EP2215617A1 - Reactor dosimetry applications using a parallel 3-d radiation transport code - Google Patents
Reactor dosimetry applications using a parallel 3-d radiation transport codeInfo
- Publication number
- EP2215617A1 EP2215617A1 EP08833567A EP08833567A EP2215617A1 EP 2215617 A1 EP2215617 A1 EP 2215617A1 EP 08833567 A EP08833567 A EP 08833567A EP 08833567 A EP08833567 A EP 08833567A EP 2215617 A1 EP2215617 A1 EP 2215617A1
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- European Patent Office
- Prior art keywords
- radiation
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- angular
- processor
- spatial
- Prior art date
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- G—PHYSICS
- G09—EDUCATION; CRYPTOGRAPHY; DISPLAY; ADVERTISING; SEALS
- G09B—EDUCATIONAL OR DEMONSTRATION APPLIANCES; APPLIANCES FOR TEACHING, OR COMMUNICATING WITH, THE BLIND, DEAF OR MUTE; MODELS; PLANETARIA; GLOBES; MAPS; DIAGRAMS
- G09B23/00—Models for scientific, medical, or mathematical purposes, e.g. full-sized devices for demonstration purposes
- G09B23/06—Models for scientific, medical, or mathematical purposes, e.g. full-sized devices for demonstration purposes for physics
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- G—PHYSICS
- G09—EDUCATION; CRYPTOGRAPHY; DISPLAY; ADVERTISING; SEALS
- G09B—EDUCATIONAL OR DEMONSTRATION APPLIANCES; APPLIANCES FOR TEACHING, OR COMMUNICATING WITH, THE BLIND, DEAF OR MUTE; MODELS; PLANETARIA; GLOBES; MAPS; DIAGRAMS
- G09B19/00—Teaching not covered by other main groups of this subclass
-
- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F30/00—Computer-aided design [CAD]
- G06F30/20—Design optimisation, verification or simulation
-
- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F9/00—Arrangements for program control, e.g. control units
- G06F9/06—Arrangements for program control, e.g. control units using stored programs, i.e. using an internal store of processing equipment to receive or retain programs
- G06F9/44—Arrangements for executing specific programs
- G06F9/445—Program loading or initiating
- G06F9/44521—Dynamic linking or loading; Link editing at or after load time, e.g. Java class loading
-
- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F9/00—Arrangements for program control, e.g. control units
- G06F9/06—Arrangements for program control, e.g. control units using stored programs, i.e. using an internal store of processing equipment to receive or retain programs
- G06F9/44—Arrangements for executing specific programs
- G06F9/455—Emulation; Interpretation; Software simulation, e.g. virtualisation or emulation of application or operating system execution engines
-
- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F9/00—Arrangements for program control, e.g. control units
- G06F9/06—Arrangements for program control, e.g. control units using stored programs, i.e. using an internal store of processing equipment to receive or retain programs
- G06F9/46—Multiprogramming arrangements
- G06F9/54—Interprogram communication
-
- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F2111/00—Details relating to CAD techniques
- G06F2111/10—Numerical modelling
Definitions
- the present invention relates generally to the calculation of radiation field distributions, and is particularly useful in predicting neutron-dosimetry responses for nuclear reactor cavities and internal components.
- LBE Linearized Boltzmann Equation
- SN discrete ordinates method
- the numerical solution of the SN equations is achieved through the concurrent discretization of the phase space, i.e., angular, spatial and energy domains.
- the concurrent discretization of the phase space leads to a large number of unknowns in the SN equations and therefore, extensive computational resources are required to solve this problem.
- a method for calculating a radiation field distribution comprising applying a 3-D radiation transport computer code, said code comprising domain decomposition algorithms including domains selected from the group consisting of angular and spatial domains, wherein said domains are allocated and processed independently on a multi-processor computer architecture.
- a computer program for calculating a radiation field distribution including a code segment that when executed divides the angular and spatial domains into subsets, independently allocates the subsets, and processes the subsets on a multi-processor architecture.
- Figure Ia shows the geometry and material distribution of a 3-D transport model for a 2-loop PWR.
- Figure 2 shows the measured versus calculated (M/C) ratios of the dosimetry data calculated using the Directional Theta Weighted adaptive differencing scheme.
- Figures 3a, 3b and 3c show the M/C ratios for a corrected reactor pressure vessel having a corrected thickness as compared to a non-corrected thickness for the core top location (3a), the core mid-plane location (3b) and the core bottom location (3c).
- Figure 4 shows the speed-up obtained for a range of processors (e.g., up to
- Figure 5 shows a flow chart of an embodiment of the present invention, wherein the 3-D radiation transport computer program is implemented to generate a radiation field distribution.
- the present invention relates to a method of calculating radiation field distribution for a system.
- the radiation fields can include neutron and gamma radiation.
- ex-vessel neutron dosimetry responses in the cavity of a nuclear reactor can be calculated.
- the type of nuclear reactor is not limiting and can include a variety of commercial designs known in the art. Suitable reactors can include but are not limited to Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). For simplicity of disclosure, this aspect of the invention will be described with reference to a 2-loop commercial PWR. Computer modeling of the PWR to generate dosimetry responses are used in the design and operation of the PWR.
- the method of the present invention includes application of a 3-D parallel radiation transport code referred to herein as RAPTOR-M3G (RApid Parallel Transport
- the transport code provides a set of parallel algorithms for solving the SN equations.
- the method is based on domain decomposition algorithms, where the spatial, angular and/or energy domains are partitioned into subsets which can be independently allocated and processed on a multi-processor architecture.
- suitable 3-D parallel deterministic transport codes which are known include PENTRANTM and PARTISN (Sjoden G. E. and Haghighat A., "PENTRAN - Parallel Environment Neutralparticle TRANport in 3-D Cartesian Geometry," Proceedings of the Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications, Vol. 1, pp. 232-234, Saratoga Springs, NY (1997)).
- the method of the present invention reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems.
- Figure Ia shows the geometry of a 3-D transport model for an embodiment of the present invention, e.g., a 2-loop PWR.
- the PWR can include a 12- foot nuclear core, thermal shield design, and a 3-inch reactor cavity air gap.
- the model geometry includes a core-water mixture, core shroud, core barrel, thermal shield, Reactor Pressure Vessel (RPV) including stainless-steel liner, and reflective insulation.
- the RPV of a PWR typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head.
- the upper and lower internals regions above and below the reactor core are modeled using a steel-water mixture.
- the lower internals of the RPV include the core barrel (i.e., a core support structure).
- the core barrel is enclosed with the thermal shield between the core barrel and an inner wall of the RPV.
- neutron pads are used in lieu of the thermal shield.
- the core shroud sets inside the core barrel.
- An annular downcomer surrounds the reactor core barrel. Cooling fluid, typically water, is circulated into the downcomer.
- the model geometry and the mesh discretization are generated using the BOT3P code, version 5.2.
- the model extends from 0.0 cm to 245.0 cm along the x-, and y-axis, and from -200.0 cm to 200.0 cm along the z-axis.
- a uniform mesh is applied throughout the model; a mesh size of 2.0 x 2.0 x 4.0 cm is specified along the x-, y- and z-axes, respectively, yielding a total of 1,464,100 meshes.
- the cross sections for the material mixtures in the transport model are processed using the BUGLE-96 cross sections library (RSICC Data Library Collection BUGLE-96, "Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Oak Ridge National Laboratory, Oak Ridge, TN (1999)) and the GIP computer code, part of the DOORS package (RSICC Computer Code Collection DOORS 3.2a, "One-, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," Oak Ridge National Laboratory, Oak Ridge, TN (2003)).
- An Sg level symmetric quadrature set and a P 3 spherical harmonics expansion of the scattering kernel is used for the transport calculations.
- a system of passive neutron detectors can be installed in the reactor cavity air gap between the reflective insulation and the pressure vessel.
- the dosimetry system can provide accurate information relative to the fast neutron exposure over the beltline region of the reactor vessel.
- Pure metal foils can be installed in the reactor cavity, encased in an aluminum shell, which minimizes distortions of the fast neutron spectrum, effectively yielding a free-field measurement.
- the neutron dosimeters installed in the reactor cavity air gap are not explicitly defined in the transport model.
- An aspect of the present invention includes a domain decomposition algorithm(s) for the discretization of the SN equations that uses an approach wherein the spatial and/or angular energy domains are partitioned into subsets which can be independently allocated and processed on multi-processor architectures.
- phase space of the SN equations is discretized, i.e., angle, space and energy.
- the resulting set of linear algebraic equations is suitable for solution on a digital computer.
- the transport equation (i.e., LBE) in the multigroup approximation is formulated in Eq. (1).
- the angular domain is discretized by considering a finite set of directions and by applying an appropriate quadrature integration scheme.
- Each discrete direction can be visualized as a point on the surface of a unit sphere with an associated surface area which mathematically corresponds to the weight of the quadrature scheme.
- the combination of the discrete directions and the corresponding weights is referred to as quadrature set.
- quadrature sets must satisfy a number of conditions in order to be accurate and mathematically determined; several approaches can be employed, e.g., level-symmetric quadrature set (LQn) and Legendre polynomial based quadrature sets (Longoni G.
- the quadrature sets developed and implemented in RAPTOR-M3G are based on the LQn method.
- the spatial variable can be discretized with several techniques, e.g., finite difference and finite element methods.
- the formulation developed in RAPTOR-M3G is based on the finite difference approach which includes partitioning the spatial domain into computational cells, e.g., fine meshes, where the cross sections are assumed constant within each cell. In 3D Carstesian geometry, the angular flux at the cell-center location is evaluated using Eq. (2).
- the angle and energy dependence are denoted by the indices m and g, respectively.
- the term q ⁇ k represents the sum of the scattering, fission and external sources at cell-center.
- the indices i, j, k represent the cell-center values, and the weights ajj,k,m,g, bij, k , m ,g, an ⁇ c ij,k,m,g are restricted to the range between 0.5 and 1.0;
- RAPTOR-M3G utilizes the Theta- Weighted (TW), Zero- Weighted (ZW), or the adaptive Directional Theta Weighted (DTW) differencing schemes to calculate the weights during the transport sweep (B.
- TW Theta- Weighted
- ZW Zero- Weighted
- DTW adaptive Directional Theta Weighted
- the S N equations are solved by going through each direction starting from the boundary of the problem domain; this solution process is also referred to as transport sweep.
- the angular flux defined at center-cell locations is evaluated starting from boundary conditions or from the boundary angular flux previously calculated in adjacent cells.
- the cell-center angular flux is calculated using Eq. (2).
- the angular flux exiting the computational cell is calculated using additional relationships referred to as the "differencing schemes".
- the transport sweep is performed within an iterative process which is termed source iteration, also known as fixed point iteration, or Richardson iteration. This process is continued until an appropriate convergence criterion is satisfied, i.e., the relative error on the scalar flux in any norm between two iterations is below a certain cutoff value (Adams M. L. and Larsen E. W., "Fast Iterative Methods for Discrete- Ordinates Particle Transport Calculations," Progress in Nuclear Energy, Vol. 40, n. 1
- RAPTOR-M3G The parallel algorithms developed in RAPTOR-M3G are based on the decomposition of the angular and/or spatial domains on a network of processors.
- RAPTOR-M3G creates a virtual topology based a number of processors allocated to the angular and spatial domains, specified as P a and P s respectively.
- the network of processors is mapped on the spatial and angular domains creating a virtual topology which associates each processor to its local sub-domain.
- the angular domain is partitioned on an octant basis, where the processors specified on the angular domain, are sequentially assigned to the local octants.
- the local number of octants allocated per processor is given by Eq. (3).
- the transport sweep is locally performed on Ni oct octants on P a processors; an MPI communicator for the angular domain is used to synchronize the angular flux among processors and to account for reflective boundary conditions.
- the spatial domain is partitioned along the z-axis by sequentially assigning the P s processors to a number of x-y planes.
- the total number of fine meshes along the z-axis, i.e., km, is partitioned on P s processors; a mapping array, i.e., kmloc, is used to assign the x-y planes to the P s processors.
- the number of x-y planes assigned to the P s processors is arbitrary; however, the condition in Eq. (4) needs to be satisfied in order to define a spatial decomposition that is topologically consistent with the problem geometry.
- the flexibility to map the processors on the spatial domain to an arbitrary number of x-y planes can depend on the fact that the number of z-planes may not be exactly divisible by the number of processors on the spatial domain.
- An uneven partitioning of the x-y planes on the P s processors can lead to processor load imbalance with consequent loss in performance.
- a hybrid angular/spatial decomposition strategy may be applied to overcome this difficulty.
- the hybrid decomposition includes the combination of the angular and spatial domains to include concurrent partitioning of these domains. Hybrid domain decomposition is further described in the Example 1 below.
- Figure 5 shows a flow chart of an embodiment of the present invention, wherein the 3-D radiation transport computer program is implemented to generate a radiation field distribution.
- This embodiment includes obtaining geometrical and material information on the system to be modeled. The information can be obtained from various sources such as but not limited to nuclear reactor drawings. An adequate S N order required for the calculation can then be selected. Further, the P n expansion order to the angular domain (i.e., scattering kernel and angular flux) can be selected. An appropriate differencing scheme (i.e., TW, ZW or DTW) can be selected as well.
- a 3-D model of the system can be generated and discretized using an appropriate mesh generator for the Cartesian XYZ or RTZ geometries.
- a cross section table is generated for each material by mixing them with an appropriate data set such as BUGLE-96.
- the number of processors and the corresponding decomposition strategy is selected for the problem to be solved.
- the calculation is then performed. Following calculation, postprocessing and analysis of the results generated may occur.
- the RAPTOR-M3G computer code used in the present invention provides an accurate and efficient (e.g., reduced computational time) solution of radiation transport problems.
- ex-vessel neutron dosimetry responses are calculated for a cavity in a nuclear reactor vessel.
- the fast neutron reactions in the reactor cavity air gap of a 2-loop PWR calculated by RAPTOR-M3G were 96% accurate on average.
- solution of the transport problem was obtained in approximately 106 minutes of clock time on a 20- processor computer cluster using a hybrid angular/spatial domain decomposition strategy.
- an aspect of the present invention described herein relates to the nuclear industry and in particular, nuclear reactors.
- the present invention can also be used in a wide range of other applications such as the medical field.
- the present invention can be used to determine the dose of radiation delivered to a patient for the treatment and/or cure of cancer. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
- RAPTOR-M3G running on a 20 processors computer cluster, i.e., EAGLE-I.
- the cluster was composed of 5 nodes with 2 dual-core dual processor AMD Opteron 64-bit architecture.
- the cluster total memory, i.e., RAM, available was 40 GByte; the network interconnection was characterized by 1 GBhVs bandwidth.
- RAPTOR-M3G completed a full 3D transport calculation for a 2-loop PWR in approximately 106 minutes on 20 processors. No significant differences in performance were observed using the DTW, TW, or ZW differencing schemes.
- the over-prediction at the 30° and 45° azimuthal positions could be reduced by using a non-uniform mesh refinement at these locations, where the curvature of the system becomes more relevant.
- the average M/C ratio over all dosimetry locations was 0.96.
- Figures 3a through 3c present the M/C values obtained with the ISI- corrected RPV thickness as compared to the value obtained without the thickness correction. The comparison was conducted and presented for all the dosimetry specimens located at the 0° azimuthal location. The corrected RPV thickness using the ISI measurements improved the accuracy of the calculated responses at every dosimetry location. Similar results were obtained also at 15°, 30°, and 45° azimuthal locations.
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Abstract
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US97552507P | 2007-09-27 | 2007-09-27 | |
US5678508P | 2008-05-28 | 2008-05-28 | |
PCT/US2008/077625 WO2009042747A1 (en) | 2007-09-27 | 2008-09-25 | Reactor dosimetry applications using a parallel 3-d radiation transport code |
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EP2215617A1 true EP2215617A1 (en) | 2010-08-11 |
EP2215617A4 EP2215617A4 (en) | 2014-11-19 |
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EP08833567.4A Ceased EP2215617A4 (en) | 2007-09-27 | 2008-09-25 | Reactor dosimetry applications using a parallel 3-d radiation transport code |
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EP (1) | EP2215617A4 (en) |
JP (1) | JP5460601B2 (en) |
KR (1) | KR101545035B1 (en) |
CN (1) | CN101861609B (en) |
WO (1) | WO2009042747A1 (en) |
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JP5289542B2 (en) * | 2011-11-25 | 2013-09-11 | 日立Geニュークリア・エナジー株式会社 | Air dose evaluation apparatus and method |
JP2013186016A (en) * | 2012-03-09 | 2013-09-19 | Hitachi-Ge Nuclear Energy Ltd | Method of measuring three-dimensional space distribution of radioactive ray data |
CN104678424B (en) * | 2013-12-03 | 2017-08-25 | 中广核(北京)仿真技术有限公司 | The neutron counting system and method for nuclear power plant analog machine |
CN106126929B (en) * | 2016-06-24 | 2018-10-19 | 西安交通大学 | Method based on the extensive interior vacuum particle transport issues of DISORT method processing |
US11060987B2 (en) | 2019-04-23 | 2021-07-13 | Canon Medical Systems Corporation | Method and apparatus for fast scatter simulation and correction in computed tomography (CT) |
CN110502813B (en) * | 2019-08-09 | 2023-04-18 | 中国舰船研究设计中心 | Nuclear power ship reactor cabin mixed radiation field calculation method based on radiation source discretization |
KR102374327B1 (en) * | 2019-10-08 | 2022-03-15 | 주식회사엔에스이 | A 3D System for Estimating Radioactivity of Equipment Elements and Representing Radiation Maps in Nuclear Power Plants |
CN112346873B (en) * | 2020-11-26 | 2022-02-11 | 中国核动力研究设计院 | Characteristic line method multistage parallel method suitable for hardware architecture of modern supercomputer |
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US4666662A (en) * | 1984-07-10 | 1987-05-19 | Westinghouse Electric Corp. | Steam generator recirculating system for a pressurized water nuclear reactor |
US5267276A (en) * | 1991-11-12 | 1993-11-30 | The University Of Chicago | Neutron transport analysis for nuclear reactor design |
US7233888B2 (en) * | 2002-07-09 | 2007-06-19 | General Electric Company | Monte Carlo criticality-mode systems and methods for computing neutron and gamma fluence in a nuclear reactor |
TWI274354B (en) * | 2002-09-03 | 2007-02-21 | Gen Electric | Methods and apparatus for determining neutron flux in a nuclear reactor |
US8109766B2 (en) * | 2003-10-03 | 2012-02-07 | Global Nuclear Fuel-Americas Llc | Method for predicted reactor simulation |
WO2006138513A1 (en) * | 2005-06-16 | 2006-12-28 | Nomos Corporation | Variance reduction simulation system, program product, and related methods |
JP2007278991A (en) * | 2006-04-11 | 2007-10-25 | Nuclear Fuel Ind Ltd | Control method for three-dimensional mesh system for parallel calculation |
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- 2008-09-25 JP JP2010527130A patent/JP5460601B2/en active Active
- 2008-09-25 CN CN2008801166435A patent/CN101861609B/en active Active
- 2008-09-25 WO PCT/US2008/077625 patent/WO2009042747A1/en active Application Filing
- 2008-09-25 EP EP08833567.4A patent/EP2215617A4/en not_active Ceased
- 2008-09-25 KR KR1020107009279A patent/KR101545035B1/en active IP Right Grant
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CN101861609B (en) | 2013-12-18 |
JP2010540936A (en) | 2010-12-24 |
EP2215617A4 (en) | 2014-11-19 |
KR101545035B1 (en) | 2015-08-17 |
CN101861609A (en) | 2010-10-13 |
WO2009042747A1 (en) | 2009-04-02 |
KR20100066572A (en) | 2010-06-17 |
JP5460601B2 (en) | 2014-04-02 |
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