RHF has a maximum power of 58 MW and a peak thermal neutron flux of about 1.5 x 1015 n/cm2-s. The core consists of a one-time-use as­sembly consisting of 280 curved plates arranged between two concentric cylindrical “sideplates.” The reactor is currently fueled with a 93 percent enriched UAlx-aluminum dispersion fuel with boron-10 burnable poisons at the tops and bottoms of the fuel plates.

This reactor is used as a neutron beam source, and a key requirement for conversion is the preservation of “brightness” (i. e., intensity) of these beams and the reactor cycle length. To meet these objectives, the fueled height of the reactor core will be increased by eliminating the burnable poi­son zones at the tops and bottoms of the fuel plates. (These poisons will be moved to another location in the reactor.) However, even with this change there will still be a 5-10 percent loss in brightness at key experimental positions. This loss of brightness can be compensated for by increasing the beam times for some experiments, but it will not affect overall throughput of experiments in the reactor.

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 to fulfill its main missions. The thermal neutron flux to the experiments cannot be degraded, and the reactor power (100 MW) and campaign dura­tion (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.[73]

UMo dispersion LEU fuel is the most likely candidate for conversion of the MIR. M1 reactor. Recent calculations have shown that to retain the re­quired 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. Dif­ferent 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 as­semblies 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 us­ing 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.

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.[74] 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.

The Russian Federation has been cooperating with the United States and the IAEA in several GTRI programs. These include the return of

Russian-origin HEU fuel to the Russian Federation from countries in East­ern Europe and Asia; reduction of fuel enrichment in research and test reactors; and enhancement of physical security for high-risk radioactive sources. Active international cooperation and collaboration are necessary for achieving the strategic objectives of GTRI.

ROSTEXNADZOR is the nuclear safety watchdog in the Russian Federation. It is responsible for regulating more than 6,000 facilities in the Russian Federation, including research and test reactors.[46] It has three primary functions: regulatory control, licensing, and supervision of atomic energy facilities.

The federal codes and standards developed by ROSTEXNADZOR are of two types: (1) general and (2) facility specific. The agency develops and promulgates federal codes and standards for atomic energy use, adminis­trative regulations, guidelines, and safety guides. The federal codes and standards provide general safety provisions for each type of atomic energy facility, for example, nuclear power plants, research reactors, icebreaker reactors, and nuclear fuel cycle facilities. These codes and standards also provide specific provisions for activities at these facilities including siting, construction, operation, and decommissioning.

There are 10 separate codes and standards for research nuclear instal­lations, which include research reactors. These include, for example:

• General Safety Assurance Provisions for Research Nuclear Instal­lations (NP-033-01)

• Requirements for the Content of Research Nuclear Facility Safety Analysis Reports (NP-049-03)

• Rules of Nuclear Safety for Research Reactors (NP-009-04)

• Requirements for a Content of Action Plan for Protection of Person­nel in Case of an Accident at a Research Nuclear Installation (NP-075-06)

Many of these codes and standards draw from IAEA documents, either in full or part, the latter being adapted to local conditions.

An effort is currently under way to enhance the regulatory framework for nuclear and radiation safety at research reactors in the Russian Fed­eration. This includes the modification of current regulatory documents and the development of new regulations. The new regulations would re­quire periodic safety reviews of research reactors, development of rules for withdrawing research reactors from state supervision, and development of procedures for modifying the design, engineering, and operating documen­tation of research reactors.

Federal Environmental, Industrial and Nuclear Supervision Service

Status of Nuclear Research Facilities

There were 74 licensed research reactors (including critical and subcriti­cal assemblies) in the Russian Federation in 2011. These are being operated by 19 organizations, including Rosatom and the Russian Academy of Sci­ences. These reactors comprise (Figure 2-10):

• 32 research reactors (24 operating, 6 decommissioned, and 2 under construction)

• 30 critical assemblies

• 12 subcritical assemblies

The average operation age of the research reactors is 24 years, but 17 reac­tors have been operating for more than 30 years.

ROSTEXNADZOR is just beginning to develop regulations for the conversion of research reactors in the Russian Federation. The regulator does not see any serious barriers or obstacles that might prevent conver­sion-related licensing activities. The USNRC’s rich experience with fuel development and conversion-related approval activities would be useful for ROSTEXNADZOR in organizing its work.

The specific issues that will need to be addressed by ROSTEXNADZOR in research reactor conversion in the Russian Federation are the following:

• R&D for design and fabrication of new LEU fuel, LEU fuel tests, and validation of LEU fuel characteristics and operating conditions.

• Safety demonstrations of fabrication, transportation, storage, and disposal of new LEU fuel.

• Analysis of flux kinetics and distribution in reactor cores with LEU fuel.

• Thermohydraulic analysis.

• Safety analysis, including certification of computer codes; justifica­tion of safe operation limits and conditions; accident initiators; and modifi­cation of Safety Analysis Reports, plans of personnel and public protection, quality assurance programs, and operational procedures.

• Modification of research nuclear installation designs.

Yuri Shiyan is the director of the Russian Academy of Sciences Office for North American Scientific Cooperation. He has worked in this capacity for more than 25 years, facilitating collaborative efforts and exchanges between international partners and Soviet/Russian scientists, engineers, and medi­cal professionals. In 2004-2005, he served as IAEA expert for the Nuclear Fuel Subcommittee, and since 1981 he has served as the coordinator of the Russian Academy of Sciences Committee on International Security and Arms Control. For the past four years, he has served as coordinator of the RAS-NAS Committees on Counterterrorism and Non-Proliferation. Further, he has assisted in several joint U. S.-Russian projects focusing on various aspects of the nuclear fuel cycle, including the storage of nuclear spent fuel. His knowledge of English and professional experience gained through assignments at several international posts have contributed to his success as an international scientific liaison.

Sarah Case (Study Director) is currently senior program officer in the Nuclear and Radiation Studies Board of the National Research Council, where she has worked since 2007. She currently manages a portfolio of consensus studies and workshops focused on technical issues related to nuclear security and non-proliferation. Previous projects have focused on nuclear security but have also addressed issues related to nuclear energy, electrical transmission and distribution, and the health effects of radiation. Dr. Case received her Ph. D. in physics from the University of Chicago and her A. B. in physics from Columbia University.

Kevin D. Crowley is senior board director of the Nuclear and Radiation Studies Board (NRSB) at the National Research Council-National Acad­emy of Sciences in Washington, D. C. He is responsible for managing the NRSB’s work on nuclear safety and security, radioactive-waste management and environmental cleanup, and radiation health effects. He is also the principal investigator for a long-standing cooperative agreement between the National Academy of Sciences and the U. S. Department of Energy to provide scientific support for the Radiation Effects Research Foundation in Hiroshima, Japan. Dr. Crowley’s professional interests and activities focus on safety, security, and technical efficacy of nuclear and radiation-based technologies. He has directed more than 20 National Research Council studies on these and other topics, including Safety and Security of Commer­cial Spent Nuclear Fuel Storage (2004, 2006); Going the Distance? The Safe Transport of Spent Nuclear Fuel and High-Level Radioactive Waste in the United States (2006); Medical Isotope Production without Highly Enriched Uranium (2009); America’s Energy Future: Technology and Transformation (2009); and Analysis of Cancer Risks in Populations near Nuclear Facili­ties (in progress). Before joining the National Academies staff in 1993, Dr. Crowley held teaching/research positions at Miami University of Ohio, the University of Oklahoma, and the U. S. Geological Survey. He holds M. A. and Ph. D. degrees, both in geology, from Princeton University.

ession 2 of the symposium (see Appendix A) focused on technical chal­lenges associated with conversion and potential solutions for overcom­ing those challenges. Three panels of Russian Federation (R. F.) and U. S. speakers were organized to address these topics:

• Panel 2.1: Technical challenges associated with conversion and potential solutions featured Russian and U. S. presentations on low enriched uranium (LEU) fuel design, core modifications, and approaches for main­taining reactor performance and missions after conversion.

• Panel 2.2: Other technical challenges associated with conversion featured presentations on ageing and obsolescence, regulatory challenges, and challenges posed by research reactors that cannot be converted.

• Panel 2.3: How challenges associated with previously converted reactors were overcome featured presentations on approaches for overcom­ing the conversion challenges identified by the other panels in this session.

These panel presentations are summarized in this chapter along with key thoughts from the participant discussions.

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 train­ing, 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

d N21-R6 (final design)

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 empty positions. SOURCE: Austin (2010).

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 x 1013 and 1 x 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 conver­sion. 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).

When converting a research reactor from HEU to LEU fuel it is impor­tant to avoid degradation of the following:

• Consumer characteristics: neutron flux level, thermal power, neu­tron spectrum, and adequacy of safety systems.

• Safety characteristics: reactivity margins, effectiveness of control rods, and peak power density.

• Performance characteristics: fuel cycle duration, number of planned reactor shutdowns, and reactor serviceability.

• Technical and economic indices: mass of uranium loading, volume of spent fuel to be reprocessed, and financial expenditure for fuel purchase and reprocessing of spent fuel.

There are two potential paths for converting a research reactor while maintaining these characteristics. One path is to design a new core that can fit into the existing reactor. The other path is to maintain the geometric configuration of the current core but change the design and arrangement of fuel and/or reflector elements.

Conversion to LEU fuel may result in decreased uranium-235 content and will result in increased uranium-238 content in the reactor core. This can change the neutronic characteristics of the core, which in turn can change its reactivity, the effectiveness of control rods, and the dynamics of fuel burnup. All of these changes can affect reactor safety. Consequently, safety analyses must be carried out to demonstrate that conversion will preserve reactor safety at required levels, including neutron-physical analy­sis, thermal-hydraulic analysis, and an analysis of transient and emergency operations.

As illustrated by the following three examples, for some Russian re­search reactors there are no developed LEU fuel elements that would en­able conversion with acceptable consumer characteristics. Moreover, some Russian research reactors are approaching the ends of their operating lives, and there is a need to consider whether to shut down these reactors or modernize them.

IRT (Moscow Engineering and Physics Institute)

The IRT is a medium-flux, 2.5 MW pool-type reactor with a square core containing 16 IRT-3M fuel elements enriched in uranium-235 to 90 percent. Initial studies have been carried out to examine the feasibility of converting this reactor to 19.75 percent enriched fuel elements of an IRT — 4M design containing a UO2-aluminum dispersion fuel meat.

These studies indicate that conversion would result in some consumer and economic penalties compared to HEU fuel: neutron flux densities in the fuel and reflector regions would decrease by 20-30 percent and 10-20 percent, respectively, and the number of fuel elements in the core would increase by 2-4 elements.[34] The economics of conversion will depend on the cost of LEU fuel elements and their reprocessing compared to the costs for HEU fuel elements.[35] However, there would be no unacceptable changes in safety characteristics, and fuel burnups would not change.

he final session (Session 4) of the symposium focused on possible futures for research reactors and how the proliferation risks as­sociated with them can be managed. Five briefings presented at the symposium (Appendix A) on these topics are summarized in this chapter:

• A. Zrodnikov (Rosatom Institute for Physics and Power Engineer­ing) discussed missions for future research reactors (Zrodnikov, 2011);

• R. P. Kuatbekov (Dollezhal Scientific Research and Design Insti­

tute of Energy Technologies [NIKIET]) and P. Lemoine (Commissariat a l’Energie Atomique) provided Russian and French perspectives on future research reactor plans and designs (Kuatbekov, 2011; Lemoine, 2011); and • A. N. Chebeskov (Rosatom Institute for Physics and Power En­gineering) and Robert Bari (Brookhaven National Laboratory) provided Russian and U. S. perspectives on proliferation risks associated with highly enriched uranium — (HEU-) fueled research reactors (Bari, 2011; Chebeskov, 2011).

Following these briefings, symposium participants engaged in a discus­sion about future opportunities for the United States and Russia related to research reactor conversion. This discussion is summarized in the last section of this chapter.

Two presentations on the challenges posed by research reactors that cannot be converted were given by Panel 2.2 speakers: Jeffrey Chamberlin (U. S. Department of Energy, National Nuclear Security Administration) provided a U. S. viewpoint (Chamberlin, 2011), and G. Pshakin (Institute for Physics and Power Engineering in Obninsk) provided a Russian view­point (Zrodnikov et al., 2011).

U. S. Viewpoint on Challenges

Jeffrey Chamberlin

GTRI is the key program within the U. S. government for implementing the U. S. policy to minimize the civilian use of HEU. GTRI’s mission is to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. Its specific goals are to: (1) convert research reactors and isotope production facilities from HEU to LEU; (2) remove and dispose of excess nuclear and radiological materials; and (3) protect high-priority nuclear and radiological materials from theft and sabotage.

GTRI’s Reactor Conversion Program is focused on converting civilian research reactors worldwide to operate on LEU fuel. Its goal is to convert or verify the shutdown of 200 civilian research reactors and HEU facilities by 2020.[47] However, GTRI does not specifically encourage the shutdown of

research reactors; such decisions are made by facility operators. A research reactor does not have to be considered to be vulnerable to be a candidate for conversion. GTRI is focused on converting civilian reactors and HEU facili­ties that use HEU fuel because it provides for permanent threat reduction.

Since the inception of GTRI in 2004, 23 HEU-fueled research reactors have been converted as part of the program, including 7 research reactors in the United States and 16 research reactors in other countries.[48] The most recent conversions were the Kyoto University Research Reactor in Japan (March 2010) and the Rez Reactor in the Czech Republic (April 2011).

As noted in Chapter 1, nearly all U. S. HEU-fueled reactors that can convert with existing LEU fuels have successfully been converted (see also Footnote 3 in this chapter). As noted in previous presentations, there are six HEU-fueled U. S. research reactors (ATR and its critical assembly, HFIR, MITR, MURR, and NBSR) that cannot be converted until a new LEU fuel is developed. Additionally, in December 2010, DOE and Rosatom signed an Implementing Agreement to perform feasibility studies for the possible conversion of six HEU-fueled research reactors in the Russian Federation.

The reduction of HEU use in civilian applications is supported at the highest levels in the U. S. and Russian governments. In a joint statement issued on July 6, 2009, Russian Federation President Dmitry Medvedev and U. S. President Barack Obama issued a joint statement expressing their strong support for HEU minimization:

We declare an intent to broaden and deepen long-term cooperation to further increase the level of security of nuclear facilities around the world, including through minimization of the use of highly enriched uranium in civilian applications and through consolidation and conversion of nuclear materials.

This call for minimization was echoed in UN Security Council Joint Reso­lution 1887, which was issued in September 2009, and in the April 2010 Nuclear Security Summit.

GTRI works in cooperation with reactor owners/operators to convert reactors to LEU fuel. This cooperation involves:

• Performance of feasibility studies to determine if reactors can be converted and still achieve their missions without major changes in reactor structures or equipment.

• Ensuring that required fuel assembly criteria for LEU conversion are satisfied; LEU fuel provides a similar service lifetime as the HEU fuel; there is no significant penalty in reactor performance; and safety criteria are satisfied.

• Development of a schedule for conversion based on operational requirements, capabilities, and regulatory processes.

• Demonstrating that conversion and subsequent reactor operations can be accomplished safely.

• Determining, to the extent possible, that overall costs associated with conversion do not significantly increase the annual operating expen­ditures for reactor owners/operators.

• Obtaining/verifying that agreements and authorities are in place to proceed with conversion.

GTRI’s starting assumption for reactor conversions is that “anything is possible.” The experience gained from previous conversions demonstrates that there are many ways to overcome technical barriers. Indeed, many of the recent successful conversions of U. S. reactors were not thought to be possible 20-30 years ago.

Although GTRI policy is to take all reasonable steps to convert facilities and reduce the use of HEU, there may be some facilities that are not feasible to convert. For example, a feasibility study for a particular reactor might indicate that conversion is not feasible because LEU fuel assembly criteria are not satisfied and a unique fuel development effort is not technically or economically feasible. This might be the case for fast reactors, fast critical assemblies, or HEU reactors with very small core volumes.

In such cases, there are four options for addressing HEU minimization at such facilities: [49]

GTRI considers each of these options to be “last resort” and does not en­dorse them as a matter of policy. These options must be considered on a case-by-case basis by the facility and the host government.

he U. S. National Academies and Russian Academy of Sciences will organize a joint symposium to discuss progress, challenges, and op­portunities for conversion of research reactors from highly enriched uranium to low enriched uranium fuel. The symposium will address the following topics:

• Recent progress on conversion of research reactors, with a focus on U. S.- and Russian-origin reactors.

• 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. • Actions that could be taken by U. S. and Russian organizations to promote conversion.

[1] This report was authored by committees of the National Academy of Sciences (NAS) and RAS. These committees are responsible for the report’s quality and accuracy.

[2] No effort was made in this report to attribute statements of fact to individual participants.

[3] In this report, the term “research reactors” is defined to include research, test, and training reactors, including critical and subcritical assemblies.

[4] By international agreement, HEU is defined as uranium enriched to a concentration of 20 percent uranium-235 or greater, whereas LEU is defined to be uranium enriched to a concentration of less than 20 percent uranium-235.

[5] This symposium focused on HEU-fueled reactors; however, some research reactors are also fueled with plutonium. The challenges of managing plutonium-fueled reactors—which will need to be accomplished through materials protection, control, and accounting measures—are mentioned in this report but were not the focus of this symposium.

[6] The terms “origin,” “supplied,” and “designed” are used interchangeably in this report to describe reactors that were developed by the United States and Russia for both domestic and third-country use.

[7] Important statements of opinion are attributed to individual workshop participants where appropriate, but no attempt has been made to attribute statements of fact.

[8] “The term ‘special nuclear material’ means plutonium, uranium enriched in the isotope

233 or in the isotope 235, and any other material that the [Nuclear Regulatory] Commission… determines to be special nuclear material.” (42 U. S.C. § 2014)

[9] Although LEU could, in principle, be enriched and converted into HEU for use in building a nuclear weapon, this process would require a significant technical infrastructure, and the mass of LEU required would be very large. The international community could track an effort to enrich LEU more effectively than one involving the theft of HEU.

[10] Assuming 50 kilograms of HEU per explosive device. This may be a conservative as­sumption. The IAEA defines the siqnificant quantitity of HEU to be 25 kilograms. Significant quantity is defined as “the approximate amount of nuclear material for which the possibility of manufacturing a nuclear explosive device cannot be excluded” (IAEA, 2001).

[11] Much research reactor used fuel is not considered to be “self-protecting” (formally defined as producing a dose rate greater than 100 rad per hour at 1 meter in air) because of its low radioactivity. However, irradiated fuel from virtually all of the high-performance reactors mentioned in this report would be considered to be self-protecting, as would irradiated fuel from commercial power reactors, for a period following removal from the reactor.

[12] www-naweb. iaea. org/napc/physics/ACTIVITIES/Research_Reactors_Worldwide. htm.

[13] A moderator is a material used to slow down neutrons (i. e., reduce their kinetic energies), which increases the probability of fission when the neutrons are captured by uranium nuclei.

Light materials such as water and graphite are commonly used as moderators.

[14] Control rods contain materials (e. g., boron) that absorb neutrons; they are used to control fission rates in the reactor fuel and hence the power levels in the reactor.

[15] Tank-type research reactors are similar to pool-type reactors in overall design, but they typically operate at higher power densities, requiring higher coolant flows and pressures, making it necessary to separate the coolant from the remainder of the pool contents.

[16] Power upgrades of U. S.- and Soviet-supplied research reactors were undertaken to in­crease neutron fluxes in experimental positions.

[17] More information about this program can be found at www. rertr. anl. gov.

[18] Research, test, and training reactors that have military or national security missions are outside the scope of DOE’s conversion program.

[19] DOE and GTRI assist reactor operators to perform feasibility studies and safety analyses required for regulatory approval to convert and procure LEU replacement fuels. GTRI also funds work to develop and qualify higher-density uranium-molybdenum (UMo) LEU fuel to convert high-performance research reactors (see Chapter 2).

[20] In 1996 the Bochvar All-Russian Research Institute of Inorganic Materials (VNIINM) became the lead Russian institute under the contract with ANL.

[21] These reactors are regulated by the U. S. Nuclear Regulatory Commission or the U. S. Department of Energy.

[22] The NTR General Electric Reactor in California and the Idaho National Laboratory’s TREAT reactor (Roglans, 2011b).

[23]A critical assembly contains sufficient fissionable and moderator material to sustain a fission chain reaction at a low (close to zero) level. It is designed so that fissionable and moderator materials can be easily rearranged in various geometries to mock up different reactor designs.

[24] Not including naval or other defense-related reactors.

[25] Chapter 1 (this chapter) provides the context for this study and introductory material from the symposium;

• Chapter 2 addresses challenges associated with conversion as well as potential solutions;

[26] As noted in Chapter 1, there are two additional HEU-fueled research reactors in the United States (NTR General Electric and TREAT; see Footnote 20 in Chapter 1) that appear to be convertible using current-type LEU fuels. The Department of Energy (DOE) is com­pleting studies to confirm the feasibility of converting these reactors using current-type LEU fuels. Additional research will be required to more fully develop the capability to fabricate these LEU fuels.

[27] The Reduced Enrichment for Research and Test Reactors (RERTR) program (see Chapter 1) also participated in the qualification of a fourth LEU fuel system: a uranium-zirconium hydride with an erbium burnable poison (UZrHx-Er) fuel system that is used for the conversion of TRIGA (Test, Research, Isotope production—General Atomics) reactors. General Atomics began developing a higher-density fuel (up to 3.7 gU/cm3) before the RERTR program was started in 1978. The RERTR program performed irradiation tests on 20/20 (i. e., 20 weight percent uranium, 20 percent enriched), 30/20, and 45/20 fuels. The 30/20 fuel was used to convert the Oregon State TRIGA Mark II reactor, discussed later in this chapter, and the University of Wisconsin Nuclear Reactor, discussed in Chapter 3, as well as a number of other TRIGA reactors in the United States and abroad.

[28] These high-performance reactors have high-power-density (i. e., high-flux-density) cores. Fuels having higher uranium densities than are available with existing LEU fuels are required to convert these reactors.

[29] That is, alloys consisting of 9 parts uranium to 1 part molybdenum by weight.

[30] Fission gas bubbles are formed in the fuel phase as a result of the production of gaseous fission products.

[31] At present, no LEU replacement fuel has been identified for the FRM II reactor.

[32] As the name suggests, a partial fuel assembly contains only portions of a full fuel assembly. For example, a partial assembly might contain fewer fuel plates than a full assembly.

[33] Extrusion processes are used to manufacture research reactor fuel in Russia, whereas rolling processes are used to produce research reactor fuels in the United States and Europe. Both processes produce suitable fuels, but fuel produced by extrusion generally has a lower density than fuel produced by rolling.

[34] This conclusion was reached by studying the use of existing IRT-4M LEU fuel. A feasibility study with the IRT-3M UMo fuel of higher density, currently under development, is under way at MEPhI (see Chapter 3) and may reach different conclusions when completed.

[35] In Russia, reprocessing costs are based on fuel mass.

[36] Develop or identify an LEU fuel assembly that is acceptable for conversion.

• Ensure that the ability of the reactor to perform its scientific mis­sion is not significantly diminished.

• Ensure that conversion can be achieved without requiring major changes in reactor structures or equipment.

• Demonstrate that the LEU fuel meets all safety requirements and that conversion and subsequent operations can be accomplished safely.

• Ensure that annual operating costs do not increase significantly as the result of conversion.

• Develop a conversion schedule that is based on operational require­ments, capabilities, and regulatory processes.

[37] Service lifetimes of LEU fuel assemblies can be increased if the uranium-235 loadings are higher than comparable loadings in the HEU fuel.

[38] This reactor is also discussed in another symposium presentation that is summarized in

Chapter 3.

[40] And under a Contract Service Agreement with the IAEA.

[41] The 77 reactors represented in the template responses range from less than 5 years to more than 50 years old. The average age was 37.8 years (Figure 2-7).

• The most frequently reported ageing problems were obsolescence and technology changes (92 out of 367 reports); corrosion (73 out of 367); and changes to regulatory requirements and standards (49 out of 367).

[42] Some of the Russian reactors described in this presentation are also discussed in another presentation that is summarized in Chapter 3.

[43] The USNRC does not regulate the High Flux Isotope Reactor at the Oak Ridge National Laboratory or the Advanced Test Reactor and its critical assembly at the Idaho National Laboratory. These reactors are the responsibility of DOE.

[44] The Cintichem Reactor was located in Tuxedo, New York. It was shut down in 1990.

[45] Depending on its design, HEU fuel is shipped to either the Savannah River Site in South Carolina (for aluminum-based fuels) or the Idaho National Laboratory (for other fuel designs), where it is stored.

[46] This number includes radiation sources at hospitals.

[47] This deadline slipped to 2022 while this report was being completed because of Fiscal Year 2011 federal budget reductions.

[48] In Chapter 1, it was noted that 35 conversions or shutdowns of HEU-fueled reactors have occurred since 2004. This larger number includes 10 reactors that were shut down and 2 reactors that were converted to LEU under domestic programs rather than GTRI.

[49] Option 1: Assess the possibility of changing the facility mission such that it can be accomplished with LEU fuel. However, GTRI does not advocate a change of reactor mission for the sole purpose of converting.

• Option 2: Reduce HEU enrichments. This may be technically fea­sible in some cases where LEU conversion is not. Note, however, that re­duced enrichments above 20 percent are not considered HEU minimization under international norms or GTRI policy.

• Option 3: Shut down the facility or consolidate it with similar facilities if it is underutilized.

• Option 4: If no other options exist for the facility other than to operate with HEU, remove all excess HEU and enhance physical protection measures to achieve threat reduction.

[50] As noted in Chapter 1, a critical assembly contains sufficient fissionable and moderator material to sustain a fission chain reaction at a low (close to zero) level. It is designed so that fissionable and moderator materials can be easily rearranged in various geometries to mock up different reactor designs.

[51] As noted in Chapter 1, a material is considered to be “self-protecting” if it produces a dose rate greater than 100 rad per hour at 1 meter in air. These high levels create substantial radiological barriers to illicit use.

[52] Research reactors will continue to be an essential tool for many applications. B. Myasoedov commented that he expected the role of re­search reactors to grow in the future to support the development of more complex reactor designs for nuclear power plants, including those based on fast reactor designs; for radiopharmaceutical production; and for analytical methods (such a neutron activation analysis) to support safety monitoring and control. He suggested that Russia and the United States should agree to work together and with third-party countries to design a standardized research reactor that could be produced on an industrial basis. This would eliminate the need to design individual, customized cores and fuel elements.

• Past experience suggests that successful conversion solutions can be found for most reactors. Jim Snelgrove commented that in view of the suc­cess that has occurred in converting reactors in the United States and some other countries, a key take-away message from this symposium should be that it is possible to find conversion solutions if one works hard enough to uncover them. Yu. S. Cherepnin added that some of the presentations in this session documented how enrichment levels could be reduced without degrading reactor performance. These examples should be publicized. H.-J. Roegler commented that conversion can result in improvements to reactors.

[53] “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.

[54] 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 de­layed fission neutrons per fission to the mean total number of neutrons per fission (prompt plus delayed).

[55] The prompt neutron lifetime is the average time between the emission of neutrons and either their absorption in or their escape from the system.

[56] The control element worth is the negative reactivity change caused by inserting a control element into the reactor. UWNR has five separate control elements.

[57] The prompt temperature coefficient is the change in reactivity per degree change in fuel temperature.

[58] This confirmatory analysis was a two-dimensional deterministic analysis coupled with a one-dimensional diffusion approximation.

[59] 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.

[60] This analysis assumed no cross-flow, i. e., no exchange of coolant with adjacent fuel or reflector assemblies.

[61] 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).

[62] 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).

[63] The fuel temperature-limiting safety setting is the temperature below which the fuel is required to be maintained to prevent fuel element failure.

[64] A “near maximum hypothetical accident” maintains an intact pool and ventilation.

[65] This is the relationship between the pressure drop and the conditions in the flow channel such as mass flow rate.

[66] 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).

[67] 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.

[70] At specified setpoints, the reactor will shut down by opening the circuit breakers that supply electrical power to control rods.

[71] COMSOL is a multiphysics engineering software package.

[72] Descriptions of some of these reactors were provided in the presentations that are summarized Chapter 2.

[73] 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.

[74] 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.

[75] Tomsk Polytechnic Institute has recently requested that this reactor be licensed to operate at 11 MW.

[76] Aluminum is now being used rather than beryllium in a number of locations because of swelling of some of the beryllium blocks.

[77] Nuclear physics of the interaction of radiation with matter.

[78] Radiation damage of metallic and nonmetallic reactor materials.

[79] For example, one facility might require only very few radiation protection measures to isolate nuclear materials, whereas another facility might require more sophisticated measures. These operational characteristics affect the proliferation risk of the facility.