• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• pp.191-196
• /
• 1997

As a part of International Criticality Safety Benchmark Evaluation Project (ICSBEP), SAXTON critical experiments were reevaluated. The effects on $K_{eff}$ of the uncertainties in experiment parameters, fuel rod characterization, soluble boron, critical water level, core structure, $^{241}$ Am and $^{241}$ Pu isotope number densities, random pitch error, duplicated experiment, axial fuel position, model simplification, etc., were evaluated and added in benchmark-model $k_{eff}$. In addition to detailed model, the simplified model for Saxton critical experiments was constructed by omitting the top, middle, and bottom grids and ignoring the fuel above water.r.r.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2001년도 추계학술발표회요약집
• /
• pp.312-312
• /
• 2001

• Nuclear Engineering and Technology
• /
• 제49권3호
• /
• pp.549-555
• /
• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• Nuclear Engineering and Technology
• /
• 제51권3호
• /
• pp.755-760
• /
• 2019

Compound samples based on the mineral-like magnesium potassium phosphate matrix $MgKPO_4{\times}6H_2O$ were synthesized by solidification of high level waste surrogate. Phase composition and structure of synthesized samples were studied by XRD and SEM methods. Compressive strength of the compounds is $12{\pm}3MPa$. Coefficient of thermal expansion of the samples in the range $250-550^{\circ}C$ is $(11.6{\pm}0.3){\times}10^{-6}1/^{\circ}C$, and coefficient of thermal conductivity in the range $20-500^{\circ}C$ is $0.5W/(m{\times}K)$. Differential leaching rate of elements from the compound, $g/(cm^2{\times}day)$: $Mg-6.7{\times}10^{-6}$, $K-3.0{\times}10^{-4}$, $P-1.2{\times}10^{-4}$, $^{137}Cs-4.6{\times}10^{-7}$; $^{90}Sr-9.6{\times}10^{-7}$; $^{239}Pu-3.7{\times}10^{-9}$, $^{241}Am-9.6{\times}10^{-10}$. Leaching mechanism of radionuclides from the samples at the first 1-2 weeks of the leaching test is determined by dissolution ($^{137}Cs$), wash off ($^{90}Sr$) or diffusion ($^{239}Pu$ and $^{241}Am$) from the compound surface, and when the tests continue to 90-91 days - by surface layer depletion of compound. Since the composition and physico-chemical properties of the compound after irradiation with an electron beam (absorbed dose of 1 MGy) are constant the radiation resistance of compound was established.

• 터널과지하공간
• /
• 제16권3호
• /
• pp.241-250
• /
• 2006

LNG지하저장 핵심기술 중 하나인 ice ring 형성을 열-수리 역학 해석을 통해 분석하였다. ice ring은 극저온물질의 누출에 대한 이차방벽 역할과 지하수의 저장공동 침투에 대한 일차방벽역할을 수행한다. 따라서 ice ring의 두께와 위치는 콘크리트와 PU foam 및 멤브레인으로 구성된 일차방호재의 무결성 유지에 있어서 핵심인자가 된다. 수치해석을 통해 이러한 ice ring의 위치와 두께를 추정하였으며 온도와 지하수위 결과를 계측치와 비교하였다. 수치해석결과는 PU foam 열전도도의 온도의존성과 지하수의 상변화를 고려한 경우 파일럿 공동에서 측정한 계측치와 잘 일치하였다. 본 논문에서 제시한 수치해석 기법은 실규모 저장시설의 설계에 있어서 ice ring 형성을 모델링하는 데에 바로 사용될 수 있을 것으로 판단된다.

• 한국수산과학회지
• /
• 제52권4호
• /
• pp.334-343
• /
• 2019

We investigated the sanitary characteristics of Alaska pollock Theragra chalcogramma roe as a raw material based on the standards of several countries. The standards for raw materials of Alaska pollock roe for lead, total mercury, $^{134}Cs+^{137}Cs$, and $^{131}I$ were those of the South Korean Ministry of Food and Drug Safety; Staphylococcus aureus, Salmonella spp., Clostridium botulinum, methyl mercury, $^{134}Cs+^{137}Cs$, $^{131}I$, $^{239}Pu$, and $^{90}Sr$ were those of the United States Food and Drug Administration; lead, methyl mercury, inorganic arsenic, chrome, $^{134}Cs+^{137}Cs$, and $^{131}I$ were those of the Ministry of Agriculture of China; nitrite ion, $^{134}Cs+^{137}Cs$, $^{239}Pu$, and $^{235}U$ were those of the Ministry of Health, Labor and Welfare of Japan; $^{134}Cs+^{137}Cs$, $^{131}I$, $^{239}Pu$, and $^{90}Sr$ were those of Codex; and $^{134}Cs+^{137}Cs$, $^{131}I$, $^{239}Pu$, $^{241}Am$, and $^{90}Sr$ were those of the European Food Safety Authority. The results for the global standard items other than C. botulinum (lead, total mercury, methyl mercury, inorganic arsenic, chrome, $^{134}Cs+^{137}Cs$, and $^{131}I$, S. aureus, and Salmonella spp.) suggest that Alaska pollock roe is safe for use as a raw material.

• Nuclear Engineering and Technology
• /
• 제47권3호
• /
• pp.380-387
• /
• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2733-2742
• /
• 2020

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P₀ scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P₅, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC²-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in ²³⁷Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for ²³⁵U, from 2.2% to 0.7% for ²³⁸Pu, and from 1.6% to 0.2% for ²⁴¹Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

• Nuclear Engineering and Technology
• /
• 제40권1호
• /
• pp.9-20
• /
• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.2803-2815
• /
• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.