to the outside yard. The piping for a liquid waste drain located in the trench at some point became severed due to soil settling and allowed contaminated ion exchange system water and/or resins to seep into the soil underneath the floor.
Since approval of the GA TRF Decommissioning Plan in January 2000, Decontamination
& Decommissioning (D&D) has proceeded for the Mark I and Mark F reactors as well as the surrounding land area, auxiliary laboratories and other facilities within the TRF.
The following sections provide an overview and history of the GA site and Mark I and Mark F reactors as well as a summary of the decommissioning activities that have taken place from reactor shutdown through the most recent Mark I and Mark F reactor pit characterization and remediation efforts.
2.1 Site Description The TRF is located on the GA main site at 3550 General Atomics Court located in the City and County of San Diego, State of California. This 60-acre GA site is roughly 20 minutes north of downtown San Diego along Interstate 5 and 1 mile east of the ocean. It is situated 300 ft. above sea level on Torrey Pines Mesa and surrounded by the University of California to the south, various medical facilities, Torrey Pines Golf Course to the west and several science parks (See Figure 1).
The land upon which GA is built is designated for scientific research by the City of San Diego as is much of the surrounding area. The GA facilities contain nearly one million square feet of office space, including engineering, sophisticated test
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 3 of 54 facilities, precision manufacturing installations, and advanced technology laboratories.
The TRF is located on the northern edge of the GA property on the Torrey Pines Mesa. In prior years there was a hot cell facility located in the vacant lot adjacent to the reactor building where material could easily be transported for examination by remote handling in the cells. The hot cell has since been decommissioned and the land released for unrestricted use in 2000. The reactor building shown in Figure 2 housed the TRIGA Mark I and Mark F reactors. The TRIGA Mark III reactor was shut down in 1973 and has been decommissioned. This reactor, along with several laboratories, was located in the TRF just north of the Mark F reactor control room.
Figure 1: General Atomics Aerial View TRIGA Reactor Facility
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 4 of 54 Figure 2: TRIGA Reactor Facility 2.2 Site History The GA property on the Torrey Pines Mesa was acquired in 1956 from the City of San Diego by the General Dynamics Corporation with the expressed purpose to establish the John J. Hopkins Laboratory for Pure & Applied Science, later named the General Atomic Division of the General Dynamics Corporation. The property consisted of roughly 290 acres. Today, GA is privately owned and occupies a smaller footprint on the Torrey Pines Mesa but retains many of the original buildings dating from the late 1950s. This includes the TRF which was one of the first buildings constructed. Other early programs developed by GA include controlled fusion research and a High Temperature Gas-Cooled Reactor (HTGR).
The fusion research program went on to become the largest and most successful fusion program in industry. The DIII-D fusion facility has been operating for over 40 years at the GA Sorrento Valley site, a 60-acre land area which adjoins the Torrey Pines site by an access road. Since its inception, GA has gone on to expand to other areas including electromagnetic systems, remotely operated aircraft, laser and radar technologies, biotechnology and many other advanced high technology areas under Department of Defense (DOD), Department of Energy (DOE), and commercial sponsorship. GA has become a major center of diversified energy research and development, engineering, fabrication and testing with extensive experience in performing innovative Research & Development (R&D) and transforming conceptual results into practical systems. In its 65 years of operation, GA has gone on to expand worldwide but to this day retains the company headquarters on the original Torrey Pines Mesa property.
One of the first goals of the newly-established General Atomic Division of General Dynamics was the development of a new family of inherently safe small nuclear reactors which could be used in both industrial and academic applications for training, research and isotope production. This was less than five years from Mark I Mark III Mark F
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 5 of 54 the date of President Eisenhowers December 1953 Atoms for Peace proposal to the United Nations General Assembly where he sought to transform the atom into a benefit for mankind. In that short amount of time the GA TRIGA reactor was first conceived, built and operated on the Torrey Pines Mesa and went on to become the most widely used research reactor in the world with 66 facilities built in 24 countries on five continents. The initial Mark I reactor, which formed the basis for the subsequent worldwide facilities, and two other reactors, the Mark F and Mark III, were constructed between 1957 and 1966 in what is now known as GA Building G21 (or the TRF). This report addresses the decommissioning and final status survey of two of those reactors, the Mark I and Mark F.
The third reactor, the Mark III, has been previously decommissioned and NRC License R-100 terminated in 1975. The last prototype built of the three, the Mark III was a 2MW steady-state reactor which served for several years in the late 1960s and early 1970s as a test bed for thermionic fuel cell development before being shut down and secured in 1973. In addition to the TRIGA Mark III reactor, other portions of GA Building 21 and surrounding land area were previously released by the US Nuclear Regulatory Commission (NRC). These non-reactor areas were under the jurisdiction of GAs SNM-696 license issued by the NRCs Office of Nuclear Materials Safety and Safeguards (NMSS). This license has since been terminated.
2.2.1 Mark I History As part of the early nuclear reactor development efforts, GA initiated plans to design, build, and operate a prototype reactor unit on the companys main site on Torrey Pines Mesa. To this end, in late 1957, GA requested and obtained a Construction Permit and Utilization Facility License from the U.S. Atomic Energy Commission (USAEC) to authorize this activity. Immediately thereafter, GA proceeded with construction of the Isotope Reactor Building, later named the TRIGA Reactor Facility, to house the prototype TRIGA Reactor and supporting systems. Following building construction and reactor hardware installation, the prototype TRIGA Reactor was brought to initial criticality on May 3, 1958. The initial charter of the prototype was to prove the inherent operational safety of the Uranium-Zirconium-Hydride (UZrH) TRIGA fuel matrix that became the foundation for all subsequent TRIGA reactors. The Mark I became particularly useful in developing and demonstrating neutron activation analysis as an extremely sensitive technique for the detection of trace amounts of impurities in a variety of sample materials. In 1960, the TRIGA King Furnace (TKF) began being used to measure in-pile behavior, including fission gas release, of various types of nuclear fuel. The TKF aluminum containment was adapted to fit an in-core TRIGA fuel element position, thus allowing exposure of the test fuel to a variable neutron flux while maintaining an elevated temperature by electrically heating a graphite tube. Other applications over the years included semiconductor irradiation, transient experiments, isotope production, and reactor operator training. A pneumatic transfer system and rotary specimen rack were other irradiation facilities available for use with the Mark I.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 6 of 54 Continuously operational from 1958 until late 1997, the prototype TRIGA Reactor was later designated as the Torrey Pines TRIGA Reactor, and later yet, as the TRIGA Mark I Reactor. The maximum steady-state reactor power level for the Mark I reactor was originally limited to 10 kW(t) but was modified roughly two years later to allow for routine steady-state operations up to 250 kW(t). Because of its inherent safety features, the TRIGA Mark I could be pulsed to power levels of over 1000 MW, after which it returned in a few thousandths of a second to a low power level. This pulsing feature of UZrH-fueled reactors, first demonstrated with the prototype GA Mark I, is a standard feature today among many TRIGA reactors.
The initial Mark I reactor at GA shown in Figure 3 was designated by the American Nuclear Society in 1986 as a Nuclear Historic Landmark. The citation highlighted its role in pioneering the use of unique, inherently safe capabilities in nuclear reactors. At GAs request, the NRC issued an amendment to the TRIGA Mark I utilization facility license on October 29, 1997, which placed the reactor in Possession-Only-License (POL) status. Shortly thereafter all irradiated TRIGA fuel in the Mark I core was transferred to the secured Mark F fuel storage canal and consolidated with all other irradiated fuel from the Mark F and Mark III cores. Later amendments to the POL allowed decontamination and decommissioning activities to proceed with the goal of release of the building and land area to unrestricted use.
2.2.2 Mark F History In March 1960, GA submitted an application to the USAEC requesting a Construction Permit and Utilization Facility License for the Flashing Advanced Irradiation Reactor (FLAIR). These documents were issued to GA by the USAEC and Building 21 was modified by the addition of Rooms21-107 and 21-108 to house the FLAIR Reactor and Reactor Instrumentation & Control Systems, respectively. The reactor was built expressly to utilize the pulsing features of the TRIGA design and to demonstrate the behavior of UZrH fuels when pulsed to power levels even above 5000 MW. It was designed to provide controlled, instantaneous pulses of intense neutron and gamma radiation for studies where high neutron flux and narrow pulse width were required. This reactor was brought to initial criticality on July 2, 1960 and was continuously maintained and operated by GA from that time until March 22, 1995. One of the more notable and long-term experiments was for an in-pile thermionic power conversion project which included round-the-clock operations from 1987 to 1993. There were also earlier efforts focused on direct power conversion experiments as well as TKF experiments and in-core testing of new fuel designs. During 1973 to Figure 3 - TRIGA Mark I Reactor Core
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 7 of 54 1974, Mark III fuel was installed in the Mark F core to provide a smaller, higher power density configuration.
Upon final shutdown in 1995, the Utilization Facility License was amended at the request of GA to authorize POL activities that allowed decontamination and decommissioning activities to proceed with the goal of release of these areas to unrestricted use. During the operating period, the reactor installation was designated as the Advanced TRIGA Prototype Reactor and later referred to as the TRIGA Mark F Reactor. The maximum steady-state reactor power level for the Mark F reactor was limited to 1500 kW(t) from initial criticality to permanent shutdown. Upon permanent shutdown, the core was defueled and consolidated in the secured Mark F storage canal with irradiated fuel from The Mark I and Mark III cores. In 2010, all irradiated fuel elements from the three TRIGA reactors located on the Torrey Pines Mesa were shipped from storage in Building 21 to an authorized off-site storage facility at the Idaho National Laboratory.
2.3 Previously Released Areas A portion of the land area and buildings considered to be part of the TRF was previously released for unrestricted use by the NRC in 2007. At the time, the land and buildings were licensed under GAs SNM-696 license issued by the NRCs NMSS Office. The balance of the TRF was referred to as the Nuclear Reactor Regulation or NRR area which includes the Mark I and Mark F reactors. This Non-NRR area released for unrestricted use in 2007 included TRF Rooms 109-115 and portions of the land area. Figure 4 illustrates these previously released areas. The former Mark III reactor was located in Room 111. The land area released included the TRIGA front lot which was used following its release to stage radioactive waste for disposal. It is being included in the current final status survey due to the potential for contamination while being used for this purpose.
Similarly, Room 112 was released in 2007 as part of the Non-NRR portion of Building G21 but was subsequently used for storage of radioactive materials and is included in the current survey. Both the front lot and Room 112 have been given their own survey unit designations for the purposes of the final status survey of the TRF.
Non-impacted areas are areas without residual radioactivity from licensed activities, i.e., no radiological impact from site operations, and are not typically considered during final status surveys. Several rooms in the TRF previously released for unrestricted use in 2007 are being considered in the current survey as non-impacted areas but will be surveyed as a conservative measure to ensure the D&D efforts performed in other areas of the facility did not adversely affect these rooms. Rooms 109-111 and 113-115 have been collected into a single survey unit designated TRIGA Non-Impacted Areas for the purpose of this final survey.
These rooms have been used for storage of non-radioactive equipment and supplies since decommissioning activities were completed and post D&D surveys were conducted.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 8 of 54 Previously, Rooms 109 and 110 served as laboratories for fuel development and fission product release studies. There was a divider between these two rooms that has since been removed. There is a mezzanine above Room 110 that was not previously released and is included as a separate survey unit here. There is also a pipe trench located in the southern portion of Rooms 109-110 which was not previously released and is being considered here as part of a larger grouping of trenches within the TRF which has been assigned a unique survey unit.
Figure 4 - Non-NRR Areas Released in 2007
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 9 of 54 2.4 Mark I Decommissioning History Upon permanent shutdown of the Mark I reactor in 1997, all irradiated fuel elements were transferred to the fuel storage canal of the Mark F reactor pool where they were consolidated with the fuel from the Mark F and Mark III reactor cores. All activated and contaminated hardware, startup neutron sources and Balance of Plant (BOP) components were removed from the facility to a waste disposal site. This included equipment in Rooms 105 (soil lab), 106 (counting room) and 102 (Mark I reactor room) as well as equipment located in the cooling tower controlled yards and Mark I controlled yard immediately north of the Mark I reactor room. For the purpose of this final status survey, the cooling tower yards and Mark I controlled yard have been combined into a single survey unit forming the TRIGA Back Lot. A portion of the Mark I controlled yard previously contained a make-up water tank for the Mark I reactor and a contaminated liquid waste tank which served various laboratories in the TRF.
The entirety of these excavations are included in the TRIGA Back Lot survey unit although the eastern portion was released in 2007 as part of the Non-NRR land.
Following removal of equipment outside of the Mark I reactor proper, all items in the reactor pool were removed, the pool drained and the aluminum tank removed and sectioned into pieces for disposal.
2.5 Mark F Decommissioning History Upon permanent shutdown of the Mark F reactor in 1995, fuel elements in the core were moved to the adjoining storage canal to await final disposition and shipment to a DOE interim storage facility in Idaho. Prior to the shipment of all TRF fuel elements, all items in and around the reactor pool not deemed necessary for handling or surveillance of the fuel elements were removed and dispositioned.
This included items such as the reactor bridge, diffusion system and shroud, fission chambers, ion chambers, beam tubes, startup sources, core support structure and control rod drive system.
Decommissioning of the Mark F included the removal of asphalt from the Mark I controlled yard and the area between this yard and the Cooling Tower Controlled Yard. Upon completion of asphalt removal, cooling lines originating in the Mark F reactor room were removed from under the soil.
Following the shipment of all fuel elements to Idaho National Laboratory in September 2010, the removal of all highly radioactive neutron activated hardware remaining in the pool was planned. In May 2012, several reinforced Type A containers were loaded and grouted for shipment to the Nevada National Security Site (NNSS) for disposal. Subsequently, the reactor pool was drained and remaining hardware such as fuel storage racks and underwater barriers were removed and dispositioned.
2.6 Characterization & Remediation of Reactor Pits Following the fuel being transferred offsite to the Idaho National Laboratory, activated hardware transported to the NNSS for disposal, the reactor pools drained
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 10 of 54 and, in the case of the Mark I, the aluminum tank removed, characterization of the concrete biological shields commenced. For the Mark F this included the steel liner, Gunite layer and Epocast surface layer. For both reactors, these surfaces were inaccessible for characterization prior to the tanks being drained of water.
These characterization surveys serve to fill the gaps where there are data deficiencies between the HSA and Final Status Surveys.
2.6.1 Mark I After the aluminum tank was removed, the concrete biological shield and underlying soil were left as the only remaining sources of residual activation.
Several campaigns were undertaken to characterize the depth of activation in the concrete and soil following removal of the tank in CY 2000. Multiple cores were drilled with sizes ranging from 2.0 to 3.5 in diameter. Cores were taken from the floor and from the walls up to 2.5 meters (8.2 ft.) in height measured from the bottom of the Mark I reactor pit. All sample cores taken were through the biological shield and into the surrounding soil. Comparisons of activity levels measured through gamma spectroscopy were made with approved radiological release criteria to determine the required remediation. This was accomplished by sectioning the cores into disks and evaluating each disk for radionuclide content.
Using this data, activation profiles were developed which in turn evolved into an excavation plan. In the case of both reactors, the dominant isotope was Eu-152 with the default screening value (DSV) or release limit being 7.0 pCi/g[6]. Based on the core drilling results, it was determined that the entire bottom two meters of the right circular cylinder that forms the Mark I reactor pit would require removal, consistent with the fact the Mark I reactor was positioned in the center of the lower region of the pit and the neutron flux incident on the surrounding surfaces was uniform. In addition, the entirety of the floor of the pit would require removal as well as several inches of soil both underneath the floor and behind the lower cylinder walls. The top of the fueled region of the Mark I reactor was just under 1 meter (3.3 ft.) above the pit floor, confirming the excavation plan was consistent with the neutron flux profile of the reactor.
Because of the extent and geometry of the excavation, a micro-pile system was put in place prior to removal of the pit concrete to provide stability to the remaining upper concrete cylinder and to minimize soil sloughing along the walls of the reactor pit once the lower concrete biological shield had been removed.
The system consists of 24 micro-piles around the circumference of the Mark I pit down to a depth of 0.9 meters (3 ft.) below the existing reactor pit foundation. As an additional stability measure, four screw jacks were installed underneath the upper concrete cylinder once the bottom 2 meters (6.6 ft.) of concrete were removed. Figures 5 and 6 show respectively the micro-pile system and a screw jack braced underneath the pit wall.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 11 of 54 Figure 5: Micro-Pile System Figure 6: Screw Jacks 2.6.2 Mark F After the water was drained and the Mark F pit and fuel storage canal were rinsed, dried and vacuumed, several core samples were collected from selected locations for initial characterization studies in which both radiological and chemical analyses were performed. These initial studies were focused on the outer layers of material and not the underlying concrete biological shield to determine the extent of surface and near surface contamination by not only radioactive but potentially hazardous materials. Unlike the Mark I, the Mark F did not have an aluminum tank but rather had a 1/4 thick carbon steel liner covered with a 10 cm (4 inch) thick layer of steel mesh-reinforced Gunite which was in turn sealed with a thin coating of Epocast. These layers formed the barrier between the pool and the underlying concrete biological shield. The initial characterization studies showed that the Epocast coating contained not only low levels of radioactive contamination, but also levels of cadmium and lead that would classify the material as a mixed waste with both radiological and hazardous components.
Lead was commonly used as a ballast material for various experiments and other reactor installations and over time plated out on the Epocast layer of the Mark F pit. Similarly, cadmium was used in the Mark F as a strong thermal neutron absorber and acted as a filter to produce a harder neutron beam for experiments.
The cadmium and lead could have potentially been deposited on any region of the Epocast layer throughout the water filled reactor pit and fuel storage canal.
After the discovery of cadmium and lead, the Epocast and the top layer of Gunite were removed throughout the Mark F reactor pit and fuel storage canal. After removal, the remaining Gunite, steel liner and concrete biological shield remained (See Figure 7). Once the cadmium and lead were removed from the surfaces of the Mark F, core samples were taken to determine the extent of activation in the Gunite, steel liner, concrete biological shield Figure 7: Mark F Reactor
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 12 of 54 and surrounding soil. Samples were taken in the pit floor, canal region and pit walls up to a level of 5 meters (16.4 ft.) up from the floor. Unlike the Mark I, the Mark F pit was not only volumetrically larger, it was surrounded by a much thicker concrete biological shield in the lower regions of the pit where the reactor was positioned. The reactor core was also skewed to the north of the pit center, leading to asymmetric activation in the walls and floor. The core samples were analyzed using gamma spectroscopy and the results confirmed that the regions with higher activation were on the north side of the pit and activation did not extend into the soil anywhere surrounding the Mark F structure. Similar to the Mark I, the cores were sectioned into disks and analyzed for radionuclide content.
The length of the cores ranged from 0.3 to 1.8 meters (1 to 6 feet) depending upon the wall thickness at the sample location. There was no activation in excess of the approved radiological release criteria in the canal area or in any of the walls above 2 meters (6.6 ft.) in elevation. For perspective, the top of the fueled region of the core is at 1.12 meters (3.7 ft.) in elevation above the floor of the reactor pit.
Material was preferentially removed from the pit to meet the approved radiological release criteria. No soil removal was required. The dominant isotope of interest was Eu-152 from activation of the biological shield concrete, with Co-60 to a lesser extent in the 1/4 thick steel liner and rebar. The Gunite layer and steel liner were removed from the reactor pit up to a height of 2 meters (6.6 ft.).
The top layer of floor concrete and the underlying steel plate were also removed.
Portions of the concrete biological shield walls were then removed from the pit floor to 2 meters (6.6 ft.) in height from the north side extending to the east and west compass points. In addition, the pit floor was removed to varying depths in the northern section of the reactor pit. Figures 8 and 9 show respectively the bare steel liner prior to removal and a wall of the reactor pit following removal of concrete.
Figure 8: Bare Steel Liner Figure 9: Reactor Pit Wall Remediation activities took place in the TRF from April 2015 through September 2020. Remedial action surveys were conducted during and after the concrete removal process for both reactor pits to ensure both reactors met the release criteria. Once completed, the Final Status survey phase of the decommissioning process was begun and is documented in the following sections.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 13 of 54 3.0 Design of the Final Status Survey The FSS was designed in accordance with GAs approved Decommissioning Plan (1998),
GAs Final Status Survey Plan[8], NUREG 1757, and MARSSIM. The FSS was performed to demonstrate that residual radioactivity in each survey unit satisfies the NRC annual dose limit criteria of less than 25 mrem/yr Total Effective Dose Equivalent (TEDE).
The FSS was conducted by performing scan surveys, total direct surveys, removable contamination surveys, exposure rate measurements, and gamma spectroscopy analysis.
The data quality objectives (DQO) process was used to optimize the final status survey design.
For surfaces and structures, the release criteria listed in GAs approved TRIGA Reactor Facility Decommissioning Plan[5] were used. For land areas and the reactor pits, the screening values listed in NUREG 1757, V.2[3], Table H.2 were used. The Unity Rule was applied to the release criteria for survey units with multiple nuclides of concern.
The nuclides of concern screening values for surfaces and structures under default conditions (generic screening levels) are provided in Table 3.1.
Table 3.1 Default Surfaces & Structures Screening Values for Nuclides of Concern Nuclide (dpm/100 cm2)
Average Maximum Removable U-nat, 235U, 238U, & assoc. decay products 5,000 15,000 1,000 Transuranics, 226Ra, 228Ra, 230Th, 228Th, 125I, 129I 100 300 20 Th nat, 232Th, 90Sr, 223Ra, 224Ra, 232U, 126I, 133I, 131I 1,000 3,000 200 Beta/gamma emitters (nuclides with decay modes other than alpha) except 90Sr and others noted above 5,000 15,000 1,000 The nuclides of concern screening values for concrete and soil under default conditions (generic screening levels) are provided in Table 3.2. GA requested and received approval from the NRC[6] to modify the release limits for concrete and soil from the values listed in GAs previously approved Decommissioning Plan[5]. This allowed GA to use NUREG-1757[3] screening DCGLs as the isotope specific radiological release criteria for release to unrestricted use of the residual soil and concrete.
Table 3.2 Default Concrete & Soil Screening Values for Nuclides of Concern Isotope Half-life Radiation Type Default Screening Value (pCi/g)
Co-60 5.27 years 3.8 Sr-90 28.8 years 1.7 Cs-137 30.1 years 11.0 Eu-152 13.516 years
-,, +, e-7.0*
Eu-154 8.593 Years
-,, +, e-8.0
- This value is lower than the value listed in NUREG-1757 (8.7 pCi/g) due to the memorandum of understanding between the EPA and the NRC (2002).
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 14 of 54 3.1 Background Determination for Surfaces and Structures Background radiation measurements in counts per minute (cpm) were determined for three different types of survey material (concrete, cinder block, and metal) plus an ambient (non-material specific background) reading. The average background value (in cpm) was subtracted from the gross cpm to determine net cpm. The measurements were taken in non-impacted areas outside the TRF boundary with similar physical, chemical, geological, and radiological characteristics as the survey units being evaluated.
3.2 Background Determination for Land Areas The use of reference background measurements for land areas was not necessary since the primary contaminants/activation products are not present in the background. MARSSIM Section 4.5 states that reference areas and reference samples are not needed when there is sufficient information to indicate there is essentially no background concentration for the radionuclide(s) being considered.
3.3 Data Quality Objectives (DQOs)
The following is a list of the major DQOs for the survey design:
Static measurements were taken to achieve an MDCstatic of less than 50% of the DCGL(s).
Scanning was conducted at a rate to achieve an MDCscan of less than 50% of the DCGL(s).
Individual measurements were made to a 95% confidence interval.
Decision error probability rates were set at 0.05 for both and.
The null hypothesis (H0) and alternate null hypothesis (HA) are:
H0 - The residual radioactivity in the survey unit exceeds the release criterion Ha - The residual radioactivity in the survey unit does not exceed the release criterion 3.4 Area Classifications Based on the results of the historical site assessment and previous characterization survey results, facility areas were classified as impacted areas or non-impacted areas.
3.5 Non-Impacted Areas Non-impacted areas are areas without residual radioactivity from licensed activities and are not typically surveyed during final status surveys. These areas have no radiological impact from site operations and are typically identified early in decommissioning. Areas with reasonable potential for residual contamination are classified as impacted areas.
The TRF has a number of rooms (109-115) that were previously released for unrestricted use in 2007. Normally these areas would be considered non-impacted; however, as a conservative measure we classified these areas as Class
- 3. The exception is Room 112, which was used for the storage of radioactive
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 15 of 54 materials subsequent to its release in 2007 and is considered a Class 2 area. Static and removable surveys were completed to ensure the remediation performed in other areas of the facility did not affect these areas.
3.6 Impacted Areas Impacted areas are areas that have potential residual radioactivity from licensed activities. Impacted areas are subdivided into Class 1, Class 2 or Class 3 areas.
Class 1 areas have the greatest potential for contamination and therefore receive the highest degree of survey effort for the final status survey using a graded approach, followed by Class 2, and then by Class 3. Impacted sub-classifications are defined as follows:
3.7 Class 1 Area Areas with the highest potential for contamination, and meet the following criteria: (1) impacted; (2) a potential for delivering a dose above the release criterion; (3) a potential for small areas of elevated activity; and (4) insufficient evidence to support classification as Class 2 or Class 3.
3.8 Class 2 Area Areas that meet the following criteria: (1) impacted; (2) low potential for delivering a dose above the release criterion; and (3) little or no potential for small areas of elevated activity.
3.9 Class 3 Area Areas that meet the following criteria: (1) impacted; (2) little or no potential for delivering a dose above the release criterion; and (3) little or no potential for small areas of elevated activity.
3.10 Survey Units (SUs)
Areas of similar construction and composition were grouped together as survey units and tested individually against the derived concentration guideline levels (DCGLs) and the null hypothesis to show compliance with the release criteria.
The survey units were homogeneous in construction, contamination potential, and contamination distribution.
Survey units were sized according to the potential for small elevated areas of residual radioactivity. MARSSIMs recommended maximum survey unit sizes for building structures, based on floor area, are Class 1: up to 100 m2, Class 2:
100 m2 to 1000 m2 and Class 3: no limit. For land areas, the maximum survey areas are Class 1: up to 2,000 m2, Class 2: 2,000 m2 to 10,000 m2, and Class 3: no limit.
The survey units and their final classification are listed in Table 3.3 below. A map of the individual survey units within the TRF boundary is shown in Attachment A.
Individual survey unit maps have been included as Attachment B.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 16 of 54 Table 3.3 - Survey Unit Descriptions Survey Unit #
Description Final Classification Area m2 (ft2) 1 Mark I Reactor Pit 1
9.0 m2 (96.7 ft2) 2 Mark F Reactor Pit & Canal 1
14.8 m2 (159.6 ft2) 3 Mark I Reactor Room (Floors and lower walls) 1 84.8 m2 (913.3 ft2) 4 Mark I Reactor Room (Upper walls and ceiling) 1 93.8 m2 (1,010 ft2) 5 Mark F Reactor Room (Floors and lower walls) 1 67.1 m2 (722.4 ft2) 5A Mark F Excavation Area 1
5.6 m2 (60 ft2) 6 Mark F Reactor Room (Upper walls and ceiling) 1 87.5 m2 (942 ft2) 7 Soil Lab 1
27.4 m2 (295 ft2) 8 Mezzanine 1 1
23.6 m2 (254 ft2) 9 Mezzanine 2 2
30.6 m2 (329 ft2) 10 TRIGA Waste Yard 1
117.7 m2 (1,267 ft2) 11 TRIGA Front Yard (Asphalt) 2 546 m2 (5,883 ft2) 12 TRIGA Back Yard (Soil) 2 886 m2, (9,539 ft2) 13 Room 112 2
36.47 m2 (392.61 ft2) 14 Rooms 109-111, 113-115 3
283.6 m2 (3,053 ft2) 15 TRIGA Mark I and Mark F Roof 2
350.4 m2 (3,771.75 ft2) 16 Trenches, Sub-Critical Pit, and Mark F Canal Track 1
Trenches 60.7 m (199.2 ft.)
Pit 3.08 m2 (33.2 ft2) 3.11 Scanning Surveys Scanning was used to identify locations within the survey unit that exceed the action level. These locations were marked and received additional investigations to determine the concentration, area, and extent of the contamination. For Class 1 areas, scanning surveys were designed to detect small areas of elevated activity that were not detected by the systematic measurement pattern described in Section 3.17.
Table 3.4 summarizes the percentage of accessible building structural surfaces scanned based on classification.
Table 3.4 - Scan Survey Coverage by Classification Structure Class 1 Class 2 Class 3 Floors 100%
100%
50%
Other Structures 100%
50%
10%
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 17 of 54 3.12 Total Surface Activity Measurements Total Surface Measurements incorporates both static and wipe surveys to determine fixed and loose contamination. Static measurements were taken to the extent reasonably possible in impacted areas utilizing instrumentation of the best geometry based on the surface at the survey location. Judgmental wipe and static survey measurements were taken at locations of elevated activity identified and marked during the scan survey.
3.13 Determining the Number of Samples A minimum number of samples are needed to obtain sufficient statistical confidence the conclusions drawn from the samples are correct. The number of samples depended on the Relative Shift (the ratio of the concentration to be measured relative to the statistical variability of the contaminant concentration).
The minimum number of samples can be obtained from MARSSIM tables or calculated using the methodology presented in Section 5 of MARSSIM[2].
3.14 Determination of the Relative Shift The number of required samples will depend on the ratio involving the activity level to be measured relative to the variability in the concentration. The ratio to be used is called the Relative Shift, /S and is defined in MARSSIM as:
Where:
DCGL
= derived concentration guideline level LBGR
= concentration at the lower bound of the gray region. The LBGR is the average concentration to which the survey unit should be cleaned in order to have an acceptable probability of passing the test S
= an estimate of the standard deviation of the residual radioactivity in the survey unit 3.15 Determination of Acceptable Decision Errors A decision error is the probability of making an error in the decision on a survey unit by failing a unit that should pass ( decision error) or passing a unit that should fail ( decision error). MARSSIM uses the terminology and decision errors; this is the same as the more common terminology of Type I and Type II errors, respectively. The decision errors are 0.05 for Type I errors and 0.05 for Type II errors.
3.16 Determination of Number of Data Points (Sign Test)
The number of measurements for a particular survey unit, employing the Sign Test, is determined from MARSSIM Table 5.5, which is based on MARSSIM equation 5-2.
S S
LBGR DCGL
/
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 18 of 54 The sign test was used because the nuclides of concern are not present in background and a nuclide-specific analysis was performed. MARSSIM recommends increasing the calculated number of measurements by 20% to ensure sufficient power of the statistical tests and to allow for possible data losses. The values in MARSSIM Table 5.5 already include an increase of 20% of the calculated value.
3.17 Determination of Sample Locations Determination of Class 1 survey unit sample locations was accomplished by first determining sample spacing and then systematically plotting the sample locations from a randomly generated starting point (See Section 3.17.1). Similar systematic spacing methods were used for Class 2 survey units because there is an increased probability of small areas of elevated activity.
The guidance in MARSSIM recommends simple random measurement patterns for Class 3 survey units to ensure that the measurements are independent and support the assumptions of the statistical tests.
Table 3.5 details the survey sample location methodology.
Table 3.5 - Survey Sample Placement Overview Survey Unit Classification DCGLw Comparison Elevated Measurement Comparison Measurement Locations Impacted Class 1 Yes No*
Systematically determined from a random starting point Class 2 Yes N/A Systematically determined from a random starting point.
Class 3 Yes N/A Judgmental Non-Impacted None None None
- - As stated in the Final Status Survey Plan Rev.2, the EMC will not be used Furniture and fixtures were removed and either free-released or sent out as radioactive waste prior to starting the final status surveys. Areas where permanent counter tops and other horizontal surfaces block the floor surface were included as a replacement to the blocked floor surface.
3.17.1 Determining Class 1 and Class 2 Sample Locations In Class 1 and Class 2 survey units, the sampling locations were established in a unique pattern beginning with the random start location and the determined sample spacing. After determining the number of samples needed in the survey unit, sample spacing was determined from MARSSIM equation 5-8.
Maps were generated of the survey units permanent surfaces included in the statistical tests (floors, walls, ceilings, fixed cabinetry, etc.). All of the survey units with the exception of the two reactor pits had their walls folded-down to create a two-dimensional map. For the reactor survey units, the cylindrical reactor
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 19 of 54 walls were un-rolled and the floor folded-up to create a two-dimensional surface (see cartoon below). A random starting point was determined using computer-generated random numbers coinciding with the x and y coordinates of the survey unit map. A rectangular grid was plotted across the survey unit map based on the random start point and the sample spacing. A measurement location was determined at each intersection of the grid.
3.17.2 Determining Class 3 Sample Locations Class 3 sample locations were chosen on floor, lower wall (<2m), and permanent surfaces at the discretion of the Project Manager. Measurement locations were biased towards areas with the highest potential of residual contamination. Each chosen location is identified on the applicable survey map.
3.17.3 Judgmental Sample Locations In addition to the systematic locations determined by Section 3.17.1, a minimum of five judgmental sample locations were added to each survey unit by Health Physics. The locations were selected by a combination of process knowledge and professional judgment.
3.18 Removable Contamination Surveys Removable contamination surveys (wipes) were collected on building structural surfaces at each sample location established by Section 3.17.1 and the judgmental locations selected by Health Physics.
3.19 Surveys of Building Mechanical System Internals In addition to building surfaces and structures, surveys of various mechanical systems internals were performed.
3.20 Ventilation Systems The Heating, Ventilation & Air Conditioning (HVAC) system (blowers, duct work, and exhaust fans) were completely dismantled and sent out as radioactive waste in April 2019. A portable ventilation system was utilized during subsequent remediation.
3.21 Trenches and Drains There are several trenches that run inside and outside of the TRF. Their locations
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 20 of 54 have been identified on the survey unit maps; however, a separate survey unit (16) was created to provide more clarity. Scan surveys were performed on 100% of the surface area and judgmental static, wipe and exposure rate measurement were collected at locations selected by Health Physics.
3.22 Buried Pipe We identified a buried drain and pipe in the Mark F reactor room with readings above our release criteria. A new survey unit (5A) was created to document the subsurface soil area inside Survey Unit 5. The pipe was broken near the drain inlet due to settling of the soil. A plume of soil contamination approximately 6 feet wide, 10 feet long, and 10 feet deep was identified and excavated after a portion of the Mark F reactor room floor was removed. The soil and concrete were sent out as radiological waste.
The soil locations in SU 5A were determined using the methodology in 3.17.1.
The nuclides of concern are Cs-137 and Sr-90.
Scanning was performed to identify elevated Cs-137 locations. Soil was sent to the TestAmerica (TA) analytical laboratories for Sr-90 analysis.
A detailed write-up of the survey methodology and subsequent dose assessment is presented in Section 7.0, Buried Pipe Dose Assessment.
3.23 Sub-Critical Pit The sub-critical pit is located inside the Mark I reactor room. It was made into a separate survey unit (SU 16) due to space limitations on the Mark I floors and lower walls map (SU 3). The sub-critical pit was constructed at the same time as the Mark I reactor with the expectation that it would be used for sub-critical experiments. However, it was never used for such experiments and there was never any neutron production in the pit. Survey locations were judgmentally selected by the Health Physics.
3.24 Survey Action Levels Investigation (action) levels were used to flag locations that require special attention and further investigation to ensure areas were properly classified and adequate surveys were performed. These locations were marked and received additional investigations to determine the concentration, area, and extent of the contamination. The survey investigation levels for each type of measurement are listed by classification in Table 3.6.
Table 3.6 - Survey Investigation Levels Survey Unit Classification Flag Static Measurement or Sample Result When:
Flag Scanning Measurement Result When:
Flag Removable Measurement Result When:
Class 1
> 50% of DCGL
> 50% of DCGL
> 100 dpm/100cm2 Class 2
> 20% of DCGL
> 20% of DCGL
> 100 dpm/100cm2 Class 3
>MDC
>MDC
> 100 dpm/100cm2
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 21 of 54 4.0 Instrumentation The following sections describe the calibration of the instrumentation, the instrument efficiency, functional checks, minimum detectable concentrations, and specifications.
4.1 Instrument Calibration All instruments (with the exception of the pipe detector) utilized during the final status survey were calibrated on a semi-annual basis in-house by GA in accordance with the American National Standards Institute (ANSI) N323B-2003 standard. These calibrations included efficiency determination with the use of National Institute of Standards and Technology (NIST) traceable standards for radiation emission types and energies that provided detection capabilities similar to the radionuclides of concern. The pipe detector was calibrated by Ludlum and the sensitivity determined by GA. The instrument calibration included data pertinent to efficiency determination based upon the current activity of the calibration source.
4.2 Instrument Efficiency Determination Instrument efficiencies were determined with the use of reference radiation sources with known emission rates per unit area. Consistent with International Standard ISO 7503-1, Evaluation of Surface Contamination, the total surface activity, As, was calculated using the following equation:
A RR W
Where:
As = Total surface activity RS+B = Gross count rate in counts per minute RB = Background count rate in counts per minute i = Instrument efficiency s = Surface efficiency W = Probe area/100 cm2 The instrument efficiency was determined by the following equation where A is the activity of the source:
RR A
As recommended in ISO 7503-1 for beta particles with energy greater than 400 keV, a value of 0.5 was used for s for Sr-90 and Cs-137 activity calculations.
4.3 Functional Checks Portable field instruments were response-tested daily when is use. Background and source readings were taken as part of the daily instrument check and compared with the acceptance range of +/- 20 percent. The background, source
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 22 of 54 check, and field measurement count times for radiation detection instrumentation were specified by Health Physics to ensure that measurements were statistically valid. Daily functional tests were conducted using NIST traceable sources.
4.4 Determination of Counting Times and Minimum Detectable Concentrations Minimum counting times for background determinations and measurements for total and removable contamination were chosen to provide a Minimum Detectable Concentration (MDC) at or below 50% of the designated DCGL for the hardest to detect nuclide. MARSSIM guidance (NUREG 1575, Section 4.7.1) recommends that it is generally considered good practice to select a measurement system with a MDC between 10 - 50% of the DCGL. Tables 4.1 and 4.2 provides the counting times, efficiencies and MDCs for each instrument.
4.4.1 Static Measurements Static measurement MDCs at a 95 percent confidence level were calculated using the following equation which is an expansion of NUREG 1507[7], Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, Equation 6-7 (Strom & Stansbury, 1992):
Where:
BR = Background count rate tS = Source count time tB = Background count time tot = Total instrument efficiency A = Probe area 4.4.2 Ratemeter Scanning of Surfaces and Structures Scanning MDC at a 95 percent confidence level was calculated using the following equation, which is a combination of MARSSIM Equations 6-8, 6-9 and 6-10:
Where:
d' = Observers criterion bi = Background counts in observation interval i i = Observation interval 3
3.29 (1
)
100 S
R S
B static S
tot t
B t
t MDC A
t
2 60 100 i
scan tot d
b i
MDC A
p cm
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 23 of 54 p = Surveyor efficiency tot = Total instrument efficiency A = Probe area 4.4.3 Scan MDCs for Soil Areas The scan MDCs for land areas are based on the area of elevated activity, depth of contamination, and the radionuclide (i.e., energy and yield of gamma emissions).
If one assumes constant parameters for each of the above variables, with the exception of the specific radionuclide of interest, the scan MDC may be reduced to a function of the radionuclide alone. These calculations are provided in Section 6 of MARSSIM. The purpose of the soil scans are to identify potential gamma contaminants in the soil, which is predominantly Cs-137.
The scan MDCs are based on site-specific background data using the equations below.
The number of source counts (si) required for a specific time interval is calculated by MARSSIM Equation 6-8:
Where:
d is the performance factor based on required true and false positives rates (1.38) bi is the number of background counts in the observation interval Assuming that the source remains under the detector for 2 seconds (i.e., i=2) and the background count rate is 8,410 cpm, the value for bi is then calculated:
bi = (8,410/60) x 2 = 280.33 and si = 1.38 x 280.33 = 23.11 The scan minimum detectable count rate (MDCR) is calculated using MARSSIM Equation 6-9:
MDCR = si x (60/i) = 23.11 x (60/2) = 693.3 cpm The MDCRsurveyor is calculated assuming a surveyor efficiency of 0.5 (see MARSSIM page 6-42):
MDCRsurveyor = MDCR/0.5 = 693.3/0.5 = 980 cpm i
i b
d s
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 24 of 54 The instrument sensitivity of a 2 x 2 NaI detector is 900 cpm/µR/hr for 137Cs (NUREG-1507). From above, the MDCRsurveyor is 980 cpm. This is equivalent to 1.09 µR/hr.
Modeling (using MicroShield Version 8.03) was used to determine the net exposure rate produced by 5 pCi/g of 137Cs in soil. The physical and geometrical factors considered in the modeling are as follows:
The dose point of 4 inches above the soil was used The density of 1.6 grams per cubic centimeter (g/cm3) was used for soil The depth of the area of elevated activity was 15 cm The circular dimension of the cylindrical area of elevated activity was 0.25 m2 Using the above input parameters, the calculated exposure rate is 0.26 R/hr for 137Cs. The radionuclide concentration of 137Cs (scan MDC) necessary to yield the minimum detectable exposure rate (1.09 R/hr) may be calculated using MARSSIM Equation 6-11:
MDCscan = (5.0 pCi/g)(1.09 µR/hr) / (1.66 µR/hr) = 3.28 pCi/g The calculation is based on an observation interval of 2 seconds and an observation area of 0.25 m2.
4.4.4 Smear Counting The smear counting MDC at a 95 percent confidence level is calculated using the following equation found in NUREG 1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, Table 3.1 (Strom & Stansbury, 1992):
Where:
MDCsmear = Minimum detectable concentration level in dpm/smear Br = Background count rate in counts per minute tb = Background count time in minutes ts = Sample count time in minutes tot = Total detector efficiency for radionuclide emission of interest (includes combination of instrument and surface efficiencies) 3 3.29 (1
)
S R
S B
Smear S
tot t
B t
t MDC t
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 25 of 54 4.5 Counting Uncertainty, Error Propagation, and Confidence Interval The counting uncertainty for both total and removable measurements were calculated using Equation 6-15 from MARSSIM:
Where:
n = Standard deviation of the net count rate result CS+B = Number of gross counts (sample)
Cb = Number of background counts T2S+B = Gross count time TB = Background count time Because calculations to determine the final static measurement results are based on dividing the net count rate by total efficiency, the uncertainty propagation formula to be used is as follows (MARSSIM Section 6.8.3):
Where:
A
=
Measurement propagated error or total uncertainty u
=
Final static result in dpm/100 cm2 b
R
=
Standard deviation or statistical counting uncertainty of the net count rate Rb
=
Net count rate E
=
Standard deviation of the instrument efficiency E
=
Instrument efficiency Referring to MARSSIM Table 6.9, a k value of +/-1.96 represents a confidence interval equal to 95 percent about the mean of a normal distribution. All total activity measurements are presented as the final result in dpm/100 cm2 +/-1.96 4.6 Instrumentation Specifications The instrumentation used for facility decommissioning surveys is summarized in the following tables. Table 4.1 lists the standard features of each instrument such as probe size and efficiency.
2 2
E R
u E
b R
A b
A
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 26 of 54 Table 4.1 - Instruments Used at the TRIGA Facility Detector Model Detector Type Detector Area Meter Model Window Thickness Instrument Efficiency/Sensitivity Ludlum 43-68 Gas Flow Proportional 100 cm2 Ludlum 2221 0.8 mg/cm2 37.5 % (Sr-90/Y-90)
Ludlum 43-37 Floor Monitor Gas Flow Proportional 434 cm2 Ludlum 2221 3.4 mg/cm2 35.7 % (Sr-90/Y-90)
Ludlum 44-10 NaI 20 cm2 Ludlum 2221 NA 900 cpm/µR/hr (Cs-137)
Ludlum 43-93 Scintillation 100 cm2 Ludlum 2221 1.2 mg/cm2 20% (Sr-90/Y-90)
Ludlum 44-159-1 CsI Scintillator 3.24 cm2 Ludlum235 0-1 NA 102.4 cpm/µR/hr* (Cs-137)
Ludlum 9-3 Ion Chamber 40 cm2 NA NA NA Canberra Tennelec Gas Flow Proportional NA NA NA 37% (Sr-90/Y-90)
- Probe sensitivity was determined by GA Table 4.2 lists the typical operational parameters such as scan rate, count time, and the associated MDCs. The scan and static MDCs for each instrument were calculated with site-specific instrument efficiencies and backgrounds. Site-specific background measurements (See Attachment D) of each surface type (e.g.,
concrete and cinder block) were taken and used in calculating each instruments MDCs. Scan rates and static counting times were adjusted as necessary to satisfy the requirements identified in Section 3.3, DQOs. MDC Calculations have been included as Attachment E.
Table 4.2 - Typical Instrument Operating Parameters and Sensitivities Measurement Type Detector Model Meter Model Scan Rate Count Time
Background
(cpm)
MDC (dpm/100cm2)
Surface Scans Ludlum 43-68 Ludlum 2221 2
in./sec.
NA 350 1,475 (Sr-90/Y-
- 90)
Surface Scans Ludlum 43-37 Ludlum 2221 4
in./sec.
NA 1,000 465 (Sr-90/Y-90)
Soil & Concrete Scans Ludlum 44-10 Ludlum 2221 2
in./sec.
NA 6,000 6.4 pCi/g (Cs-137)
Total Surface Activity Ludlum 43-93 Ludlum 2221 NA 1 Minute 300 418 (Sr-90/Y-90)
Gamma Pipe Detector Ludlum 44-159-1 Ludlum Model 2350-1 NA 1 Minute 936 805 (Cs-137) dpm/cm2 Exposure Rate Ludlum Model 9-3 Ludlum Model 9-3 NA NA NA NA Removable Activity Canberra Tennelec NA NA 5 Minutes 0.1 (Alpha) 1.0 (Beta) 4 (Sr-90/Y-90)
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 27 of 54 5.0 Survey Unit Documentation, Validation, and Descriptions The following sections describe the survey packages provided to the HP Technician, how the field data is validated, and a detailed description of each survey unit.
5.1 Survey Documentation A survey package was developed and approved by the Radiological Engineer for each survey unit and contains the following:
Survey instruction sheet General survey requirements Instrument requirements with associated MDCs, count times and scan rates Survey maps detailing survey locations and placement methodology Survey data sheets An example survey package has been included as Attachment F.
5.2 Data Validation Field data was reviewed by Health Physics and validated to ensure:
Completeness of forms Proper types of surveys were performed The MDCs for measurements met the established data quality objectives Independent calculations were performed on a representative sample of data sheets Satisfactory instrument calibrations and daily functionality checks were performed as required Additionally, all final status survey data was entered into the Final Status Survey data sheets. This provided the means to sort survey data, verify activity calculations and to compute the associated MDC and counting errors.
5.3 Survey Unit Overview The Final Status Survey was comprised of 16 survey units and one sub-unit.
Their descriptions have been included below.
5.3.1 Survey Unit 1 (Surveyed on 10-12-19)
Survey Unit 1 is a Class 1 area that encompasses the Mark I reactor pit which is a below grade right cylinder with a diameter of 6 1/2 feet and a depth of 21 feet.
Based on the results of the characterization survey, the nuclides of concern for this survey unit are Eu-152 (from neutron activated concrete) and Co-60 (from neutron activated rebar). No activation was identified above two meters. The concrete floor and walls up to a height of two meters were removed exposing the underlying soil. A 2 x 2 NaI detector was used to scan the concrete and exposed soil. Remediation was performed in all areas that showed elevated readings.
The survey map was created by unrolling the walls starting at 0 degrees (north) and flipping up the floor. Concrete samples were approximately 500 grams, with
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 28 of 54 each sample being taken to a depth of no more than two inches at 20 locations above the two-meter mark and analyzed using GAs gamma spectroscopy system.
Approximately 1000 grams of soil were collected at each of 19 locations below the two-meter mark and also analyzed using GAs gamma spectroscopy system.
270° 0° 90° 180° Circumference=78in.
Figure 10: Mark I Reactor Mapping Methodology Survey Unit Area = 989.9 ft2 (92.0 m2)
Radionuclides of Concern = Eu-152, Co-60 Relative Shift = 6.7 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements 39 (20 concrete, 19 soil) 5.3.2 Survey Unit 2 (Surveyed on 11-4-19)
Survey Unit 2 is a Class 1 area that encompasses the Mark F reactor pit and storage canal. The pit is a below grade right cylinder (10-foot diameter, 25 feet deep) connected to an L-shaped storage canal with a depth of 14 feet. Based on the results of the characterization survey, the nuclides of concern for this survey unit are Eu-152 (from neutron activated concrete), Co-60 (from neutron activated rebar) and Cs-137 (fission product). The activated concrete was removed on the north side of the reactor pit from the floor up to a height of 2 meters. Activation was not present on the south side and the concrete remained intact. No soil was exposed as a result of removing the activated concrete.
To create a two-dimensional map the walls of the reactor pit were un-rolled and canal and pit floors flipped up. The map starts at the southern end of the reactor (168 degrees and goes clockwise around the pit and canal). For each sample, approximately 500 grams of concrete was removed to a depth of no more than two inches at 56 locations and analyzed using GAs gamma spectroscopy system.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 29 of 54 L6ft.(1.83m)
L4ft.(1.21m)
L10ft.(3.05m)
L7ft.5in.(2.26m)
L3ft.5in.(1.07m) 168°StartingPoint GoingClockwise 270° 0° 90° 180° Circumference Above2M-93.75in.
Below2M-93.75in.
Circumference Above2M-93.75in.
Below2M-114in.
Circumference Above2M-93.75in.
Below2M-114in.
Circumference Below2M-93.75in.
15ft.Depth 26ft.Depth OriginalPitCircumference=375in.
ExcavatedArea Figure 11: Mark F Reactor Mapping Methodology Survey Unit Area = 1,884 ft2 (175 m2)
Radionuclides of Concern = Eu-152, Co-60, Cs-137 Relative Shift = 2.1 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 56 5.3.3 Survey Unit 3 (Surveyed on 8-30-19)
Survey Unit 3 is a Class 1 area that encompasses the floors and lower walls in the Mark I reactor room.
A total of 90 systematic and 5 judgmental measurements/samples were analyzed.
The survey unit included a couple of small alcoves that may be difficult to read on the map, however, the walls on the map have been folded-down and appropriately labeled.
Survey Unit Area = 1,011 ft2 (93.9 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 2.2
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 30 of 54 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 95 5.3.4 Survey Unit 4 (Surveyed on 9-5-19)
Survey Unit 4 is a Class 2 area that encompasses the upper walls and ceiling in the Mark I reactor room.
A total of 40 systematic radiological measurements were taken.
Survey Unit Area = 1,011 ft2 (93.9 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 5.7 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 40 5.3.5 Survey Unit 5 (Surveyed on 1-8-20, Re-surveyed on 9-11-20)
Survey Unit 5 is a Class 1 area that encompasses the floors and lower walls in the Mark F reactor and control rooms. A total of 85 systematic and 10 judgmental measurements/samples were analyzed.
After the survey was completed, some sub-surface soil in the survey unit was identified that exceeded our release criteria. A new survey unit (SU 5A) was created for the sub-surface soil. After the remediation of SU 5A was completed, GA performed a follow-up survey in SU 5 to ensure the area was not contaminated during the remediation of 5A.
Survey Unit Area = 1,064 ft2 (98.84 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 3.6 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 95 5.3.6 Survey Unit 5A (Surveyed on 7-27-20)
Survey Unit 5A is a Class 1 area. A liquid waste drain line was identified in the overflow trench on the east side of the Mark F reactor. After removing the concrete surrounding the drain, it became apparent that the drain line had cracked due to soil settling. Approximately 10 cubic yards of concrete and soil were removed in order to access the remaining contaminated pipe. The remediation of the buried pipe will be discussed in Section 7.0. The nuclides of concern were Sr-90 and Cs-137. Using a 2 x 2 NaI detector GA scanned the excavated area for Cs-137. Soil samples were processed in-house to determine the Cs-137 contamination levels in units of pCi/g. Unfortunately, the Sr-90 migrated to a greater depth than Cs-137 and we were unable to scan for it in the field. A series of soil samples was sent to TestAmerica to determine when the concentration of Sr-90 in the soil was below the release criteria. Once the remediation was successful, a new set of 15 soil samples was collected and sent to TestAmerica for Cs-137 and Sr-90 analysis. All final samples were below the release criteria.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 31 of 54 October through December of 2019 was spent excavating the concrete and soil on the east side of the reactor pit, locating the contaminated pipe, documenting the contamination levels, and remediating the pipe. Below is a list of soil sample results by TestAmerica report number showing how the process evolved following completion of pipe remediation:
37873-1 Sixteen samples (12 systematic, 4 biased) were collected on 4-14-20 and analyzed after no elevated Cs-137 activity was seen with the 2 x 2 NaI detector. The results received on 5-12-20 showed four sample locations (3, 5, 7, 16) with elevated Sr-90 levels. The goal was to correlate the Cs/Sr ratio. Unfortunately, there was no pattern to the results. It became obvious the Sr had migrated deeper than the Cs and therefore we could not use Cs as an indicator for the presence of Sr.
38284-1 Additional remediation was performed at and around locations 3, 5, 7, and 16. Samples were collected on 5-13-20 at these locations. Results received on 6-23-30 continued to show locations 7 & 16 with elevated Sr-90 levels.
38590-1 Additional remediation was performed at locations 7 and 16 and the surrounding areas. A total of 61 cubic feet of soil was removed. Two additional samples were collected on 6-23-20 & 6-30-20 from locations 7 and 16. Sample results were received on 7-29-20 showing acceptable levels of Sr-90 and Cs-137, allowing GA to proceed with the Final Status Survey of SU 5A.
39043-1 A new Final Status Survey map with 15 systematic locations was created. Samples were collected at these locations from 7-30-20 to 7-31-20. All sample results were received on 9-11-20 and were below unity.
5.3.7 Survey Unit 6 (Surveyed on 10-15-19)
Survey Unit 6 is a Class 2 area that encompasses the upper walls and ceiling in the Mark F reactor room.
A total of 34 systematic measurements/samples were analyzed.
Survey Unit Area = 1064 ft2 (98.84 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 4.2 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 34 5.3.8 Survey Unit 7 (Surveyed on 8-15-19)
Survey Unit 7 is a Class 2 area encompasses the floors and lower walls in the Soil Lab. The Soil Lab is located at the entrance to the Mark I reactor and contains a bathroom. At one time the room contained a number of microwave ovens used to dry soil samples but that process was discontinued prior to the commencement of decommissioning. More recently, the room was used to section core samples into
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 32 of 54 disks for gamma analysis. This work was completed several years ago and the room has since been empty with the exception of a hand & foot monitor.
A total of 29 systematic and 5 judgmental measurements/samples were analyzed.
Survey Unit Area = 298 ft2 (27.7 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 2.3 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 34 5.3.9 Survey Unit 8 (Surveyed on 8-8-19)
Survey Unit 8 is a Class 1 area that encompasses the floors and walls in Mezzanine 1. Mezzanine 1 is located directly above the Mark F Control room and was used to house the Mark F HEPA exhaust fans. A couple of areas with elevated activity were identified during the scan surveys. Remediation was performed and these areas were included in the survey as judgmental locations.
A total of 41 systematic and 10 judgmental measurements/samples were analyzed.
Survey Unit Area = 254 ft2 (23.6 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 2.4 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 51 5.3.10 Survey Unit 9 (Surveyed on 8-8-19)
Survey Unit 9 is a Class 1 survey unit that encompasses the floors and walls in Mezzanine 2. Mezzanine 2 is located above a portion of Room 109 which was previously released in 2007 when the non-reactor portion of the site was decommissioned. The Mezzanine was added to the Final Status Surveys because of concern that equipment used with previous remediation work was stored in the room.
A total of 53 systematic and 6 judgmental measurements/samples were analyzed.
Survey Unit Area = 329 ft2 (30.6 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 3.0 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 59 5.3.11 Survey Unit 10 (Surveyed on 2-4-20)
Survey Unit 10 encompasses the concrete outdoor lot and lower outside exterior walls in the East Yard which includes a trench along the east wall of the reactor building. As was discussed in the HSA, the concrete became contaminated with Sr-90 and Cs-137 from an ion exchange resin change out process. The contamination in SU 5A is also thought to have originated in the East Yard and was transported there through the trench and subsequently the broken liquid waste drain line. Accidental spillage during maintenance or resin replacement likely resulted in the deposition of Cs-137 and Sr-90 onto surfaces in the yard and into
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 33 of 54 the Mark F trench. Significant remediation was performed in order for the area to meet the release criteria. Since Sr-90 was a nuclide of concern, the release criteria was lowered to 1,000 dpm/100cm2.
A total of 42 systematic and 9 judgmental measurements/samples were analyzed.
Survey Unit Area = 1072 ft2 (99.59 m2)
Radionuclides of Concern = Cs-137, Sr-90 Relative Shift = 4.1 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 51 5.3.12 Survey Unit 11 (Surveyed on 10-24-19)
Survey Unit 11 encompasses the pavement of the front lot directly in front of the Mark F reactor building. This area was previously released but was re-surveyed due to waste containers from the remediation phase being stored in this location.
A total of 47 systematic and 5 judgmental measurements/samples were analyzed.
Survey Unit Area = 6,045 ft2 (561.6 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 3.8 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 52 and 5 judgmental samples were analyzed.
5.3.13 Survey Unit 12 (Surveyed on 9-16-19)
Survey Unit 12 encompasses the TRIGA back yard. Portions of the back yard were released previously during the decommissioning of the Non-Reactor portion land area which left an irregular shape. The location of the former make-up water tank was included in the survey. Soil samples were collected from the land area and concrete samples were collected from the concrete pad where the make-up tank was located.
A total of 34 systematic measurements/samples were analyzed.
Survey Unit Area = 9,539 ft2 (886 m2)
Radionuclides of Concern = Cs-137 Relative Shift = N/A Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 34 5.3.14 Survey Unit 13 (Surveyed on 8-14-19)
Survey Unit 13 encompasses the floor and lower walls of Room 112. This room was previously released; however, several neutron activated bolts and neutron startup sources were stored in lead caves behind a brick wall in the room. The bolts and sources were shipped offsite and a survey was performed due primarily to the potential of contamination from extracting the neutron sources from their holders prior to shipment. The room was also used to temporarily store various other radioactive materials following its release in 2007.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 34 of 54 A total of 35 systematic and 5 judgmental measurements/samples were analyzed.
Survey Unit Area = 397 ft2 (36.9 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 2.7 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 40 5.3.15 Survey Unit 14 (Surveyed on 11-6-19)
Survey Unit 14 encompasses the floors and lower walls in the previously decommissioned areas of the TRIGA facility and includes rooms 109, 110, 111, 113, 114 and 115.
A total of 40 systematic measurements/samples were analyzed.
Survey Unit Area = 3,189 ft2 (296.3 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 2.5 Minimum Number of Samples/Measurements = 15 Actual number of Samples/Measurements = 40 5.3.16 Survey Unit 15 (Surveyed on 9-27-19)
Survey Unit 15 encompasses the contiguous gravel roof of the Mark I and Mark F reactor rooms.
A total of 63 systematic and 10 judgmental measurements/samples were analyzed.
Survey Unit Area = 3,649 ft2 (339 m2)
Radionuclides of Concern = Cs-137 Relative Shift = 3.2 Minimum Number of Samples/Measurements = 14 Actual number of Samples/Measurements = 73 5.3.17 Survey Unit 16 (Sub-Critical Pit [9-11-19], Mark F Canal [3-6-20], Trenches [3-6-20])
Survey Unit 16 is a Class 1 area that encompasses the sub-critical pit in the Mark I reactor room, the trench in Rooms 109-110 and a track in the Mark F canal used to accommodate installation of a barrier separating the pit from the canal.
A total of 44 measurements/samples were analyzed including 16 each for the subcritical pit and canal track and 12 for the Room 109-110 trench. All sample located were chosen judgmentally.
Survey Unit Area (subcritical pit) = 294.6 ft2 (27.4 m2)
Survey Unit Area (room 109-110 trench) = 37 ft (11.3 m)
Survey Unit Area (canal track) = 92 ft2 (8.5 m2)
Radionuclide of Concern = Cs-137
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 35 of 54 6.0 Final Status Survey Results The final status surveys for the GA TRF were performed under the guidance of the Final Status Survey Plan, Rev 2[8]. There are 16 survey units, each described in Section 5.3 of this report. The results of these surveys are grouped into similar survey types with specific release criteria; for example, surfaces and structures will include wipes for removable contamination, scanning and static measurements for total activity and exposure rate measurements. Open land and soil areas include volumetric sampling with release criteria in pCi/g. The 16 survey units are conveniently grouped under three headings, 6.1) Mark I and Mark F pits, 6.2) Surfaces and Structures, and 6.3) Open Land/Soil Areas. One area of concern that did not fit approved release criteria was addressed by performing a dose assessment to demonstrate that the potential yearly TEDE to a representative person of the critical group following license termination is well below the release criteria in 10CFR20 Subpart E. This analysis was for an abandoned liquid waste drain pipe uncovered in the SU5A excavation. The results of the dose assessment for this unique circumstance are discussed in Section 7.0, Buried Pipe Dose Assessment.
6.1 Mark I and Mark F Reactor Pits The Mark I reactor pit (SU 1) and the Mark F reactor pit (SU2) are Class 1 survey units and differ from the Surfaces and Structures and Open Land/Soil Area survey units in that they have volumetric gamma spectroscopy results as well as wipes for removable contamination, scanning and static measurements for total activity and exposure rate measurements.
Mark I Reactor Pit The Mark I reactor pit (SU 1) contains both concrete surfaces and soil. A total of 39 systematic and judgmental sample locations were chosen, 20 of which were concrete and 19 soil. The radionuclide of concern was the activation product Eu-152. The release criteria for this radionuclide is 7.0 pCi/g. Results from GA gamma spectroscopy analysis are given in Table 6.1. Only results above detectable limits are shown and are all well below the release limit of 7.0 pCi/g.
Complete reports from GAs gamma spectroscopy analysis are provided in Attachment G. The survey unit map is provided in Attachment B.
Table 6.1 - Survey Unit 1 Sample Analysis Results.
SampleNumber Analyte Result Units MkI1Soil22 Europium152 0.976 pCi/g MkI1Soil34 Europium152 0.402 pCi/g MkI1Soil35 Europium152 1.035 pCi/g MkI1Soil36 Europium152 0.587 pCi/g MkI1Soil37 Europium152 0.608 pCi/g MkI1Soil38 Europium152 0.865 pCi/g MkI1Soil39 Europium152 0.409 pCi/g
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 36 of 54 Results for static measurements of total surface activity are summarized in Table 6.3 in Section 6.2 for Surfaces and Structures. These results are for the 20 concrete sample locations. All wipes (removable alpha and beta) and all exposure rate values were at or near background levels. These results are presented in Attachment H.
MarkFReactorPit The Mark F reactor pit (SU 2) contains concrete only. A total of 56 systematic and judgmental sample locations were chosen. The radionuclides of concern were the activation products Eu-152 and Co-60 and the fission product Cs-137. The release criteria for these radionuclides are 7.0 pCi/g, 3.8 pCi/g and 11 pCi/g respectively. For all survey units where more than one radionuclide exists, the unity rule is used. The presence of Co-60 in the Mark F reactor and not the Mark I reactor is due to a significant amount of rebar in the concrete. The Mark F GA gamma spectroscopy analysis results above detectable limits are given in Table 6.2 and are all below release limits. Complete reports from GAs gamma spectroscopy analysis are provided in Attachment G. The survey unit map is provided in Attachment B.
Table 6.2 - Survey Unit 2 Sample Analysis Results SampleName Analyte Result Units MkF1Conc12 Cesium137 0.110 pCi/g MkF1Conc32 Europium152 0.928 pCi/g MkF1Conc45 Europium152 2.730 pCi/g MkF1Conc47 Europium152 3.070 pCi/g MkF1Conc47 Cobalt60 0.265 pCi/g MkF1Conc48 Europium152 2.270 pCi/g MkF1Conc50 Europium152 2.240 pCi/g MkF1Conc51 Europium152 6.260 pCi/g MkF1Conc53 Cesium137 0.179 pCi/g Results for static measurements for total surface activity are summarized in Table 6.3 in Section 6.2 for Surfaces and Structures. All wipes (removable alpha and beta) and all exposure rate values were at or near background levels. These results are presented in in Attachment H.
6.2 Surfaces and Structures All survey units except for SU 12 and SU 5A are considered surfaces and structures and can be characterized as floors, walls, ceilings and concrete areas which makeup the reactor facility building. The Mark I reactor pit (SU 1) and the Mark F reactor pit (SU2) addressed in the previous section are Class 1 survey units, however, they have volumetric gamma spectroscopy results as well as results for removable contamination, scanning and static measurements for total surface activity and exposure rate measurements. The remaining survey units
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 37 of 54 only received wipes for removable contamination, scanning and static measurements for total activity and exposure rate measurements with no volumetric component requiring gamma spectroscopy analysis.
Determination of Class 1 survey unit sample locations was accomplished by first determining sample spacing and then systematically plotting the sample locations from a randomly generated starting point. Similar systematic spacing methods were used for Class 2 survey units. In addition to the systematic locations, a minimum of five judgmental sample locations were added to each survey unit by Health Physics. The locations were selected by a combination of process knowledge and professional judgment.
For surfaces and structures, the release criteria listed in GAs approved TRIGA Reactor Facility Decommissioning Plan were used and are provided in Section 3.0 of this report. A summary of the survey results for surfaces and structures is provided in Table 6.3 shown below. A complete listing of the all the results including the associated maps is provided in Attachments H and B respectively.
Table 6.3 Summary of Survey Results for Surfaces and Structures SU#
Relative Shift Min # of Samples/
Measurements Actual # of Samples/
Measurements Average (DPM/100cm2)
Standard Dev.
(DPM/100cm2)
Minimum (DPM/100cm2)
Maximum (DPM/100 cm2) 1 6.7 14 20 1,107 372 200 1,737 2
2.1 15 56 140 1,217
-1,427 3,713 3
2.2 15 96 366 1,128
-2,255 3,603 4
5.7 14 40
-3 441
-938 1,108 5
3.6 14 95
-1,236 677
-2,618 350 6
4.2 14 33
-434 591
-1,946 739 7
2.3 15 34
-212 1,070
-1,717 1,856 8
2.4 15 51
-85 1,034
-1,587 3,383 9
3.0 14 59 74 821
-2,176 1,337 10 4.1 14 51
-596 564
-1,866 56 11 3.8 14 52 1,192 655
-230 2,056 13 2.7 15 40 142 920
-1,457 2,156 14 2.5 15 40 50 1,014
-2,964 1,786 15 3.2 14 73 135 784
-2,325 2,246 16 3.4 14 44
-464 731
-1,722 752 6.3 Open Land/Soil Areas The FSS includes two Class 1 open land/soil areas. Sample locations in the survey units were assigned by first determining sample spacing and then systematically plotting the sample locations from a randomly generated starting point. The volumetric release criteria in pCi/g for the open land/soil areas are given in NUREG-1757, Volume 2, Table H.2. The two Survey Units are the excavation in the Mark F reactor room (SU 5A) and the TRIGA back yard (SU 12).
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 38 of 54 Table 6.4 presents a summary of the results for SU 5A (Mark F excavation) where an abandoned liquid waste drain line was discovered. The two radionuclides of concern are Cs-137 and Sr-90 with release criteria of 11.0 pCi/g and 1.7 pCi/g respectively. The unity rule was used to assess whether the sample results were acceptable. The results show all values for the 15 sample locations are well below unity. The results were taken from reports received from TestAmerica which are provided in Attachment I. The survey unit map is provided in Attachment B.
Table 6.4 Survey Unit 5A Sample Analysis Results SampleName Analyte Result Units Unity MkF5ASoil0106 Cesium137 0.012 pCi/g 0.126 MkF5ASoil0106 Strontium90 0.213 pCi/g MkF5ASoil0206 Cesium137 0.023 pCi/g 0.049 MkF5ASoil0206 Strontium90 0.080 pCi/g MkF5ASoil0306 Cesium137 0.017 pCi/g 0.070 MkF5ASoil0306 Strontium90 0.116 pCi/g MkF5ASoil0406 Cesium137 0.296 pCi/g 0.085 MkF5ASoil0406 Strontium90 0.098 pCi/g MkF5ASoil0506 Cesium137 0.070 pCi/g 0.059 MkF5ASoil0506 Strontium90 0.089 pCi/g MkF5ASoil0606 Cesium137 0.342 pCi/g 0.165 MkF5ASoil0606 Strontium90 0.228 pCi/g MkF5ASoil0706 Cesium137 0.305 pCi/g 0.211 MkF5ASoil0706 Strontium90 0.312 pCi/g MkF5ASoil0806 Cesium137 0.109 pCi/g 0.045 MkF5ASoil0806 Strontium90 0.059 pCi/g MkF5ASoil0906 Cesium137 0.151 pCi/g 0.060 MkF5ASoil0906 Strontium90 0.126 pCi/g MkF5ASoil1006 Cesium137 0.072 pCi/g 0.014 MkF5ASoil1006 Strontium90 0.012 pCi/g MkF5ASoil1106 Cesium137 0.147 pCi/g 0.279 MkF5ASoil1106 Strontium90 0.451 pCi/g MkF5ASoil1206 Cesium137 0.195 pCi/g 0.151 MkF5ASoil1206 Strontium90 0.227 pCi/g MkF5ASoil1306 Cesium137 0.093 pCi/g 0.387 MkF5ASoil1306 Strontium90 0.643 pCi/g MkF5ASoil1406 Cesium137 0.291 pCi/g 0.036 MkF5ASoil1406 Strontium90 0.016 pCi/g MkF5ASoil1506 Cesium137 0.001 pCi/g 0.137 MkF5ASoil1506 Strontium90 0.232 pCi/g
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 39 of 54 Table 6.5 presents a summary of the results for SU 12. SU 12 (TRIGA back yard) includes the land area associated with the TRF that was not previously released in 2007[1]. The radionuclide of concern is Cs-137 with a release criteria of 11.0 pCi/g. The results show all values that were above detectable limits, none of which approach the release limit of 11.0 pCi/g. Complete reports generated using the GA gamma spectroscopy system and showing all radionuclides identified are provided in Attachment G. The survey unit map is provided in Attachment B.
Table 6.5 - Survey Unit 12 Sample Analysis Results.
SampleName Analyte Result Units GA12SOIL01 Cesium137 0.449 pCi/g GA12SOIL02 Cesium137 0.313 pCi/g GA12SOIL03 Cesium137 0.330 pCi/g GA12SOIL04 Cesium137 0.116 pCi/g GA12SOIL05 Cesium137 0.421 pCi/g GA12SOIL07 Cesium137 0.030 pCi/g GA12SOIL08 Cesium137 0.671 pCi/g GA12SOIL09 Cesium137 0.123 pCi/g GA12SOIL10 Cesium137 0.302 pCi/g GA12SOIL17 Cesium137 0.052 pCi/g GA12SOIL19 Cesium137 0.319 pCi/g GA12SOIL24 Cesium137 0.179 pCi/g GA12SOIL30 Cesium137 0.448 pCi/g
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 40 of 54 7.0 Buried Pipe Dose Assessment The purpose of this dose assessment is to demonstrate that the calculated TEDE to the representative person of the critical group from residual radioactive material contained in the buried drain pipe at the GA TRIGA Mark F reactor is well below the release criteria stated in 10CFR20 Subpart E.
7.1 Background
In early 2019, GA Health Physics staff were writing the Final Status Survey Plan while the TRIGA staff were preparing waste shipments from prior remediation work at the facility. A delay in the shipment of the remediation waste pushed back commencing the Final Status Survey (FSS) until early August 2019. Since the NRC inspection and confirmatory surveys by Oak Ridge Associated Universities (ORAU) were already scheduled for August 5 through August 8, only one of 16 survey units was completed when they arrived.
During the inspection, ORAU conducted surveys of all 16 survey units in accordance with their project specific plan dated August 2, 2019[9]. On August 7, 2019 GA and ORAU personnel discovered a drain located in the trench on the east side of the Mark F reactor pit. After removing the drain plug, ORAU personnel placed a 2x2 NaI probe in the drain and measured 350,000 cpm. The measurement was confirmed by GA Health Physics.
Facility blueprints show the drain located in the Mark Feast trench was piped to a liquid waste tank in the TRIGA Facility backyard. In 1984 the liquid waste tank was removed as part of a remediation effort following the license termination of the Mark III reactor. Inquiries with current and former TRIGA staff indicate that this drain was last used in the 1960s, as each person who was told about the waste drain line was surprised by its discovery. In the 1960s, 70s and 80s, the liquid waste tank was also fed by radiochemistry labs located behind the Mark III reactor. Additionally, experiments and projects detailed in the HSA support this conclusion. During the Thermionics project, the Mark F East trench was filled with piping and covered by instrumentation cabinets which would preclude anyone from using the liquid waste drain line. Furthermore, the GA TRF Process Knowledge Report[4], dated December 1996, did not mention the liquid waste line.
Following the inspection, GA staff removed a square foot section of concrete around the drain to try to determine the extent of contamination. Immediately, it was discovered the drain pipe was broken and there was a significant settling of soil, approximately 18 to 24 inches, under the Mark F floor around the reactor pit.
A small section of pipe was removed. It became apparent the settling soil created bending stress at the Mark F East trench connection, eventually resulting in a broken pipe (See Figure 12).
Note: Soil contamination resulting from the broken pipe (Survey Unit 5A) is addressed in Section 6.3 of this report.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 41 of 54 An outside contractor was brought in to remove a 10 x 7 section of the concrete floor so GA could further examine the level of soil contamination and locate the remaining drain line. Approximately 300 ft3 of soil was removed and the pipe was located at a depth of 1.38 meters below the reactor room floor next to the Mark F reactor pit wall (See Figure 13).
The newly excavated pit area was assigned as Survey Unit 5A, however, no additional work would occur until a proper contamination survey of the remaining pipe was completed.
The pipe is a standard cast-iron drain line with a 3 inch outer diameter and 1/8 inch wall thickness. A pipe detector, Ludlum Model 44-159-1 Cesium Iodide (CsI) scintillator, mounted on a small camera with a 50 foot cable was pushed through the pipe. A video of the inside of the pipe indicated that considerable pipe scale had flaked off and could be easily removed, possibly below the release criteria.
Additionally, removal of the contaminated pipe scale would also reduce the potential for contamination in the new survey unit, reduce employee exposure and align with GAs As Low As Reasonably Achievable (ALARA) policy.
TRIGA and Health Physics personnel used a ball hone and chain knocker attachment welded to the end of a power driven sewer line snake to clean the inside of the pipe. After each pass, a High-Efficiency Particulate Air (HEPA) vacuum with a 30 foot hose was used to remove the debris. This process was continued in the spirit of ALARA until the reduction in detector count rate plateaued and we began to see diminishing returns with efforts to remove contamination. A considerable reduction was noted, with initial count rates ranging from 20,000 cpm to 40,000 cpm reduced to less than 10,000 cpm with an average reading (not background subtracted) of 3,300 cpm, see Table 7.1.
Background readings in cpm were determined by placing a clean section of cast iron pipe, at depth, in non-impacted soil on the TRIGA site. Ten measurements were taken yielding an average reading of 936.1 cpm as shown in Table 7.1.
Figure 12 - Broken Pipe Figure 13 - Pipe Opening
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 42 of 54 The configuration of the pipe is shown in Figure 14. It is 1.38 meters (4.53 ft.) below the floor of the Mark F reactor room and travels northwest 9.15 meters (30 ft.),
terminating in a large concrete slab poured after the release of the Non-Reactor portion of TRIGA[1]. The pipe travels through the footings of the load bearing walls of the reactor room and control room. A video of the pipe interior following decontamination efforts show a straight pipe with no bends or tees. The video was shown to the NRC inspectors during the March 2020 inspection.
A final step in reducing exposure from this abandoned pipe was filling the pipe with a concrete slurry, fixing the removable contamination to the inside of the pipe and adding a bit of shielding to reduce external exposure. This was completed in October 20, 2020.
7.2 Determination of Source Term To demonstrate the abandoned pipe meets the regulatory release criteria of less than 25 mrem/yr to a representative person, a dose assessment must be performed.
Determining the source term is the first step. Without dates of when the waste line was used and in the absence of historical documentation of discharged waste, samples of the surrounding soil and pipe scale were sent to TestAmerica analytical laboratories for comprehensive gamma spectroscopy and Sr-90 analysis. The analytical report confirmed the presence of Sr-90 and showed Cs-137 as another radionuclide of concern.
7.2.1 Detector Efficiency and Sensitivity The Ludlum Model 44-159-1 CsI scintillator provided survey measurements in counts per minute (cpm) which were collected approximately every six inches throughout the pipe. Determining the radioactivity inside the pipe for a given area typically requires knowledge of the detector efficiency for the radionuclide of interest. Dividing the measured result in cpm by the efficiency (represented as a fraction detected of known emissions) will yield disintegration per minute (dpm).
Typically, the manufacturers published efficiency is acquired by placing a NIST traceable check source inside a jig to ensure a consistent distance from the source to the detector. These check sources are flat and provide a 2 geometry which may be acceptable in most cases of surface contamination but it is not consistent with the contaminated pipe geometry.
Using detector efficiency to determine contamination levels from count rate measurements presents more than just geometry challenges. These challenges include: acquiring a proper NIST traceable source, applying dpm to a 100 cm2 section of pipe when the count rate is influenced by other sections of pipe, sliding the detector through the pipe at a constant rate to determine concentration and accounting for differences at the pipe ends. All these factors will introduce measurement errors and lead to a large uncertainty in the measurement. After Figure 14 - Pipe Configuration Broken Section of Pipe
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 43 of 54 carefully considering these challenges, GA decided to evaluate detector sensitivity as a method to confidently determine the source term.
As stated previously, the pipe detector used for this project is the Ludlum Model 44-159-1 CsI scintillator. Ludlums published Cs-137 sensitivity for that detector is 120 cpm/uR/hr. To take measurements along the entire length of pipe, the pipe detector was connected to a Ludlum Model 2350-1 portable survey instrument using a 50 foot long cord. Anticipating some signal loss, GA-HP decided to use the GA Nuclear Calibration lab to determine the in the field detector sensitivity.
With the detector placed on the calibrated GA irradiator range, connected to the survey instrument using the 50 foot cord, the detector sensitivity was determined to be 102 cpm/uR/hr. Applying this detector sensitivity to the net cpm values in Table 7.1, average and maximum exposure rates were calculated. The average and maximum exposure rates are 21.57 uR/hr and 51.76 uR/hr respectively.
7.2.2 MicroShield Using MicroShield to model the 30 foot pipe, the total Cs-137 concentration can be adjusted until the average exposure rate in the center of the pipe is equal to 21.57 uR/hr. This exercise is repeated for the maximum exposure rate in Table 7.1 of 51.76 uR/hr. The corresponding total Cs-137 activity for the average and maximum exposure rates is 7.922 uCi and 19.01 uCi respectively. The MicroShield output files are included in Attachment J.
7.2.3 Pipe Weight To determine the weight of the buried pipe, a 1.79 inch section of pipe was cut from a 20 inch broken section of pipe removed during remediation of soil in Survey Unit 5A. The 1.79 inch section of pipe had an interior surface area of 100 cm2 and weighed 426.28 grams using the calibrated scale in the GA Health Physics lab. There are 201.1 sections of 1.79 inch pipe in a 30 foot length of pipe.
Therefore, the total weight of the buried pipe is 85.72 kg. Dividing the total radioactivity of Cs-137 by the weight of the buried pipe, the average and maximum concentration was determined to be 92.41 pCi/g and 221.8 pCi/g respectively.
Area of Sectioned Pipe interior:
Interior diameter = 2.75 inches x 2.54 cm/inch = 6.985 cm Area of interior of pipe = 2r x L where r is the radius and L is the length.
By setting the area to 100 cm2 and solving for L:
L = 100 cm2 / 2(6.985 cm/2)
L = 4.55 cm or 1.79 inches 7.2.4 Radionuclide Scaling Factors The final step in determining the source term is applying the scaling factors to Cs-137 to account for the Sr-90 contaminate in the buried pipe. It was well known from characterization surveys and soil samples sent to Test America that Sr-90 is prevalent throughout the survey unit and must be accounted for in determining the source term. Analytical data from TestAmerica for soil samples taken from
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 44 of 54 Survey Unit 5A did not provide a consistent radionuclide ratio between Cs-137 and Sr-90. It is well documented in professional journals that Sr-90 behaves differently in soil than Cs-137 and concentration is dependent on depth.
Additionally, analytical data from the pipe scale removed during remediation did not provide a consistent radionuclide ratio. Therefore, the scaling factor of 0.803, provided by the TRIGA Decommissioning Project Manager in a memorandum dated April 18, 2019[10] was used to determine the source term average and maximum Sr-90 concentration of 74.25 pCi/g and 178.2 pCi/g respectively.
7.3 Dose Assessment Methodology Once the source term is properly calculated, the concentration for each radionuclide of concern is entered into RESRAD along with appropriate parameters to accurately characterize each dose assessment scenario. The TEDE for the representative person (receptor) is then calculated. In the TRIGA Mark F buried pipe dose assessment, three scenarios will be presented to demonstrate the receptor TEDE is less than 25 mrem/yr.
7.3.1 RESRAD Pathways and Assumptions For the scenarios using RESRAD, dose assessment default pathways were modified to account for City of San Diego zoning, local geology and site characteristics. The GA main site is zoned SR (Scientific Research) in the City of San Diego. The SR zone is intended to provide areas for scientific research and administration and for limited manufacturing of related products. No significant fresh water recreation areas exist within the local hydrological area, nor is there significant agricultural activity. Therefore, the following RESRAD pathways for Scenarios 1 and 2 will be considered: External Gamma, Inhalation and Soil Ingestion.
Other conservative assumptions applied in Scenarios 1 and 2 assume the pipe disintegrates instantaneously, the media becomes soil and the volume is assumed to be equal to the pipe volume. Standard cylindrical geometric formulas were used to calculate the surface area and volume of the buried pipe. Although the pipe is filled with a concrete slurry, that slurry is not used in the determination for pipe volume and therefore not used for the concentration of contaminates in the soil. For the external exposure pathway, it is assumed that the TRIGA building and floor are removed and backfilled to current floor height, although there is no plan to do so at the time of this report. Details of Scenario 3 are provided in Section 7.3.4.
7.3.2 Scenario 1 - Buried Pipe at Depth Under this scenario, the TRIGA Mark F buried pipe is assumed to disintegrate instantaneously, the pipe media is assumed to be soil with a volume equal to the piping volume of 6,583 cm3. The contamination zone is located in subsurface soil at the buried pipes current depth of 1.38 meters below the TRIGA building floor.
As stated earlier, this scenario assumes the floor is removed and backfilled with clean soil to the current floor height. The contamination zone is assumed to be a flat plate geometry with the length equal to the pipe length of 360 inches or 914.4
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 45 of 54 cm, the thickness is equal to the pipe material thickness of 0.25 inches or 0.635 cm and the width equal to 11.33 cm. The area is equal to 914.4 cm x 11.33 cm or 1.036 square meters. RESRAD rounds these values up so the contamination zone has an area of 1.04 square meters at 0.01 meters thick.
Scenario 1: RESRAD Input Parameters Pipe Volume:
6,583 cm3 Pipe Weight :
85,720 grams (85.72 kg)
Cs-137 Activity:
7.922 uCi (determined using MicroShield; Attachment J)
Cs-137 Concentration: 92.41 pCi/g Sr-90 Concentration:
74.25 pCi/g Cover Depth:
1.38 meters Contamination Zone:
1.04 square meters; 0.01 meters thick The Maximum Total Dose as calculated by RESRAD-ONSITE Version 7.2 is 1.854E-06 mrem/yr. See Attachment K.
7.3.3 Scenario 2 - Buried Pipe in Surface Soil In this scenario, the contamination zone is assumed to be located in surface soil (0-15 cm) at a depth of only 5 cm. All other parameters are the same as Scenario
- 1. This approach provides an opportunity to bound the total dose to the receptor by conservatively assuming that the cover material has been removed.
Scenario 2: RESRAD Input Parameters Pipe Volume:
6,583 cm3 Pipe Weight :
85,720 grams (85.72 kg)
Cs-137 Activity:
7.922 uCi (determined using MicroShield; Attachment J)
Cs-137 Concentration:
92.41 pCi/g Sr-90 Concentration:
74.25 pCi/g Cover Depth:
0.05 meters Contamination Zone:
1.04 square meters; 0.01 meters thick The Maximum Total Dose as calculated by RESRAD-ONSITE Version 7.2 is 6.647 mrem/yr. See Attachment K.
7.3.4 Scenario 3 - External Exposure to Worker In this scenario, the representative person (receptor) is a construction or utility worker who may discover the pipe during demolition of the TRIGA building. It is assumed the worker has no prior knowledge of the pipe and associated use. The pipe was modeled in MicroShield using the Cylinder Surface-External Dose Point geometry. This configuration accounts for the pipe being filled with a concrete slurry and the 1/8 inch pipe wall thickness as the cladding. The exposure rate at 1 foot was determined using MicroShield and the maximum Cs-137 activity of 19.01 uCi (See Section 7.2.2) for the entire pipe, resulting in a concentration of 9.47E-4 uCi/cm2 or 2.10E+05 dpm/100cm2 (see Attachment J). The Cs-137 contamination is uniformly distributed and fixed on the inside of the pipe. Only
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 46 of 54 the direct gamma dose from the pipe is considered which is 30 feet long and 3 inches in diameter. It is assumed under this scenario that the worker is 30 cm (1 foot) from the pipe for 1,040 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> during the year. Considering a full time employee works 2,080 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />s/yr, it is extremely conservative to assume that the same worker would spend half of their working year on the same pipe.
Additionally, no cover material is included in this case.
Scenario 3: MicroShield Input Parameters Geometry MicroShield #10 - Cylinder Surface - External Dose Point Pipe length:
30 ft (914.4 cm)
Pipe Radius:
1.4 inches (3.49 cm)
Pipe Wall Thickness 0.125 inches (0.3175 cm)
Pipe Material Iron (7.86 g/cm3)
Cylinder Material Concrete (2.35 g/cm3)
Cs-137 Activity:
19.01 uCi (determined using MicroShield; Attachment J)
Cs-137 Conc:
9.47E-4 uCi/cm2 Cover Depth:
None Distance to worker 12 inches (30.48 cm)
The result of the MicroShield run for a receptor at 1 foot from the pipe is 2.94E-3 mR/hr. From MicroShields Dose Equivalent Report using the anterior/posterior geometry the effective dose equivalent rate, with buildup, is 4.05E-3 mrem/hr.
For bounding purposes, the worker occupancy is conservatively estimated at 1,040 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />s/yr for an upper limit and at 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />s/yr for a lower limit, resulting in a dose of 4.21 mrem/yr and 0.162 mrem/yr respectively.
7.4 Results The dose assessment methodology for evaluating exposure to a representative person (receptor) from residual contamination in the buried pipe at the GA TRIGA Mark F reactor facility is based on sound Health Physics principles. In all three scenarios shown above, conservativism has been incorporated whenever possible. With this in mind the following summary of results are listed:
Scenario #1 Maximum dose per year is H = 1.854E-6 mrem/yr Scenario #2 Maximum dose per year is H = 6.647 mrem/yr Scenario #3 Maximum dose per year is H = 4.21 mrem/yr 7.5 Conclusion Throughout the process from discovery of the buried pipe, cleaning the interior and to filling the pipe with a concrete slurry, GA implemented ALARA principles whenever possible. This dose assessment also illustrates the GA effort to use conservative approaches to ensure the end state of the pipe upon license termination will not exceed federal release criteria for unrestricted use. The dose assessment methodology for evaluating exposure to a representative person (receptor) from residual contamination in the buried pipe at the GA TRIGA Mark F reactor facility is based on sound Health Physics principles. Results demonstrate the dose to the receptor is less than the 25 mrem/yr release criteria.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 47 of 54 Table 7.1 - Buried Pipe Survey Results
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 48 of 54 8.0 Data Quality Assessment & Interpretation of Survey Results The statistical guidance contained in Section 8 of MARSSIM was used to determine if areas were acceptable for unrestricted release and whether additional surveys or sample measurements were required.
8.1 Preliminary Data Review A preliminary data review was performed for each survey unit to identify any patterns, relationships or anomalies. Additionally, measurement data was reviewed and compared with the DCGLs and investigation levels to confirm the correct classification of survey units. All calculations of means, standard deviations, minimum and maximum values and comparisons between survey data and investigation levels are presented in Table 6.3 (Section 6.2) 8.2 Determining Compliance for Surfaces and Structures Surveys Scan surveys were completed for all survey units at the prescribed coverage.
Removable contamination measurements (wipes) were compared directly to the applicable investigation levels and DCGLs to determine if an area required further actions or surveys. All removable contamination measurements collected during the final status surveys were less than the applicable investigation levels and significantly less than the established DCGLs for Sr-90 and Cs-137 removable activity.
All total surface activity measurements (statics) were compared directly to the DCGLs and investigation levels to determine if an area required further surveillance. All total surface activity measurements collected during final status surveys were less than the DCGLs for total surface activity. No final status measurements exceeded the investigation level for the applicable DCGLs.
Consequently, it was determined that these areas were properly classified and no further action was necessary.
8.3 Verification of Required Number of Samples for Surfaces and Structures A minimum number of samples are needed to obtain sufficient statistical confidence that the conclusions drawn from sample analysis results are correct.
The number of samples depends on the relative shift (the ratio of the concentration to be measured relative to the statistical variability of the contaminant concentration). The minimum number of samples was obtained from MARSSIM tables or calculated using equations in Section 5 of MARSSIM. To calculate the initial relative shift, data from the characterization survey was used.
After the final relative shift was calculated, the number of samples required by MARSSIM was compared to the actual number of samples collected.
Table 6.3 (Section 6.2) compares the actual number of locations sampled to the number required by MARSSIM Table 5.5 Values of N for Use with the Sign Test.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 49 of 54 8.4 Determining Compliance for Open Land/Soil Area Survey Units Scan surveys were completed for all survey units at the prescribed coverage. The nature of soil prevents the testing of removable contamination because it is all potentially removable. The isotopic average sum of fractions data for soil samples was used as the basis when testing for the presence of contamination.
The sum of fractions results were compared with the critical value in the Sign Test. The unity rule of the test was a check for determining if an area required further actions or surveys. All soil sample measurements collected during the final status surveys were less than 1 and thus below the investigation level for the established DCGLs.
There was no need to perform the Sign Test for our open land/soil areas because we stated in our Final Status Survey Plan[8] that we would remediate all areas to less than the DGCL. By MARSSIM definition, if all values are below the DCGL, the survey unit meets the release criteria and the Sign Test does not need to be performed.
8.5 Verification of Required Number of Samples for Open Land/Soil Areas The number of required samples is determined based upon analysis of final data collected. Initially, an estimate of the number of samples was completed based upon the characterization data performed in the early 2000s and it was estimated that 20 samples would be required to meet the statistical tests for open land areas.
Additionally, in each survey unit package there is a Random Sample Start Location and Sample Spacing Worksheet that does state required number of samples. This is actually the desired number of samples. Based upon the requirement to have at least 20 samples in the survey unit and within the bounds prescribed by the random starting point and the sample spacing, it often results in a number higher than 20. The required number of samples was verified after the final status survey collection to determine compliance with the statistical tests.
The comparison of samples taken versus samples required is included in Attachment L.
After the soil samples were obtained and the data analyzed, each survey unit was evaluated based on the final status survey data to determine if the relative shift used to estimate the required numbers of samples still applied. Section 3 provides the details for the average, minimum and maximum values for each survey unit and also calculates the relative shift. From the relative shift, we can use the tables in MARSSIM to determine the required number of samples that need to be taken in the survey unit. That number is compared to the actual number of samples taken.
In all of the open land/soil area survey units it was determined that the number of samples taken exceeded the number required and no additional sampling was necessary.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 50 of 54 9.0 Waste Management The operational history of the two GA TRIGA reactors spans over 60 years, during which time the TRF structure, equipment, reactor hardware, experimental hardware and soil had the potential to become activated and/or contaminated. Once operations ceased in the mid-1990s and the D&D process began, material identified as potential waste was characterized and either consolidated for disposal or surveyed and released from the TRF as non-radioactive material.
GA disposed of radioactive and mixed waste using two primary waste disposal options.
GA maintained a relationship for many years as a waste generator with the US DOE NNSS in Mercury, Nevada. Approved waste profiles were developed and used to send low-level radioactive waste in the form of activated hardware, concrete, steel, solid debris, soil and asphalt to the NNSS for disposal. In addition to annual audits by the NNSS, the waste disposal program that GA maintained underwent periodic review and inspection by the NRC and GA Quality Assurance. Mixed low-level radioactive waste (MLLW), sludge residue and soil was brokered by Philotechnics, LTD, a full service waste management company, and disposed at approved licensed facilities.
9.1 Waste Types, Volumes, and Activity All waste shipments sent to the NNSS disposal facility adhered to a waste profile developed by GA and approved by NNSS. Two primary waste profiles were developed for radioactive waste generated at the TRF and destined for NNSS.
One profile allowed for building debris such as concrete, wood and metal as well as other solid waste. A second profile was for soil and asphalt and used early in the decommissioning process to dispose of these materials from areas exterior to the TRF structure. The waste profiles included a list of radionuclides and maximum concentrations allowed. GA did not request waste profiles for MLLW and absorbed liquids although NNSS allows for them. These materials were sent to approved disposal facilities through our waste broker Philotechnics. In 2012, GA shipped highly activated reactor hardware that required the use of specially reinforced containers to support the necessary shielding material. This shipment required the development of a separate NNSS waste profile for activated metal due to the weight of the containers and activity level. The last shipments to the NNSS occurred in 2019 and primarily contained the concrete and steel from the excavation of the two reactor pits. The table below lists all shipments sent to the NNSS since 2012.
Table 9.1 - NNSS Waste Shipments Shipment Number (Date)
Classification Activity (Ci)
Volume (ft3)
Weight (lbs) 12001 (05/2012)
UN2915, Type A 9.96 134 17,600 12002 (05/2012)
UN2915, Type A 9.90 134 17,740 12003 (05/2012)
UN2915, Type A 10.46 134 18,420 12004 (05/2012)
UN2915, Type A 6.77 134 18,180 13001 (07/2013)
UN3321, LSA-II 7.25E-03 630 16,840 17001 (11/2016)
UN2912, LSA-I 2.37E-03 630 12,106
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 51 of 54 Shipment Number (Date)
Classification Activity (Ci)
Volume (ft3)
Weight (lbs) 19001 (06/2019)
Non-Regulated per DOT 1.85E-04 630 37,766 19002 (06/2019)
Non-Regulated per DOT 1.67E-04 504 31,290 19003 (06/2019)
Non-Regulated per DOT 1.25E-04 504 31,604 19004 - Part 1 (06/2019)
Non-Regulated per DOT 4.62E-05 378 11,914 19004 - Part 2 (06/2019)
UN2912, LSA-I 1.52E-03 252 9,590 Totals 37.10 4,064 223,050 Philotechnics, LTD primarily handled waste shipments that contained MLLW or that had significant moisture content such as sludge and in later phases of the decommissioning process, contaminated soil. The initial phases of the Mark F pit remediation saw removal of the outer layer of Epocast and Gunite containing lead and cadmium. This, along with other lead waste including lead bricks, a single element transfer cask and experimental shielding material comprised all of the MLLW removed from the facility. In the later phases of the TRF D&D project, soil removed from the TRF came from the Mark I excavation and the excavation in the Mark F reactor room east of the reactor pit where a contaminated liquid waste drain was located. Disposal of this soil was handled exclusively by Philotechnics.
9.2 Classification All potential radioactive waste from the TRF was classified using either direct gamma spectroscopy, analytical laboratory analysis or hand surveyed as applicable. Eurofins TestAmerica, a commercial analytical laboratory pre-approved by GA Quality Assurance, was used for all sample analysis. In the case of MLLW, a hand-held X-ray fluorescent (XRF) device was used to identify potentially hazardous elements and laboratory analysis used to quantify the activity level of the hazardous component. Based on process knowledge, the two elements of concern were cadmium and lead, both of which were used over time in experiments or testing. Laboratory analysis was also used where quantification of beta emitters such as Sr-90 was necessary. In most cases, the quantity of beta emitters was inferred from gamma analysis using facility specific radionuclide scaling factors developed from commercial laboratory analysis of representative samples from the TRF[10]. For all containers sent to the NNSS, radiological characterization and classification of the waste was done using direct gamma spectroscopy analysis performed at the TRF using a system specially configured for that geometry. The geometry configuration was for a Type 1 Industrial package (IP-1) metal box. Gamma analysis of material in drums was accomplished using the gamma spectroscopy system at the GA Health Physics laboratory.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 52 of 54 9.3 Storage & Transportation Depending on activity level, material identified as radioactive waste was stored either underwater (activated solid materials), in a shielded container, an intermediate container or in a container to be used for final transport to a disposal site.
The highly radioactive waste transported to NNSS for disposal in 2012 was stored underwater until it was ready to be placed into specially reinforced IP-1 containers. Once this waste was removed, the Mark F pool water was drained and remediation of the Mark F reactor pit began.
Waste was generally placed directly into the container that would be picked up for transport to the final disposal site or interim processing facility as the case may be. On occasion, waste was consolidated into fewer containers prior to removal.
IP-1 containers (metal boxes) were used for waste destined for the NNSS. For other waste not destined for the NNSS, 55-gallon drums or supersacks were used.
Once loaded or partially loaded, the containers were secured and stored in the TRF controlled yard. The final status survey of the TRF was completed once all waste containers were removed.
For shipments to the NNSS in Mercury Nevada, Tri-State Motor Transit was chosen from the approved carrier list NNSS maintains for transport of low-level radioactive waste to their facility. Up to five IP-1 metal boxes were loaded and secured on either an open flat bed or Conestoga trailer. As a full-service waste broker, Philotechnics maintains its own transportation network to accommodate shipments leaving the TRF.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 53 of 54 10.0 Executive Summary It was GAs intention to demonstrate in this report that carefully planned remediation efforts have systematically removed contaminated and/or activated materials from the facility. Furthermore, radiological measurements and surveys that followed NRC published guidance while using sound Health Physics practices have met approved release criteria and shown that the facility is suitable for unrestricted use. GAs conservative approach to dose assessments and statistical tests has provided defensible assurance that results presented in this report clearly show minimal risk to GA employees, the public and environment upon termination of the TRIGA licenses and release of the facility for unrestricted use. GA did not rely on statistical tests to demonstrate the TRF meets release criteria but instead remediated the facility to a level where radiological measurements and surveys results are all below DCGLs. In addition, when dose assessments were required, conservative scenarios show dose to the general public 4 times less than the 25 mrem per year TEDE prescribed by 10CFR20, Subpart E.
Based on these results, no further remediation is needed to assure protection of human health and the environment from the radiological activation and/or contamination that occurred during reactor operations at the GA TRF.
The TRIGA Reactor Facility has 2,728 m2 (29,365 ft2) of indoor and outdoor surface area licensed by the NRC under Facility Licenses No. R-38 and R-67. This square footage was divided into 16 survey units that are covered under this Final Status Survey Report.
Throughout the decommissioning process, GA has worked closely with the NRC, providing opportunities to share 3rd party radiological measurement results, several in-process inspections, split sampling of materials, side by side surveys with ORAU and complete transparency throughout remediation and final status surveys. This cooperative approach furthered GAs ability to remediate successfully the TRF to levels supporting the release of the facility for unrestricted use and license termination.
License Nos. R-38 & R-67 December 2020 General Atomics TRIGA Final Status Survey Report Page 54 of 54 11.0 References
- 1. Docket No.70-734; SNM-696: Request to Release the Non-Reactor Portion of GAs TRIGA (Building 21) Reactor Site to Unrestricted Use and Delete from License, Letter 696-4039 GA (Asmussen) to USNRC (Baker), November 2006
- 2. NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), August 2000
- 3. NUREG-1757, Vol. 2, Rev. 1 Consolidated Decommissioning Guidance: Characterization, Survey, and Determination of Radiological Criteria, USNRC Office of Nuclear Material Safety and Safeguards, September 2006
- 4. General Atomics TRIGA Reactors Facility Process Knowledge Report, GA Report GA-D22520, GA Private Data, December 1996
- 5. General Atomics TRIGA Reactor Facility Decommissioning Plan, GA Report PC-000482, Rev. 3, July 1999
- 6. General Atomics TRIGA Reactor Facility - Update of the Isotope Specific Radiological Release Criteria for the Decommissioning of the Mark I and Mark F Reactors (CAC Nos.
L53111 and L53112), Letter USNRC (Watson) to GA (Engstrom), February 2017
- 7. NUREG-1507, Rev. 1 Minimum Detectable Concentrations with Typical Radiation Survey for Instruments for Various Contaminants and Field Conditions, USNRC Office of Nuclear Material Safety and Safeguards, August 2020
- 8. TRIGA Reactor Facility Final Status Survey Plan, Rev. 2, General Atomics, February 2020
- 9. Project-Specific Plan for Confirmatory Survey Activities for the General Atomics TRIGA Reactor Facility Building G21 and Associated Land Area, San Diego, California, Oak Ridge Institute for Science and Education (ORISE), August 2019.
- 10. Update to 06/30/2019 and Application of Radionuclide Scaling Factors for Bldg. 21, GA Memo DDI:001:JSG:19, April 2019.