BUGLE-96 Cross-section Library Research Articles

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Analysis of the H. B. ROBINSON-2 reactor pressure vessel dosimetry benchmark using TRIPOLI-4® Monte Carlo code

This paper presents an analysis of the H.B. Robinson-2 reactor pressure vessel dosimetry benchmark. This benchmark is based on an experiment performed at an operating PWR (Pressurized Water Reactor). It includes measurements for both in-vessel and ex-vessel dosimeters. To enlarge the TRIPOLI-4 ® validation database of neutron attenuation in PWR, an analysis of H.B. Robinson-2 benchmark was carried out using the TRIPOLI-4 ® Monte Carlo code, the DARWIN/PEPIN2 depletion code, the general-purpose point-wise library JEFF3.1.1, and the International Reactor Dosimetry File IRDF-2002. Calculated activity values are compared with measured ones provided in this benchmark. Furthermore, calculated activity values are also compared with results obtained using two calculation schemes: the deterministic code DORT and the BUGLE-96 cross section library on the one hand, and the Monte Carlo code MCNPX and the JEFF3.1.1 cross section library on the other hand.

Neutron Flux Reduction Programs for Reactor Pressure Vessel of Korea Nuclear Unit 1

Abstract The objective of this work is to implement various fast neutron flux reduction programs on the beltline region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RTPTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RTPTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RTPTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RTPTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library.

VVER-440 and VVER-1000 Reactor Dosimetry Benchmark —BUGLE-96 Versus ALPAN VII.0

Abstract Analytical results of the vodo-vodyanoi energetichesky reactor- (VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Institute Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10 %) between BUGLE-96 and ALPAN VII.0 libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15 % with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements.

On Neutron Spectroscopy Using Gas Proportional Detectors Optimized by Transport Theory

We explore in this study the practical limits in designing a neutron detector array to resolve the spectra from special nuclear material (SNM) neutron sources using 3He detectors. We demonstrate that radiation transport analysis yielded a spectrum unfolding strategy based on the energy structure of the BUGLE-96 cross-section library, with 47 neutron energy groups. The initial computational model used is an isotropic planar source incident on a block of high-density polyethylene moderator. Assuming 3He is diluted throughout the moderator, the 3He(n,p) reaction rate energy group matrix in the block was computed using a completely “flat” neutron source spectrum. Analyzing the energy group matrix, there are neutrons from specific collections of energy groups (energy “bands”) that induce a maximum reaction rate in specific locations; we determined that these groups cannot be further differentiated within the energy band using 3He detectors. It was determined that optimal spectral fidelity for SNM detection and characterization is achievable using four spectral bands spanning among groups 1 through 29 (31.8 keV to 17.3 MeV). Using ideal-filter materials to remove the neutrons from different regions of the spectrum, we predicted the maximum neutron spectral resolution obtainable using this approach. To demonstrate our method, we present the optimally detected spectral differences between SNM materials (plutonium and uranium), metal and oxide, using ideal-filter materials. We have also selected a number of candidate filtering materials and, by replacing the ideal filters with real materials, we exemplified our approach with a design of a neutron detector array capable of resolving the spectra from SNM neutron sources using 3He detectors.

Ex-Core Detector Response Evaluation of the SMART Reactor by Using the DORT Code

The development of a 660 MWt integral reactor SMART (System integrated Modular Advanced ReacTor), called SMART-660, has been initiated in Korea. A preliminary evaluation of the detector response functions was performed for the ex-core detector of the SMART-660. First, a candidate position to install the ex-core detector was selected by considering the reactor configuration, the detector performance, and the thermal neutron flux distribution in the reactor assembly. Second, the detector response functions such as the shape annealing function (SAF) and assembly weighting factor (AWF) for the candidate ex-core detector were calculated by using an adjoint transport calculation with the DORT code and the BUGLE-96 cross-section library. The GEOSHIELD program was used to generate the DORT input file and to process the DORT output file with a graphic visualization. This study has shown that the response functions of the ex-core detector are only sensitive to some parts of the reactor and additional efforts are required to reduce the locality of the ex-core detector response.

Radiation Shielding Analysis for the Reactor Assembly of the SMART Reactor

A 660 MWt integral reactor SMART (System integrated Modular Advanced ReacTor), called SMART-660, is under development in Korea as a national research and development project to supply energy for a seawater desalination as well as an electricity generation. A radiation shielding analysis has been carried out to evaluate the shielding design of the SMART-660. Two-dimensional discrete ordinates transport code DORT was used to calculate the neutron and gamma dose rates and the fast neutron fluences in the reactor pressure vessel assembly (RPV). The reactor assembly of the SMART-660 was modeled as an R-Z geometry through an azimuthal homogenization of the reactor components. The GEOSHIELD program was used to generate the DORT input file and process the DORT output file with a graphic visualization. The neutron 47-group and gamma 20-group BUGLE-96 cross-section library was used for the DORT calculation. It is found that the integrity of the RPV is preserved throughout the reactor lifetime and the shielding system of the SMART-660 is satisfactory to prevent radiation escaping from the reactor core.

Inversion of time‐dependent nuclear well‐logging data using neural networks

ABSTRACTThe purpose of this work was to investigate a new and fast inversion methodology for the prediction of subsurface formation properties such as porosity, salinity and oil saturation, using time‐dependent nuclear well logging data. Although the ultimate aim is to apply the technique to real‐field data, an initial investigation as described in this paper, was first required; this has been carried out using simulation results from the time‐dependent radiation transport problem within a borehole. Simulated neutron and γ‐ray fluxes at two sodium iodide (NaI) detectors, one near and one far from a pulsed neutron source emitting at ∼14 MeV, were used for the investigation. A total of 67 energy groups from the BUGLE96 cross section library together with 567 property combinations were employed for the original flux response generation, achieved by solving numerically the time‐dependent Boltzmann radiation transport equation in its even parity form. Material property combinations (scenarios) and their correspondent teaching outputs (flux response at detectors) are used to train the Artificial Neural Networks (ANNs) and test data is used to assess the accuracy of the ANNs. The trained networks are then used to produce a surrogate model of the expensive, in terms of computational time and resources, forward model with which a simple inversion method is applied to calculate material properties from the time evolution of flux responses at the two detectors.The inversion technique uses a fast surrogate model comprising 8026 artificial neural networks, which consist of an input layer with three input units (neurons) for porosity, salinity and oil saturation; and two hidden layers and one output neuron representing the scalar photon or neutron flux prediction at the detector. This is the first time this technique has been applied to invert pulsed neutron logging tool information and the results produced are very promising. The next step in the procedure is to apply the methodology to real data.