Control rod material effect on safety parameters of research reactors

Kerntechnik ◽  
2019 ◽  
Vol 84 (3) ◽  
pp. 200-207
Author(s):  
M. Hassan ◽  
H. K. Louis
Keyword(s):  
Control Rod ◽  
Research Reactors ◽  
Safety Parameters

scholarly journals Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor

2018 ◽  
Vol 50 (7) ◽  
pp. 1017-1023 ◽  
Author(s):  
Mina Torabi ◽  
A. Lashkari ◽  
Seyed Farhad Masoudi ◽  
Somayeh Bagheri
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Tehran Research Reactor ◽  
Safety Parameters

Development of REFT Code System for the Analysis of Highly Heterogeneous Research Reactors

2014 ◽  
Author(s):  
Tengfei Zhang ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Yunzhao Li
Keyword(s):  
Least Square Method ◽  
Least Square ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
User Friendly ◽  
Control Rods

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal SN method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT. In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

scholarly journals Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

Nukleonika ◽  
2014 ◽  
Vol 59 (2) ◽  
pp. 67-72 ◽  
Author(s):  
Farahnaz Saadatian-derakhshandeh ◽  
Omid Safarzadeh ◽  
Amir Saiid Shirani
Keyword(s):  
Control Systems ◽  
Systems Design ◽  
Reactor Physics ◽  
Control Rod ◽  
Research Reactors ◽  
Best Estimate ◽  
Hazard Control ◽  
Control Systems Design ◽  
And Control ◽  
Good Agreement

Abstract One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

scholarly journals Truly-Optimized PWR Lattice for Innovative Soluble-Boron-Free Small Modular Reactor

2021 ◽  
Author(s):  
Xuan Ha Nguyen ◽  
Seongdong Jang ◽  
Yonghee Kim
Keyword(s):  
High Performance ◽  
Local Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Low Leakage ◽  
Small Modular Reactor ◽  
Burnable Absorber ◽  
Safety Parameters ◽  
Atom Core

Abstract A novel re-optimization of fuel assembly (FA) and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (<1,000 pcm) as well as a long cycle length and high fuel burnup. For the SBF ATOM core, safety parameters are evaluated and the moderator temperature coefficient is shown to remain sufficiently and similarly negative throughout the whole cycle. It is demonstrated that the small excess reactivity can be well managed by mechanical shim rods with a marginal increase in the local power peaking, and a cold-zero shutdown is possible with a pseudo checker-board control rod pattern. In addition, a thermal-hydraulic-coupled neutronic analysis of the ATOM core is discussed.

Control rod drop-time reduction in typical swimming pool research reactors

2018 ◽  
Vol 106 ◽  
pp. 146-152
Author(s):  
M.R. Salmanpour Paean Afrakati ◽  
M. Gharib ◽  
S.M. Mirvakili
Keyword(s):  
Swimming Pool ◽  
Control Rod ◽  
Research Reactors ◽  
Drop Time ◽  
Time Reduction

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 Truly-optimized PWR lattice for innovative soluble-boron-free small modular reactor

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Xuan Ha Nguyen ◽  
Seongdong Jang ◽  
Yonghee Kim
Keyword(s):  
High Performance ◽  
Local Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Low Leakage ◽  
Small Modular Reactor ◽  
Burnable Absorber ◽  
Safety Parameters ◽  
Atom Core

AbstractA novel re-optimization of fuel assembly and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (< 1000 pcm) as well as a long cycle length and high fuel burnup. For the SBF ATOM core, safety parameters are evaluated and the moderator temperature coefficient is shown to remain sufficiently and similarly negative throughout the whole cycle. It is demonstrated that the small excess reactivity can be well managed by mechanical shim rods with a marginal increase in the local power peaking, and a cold-zero shutdown is possible with a pseudo checker-board control rod pattern. In addition, a thermal–hydraulic-coupled neutronic analysis of the ATOM core is discussed.

scholarly journals Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

2022 ◽  
Vol 2022 ◽  
pp. 1-13
Author(s):  
Izza Shahid ◽  
Nadeem Shaukat ◽  
Amjad Ali ◽  
Meer Bacha ◽  
Ammar Ahmad ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Distribution ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Pressurized Water ◽  
Control Rod ◽  
Operational Conditions ◽  
The Core ◽  
Safety Parameters

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

scholarly journals BANDUNG TRIGA 2000 REACTOR POWER ANALYSIS AS A FUNCTION OF THE NUMBER OF FUEL ELEMENTS AND THE POWER PEAKING FACTOR

2020 ◽  
Vol 22 (2) ◽  
pp. 68
Author(s):  
Sudjatmi K. Alfa ◽  
Endiah Puji Hastuti ◽  
Prasetyo Basuki ◽  
Santiko T. Sulaksono ◽  
Rian Fitriana
Keyword(s):  
Fuel Element ◽  
Power Analysis ◽  
Reactor Core ◽  
Reactor Power ◽  
Reactor Safety ◽  
Control Rod ◽  
Safety Parameters ◽  
Operating Limits ◽  
Fuel Elements ◽  
Safety Limit

The reactivity value of the Bandung TRIGA 2000 reactor core has decreased over time, so the power generated by the reactor is also getting smaller, despite the control rod position is fully withdrawn. Therefore, it is necessary to reshuffle and refuel the fuel element to increase the excess reactivity by considering the safety parameters, such as axial and radial power peaking factors, DNBR, dTsat, and temperature on the cladding and in the center of the fuel element. The analyzed reactor safety parameters are the number of fuel elements, which varied at 105, 110, and 115 elements, as well as power peaking factor, which varied at 1.55, 1.65, 1.75, 1.85, and 1.95. The calculations were done using MCNP and COOLOD-N2 programs. If DNBR ≈ 1.3 is determined as the safety limit for the operation of the Bandung TRIGA 2000 reactor, at PPF 1.95 (105, 110, and 115 fuel elements), it can be considered to operate the reactor at the power of 600-700 kW. However, at PPF of 1.75 (105, 110, and 115 fuel elements), the reactor can be operated at the power of 700-800 kW, and at PPF of 1.55 (105, 110, and 115 fuel elements), the reactor can be considered for operation at the power of 800-900 kW. The results of these calculations can be used for consideration in determining the operating limits of the Bandung TRIGA 2000 reactor.Keywords: TRIGA 2000, fuel element, power peaking factor, DNBR, boiling

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