• Nuclear Engineering and Technology
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• v.37 no.4
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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.

• Analytical Science and Technology
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• v.11 no.6
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• pp.421-428
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• 1998

A method has been studied to separate Zr from various fission products in PWR spent nuclear fuels. A solution containing metal ions in place of radioactive fission products was prepared. The Zr was separated with 5 M HCl followed by eluting metal ions such as Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag and Cd with 12 M HCl on Dowex $1{\times}8$, anion exchange resin. The recovery of Zr was more than 95%. The purification of Zr was carried out on anion exchange resin, Dowex $1{\times}8$, in 5 M HCl in order to remove Mo causing isobaric effect during mass spectrometry. The method was applied to separate Zr from a spent PWR fuel. From mass spectrometric measurement, the purified Zr portion was not showed the isobars from other elements such as Mo and Sr.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2003.05a
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• pp.327-330
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• 2003

In this paper, a viscoelastic structural analysis for the spent pressurized water reactor(PWR) nuclear fuel disposal canister is carried out to predict the collapse of the canister while the canister is stored in a deep repository for long time. There may exist some subterranean heat in a deep repository while the nuclear fuel disposal canister is stored for long time. Then, a time-dependent viscoelastic structural deformation may occur in the canister due to the subterrnean heat Hence, the viscoelastic stress variation according to time should be computed to predict the structural strength of the canister. A viscoelastic material model is adopted. Analysis results show that even though some subterrnean heat may exist for quite a long time, the canister structure still endures stresses below the yield strength of the canister. Hence, some subterranean heat cannot seriously affect the structural strength of the canister.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.299-309
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• 1997

The neutronic feasibility of typical Korean three-loop 900MWe class PWR core loaded with mixed oxide fuels for both annual and 18-month cycle strategies has been investigated as a means for spent fuel management. For this study, a method of determining equivalent plutonium content was developed under the equivalence concept which gives the same cycle length as uranium fuel. Optimal plutonium zoning within the MOX assembly was also designed with the aim of minimizing the peak md power. Conceptual core designs hate hen developed for equilibrium cycle with the following variations: annual and 18-month cycle, 1/3 and full MOX loading schemes, and typical and high moderation lattice. The analysis of key core physics parameters shows that in all cases considered satisfactory core designs seem to be feasible, though addition of control rod system and change in Technical Specification for soluble boron concentration are required for full MOX loading in order to meet the current design requirements.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.317-328
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• 2011

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Analytical Science and Technology
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• v.14 no.4
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• pp.291-299
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• 2001

A method of separation and purification of Sm for quantitation of Sm isotopes from various fission products in PWR spent nuclear fuels has been studied. Simulated solution containing inactive metal ions(Cs, Ba, Gd, Eu, Sm and Nd) in place of radioactive fission products was prepared. Sm was separated with 0.5 M $HNO_3$/80% MeOH after washing with 1 M $HNO_3$/90% MeOH on AG $1{\times}8$, anion exchange resin. Sm was purified on cation exchange resin, AG $50W{\times}8$, pretreated with 0.2 M alpha-hydroxisobutyric acid(pH 4.5-4.6) to remove Ba causing isobaric effect Sm from PWR spent fuel. As a result of mass spectrometric measurement, eluted Sm portion did not include isobars form other elements such as Gd, Eu, Pm, Nd and BaO. The contents of Sm and its isotopes($^{147}Sm$, $^{148}Sm$, $^{149}Sm$, $^{150}Sm$, $^{151}Sm$, $^{152}Sm$ and $^{154}Sm$) in spent fuel were determined by isotope dilution mass spectrometric method spiking $^{154}Sm$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.167-173
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• 2004

The amount of cesium released from the leaching of spent fuels in contact with and without the compacted bentonite bloc]t which was compacted as the density of $1.4g/\textrm{cm}^3$, up to 5.7 years were measured and the empirical formula of the fractional release rate of cesium were derived from these measured values. The empirical formulas show that the long-term release rate of cesium under a repository would become a constant, as about $3{\times}10_{-6}$ fraction/day, after a certain period. The cumulative fractions of cesium released from the spent fuel with bentonite and with copper and stainless steel sheets were steadily increased, but the fraction from bare fuel was rapidly increased and then sluggishly increased. However, the remained value except its gap inventory from the cumulative fraction of cesium released from bare fuel was almost very close to the others. This suggests that the initial release of cesium from bare fuel might be dependant on its gap inventory.

• Bulletin of the Korean Chemical Society
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• v.24 no.11
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• pp.1686-1688
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• 2003