• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.275-289
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• 2020

In 2017, multimodal transportation tests for evaluating road, sea, and rail transport were performed by research institutes in the US, Spain, and the Republic of Korea. In this study, acceleration and strain data determined through road and sea tests were analyzed. It was investigated whether the load generated for each transport mode was amplified or attenuated according to the load transfer path. From the results, it was confirmed that the load transfer characteristics differed according to the transportation mode and loading path. The effects of strain determined through each test on the structural integrity of the spent nuclear fuel were also investigated. It was found that the magnitude of the measured strain had a negligible effect on the structural integrity of the spent nuclear fuel, considering its fatigue strength. The results for the acceleration and strain data analyses obtained in this study will be useful for scheduled domestic transportation tests under normal transport conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.97-102
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• 1984

Criticality analyses of the KSC-1(Korean Shipping Cask-1) spent fuel shipping cask have been performed with the help of KENO-IV Monte Carlo computer code and 19-group CSLIB 19 cross section set which was generated from AMPX modular system. The analyses followed a benchmark calculation which has been made regard to the B & W CX-10 criticality facility in order to validate the Monte Carlo code cross section set described above. The KSC-1 shipping cask seems to be safe in the criticality point of view for the transport of one PWR spent fuel assembly under the normal conditions as well as the hypothetical accident conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.347-353
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• 2010

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

• Resources Recycling
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• v.21 no.5
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• pp.18-30
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• 2012

Zirconium is one of the most important material used as cladding of fuel rods in nuclear reactors because of its high dimensional stability, good corrosion resistance and especially low neutron-absorbing cross section. However, Hf free nuclear grade Zr sponge is commercially produced by only three countries including USA, France and Russia. So, Zr has been thoroughly managed as a national strategic material in Korea. Most of the zirconium is used for Korean nuclear industry as nuclear fuel cladding materials manufactured from Hf free Zr alloy raw material. Also, there are some other applications such as alloying element and detonator. In this review, zirconium production and recycling technologies have been reviewed and current industrial status was also analyzed. And recent achievements in innovative reduction technologies such as electrolytic reduction process and molten oxide electrolysis were also introduced.