Flux Rate Calculation and Analysis of the Integrated Small Pressurized Water Reactor Based on Monte Carlo Method

2021 ◽  
Author(s):  
Wen Yang ◽  
Fei Chao ◽  
Yun Tai ◽  
Longze Li
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Pressure Vessel ◽  
Normal Operation ◽  
Photon Flux ◽  
Flux Rate ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Inner Surface ◽  
Rate Calculation

Abstract The neutron and photon flux rates are important parameters for safe reactor operation, refueling and decommissioning, scientific applications and radiation protection. For the Integrated small pressurized water reactor, the advanced reactor core analysis code CORAL, the source calculation code ORIGEN-II and the Monte Carlo code SuperMC are used to establish the reactor flux rate calculation model under normal operation and shutdown refueling condition. The results show that (1) In the normal operation of the reactor, the neutron flux rate is attenuated by 10 orders of magnitude from the outermost component to the inner surface of the pressure vessel, and the shielding effect of the coolant on neutrons is more significant. The neutron flux of the inner surface of the pressure vessel in 40 years is 3.723 × 1014 ncm2; the neutron flux in 60 years is 5.585 × 1014 ncm2. The photon flux rate is reduced by 10 orders of magnitude from the periphery of the core to the outer surface of the pressure vessel. High-quality density materials have better photon shielding effects. (2) In the case of reactor shutdown and refueling, the neutron flux rate is much smaller than the photon flux rate. On the outer surface of the pressure vessel, the maximum neutron and photon dose rates are 7.74 × 10−10 mSv · h−1 and 6.97 × 10−5 mSv · h−1, respectively, which belong to the supervision area. When the cover is opened, the radiation dose value of the workplace at the top of the reactor is less than 0.0025 mSv · h−1, which can ensure the radiation safety of the operation.

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

Kinetic Monte Carlo simulations of cascades in Fe alloys

2000 ◽  
Vol 650 ◽  
Author(s):  
C. Domain ◽  
C.S. Becquart ◽  
J.C. van Duysen
Keyword(s):  
Monte Carlo ◽  
Kinetic Monte Carlo ◽  
Md Simulations ◽  
Atom Probe ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Displacement Cascades ◽  
Cu Content ◽  
Primary Damage ◽  
Eam Potential

ABSTRACTThe Pressurized Water Reactor vessel steels are embrittled by neutron irradiation. Among the solute atoms, copper play an important role in the embrittlement and different Cu-rich defects have been experimentally observed to form. We have investigated by Kinetic Monte Carlo (KMC) on rigid lattices the evolution of the primary damage. Since the point defects created by the displacement cascades have very different kinetics, their evolution is tracked in two steps. In a first step, we have studied their recombination in the cascade region and the formation of interstitial clusters using “object diffusion”. The parameters of this model are based on MD simulations, or on first principles calculations. In a second part, we have investigated the subsequent evolution of the primary damage with a model based on a vacancy jump mechanism. These simulations which rely on an adapted EAM potential show the formation of copper rich defects. Some of the potential's predictions that played a key role in the model were checked by ab initio calculations. The defects obtained from these simulations, subsequent to the primary damage created by displacement cascades, exhibit similarities with the ones observed by atom probe. The influence of temperature and Cu content on the final damage was investigated.

Application of Monte Carlo and discrete ordinate coupling method to pressurized water reactor cavity radiation streaming calculation

2012 ◽  
Vol 24 (12) ◽  
pp. 2946-2950
Author(s):  
郑征 Zheng Zheng ◽  
吴宏春 Wu Hongchun ◽  
曹良志 Cao Liangzhi ◽  
郑友琦 Zheng Youqi ◽  
张宏博 Zhang Hongbo ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Coupling Method ◽  
Pressurized Water Reactor ◽  
Discrete Ordinate ◽  
Pressurized Water ◽  
Water Reactor ◽  
Cavity Radiation

Radionuclide Composition of Airborne Discharges of European NPPS With WWER, Pressurized Water Reactor, and Boiling Water Reactor Facilities

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Angelina-Nataliya V. Vukolova ◽  
Andrei A. Rusinkevich
Keyword(s):  
Power Plants ◽  
Boiling Water ◽  
Normal Operation ◽  
Boiling Water Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Operation Conditions ◽  
Discharge Activity ◽  
Radionuclide Composition

Abstract The article presents the analysis of the data on radionuclide composition of airborne discharges of 52 European nuclear power plants (NPPs) with water–water energetic reactor facilities (WWER), pressurized water reactor facilities (PWR), and boiling water reactor facilities (BWR) under normal operation conditions. It contains lists of radionuclides, registered in discharges of researched NPPs, and gives estimation of contributions of radionuclides, forming the discharge, into total activity of discharge and into total effective dose, created by the discharge activity. It was determined that the maximal contribution into discharge activity of all researched NPPs make noble gases, tritium, and carbon-14, while the latter is the main dose-making radionuclide.

Fast Neutron Irradiation Optimization of Thorium-Fueled SCWR Reactor Pressure Vessel

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Petra Pónya ◽  
Gyula Csom ◽  
Sándor Fehér
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Flow Direction ◽  
Reactor Pressure Vessel ◽  
Fast Neutron Flux ◽  
Inner Surface ◽  
Fast Neutron Irradiation ◽  
Reactor Pressure

Abstract Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at ∼500 °C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Thermal Expansion ◽  
Residual Stresses ◽  
Base Metal ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Normal Operation ◽  
Inner Surface ◽  
Reactor Pressure

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

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