Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F46356088%3A_____%2F18%3AN0000013" target="_blank" >RIV/46356088:_____/18:N0000013 - isvavai.cz</a>

Result on the web

<a href="http://dx.doi.org/10.1016/j.anucene.2018.03.040" target="_blank" >http://dx.doi.org/10.1016/j.anucene.2018.03.040</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.anucene.2018.03.040" target="_blank" >10.1016/j.anucene.2018.03.040</a>

Result language

angličtina

Original language name

Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

Original language description

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen Ill + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for WER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the "3 layers" configuration, where the "focusing effect" may cause higher heat fluxes than in steady-state (due to transient "thin" metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m(2)) whereas the first type provides the lowest heat fluxes (around 500 MW/m(2)) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m(2) which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

Czech name

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Czech description

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Type

J_imp - Article in a specialist periodical, which is included in the Web of Science database

CEP classification

—

OECD FORD branch

20305 - Nuclear related engineering; (nuclear physics to be 1.3);

Project

—

Continuities

R - Projekt Ramcoveho programu EK

Publication year

2018

Confidentiality

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Name of the periodical

Annals of Nuclear Energy

ISSN

0306-4549

e-ISSN

—

Volume of the periodical

119

Issue of the periodical within the volume

September

Country of publishing house

GB - UNITED KINGDOM

Number of pages

10

Pages from-to

36-45

UT code for WoS article

000437819900004

EID of the result in the Scopus database

2-s2.0-85046369762