High Conversion Light Water Reactor Research Articles

Toward Reanalysis of the Tight-Pitch HCLWR-PROTEUS Phase II Experiments

The HCLWR-Proteus Phase II experiments were conducted from 1985 to 1990 in the zero-power reactor Proteus at PSI in Switzerland. The experimental program was dedicated to the physics of high conversion light water reactors and in particular to the measurement of reactor parameters such as reaction rate traverses, spectral indices, absorber reactivity worths and void coefficients. The HCLWR experiments are especially interesting because they generated knowledge in the epithermal range of the neutron flux spectrum, for which little integral experimental data is available. In an effort to assess the interest of this experimental data to validate modern nuclear data and improve their uncertainties, a preliminary re-analysis of selected configurations was conducted with Monte-Carlo codes (MCNP6/SERPENT2) and modern nuclear data libraries (ENDF/B-VII.0, JEFF-3.1.1 and JENDL-4.0). The spectral indices, flux spectra and sensitivity coefficients on k(infinity) were calculated using cell models representative of the tight-pitch measurement configurations containing 11% PuO2-UO2 fuel rods in different moderation conditions (air, water and dowtherm). Spectral index predictions using the three nuclear data libraries agreed within two standard deviations with the measured values. The only exception is the Pu-242-captureto- Pu-239-fission ratio, which was overestimated with all libraries by more than four standard deviations, i.e. 13%, in the non-moderated configuration. In this configuration, Pu-242 captures are few since the flux spectrum in the Pu-242 capture resonance region (between 1eV and 1keV) is small making this spectral index hard to measure. Sensitivity coefficient predictions with both MCNP6 and SERPENT2 were in good agreement.

Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

Neutron imaging of annular flows in a tight lattice fuel bundle model

High conversion light water reactors (HCLWRs), featuring tight lattice fuel bundles, could fulfill the transition between present-day and future, fast reactor based technology allowing an improved fuel utilization. Nevertheless, the cooling of the tight lattice and the thermal-hydraulic feasibility should be thoroughly examined. We have studied annular flows and the influence of functional spacer on them in a tight lattice fuel bundle model using cold-neutron imaging, a non-intrusive technique. The investigation focused on adiabatic, air–water annular flows in a scaled-up model of four neighboring subchannels of a tight lattice fuel bundle model. Liquid film thickness distributions have been measured for different flow conditions with and without spacer to quantify the influence of the latter. The film thickness is determined with a minimal bias and a reasonable statistical uncertainty (7–10%) at a high spatial resolution (∼60μm). The entrained liquid hold-up fraction and the effect of the spacer on it is given as well.

Fluctuations of Mass Yield Distribution for the Epithermal Fission of 235U and Its Impact on HCLWR Calculations

The need for accurate nuclear data represents a permanent challenge to improve the calculation tools used in reactor physics. Uranium-235 is one of the most important isotopes, and its related nuclear data need to be known with a high degree of accuracy. In this context, many studies have been undertaken to improve the fission yields of 235U for the main fission products in spent fuels.For a few years, an increasing interest has been observed for high-conversion light water reactors, for better use of the fuel and for nonproliferation considerations. These concepts are based on a low moderation ratio (˜0.9) and the use of highly enriched mixed oxide fuels (>8%). Because of a neutron flux much harder than in pressurized water reactors, calculations require a good knowledge of nuclear data in the epithermal range to accurately predict the fuel depletion with burnup. In particular, the energy dependence of the fission yields must be considered, because of the existence of fluctuations of the fragment fission yields in the resonances of 235U(n,f). Unfortunately, the fluctuations are incorrectly taken into account in calculation codes because of the rough energy description in the usual nuclear data libraries.In this paper, integral experiments are presented for the identification and the quantification of this effect in an almost 1/E neutron spectrum. The experiments consist of the irradiation of 235U samples in the MINERVE reactor (CEA Cadarache), first in a mostly thermal neutron spectrum and second in a mostly epithermal one. Measurements of some abundant fission products are realized with gamma-ray spectrometry and show fluctuations of the 235U cumulative fission yields between the two experiments. Results are interpreted with the multimodal random neck rupture model as a variation of 5% of the weight ratio w1/w2 between the standard I and standard II fission modes, and compared with the differential experiments.

Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups – “equivalence in dilution” and “subgroup approach”. Under the “equivalence in dilution” approach we have analysed the generalized Stamm’ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant–Jeanpierre normalization. Under the “subgroup approach”, we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models.

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.

Analysis of Differences in Void Coefficient Predictions for Mixed-Oxide—Fueled Tight-Pitch Light Water Reactor Cells

Analysis of the benchmark problems on the void coefficient of mixed-oxide (MOX)-fueled tight-pitch cells has been performed using the Japanese SRAC code system with the JENDL-3.2 library and the French APOLLO-2 code with the CEA93 library based on JEF-2.2. The benchmark problems have been specified to investigate the physical phenomena occurring during the progressive voidage of MOX-fueled tight-pitch lattices, such as high conversion light water reactor lattices, and to evaluate the impact of nuclear data and calculational methods. Despite the most recently compiled nuclear data libraries and the sophisticated calculation schemes employed in both code systems, the k∞ and void reactivity values obtained by the two code systems show considerable discrepancy especially in the highly voided state. The discrepancy of k∞ values shows an obvious dependence on void fraction and also has been shown to be sensitive to the isotopic composition of plutonium. The observed discrepancies are analyzed by being decomposed into contributing isotopes and reactions and have been shown to be caused by a complicated balance of both negative and positive components, which are mainly attributable to differences in a limited number of isotopes including 239Pu, 241Pu, 16O, and stainless steel.

China wants more power for the people

With the aim of obtaining a fundamental data base with regard to HCLWR (high conversion light water reactor) core thermo-hydraulics, the effect of channel gaps on thermo-hydraulic behavior is investigated, particularly at low velocities as these may be assumed to occur under various abnormal operating conditions in HCLWRs. Boiling heat transfer and frictional loss are investigated experimentally, and the applicability of the drift flux model to narrow channels is examined. Finally, the reflooding behavior is investigated experimentally in relation to a postulated loss-of-coolant accident in HCLWRs. It has been made clear from the results that the incidence of a pressure drop during the reflooding phase is much higher than that in current LWRs.

Neutron Spectrum Calculation Code HARPY for Hexagonal Assembly

Abstract Hexagonal assemblies are used in a high conversion light water reactor (HCLVVR) and a nuclear reactor for cancer therapy. A cross section generation code for hexagonal assemblies is needed to design these cores precisely. For this purpose, a two-dimensional spectrum calculation code HARPY which can treat a hexagonal assembly has been developed. The HARPY code consists of a nodal code for pin-cell coupling in hexagonal geometry and a hexagonal mesh SN transport code and calculates detailed neutron spectra of fuel cells in the assembly. Verification calculations have been performed on a small hexagonal assembly by using SRAC and a Monte Carlo code GMVP to check the calculational accuracy of HARPY. The comparisons of k ∞'s, rod worths and pin power distributions have shown good agreement. This shows the adequacy of HARPY calculation.

Thermo-hydraulics of boiling two-phase flow in high conversion light water reactors (thermo-hydraulics at low velocities)

With the aim of obtaining a fundamental data base with regard to HCLWR (high conversion light water reactor) core thermo-hydraulics, the effect of channel gaps on thermo-hydraulic behavior is investigated, particularly at low velocities as these may be assumed to occur under various abnormal operating conditions in HCLWRs. Boiling heat transfer and frictional loss are investigated experimentally, and the applicability of the drift flux model to narrow channels is examined. Finally, the reflooding behavior is investigated experimentally in relation to a postulated loss-of-coolant accident in HCLWRs. It has been made clear from the results that the incidence of a pressure drop during the reflooding phase is much higher than that in current LWRs.

Neutron Spectrum Calculation Code HARPY for Hexagonal Assembly.

Abstract Hexagonal assemblies are used in a high conversion light water reactor (HCLVVR) and a nuclear reactor for cancer therapy. A cross section generation code for hexagonal assemblies is needed to design these cores precisely. For this purpose, a two-dimensional spectrum calculation code HARPY which can treat a hexagonal assembly has been developed. The HARPY code consists of a nodal code for pin-cell coupling in hexagonal geometry and a hexagonal mesh SN transport code and calculates detailed neutron spectra of fuel cells in the assembly. Verification calculations have been performed on a small hexagonal assembly by using SRAC and a Monte Carlo code GMVP to check the calculational accuracy of HARPY. The comparisons of k ∞'s, rod worths and pin power distributions have shown good agreement. This shows the adequacy of HARPY calculation.

Development of Transport Code for Hexagonal Geometry

An improved coarse-mesh discrete ordinates method has been developed for three-dimensional hexagonal transport calculations of high-conversion light water reactors and fast reactors. This method employs a new weighted diamond difference approximation which is obtained by using the neutron balance equations in divided submeshes. The weight is a function of neutron direction and scaler flux, and this method can be easily incorporated into conventional discrete ordinates transport codes. The present method was applied to hexagonal fuel assembly calculations of high-conversion reactor and fast reactor core calculations, and the results were compared with those of Monte- Carlo calculations. The values of kefi and power distributions agreed with each other within 0.5 and 3%, respectively, verifying accuracy of the present improved coarse-mesh discrete ordinates transport calculation method.

Development of Transport Code for Hexagonal Geometry. Improved Coarse-Mesh SN Calculation Code HEXTR.

An improved coarse-mesh discrete ordinates method has been developed for three-dimensional hexagonal transport calculations of high-conversion light water reactors and fast reactors. This method employs a new weighted diamond difference approximation which is obtained by using the neutron balance equations in divided submeshes. The weight is a function of neutron direction and scaler flux, and this method can be easily incorporated into conventional discrete ordinates transport codes. The present method was applied to hexagonal fuel assembly calculations of high-conversion reactor and fast reactor core calculations, and the results were compared with those of Monte- Carlo calculations. The values of kefi and power distributions agreed with each other within 0.5 and 3%, respectively, verifying accuracy of the present improved coarse-mesh discrete ordinates transport calculation method.

Thermal-hydraulic study on a double-flat-core high-conversion light water reactor

Research and developmental work on the high-conversion light water reactor (HCLWR) at JAERI aims at attractive design in terms of energy source and economy to ensure a safety level equal to or higher than that of current LWRs. In thermal hydraulics, efforts are being devoted to understand the phenomena in a tight lattice core and to develop analytical methods for optimum design. In critical heat flux (CHF) experiments, several existing CHF correlations were evaluated. It was found that the KfK correlation can be applied to the triangular tight lattice rod array used for the experiment reported here. In the hydraulic experiment the core pressure drop was investigated with existing correlations in various types of core geometries. It was found that Rehme's correlation for spiral wire and integral spiral rib can be applied and that the drag coefficient of grid spacers is noticeably affected by the height of spacers. On the basis of these experiments, the thermal-hydraulic characteristic was investigated for the double-flat-core HCLWR, which potentially has advantages such as negative void reactivity coefficient, high conversion ratio, and high burnup. It was found that the double-flat-core HCLWR is highly feasible, having enough margin in CHF, a core pressure drop comparable to current PWRs, and a fairly low peak clad temperature in LOCA reflood as calculated by REFLA code.

Evaluation of DNBR under operational and accident conditions for double-flat-core type HCLWR.

A double-flat-core type high conversion light water reactor (HCLWR) has been developed at JAERI to improve fuel utilization. Experimental and analytical studies on the departure from nucleate boiling ratio (DNBR) under operational and accident conditions for the HCLWR have been performed. It was found by comparing several critical heat flux (CHF) correlations with the CHF data obtained at JAERI and Bettis Atomic Power Laboratory that the KfK correlation has the most promising features for the application to the triangular tight lattice rod array. The minimum allowable DNBR (MDNBR) for the HCLWR was determined to be 1.28 by comparing the Bettis CHF data with the KfK correlation. The best-estimate code J-TRAC was used for system transient calculations under the primary coolant pump trip and locked rotor accident conditions. The subchannel code COBRA-IV-I was then used to obtain local flow conditions and fuel rod surface heat flux. The analytical results indicated that an enough safety margin was assured und...

Experimental Study on Reactivity Worth for Absorber Material in High Conversion Light Water Reactor Using FCA-HCLWR Core Fueled with Enriched Uranium

Experimental Study on Reactivity Worth for Absorber Material in High Conversion Light Water Reactor Using FCA-HCLWR Core Fueled with Enriched Uranium

Accuracy of Cell Calculation Methods Used for Analysis of High Conversion Light Water Reactor Lattice

Validation tests were made for the accuracy of cell calculation methods used in analyses of tight lattices of a mixed-oxide (MOX) fuel core in a high conversion light water reactor (HCLWR). A series of cell calculations was carried out for the lattices referred from an international HCLWR benchmark comparison, with emphasis placed on the resonance calculation methods; the NR, IR approximations, the collision probability method with ultra-fine energy group. Verification was also performed for the geometrical modelling; a hexagonal/cylindrical cell, and the boundary condition; mirror/white reflection. In the calculations, important reactor physics parameters, such as the neutron multiplication factor, the conversion ratio and the void coefficient, were evaluated using the above methods for various HCLWR lattices with different moderator to fuel volume ratios, fuel materials and fissile plutonium enrichments. The calculated results were compared with each other, and the accuracy and applicability of each met...

高転換型軽水炉研究における転換比の用語および定義

Conversion ratio is an important index as a term to indicate fuel source utilization. It was used to indicate breeding ratio in FBR and its definition was discussed in the paper. Studies for High Conversion Light Water Reactor have been started and the conversion ratio has been used as an index of fuel utilization. Its definition is not unique and there are some confusions among the papers.In this data the author tries to review various definitions of conversion ratio and clarify their physical meaning and its relation among the definitions.

Accuracy of cell calculation methods used for analysis of high conversion light water reactor lattice.

Accuracy of cell calculation methods used for analysis of high conversion light water reactor lattice.

Experimental study on reactivity worth for absorber material in high conversion light water reactor using FCA-HCLWR core fueled with enriched uranium.

Experimental study on reactivity worth for absorber material in HCLWR core has been carried out in a series of experiments using the Fast Critical Assembly (FCA) in Japan Atomic Energy Research Institute (JAERI). The central reactivity worth as well as the simulated control rod worth of B4C with different 10B content and of Hf was measured in FCA-HCLWR core fueled with enriched uranium. Both reactivity worths of B4C increase with 10B content. These increasing trends do not saturate to 90% enriched B4C. The Hf has the smaller reactivity worth than the 20% B4C. The experimental values are compared with the calculated ones which obtained from JENDL-2 data and the SRAC system. The calculation predicts well the dependence of reactivity worth on 10B content and underestimates the reactivity worth ratios of the Hf to the 20% B4C.