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Abstract
In 2007, 334 nuclear reactors (including for naval propulsion) and isotope production facilities employed highly enriched uranium (HEU) fuel or target material. One year of operations at these reactors and facilities required more than 3,100 kilograms (kg) of HEU for naval propulsion, more than 750 kg for research reactors, and 40?–50 kg for isotope production in civilian facilities—in addition to several tons used in other types of reactors. Material with high enrichment levels and low radiation barriers stored or handled in large batches, such as HEU target waste and certain types of fuel from isotope production, research reactors/critical assemblies, and naval fuel, presents serious safety and security concerns. Forty-eight civilian research reactors have converted to low-enriched uranium as a result of a three-decade international effort to minimize HEU use, resulting in a decrease in HEU consumption of 278 kg per year. This article's establishment of baseline measurements for assessing the results of HEU minimization efforts calls for additional focus on the scope and methodology of HEU minimization. Facility decommissioning and dismantling should play a larger role in the future HEU minimization effort, materials with specific weapons-relevant properties should be given higher priority compared to bulk HEU material, and the use of large quantities of weapon-grade HEU fuel for naval propulsion should be reconsidered.
Acknowledgements
This work has been supported by the Norwegian Research Council. This article was written as part of the effort by the International Panel on Fissile Materials (IPFM) to develop a technical basis for practical and achievable policy initiatives to secure, consolidate, and reduce HEU and plutonium stockpiles. The authors would like to express their gratitude for the continued exchange of opinions and knowledge with the members of IPFM, in particular Alexander Glaser and Frank N. von Hippel of Princeton University, in addition to Cristina Hansell of the James Martin Center for Nonproliferation Studies for her helpful comments in the preparation of this paper. Any mistakes, errors, and inconsistencies, however, are the sole responsibility of the authors.
Notes
1. See International Nuclear Fuel Cycle Evaluation (INFCE), Report of INFCE Working Group 8, Advanced Fuel Cycle and Reactor Concepts (Vienna: IAEA, 1980), p. 43.
2. For more information, see the article by Anya Loukianova and Cristina Hansell in the special section in this issue.
3. N. Arkhangelsky, “Twenty Years of RERTR: Past, Present and Future,” presented at RERTR 2000 International Meeting, October 1–6, 2000, Las Vegas, p. 1, <www.rertr.anl.gov/Web2000/Title-Name-Abstract/Arkhang00.html>.
4. In May 2004, the United States established GTRI as an umbrella program for RERTR and efforts to retrieve HEU fuels exported by Russia and the United States, as well as a program to retrieve proliferation-sensitive radioactive sources.
5. See Ole Reistad and Styrkaar Hustveit, Appendix II (online only), Nonproliferation Review 15 (July 2008), <cns.miis.edu/pubs/npr/vol15/152_reistad_appendix2.pdf>.
6. “Burnup” describes the degree to which the content of U-235 has been reduced. It varies between 1–4 percent for target material—material placed in reactors for isotope production—and up to 70–75 percent for reactor fuel. Target burnup is extremely low due to the plateauing of Mo-99 in the targets as irradiation time increases toward the Mo-99 half-life. Thus, the HEU contained in target waste is reduced by only 1–2 percent less relative to the original amount. Furthermore, HEU target waste tends not to be highly radioactive. Depending on the producer, it can also be in a form that is relatively easy to process into metal via a well-known chemical process (in the case of liquid waste) or is relatively easily handled (small containers of solid wastes). In general, reactor fuel stored more than ten to fifteen years is not self-protecting. This is dependent on the mass of U-235 burned, the fraction of U-235 burned, and the specific power density. See R.B. Pond and J. Matos, “Photon Dose Rates from Spent Fuel Assemblies with Relation to Self-Protection (Rev. 1),” ANL/RERTR/TM-25, RERTR Program, February 1996.
7. Determinations of annual consumption can be made on the basis of refueling frequency and batch size. This can be done, for example, for the German FRM-II reactor (8 kg, 90 percent enriched, supplied five times a year). In the cases of most research reactors, however, this information is unavailable. Therefore, we have estimated refueling frequency on the basis of core inventory, power, capacity factor, and burnup (g (U-235)/ MWd). We specify the calculation method used for each fuel cycle in the report, as it varied by each reactor type.
8. Ole Reistad, Morten Bremer Mærli, and Nils B⊘hmer, “Russian Naval Nuclear Fuel and Reactors: Dangerous Unknowns,” Nonproliferation Review 12 (Spring 2005), pp. 163–197.
9. These sources include Mitchell K. Meyer, “A Global Overview of High Density UMo Fuel Development Efforts,” paper presented at Minimization of Highly Enriched Uranium (HEU) in the Civilian Sector, technical workshop, Oslo June 18, 2006, <www.nrpa.no/symposium>; J.E. Matos, “Relationships Between the RERTR Program and the U.S. Spent Fuel Acceptance Policy,” Waste Management Conference, Tucson, Arizona, March 15, 1998. For similar studies, see Robert L. Civiak, “Closing the Gaps, Securing Highly Enriched Uranium in the Former Soviet Union and Eastern Europe,” Federation of American Scientists, May 2002. For additional estimates, consult David Albright, Frans Berkhout, and William Walker, Plutonium and Highly Enriched Uranium 1996: World Inventories, Capabilities and Policies (Oxford: Oxford University Press, 1997), pp. 486–489.
10. For the ATR and HFIR facilities, IAEA figures on burnup and operational availability are not entirely credible because they indicate that ATR is never shut down, which is impossible, while in the case of HFIR, the given information in the IAEA reactor database for calculating fuel consumption is not consistent with given data from the operator. Annual consumption numbers used in the figure for these facilities are from presentations at the RERTR 2007 International Meeting. On ATR, see, G. Chang, “ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Plate Thickness Variation,” and on HFIR, see, J.D. Sease, R.T. Primm, and J. Miller, “Considerations in the Development of a Process to Manufacture Low-enriched Uranium Foil Fuel for the High Flux Isotope Reactor,” both papers presented at the RERTR 2007 International Meeting, Prague, September 23–27, 2007.
11. This estimate is based on an assumption that deviations between the calculated and given values as discussed above exhibit a normal distribution.
12. See the article on isotopes by Cristina Hansell in the special section in this issue.
13. Frank N. von Hippel and Laura Kahn, “Feasibility of Eliminating the Use of Highly Enriched Uranium in the Production of Medical Isotopes,” Science and Global Security 14 (2006), pp. 151–162.
14. “Cintichem” refers to the method for processing targets developed by the U.S. firm Union Carbide, which formed the basis for the production of Mo-99 in the United States before 1989. The commercial rights to the Cintichem process were later procured by the Department of Energy (DOE), which made this technology the foundation for U.S. technical assistance for developing LEU-based Mo-99 production processes.
15. The assumed yield is taken from the Cintichem process: 540 curies Mo-99/g HEU (90 percent enriched), as given in George Vandegrift, “Facts and Myths Concerning 99Mo Production with HEU and LEU Targets,” paper presented at the RERTR 2005 International Meeting, Boston, November 6–10, 2005.
16. IAEA, “Management of High Enriched Uranium for Peaceful Purposes: Status and Trends,” IAEA-TECDOC-1452, June 2005, p. 6.
17. U.S. National Nuclear Security Administration (NNSA)–Australian Nuclear Science and Technology Organization, Global Initiative to Combat Nuclear Terrorism (workshop report), Workshop on the Production of Mo-99 Using LEU, Sydney, Australia, December 2–5 2007, p. 1.
18. George F. Vandegrift, Allen J. Bakel, and Justin W. Thomas, “Overview of 2007 ANL Progress for Conversion of HEU Based Mo-99 Production as Part of the U.S. Global Threat Reduction—Conversion Program,” paper presented at the RERTR 2007 International Meeting, Prague, September 23–27, 2007.
19. George F. Vandegrift, Allen J. Bakel, and Justin W. Thomas, “Overview of 2007 ANL Progress for Conversion of HEU Based Mo-99 Production as Part of the U.S. Global Threat Reduction—Conversion Program,” paper presented at the RERTR 2007 International Meeting, Prague, September 23–27, 2007.,p. 5. See also the discussion in the article by Cristina Hansell on nuclear medicine in the special section in this issue.
20. Between 1955 and 1972, U.S. critical facilities at Los Alamos concentrated on supporting the naval propulsion program. See David Loaiza and Daniel Gehman, “End of an Era for the Los Alamos Critical Experiment Facility: History of Critical Assemblies and Experiments (1946–2004),” Annals of Nuclear Energy, 2006, p. 1357.
21. A number of critical assemblies are used for training and practice to avoid criticality accidents. The United States recently established a five-year plan for its efforts on nuclear criticality safety, with various configurations involving HEU. DOE, “Nuclear Criticality Safety Program: Five-Year Plan,” October 2004, <www.hss.energy.gov/deprep/2005/AttachedFile/tb05f01b_enc2.pdf>.
22. In France, for instance, the available plutonium rodlets and plates inventory for simulating large fast neutron reactor cores proved insufficient, so the core loading was completed with HEU enriched to 30–35 percent. See Massimo Salvatores et al., “Advanced Fast Reactor Development Requirements: Is There Any Need for HEU?” presented in Oslo at the International Symposium (technical workshop) on Minimization of HEU in the Civilian Sector, June 18, 2006. Mocking up fast reactor core designs as proposed today using enrichment levels lower than 20 percent will not provide the required minimum reactivity needed to complement the plutonium loading, while enrichment higher than 30–35 percent does not seem to be needed. Russia has used disks of 36 percent and 90 percent HEU interleaved with disks of depleted uranium to simulate low- and medium-enriched uranium fuel in a fast neutron reactor core. See Frank N. von Hippel, “Future Needs For HEU-Fueled Critical Assemblies,” proceedings of the International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, November 6–10, 2005.
23. The pulsed reactors with the biggest HEU inventories are used by the nuclear weapon states to simulate the effects of neutrons from nuclear explosions on missile warheads and satellites. Most pulsed reactors have lifetime cores. That is, despite their high power, their pulses are so short and infrequent that their integrated energy output does not greatly deplete the fuel's U-235 inventory.
24. I.P. Matveyenko et al., “Physical Inventory of Nuclear Materials on BFS Facility,” paper presented at the Materials Protection, Control, and Accounting Conference, Obninsk, Russia, May 22–26, 2000.
25. For further discussion of Russian critical installations, see the article by Elena Sokova in the special section in this issue.
26. Most naval reactor cores have a volume of about 1 cubic meter (m3) and a power density of up to 200 MW/m3. High power densities require efficient heat removal from the reactor core. This is achieved through the use of large heat transfer surfaces, or in some cases, by using liquid metal instead of water as coolant.
27. HEU was used in space reactors during the Cold War by both the Soviet Union and the United States. While no space reactors are currently in operation, they are under consideration for future space operations. According to the DOE's five-year plan, NASA was interested in using a space reactor to power the Jupiter Icy Moons Orbiter. This program has since been canceled. See DOE, “Nuclear Criticality Safety Program,” p. 34. In space missions, the use of HEU can reduce the size and weight of the reactor and its associated radiation shielding.
28. This approach demonstrates that increased understanding can be achieved by systemizing available information on similar power systems. As there are no distinct tendencies in this data set between small versus large reactors, old versus new vessels, U.S. versus Russian facilities etc., normal distribution has been assumed. The main exception is the propulsion system in the Lenin icebreaker installed after an accident in 1967; in this case, while keeping the vessel shaft power, the new system had an additional 140 MWt compared with the old system. This system therefore has not been included in the calculations.
29. Fifty percent burnup has been assumed on the basis of DOE, Highly Enriched Uranium: Striking a Balance, A Historical Report on the United States Highly Enriched Uranium Production, Acquisition, and Utilization Activities from 1945 through September 30, 1996 (Washington, DC: DOE, 2001). In that report, the average enrichment level for spent U.S. submarine fuel has been given as 83 percent, a level that represents approximately 50 percent burnup when assuming an initial enrichment level on 97 percent. Alexander Glaser, “On the Proliferation Potential of Uranium Fuel for Research Reactors at Various Enrichment Levels,” Science and Global Security 14 (2006), pp. 1–24, discusses the specific consumption of U-235 as a function of various enrichments levels. In Ole Reistad, R. Stamm'ler, and K. Gussgard, “Modeling Design and Operating Parameters for Russian Naval PWR Reactors Using the HELIOS Reactor Code,” submitted for publication in Science and Global Security, February 4, 2008, p. 22, the specific consumption for HEU-fueled vessels was calculated at 1.24–1.27 gU-235/MWd. On the EFPDs, National Academy of Sciences, Committee on International Security and Arms Control, Management and Disposition of Excess Weapons Plutonium — Reactor-Related Options (Washington, DC: National Academy Press, 1995), p. 78. The uncertainty was established on the basis that the operation may possibly, but rather unlikely, be as low as eighteen EFPD, as was the case, for example, with the first Russian submarines in the early 1960s. Operations above the upper limit of fifty-four EFPD should be equally rare, as cruising speed is much less than full power.
30. This estimate, equaling 580 GWt over four and a half years, is taken from G.A. Gladkov and Yu. V. Sivintsev, “Radiation Conditions Near the Sunken Submarine ‘Komsomolets,’” Atomnaya Énergiya 77 (November 1994), p. 379, specifying the operational experience for Komsomolets, the only modern Russian submarine for which operating figures have been published.
31. U.S. Navy, “Report on Use of Low Enriched Uranium in Naval Nuclear Propulsion,” June 1995, p. 9.
32. V.I. Makarov et al., “Experience in Building and Operating Reactor Systems for Civilian Ships,” Atomic Energy 89 (1996), pp. 691–700.
33. Chunyan Ma and Frank von Hippel, “Ending the Production of Highly Enriched Uranium for Naval Reactors,” Nonproliferation Review 8 (Spring 2001), pp. 86–101.
34. The Japanese JOYO reactor may be HEU-fueled; if so, with a fuel enrichment level just above 20 percent.
35. IAEA, “Management of High Enriched Uranium for Peaceful Purposes: Status and Trends,” p. 16.
36. Kang Ya-lun, “Progress of Fuel Assembly Design and Test for China Advanced Research Reactor,” Atomic Energy Science and Technology 37 (July 2003); Yin Chang-geng “Development of Fuel Element for Research Reactor in Nuclear Power Institute of China,” Atomic Energy Science and Technology 39 (July 2005), pp. 20–24.
37. Oleg Bukharin, “Securing Russia's HEU Stocks,” Science and Global Security 7 (1998), p. 328, footnote 7.
38. Oleg Bukharin, “Securing Russia's HEU Stocks,” Science and Global Security 7 (1998), p. 328, footnote 7., p. 328, footnote 8.
39. According to Olivier Caron, French governor to the IAEA, in his statement in Oslo at the International Symposium on Minimization of HEU in the Civilian Sector, June 19, 2006, none of the designs under development today, whether through the IAEA's International Project on Innovative Nuclear Reactors and Fuel Programs or through the Generation IV International Forum program, call for the use of HEU; however, the research and development may involve experiments where HEU presently plays a vital role.
40. Today Cadarache does not have enough plutonium to undertake these experiments without HEU. It should be noted, however, that “enrichment higher than 30–35% does not seem to be needed to mock-up conceivable core design as proposed today.” Salvatores et. al, “Advanced Fast Reactor Development Requirements.”
41. IAEA, “Management of High Enriched Uranium for Peaceful Purposes: Status and Trends,” p. 11.
42. INFCE, Report of INFCE Working Group 8, Advanced Fuel Cycle and Reactor Concepts, p. 43, stated that the amount of HEU consumed in civilian steady-state reactors in 1978 was “over 1200 kg U-235.” Our calculations show that the total amount for all civilian research reactors in 1978 was 1,161 kg U-235/1,351 HEU. Given the uncertainties, these figures are consistent with each other. Regarding the number of facilities and nominal power, INFCE mentioned “more than 150 HEU-fueled facilities of significant power” with total nominal power of more than “1700 MW.” Not all types of research facilities were included in the INFCE scope, however. Our count of all HEU-fueled facilities in all regions shows that there were at least 244 HEU-fueled research reactors in 1978 with a total nominal power of 1,919 MW.
43. NNSA, Criticality Experiments Facility Project, Nevada Site Office Fact Sheet DOE/NV 1063, 2005, p. 1, <www.nv.doe.gov/library/factsheets/DOENV_1063.pdf>.
44. For more information on Russian HEU policy, see the article by Elena Sokova in the special section in this issue.
45. Parrish Staples and Nicholas Butler, “Conversion of Research and Test Reactors: Status and Current Plans,” paper presented at the Meeting of the International Group on Reactor Research, Lyon, France, March 11–15, 2007.
46. Ann MacLachlan, “NRG to Study Potential for Use of LEU for Mo-99,” Nuclear Fuel 32 (December 17, 2007).
47. For an in-depth discussion of the politics of converting Mo-99 production, including the status of current plans for non-HEU production, see the article on nuclear medicine by Cristina Hansell in the special section in this issue.
48. Von Hippel and Kahn, “Feasibility of Eliminating the Use of Highly Enriched Uranium in the Production of Medical Isotopes,” p.159.
49. These results are sensitive to assumptions concerning core inventory, which have error margins of at least 40 percent. The suggested uncertainty is based on a variation in the burnup figures/EFPDs of up to 40 percent; however, the variability is probably less for current reactor types than for first- and second-generation propulsion reactors. A major component of the HEU consumption is due to the U.S. aircraft carriers, as they have reactors with significantly higher maximum power output and consequently larger inventories. Similar variability for the first generations of Russian submarines has been assumed as for U.S. submarines; in this case data on the core size of the early naval reactors have been assessed using data on reprocessed naval fuel. See Thomas Cochran et al., Nuclear Weapons Databook, Volume II: U.S. Nuclear Warhead Production (Cambridge: Ballinger, 1987), p. 186. As for the third- and fourth-generation submarine reactors, the proposed core inventory is based on the assumptions of 50 percent burnup, thirty-six EFPDs, lifetime cores (thirty-three years) for the Virginia class, and power levels as suggested in Table 3.7, some of which was also suggested in earlier sources. The suggested data on power levels are taken from Ma and von Hippel, “Ending the Production of Highly Enriched Uranium for Naval Reactors.” With respect to the core size calculation, the figures are taken from National Academy of Sciences, Management and Disposition of Excess Weapons Plutonium—Reactor-Related Options, p. 12.
50. To preserve the longevity of the core, core volume would have to be increased threefold, and the pressure vessel, the reactor compartment, and the size and cost of the vessel itself would have to increase correspondingly. According to the assessment, construction costs would rise “about 28% for aircraft carriers and 26% for submarines—about $1.1 billion pr. year.” U.S. Navy, “Report on Use of Low Enriched Uranium in Naval Nuclear Propulsion,” p. 22.
51. For a broader discussion see Ma and von Hippel, “Ending the Production of Highly Enriched Uranium for Naval Reactors.”
52. T.D. Ippolito Jr., “Effects of Variation of Uranium Enrichment on Nuclear Submarine Reactor Design,” master's thesis, Department of Nuclear Engineering, Massachusetts Institute of Technology, May 1990.
53. A.M. Dmitriev, A.S. Diakov, and A.M. Shuvaev, “Assessment of Feasibility of Converting Russian Icebreaker KLT-40 Reactors from HEU to LEU Fuel,” paper presented at the RERTR 2005 International Meeting, Boston, November 6–10 2005.
54. Cristina Hansell Chuen and Ole Reistad, “Sea Fission: Russia's Floating Nuclear Power Plants,” Jane's Intelligence Review, December 2007, pp 1–7.
55. With its recent introduction of lifetime cores, the United States probably enjoys a considerable advantage regarding optimization of operational costs and, perhaps, properties such as diving depth because of the lack of large refueling hatches. The U.S. long-term plan is to introduce one Virginia-class submarine every year until 2020, while reducing the number of Los Angeles-class subs by one every year.
56. The United Kingdom is currently trying to decide what to do when the operational lifetimes of the naval reactors in its Vanguard-class vessels come to an end by the mid-2020s.
57. Regarding the annual use of HEU, the main issue is continued phase-out of the PWR-1 (Swiftsure and Trafalgar submarine classes) and the introduction of the PWR-2 reactor in the Astute class, one vessel every other year beginning in 2009. A reasonable assumption about the differences between these two installations—PWR-1 and PWR-2—is that the latter has a considerably larger core inventory. However, the net effect on the overall use of HEU is limited.
58. Stephen Saunders, ed., Jane's Fighting Ships 2005–06 (Alexandria, VA: Jane's Information Group, 2005), p. 602. The attack submarines (Sierra I and II, Akula, Oscar II, Victor) are assumed to have a lifespan of twenty years, somewhat longer than earlier versions of similar vessels. Nevertheless, Russia's nuclear navy will be down to fewer than fifteen vessels after 2015. According to current estimates made by the authors, the strategic submarines (Delta III, Delta IV, Typhoon) are assumed to have a lifespan of thirty years after retrofitting.
59. IAEA, “The Physical Protection of Nuclear Material and Nuclear Facilities,” INFCIRC/225/Rev.4 (Corrected).
60. Out of a remaining 1,414 kg not repatriated to the United States, 101 kg had been acquired as of August 2007, and there have been plans to repatriate the remainder as part of the GTRI “gap materials” program. The Russian Research Reactor Fuel Return Program started in 2002 with U.S. funding and has since come under the GTRI framework. The IAEA gives substantial support to this program. In May 2004, Russia and the United States signed an Intergovernmental Agreement concerning cooperation for the transfer and repatriation of Soviet-origin research reactor nuclear fuel to the Russian Federation. Thirteen countries have joined the program to date. As of August 2007, 433 kg of U-235 in fresh fuel and 63 kg of irradiated fuel had been returned to Russia by various countries. Further repatriation of both fresh and spent fuel is planned through 2009.
61. Burnup varies between 1–4 percent for targets and can be more than 50 percent for research reactor fuel. (The U.S. HFIR reactor has a maximum burnup of 70 percent.)
62. After the Russian military fleet experienced numerous reactor and fuel failures, a new fuel matrix was developed consisting of a uranium-zirconium alloy with zirconium cladding for the new civilian reactor models OK 900, OK 900 A, and KLT-40. However, given the current technology available at Russian reprocessing facilities, this fuel is not reprocessable. Different solvents are necessary to extract the fission products and fissile material. As a result, this fuel, containing weapons-quality HEU, is accumulating on service ships and at the icebreaker base in Northwest Russia. The question is where and how the spent cores are stored. The Murmansk Shipping Company indicated in 1996 that 3,100 fuel assemblies are stored on the service ship Lotta, which is moored in the harbor area of the Murmansk Shipping Company (Murmansk Shipping Company, “Information on the Non-Reprocessable Spent Fuel from the Nuclear Icebreaker Fleet Stored on the MSCO's Service Ships,” unpublished, undated). Additional fuel of this type is also stored on the service vessel Lepse. Together these facilities hold some fourteen spent icebreaker cores, as mentioned above. Icebreaker fuel may also be held at the storage facility in Andreeva Bay. However, as there is no complete inventory of Russia's spent naval nuclear fuel, this remains a matter of speculation.
63. IAEA Research Reactor Database, <www.iaea.org/worldatom/rrdb/>.
64. Armando Travelli, “Proceedings of the 1978 International Meeting on Reduced Enrichment for Research and Test Reactors,” Argonne National Laboratory, ANL/RERTR/TM-1 CONF-781151, p. 3. The argument was that it is reasonable to consider as an average figure that every kilogram introduced to an HEU fuel cycle stays in the fuel cycle for four years due to storage before it is loaded into the reactor and after being irradiated for cooling purposes.
65. A “significant quantity” is the approximate quantity of nuclear material in respect of which, taking into account any conversion process, the possibility of manufacturing a nuclear explosive device cannot be excluded. The IAEA considers 25 kg a “significant quantity” of U-235.
66. Cores/year is a measure of the length of a core-life, e.g., the annual consumption of HEU has been averaged over all operational years. With respect to the United States, the core-life is a reasonably known measure because the development and vessel upgrades of the U.S. Navy are part of a systematic process.
67. Reistad et al., “Modeling Design and Operating Parameters for Russian Naval PWR Reactors Using the HELIOS Reactor Code,” p. 22.
68. Ma and von Hippel, “Ending the Production of Highly Enriched Uranium for Naval Reactors,” p. 89.
69. Estimated based on the amount of spent naval fuel reprocessed from 1954 to 1984. See Cochran et al., Nuclear Weapons Databook. We have assumed an increase in core inventory to 350 kg for the U.K. model PWR-2.
70. The United Kingdom uses its base in Scotland for these operations. The Russian icebreaker fleet has only one site, inside the city limits of Murmansk, to and from which all fresh and spent fuel is transported. However, in addition to being transported on service ships to the bases for refueling, significant amounts of spent fuel are stored at similar service ships not in operation. The relevant sites are Severomorsk in Northwest Russia, and Site 34 at Chazhma, Russian Far East (Kamchatka). As the Russian Navy previously refueled submarines at its bases, this situation represents an important step in fresh fuel consolidation implemented during the last ten years. See James Clay Moltz, “Russian Nuclear Submarine Dismantlement and the Naval Fuel Cycle,” Nonproliferation Review 7 (Spring 2000), pp. 76–87.