Soo-Yong Park

Diagnostics of Loss of Coolant Accidents Using SVC and GMDH Models

As a means of effectively managing severe accidents at nuclear power plants, it is important to identify and diagnose accident initiating events within a short time interval after the accidents by observing the major measured signals. The main objective of this study was to diagnose loss of coolant accidents (LOCAs) using artificial intelligence techniques, such as SVC (support vector classification) and GMDH (group method of data handling). In this study, the methodologies of SVC and GMDH models were utilized to discover the break location and estimate the break size of the LOCA, respectively. The 300 accident simulation data (based on MAAP4) were used to develop the SVC and GMDH models, and the 33 test data sets were used to independently confirm whether or not the SVC and GMDH models work well. The measured signals from the reactor coolant system, steam generators, and containment at a nuclear power plant were used as inputs to the models, and the 60 sec time-integrated values of the input signals were used as inputs into the SVC and GMDH models. The simulation results confirmed that the proposed SVC model can identify the break location and the proposed GMDH models can estimate the break size accurately. In addition, even if the measurement errors exist and safety systems actuate, the proposed SVC and GMDH models can discover the break locations without a misclassification and accurately estimate the break size.

OVERVIEW OF CONTAINMENT FILTERED VENT UNDER SEVERE ACCIDENT CONDITIONS AT WOLSONG NPP UNIT 1

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

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.

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.

The iROCS Approach to Mitigating Beyond-Design-Basis External Events

An extended loss of all electric power occurred at the Fukushima Dai-ichi nuclear power plant by a large earthquake and subsequent tsunami. This event led to a loss of reactor core cooling and containment integrity functions at several units of the site, ultimately resulting in large release of radioactive materials into the environment. In order to cope with beyond-design-basis external events (BDBEEs), this study proposes the iROCS (integrated, RObust Coping Strategies) approach. The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. The plant damage conditions considered in the iROCS approach include combinations of the following conditions of the critical safety functions: (1) an extended loss of AC power, (2) an extended loss of DC power (loss of the monitoring and control function at control rooms), (3) a loss of RCS inventory, and (4) a loss of secondary heat removal. From a case study for an extreme damage condition, it is shown that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against extreme damage conditions of the site and accident scenarios of the plant.Copyright