• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• Nuclear Engineering and Technology
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• v.13 no.4
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• pp.264-276
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• 1981

A simulation procedure which can represent time-dependent nuclear characteristics of TRIGA Mark-III reactor is developed. CITATION, a multi-group diffusion-depletion program, has been utilized as calculational tool. The group structure employed in this study consists of 7 groups: -3-fast and 4-thermal-which is conventionally utilized in TRIGA type reactor analysis. Three-dimensional nuclear characteristics are synthesized by combining results from two-dimensional plane calculation and two-dimensional cylinder calculation, since direct three-dimensional approach is not yet possible. An effort ia made to develope a method which can extract effective zone and group dependent bucklings by neutron diffusion theory rather than conventional zone and/or group independent Ducklings by neutron transport theory, since neutron leakage is quite high for small core such as research reactors. It is turned out that the method developed in this study gives satisfactory results. The calculation is performed under assumptions that all control rods are fully withdrawn, that no samples are inserted in the irradiation holes and that the core is located in the center of the reactor pool. Burnup-dependent variation of core excess reactivity, time dependent change of Xe-135 poisoning and reactivity worth of rotary specimen rack are calculated and compared with operation records. Neutron flux and power distribution as well as neutron spectrum in each irradiation .facility are presented.

• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• v.23 no.2
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• pp.165-173
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• 1991

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is Investigated in this study. This scheme of utilizing Pm spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification to the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories： refabrication and reconfiguration For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burnup and power distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The result show that most tandem fuel cycle options considered in this study are technically feasible as well as economically viable.

• Nuclear Engineering and Technology
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• v.16 no.4
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• pp.181-194
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• 1984

The in-core fuel management problem was studied by use of the calculus of variations. Two functions of interest to a public power utility, the profit function and the cost function, were subjected to the constraints of criticality, the reactor turnup equations and an inequality constraint on the maximum allowable power density. The variational solution of the initial profit rate demonstrated that there are two distinct regions of the reactor, a constant power region and a minimum inventory or flat thermal flux region. The transition point between these regions is dependent on the relative importance of the profit for generating power and the interest charges for the fuel. The fuel cycle cost function was then used to optimize a three equal volume region reactor with a constant fuel enrichment. The inequality constraint on the maximum allowable power density requires that the inequality become an equality constraint at some points in the reactor. and at all times throughout the core cycle. The finite difference equations for reactor criticality and fuel burnup in conjunction with the equality constraint on power density were solved, and the method of gradients was used to locate an optimum enrichment. The results of this calculation showed that standard non-linear optimization techniques can be used to optimize a reactor when the inequality constraints are properly applied.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.