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«Prospects for Conversion of HEU-Fueled Research Reactors in Russia Anatoli S. Diakov Center for Arms Control, Energy, and Environmental Studies, ...»

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Science & Global Security, 22:166–187, 2014

Copyright C Taylor & Francis Group, LLC

ISSN: 0892-9882 print / 1547-7800 online

DOI: 10.1080/08929882.2014.952136

Prospects for Conversion

of HEU-Fueled Research

Reactors in Russia

Anatoli S. Diakov

Center for Arms Control, Energy, and Environmental Studies, Dolgoprudny, Moscow

The importance of converting research reactors from highly enriched uranium (HEU)

fuel, with enrichment levels as high as 90–93 percent uranium-235, to low-enriched uranium (LEU) fuel, was recognized in the 1970s. Russia has developed and produced fuel enriched to below 20 percent to replace HEU-fuel for research reactors it had sup- plied to Hungary, Ukraine, Vietnam, the Czech Republic, Uzbekistan, Libya, Bulgaria, and North Korea, but until recently, has not given priority to the task of converting its own research reactors, despite the fact that Russia now has more HEU-fueled re- search reactors than any other country. In December 2010, Russia and the United States agreed to conduct a preliminary study on the possibility of converting six Rus- sian research reactors. This article assesses the prospects for their conversion.

INTRODUCTION

Nuclear research reactors, including steady state and pulsed reactors, and crit- ical and sub-critical assemblies, offer a source of neutrons and are a unique tool for experimental research in various fields of science and technology. Without them the development of nuclear weapons and nuclear energy would be impos- sible. Over time, the use of research reactors has been adopted in other sectors such as medicine and biology. The number of research reactors in the world increased rapidly starting in the 1950s and reached a maximum of 390 in the 1970s.1 According to the International Atomic Energy Agency, as of mid-2014, more than 747 research reactors, critical and subcritical assemblies of differ- ent types and different capacities, have been built.2 By the early 1980s the growth in the number of nuclear research reactors in the world had stopped.

Received 15 May 2013; accepted 5 March 2014.

Address correspondence to Anatoli S. Diakov, Center for Arms Control, Energy and Environmental Studies, 3 Zhukovskogo Street, #301, Dolgoprudny, Moscow Region, 141700, Russia. E-mail: diakov@armscontrol.ru Color versions of one or more of the figures in the article can be found online at www.tandfonline.com/gsgs.

166 167 Prospects for Conversion of HEU-Fueled Research Reactors in Russia The most powerful research reactors had achieved a neutron flux density of 0.5 × 1015 n/cm2 per second, and attempts to increase flux density beyond these limits ran into problems due to the instability of reactor materials. Solving this materials science problem required significant research and funding. By this time, however, large amounts of experimental data had been accumulated which allowed for the development and verification of computer programs to solve many research problems through calculation and modeling rather than by experimentation. As a result, construction of new research reactors has virtually ceased and increasing numbers of reactors are being decommissioned.

As of mid-2014, there were 247 active research reactors worldwide, with only 6 under construction and 12 planned, while 143 are shut down and 338 decommissioned.3 The principal characteristic of a research reactor is the ratio of the neutron flux density to the reactor power. A priority for researchers and designers has been to maximize the neutron flux density available in experimental channels while minimizing reactor power. Achieving this goal has traditionally required a small core fueled with uranium at the highest possible enrichment level. For this reason, the majority of research reactors in Russia and in the United States were designed to be fueled with HEU, enriched to 90–93 percent uranium-235, a level typically reserved for nuclear weapons.4 In the late 1970s, both in the United States and the Soviet Union, it was recognized that using HEU in civilian research reactors, especially where the fuel was exported to other countries, created proliferation risks. To counter these risks both countries initiated programs to lower the enrichment level of fuel supplied to other countries from 80–90 percent to 20–36 percent uraniumThe two-stage Soviet program to reduce fuel enrichment in research reactors was adopted in the early 1980s5: the first stage reduced the enrichment level to 36 percent, the second—to below 20 percent.

In 1993, Russia and the United States began collaborating on the development of low-enriched fuel (less than 20 percent) for research reactors supplied by Russia (USSR) abroad. This ongoing program is part of the Reduced Enrichment in Research and Test Reactors program (RERTR).

In 1994, the Ministry of Atomic Energy of the Russian Federation initiated the program “Creation of fuel rods and fuel assemblies with fuel enriched to 20 percent uranium-235 for the cores of research reactors.”6 The main goal of the program is the development and organization of the production of fuel assemblies for Soviet- supplied reactors in third countries. This program participants include TVEL Fuel Company (JSC TVEL), N. A. Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET), A.

A. Bochvar All-Russian Scientific Research Institute for Inorganic Materials (VNIINM), Novosibirsk Chemical Concentrate Plant (JSC NCCP), Scientific Research Institute of Atomic Reactors (JSC State Scientific Center - NIIAR), A.I. Leipunski Institute of Physics and Power Engineering (IPPE), Institute of 168 Diakov Reactor Materials (JSC IRM), National Research Center Kurchatov Institute (NRC KI), and the V.P. Konstantinov Petersburg Nuclear Physics Institute.





The program had three phases:

1. Development of fuel rods and fuel assemblies with fuel using UO2 -Al.

2. Development of fuel rods and fuel assemblies with high-density fuel based on uranium-molybdenum alloys.

3. Development of fuel rods and fuel assemblies for the new generation of research reactors.

As of 2006, the work on the first phase had been practically completed.

The production of fuel assemblies VVR-M2 and IRT-4M with enrichment below 20 percent was initiated at the Novosibirsk Chemical Concentrate Plant for research reactors supplied to Hungary, Ukraine, Vietnam, the Czech Republic, Uzbekistan, Libya, Bulgaria, and North Korea.

The success of the first phase laid the foundation for the May 2004 United States-Russian agreement on fuel removal and repatriation and the Russian Research Reactor Fuel Return Program (RRRFR program). This gave a boost to the efforts to remove Russian-made fresh and spent HEU fuel from third countries to Russia and convert those research reactors to LEU fuel. Fourteen countries participated in the program: Belarus, Bulgaria, Czech Republic, Hungary, Germany, Kazakhstan, Latvia, Libya, Poland, Romania, Serbia, Ukraine, Uzbekistan, and Vietnam.

Approximately 1,930 kg of fresh and spent HEU fuel was returned to the Russian Federation by the end of 2012.7 The entire stockpile of HEU fuel was removed from Latvia, Bulgaria, Romania, Libya, Serbia, Ukraine, and Vietnam.8 It is important to note that United States-Russian cooperation on research reactor conversion and return of fresh and spent HEU fuel was supported by joint statements of Russian and U.S. presidents Vladimir Putin and George W. Bush in 2005, and Dmitry Medvedev and Barack Obama in 2009.

Despite the fact that Russia has the largest number of HEU-fueled research reactors inside its borders, the task of converting Russian reactors to LEU fuel was not considered until very recently. Discussions among Russian experts started in connection with the December 2010 Agreement

between Rosatom and the U.S. Department of Energy to conduct a preliminary study on the possibility of converting six Russian research reactors:

Argus, OR, and I-8 at the Kurchatov Institute (Moscow), IRT-MEPhI (Moscow Institute of Physics and Engineering), IRT-T (Tomsk Technical University), and MIR-M1 (Research Institute of Atomic Reactors, Dimitrovgrad).9 The data on this program presented in this article is based on available information about the current status of the civilian research reactors and plans for their use.10 169 Prospects for Conversion of HEU-Fueled Research Reactors in Russia Russia’s Research Reactor Fleet At the end of 2013, there were 30 research reactors in Russia not including the reactors belonging to VNIIEF and VNIITF, which are used in defense programs. The possibility and necessity of conversion of each civilian reactor is determined by its purpose, core design, and plans for its future use (see Appendix, Russian Civilian Research Reactors).

According to Russian Federal Service for Ecological, Technological, and Nuclear Supervision data for 2013, of the 30 civilian research reactors in the country, only 20 had operating licenses. Seven reactors had decommissioning licenses or are in the final stages of shutdown (MR, RBT-10-1, Arbus, BR-10, AM-1, BARS-6 and TVR). One reactor (IRV-M2) has a construction license, and two reactors (IRT-MEPhI, Gamma-KI) are unlicensed. A previous license for IRT MEPhI has expired and an application for a new license had not been submitted.

The fuel for the IBR-2M pulse reactor is produced from plutonium oxide and two others (IR-50 and VK-50) use LEU fuel. Therefore, there are only fifteen licensed research reactors fueled by HEU, one HEU-fueled reactor with a construction license (IRV-M2) and one whose license application has not yet been submitted (IRT-MEPhI). These reactors are described below.

BOR-60: Research Institute of Atomic Reactors (NIIAR), Dimitrovgrad BOR-60 is designed to test fuel elements with different compositions containing plutonium; it was commissioned in 1969. It is a large sodium-cooled experimental fast reactor with a thermal power of 60 MW. It is also used for engineering and safety studies to support the development of sodium-cooled fast neutron reactors and for irradiation of structural materials for nuclear and thermonuclear reactors by neutrons with hard spectrum in the temperature range 300 to 1000 ◦ C.

The reactor core may consist of 85 to 124 fuel assemblies, with fuel composed of either uranium dioxide enriched to 90 percent or a mixture of uranium and plutonium. The uranium enrichment is in the range 45–90 percent and the concentration of plutonium reaches 30 percent. In recent years, the reactor has operated at a power of 53 MW for about 220–230 days per year.

The time factor of utilization (the ratio of the number of full days of full power operation to the number of days in the calendar year) has recently been in the range of 0.60–0.65. Assuming a discharged fuel burn-up of 30 percent, annual uranium-235 consumption is estimated to be up to 39 kg.

The 20 years design lifetime of the reactor has been exceeded twice. In 2009, the reactor was supposed to be renovated and the life extended until

2030. However, an assessment of the performance of various reactor systems 170 Diakov showed that the planned reconstruction was inappropriate and it was decided to extend the life of the BOR-60 only for the period from 2010 to 2015. However, in 2010 the decision was made to extend the operation of the BOR-60 to 2020, because it was seen as critical for realizing the goals of the Federal Targeted Program, “The New Generation for Nuclear Power Technologies for the Period 2010–2015 and Further to 2020.” BOR-60 is expected to operate until the completion of the multipurpose fast neutron research reactor (MBIR) which is to be commissioned in 2019–2020.11 The unique characteristics of BOR-60, the scientific and practical problems that can be studied with it, and its decommissioning timeline exclude the possibility of converting it to LEU fuel.

SM-3: Research Institute of Atomic Reactors (NIIAR), Dimitrovgrad A high-flux, water-cooled and water-moderated tank-type reactor with a thermal capacity of 100 MW, SM-3 is designed primarily for the production of trans-uranium elements and radioactive isotopes of light elements, as well as for irradiation studies of reactor material samples.12 The reactor has an extremely compact core consisting of 28 fuel assemblies and a metal beryllium reflector in a steel vessel. The fuel assemblies consist of fuel rods with a cruciform cross-section. The fuel meat is 90 percent uranium dioxide dispersed in a copper matrix with added beryllium bronze. The mass of uranium-235 in each fuel assembly is 1.128 kg and the average annual fuel consumption is 70 fuel assemblies or 79 kg of uranium-235.13 The utilization factor of the reactor is high, at about 0.7. The design service life of the reactor is 25 years and runs until 2017. However, technical improvements of various reactor systems and experimental studies may allow operation beyond its design service life.

The work on expanding the experimental capabilities of the reactor is continuing in order to allow long-term irradiation of large-size samples of materials for nuclear power plants. For this purpose, the content of uranium-235 in the fuel rods was increased from 5 to 6 grams. Work to replace the reactor core is reported to be underway (Figure 1).14 Russian research reactor experts believe that the SM-3 cannot be converted to LEU fuel and still maintain its key operating characteristics, and the reactor is unlikely to be converted.15

–  –  –

Figure 1: Schematic map of the active core of SM-3 reactor.

for research that does not require high neutron fluence but requires that its neutron flux parameters remain stable over the long term.

The core of RBT-6 consists of 56 spent fuel assemblies of the SM-3 reactor.

The average burnup of loaded fuel assemblies is not less than 35 percent, and the burnup of discharged fuel assemblies is not less than 50 percent. The total mass of uranium-235 in the reactor core at the beginning of a campaign is 32–34 kg. The average duration of a campaign is about 40 days.

The RBT-10/2 reactor core consists of 78 spent fuel assemblies of the SM-3 reactor. Usually, the core is formed from fuel assemblies with a burnup of 10–30 percent, but not more than 50 percent. The average burnup of discharged fuel assemblies is 37–39 percent. Very pure distilled water is used as moderator.



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