Korean Nuclear Power Plants Research Articles

국내 복수호기 원전 운영을 고려한 일반인 선량제약치 설정에 대한 고찰

In the 2007 recommendation, the ICRP evolves from the previous process-based system of practices and intervention to the system based on the characteristics of radiation exposure situation. In addition, ICRP recommends the application of source-related dose constraints under the planned exposure situation as a tool for the optimization of protection to workers and the member of public. In this study, the analysis of radioactive effluents from Korean nuclear power plants and the public dose assessment were conducted in reference with the use of dose constraints. Finally, the measure to implement the dose constraints for the member of public was suggested taking into account multi-unit reactors operating at a single site in Korea.

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.

An analysis of the radiation field characteristics for extremity dose assessment during maintenanceperiods at nuclear power plants in Korea

Workers who maintain the water chambers of steam generators during maintenance periods in nuclear power plants (NPPs) have a higher likelihood of high radiation exposure, even if they are exposed for a short period of time. In particular, it is expected that the hands of workers would receive the highest radiation exposure as a consequence of hand contact with radioactive materials. In this study, a characteristic analysis of inhomogeneous radiation fields for contact operations was conducted using thermoluminescent dosemeters for the whole body and extremities during maintenance periods at Korean NPPs. It was observed that inhomogeneous radiation fields for contact operations at NPPs were dominated by high-energy photons.

IDENTIFICATION AND ASSESSMENT OF AGING-RELATED DEGRADATION OCCURRENCES IN NUCLEAR POWER PLANTS

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

Application of element correction and personal dose equivalent response to extremity dosimeters used at Korean nuclear power plants

Radiation workers who inspect and repair steam generators during maintenance periods in nuclear power plants (NPPs) have a likelihood of high radiation exposure to their extremities, especially the hands, because of contact with radioactive materials. To estimate the personal dose equivalent, Hp(0.07) to an extremity, it is necessary for workers to wear extremity dosimeters, thermoluminescent dosimeters (TLD), on their fingers. In this article, the element correction and the personal dose equivalent response for two types of extremity dosimeters, Harshaw EXTRAD dosimeters and Panasonic UD‐807 dosimeters, which are currently available at Korean NPPs, were calculated by calibration and irradiation tests. To provide the characteristics of the incident radiation field for contact tasks, the results of a radiation field test for inhomogeneous radiation fields in Korean NPPs were also introduced. Finally, on the basis of the results of calibration, irradiation, and radiation field tests, the application of element correction and personal dose equivalent response to extremity dosimeters for radiation workers at Korean NPPs was evaluated in terms of practical use.

A bioassay experience and lessons learned on the internal contamination of 131I during a maintenance period in a Korean nuclear power plant

During a maintenance period at a Korean nuclear power plant, internal exposure of radiation workers occurred by the inhalation of (131)I that was released into the reactor building from a primary system opening due to defective fuels. The internal activity in radiation workers contaminated by (131)I was immediately measured using a whole body counter (WBC). A whole body counting was performed again a few days later, considering the factors of equilibrium in the body. The intake and the committed effective dose were estimated based on the WBC results. The intake was also calculated by hand, based on both the entrance records to the reactor building, and the counted results of the air concentration for (131)I were compared with the whole body counting results.

Determination of a representative composition of dry active waste from Korean nuclear power plants

As part of the following-up action pertaining to the construction and operation permit for the first stage of the LILW (Low- and Intermediate-Level Radioactive Waste) repository, the preparation for a large-scale in situ experiment is underway by the Korea Radioactive Waste Management Corporation (KRMC) for a realistic assessment of the characteristics of gas generation after the post-closure phase of a repository. In this paper, we discuss a method of determining the representative composition of simulated dry active waste and a related fabrication plan for this material. After a comparison with experimental gas generation results from Finland, dry active waste content was chosen for a large-scale gas generation experiment. Six different types of materials and details on their simulated dry active waste contents are derived with the total mass and density for both a 200L and a 320L drum.

The monetary value of the man-mSv for Korean NPP radiation workers assessed by the radiation aversion factor

The monetary value of the man-mSv for operators of Korean nuclear power plants (NPPs) was calculated using a radiation aversion factor based on a survey of NPP workers. Initially, the life expectancy in the population is 79.4 y, the average age of cancer occurrence is 60 y, the average annual wage for an electric worker is 56 000 $ y(-1) and the nominal risk coefficient induced by radiation is 4.2E(-5) mSv were used to evaluate the basic monetary value (α(base)) resulting in 45.6 $ mSv(-1). To investigate the degree of radiation aversion, the subject of the investigation was selected as the working radiation workers in 10 NPPs in Korea (Kori 1-2, Yeonggwang 1-3, Ulchin 1-3 and Wolseong 1-2). In August 2010, with the cooperation of KHNP and partner companies, a total of 2500 survey questionnaires to 10 NPPs (or 250 surveys to each NPP) were distributed to currently employed radiation workers. From these, 2157 responses were obtained between August and October 2010. The assessed radiation aversion factor and the monetary value of the man-mSv from the calculated radiation aversion factor were 1.26 and ∼50 $ in the 0-1 mSv range, 1.38 and ∼200 $ in the 1-5 mSv range, 1.52 and ∼1000 $ in the 5-10 mSv range, 1.65 and ∼4000 $ in the 10-20 mSv range and 1.74 and ∼8500 $ >20 mSv.

ABRASIVE BLASTING TECHNOLOGY FOR DECONTAMINATION OF THE INNER SURFACE OF STEAM GENERATOR TUBES

The inner surfaces of bundled inconel tubes from steam generators in South Korean nuclear power plants are contaminated with cobalt and abrasive blasting equipment has been developed to efficiently remove the cobalt. The principal parameters related to the efficient removal using this equipment are the type of abrasive, the distance from the nozzle, and the blasting time. Preliminary tests were performed using oxidized inconel samples which enabled the simulation of cobalt removal from the radioactive inconel samples. The oxygen in the oxidized samples and the cobalt in the radioactive inconel were removed more effectively using the blasting distance, blasting time, and a silicon carbide abrasive. Using the developed abrasive blasting equipment, the optimum decontamination conditions for radioactive inconel samples were blasting for more than 6 minutes using silicon carbides under 5 atmospheric pressures.

고리 1발전소 부지 내 지하수 유동 및 삼중수소 이동 모델링

원전 운영자는 계통 및 기기의 열화 등에 의해서 발생할 수 있는 비계획적 방사성물질의 방출로 인한 환경 및 주변 주민에의 영향을 합리적으로 달성 가능한 한 낮게 유지하기 위해서 비계획적 방출을 조기에 감지할 수 있는 부지에 적합한 지하수 감시 프로그램을 수립해야 하며, 이를 위해서는 해당 부지의 수문지질 특성의 파악을 통해 지하수 유동을 평가해야 한다. 본 논문에서는 고리 1호기에서 계획되고 있는 지하수 감시 프로그램에 필요한 자료를 제공하기 위해, 고리 1발전소 부지의 기존 수문지질 조사 및 관련 자료의 조사를 통해 부지 내 지하수 유동특성을 파악하고 이를 바탕으로 가상의 비계획적 방출에 의한 삼중수소의 오염운(汚染雲)의 거동을 모의하였다. 모의 결과 지하수의 주 유동 방향은 남서방향이었으며 지하수의 대부분이 남쪽 및 동쪽 바다로 유입되었다. 삼중수소 오염운 역시 바다로 향하였으나 지하집수조(dewatering sump)에 의해 그 속도가 지연되는 것을 확인할 수 있었다. Nuclear power utilities should establish a site-specific groundwater monitoring program for early detection of unplanned radioactive material's releases which can occur due to degradation of systems, structures and components of the nuclear power plants in order to keep the impact of the unplanned releases on the environment and the residents as low as reasonably achievable. For this end, groundwater flow on site should be evaluated based on characterization of the hydrogeology of a site of concern. This paper aims to provide data necessary for establishing groundwater monitoring program which is currently considered at Kori nuclear power plant 1 by characterizing groundwater flow system on the site based on the existing hydrogeological studies and related documents, and by modeling tritium transport. The results showed that the major groundwater flow direction was south-west and that most of groundwater entered a southern and eastern seas. Although the tritium plume also released into the sea, its rate was delayed by dewatering sump.

6시그마 기법을 적용한 원자력 터빈 시뮬레이터의 발전기 출력 연산오차 저감

This paper describes the improvement of the calculation by using 6σ method on steam turbine simulator in a nuclear power plant. The simulator is essential to not only verification and validation of control logic but also making sure of control constants in upgrading the long time used control system into the new one. And the dynamic model is a key point in that simulator. The model used during the retrofit period of the turbine controller in Kori Nuclear Power Plant makes difference in calculating generator output and control valve positions. That is because such operating data as the main steam pressure, the main steam temperature and control valve positions of Yongkwang #3 are different from those of Kori #4. Therefore, the model parameters must be tuned by using actual operating data for the high fidelity of simulator in calculating the dynamic characteristic of the model. This paper describes that the 6σ method is used in improvement of precision of generator output calculation in the steam turbine model of the simulator.

Radiation Dose Distribution for Workers in South Korean Nuclear Power Plants

Radiation Dose Distribution for Workers in South Korean Nuclear Power Plants

Design process of the nanofluid injection mechanism in nuclear power plants

Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs). This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs) assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD) in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

Radiation fields and dose assessments in Korean nuclear power plants

In the primary systems of nuclear power plants (NPPs), various radionuclides including fission products and corrosion products are generated due to the complex water chemistry conditions. In particular, (3)H, (14)C, (58)Co, (60)Co, (137)Cs, and (131)I are important or potential radionuclides with respect to dose assessment for workers and the management of radioactive effluents or dose assessment for the public. In this paper, the dominant contributors to the dose for workers and the public were reviewed and the process of dose assessment attributable to those contributors was investigated. Furthermore, an analysis was carried out on some examples of dose to workers during NPP operation.

Estimates of Mechanical Properties and Residual Stress of Narrow Gap Weld for Leak-Before-Break Application to Nuclear Piping

The present paper addresses the mechanical characteristics of the stainless steel narrow gap weld (NGW) for the leak-before-break (LBB) application to the main loop piping of a nuclear power plant. Recently, in Korea, the connection with the reactor coolant main loop piping and the steam generator has been welded with ER308L NGW after the replacement of a steam generator of a Korean nuclear power plant. The NGW technique has many merits, for instance, the reduction of construction time and the reduction of shrinkage and deformation after welding due to its small groove angle and welding bead width compared with the conventional welds. In this paper, the tensile and fracture toughness test results of the three ER308L test coupons from NGW were presented and compared with those from conventional welds at the operating condition of the nuclear power plant. In addition, the distribution of the welding residual stress as well as the deformation behavior of the ER308L weld due to NGW was predicted through the nonlinear two-dimensional finite element analysis in which the detailed actual welding process of NGW was simulated. The results presented in this paper can be used to evaluate LBB application to nuclear piping with NGW and to provide the important information to perform the flaw evaluation as well as improve the weld procedure of NGW.

Analysis of Metabolism and Effective Half-life for Radiation Workers’ Tritium Intake at Pressurized Heavy Water Reactors

Tritium is the one of the dominant contributors to the workers of Pressurized Heavy Water Reactors’ (PHWRs) internal radiation exposure. This nuclide is likely to be released in work places as tritiated water vapor (HTO) from the primary system and can get relatively easily absorbed into the body of workers by inhalation. Inhaled tritium usually reaches the equilibrium of concentration after approximately 2 hours inside the body and is then excreted from the body with an effective half-life of 10 days. Because the tritium inside the body is transported within the body, the whole body can be exposed to radiation. Internal radiation exposure at PHWRs accounts for approximately 20-40 % of the total radiation exposure, where most of the internal radiation exposure is attributed to tritium. Thus, tritium is an important nuclide that needs to be necessarily monitored for radiation management safety. In this study, metabolism for tritium was established by using its excretion rate results in urine samples of workers at PHWRs. An effective half-life, a key parameter that estimates the radiation exposure, was derived from these results. As a result, it was found that the effective half-life for workers at Korean Nuclear Power Plants (NPPs) was 7.43 2.22 days, which is shorter than that of International Commission on Radiological Protection guides.

A Whole Body Counting Experience on the Internal Contamination of Iodine-131 at a Korean Nuclear Power Plant

A Whole Body Counting Experience on the Internal Contamination of Iodine-131 at a Korean Nuclear Power Plant

Development of Dual Counting Analysis and Exposure Possibility Evaluation for Carbon-14 at Korean Nuclear Power Plants

Development of Dual Counting Analysis and Exposure Possibility Evaluation for Carbon-14 at Korean Nuclear Power Plants

Comparison of vessel failure probabilities during PTS for Korean nuclear power plants

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of <TEX>$8{\times}10^{19}\;n/cm^2$</TEX>, generating sufficient margin beyond the design life.

Estimation of common cause failure parameters for essential service water system pump using the CAFE-PSA

This paper presents the estimation results of the common cause failure (CCF) parameters of essential service water system (ESWS) pump failure to run for KX nuclear power plant (NPP) in Korea. Until now, the generic values of the CCF parameters have been mainly used in most probabilistic safety assessment (PSA) projects for the Korean NPPs. The PSA results for KX NPP showed that the CCF events of ESWS pump failure to run was identified as one of dominant contributors to its internal event core damage frequency (CDF). Thus, we performed the plant specific detailed CCF analysis to estimate CCF parameters of ESWS pump failure to run for KX NPP with the CAFE-PSA, a program to analyze CCF events in the ICDE database. Reasonable values of CCF parameters were obtained through performing plant specific detailed CCF analysis. The estimated Alpha Factor with three out of three failure criterion was about one half of that for recent US NRC CCF parameters. The re-quantification results on the CDF of KX NPP with the new estimated Alpha Factor showed that originally estimated CDF with generic Alpha Factor decreased by 16.84% and the contribution of the sum of cutsets for the CCF events of ESWS pump failure to run to internal event CDF decreased from 20% to 3.29%.