• Title, Summary, Keyword: Nuclear research reactor

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STATUS OF FACILITIES AND EXPERIENCE FOR IRRADIATION OF LWR AND V/HTR FUEL IN THE HFR PETTEN

  • Bakker Klaas;Klaassen Frodo;Schram Ronald;Futterer Michael
    • Nuclear Engineering and Technology
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    • v.38 no.5
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    • pp.417-422
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    • 2006
  • The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

Vessel failure sensitivities of an advanced reactor for SBLOCA

  • Jhung, Myung Jo;Oh, Chang-Sik;Choi, Youngin;Kang, Sung-Sik
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.185-191
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    • 2020
  • Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×1019 n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

Risk-informed approach to the safety improvement of the reactor protection system of the AGN-201K research reactor

  • Ahmed, Ibrahim;Zio, Enrico;Heo, Gyunyoung
    • Nuclear Engineering and Technology
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    • v.52 no.4
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    • pp.764-775
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    • 2020
  • Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.

Safety Classification of Systems, Structures, and Components for Pool-Type Research Reactors

  • Kim, Tae-Ryong
    • Nuclear Engineering and Technology
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    • v.48 no.4
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    • pp.1015-1021
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    • 2016
  • Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

Measurement of safety rods reactivity worth by advanced source jerk method in HWZPR

  • Nasrazadani, Z.;Ahmadi, A.;Khorsandi, J.
    • Nuclear Engineering and Technology
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    • v.51 no.4
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    • pp.963-967
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    • 2019
  • Accurate measurement of the reactivity worth of safety rods is very important for the safe reactor operation, in normal and emergency conditions. In this paper, the reactivity worth of safety rods in Heavy Water Zero Power Reactor (HWZPR) in the new lattice pitch is measured by advanced source jerk method. The average of the results related to two different detectors is equal to 29.88 mk. In order to verify the result, this parameter was compared to the previously measured value by subcritical to critical approach. Different experiment results are finally compared with corresponding calculated result. Difference between the average experimental and calculated results is equal to 2.2%.

Design of a Nuclear Reactor Controller Using a Model Predictive Control Method

  • Na, Man-Gyun;Jung, Dong-Won;Shin, Sun-Ho;Lee, Sun-Mi;Lee, Yoon-Joon;Jang, Jin-Wook;Lee, Ki-Bog
    • Journal of Mechanical Science and Technology
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    • v.18 no.12
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    • pp.2080-2094
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    • 2004
  • A model predictive controller is designed to control thermal power in a nuclear reactor. The basic concept of the model predictive control is to solve an optimization problem for finite future time steps at current time, to implement only the first optimal control input among the solved control inputs, and to repeat the procedure at each subsequent instant. A controller design model used for designing the model predictive controller is estimated every time step by applying a recursive parameter estimation algorithm. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), was used to verify the proposed controller for a nuclear reactor. It was known that the nuclear power controlled by the proposed controller well tracks the desired power level and the desired axial power distribution.

Investigating Heavy Water Zero Power Reactors with a New Core Configuration Based on Experiment and Calculation Results

  • Nasrazadani, Zahra;Salimi, Raana;Askari, Afrooz;Khorsandi, Jamshid;Mirvakili, Mohammad;Mashayekh, Mohammad
    • Nuclear Engineering and Technology
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    • v.49 no.1
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    • pp.1-5
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    • 2017
  • The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor

  • Torabi, Mina;Lashkari, A.;Masoudi, Seyed Farhad;Bagheri, Somayeh
    • Nuclear Engineering and Technology
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    • v.50 no.7
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    • pp.1017-1023
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    • 2018
  • The measurement and calculation of neutronic parameters in nuclear research reactors has an important influence on control and safety of the nuclear reactor. The power peaking factors, reactivity coefficients and kinetic parameters are the most important neutronic parameter for determining the state of the reactor. The position of the control shim safety rods in the core configuration affects these parameters. The main purpose of this work is to use the MTR_PC package to evaluate the effect of the partially insertion of the control rod on the neutronic parameters at the operating core of the Tehran Research Reactor. The simulation results show that by increasing the insertion of control rods (bank) in the core, the absolute values of power peaking factor, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time decreased. In addition, the results show that the changes of moderator temperature coefficients value versus the control rods positions are very significant. The average value of moderator temperature coefficients increase about 98% in the range of 0-70% insertion of control rods.

Power upgrading of WWR-S research reactor using plate-type fuel elements part I: Steady-state thermal-hydraulic analysis (forced convection cooling mode)

  • Alyan, Adel;El-Koliel, Moustafa S.
    • Nuclear Engineering and Technology
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    • v.52 no.7
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    • pp.1417-1428
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    • 2020
  • The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MWth. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MWth.

Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

  • Kim, Wonkyeong;Lee, Hyun Chul;Pyeon, Cheol Ho;Shin, Ho Cheol;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.48 no.2
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    • pp.304-317
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    • 2016
  • An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.