• Title, Summary, Keyword: Burnup reactivity

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Void Reactivity of DUPIC Fuel Bundle

  • Hari P. Gupta;Park, Hangbok;Bo W. Rhee;Park, Hyungsoo
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.52-57
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    • 1996
  • The coolant void reactivity is positive for CANDU reactor loaded with DUPIC fuel which has more fissile content compared to natural uranium. A parametric study was done to reduce the void reactivity of the fuel bundle and loss in discharge burnup was estimated. It is observed that the burnable absorbers like gadolinium, boron, europium are not able to keep the reduction in void reactivity uniform throughout fuel burnup. Dysprosium and erbium can keep the void reactivity reduction uniform throughout. fuel burnup but toss in discharge burnup for erbium case is more compared to that of dysprosium case.

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Neutron Spectrum Effects on TRU Recycling in Pb-Bi Cooled Fast Reactor Core

  • Kim Yong Nam;Kim Jong Kyung;Park Won Seok
    • Nuclear Engineering and Technology
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    • v.35 no.4
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    • pp.336-346
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    • 2003
  • This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction.

A Criticality Analysis of the GBC-32 Dry Storage Cask with Hanbit Nuclear Power Plant Unit 3 Fuel Assemblies from the Viewpoint of Burnup Credit

  • Yun, Hyungju;Kim, Do-Yeon;Park, Kwangheon;Hong, Ser Gi
    • Nuclear Engineering and Technology
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    • v.48 no.3
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    • pp.624-634
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    • 2016
  • Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

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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    • v.51 no.7
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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.

Evaluation of Ductility During Reactivity Initiated Accident for Zirconium Cladding using Ring Tension Test (링 인장시험을 이용한 지르코늄 피복관의 반응도 사고(RIA) 시 연성 평가)

  • Kim Jun Hwan;Lee Myoung Ho;Choi Byoung Kwon;Bang Je Geon;Jeong Yong Hwan
    • Korean Journal of Materials Research
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    • v.15 no.2
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    • pp.126-133
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    • 2005
  • Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.

Fission-product Burnup Chain Model for Research Reactor Application (연구로용 핵분열 생성물 연소 체인 모델)

  • Kim, Jung-Do;Gil, Choong-Sup;Lee, Jong-Tai
    • Nuclear Engineering and Technology
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    • v.22 no.4
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    • pp.351-358
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    • 1990
  • A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the pseudo-element and the pseudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.

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Criticality effect according to axial burnup profiles in PWR burnup credit analysis

  • Kim, Kiyoung;Hong, Junhee
    • Nuclear Engineering and Technology
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    • v.51 no.6
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    • pp.1708-1714
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    • 2019
  • The purpose of the critical evaluation of the spent fuel pool (SFP) is to verify that the maximum effective multiplication factor ($K_{eff}$) is less than the critical safety limit at 100% stored condition of the spent fuel with the maximum reactivity. At nuclear power plants, the storage standard of spent fuel, ie, the loading curve, is established to prevent criticality from being generated in SFP. Here, the loading curve refers to a graph showing the minimum discharged burnup versus the initial enrichment of spent fuel. Recently, US NRC proposed the new critical safety assessment guideline (DSS-ISG-2010-01, Revision 0) of PWR SFPs and most of utilities in US is following it. Of course, the licensed criterion of the maximum effective multiplication factor of SFP remains unchanged and it should be less than 0.95 from the 95% probability and the 95% confidence level. However, the new guideline is including the new evaluation methodologies like the application of the axial burnup profile, the validation of depletion and criticality code, and trend analysis. Among the new evaluation methodologies, the most important factor that affects $K_{eff}$ is the axial burnup profile of spent fuel. US NRC recommends to consider the axial burnup profiles presented in NUREG-6801 in criticality analysis. In this paper, criticality effect was evaluated considering three profiles, respectively: i) Axial burnup profiles presented in NUREG-6801. ii) Representative PWR axial burnup profile. iii) Uniform axial burnup profile. As the result, the case applying the axial burnup profiles presented in NUREG-6801 showed the highest $K_{eff}$ among three cases. Therefore, we need to introduce a new methodology because it can be issued if the axial burnup profiles presented in NUREG/CR-6801 are applied to the domestic nuclear power plants without any other consideration.

Sensitivity studies in spent fuel pool criticality safety analysis for APR-1400 nuclear power plants

  • Al Awad, Abdulrahman S.;Habashy, Abdalla;Metwally, Walid A.
    • Nuclear Engineering and Technology
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    • v.50 no.5
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    • pp.709-716
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    • 2018
  • A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

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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    • v.48 no.3
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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.