3.4. Future applications of BUC
3.4.4 Advanced reactor fuels
3.4.4.1 Types
There are a number of advanced reactor and advanced reactor fuel (ARF) designs being considered by the US DOE Generation IV road mapping Group and the IAEA INPRO Group.
Additional road mapping activities have been performed; reports issued include [9] and [10].
A partial list of some of the better-known concepts are summarized in Table VII and listed below.
Molten Salt Fluoride - salt
Reprocessing Recycle Need Fission Product BUC
Pebble bed reactor
The pebble bed modular reactor (PBMR) is a graphite helium cooled reactor, which uses the Brayton direct gas cycle to convert the heat that is generated in the core nuclear fission. The heat is transferred to the coolant gas (helium), where it is converted into electricity by the means of a gas turbine-generator. The PBMR core is based on the German high temperature gas reactor technology and uses spherical fuel elements. The fuel is typically not reprocessed due to its low fissile loading. The fuel typically has an enrichment of about 8 wt% with a typical discharge burnup of 80 GWd/kg U.
Gas turbine-high temperature gas reactor
The Gas Turbine-Modular Helium Reactor (GT-MHR) is an advanced nuclear power system designed to provide very high safety, high thermal efficiency, environmental compliance, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR), contained in one vessel, with a high efficiency Brayton cycle gas turbine (GT) energy conversion system contained in an adjacent vessel. The reactor and power conversion vessels are interconnected with a short cross-vessel and are located in underground concrete silo. The MHR refractory coated particle fuel consists of a spherical kernel of fissile or fertile material, as appropriate for the application, encapsulated in multiple coating layers.
The overall diameter of the standard particles varies from 650 microns to about 850 microns.
For the GT-MHR the particle size vary from 13 mm in diameter to 51 mm in length. The fuel particles are loaded in a “compact” and the compacts are loaded into hexagonal channels. The hexagonal channels are made of graphite, 793 mm long and 360 mm across the flat. The fuel is anticipated to be in the low enrichment range with discharge burnups of greater than 100 GWd/kg U.
Advanced light water reactor
The advanced light water reactor is the next generation LWR, using the once through UO2
fuel cycle. The joint U.S. DOE and industry funded program resulted in three NRC-certified ALWR designs, the General Electric Advanced Boiling Water Reactor (ABWR), the Combustion-Engineering (now Westinghouse) System 80+, and the Westinghouse AP600.
The last design is a passive-safety mid size plant (approx. 600 MW(e), ABWR and System 80+ are large 1300 MW(e) evolutionary designs. The fuel type is similar to the LWRs now in operation, with a few basic improvements. The improvements are based in the materials of construction of the fuel. The materials have migrated to higher and higher purity with the selection of higher temperature alloys. The enrichments for the new generation fuel assemblies are focused on a maximum of approximately 6-wt% U-235.
Advanced fast reactor
The advanced fast reactor (AFR) is an extension of the existing fast reactor technology. Three different coolants are being considered: sodium (reference design), lead and helium. Two different fuel forms are being considered: oxide and metal. The fuel is encased in stainless steel cladding on a triangular array and the enrichment ranges from 20 to 35%
plutonium/uranium, with an anticipated burnup in excess of 100 GWd/kg U. It is envisioned that the AFR will be operated under a closed fuel cycle (reprocessing). The oxide fuel can be reprocessed by an aqueous (e.g. PUREX) process or by the combined oxide reduction/electrochemical process. The metal fuel will be electrochemically reprocessed.
Molten salt reactor
The molten salt reactor (MSR) is a fluid fuelled reactor that uses a homogeneous mixture of molten salt as the fuel and coolant. The UF4 fuel is dissolved in a mixture of fluoride based salts, forming a clear liquid at reactor operating temperatures of about 650oC. MSR can be operated with U-235 or U-233 as the fissile component of the fuel. The fuel salt circulated through the graphite-moderated reactor vessel to a circulating pump and then through a heat exchanger before returning to the reactor vessel. Because of the fluid nature of the fuel some form of on-site or off-site treatment/reprocessing is envisioned. Since there are no traditional fuel assemblies, an online salt slurry treatment process is used to remove the fission products and impurities. BUC is not anticipated for this reactor fuel type.
3.4.4.2 Specific applications
This application of advanced reactor fuels (ARF) does not differ generally from the BUC applications for conventional LWR fuels. The potential applications for BUC include:
• Wet storage at or away from the reactor,
• Dry storage at or away from the reactor,
• Transportation wet or dry,
• Reprocessing,
• Disposal.
The specific application of BUC depends on the specific ARF, as well as technical, economical, political decisions and considerations. In general BUC application can be expected for any large-scale operations.
3.4.4.3 Benefits and motivations
From the point of view of the nuclear industry, the reduction of cost is the driving force. BUC application can cause an increase of storage capacity per cask or and increase of the capacity of a storage pool.
The increase of cask capacity reduces the number of the cask to be handled and shipped. This will result in a decreased impact for the environment from handling and shipping activities. It also decreases the risk of an accident since the number of shipments is less. Furthermore, dose to workers and the public may be reduced.
In critical analyses for long-term storage and disposal, specific aspects may be important that make it necessary to consider burnup. Finally, the investigation of burnup effects on application of the spent fuel may result in an increase of safety margin even if BUC is not applied in any specific case.
3.4.4.4 Status
For the most part the advanced reactor fuel (ARF) types are at the conceptual development stage with limited prototyping. There is little work on BUC for ARF types at this time.
The ongoing development activities for the advanced LWR fuels with very high burnup (100 GWd/tU), include activities that will support BUC application. The material testing, benchmarking and PIE/radiochemical assay work will be useful for BUC applications.
3.4.4.5 Future plans
There are few plans for BUC-activities with the ARF. Until a significant number of AFRs exist, there is no benefit to make definite plans.
3.4.4.6 R&D and operational needs
Implementing BUC for new fuel types requires the following tasks:
(1) Identifying the relevant activities, nuclides, fissile nuclides further actinides, fission products;
(2) Studying the influencing parameters e.g. from reactor operation and cooling, by calculation and measurement;
(3) Validation of codes for inventory and criticality calculations according to specific fuel types;
(4) Development of methods for practical implementation, e.g. verification or measurement, further safety measurement to exclude inadvertent interchange.
Further needs depend on the specifics of the fuel under consideration 3.4.4.7 Regulatory consideration
The regulatory concerns for BUC in ARF are the same as for BUC in current reactor fuels.
Protection of the public and workers, health and safety, and protection of the environment are still the underlying considerations.
The potential benefits of BUC to health and environmental protection would continue to be a consideration to be accounted for by regulators.
3.4.4.8 Conclusions and recommendations MOX - LWR MOX and fast MOX
The type of MOX fuel under consideration determines whether or nor it is advisable to claim for BUC. For standard RG MOX, as used in Europe and in Japan, BUC is of major interest for reprocessing, and to a lesser extent for storage, transportation and disposal. For WG MOX, BUC procures the same advantages as UO2 fuel. Fast reactor MOX fuel could benefit from fission products credit for wet storage.
Future developments for BUC for MOX should cover calculation methodologies, PIE, for nuclide inventory of the spent fuel and critical experiments.
Advanced reactor fuel designs
There are a large number of advanced reactor fuel designs being considered as concepts. In general, none are sufficiently developed so that there are clear motivations or benefits for implementing BUC at this time.
However, from experience with applications of BUC with current LWR fuel, benefits can be expected if any of the ARF is developed to an operational stage.
Also from experience with implement BUC for current LWR fuels minor additions to developmental activities will allow easier implementation of BUC for the AFRs.
Accordingly, consideration of BUC implementation issues should be included in the development (testing/experimenting) for ARF concepts.
REFERENCES
[1] ORGANIZATION OF ECONOMIC CO-OPERATION AND
DEVELOPMENT/NUCLEAR ENERGY AGENCY, JEFF Report 17, "The JEF2.2 Nuclear Data Library", (2000).
[2] ORGANIZATION OF ECONOMIC CO-OPERATION AND
DEVELOPMENT/NUCLEAR ENERGY AGENCY, ICSBEP Guide to the Expression of Uncertainties Revision, International Handbook of Evaluated Criticality Safety Benchmark Experiments, OECD, NEA/NSC/DOC(95)03, September 2001 Edition.
[3] INTERNATIONAL ATOMIC ENERGY AGENCY, “Implementation of burnup credit in spent fuel management systems” (Proc. of a TCM Vienna, 2000), IAEA-TECDOC-1241, Vienna (2001).
[4] INTERNATIONAL ATOMIC ENERGY AGENCY, “Implementation of burnup credit in spent fuel management systems” (Proc. of a AGM Vienna, 1997), IAEA-TECDOC-1013, Vienna (1998).
[5] UNITED STATES NUCLEAR REGULATORY COMMISSION, Interim Staff Guidance (ISG-8), accessed at http://www.nrc.gov/reading-rm/doc-collections/isg/spent-fuel.html, (2002).
[6] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary (Terminology used in Nuclear, Radiation, Radioactive Waste and Transport Safety) IAEA-Working Material (2000).
http://www.iaea.org/ns/CoordiNet/safetypubs/iaeaglossary/glossaryhomepage.htm) [7] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport
of Radioactive Material, 1996 Edition (Revised), Safety Standards Series No. TS-R-1 (2000).
[8] UNITED STATES DEPARTMENT OF ENERGY, Surplus Plutonium Disposition Environmental Impact Statement (SPD EIS), Report, DOE/EIS-0283, (1999).
[9] UNITED STATES DEPARTMENT OF ENERGY, Technology Roadmap for Generation IV Nuclear Energy Systems, GIF-002-00 (2002).
http://www.ne.doe.gov/
[10] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidance for the Evaluation of Innovative Nuclear Reactors and Fuel, IAEA-TECDOC-1362, Vienna (2003).
http://www-pub.iaea.org/MTCD/Meetings/Announcements.asp?ConfID=108
ABBREVIATIONS
AR at-reactor
ARF advanced reactor fuel
ARIANE an experimental programme of radiochemistry analysis providing data on MOX and UO2 fuels irradiated in PWR and BWR conditions, managed by Belgonucleaire
BUC burnup credit
BUC-FP burnup credit – fission product level
BWR boiling water reactor
CERES international programme to validate cross section data and inventory predictions for actinides and fission products important for burnup credit, using the MINERVE reactor, CEA Cadarache.
EFR European fast reactor
GT-MHR gas turbine modular helium reactor
ICSBEP International Criticality Safety Benchmark Experiment Project
IE initial enrichment
JEF joint evaluated files
JEFF Joint Evaluated Data Library based on co-operation between JEF project and European Fusion File (EFF) project.
LWR light water reactor
MALIBU PIE programme supporting depletion calculations
MOX mixed oxide fuel
MSR molten salt reactor
OM optimal moderation
PA package array
PIE post irradiation examination
PROTEUS zero-power, fast/thermal mixed critical reactor facility in Switzerland
PUREX plutonium, uranium extraction (reprocessing) process
PWR pressurized water reactor
RBMK Russian graphite moderated reactor
REBUS reactivity tests for a direct evaluation of BurnUp credit on Selected irradiated LR fuel bundles
SFCOMPO LWR spent fuel isotopic composition database
SFDS spent fuel dry storage
SP single package
UOX uranium oxide
WG weapons grade
WWER Russian type of PWR