Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC

The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.

Approach to Criticality

Source neutrons are essential for reactor restart after a long shutdown. The term “source neutrons” applied to a particular time interval refers to a steady supply of neutrons, constant over the time interval of interest. This supply must be independent of the current or very recent fission rate, which can vary over the time interval. Thus, source neutrons exclude prompt neutrons and even delayed neutrons which originate in the fuel (i.e., those born in the fuel itself). This exclusion does not apply to delayed photoneutrons, which come from fissions that have occurred a long time before, and whose numbers are quite constant over the current time interval (further discussion of this point below).

Reactor Shutdown and Reactor Restart

The power referred to most frequently in reactor physics is neutron power. Neutron power is essentially the fission rate multiplied by the average prompt energy released and recovered per fission (see Section 2.1.2). It is also called “prompt” power, as it appears very quickly following fission. We cannot measure neutron power directly, but we do monitor the neutron flux with ion chambers located outside the calandria and in-core flux detectors. These neutronic signals are calibrated to the thermal-power measurement which allows neutron power to be derived.

Impact of Approximations in Operating History Data on Spent Fuel Properties with Serpent 2

In this work, the effect of averaging operating history parameters such as power history, boron concentration and coolant density and temperature on spent nuclear fuel properties was investigated. The examined properties were assembly activity, decay heat, photon emission rate, spontaneous fission rate and the concentration of some mobile nuclides and fissile nuclides. Calculations were performed on two similar VVER-440 fuel assemblies irradiated in different positions of the core using Serpent 2. Averaging power history over the entire irradiation history had a significant effect on assembly activity, decay heat and photon emission rate overestimating these properties approximately 70 % right after irradiation. However, the effect quickly died out and after 10 years of cooling the effect was less than 1 %. If the last cycle (3rd cycle) was modelled accurately and the power density of only the first two cycles were averaged, the differences remained always below 1 %. The effect of operating history approximations on spontaneous fission rate and the nuclide concentrations was much smaller reamaining mostly below 1.5 %. The sensitivity of nuclide concentrations to approximations in individual operating history parameters was dependent on the nuclide in question and no trend applying to all studied nuclides could be observed.

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, choice of calculation parameters etc. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of Serpent. The effect of the following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy dependent branching ratios. As a test case a 2D BWR fuel assembly was modelled by first running a burnup calculation followed by a decay calculation. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6 % and for the photon emission rate approximately 1.1 %. For the spontaneous fission rate maximum absolute relative difference of nearly 8 % was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy dependent branching ratios had an insignificant effect on the studied source term components.

GENERATION AND INITIAL VALIDATION OF A NEW CASMO5 ENDF/B-VIII.0 NUCLEAR DATA LIBRARY

A new nuclear data library for the CASMO5 advanced lattice physics code has been generated based on the recently-released ENDF/B-VIII.0 evaluation. The ENDF/B-VIII.0 evaluation represents the state-of-the-art in nuclear data and features new evaluations from the CIELO project for 1H, 16O, 56Fe, 235U, 238U and 239Pu. A description of the library generation procedure used to process these data into the CASMO5 586 energy group structure is provided. Initial validation of the new ENDF/B-VIII.0-based library, referred to as the E8R0 library, is also presented and involves the comparison of predicted k–eff and fission rate distributions to measurements from various critical experiments. The critical experiments used in the initial validation of the E8R0 library consist of the B&W 1810 series, B&W 1484 series, DIMPLE S06A/B, and TCA reflector experiment with iron plates. The results from the initial validation indicate that the new E8R0 library provides a satisfactory performance in terms of CASMO5 predicted k–eff and fission distributions.

FUEL-CYCLE SCENARIO TO REDUCE RADIOACTIVE WASTE FROM LIGHT-WATER REACTOR

It is important to reduce the amount of trans-uranium (TRU) produced from the existing nuclear power plants to realize sustainable nuclear energy since the some TRU nuclides remain for a long time and have high radioactivity and radiotoxicity. One of the promising solutions is to transmute the TRU nuclides to those with lesser radioactivity and radiotoxicity in the existing nuclear reactors. In the current scheme, the TRU nuclides are transmuted in fast reactors and/or accelerator-driven-systems, however, this scenario seems unpromising in Japan: after the Fukushima Daiichi accident, it is required to reduce the production of TRU nuclides from the light-water reactors. In the previous studies, a concept of FORSETI was investigated, and a nuclear-fuel cycle simulation code ATRUNCYS was developed to study the low TRU production scenario. The FORSETI concept consists of two types of fuels: 1) UO2 fuels with high-assay low-enriched-uranium, and 2) MOX fuels with highly fissile concentrated plutonium reprocessed from the FORSETI-UO2 fuels. The current paper focuses on the following two scenarios: a) once-recycled scenario with the current fuel design, and b) once-recycled scenario with the FORSETI concept. The two scenarios were compared by using the ATRUNCYS code where the simulation studies showed that the amount, radioactivity, and radiotoxicity of resulting waste can be decreased in the FORSETI concept: In the case 1), the production of TRU nuclides decreased in the UO2 fuel; In the case 2), the fission rate increased and neutron-capture reactions of 240Pu and 241Pu decreased in the MOX fuels.

APPLICATION OF RECENT DEVELOPMENTS IN INTEGRAL DATA ASSIMILATION TO IN-DEPTH ANALYSIS OF UH1.2 EXPERIMENT AND TRANSPOSITION TO WHOLE PWR CORE

Recent developments in the Integral Data Assimilation (IDA) methods within Bayesian framework have been achieved at CEA to tackle the problem of correlated experiments (through technological uncertainties) and neutron transport model numerical effects. Hence, reference Monte-Carlo and deterministic calculations (TRIPOLI4® and APOLLO3®) are used together to solve neutron transport equations and get the sensitivity profiles. Furthermore, the analysis of the mock-up experiments technological parameters is performed to get accurate uncertainties and correlations between the experiments (finally the covariance experimental matrix required for IDA). We apply here the IDA approach with a new, extend set of statistical indicators (Cook’s distance, Bayesian and Aikike Information criteria (BIC, AIC)) implemented in the nuclear physics CEA CONRAD code, to the integral experiments UH1.2 in reference and voided configurations (standard PWR fuel assembly in the EOLE mock-up reactor). The adjusted multigroup cross-sections and posterior covariances are compared by choosing different ingredients in the assimilation process. Finally, the investigated key neutron parameters; reactivity, reactivity worth (void effects) and fissions rates are transposed (with the same CONRAD code) to a standard PWR core. This in-depth analysis enables us to predict the residual uncertainties and biases due to the multigroup cross-section adjustments assessing at the same time the similarity of these integral experiments for the main PWR neutronic safety parameters. In addition, technological parameters uncertainties and their impact on Bayesian adjustment process are taken into account through a global experimental covariance matrix. We point out that the UH1.2 experiments bring relevant additional information to PWR keff calculations reducing significantly the posterior results but are less relevant for fission rate distribution in reference and voided configurations.