The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years

The GIF Proliferation Resistance and Physical Protection (PR&PP) Evaluation Methodology: Status, Applications and Outlook

Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas of sustainability, safety and reliability, economics, proliferation resistance and physical protection (PR&PP). The last of these four areas was assigned to the GIF Working Group on Proliferation Resistance and Physical Protection (PRPPWG). The PRPPWG developed the methodology through a series of development and demonstration case studies, by use of a hypothetical “Example Sodium Fast Reactor” (ESFR). This is a generic design of Generation IV reactor based on the US Advanced Fast Reactor (AFR) developed by Argonne National Laboratory. The PR&PP ESFR assessment was the first opportunity to exercise the full methodology on a complete system, and many insights were gained from the process. In particular, the approach of breaking the assessment into subtasks, each focusing on a separate area of PR&PP (for PR: diversion, misuse, breakout; for PP: theft and sabotage) handled by a dedicated subgroup with diverse international membership, was useful in generating new insights and concept development. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the PR&PP methodology, its application to the ESFR case study, and other accomplishments of the PRPPWG. Current challenges with the efficient implementation of the methodology are outlined, along with the path forward for increasing its accessibility to a broader stakeholder audience - including supporting the next generation of skilled professionals in the field of nuclear non-proliferation and security.JRC.G.II.7-Nuclear securit

Status of the Gen-IV Proliferation Resistance and Physical Protection (PRPP) Evaluation Methodology

Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas safeguards, security, economics and safety. Of these four areas, safeguards and security are the subjects of the GIF working group on Proliferation Resistance and Physical Protection (PRPP). Since the PRPP methodology (now at Revision 6) represents a mature, generic, and comprehensive evaluation approach, and is freely available on the GIF public website, several non-GIF technical groups have chosen to utilize the PRPP methodology for their own goals. Indeed, the results of the evaluations performed with the methodology are intended for three types of generic users: system designers, program policy makers, and external stakeholders. The PRPP Working Group developed the methodology through a series of demonstration and case studies. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the technical progress and accomplishments of the PRPP evaluation methodology, including applications outside GIF, and we outline the PRPP methodology’s relationship with the IAEA’s INPRO methodology. Current challenges with the efficient implementation of the methodology are outlined, along with our path forward for increasing its accessibility to a broader stakeholder audience – including supporting the next generation of skilled professionals in the nuclear non-proliferation field.JRC.E.8-Nuclear securit

The IAEA CRP on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste

In 2003, the IAEA has initiated the Coordinated Research Project (CRP) on ¿Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste¿. The overall objective of the CRP, performed within the framework of IAEA¿s Nuclear Energy Department¿s Technical Working Group on Fast Reactors, is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. Twenty institutions from 15 Member States and one international organization participated in this CRP. The CRP concentrated on the assessment of the dynamic behavior of various transmutation systems. The reactor systems investigated comprise critical reactors, sub-critical accelerator driven systems with heavy liquid metal and gas cooling, critical molten salt systems, and hybrid fusion/fission systems. Both fertile and fertile-free fuel options have been investigated. Apart from the benchmarking of steady state core configurations (including the investigation of transmutation potential, burn-up behavior and decay heat of minor actinide (MA) bearing fuels), the CRP participants determined the safety coefficients for the individual systems and, in a second stage, performed transient analyses which reflected the generic safety related behavior of the various reactors types.JRC.F.4-Safety of future nuclear reactor

REPORT ON INTERMEDIATE RESULTS OF THE IAEA CRP ON STUDIES OF ADVANCED REACTOR TECHNOLOGY OPTIONS FOR EFFECTIVE INCINERATION OF RADIOACTIVE WASTES

In 2003 the IAEA has initiated a Coordinated Research Project (CRP) on ‘‘Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste”. Major intermediate results have been obtained and will be reported here. The overall objective of the CRP, performed within the framework of IAEA’s Nuclear Energy’s Department Technical Working Group on Fast Reactors, is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. Sixteen institutions from 12 member states and one international organization participated in this CRP. The CRP concentrated on the assessment of the dynamic behaviour of various transmutation systems. The reactor systems investigated comprise critical reactors, subcritical accelerator driven systems with heavy liquid metal and gas cooling, critical molten salt systems and hybride fusion/fission systems. Both fertile and fertile-free fuel options have been investigated. For a deep assessment of the transient and safety behaviour, the analytical capabilities have to be qualified. A major effort of the CRP consisted in the benchmarking of steady state core configurations and performing transient/accident simulations. For a general assessment and comparison, the safety coefficients were determined for the individual systems. In a second step transient analyses were performed which reflected the generic behaviour of the various reactors types. In addition the transmutation potential, burn-up behaviour and decay heat of minor actinide bearing fuels were investigated

REPORT ON INTERMEDIATE RESULTS OF THE IAEA CRP ON STUDIES OF ADVANCED REACTOR TECHNOLOGY OPTIONS FOR EFFECTIVE INCINERATION OF RADIOACTIVE WASTES

In 2003 the IAEA has initiated a Coordinated Research Project (CRP) on ‘‘Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste". Major intermediate results have been obtained and will be reported here. The overall objective of the CRP, performed within the framework of IAEA's Nuclear Energy's Department Technical Working Group on Fast Reactors, is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. Sixteen institutions from 12 member states and one international organization participated in this CRP. The CRP concentrated on the assessment of the dynamic behaviour of various transmutation systems. The reactor systems investigated comprise critical reactors, subcritical accelerator driven systems with heavy liquid metal and gas cooling, critical molten salt systems and hybride fusion/fission systems. Both fertile and fertile-free fuel options have been investigated. For a deep assessment of the transient and safety behaviour, the analytical capabilities have to be qualified. A major effort of the CRP consisted in the benchmarking of steady state core configurations and performing transient/accident simulations. For a general assessment and comparison, the safety coefficients were determined for the individual systems. In a second step transient analyses were performed which reflected the generic behaviour of the various reactors types. In addition the transmutation potential, burn-up behaviour and decay heat of minor actinide bearing fuels were investigate