3

Reactor Conversion Case Studies

Session 3 of the symposium focused on technical challenges associated with conversion of specific U.S. and Russian reactors. Eight case studies of individual research reactors’ potential for conversion—three U.S. reactors and five Russian reactors—were presented in this session. These presentations and some key thoughts from the participant discussions are summarized in this chapter.

As was discussed in Chapter 2, there are several analyses that need to be performed to enable conversion of a research reactor from HEU fuel to LEU fuel:

1. Neutronics analysis1 is performed to determine neutron fluxes in various regions of the new LEU core, reactivity effects, including burnup effects, and various reactor safety parameters.

2. Thermal and hydraulic analysis is performed to ensure that the new LEU core can be adequately cooled during normal and accident conditions.

3. Accident analysis is performed to analyze the potential for fission product release under hypothetical accident conditions.

Because of the uniqueness of many research reactors, conversion studies

________________

1“Neutronics” refers to an analysis of the neutron flux throughout the core, which entails analysis of fission and neutron capture events caused by absorption of neutrons by the reactor core, scattering of the neutrons, and losses of the neutrons from the reactor.



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3 Reactor Conversion Case Studies S ession 3 of the symposium focused on technical challenges associated with conversion of specific U.S. and Russian reactors. Eight case stud- ies of individual research reactors’ potential for conversion—three U.S. reactors and five Russian reactors—were presented in this session. These presentations and some key thoughts from the participant discussions are summarized in this chapter. As was discussed in Chapter 2, there are several analyses that need to be performed to enable conversion of a research reactor from HEU fuel to LEU fuel: 1. Neutronics analysis1 is performed to determine neutron fluxes in various regions of the new LEU core, reactivity effects, including burnup effects, and various reactor safety parameters. 2. Thermal and hydraulic analysis is performed to ensure that the new LEU core can be adequately cooled during normal and accident conditions. 3. Accident analysis is performed to analyze the potential for fission product release under hypothetical accident conditions. Because of the uniqueness of many research reactors, conversion studies 1 “Neutronics” refers to an analysis of the neutron flux throughout the core, which entails analysis of fission and neutron capture events caused by absorption of neutrons by the reactor core, scattering of the neutrons, and losses of the neutrons from the reactor. 61

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62 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS need to be carried out for each individual reactor, and the challenges en- countered can be very different for different reactors. U.S. REACTOR CONVERSION CASE STUDIES The following three case studies of U.S. research reactor conversions are summarized in this chapter: • Paul Wilson (University of Wisconsin) reported on the success- ful conversion of the University of Wisconsin research reactor (UWNR) (Wilson, 2011). • Thomas Newton (Massachusetts Institute of Technology) reported on the status of conversion plans for the Massachusetts Institute of Technol- ogy Reactor (MITR) (Newton, 2011). • David Cook (Oak Ridge National Laboratory; ORNL) reported on the status of conversion plans for the High Flux Isotope Reactor (HFIR) (Cook, 2011). These reactors are quite different: MITR is planned to be the first research reactor to convert to using high-density uranium-molybdenum (UMo) monolithic LEU fuel and is considered to be a relatively straight- forward conversion for a high-performance reactor. In contrast, HFIR is planned to be the last U.S. domestic reactor to convert to LEU fuel and is likely to pose far greater conversion challenges. Current approaches and plans for converting these reactors are described in the following sections. University of Wisconsin Nuclear Reactor Paul Wilson UWNR is a 1 megawatt (MW) TRIGA pool reactor (see Chapter 1) housed on the University of Wisconsin campus in Madison, Wisconsin. Its primary mission is the training of undergraduate and graduate nuclear engineering students; however, it is also used to perform research, including irradiation for neutron activation analysis. The reactor first went critical as a 10 kilowatt (kW) LEU-fueled reac- tor in 1961 and, following several power upgrades, was converted to HEU fuel in 1979. It was converted back to LEU fuel 30 years later, successfully achieving criticality in 2009. At that time, UWNR was converted from us- ing 70 percent enriched TRIGA-FLIP (Fuel Life Improvement Program) fuel to TRIGA LEU 30/20 (30 percent uranium by weight, 20 percent enriched) fuel. The new LEU fuel is, like the previous FLIP fuel, a standard TRIGA-

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63 REACTOR CONVERSION CASE STUDIES type fuel element containing erbium-doped uranium-zirconium hydride (UZrHx-Er) fuel (see Chapter 2). Neutronics Analysis A number of key neutronics analyses were performed for a range of reactor core states, including the beginning-of-life, middle-of-life, and end- of-life states. These studies included analyses of: • Power distributions (for use in the thermal/hydraulic analyses), in- cluding (1) total fuel assembly power and core power distributions; and (2) axial and radial power distributions in the maximum power fuel assembly; • Shutdown margins as a function of fuel burnup; and • Key reactivity parameters, including (1) delayed neutron fraction2; (2) prompt neutron lifetime3; (3) control element worth4; and (4) prompt temperature coefficient.5 The neutronics of the reactor core were modeled using Los Alamos National Laboratory’s Monte Carlo n-Particle code, version 5 (MCNP5) with the core nuclear reaction database ENDF/B-VII maintained by the National Nuclear Data Center. In addition, Argonne National Laboratory’s REBUS codes for analysis of fuel cycles were used for the burnup analysis. Finally, some confirmatory analysis was performed using the HELIOS two- dimensional generalized-geometry lattice physics transport code.6 Several challenges were associated with performing these analyses at Wisconsin. First, sufficient information was not available on the operational history of the HEU core to be able to calculate fuel composition for use in benchmarking the model. As a substitute, analysts worked backwards to estimate the composition of the fuel using the current critical conditions for the core. This does not provide a benchmark but gives some confidence in the validity of the model. Second, large computing resources were required 2 Delayed fission neutrons are neutrons emitted spontaneously from decay of a fission product from a prior fission event, whereas prompt neutrons are neutrons emitted from the fission process directly. The delayed neutron fraction is the ratio of the mean number of de- layed fission neutrons per fission to the mean total number of neutrons per fission (prompt plus delayed). 3 The prompt neutron lifetime is the average time between the emission of neutrons and either their absorption in or their escape from the system. 4 The control element worth is the negative reactivity change caused by inserting a control element into the reactor. UWNR has five separate control elements. 5 The prompt temperature coefficient is the change in reactivity per degree change in fuel temperature. 6 This confirmatory analysis was a two-dimensional deterministic analysis coupled with a one-dimensional diffusion approximation.

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64 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS for some of the analyses, beyond what was easily available at the university. Finally, the university had only a modest existing capacity for performing such reactor analyses. This capacity had to be built up for the analyses to be carried through successfully. The major difficulty associated with conversion was related to system reactivity. The like-for-like replacement of HEU-FLIP fuel with LEU 30/20 fuel increased the reactivity of the core. The modeled core with LEU fuel could not be shut down even with all control elements fully inserted. To reduce system reactivity and meet shutdown margin requirements, the core design was changed from a 23 fuel assembly/10 reflector configura- tion (in which the assemblies are arranged in an “H” pattern) to a 21 fuel assembly/14 reflector configuration (in which the assemblies are arranged in an “X” pattern) (see Figure 3-1). However, the reduction in the number of fuel assemblies resulted in a slightly reduced core lifetime following conversion.7 Thermal/hydraulic Analysis The thermal/hydraulic analysis focused on the behavior of the high- power channel at steady state, low-power pulse, and high-power pulse.8 The analysis yielded estimates of: • Coolant flow rate; and • Temperatures at the fuel centerline, the axial/radial temperature profile, and the minimum departure from nucleate boiling ratio (DNBR).9 The U.S. Nuclear Regulatory Commission’s (USNRC’s) RELAP5/ Mod3.3 code was used to perform the thermal/hydraulic analysis. A single channel analysis was performed with the highest-power channel, involving 20 axial nodes (15 in the fuel meat) and 27 radial nodes (21 in the fuel meat). To model the reactor pulsing mode, a two-channel model was used, with the two channels defined as (1) the hot channel and (2) the rest of the core. For the pulsing analysis, a RELAP point reactor kinetics model was used, with temperature coefficients obtained from the MCNP5 analysis that was described previously. Finally, a two-channel model was used to model 7 When preparing for conversion, the University of Wisconsin was provided with two additional LEU fuel assemblies so that the reactivity could be boosted if required. 8 This analysis assumed no cross-flow, i.e., no exchange of coolant with adjacent fuel or reflector assemblies. 9 DNBR is the ratio of the heat flux needed to cause departure from nucleate boiling to the actual local heat flux of a fuel rod. Departure from nucleate boiling is the point at which the heat transfer from a fuel rod rapidly decreases because of the insulating effect of a steam blanket that forms on the rod surface when the temperature continues to increase (USNRC, 2011).

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FIGURE 3-1 Core map of the University of Wisconsin reactor before (left) and after (right) conversion from HEU to LEU fuel. Fuel elements are shown in red, and beryllium reflector elements are shown in grey. SOURCE: Austin (2010). Figure 3-1.eps bitmap, landscape 65

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66 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS a loss-of-coolant accident using three phases to represent different water levels remaining in the pool and assuming axial conduction in the fuel. The thermal/hydraulic analysis faced four major challenges. First, the analysis was very sensitive to gap thickness, so additional sensitivity analy- ses needed to be carried out. Second, a discrepancy was found between the two critical heat flux correlations used to analyze the natural circulation mode.10 Third, there was some uncertainty in the natural convection heat transfer models. Finally, it was challenging to determine appropriate air- cooled temperature safety limits for the new LEU 30/20 fuel type. The overall outcome of the thermal/hydraulic analysis was encourag- ing. The average fuel assembly power increase associated with the use of fewer assemblies caused small changes to appear in the models of the steady-state operation of the reactor following conversion. However, the definition of the fuel temperature-limiting safety setting11 was updated, ensuring that the fuel temperatures remained below the set points. The temperatures were calculated to be within technical specifications for the reactor: The maximum fuel temperature under pulsing operation at 1 kW and at 1.3 MW was calculated to be within maximum allowable tempera- tures, and the loss-of-coolant fuel temperatures were less than 700 oC. Accident Analysis The potential for a fission product release under accident conditions was analyzed for a maximum hypothetical accident consisting of cladding failure in the high-power fuel assembly (25 kW) after continuous full-power operation. The accident analysis was carried out with and without an intact water pool and operating ventilation system.12 Reactivity insertion was also analyzed. Additionally, a loss-of-cooling accident was analyzed to determine the fuel temperature and radiation dose from the exposed core. The accident analysis used Oak Ridge National Laboratory’s ORIGEN code to calculate the fission product inventory in case of accident. An analy- sis of release fractions used a Gaussian plume model, and radiation doses were calculated using MCNP5. 10 This is the relationship between the conditions in a heated channel and the heat flux at which the heat transfer becomes impaired as a result of the transition from nucleate boiling to film boiling. These conditions may include the mass flow rate, channel geometry, and thermal properties of the fluid (e.g., the density of liquid and vapor, heat of vaporization, specific heat). The critical heat flux correlations used in this analysis were the Groeneveld 2006 look up tables and the Bernath correlation. More information on these correlations can be found in Vitiello (2008). 11 The fuel temperature-limiting safety setting is the temperature below which the fuel is required to be maintained to prevent fuel element failure. 12 A “near maximum hypothetical accident” maintains an intact pool and ventilation.

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67 REACTOR CONVERSION CASE STUDIES The accident analysis had an overall positive outcome. LEU conversion required no changes in response to any accident. The reactor remained within regulatory limits under all variations to the maximum hypothetical accident. This analysis had another positive benefit: The university’s capabilities to analyze core accidents increased significantly; previously, only simple methods and models had been used to analyze such accidents. As a result, a more detailed understanding of the potential radiation dose was gained, including the time-dependent behavior and the spatial distribution of dose. Results of Conversion and Future Plans Although the overall conversion experience was positive, the converted reactor core behaved somewhat differently than the calculated core. In particular, the converted reactor was substantially less reactive than was calculated. The reason for this difference is still not fully understood. In the near term, the UWNR staff is pursuing a plan to shuffle the core and reduce the number of reflectors. This will cause a slight reduction of the neutron flux in some positions; however, the reshuffling should increase the flux in other positions. This reshuffling would take the core from the “X” configuration of 21 fuel assemblies and 14 reflectors to a “+” configuration of 21 fuel assemblies and 6 reflectors (see Figure 3-2). Overall, the conversion-related work enabled a widespread upgrade in UWNR staff’s analysis capabilities, and it has also provided opportunities for further analysis. For example, experimental research is ongoing to better understand natural circulation heat transfer in TRIGA-relevant conditions, and the fresh LEU core provides a wide variety of benchmark data for continuing to improve analytical capabilities. Massachusetts Institute of Technology Reactor Thomas Newton MITR is a 6 MW research reactor that is currently operating using aluminide (UAlx) dispersion fuel that is 93 percent enriched in uranium-235. Its primary mission is research, although it is also used for student train- ing, particularly for nuclear engineers. The research performed at MITR focuses primarily on fast neutron experiments, including irradiation testing of cladding for next-generation light-water reactors and advanced nuclear fuel experiments. The reactor core is highly compact and has a hexagonal geometry with 27 fuel assembly positions. Twenty-four of these positions contain fuel; the remaining three positions are reserved for experiments (see Figure 3-3). The

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68 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS Figure 3-2.eps FIGURE 3-2 Planned future core map for the UWNR reactor. Fuel elements are shown in red, beryllium reflector elements are shown in grey, and white boxes are bitmap empty positions. SOURCE: Austin (2010). FIGURE 3-3 Overhead view of the MITR reactor core. The 27 fuel assembly posi- tions are labeled A-1 through C-15. Twenty four of these positions hold fuel. A fuel element is shown in dark blue in position C-9. SOURCE: Newton (2011). Figure 3-3.eps bitmap

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69 REACTOR CONVERSION CASE STUDIES FIGURE 3-4 MITR’s unique finned fuel elements. A complete fuel assembly consists of 15 stacked fuel plates in an Figure 3-4.eps fins can be seen on the indi- aluminum shell. The vidual fuel plate on the right. The 2 bitmaps percent enriched UAlx dispersed fuel meat is 93 in aluminum. SOURCE: Newton (2011). aluminum-clad fuel plates (15 per assembly) are designed with longitudinal fins to increase the heat transfer area (see Figure 3-4). The thermal and fast neutron fluxes in the core region are approximately 3 × 1013 and 1 × 1014 neutrons per square centimeter per second (n/cm2-s), respectively. The core is light-water cooled and moderated, with six control blades located around the periphery. MITR has not yet been converted to LEU fuel because an appropriate fuel has not yet completed development and qualification. In fact, MITR’s unique fuel assembly design and highly compact core complicate conver- sion. Currently available LEU fuels were judged not to be appropriate for use in MITR because they would not allow criticality to be maintained and would also require a complete redesign of the core. However, the use of high-density UMo monolithic LEU fuel (discussed in Chapter 2) is likely to allow conversion of the reactor core to LEU. It is the reference fuel used in the conversion analyses. This fuel is 19.75 percent enriched in uranium-235 and has a density of 15.5 grams of uranium per cubic centimeter (gU/cm3). Neutronic and Thermal/hydraulic Analyses A major challenge for MITR is to convert to LEU while still meeting the performance requirements for experiments in the reactor. Meeting these re- re- quirements will entail optimizing the fuel design to maximize heat transfer

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70 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS and neutron flux. In particular, the neutron flux optimization is focused on maintaining HEU-equivalent fast neutron fluxes in in-core materials experi- ments and thermal neutron fluxes in out-of-core experiments. To prepare for conversion, the existing neutronics and thermal/hydraulics models for the MITR core were improved in several ways. Several major improvements were made to the neutronics codes. The primary change enabled more accurate burnup modeling and benchmark- ing. The improvements entailed an extensive review of the model’s core structure and dimensions as well as an update of the cross-section li - braries, homogenized volume fractions, and discrete structures. Two ini- tial HEU cores were modeled, and the results compared favorably with measurements. The neutronics codes were improved in other ways as well. A graphi- cal user interface was added, as was the capability to model HEU, LEU, and mixed core geometries. In addition, it is now possible to model all fuel movements, including flipping, rotating, and fuel storage. The models are also now able to track and plot the mass of isotopes as well as the power distribution in the core. The improved burnup modeling has shown good results. Twelve recent cores have been modeled; the results were in good agreement with measured beginning- and end-of-cycle control blade positions. There was also good agreement among different models. In addition to the neutronics modeling, thermal/hydraulics models were also updated and modified. Specifically, the models were modified to include the fuel’s longitudinal fins for the steady-state and loss-of-flow analyses. Initial results have shown that the LEU design core has a higher margin to onset of nucleate boiling and a lower peak cladding temperature with loss of flow. The results of the neutronic and thermal-hydraulic analyses have been used to design an LEU fuel for this reactor. The LEU fuel assembly will contain more plates and use a thinner fuel meat (0.51 mm for UMo LEU fuel versus 0.76 mm for HEU fuel) and cladding (0.28 mm for UMo LEU fuel versus 0.35 mm for HEU fuel). Fuel developers have informed MITR staff that fuel and cladding of this thickness is feasible to manufacture. As was noted in Chapter 2, the reactor’s operational power will need to be increased from 6 MW to 7 MW to counter the expected loss of per- formance after conversion. This will result in an increase in the cycle length from 40-50 days for the HEU fuel to 60-70 days for the LEU fuel. Safety Analysis Prior to beginning the conversion analysis, some safety analysis param- eters at MITR were not well known, particularly for the finned cladding.

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71 REACTOR CONVERSION CASE STUDIES To complete the safety analysis, further information is being gathered on the following three topics: • Finned channel hydraulic pressure drop. A flow experiment has been built, measurements have been made, and a finned channel correla- tion13 describing the relationship between the pressure drop and the flow rate has been developed. • Adequacy of the onset of nucleate boiling (ONB) correlation for finned channels.14 For this purpose, the MITR boiling flow loop is being constructed to measure ONB for the LEU channel geometry and validate the Bergles-Rohsenow ONB correlation15 for finned channels. This facility will be operational later this year. • Oxide distribution in the finned cladding. The current burnup limit of 1.7 × 1021 fissions per cubic centimeter is based on an even 50 microm- eter-thick aluminum oxide distribution on the cladding. However, particu- larly within the finned region, the actual oxide distribution is unknown. MITR is currently using an eddy current probe for fin-tip measurements of oxide thickness. This thickness will then be compared with the operational history of the fuel element. Finally, a selected fuel element will be shipped to Idaho National Laboratory for evaluation of the oxide distribution in the areas between the fins. Other Potential Challenges Aside from the challenges discussed above, there are several other po- tential challenges that MITR will face during the transition to an LEU-fueled core. First, MITR is likely to be the first reactor to convert using UMo monolithic LEU fuel. MITR staff is presently working to better understand how best to introduce this first-of-a-kind fuel into the reactor. The current plan is to gradually introduce LEU fuel into the HEU core. To evaluate this plan, a mixed-core analysis will be carried out prior to conversion. Two challenges are foreseen in the mixed-core transition: Power peaking is gener- ally higher in new LEU elements, and steady-state HEU and LEU margins to ONB decrease with an increasing number of LEU fuel elements in the mixed 13 This is the relationship between the pressure drop and the conditions in the flow channel such as mass flow rate. 14 This is the relationship between the conditions in the flow channel at the time when there is net vapor generation. It is a relationship between parameters such as mass flow rate, heat flux, and thermal properties of the liquid (e.g., thermal conductivity, specific heat, saturation temperature). 15 The Bergles-Rohsenow correlation is commonly used for prediction of ONB in narrow rectangular coolant channels and relates cladding local heat flow at ONB to local heat flux and pressure.

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78 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS • Yu.A. Tzibulnikov (Tomsk Polytechnic Institute) discussed the con- version potential of IRT-T (Tzibulnikov, 2011). • E.A. Kryuchkov (MEPhI) discussed the conversion potential of IRT (Kryuchkov, 2011). Because the feasibility studies of these reactors were at an earlier stage of development than the U.S. studies when the symposium was held, less detail is provided in presentation summaries than was given for the U.S. reactor conversions. MIR.M1 V.A. Starkov The MIR.M1 reactor is a 100 MW pool-type research reactor located at RIAR in Dimitrovgrad. It has a maximum thermal neutron flux at the experimental positions of 5 × 1014 n/cm2-s. Its primary mission is to test experimental fuel assemblies and fuel rods under normal, abnormal, and accident conditions. The core and beryllium reflector blocks are stacked in a hexagonal grid comprising 127 hexagonal blocks 148.5 mm in size, installed at a pitch of 150 mm (see Figure 3-7). Four central rows of beryllium blocks operate as Figure 3-7.eps FIGURE 3-7 Diagram of the MIR.M1 reactor core. The core is composed of hex- agonal beryllium blocks with channels cut through their centers. Individual fuel bitmap assemblies can be seen (silver circles), as can experimental positions (black and white). Each experimental position is surrounded by six fuel assembly channels. SOURCE: Starkov (2011).

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79 REACTOR CONVERSION CASE STUDIES FIGURE 3-8 Diagram of a MIR.M1 fuel assembly. Each fuel plate is cylindrical Figure 3-8.eps and has a fuel meat thickness of 0.56 mm and a cladding thickness of 0.72 mm. bitmap SOURCE: Starkov (2011). a moderator, and two external rows of beryllium blocks act as a neutron reflector. The core also contains 11 loop channels where experiments are placed. Each experimental channel is surrounded by six fuel assemblies to maximally isolate each experiment from neighboring experiments. Each fuel assembly consists of four cylindrical fuel tubes arranged concentrically (see Figure 3-8). Absorbing rods are located along the edges of the blocks. For every channel there are two to three such absorbers for a total of about 30. This core design is very flexible and allows for the simultaneous irradia- tion of multiple experiments in different power regimes. Potential and Plans for Conversion The MIR.M1 reactor has had a long-running research program focused on HEU minimization. In addition, further work is being undertaken as part of the contract (described previously) that was recently signed with the United States to study the feasibility of converting MIR.M1 from HEU to LEU. If MIR.M1 is converted from HEU to LEU, several key performance characteristics will need to remain the same to allow the reactor to continue

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80 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS to fulfill its main missions. The thermal neutron flux to the experiments cannot be degraded, and the reactor power (100 MW) and campaign dura- tion (30 days) will also need to remain constant. Two fuel types were considered as candidates for converting the MIR. M1 reactor: (1) a UMo dispersion LEU fuel (described in Chapter 2), and (2) a uranium dioxide (UO2) dispersion LEU fuel (the existing technology). A uranium silicide fuel type was considered at an earlier stage but was ruled out because the technology for producing UO2 and UMo dispersion LEU fuels is better understood in Russia.19 UMo dispersion LEU fuel is the most likely candidate for conversion of the MIR.M1 reactor. Recent calculations have shown that to retain the re- quired performance characteristics after conversion, the density of uranium in the core will need to be higher than is possible technologically for UO2 LEU fuel but that is obtainable using UMo dispersion LEU fuel. UMo dispersion LEU fuel has been tested extensively in Russia. Dif- ferent material compositions (e.g., additions of silicon to the aluminum matrix) as well as different fuel fabrication technologies have been tested both with and without coatings. The results have been positive, particularly when the fuel is coated with titanium nitride. Four tests on full-scale as- semblies have been performed so far—primarily to validate the conversion of the research reactor in Tashkent, Uzbekistan—and the findings have been reported by Russian scientists at conferences on enrichment reduction (Chernyshov et al., 2002). Post-irradiation materials science studies have been performed and are still ongoing. MIR.M1 staff has found that changes in the thermal loading will require the fuel assemblies to be changed slightly from the original HEU design. Preliminary analysis has shown that using UMo dispersion LEU fuel is feasible if the fuel meat thickness is increased from 0.56 mm to 0.94 mm. Under this scenario, the annual fuel consumption for LEU would be four times higher than for HEU, but the number of fuel assemblies used would decrease by a factor of approximately 1.75. Overall, it appears that the quality of the core can be improved by us- ing UMo dispersion LEU fuel and changing the fuel meat thickness. The next stage of the feasibility analysis will involve verification using precision programs. However, some outstanding problems remain to be solved for the UMo dispersion LEU fuel before adopting it for use in MIR. RIAR (working in collaboration with Argonne National Laboratory) expects to complete the feasibility study for MIR.M1 by the end of 2011. 19 In addition, Dr. Starkov stated during the symposium that he believed there have been some problems in reprocessing silicide fuels. As was noted previously, Russia reprocesses its research reactor fuel unlike in the United States.

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81 REACTOR CONVERSION CASE STUDIES Argus V.A. Pavshuk The Argus reactor at the Kurchatov Institute in Moscow is one of three HEU-fueled research reactors at the Institute to be included in the U.S.- Russia conversion feasibility study agreement.20 The Argus reactor is a 20 kW light-water cooled and moderated solution reactor with a core volume of 22 liters of UO2SO4 solution containing 1.71 kg of 90 percent enriched uranium. The reactor is used for neutron radiography, neutron activation analysis, and production of isotopes and nuclear filters. Potential and Plans for Conversion Concrete plans are in place to convert the Argus reactor to LEU fuel. At present, work is underway to assess the feasibility of converting the reactor from HEU to 17.5 percent enriched LEU. The neutronics and thermal hy- draulics calculations have been completed and will be sent to the customer by the end of 2011. Once this has been done, the documentation will need to be completed, the fuel will need to be qualified, and a license to oper- ate with LEU fuel will need to be obtained. These activities are planned to occur in 2012. Also throughout 2012, Argus staff will begin preparations for reloading the reactor with LEU fuel. Presently, reloading is planned to be completed by the end of 2012. After reloading, the reactor will restart and a safety validation will be performed through 2013. The conversion is planned to be completed in 2014. IR-8 V.A. Nasonov IR-8 is a pool-type reactor operating at 6 MW (but rated to 8 MW) with 90 percent enriched HEU fuel. The 60-cm-high IR-8 reactor core contains 16 IRT-3M fuel assemblies with a beryllium reflector. There are 12 horizontal experimental channels and a number of vertical experimen- tal channels located in the core, the reflector, and the vessel. IR-8 is used to perform research in a broad range of fields, including nanotechnology, materials science, solid-state physics, nuclear physics, and medicine. 20 The other reactors are IR-8, discussed in the next section, and OR (OP-M in Table 1-2), a 0.3 MW pool-type reactor. See Chapter 2.

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82 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS Potential and Plans for Conversion Maintaining the current fast and thermal neutron fluxes at IR-8 after conversion is essential for carrying out the facility’s primary missions. Although the reactor is currently operating at 6 MW, there are plans in the near future to increase power to the rated 8 MW. If neutron fluxes are insufficient after this power increase, further uprating will not be pos- sible; although the reactor can in principle operate to 30 MW, its power is restricted because of its location in the heart of Moscow. Maintaining the high fluxes with limited power will require a high-density LEU fuel. Kurchatov has proposed to convert this reactor using IRT-3M fuel as- semblies with UMo dispersion LEU fuel having 19.7 percent enrichment. Similar to the MIR.M1 reactor, two options were initially considered for transitioning to LEU: the IRT-3M assemblies with UMo dispersion LEU fuel, and IRT-4M fuel assemblies with UO2 fuel. The IRT-4M assemblies were determined to be inadequate to maintain the needed neutron fluxes because the 3 gU/cm3 uranium density in the UO2 fuel is too low. Although it is possible to increase the uranium density in the UO2 fuel to high-enough levels to obtain the needed fluxes, fuel reliability is likely to decrease. Previous experiments were conducted with an enhanced UO2 density (3.85 gU/cm3) in the IRT-3M fuel; however, some fuel elements failed after being installed in the reactor for testing, and the satisfactory fuel elements achieved burnups of only 40 percent. Kurchatov has developed a set of full-scale models to describe the reactor geometry in detail, including the reflector, core, fuel elements, and experimental channels. The staff at Kurchatov has also carried out a neu- tronics analysis to assess the feasibility of conversion. This analysis relies on recently developed Monte Carlo codes (MCU-PTR codes with the data- base MDBPTR50, along with ASTRA for thermal hydraulics calculations). These codes were benchmarked during reactor operation by measuring the isotopic composition of materials in the core and reflector. The Kurchatov Institute (working in collaboration with Argonne Na- tional Laboratory) expects to complete the feasibility study for IR-8 by the end of 2011. IRT-T Yu.A. Tzibulnikov The IRT-T reactor at Tomsk Polytechnic Institute (TPU) in Tomsk is a 6 MW21 pool-type research reactor. The reactor is primarily used for training 21 Tomsk Polytechnic Institute has recently requested that this reactor be licensed to operate at 11 MW.

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83 REACTOR CONVERSION CASE STUDIES engineers and managerial staff for nuclear power facilities as well as other specialists, but it also supports a significant amount of research, particularly industrial research. As of the time of the symposium (June 2011), up to a third of the experiments performed at IRT-T were for industrial purposes, which allowed the reactor to operate as a source of revenue for TPU. The facility plans to become financially self-sustaining between 2015 and 2017. In 2011, TPU began implementing a growth and development program for nuclear physics research. The institute also maintains a silicon alloy laboratory (for silicon doping studies), a laboratory devoted to radioactive pharmaceuticals, and a laboratory devoted to instrumental neutron activa- tion and analysis. A large part of the industrial production involves silicon doping, but the facility also produces some medical isotopes, including molybdenum-99. Because of the importance of industrial revenue to the operation of the reactor, it is essential to understand how the conversion might affect the technological capability of IRT-T before proceeding—not simply the cur- rent capability, but also projected increases in technological capability. For example, current plans call for an improved instrumental and technical base to boost production to 8 tonnes per year of doped silicon from the current production of 2 tonnes (with a current capacity of 4 tonnes). Potential and Plans for Conversion TPU has an agreement with Argonne National Laboratory to perform computations to study the feasibility of converting IRT-T to LEU fuel. At the same time, TPU is performing independent calculations. In early 2012, TPU plans to complete an analysis comparing the performance of IRT-T using both HEU and UMo dispersion LEU fuel. At present, only initial computations and analytical work on the potential impacts of reactor con- version have been performed. The results of these initial computations have been alarming. The use of UMo dispersion LEU fuel results in a harder neutron spectrum compared to HEU fuel, which could create problems for silicon doping applications as well as for the production of radiopharmaceuticals. Another issue of concern is the potential for higher fuel costs for LEU relative to HEU. TPU purchases its own fuel, so a significant increase in fuel costs could negatively impact revenues. IRT E.A. Kryuchkov The IRT reactor at MEPhI in Moscow is a 2.5 MW pool-type research reactor. The IRT facility was designed primarily as a student training facility

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84 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS and secondarily to conduct a wide range of research activities. For example, MEPhI performs scientific experiments for producing short-lived isotopes, tests sensors for power stations, and hosts medical physics research, par- ticularly the development of equipment for neutron therapy. Because of the relatively low neutron flux densities (described below), materials testing and industrial-scale isotope production are not performed at this facility. Beyond the training and research missions, MEPhI also hosts visits to the facility by members of the public. These visits are intended to improve public relations and demonstrate the safety and reliability of the reactor. The IRT reactor uses IRT-3M type fuel enriched to 90 percent ura- nium-235. The reactor has a maximum fast neutron flux in the core of 4.3 × 1013 n/cm2-s and a maximum thermal neutron flux in the core of 4.8 × 1013 n/cm2-s. The reactor uses a beryllium reflector.22 IRT has 48 vertical experimental locations, with 6 of these locations occupied by fuel. IRT also has 10 horizontal experimental channels, allowing for a range of training and scientific work to be conducted (see Figure 3-9). Two horizontal experimental channels are currently being used for neutron therapy. The first is used to irradiate animals, and the second is being reconfigured for human testing. These channels require specific pa- rameters that would not be changeable if the reactor were to be converted to LEU fuel. In particular, it is important to MEPhI to address the following issues in converting to LEU fuel: • Ensure that safety parameters will continue to conform to existing regulations. The radiation safety parameters for IRT are set to be stricter than for many other research reactors because of the public tours. This will continue to be true after conversion. • Retain current neutron fluxes. Although IRT is not used to produce isotopes—meaning that the neutron flux in the vertical channels is not the key parameter—it is important to maintain the capabilities required for neutron capture therapy and to enable research by maintaining current neutron flux densities in some locations. • Estimate the annual operation costs when working with LEU. Op- erating funds for the IRT reactor are limited, so economic parameters will be essential when considering conversion. 22 Aluminum is now being used rather than beryllium in a number of locations because of swelling of some of the beryllium blocks.

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85 REACTOR CONVERSION CASE STUDIES Figure 3-9.eps FIGURE 3-9 Core of the IRT research reactor. The diagram on the left shows the IRT core from overhead, and the diagram on the right shows the core from the side. bitmap The fuel assembly positions (1) are shown as yellow boxes; the fuel assemblies with channels for control rods (2) are shown as yellow boxes with gray circles (3) at the center, representing the control rods; the beryllium reflector positions (6) are shown as blue boxes; and the aluminum reflector positions (7) are shown as green boxes. The large green object on the left of each diagram is aluminum, and the aquamarine object to the left is graphite. SOURCE: Kryuchkov (2011). Potential and Plans for Conversion MEPhI is in the early stages of conversion analysis for the IRT facil- ity. An initial neutronics analysis of the HEU core has been completed and further analysis on the neutronics and thermal-hydraulics of the core is currently under way. For these two tasks, MEPhI has used an application developed within its institute and qualified by the Russian nuclear regulator. Using this application, the safety, experimental performance, and fuel assembly consumption parameters of an HEU core were determined for comparison with the proposed LEU core. The key parameter used for the analysis is the neutron flux density in two channels. MEPhI staff has modeled the reactor’s lifetime (with the HEU core) in great detail, including the isotopic composition of every unloaded bundle. MEPhI staff has determined that high-density UMo dispersion LEU fuel will be necessary to successfully convert the IRT reactor, as is the case for most of the other Russian reactors discussed in this chapter. However, staff

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86 CONVERTING U.S. AND RUSSIAN RESEARCH REACTORS remains concerned about the economic uncertainties associated with using this fuel. At this time, UMo dispersion LEU fuel has not yet been licensed in Russia, so there is no answer at this time as to what the allowed burnup will be. The IRT reactor has been in operation for 44 years and is in need of some refurbishment. For example, the beryllium reflectors should be re- placed and soon the control rods will also need to be replaced. Although the need for the reactor refurbishment is not directly connected with the con- version to LEU fuel, modifying the fuel enrichment without updating the reactor would be problematic; it would be best to combine these two tasks. DISCUSSION Following the individual case study briefings some time was set aside for free discussion among the workshop participants. The major points made by individuals (sometimes multiple individuals) over the course of this discussion are summarized in the paragraphs below. • Much work remains to be done to convert Russian reactors. The feasibility of converting from HEU fuel to LEU fuel has been studied for a number of Russian reactors, but some participants noted that a significant amount of work still remains to be done to successfully convert them. On the other hand, it was also noted that although the United States has suc- cessfully converted a number of domestic reactors, challenges still lie ahead as the United States continues to research what will be needed to convert its high-performance research reactors. • Fuel development activities in both the United States and Russia are progressing quickly. Throughout the first two days of the symposium, particularly during discussions of the case studies, one of the most fre- quently discussed issues involved fuel development. It was noted that the focus of U.S. efforts was on the development and qualification of UMo monolithic LEU fuel with densities of up to 15.5 gU/cm3. Efforts in Russia are focused on the development and qualification of UMo dispersion LEU fuel with densities of more than 5 gU/cm3. Argonne National Laboratory has been working closely with the Bochvar Institute to develop and qualify UMo dispersion LEU fuel for use in Russian research reactors. The opin- ion was expressed by several individuals on the U.S. side that Russian fuel development is progressing quickly. • The technical feasibility of reactor conversion appears to be high, but the economics of conversion in Russia still needs study. It was observed that it may be possible in the near future to successfully convert many of the reactors discussed in this chapter without significant degradation in mission. However, particularly on the Russian side, it appears that the

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87 REACTOR CONVERSION CASE STUDIES economics of conversion have yet to be studied in detail. Several Russian participants noted that such an economic analysis will be essential in the coming years. REFERENCES Austin, K.T. 2010. LEU Core Design for the Conversion of University of Wisconsin Nuclear Reactor. Paper presented at International Meeting on Reduced Enrichment for Research and Test Reactors, October 10-14, 2010. Available at www.rertr.anl.gov/RERTR32/pdf/ S2-P3_Austin.pdf. Cook, D. 2011. Challenges Associated with Conversion of the High Flux Isotope Reactor. Presentation to the Research Reactor Conversion Symposium. June 9. Chernyshov, V.M., Ryazantsev, E.P., Egorenkov, P.M., Nassonov, V.A., Yuldashev, B.S., Karabaev, Kh.Kh., Dosimbaev, A.A., Aden, V.G., Kartashev, E.F., Lukichev, V.A., Aleksandrov, A.B., and Yenin, A.A. 2002. Results of IRT-4M Type FA’s Testing in the WWR-CM Reactor (Tashkent). Paper presented at International Meeting on Re- duced Enrichment for Research and Test Reactors, 2002. Available at www.rertr.anl.gov/ web2002/2003web/fullpapers-pdf/chernyshov.pdf. Kryuchkov, E.F. 2011. Problems (peculiarities) of IRT MEPhI Research Reactor Conversion. Presentation to the Research Reactor Conversion Symposium. June 9. Nasonov, V. 2011. Conversion of the IR-8 Reactor. Presentation to the Research Reactor Conversion Symposium. June 9. Newton, T. 2011. Challenges with the Conversion of the MITR. Presentation to the Research Reactor Conversion Symposium. June 9. Pavshuk, B.A. 2011. Classification of Reactors by Type of Task. Presentation to the Research Reactor Conversion Symposium. June 9. Starkov, V.A. 2011. The Status of Testing LEU U-Mo Full-Size IRT Type Fuel Elements and Mini-Elements in the MIR Reactor. Presentation to the Research Reactor Conversion Symposium. June 9. Tzibulnikov, Yu.A. 2011. Business Plans and Key Projects at TPU. Presentation to the Research Reactor Conversion Symposium. June 8. USNRC (U.S. Nuclear Regulatory Commission). 2011. Nuclear Glossary. Available at: www. nrc.gov/reading-rm/basic-ref/glossary/. Vitiello, B. 2008. Thermal Hydraulic and Safety Analysis of the University of Wisconsin Nuclear Reactor. Masters Thesis, University of Wisconsin-Madison. Wilson, P. 2011. LEU Conversion of the University of Wisconsin Nuclear Reactor. Presentation to the Research Reactor Conversion Symposium. June 9.

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