Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

Measured isotopic compositions of UO₂ and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO₂ fuel assemblies. The average and standard deviation of the biases (C/E ? 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO₂ fuel samples. The averaged biases for ²³⁵U, ²³⁶U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu were 2.7%, ?0.9%, 0.3%, 0.7%, ?2.4% and ?1.7% for PWR UO₂ samples, and 6.7%, ?1.5%, 2.5%, ?0.6%, 0.4% and ?0.1% for BWR UO₂ samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of ²³⁹Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.     

Fuel Temperature Coefficients of Reactivity for Heavy-Water-Moderated,Cluster-Type Fuel Lattices

Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO₂ fuel for three kinds of ²³⁵U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding land pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, ?3.48±0.13 and - 6.36±0.25 in the unit of l0^-5% Δk/k.°C for 0.2%, 0.7% and 1.5%²³⁵U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2x10^-5 Δk/k.°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) x 10^-5 Δk/k.°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement.     

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% ²³⁵U enriched UAl_x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% ²³⁵U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE.Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.     

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.     

Performance improvement analysis of boiling water reactors by incorporation of hydride fuel

The feasibility of improving the neutronic characteristics of boiling water reactors (BWR) by using U–Zr hydride fuel is studied. Several modified BWR fuel assembly designs are considered. These include designs in which hydride fuel rods replace water rods only, replace water rods and a fraction of the oxide fuel rods, replace oxide fuel in the upper half of all the fuel rods, and replace all the oxide fuel in the assembly. It is found that replacement of at least half of the oxide fuel rods in the fuel assembly by U–ZrH_1.6 fuel might simultaneously improve the performance of BWR in three ways: (a) Increasing the energy extracted per fuel assembly and the cycle length by up to 10%. (b) Reducing the uranium ore and SWU requirements by approximately 10%. (c) Reducing the negative void coefficient of reactivity by, at least, 50%. It is also found that replacement of all the oxide fuel by hydride fuel opens interesting new options for the design of BWR fuel assemblies. The net result might be simplified assembly designs that can generate significantly more energy while featuring small negative void coefficient of reactivity. U–ThH₂ fuel appears to be even more promising than U–ZrH_1.6. For the potential benefits from hydride fuel to be realized, a clad material that is not permeable to hydrogen and is not as neutron absorbing as stainless steel needs to be developed.     

Neutron-physical and thermohydraulic calculations of VVR-SM with high-and low-enrichment uranium fuel assemblies

The VVR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences of Uzbekistan is being converted from fuel assemblies with high-enrichment uranium (36% ²³⁵U) to fuel assemblies with low-enrichment uranium (19.7% ²³⁵U). During the conversion process consisting of nine cycles, the IRT-3M fuel assemblies with high-enrichment uranium, which are removed at the end of each cycle, will be replaced with IRT-4M fuel assemblies with low-enrichment uranium. This will require increasing the core size up to 20 fuel assemblies and increasing the power of the reactor to 11 MW. The methods used for and the results of neutron-physical calculations (burnup, power distribution, subcriticality), thermohydraulic analysis, and calculations of the kinetic parameters of a stable state are described for a core with high-enrichment uranium, a mixed core, and the first full core with low-enrichment uranium. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 269–273, May, 2008.     

Additional studies of criticality safety of failed used nuclear fuel

Commercial used nuclear fuel (UNF) in the USA is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high burnup values (>45 GWd t^?1) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF are not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on the criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories, and specific configurations were evaluated to understand trends and quantify the consequences of worst case potential reconfiguration progressions. These results are summarised here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g. >20% Δk_eff). However, for credible fuel failure configurations from ES or transportation following ES, the consequences are judged to be manageable (e.g. <5% Δk_eff). The current work expands on the previous efforts by including part length rods in fresh boiling water reactor fuel assemblies and studying the effect of damage in varying numbers of fuel assemblies.     

Modeling for Void Reactivity Effect by Subcooled Boiling under Reactivity Initiated Accident Condition of BWR

A simple mechanistic model is presented to evaluate the subcooled void reactivity effect under a Reactivity Initiated Accident (RIA) at cold critical condition of BWR. This model consists of a drift flux model for vapor velocity and a vapor mass conservation model with a term of vapor source on a heated wall, and it was incorporated into a homogeneous and equilibrium thermal-hydraulic code EUREKA-JINS. A sample analysis by this model showed that the subcooled void reactivity effect leads to reduction of the maximum fuel enthalpy by about 20 cal/g UO₂ in the case of RIA at cold critical condition. Though the reduced value is dependent on the reactor core condition, this result indicates the significance of subcooled void reactivity effect in the accident, while the effect can be neglected in the hot stand-by case where, at most, only 4 cal/g UO₂ is reduced for the maximum fuel enthalpy.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Applying burnable poison particles to reduce the reactivity swing in high temperature reactors with batch-wise fuel loading

Burnup calculations have been performed on a standard HTR fuel pebble with a radius of 3 cm containing 9 g of 8% enriched uranium and burnable poison particles (BPP) made of B₄C highly enriched in ¹⁰B. The radius of the BPP and the number of particles per fuel pebble have been varied to find the flattest reactivity-to-time curve. It was found that for a k∞ of 1.1, a reactivity swing as low as 2% can be obtained when each fuel pebble contains about 1070 BPP with a radius of 75 μm. For coated BPP that consist of a graphite kernel with a radius of 300 μm covered with a B₄C burnable poison layer, a similar value for the reactivity swing can be obtained. Cylindrical particles seem to perform worse. In general, the modification of the geometry of BPP is an effective means to tailor the reactivity curve of HTRs.     

Simulating control rod and fuel assembly motion using moving meshes

A prerequisite for designing a transient simulation experiment which includes the motion of control and fuel assemblies is the careful verification of a steady state model which computes k_eff versus assembly insertion distance. Previous studies in nuclear engineering have usually approached the problem of the motion of control rods with the use of nonlinear nodal models. Nodal methods employ special approximations for the leading and trailing cells of the moving assemblies to avoid the rod cusping problem which results from the naive volume weighted cell cross-section approximation. A prototype framework called the MOOSE has been developed for modeling moving components in the presence of diffusion phenomena. A linear finite difference model is constructed, solutions for which are computed by SLEPc, a high performance parallel eigenvalue solver. Design techniques for the implementation of a patched non-conformal mesh which links groups of sub-meshes that can move relative to one another are presented. The generation of matrices which represent moving meshes which conserve neutron current at their boundaries, and the performance of the framework when applied to model reactivity insertion experiments is also discussed.     

基于SCIENCE程序包的IFBA组件模型的可行性研究

表面涂有一薄层硼化锆的一体化燃料可燃吸收体(IFBA)被用作轻水堆UO2燃料组件的反应性控制。法国AREVA公司开发的SCIENCE程序包具有模拟IFBA组件的能力,但其模拟精度需经标定。本文利用APOLLO2-F程序建立IFBA组件模型和不含IFBA组件模型,研究了组件的无限增殖因数k∞及IFBA价值,并与西屋公司结果进行比较。分析了燃料和包壳温度的处理方法以及数据库的差异对结果的影响。利用硼化锆密度修正因子评估IFBA价值偏差对堆芯参数和功率分布等的影响。结果表明:SCIENCE计算的k∞及IFBA价值与西屋公司的结果符合较好,低燃耗区SCIENCE计算的价值偏小2%。装载8个104根IFBA棒组件的堆芯,组件相对功率最大偏差约为1%;硼浓度、功率峰因子FQ和焓升因子FΔH的变化均不到0.1%,可忽略。先导组件采用28根或更少的IFBA棒时,可直接采用SCIENCE程序进行计算。     

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg^–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium–thorium fuels

We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in ²³⁵U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides ²⁴⁰Pu, ²³⁸U and ²³²Th and it requires the esteem of the Dancoff–Ginsburg factor for a pebble bed or prismatic core. The Dancoff–Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057–5692, 6.674–14485, 21.78–3472 eV for ²⁴⁰Pu, ²³⁸U and ²³²Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.     

Enhancing BWR proliferation resistance fuel with minor actinides

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the ²³⁸Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (²³⁷Np and ²⁴¹Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of ²³⁸Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO₂ fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm³) to the top (0.35 g/cm³) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (²³⁷Np and/or ²⁴¹Am), significantly increases the ²³⁸Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.     

Transient hydride fuel behavior in LWRs

The transient response of a uranium zirconium hydride (U_0.31ZrH_1.6) fuel element to conditions typical of a light water reactor reactivity insertion accident has been studied. Hydrogen diffusion within the fuel is treated as a function of the time-dependent temperature profile. Temperature and hydrogen concentration dependence of thermal properties of the fuel is also considered. The set of coupled equations describing the transient conduction of heat and diffusion of hydrogen are solved with a Crank-Nicolson discretization scheme for heat diffusion and an explicit Euler scheme for hydrogen diffusion.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.