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.

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 TRANSIENT ANALYSIS OF SIMULTANEOUS LOFA AND RIA IN RSG-GAS REACTOR AFTER 32 YEARS OPERATION

2020 ◽  
Vol 22 (3) ◽  
pp. 111
Author(s):  
Muhammad Darwis Isnaini ◽  
Iman Kuntoro ◽  
Muhammad Subekti
Keyword(s):  
Research Reactor ◽  
Safety Margin ◽  
Low Flow ◽  
Maximum Temperature ◽  
Design Parameters ◽  
Transient Condition ◽  
Trip Time ◽  
Safety Margins ◽  
Fuel Meat ◽  
The Impact

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip.  From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.Keywords: RSG-GAS, Simultaneous, LOFA, RIA, PARET

Control Rod Withdrawal Analysis of the Supercritical Water Reactor-Fuel Qualification Test Facility in the LVR-15 Research Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Csaba Maráczy ◽  
György Hegyi ◽  
István Trosztel ◽  
Emese Temesvári
Keyword(s):  
Supercritical Water ◽  
Research Reactor ◽  
Three Dimensional ◽  
Reactor Fuel ◽  
Test Facility ◽  
Qualification Test ◽  
Control Rod ◽  
Water Reactor ◽  
Reactivity Coefficients ◽  
Supercritical Water Reactor

The aim of the supercritical water reactor-fuel qualification test (SCWR-FQT) Euratom-China collaborative project is to design an experimental facility for qualification of fuel for the supercritical water-cooled reactor. The facility is intended to be operated in the LVR-15 research reactor in the Czech Republic. The pressure tube of the FQT facility encloses four fuel rods that will operate in similar conditions to the evaporator of the HPLWR reactor. This article deals with the three-dimensional (3D) coupled neutronic-thermohydraulic steady-state and transient analysis of LVR-15 with the fueled loop. Conservatively calculated enveloping parameters (e.g., reactivity coefficients) were determined for the safety analysis. The control rod withdrawal analysis of the FQT facility with and without reactor SCRAM was carried out with the KIKO3D-ATHLET-coupled dynamic code.

A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System

2018 ◽  
Author(s):  
Heng Yu ◽  
Guan-bo Wang ◽  
Da-zhi Qian ◽  
Yu-chuan Guo ◽  
Bo Hu
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
Core Damage ◽  
The World ◽  
Safety Features ◽  
Control Rods ◽  
Pool Type

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.

A New Low Enrichment Uranium Core Design of MNSR

2017 ◽  
Author(s):  
Wang Mengjiao ◽  
Li Yiguo ◽  
Wu Xiaobo ◽  
Peng Dan ◽  
Hong Jingyan ◽  
...  
Keyword(s):  
Optimization Design ◽  
Research Reactor ◽  
Nuclear Safety ◽  
Fuel Pellet ◽  
Design Parameters ◽  
Control Rod ◽  
The Core ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Better Than

The Miniature Neutron Source Reactor (MNSR) is a low-power research reactor, which uses 90% high enriched uranium (HEU) fuel. However, due to the nuclear safety risk, and according to the principle of nuclear non-proliferation, MNSR must be gradually converted from HEU to low enriched uranium (LEU), which means the LEU fuel with U-235 enrichment less than 20% should be used. The prototype MNSR of China Institute of Atomic Energy has completed the transformation, but other commercial MNSRs have not finished, which is different with the prototype in the application and structure. Therefore, using MCNP code to simulate, calculate and optimization design LEU core has been done in this issue. Firstly, UO2 with U-235 enrichment of 12.5% was selected as the fuel pellet of LEU core, keeping the rest of the core unchanged. The Φ, excess reactivity and the worth of the central control rod are calculated and analyzed. The results show that the commercial MNSR of LEU conversion is feasible. Secondly, in this paper, through changing the fuel elements and the arrangement method, the new low enriched uranium (NLEU) core was designed to improve Φ/P ratio of the core, the proportion of thermal neutrons and the worth of the control rod. UO2 with U-235 enrichment of 19.75% was selected as the fuel pellet of the NLEU, NLEU not only meets the design parameters, but in many parameters, NLEU is better than LEU. The fuel element quantity is reduced by 43%, from original 344 to 196; reducing the amount of U-235 loading; improving the Φ/P ratio and the thermal neutron fraction is increased. The results show that the NLEU optimizes some parameters, simplifies the core structure, saves the construction cost, improves the nuclear safety and is more suitable for the application of MNSR.

scholarly journals Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

2021 ◽  
Vol 8 (1) ◽  
pp. 10-16
Author(s):  
Nguyen Thanh Vinh Ho ◽  
Vinh Vinh Le ◽  
Nhi Dien Nguyen ◽  
Kien Cuong Nguyen ◽  
Ton Nghiem Huynh ◽  
...  
Keyword(s):  
Power Distribution ◽  
Nuclear Physics ◽  
Research Reactor ◽  
Reactor Core ◽  
Calculated Data ◽  
Thermal Hydraulics ◽  
Fuel Type ◽  
Fuel Assemblies ◽  
Enriched Uranium ◽  
New Research

VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions.

INNOVATION CAPABILITY, ABSORPTIVE CAPACITY AND SMES PERFORMANCE IN PAKISTAN: THE MODERATING EFFECT OF BUSINESS STRATEGY

2018 ◽  
Vol 13 (Number 2) ◽  
pp. 1-11
Author(s):  
Muhammad Zulqarnain Arshad ◽  
Darwina Arshad
Keyword(s):  
Economic Growth ◽  
Absorptive Capacity ◽  
Business Strategy ◽  
Innovation Capability ◽  
Research Directions ◽  
Conceptual Paper ◽  
And Performance ◽  
New Research ◽  
The Relationship ◽  
Crucial Part

The small and medium-sized enterprises (SMEs) play a crucial part in county’s economic growth and a key contributor in country’s GDP. In Pakistan SMEs hold about 90 percent of the total businesses. The performance of SMEs depends upon many factors. The main aim for the research is to examine the relationship between Innovation Capability, Absorptive Capacity and Performance of SMEs in Pakistan. This conceptual paper also extends to the vague revelation on Business Strategy in which act as a moderator between Innovation Capability, Absorptive Capacity and SMEs Performance. Conclusively, this study proposes a new research directions and hypotheses development to examine the relationship among the variables in Pakistan’s SMEs context.

Research reactor model for isotope production based on the design of VVER-440 reactors

Kerntechnik ◽  
2020 ◽  
Vol 85 (2) ◽  
pp. 105-108
Author(s):  
A. Terekhova ◽  
A. Mahdi ◽  
R. Zykova
Keyword(s):  
Research Reactor ◽  
Isotope Production ◽  
Reactor Model

scholarly journals A modular design method based on TRIZ and AD and its application to cutter changing robot

2021 ◽  
Vol 13 (7) ◽  
pp. 168781402110343
Author(s):  
Mei Yang ◽  
Yimin Xia ◽  
Lianhui Jia ◽  
Dujuan Wang ◽  
Zhiyong Ji
Keyword(s):  
Modular Design ◽  
Design Method ◽  
Idea Generation ◽  
Design Parameters ◽  
Problem Analysis ◽  
Concept Design ◽  
Functional Requirements ◽  
Shield Tunneling ◽  
Structural Hierarchy ◽  
And Performance

Modular design, Axiomatic design (AD) and Theory of inventive problem solving (TRIZ) have been increasingly popularized in concept design of modern mechanical product. Each method has their own advantages and drawbacks. The benefit of modular design is reducing the product design period, and AD has the capability of problem analysis, while TRIZ’s expertise is innovative idea generation. According to the complementarity of these three approaches, an innovative and systematic methodology is proposed to design big complex mechanical system. Firstly, the module partition is executed based on scenario decomposition. Then, the behavior attributes of modules are listed to find the design contradiction, including motion form, spatial constraints, and performance requirements. TRIZ tools are employed to deal with the contradictions between behavior attributes. The decomposition and mapping of functional requirements and design parameters are carried out to construct the structural hierarchy of each module. Then, modules are integrated considering the connections between each other. Finally, the operation steps in application scenario are designed in temporal and spatial dimensions. Design of cutter changing robot for shield tunneling machine is taken as an example to validate the feasibility and effectiveness of the proposed method.

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