Wolsong Nuclear Power Plant Research Articles

DEVELOPMENT OF AN IMPROVED FARE TOOL WITH APPLICATION TO WOLSONG NUCLEAR POWER PLANT

In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.

Calibration method of liquid zone controller using the ex-core detector signal of CANDU 6 reactor

The Phase-B test (low-power reactor physics test) is one of the commissioning tests for Canada Deuterium Uranium (CANDU) reactors that ensures the safe and reliable operation of the core during the design lifetime. The Phase-B test, which includes the approach to the first criticality at low reactor powers, is performed to verify the feasibility of the reactor’s physics design and to ensure the integrity of the control and protection facilities. The commissioning testing of pressurized heavy water moderated reactors (PHWRs) is usually performed only once (at the initial commissioning after construction). The large-scale facilities of the Wolsong nuclear power plant (NPP) Unit 1 have been gradually improved since May 2009 to extend its lifetime. The refurbishment was completed in April 2011 – then this NPP has been in operation again. We discusses the new methodology and measurement system that uses an ex-core detector signal for liquid zone controller (LZC) calibration of the Phase-B test instead of conventional methods. The inverse kinetic equation in the reactivity calculator is modified to treat the 17 delayed neutron groups including 11 photo-neutron fractions. The signal acquisition resolution of the reactivity calculator was enhanced and installed reactivity calculating module by each channel. The ex-core detector was confirmed to be applicable to a large reactor core, such as the CANDU 6 by comparison with the in-core flux detector signal. A preliminary test was performed in Wolsong NPP Unit 2 to verify the robustness of the reactivity calculator. This test convincingly demonstrated that the reactivity calculator works well. Commissioning testing of Wolsong NPP Unit 1was performed after successful refurbishment in June 2011. The LZC calibration test was completed successfully, and the requirement of the test was satisfied. The test difference was biased by approximately −6% between 20% and 70% of the liquid zone level. Moreover, the test was completed within 4h compared with about 48h required for Wolsong NPP Units 2, 3, and 4.

월성 원자력발전소 중수 누출에 대한 언론의 부정적 보도 분석 : 주요 일간지의 1999년 10월 한 달간 월성 원자력발전소 중수 노출 사고 기사 내용 분석

원자력발전은 우리나라 전력의 약 30%를 공급하는 중요한 전력 공급 수단중의 하나이다. 그러나 2011년 3월 11일 일본 동북부를 강타한 규모 9.0의 대지진으로 인해 발생한 후쿠시마 원전사고는 '국내 원자력 발전은 불안하다'는 불안감을 조성하고 있다. 이러한 불신의 요인은 여러 가지가 있지만 부정확한 언론보도도 일조하고 있다. 이 논문은 1999년 10월 4일 발생한 월성원전 3호기 중수누출에 관한 당시 언론보도 내용 분석을 통해 문제점을 파악하고, 바람직한 언론보도에 대해 알아보기로 한다. Nuclear power provides 30% of our country's power, which acts as one of the most important power sources. But on March 11, 2011, the earthquake that hit Northeast Japan with a 9.0 magnitude, known as the Fukushima Reactor Leak Incident has created fear in the public's mind that 'nuclear power is unstable'. The reason for such distrust are many but inaccurate reports of the incident by the media has added to the fear. This paper will analyze the contents of the media report of the heavy water leakage in reactor 3 at the Wolsong Nuclear Power Plant on October 4, 1999 to discover the problematic areas and ascertain a more appropriate method of media coverage.

Three-Dimensional Dispersion Modeling of Liquid Radioactive Effluents around the Wolsong Nuclear Power Plant Site in Korea

Three-Dimensional Dispersion Modeling of Liquid Radioactive Effluents around the Wolsong Nuclear Power Plant Site in Korea

중수로 연료관 건전성 평가시스템(WIES) 개발

가압중수로 연료관은 CSA N 285.4 기술기준에 따라 주기적인 가동중검사가 수행되고 있다. 가동중검사시 발견된 결함이 CSA N 285.4 의 허용기준을 초과하는 경우, 결함 연료관의 계속 운전을 위해 가동적합성 평가를 허용하고 있다. 캐나다 COG(CANDU Owner's Group)를 중심으로 중수로 연료관의 결함 건전성 평가 기술기준인 CSA N285.8 이 개발되었다. 본 논문에서는 CSA N285.8 을 기반으로 연료관의 건전성 평가시스템 WIES(Wolsong In-service Evaluation System)를 개발 하였다. 중수로 연료관의 가동중검사시 결함이 발견되는 경우, 개발된 시스템은 신속하고 정확한 건전성 평가를 수행하여 계획예방 정비기간의 연장을 방지하여 원전 이용률 향상에 기여할 것으로 판단된다.

Gasification characteristics to 14CO2 of 14C radionuclide desorbed from spent resin by phosphate solutions

The removal characteristics of H14CO3− ions from IRN-150 mixed resin contaminated with 14C radionuclide and the gasification effects of 14C radionuclide on 14CO2 are investigated in this study. The stripping solutions used for the removal of 14C from spent resin are NaNO3, Na3PO4, NH4H2PO4, and H3PO4. The influence of the stripping solution concentration on the desorption characteristics of an inactive HCO3 ion into a stripping solution from IRN-150 mixed resin and the gasification of this ion to CO2 is analyzed. The gasification behavior to CO2 using NaOH, HNO3, and HCl was also compared to that of phosphate solution. Spent resin stored in Wolsong nuclear power plant is used to evaluate the gasification characteristics of 14C radionuclide to 14CO2. Gamma radionuclides such as 137Cs and 60Co in residual striping solutions after desorption experiments are analyzed.

Estimation of Tritium and Helium Inventory in the Tritium Handling System in Korea

In Korea, the Wolsong Tritium Removal Facility (WTRF) is under construction to reduce the amount of tritium present in the moderator and coolant of the CANDU type Wolsong nuclear power plants. Recently, a study on the tritium handling system for recovery of the tritium collected from the WTRF was started. Some tritium would enter the steel of the container walls and subsequently decay to helium. This helium can deteriorate the mechanical properties of the material of the tritium handling system. To evaluate the tritium and helium inventory in the stainless steel wall of this system, the time-dependent diffusion equation was developed, solved and the results are presented in this paper. These results were compared to previous work that evaluated the tritium inventory in the stainless steel wall of 50-L tritium containers. Tritium and helium concentration profiles and the corresponding inventories were evaluated with respect to the various parameters such as exposure time, temperature, and partial pressure. After 24 years, the helium inventory in the wall of the tritium handling system exceeds the tritium inventory.

Benchmark Calculation of WIMS/RFSP against Physics Measurement Data of Wolsong Nuclear Power Plants

Benchmark calculations of Canada deuterium uranium (CANDU) reactor physics design and analysis codes have been performed for a lattice code WIMS-AECL, a supercell code DRAGON, and a core analysis code RFSP by using the physics measurement data of Wolsong nuclear power plants. In this study, the lattice and reactivity device models were examined based on Wolsong-2 measurement data for the criticality and reactivity device worth. Sensitivity calculations were also performed for the number of energy groups and the cross-section library. Using the lattice and reactivity device models obtained from the Wolsong-2 calculation, the benchmark calculations were extended to the Wolsong-3 and Wolsong-4 plants. Compared to a previous study, this study showed that the results of the criticality and reactivity device worth calculations were improved when the material data were updated and the exact two-group cross sections were used. For the three nuclear power plants, the calculated core reactivity was within 0.2% Δk of criticality. The zone controller unit reactivity worth was estimated to have a maximum error of ~8%. The total reactivity worth of other reactivity control devices was consistent with the measurement data within 13%. The root-mean-square error of the flux distribution calculation was <12% when compared with flux scans performed during Phase B physics tests. In conclusion, the CANDU physics design and analysis codes used in this benchmark study predicted the physics parameters within the allowed uncertainty level of the measurement data.

SMV model-based safety analysis of software requirements

Fault tree analysis (FTA) is one of the most frequently applied safety analysis techniques when developing safety-critical industrial systems such as software-based emergency shutdown systems of nuclear power plants and has been used for safety analysis of software requirements in the nuclear industry. However, the conventional method for safety analysis of software requirements has several problems in terms of correctness and efficiency; the fault tree generated from natural language specifications may contain flaws or errors while the manual work of safety verification is very labor-intensive and time-consuming. In this paper, we propose a new approach to resolve problems of the conventional method; we generate a fault tree from a symbolic model verifier (SMV) model, not from natural language specifications, and verify safety properties automatically, not manually, by a model checker SMV. To demonstrate the feasibility of this approach, we applied it to shutdown system 2 (SDS2) of Wolsong nuclear power plant (NPP). In spite of subtle ambiguities present in the approach, the results of this case study demonstrate its overall feasibility and effectiveness.

Deactivation of hydrophobic catalysts for a hydrogen isotope exchange: Application of the time-on-stream theory

A recycle reactor was built for the purpose of characterizing newly developed hydrophobic catalysts for a hydrogen isotope exchange. The catalytic rate constants of two types of hydrophobic catalysts were measured at a 100% relative humidity. The catalytic rate constants were measured at 60 °C for 28 days and both the catalysts showed very high initial catalytic rate constants. The measured deactivation profile showed that the catalytic rate constants of both the catalysts were almost identical for 28 days. The deactivation of the catalysts was modelled based upon the time-on-stream theory. The deactivation profiles of the catalysts were estimated by using the model for a period of three years. The results showed that both the catalysts had a good exchange capacity for hydrogen isotopes and they could be applicable to a tritium removal facility that will be built at the Wolsong nuclear power plants in the near future.

Benchmarking MCNP and WIMS/RFSP Against Measurement Data—II: Wolsong Nuclear Power Plant 2

Benchmark calculations of the Canada deuterium uranium reactor design and analysis codes were performed for the Monte Carlo and conventional methods using Phase-B measurement data of the Wolsong Nuclear Power Plant 2. In this study, the benchmark calculations were done for the criticality, boron worth, reactivity device worth, and flux scan. For the benchmark calculation of the Monte Carlo method by MCNP-4B, the criticality was estimated within 4 mk. The reactivity worth of the control devices was consistent with the measurement data within 15%. For the benchmark calculation of the conventional method composed of WIMS-AECL, SHETAN, and RFSP, the criticality was also predicted within 4 mk. The reactivity device worth was generally consistent with the measured data except for the strong absorbers such as shutoff rods and mechanical control absorbers. The results of the flux distribution calculations were also satisfactory for both code systems.

Software Safety Analysis of Digital Protection System Requirements Using a Qualitative Formal Method

The safety analysis of requirements is a key problem area in the development of software for the digital protection systems of a nuclear power plant. When specifying requirements for software of the digital protection systems and conducting safety analysis, engineers find that requirements are often known only in qualitative terms and that existing fault-tree analysis techniques provide little guidance on formulating and evaluating potential failure modes. A framework for the requirements engineering process is proposed that consists of a qualitative method for requirements specification, called the qualitative formal method (QFM), and a safety analysis method for the requirements based on causality information, called the causal requirements safety analysis (CRSA). CRSA is a technique that qualitatively evaluates causal relationships between software faults and physical hazards. This technique, extending the qualitative formal method process and utilizing information captured in the state trajectory, provides specific guidelines on how to identify failure modes and the relationship among them. The QFM and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital protection system example.

Fault tree construction of hybrid system requirements using qualitative formal method

When specifying requirements for software controlling hybrid systems and conducting safety analysis, engineers experience that requirements are often known only in qualitative terms and that existing fault tree analysis techniques provide little guidance on formulating and evaluating potential failure modes. In this paper, we propose Causal Requirements Safety Analysis (CRSA) as a technique to qualitatively evaluate causal relationship between software faults and physical hazards. This technique, extending qualitative formal method process and utilizing information captured in the state trajectory, provides specific guidelines on how to identify failure modes and relationship among them. Using a simplified electrical power system as an example, we describe step-by-step procedures of conducting CRSA. Our experience of applying CRSA to perform fault tree analysis on requirements for the Wolsong nuclear power plant shutdown system indicates that CRSA is an effective technique in assisting safety engineers.

Analysis of Microdosimetric Quantities of Mixed Radiation Fields in the Reactor Building at the Wolsong Nuclear Power Plant

A commercially available tissue equivalent proportional counter (TEPC), model FW-AD1, was used as microdosimetric detector to measure the dose equivalent rates at several workplaces in the reactor building of a pressurized heavy water reactor, the Wolsong Nuclear Power Plant unit 3 in Korea. The microdosimetric quantities including the lineal-energy distributions and the dose-mean lineal energy were determined by unfolding the measured data, and then the mean quality factors and the dose equivalent rates were evaluated. The mean quality factors at places where neutron dose rates were significant were in the range of 2.2 to 7.1, while those values were near unity at the places where gamma-ray dose rates dominated. The lineal energy spectra of the reference radiation fields at the Korea Atomic Energy Research Institute (KAERI) were also measured for comparisons. Particularly in the vicinity of the primary heat transport pump, the dose-mean lineal energy and the mean quality factor appeared similar to those values of the D20 moderated 252Cf reference field. The tool and the analysis method can be applied to characterization of other mixed radiation fields.

Approaches for Reducing Carbon-14 Stack Emissions from Korean CANDU® Nuclear Power Plant

Relatively high carbon-14 emissions, which occurred at Wolsong Nuclear Power Plant during 1998 and 1999, made the site staff to implement several operational improvements: (1) restriction of the maximum in-service time of moderator ion exchange (IX) resin columns to 80 days, (2) discontinuation of the practice of re-using ‘used’ IX resin columns for removing the gadolinium nitrate at star-up, and (3) reduction of the frequency and volume of the moderator cover gas purging through increased O2 additions to the cover gas. In this way, the emission rates returned to normal levels during the remainder of 1999 and 2000. To optimize the use of the moderator IX resin column, the special surveillance was carried out at Wolsong Unit 3. This showed that only the 14C stack emission data were a practical indicator for exhaustion of IX resin column. From this data, a criterion was derived: IX resin column can be put in service for 100 days unless a 14C emission rate higher than about 20GBq-month-1 (0.4Ci-month-1) takes place for two consecutive weeks. Finally it was suggested that using a portable 14C scrubber could benefit for reducing the 14C stack emissions, especially those due to leakage of the moderator cover gas.

Approaches for Reducing Carbon-14 Stack Emissions from Korean CANDU Nuclear Power Plant

Relatively high carbon-14 emissions, which occurred at Wolsong Nuclear Power Plant during 1998 and 1999, made the site staff to implement several operational improvements: (1) restriction of the maximum in-service time of moderator ion exchange (IX) resin columns to 80 days, (2) discontinuation of the practice of re-using ‘used’ IX resin columns for removing the gadolinium nitrate at star-up, and (3) reduction of the frequency and volume of the moderator cover gas purging through increased O2 additions to the cover gas. In this way, the emission rates returned to normal levels during the remainder of 1999 and 2000. To optimize the use of the moderator IX resin column, the special surveillance was carried out at Wolsong Unit 3. This showed that only the 14C stack emission data were a practical indicator for exhaustion of IX resin column. From this data, a criterion was derived: IX resin column can be put in service for 100 days unless a 14C emission rate higher than about 20GBq-month-1 (0.4Ci-month-1) takes place for two consecutive weeks. Finally it was suggested that using a portable 14C scrubber could benefit for reducing the 14C stack emissions, especially those due to leakage of the moderator cover gas.

An Estimate of Carbon-14 Inventory at Wolsong Nuclear Power Plant in the Republic of Korea

In this paper, carbon-14 inventory of Wolsong Nuclear Power Plant (NPP) which has CANDU type reactors was estimated for 1984–2002. The carbon-14 inventory can be estimated based on a simple relation: [carbon-14 inventory]=[carbon-14 production]—[carbon-14 emission]. The second term can be calculated whereas the third term should be measured. However, all the measured data for carbon-14 emissions at Wolsong NPP are not available. So an empirical relation is employed: [98.283%]=[100%]—[1.717%]. From this relation, the theoretical values for the total production, emission and inventory of carbon-14 at Wolsong NPP for 1984–2002 are estimated to be 694.2 TBq, 11.9 TBq and 682.3 TBq, respectively. Comparison of carbon-14 emission levels at Wolsong NPP and Pickering A NPP in terms of rates for theoretical productions and measured emissions of carbon-14 per unit power generated shows that although more carbon-14 is produced, less carbon-14 is released at Wolsong NPP except for 1999. Furthermore, comparison of the extend of carbon-14 emissions at both NPPs in terms of ratios of measured to theoretical carbon-14 emissions indicates that the control and regulation of carbon-14 emission are more effectively performed at Wolsong NPP.

An Estimate of Carbon-14 Inventory at Wolsong Nuclear Power Plant in the Republic of Korea

In this paper, carbon-14 inventory of Wolsong Nuclear Power Plant (NPP) which has CANDU type reactors was estimated for 1984–2002. The carbon-14 inventory can be estimated based on a simple relation: [carbon-14 inventory]=[carbon-14 production]—[carbon-14 emission]. The second term can be calculated whereas the third term should be measured. However, all the measured data for carbon-14 emissions at Wolsong NPP are not available. So an empirical relation is employed: [98.283%]=[100%]—[1.717%]. From this relation, the theoretical values for the total production, emission and inventory of carbon-14 at Wolsong NPP for 1984–2002 are estimated to be 694.2 TBq, 11.9 TBq and 682.3 TBq, respectively. Comparison of carbon-14 emission levels at Wolsong NPP and Pickering A NPP in terms of rates for theoretical productions and measured emissions of carbon-14 per unit power generated shows that although more carbon-14 is produced, less carbon-14 is released at Wolsong NPP except for 1999. Furthermore, comparison of the extend of carbon-14 emissions at both NPPs in terms of ratios of measured to theoretical carbon-14 emissions indicates that the control and regulation of carbon-14 emission are more effectively performed at Wolsong NPP.

Environmental distribution and behavior of 3H and 14C around Wolsong Nuclear Power Plants.

To estimate of the dispersion trend of 3H and 14C discharged from the Wolsung Nuclear Power Plants, the present level of 3H and 14C in environmental samples in the vicinity of the Wolsong site was studied. Tree-ring cellulose analyses were carried out for retrospective evaluation of 3H and 14C in the environment around the Wolsong Nuclear Power Plants. 3H released from the Wolsong Nuclear Power Plants has affected an area up to a 25-km radius from the site, while almost all 14C was deposited within a 2-km radius, reaching to a natural level over 2 km. Organically bound tritium concentrations in tree rings were strongly correlated with the gaseous tritium discharge rate, while delta14C (excess) in tree rings ranged from 204 per thousand to 460 per thousand, which did not significantly vary with year.

Qualitative formal method for requirements specification and validation of hybrid real-time safety systems

The difficulties that engineers have in understanding and applying the quantitative methods in an abstract requirements phase are major obstructions in using formal methods for hybrid real-time safety systems. While formal methods technology in safety-critical systems can help increase confidence of software, the difficulty and complexity in using them can cause another hazard. The authors have proposed a framework for requirements engineering, called the qualitative formal method (QFM) for the specification and validation of hybrid real-time safety systems. The QFM emphasises the idea of a causal and qualitative reasoning in formal methods to reduce the difficulty of specifying and validating the software requirements of hybrid safety systems. They use the qualitative formal languages, Compositional Modelling Language, and Causal Functional Representation Language in particular, to specify hybrid system dynamics and the required behaviour, respectively. The system behaviour has been simulated by the Device Modelling Environment, and validated against the required behaviour. Using the Shutdown System 2 of Wolsong nuclear power plants as a realistic example, they demonstrate the effectiveness of their approach.