Experimental validation of control rod related perturbations of moderator regions in an SCWR-like fuel lattice

Two different types of perturbations of an SCWR-like fuel lattice have been investigated experimentally in the central test zone of the PROTEUS zero-power research reactor at the Paul Scherrer Institute in Switzerland. In each case, a campaign of high-resolution gamma-ray spectroscopy measurements was carried out on 34 fuel pins of the test lattice. In the first case, the test lattice was perturbed by inserting aluminum rods into the four central moderator regions, while in the second case, the perturbation was affected using steel absorber rods (instead of aluminum). The derived reaction rates are the capture rate in ²³⁸U (C₈) and the total fission rate (F_tot), as also the reaction rate ratio C₈/F_tot. Each of these has been mapped on the lattice and compared to calculated results from whole-reactor Monte Carlo simulations with MCNPX. Excellent agreement has been obtained, for both perturbed lattices, between the calculated and experimental distributions of C₈, F_tot and C₈/F_tot. Considering that control rods in an SCWR assembly are foreseen to be inserted into the central moderator regions, these results may be considered as generic validation of Monte Carlo simulations for the two different types of lattice perturbations which inserted control rods imply, viz. moderator displacement and strong neutron absorption.In a second step, calculated C₈, F_tot and C₈/F_tot distributions for the two perturbed lattices (as well as for the unperturbed lattice) have been compared, at assembly level, between MCNPX and the deterministic LWR lattice code CASMO-4E. In the case of the unperturbed lattice, as well as for the lattice with steel rods, the agreement between the codes is found to be within ～1% for all pins and each reaction rate. However, for the lattice with aluminum rods, i.e. the case with mainly just moderator displacement involved, CASMO overestimates the reaction rates in the vicinity of the perturbations by up to 2–3%, when employing the standard input options. The reason for this discrepancy has been found to be the leakage treatment, which uses the fundamental-mode buckling applied in a homogenized sense across the lattice. In this way, global leakage gradients get averaged out over the entire assembly. The optional input card BZ2 for CASMO resolves this problem, and the codes then agree within 1% even for the aluminum case.     

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades in boiling water reactors (BWRs) is important for safety assessment and for achieving flexible operation during the cycle. Characteristics that are affected significantly include distributions of the total fission (F_tot) and ²³⁸Ucapture (C₈) rates for controlled regions. Representative experimental investigations have been performed in the framework of the LWR-PROTEUS programme. In particular, the LWRPROTEUS I-2A experiments concerned the neutronics characterisation of a SVEA-96+ BWR assembly controlled with a hafnium (Hf) blade under full-density water moderation conditions. The current paper presents the comparisons of the measured F_tot and C₈ pin-wise distributions with a variety of stochastic and deterministic calculations: (a) MCNPX2.5 using recent nuclear data libraries (JEFF-3.1, ENDF/BVI. 8, and JENDL-3.3), (b) PHOENIX4 using ENDF/B-VI.3, (c) BOXER using JEF-1, (d) CASMO₄ using JEF-2.2, and (e) HELIOS1.6 using ENDF/B-VI.1. The calculation/experiment comparisons show standard deviations from 1.2% (MCNPX2.5) up to 1.9% (BOXER) for the prediction of the F_tot distribution, the highest individual discrepancy (7.6% with BOXER) being seen close to the “Hf-vertex.” The C₈ comparisons show systematically better agreement than those of F_tot, the lowest standard deviations being 1.0% (BOXER) and the highest only 1.4% (HELIOS). In addition, sensitivity studies highlight the greater importance of modelling aspects, compared with that of nuclear data libraries, for the achievement of satisfactory and validated F_tot and C₈ predictions.     

Generalized Bias Factor Method for Accurate Prediction of Neutronics Characteristics

A generalized bias factor method is proposed to improve the prediction accuracy of neutronics characteristics of a target core. The generalized bias factor method uses conventional bias factors calculated for several critical assemblies. The weighting factors for individual bias factors are determined to minimize the variance of neutronic characteristics of the target core. Numerical calculations are performed to investigate the uncertainty reductions of neutronics characteristics for a tight-lattice core. Though the uncertainty is not remarkably reduced for k_eff , that for the reaction rate ratio of ²³⁸U capture/²³⁹Pu fission is remarkably reduced: For example, the uncertainty reduction of the reaction rate ratio in the upper core is 0.871 for the present method, and 0.657 for the conventional bias factor method.     

Determination of the 238U capture to total fission ratio in alternate depleted uranium/polyethylene shells with D-T neutrons

Aiming at checking the conceptual design of the subcritical blanket in the fusion–fission hybrid reactor, an integral experiment was carried out on an alternate depleted uranium/polyethylene-shell setup with D-T neutrons using activation technique. 18 depleted uranium foils were placed at 90° direction to the incident D beam, and the distribution of the ²³⁸U capture to total fission ratio was determined by measuring the 277.6 keV γ ray generated by neutron capture of ²³⁸U and the 293.3 keV γ ray generated by fission of ²³⁵U and ²³⁸U. The ratios were generally between 1 and 2 in the depleted uranium shells, with relative uncertainties between 3.0% and 5.5%. The ratios were calculated by the MCNP4B code employing ENDF/B-VI nuclear data library, the discrepancies between calculations and experiments were all within 6%, and the average calculation to experiment(C/E) ratio was 0.998.     

Comparisons of Deterministic Neutronic Calculations with Monte Carlo Results for an Advanced BWR Fuel Assembly with Hafnium Control Blades

From the neutronic viewpoint, the optimization of BWR core designs is strongly related to the accurate determination of flux variations inside and around fuel assemblies. These fluctuations, which are mainly due to the high heterogeneity of the fuel and moderator regions, as additionally to the presence of cruciform absorber blades, have a direct impact on reactor safety and performance. Of particular importance is the pin power distribution, leading to the need of assessing the capabilities of design tools in a sufficiently rigorous manner. The basic configuration chosen for the code comparisons corresponds to a SVEA-96 fuel assembly under full-density water moderation conditions, with inserted hafnium absorber blades. The calculational schemes employed are the Monte Carlo code MCNPX2.5, in conjunction with various nuclear data libraries (ENDF/B-VI, JEF2.2, JEFF3.0, JENDL-3.2 and JENDL3.3), and the deterministic codes CASMO4 with JEF2.2, BOXER with JEF1.0 and HELIOS 1.6 with ENDF/B-VI based libraries, respectively. The significant discrepancies observed in k _∞ predictions (>500pcm) are found to be mainly nuclear data related. On the other hand, data library effects have been found to be quite small for the prediction of pin-wise distributions of total fissions (F_tot), ²³⁸U captures (C8), as also of the C₈=F_tot ratio. Significant differences in these reaction rate distributions (up to several percent) have, however, been observed between the Monte Carlo and deterministic calculations, particularly in the vicinity of the hafnium blades and in the gadolinium pins.     

Measurement of thermal neutron and resonance integral cross sections of the reaction V(n,γ)V using a 20 Ci Am–Be isotopic neutron source

The thermal neutron capture cross section (σ_o) and the resonance integral (I_o) of the ⁵¹V(n,γ)⁵²V reaction were measured with an activation method to provide fundamental data for reactor calculation, activation analysis, and other theoretical and experimental uses concerning the interaction of neutron with matter. The vanadium and manganese samples were irradiated within and without a Cd shield case using a 20 Ci Am–Be neutron source. The activities of the samples were measured using gamma-ray spectroscopy. The thermal neutron capture cross section and the resonance integral were determined relative to the reference reaction ⁵⁵Mn(n,γ)⁵⁶Mn and the values obtained are 5.16 ± 0.19 barns and 2.53 ± 0.1 barns respectively. The previous measurements of the σ_o and I_o of the reaction ⁵¹V(n,γ)⁵²V were reviewed and the difference between the present values and the previous results were discussed.     

Reaction Rate Calculation in Fast Reactor Blanket Using Multiband Sn Theory

A new method was applied to calculation of reaction rates in blanket of LMFBR using the multiband S_n theory. This procedure leads to the use of direction dependent total micro cross sections, which advances the penetration of neutrons into the blanket.

Test calculation with RZ model of a prototype fast reactor shows that reaction rates tend to rise with the penetration into blanket compared to the conventional multigroup calculation: the maximum difference was about 3% for ²³⁸U capture, 4% for ²³⁵U fission, 4% for ²³⁹Pu fission, and less than 1% for ²³⁸U fission. This tendency shows the same direction as the difference observed between the continuous energy Monte Carlo method and the conventional method.     

Measured activation cross sections below 10 MeV for the 51V(n,p)51Ti and 51V(n, α)48Sc reactions

The activation method is used to measure cross sections for the ⁵¹V(n, p)⁵¹Ti reaction from E_n = 2.856 to 9.267 MeV and for the ⁵¹V(n, α)⁴⁸Sc reaction from 5.515 to 9.567 MeV. Both measurements utilize ENDF/B-V evaluated neutron-induced fission cross sections of ²³⁸U as a standard. The experimental results from this work are compared with corresponding ENDF/B-V evaluated cross sections for V and substantial differences are evident. The most significant difference is a tendency for the measured values to exceed evaluated ones by as much as 50% in the vicinity of 8 MeV.     

Measurement of Neutron Capture Cross Sections with Fe-Filtered Beam

The neutron capture cross sections of ⁹³Nb, ¹¹⁵In, ¹²⁷I, ¹⁶⁵Ho, ¹⁸¹Ta, ²³²Th and ²³⁸U were measured using the Fe-filtered beam. A 15-cm thick Fe filter was placed in the neutron beam produced by the KUR 46-MeV electron Linac and capture prays were detected by two C₆F₆ scintillation detectors located at an 11.7 m-flight path. The pulse-height weighting technique was used to determine the relative capture pray detection efficiency. The neutron flux was measured by the same detectors, whose detection efficiency for the 480-keV pray from the ¹⁰B(n, α₁) reaction was calibrated by the saturated resonance capture in Ag at 5.2-eV. Self-shielding and multiple scattering corrections were applied to the data. The results of 24-keV capture cross sections are 340, 770, 780, 1,280, 880, 520 and 520 mb for ⁹³Nb, ¹¹⁵In, ¹²⁷I, ¹⁶⁵Ho, ¹⁸¹Ta, ²³²Th and ²³⁸U, respectively. Total errors are 5 to 8%, with an estimated systematic error of 4%. The discrepancy between the present results and other data measured recently is within 10%.     

Characterisation of radial reaction rate distributions across the 92-pin section of a SVEA-96 Optima2 assembly

Characterisation of the SVEA-96 Optima2 boiling water reactor assembly, in terms of the radial distributions of normalised total fission and ²³⁸U capture rates, is reported at its central elevation, i.e. at the 92-pin section, where the one-third part-length pins are replaced by water. Measurements performed in the PROTEUS facility are compared with MCNPX predictions. The calculation model included the measured locations of the SVEA-96 Optima2 assemblies and sub-assemblies, within the PROTEUS test zone. Predicted and experimental fission and ²³⁸U capture rates are found to agree, respectively, within 3.5% and 4% for all pins. Fission rates in the burnable-absorber UO₂–Gd₂O₃ fuel pins have been predicted without bias using the ENDF/B-VI data library but show an average 1.4% under-prediction with the JEFF-3.1 data library. A slight overestimation of the total fission rate in the pins located at the periphery of the assemblies was observed and has been attributed to an inaccurate modelling of the pin positions. However, there was no systematic bias observed due to the absence of the one-third pins at the corners of the assembly.     

Measurement of thermal neutron cross-section and resonance integral for the Ho(n,γ)Ho reaction using electron linac-based neutron source

The thermal neutron cross-section and the resonance integral of the ¹⁶⁵Ho(n,γ)^166gHo reaction have been measured by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th), the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. The thermal neutron cross-section for the ¹⁶⁵Ho(n,γ)^166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ¹⁶⁵Ho(n,γ)^166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.     

Permeability of hydrogen and deuterium of Hastelloy XR

Permeation of hydrogen isotope through a high-temperature alloy used as heat exchanger and steam reformer pipes is an important problem in the hydrogen production system connected to be a high-temperature engineering test reactor (HTTR). An experiment of hydrogen (H₂) and deuterium (D₂) permeation was performed to obtain permeability of H₂ and D₂ of Hastelloy XR, which is adopted as heat transfer pipe of an intermediate heat exchanger of the HTTR. Permeability of H₂ and D₂ of Hastelloy XR were obtained as follows. The activation energy E₀ and pre-exponential factor F₀ of the permeability of H₂ were E₀=67.2±1.2 kJ mol⁻¹ and F₀=(1.0±0.2)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5, respectively, in the pipe temperature ranging from 843 K (570 °C) to 1093 K (820 °C). E₀ and F₀ of the permeability of D₂ were respectively E₀=76.6±0.5 kJ mol⁻¹ and F₀=(2.5±0.3)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5 in the pipe temperature ranging from 943 K (670 °C) to 1093 K (820 °C).     

Measured activation cross sections below 10 MeV for the V(n, p)Ti and V(n, α)Sc reactions

The activation method is used to measure cross sections for the ⁵¹V(n, p)⁵¹Ti reaction from E_n = 2.856 to 9.267 MeV and for the ⁵¹V(n, α)⁴⁸Sc reaction from 5.515 to 9.567 MeV. Both measurements utilize ENDF/B-V evaluated neutron-induced fission cross sections of ²³⁸U as a standard. The experimental results from this work are compared with corresponding ENDF/B-V evaluated cross sections for V and substantial differences are evident. The most significant difference is a tendency for the measured values to exceed evaluated ones by as much as 50% in the vicinity of 8 MeV.     

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.     

Thermal neutron cross-section and resonance integral of the Mo(n,γ)Mo reaction

We measured the thermal neutron cross-section and the resonance integral of the ⁹⁸Mo(n,γ)⁹⁹ Mo reaction by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸ Au monitor reaction as a single comparator. The high-purity natural Mo and Au metallic foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The necessary correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th) and the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. In addition, for the ⁹⁹Mo activity measurements, the correction for true coincidence summing effects was also taken into account. The thermal neutron cross-section for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction has been determined to be 0.136 ± 0.007 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸ Au reaction. The present result is, in general, in good agreement with most of the experimental data and the recently evaluated values of ENDF/B-VII.0, JENDL-3.3, and JEF-2.2 by 5.1% (1σ). By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction is 7.02 ± 0.62 barn, which is determined relative to the reference values of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present resonance integral value is in general good agreement with the previously reported data by 8.8% (1σ).     

Experimental Determination of Spectral Indices in Low Enriched UO2-H2O Lattices

To get information about the neutron spectrum in low enriched UO₂-H₂O lattices, the spectral indices SI(U^8c/Dy) and SI(U^8c/U^5f) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO₂ was used for the measurement of activities caused by the neutron capture of ²³⁸U and the fission of ²³⁵U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.     

Lattice parameters of (U, Pu, Am, Np)O2−x

The solid solutions of (U_{1−z−y’−y”}Pu_zAm_y’Np_y”)O_2−x (z = 0-1, y’ = 0-0.12, y” = 0-0.07) were investigated by X-ray diffraction measurements, and a database for the lattice parameters was updated. A model to calculate the lattice parameters was derived from the database. The radii of the ions present in the fluorite structure of (U, Pu, Am, Np)O_2−x were estimated from the lattice parameters measured in this work. The model represented the experimental data within a standard deviation of σ = ±0.025%.     

Densification behaviour and sintering kinetics of ThO2-4%UO2 pellet

ThO₂-?4% ²³³UO₂ fuel will be the driver fuel for the forthcoming Advanced Heavy Water Reactor (AHWR) in India. Densification behaviour such as shrinkage and shrinkage rates of the green pellets of ThO₂-4wt.% UO₂ (natural ‘U’) fabricated by Coated Agglomerate Pelletization (CAP) process were studied using a vertical dilatometer at different heating rates. Activation energy of sintering, ‘Q’, was estimated in the initial stages of sintering by continuous rate of heating (CRH) technique as proposed by ‘Wang and Rishi Raj’ and ‘Young and Cutler’. The sintering mechanism was identified to be as the grain boundary diffusion (GBD) and the average ‘Q’ value obtained by these two methods were found to be 350 ± 16 kJ/mole and 358 ± 5 kJ/mole, respectively.     

Approximate Calculation Method for Second Order Sensitivity Coefficient

A simple method has been developed for calculating the second order sensitivity coefficient of static and burnup-dependent core performance parameters. The method is applied to a small and a large fast breeder reactors. Changes in core performance parameters due to 10% cross section changes are compared with that predicted by the first and the second order sensitivity analyses. Numerical results reveal that the changes in breeding ratio, reaction rate ratio of the ²³⁸U capture to the ²³⁹Pu fission rate and burnup reactivity loss due to the 10% change in the ²³⁹Pu fission cross section and/or the ²³⁹Pu v-value show nonlinear behavior, and the second order sensitivity can predict the changes accurately.     

Measurement of keV-neutron capture cross-sections and capture γ-ray spectra of Fe and Fe

The neutron capture cross-sections and the capture γ-ray spectra of ⁵⁶Fe and ⁵⁷Fe have been measured in the neutron energy range from 10 to 90 keV. Pulsed keV-neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with a 1.5-ns bunched proton beam from a 3 MV Pelletron accelerator. The incident neutron spectrum on the capture sample was measured using a time-of-flight method with a ⁶Li-glass detector. The capture γ-rays emitted from an iron or standard gold sample were detected with a large anti-Compton NaI(Tl) spectrometer. The capture yield of the iron or gold sample was obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The capture cross-sections of ^56,57Fe were derived with errors less than 5% using the standard capture cross-sections of ¹⁹⁷Au. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. The present results for the capture cross-sections were compared with the previous measurements and the evaluated values of ENDF/B-VII.0 and JENDL-3.3. The Maxwellian-averaged capture cross-sections of ⁵⁶Fe and ⁵⁷Fe at 30 keV are derived as 12.22 ± 2.06 mb and 44.48 ± 7.56 mb, respectively.