• Title/Summary/Keyword: Reactor core calculation

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RADIOLOGICAL DOSE ASSESSMENT ACCORDING TO METHODOLOGIES FOR THE EVALUATION OF ACCIDENTAL SOURCE TERMS

  • Jeong, Hae Sun;Jeong, Hyo Joon;Kim, Eun Han;Han, Moon Hee;Hwang, Won Tae
    • Journal of Radiation Protection and Research
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    • v.39 no.4
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    • pp.176-181
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    • 2014
  • The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.

APOLLO2 YEAR 2010

  • Sanchez, Richard;Zmijarevi, Igor;Coste-Delclaux, M.;Masiello, Emiliano;Santandrea, Simone;Martinolli, Emanuele;Villate, Laurence;Schwartz, Nadine;Guler, Nathalie
    • Nuclear Engineering and Technology
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    • v.42 no.5
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    • pp.474-499
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    • 2010
  • This paper presents the mostortant developments implemented in the APOLLO2 spectral code since its last general presentation at the 1999 M&C conference in Madrid. APOLLO2 has been provided with new capabilities in the domain of cross section self-shielding, including mixture effects and transfer matrix self-shielding, new or improved flux solvers (CPM for RZ geometry, heterogeneous cells for short MOC and the linear-surface scheme for long MOC), improved acceleration techniques ($DP_1$), that are also applied to thermal and external iterations, and a number of sophisticated modules and tools to help user calculations. The method of characteristics, which took over the collision probability method as the main flux solver of the code, allows for whole core two-dimensional heterogeneous calculations. A flux reconstruction technique leads to fast albeit accurate solutions used for industrial applications. The APOLLO2 code has been integrated (APOLLO2-A) within the $ARCADIA^{(R)}$ reactor code system of AREVA as cross section generator for PWR and BWR fuel assemblies. APOLLO2 is also extensively used by Electricite de France within its reactor calculation chain. A number of numerical examples are presented to illustrate APOLLO2 accuracy by comparison to Monte Carlo reference calculations. Results of the validation program are compared to the measured values on power plants and critical experiments.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

  • Hyunjoon Jeong;Taewan Kim
    • Nuclear Engineering and Technology
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    • v.56 no.6
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    • pp.2002-2010
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    • 2024
  • Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Application of data-driven model reduction techniques in reactor neutron field calculations

  • Zhaocai Xiang;Qiafeng Chen;Pengcheng Zhao
    • Nuclear Engineering and Technology
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    • v.56 no.8
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    • pp.2948-2957
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    • 2024
  • High-order harmonic techniques can be used to recreate neutron flux distributions in reactor cores using the neutron diffusion equation. However, traditional source iteration and source correction iteration techniques have sluggish convergence rates and protracted calculation periods. The correctness of the implicitly restarted Arnoldi method (IRAM) in resolving the eigenvalue problems of the one-dimensional and two-dimensional neutron diffusion equations was confirmed by computing the benchmark problems SLAB_1D_1G and two-dimensional steady-state TWIGL using IRAM. By integrating Galerkin projection with Proper Orthogonal Decomposition (POD) techniques, a POD-Galerkin reduced-order model was developed and the IRAM model was used as the full-order model. For 14 macroscopic cross-section values, the TWIGL benchmark problem was perturbed within a 20% range. We extracted 100 sample points using the Latin hypercube sampling method, and 70% of the samples were used as the testing set to assess the performance of the reduced-order model The remaining 30% were utilized as the training set to develop the reduced-order model, which was employed to rebuild the TWIGL benchmark problem. The reduced-order model demonstrates good flexibility and can efficiently and accurately forecast the effective multiplication factor and neutron flux distribution in the core. The reduced-order model predicts keff and neutron flux distribution with a high degree of agreement compared to the full-order model. Additionally, the reduced-order model's computation time is only 10.18% of that required by the full-order model.The neutron flux distribution of the steady-state TWIGL benchmark was recreated using the reduced-order model. The obtained results indicate that the reduced-order model can accurately predict the keff and neutron flux distribution of the steady-state TWIGL benchmark.Overall, the proposed technique not only has the potential to accurately project neutron flux distributions in transient settings, but is also relevant for reconstructing neutron flux distributions in steady-state conditions; thus, its applicability is bound to increase in the future.

A SENSITIVITY STUDY ON NEUTRONIC PROPERTIES OF DUPIC FUEL

  • Park, Hangbok;Roh, Gyu-Hog
    • Proceedings of the Korean Nuclear Society Conference
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    • 1998.05a
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    • pp.124-129
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    • 1998
  • A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

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Improvement and verification of the DeCART code for HTGR core physics analysis

  • Cho, Jin Young;Han, Tae Young;Park, Ho Jin;Hong, Ser Gi;Lee, Hyun Chul
    • Nuclear Engineering and Technology
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    • v.51 no.1
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    • pp.13-30
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    • 2019
  • This paper presents the recent improvements in the DeCART code for HTGR analysis. A new 190-group DeCART cross-section library based on ENDF/B-VII.0 was generated using the KAERI library processing system for HTGR. Two methods for the eigen-mode adjoint flux calculation were implemented. An azimuthal angle discretization method based on the Gaussian quadrature was implemented to reduce the error from the azimuthal angle discretization. A two-level parallelization using MPI and OpenMP was adopted for massive parallel computations. A quadratic depletion solver was implemented to reduce the error involved in the Gd depletion. A module to generate equivalent group constants was implemented for the nodal codes. The capabilities of the DeCART code were improved for geometry handling including an approximate treatment of a cylindrical outer boundary, an explicit border model, the R-G-B checker-board model, and a super-cell model for a hexagonal geometry. The newly improved and implemented functionalities were verified against various numerical benchmarks such as OECD/MHTGR-350 benchmark phase III problems, two-dimensional high temperature gas cooled reactor benchmark problems derived from the MHTGR-350 reference design, and numerical benchmark problems based on the compact nuclear power source experiment by comparing the DeCART solutions with the Monte-Carlo reference solutions obtained using the McCARD code.

NTP-ERSN verification with C5G7 1D extension benchmark and GUI development

  • Lahdour, M.;El Bardouni, T.;El Hajjaji, O.;Chakir, E.;Mohammed, M.;Al Zain, Jamal;Ziani, H.
    • Nuclear Engineering and Technology
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    • v.53 no.4
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    • pp.1079-1087
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    • 2021
  • NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

Simulation, design optimization, and experimental validation of a silver SPND for neutron flux mapping in the Tehran MTR

  • Saghafi, Mahdi;Ayyoubzadeh, Seyed Mohsen;Terman, Mohammad Sadegh
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2852-2859
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    • 2020
  • This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

Radiochemical behavior of nitrogen species in high temperature water

  • Young-Jin Kim;Geun Dong Song;Seung Heon Baek;Beom Kyu Kim;Jin Sik Cheon;Jun Hwan Kim;Hee-Sang Shim;Soon-Hyeok Jeon;Hyunmyung Kim
    • Nuclear Engineering and Technology
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    • v.55 no.9
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    • pp.3183-3193
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    • 2023
  • The water radiolysis in-core at light water reactors (LWRs) produces various radicals with other ionic species/molecules and radioactive nitrogen species in the reactor coolant. Nitrogen species can exist in many different chemical forms and recirculate in water and steam, and consequently contribute to what extent the environmental safety at nuclear power plants. Therefore, a clear understanding of formation kinetics and chemical behaviors of nitrogen species under irradiation is crucial for better insight into the characteristics of major radioactive species released to the main steam or relevant coolant systems and eventually development of advanced processes/methodologies to enhance the environmental safety at nuclear power plants. This paper thus focuses on basic principles on electrochemical interaction kinetics of radiolytic molecules and various nitrogen species in high temperature water, fundamental approaches for calculating thermodynamic values to predict their stability and domain in LWRs, and the effect of nitrogen species on crevice chemistry/corrosion and intergranular stress corrosion cracking (IGSCC) susceptibility of structure materials in high temperature water.

Application of the HELIOS-MASTER Code System on the Criticality Analysis for the SMART-P Spent Fuel Storage (SMART연구로 사용후 연료 저장조의 임계해석에 HELIOS-MASTER계산체계의 적용)

  • Kim, Ha-Yong;Koo, Bon-Seung;Kim, Kyo-Youn;Lee, Chung-Chan;Zee, Sung-Quun
    • Journal of Radiation Protection and Research
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    • v.30 no.2
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    • pp.61-67
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    • 2005
  • The criticality analysis method using HELIOS-MASTER code system, which is the nuclear core analysis code system, was developed for the spent fuel storage of SMART-P reactor. We generated the macroscopic cross section of the geometric model with HELIOS and estimated the criticality of the 3-dimensional model with MASTER for SMART-P spent fuel storage. The validity of criticality analysis method for SMART-P spent fuel storage with the HELIOS-MASTER code system by 3-D MCNP calculation was also verified. The result of the criticality analysis with the HELIOS-MASTER code system is more conservative than that with the MCNP and the accuracy of this result is within the range of an allowable error. Because HELIOS-MASTER can perform the 3-D depletion calculation lot a spent fuel storage, it will be useful to perform the criticality analysis including a burnup credit in future.