Development and Application of Dose Assessment Methodology During IS-LOCA in BWR Plant

A dose assessment methodology was developed for IS-LOCA (Interface System – Loss of Coolant Accident) event, which is one of the design basis accidents in BWR plants, and a dose assessment at site boundary was implemented for the accident scenario by applying the evaluation method. The main objective of the assessment is to review a conventional conservative dose assessment model in the accident scenario. The conventional conservative model is a method due to considering all fission products (FP) in liquid migrate to airborne (i.e. 100%) when reactor water is discharged from piping break area, while the developed new methodology is a method to evaluate transition behavior of fission products precisely by using flash fraction (FF) from liquid to airborne and consider it. As the result of dose assessment, it was confirmed that the calculated dose during IS-LOCA event was about one-fifteith reduced by adopting the new model from the conventional model. The content of this paper includes: • Background and purpose that led to the dose assessment methodology development during IS-LOCA. • An evaluation method by using discharged amount of reactor water at piping break area and flash fraction (FF) from liquid to airborne based on Regulatory guide 1.183 published by U.S. NRC. • Another evaluation method that takes into account the fluctuation due to depressurization of reactor pressure per time step from the start of accident until completion of isolation at piping break area in addition to the above method. • Comparison between dose results at site boundary based on these methods and conventional conservative model. This results will give a precise information for public and it will be useful for making emergency evacuation plan.

Application of Environmentally Assisted Fatigue Transient Ramp Times Optimization in Design Evaluations and Fatigue Monitoring Algorithms

As part of the effort for a nuclear plant to undergo license renewal, the effect of reactor water environment on fatigue life must be addressed for limiting component locations. One method to incorporate the effects of reactor water environment into the fatigue evaluations of metal components is to apply an environmental fatigue penalty factor (Fen) to the partial usage factor obtained from the design fatigue curve for each stress cycle. Fatigue evaluations have historically been performed by assuming that temperature transient loads occur at conservatively high rates to maximize the stress response and corresponding fatigue usage values. However, with consideration of reactor water environmental effects on fatigue, transients with slower rates generally produce higher Fen values that could potentially result in higher environmental fatigue usage values than transients with identical temperature changes but faster rates. A generic parametric study was performed in MRP-218 to characterize limiting transient ramp rates with respect to environmental fatigue usage for a range of piping geometry and material configurations. This paper describes the application of the parametric study results to optimize thermal hydraulic and stress response modeling assumptions with respect to transient rates and downstream effects on environmental fatigue results in both design and monitoring fatigue evaluations.

Upgrade of Accident Sequence Model of PRA for LOCA Inside PCV

In Japan, Risk-informed activity using PRA is becoming increasingly active and Reactor Oversight Process (ROP) will be introduced in 2020. Then, PRA has attracted more attention, and it is necessary to immediately develop its infrastructures. Therefore, as the first step, TEPSYS/TEPCO conduct to upgrade the internal event level 1 PRA that is the basis of all events targeting unit 7 (ABWR) of TEPCO Kashiwazaki-Kariwa Nuclear Power Plant. In particular, LOCA inside PCV for all piping connecting RPV including instrumentation and control piping reclassified based on its break size, break location, success or failure of the vapor suppression function, the accident progression during mission time, and PCV control procedures. Especially, the finding presented in this paper address the newly grouping that break location is classified into not 2 (steam phase/liquid phase) groups [1] but 3 groups as below. ✓ RPV nozzles are located above L8 reactor water level (steam phase) ✓ RPV nozzles are located between TAF and L8 reactor water level (steam phase / liquid phase) ✓ RPV nozzles are located below TAF reactor water level (liquid phase) In the Results, completeness of the accident sequence for LOCA inside PCV was improved to be applied to risk-informed activity and the proactive safety review.