• Nuclear Engineering and Technology
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• 제20권1호
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• pp.19-26
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• 1988

저누출 장전 모형은 새 연료를 안에서부터 넣는 in-out 형태를 취하여 격납 용기의 fluence를 줄이고 중성자 경제성을 높이고자 하는 것으로, 이 경우에는 노심내의 전체적인 중성자 경제성은 좋아지지만 노심 중앙부에서의 새연료의 과다 반응도 때문에 안전성 여유를 줄이게 되므로 많은 수의 가연성 독붕봉을 사용하여 첨두 인자를 조절해야만 한다. 본 논문에서는 가연성 독붕봉 연소에 따른 출력 변화를 섭동으로 취급하며, 이를 출력감도 계수(Power Sensitivity Coefficient)로 표시한다. 최적화된 가연성 독붕봉의 분포를 구하기 인하여 알고 있는 주기말상태로부터, 노심 내의 출력과 과다 반응도를 제어하면서 주기초로 추적해 나가는 역연소법(Reverse Depletion Method)의 도입에 대한 타당성을 출력 민감도 계수개념과 선형 계획법을 이용하여 원자력 7호기 제1주기에 응용하여 검증했으며, 가연성 독붕봉의 추정량과 실제량과의 차이는 최대 4.5%의 오차를 보였다.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2007년도 춘계학술대회논문집
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• pp.197-201
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• 2007

Vibration characteristics of an annular cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.249-258
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• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2616-2628
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• 2021

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.