• 제목/요약/키워드: burnup measurement

검색결과 21건 처리시간 0.017초

iBEST: A PROGRAM FOR BURNUP HISTORY ESTIMATION OF SPENT FUELS BASED ON ORIGEN-S

  • KIM, DO-YEON;HONG, SER GI;AHN, GIL HOON
    • Nuclear Engineering and Technology
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    • 제47권5호
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    • pp.596-607
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    • 2015
  • In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems.

DISSOLUTION AND BURNUP DETERMINATION OF IRRADIATED U-Zr ALLOY NUCLEAR FUEL BY CHEMICAL METHODS

  • Kim, Jung-Suk;Jeon, Young-Shin;Park, Soon-Dal;Song, Byung-Chul;Han, Sun-Ho;Kim, Jong-Goo
    • Nuclear Engineering and Technology
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    • 제38권3호
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    • pp.301-310
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    • 2006
  • Destructive methods were used for the burnup determination of U-Zr alloy nuclear fuel irradiated in the High-flux Advanced Neutron Application Reactor (HANARO) at KAERI. The dissolution rate of unirradiated U-Zr alloy fuel in $HNO_3$/HF mixtures was investigated for the experimental conditions of a different temperature, and initial concentrations of HF and $HNO_3$. The irradiated U-Zr alloy fuel specimen was dissolved in a mixed acid condition of 3 M HNO3 and 1 M HF at $90^{\circ}C$ for 8 hours under reflux. The total burnup was determined from measurement of the Nd isotope burnup monitors. The method includes U, Pu, $^{148}Nd,\;^P{145}Nd+^{146}Nd,\;^{144}Nd+^{143}Nd$ and total Nd isotopes determination by the isotope dilution mass spectrometric method (IDMS) using triple spikes $(^{233}U,\;^{242}Pu\;and\;^{150}Nd)$. The effective fission yield was calculated from the weighted fission yields averaged over the irradiation period. The results are compared with that obtained by the destructive -spectrometric measurement of the $^{137}Cs$ monitor.

Determination of burnup limit for CANDU 6 fuel using Monte-Carlo method

  • Lee, Eun-ki
    • Nuclear Engineering and Technology
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    • 제53권3호
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    • pp.901-910
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    • 2021
  • KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that a fuel bundle can stay in the CANDU reactor without a time limit. However, the regulator requested traditional design values as well as the burnup limit reflecting the computer code uncertainty. The method for the PWR burnup limit was not applied to the CANDU fuel bundle. Since there was no approved methodology to build the burnup limit with uncertainties, KHNP introduced a Monte-Carlo method coupled with a 95/95 approach to determine the conservative burnup limit from the viewpoint of the centerline temperature, internal pressure, strain measurement deviation. Moreover, to consider the uncertainties of various computing models, a converted power uncertainty was introduced. This paper presents the methodology and puts forward the limits on burnup, evaluated for each of the existing and modified fuel bundles, in consideration of the pressure tube aging condition.

SCALE-ORIGEN-ARP를 이용한 사용후핵연료 내 중성자 및 감마선원 분석 (An analysis of neutron sources and gamma-ray in spent fuels using SCALE-ORIGEN-ARP)

  • 차소희;박광헌
    • 한국표면공학회지
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    • 제56권1호
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    • pp.84-93
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    • 2023
  • The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

조사후핵연료의 연소도 측정을 위한 동적이온교환체에 의한 우라늄 매질로부터 Pu 및 Nd의 분리 (Separation of Pu and Nd from Uranium Matrix by Equilibrated Cation Exchanger for Burnup Measurement of Irradiated Nuclear Fuel)

  • Joe, Kih-Soo;Kim, Jung-Suk;Jeon, Young-Shin;Han, Sun-Ho;Eom, Tae-Yoon
    • Nuclear Engineering and Technology
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    • 제25권2호
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    • pp.259-264
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    • 1993
  • 조사후핵연료의 연소도측정에 1-octanesulfonate 를 양이온 교환체로 사용하고 $\alpha$-hydroxyisobutyric acid를 용리액으로 사용하는 동적계의 이온크로마토그래피를 적용하였다 Pu, U 및 Nd의 최적 분리조건을 찾기위해 분리조건들을 변화하였다. 이들 원소들을 $\alpha$-hydroxyisobutyric acid 용리액을 0.05 M과 0.40 M을 혼합시키는 기울기용리법으로 개별 분리한후 분취하여 동위원소희석 질량분석법으로 각각 정량하였다. 본 방법에 의래 구한 연소도 값을 기존의 음이온교환수지법에 의한 값과 비교한 결과 3.5 %차이 이내에서 두 값이 서로 일치하였다.

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감마선 분광분석에 의한 조사후 핵연료 집합체(PWR)의 연소분포 및 냉각시간 결정 (Gamma-Ray Spectrometric Determination of Burnup Distribution and Cooling Time of Spent PWR Fuel Assemblies)

  • Young-Gil Lee;Jae-Shik Jun
    • Nuclear Engineering and Technology
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    • 제17권1호
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    • pp.1-7
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    • 1985
  • 원자력 발전소의 조사후 핵연료 저장풀에시 조사후 핵연료 집합체에 대한 감마선 분광분석 실험을 비파괴적인 방법으로 수행하였다. 조사후 핵연료 집합체가 갖는 연소분포를 알기 위해서 1차핵분열 생성물과 2차핵분열 생성물간의 감마선 강도비인 $^{134}$ Cs$^{137}$Cs을 사용했으며 그 결과는 이들 집합체가 노심내에서 연소시에 가졌든 중성자 분포의 기대치와 상응하였다. 이로부터 감마선 강도비 $^{134}$ Cs$^{137}$Cs은 연소도 해석을 위한 좋은 인디케이터임을 확인하였다. 핵물질의 안전관리면에서 중요시되고 있는 조사후 핵연료의 냉각시간을 감마선 강도비 $^{144}$ Ce$^{137}$Cs을 사용하여 구했으며 이를 핵연료 관리기록에 의한 냉각시간과 비교해 본 결과 각각 2%, 10%이내의 차이를 나타내었다. 이로부터 본 실험에서 냉각시간을 하기 위해서 유도한 방정식을 단일 주기로 연소된 핵연료에 대해서 사용할 수 있음을 실증하였다.

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DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL

  • Lee, Yongdeok;Park, Chang Je;Kim, Ho-Dong;Song, Kee Chan
    • Nuclear Engineering and Technology
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    • 제45권7호
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    • pp.921-928
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    • 2013
  • A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

Burnup Measurement of Spent $U_3$Si/Al Fuel by Chemical Method Using Neodymium Isotope Monitors

  • Kim, Jung-Suk;Jeon, Young-Shin;Park, Kwang-Soon;Song, Byung-Chul;Han, Sun-Ho;Kim, Won-Ho
    • Nuclear Engineering and Technology
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    • 제33권4호
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    • pp.375-385
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    • 2001
  • The total burnup in the spent U$_3$Si/Al fuel samples from Hanaro reactor was determined by destructive methods using $^{148}$ Nd, the sum of $^{143}$ Nd and $^{144}$ Nd, the sum of $^{145}$ Nd and $^{146}$ Nd, and the sum of total Nd isotopes($^{143}$ Nd, $^{144}$ Nd, $^{145}$ Nd, $^{146}$ Nd, $^{148}$ Nd and $^{150}$ Nd) monitors. The fractional($^{235}$ U) turnup in the spent fuel samples was also determined by U and Pu mass spectrometric method. The samples were dissolved in a mixture of 4 M HCI and 10 M HNO$_3$ without any catalyst. The separation of U, Pu and Nd from the spiked and unspiked sample solutions was achieved by two sequential anion exchange separation methods. The isotope compositions of these elements, after their separation from the fuel samples were measured by mass spectrometry. The contents of the elements in the spent fuel samples were determined by isotope dilution mass spectrometric method(IDMS) using $^{233}$ U, $^{242}$ Pu and $^{150}$ Nd as spikes. The effective fission yield was calculated from the weighted fission yields averaged over the irradiation period. The difference between total turnup values determined by various Nd monitors were in the range of 1.8%.

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Behavour of Hold-down Springs in Kori Nuclear fuels

  • Chun, Yong-Bum;Park, Kwang-June
    • 한국원자력학회:학술대회논문집
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    • 한국원자력학회 1995년도 추계학술발표회논문집(2)
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    • pp.674-679
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    • 1995
  • The hold-down spring forces of Kori nuclear fuels were measured for seven fuel assemblies having 1 to 4 cycles of irradiation histories in the Kori Unit-1 and -2 reactor. The fuel assemblies examined had burnup from 17 to 38 GWD/MTU and the examination was conducted in KAERI PIEF spent fuel storage pool with the newly developed underwater hold-down suing force measuring device. The measurement was made within the elastic deformation ranges and the trends of hold-down spring force relaxation behavour were examined.

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