Lead Test Assemblies Research Articles

Study on fuel management strategy for accident-tolerant fuel on in-pile irradiation test

Within the context of the long-term development plan of batch loading of accident-tolerant fuel assemblies, an equivalent physical model of lead test assemblies is established based on the requirements of radiation testing and the applicability of analysis software. The quantitative assessment has been conducted on the impacts of parameters including zirconium alloy cladding coating and pellet additives on reactivity of fuel assembly. Concurrently, the fuel management strategy is optimized in alignment with the irradiation test objectives of ATF assemblies, leading to the creation of an evaluation process tailored for ATF assembly loading pattern. Focusing on the irradiation of the lead test assemblies into a nuclear power plant in China, three-cycle nuclear reactor loading pattern are designed, which are taking into account the economic, safety, and irradiation testing goals of the nuclear power plant. Simultaneously, the neutronic characteristics are also evaluated, quantitatively. The results show that the established equivalent physical model can effectively represent the characteristics of the lead test assemblies. The loading patterns designed based on the Optimised fuel management strategy achieve the desired goals in terms of economy, safety and radiation test requirements. The designed loading pattern evaluation process and equivalent model analysis method in this work provide guidance for the subsequent engineering design and application of irradiation testing for lead test assemblies.

Implications of accident tolerant fuels on thermal-hydraulic research

Accident tolerant fuel deployment has been rapidly accelerating in the US with lead test rods and assemblies already inserted in commercial reactors. With the current deployment target of mid 2020s for near term concepts and long term concepts thereafter, the implication of these fuels on thermal-hydraulic (TH) research should be documented to support their development and value proposition. Without economic argument, once the US government ATF program ends in ~2021, the US utilities will not adopt the more expensive ATF in their already NRC licensed safe and robust LWRs. Moreover, only limited severe accident gain can be realized by adopting ATFs alone. Thus economic opportunities with ATF cladding lies within challenging the current conservative technical specifications and safety procedures and pairing with more cost-effective high density fuel forms. Opportunities exits to go higher that current Zircaloy peak clad temperature (PCT) of ~1200 °C and survive a departure from nuclear boiling (DNB) or dryout event for considerable longer time with coated clad or SiC fiber composite ATF concepts. Higher PCTs along with longer durations at high PCTs emphasizes more importance on radiation heat transfer and fuel collapse limits during severe accidents. The tolerance to DNB and dryout for longer periods motivates their replacement by a fuel performance based limit. This path enabled by ATFs requires accurate prediction of post-DNB or dryout heat transfer and rewet characteristics. TH research is also needed in exploring new failure modes especially for SiC cladding where more accurate local temperature, flow and quenching is required to assess its thermomechanical performance. At the same time, adoption of ATF cladding could bring about less emphasis on GTRF or CRUD deposition that is being tackled by the TH community.

Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data

Abstract After a preparation period with VERONA upgrade and lead test assembly program, a new fuel type was introduced at MVM Paks NPP Ltd. This 4.7% average uranium enriched assembly type, together with the former 4.2% uranium enriched fuels, allowed us to lengthen the operating cycles to 15 months. Both fuel types contain gadolinium burnable poison, in six and three pins respectively. All of the four units have been converted to the C15 cycles, and have been being operated without any problems in the last few years. In this paper the test results of core design code HELIOS/C-PORCA, which is the basic model of VERONA, are outlined. C15 cycles were entirely investigated with the comprehensive study of measured and predicted (calculated) values of different reactor states. At first step, in order to prove the capability of the reactivity calculation of the nodal diffusion model, critical boric acid concentrations of different burnup and start up states were calculated and compared with measured values. During the next step of the verification process local in-core parameters were investigated. Measured neutron flux distributions (SPND signals) and coolant outlet temperatures were examined. SPND and thermo couple (TC) signals were predicted as a part of the monitoring system. The results of statistical investigations (average differences and standard deviations) for the applied fuel types are also presented.

Distribution characteristics of fission gas along the axial direction for an irradiated fuel rod of a pressurized water reactor (PWR)

In this work, the fission gas distribution of an irradiated oxide fuel was investigated by measuring the fission gas release (FGR) and retained gas from and in an oxide fuel along the axial height of the fuel rod. In addition to the fission gas measurements, other destructive and non-destructive post irradiation examination tests such as gamma scanning, eddy current testing, density and oxide layer thickness measurements, and ceramography/metallography were conducted to review their impact on the FGR and gas retention. A lead rod was extracted from a lead test assembly, which was irradiated to 56.9 GWd/tU during three irradiation cycles for a total exposure of 1,406 effective full power days. Considering the water-side oxide layer thickness and local burnup of the fuel rod, the rod was sectioned into four positions along the axial height and four samples were prepared from these positions. A sample fragment of around 0.1–0.2 g was individually melted to measure its retained krypton and xenon concentration. According to the measurement results, the retained krypton and xenon concentration ranges were 0.114–0.139 cc/gU and 1.073–1.338 cc/gU, respectively, and their retention percentage after normalization of the sample local burnup showed a decreasing trend as the axial height of the fuel rod raised. The water-side oxide layer thickness scope of the tested fuels measured by an observation of the optical microscope images was 13–42 μm, and the thickness was increased along the axial height of the rod. The amount of released krypton and xenon into the rod-free volume was measured as 5.7 and 54.6 cc, respectively, by a rod puncturing/collection procedure, which corresponds to a 1.94 % fractional fission gas release referring to fission gas generation by a code calculation.

A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10–15 years and would take 3–5 years to construct.

Nuclear fuel performance: Trends, remedies and challenges

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. ‘Fuel reliability’ has a much broader definition than just maintaining mechanical integrity and being leaker free – fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

In-Reactor Performance of an Advanced PWR Fuel, PLUS7, for OPR1000s in Korea

An advanced Pressurized Water Reactor (PWR) Fuel, PLUS7, for Optimized Power Reactors (OPR1000s) in Korea has been developed to achieve peak rod average burnup of 72GWD/MTU, overpower margin increase of 10% with respect to a reference fuel used in the OPR1000s, fuel integrity maintenance even with 0.3 g of seismic load, zero defect against debris, and grid-to-rod fretting wear. All the mechanical and thermal hydraulic performances of PLUS7 had been verified through a wide spectrum of out-of-pile verification tests, whereas the burnup-dependent performances have been verified through in-reactor verification tests of four PLUS7 Lead Test Assemblies (LTAs) in one of the OPR1000s. During each refueling outage period, burnup-dependent parameters were measured, which include fuel assembly, fuel rod and spacer grid dimensional changes, fuel assembly bow and twist, fuel rod bow, and fuel rod oxide layer thickness. Based on the measured data, it is said that the PLUS7 fuel will maintain integrity at least up to a peak rod average burnup of 58 GWD/MTU. In addition, from the coolant activity analyses and ultrasonic tests, the PLUS7 fuel was found to be free from any fuel damage including debris-induced and grid-to-rod fretting-wear-induced fuel failures. This successful in-reactor operation of the PLUS7 LTAs has made the Korea Nuclear Fuel (KNF) initiate the supply of PLUS7 fuel on a region basis for all OPR1000s from the year 2006.

Plutonium oxide polishing for MOX fuel fabrication

Los Alamos National Laboratory (LANL) successfully polished 120 kg of plutonium from surplus nuclear weapons for the European Mixed Oxide (MOX) Lead Test Assembly (LTA), managed by Duke, COGEMA, and Stone & Webster (DCS). The purified oxide was fabricated into fuel pellets, which comprised the LTAs. The LTAs have been inserted into the Catawba (SC, USA) nuclear reactor, where they are presently being used to generate electricity. The material used in the MOX LTA represents the first plutonium oxide from LANL to be polished under the quality requirements set forth by the Nuclear Regulatory Commission (NRC). In addition, the MOX project has been hailed as the “largest single nonproliferation project in history” by US Ambassador Linton Brooks (23/9/04). The process flow for PuO 2 purification was based on aqueous recovery that included various unit operations (dissolution, ion exchange, oxalate precipitation, and calcination). Data from a variety of chemical and physical analyses demonstrate product quality and process consistency. Further process optimization experiments will be performed prior to polishing an additional 330 kg of PuO 2 for the Mixed Oxide Fuel Fabrication Facility (MFFF) at Savannah River Site. During LTA production in 2004, approximately 500 L of 7 M nitric acid was used for washing impurities from the plutonium in order to meet product purity specifications. The majority of this acid was sent to the LANL Radioactive Liquid Waste Treatment Facility (TA-50) as effluent. Optimization experiments during 2006 and 2007 will investigate the use of recycled HNO 3 for PuO 2 washing and ion exchange. If successful, aqueous polishing processes will re-use almost all of the nitric acid that would otherwise have been sent to TA-50. In addition, an overall reduction in the volume of nitric acid used during ion exchange will be tested. The use of new Inconel calcination boats will also be verified. Additional process optimization activities will be conducted to validate the use of 100% quaternized Reillex ion exchange resin. Qualification exercises will include a series of process runs, followed by chemical analyses to assess product oxide impurity levels. Once the optimized processes have been qualified, the production phase of the project will begin.

Plutonium Oxide Polishing for MOX Fuel Production

:Los Alamos National Laboratory (LANL) has been tasked to produce 120 kg of plutonium as highly purified PuO2 for the European Mixed Oxide (MOX) Lead Test Assembly managed by Duke, COGEMA, Stone & Webster (DCS). To meet stringent and challenging technical requirements for PuO2 production, the LANL aqueous polishing team recently established consistency in generating quality material from weapons-grade PuO2. Polishing was required to remove undesirable quantities of trace-metal impurities, particularly gallium, and to produce a material with appropriate powder characteristics, such as particle size and surface area. The process flow sheet for PuO2 polishing was based on aqueous purification and included the unit operations of dissolution, ion exchange, oxalate precipitation, and calcination.

Operating and test experience with EBR-II, the IFR prototype

The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core.

ASEA-ATOM results on pci experiments on fuel rods; a review

ASEA-ATOM results on PCI-behaviour of LWR fuel generated since 1970 are reviewed and discussed. It is shown that standard LWR fuel is sensitive to PCI failures and that minor design changes such as variation in cladding thickness, pellet density etc. have a negligible effect on this sensitivity. PCI failures have been observed without excessive release of fission products from the UO 2 pellets. Further results generated are that the PCI failure probability can be strongly influenced by certain operating restrictions. The ramp rate dependence clearly dominates over effects from minor design modifications. This fact is utilized in ASEA-ATOM constructed reactors, thus avoiding PCI failures at a very low penalty (0.4% loss in capacity factor). The current ASEA-ATOM development work in the PCI area is directed towards PCI immune designs. One promising remedy design is internally copper-plated cladding tubes. Lead test assemblies are in the 2nd cycle of irradiation in a power reactor.

Ambulatory Pneumothorax in Pulmonary Tuberculosis

Los Alamos National Laboratory (LANL) successfully polished 120 kg of plutonium from surplus nuclear weapons for the European Mixed Oxide (MOX) Lead Test Assembly (LTA), managed by Duke, COGEMA, and Stone & Webster (DCS). The purified oxide was fabricated into fuel pellets, which comprised the LTAs. The LTAs have been inserted into the Catawba (SC, USA) nuclear reactor, where they are presently being used to generate electricity. The material used in the MOX LTA represents the first plutonium oxide from LANL to be polished under the quality requirements set forth by the Nuclear Regulatory Commission (NRC). In addition, the MOX project has been hailed as the “largest single nonproliferation project in history” by US Ambassador Linton Brooks (23/9/04). The process flow for PuO2 purification was based on aqueous recovery that included various unit operations (dissolution, ion exchange, oxalate precipitation, and calcination). Data from a variety of chemical and physical analyses demonstrate product quality and process consistency. Further process optimization experiments will be performed prior to polishing an additional 330 kg of PuO2 for the Mixed Oxide Fuel Fabrication Facility (MFFF) at Savannah River Site. During LTA production in 2004, approximately 500 L of 7 M nitric acid was used for washing impurities from the plutonium in order to meet product purity specifications. The majority of this acid was sent to the LANL Radioactive Liquid Waste Treatment Facility (TA-50) as effluent. Optimization experiments during 2006 and 2007 will investigate the use of recycled HNO3 for PuO2 washing and ion exchange. If successful, aqueous polishing processes will re-use almost all of the nitric acid that would otherwise have been sent to TA-50. In addition, an overall reduction in the volume of nitric acid used during ion exchange will be tested. The use of new Inconel calcination boats will also be verified. Additional process optimization activities will be conducted to validate the use of 100% quaternized Reillex ion exchange resin. Qualification exercises will include a series of process runs, followed by chemical analyses to assess product oxide impurity levels. Once the optimized processes have been qualified, the production phase of the project will begin.