Korean Pressurized Water Reactor Research Articles

Development of CFD porous model parameters of Korean PWR spent nuclear fuel assemblies for thermal analysis

This study focused on developing and validating porous model parameters to accurately estimate the thermal behavior of spent nuclear fuel assemblies for computational fluid dynamics simulations. These assemblies are stored in Korean pressurized water reactors (PWRs) under dry storage conditions. Korean PWR fuel assemblies were categorized into four distinct types based on their configuration, facilitating the calculation of effective thermal conductivities and flow resistance coefficients. These parameters are crucial for simulating heat transfer and determining the peak cladding temperature (PCT), ensuring the thermal safety of stored nuclear fuel. The validation of the proposed porous model against explicit two-dimensional and three-dimensional simulations demonstrated its efficacy and reliability in enhancing the thermal analysis of spent fuel assemblies, highlighting the significance of accurately calculating the porous model parameters in the safety assessment of nuclear fuel storage. Furthermore, the study assessed the sensitivity of variables influencing temperature during spent nuclear fuel storage, focusing on flow resistance coefficients, control rod effect, and basket size. The findings revealed that variations in flow resistance and the presence of control rods minimally impacted PCT, whereas basket size significantly influenced temperature, underscoring its importance in the thermal analysis of nuclear fuel storage systems.

A new approach to determine abnormality of radioactive discharges from pressurized water reactors and to derive abnormality indicators correlated with a specific causal event

Abstract A novel methodology to analyze the trend of radioactive effluent data from nuclear power plants (NPPs) has been proposed to estimate the possibility of abnormal releases of radioactive materials, and to identify sensitive radionuclides with high correlation to specific events and abnormal releases. Actual discharge data including abnormal releases for 21 years (1,610 reactor-years) from 62 US and 22 Korean pressurized water reactors were collected. A release correlation factor (RCF) was newly proposed as a quantitative indicator of the abnormality of radioactive discharges based on annual discharge data from each NPP. The probability of correct estimation of abnormality using the RCF for a test case ranges 27–36 %, and such a modest estimation power was ascribed to the relatively low contribution (i.e., about 12 %) of a specific causal event to the annual radioactivity release. Through this, the feasibility of the RCF for estimating the abnormality of radioactive discharges was demonstrated. The RCF was derived for three types of causal events, and subsequently the sensitive radionuclides to each specific event were derived as 88Kr, 137Xe, and 138Xe for leakage from gas decay tank, 3H for leakage from steam generator’s power operated relief valve, and 85Kr, 88Kr, 133Xe, 133mXe, 3H and 14C for leakage from defected spent fuel during handling. The new approach proposed in this study is to be used to suppose the significant unreported abnormality in radioactive effluent data and the potential causal event, which may contribute to improving the safety and performance of radioactive effluent control as necessary.

Analysis of carbon‐14 discharges from Korean nuclear power plants

AbstractThe routine practice accompanying the operation of nuclear facilities involves the discharge of radioactive effluents from nuclear power plants (NPPs). Regulation of this discharge to the environment hinges on three criteria: radioactivity concentration, public dose, and radioactivity. Among these, radioactive carbon‐14 holds particular significance as it possesses an extensive half‐life of 5730 years, making it a primary source of radiation dose to communities residing around NPPs. In Korea, the monitoring of carbon‐14 discharges from pressurized water reactors (PWRs) in gaseous effluents has been ongoing since 2012, whereas before 2012, monitoring exclusively focused on carbon‐14 discharges from pressurized heavy water reactors (PHWRs). Analysis of carbon‐14 discharges from Korean PHWRs indicates that their emission constituted less than 1% of total radioactive effluents over the past two decades. In the context of Korean PWRs, carbon‐14 discharge monitoring was absent from 2002 to 2011, resulting in an absence of data regarding such discharges during that period. After introducing carbon‐14 discharge monitoring in gaseous effluents from Korean PWRs, emissions from 2012 to 2021 contributed 3% of the total gaseous effluents. These findings indicate that despite being the primary contributor to public dose, carbon‐14 discharges from NPPs constitute a minor portion of the radioactive effluent discharge.

An estimation and radioactivity measurement for radiocarbon(14C) in the Korean nuclear power plants

Radiocarbon (14C), with a radioactive half-life of approximately 5730 years, poses a long-term environmental contamination risk when released into the atmosphere. The quantification analysis of its release estimates plant-specific generation rates based on factors such as plant power, core neutron flux distribution, and the volume of water exposed to this flux. Utilizing the improved estimation method, the 14C production rate for several Korean Pressurized Water Reactors (PWRs) was calculated. Also, improvements in measurement methods through sampling have also been made. These enhancements include the verification of the absorption method versus the mixing method.The results of this study indicate that plant-specific 14C production rates range from 0.213 to 0.317 TBq/yr, which are comparable to the global range observed in PWRs. Furthermore, the study evaluated a quenching correction curve for a liquid scintillation counter using two quenching correction methods: the external standard method and the internal standard method. The accuracy of these methods with 72 samples was validated with an average relative error within ±2.5%. The relative error of the mixing method, when compared to the direct absorption method, was found to be within ±20%. This finding underscores the validity of the improved measurement technique.

High-radiation-exposure work in Korean pressurized water reactors

Owing to strict radiation safety management in Korean nuclear power plants (NPPs), most radiation workers receive very low radiation doses, even lower than the annual dose limit for the general public. However, the occupational dose distribution indicates that some Korean NPP workers receive a relatively higher dose than the average dose. This inequity in radiation exposure could be reduced by providing customized radiation protection measures, such as dose constraints, to workers receiving relatively higher doses. In this study, dose normalization was performed to identify the highest radiation exposure work in Korean pressurized water reactors (PWRs). The results show that most of the occupational exposure in Korean PWRs occurs during the planned maintenance period. Finally, the three highest radiation exposure tasks in Korean PWRs were identified: nozzle dam installation and removal, eddy current testing, and man-way opening and closing.

A methodology to designate radiation‐controlled areas in decommissioning nuclear power plants

AbstractKori Unit 1, which was permanently shut down in 2017, is expected to be the first decommissioned commercial nuclear power plant (NPP) in Korea. Operating NPPs designate radiation‐controlled areas (RCAs) to protect workers from radiological hazards and provide appropriate measures. Decommissioning RCAs (DRCAs) in NPPs being decommissioned should also be designated by considering the radiological characteristics of workplaces and exposure routes of decommissioning workers, such as external and internal exposure. However, the criteria for the DRCAs in decommissioning Kori Unit 1 are presently the same as those during normal operation, according to the public‐hearing material of the final decommissioning plan for Kori Unit 1. This study analyzed criteria for RCAs in all Korean NPPs to propose criteria for DRCAs in decommissioning NPPs. Analysis results show that RCAs are classified into four, five, six, and eight zones in Korean pressurized water reactors with respective external radiation dose rates. The RCAs in Korean pressurized heavy water reactors are classified into six zones with external dose rates and derived air concentrations of tritium. This study proposes three classifications of DRCAs to keep occupational exposure as low as reasonably achievable while considering potential exposure routes, that is, external exposure, airborne contamination, and surface contamination areas. First, the external exposure areas are classified into several zones according to the levels of external radiation dose rates. Second, the airborne contamination areas are designated to prevent internal exposure to workers from the inhalation of radioactive material in the air during cutting and demolition. Last, the surface contamination areas are designated to minimize the skin contamination of workers as radioactive dust in the air is deposited on the surface of facilities and equipment over time.

Illustration of Nagra’s AMAC approach to Kori-1 NPP decommissioning based on experience from its detailed application to Swiss NPPs

This work presents an illustration of Nagra’s AMAC (Advanced Methodology for Activation Characterization) approach to the South Korean pressurized water reactor Kori-1 decommissioning. The results achieved are supported by the existing experience from the detailed AMAC applications to Swiss NPPs and are used not only for a demonstration of the applicability of AMAC to South Korean NPPs, but also for a first approximation of the activated waste volumes to be expected from Kori-1. A packaging concept based on the above activation characterization is also presented, using the AMAC algorithmic optimization software ALGOPACK leading to the minimum number of waste containers needed given the selected packaging constraints. Nagra’s AMAC enables effective planning before and during NPP decommissioning, including recommendations for cutting profiles for diverse reactor components and building structures. Finally, it is expected to lead to significant cost savings by reducing the number of expensive waste containers, by optimizing a potential melting strategy for metallic waste as well as by significantly limiting the number of radiological measurements. All information about Kori-1 used for the purpose of this study was collected from publicly available sources.

A Study on the Crystalline Boron Analysis in CRUD in Spent Fuel Cladding Using EPMA X-ray Images

Chalk River Unidentified Deposits (CRUDs) were collected from the Korean pressurized water reactor (PWR) plant (A, B, and C) where the axial offset anomaly (AOA) occurred. AOA, also known as a CRUD-induced power shift, is one of the key issues in maintaining stable PWR plant operations. CRUDs were sampled from spent nuclear fuel rods and analyzed using an electron probe micro-analyzer (EPMA). This paper describes the characteristics of boron-deposits from the CRUDs sampled from twice-burnt assemblies from the Korean PWR. The primary coolant of a PWR contains boron and lithium. It is known that boron deposition occurs in a thick CRUD layer under substantial sub-cooled nucleate boiling (SNB). The results of this study are summarized as follows. Boron was not found at the locations where the existence was confirmed in simulated CRUDs, in other words, the cladding and CRUD boundaries. Nevertheless, we clearly observed the presence of boron and confirmed that boron existed as a lump in crystalline form. In addition, the study confirmed that CRUD existed in a crystal form with a unique size of about 10 μm.

Operating margin for radioactive effluent released to the environment from korean pressurized water reactors compared to the effluent control limits, obtained by monitoring the concentration of radioactive effluent with radioactive effluent monitors

ABSTRACT Korean pressurized water reactors (PWRs) generally use radioactive effluent monitors for monitoring the concentration of radioactive effluents released to the environment. In this study, the operating margins for radioactive effluent monitors were analyzed to determine the levels of real-time concentration of effluents compared to effluent control limits (ECLs), the regulatory limits. The results show that the concentration of radioactive effluents released from Korean PWRs complied with the ECLs during the years 2012–2016. It was also found that outages at Korean PWRs did not impact the operating margins for radioactive effluent monitors; that is, there was no remarkable difference of the concentration of effluents between normal operation and maintenance periods. In terms of simultaneous effluent releases, the results demonstrate that exceeding the ECLs is unlikely to occur even under the hypothetical condition of coincident effluent releases from multiple discharge points at a Korean PWR.

PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs

Abstract This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

Methodology for determination of alarm and warning set-points for radioactive effluent monitors in Korean pressurized water reactors

All radioactive gaseous and liquid effluents discharged from Korean nuclear power plants are monitored by effluent monitors to prevent effluent releases to the environment under uncontrolled conditions. This paper provides the methodology and parameters used in the calculation of alarm (high) and warning (low) set-points for gaseous and liquid effluent monitors in Korean pressurized water reactors. Alarm set-points are determined to assure compliance with the Korean regulatory limits of concentration of radioactive effluents. Even though warning set-points are not required by the regulatory body, Korean pressurized water reactors determine the warning set-points of effluent monitors not only to take an active management of effluent discharge but also to keep radiation doses to members of the public living around nuclear power plants as low as reasonably achievable.

Evaluation of the 14C release characteristics of a reactor building using high-compression collection equipment in Korean pressurized water reactors

In general, most nuclear power plants (NPPs) have one exhaust duct. However, each Korean NPPs specifically equip four exhaust ducts that release gas into the environment. Thus, Korean NPPs have difficulty in managing the radioactive effluents.The gas from the reactor building at a nuclear power plant is released into the environment through an exhaust system while fresh air is supplied into the building at the same time to satisfy the access conditions of the reactor building. It is important to secure measurable samples while the emission is in progress for reliable measurements of 14C in the effluent gas from the reactor building and for reliable assessments of the radiation exposure dose to residents who live in the area. The emission from the reactor building is done in a batch-wise manner in a limited time. It is difficult to measure 14C with conventional gaseous 14C collection equipment due to the typically small sampling volume of this type of equipment. To address this problem, a device was developed that can highly compress and store gas released over a short time in a tank that can be connected to gaseous 14C collection equipment for later analysis. This device was successfully used to capture samples to assess the characteristics of the 14C emissions from a reactor building.In this paper, a high-compression sampling device developed for the analysis of 14C from a reactor building is introduced and the result of an assessment of the characteristics of 14C emission from the reactor building with this device is discussed.

Pitting Corrosion and its Countermeasures for Pressurized Water Reactor Steam Generator Tubes

Abstract Pitting corrosion was the primary cause of the Alloy 600 (UNS N06600) steam generator tube degradation in a Korean pressurized water reactor (PWR) plant. Pulled tube examinations and remedial measures were carried out to mitigate the pitting. Based on the destructive examinations, the main causes of pitting corrosion were considered to be the following: accumulated sludge with a high copper content due to corrosion of copper alloys in the secondary system, acidic crevice conditions caused by chloride from condenser leakage, and ingress of air during layup. Countermeasures such as copper alloy replacement, water chemistry control, and chemical cleaning were implemented to mitigate the pitting. Chemical cleaning was evaluated as the most effective.

Investigation of PWR Hull with a View to Downgrade

The cladding materials remaining after the reprocessing of nuclear fuel, generally called hulls, are classified as high‐level radioactive waste. They are usually packaged in a container for disposal after being compacted, melted, or solidified into a heterogeneous matrix. Efforts to fabricate a better waste form from an environmental perspective have failed due to the technical difficulties encountered in the chemical decontamination of cladding hulls. In the early 1990's, the accumulation of radiochemical data on hulls and the advent of new technology such as laser or plasma have made the decontamination of hulls a viable option. This paper summarizes information regarding the radiochemical analysis of spent nuclear fuel hulls through a literature survey, including the characteristics of the hulls of 32,000 MWd/tU burn‐up and 15 years cooling of Korean pressurized water reactor. The reduction of the radioactivity by peeling off the inner surface of the hulls via laser technology was evaluated.

Evaluation of Nuclear Piping Failure Frequency in Korean Pressurized Water Reactors

The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.

Modeling the activity and scaling factor of 129I in the primary coolant and CVCS resin of operating PWR

To estimate the activity of 129I at the primary coolant and chemical and volume control system (CVCS) resin in Korean pressurized water reactor (PWR) plants, a theoretical methodology was developed on the basis of an existing model of primary coolant activity and new model of CVCS resin activity. In order to reflect the difference between 129I and 137Cs, the different power-related diffusivities in the defective fuel were derived, and the variable removal efficiency of the CVCS resin for 137Cs was applied as a function of the coolant activity ratio of 131I/ 137Cs. The current computational method was validated by using the measured coolant activities of 137Cs, and the results show better agreement than a previously suggested parameter correlation method between 129I and 137Cs. There was also reasonable agreement in a comparison of the results of the test resin columns of the coolant from the PWR plants of other countries. It was shown that the ratio of the effective removal efficiency of 129I and 137Cs in the CVCS resin linearly influences the activity ratio of 129I/ 137Cs in the coolant, but on the other hand, its influence on the activity ratio in the CVCS resin is relatively less sensitive compared with that in the coolant.