NOVEL UTILIZATION OF TRIGA REACTORS

Transcription

1 Preliminary Safety Analysis Report (PSAR): NOVEL UTILIZATION OF TRIGA REACTORS FOR ISOTOPE PRODUCTION (NUTRIP) MO- 99 PRODUCTION USING TRIGA REACTORS NE 267: Nuclear Reactor Safety Professor Per F. Peterson Final Report 09 MAY 2011 Ross Barnowski Joseph Miller Kevin Tirohn Nicolas Zweibaum

2 ABSTRACT This report considers the safety analysis and licensing approach for co- locating a TRIGA Reactor with a Molybdenum- 99 processing facility in order to provide Technetium- 99m used in medical diagnostic procedures. In this report we propose using the flux from a 4MW TRIGA Reactor to irradiate LEU foils to produce Molybdenum- 99. This TRIGA design is capable of producing day Ci/week, or half of the US demand. This report also considers safety concerns and risk mitigation associated with both the TRIGA reactor and the associated processing facility. Licensing basis events are identified and their associated frequencies and consequences are calculated. The proposed site for the project is North Billerica, MA where domestic capacity for 99m Tc generators is expected to be sited. A preliminary analysis of the site including seismic characterization is conducted. 2

3 REPORT OUTLINE CHAPTER 1: INTRODUCTION 12 CHAPTER 2: PLANT AND SITE DESCRIPTION 16 CHAPTER 3: DESIGN BASIS EVENTS 48 CHAPTER 4: BEYOND DESIGN BASIS EVENTS 65 CHAPTER 5: RISK ASSESSMENT AND MANAGEMENT 81 CHAPTER 6: SITE SAFETY AND SEISMIC ANALYSIS 86 CHAPTER 7: CONCLUSIONS AND RECOMMENDATIONS 94 3

4 Table of Contents ABSTRACT... 2 REPORT OUTLINE INTRODUCTION BUSINESS MOTIVATION MOLYBDENUM- 99/ TECHNETIUM- 99M ISOTOPE PRODUCTION OVERVIEW MOLYBDENUM- 99 PRODUCTION METHOD HOW WE INTEND TO MEET THE NRC POLICY ON ADVANCED REACTORS REFERENCES PLANT AND SITE DESCRIPTION REACTOR DESIGN DECISIONS GENERAL DESIGN CONSIDERATIONS TRIGA REACTOR CORE CONFIGURATION ANALYSIS OF CORE SAFETY PARAMETERS WITH FOIL INTRODUCTION Rod Worth Core Peaking Factors Foil Temperature Temperature coefficient of reactivity Conclusions MOLYBDENUM- 99 FISSION TARGET DESIGN FEASIBILITY STUDY IRRADIATION AT MCCLELLAN NUCLEAR RADIATION CENTER GENERAL DESIGN SPECIFICATIONS OF UCD MNRC SAFETY LIMITS CONCLUSION IRRADIATED TARGET PROCESSING LOCATION REQUIREMENTS TRANSITION PHASE BETWEEN IRRADIATION AND PROCESSING PROCESSING PHASE 1: DISASSEMBLY OF TARGETS PROCESSING PHASE 2: FOIL DISSOLUTION MO- 99 RECOVERY AND PURIFICATION SITE DESCRIPTION PLANT LOCATION AND SITE DESCRIPTION INTRODUCTION SIGNIFICANCE OF BROWN- FIELD SITING SITE BOUNDARY

5 POPULATION AND INDUSTRY TRANSPORTATION Highway Transportation Railway Transportation Air Traffic EVALUATION OF POTENTIAL LOCAL ACCIDENTS LOCAL METEOROLOGY General Climate Hurricanes Tornadoes Floods Geology and Seismology REFERENCES DESIGN BASIS EVENTS DESIGN BASIS EVENT DURING OPERATION OF A TRIGA REACTOR: LOSS OF COOLANT ACCIDENT (LOCA) MITIGATION ELEMENTS RADIATION LEVELS FROM THE UNCOVERED CORE CONCLUSION DESIGN BASIS ACCIDENTS FOR PROCESSING FACILITY ANTICIPATED OPERATION OCCURRENCE - HYDRAULIC TUBE TRANSFER SYSTEM FAILURE Transfer Tube Layout Overview Potential Failure Modes and Estimated Event Frequency Consequence of Transfer Tube Failure Design Safety Features DESIGN BASIS EVENT - LOSS OF FISSILE SOLUTION WITHIN PROCESSING HOT CELL Foil Processing Facility Overview Potential Failure Modes and Event Frequency Consequence of Solution Egress REFERENCES BEYOND DESIGN BASIS EVENTS BEYOND DESIGN BASIS EVENT DURING OPERATION OF A TRIGA REACTOR: MAXIMUM HYPOTHETICAL ACCIDENT (MHA) ACCIDENT INITIATING EVENT AND SCENARIO ACCIDENT ANALYSIS AND DETERMINATION OF CONSEQUENCES CONCLUSION BDBE FOR PROCESSING FACILITY - RE- CRITICALITY OF SOLUTION ACCIDENT SCENARIO DESCRIPTION ESTIMATION OF CRITICAL MASS AND ACCIDENT FREQUENCY ESTIMATION OF CONSEQUENCE OF RE- CRITICALITY BDBE Prompt Dose Calculation Delayed Dose Calculation

6 Delayed Dose from Fission Products REFERENCES RISK ASSESSMENT AND MANAGEMENT SAFETY GOALS AND SOURCES OF RISK SAFETY GOALS INTERNAL INITIATING EVENTS EXTERNAL INITIATING EVENTS CINTICHEM PROCESSING ACCIDENTS RISK FROM TRIGA OPERATION TRIGA FAULT TREE TRIGA PROBABILITIES RISK FROM MO- 99 PROCESSING PNEUMATIC TRANSFER SYSTEM FAILURE REFERENCES SITE SAFETY AND SEISMIC ANALYSIS DEFINITION AND SCOPE GEOLOGICAL ANALYSIS SURFACE AND SUB- SURFACE GEOLOGY FAULT STRUCTURE SEISMIC ANALYSIS SEISMIC HISTORY OF REGION REGIONAL SEISMIC HAZARD DETERMINATION OF SEISMIC SAFETY OF SITE IDENTIFY SAFE SHUTDOWN AND OPERATION BASIS EARTHQUAKES OTHER SEISMIC DESIGN BASES Surface Faulting Induced Floods and Water waves Soil Stability PLANT SEISMIC DESIGN BASIS REFERENCES CONCLUSIONS AND RECOMMENDATIONS APPENDIX A. SUMMARY OF CONTRIBUTIONS BY REPORT AUTHORS APPENDIX B. DETAILED CALCULATIONS APPENDIX C. COMPARISON OF MO- 99 PRODUCTION METHODS

7 List of Figures Figure 1 - Reactor Unavailability Impacts on Molybdenum- 99 Supply [2] Figure 2 - Projected Supply vs. Demand for Molybdenum- 99 [2] Figure 3 - Side and Top View of a TRIGA Reactor Figure 4: Mo- 99 Activity produced from foil irradiation vs. irradiation time. Each line represents a different number and layout of foils. 18- R was selected for this project and is the only configuration considered moving forward Figure 5 - Core Configuration Figure 6: Results of thermal analysis with two fuel channels and an adjacent target channel Figure 7: Reactivity versus temperature for benchmark core. Notice the coefficient of reactivity is cents per degree celsius Figure 8: Reactivity versus temperature for modified 18- R core. Notice the coefficient of reactivity is -.23 cents per degree celsius Figure 9 - CAD Drawing of the ANL Target Design Figure 10 - Cross- Sectional View of the ANL Target Figure 11 - McClellan Nuclear Radiation Center Figure 12 - MNRC Pneumatic Transfer System [14] Figure 13 - Pressure vs. Time for Dissolution of the Irradiated Uranium Foil with the Nickel- Foil Fission Barrier [16] Figure 14 - Zoomed aerial view of Iron Horse Park, Proposed Site for NuTRIP production facilities. The blue, green, and red bubbles indicate areas that have been identified as brown- fields by MassDEP Figure 15 - Aerial View of Iron Horse Park showing proximity to population centers and major transportation routes Figure 16 - Artist's Depiction of MNRC. The central building contains the 2 MWt TRIGA, while the perimeter fence represents the EAB Figure 17 - Railway infrastructure related to the Iron Horse site. The lines entering the north- east of the site are no longer railed

8 Figure 18 - Airports within 20 miles of proposed site (blue bubble) Figure 19 - General Climate Information for Billerica, MA [22] Figure 20 - Wind Rose Plot for Logan International Airport [23], just southeast of the Iron Horse site Figure 21 - Maximum and Average Fuel Temperature during Air Cooling Cycle for various Spray Cooling Times [3] Figure 22 - Example Farmers Curve Displaying the frequency and consequences of several classes of accident scenarios [5] Figure 23 - Axial sketch of proposed tube transfer system. Not to scale. Dimensions of Cooling Pond and TRIGA reactor not finalized Figure 24 - Sketch of proposed core layout. Each spoke represents a transfer tube, illustrating the possible radial distribution. Green = LEU foil, Red = Fuel Element, Gray = Control Rod, Blue = Reflector Figure 25 - Indication of common failure points for pressurized transfer systems Figure 26 - Hot- cell laboratory at Argonne's Alpha- Gamma Laboratory [12]. The NuTRIP hot cell facilities have a similar proposed design Figure 27 - Schematic representation of a liquid criticality accident for a 2- phase chemical process. Adopted from description of accident at Tokaimura fuel processing plant [4] Figure 28 - Hypothetical Event Tree for accident pathway leading to a re- criticality accident Figure 29 - Dose "Slide- Graph" Charts for estimating dose due to criticality accidents at various times and distances from the event Figure 30 - Trend Based on Data Shown in Table Figure 31 - Empirical Estimates for Dose Reduction Factors due to Shielding from Various Sources [8].. 77 Figure 32 - Trend in Dose Rate Data Shown in Table Figure 33 - Logic Tree for TRIGA Events. See Table 21 for Known Event Frequencies Figure 34 - Geologic map of Massachusetts from the Office of the Massachusetts State Geologist Figure 35 - Map showing postulated fault region in Massachusetts [4] Figure 36 - Historical record of New England Earthquakes from Data on the left is from the USGS while data on the right is from the Weston Geological Observatory [6]

9 Figure 37 - Seismic Hazard map for region around the Iron Horse site [8]. The site is represented by the red star Figure 38 - Seismic zones defined by the 1997 UBC. Massachusetts is a 2A zone while the MNRC is in a zone 3 [11]

10 List of Tables Table 1 - Current Molybdenum- 99 Production Reactors [2] Table 2: Calculated Keff values of the cores for different rod positions. Calculations were done with MCNP5 [10] Table 3: Peaking Factor Comparison for benchmark (mark II) and the final 18- R modified core [10] Table 4 - Activity from Impurities Following the CINTICHEM Process for HEU vs. LEU Foils Table 5 - Airport Information for Non- International Airfields within 20 miles of Iron Horse Site Table 6 - Record of Hurricanes Making Landfall in Southeast New England during the 20th Century [24] Table 7 - General Characterization of Hurricanes Table 8 - Fujita Scale for Tornado Severity Table 9 - Dose Rates on the Reactor Top after a Loss of Pool Water Accident following 2 MW Operations [3] Table 10 - Scattered Radiation Dose Rates in the Reactor Room After a Loss of Pool Water Accident Following 2 MW Operations [3] Table 11 - Scattered Radiation Dose Rates at the Facility Fence After a Loss of Pool Water Accident Following 2 MW Operations [3] Table 12 - Summary of Physical Properties for Additional Materials Present in the Uranium Target Cells Table 13 - Radiation Dose to Members of the General Public under the Most Conservative Atmospheric Conditions (Pasquill F) at Different Distances from NuTRIP Facility following Total Loss of Solution Containing 18 Foils. EAB is 10m Table 14 - Occupational Radiation Doses in the Reactor Room following the Maximum Hypothetical Accident [2] Table 15 - Radiation Doses to Members of the General Public Under the Most Conservative Atmospheric Conditions, at Different Distances from the Reactor, Following a Fuel Element Cladding Failure in Air with no Decay (MHA) [2] Table 16 - Experimental Results from Several Applications of the CINTICHEM Process in a Lab Setting at ANL [3]

11 Table 17 - k- infinity Estimation for each of the Experiments Listed in Table Table 18 - Prompt Dose Due to 5.5E15 Fission Criticality Event. No Decay Time or Shielding is Taken into Account Table 19 - Dose Rate vs. Time for 5.5E15 Fission Criticality Event at 5 ft from the Event Location. No Shielding is Taken into Account Table 20 - Dose Rate vs. Time Elapsed since Accident at EAB Table 21 - Quantitative Information Regarding the Frequency of Events that May Lead to Radiation Releases. Numbers Correspond to Figure Table 22 - Earthquake Effects and Frequencies Associated with Richter Scale Measurements According to the USGS Table 23 - List of Authors and their Contributions to this Report

12 1. INTRODUCTION 1.1. BUSINESS MOTIVATION MOLYBDENUM- 99/ TECHNETIUM- 99M Currently, there are approximately 20 million nuclear medicine diagnostic procedures each year in the United States. Technetium- 99m is used in about 85% of these procedures and has over 70 medical diagnostic applications including brain, bone, and heart scans [1]. Technetium- 99m has a half- life of only 6.1 hours and is produced by beta- decay from Molybdenum- 99, which has a half- life of 66 hours. Due to its short half- life, it is not possible to stockpile Molybdenum- 99. At present, the United States demand for Molybdenum- 99 is estimated to be day Ci/week. The 6- day Ci unit represents the activity of a shipment of Mo days after it leaves the producer s facilities, and is widely used by producers to calibrate their sale prices ISOTOPE PRODUCTION OVERVIEW Despite the high demand for medical isotopes, the United States currently has no domestic source of Molydenum- 99 and imports most of it from Canada and Europe. The current Molybdenum- 99 production reactors are listed in Table 1 along with their dates of first commissioning. TABLE 1 - CURRENT MOLYBDENUM- 99 PRODUCTION REACTORS [2] Reactor Name Location Weekly % of World Demand Date of First Commissioning BR- 2* Belgium HFR* Netherlands LVR- 15 Czech Republic MARIA Poland NRU* Canada OPAL Australia OSIRIS France

13 SAFARI- 1* South Africa RA- 3 Argentina < = Domestic production only, * = major US supplier. From OECD/NEA HLG- MR Interim Report The international Molybdenum- 99 suppliers currently operate an aging fleet of reactors, placing the United States supply at risk. The effects of the unscheduled shutdown of NRU the main isotope production reactor and world supplier, located in Canada in 2009 and into 2010 and subsequent shutdowns of HFR and BR- 2 (Europe) on the United States supply of Molybdenum- 99 are shown in Figure 1. FIGURE 1 - REACTOR UNAVAILABILITY IMPACTS ON MOLYBDENUM- 99 SUPPLY [2] During this timeframe, there were an estimated 50,000 procedures missed per day due to supply shortages. The lack of Technetium- 99m has been estimated to have cost 6000 years of life per day for patients with history of coronary disease (see Appendix B). With the upcoming decommissioning of many isotope production reactors and the expected demand growth, supply will be unable to keep pace with demand as seen in Figure 2. Each of the sharp drops in the figure represents the scheduled decommissioning of the NRU, BR- 2, and HFR respectively. 13

14 FIGURE 2 - PROJECTED SUPPLY VS. DEMAND FOR MOLYBDENUM- 99 [2] MOLYBDENUM- 99 PRODUCTION METHOD Molybdenum- 99 is produced by irradiating Uranium- 235 foils with neutrons to induce fissions. This method has been calculated to have a yield a hundred times larger than accelerating neutrons into Molybdenum- 98 (see Appendix C). The irradiated foils are then chemically processed to extract Molybdenum- 99. A TRIGA reactor is proposed for the irradiation of LEU foils due the low capital cost, very high inherent safety, vast operational experience, and sufficient neutron flux. Additionally, the open pool design and pneumatic/hydraulic transfer tubes allow for online manipulation of the core contents. The fuel is composed of UZrH, a unique hydrogenated ceramic which has 3-4 times the burnup tolerance of conventional reactor fuel, as well as excellent fission product retention properties. This provides better supply stability due to the increase in capacity factor and reduces the amount of waste generated. The fuel composition also makes the reactor inherently safe due to the prompt negative temperature coefficient of reactivity. The reactor design also has pulsing capability and can be pulsed to powers in the gigawatt range, causing an increase in neutron flux and higher production rate of Molybdenum- 99. However, the pulsed mode of operation is not considered in this PSAR; only steady- state operation is evaluated. A single TRIGA reactor is capable of producing half of the United States demand for Molybdenum More details are provided in

15 1.2. HOW WE INTEND TO MEET THE NRC POLICY ON ADVANCED REACTORS The TRIGA reactor will be licensed under the existing NRC framework since the power will be under 10 MW thermal. The reactor will be licensed under 10 CFR 50 Part 21 as a class 104 facility [3]. The fuel temperature is limited in both steady- state and pulse mode operation to 1000 C and 930 C respectively. The nine credible accidents for research reactors identified by NUREG are the maximum hypothetical accident (MHA), insertion of excess reactivity, loss of coolant accident (LOCA), loss of coolant flow, mishandling or malfunction of fuel, experiment malfunction, loss of normal electrical power, external events, and mishandling or malfunction of equipment [4]. The reactor will be sited on a brown field in North Billerica, Massachusetts in order to reduce concern of further site contamination and follow the exclusion area boundary conditions set forth in 10 CFR [3]. Transport of Molybdenum- 99 will be done in accordance with 10 CFR 71 [3]. Criticality safety of the reactor complies with the requirements in 10 CFR 50, while criticality safety of the processing plant is largely derived from 10 CFR 70 [3]. Further considerations for the handling of aqueous special nuclear materials are found in SECY [5] REFERENCES [1] Idaho State University. (2009, February 10). ISU Headlines. Retrieved from Idaho State researchers experimenting with using accelerators to produce medical isotopes: [2] OECD/NEA. (2010). The Supply of Medical Radioisotopes - Interim Report of the OECD/NEA High- level Group on the Security of Supply of Medical Radioisotopes. OECD [3] U.S. Nuclear Regulatory Commission. (2011). U.S. NRC Regulations: Title 10, Code of Federal Regulations. [4] U.S. Nuclear Regulatory Commission. (1996). NUREG- 1537: Guidelines for Preparing and Revewing Applications for the Licensing of Non- Power Reactors. [5] Borchardt, R. (2009). SECY Licensing of a Babcock & Wilcox Medical Isotope Production System. NRC Commission Papers. 15

16 2. PLANT AND SITE DESCRIPTION The objective of this section is to discuss the co- located plant and site description of the TRIGA reactor and Mo- 99 production facility. Coupling attributes of the design will be discussed. Furthermore, a preliminary non- seismic characterization of the site is conducted. Seismic site characterization can be found in Chapter REACTOR DESIGN DECISIONS In this first part, decisions concerning an optimized design for the TRIGA reactor used for Mo- 99 production are detailed GENERAL DESIGN CONSIDERATIONS Based on the preliminary research on existing reactors [1], several conclusions about the general structure for the construction of a Mo- 99 production reactor were made. The preliminary research on other medical isotope reactors yielded a general design criterion for the size of the core itself. The NRU reactor (Chalk River, Canada), the largest of any of the cores researched, both in terms of heat production (135 MW) and physical dimension, has a core size about 27 m 3 [2]. The other two reactors, OPAL (Australia) and the High Flux Reactor (HFR) at Petten, Netherlands, have core dimensions similar to the size of a washing machine. Following 10 CFR 50 [3], any reactor with a rated power over 5 MW is required to have a secondary cooling loop. In order to avoid the complication of the design and implementation of the cooling loop, a core with a rated power less than 5 MW is pursued. A simple dry cooling tower with a heat exchanger to the primary coolant loop is then sufficient to extract heat from the core. Therefore, in order to attain a rated power less than 5 MW, the fuel elements are expected to be similar in dimension to those listed for the HFR (8x7cm along the base and nearly 1m tall [4]) as opposed to the larger NRU reactor (about 3m wide by 2.5m tall [2]). For the core housing structure, an open- pool design is chosen due to the ease of unloading and reloading the targets. The HFR s target loading cycle is 11 days long [4]. Assuming our loading cycle is on a similar time frame, an open pool design with no closed vessel will facilitate the loading/unloading procedure. Also, the water provides significant neutron shielding minimizing the amount of concrete and other structural material used for neutron shielding. The reactor is designed to be a thermal reactor. The designs for all of the existing medical isotope production reactors utilize a thermal spectrum, so there is a wide knowledge base for the use of a 16

17 thermal spectrum to generate medical isotopes. Furthermore, the design of a fast reactor is not necessarily conducive to Mo- 99 production as data for the fission spectrum yields for fast fission are far less complete than for thermal fission. Furthermore, the designs for thermal spectrum research reactors are readily licensable in the US. No licensing procedure has been established for designs for fast- spectrum reactors (such as the sodium cooled fast reactor design). Therefore, a thermal neutron spectrum is utilized in the reactor design. Finally, as discussed above, the reactor is a low- power (< 5 MW) reactor so that the design of a complex secondary loop can be avoided, and a dry cooling tower is sufficient to extract heat from the primary coolant loop. Also, the reactor will be licensed under 10 CFR 50 Part 21 as a class 104 facility for the purpose of producing medical isotopes [3]. As such, no secondary turbine loop will be applied to utilize the waste heat for power production TRIGA REACTOR The concept of a TRIGA (Training, Research, Isotope, General Atomics) reactor was developed in the 1950 s by General Atomics (GA). FIGURE 3 - SIDE AND TOP VIEW OF A TRIGA REACTOR A TRIGA reactor is ideal for the production of medical isotopes for several reasons. GA's TRIGA reactor is the most widely used non- power nuclear reactor in the world. GA has sold 66 TRIGA reactors, which are in use or under construction at universities, government and industrial laboratories, and medical centers in 24 countries. GA's reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, nondestructive testing, basic research on the properties of matter, and for education and training [5]. Because of this, the operational history is well documented. This documentation facilitates in the benchmarking process for any reactor design. 17

18 With this many reactors operating globally, it is also indicative of the amount of success that TRIGA reactors have during operation. The second benefit of TRIGA reactors is their inherent safety. Low- enriched, long- lifetime uranium zirconium hydride (UZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, ruggedness, dependable performance, economy of operation, and its acceptance worldwide. The large prompt negative temperature coefficient of reactivity characteristic of UZrH fuel results in safety margins far above those achieved by any other reactor fuel. Large reactivity insertions are readily accommodated and are routine operation for some applications. Inadvertent reactivity insertions have been demonstrated to produce no fuel damage in TRIGA cores. Power coast- down from full power after loss of forced flow cooling (and resultant power scram) has been demonstrated to be a very benign event with the reactor immediately available to return to full power [5]. Long core life of the TRIGA fuel results from the fact that a large amount of uranium can be readily accommodated in the fuel matrix, occupying a relatively small volume fraction of the mixture. Major operating cost as well as total fuel cycle cost savings result from the much longer core lifetimes resulting from the higher Uranium loading in TRIGA fuels compared to competing fuels. The long fuel cycle times for UZrH fuel also result in the greatest possible operational flexibility for the system in that reactor shutdowns can most always be determined by user requirements rather than fuel cycle requirements. The UZrH material also has fission product retention capabilities far superior to competing research reactor fuel. Aluminum clad plate type fuel melts at about 650 C, releasing essentially 100% of the volatile fission products. At this same temperature, UZrH retains about 99.9% of these fission products even with the rugged clad removed [5]. Fuel design options include low density LEU fuels containing 8.5 wt% of uranium, to high density fuels containing 45 wt% uranium with burnable poisons. For our design, LEU fuel at wt% enrichment was selected in order to meet the goals of the U.S. Department of Energy s Reduced Enrichment for Research and Test Reactors (RERTR) program [6]. When irradiating targets in a reactor, it is important for the operators to have easy access to the core. Because of this, an open pool reactor was considered an ideal choice of reactor for this project. This was another condition that was filled by the TRIGA design. The large amounts of available documentation, the inherent safety of the core, and the open pool design were all enticing incentives when making the decision to use a TRIGA reactor for the production of medical isotopes. From business perspective, the perceived ease of licensing is also a valuable asset when proposing this design. All of these properties led to the selection of the TRIGA reactor. 18

19 CORE CONFIGURATION The Bangladesh 3 MW TRIGA Mark II reactor [7, 8, 9] was chosen as a model for preliminary design analysis of the medical isotope reactor. Design considerations demand a relatively high power TRIGA reactor. This reactor uses a hexagonal fuel arrangement with 95 fuel pins and 5 fuel- follower control rods. A prismatic design was desirable due to the larger power density attainable in a hexagonal core layout. The initial core design also easily allowed for replacement of center graphite elements with molybdenum targets. This made the Bangladesh TRIGA an ideal core to model to investigate feasibility of molybdenum production within the constraint of existing operating limitations. The goals of the design process were to maximize the production of molybdenum in the LEU targets, while still keeping as much of the original core intact to allow for an easier licensing process. It is known that the highest molybdenum production rates will be produced with the highest fission rates and the targets will therefore initially be placed in the core regions with the highest thermal flux. The decision was made to attempt meeting half of the United States demand on a per week basis, i.e day Ci of Mo- 99. The estimation of one day for back end processing and shipping limits the reactor to a six day cycle for target burn. It was calculated that equilibrium production rates would be reached after approximately 12 days and would from then on remain at equilibrium. Since diminishing returns would not be reached it was decided to use the longest possible cycle length based on original criteria. The cycle used for total molybdenum production was a six day cycle based on previous criteria. Figure 4: Mo- 99 Activity produced from foil irradiation vs. irradiation time. Each line represents a different number and layout of foils. 18- R was selected for this project and is the only configuration considered moving forward. shows the Mo- 99 production curve as a function of irradiation time for several different core configurations. The configurations are labeled based on the number and locations of the targets in the core. The 18- R configuration is selected based on the fact it results in the highest yield of Mo

20 Molybdenum-99 Activity (Ci) R 7-R 12-R 12-O 6-O 13-O 13-R 18-R Irradiation Time (Days) FIGURE 4: MO- 99 ACTIVITY PRODUCED FROM FOIL IRRADIATION VS. IRRADIATION TIME. EACH LINE REPRESENTS A DIFFERENT NUMBER AND LAYOUT OF FOILS. 18- R WAS SELECTED FOR THIS PROJECT AND IS THE ONLY CONFIGURATION CONSIDERED MOVING FORWARD. The final design was determined using the maximum total production of molybdenum. Using this criterion, the configuration presented in Figure 5 yielded the most molybdenum. It consists of 18 LEU targets placed at the center of the core, where the neutron flux is highest (~ neutrons/cm 2 s). FIGURE 5 - CORE CONFIGURATION 20

21 ANALYSIS OF CORE SAFETY PARAMETERS WITH FOIL INTRODUCTION The benchmark TRIGA Mark II design contains many inherent safety features and easily meets all the licensing requirements for a class 104 facility. However, for this project, the graphite reflectors in the central part of the benchmark core are removed, and replaced by the fissile foils within target assemblies. The graphite rods in the center of the core will tend to smooth out the thermal spectrum and increase the thermal flux in the center of the core. By removing these rods, the flux distribution in the center of the core is changed. Furthermore, the rods are being replaced by fissile foils, which may contribute to the reactivity of the core. In order to assure these changes don t change any of the safety characteristics of the TRIGA core, some of the safety aspects of the core are analyzed. While design analysis falls outside of the scope of the course, pre- existing work has been done to analyze the effects of the design modifications [10]. A brief summary of some of the analysis of the design changes is given below. Full detailed analysis can be found in [10] ROD WORTH Table 2 shows the MCNP calculation of the criticality eigenvalue for fully inserted and fully extracted control rod positions in both the benchmark and the modified cores. Notice that there is less excess reactivity in the modified core case, although the rod worth increases slightly. This data suggests that the insertion of the foils in the core does not diminish control rods ability to regulate the reactivity of the core. TABLE 2: CALCULATED KEFF VALUES OF THE CORES FOR DIFFERENT ROD POSITIONS. CALCULATIONS WERE DONE WITH MCNP5 [10]. Control Benchmarked 18- R configuration Rod Position Configuration Fully ± ± Withdrawn Fully ± ± Inserted CORE PEAKING FACTORS One of the most important comparisons that can be made is that of peaking factors. The goal was to obtain a core peaking factor that was lower or comparable to that of the benchmark reactor. The 18- R case accomplishes this, lowering the total peaking factor by 30.3%. The final value obtained is , compared to the benchmark value of This decrease is somewhat due to a drop in the hot channel (HC) peaking factor that was a result of removing the fuel elements from the 3 rd ring. The other 21

22 major contribution is from a large decrease in the radial peaking factor. The peaking factors between the 18- R and the benchmarked case are compared below in Table 3. TABLE 3: PEAKING FACTOR COMPARISON FOR BENCHMARK (MARK II) AND THE FINAL 18- R MODIFIED CORE [10] Peaking Factors Type Benchmark Final Design HC Axial Radial Total FOIL TEMPERATURE An additional analysis was performed on a single channel of fuel elements with an adjacent target. A model reflecting the dimensions of the targets was constructed in Solidworks and the material definitions for uranium were added to the library. The power generated by the fuel elements and the uranium in the targets was ascertained from the MCNP simulations of this updated core. Because the hot channel was adjacent to one of the targets, 52 kw was used applied to the fuel cells. MCNP showed that the target was producing 21.2 kw. The results of this analysis are displayed in Figure 6 below. FIGURE 6: RESULTS OF THERMAL ANALYSIS WITH TWO FUEL CHANNELS AND AN ADJACENT TARGET CHANNEL 22

23 Figure 6 shows that the heat produced by the target cell creates a negligible temperature effect on the target channel. This is because the target is electroplated to layers of nickel, which is inside layers of aluminum. The thin layers of uranium and nickel conduct heat well enough to discharge the heat to the aluminum, which is an excellent thermal conductor. As can be seen in Figure 10, the target is a hollow cylinder. This allows for coolant flow on both the inner and outer surfaces of the target. Coupled with the high thermal conductivity of aluminum, this led to the negligible temperature gradient across the target (ΔT < 2 C). The results show that the power produced in the foil is not sufficient to raise the coolant temperature in adjacent cells. The coolant in the target cell provides more than adequate heat rejection for the heat generated by fission in the foil. Furthermore the minimal temperature gradient across the foil suggests the chances of foil failure and rupture due to thermal stress are low TEMPERATURE COEFFICIENT OF REACTIVITY Arguably the most important safety advantage of the TRIGA reactor is the prompt negative temperature coefficient of reactivity inherent to the UZrH fuel. The foils that are being added to the center of the core are pure LEU foils thus they do not share the inherent prompt negative temperature coefficient. MCNP analysis was done to evaluate the reactivity as a function of the fuel temperature to ensure that the TRIGA core maintains its robustness against reactivity transients. Figure 7 shows a plot of reactivity versus temperature for the benchmark core, while Figure 8 shows a plot of reactivity versus temperature for the modified 18- R core. Reactivity ($) y = - 0,0168x + 5,6433 R² = 0,98249 Temperature (C) FIGURE 7: REACTIVITY VERSUS TEMPERATURE FOR BENCHMARK CORE. NOTICE THE COEFFICIENT OF REACTIVITY IS CENTS PER DEGREE CELSIUS. 23

24 0,5 Reactivity ($) 0-0, , , y = - 0,0023x + 0,6765 R² = 0,9601 Temperature (C) FIGURE 8: REACTIVITY VERSUS TEMPERATURE FOR MODIFIED 18- R CORE. NOTICE THE COEFFICIENT OF REACTIVITY IS -.23 CENTS PER DEGREE CELSIUS While the temperature coefficient of reactivity becomes less negative for the 18- R case, it is still prompt negative. Thus the robustness of the core against reactivity insertion accident scenarios is maintained CONCLUSIONS The analysis shown in section presents a strong preliminary case that the safety features of the TRIGA reactor are not diminished by the addition of the LEU foils. The rod worth and peaking factors both improved over the benchmark core, while the temperature coefficient of reactivity remained prompt negative, although to a lesser degree. The feasibility study discussed in section 2.2 is required to garner more safety data before design finalization can occur MOLYBDENUM- 99 FISSION TARGET DESIGN The target that was selected for design analysis was first developed at Argonne National Laboratory (ANL) as part of the Reduced Enrichment in Research and Test Reactors (RERTR) program [6]. This target had a number of important characteristics that were consistent with our goals: LEU target material, excellent thermal properties, and compliance with current back- end CINTICHEM processes (detailed in 2.3) to meet molybdenum purity standards. The ANL target design, shown in Figure 9 with a cross- sectional view in Figure 10, is an annular target that consists of concentric aluminum 3003 cylinders that surround a ~125μm LEU target foil. The LEU target foil is a 99% uranium foil that has been enriched to 19.75% Uranium The manufacturing 24

25 process for these foils is a low- tech cooling- roll casting process (consistent with the RERTR program objectives) demonstrated by researchers at the Korea Atomic Energy Research Institute [11]. After fabrication, the foil is electroplated with a ~15μm nickel fission barrier in order to prevent diffusion coupling to the inner aluminum cylinder. Diffusion coupling of the target foil to the aluminum cylinders prohibits separation for the back- end chemical processing of the target. Electroplating the nickel barrier onto the uranium foil also creates an excellent conduction barrier for heat rejection. The target cylinder is inserted via a smaller diameter aluminum 6061 insertion rig that allows for multiple targets to fit inside of a single irradiation position within the reactor (dependent upon the uranium foil length). This design has optimum thermal properties as heat rejection from the uranium fission power is across two water barriers. Annular Aluminum Target Target Insertion Rig FIGURE 9 - CAD DRAWING OF THE ANL TARGET DESIGN FIGURE 10 - CROSS- SECTIONAL VIEW OF THE ANL TARGET 25

26 Currently, major producers of Mo- 99 produce it using an HEU target coupled with a CINTICHEM chemical extraction process. Using LEU in place of HEU requires that approximately 5 times the amount of uranium be used in order to produce the same amount of molybdenum from Uranium- 235 fission (assuming LEU at 20% enrichment). The United States Pharmacopeia (USP) and the Food and Drug Administration (FDA) have strict standards for molybdenum purity, specifically regulations on the gamma- ray and α- particle emitting impurity concentrations. Several design feasibility studies have been performed upon this target design analyzing the use of the current CINTICHEM process for LEU Molybdenum- 99 targets. The results of these studies are that the CINTICHEM process needs little modification and that the molybdenum purity is not compromised by the excess plutonium present from Uranium- 238 transmutation (Table 4) [12]. TABLE 4 - ACTIVITY FROM IMPURITIES FOLLOWING THE CINTICHEM PROCESS FOR HEU VS. LEU FOILS Reference Target HEU (93% U- 235) LEU (19.8% U- 235) Mo- 99 yield [Ci] Total U (U- 235) [g] 16 (15) 94 (18) Pu- 239 [μci] U- 234,235,238 [μci] Total α [μci] FEASIBILITY STUDY IRRADIATION AT MCCLELLAN NUCLEAR RADIATION CENTER Because of the large number of TRIGA reactors already operating in the US, the first phase for licensing our reactor will be the irradiation of LEU foils in the existing TRIGA reactor of McClellan Nuclear Radiation Center (UC Davis, California). The UC Davis McClellan Nuclear Research Center (UCD MNRC) is owned and operated by the University of California, Davis. The UCD MNRC was originally developed by the US Air Force to detect low- level corrosion and hidden defects in aircraft structures using neutron radiography. Since then, UCD MNRC service has expanded to include computer tomography (three- dimensional neutron radiography), silicon doping, isotope production, neutron activation analysis, and radiation effects testing. There is capability to move materials and parts into the central core facility, and locations adjacent to the core while the reactor is operating [13]. 26

27 GENERAL DESIGN SPECIFICATIONS OF UCD MNRC The reactor, which began operation in 1990, is the newest research reactor in the United States. It is also the highest power TRIGA reactor in the United States, rated at 2 MW in steady state, and can pulse to approximately 1000 MW for 20 milliseconds. The McClellan TRIGA is a heterogeneous, tank- type reactor. The core is immersed in highly purified water in an open aluminum tank that holds approximately 26,500 L of water. The tank is surrounded by concrete. The core is cooled by natural convection flow. The coolant/moderator is light water, and the reactor core is reflected by light water or graphite. The reactor coolant is circulated through an external heat removal and purification system. The reactor facility includes the space next to the reactor core, a pneumatic transfer system, beam tubes and irradiation bays for larger material. The McClellan fuel design is similar to that used by other NRC- licensed TRIGA reactors, except that the top and bottom end fittings were modified to enhance coolant flow. The uranium is enriched to less than 20 % in U The reactor exhibits a large prompt negative temperature coefficient typical of all TRIGAs. Reactivity is controlled by six control rods. The McClellan TRIGA reactor is similar to 19 other TRIGA research reactors licensed to operate by the NRC. The instruments and controls are similar to the newer non- power TRIGA reactors licensed by the NRC. FIGURE 11 - MCCLELLAN NUCLEAR RADIATION CENTER 27

28 The standard features of McClellan TRIGA reactor compared to other operating TRIGA reactors, and its availability for LEU foils irradiation, make it an ideal choice for preliminary licensing analysis. The use of this reactor in the feasibility study phase will allow NRC reviewers to assess: safety modifications due to the insertion of LEU foils in the core, procedure safety while continuously inserting and removing the foils into/from the core, equipment reliability for moving the irradiated foils SAFETY LIMITS The limiting criterion for safety is the assurance that the fuel cladding will remain intact and not allow the escape of fission products. Therefore, for purposes of the safety analysis, a safety limit is proposed for the temperature of the fuel that will not result in failure of the fuel cladding as a result of internal pressure or clad melting [14]. As the temperature of a fuel element increases, the internal pressure inside the fuel cladding also increases because of the presence of air, fission product gases, and hydrogen from the disassociation of hydrogen and zirconium in the fuel moderator with hydrogen being the most important contributor to the internal pressure. If the temperature becomes high enough, the stress on the cladding as a result of the internal pressure can exceed the ultimate strength of the stainless steel cladding, and the cladding will fail, releasing fission products from the fuel. The ultimate strength of the cladding material is also temperature- dependent and decreases with increasing temperature. A safety limit of 930 C on fuel temperature (for cladding temperature above 500 C) is proposed, where internal pressure is slightly less than the ultimate cladding strength. For the pulse mode of operation, a safety limit of 1100 C is proposed (for clad temperature less than 500 C). During a pulse, the clad temperature is well below the fuel temperature. The cooler clad temperature results in a higher ultimate stress for the stainless steel cladding. This allows a higher internal pressure to be present inside of the fuel cladding, which allows a higher fuel temperature safety limit. Also, the diffusion of hydrogen inside the fuel element reduces the peak pressure inside the fuel element as contrasted with that predicted at equilibrium at peak fuel temperature. This also allows for a higher safety limit for fuel temperature during pulsing. In accordance with 10 CFR 50.36, a limiting safety system setting (LSSS) is proposed, designed to ensure that automatic protective action (reactor shutdown) will occur in sufficient time to prevent safety limits from being exceeded. The value used to set the reactor instrumentation is a fuel temperature of 750 C. The instrumented fuel element is located in the analyzed peak power location of the operational core. This temperature provides a significant safety margin to allow for any difference between true and measured values (estimated to be only a few degrees). During actual operation at 2 MW, the maximum fuel temperatures remain below 500 C; therefore, operating experience would indicate that an additional safety margin exists. 28

29 Furthermore, the limit on the minimum shutdown margin ensures that the reactor can be safely shutdown from any operational configuration, even if the highest worth control rod remains stuck out of the core. A minimum shutdown margin of $0.50 ensures that the reactor can be shut down and remain shut down. This minimum shutdown margin must be met with the reactor in any core condition, with the most reactive control rod assumed to be fully withdrawn, and with the absolute value of all moveable experiments in their most reactive condition or $1.00, whichever is greater. The value of $0.50 is a standard value for shutdown margin. The limit of $1.00 for the absolute value of all moveable experiments in their most reactive condition is expected to be met when irradiating LEU foils in the reactor, since the total excess of reactivity from the insertion of fresh LEU foils has been calculated to be $0.94 [10] CONCLUSION The TRIGA reactor at McClellan Nuclear Radiation Center has already been licensed for operation by the NRC. Its main features will be used in our reactor after assessing the feasibility of Mo- 99 production from LEU foils irradiated in that reactor, which will simplify the licensing process. The main concern in our utilization of the reactor is the reactivity insertion due to the presence of LEU foils in the core. Preliminary analysis has shown that this positive reactivity insertion should not exceed $0.94, thus consistent with the conservative limit of $1.00 reactivity insertion (for the absolute value of all moveable experiments in their most reactive condition) in the NRC recommendations to MNRC IRRADIATED TARGET PROCESSING In this part, the series of processes, from the end of LEU foil irradiation to the production of Mo- 99, is described LOCATION REQUIREMENTS One of the key limitations to processing capacity is the location requirements of processing facilities they should be located close to the reactor. In the current situation (without any US- based production), irradiated targets have to be shipped to the processing facility in secure containers that weigh approximately four tons. These containers can only be transported at reasonable costs and under current regulations via road transportation (10 CFR 71). In order to minimize the decay of the Mo- 99 that would occur during transportation, the processor should be located as close to the reactor as 29

30 possible. Recognizing the time required for transportation, 1000 km (on land) is considered to be the maximum acceptable distance for transporting irradiated targets from the reactor to the processing facility (with much shorter distances being preferred) [15]. Transportation via roads is required, as air transportation would not be cost effective and would require dedicated cargo airplanes. In addition, there are no containers that are widely licensed for transporting irradiated targets via air and it is expected that it would be a challenge to license such containers. Currently there is no air transportation of irradiated targets for the production of Mo- 99. In terms of limitations, if the processing facility was further away than 1000 km, the decay of the Mo- 99 during transportation time would create a meaningful loss of product. This would result in an overproduction of material, resulting in an increase in radioactive waste volumes and increased waste management costs, increased use of valuable reactor fuel and an increase in safety risks as more radioactive material is required to be handled and transported than would otherwise be necessary. In addition, the further away that processing facilities are located from reactors, the more complicated the transportation logistics and regulatory requirements. For example, crossing multiple jurisdictions requires approval from all the jurisdictions that are transited. There is also an increased risk of delays (regulatory delays, etc.). These arguments are a strong incentive to co- locating the reactor where LEU foils are irradiated and the processing facility itself, which is the purpose of our project TRANSITION PHASE BETWEEN IRRADIATION AND PROCESSING After being irradiated for 6 days, the foils are cooled for 8 hours in a dedicated tank to allow for the decay of the most short lived fission products that built up in the irradiated uranium. In order to limit the amount of radioactive material being handled at a given time, a limit to the number of LEU foils that can be present in the cooling tank is limited. After further analysis, this number could be increased in order to accommodate foils coming from more than one reactor. Foils are moved from the irradiation site to the cooling tank through a pneumatic transfer system. The McClellan Nuclear Radiation Center Pneumatic Transfer System is used as a design reference. It was designed to accommodate the transfer of individual small specimens into and out of the reactor core (Figure 12). 30

31 FIGURE 12 - MNRC PNEUMATIC TRANSFER SYSTEM [14] Specimens are placed in a small polyethylene holder or 'rabbit,' which in turn is placed into the receiver. The rabbit travels through aluminum tubing to the terminus at the reactor core centerline. After the irradiation is completed, it returns along the same path to the receiver. Directional air flow moves the rabbit between the receiver and terminus. A blower assembly provides air flow in the system, and a solenoid valve directs air flow. Controls to operate the blower and solenoid valve are mounted to the wall adjacent to a fume hood that contains the receiver. The air flow design uses a blower to evacuate air, allowing atmospheric air pressure to push the rabbit into position, either at the irradiation terminus or at the receiver. This approach tends to decrease the likelihood of fragments from a shattered rabbit becoming trapped in the terminus. The rabbit- based pneumatic system at the MNRC serves as a design basis for a similar transfer system for the LEU target assemblies. The only radiation concern identified related to the use of this system is the production of Ar- 41 in the section of the pneumatic transfer system that is located in the reactor core. During operation of the transfer system, air containing very small amounts of Ar- 41 is exhausted from the system through a HEPA filter to the facility stack. Previous operation has not shown any significant increase in Ar- 41 releases, as measured by the stack monitor. Therefore, the Ar- 41 from the pneumatic transfer system is not considered to be a measurable contributor to the doses associated with MNRC operations, and the same conclusions can be drawn when reproducing the MNRC pneumatic transfer system design in our own reactor. 31

32 The cooling tank, where the foils remain for 8 hours after irradiation for cooling, needs further design. It serves the same goal as spent fuel pools in power reactors, and should therefore be designed in a similar way. However, no re- criticality risks are involved here, since the tank will be designed to accommodate few foils only, with the total amount of Uranium in the pool much less than that present in a single fuel rod. The only risk is that of foil failure and release of radioactive material if the foil s temperature were to exceed material integrity limits, which will be prevented by constant monitoring of the water level and temperature in the cooling tank PROCESSING PHASE 1: DISASSEMBLY OF TARGETS After cooling for eight hours, the targets are moved to a hot cell at the processing facility for disassembly. Disassembly consists of cutting off both ends of the target and pushing the inner tube, and uranium foil, from the outer tube. A slight taper in the cylinders facilitates this operation. Preliminary experiments have shown that the target with an aluminum inner tube can be extracted, though with some difficulty, because the combination of high temperature from decay heat and the large thermal expansion coefficient for aluminum result in a tight mechanical fit. A larger taper might solve this problem. All previous experiments [16] have shown that foils with fission fragment barriers do not bond to zirconium tubes. In particular, uranium foils with nickel fission- barriers are easily handled PROCESSING PHASE 2: FOIL DISSOLUTION Following target disassembly, one target per week is processed. The process currently used for Mo- 99 extraction from HEU foils is the CINTICHEM process, developed by Cintichem Inc. Investigation about the feasibility of Mo- 99 extraction from LEU foils using this same process has been performed. One foil is dissolved in 40 ml of 6M nitric acid in a closed dissolver. The low dissolver weight (0.8 kg) allows for easy handling by remote manipulators in a shielded- cell facility. Figure 13 shows the dissolution profile (pressure buildup vs. time) for the irradiated uranium foil with the nickel- foil fission barrier. The pressure peak is due to heating of the constituents caused by heat released in the reaction between the metals and nitric acid. 32

33 FIGURE 13 - PRESSURE VS. TIME FOR DISSOLUTION OF THE IRRADIATED URANIUM FOIL WITH THE NICKEL- FOIL FISSION BARRIER [16] Once the metals are dissolved, the temperature (and, consequently, the pressure) is decreased to that controlled by the heating unit 103 C. Two moles of NO gas are released for every mole of uranium metal dissolved and 2/3 mole for every mole of nickel. The next steps in processing are to allow the dissolver to cool and then to evacuate the dissolver gases by connecting it to an evacuated cold trap (liquid nitrogen) MO- 99 RECOVERY AND PURIFICATION The recovery and purification of an LEU foil involves 20 ml of a KMnO 4 - reagent solution. This step of the process is proprietary information of Cintichem Inc. However, since the process is currently used by producers in Canada, its licensing should not be an issue. Several studies have set out to determine whether the CINTICHEM process will be equally viable for the processing of LEU foils. The results of a study done at Argonne National Laboratory have shown that the purity of the Mo- 99 extracted using the CINTICHEM process on LEU foils is not degraded compared to HEU foils. In other words, the Mo- 99 produced from the LEU foils is expected to meet the same purity standards set by the FDA as the HEU- produced Mo- 99 conforms to. Similarly, the efficiency of Mo- 99 extraction from LEU foils is not reduced when compared to HEU foils (about 90-95% of the Mo- 99 present in the foil is 33

34 recovered). Finally, the results of a National Academy of Sciences study [1] estimates that the total cost increase of extracting Mo- 99 from LEU foils as opposed to HEU is estimated to be less than 10%. Therefore, the CINTICHEM process is both physically and economically viable for use with LEU foils. One factor that must be considered is what will happen to the waste from the CINTICHEM process. Since the foils are LEU, the plutonium production in the foils is expected to be higher in LEU foils than in HEU foils of the same dimension (placed in the same flux field). A study needs to be conducted to ascertain whether special procedures need to be developed for the disposal of this waste SITE DESCRIPTION PLANT LOCATION AND SITE DESCRIPTION INTRODUCTION The proposed site for the reactor and processing facility is Iron Horse Park, a brown- field 553 acre industrial park in North Billerica, Massachusetts. The site was selected for several reasons, namely proximity to proposed Tc- 99m generator manufacturers; a high volume of regional hospitals and Tc- 99m demand; and the brown- field nature of the site. North Billerica is also home to the offices of Lantheus Medical Imaging, one of two companies with plans to construct Tc- 99m generator manufacturing facilities in the US. While preliminary, current siting research suggests that the Tc- 99m generator manufacturing facility from Lantheus Medical Imaging will be located near North Billerica [17]. Iron Horse Park itself is a 553 acre industrial park that has previously played host to railroad maintenance facilities and open area storage, landfills, and waste- water lagoons. The site has been declared a brown- field by the Massachusetts Department of Environmental Protection (MassDEP) due to ground water and surface water contamination by asbestos, heavy metals, and other inorganic chemicals [18]. Many waste- water lagoons and other sources of contamination have been cleared, while it is estimated there are still six significant sources of contaminants on the site. As such, an extensive monitoring and ground- water diversion plan is already in place for the site and overseen by the MassDEP. All active (non- storage) industrial activities including railroad maintenance and waste- water treatment have ceased. The site is declared closed to further industrial development without entering a site- clearing agreement with MassDEP. 34

35 FIGURE 14 - ZOOMED AERIAL VIEW OF IRON HORSE PARK, PROPOSED SITE FOR NUTRIP PRODUCTION FACILITIES. THE BLUE, GREEN, AND RED BUBBLES INDICATE AREAS THAT HAVE BEEN IDENTIFIED AS BROWN- FIELDS BY MASSDEP FIGURE 15 - AERIAL VIEW OF IRON HORSE PARK SHOWING PROXIMITY TO POPULATION CENTERS AND MAJOR TRANSPORTATION ROUTES The site itself is roughly five miles SSW of Lowell Massachusetts (population: 104,400 in 2009 [19]), and within 8 miles of several large highways. I- 93 is approximately 6 miles to the east, while I- 495 (also an interstate highway) is roughly 2.5 miles to the north. The site is relatively flat with an elevation above 35

36 sea level ranging from ft. The site is roughly 27 miles inland from the Atlantic coast, 25 miles NNW of Boston. The significance of the population density and nearby highways is discussed in the subsequent population, industry, and transportation section ( & ) SIGNIFICANCE OF BROWN- FIELD SITING As previously noted, the Iron Horse Park has been identified as a brown- field site by MassDEP due to significant ground and surface water contamination from previous industrial activities. As such, the MassDEP has invested $5.4 million in ground- water isolation and monitoring programs to ensure the contaminated water is isolated from sources used for drinking water. The existing isolation and monitoring infrastructure will be advantageous in the monitoring of the construction and operation of the NuTRIP facility. The CINTICHEM reactor and associated processing facility operated by Hoffman & LaRoche had a series of unauthorized radioactivity releases in The radioactivity egresses included I- 131 and low levels of tritium contamination of the nearby Indian Kill Reservoir in Tuxedo, NY. The releases were small in magnitude and diluted in the reservoir as to be well below the limits set forth for drinking water in 40 CFR 141 by the EPA [20]. Political pressure from the resulting escape of radioactive material from the plant is thought to have contributed to the decommissioning of the site. NuTRIP will take advantage of the existing monitoring and isolation infrastructure in place at Iron Horse Park to ensure that a similar situation does not occur with this facility. In accordance with NRC guidelines, the isolation program on the site and the restricted- use nature of water derived from the site guarantees that the limit for the maximum permissible concentration for release will not be crossed [21]. Furthermore, by entering into an agreement with MassDEP to assist with the cleanup of the site, NuTRIP hopes to garner public support for the project SITE BOUNDARY The exclusion area boundary is determined in accordance with the guidelines set forth in 10 CFR The maximum hypothetical accident (or MHA - as defined in NUREG- 1537, see chapter 4) was used as the initiating event for determining the exclusionary area. The MHA for the TRIGA reactor is a full clad failure of a single fuel element at 2 MWt and full exposure of the rod to air. Table 15 in 4.1 shows the TEDE and CDE to the thyroid as a function of the distance from the reactor boundary for the TRIGA reactor at McClellan operating at 2 MWt. Figure 16 shows the exclusionary area boundary as defined in the SAR for the McClellan reactor. The minimum distance between the reactor building and the EAB fence is about 30 ft. 36

37 Distance to EAB fence ~ 10m FIGURE 16 - ARTIST'S DEPICTION OF MNRC. THE CENTRAL BUILDING CONTAINS THE 2 MWT TRIGA, WHILE THE PERIMETER FENCE REPRESENTS THE EAB. From Table 15, it is clear that the guidelines for the EAB as set forth by 10 CFR are met. In fact, the estimated dose to the public beyond the exclusionary boundary for the worst- case MHA does not exceed the limit of 100 mrem established in 10 CFR for non- accident scenarios. Despite the EAB being much less than the recommended 10 miles for a nuclear power plant, the dose to the public for the MHA is still far less than the guidelines set forth in 10 CFR Extremely small EAB's is a characteristic of TRIGA reactors; over 19 of which have been operated on college campuses [5]. By drawing a parallel to the MNRC, the EAB for the NuTRIP facility will be negligibly small, i.e. the EAB will be fully contained within the industrial park. Furthermore, the EAB may be extended beyond 30 feet from the facility to offer significant conservatism for the processing facility. The full radioactive inventory contained in the irradiated foil is much less than that for the release discussed for the TRIGA MHA (only about 2 grams of U- 235 irradiated at roughly MWdt per foil). As such, the exclusionary boundary will not need to be extended beyond that for the reactor facility. 37

38 POPULATION AND INDUSTRY The town of Billerica, Ma has an estimated population around 38,981 according to the 2010 census. The total estimated population within a 3- mile radius around the site is around 61,000 due to the site's proximity to South Lowell. Zoning around the site is generally for low- density residential housing (population density < 1,540 people per square mile), with some commercial areas as well as light industry, such as that which used to be contained within the park itself. The 3- mile radius around the site includes four nursery schools/day- care centers and two elderly live- in communities. There is a trailer park and a condominium complex within one mile of the site. Despite the high density of population around the site and the relative proximity to large population centers such as Lowell and Boston, the minimal EAB (contained well within the boundaries of the industrial park itself) as well as the groundwater isolation and monitoring of the site ensure the NRC standards in 10 CFR 100 are met. Furthermore, no emergency preparedness zone (EPZ) is required for the facility due to the lack of credible accident scenarios for emitting radioactivity beyond the EAB. Again, this is a common feature of TRIGA reactors. Other research reactors in the area include the MNTR (MIT Nuclear Test Reactor), which operates at 5 MWt and is located in downtown Cambridge (approximately 18 miles from the Iron Horse site) TRANSPORTATION In order to minimize the loss of Mo- 99 due to decay during transportation, it is important to ensure adequate access to ground transportation as well as for the Tc- 99m generator manufacturer to be nearby the processing site. Lantheus Medical Imaging Inc. is headquartered in North Billerica and plans on developing Tc- 99m generator production capacity in the region. The NuTRIP design consists of vertical integration of irradiation and Mo- 99 extraction facilities, eliminating the need to transport irradiated materials off- site. The off- site transport of Mo- 99 will occur in accordance with the guidelines set forth in 10 CFR HIGHWAY TRANSPORTATION The Iron Horse site sits between two major highways: I- 93, which spans the eastern seaboard, and I- 495, which spans the eastern parts of Massachusetts and New Hampshire. I- 495 can be accessed by travelling west from the site about 1 mile along CR- 129, which intersects I- 495 about 2 miles from the site. I- 93 cannot be directly accessed from the site without travelling along several local roads. The easiest access to I- 93 is taking I- 495 east. The intersection is about 6 miles east of the site. Access to CR- 129 is only 38

39 available from the south west corner of the industrial park. The remaining three sides of the park are bounded by Salem Rd, Pond Rd, and High St, all local two- lane roads. The site is hidden from view by hundreds of feet of wooded area along each of the three roads RAILWAY TRANSPORTATION FIGURE 17 - RAILWAY INFRASTRUCTURE RELATED TO THE IRON HORSE SITE. THE LINES ENTERING THE NORTH- EAST OF THE SITE ARE NO LONGER RAILED. As can be seen in Figure 17, the Iron Horse site used to be home to a railway maintenance station. The rail line is a commuter line operated by the Massachusetts Bus and Transit Authority (MBTA), originally owned and operated by Pam- Am railways. No information on freight activity along the rail line could be found. The nearest station is roughly 1 mile north in downtown North Billerica. The rail line that used to extend into the Iron Horse site has been cut and is no longer in use. The North- East portion of the site contains a storage area for rail- ties, which cannot be burned due to high levels of carcinogenic fire retardants in the wood AIR TRAFFIC There are roughly 10 airports in a roughly 20 mile radius around the site, including Logan International in downtown Boston. Figure 18 shows the airports in the vicinity of the site, while Table 5 lists the type and volume of air traffic. 39

40 FIGURE 18 - AIRPORTS WITHIN 20 MILES OF PROPOSED SITE (BLUE BUBBLE). Logan Intl TABLE 5 - AIRPORT INFORMATION FOR NON- INTERNATIONAL AIRFIELDS WITHIN 20 MILES OF IRON HORSE SITE Letter Airport Name Type Aircraft Type A Minute Man Airfield Public Access Single- Engine B, F Lawrence Municipal Airport Municipal Regional Jet C Sterling Airfield Public Access Single- Engine D Hanscom Field Regional Regional Jet E Beverly Municipal Airport Municipal Regional Jet G Fitchburg Muni Municipal Regional Jet * Logan International International Jumbo Jet Logan International Airport is the 12th busiest airport in the US based on international traffic, with about 25 million passengers passing through the airport per year. The airport records an average 924 flights per day, with 95% of the traffic being commercial or air- taxiing. Roughly 16% of the air traffic at Logan is from international flights. The airport itself is over 25 miles from the Iron Horse site EVALUATION OF POTENTIAL LOCAL ACCIDENTS There are no nearby industrial or military facilities with credible potential for causing an accident that would result in a release of radioactivity beyond the general public exposure limits set forth in 10 CFR 20. External accident sources due to proximity to airports may be considered further. For the MNRC, which is located on an ex- Airforce base with a runway very near the reactor building, the probability of an 40

41 external aircraft impact was calculated to be less than 10-8 per year [14]. This falls well below the region of BDBE and is considered an incredible external event. Since the Iron Horse site is situated further from the nearest airport and has comparable air traffic than the Sacramento area, the probability for such an event at the Iron Horse site can be considered to be less than 10-8 per year LOCAL METEOROLOGY GENERAL CLIMATE FIGURE 19 - GENERAL CLIMATE INFORMATION FOR BILLERICA, MA [22] The climate of Billerica, Massachusetts does not deviate from the New England regional climate [22]. The average temperature during the winter is about 25 F from Nov- Feb. and 70 F in the summer 41

42 months. The proximity to the Atlantic coast yields a higher degree of precipitation than the US average as well as higher wind speeds. The prevailing wind direction is WNW in the summer months, and east for the rest of the year, typical of North Atlantic coastal regions. Figure 20 shows the prevailing wind directions as well as wind speeds at Logan International Airport, roughly 20 miles southeast of the site. Wind patterns at the site are expected to mimic those shown in Figure 20, although their intensity will be less due to the location being further inland. FIGURE 20 - WIND ROSE PLOT FOR LOGAN INTERNATIONAL AIRPORT [23], JUST SOUTHEAST OF THE IRON HORSE SITE HURRICANES Table 6 shows all of the hurricanes that have made landfall in southeast New England during the 20th century. Notice there have been several storms to make landfall as category 3 storms. The rate of hurricane landfall in New England is roughly one per ten years with a category 3 or above one in every 30 years. Table 7 shows the sustained wind speeds, rainfall, and building damage potential for the different categories of hurricanes [25]. Due to the fact that Iron Horse is about 27 miles inland from the coast, exposure to prolonged category 1+ winds is predicted to be negligible. One consequence of 42

43 hurricanes and tropical storms is large amounts of precipitation, which could lead to threats from flooding. Flooding characteristics of the site will be discussed in section TABLE 6 - RECORD OF HURRICANES MAKING LANDFALL IN SOUTHEAST NEW ENGLAND DURING THE 20TH CENTURY [24] 43

44 TABLE 7 - GENERAL CHARACTERIZATION OF HURRICANES Category Wind Speed Barometric Pressure Storm Surge Conventional Building Damage Potential 1 (weak) mph kts m/s > in. Hg > mb > 97.7 kpa ft m minimal damage to vegetation 2 (moderate) mph kts m/s in. Hg mb kpa ft m moderate damage to houses 3 (strong) mph kts m/s in. Hg mb kpa ft m extensive damage to small buildings 4 (very strong) mph kts m/s in. Hg mb kpa ft m extreme structural damage 5 (devastating) > 155 mph > 135 kts > 70 m/s < in Hg < mb < 91.7 kpa > 18.0 ft > 5.5 m catastrophic building failures possible TORNADOES Middlesex county in northeastern Massachusetts is hit by F2 or greater tornadoes at a rate of about one every two years, with F3 tornadoes once every 15 years and two recorded F4 tornadoes since 1950 [26]. Table 8 demonstrates the strength scale for tornadoes. TABLE 8 - FUJITA SCALE FOR TORNADO SEVERITY F- Scale Number Intensity Phrase Wind Speed Type of Damage Done F0 Gale tornado mph Some damage to chimneys; breaks branches off trees; pushes over shallow- rooted trees; damages sign boards. F1 Moderate tornado mph The lower limit is the beginning of hurricane wind speed; peels surface off roofs; mobile homes pushed off foundations or overturned; 44

45 moving autos pushed off the roads; attached garages may be destroyed. F2 Significant tornado mph Considerable damage. Roofs torn off frame houses; mobile homes demolished; boxcars pushed over; large trees snapped or uprooted; light object missiles generated. F3 Severe tornado mph Roof and some walls torn off well constructed houses; trains overturned; most trees in fores uprooted F4 Devastating tornado mph Well- constructed houses leveled; structures with weak foundations blown off some distance; cars thrown and large missiles generated. Tornadoes are more frequent during the autumn months, although they can occur at any time. The winds from the F4 tornadoes that occurred in Middlesex county were more severe than a category 5 hurricane. Each of the F4 tornadoes occurred at least 20 miles from the Iron Horse site FLOODS The most significant risk posed by hurricanes and tropical storms is the increased levels of precipitation and the risk of flooding. The Concord River passes through Billerica and is only about 2000 ft from the edge of the industrial park at its closest point. The floodplains for Billerica however are located further downstream near downtown Billerica. Floodplains maps are available at [27]. The maps show the border of the nearest floodplain from the Concord River slightly encroaching on the southwest corner of the site, where the parking lot can be seen in Figure 14. The east end of the site is at slightly higher elevation. There are two ponds within a mile of the northeast corner of the site, but each is at a lower elevation than the site itself. Marshland immediately north and east of the northeast corner of the site is separated from the site by earthen dykes. In the event of very heavy precipitation from multiple tropical storms it is possible that the marshland could encroach upon the northeast boundary of the site. Therefore, there are slight flooding risks for the extreme southwest and northeast corners of the site. The central area of the site is several feet higher in elevation than that immediately adjacent to the marshland and thus is under very little risk from flooding. No records of flooding on the Iron Horse site could be found. 45

46 GEOLOGY AND SEISMOLOGY The geologic and seismologic characteristics of the site will be discussed in Chapter REFERENCES [1] Nuclear and Radiation Studies Board. (2009). Medical Isotope Production Without Highly Enriched Uranium. National Academy Press. [2] National Research Universal. (n.d.). Retrieved from The NRU Reactor: [3] U.S. Nuclear Regulatory Commission. (2011). U.S. NRC Regulations: Title 10, Code of Federal Regulations. [4] Wikipedia. (2011, March 15). Petten Nuclear Reactor. Retrieved from [5] General Atomics. (n.d.). TRIGA Nuclear Reactors. Retrieved from esi.com/triga/ [6] Argonne National Laboratory. (2010, May 19). Retrieved from Reduced Enrichment for Research and Test Reactors: [7] Bhuiyan, S. M. (2000, May). Criticality and Safety Parameter Studies of a 3- MW TRIGA Mark- II Research Reactor and Validation of the Generated Cross- Section Library and Computational Method. Nuclear Technology, 130, pp [8] Huda, M. (2006, July 31). Computational Analysis of Bangladesh 3 MW TRIGA Research Reactor Using MCNP4C, JENDL- 3.3 and ENDF/B- VI Data Libraries. Annals of Nuclear Energy, pp [9] Huda, M. S. (2001, July). Thermal- Hydraulic Analysis of the 3- MW TRIGA Mark- II Research Reactor under Steady- State and Transient Conditions. Nuclear Technology, 135, pp [10] Barnowski, R. Hunter, A. Laird, J. Pfeffer, S. Wagner, S. Investigation of Molydbenum- 99 in a TRIGA Reactor. Senior Design Project, University of Michigan, [11] Kim, K. S. (2003). An Investigation of the Fabrication Technology for Uranium Foils by Cooling- roll Casting. International Meeting on Reduced Enrichment for Research and Test Reactors. Chicago, IL. [12] Vandegrift, G. (2006). HEU vs. LEU Targets for 99Mo Production Facts and Myths. Argonne National Laboratory. 46

47 [13] UC Davis. (2003). Retrieved from McClellan Nuclear Research Center: [14] U.S. Nuclear Regulatory Commission. (1998). NUREG- 1630: Safety Evaluation Report Related to the Issuance of a Facility Operating License for the Research Reactor at McClellan Air Force Base. [15] OECD/NEA. (2010). The Supply of Medical Radioisotopes - Interim Report of the OECD/NEA High- level Group on the Security of Supply of Medical Radioisotopes. OECD. [16] Indonesian National Atomic Energy Agency & Argonne National Laboratory. (1999). Full- Scale Demonstration of the Cintichem Process for the Production of Mo- 99 Using a Low- Enriched Target. [17] Press Release Lantheus Medical Imaging Completes Private Offering of $150 Million of 9.750% Senior Notes Due Press html. (2011, March 13). [18] U.S. Environmental Protection Agency. (2010, December 09). Iron Horse Park. Retrieved from Waste Site Cleanup & Reuse in New England: 691f0063f6d0!OpenDocument [19] City Data. (2010). Lowell, Massachusetts. Retrieved from data.com/city/lowell- Massachusetts.html [20] Cintichem, Inc. (1991). Cintichem Response to NRC Request for Additional Information Regarding Decommissioning Plan - Cintichem, Inc. [21] Shapiro, J. (2002). Radiation Protection: A Guide for Scientists, Regulators and Physicians. Harvard University Press. [22] City Data. (2010). Billerica, Massachusetts. Retrieved from data.com/city/billerica- Massachusetts.html [23] Windfinder. (2011). Wind & weather statistic Boston Logan Airport. Retrieved from [24] Wikipedia. (2011, April 02). List of New England Hurricanes. Retrieved from [25]Elert, Glenn. Speed of the Winds in a Hurricane. [Online] [26] Homefacts. (2011). Billerica Tornado Information. Retrieved from County/Billerica.html [27] Billerica Department of Public Works. (2007). Geographic Information Systems. Retrieved from 47

48 3. DESIGN BASIS EVENTS Design Basis Events (DBEs) are events not expected to occur within the lifetime of a single reactor, but might happen in the lifetime of a fleet of reactors. This condition implies frequencies between 1E- 02/year and 1E- 04/year. At these event frequencies, an accident should release a dose lower than 25 rem (250 msv) for Total Effective Dose Equivalent (TEDE) [1]. On our proposed site, DBEs can occur at every step of the Mo- 99 production process, from the TRIGA reactor where LEU foils are irradiated to the processing facility where Mo- 99 is extracted and purified. In this Preliminary Safety Analysis Report, one DBE is highlighted for the reactor, and one is detailed at the processing stage DESIGN BASIS EVENT DURING OPERATION OF A TRIGA REACTOR: LOSS OF COOLANT ACCIDENT (LOCA) An earthquake of much greater intensity than the Uniform Building Code Zone 3 [2] earthquake appears to be the only credible mechanism for causing a large rupture in the tank of a TRIGA reactor, since the tank when supported by its associated biological shield structure was designed to withstand this magnitude of earthquake. Records of earthquakes occurring in the area have been collected for the past 300 years, and the largest recorded earthquake was an estimated magnitude 4.6 earthquake in 1755 (more details in Chapter 6). However, to be conservative, we assume that such a large earthquake could happen in a 10,000 year time frame, which makes this event fall into the DBE category. Even if such an event is assumed to cause very rapid loss of water while the reactor is operating at peak power, a reactor shutdown would be caused by voiding of water from the core, even if there were no scram. A large rupture of the tank would obviously result in a more rapid loss of water than a leak due to corrosion or a minor mechanical failure in the tank wall. The reactor tank has no breaks in its structural integrity (i.e., there are no beam tube protrusions or other discontinuities in the reactor tank surface). In addition, the reactor core is below ground level. Thus the potential for most types of leaks is minimized. A cut into the biological shield exposes the reactor tank wall below the reactor core level, and this introduces an increased possibility of draining water from the core area. While steps have been taken to minimize the probability of a tank rupture in this location, and it is believed that the likelihood of such a rupture is very low, an unplanned occurrence could nevertheless initiate such an event. Therefore, an Emergency Core Cooling System (ECCS) has been installed to cool the core until the fuel has decayed to a level where air cooling is adequate to maintain fuel temperatures below the design basis limit. 48

49 MITIGATION ELEMENTS A loss- of- coolant accident (LOCA) is postulated, in which the reactor pool is rapidly drained of water during operation at 2 MW (it is assumed that the reactor has been running at 2 MW for an infinitely long time). Because the LOCA uncovers the core quickly, the fuel clad temperature in some of the centrally located fuel elements could exceed the design basis temperature limit of 930 C after a period of at least 20 minutes. When the reactor tank water level drops below the normal operating range, a tank low- level alarm sounds. This alerts the operator that action must be taken. Depending upon the rate of water loss, the suspected cause of the loss, and other considerations, several different actions may be taken by the operator in response to a reduction in the tank water level. One such action could be activation of the ECCS. Upon activation of the ECCS, cooling water from the domestic water supply will be introduced into the reactor tank and maintained until the fuel no longer contains sufficient decay heat to present a threat to the fuel cladding or water is restored to a level above the core. If the tank water level has dropped to less than about two 2 ft above the core, water from the ECCS will be sprayed onto the top of the remaining water column above the core; however, if the tank water has dropped below or partially below core level, the ECCS water will be sprayed directly onto the core. During this time, the decay heat will be removed by the remaining tank water or by the water spray and the maximum fuel temperature will be reduced rapidly from an elevated operating temperature down to about 200 C and then gradually to 100 C with continued spray cooling. At the end of spray cooling, natural air convection will be established in the core. During this cooling phase, the temperature of the fuel will rise slowly over several hours to a maximum and then decrease with continued air cooling. The maximum fuel and cladding temperature is controlled by the length of spray cooling and by the natural air cooling. Under the preceding conditions, no fuel cladding will be ruptured. Figure 21 shows the maximum and average fuel temperature during air cooling cycle for various spray cooling times. It can be seen that for spray cooling times above 3.7 hrs, followed by air cooling, the safety limits are met. 49

50 FIGURE 21 - MAXIMUM AND AVERAGE FUEL TEMPERATURE DURING AIR COOLING CYCLE FOR VARIOUS SPRAY COOLING TIMES [3] RADIATION LEVELS FROM THE UNCOVERED CORE Even though there is a very remote possibility that the primary coolant and reactor shielding water will be totally lost, direct and scattered radiation doses from an uncovered core following 2 MW operations were calculated. Direct radiation doses were calculated for a person standing on the grating directly above the reactor core. The core, shut down and drained of water, was treated as a bare cylindrical uniform source of 1 MeV photons. No accounting was made of sources other than fission product decay gammas, and no credit was taken for gamma attenuation through the fuel element end pieces and the upper grid plate. The first of these assumptions is optimistic, the second conservative, and the net effect is conservative. 50

51 TABLE 9 - DOSE RATES ON THE REACTOR TOP AFTER A LOSS OF POOL WATER ACCIDENT FOLLOWING 2 MW OPERATIONS [3] Dose Rates on the Reactor Top After a Loss of Pool Water Accident Following 2 MW Operations Time After Shutdown Effective Dose Equivalent Rate (rem/h) 10 seconds 3.64E04 1 hour 3.77E03 1 day 1.69E03 1 week 8.96E02 1 month 4.70E02 A second calculation was made to determine the dose rate to a person in the reactor room who is not in the direct beam from the exposed core but is still subject to scattered radiation from the reactor room ceiling. The dose point was chosen to be three feet above the reactor room floor at a distance of six feet away from the edge of the reactor tank. This is the furthest distance a person can get from the edge of the tank and still remain in the reactor room. The ceiling of the reactor room is about twenty four feet from the reactor top and is assumed to be a thick concrete slab. The concrete slab assumption gives the worst case scattering, but it should be carefully noted that the roof over the reactor is only corrugated metal and not a thick concrete slab. Therefore, in reality the scattering will not be as great as calculated because the radiation from the unshielded core will be collimated upward by the shield structure and will undergo minimal interaction with the roof, greatly reducing the actual dose rates away from the edge of the tank. The results of the calculated dose rates due to scatter in the reactor room are found in Table 10. These dose rates show that personnel could occupy areas within the reactor room shortly after the accident for a sufficient period of time to undertake mitigating actions without exceeding NRC occupational dose limits (10 CFR : TEDE limit for workers: 5 rem/yr). TABLE 10 - SCATTERED RADIATION DOSE RATES IN THE REACTOR ROOM AFTER A LOSS OF POOL WATER ACCIDENT FOLLOWING 2 MW OPERATIONS [3] Scattered Radiation Dose Rates in the Reactor Room After a Loss of Pool Water Accident Following 2 MW Operations Time After Shutdown Effective Dose Equivalent Rate (rem/h) 10 seconds

52 1 hour day week month A final calculation was carried out to estimate the dose rates to a person at the facility fence due to scattered radiation from the reactor room ceiling. The dose point was chosen to be three feet above the ground at the facility fence. This is the closest point a member of the public would be able to occupy. The calculated dose rates are presented in Table 11, but once again are overestimates because scatter off of the reactor room ceiling will be much less than assumed. They do not exceed NRC dose limits (10 CFR : TEDE limit for the public: 0.1 rem/yr). TABLE 11 - SCATTERED RADIATION DOSE RATES AT THE FACILITY FENCE AFTER A LOSS OF POOL WATER ACCIDENT FOLLOWING 2 MW OPERATIONS [3] Scattered Radiation Dose Rates at the Facility Fence After a Loss of Pool Water Accident Following 2 MW Operations Time After Shutdown Effective Dose Equivalent Rate (rem/h) 10 seconds hour day week month CONCLUSION From conservative calculations, both in terms of frequency and dose to the workers and the public, it has been assessed that a LOCA would not exceed the limit of a 25 rem release set for DBEs. Furthermore, expected releases will not exceed limits as set in 10 CFR 20, thus meeting the NRC licensing requirements. 52

53 3.2. DESIGN BASIS ACCIDENTS FOR PROCESSING FACILITY Unlike the previous safety analysis of the TRIGA reactor portion of the facility, there is no existing design for the Mo- 99 post processing facility. As such, the NuTRIP team has approached the issue by identifying potential accidents for such a facility, and estimating their frequency of occurrence and the associated consequences. All values for frequency of occurrence are best- estimates based on engineering judgment or comparative studies. Values for estimates of dose- release consequences of the accidents are framed in terms of releases from identified accidents with TRIGA fuel, for which dose calculations exist. Dose release estimates will be analyzed within the framework of 10 CFR 20 for dose to the public and to workers at the facility. 10 CFR contains a higher limit for dose to the public during accident scenarios (.25 Sv for 2- hour exposure on the site boundary) [4]; however, due to the smaller radioactive inventory of the core and low power- level of the facility, NuTRIP proposes to meet the safety goals set forth in 10 CFR 20 for non- accident scenarios. FIGURE 22 - EXAMPLE FARMERS CURVE DISPLAYING THE FREQUENCY AND CONSEQUENCES OF SEVERAL CLASSES OF ACCIDENT SCENARIOS [5] Hypothetical accident scenarios are identified by their estimated frequency of occurrence. Figure 22 shows an example Farmer's curve for accident scenarios identified for nuclear facilities. Notice the ordinate axis contains the frequency domains for which the different accidents are defined. In the following sections, one credible hypothetical accident will be identified for each of the frequency domains. 53

54 ANTICIPATED OPERATION OCCURRENCE - HYDRAULIC TUBE TRANSFER SYSTEM FAILURE TRANSFER TUBE LAYOUT OVERVIEW One possible accident scenario that may occur is a failure of the hydraulic/pneumatic tube transfer system. The tube transfer is designed to use pressurized fluid to transport the foils into and out of their irradiation positions in the core. The transfer tubes will connect a cooling/loading pond area to the center of the TRIGA core. Figure 23 shows an example diagram of the expected transfer system and reactor layout. 10 ft 27 ft 4 ft FIGURE 23 - AXIAL SKETCH OF PROPOSED TUBE TRANSFER SYSTEM. NOT TO SCALE. DIMENSIONS OF COOLING POND AND TRIGA REACTOR NOT FINALIZED. 54

55 The cooling and loading pond is within the same tank as the TRIGA reactor with all the same dimensions for the steel pool lining. Figure 23 shows a representative sketch only, not a design schematic. For instance, only four transfer tubes are shown, while the final number will be 18. Furthermore, there will be no ninety degree angles in the transfer tubes. A higher degree of curvature is required to allow for the passage of the foils. The pool depth is over 27 feet to allow for N- 16 (t 1/2 = 7 seconds) from the TRIGA to decay before bubbles from the bottom of the core reach the surface of the water and release the gas; the cooling pond requires no such consideration. The cooling and loading pond will extend radially around the top of the reactor pool. Each tube transfer line will be at a different radial position in the pool, as indicated by the sketch in Figure 24. This ensures the independence of the transfer lines so that insertion and extraction procedures can be operated at the same time in multiple lines. FIGURE 24 - SKETCH OF PROPOSED CORE LAYOUT. EACH SPOKE REPRESENTS A TRANSFER TUBE, ILLUSTRATING THE POSSIBLE RADIAL DISTRIBUTION. GREEN = LEU FOIL, RED = FUEL ELEMENT, GRAY = CONTROL ROD, BLUE = REFLECTOR 55

56 The green elements in Figure 24 represent reactor positions devoted to transfer tubes for foil irradiation. Each of the black spokes in Figure 24 represents a proposed radial distribution for the individual transfer tubes. The spokes do not reflect the actual path of the transfer tubes; axial design, tube curvature, and mechanical interference from control rod placement are not considered in Figure 24. At steady state operation, the facility is expected to process three foils per day. Figure 24 shows how the reactor may be split into quadrants, with the activities occurring in one quadrant not affecting those in adjacent quadrants POTENTIAL FAILURE MODES AND ESTIMATED EVENT FREQUENCY Several potential failure modes are identified for the pressurized tube transfer design described in section The consequence for the failure mode which is most likely to contribute to the largest release of radioactivity is discussed in The material for the pressurized transfer system has not been identified, and failure modes may differ depending on the material type. For this preliminary analysis, a metal such as aluminum or stainless steel is assumed. Mechanical sticking - One possible failure mode is that the LEU target assembly gets physically stuck within the transfer tube. This is likely to occur within the part of the tube that is in the core due to thermal expansion in the assembly. The target assembly will be used multiple times and may also be subject to creep or other strain- inducing phenomena that may lead to a deformation in the assembly shape over time [6]. Similarly, the section of the transfer tube within the core will be subject to temperatures of 300 C during normal operation, as well as neutron fluxes of about 2x10 13, potentially leading to changes in the material. All bends in the transfer tube are potential failure points due to this mode, since small changes in radius of curvature or the size of the target can lead to sticking [7]. Loss of Tube Integrity - The pressurized transfer system can only function if the tube is capable of maintaining a pressure differential, therefore any cracks in the transfer tube will lead to a failure in transfer capabilities. The transfer tubes will be subject to low levels of vibration during steady state operation due to boiling and other currents in coolant near the reactor core. Despite the low stress levels associated with such motion, the repetitive nature subjects the transfer tubes to fatigue failure. Furthermore, although the water within the reactor cool is in a closed loop and constantly circulated through demineralizers, the fission products produced in the core and the target foils may contribute to stress corrosion cracking. Structural elements with curvature, such as the transfer tubes, are especially susceptible to corrosion, especially if the tube comes to any sharp points, as shown in Figure

57 FIGURE 25 - INDICATION OF COMMON FAILURE POINTS FOR PRESSURIZED TRANSFER SYSTEMS Sudden Pressure Loss or Target Assembly Impact - Another possible mode of failure is that the target assembly is subject to mechanical shock due to a sudden loss of pressurization or impact due to loss of velocity control. Either of these scenarios could subject the target assembly and the LEU foil to a mechanical shock that might damage the integrity of the foil. Without an existing design for the transfer system, it is impossible to know the failure rate of the components. However, based on the fact that there are 18 transfer lines that will be tasked with insertion/extraction operations on a six to seven day schedule; and that it is the element of the reactor with the most moving parts, it is safe to estimate a transfer tube failure on the order of one per year. Thus it is expected transfer tube failures will occur within the frequency region defined by the AOO. The sudden pressure loss scenario can only occur if there is a sudden failure of the tube pressure boundary, which is most likely the result of stress corrosion cracking, material embrittlement due to high neutron flux, or fatigue from continued operation. Each of these failure mechanisms occurs on a longer time scale (on the order of decades) depending on the material selected for the transfer tube. As such, the frequency of such failures may be more suited to the DBE region (one such failure per the lifetime of the reactor). However, due to the preliminary nature of this analysis and the uncertainty in identifying possible failure modes, it is conservatively assumed that this event has a higher return frequency and falls within the AOO region. Further design and PRA analysis will help clarify the return frequency of such an event CONSEQUENCE OF TRANSFER TUBE FAILURE The first two accident scenarios identified in the previous section would have an ultimate consequence of the target being stuck somewhere within the transfer tube with the failure of the primary means of extraction. Loss of extraction capability on its own is not expected to contribute to any radiation release. 57

58 The LEU foils and target assemblies can withstand significantly higher burn- ups than the six- day irradiation period they will be used for and are in no danger of releasing radionuclides due to a prolonged exposure to the core. The reactor may need to be shut down in order to activate a secondary means of extraction from the tube. For the third failure mode, a loss of foil transport capability is coupled with possible damage to the target assembly and LEU foil due to mechanical shock. One possible consequence of the mechanical shock is the rearrangement of the internal target geometry, i.e. the foil may come into contact with the aluminum casing of the target. In this event, we must ensure that the loss of one of the cooling channels for the target does not result in melting of the foil or the target assembly. TABLE 12 - SUMMARY OF PHYSICAL PROPERTIES FOR ADDITIONAL MATERIALS PRESENT IN THE URANIUM Material TARGET CELLS Thermal Conductivity (W/m- k) Specific Heat (J/kg- K) Density (kg/m^3) Aluminum (6061) Nickel Lightly Enriched Uranium Table 12 shows the physical properties associated with materials used in the target assembly. MCNP modeling was used to calculate the power in the LEU foil, resulting in a maximum value of 21.2 kw [8]. Due to the much higher thermal conductivity of Aluminum than LEU, as well as the relatively low power of the foil, no further damage due to melting is expected. In the interest of conservatism, it may be assumed that a temperature increase due to the aforementioned accident scenario causes a full instantaneous release of all noble- gas fission products. The maximum power in the foil listed above, and no scrubbing due to the pool water is assumed. Furthermore, assuming a collection efficiency of zero for all charcoal and HEPA filters, such an accident could produce a release of Kr- 85 at about 2.8 µci/s (see appendix B.2). Conservatively assuming the accident occurred for an hour, the total amount of Kr- 85 released would be roughly 9 mci. Without doing a detailed dose calculation from this activity, it is easy to see that the dose to a member of the public 30 feet from the plant due to the dispersion of 9 mci of beta- emitting radiation over the course of an hour into the atmosphere above the plant will be minimal. Due to the low activity results from exceedingly conservative estimates for radionuclide egress from the above accident scenario, the maximum amount of dose to be expected by a member of the public at the site boundary is expected to fall far below the 100 mrem limit given in 10 CFR 20. More importantly, we are confident that a more detailed analysis of this accident scenario with a finalized design will fall well below 10-6 mrem at the site boundary due to the excessive levels of conservatism in the above estimation. 58

59 DESIGN SAFETY FEATURES In accordance with the principle of defense in depth, several design safety features should be implemented to mitigate the effects of the failure modes that have been preliminarily identified. Secondary Target Extraction Mechanism - The most likely result of a failure of the pressurized transfer tube system is the loss of the ability to insert or extract target assemblies. While the safety consequences of this scenario are negligible, such a situation may lead to a higher frequency of reactor shutdown, or a loss of Mo- 99 inventory. To avoid these consequences, a second insertion and extraction mechanism should be employed concurrently with the pressurized system. For instance, the target assembly may be placed in a "basket" that is attached by a cable to an external winch. If the pressurized system fails, the tube may be extracted by using the winch to physically pull the basket back out of the tube. Furthermore, it would be beneficial to have the transfer tube itself be modular and extractable. That way, if the first two extraction methods for the target assembly fail, the transfer tube itself may be extracted. If the tube is modularly designed, a failure in one part of the tube will not necessitate the disassembly of the entire transfer line. Temperature Monitoring - Each transfer tube should be fitted with a thermocouple so that the temperature of the tube may be actively monitored. If the mechanical shock accident discussed above were to occur, the operator would be able to monitor the temperature in the tube (i.e. due to fission in the foil). Decoupling of Individual Lines - To eliminate a common mode of failure, it is recommended that each of the transfer tubes be run by independent pneumatic/hydraulic systems. That way, the failure of a single pressurizer does not knock out the capability to use multiple tubes. Since there are no immediate safety risks due to this AOO, this is more of an economic consideration DESIGN BASIS EVENT - LOSS OF FISSILE SOLUTION WITHIN PROCESSING HOT CELL As mentioned in Chapter 2, the CINTICHEM (or modified CINTICHEM, for LEU foils) process is used to extract the Mo- 99 from the fission foil targets. While the process as a whole is proprietary and not freely available, the first step involves the dissolution of the uranium foil in nitric acid [9]. Therefore, for at least part of the processing procedure, the fissile material (along with all non- volatile fission products) contained in the irradiated foil will be in an aqueous form. The loss of the dissolved target solution from its primary container is considered a design basis event. Loss of solution from the primary trough will contaminate the hot cell interior. Furthermore, the possibility of criticality must be taken into consideration. While there is no existing NRC regulation on criticality safety for fissile solutions, 10 CFR part 70 contains regulations for the handling of special 59

60 nuclear materials. While 10 CFR 70 contains no specific regulations for liquid materials, it will be used as the basis for material accounting and criticality safety for material at the NuTRIP facility. Further regulations regarding the handling of liquid fissile materials are derived from SECY FOIL PROCESSING FACILITY OVERVIEW After irradiation, the target assembly is transported by means of the pressurized transfer system to the cooling pond where it will stay for roughly 8 hours [10]. After the cooling period, it is transferred via another system to a hot- cell laboratory. The transfer method for this portion has yet to be determined. One possible solution is the extension of the pneumatic system discussed previously. At the Cintichem reactor, a flume- like coolant "trench" was used [11]. Since the transfer system has not been decided, it will not be the primary focus of this design basis event analysis. The hot cell laboratory will consist of a central trough or drum where the foil, having been extracted from the aluminum target assembly, will be placed. The ensuing chemical separation will occur in a series of segments within the trough. The trough itself needs to be pressurized during the dissolution to ensure an adequate reaction rate. Figure 26 shows a representative hot- cell with a similar design to one for the proposed NuTRIP facility [12], except with a drum instead of a central trough design. FIGURE 26 - HOT- CELL LABORATORY AT ARGONNE'S ALPHA- GAMMA LABORATORY [12]. THE NUTRIP HOT CELL FACILITIES HAVE A SIMILAR PROPOSED DESIGN. 60

61 POTENTIAL FAILURE MODES AND EVENT FREQUENCY External Initiating Events - The solution may egress from the primary trough structure by means of mechanical motion of the trough. To induce vibratory motion of the trough, the facility would need to undergo prolonged vibratory motion. An earthquake of significantly prolonged duration and high period (low frequency) may be able to induce resonance sloshing of the solution within the trough [13], leading to the solution spilling over the edges out of the primary container. As discussed in Chapter 6, the proposed site has had earthquakes with magnitudes up to M 4.6. For sloshing out of the container, the duration of the earthquake is more important than the magnitude; unfortunately, the highest magnitude quakes occurred in the 1700's, and no accurate records on their duration can be found. For conservatism, it is assumed that these earthquakes occurred with significantly long duration as to be capable of inducing resonant sloshing in the primary trough. From the seismic history of the site (Chapter 6), there have been two earthquakes over the past 300 years with a magnitude great enough to induce ground motion significant enough to affect structures. Based on the low number of events of this magnitude, a statistically significant estimate of their return frequency is not possible. However, based on the history of these events alone, and the conservative assumption that each was capable of producing sloshing, the frequency can be predicted to be about 10-4 per plant- year. Internal Initiating Events - Another possible pathway by which the solution could egress from the trough is failure of the trough itself. The troughs in the hot cells will contain highly corrosive chemicals. Furthermore, the frequency of trough use will subject it to mechanical fatigue. A combination of these factors could lead to a breach in the trough allowing any solutions contained to escape. The frequency of trough failure is currently unknown. However, based on engineering judgment for fatigue and corrosion of metals in acidic environments, this mode of failure is expected to have a frequency of <10-2 per plant- year CONSEQUENCE OF SOLUTION EGRESS In order to determine the overall risk from a radioactive solution egress accident, the consequence, in terms of dose at the EAB, needs to be quantified. There currently does not exist any accident data on full release of radioactive solutions from a target. The dose at the EAB can be estimated by making several conservative assumptions, and comparing the total radioactive inventory of the solution to that of a single TRIGA fuel element. The maximum hypothetical accident identified for the TRIGA reactor is the full release of all radionuclides from a single spent fuel element. Previous safety analysis of the MNRC included dose calculations for full release of radioactive inventory from a burned fuel element. Table 14 & Table 15 61

62 summarize the calculation for the fuel element for reactor staff as well as at the EAB (10 m from the reactor building). If the same assumptions are made for full release from radioactive foils as were made for the maximum hypothetical accident, the dose from the loss of solution accident can be estimated. Since the fuel and foil compositions are the same, dose is assumed to be proportional to the amount of fuel, with all other factors constant. Fuel and foil dimensions are taken from [8]. Mass!"# Fuel Element = Vol!"# ρ!"# = 492g LEU Mass!"# Foil = Vol!"# ρ!"# = 3.2g LEU The amount of LEU in the fuel is about 150 times the mass of LEU in one radioactive foil. For conservatism, it is assumed all eighteen of the foils are processed at the same time in the same hot cell. Thus the mass ratio of LEU in the Fuel:Foil becomes 8.5:1. Applying this scaling factor, Table 13 represents the estimates for the maximum hypothetical design basis even where the radioactive solution from 18 foils escapes the reactor hot cell. TABLE 13 - RADIATION DOSE TO MEMBERS OF THE GENERAL PUBLIC UNDER THE MOST CONSERVATIVE ATMOSPHERIC CONDITIONS (PASQUILL F) AT DIFFERENT DISTANCES FROM NUTRIP FACILITY FOLLOWING TOTAL LOSS OF SOLUTION CONTAINING 18 FOILS. EAB IS 10M. CDE Distance CEDE TEDE Thyroid DDE (mrem) (Meters) (mrem) (mrem) (mrem) From Table 13, the CDE to the Thyroid as well as the TEDE values to the public at the site boundary fall well below the standards set in 10 CFR for non- accident situations. The total dose given in the above tables is assumed to occur during a 9.2 minute period during which the radioactive contents are vented from the facility. The analysis relies heavily on assumption; however, there is a significant degree of conservatism in the assumptions to afford confidence that the results in Table 13 represent the upper bound of dose from this DBE. For instance, the number of foils being processed at any time would not exceed three. More 62

63 significantly, the processing facility will be designed with an off- gas system [14] thus the percentage of volatile radiation expected to escape would be less than the assumed values. Another passive safety feature would be to coat the floors of the hot cell with water- soluble resins, similar to those used in Fukushima [15], to immobilize dissolved fission products. Another feature which would simplify cleanup after such an accident would be to have a catch- basin separating the trough from the hot cell floor. Thus, in the event of this DBE, any non- volatile solution will be contained within the catch- basin, thus allowing it to be replaced and simplifying subsequent decontamination. Care must also be taken to avoid re- criticality of the solution, which is covered in REFERENCES [1] Peterson, P. (n.d.). Risk- based and Risk- informed Regulation. UC Berkeley, Lecture for NE 167/267 Class. [2] Uniform Building Code (Vol. I). (1997). [3] U.S. Nuclear Regulatory Commission. (1998). NUREG- 1630: Safety Evaluation Report Related to the Issuance of a Facility Operating License for the Research Reactor at McClellan Air Force Base. [4] U.S. Nuclear Regulatory Commission. (2011). U.S. NRC Regulations: Title 10, Code of Federal Regulations, Part [5] Stojadinovic, B. (2011, February 26). Earthquake Hazard and Risk to NPP (and other) Structures. UC Berkeley, Lecture for NE 167/267 Class. [6] Callister, William D. Materials Science and Engineering An Introduction. Wiley, [7] Klinzing, G. (2010). Pneumatic conveying of solids: a theoretical and practical approach (3rd Edition ed.). Springer. [8] Barnowski, R. Hunter, A. Laird, J. Pfeffer, S. Wagner, S. Investigation of Molydbenum- 99 in a TRIGA Reactor. Senior Design Project, University of Michigan, [9] Vandegrift, G. (2006). HEU vs. LEU Targets for 99Mo Production Facts and Myths. Argonne National Laboratory. [10] OECD/NEA. (2010). The Supply of Medical Radioisotopes - Interim Report of the OECD/NEA High- level Group on the Security of Supply of Medical Radioisotopes. OECD. [11] Cintichem, Inc. (1991). Cintichem Response to NRC Request for Additional Information Regarding Decommissioning Plan - Cintichem, Inc. 63

64 [12] Argonne National Laboratory. (n.d.). ARRA- funded work at Alpha Gamma Hot Cell Facility. Retrieved from /lightbox/ [13] Ibrahim, R. (2005). Liquid Sloshing Dynamics: Theory and Applications. Cambridge University Press. [14] Ahn, J. Volume Reduction, Solidification, and Storage of Radioactive Waste. UC Berkeley, Lecture for NE 124, Fall [15] Kyodo News Network. Workers to Spray Resin over Debris at Fukushima Nuclear Plant. to- spray- resin- over- debris- at- fukushima- nuclear- plant. (2011, April 10th). 64

65 4. BEYOND DESIGN BASIS EVENTS Severe accidents or Beyond Design Basis Events (BDBEs) differ from Design Basis Events based on their frequencies and dose release. These events are not expected to ever occur over the lifetime of a fleet of reactors or with a frequency of between 1E- 04/year and 5E- 07/year. At these event frequencies, an accident should release a dose lower than 800 rem (8 Sv) for Total Effective Dose Equivalent (TEDE) [1]. Once again, on our proposed site, BDBEs (like DBEs) can occur at every step of the Mo- 99 production process, from the TRIGA reactor where LEU foils are irradiated to the processing facility where Mo- 99 is extracted and purified. In this Preliminary Safety Analysis Report, one BDBE is highlighted for the reactor, and one is detailed at the processing stage BEYOND DESIGN BASIS EVENT DURING OPERATION OF A TRIGA REACTOR: MAXIMUM HYPOTHETICAL ACCIDENT (MHA) In this part, the Maximum Hypothetical Accident (MHA) encountered at a TRIGA reactor is described, and its consequences are analyzed. It is important to notice that this event is hypothetical, and never expected to occur at an operating TRIGA reactor, but at frequencies so low that it can be classified as a BDBE ACCIDENT INITIATING EVENT AND SCENARIO A single fuel element could fail at any time during normal reactor operation or while the reactor was shutdown, owing to a manufacturing defect, corrosion, or handling damage. This type of failure is infrequent, based on many years of operating experience with TRIGA fuel, and such a failure would not normally incorporate all the necessary operating assumptions required to obtain a worst case fuel failure scenario. For our TRIGA reactor, the MHA has been defined as a cladding rupture of one highly irradiated fuel element with no decay followed by instantaneous release of fission products into the air. The failed fuel element was assumed to have been operated at the highest core power density for a continuous period of 1 year at 2 MW. This is the most severe accident for a TRIGA and is analyzed to determine the limiting or bounding potential radiation doses to the reactor staff and to the general public in the unrestricted area. 65

66 A realistic scenario for the MHA is difficult to establish since fuel handling, the activity frequently associated with this accident, would be unlikely to occur immediately after reactor shutdown, and fuel elements would not be moved out of the reactor tank into air with no time to decay. Nevertheless, the accident has been analyzed and the results are summarized in this section ACCIDENT ANALYSIS AND DETERMINATION OF CONSEQUENCES The inventory used for the volatile fission products present at shutdown in a fuel element run to saturation at the highest core power density. A fission product release fraction of 7.7E- 05 is assumed for the release of noble gases and halogens from the fuel to the cladding gap, based on calculation of fuel temperature in the hottest core element. In addition, it is assumed that 100% of the noble gases ultimately reach the unrestricted environment outside the reactor building and that 25% of the halogens released to the cladding gap are eventually available for release from the reactor room to the outside environment. This value for the halogens is based on historical usage and a recommended value of 50% release of the halogens, with a natural reduction factor of 50% due to plateout in the building. This latter 50% applied to the 50% of the inventory released from the fuel element cladding gap results in 25% of the available halogen inventory reaching the outside environment. It should be noted, however, that this value appears to be very conservative based on the 1.7% gap release fraction for halogens quoted in the literature. It was assumed that all of the fission products were released to the unrestricted area at 10 m from the reactor room by a single reactor room air change, which maximizes the dose rate to persons exposed to the plume during the accident and minimizes the exposure time to receive the highest estimated dose from this accident. These latter assumptions regarding release are extremely conservative since the reactor room is not at ground level and, rather than 10 m, is approximately 30 m from the perimeter fence. Shown in Table 14 & Table 15 are doses inside the reactor room and doses at several locations in the unrestricted area outside the reactor (10 to 100 m from the building) as a function of weather class. Results are reported for the Committed Dose Equivalent (CDE) to the thyroid, the Committed Effective Dose Equivalent (CEDE) due to inhalation, the Deep Dose Equivalent (DDE) due to air immersion, and the Total Effective Dose Equivalent (TEDE) resulting from adding the CEDE and the DDE. As indicated by the results in Table 14, the occupational dose to workers who evacuate the reactor room within 5 minutes following the MHA should be approximately 454 mrem Total Effective Dose Equivalent and 11,500 mrem Committed Dose Equivalent to the thyroid. If evacuation were to occur within 2 minutes, as it no doubt would because the reactor room is small and easy to exit, the doses drop to 180 mrem TEDE and 4,640 mrem CDE. All of these doses are well within the NRC limits for occupational exposure as stated in 10 CFR (5 rem/yr). 66

67 TABLE 14 - OCCUPATIONAL RADIATION DOSES IN THE REACTOR ROOM FOLLOWING THE MAXIMUM HYPOTHETICAL ACCIDENT [2] Accident: Cladding Failure in Air (MHA) CDE Thyroid [mrem] CEDE [mrem] DDE [mrem] TEDE [mrem] 2 min room occupancy 5 min room occupancy 4, , Projected doses to the general public in the unrestricted area around the reactor following the MHA are shown in Table 15. To receive the indicated dose, a person must be exposed to the airborne plume from the reactor room for the entire 9.2 minute period it is being vented. Even using this exposure requirement at the closest distance to the reactor building (10 m), and assuming the most unfavorable atmospheric conditions, the maximum TEDE to a member of the general public would be 66 mrem. TABLE 15 - RADIATION DOSES TO MEMBERS OF THE GENERAL PUBLIC UNDER THE MOST CONSERVATIVE ATMOSPHERIC CONDITIONS, AT DIFFERENT DISTANCES FROM THE REACTOR, FOLLOWING A FUEL ELEMENT CLADDING FAILURE IN AIR WITH NO DECAY (MHA) [2] Distance [m] CDE Thyroid [mrem] CEDE [mrem] DDE [mrem] TEDE [mrem] 10 1, , CONCLUSION Although this accident and the corresponding radiation doses are never expected to occur, the maximum estimated dose of 66 mrem to the general public is still within the 100 mrem TEDE limit for the general public published in the NRC's 10 CFR Furthermore, the above analysis clearly shows 67

68 that the reactor can be subjected to current MHA criteria and remain within dose limits established by the NRC for occupational radiation exposure and exposure of the general public. As a point of interest, should the MHA occur after 48 hours of decay, the maximum TEDE to the public drops to approximately 34 mrem BDBE FOR PROCESSING FACILITY - RE- CRITICALITY OF SOLUTION As part of the Cintichem process, the LEU foil must be extracted from the target assembly and dissolved in nitric acid post irradiation. The specific volumes and concentrations of the solutions used in the Cintichem procedure are not publicly available, as the DOE holds the license to the process and much of the information on the chemical process is proprietary or under development [3]. Nevertheless, a rudimentary study of the expected frequency and consequences of solution re- criticality is considered ACCIDENT SCENARIO DESCRIPTION It is assumed that the re- criticality event will occur while the fissile material is dissolved in an aqueous phase. The critical mass of uranium is significantly reduced in solution. As a result of this assumption, the criticality accident will take place in the hot- cell facility, which is the only portion of the separation process in which the fissile material is in solution. It is assumed for the accident scenario that any active reactivity control systems within the hot cell fail to perform properly. Furthermore, conservative assumptions are made about passive reactivity control measures. Since the details of the Cintichem process are not publicly available, the conservative assumption is made that none of the chemicals used contain neutron poisons or materials with significant neutron absorption cross sections. Also, it is conservatively assumed that the processing trough is perfectly reflective (albedo = 1) such that there are no neutronic losses due to leakage from the container. This assumption assures that the analysis will be valid for any trough design, regardless of surface- to- volume ratio. Figure 27 shows a schematic for a solution- based criticality accident pathway. The critical mass of uranium is reduced when in homogenous solution, allowing for criticality accidents even at much lower amounts of fissile material than in solid lattices. Figure 27 is adopted from a criticality accident that occurred in 1999 at a Japanese fuel processing facility in Tokaimura. While the Cintichem process is unlikely to need mixers, the multiple phases of the purification process may be represented by the different phases in Figure

69 FIGURE 27 - SCHEMATIC REPRESENTATION OF A LIQUID CRITICALITY ACCIDENT FOR A 2- PHASE CHEMICAL PROCESS. ADOPTED FROM DESCRIPTION OF ACCIDENT AT TOKAIMURA FUEL PROCESSING PLANT [4] ESTIMATION OF CRITICAL MASS AND ACCIDENT FREQUENCY The four factor formula is used in this preliminary analysis to estimate the concentration of uranium in solution required to achieve criticality. k! = ηεfp [1] Assumptions for criticality analysis 1) No leakage is assumed, in accordance with the assumption discussed in section 4.2.1, thus k will be calculated, representing the upper- bound of possible criticality values 2) It is assumed p = 1. The resonance escape probability actually tends to be lower for solutions than for lattices due to the homogeneity of the system [5]. Nonetheless, the resonance escape probability is assumed to be one to maintain conservatism. 3) ε = The fast fission factor is assumed to be 1.03, consistent with the value for natural uranium. This too is conservative since the uranium used in the foils is enriched to 20%, reducing the inventory of 238 U. 4) For subsequent analysis, it is assumed that water is used as the dissolving agent instead of nitric acid. This is a conservative estimate due to the higher mass of nitric acid as well as the lower absorption cross- section [6]. 69

70 Calculation of η η depends only on the enrichment of the fuel and the fuel type, thus can be calculated for the 19.75% enriched LEU foil: where thus σ! η = ν σ! + σ! ν = 2.5 σ! = σ!!" = barns σ! = σ!!" σ!!" = barns σ! η = ν = 2.5 σ! + σ! = 2.11 The mean number of neutrons produced per original neutron is estimated to be 2.11 for the LEU foil. The fission utilization factor is calculated using the following equation [6] where M represents the concentration (molarity), and σ represents the cross- section. [2] Table 16 shows the results of several applications of the Cintichem process [3]. Although not specified, for this analysis it is assumed each metal sample is 19.75% enriched LEU with a density of

71 TABLE 16 - EXPERIMENTAL RESULTS FROM SEVERAL APPLICATIONS OF THE CINTICHEM PROCESS IN A LAB SETTING AT ANL [3] Final Final Initial Conc Dissol. Steady Total Initial Test Solution Metal Acid of HNO3 Angle, State Temp Time Metal Foil ID Vol (ml) Conc. Conc. (M) deg. ( C) (min) Mass (M) (M) Equation 2 is used in conjunction with the data presented in Table 16 to estimate k for each test. The results are shown in Table

72 TABLE 17 - K- INFINITY ESTIMATION FOR EACH OF THE EXPERIMENTS LISTED IN TABLE 16 Initial Final Solution Metal Metal Test ID f k- inf Vol (ml) Foil Conc. Mass (M) The results in Table 17 show that even with very conservative assumptions about leakage and resonance absorption, the multiplication factor is much less than unity. By applying equation 2 in reverse, the mass ratio of LEU:water that results in a k of 1 can be calculated. The result of the calculation (see appendix B.3) are given below: Critical Ratio =!!"# = 1.03!"#$%!"#!"#!"#$!" [3] Equation 3 represents a very conservative estimate of the critical ratio for LEU in solution for the Cintichem process. LEU must be present in solution in a ratio of roughly 1 gram per ml of water in order for criticality to be achieved. Based on the degree of conservatism in the assumptions, the result presented in equation three represents the lower limit of the critical ratio. The frequency of occurrence for a criticality event can be estimated from the calculated critical ratio. Possible processes by which a ratio of 1 gram of LEU per ml of solution will be postulated, and the frequency of the processes estimated. External Initiating Events - There are currently no possible pathways by which an external event identified in chapters 2 or 6 could trigger a criticality event. Even for beyond- design- basis seismic and 72

73 meteorological events, the most severe expected consequence is loss of solution from the trough (see chapter 3). Losing solution from the trough decreases the chance of a criticality event because it disperses the solution over a larger area, decreasing the concentration and reducing the chance of arriving at the critical ratio. One possible consequence of beyond design basis external events is a station blackout. Even in blackout conditions however, the possibility of a criticality accident does not increase. No active coolant or solution addition is required to maintain solution levels. The steady state temperatures of the dissolution reactions are listed in Table 16. Although the temperatures are above the boiling point of water, the dissolution is to occur under higher pressure, preventing significant loss of solution by boiling. Based on these considerations, the probability of an externally initiated event resulting in a criticality accident is expected to be much less than 10-8 per reactor- year. Internal Initiating Events - The most likely cause of a potential criticality event is a combination of station blackout, unexpected material failure, and operator error. Figure 28 shows a hypothetical event tree that could lead to the re- criticality BDBE. Notice that there is no consequence unless all events are combined, because if criticality is not achieved, then there is no hazard, and each event must occur in order to achieve criticality. FIGURE 28 - HYPOTHETICAL EVENT TREE FOR ACCIDENT PATHWAY LEADING TO A RE- CRITICALITY ACCIDENT. All probabilities from Figure 28 are estimated values and expected to be conservative. It is assumed that the operators of the processing facility will fail to comply with proper procedure and increase the foil inventory per hot- cell on the order of once a year. With proper monitoring and maintenance, the processing trough is not expected to fail; the estimated unexpected (i.e. due to lack of monitoring and irregular maintenance) failure of the trough pressure boundary is expected to occur infrequently (10-3 per reactor- year). The failure of the pressure boundary allows for a situation in which the solution 73

74 containing the LEU may begin to boil away, leaving the LEU behind in a process similar to distillation. This condition is necessary in order to achieve the critical ratio. The reactor does not generate any power, and as such is subject to the same blackouts and brownouts that New England has faced in recent years [7]. Finally, in order for the critical ratio to be achieved, there must be a failure in all secondary and tertiary reactivity control systems. The design and nature (i.e. active or passive) of these systems is not finalized, therefore, a conservative estimate for compound failure rate is set at 10-4 per reactor year. Based on these considerations, the frequency of the re- criticality event is estimated to be on the order of 10-8 per reactor- year. Based on the regions defined by the Farmer's curve and 10 CFR 50.34, a re- criticality accident falls under beyond design basis events ESTIMATION OF CONSEQUENCE OF RE- CRITICALITY BDBE NUREG/CR "An Updated Nuclear Criticality Slide Rule" contains conversion factors and sliding graphs from which the doses due to criticality events may be estimated. The first step is to determine the number of fissions expected to occur. It requires approximately fissions to evaporate 1 L of water that was originally at room temperature [8]. Assume the amount of solution used to dissolve a single foil is the largest value given in Table 16, which is 55 ml. Furthermore, it is conservatively estimated that all 55 ml are required to be boiled away before the criticality event ceases (very conservative, given the criticality ratio calculated in 4.2.2). Thus, the total expected number of fissions from this criticality accident is roughly 5.5 x Based on the determination of the number of fissions occurring during the criticality event, the charts in Figure 29 can be used to estimate the dose. The results from Figure 29 are for HEU (93.2% enriched); however, no down scaling due to the difference in enrichment will be done in the interest of conservatism. 74

75 FIGURE 29 - DOSE "SLIDE- GRAPH" CHARTS FOR ESTIMATING DOSE DUE TO CRITICALITY ACCIDENTS AT VARIOUS TIMES AND DISTANCES FROM THE EVENT. 75

76 PROMPT DOSE CALCULATION Unfortunately, in order to have access to the full "slide- rule" feature of the dose estimator, one needs special authorization which none of the members of the NuTRIP team have (requests must be filed by an approved liaison to the NEA). Thus accurate values for the identified magnitude of the NuTRIP criticality event (i.e. 5.5E15 fissions) are not available. Instead, the information shown in Figure 29 is collected for the given event magnitude of fissions. These results are then scaled down linearly to match the hypothetical magnitude of the BDBE. TABLE 18 - PROMPT DOSE DUE TO 5.5E15 FISSION CRITICALITY EVENT. NO DECAY TIME OR SHIELDING IS TAKEN INTO ACCOUNT. 10E17 Fissions 5.5E15 Fissions Distance From Incident (ft) Total Dose including Skyshine (Rad) Total Dose including Skyshine (Rad) E E E E E E E- 03 1,00E+03 1,00E+02 Total Prompt Dose including Skyshine (Rad) vs. Distance - 5E15 Fissions Dose (Rad) 1,00E+01 1,00E+00 1,00E ,00E- 02 1,00E- 03 Distance from Accident (Unshielded) FIGURE 30 - TREND BASED ON DATA SHOWN IN TABLE 18 76

77 From this information, the dose to workers and the dose at the site boundary can be estimated. The information contained in Table 18 and Figure 30 represents a worst case scenario - prompt dose (no cooling time) is plotted and no shielding is assumed. In order to better reflect the prompt dose received by a worker or a member of the public at the site boundary, proper shielding and distance estimates must be made. Dose reduction factors due to shielding can be calculated as follows [8] FIGURE 31 - EMPIRICAL ESTIMATES FOR DOSE REDUCTION FACTORS DUE TO SHIELDING FROM VARIOUS SOURCES [8] Maximum Prompt Dose to Worker It is assumed that the worker is at the hot- cell boundary when the accident occurs - a distance of 5 feet from the accident location. Also, the hot cell wall is conservatively estimated to be composed of 6 inches of concrete. Using these values, the maximum expected prompt dose to the worker is calculated. Maximum Prompt Dose to Worker = 21 Rem This dose falls within the regime of acute dose effects and falls beyond the dose guidelines set forth in 10 CFR 20. However, this maximum prompt dose is within the allowable doses to radiological workers in accident scenarios as defined by 10 CFR 50. Furthermore, this value for dose represents a maximum due to the multiple levels of conservatism built into the calculation. Maximum Prompt Dose at EAB Recall the EAB for the site is only 10m from the reactor building. For this analysis, it is assumed the location of the hot labs is an additional 5 meters from the outer wall of the facility, giving a total distance of 15 m or about 50 feet between the accident location and the EAB. Also, it is assumed that there is two feet of concrete between the accident location and the EAB. Using these values, the maximum expected prompt dose to a member of the public at the EAB is calculated. Maximum Prompt Dose at EAB =.6 mrem This value also represents the upper bound on the expected dose at the EAB due to the high level of conservatism in the estimation. Furthermore, this value falls well within the allowable region as defined by 10 CFR

78 DELAYED DOSE CALCULATION As stated previously, the criticality event is assumed to be instantaneous; due to the very small amounts of solution, sustained criticality is not possible. Therefore there is no buildup or continuing production of fission products. Table 19 and Figure 32 show the dose rate at 5 feet away from the accident location as a function of time. No shielding is assumed. TABLE 19 - DOSE RATE VS. TIME FOR 5.5E15 FISSION CRITICALITY EVENT AT 5 FT FROM THE EVENT LOCATION. NO SHIELDING IS TAKEN INTO ACCOUNT 10E17 Fissions 5.5E15 Fissions Time Elapsed Since Incident (s) Gamma Dose Rate (R/Hr) Gamma Dose Rate (R/Hr) E E E E E E E E E E- 02 1,00E+03 Gamma Dose Rate (R/Hr) vs. Time Elapsed Since Criticality Event Gamma Dose Rate (R/Hr) 1,00E+02 1,00E+01 1,00E+00 1,00E ,00E- 02 Time Since Event (s) FIGURE 32 - TREND IN DOSE RATE DATA SHOWN IN TABLE 19 78

79 Figure 32 shows that the dose rate remains about 10 rem per hour at 5 feet from the accident an hour after the event occurred. The hot cell laboratory should be evacuated until radiation levels drop significantly in accordance with ALARA. There is no danger of further criticality accidents occurring due to a lack of foil monitoring from evacuation. Table 20 shows the dose rate results at the EAB. The same assumptions were used for this analysis as for the Prompt Dose calculations in the previous section. TABLE 20 - DOSE RATE VS. TIME ELAPSED SINCE ACCIDENT AT EAB 10E17 Fissions 5.5E15 Fissions Time Elapsed Since Incident (s) Gamma Dose Rate (R/Hr) Gamma Dose Rate (R/Hr) E E E E E E E E E E- 05 Note that the dose rate at the EAB is well below the limit specified in 10 CFR 50.34, and falls below background after about 8 hours DELAYED DOSE FROM FISSION PRODUCTS There is no expected dose due to the additional fission products produced from the re- criticality. The hot- cell lab is sealed and contains an off- gas system to isolate all volatile fission products from release to the outside environment. In the event of a criticality accident combined with a hot- cell breach, fission products may spread from the lab to the rest of the facility. The facility itself however has a closed- loop air circulation and purification system which is capable of scrubbing many of the fission products before release. Furthermore, the fission product inventory from the re- criticality is similar to that of the irradiated rod, with the exception of the presence of the short- lived fission products (t 1/2 < 8 hrs) which decayed away during the cooling time. These short- lived isotopes will likely be circulated long enough by the closed loop system that they are likely to decay away before release to the atmosphere. Thus the fission product release from the re- criticality BDBE is similar to the solution egress DBE discussed in Chapter 3. This event is a candidate for further safety analysis. 79

80 4.3. REFERENCES [1] Peterson, P. (n.d.). Risk- based and Risk- informed Regulation. UC Berkeley, Lecture for NE 167/267 Class. [2] U.S. Nuclear Regulatory Commission. (1998). NUREG- 1630: Safety Evaluation Report Related to the Issuance of a Facility Operating License for the Research Reactor at McClellan Air Force Base. [3] Indonesian National Atomic Energy Agency & Argonne National Laboratory. (1999). Full- Scale Demonstration of the Cintichem Process for the Production of Mo- 99 Using a Low- Enriched Target. [4] Goss Levi, B. (1999). What happened at Tokaimura? Retrieved from Physics Today: [5] Krane, K. (1988). Introductory Nuclear Physics. John Wiley & Sons. [6] Cember, H. T. (2008). Introduction to Health Physics (4th Edition ed.). McGraw- Hill Medical. [7] Holguin, J. (2003, August 15). Biggest Blackout In U.S. History. CBS News. [8] Hopper, C. B. (1998). NUREG/CR- 6504: An Updated Nuclear Criticality Slide Rule. Oak Ridge National Laboratory. 80

81 5. RISK ASSESSMENT AND MANAGEMENT It is necessary to assess the risk involved in reactor operation probabilistically. The objectives of this section are to discuss risk and establish safety goals for the plant. TRIGA have been built and operated for many years, and reactor risk is expected to be similar to established TRIGA designs. Discussion of risk involved in on- site processing of radioactive materials is also to be provided in this section SAFETY GOALS AND SOURCES OF RISK SAFETY GOALS The plant is expected to conform to these safety goals so the facility does not pose significant threat to society or the health of individuals [1]. Control any release of reactor material beyond exclusion area boundary. Control emission of radiation due to failure of pneumatic foil transfer system. Control all other emissions of radioactive material. Control emission of hazardous materials involved in chemical processing INTERNAL INITIATING EVENTS 1) Pneumatic System Failure a) Pneumatic Tube Failure i) Damage by impact with rabbit ii) Damage by pressure transient iii) Damage by pipeline blockage iv) Unexpected loss of power to pneumatic system b) Rabbit Failure i) Damage by impact with transfer tube ii) Mechanical accidents while extracting foils c) Pneumatic pressure boundary failure 2) Target failure during irradiation 3) Hot cell failure a) Failure of processing trough 81

82 5.1.3 EXTERNAL INITIATING EVENTS 1) Earthquakes 2) Storms 3) Impact by Aircraft 4) Flooding 5) Theft of materials 6) Large explosions/ Sabotage CINTICHEM PROCESSING ACCIDENTS 1) Accidental Cutting of Activated Reactor Components [2] 2) Resuspension of hot cell concrete dust [2] 5.2. RISK FROM TRIGA OPERATION The risk involved in operation of the TRIGA is expected to be similar to reactors of similar design so for reference data from the TRIGA Mark II reactor in Vienna will be provided. The risk can be portrayed as below using a fault tree TRIGA FAULT TREE The fault tree provided below is adapted from known studies on the Vienna TRIGA [3]. 82

83 FIGURE 33 - LOGIC TREE FOR TRIGA EVENTS. SEE TABLE 21 FOR KNOWN EVENT FREQUENCIES. 83

84 5.2.2 TRIGA PROBABILITIES Probability values for Vienna TRIGA events were assessed and assigned relative consequences [3]. The data specifically reflects 26 years of operation of the TRIGA reactor so some differences could exist for the proposed reactor. Due to the lack of a centralized collection of component failure data, these figures are provided, and it is suggested that a similarly constructed TRIGA Mark II could be expected to perform in a similar way. 1) Low consequences: releases which are of no consequence to facility staff or the public [3]. 2) Medium consequences: releases with low consequence on facility staff and no consequence to the public [3]. 3) High consequences: releases with consequences to facility staff and possible public exposure [3]. The following table is adapted from the Vienna TRIGA data [3]. TABLE 21 - QUANTITATIVE INFORMATION REGARDING THE FREQUENCY OF EVENTS THAT MAY LEAD TO RADIATION RELEASES. NUMBERS CORRESPOND TO FIGURE 33. Event Frequency [year - 1 ] Consequences A 7 release of FP from fuel elements ~ Low/Medium A 9 disturbances in reactivity 0 A 10 fault of fuel element transfer cask ~5 x 10-7 High A 11 direct gamma exposure from reactor pool ~8 x 10-5 High A 12 direct gamma exposure from other sources ~ High A 13 release of activity from experiments 0 if only used for isotope production A 14 release of ion exchange resin 0 A 16 loss of primary coolant from reactor tank ~10-5 Low A 17 loss of primary coolant by failure of primary circuit components other than tank ~3 x 10-2 Low A 4 fault in ventilation system >1 No consequence by itself 84

85 5.3. RISK FROM MO- 99 PROCESSING There is not an existing analogue to some of the design features proposed, which is why further studies will be required on components such as the Pneumatic Transfer Tube system. Reasonable estimates can be made about the frequency of issues that arise PNEUMATIC TRANSFER SYSTEM FAILURE 1) Discussed at length in , the failure modes for the Pneumatic system include : a) Mechanical sticking b) Loss of tube integrity c) Sudden pressure loss d) Target assembly impact 2) Frequency of Pneumatic Failure a) Foils expected to be extracted from core at rate of 3 per day. i) Foils can be extracted at intervals such that any accident involving pneumatic transfer is limited to damage to a single foil. b) Estimated rate of occurrence is less than 1 per year 3) Worst case scenario for this type of failure would be release of an entire foil s worth of radioactivity 4) Assume that there must also be a failure of the plant s ventilation system to cause any releases 5.4. REFERENCES [1] Peterson, P. (n.d.). Risk- based and Risk- informed Regulation. UC Berkeley, Lecture for NE 167/267 Class. [2] Cintichem, Inc. (1991). Cintichem Response to NRC Request for Additional Information Regarding Decommissioning Plan - Cintichem, Inc. [3] Kirchsteiger, C. H. (1988). Probabilistic Safety Assessment of the Vienna TRIGA Reactor. Atominstitut der Österreichischen Universitäten, Vienna, Austria. 85

86 6. SITE SAFETY AND SEISMIC ANALYSIS 6.1. DEFINITION AND SCOPE The purpose of this section is to characterize the seismological nature of the site and identify the relevant safety impacts in accordance with Appendix A to CFR 100, as well as the IPEEE identification standards reported in NUREG Note that characterization of non- seismic external hazard relevant to the site are documented and discussed in 2.4. Chapter 6 focuses solely on the seismic safety features of the site. Preliminary geologic analysis is conducted to characterize the seismology around the Iron Horse site. The safe shutdown earthquake [1] is identified according to the seismological history of the region and the geologic structure near the site. Similarly, the operating basis earthquake condition is estimated based on the record of seismological events in the area. It was brought to our attention during the meeting with the ACRS panel that several aspects of the seismic analysis should be re- evaluated before submittal for NRC approval. For instance, the site stratigraphy is not optimal due to a sandy layer which is subject to liquefaction, as well as a clay upper layer that is subject to swelling during precipitation. The site data presented in this analysis are specific to the region only, not the site itself. Therefore, further study needs to be conducted at the iron horse site to determine the exact geologic and stratigraphic make- up of the site. Furthermore, although the MNRC was licensed using the Universal Building Code seismic standards, it was recommended to NuTRIP that the NRC favors the IPC seismic standards. Significant seismic analysis within the framework of the new seismic code will need to be done prior to final SAR submittal. However, NuTRIP does not have access to any detailed information on how the seismic analysis for the MNRC was conducted. Finally, it should be noted that despite sub- optimal stratigraphic characteristics for the region in general, it is not prohibitive to siting research reactors in the region. For instance, the MNTR rests on a site in Cambridge Massachusetts with similar stratigraphic features as the rest of the region GEOLOGICAL ANALYSIS SURFACE AND SUB- SURFACE GEOLOGY The geologic condition of Middlesex county is underlain by late Pleistocene glaciomarine deposits [2]; generally surface clay underlain by sand and finally by bedrock. The prime geomorphic expression due to seismic activity in the area is from liquefaction of the sandy stratum between the bedrock and surface clay. Forensic geologic studies of the area from a large (estimated magnitude 4.8) earthquake in

87 led to sandblows at the surface and sand dikes in the excavations. Furthermore, the formation of two new springs and a lateral spread of marshland is thought to be due to liquefaction in the sand layer [2]. Figure 34 shows the composition of the bedrock; the white star marks the approximate location of the Iron Horse site. Note that the Iron Horse site is located on bedrock composed mainly of granite and metamorphic rock. The main geomorphic expression resulting from seismic activity in the area is due to liquefaction of the sandy sub- stratum, not catastrophic fissuring of the bedrock. Note also that the significant evidence of liquefaction resulting from the 1727 earthquake was located in northeastern- most Massachusetts, approximately 25 miles from the Iron Horse site. No evidence of liquefaction has been found nearer to the site [2]. FIGURE 34 - GEOLOGIC MAP OF MASSACHUSETTS FROM THE OFFICE OF THE MASSACHUSETTS STATE GEOLOGIST FAULT STRUCTURE No large tectonic faults exist near the site. New England resides squarely within the North American Plate with the nearest tectonic boundary several hundred miles off the Atlantic coast [3]. Fault and folds may exist near the area due to the volcanic nature of the formation of the bedrock; however, the propensity for large earthquakes due to the thrust faults sometimes encountered at tectonic boundaries 87

88 does not exist. Furthermore, precise location of the fault structure in New England has yet to be determined. The [New England] area is underlain by Paleozoic and Precambrian metamorphic and igneous rocks that were folded and juxtaposed by slip on numerous mapped faults during the assembly of the northern Appalachians (Zen and others, 1983 #1960). However, the locations of both earthquakes and faults at depth have large uncertainties. Thus, to date no New England earthquakes have been convincingly associated with known faults. [2] Figure 35 shows an estimate of the postulated location of a Quaternary fault based on historical accounts of seismic activity from Colonial times. Note that the fault location is estimated, and its features unidentified. No capable faults have been identified in the region. Furthermore, the Iron Horse site (indicated by the star) is about 25 miles from the fault region shown in Figure 35. All earthquakes in the region are classified as intraplate earthquakes, none of which are empirically tied to a capable fault. FIGURE 35 - MAP SHOWING POSTULATED FAULT REGION IN MASSACHUSETTS [4] 88

89 6.3. SEISMIC ANALYSIS Table 22 shows the approximate frequency and earthquake effects correlated to their measured value on the Richter scale [5]. TABLE 22 - EARTHQUAKE EFFECTS AND FREQUENCIES ASSOCIATED WITH RICHTER SCALE MEASUREMENTS ACCORDING TO THE USGS SEISMIC HISTORY OF REGION Figure 36 shows the seismic history of New England for earthquakes with a magnitude between 3.0 and 6.6, or minor to strong according to Table 22. The star on each map indicates the approximate location of the Iron Horse site. There have been five recorded earthquakes with an epicenter within 25 miles of the Iron Horse site since Each of these earthquakes registered less than 3.6 on the Richter scale, falling within the range of minor earthquakes - no reports of structural damage or injury could be found [6]. FIGURE 36 - HISTORICAL RECORD OF NEW ENGLAND EARTHQUAKES FROM DATA ON THE LEFT IS FROM THE USGS WHILE DATA ON THE RIGHT IS FROM THE WESTON GEOLOGICAL OBSERVATORY [6]. 89

90 The largest reported earthquake within 25 miles of the site was in Seismologists estimate its magnitude to be about 4.8 based on anecdotal accounts and geologic evidence [7]. Structural damage was reported in Cape Ann and Boston, but was confined only to chimneys and other red- brick structures REGIONAL SEISMIC HAZARD Figure 37 shows the seismic hazard map for the region encompassing the Iron Horse site. The peak ground acceleration for the region surrounding the site is 12% g, with a 2% chance in 50 years to exceed this threshold. FIGURE 37 - SEISMIC HAZARD MAP FOR REGION AROUND THE IRON HORSE SITE [8]. THE SITE IS REPRESENTED BY THE RED STAR DETERMINATION OF SEISMIC SAFETY OF SITE IDENTIFY SAFE SHUTDOWN AND OPERATION BASIS EARTHQUAKES Based on the seismic history and hazard of the Iron Horse site, the safe shutdown earthquake is identified to be an M 5.0 earthquake with a peak ground acceleration less than 20% g. There have been no reported earthquakes with a magnitude greater than 4.8 since the beginning of written records in the area (1650). Furthermore, the maximum recorded ground acceleration and peak velocity due to seismic activity in the area were 3.9 %g and 3.4 cm/s respectively since Despite these historical intensity and magnitude values, the safe shutdown earthquake was identified based on the seismic hazard shown in Figure 37, and with a significant amount of conservatism to ensure the maximum magnitude of 90

91 seismic activity would not be exceeded at the site. Fault location and length was not considered for the determination of the safe shutdown earthquake due to the lack of identified faults in the region. Based on the characterization of the safe shutdown earthquake for the site, the operation basis earthquake is determined to be an M 4.0 or above earthquake. The threshold for the operation basis earthquake is conservatively determined in order to ensure any vibratory ground motion does not interfere with the handling of radioactive materials in solution. In the event of seismic activity at or above the operation basis earthquake, post- processing will be halted until successful containment of liquid radioactive inventory is confirmed. No structural damage is expected as a result of the operation basis or safe shutdown earthquake OTHER SEISMIC DESIGN BASES SURFACE FAULTING Based on the lack of faults in the region, the effects of surface faulting need not be considered for plant design. Furthermore, due to the lack of a capable fault in the area, the site is not considered a zone requiring detailed fault investigation INDUCED FLOODS AND WATER WAVES The site is located over 20 miles from the Atlantic Coast, thus seismically induced Tsunami activity need not be considered. Similarly, the possible risk of flooding as a result of seismic activity are not likely to be great as a result of a small inventory of standing water and a lack of dams near the site. The risk of seismically induced flooding is determined to be far less than the flooding risk due to extended periods of precipitation as discussed in SOIL STABILITY Section discusses the geomorphic expression due to sub- soil liquefaction from seismic activity. Due to the stratigraphic makeup of the site and the seismic history of the site, further study of the soil stability is recommended prior to construction. The existence of a sandy layer at the site needs to be confirmed, and the risks associated with liquefaction quantified. Furthermore, topsoil and shallow ground surveys need to be taken in order to determine the surface quality. If the surface is largely clay 91

92 based and there are water retention issues, the site may require additional excavation. The site cannot be fully evaluated until these local studies are conducted PLANT SEISMIC DESIGN BASIS The seismic characterization of the Iron Horse site is completed in sections The result of the seismic characterization shows that the seismic hazard of the site is minimal. The conservatively identified safe shutdown earthquake still falls below the minimum requirement for further fault investigation as defined in Appendix A to 10 CFR 100. Nevertheless, the seismic design of the plant should be sufficiently robust as to offer a very large margin of seismic safety which may aid in the licensing of future facilities in more seismically active areas. FIGURE 38 - SEISMIC ZONES DEFINED BY THE 1997 UBC. MASSACHUSETTS IS A 2A ZONE WHILE THE MNRC IS IN A ZONE 3 [11]. The Iron Horse site and the surrounding region is classified as being in Seismic Zone 2A as defined in the Uniform Building Code (UBC). The structures for the TRIGA reactor, which are based on the MNRC design have been designed and constructed in accordance with UBC zone 3 and an importance factor of 1.5 [10]. Based on this design, there is more than ample conservatism in the design for the maximum expected event at the Iron Horse site. Adhering to the UBC zone 3 requirements ensures that the reactor can be returned to operation without structural repairs following an earthquake likely to occur during the plant lifetime (estimated ~ 80 years for conservatism). While the design of the processing and 92

93 waste storage elements of the facility have not been finalized, it is recommended that these too conform to UBC zone 3 requirements with an importance factor of 2.0. The higher degree of importance factor is due to the fact that there will be fissile material in solution in these areas of the plant. Specific design features to mitigate the consequences from externally initiated events, such as earthquakes, are discussed in more detail in chapters 2 and REFERENCES [1] Office of the Massachusetts State Geologist. (2008). The Bedrock of Massachusetts. Retrieved from [2] Perley, Sydney. Massachusetts Historic Earthquakes. (2011 March 17 th ). [3] World Atlas. (n.d.). Tectonic Plates. Retrieved from [4] USGS. (2010, April 13). Eastern U.S. Quaternary Faults. Retrieved from [5] USGS. (2010, April 13). Measuring Earthquakes - FAQs. Retrieved from [6] Kafka, A. (2008, September 16). Why does the Earth Quake in New England? Retrieved from [7] USGS. (2009, October 21). Historic Earthquakes. Retrieved from [8] USGS. (2009, October 21). Massachusetts Seismic Hazard Map. Retrieved from [9] U.S. Nuclear Regulatory Commission. (1998). NUREG- 1630: Safety Evaluation Report Related to the Issuance of a Facility Operating License for the Research Reactor at McClellan Air Force Base. [10] Uniform Building Code (Vol. I). (1997). 93

94 7. CONCLUSIONS AND RECOMMENDATIONS The neutron flux provided by a TRIGA reactor is sufficient to produce Molybdenum- 99 by causing fissions in LEU foils. Using a TRIGA reactor for production is more efficient than bombarding Molybdenum- 98 with neutrons. Having a domestic source of production eliminates the reliance the US has on foreign reactors for medical isotopes. Additionally, the US currently ship HEU foils to other countries for medical isotope, presenting a large proliferation risk. With many of the current isotope production reactors aging and nearing decommissioning, it is important for the US to develop its own means of producing Molybdenum- 99 to ensure a stable supply. We present a TRIGA reactor collocated with a Molybdenum- 99 processing facility located on a brown field in North Billerica, Massachusetts. This collocation reduces the need for wasteful production contracts and consolidates waste management. By locating on an existing brown field, there is a reduction of concern for further site contamination from solution egress accidents. Additionally, collocation allows for closed- loop pneumatic transfer between the reactor and the processing facility, eliminating the need for external handling of hot foils. Site location in Massachusetts is near numerous population centers and allows for processing and transport to occur before much of the medical isotopes have decayed. Despite the close proximity to numerous population centers, the standards set forth by 10 CFR 100 are met by the exclusion area boundary, groundwater isolation and site monitoring. The TRIGA reactor at McClellan Nuclear Radiation Center will be used to assess the feasibility of irradiating LEU foils to produce Mo- 99 in order to streamline the licensing process. The primary concern with the TRIGA design is not exceeding a positive reactivity insertion of $0.94 in order to be compliant with the NRC recommendation of $1.00 for MNRC. Based upon conservative calculations, it has been determined that LOCA would not exceed the limit of 25 rem release set for DBEs. Expected releases would be compliant with the limits set forth by 10 CFR 20 and meet NRC licensing requirements. By using a design conforming to the Uniform Building Code Zone 3, there is a large seismic safety margin relative to the seismic characteristics of the site. Additionally, the non- seismic hazards for the proposed site are well characterized. The TRIGA reactor falls under the NRC framework for research reactors and will be licensed under 10 CFR as a class 104 reactor. The emergency planning zone for this design is minimal and numerous other TRIGA designs are currently located on college campuses. The identified accident scenario risks were well below the limits set forth in 10 CFR 20 and 10 CFR 50. Lastly, we recommend that the NRC develop expertise in licensing nuclear reactors for medical isotope production. With the projected increase in demand for Molybdenum- 99 and unstable foreign supply, it is important to develop this expertise to avoid supply shortcomings for medical procedures in the not so distant future. 94

95 APPENDIX A. REPORT AUTHORS SUMMARY OF CONTRIBUTIONS BY Name TABLE 23 - LIST OF AUTHORS AND THEIR CONTRIBUTIONS TO THIS REPORT Graduate / Undergraduate Main Contributions Ross Barnowski Graduate 2.4. Site Description 3.2. AOO and DBE for Processing Facility 4.2. BDBE for Processing Facility Re- criticality of Solution 6. Site Safety and Seismic Analysis Appendix B. Detailed Calculations Appendix C. Comparison of Mo- 99 Production Methods Joseph Miller Undergraduate 5. Risk Assessment and Management Kevin Tirohn Nicolas Zweibaum Undergraduate Graduate 1. Introduction 7. Conclusions and Recommendations 2.1. Reactor Design Decisions 2.2. Feasibility Study Irradiation at McClellan Nuclear Radiation Center 2.3. Irradiated Target Processing 3.1. Design Basis Event during Operation of a TRIGA Reactor: Loss of Coolant Accident (LOCA) 4.1. Beyond Design Basis Event during Operation of a TRIGA Reactor: Maximum Hypothetical Accident (MHA) 95

96 APPENDIX B. DETAILED CALCULATIONS B.1. ESTIMATION OF LIFE LOST DUE TO Tc- 99m SHORTAGE B.2. ACTIVITY RELEASE OF Kr

97 B.3. ESTIMATED CRITICAL RATIO BASED ON THE 4- FACTOR FORMULA 97