Category Archives: Progress, Challenges, and Opportunities for. Converting U. S. and Russian Research Reactors
Two presentations discussing performance and missions of reactors after conversion were given by Panel 2.1 speakers: Jordi Roglans (Argonne National Laboratory) provided a U. S. viewpoint on maintaining performance and missions (Roglans, 2011), and A. L. Petelin (Research Institute of Atomic Reactors [RIAR]) provided a description of several Russian research reactors at RIAR and their missions (Svyatkin et al., 2011).
The Global Threat Reduction Initiative (GTRI) strives to achieve several goals when converting research reactors: 
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As noted in John Stevens’ presentation (summarized elsewhere in this chapter), a fuel assembly is considered to be acceptable for use in a conversion project when it meets the following criteria and the reactor operator and regulator agree to accept fuel assembly for conversion:
• Qualified: The fuel assembly has been successfully irradiation — tested and is licensable.
• Commercially available: The fuel assembly is available from a commercial manufacturer.
• Suitable: The fuel assembly satisfies the criteria for LEU conversion of a specific reactor; safety criteria are satisfied; fuel service lifetime is comparable to current HEU fuel; and the performance of experiments is not significantly lower than for HEU fuel.
When converting from an HEU to LEU fuel, one should strive to make as few changes as possible in the fuel assembly and core geometries. Conversion should also be carried out in a way that has the least possible effect on scientific operations in the facility.
The annual operating costs of a reactor will be affected by the costs of the LEU fuel assemblies compared to the HEU fuel assemblies they are replacing. The new very-high-density UMo fuels will likely cost more to fabricate because there are more manufacturing steps. However, work is under way to minimize those cost differences with the goal of maintaining or even reducing when feasible the number of LEU fuel assemblies that are consumed in a reactor each year compared to HEU fuel assemblies. The number of fuel assemblies consumed per year dominates costs when LEU and HEU fuel assemblies are of similar cost.
Analytical studies are typically needed to determine whether conversion can be accomplished without a significant impact on reactor performance and missions. However, such formal studies may not be required for HEU — fueled reactors that are of a similar type and performance to reactors that have already been converted to LEU.
The analytical studies needed to assess the potential for conversion include:
• Feasibility studies that identify suitable LEU fuel assemblies (either existing qualified fuels or new fuels under development), compare reactor performance with HEU and LEU fuels, and calculate key safety parameters.
• Operational and safety analyses to demonstrate that the transition from HEU to LEU fuel can be done safely and without interrupting normal reactor operations, and also that the converted reactor satisfies all safety requirements.
One also needs to formulate safety requirements and resolve any issues raised by regulators regarding the reactor’s safety documentation. Additionally, economic impact studies may be required to determine the overall impact and acceptability of conversion.
A feasibility study entails many activities. Initially, fuel requirements and experimental performance indicators must be defined. With respect to the latter, it is important to determine what the most important experimental positions are in the reactor and what performance characteristics (e. g., flux densities and neutron energy distributions) are required in those positions. Iterative modeling studies are used to determine these characteristics as well as other operating criteria such as shutdown margins. Fuel assembly and reactor core designs are adjusted, and the models are rerun until acceptable performance and other important reactor characteristics are achieved. The final LEU fuel assembly design can be selected once these studies are completed.
Some high-performance reactors may require fuel-design optimization and possibly facility-specific mitigation measures to address any performance penalties arising from conversion. For U. S. high-performance reactors, the anticipated unmitigated decreases in performance resulting from conversion do not preclude any current applications but could affect application throughputs. The high demand for these reactors is already limiting scientific output and isotope production. Consequently, several mitigation strategies are being pursued to avoid throughput penalties.
For the U. S. high-performance reactors, the following mitigation strategies are being pursued:
• HFIR: The anticipated performance penalty of 10-15 percent will be mitigated by increasing reactor power from 85 MW to 100 MW. This could result in small gains in performance.
• MITR: The anticipated performance penalty of 5-10 percent will be mitigated by increasing reactor power from 6 MW to 7 MW.
• MURR: The anticipated performance penalty of 15 percent will be mitigated by changing LEU plate thickness (see the presentation by John Stevens elsewhere in this chapter) and by increasing reactor power from 10 MW to 12 MW.
• NBSR: The anticipated performance penalty of 10 percent will be mitigated by upgrading the cold neutron source.
Power increases in HFIR, MITR, and MURR are possible because their existing cooling systems are adequate to handle the increased heat loads. As a result of these mitigation strategies, no current applications are expected to be precluded by conversion. In ATR, preliminary studies indicate that there could be a 5-10 percent performance penalty after conversion. A strategy to mitigate this penalty has not yet been identified.
The key to successful conversion is collaboration. In the case of high — performance reactors or reactors with unique designs, iterative collaborations among facility operators, fuel designers, and conversion analysts are essential to optimize fuel and core design and minimize performance impacts.
The Jules Horowitz Reactor (JHR) is a 100 MW multipurpose materials testing reactor that was commissioned to replace another reactor, OSIRIS, which was built in the 1960s. JHR was initially designed to operate with a new high-density LEU fuel; however, because of difficulties in the development and qualification of this fuel, the reactor will begin operation with HEU fuel instead as described in the paragraphs to follow.
The JHR fuel elements consist of eight circular rings of curved fuel plates, each 1.37 mm thick (see Figure 4-1). The fuel elements have a 98 mm external diameter and a 600 mm active height. The nominal hydraulic gap (“coolant gap” in Figure 4-1) between the fuel plates is 1.95 mm; light water, which streams upward through the gap at a speed of 15 meters per second, is used for both cooling and moderating the core.
The core can contain 34 to 37 fuel elements and has up to 10 experimental positions (see Figure 4-2). The designed neutron fluxes are 5.5 x 1014 fast neutrons per square centimeter per second (n/cm2-s) in the core and up to 4.5 x 1014 thermal n/cm2-s in the reflector.
fuel plate (1.37 mm thick.)
coolant gap (1.95 mm thick.)
Aluminium filler or Hf control rod or Irradiation device (0 37 mm)
FIGURE 4-1 Schematic illustration of a JHR fuel element. The 1.37-mm-thick fuel plates form eight concentric rings, with coolant gaps of 1.95 mm between the plates. The center of the fuel element contains aluminum filler, a hafnium control rod, or an experimental position. SOURCE: Lemoine (2011).
The reactor was designed in 2002 using a reference fuel of high-density (8 grams uranium per cubic centimeter [gU/cm3]) UMo dispersion LEU fuel. Original plans had called for this fuel—in development under the RERTR program—to be qualified in 2006. In 2004, however, problems with the fuel’s irradiation behavior indicated that it would be unlikely to be available in time for JHR’s completion. At the time of this symposium, UMo dispersion LEU fuel was still under development by the European initiative LEONIDAS, which is supported in part by the U. S. Department of Energy (DOE). Further optimization still needs to be done to qualify this fuel and demonstrate that it will be available at reasonable cost.
JHR still intends to use UMo dispersion LEU fuel when it becomes available. However, for the time being, JHR plans to use a neutronically equivalent uranium silicide (U3Si2) dispersion fuel enriched to 27 percent uranium-235. The higher enrichment of the silicide fuel is intended to balance its lower density (4.8 gU/cm3) relative to UMo dispersion LEU fuel. The neutron-equivalent U3Si2 fuel is currently under qualification. Although this fuel has been used in other reactors, qualification for JHR is needed because its operating level is much higher than the operating levels of other reactors that use this fuel.
FIGURE 4-2 Schematic illustration of the JHR core. The fuel elements are shown in purple. Ten experimental positions are shown in yellow, with seven located in the center of individual fuel elements. Three“triple” experimental positions are available in fuel element positions. The core is surrounded by a beryllium reflector with additional fixed experimental positions and eight cross water channels for mobile devices. SOURCE: Lemoine (2011).
ighly enriched uranium (HEU) is used for two major civilian purposes: as fuel for research reactors and as targets for medical isotope production. This material can be dangerous in the wrong hands. Stolen or diverted HEU can be used—in conjunction with some knowledge of physics—to build nuclear explosive devices. Thus, the continued civilian use of HEU is of concern particularly because this material may not be uniformly well-protected.
This report focuses on the civilian use of HEU for research reactor fuel. It summarizes the proceedings of a joint symposium organized by the National Research Council of the U. S. National Academies and Russian Academy of Sciences (RAS) to address progress, challenges, and opportunities for converting U. S. and Russian research reactors from HEU to low enriched uranium (LEU) fuel. This symposium—held in Moscow on June 8-10, 2011—was sponsored by the U. S. Department of Energy-National Nuclear Security Administration’s (NNSA) Office of Defense Nuclear Nonproliferation.
This report provides a summary of the symposium presentations and discussions; it does not represent a consensus of the symposium participants or the authoring committees.1 Many important points were made by individual participants during the symposium,  particularly regarding possible future actions for reducing or managing the proliferation risks posed by
HEU-fueled U. S. and Russian research reactors. These points include but are not limited to the following:
• Many symposium participants from both the United States and Russia emphasized the importance of reducing and, where possible, eliminating the use of HEU in research reactor fuel. Participants noted that conversion of research reactors to LEU fuel provides for permanent threat reduction and may reduce the requirements for (and potentially the costs of) facility security.
• Research reactors currently serve important purposes for research and industry, and they will continue to serve important purposes into the future. Prominent examples include medical isotope production and research associated with the design of next-generation nuclear plants.
• The United States and other nations have been able to convert research reactors to LEU fuel while maintaining their performance for key missions. In fact conversions of research reactors in the United States have resulted in improved understanding of their operating characteristics and, in some cases, improved performance. In the United States, all reactors that can be readily converted with existing LEU fuels have been converted. Many symposium participants observed that conversion studies of research reactors in Russia has started but conversion is lagging behind the United States.
• The economic and performance challenges associated with conversion are likely to be surmountable in many cases, particularly with government assistance and the involvement of reactor operators and customers.
The development of higher-uranium-density LEU fuels could reduce fears of loss of performance by reactor customers.
• Collaboration between the United States and Russia on conversion of research reactors has been and is likely to continue to be valuable.
Several participants noted that collaborative U. S.-Russian work on fuel development has provided opportunities to advance conversion of both countries’ reactors. Additionally, the United States has already confronted regulatory challenges associated with conversion. This experience could be useful to Russia.
• Some facilities may not be easily convertible to LEU fuel, including fast spectrum reactors, fast critical assemblies, reactors with small core volumes, and reactors with high specific power per unit volume of active core. The feasibility of conversion depends to some extent on policy choices by host nation governments. Several workshop participants suggested that one way of minimizing the use of HEU for essential or unique missions would be to create major international nuclear centers to house these reactors and to ensure that those facilities have strong security and safeguards protections.
ession 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 analysis 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
need to be carried out for each individual reactor, and the challenges encountered can be very different for different reactors.
Most Russian research and test reactors use HEU fuels consisting of UO2-aluminum dispersions fabricated as thin-walled tubular elements of various enrichments and configurations. A Russian program was started in the 1990s to further reduce the enrichment of fuel used in Russian-origin research reactors that are located outside of the Russian Federation. This work has been led by three Russian organizations (NIKIET, Bochvar All-
Russian Scientific Research Institute for Inorganic Materials [VNIINM], and Novosibirsk Chemical Concentrates Plant [NZKhK]) with the collaboration of several other organizations and customers (i. e., research reactor operators) and has resulted in the development of LEU fuels.
The initial phase of this program created UO2-Al LEU fuel assemblies for conversion of all existing Russian-origin research reactors that are located outside of the Russian Federation. The aim was to reduce the enrichment of uranium in the fuel elements without changing fuel element geometry. LEU fuel assemblies of several designs have been developed (Figure 2-3):
• VVR-M2 fuel assembly. This assembly has a tubular geometry and contains a UO2-aluminum dispersion fuel meat with a density of 2.5 gU/cm3. These fuel assemblies have undergone a full cycle of design, testing, and licensing and are currently being manufactured at the fuel production facility at NZKhK in Novosibirsk. This fuel is being supplied to Russian — origin research reactors in Hungary, Vietnam, and Romania.
• IRT-4M fuel assembly. This assembly has a square geometry and contains a UO2-aluminum dispersion fuel meat with a density of 3.0 gU/cm3. This fuel, which is fully licensed, is the highest-demand fuel for Russian-origin research reactors located outside of the Russian Federation. This fuel is being supplied to Russian-origin research reactors in the Czech Republic, Uzbekistan, and Libya.
• VVR-KN fuel assembly. This assembly has a hexagonal geometry and is being developed for use in a Russian-origin research reactor in Kazakhstan. It will replace a 36 percent enriched assembly that is now in use. Three assemblies have been manufactured and are now being irradiated in the reactor. Conversion studies and fuel qualification activities for this reactor are proceeding in close cooperation with the reactor operator, producing good results.
• MR fuel assembly. Design work is about to begin to develop a UO2-aluminum dispersion fuel for this tubular fuel assembly. The fuel meat (which currently has an enrichment of 36 percent) is expected to have an enrichment of 19.5 percent with a density no less than 3.5 g U/cm3. It is expected to take about a year to complete this design work and manufacture fuel assemblies for testing. The 19.5 percent enriched fuel will be used in the Russian-origin MARIA research reactor in Poland.
The transition to these LEU fuel assemblies has proceeded using the same fabrication technologies and equipment for producing HEU fuel. However, the use of LEU fuels can reduce reactor “performance” (i. e., reduce neutron flux densities in the core and reflector regions) by up to about 15 percent and shorten fuel replacement cycles. Consequently, the develop-
ment of higher-density LEU fuels is needed to maintain reactor performance and fuel cycle length and also to increase fuel robustness by allowing an increase in cladding thickness.
The development of higher-density fuels is being carried out in a second phase of the Russian program to reduce fuel enrichments. Work is proceeding on a UMo dispersion LEU fuel with a density of about 5 gU/cm3. Test irradiations of this material have been carried out to burnups of 40-60 percent. Design efforts are under way for two fuel assembly types: IRT-3M (which has a tubular geometry) and IRT-U (which has a pin geometry).
The third phase of the reduced enrichment program is envisaged to involve the development of completely new fuel designs for future reactors. These new designs should be safe, reliable, easy to fabricate, and economically efficient compared to current designs. UMo monolithic LEU fuels manufactured in the form of pins appear to be a promising future design concept. These could be arranged in geometries to mimic the tubular, square, and hexagonal geometries of current-generation fuel assemblies that are used in Russian-origin research reactors.
Once the UMo monolithic LEU fuel is qualified, MITR has a number of future plans to prepare for conversion that draw on the technical work described in the previous sections. First, a preliminary safety analysis report will be prepared and approved prior to conversion. Second, because the fuel fabrication requirements are unique to the MITR reactor, fuel manufacturing tolerances will need to be determined. Finally, all cores up to the last full HEU core will need to be analyzed using the newly upgraded neutronics model; this model will also be used for fuel management.
MITR staff is using the requirement to convert to LEU fuel to improve its analysis capabilities and obtain a greatly improved understanding of the reactor. These improved capabilities have resulted in an optimized LEU fuel design for the MITR reactor.
A. L. Petelin
The Russian presentation focused on current characteristics and missions of the research reactors at RIAR in Dimitrovgrad. RIAR is Russia’s largest complex for examinations of full-scale components of nuclear reactors and irradiated materials. It also has equipment and facilities for fuel cycle research and a radiochemical complex for investigation and production of transuranic elements and radioisotopes.
RIAR currently operates five research reactors. A sixth reactor is being decommissioned. The characteristics and missions of the operating reactors are described briefly in the following sections.
SM-3 is a 100 MW pressurized water flux trap-type reactor containing 32 fuel elements enriched in uranium-235 to 90 percent. The reactor has a compact square core (420 mm in plan dimension and 350 mm in height) with a central trap. Up to 41 positions are available for irradiation experiments in the central trap, core, and reflector. The maximum thermal neutron flux density in the central trap is 5 x 1015 n/cm2-s. Thermal neutron flux densities of 1.5 x 1013 to 1.5 x 1014 n/cm2-s can be obtained in the reflector.
The reactor has two low-temperature coolant water loops and a high — temperature loop that can be used for fuel testing, examination of fission — product releases from leaky fuel rods and their removal from primary cooling circuits, and the irradiation of structural and absorbing materials. The spectral characteristics and neutron-flux-density variability in this reactor also make it useful for producing a range of isotopes, including transplutonium elements and industrial isotopes such as cobalt-60.
This reactor is potentially useful for other high-dose irradiation applications, for example, testing of fuel and structural materials for high- temperature reactors, fast-boiling reactors, and supercritical reactors, as well as new designs for research reactors. In particular, new LEU fuel compositions can be examined for applications in high-flux reactors. The reactor can also be used for training.
MIR. M1 is a 100 MW loop-type reactor that uses 48-58 fuel elements enriched in uranium-235 to 90 percent (see Figure 3-8 in Chapter 3). It has seven loop facilities: Two with water coolant (PV-1, PV-2), two with water/ boiling-water coolant (PVK-1, PVK-2), two with water/boiling water and steam coolant (PVP-1, PVP-2), and one with nitrogen and helium coolant (PG). The facility also contains hot cells and cooling pools. The maximum thermal neutron flux in the loop channel is 5-7 x 1014 n/cm2-s.
A variety of experimental activities are currently performed in this reactor. These include the examination of advanced VVER-1000 fuel, testing of VVER-1000 fuel with high burnup (greater than or equal to 60 megawatt days per kilogram of uranium [MWd/KgU]), testing of new VVER cladding materials, and examination of fission-product releases from VVER-1000 fuel rods containing artificial defects. The reactor is also used to test LEU fuel and produce the industrial isotope iridium-192.
This reactor is potentially useful for other types of experimental applications, including high-temperature and high-pressure testing of reactor materials, simulation of severe reactor accidents, testing of innovative fuel and cladding materials, and expanded production of isotopes. Realizing some of these activities would require upgrades to some of the reactor loops.
If research reactors will continue to be needed in the foreseeable future it is important to understand as clearly as possible their risks. As noted previously, conversion of research reactors from HEU to LEU lowers risk. However, some reactors may not be able to be converted, so it is important to understand the risks associated with their continuing operation. This risk goes beyond the reactor itself to involve all facilities and associated infrastructures, including fuel manufacturing; transportation; fresh fuel storage; irradiated fuel storage; and reprocessing or final repository placement.
Robert Bari described two different types of risk associated with research reactor facilities and infrastructures (systems) as follows (Bari, 2011):
• Proliferation risk of an HEU-fueled research reactor’s fuel cycle is associated with the diversion or undeclared production of nuclear material or misuse of technology by a host state seeking to acquire nuclear weapons or other nuclear explosive devices.
• Terrorism risk of an HEU-fueled research reactor’s fuel cycle is associated with the theft of materials suitable for nuclear explosives or radiation dispersal devices and the sabotage of facilities and transportation by sub-national entities and/or non-host states.
The following sections describe two methodologies to structure and improve the understanding of proliferation and terrorism risk: First, assessing the relative attractiveness of various nuclear materials; and second, proliferation risk assessment methods.
his report is a summary of a joint symposium held on June 8-10, 2011, by the National Research Council (NRC) of the U. S. National Academies and the Russian Academy of Sciences (RAS) on progress, challenges, and opportunities for converting United States and Russian Federation (R. F.) research reactors1 from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel.2,3 This symposium was organized in response to a 2010 request from the U. S. Department of Energy (DOE), National Nuclear Security Administration’s (NNSA) Office of Defense Nuclear Nonproliferation.
NNSA requested that a symposium be organized and a subsequent summary document be produced to address:
• Lessons learned for overcoming conversion challenges, increasing the effectiveness of research reactor use, and enabling new reactor missions;
• Future research reactor conversion plans, challenges, and opportunities; and
• Actions that could be taken by U. S. and Russian organizations to promote conversion.
The statement of task for the project is included as Appendix C.
The preparation of the symposium agenda and the production of this summary report were carried out by a committee of U. S. experts appointed by the National Academies and a committee of Russian experts appointed by the Russian Academy of Sciences. Biographical sketches of the committee members are provided in Appendix B. These organizing committees met jointly three times over the course of the project: First, in November 2010 to plan the symposium; second, in June 2011 to hold the symposium; and third, in September 2011 to finalize the symposium report. The agenda for the symposium is provided in Appendix A, along with a list of briefings presented at the November 2010 meeting.
NNSA and the NRC agreed that the symposium would not produce consensus findings or conclusions but would instead be used to encourage discussion among U. S. and Russian participants. For this reason, this symposium summary does not contain findings, conclusions, or recommendations and does not represent a consensus of symposium participants. This report represents a summary record of the briefings and discussions that occurred during the symposium. Although the U. S. and Russian organizing committees are responsible for the content of this report, any views contained in the report are not necessarily those of these committees, the National Academies, or the Russian Academy of Sciences.
The remainder of the chapter provides background information on proliferation risks associated with civilian use of HEU; basic operating principles and terminology associated with research reactors; and potential impacts of reducing HEU use in research reactors. Much of the content of this discussion is drawn from symposium briefings (Adelfang, 2011; Arkhangelsky, 2011; D’Agostino, 2011; Dragunov, 2011; Matos, 2011; Roglans, 2011a). Additionally, some basic concepts and definitions were added for the benefit of non-expert readers.
The following three case studies of U. S. research reactor conversions are summarized in this chapter:
• Paul Wilson (University of Wisconsin) reported on the successful 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 Technology 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 straightforward 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.