• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05a
• /
• pp.1054-1059
• /
• 1995

국제원자력기구에서는 현재 적용되고 있는 보장조치(Safeguards) 방법을 보다 효과적이고 효율적으로 적용하기 위하여 1993년부터 'Program 93+2'라는 사업을 수행하고 있다. 이중 하나의 과제로 수행되고 있는 구역 보장조치는 기존의 보장조치 개념이 하나의 시설을 대상(Facility-Oriented Safeguards)으로 개발된 것과는 달리 동일한 범주의 핵물질을 취급하는 여러 개의 시설을 하나의 가상적인 구역(Fuel Cycle-Oriented Safeguards)으로 설정하여 보장조치를 적용하는 개념으로, 보다 강화된 사찰 활동에 의하여 보장조치 신뢰도를 향상시키면서 사찰 횟수 및 사찰량은 절감되고 있다. 우리나라는 한국원자력연구소의 중수로핵연료 가공시설과 월성 1호기를 천연우라늄 구역(Natural Uranium Zone)으로, 한국원전연료(주)의 경수로핵연료 가공시설과 국내의 모든 경수로를 저농축우라늄 구역(Low Enriched Uranium Zone)으로 설정하여 성공적으로 구역 보장조치를 적용하고 있다. 그러나 이러한 구역 보장조치의 적용에는 원자력산업 체제의 단순화와 같은 제약조건이 따른다. 앞으로 우리나라에서는 현재 적용되고 있는 구역 보장조치 방법이 보다 효율적으로 운영되고 시설 운영에 대한 방해를 최소화시키는 방안을 고려하여야 하며 이에 는 가공시설에서의 생산 및 수송 일정을 발전소 운영 및 사찰 일정과 적절히 조화시키는 방법, 가공시설에서 검증된 핵연료에 대하여 적절한 감시 및 봉인 장비를 적용하는 방법, 현재의 구역 이외의 시설 또는 핵물질에 새로운 구역을 설정, 적용하는 방안 등을 고려할 수 있다.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.469-476
• /
• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• Nuclear Engineering and Technology
• /
• v.38 no.5
• /
• pp.443-448
• /
• 2006

The OPAL (Open Pool Australian Light-water) reactor is currently being constructed to replace HIFAR (HI-Flux Australian Reactor, commissioned in 1958) in mid-2006. HIFAR will be shutdown for decommissioning after several months of simultaneous operation with OPAL for smooth transition of operating systems and business. OPAL is a 20 MW multipurpose research reactor for radioisotope production, irradiation services and neutron beam research. The OPAL reactor uses low enriched uranium fuel in a compact core, cooled by light water and moderated by heavy water, yielding maximum thermal flux not less than $4{\times}10^{14}ncm^{-2}s^{-1}$. The reactor containment building is constructed of reinforced concrete and has been designed to protect the reactor from all external events such as seismic occurrences and impact from a hypothetical light aircraft crash. This paper describes the main elements of the reactor design and its applications.

• Nuclear Engineering and Technology
• /
• v.22 no.4
• /
• pp.351-358
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.47 no.7
• /
• pp.875-883
• /
• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.125-146
• /
• 2014

This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

• Nuclear Engineering and Technology
• /
• v.48 no.2
• /
• pp.351-359
• /
• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3913-3923
• /
• 2023

Todays, medium energy resolution detectors are preferably used in radioisotope identification devices(RID) in nuclear and radioactive material categorization. However, there is still a need to develop or enhance « automated identifiers » for the useful RID algorithms. To decide whether any material is SNM or NORM, a key parameter is the better energy resolution of the detector. Although masking, shielding and gain shift/stabilization and other affecting parameters on site are also important for successful operations, the suitability of the RID algorithm is also a critical point to enhance the identification reliability while extracting the features from the spectral analysis. In this study, a RID algorithm based on Bayesian statistical method has been modified for medium energy resolution detectors and applied to the uranium gamma-ray spectra taken by a LaBr3:Ce detector. The present Bayesian RID algorithm covers up to 2000 keV energy range. It uses the peak centroids, the peak areas from the measured gamma-ray spectra. The extraction features are derived from the peak-based Bayesian classifiers to estimate a posterior probability for each isotope in the ANSI library. The program operations were tested under a MATLAB platform. The present peak based Bayesian RID algorithm was validated by using single isotopes(²⁴¹Am, ⁵⁷Co, ¹³⁷Cs, ⁵⁴Mn, ⁶⁰Co), and then applied to five standard nuclear materials(0.32-4.51% at.²³⁵U), as well as natural U- and Th-ores. The ID performance of the RID algorithm was quantified in terms of F-score for each isotope. The posterior probability is calculated to be 54.5-74.4% for ²³⁸U and 4.7-10.5% for ²³⁵U in EC-NRM171 uranium materials. For the case of the more complex gamma-ray spectra from CRMs, the total scoring (ST) method was preferred for its ID performance evaluation. It was shown that the present peak based Bayesian RID algorithm can be applied to identify ²³⁵U and ²³⁸U isotopes in LEU or natural U-Th samples if a medium energy resolution detector is was in the measurements.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.499-507
• /
• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• Economic and Environmental Geology
• /
• v.24 no.3
• /
• pp.245-260
• /
• 1991

Surficial dispersion patterns of heavy metals and toxic elements (U, Mo, Cu, Zn, Fe, Mn, Co, Cr, V, Ni, Pb, and Cd) were investigated in the Dukpyungri, Goesan area covered with low grade uranium-bearing black shales. Maximum abundance of U in the black shale was 455ppm. Radioactivity was counted at a maximum of 7cps in black shales, and was less than 0.5cps in shales, slates, and oil shales of the control areas. Enrichment of Mo, V, Cu, Zn, Cd, and Pb in black shales is particularly characteristic compared with shales, slates, and oil shales of the control areas, whereas contents of Mn, Cr, Co, and Th in all rock samples tend to be almost similar. Residual top soils (0~15cm depth) over black shales show high contents of Mo, Cu, Zn, Ni, Cd, and V in comparison with the control areas. Contents of trace elements in subsoils (15~30cm depth) were higher about one and half times than those in topsoils. Average contents of Mo, Cu, Pb, Zn, Cd and V in garden soil and playground soil of an elementary school in Dukpyungri, Goesan area, were high about two to fifteen times compared with the control areas. Contents of trace elements in stream sediments were higher from two to eight times than those in residual soils. Sodium, AI, K, V, Cr, and Fe were more enriched in the roots of pine than in the twigs of pine. Contents of Li, AI, V, Ni, Cd, Fe, and Co were higher in the roots of azalea than in the twigs of azalea. Enrichment of P, Ca, and Mg was remarkable in the twigs of both pine and azalea. Biological absorption coefficients for essential elements (Zn, P, Mn, Ca and K)tend to be high, whereas those for the non-essential elements.(Ba, Ti, V, and Mo) and toxic elements(Cr, Co, Pb and Ni) be low. Less mobile elements (Pd, Cd, and Co) tend to show anomalies with higher contrast than more mobile elements(Mo, V, Zn, Cu and Ni) in the area covered with black shales.