• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.771-776
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• 1998

TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10$^{18}$ n/$\textrm{cm}^2$. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60cm below the midplane at zero degree.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.153-158
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• 2000

Nuclear reactor pressure vessel should be safety even in the case that hypothetical defects with allowable size are in vessel. Therefore, the materials should have excellent fracture resistance characteristics. The purpose of this study is to analyze the results of preirradiation baseline test of nuclear pressure vessel for Yonggwang Unit 5/6. In experiments, drop weight tests and impact tests are carried out to obtain nil-ductility transition reference temperature, $RT_{NDT}$ and static and dynamic fracture toughness tests are performed to compare with $K_{IR}$ curve in accordance with ASME Sec.III. The test results show that the materials had sufficiently fracture resistance characteristics for 40 years of design life.

• Transactions of Materials Processing
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• v.14 no.6 s.78
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• pp.565-569
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• 2005

Reactor Vessel is one of the most important structural parts of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings have been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the open die forging process and manufacturing experience of large head forgings which can be used for the reactor vessel of 1,000MW nuclear power plant.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2005.06a
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• pp.131-139
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• 2005

Reactor Vessel is one of the most important structural part of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings has been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the development trend of the open die forging process and manufacturing experience of large head forgings which canl be used for the reactor vessel of nuclear power plant.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2005.05a
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• pp.397-400
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• 2005

Reactor Vessel is one of the most important structural part of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings has been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the open die forging process and manufacturing experience of large head forgings which cant be used for the reactor vessel of 1,000MW nuclear power plant.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.355-361
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• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.1-10
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• 2006

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.509-516
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• 1995

The uncertainty analysis for the in-vessel steam explosion during severe accidents at a nuclear power plant is performed using a probabilistic approach. This approach consists of four steps; 1) screening, 2) quantification of uncertainty 3) propagation of uncertainty, and 4) output analysis. And the specific methods which satisfy the sub-objectives of each step are prepared and presented. Compared with existing ones, the unique feature of this approach is the improved estimation of uncertainties through quantification, which ensures the defensibility of the resultant failure probability distributions. Using the approach, the containment failure probability due to in-vessel steam explosion is calculated. The results of analysis show that 1) pour diameter is the most dominant factor and slug condensed phase fraction is the least and 2) fraction of core molten is the second most dominant factor, which is identified as distinct feature of this study as compared with previous studies.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.906-914
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• 2016

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a $CuSO_4-H_2SO_4$ electroplating system were performed based on the analogy between heat and mass transfer. The $Ra^{\prime}_H$ of $10^{14}$ order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.337-337
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• 2002