David J. Diamond

Calculation of the prompt neutron lifetime in the NBSR

Abstract The prompt neutron lifetime was calculated for the NBSR, a heavy water-cooled and -moderated research reactor at the National Institute of Standards and Technology. The method is based on the fact that the decay of a pulse of fast neutrons is related to the prompt neutron lifetime (and the multiplication constant for the reactor and the delayed neutron fraction). A Monte Carlo simulation of the decay is then used to calculate the prompt neutron lifetime at two points in the fuel cycle. At the start-up of a new cycle, the prompt neutron lifetime was calculated to be 774 ± 35 μs, and at the end of a cycle, it was calculated to be 819 ± 48 μs.

Higher Order Effects in Calculating Boiling Water Reactor Doppler and Void Reactivity Feedback

The effect of steam voids and control rods on the Doppler feedback and of bypass voids on the total void feedback has been calculated for a gadolinia-shimmed boiling water reactor. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients, the inclusion of the void effect will make the Doppler feedback stronger. The effect of control rods is to reduce Doppler feedback. For overpressurization transients, the inclusion of the effect of bypass void will increase the reactivity due to void collapse.

Effect of Shim Arm Depletion in the NBSR

The cadmium shim arms in the NBSR undergo burnup during reactor operation and hence, require periodic replacement. Presently, the shim arms are replaced after every 25 cycles to guarantee they can maintain sufficient shutdown margin. Two prior reports document the expected change in the 113Cd distribution because of the shim arm depletion. One set of calculations was for the present high-enriched uranium fuel and the other for the low-enriched uranium fuel when it was in the COMP7 configuration (7 inch fuel length vs. the present 11 inch length). The depleted 113Cd distributions calculated for these cores were applied to the current design for an equilibrium low-enriched uranium core. This report details the predicted effects, if any, of shim arm depletion on the shim arm worth, the shutdown margin, power distributions and kinetics parameters.

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

A Neutronic Analysis of Boiling Water Reactor In-Core Detector Noise

The response of boiling water reactor in-core detectors undergoing vibration has been calculated. A neutronic model based on calculating the fission activity at a detector position in a planar multibundle environment was employed. The model used eight energy groups and two-dimensional Cartesian geometry in a discrete-ordinates transport approximation. The in-core detector responses due to various detector displacements were calculated as a function of channel box corner wear with different effective in-channel voids, bypass voids, and instrument tube voids. The calculated noise was found to have a linear dependence on channel box wear. This was corroborated by measurements. An increase in in-channel voids was found to increase the noise, while an increase in bypass and instrument tube voids decreased the noise. The presence of a nearby control blade increased the noise.

Analysis of Reactivity Insertion Accidents for the NIST Research Reactor Before and After Fuel Conversion

Reactivity insertion accidents have been analyzed for the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains highly enriched uranium fuel, and for a proposed equilibrium core with low-enriched uranium fuel. The time-dependent analysis of the primary system is performed with a RELAP5 model that includes the reactor vessel, primary coolant pump, heat exchanger, fuel element geometry, and flow channels for both the 6 inner and 24 outer fuel elements. Postprocessing of the simulation results has been conducted to evaluate minimum critical heat flux (CHF) ratio and minimum onset of flow instability (OFI) ratio using the Sudo-Kaminaga correlations and Saha-Zuber criteria, respectively. Evaluations are carried out for the control rod withdrawal start-up accident and the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that no damage to the fuel will occur and there is adequate margin to CHF and OFI because of sufficient coolant flow through the fuel channels and the negative reactivity insertion due to scram.

Nuclear Power and Energy Security

Preface.- RA Presidents Message.- Opening Remarks NATO Advanced Research Workshop D. Diamond.- -Developing The Necessary Infrastructure. IAEA Activities in Support of Countries Considering Embarking on Nuclear Power Programme O. Akira.-Creating a National Nuclear Regulatory Authority R. Way.- Building Safeguards Infrastructure J. McClelland-Kerr et al.- Regulatory Challenges Related To The Licensing Of A New Nuclear Power Plant M. Maris.- Infrastructure Development Through Civil Nuclear Cooperation M. Humphrey, A. Burkart.-Nuclear Safety Infrastructure R. Moffitt.- Upgrading Nuclear Regulatory Infrastructure in Armenia A. Martirosyan et al.- Seven Principles of Highly Effective Nuclear Energy Programs Ch. Ferguson, Ph. Reed.- The Case For Nuclear Energy. Nuclear Safety and Energy Security G. Trosman.-Nuclear Energy and Social Impact N. Carpintero-Santamaria.-The Role of Nuclear Power in the Reduction of Greenhouse Gas Emissions A. Baratta.-Nuclear Energy & Energy Security J. Mamasakhlisi.- Overview Of The Electricity Market Of Estonia And The Plausibility Of Nuclear Energy Production M. Lehtveer, A. Tkaczyk.-The role of Small and Medium Reactors in the Energy Security of a Country, IRIS Example N. Cavlina.-- Applicable Technology. Status Report on the Safety of Operating US Nuclear Power Plants R. Budnitz.- NATO-ASTEC-MATRIX -Research Environment, Information Sharing and MC S. Apikyan et al.- Establishing Control Over Nuclear Materials And Radiation Sources In Georgia G. Basilia.- Nuclear Energy In Armenia History, Problems, Possibilities And Outlook G. Sevikya et al.-Development of Nuclear Energy in Armenia A. Gevorgyan, A.Galstyan.- Some Neutron Absorbing Elements and Devices for Fast Nuclear Reactors Regulation Systems P. Kervalishvili.- Development of Design of a Radioisotope Switchable Neutron Source and New Portable Detector of Smuggling L. Meskhi, L.Kurdadze.- New designs of medium power VVER reactor plants S.B.Ryzhov et al.-National Assessment Study In Armenia Using Innovative Nuclear Reactors And Fuel Cycles Methodology For An Innovative Nuclear Systems In A Country With Small Grid V. Sargysan et al.- CANDLE Reactor: An Option For Simple, Safe, High Nuclear Proliferation Resistant, Small Waste And Efficient Fuel Use Reactor H. Sekimoto.- Emissions of the corrosion radionucides in an atmosphere M. Vardanyan.- IAEA Support for Operating Nuclear Reactors O. Akira.-The Solid Coolant and Prospects of Its Use in Innovative Reactors A.M Dmitriev, V. Deniskin.- Innovation Projects of Atomic Energy Institute of National Nuclear Center RK in the Area of Peaceful Use of Atomic Energy E.Kenzhin et al.- Innovative Designs of Nuclear Reactors B. Gabaraev, Y. Cherepnin.- Development of Devices for Handling with BN-350 Radioactive Waste A.G. Ikanov et al.- Institutional Support to the Nuclear Power Based on Transportable Installations V. Kuznetsov, Y. Cherepnin.-International Cooperation and Security in the Field of Nuclear Energy in Armenia D. Khachatryan et al.- Applied Model of Through-Wall Crack of Coolant Vessels of VVER-type Reactors V.Petrosyan et al.-Index.

Using Fuel Performance to Determine Operating Limit Uncertainty Allowance

An analytic relationship between fuel failure (defined as exceeding a prescribed fuel limit) due to measurement uncertainties and operating limit uncertainty allowance is derived. The expected number of fuel failures due to measurement uncertainties is evaluated for a set of selected power distributions and a general bounding power distribution. The expected number of fuel pins challenging limits is determined as a function of uncertainty allowance for selected core power distributions. 5 refs.

TRACE/PARCS Analysis of Anticipated Transient Without Scram with Instability for a MELLLA+ BWR/5

Abstract A TRACE/PARCS model has been developed to analyze anticipated transient without scram (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands the allowable operation in the power/flow map of a BWR to low flow rates at high-power conditions. Such operation exacerbates the likelihood of large-amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large-amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural-circulation flow in the reactor pressure vessel after the trip of the recirculation pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heatup that could result in localized fuel damage. TRACE predicts that heatup will occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet, and the fuel becomes locked into a boiling-film regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.

ANALYSIS OF LOSS-OF-FLOW ACCIDENTS FOR THE NIST RESEARCH REACTOR WITH FUEL CONVERSION FROM HEU TO LEU

Abstract A program is underway to convert the current high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel in the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology. A RELAP5 model has been developed to analyze postulated accidents in the NBSR with the present HEU fuel and a proposed LEU fuel. The model includes the reactor vessel, primary pumps, shutdown pumps, various valves, heat exchangers, and average and hottest fuel elements and flow channels in the region where flow enters through an inner plenum (6 fuel elements) and a region where flow enters through an outer plenum (24 elements). The equilibrium cycle power distributions in the fuel elements were determined based on three-dimensional Monte Carlo neutron transport calculations performed with the MCNPX code. In this paper we discuss safety analyses conducted for the loss-of-flow accidents resulting from either loss of electrical power or inadvertent throttling of flow control valves at the inlets to the inner and outer plena. The analysis shows that the fuel conversion will not lead to significant changes in the safety analysis and that the calculated maximum clad temperatures, minimum critical heat flux ratios, and minimum onset of flow instability ratios assure that there is adequate margin to fuel failure.