R. Chawla - Academia.edu (original) (raw)

Papers by R. Chawla

Progress in Nuclear Energy

Nuclear Technology

A consistent analytical comparison has been made of the transient behavior of critical and subcri... more A consistent analytical comparison has been made of the transient behavior of critical and subcritical fastspectrum reactor systems, the basic core design assumed in each case being that of the 80-MW(thermal) mixedoxide-fueled, Pb-Bi-cooled, Experimental Accelerator Driven System (XADS). The transient calculations were performed using the FAST code system developed at the Paul Scherrer Institute. The present study demonstrates a high level of self-protection of both the critical and subcritical systems over a wide range of postulated events, including transient overpower due to reactivity insertion, loss of flow, station blackout, loss of coolant, and core overcooling accidents. The relative advantages and shortcomings of the two system types, from the viewpoint of transient behavior, are discussed on the basis of the corresponding simulation results obtained.

Nuclear Science and Engineering

ABSTRACT The reaction-rate ratio C8/Ftot, neutron captures in 238U to total fissions, has been me... more ABSTRACT The reaction-rate ratio C8/Ftot, neutron captures in 238U to total fissions, has been measured in 80 out of 96 fuel rods of a Westinghouse SVEA-96+ boiling water reactor fuel assembly. High resolution gamma spectroscopy was performed on individual fuel rods, withdrawn from the SVEA-96+ assembly after irradiation at low power in the center of the LWR-PROTEUS reactor core. Absolute experimental errors of 1.7% and relative errors of 0.6% (for rod-to-rod ratios) were achieved. The experimental results were used as a database for validation of four different calculational tools: CASMO-4 and HELIOS as commercial assembly codes, the Paul Scherrer Institute in-house code BOXER, and the Monte Carlo transport code MCNPX. In general, on the level of a few percent, there is good agreement between experiment and calculations, the use of a recently proposed 239Np gamma-ray emission probability improving even further the agreement. However, the highly heterogeneous design of the SVEA-96+ assembly (both in terms of material compositions and neutron moderation conditions) causes some problems. Clear deviations from assembly mean values are found among the burnable absorber fuel rods that are grouped in clusters (direct neighbors), a unique feature of this assembly design. For these rods the codes overpredict C8/Ftot by several percent, including MCNPX. Additional trends, not present in the results from the Monte Carlo calculation which generally shows the best overall agreement with experiment, are identified for the deterministic codes.

Several new fuel designs are currently being considered for the Generation-IV gascooled fast reac... more Several new fuel designs are currently being considered for the Generation-IV gascooled fast reactor. Two designs have been analysed in the paper: 1) a ceramic honeycomb plate matrix containing fuel cylinders (slab geometry) and 2) a pin with fuel pellets in a ceramic cladding (rod geometry). Mixed uranium-plutonium carbide is the fuel in both cases, while the matrix/cladding material is silicon carbide. A simplified approach to the transient simulation of the heterogeneous plate design has been developed and benchmarked against detailed finiteelement code calculations. The limits of applicability of the simplified approach to transient calculations have been estimated and directions for its further improvement have been formulated. The slab and rod fuel designs were then compared from the viewpoint of their thermal transient response to a number of hypothetical accidents, including loss of heat sink, core overcooling, loss of flow and transient overpower.

Transactions of the American Nuclear Society

Nuclear Physics A, 2006

A high-fluence proton irradiation of neptunium was the last experiment in PSIs programme ATHENA r... more A high-fluence proton irradiation of neptunium was the last experiment in PSIs programme ATHENA related to accelerator-based transmutation. The principal aim of the programme has been to provide experimental data for the validation of theoretical models in nucleon-meson transport codes, with emphasis on the mass yield distribution of fission and spallation products. An improved mass spectrometry method has allowed the direct derivation of isobaric production cross sections with only minor corrections and an estimate of the fission cross section by integration in the fission hump. In a second sample position of the irradiation head, a repetition of the previous ATHENA experiment with thorium was possible, profiting from the improved mass spectrometry technique. The experimental results are better predicted by the code FUSSPOT than by HETC/RAL, both used at PSI.

Progress in Nuclear Energy, 2001

Zirconium is employed as the “inert” component in several different inert matrix fuel concepts cu... more Zirconium is employed as the “inert” component in several different inert matrix fuel concepts currently under development (cermet, oxide solution, etc.). Relative to the situation in standard light water reactors, this implies a very significant increase in the zirconium inventory, with a correspondingly enhanced importance of the nuclear data for zirconium in calculating safety related parameters. The present paper discusses new numerical results for various reactivity effects. On the basis of the current findings, it is recommended that a high priority be assigned to both the reduction of uncertainties in the epithermal data for zirconium and to the improvement of methods for resonance self-shielding of these data.

Fuel and Energy Abstracts, 2002

Nuclear Science and Engineering, 2009

ABSTRACT Radial distributions of the total fission rate and the 238U-capture-to-total-fission (C8... more ABSTRACT Radial distributions of the total fission rate and the 238U-capture-to-total-fission (C8/Ftot) ratio were measured in SVEA-96+ and SVEA-96 Optima2 assemblies during the LWR-PROTEUS program. Fission rates predicted using MCNPX with JEFF-3.1 cross sections underestimated the measured values in the gadolinium-poisoned pins of the SVEA-96 Optimal assembly; similarly, C8/Ftot ratios were overestimated in some gadolinium-poisoned pins of the SVEA-96+ assembly. A considerable effort was invested at the Paul Scherrer Institut to explain the discrepancies in gadolinium pins, without success. Recently, gadolinium cross sections were measured at the Rensselaer Polytechnic Institute by Leinweber et al. and differed significantly from current library values. ENDF/BVII.0 gadolinium cross sections have currently been modified to include the new measurements, and these data have been processed with NJOY to yield files usable by MCNPX. Fission rates in the gadolinium-poisoned fuel pins of the SVEA-96 Optimal pins were increased by 1.4 to 2.0% using the newly produced cross sections, yielding to a better agreement with the experimental values. Predicted C8/Ftot ratios were decreased on average by 1.7% in both clustered and unclustered groups of gadolinium-poisoned fuel pins of the SVEA-96+ assembly correcting the overpredictions previously reported in the clustered gadolinium pins. Earlier reported discrepancies observed in PROTEUS integral experiments, between measured and calculated reaction rates in the gadolinium-poisoned pins, might thus be due to inaccurate gadolinium cross sections. The PROTEUS results support the new thermal and epithermal gadolinium data measured by Leinweber et al.

Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2004

Recently, the original Lawrence Livermore National Laboratory Evaluated Electron Data Library [S.... more Recently, the original Lawrence Livermore National Laboratory Evaluated Electron Data Library [S.T. Perkins, D.E. Cullen, S.M. Seltzer, LLNL, CA, UCRL-50400, 31 (1991)] was extended down from 10 eV to 1 eV. Efforts have also been made to generate a similarly formatted data library for positrons-EPODL. These libraries now include data in the energy range from 1 eV to 100 GeV for electron and positron interactions with neutral atoms of the elements H through Fm. Validation of the new data has been carried out by comparison with experimental results for single ionization and differential elastic scattering cross-sections. The calculated cross-section data are generally found to be in good agreement with reported measurements.

Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2006

High-resolution gamma spectroscopy was performed on individual fuel rods of a fresh, highly heter... more High-resolution gamma spectroscopy was performed on individual fuel rods of a fresh, highly heterogeneous Boiling Water Reactor (BWR) fuel assembly, after irradiation at low power in the PROTEUS reactor at PSI, to determine the ratio of neutron captures in 238 U ðC 8 Þ to total fissions ðF tot Þ. The methods for correcting the measured gamma count-rates for attenuation within the fuel rods and the collimator shielding are now so advanced that one can obtain several independent estimates of F tot for each fuel rod by using gamma lines from different fission products with a wide range of energies. The recording of a detailed irradiation history further allowed the gamma activities to be precisely corrected for radioactive decay between the time of irradiation and measurement. Due to these facts, and the good statistical quality of the present data, the main limitation on the accuracy of C 8 =F tot comes from uncertainties on the basic nuclear quantities: fission yields and beta-gamma branching ratios. The various steps involved in determining C 8 =F tot from recorded gamma spectra, and their contributions to the final error, are analysed. The results for 80 fuel rods of varying material composition exposed to different neutron moderation conditions in the reactor core are compared with the results of a Monte Carlo full-assembly calculation.

Nuclear Engineering and Design, 2012

This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated... more This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat à l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process-which is foreseen for operation at high and medium pressures-with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called "Brayton scoping model", is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the "Brayton scoping" model, e.g., that the turbo-machine conveniently accelerates during the depressurization process to tend towards a steady rotational speed value, the speed rise being inversely proportional to the experienced pressure drop. Finally, CATHARE simulations are presented for complete DHR scenarios for the GFR, involving lossof-coolant-accidents (LOCAs) in conjunction with loss of backup pressure (LOBP). Thereby, it is shown that, in each of the investigated cases, incorporation of the Brayton loop turbo-machine with nitrogen indeed leads to fuel temperatures remaining considerably below Category 4 accident limits.

Nuclear Engineering and Design, 2006

The paper presents a comparison of transient calculations for two 80-MWt MOX-fuelled experimental... more The paper presents a comparison of transient calculations for two 80-MWt MOX-fuelled experimental accelerator-driven systems (XADS), one cooled by lead-bismuth eutectic and the other by helium. The results for protected (with accelerator beam trip) and unprotected (without accelerator beam trip) transient overpower, spurious beam trip, loss of flow, loss of heat sink, and loss of coolant accidents, as well as the failure of heat exchanger tubes, are analysed for the two systems. The analysis was performed using TRAC/AAA, which is part of the PSI FAST code system. The advantages and shortcomings of the two designs from the viewpoint of their transient behaviour are discussed.

Nuclear Instruments and Methods, 1980

Journal of Nuclear Materials, 1999

ABSTRACT Light water reactor (LWR) neutronics codes and cross-section libraries need further qual... more ABSTRACT Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO2–Er2O3–ZrO2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO2–UO2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu–Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

Journal of Nuclear Materials, 2006

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is ... more The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO 2 fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.

Nuclear Engineering and Design, 2000