Burn-Up Dependency of Control Rod Position at Zero-Power Criticality in the High-Temperature Engineering Test Reactor

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR), which was constructed in Japan. The operating data of HTTR with burn-up to about 370 EFPD (effective full-power days), which are very important for the development of HTGRs, have been collected in both zero-power and powered operations. In the aspects of code validation, the detailed prediction of temperature distribution in the core makes it difficult to validate the calculation code because of difficulty in measuring the core temperature directly in powered operation of the HTTR. In this study, the measured data of the control rod position, while keeping the temperature distribution in the core uniform at criticality in zero-power operation at the beginning of each operation cycle were compared with the calculated results by core physics design code of the HTTR. The measured data of the control rod position were modified based on the core temperature correlation. At the beginning of burn-up, the trends of burn-up characteristics are slightly different between experimental and calculation data. However, the calculated result shows less than 50 mm of small difference (total length of control rod is 4060 mm) to the measured one, which indicates that the calculated results appropriately reproduced burn-up characteristics, such as a decrease in uranium-235, accumulation in plutonium, and decrease in burnable absorber.

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Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4033813 ◽

2016 ◽

Vol 3 (1) ◽

Author(s):

Yuki Honda ◽

Nozomu Fujimoto ◽

Hiroaki Sawahata ◽

Shoji Takada ◽

Kazuhiro Sawa

Keyword(s):

High Temperature ◽

Cross Section ◽

Flux Distribution ◽

Cell Model ◽

Control Rod ◽

The Core ◽

Macroscopic Cross Section ◽

Conventional Cell ◽

Test Reactor ◽

Control Rods

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

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PRELIMINARY ANALYSIS OF CORE TEMPERATURE DISTRIBUTION OF EXPERIMENTAL POWER REACTOR USING RELAP5

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2018.20.3.4665 ◽

2018 ◽

Vol 20 (3) ◽

pp. 159

Author(s):

Andi Sofrany Ekariansyah ◽

Surip Widodo ◽

Hendro Tjahjono ◽

Susyadi Susyadi ◽

Puradwi Ismu Wahyono ◽

...

Keyword(s):

High Temperature ◽

Temperature Distribution ◽

Core Temperature ◽

Power Reactor ◽

Pebble Bed ◽

Pebble Bed Reactor ◽

The Core ◽

Detail Design ◽

Core Section ◽

High Temperature Gas

High Temperature Gas Cooled Reactor (HTGR) is a high temperature reactor type having nuclear fuels formed by small particles containing uranium in the core. One of HTGR designs is Pebble Bed Reactor (PBR), which utilizes helium gas flowing between pebble fuels in the core. The PBR is also the similar reactor being developed by Indonesia National Nuclear Energy Agency (BATAN) under the name of the Reaktor Daya Eksperimental (RDE) or Experimental Power Reactor (EPR) started in 2015. One important step of the EPR program is the completion of the detail design document of EPR, which should be submitted to the regulatory body at the end of 2018. The purpose of this research is to present preliminary results in the core temperature distribution in the EPR using the RELAP5/SCDAP/Mod3.4 to be complemented in the detail design document. Methodology of the calculation is by modelling the core section of the EPR design according to the determined procedures. The EPR core section consisting of the pebble bed, outlet channels, and hot gas plenum have been modelled to be simulated with 10 MWt. It shows that the core temperature distribution under assumed model of 4 core zones is below the limiting pebble temperature of 1,620 °C with the highest pebble temperature of 1,477.0 °C. The results are still preliminary and requires further researches by considering other factors such as more representative radial and axial power distribution, decrease of core mass flow, and heat loss to the reactor pressure vessel.Keywords: Pebble bed, core temperature, EPR, RELAP5 ANALISIS AWAL DISTRIBUSI TEMPERATUR TERAS REAKTOR DAYA EKSPERIMENTAL MENGGUNAKAN RELAP5. High Temperature Gas Cooled Reactor (HTGR) adalah reaktor tipe temperatur tinggi yang memiliki bahan bakar nukir dalam bentuk bola-bola kecil yang mengandung uranium. Salah satu desain HTGR adalah reaktor pebble bed (Pebble bed reactor/PBR) yang memanfaatkan gas helium sebagai pendingin yang mengalir di celah-celah bahan bakar bola di dalam teras. PBR juga merupakan tipe reaktor yang sedang dikembangkan oleh BATAN dengan nama reaktor daya eksperimental (RDE) yang dimulai pada 2015. Salah satu tahapan penting dalam program RDE adalah penyelesaian dokumen desain rinci yang harus dikirimkan ke badan pengawas pada akhir 2018. Tujuan penelitian adalah untuk menyajikan hasil-hasil awal pada distribusi temperatur di teras RDE menggunakan RELAP5/SCDAP/Mod3.4 sehingga dapat melengkapi isi dokumen desain rinci. Metode perhitungan adalah dengan memodelkan bagian teras RDE sesuai hasil penelitian sebelumnya. Bagian teras RDE yang dimodelkan terdiri dari pebble bed, kanal luaran, dan plenum gas bawah yang disimulasikan pada daya 10 MWt. Hasil simulasi menunjukkan bahwa distribusi temperatur teras dengan asumsi pembagian 4 zona teras mendapatkan temperatur tertinggi sebesar 1477 °C yang masih di bawah batasan temperatur di bola bahan bakar yaitu 1620 °C. Hasil yang diperoleh masih estimasi awal dan membutuhkan penelitian lebih lanjut dengan mempertimbangkan faktor-faktor lainnya seperti distribusi daya aksial dan radian yang lebih representatif, pengurangan aliran teras, dan kehilangan panas teras yang diserap oleh bejana reaktor.Kata kunci: Pebble bed, temperatur teras, RDE, RELAP5

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Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

Nuclear Engineering and Design ◽

10.1016/s0029-5493(97)00030-7 ◽

1997 ◽

Vol 172 (1-2) ◽

pp. 93-102 ◽

Cited By ~ 5

Author(s):

Y. Tachibana ◽

S. Shiozawa ◽

J. Fukakura ◽

F. Matsumoto ◽

T. Araki

Keyword(s):

High Temperature ◽

High Temperatures ◽

Control Rod ◽

Integrity Assessment ◽

Test Reactor ◽

Very High

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Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4032873 ◽

2016 ◽

Vol 2 (3) ◽

Cited By ~ 12

Author(s):

Kaichao Sun ◽

Lin-Wen Hu ◽

Charles Forsberg

Keyword(s):

High Temperature ◽

Peak Power ◽

Fuel Cycle ◽

Thermal Power ◽

Combined Cycle ◽

Fluoride Salt ◽

Control Rod ◽

The Core ◽

Liquid Salt ◽

High Temperature Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature tristructural isotropic (TRISO) particle fuel. The refractory TRISO particle coating system and the dispersion in graphite matrix enhance safeguards (nuclear proliferation resistance) and security. Compared to the conventional high-temperature reactor (HTR) cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a nuclear air-Brayton combined cycle (NACC) that provides base-load and peak-power capabilities. Added peak power is produced using jet fuel or locally produced hydrogen. The FHR is, therefore, considered as an ideal candidate for the transportable reactor concept to provide power to remote sites. In this context, a 20-MW (thermal power) compact core aiming at an 18-month once-through fuel cycle is currently under design at Massachusetts Institute of Technology (MIT). One of the key challenges of the core design is to minimize the reactivity swing induced by fuel depletion, since excessive reactivity will increase the complexity in control rod design and also result in criticality risk during the transportation process. In this study, burnable poison particles (BPPs) made of B4C with natural boron (i.e., 20% B10 content) are adopted as the key measure for fuel cycle optimization. It was found that the overall inventory and the individual size of BPPs are the two most important parameters that determine the evolution path of the multiplication factor over time. The packing fraction (PF) in the fuel compact and the height of active zone are of secondary importance. The neutronic effect of Li6 depletion was also quantified. The 18-month once-through fuel cycle is optimized, and the depletion reactivity swing is reduced to 1 beta. The reactivity control system, which consists of six control rods and 12 safety rods, has been implemented in the proposed FHR core configuration. It fully satisfies the design goal of limiting the maximum reactivity worth for single control rod ejection within 0.8 beta and ensuring shutdown margin with the most valuable safety rod fully withdrawn. The core power distribution including the control rod’s effect is also demonstrated in this paper.

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Characteristic Tests on Transition Core of HTR-10

Volume 1: Operations and Maintenance, Engineering, Modifications, Life Extension, Life Cycle, and Balance of Plant; Instrumentation and Control (I&C) and Influence of Human Factors; Innovative Nuclear Power Plant Design and SMRs ◽

10.1115/icone26-81797 ◽

2018 ◽

Author(s):

Liqiang Wei ◽

Dongmei Ding ◽

Ling Liu ◽

Yucheng Wang ◽

Xiaoming Chen ◽

...

Keyword(s):

High Temperature ◽

Control Rod ◽

Safety Requirements ◽

Operation Characteristic ◽

Power Operation ◽

Test Reactor ◽

High Temperature Gas

After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.

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Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54041 ◽

2012 ◽

Author(s):

Yanhua Zheng ◽

Lei Shi ◽

Fubing Chen

Keyword(s):

High Temperature ◽

Temperature Distribution ◽

Normal Operation ◽

Helium Temperature ◽

Fuel Temperature ◽

Pebble Bed ◽

Decay Heat ◽

The Core ◽

Core Cooling ◽

High Temperature Gas

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.

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Experience of Xenon Oscillation During an Initial Physics Test of UCN Unit 5 in Korea

ASME 2010 Pressure Vessels and Piping Conference: Volume 9 ◽

10.1115/pvp2010-25237 ◽

2010 ◽

Author(s):

Yong Rae Kim ◽

Tae Young Choi ◽

Sun Ho Shin ◽

Ki Bong Seong

Keyword(s):

Steady State ◽

Annual Cycle ◽

Stability Index ◽

Control Rod ◽

The Core ◽

Axial Stability ◽

Xenon Oscillation ◽

Burnable Absorber ◽

Control Rods ◽

Initial Cycle

Initial core of Ulchin Nuclear Unit 3 (UCN3), which is one of earlier OPR1000 model, was 4 batches and designed as annual cycle after second cycle. The utility requested that UCN Unit 5 (UCN5), which is another of OPR1000 model, had capability of a longer cycle operation from second cycle. Therefore, KNF modified the number of batches from 4 to 3 for OPR1000 initial core, as well as, the number of burnable absorber, and the cutback length of the absorber. However, due to these changes, Xenon oscillation was slightly increased at 100% power during the physics test of UCN5, while that oscillation at 100% power in UCN3 had been gone down without any control rod motion. The xenon oscillation direction is related to axial stability index. The index of UCN3 increased from a slightly negative at BOC to positive at EOC, the index of UCN5 was positive even at BOC, which meant that the core does not go to be stable without the control rod motion. The core of UCN5 became the steady state by the insertion of control rods into the core. To meet the physics test condition, the oscillation was controlled by control rods immediately. After the happening, KNF optimized the cutback length of burnable absorber rods and applied to APR1400, which will keep being stable in xenon oscillation during physics test at the initial cycle.

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Thermal Analysis of a PWR Core Internal Baffle Structure

Volume 4: Fluid Structure Interaction, Parts A and B ◽

10.1115/pvp2006-icpvt-11-93299 ◽

2006 ◽

Cited By ~ 3

Author(s):

Christophe Pe´niguel ◽

Isabelle Rupp ◽

Nathalie Ligneau ◽

Michel Tommy-Martin ◽

Laurent Beloeil ◽

...

Keyword(s):

High Temperature ◽

Temperature Distribution ◽

Mesh Size ◽

Internal Heating ◽

The Core ◽

Internal Core ◽

Element Approach ◽

Temperature Levels ◽

Set Up ◽

Reactor Center

The internal core baffle structure of a PWR consists in baffles and formers attached to the barrel. Each baffle being independent, the connection between the core baffle sheets, the formers and the core barrel is done thanks to a large number of bolts (about 1500). After inspection, some baffle bolts have been found cracked. This behaviour is attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). In order to compute accurately the temperature distribution affecting these bolts, EDF has set up a research program. Due to symmetry reasons, only a 45° sector has been accounted for. The three-dimensionnal neutron flux and the gamma induced internal heating are calculated with a Monte-Carlo particle transport code named Tripoli-4. The by-pass flow inside the cavities is computed with the CFD code Code_Saturne with a finite volume technique. Finally, the temperature distribution inside the structure (including all bolts which leads to a considerable solid mesh size — about 236 millions tetraedra) is computed by the thermal code Syrthes using a finite element approach, taking into account both the heating due to the gamma heating deposit and the cooling by the by-pass flow. Calculations show that the solid thermal field obtained exhibit strong temperature gradients and high temperature levels but in very limited zones located inside the material. As expected mainly very limited regions located inside the material and near the corner close to the reactor center are exposed to high temperature levels. On the other hand, calculations clearly confirm that external bolts thightening the core barrel and the formers see temperature much lower than those thightening the baffles.

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