atw 2018-02

inforum

atw Vol. 63 (2018) | Issue 2 ı February

Investigation of Conditions Inside the

Reactor Building Annulus of a PWR

Plant of KONVOI Type in Case of Severe

Accidents with Increased Containment

Leakages

Ivan Bakalov and Martin Sonnenkalb

1 Introduction and analysis method The severe accident at Fukushima Daiichi NPP resulted in

severe core damage and significant releases of hydrogen and radioactive materials from primary containment boundary

into or through the reactor buildings of three out of the six reactors (units 1 to 3). Based on analyses of the accident

progression it was realized that accidentally increased leaks from the inertized containment contributed to the

radionuclide and hydrogen release into the reactor building, thus leading to hydrogen explosions, severely damaging

the reactor building constructions.

The Fukushima Daiichi accident triggered

worldwide stress tests and

re-assessments of the NPP plant

safety. In Germany the process

resulted in an improvement and

extension of the existing severe accident

management (SAM) concept

by both additional preventive and

mitigative measures. The main improvements

in the mitigative domain

is a new concept of severe accident

management guidelines (SAMG) with

strategies and procedures intended to

be used by the plant crisis team for

mitigation of the consequences of

severe accidents. The SAMG concept

follows relevant recommendations

of the German Reactor Safety Commission

RSK [1].

Analyses of the hydrogen as well

as aerosol and noble gas behaviour

in case of increased containment

leakages into the reactor building

annulus of a German PWR KONVOI

reference plant under severe accident

conditions have been performed using

the GRS lumped parameter code

COCOSYS. The investigation carriedout

focusses on the assessment of the

efficiency of newly developed SAM

measures as described in the new

SAMG handbook or some measures

proposed in addition for a PWR

reference plant of KONVOI type. The

assessed strategies are related to the

mitigation of challenging conditions

inside the reactor building (RB)

annulus due to design based and

increased containment leakages

during severe accidents.

The analyses are based on previous

GRS investigations of the hydrogen

mitigation concept with passive autocatalytic

recombiners (PAR) inside

the PWR KONVOI containment [2] as

well as the reassessment of the effectiveness

of the filtered containment

venting concept of PWR KONVOI [3].

The main findings contribute to

further improvement of the planned

mitigative SAM measures in case of

enhanced containment leakages into

the reactor building annulus under

severe accident conditions.

1.1 COCOSYS plant model

The COCOSYS nodalisation scheme

of the PWR KONVOI plant with focus

on the RB annulus is presented in

Figure 1. The nodalisation of the

containment and the RB annulus is

developed in such a way that thermal

and gas stratification processes

expected under accident conditions,

local and global convection flows

between the compartments, and longterm

convection processes inside

the containment could be simulated

appropriately. Therefore, a refined

subdivision of the containment compartments

and RB annulus rooms and

free space was chosen. The model

considers all relevant gaseous and

liquid flows through different compartment

connections such as free

openings, fire protection doors, burst

membranes, drainages, etc. For the

purpose of heat and mass transfer

modelling inside the containment

and the RB annulus heat structures

representing the walls, floors, ceilings

and metal internals are introduced

into the model. With all these features

the model adequately represents all

relevant design specific features of the

PWR KONVOI reference plant – both

inside the containment as well as the

RB annulus.

The containment has a total free

volume of 70,000 m 3 . It is subdivided

into four areas which can have

different convection flow regimes

depending on the initial event of a

sequence and the break/discharge

location. The first area represents the

containment compartments, in which

the reactor pressure vessel and the

steam generators are located. The

| | Fig. 1.

COCOSYS nodalisation scheme of the RB annulus and location of containment penetrations through the containment steel shell.

85

ENVIRONMENT AND SAFETY

Environment and Safety

Investigation of Conditions Inside the Reactor Building Annulus of a PWR Plant of KONVOI Type in Case of Severe Accidents with Increased Containment Leakages ı Ivan Bakalov and Martin Sonnenkalb

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