• 제목/요약/키워드: fuel burnup

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Neutronics design of VVER-1000 fuel assembly with burnable poison particles

  • Tran, Hoai-Nam;Hoang, Van-Khanh;Liem, Peng Hong;Hoang, Hung T.P.
    • Nuclear Engineering and Technology
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    • 제51권7호
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    • pp.1729-1737
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    • 2019
  • This paper presents neutronics design of VVER-1000 fuel assembly using burnable poison particles (BPPs) for controlling excess reactivity and pin-wise power distribution. The advantage of using BPPs is that the thermal conductivity of BPP-dispersed fuel pin could be improved. Numerical calculations have been conducted for optimizing the BPP parameters using the MVP code and the JENDL-3.3 data library. The results show that by using $Gd_2O_3$ particles with the diameter of $60{\mu}m$ and the packing fraction of 5%, the burnup reactivity curve and pin-wise power distribution are obtained approximately that of the reference design. To minimize power peaking factor (PPF), total BP amount has been distributed in a larger number of fuel rods. Optimization has been conducted for the number of BPP-dispersed rods, their distribution, BPP diameter and packing fraction. Two models of assembly consisting of 18 BPP-dispersed rods have been selected. The diameter of $300{\mu}m$ and the packing fraction of 3.33% were determined so that the burnup reactivity curve is approximate that of the reference one, while the PPF can be decreased from 1.167 to 1.105 and 1.113, respectively. Application of BPPs for compensating the reduction of soluble boron content to 50% and 0% is also investigated.

Isotopic Analysis of Decay Heat Contributors From Actinides and Fission Fragments of Spent Nuclear Fuel for Intermediate- and Long-Term Storage Times

  • Amir Mohammad Al-Ramady
    • 방사성폐기물학회지
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    • 제22권1호
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    • pp.1-7
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    • 2024
  • In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM-1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

  • Meyer, M.K.;Gan, J.;Jue, J.F.;Keiser, D.D.;Perez, E.;Robinson, A.;Wachs, D.M.;Woolstenhulme, N.;Hofman, G.L.;Kim, Y.S.
    • Nuclear Engineering and Technology
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    • 제46권2호
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    • pp.169-182
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    • 2014
  • High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

파이로공정 시설 개념설계를 위한 기준 사용후핵연료 선정 (Reference Spent Nuclear Fuel for Pyroprocessing Facility Design)

  • 조동건;윤석균;최희주;최종원;고원일
    • 방사성폐기물학회지
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    • 제6권3호
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    • pp.225-232
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    • 2008
  • 제3차 전력수급기본계획에 근거하여 현재 운영중이거나 계획중인 원자력발전소에서 발생할 사용후핵연료의 양과 특성을 추정하였다. 본 연구에서 고려된 대상 특성은 핵연료집합체에 대한 제원, 핵연료봉 배열, 무게, $^{235}U$ 초기 농축도 및 방출연소도이다. 이들은 파이로공정 시설을 설계하는데 필수적인 것이다. 2077년말까지 가압경수로 사용후핵연료의 예상발생량은 약 23,000 tU이 될 것으로 보인다. $^{235}U$ 초기 농축도 4.5 wt.% 이하를 갖는 사용후핵연료의 비율은 전체 발생량의 약 95%를 차지할 것이며, 16$\times$16 배열을 갖는 핵연료집합체는 74%를 차지할 것 같다. 현재 사용후핵연료의 평균연소도는 45 GWd/tU인데 반해, 2010년대 중 후반 이후 발생할 사용후 핵연료의 평균연소도는 55 GWd/tU이 될 것 같다. 이상의 결과에 따라 파이로공정 시설의 설계를 위한 기준 사용후핵연료를 도출하였다. 예상 사용후핵연료는 21.4 cm $\times$ 21.4 cm의 단면적, 453 cm의 길이, 672 kg의 질량, 4.5 wt.%의 $^{235}U$ 초기 농축도 및 55 GWd/tU의 방출연소도를 갖는 16$\times$16 한국표준형연료가 타당할 것으로 판단된다.

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A CLASSIFICATION OF UNIQUELY DIFFERENT TYPES OF NUCLEAR FISSION GAS BEHAVIOR

  • HOFMAN GERARD L.;KIM YEON SOO
    • Nuclear Engineering and Technology
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    • 제37권4호
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    • pp.299-308
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    • 2005
  • The behavior of fission gas in all major types of nuclear fuel has been reviewed with an emphasis on more recently discovered aspects. It is proposed that the behavior of fission gas can be classified in a number of characteristic types that occur at a high or low operating temperature, and/or at high or low fissile burnup. The crystal structure and microstructure of the various fuels are the determinant factors in the proposed classification scheme. Three types of behavior, characterized by anisotropic $\alpha$-U, high temperature metallic $\gamma$-U, and cubic ceramics, are well-known and have been extensively studied in the literature. Less widely known are two equally typical low temperature kinds: one associated with fission induced grain refinement and the other with fission induced amorphization. Grain refinement is seen in crystalline fuel irradiated to high burnup at low temperatures, whereas breakaway swelling is observed in amorphous fuel containing sufficient excess free-volume. Amorphous fuel, however, shows stable swelling if insufficient excess free-volume is available during irradiation.

Impact of fuel temperature on nuclear core design calculations

  • Dusan Calic;Luka Snoj;Marjan Kromar
    • Nuclear Engineering and Technology
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    • 제56권9호
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    • pp.3668-3685
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    • 2024
  • The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel pellet can be obtained by coupling of Monte Carlo neutron transport codes with thermal-hydraulics solvers. However, this approach is not practical for standard nuclear design calculations, and computationally cheaper and faster methods must be used. In nuclear core simulators, a concept of a single "effective temperature" that yields the same neutron response as in the case of the actual temperature shape is mainly applied. This paper evaluates various fuel temperature models used in nuclear core simulation calculations, ultimately recommending a new effective temperature model that considers the burnup correction.

A Lattice-Based Monte Carlo Evaluation of Canada Deuterium Uranium-6 Safety Parameters

  • Kim, Yonghee;Hartanto, Donny;Kim, Woosong
    • Nuclear Engineering and Technology
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    • 제48권3호
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    • pp.642-649
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    • 2016
  • Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

Simulation of Pore Interlinkage in the Rim Region of High Burnup $UO_2$Fuel

  • Koo, Yang-Hyun;Oh, Je-Yong;Lee, Byung-Ho;Cheon, Jin-Sik;Joo, Hyung-Koo;Sohn, Dong-Seong
    • Nuclear Engineering and Technology
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    • 제35권1호
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    • pp.55-63
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    • 2003
  • Threshold porosity above which fission gas release channels would be formed in the rim egion of high burnup UO$_2$ fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cube, pore distribution in the rim was simulated 3-dimensionally by the Monte Carlo method according to porosity and pore size distribution. Then, using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of rim porosity. The simulation showed that porosity of 24-25% is the threshold above which the number of rim pores forming release channels increases very rapidly. On the other hand, channels would not be formed if the porosity is less than about 23.5%. This is consistent with the observation that, for porosity less than 23.5%, almost no fission gas is released in the rim. However, once the rim porosity reaches beyond 25%, extensive open paths would be developed and considerable fission gas release would start in the rim.

SARAPAN-A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

  • Kastanya, Doddy
    • Nuclear Engineering and Technology
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    • 제49권1호
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    • pp.267-276
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    • 2017
  • In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the $^*SIMULATE$ module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the $^*INSTANTAN$ module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the $^*INSTANTAN$ module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

Development and validation of isotope prediction module for VVER spent nuclear fuel analysis

  • Jaerim Jang;Deokjung Lee
    • Nuclear Engineering and Technology
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    • 제56권5호
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    • pp.1762-1776
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    • 2024
  • A spent nuclear fuel (SNF) analysis module for the Vodo-Vodyanoi Energetichesky Reactor (VVER) was developed and validated in this study. This advancement expands the application area of the existing nodal diffusion code, RAST-V, and reduces the need for additional code during 3D core simulations for SNF analysis, leading to increased efficiency in simulation time. RAST-V uses Lagrange interpolation and a power correction factor derived from the Bateman equation to bypass the re-depletion calculations, which are used to solve the microdepletion chain. This approach improved the efficiency of analysis. To mirror the conditions during the 3D core simulations, the module used history indices related to the moderator temperature, fuel temperature, and boron concentration. The module can predict 1620 isotopes. This paper presents the validation of this isotope inventory prediction and the application of burnup credit. The VVER analysis module was validated using 28 samples discharged from the Novovoronezh-4. Most isotopes were within 10 % of the boundaries of the measurements. This study successfully offers verification results using VVER benchmarks and discusses the application of burnup credit using a VVER-440 cask.