[en] The signal production mechanism in a rhodium (Rh) fixed in-core detector emitter relies primarily on the beta particles resulting from neutron absorptions in either of two Rh isotopes to produce an electric current. As the neutron transmutation process depletes the Rh isotopes, the signal output per unit neutron flux from an Rh detector emitter will decrease. A vanadium detector is primarily sensitive to neutrons, but with a somewhat slower reaction time as that of a Rh detector. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. Platinum detectors are very sensitive to gamma flux, but only mildly sensitive to neutron flux. Because the depletion rate of platinum is very small, it can be neglected. Generally, both gamma and neutron signals are proportional to the assembly power. The characteristics of a new detector are the long life time due to the low depletion of emitter materials and the capability of reactor protection as well as reactor monitoring. The new detector uses vanadium and platinum as the emitter materials to meet the long life time and reactor protection capability. Vanadium detector is used for reactor monitoring and platinum detector is used for reactor protection. To determine the number of emitter strings, a comparative study of the power peaking factor monitoring accuracy for various self-powered fixed in-core detector geometries was made, and the configuration of the optimal detector design was also established and verified. The design of a new detector consists of five-string vanadium detector elements, and three-string platinum detector elements. The detector assembly also contains a background wire for compensation of noise signal and a thermocouple for use in the post-accident monitoring system. This new hybrid detector can be used for both reactor Monitoring And reactor Protection (MAP)

[en] Spent fuels for PWRs have been routinely transported and stored to the neighboring NPPs and to the dry storage facilities for PHWRs. Because spent fuel transportation may involve potential radiological risks such as dose risk and health effects for the public and the workers, both on-site and off-site spent fuel transportation need to be radiologically safety analyzed. Studies of SNF transportation risk assessment to Yucca Mountain have conducted by mainly Sandia National Laboratories using RADTRAN code since 1970s. RADTRAN code is no longer service and distributable due to their internal QA problems. Radiological transport risk assessment using INTERTRAN2 to benchmark RADTRAN4 was conducted and turned out no significant different between 2 codes. Risk assessment with on-site information in Korean NPPs will be made using INTERTRAN2 code in the near future.

[en] The electron flow from the emitter to the outer sheath produces an electrical current that is proportional to the number of neutron interactions in the emitter. Vanadium is one of several well known neutron detector materials and has been used in CANDU type reactors for many years. The vanadium based In-core Instrumentation (ICI) assembly design has been developing to displace the rhodium based ICI assemblies currently used in OPR1000. Rhodium has been commonly used in Light Water Reactors because it produces a relatively large output signal. The magnitude of the output signal from the rhodium detector minimizes the need to use very sensitive signal processing electronics to measure the output signal. The benefit of vanadium is its low depletion rate, which is about a factor of seven times less than rhodium, at the expense of smaller output current from the detector. The detector current can be increased by increasing the detector wire diameter and/or length. Due to a relatively small absorption cross section, and the lack of neutron resonance structure, the diameter can be increased efficiently and a simple self-powered detector model can still predict the output current very accurately. The purpose of this study is to show that vanadium signals can be used as effectively as signals from the rhodium detectors to reconstruct the core power distribution and the peaking factors, and that the validation of a nodal three-dimensional code can be based on the analysis of vanadium detector signals as well as rhodium detector signals. This study represents one step in the overall validation of vanadium detectors

[en] Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the rhodium SPND has been used in many nuclear power plants. The lifetime of rhodium is too short (about 3∼5 years) to operate the nuclear power plant economically. The vanadium (V) SPND is also primarily sensitive to neutrons like rhodium, but is a somewhat slower reaction time as that of a rhodium SPND. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. For this reason, a vanadium SPND has been being developed to replace the rhodium SPND which is used in OPR1000. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina (Al₂O₃) insulator with a cylindrical geometry. An MCNP-X code was used to simulate some factors (neutron self shielding factor and electron escape probability from the emitter) necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND

[en] The ACE7^TM fuel has been developed for Westinghouse type reactors in Korea and has several outstanding benefits against the existing fuel. For commercial use of ACE7^TM fuel, the criticality safety evaluation for fresh and spent fuel storage racks should be performed and licensed by regulatory. Benchmark calculations of an analysis code system and a cross section library, and criticality safety evaluation of the fuel storage racks in Kori unit 2 were performed in this study

[en] MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects

[en] Rh ICI (In-Core Instrumentation) used in OPR1000 generates the relatively large signal but its lifetime is below 6 years. Rh ICI consists of 5 detectors which is a type of SPND (Self Powered Neutron Detector), a couple of thermo-couple, one background wire and several fillers. The short lifetime of Rh detector causes increase of procurement price and space shortage of spent fuel pool. Also, it makes operators be exposed by more radiations. KHNP (Korea Hydro and Nuclear Power Co., Ltd.) CRI (Central Research Institute) is developing the LLICI (Long-Lived In-Core Instrumentation) based on vanadium to solve these problems. LLICI is the detector which is a type of SPND based on Vanadium and has the lifetime of about 10 years. The short lifetime of OPR1000's Rh ICI and long cycle operation strategy cause increase of procurement price, space shortage of spent fuel pool and more radiation exposed to operators. KHNP (Korea Hydro and Nuclear Power Co., Ltd.) CRI (Central Research Institute) is developing the LLICI (Long-Lived In-Core Instrumentation) to solve these problems

[en] Two platinum detector assemblies were installed in YGN Unit 4 during the cycle 5, which have the same mechanical design and dimensions as the existing rhodium detectors except the detector emitter material is platinum. By comparing the platinum signals with their symmetric location rhodium signals, the characteristics of raw platinum signals were analyzed. It is found that because the platinum signals are very small the signal response characteristics are strongly impacted by the background signal correction level and method. PHOENIX-4 calculation is performed to evaluate the gamma sensitivity to several physics parameters. Nodal Weighting Factors are introduced to represent the effect from the neighboring assemblies on the platinum detector response by using the Monte Carlo calculation code (MCNP). The normalized calculated and measured detector powers are compared to represent the behavior of platinum detector signals. The results indicate that platinum detector sensitivity has an elevation dependent behavior due to background and leakage current of the cable and detector. Also, the platinum detector sensitivity in fresh assembly appears to gradually increase due to the buildup of the fission product of long decay constants

[en] A series of MCNP runs have been performed to compare the results of radiation dose calculated from several standards. A point source and Watt fission energy spectrum were defined for the calculations. For conversion from flux to rem/rad-in-tissue, ANSI/ANS-6.1.1-1977 has been used. The new standards, ANSI/ANS-6.1.1-1991 and ICRU-57 are introduced to compare the converted dose/dose rate. For dose/dose rate in rad-in-air, F5 option in MCNP has been used. ICRU-44 and ICRP-74 standards are introduced to check the MCNP results

[en] In CANDU reactors, the Regional Overpower Protection Trip (ROPT) systems protect the reactor against overpowers in the reactor fuel, whether these are due to localized peaking within the core or to a general increase in the core power levels. The ROPT systems each have detectors distributed about the core to ensure coverage of the local flux and power peaks that could arise due to maneuvering, or to normal or abnormal combinations of reactivity devices inserted or withdrawn. For this purpose, the ROPT setpoint is calculated using design and non-design basis flux shapes for slow loss of regulation (SLOR), critical channel power (CCP) related flux shapes, and uncertainties of detector and thermal hydraulic conditions. However, these flux shapes were generated on the basis of the operating procedure for the Gentily-2 or Point Lepreau nuclear power plants in 1995. In general, some of these flux shapes are thought to match with the operation of the Wolsong nuclear power plant, but some have aspects that need to be improved. In this study, the re-classification of flux shapes has been performed to match them with the operation of the Wolsong NPP