scholarly journals INSPECTION OF EUROPIUM-BEARING CONTROL ROD ABSORBER OM-2 FROM CORE II AND CORE III OF THE ARMY SM-1A STATIONARY MEDIUM POWER REACTOR.

1972 ◽  
Author(s):  
A. E. Richt ◽  
V. O. Haynes
Keyword(s):  
Power Reactor ◽  
Control Rod ◽  
Stationary Medium

scholarly journals Control Rod Calibration and Worth Calculation for Optimized Power Reactor 1000 (OPR-1000) Using Core Simulator OPR1000

2017 ◽  
Vol 07 (01) ◽  
pp. 15-23 ◽  
Author(s):  
Nguyen An Son ◽  
Nguyen Duc Hoa ◽  
Tran Trung Nguyen ◽  
Tran Quoc Tuan ◽  
Osvaldo Camueje Raul
Keyword(s):  
Power Reactor ◽  
Control Rod

scholarly journals Chain and Sprocket Analysis of Control Rod Drive Mechanism of HTGR Experimental Power Reactor

2019 ◽  
Vol 1198 (2) ◽  
pp. 022053 ◽  
Author(s):  
M. Awwaluddin ◽  
Sri Hastuty ◽  
Z. Petrus ◽  
H. S. Putut ◽  
Krismawan ◽  
...  
Keyword(s):  
Power Reactor ◽  
Control Rod ◽  
Drive Mechanism

scholarly journals Delayed gamma fraction determination in the zero power reactor CROCUS

2021 ◽  
Vol 7 ◽  
pp. 16
Author(s):  
Oskari Pakari ◽  
Tom Mager ◽  
Vincent Lamirand ◽  
Pavel Frajtag ◽  
Andreas Pautz
Keyword(s):  
Gamma Rays ◽  
Power Reactor ◽  
Radiation Transport ◽  
Nuclear Facilities ◽  
Validation Data ◽  
Control Rod ◽  
Analysis Technique ◽  
Gamma Sources ◽  
Detection Capabilities

Gamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.

scholarly journals ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM

2016 ◽  
Vol 19 (2) ◽  
pp. 75
Author(s):  
Syarip, Khoirul Anam, Dwi Priyantoro
Keyword(s):  
High Temperature ◽  
Power Reactor ◽  
Operation Time ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Reactor Operation ◽  
Control Rod ◽  
Negative Reactivity ◽  
Control Rods ◽  
High Temperature Gas

ANALISISPENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. ABSTRAK ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM. Telah dilakukan analisis simulasi pengaturan posisi batang-batang kendali untuk melanjutkan operasi reaktor daya eksperimental (RDE) paska scram setelah beroperasi pada periode waktu tertentu. Pengendalian reaktivitas pada reaktor RDE yang akan dibangun di Indonesia dengan rujukan high temperature gas reactor (HTR) 10 MWt, dilakukan dengan 10  pasang batang-batang kendali atau control rod (CR). Apabila terrjadi kondisi abnormal maka CR secara otomatis akan jatuh tersisip ke dalam reflektor  reaktor sehingga reaktor scram dan berada pada kondisi subkritis. Untuk melanjutkan operasi reaktor pasca scram diperlukan analisis terkait pengaruh reaktivitas negatif dari Xenon dan suhu. Pada makalah ini disajikan hasil simulasi yang dilakukan untuk penentuan posisi CR paling optimum untuk melanjutkan operasi reaktor, menggunakan simulator PCTRAN-HTR. Simulasi dilakukan pada variasi 70%, 85% dan 100% dari tingkat daya penuh dan dengan variasi waktu operasi 50 s, 10.000 s, dan 20.000 s di mana setelah reaktor beroperasi pada tingkat-tingkat daya dan waktu operasi tersebut reaktor mengalami scram. Untuk melanjutkan operasi lagi maka CR harus dinaikkan lagi dan diatur ke posisi tertentu sampai   reaktor mencapai kondisi kritis lagi pada tingkat daya nominal tersebut. Hasil yang telah diperoleh menunjukkan bahwa dengan posisi CR naik 52 % sudah bisa menghasilkan kondisi kritis dan mampu mengatasi reaktivitas negatif peracunan xenon maupun suhu. Kata kunci: RDE, HTR, operasi reaktor, batang kendali, reaktivitas, scram ABSTRACT POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. Analytical study using PC-based simulator has been carried out on control rods position adjustment of the Indonesian experimental power reactor concept or reaktor daya ekperimental (RDE) in a post reactor scram to continue operation after a certain operation period. Reactivity control of the RDE uses 10 pairs of control rods (CRs), which is based on that applied in the high temperature gas reactor (HTR) 10 MW(t). If an abnormal operating condition occurs, these control rods automatically dropped to the reflector that bring the reactor into a scram and subcritical condition. To continue reactor operation after a period of time, the CRs should be withdrawn to achieve recriticality. Prior to any CRs withdrawal, an analysis of negative reactivity effects of Xenon (poissoning) and fuel temperature coefficient should be done. Simulations using PCTRAN-HTR simulator to determine the optimum CRs positions in achieving reactor criticality for continuation of reactor operation is presented in this paper. The simulations were conducted by varying the reactor power levels at 70%, 85% and 100% of full power, respectively. The reactor operation time was varied at 50s, 10000s, and 20000 s prior to the reactor scram. Adjustment of CRs position should be done to continue reactor operation at those nominal power levels by withdrawing the CRs to the proper positions. The simulation results show that recriticality can be achieverd by whitdrawing the CRs 52% of farther and the negative reactivity from xenon poisoning and temperature could be overcome. Keywords : RDE, HTR, reactor operation, control rod, reactivity, scram.

Study on a Small Power Reactor With Compact Pressure Vessel and Natural Circulation

2014 ◽  
Author(s):  
Kenya Takiwaki ◽  
Shungo Sakurai ◽  
Yutaka Takeuchi ◽  
Yasushi Yamamoto
Keyword(s):  
Power Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Cooling Water ◽  
Heat Removal ◽  
Small Power ◽  
Control Rod ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

There is movement which is developing the small reactor for the small electricity grid in place of a big power reactor which requires the high capital cost. This paper introduces a small power reactor whose purpose is to achieve high economic competitiveness and advanced safety. In order to attain high economic competitiveness, it is designed to be small and simple and uses natural circulation and high pressure. A steam generator is integrated into the reactor pressure vessel (RPV), thus dispensing with a primary system and preventing radiation leakage from the reactor core. The small core is designed to have a high power density (100 MW/m3, almost twice that of a conventional boiling water reactor). The concept of a 300 MWt (100 MWe) core design is established by introducing a boiling heat transfer system. By boiling cooling water, the cooling-water circulating flow quantity in a reactor core is enlarged. By increasing a flow, the minimum critical power ratio is improved, which is an important core characteristic. Furthermore, using a burnable poison (Gd2O3), the excess reactivity of a reactor core is reduced and excess reactivity is controlled only by the control rod. Moreover, the maximum linear power density is improved and the critical power ratio is minimized by optimizing the burnable poison arrangement and the control rod pattern. In order to attain high safety, our small reactor has an advanced decay heat removal system that can cool the core without external support. This decay heat removal system is part of the secondary cooling system and combined with a cooling tower. As a result, the quantity of cooling water stored in the decay heat removal system is reduced, and longtime decay heat removal is possible by small equipment.

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