Conceptual Core Design of HAPPY200 Reactor

2018 ◽  
Author(s):  
Xiaosheng Li ◽  
Linsen Li ◽  
Lianghui Peng ◽  
Xiaosong Chen ◽  
Zhaocan Meng ◽  
...  
Keyword(s):  
Temperature Coefficient ◽  
Power Distribution ◽  
Reactor Core ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Control Rod ◽  
The Core ◽  
Passive Safety System ◽  
Reactivity Coefficients ◽  
Full Power

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.

scholarly journals SPX Benchmark Part I: Results of Static Neutronics

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk ◽  
Liang Zhang ◽  
Evgeny Nikitin ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Thermal Conditions ◽  
Reaction Rates ◽  
Second Phase ◽  
Reference Solution ◽  
Transient Behaviour ◽  
The Core ◽  
Reactivity Coefficients

Abstract In the paper, the specification of a new neutronics benchmark for a large Sodium cooled Fast Reactor core and results of modelling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and results of the reference solution. Different core geometries and thermal conditions from cold “as fabricated” up to full power were considered. The reference Monte Carlo solution of Serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modelling with seven other solutions using deterministic and Monte Carlo methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for evaluation of transient behaviour of the core.

scholarly journals Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 7377
Author(s):  
Michał Górkiewicz ◽  
Jerzy Cetnar
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Reactor Performance ◽  
Fission Rate ◽  
Control Rod ◽  
Structured Design ◽  
The Core ◽  
Stable Core ◽  
Control Rods ◽  
Burnup Calculation

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

scholarly journals Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

2021 ◽  
Vol 927 (1) ◽  
pp. 012037
Author(s):  
Daddy Setyawan
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
The Core ◽  
Full Power ◽  
Double Heterogeneity ◽  
High Temperature Gas

Abstract In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.

Startup of In-Hospital Neutron Irradiator

2013 ◽  
Author(s):  
YiGuo Li ◽  
Pu Xia ◽  
XiaoBo Wu ◽  
ShuYun Zou ◽  
Dan Peng ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Neutron Beam ◽  
Reactor Core ◽  
Operation Time ◽  
Negative Temperature ◽  
Prompt Gamma ◽  
Control Rod ◽  
The Core ◽  
Full Power ◽  
Thermal Neutron Beam

In-hospital Neutron Irradiator (IHNI) was specially designed for Boron Neutron Capture Therapy (BNCT), the rated power of IHNI is 30kW, corresponding to the neutron flux density 1×1012n·cm−2·s−1 in reactor core. IHNI is an undermoderated reactor of pool-tank type, and UO2 with enrichment of 12.5% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection. On the both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the Prompt Gamma Neutron Activation Analysis (PGNAA). The decay constants and shares of six group of ordinary delayed neutron and nine group of photoneutron were obtained by WIMS code. Based on that, the relationship between the reactivity and the reactor period was calculated through the inhour equation. In this way, the excess reactivity and the reactivity worthies of the components (control rod, water, etc) in the core are obtained by periodic method during the startup of the reactor. The six test experiments were completed during startup, The test results show that the maximum continuous operation time at full power is 12h; the excess reactivity at cold clean state of the core is 4.2mk; The radiation levels at technical rooms are within the specified values at full power operation. When the positive reactivity with 4.2 mk is inserted into the reactor suddenly, the power will be increased to peak power, and then, it will turn to the normal value due to the negative temperature effect, this experiment shows the inherent safety of IHNI.

scholarly journals Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

2014 ◽  
Vol 4 (1) ◽  
pp. 26-35
Author(s):  
Nhi Dien Nguyen ◽  
Ton Nghiem Huynh ◽  
Vinh Vinh Le ◽  
Hai Dang Vo Doan ◽  
Chulgyo Seo ◽  
...  
Keyword(s):  
Power Distribution ◽  
Research Reactor ◽  
Design Parameters ◽  
Reactor Model ◽  
Control Rod ◽  
Reactivity Coefficients ◽  
And Performance ◽  
Fuel Meat ◽  
New Research ◽  
Pool Type

This paper presents some of studied results of a pre-feasibility project on a new research reactor for Vietnam. In this work, two conceptual nuclear designs of 20 MW multi-purpose research reactor have been done. The reference reactor is the light water cooled and heavy water reflected open-tank-in-pool type reactor. The reactor model is based on the experiences from the operation and utilization of the HANARO. Two fuel types, rod and flat plate, with dispersed U3Si2-Al fuel meat are used in this study for comparison purpose. Analyses for the nuclear design parameters such as the neutron flux, power distribution, reactivity coefficients, control rod worth, etc. have been done and the equilibrium cores have been established to meet the requirements of nuclear safety and performance.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

Preliminary Large Break LOCA and SGTR Accident Analysis of the Compact Small Reactor

2017 ◽  
Author(s):  
Mian Xing ◽  
Linsen Li ◽  
Feng Shen ◽  
Xiao Hu ◽  
Zhan Liu ◽  
...  
Keyword(s):  
Cooling System ◽  
Thermal Power ◽  
Reactor Core ◽  
Safety System ◽  
Passive Safety ◽  
Primary System ◽  
The Core ◽  
Passive Safety System ◽  
Safety Margins ◽  
Guideline Values

This paper gives a brief introduction of the Compact Small Reactor (CSR). It is a simplified two-loop reactor with thermal power of 660MW and with compact primary system and passive safety feature. Preliminary safety analysis of the CSR is conducted to evaluate and further optimize the design of passive safety system, especially the passive core cooling system. Large Break Loss Of Coolant Accident (LBLOCA) and Steam Generator Tube Rupture (SGTR) are selected as two reference accidental scenarios. Each scenario is modeled and computed by RELAP5/MOD3.4. For the LBLOCA analysis, a guillotine break happens in the cold leg of the loop containing the core makeup tanks balance lines. The results show certain safety margins from the guideline values, and the passive safety system could supply enough cooling of the core. For the SGTR analysis, the results show the robustness of the design from the safety perspective. It is concluded that the safety systems are capable of mitigating the accidents and protecting the reactor core from severe damage.

Sign in / Sign up

Export Citation Format

Share Document