• Title/Summary/Keyword: In-core neutron detector

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Reactor Noise Analyses in Yonggwang 3&4 Nuclear Power Plants (영광 3&4 호기의 원자로잡음신호 해석)

  • Park, Jin-Ho;Ryu, Jeong-Soo;Sim, Woo-Gun;Kim, Tae-Ryong;Park, Jong-Beom
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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    • 2000.06a
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    • pp.679-686
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    • 2000
  • Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

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Measurements of Thermal Neutron Spectrum Parameters in the TRIGA Mark II Reactor

  • Yang, Jae-Choon
    • Nuclear Engineering and Technology
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    • v.11 no.1
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    • pp.21-27
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    • 1979
  • The relative reaction rates were measured in the TRIGA Mark II reactor core and analyzed to obtain the neutron spectrum parameters; relative neutron temperature T$^{n}$ and epithermal index (equation omitted) Measurements were made with the central thimble and the F2 position containing the light water. The relative neutron temperature was represented by the activation ratio of Lu-Mn, and the epithermal index was measured by Au-Mn foil activation. The multichannel analyzer was used to measure the relative ${\gamma}$-rays of the detector foils. The results were compared with the calculated values.

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Research on Mechanical Shim Application with Compensated Prompt γ Current of Vanadium Detectors

  • Xu, Zhi
    • Nuclear Engineering and Technology
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    • v.49 no.1
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    • pp.141-147
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    • 2017
  • Mechanical shim is an advanced technology for reactor power and axial offset control with control rod assemblies. To address the adverse accuracy impact on the ex-core power range neutron flux measurements-based axial offset control resulting from the variable positions of control rod assemblies, the lead-lag-compensated in-core self-powered vanadium detector signals are utilized. The prompt ${\gamma}$ current of self-powered detector is ignored normally due to its weakness compared with the delayed ${\beta}$ current, although it promptly reflects the flux change of the core. Based on the features of the prompt ${\gamma}$ current, a method for configuration of the lead-lag dynamic compensator is proposed. The simulations indicate that the method can improve dynamic response significantly with negligible adverse effects on the steady response. The robustness of the design implies that the method is of great value for engineering applications.

Experimental Determination of Differential Fast Neutron Spectra in a Reactor using Threshold Detectors

  • Kim, Dong-Hoon
    • Nuclear Engineering and Technology
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    • v.4 no.4
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    • pp.280-293
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    • 1972
  • The differential fast neutron spectra above 0.5 Mev at particular spatial positions in tile reactor(TRIGA MARK-II) core has been determined experimentally using several threshold activation detectors. The series expansion technique utilizing the concept of least squares optimization was used to obtain an approximate solution to the set of integral equations which are defined by the experimentally determined activation data. The influence of use of different weighting functions in the solution was analyzed in each measurement. To carry out the necessary mathematical calculations, a computer code for the UNIVAC 1106 digital computer has been prepared. Good agreement was achieved between the differential fast neutron spectra determined in this work and the computed flux determined independently using space-independent multigroup transport theory.

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A complete 3D map of Bell Glasstone spatial correction factors for BRAHMMA subcritical core

  • Shukla, Shefali;Roy, Tushar;Kashyap, Yogesh;Shukla, Mayank;Singh, Prashant
    • Nuclear Engineering and Technology
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    • v.54 no.9
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    • pp.3488-3493
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    • 2022
  • Accelerator driven subcritical systems have long been discussed as facilities which can be used for solving the nuclear waste problem. The physics of these systems is very different from conventional reactors and new techniques had to be developed for reactivity monitoring. One such technique is the Area Ratio Method which studies the response of a subcritical system upon insertion of a large number of neutron pulses. An issue associated with this technique is the spatial dependence of measured reactivity which is intrinsic to the sub criticality of the system since the reactor does not operate on the fundamental mode and measured reactivity depends on the detector position. This is generally addressed by defining Bell-Glasstone spatial correction factor. This factor upon multiplication with measured reactivity gives the correct reactivity which is independent of detector location. Monte Carlo Methods are used for evaluating these factors. This paper presents a complete three dimensional map of spatial correction factors for BRAHMMA subcritical system. In addition, the dataset obtained also helps in identifying detector locations where the correction factor is close to unity, thereby implying no correction if the detector is used at those locations.

MASTER - An Indigenous Nuclear Design Code of KAERI

  • Cho, Byung-Oh;Lee, Chang-Ho;Park, Chan-Oh;Lee, Chong-Chul
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.05a
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    • pp.21-27
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    • 1996
  • KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.

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A Study on the Radioactivity Analysis of Decommissioning Concrete Using Monte Carlo Simulation (Monte Carlo 모사기법을 이용한 해체 콘크리트의 방사능 분석법 연구)

  • 서범경;김계홍;정운수;이근우;오원진;박진호
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2004.06a
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    • pp.43-51
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    • 2004
  • In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1&2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

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A Study on the Effect of Gamma Background in Low Power Startup Physics Tests (저출력 노물리 시험에서의 감마 Background의 영향에 관한 연구)

  • Bae, Chang-Joon;Lee, Ki-Bog
    • Nuclear Engineering and Technology
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    • v.25 no.3
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    • pp.361-370
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    • 1993
  • Low power physics tests should be peformed for the domestic pressurized light water reactors (PWRs) after refueling. The tests are peformed to ensure that operating characteristics of the core are consistent with predictions and that the core can be operated as designed. But in some low power physics tests, slow but steady reactivity increasing phenomena were noticed after step reactivity insertion by the control rod movement. These reactivity increasing phenomena are due to the low flux level and the gamma background because an uncompensated ion chamber (UIC) is used as the ex-core neutron detector. The gamma background may affect the results or the lour power physics tests. The aims or this paper are to analyze the grounds of such phenomena, to simulate a reference bank worth measurement test and to present a resolution quantitatively. In this study, the gamma background level was estimated by numerically solving the point kinetics equations accounting the gamma background effect. The reactivity computer check test was simulated to verify the model. Also, an appropriate neutron flux level was determined by simulating the reference bank worth measurement test. The determined neutron flux level is approximately 0.3 of the nuclear heating flux. This level is about 3 times as high as the current test upper limit specified in the test procedure. Then, the findings from this work were successfully applied to Kori unit 4 cycle 7 and Yonggwang unit 1 cycle 7 physics tests.

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Simulations of BEAVRS benchmark cycle 2 depletion with MCS/CTF coupling system

  • Yu, Jiankai;Lee, Hyunsuk;Kim, Hanjoo;Zhang, Peng;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.52 no.4
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    • pp.661-673
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    • 2020
  • The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.