Pressure suppression system influence on vacuum vessel thermal-hydraulics and on source term mobilization during a multiple first Wall – Blanket pipe break

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Abstract

Being the Vacuum Vessel Pressure Suppression System (VVPSS) one of the most important passive safety systems to be foreseen in DEMO plant, design and integration challenges have to be faced to ensure that best performance within safety requirements are always achieved. In this framework, parametric safety analyses have been performed to support VVPSS design activities; in particular to determine the minimum flow area required by the suppression system pipework to limit the vacuum vessel pressure below the limit imposed as a requirement by design.

The selected Postulated Initiating Event (PIE) is a double-ended guillotine break in the Primary Heat Transfer System (PHTS) feeding pipe of the Breeding Zone (BZ) during plasma activity. Coolant is discharged inside the VV upper port volume and an unmitigated disruption occurs, causing further breaks in the first wall (FW) cooling channels. Considering that limiters could be introduced in the future design of the EU DEMO reactor to prevent damages to plasma-facing components, the same parametric study has been performed considering limiters accident mitigation effects.

Because discharge flow area toward the suppression pool could also affect the source terms mobilization, the transport of radioactive products (e.g. tritium, tungsten dust and activated corrosion products) has been simulated with the fusion version of the MELCOR code (ver. 1.8.6).

Introduction

The VV pressure suppression system is one of the most important safety passive systems to be foreseen in the DEMO plant, since it limits the allowable VV pressure in the event of in-vessel LOCA and confines radioactive sources in the system [1]. For this reason, it is classified as SIC-1 system [2] and its operation should be optimized, ensuring that the evolving system design will satisfy safety requirements and that it is optimized for the safety of personnel, public and environment [[3], [4], [5]].

A Functional Failure Modes and Effects Analysis (FFMEA), performed for all EU-DEMO key systems, led to the selection of 21 Postulated Initiating Events (PIEs) that envelope all identified failures [6]. In particular, in-vessel LOCA has been classified among the most representative events in terms of challenging conditions for plant safety, because it could cause substantial damage to the VV structure. In past activities, some in-vessel LOCA analyses have been conducted in this sense, choosing breaks in the FW channels [7,8] or in the divertor cassettes cooling system [9] as PIE.

In the present work, the selected postulated initiating event is a break in a BZ-PHTS feeding pipe during plasma activity. This PIE, rather than FW-PHTS LOCA, has been considered because of the larger inventory in the BZ-PHTS and because of the larger break flow area, which could cause severe overpressure conditions in the VV. Parametric analyses have been performed to determine the minimum flow area of VVPSS rupture disk pipes needed to keep the VV pressure below the design limit of 2 bar.

Section snippets

The WCLL EU DEMO reference design

The EU-DEMO reference design adopted for this LOCA analysis has 1923 MWth of fusion power. The WCLL breeding blanket concept, one of the candidate option for the future EU DEMO [10], consists of 16 sectors in the toroidal direction. Each sector includes 3 segments in the outboard blanket (OB) and 2 segments in the inboard (IB). The single segment is constituted of about 100 breeding cells distributed along the poloidal direction, following a Single Module Segment (SMS) approach. The reference

Accident description

Parametric accident analyses of an in-vacuum vessel LOCA have been performed to determine the minimum flow area of VVPSS rupture disk pipes needed to maintain VV pressure below 2 bar. Two different accident scenarios have been investigated:

  • A “worst case” accident scenario involving the simultaneous failure of both BZ and FW PHTS;

  • A “baseline case” scenario in which limiters are foreseen to adsorb the plasma energy deposited by an unmitigated plasma shutdown, preventing the failure of

MELCOR modelling

The fully integrated, engineering-level thermal hydraulics analysis code, MELCOR (v. 1.86) [18] with modifications for fusion reactor safety applications [19] has been used to evaluate accident consequences for the selected scenarios. MELCOR has been chosen because of its capability of consistently simulating coolant thermal-hydraulic behavior and radionuclide and aerosol transport in nuclear facilities and reactor cooling systems during severe accident scenarios.

Results of accident analysis

The parametric study includes 11 simulations performed by varying the rupture discs flow area from 1 m2 (RD_1.0) to 2 m2 (RD_2.0) for both “worst case” and “baseline case” accident scenarios.

Summary and conclusions

The aim of these two simulations, which events have been classified as DBA, was to define the needed flow area of VVPSS suppression pipework to limit the VV pressure below the limit imposed by safety requirements. In both cases, a pipe rupture was initiated by opening a connection between the BZ feeding pipe and the upper port volume. Water injection inside the plasma volume, in turn, causes an unmitigated plasma shutdown transient. In the case in which limiters can mitigate the effects of a

Declaration of Competing Interest

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

Acknowledgments

This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

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