Wolsong Plant Research Articles

Best-estimate analysis for a MSGTR accident of CANDU-6 plants using the MAAP-ISAAC code

In CANDU-6 type reactors, a multiple steam generator tube rupture (MSGTR) accident is characterized as a reactor building (RB) bypass scenario. Although the probability of an MSGTR accident in the CANDU-6 type Wolsong plants is very low, a direct release of radioactive nuclides to the environment can cause severe radiation exposure to residents around the plants. For this reason, a best-estimate analysis has been carried out based on the currently available best-practice knowledge in the Wolsong plants. The present study is divided into three steps. First, the existing parameter file of generic CANDU-6 plants for the MAAP-ISAAC 4.03 code was revised to incorporate its design-specific features for the Wolsong plants. A comparative analysis between the existing and modified parameter files showed that the progress of a severe accident was delayed or mitigated in the case of the modified parameter file. Secondly, variables related to the steam generator decontamination factors (DFs) were improved. It was confirmed that applying the steam generator DFs based on the ARTIST test significantly reduced the release fraction of each nuclide element to the environment. Finally, severe accident management guidance (SAMG) developed for the Wolsong plants was used to manage the progression of severe accidents and to limit the release of fission products during a MSGTR accident. A mitigation action of closing the main steam safety valves (MSSVs) after meeting a SAMG entry condition can significantly reduce the amount of released fission products compared with simply supplying feedwater to a broken steam generator (SG). To meet the domestic regulatory requirement for the release of Cs-137 to the environment (100 TBq), it was identified that an operator should supply feedwater to the broken SG within 1.5 h after meeting the SAMG entry conditions. The best-estimate analysis methodology and relevant accident management strategy for the mitigation of severe accidents, which have been developed through the study, can be applied to all similar CANDU plants and provide valuable insights into Level 2 probabilistic safety assessment and development of relevant SAMG.

Numerical simulation of tritium migration in groundwater at Wolsong Plant 1 site

On-site groundwater has been reported to be contaminated by the unplanned release of liquid radioactive material at some nuclear power plants. Thus, it is of utmost importance to implement a timely and effective groundwater protection program at a site based on a good understanding of groundwater flow and a reasonable prediction for the potential impact of the unplanned release. In this study, the tritium migration has been simulated based on the modeling result of groundwater at Wolsong Plant 1 site to assess the potential impact of an unplanned release in the form of a leak. The results indicate that the groundwater eventually flowed into the sea, leaving marine activities as the only possible exposure pathway for receptors. Furthermore, the tritium concentrations in groundwater were simulated to be lowered very quickly with groundwater approaching the sea. Therefore, an additional radiation dose contribution due to the discharge of the contaminated groundwater was estimated to be negligible. Particularly, the draining effect of the dewatering sumps was shown to have a strong influence on the groundwater flow characteristics and the simulation of tritium migration, mainly decreasing the spread rate of the contaminated groundwater, which is advantageous to the protection of groundwater.

ESTABLISHMENT OF A SEVERE ACCIDENT MITIGATION STRATEGY FOR AN SBO AT WOLSONG UNIT 1 NUCLEAR POWER PLANT

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

중수로 사용후연료 건전성 검사장비 개발

핵연료는 원자로 운전 중 예기치 못한 상황에서 연료 결함을 초래할 수 있다. 핵연료 결함은 연료봉의 수소화나 이물질에 의한 금속 마모, 그리고 펠렛과 피복관의 상호작용에 의해 피복관이 손상된다. 이렇게 손상된 핵연료의 결함원인을 규명하는 것은 원자력발전의 안전운전에 중요하다고 사료된다. 핵연료가 손상되면 원자로 냉각재가 오염되어 원자로 출력을 낮추거나, 발전소를 정지할 수도 있다. 모든 사용후연료는 건식저장고로 이동 보관되어야 하나, 결함연료는 이동할 수 없으므로 이 연구의 목적은 중수로형 원자로에서 연료가 인출된 후 사용후연료 저장조에서 보관된 연료에 대하여 결함 여부를 판단할 수 있는 기술을 개발하고자 하였다. 이 연구를 통하여 핵종 누설 검출 기술을 이용한 사용후연료 검사기술을 개발하였으며, 이 기술을 월성발전소에 적용함으로써, 검사기술 및 검사시스템에 대한 성능을 입증하였다. Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.

APPLICATION OF SEVERE ACCIDENT MANAGEMENT GUIDANCE IN THE MANAGEMENT OF AN SGTR ACCIDENT AT THE WOLSONG PLANTS

A steam generator tube rupture (SGTR) accident, which is a partial reactor building bypass scenario, has a low probability and high consequences. SAMG has been used to manage the progression of severe accidents and the release of fission products induced by an SGTR at the Wolsong plants. Four of the six SAGs in the SAMG are used to manage the progression of a severe accident induced by an SGTR at the Wolsong plants. The results of the ISAAC code calculation have shown that the proper use the SAMG can stop a severe accident from progressing and keep the reactor building intact during a severe accident. These results confirm that the SAMG is an effective means of managing the progression of severe accidents initiated by an SGTR at the Wolsong plants.

Effects of Time Passage and Distance on Distribution of Carbon-14 among Parts of Rice during Growth Period at Korean CANDU Plant

The pattern of carbon-14 (C-14) distribution among parts of rice growing in the vicinity of the Korean CANDU plant during the growth period was investigated. Six-time samplings of rice and air were performed in seven fields from rice planting to harvest, and the measurements of C-14 content were made by using liquid scintillation counter on the air and each of available parts of the rice such as root, stem, crust and ear. The results illustrated that C-14 showed a relatively even distribution among parts of rice during the growth period implying C-14 accumulation was more dependent on interactions among the parts such as transportation of nutrients than on photosynthesis occurring only in stem that has chlorophyll. Also it was observed that the difference of C-14 concentration between each part of rice and the air decreased with time indicating that the time was needed for C-14 to reach equilibrium between both sides. The radius within which C-14 released from the Wolsong plant could have a significant effect on the C-14 concentrations of the parts was observed to be about 5 km.

An evaluation of a general venting strategy in a CANDU 6 reactor building

If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the containment filtered vent system is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D 2O vapour recovery system which has a 0.76 m (30 in.) diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a Small LOCA and a Large LOCA. The reactor building pressure behavior was analyzed with the ISAAC computer code for four different cases: without venting, 379 kPa(g)/345 kPa(g) (55 psig/50 psig), 345 kPa(g)/276 kPa(g) (50 psig/40 psig) and 345 kPa(g)/207 kPa(g) (50 psig/30 psig) valve open/close pressures. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D 2O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 379 kPa(g)/345 kPa(g) (55 psig/50 psig) is safe and effective. Based on the current study, the strategy of reactor building venting is involved in severe accident management guidance-5.

An Investigation of an In-Vessel Corium Retention Strategy for the Wolsong Pressurized Heavy Water Reactor Plants

An external reactor vessel cooling as a means for an in-vessel retention has been selected as one of the tentative severe accident management strategies for the Wolsong plants, which are typical CANDU 6 reactors. The strategy takes advantage of the plant-specific features: (a) the power density is low, (b) the calandria vessel and the calandria vault have large water volumes, (c) the calandria is always submerged in the water of the calandria vault during a normal operation, (d) the stainless steel layer of the molten corium is negligible even though the unoxidized Zircaloy could form a metal layer, (e) no insulation structure is designed around the calandria vessel, (f) the bottom area of the calandria is large enough to transfer a sufficient amount of the corium decay heat into the calandria vault water, and (g) the water supply from the backup water sources into the calandria vault is available for a long-term external cooling of the calandria. The above design features cause a severe accident progression to be considerably delayed, and they minimize the in-vessel retention issues applied to a certain pressurized light water reactor. Furthermore, the thermal analysis demonstrates that the molten corium on the bottom of the calandria is externally coolable in terms of the critical heat flux, although phenomenological uncertainties still exist. This paper shows the feasibility and the evaluation results of the in-vessel retention strategy via an external vessel cooling for the CANDU 6-type plants, which have not been addressed as yet.

Tritium inventory prediction in a CANDU plant

The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983–1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors.