SRAC Code Research Articles

Measurement of Overall Temperature Coefficient of Reactivity of VHTRC-1 Core by Pulsed Neutron Method

The overall temperature coefficient of reactivity of VHTRC-1 core was measured at the Very High Temperature Reactor Critical Assembly (VHTRC) to examine the calculation accuracy of temperature dependent neutronic characteristics of the High Temperature Engineering Test Reactor (HTTR). The core of VHTRC consists of pin-in-block type fuel using coated particles of low-enriched UO2. The whole assembly was heated by electric heaters and the reactivity change due to the temperature rise up to 200°C was measured by the pulsed neutron method to determine the overall temperature coefficient of reactivity in isothermal condition. The experimental results indicated that the temperature coefficient was −1.71 × 10−4 Δk/k/°C on the average in the measured range from 25 to 200°C and the absolute value was smaller by 20% near room temperature than at the higher temperature. The calculation by the SRAC code system using the ENDF/B-IV nuclear data well predicted the experimental results.

Evaluation methods of resonance absorption for system with pellet surrounded by fuel solution.

KEYWORDS: uranium dioxidebenchmarkscriticalityfuel pelletsfuel solutionsresonance absorptionMonte Carlo methodcalculation methodsDancoff factor methodTone's methodPEACO methodcomparative evaluations

Measurements of Reaction Rates in Zone-Type Cores of Fast Critical Assembly Simulating High Conversion Light Water Reactor

Measurements of reaction rates have been performed in three uranium-fueled zone-type cores of the FCA constructed for a series of experiments on a high conversion light water reactor (HCLWR). These cores possess central test zones of different fuel enrichments and moderator to fuel volume ratios. Radial and axial fission rates of 236U, 239Pu, 238U and 23,Np were measured in each test zone by means of the micro-fission counter traverse. A region where the fundamental mode spectrum is established in the test zone were determined by utilizing these fission rate distributions. Central reaction rate ratios relative to the 235U fission rate were obtained from the measurements by the micro-fission counters and metallic uranium foils to examine changes in the reaction rate ratios among the three cores. The measured data were analyzed by the SRAC code system on the basis of the nuclear data file JENDL-2. The calculated fission rate distributions agree well with the experimental results for the all cases. The results of reaction rate ratios show that the calculations over- predict the experimental values of the 238U capture/235U fission and 238U fission/235U fission rate ratios in the three cores.

Computational Breakdown of the High Conversion Light Water Reactor Infinite Lattice Void Reactivity into Contributing Nuclides and Energy Groups

By cell calculation with the SRAC code system, void reactivity is evaluated for a high conversion light water reactor tight lattice, with an emphasis on the breakdown of the void effect into compon...

Measurements of reaction rates in zone-type cores of fast critical assembly simulating high conversion light water reactor.

Measurements of reaction rates have been performed in three uranium-fueled zone-type cores of the FCA constructed for a series of experiments on a high conversion light water reactor (HCLWR). These cores possess central test zones of different fuel enrichments and moderator to fuel volume ratios. Radial and axial fission rates of 236U, 239Pu, 238U and 23,Np were measured in each test zone by means of the micro-fission counter traverse. A region where the fundamental mode spectrum is established in the test zone were determined by utilizing these fission rate distributions. Central reaction rate ratios relative to the 235U fission rate were obtained from the measurements by the micro-fission counters and metallic uranium foils to examine changes in the reaction rate ratios among the three cores. The measured data were analyzed by the SRAC code system on the basis of the nuclear data file JENDL-2. The calculated fission rate distributions agree well with the experimental results for the all cases. The resul...

高温ガス炉臨界実験装置（ＶＨＴＲＣ）の初装荷炉心における臨界実験

A critical assembly VHTRC (Very High Temperature Reactor Critical Assembly) was constructed by modifying SHE (Semi-Homogeneous Experiment) in order to verify the nuclear design accuracy of the HTTR (High Temperature Engineering Test Reactor).The initial core named as VHTRC-1 core was made by loading fuel rods which contained fuel compacts of 4W/O enriched uranium coated particles. The following measurements were carried out in the VHTRC-1 core: (1) critical mass, (2) reactivity worths of mockup control rod and burnable poison rod, (3) neutron flux distribution and (4) kinetic parameter βBrr/Λ.Calculations were performed with the SRAC code system using the nuclear data based on the ENDF/B-IV. The agreement between experiment and calculation was fairly good for the most experimental items. The calculation, however, cannot predict accurately the value of the kinetic parameter.

Analyses of Reactor Physics Experiments in the Kyoto University Critical Assembly

Both experimental and analytical studies have been performed on the temperature coefficient of reactivity in a light water moderated and reflected core loaded with highly enriched uranium fuel at the Kyoto University Critical Assembly. The temperature effect on reactivity was measured for the 20 to 70°C range to investigate separately the effects of the H/235U atomic ratio and the core shape on this quantity. The results of both the eigenvalue and perturbation calculations by the SRAC code system approximately reproduced the experimental data. It was found that the contribution of the core region to the temperature coefficient was negative due to the degradation of moderation, whereas that of the reflector region was positive due to the decrease in neutron absorption. The positive contribution of the reflector region became larger as the H/235U atomic ratio became smaller and the core shape became more slender.

Study on Criticality of a Light Water Moderated and Reflected Coupled Core with Highly Enriched Uranium Fuel

A criticality safety study on a light water moderated and reflected coupled core loaded with highly enriched uranium fuel was performed in the Kyoto University Critical Assembly. The critical mass and neutron flux distribution were measured systematically as a function of the separation distance between the two cores, varying the H/235U atomic ratio (i.e., the moderator-to-fuel volume ratio). These data were analyzed with the SRAC code system to assess the capability of diffusion theory to analyze the coupled-core system. It was found that the critical mass of the coupled core showed the minimum when the two cores were separated by a certain distance depending on the neutron spectrum in the core region. The neutron flux peak value at the water gap region reached the maximum when the separation distance was 5 to 6 cm. The results calculated with the diffusion code installed in the SRAC system agreed well with the experimental data.

Temperature Coefficient of Reactivity in Light-Water Moderated and Reflected Cores Loaded with Highly-Enriched-Uranium Fuel

Both experimental and analytical studies have been performed on the temperature coefficient of reactivity in a light-water moderated and reflected core loaded with highly-enriched-uranium fuel, which was constructed in the Kyoto University Critical Assembly (KUCA). The temperature effect on reactivity was measured for the range of 20∼70°C to investigate separately the effects of (1) the fuel pitch (H/235U atomic ratio) and (2) the core shape on this physical quantity. The experimental data were analyzed with use of the SRAC code system. The calculated eigenvalue k eff agreed with the measured one within 0.5% in the C/E ratio for both the 2- and 3-dimensional diffusion calculations. The experimental data were approximately reproduced by both the eigenvalue and perturbation calculations. It was found that the contribution of the core region was negative to the temperature coefficient of reactivity, whereas that of the reflector region was positive. The synthesis of these contributions made the temperature c...

Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO2 kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core. The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of 238U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.

Measurements of doppler effect of coated particle fuel rod in SHE-14 core using sample heating device.

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO2 kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core. The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of 238U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.

Temperature coefficient of reactivity in light-water moderated and reflected cores loaded with highly-enriched-uranium fuel.

Both experimental and analytical studies have been performed on the temperature coefficient of reactivity in a light-water moderated and reflected core loaded with highly-enriched-uranium fuel, which was constructed in the Kyoto University Critical Assembly (KUCA). The temperature effect on reactivity was measured for the range of 20∼70°C to investigate separately the effects of (1) the fuel pitch (H/235U atomic ratio) and (2) the core shape on this physical quantity. The experimental data were analyzed with use of the SRAC code system. The calculated eigenvalue k eff agreed with the measured one within 0.5% in the C/E ratio for both the 2- and 3-dimensional diffusion calculations. The experimental data were approximately reproduced by both the eigenvalue and perturbation calculations. It was found that the contribution of the core region was negative to the temperature coefficient of reactivity, whereas that of the reflector region was positive. The synthesis of these contributions made the temperature coefficient negative in total. The degradation of moderation was the main contributor in the core region, whereas the decrease in the neutron absorption in the reflector region. The positive contribution of the reflector region became larger as the atomic ratio became smaller and the core shape became more slender.

Analysis of reactivity coefficients of Chernobyl reactor by cell calculation.

A series of cell calculations for the Chernobyl reactor was performed using the SRAC code system to provide its fundamental neutronic characteristics for the accident analysis at JAERI. The calculations are based on a two-step cell modelling. The primary cell is supposed on a unit square graphite block of 25cm × 25cm which contains a fuel assembly or a control channel. The secondary cell IS supposed on a unit of 16 channels where 14 fuel and 2 control channels are located so as to simulate the whole core. Detailed Investigation on fractional change of reaction rates for each nuclide along with increase of void fraction was carried out. The analysis clarified the mechanism to induce the positive void coefficient, together with its burn-up dependence and the increase due to withdrawal of control rod. A comparison of the effect of void fraction on the reaction rates in the primary cell between a Monte Carlo code VIM and SRAC shows consistent results within the statistical error. The calculated results for th...