scholarly journals Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

2017 ◽  
Vol 2017 ◽  
pp. 1-10 ◽  
Author(s):  
Giang Phan ◽  
Hoai-Nam Tran ◽  
Kien-Cuong Nguyen ◽  
Viet-Phu Tran ◽  
Van-Khanh Hoang ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Maximum Deviation ◽  
Control Rod ◽  
The Core ◽  
Nuclear Research Reactor ◽  
The Difference ◽  
Enriched Uranium ◽  
Analysis Models ◽  
Control Rods

Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

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

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Kien-Cuong Nguyen ◽  
Vinh-Vinh Le ◽  
Ton-Nghiem Huynh ◽  
Ba-Vien Luong ◽  
Nhi-Dien Nguyen
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Nucleate Boiling ◽  
Fuel Cladding ◽  
Hydraulic Analysis ◽  
The Core ◽  
Nuclear Research Reactor ◽  
Calculation Results ◽  
Enriched Uranium ◽  
Thermal Hydraulic Analysis

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°C and 1.5°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 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°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 °C, for global only, and 114.3°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.

scholarly journals Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel

2014 ◽  
Vol 4 (1) ◽  
pp. 70-75
Author(s):  
D. Vu C. ◽  
Q. Thien T. ◽  
V. Doanh H. ◽  
D. Quyet P. ◽  
T. Anh T.T. ◽  
...  
Keyword(s):  
Neutron Activation Analysis ◽  
Neutron Flux ◽  
Neutron Spectrum ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Nuclear Research ◽  
Fast Neutron Flux ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Full Core

In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75% 235U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron spectrum parameters which are used in k0-NAA such as thermal neutron flux (fth), fast neutron flux (ffast), f factor, alpha factor (a), and neutron temperature (Tn) have been re-characterized at four different irradiated channels in the core. Based on the experimental results, it can be seen that the thermal neutron flux decreases by 6÷9% whereas fast neutron flux increases by 2÷6%. The neutron spectrum becomes‘harder’ at most of irradiated positions. The obtained neutron spectrum parameters from this research are used to re-establish the procedures for Neutron Activation Analysis (NAA) according to ISO/IEC 17025:2005 standard at NuclearResearch Institute.

scholarly journals Thermal-hydraulically calculation for the Dalat nuclear research reactor

1991 ◽  
Vol 13 (4) ◽  
pp. 14-19
Author(s):  
Ngo Phu Khang
Keyword(s):  
Nuclear Reactor ◽  
Research Reactor ◽  
Nuclear Research ◽  
Core Domain ◽  
The Core ◽  
Nuclear Research Reactor

This paper concerns with thermal-hydraulically calculation for the core domain of the Dalat nuclear reactor in the nominal regime of works. On the base of this calculation some conclusions. On thermal safely of the reactor are conducted.

scholarly journals Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte Carlo code

2019 ◽  
Vol 9 (3) ◽  
pp. 21-29
Author(s):  
Cuong Nguyen Kien ◽  
Dung Nguyen Thi ◽  
Phu Tran Viet ◽  
Tiep Nguyen Huu ◽  
Ha Pham Nhu Viet
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Research Reactor ◽  
Model Development ◽  
Nuclear Research ◽  
Monte Carlo Code ◽  
Reactor Kinetics ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Comparative Results

This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code. The DNRR Serpent 2 model was verified through comparing with the MCNP6 criticality calculations under different reactor conditions. The parameters to be compared include the effective neutron multiplication factor, radial and axial powerdistributions, and thermal neutron flux distributions. The comparative results generally show a good agreement between Serpent 2 and MCNP6 and thus indicate that the DNRR Serpent 2 model can be used for further calculations of the DNRR.

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.

scholarly journals Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

2020 ◽  
Vol 2020 ◽  
pp. 1-11
Author(s):  
Nhi-Dien Nguyen ◽  
Kien-Cuong Nguyen ◽  
Ton-Nghiem Huynh ◽  
Doan-Hai-Dang Vo ◽  
Hoai-Nam Tran
Keyword(s):  
Conceptual Design ◽  
Research Reactor ◽  
Nuclear Research ◽  
The Core ◽  
Horizontal Beam ◽  
Fuel Assemblies ◽  
Power Rate ◽  
Safety Parameters ◽  
Enriched Uranium ◽  
Maximum Temperatures

The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

scholarly journals Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Kien-Cuong Nguyen ◽  
Vinh-Vinh Le ◽  
Ton-Nghiem Huynh ◽  
Ba-Vien Luong ◽  
Nhi-Dien Nguyen ◽  
...  
Keyword(s):  
Research Reactor ◽  
Radiation Safety ◽  
Nuclear Research ◽  
Severe Case ◽  
Safety Analysis ◽  
Water Levels ◽  
Spent Fuel ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Interim Storage

Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4  μ Sv / h is small enough for people to work normally.

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 Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel

2014 ◽  
Vol 4 (1) ◽  
pp. 36-45
Author(s):  
Nhi Dien Nguyen ◽  
Ba Vien Luong ◽  
Van Lam Pham ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Experimental Results ◽  
Design Calculation ◽  
Uranium Fuel ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Design Calculations ◽  
Dalat Research Reactor ◽  
Good Agreement

After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements.

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