Following U. S. President Dwight Eisenhower’s 1953 Atoms for Peace speech to the United Nations (Eisenhower, 1953), the U. S. and Russia exported research reactors to about 40 countries. At present, the IAEA lists 254 operational research reactors in 55 countries (Adelfang, 2011; see Figure 1-1). According to the IAEA, 75 civilian research reactors (excluding defense and icebreaker reactors) are currently operating using HEU fuel (see Figure 1-2). Nearly all HEU-fueled research reactors are supplied with HEU of U. S. or Russian origin, with the exception of a very few that are supplied with Chinese-origin HEU. About 25 percent of all research reactors are located in developing countries, including Bangladesh, Algeria, Colombia, Ghana, Jamaica, Libya, Thailand, and Vietnam.[12]

Civilian research reactors are used for a wide variety of missions, for example, to perform research in a broad range of scientific and engineer-

Planned 2

Under Construction
3

Total =

672 Research Reactors

ing disciplines, including research related to nuclear engineering, nuclear physics and chemistry, materials science, and biology. In addition, research reactors have become indispensible for the production of medical isotopes for diagnostic and therapeutic procedures and are also used for industrial purposes such as silicon doping.

Research reactors’ key missions require them to be designed differ­ently from commercial power reactors. Most notably, research reactors are typically designed to produce higher thermal neutron fluxes at much lower thermal outputs than power reactors. Most research reactors are also physi­cally much smaller than power reactors (typically having core volumes of less than a cubic meter versus tens of cubic meters) and require far less fuel (typically a few kilograms versus thousands of kilograms).

Research reactors have a broad range of designs in terms of power levels, moderators,11 fuel types, and cooling systems, among other design features. In many cases, these reactors are one-of-a-kind or few-of-a-kind, complicating efforts to convert them to LEU fuel. For illustrative purposes, one common broad category of research reactor—the pool — or tank-type water-moderated reactor—is described in the following paragraphs. A broad range of other designs exist, including fast research reactors, which require no moderator and use plutonium as fuel, and “homogeneous reac­tors,” in which the reactor core is a solution of dissolved uranium salts contained in a tank.

Pool-type or tank-type research reactors (see Figure 1-3) comprise a cluster of fuel assemblies and control rods[13] [14] in a pool or tank of water, which serves as both a moderator and a coolant.[15] The core is often sur­rounded by graphite, beryllium, or heavy water (the “reflector”) that is used to slow down (moderate) neutrons and reflect them into the core to maximize the neutron flux. The core and reflector typically contain empty channels for irradiation of targets and test materials, and some reactors are designed with apertures in their pool or tank walls through which neutron beams can be accessed. Figures showing the core configurations for a num­ber of different research reactors can be found in Chapters 2 and 3.

Fuel assemblies (also referred to as “fuel elements”) contain the ura­nium fuel that powers the reactor. A fuel assembly is comprised of indi­vidual fuel plates, tubes, or rods, the latter of which is also referred to as

“pins.” Each fuel plate or tube consists of the uranium fuel itself (the “fuel meat”) sealed in a “cladding” most typically constructed of aluminum. The number of fuel plates or tubes in an individual fuel assembly can vary widely. For example, a Russian MIR. M1 fuel assembly contains four tubes, whereas the outer fuel assembly of the U. S. High Flux Isotope Reactor contains 369 plates. An illustration of a Russian IRT-4M fuel assembly is shown in Figure 1-4.

Plate-type and TRIGA pin-type fuel is most commonly used in pool- and tank-type research reactors of U. S. origin, whereas tubular or pin-type fuel is used in Russian-origin reactors. Different fuel production methods— rolling in the United States and extrusion in Russia—are used as well.

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), in­cluding (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 fraction[54]; (2) prompt neutron lifetime[55]; (3) control element worth[56]; and (4) prompt temperature coefficient.[57]

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 two­dimensional generalized-geometry lattice physics transport code.[58]

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 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 configura­tion (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.[59]

The RPI research reactor is licensed to operate at 1 MW power and has a peak flux of about 3.1 x 1013 n/cm2-s. The core was converted from a 93 percent enriched UAlx-aluminum dispersion fuel to 19.75 percent enriched uranium silicide (U3Si2)-aluminum dispersion fuel in 2007. The LEU fuel contains slightly more uranium-235 than the HEU fuel it replaced to ac­count for the increased neutron absorption by uranium-238.

The conversion goal for this reactor was to allow for 10 years of op­eration at acceptable neutron flux density levels using the same number or fewer fuel assemblies. A silicide fuel with the same fuel meat thickness as the original HEU fuel met this goal when the core contained 17 fuel assemblies. However, by increasing the thickness of the fuel meat by 0.1 millimeters, the conversion goal could be met using only 13 fuel assemblies, a savings of 4 assemblies. Additionally, by changing the locations of some of the beryllium reflector blocks, designers were able to increase neutron flux densities in key locations in the reactor core to better suit experimental needs.

There are remaining fuel development and fabrication challenges asso­ciated with producing the UMo monolithic LEU fuel that will be required to convert HFIR:

• Fuel development is still under way, and the results will affect the final LEU fuel design. A fuel irradiation test series is currently ongoing at Idaho National Laboratory (INL); ORNL expects that this testing (which includes fuel failure testing) will guide safety and other calculations. The analysis of the fission product release from these tests will be most helpful.

• Fuel fabrication methods still need to be developed for producing variable radial and axial fuel thicknesses. In addition, the safety analyses rely on precise manufacturing tolerances, so for HFIR more than other reactors, fuel fabrication will need to be very precise.

• Criticality testing of the fuel will need to be completed. A facility will need to be identified for this purpose.

• Reactor startup testing will need to be conducted both at low power (to validate the analyses) and at full power (to demonstrate fuel performance and the preservation of key mission capabilities).

• Finally, ORNL will need to plan and assess the impacts of initial commissioning as well as the transition from HEU operation to full LEU operation.

The practical use of atomic energy for civilian and military purposes in the Soviet Union began with the launching of research reactor F-1 in December 1946. The reactor is graphite moderated and is fueled with 50 tonnes of natural uranium. Its operational range extends from 25 kW to 4 MW. This reactor is still operating today and is used as a reference source for neutron fluxes.

There have been a total of 80 research reactors constructed by the Soviet Union, including the following 15 reactors that were constructed in foreign countries:

• VVR-S (2-10 MW power): Constructed in East Germany, Czecho­slovakia, Romania, Poland, Hungary, and Egypt between 1957 and 1961.

• IRT-2000 (2-10 MW): Constructed in China, Bulgaria, North Ko­rea, and Iraq between 1961 and 1967.

• TBP-C (10 MW): Constructed in China in 1959.

• RA (10 MW): Constructed in Yugoslavia in 1959.

• IRT-10000 (10 MW): Constructed in Libya in 1981.

• MARIA (30 MW): Constructed in Poland in 1974.

• IVV-9 (0.5 MW): Constructed in Vietnam in 1983.

Eleven research reactors besides F-1 have been constructed at the Kurchatov Institute:

• RFT: Channel graphite reactor; initial power 10 MW, later up­graded to 20 MW; began operations in 1957 and was partially demolished in 1962.

• VVR-2: Pool-type reactor; initial power 0.3 MW, later upgraded to 3 MW; began operations in 1954 and was dismantled in 1983.

• IRT: Pool-type reactor; initial power 2 MW, later upgraded to 5 MW; began operation in 1957 and was dismantled in 1979.

• MR: Channel-type reactor immersed in a pool; initial power of 20 MW, later upgraded to 50 MW; began operation in 1963 and was shut down in 1993.

• Chamomile: High-temperature neutron thermoionic converter; 0.1 MW; began operation in 1964 and was shut down in 1996.

• Hydra: Homogeneous pulse reactor; 0.01 MW (30 mega Joules per pulse); began operations in 1972 and is currently operational.

• Yenisei: High-temperature neutron thermoionic converter; 0.1 MW; began operation in 1973 and was dismantled in 1986.

• IR-8: Pool-type reactor; 8 MW; began operation in 1981 and is currently operational (Figure 2-9).

• Argus: Homogeneous reactor; 0.02 MW; began operations in 1981 and is currently operational.

• Gamma: Cabinet water-cooled reactor; 0.125 MW; began opera­tion in 1982 and is currently operational.

• OR (referred to as OP-M in Table 1-2 in Chapter 1): Pool-type reactor; 0.3 MW; began operation in 1989 and is currently operational.

These reactors created an experimental base for nuclear and materials re­search at the Kurchatov Institute.

The remainder of this presentation focused on the characteristics of the MR and IR-8 reactors at the Kurchatov Institute and activities at a branch institute in Sosnony Bory (Leningrad region).

MR was equipped with 10 experimental loops, each of which func­tioned as a small prototype power reactor. Several coolants were used in these loops, including pressurized water, steam-water mixtures, helium, carbon dioxide, and liquid lead bismuth. The neutron flux density in the reflector was 5 x 1014 n/cm2-s. This reactor was used to work out the struc­ture of active zones of nuclear reactors and test 400 fuel assemblies and more than 8,000 fuel rods for VVER, RBMK, ACT, high-temperature, and naval reactors.

IR-8 has a compact core with an effective reflector that provides for large thermal neutron densities of 2.3 x 1014 n/cm2-s. The core contains 12

experimental channels in a horizontal orientation. This reactor is used to carry out fundamental research in nuclear physics, solid state physics and superconductivity, and other experiments.

The Scientific Research Technological Institute (NITI), a branch of the Kurchatov Institute, was created in Sosnovy Bor in 1964. It has a full-scale prototype submarine reactor.

Near the close of the symposium, participants were asked to summarize important ideas that had been mentioned over the preceding three days and to identify potential future opportunities for both the United States and Russia on the conversion of research reactors from HEU to LEU fuel. During this discussion, many key points were brought up by individuals in attendance at the symposium. These points include 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. Over the past few decades, the trend in research reactor development—as well as in civilian applica­tions as a whole—has been to reduce the use of HEU.

• Research reactors currently serve important purposes for research and industry, and they will to continue to serve important purposes into the future. In some cases, accelerators or other sources of neutrons could be used to replace research reactors for medical isotope production and other applications. However, for scientific research, some types of irradiation phenomena, and advanced fuel cycle work, research reactors will continue to be invaluable into the foreseeable future. Several workshop participants stated that these reactors must continue to operate safely and cost effec­tively and fulfill their missions in ways that meet the needs of customers.

• Collaboration between the United States and Russia on conver­sion of research reactors will continue to be essential and fruitful. Daniel Wachs observed that past collaborative U. S.-Russian work on fuel develop­ment has provided opportunities to advance conversion of both countries’ reactors; he stated that the cross-fertilization of ideas, lessons learned, and technological advances has been helpful and should continue to be encour­aged. In addition to technical collaboration, one participant observed that there is significant potential for collaboration on the regulatory aspects of conversion as well. Alexander Adams and V. Bezzubtev noted that Russia will face many challenges in regulating its to-be-converted reactors; the United States has previously faced many similar challenges and may have helpful advice for Russia on this issue.

• The United States and other nations have been able to convert re­search reactors to LEU fuel while maintaining performance required for key missions, e. g., research as well as medical and industrial applications. H.-J. Roegler observed that prior to conversion of many research reactors in Eu­rope there were a number of concerns about maintaining needed functional­ity after conversion. However, in the end, the performance of many research reactors was improved as a result of the conversion process through design changes and better understanding of reactor behavior. P. Adelfang added that an analogy might be made to molybdenum-99 production. In 2001, Argentina’s Comision Nacional de Energia Atomica (CNEA) made the deci­sion to convert its domestic production from HEU targets to LEU targets. At that time, it was considered to be infeasible to produce molybdenum-99 in significant quantities using LEU; however, CNEA showed that it could be done. After nine years it has become abundantly clear that high-quality molybdenum-99 production is possible with LEU targets.

• The economic and performance challenges associated with conver­sion are likely to be surmountable, particularly with government assistance and the involvement of reactor operators and customers. Research reactor conversions have been successfully completed in many countries, but many of these efforts would have been unlikely to occur without U. S. government support. B. Myasoedov and Jeffrey Chamberlin agreed that government in­volvement is critical to future conversion successes in Russia and the United States. Jordi Roglans noted that governments’ decisions regarding future HEU use would likely be influenced by the potential for economic and other upheavals if a terrorist event involving HEU occurred related to research reactors or otherwise.

• Some facilities may not be easily convertible to LEU fuel, includ­ing fast 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 the host nation’s government. 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 the few reactors needed for these missions and to ensure that those facilities have strong se­curity and safeguards protection. A. Zrodnikov observed that international centers would complement conversion, because a large international facility would allow research to be done that would be more challenging than at a smaller facility. In addition, he observed that at such facilities it would be easier to manage high-quality MPC&A as well as physical protection because of the international attention that such facilities would receive, especially if such facilities were placed in nations with well-developed nu­clear infrastructures. The suggestion regarding major international centers received support from several Russian participants.

The United States and the Russian Federation have had active efforts to convert research reactors from HEU fuel to LEU fuel for more than 30 years. The history of these conversion efforts is outlined in the following section, followed by a brief discussion of the current state of research reac­tor conversion efforts in both countries.

History of Research Reactor Conversion Efforts

The first U. S.- and Soviet-supplied research reactors, which were con­structed beginning in the 1950s, were designed to operate on LEU fuel. During the 1960s and 1970s, power upgrades[16] in U. S.-supplied reactors required increased uranium-235 element loadings to reduce fuel consump­tion and contain fuel fabrication costs. HEU fuel enriched to 93 percent uranium-235 became standard in these reactors. During the same time period, power upgrades in Soviet-supplied research reactors also required increased uranium-235 element loadings; HEU fuel enriched to 80 to 90 percent uranium-235 became standard in these reactors (Arkhangelsky, 2011).

However, in the 1970s, concerns in both the United States and Soviet Union about potential links between the civilian trade in HEU and nuclear proliferation began to increase following a nuclear weapons test in India, unsafeguarded nuclear activities in other countries, and growing terror­ist activities around the world. In 1978, the U. S. Department of Energy (DOE) established the Reduced Enrichment for Research and Test Reac­tors (RERTR) program to develop technologies to minimize and eventually

eliminate the civilian use of highly enriched uranium.[17] At present, all of DOE’s HEU elimination efforts for civilian research and test reactors[18] are currently being carried out under the Global Threat Reduction Initiative (GTRI), into which RERTR was absorbed in May 2004.[19]

Also around 1978, the U. S.S. R. Ministry of Atomic Energy initiated a similar program, the Russian Program of Reducing of Enrichment in Research Reactors (RPRERR), to reduce the enrichment of fuel for re­search reactors in its client states from 80-90 percent enriched uranium to 36 percent enriched uranium. At this time, the U. S.S. R. began work on high-density LEU research reactor fuels for use in foreign research reactors operating with Soviet fuel (Arkhangelsky, 2011). However, there was no contact or collaboration between these U. S. and Soviet conversion pro­grams until 1993.

The first formal contact to discuss collaboration on research reactor conversions took place in Moscow in March 1993. At that meeting it was decided to initiate a contract between Argonne National Laboratory (ANL) and the Dollezhal Scientific Research and Design Institute of Energy Tech­nologies (NIKIET) on conversion studies and fuel development. Following these interactions, the Russian program began to develop fuel with a less than 20 percent enrichment based on uranium dioxide fuel for the conver­sion of foreign research reactors.[20]

Significant progress has been made to convert HEU-fueled research and test reactors around the world. As of June 2011, a total of 74 research reactors have been converted from HEU fuel to LEU fuel or shut down since 1978. Of these, 35 have been converted or shut down since 2004, including seven U. S. domestic conversions; 18 foreign conversions; and 10 domestic and foreign shutdowns prior to conversion (Chamberlin, 2010; Roglans, 2011b).

At present, the United States and Russia are cooperating on the conver­sion of U. S.- and Russian-designed reactors in other countries. The February 2005 Joint Statement by President George W. Bush and President Vladimir V. Putin on nuclear security cooperation affirmed this cooperation:

The United States and Russia will continue to work jointly to develop low-enriched uranium fuel for use in any U. S.- and Russian-design research reactors in third countries now using high-enriched uranium fuel, and to return fresh and spent high-enriched uranium from U. S.- and Russian — design research reactors in third countries. (Bush-Putin, 2005)

This cooperation was reaffirmed and expanded by U. S. President Barack Obama and Russian President Dmitry Medvedev in a July 2009 joint state­ment (Obama-Medvedev, 2009). To implement the Obama-Medvedev Joint Statement, Rosatom Director General Sergey Kiriyenko and DOE Deputy Secretary Daniel Poneman signed an agreement during their December 6-7, 2010, meeting to begin studies to determine the technical feasibility and economic impact of converting six HEU-fueled research reactors in Russia (Arkhangelsky, 2011; D’Agostino, 2011).

The thermal/hydraulic analysis focused on the behavior of the high- power channel at steady state, low-power pulse, and high-power pulse.[60] 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).[61]

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

Figure 1, tori’ Мэр: HEU on Left, LEU on Right

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).

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 analy­ses needed to be carried out. Second, a discrepancy was found between the two critical heat flux correlations used to analyze the natural circulation mode.[62] [63] Third, there was some uncertainty in the natural convection heat transfer models. Finally, it was challenging to determine appropriate air­cooled temperature safety limits for the new LEU 30/20 fuel type.

The overall outcome of the thermal/hydraulic analysis was encourag­ing. 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 tempera­tures, and the loss-of-coolant fuel temperatures were less than 700 oC.

MURR is a high-performance research reactor with a very compact core (core volume of only 33 liters with 4.3 liters of fuel meat) with a peak thermal flux of about 6.0 x 1014 n/cm2-s (Figure 2-5). The reactor is

refueled weekly to maintain a greater than 90 percent capacity factor for efficient production of medical isotopes.

Conversion studies for this reactor showed that if the fuel geometry was unchanged, conversion using UMo monolithic LEU fuel would result in a harder neutron spectrum and, thus, increased power in some regions of the reactor. A means to control this higher power density needed to be identified for conversion to become possible.

The reactor fuel plates are curved, and there is no flexibility to rear­range them to reduce power peaking. However, it was determined that by using four distinct thicknesses of fuel meat in the assemblies (ranging from 0.23-0.43 millimeters), peaking factors could be reduced to acceptable levels.

The identified changes in power, nuclear characteristics, and fuel weight will affect the HFIR facility infrastructure. At present, it appears that ORNL will need to: (1) increase the capacity of the cooling tower and the cold source helium refrigerator; (2) modify the reactor instrumentation and control systems to manage the increased heat load; (3) modify the fuel handling tools; and (4) reevaluate the core structural and seismic analyses. In addition, the spent fuel systems will need to be reevaluated, including the design analyses for the pool storage and fuel shipping containers.