This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8&#xb0;C and 1.5&#xb0;C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500&#x2009;kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4&#xb0;C, which is lower than its limit value of 103&#xb0;C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 &#xb0;C, for global only, and 114.3&#xb0;C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Kien-Cuong Nguyen

Vinh-Vinh Le

Ton-Nghiem Huynh

Ba-Vien Luong

Nhi-Dien Nguyen

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

The radiation safety design and emergency analysis of an advanced nuclear system highly depends on the source term analysis results. In modular high-temperature gas-cooled reactors (HTGRs), the release rates of fission products (FPs) from fuel elements are the key issue of source term analysis. The FRESCO-II code has been established as a useful tool to simulate the accumulation and transport behaviors of FPs for many years. However, it has been found that the mathematical method of this code is not comprehensive, resulting in large errors for short-lived nuclides and large time step during calculations. In this study, we used the original model of TRISO particles and spherical fuel elements and provided a new method to amend the FRESCO-II code. The results show that, for long-lived radionuclides (Cs-137), the two methods are perfectly consistent with each other, while in the case of short-lived radionuclides (Cs-138), the difference can be more than 1&#x0025;. Furthermore, the matrix method is used to solve the final release rates of FPs from fuel elements. The improved analysis code can also be applied to the source term analysis of other HTGRs.

Chao Fang

Chuan Li

Jianzhu Cao

Ke Liu

Sheng Fang

Analysis of Fission Products&#x2019; Release in Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel Elements Using a Modified FRESCO II Numerical Model

To prevent the formation of salt from fission products of high-level radioactive liquid wastes (HLWs), a certain amount of acid is added to maintain the acidity of liquid waste. This study analyzes the accidents associated with the addition of acids in a factory by using the hazard and operability analysis (HAZOP), while elucidating the corresponding defects and risks of this approach. By improving the design of the system, the possibility of an accident is significantly reduced. This study can provide guidance for adding acids to treat other high-level waste liquids.

Zan Guo

Shuliang Zou

Wenge Ma

Haiyin Dai

HAZOP Analysis and Research of Temporary Acid Adding System for High-Discharge Waste Liquid

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n&#x2032;)115mIn and 27Al(n,&#x3b1;)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20&#x0025;, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

Jiaju Hu

Bin Zhang

Zhiwei Zong

Cong Liu

Yixue Chen

Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

Power level control is one of the critical functions in the instrument and control system of nuclear power plants (NPPs). In most power level control systems of NPPs, the power level or average neutron flux in reactor cores provided by out-of-core neutron sensors are usually measured as feedback of power control systems, while, as critical measuring devices, there is a risk of damage to out-of-core neutron sensors. For improving the operation reliability of NPPs under the neutron sensors&#x2019; failure, a power control system based on power observer is developed in this work. The simulation based on NPP simulator shows the power control system based on the observer is effective when neutron sensors fail.

Liming Zhang

Hongyun Xie

Qizhi Duan

Chao Lu

Jixue Li

Zhihong Lv

Power Level Control of Nuclear Power Plant Based on Asymptotical State Observer under Neutron Sensor Fault

In the field of radiation protection, the point-kernel code method is a practical tool widely used in the calculation of 3-D radiation field, and the accuracy of the point-kernel integration method strongly depends on the accuracy of the build-up factor. It is well known that calculation of the build-up factor for single-layer shields is composed of single material, but it is very complicated to calculate the build-up factor for multilayer shields (MLBUF). Recently, a novel and high-precision method based on the deep neural network (DNN) for calculating MLBUF has been proposed. In this paper, the novel method is described completely by slab models. Through the study of photon transport in multilayer shields, the parameters that mainly affect the calculation of build-up factor are analyzed. These parameters are trained by DNN as the input vectors, and the build-up factor for multilayer shields is predicted based on the trained DNN. The results predicted by DNN confirm that the method can calculate the build-up factor for multilayer shields quickly and accurately. The method has been preliminarily applicated into a 3-D radiation field calculation software, and it has proved that the method for calculating MLBUF has a broad application prospects in 3-D radiation field calculation.

Chao Li

Yingming Song

Zehuan Zhang

Jie Mao

Weiwei Yuan

Bo Wang

A Novel and High-Precision Method for Calculating the &#x3b3;-Ray Build-Up Factor for Multilayer Shields

It has been well known that nuclear power plant and coal-fired power plant release some amount of radioactive materials during their normal operations. The purpose of this study was to compare radiation exposure doses to the public as a consequence of airborne effluents released from nuclear and coal-fired power plants under the normal operation. NRCDose3 was used to estimate radiation exposure doses to the public from gaseous effluents of nuclear power plant during its normal operation while CAP88-PC was used to calculate doses to the public living around coal-fired power plant. The results showed that radiation exposure doses from nuclear power plant were less than those from coal-fired power plant and regulatory annual limits. Effective dose by external exposure, skin equivalent dose, and organ equivalent dose from gaseous effluents of nuclear power plant were 2.93&#x2009;&#xd7;&#x2009;10&#x2212;4&#x2009;mSv/y, 2.90&#x2009;&#xd7;&#x2009;10&#x2212;3&#x2009;mSv/y, and 1.78&#x2009;&#xd7;&#x2009;10&#x2212;2&#x2009;mSv/y, respectively. On the contrary, the corresponding effective dose by external exposure, external skin dose, and organ dose from coal-fired power plant were 1.13&#x2009;&#xd7;&#x2009;10&#x2212;2&#x2009;mSv/y, 5.33&#x2009;&#xd7;&#x2009;10&#x2212;2&#x2009;mSv/y, and 1.17&#x2009;&#xd7;&#x2009;10&#x2212;1&#x2009;mSv/y, respectively.

Boldsaikhan Purevsuren

Juyoul Kim

Public Health Effects of Radioactive Airborne Effluents from Nuclear and Coal-Fired Power Plant

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations