atw 2018-05v6


atw Vol. 63 (2018) | Issue 5 ı May


and its both follow up phases (latest

on THAI TH-27 test and THAI HR-49).

The past activities in the severe

accident for both Czech NPPs are were

mostly focused on the enhancement

of the safety using only existing

systems on site, but out of their operational

conditions. The Fukushima Daiichi

accident significantly changed the

approach to the solution of severe

accidents at the Czech NPPs and the

implementation of new, and only for

severe accident conditions dedicated,

systems was fully accepted as the

necessary step forward in the safety

enhancement. The proposals were not

started from the scratch, but both NPP

were evaluated within the Stress

Tests, which reports were issued to

the Czech Republic State Office for

Nuclear Safety (SONS), and based on

these reports the national Stress Test

report [A] was prepared and evaluated

within the evaluation process

under the ENSREG leadership resulting

in the forming of the National

Action Plan [B]. Concerning the

severe accident issues the eight main

areas were defined (as a selection of

those most important ones from much

higher number of issues)

• Increase of the capacity of the system

for liquidation of emergency

hydrogen (action 46 – Dukovany

NPP, action 47 – Temelín NPP) –

this title is undertaken from the

official National Action Plan and it

means increase of recombination

power of emergency hydrogen as

the term “liquidation” is unusual in

this meaning

• Cooling of the melt from the outside

of RPV (action 48 – Dukovany


• Recriticality (action 61 – both


• Control room habitability (actions

58, 31, and 51 – both NPPs)

• The means for maintaining containment

integrity due to overpressure

(actions 46 – 50 – both


• Corium in/ex vessel cooling

( actions 48, 49, 50 – Temelín NPP)

• Extension of SAMGs for shutdown/

severe accident in SFP ( action 56 –

for both NPPs)

• System setup of training (drills),

exercises and training for severe

accident management according to

SAMG, including possible solution

of multi-unit severe accident

The solutions of some topics listed

above are independently described

in following subchapters with the

pointing out the contribution of the

UJV to their solution.

3.1 Increase of capacity of

system for emergency

hydrogen removal

The UJV performed a full analytical

support for the increased design of

the hydrogen removal system. The

analytical program consisted from

several steps. The first step contained

the integral analysis of the severe

accident progression with the aim to

define mass and energy source into

the containment for selected severe

accident scenarios. The scenario

selection was based on several conditions

like – a location of the hydrogen

source in the containment, its

potential intensity, an operation of

the containment spray system, or conditions

of the molten corium concrete

interaction. Generally three initiating

events were selected and overall six

integral analyses performed with the

MELCOR 1.8.6 code. The sources to

the containment were later used in

the stand alone analyses of the containment

response using the very

detailed model of the containment

again for the MELCOR 1.8.6 code.

Two kinds of analyses were performed

– first the hydrogen risk evaluation

based on the Sigma and Lambda

criteria, which were used for the first

proposal of PAR design, second the

optimization analyses with the aim to

fullfil predefined succes criteria – an

elimination of Lambda criterion in all

parts of containment, an elimination

of Sigma in practically all parts of

containment (small individual spaces

allowed for temporary occurrence),

global and local concentration limits

after recalculation of hydrogen concentration

in dry air. The design of

the PARs of the NIS vendor was

succesfully implemented at both units

of the Temelín NPP with finalization

and starting of its full operation after

outages in 2015.

3.2 Recriticality of degraded

core due to reflooding

with demi water

The UJV is recently performing

analytical investigation of this topic

independently for each of Czech

NPPs, because of their principal

design differences. The methodological

approach is identical and consists

of the integral analyses of the severe

accident progression using the

MELCOR 2.2 code with an externaly

defined boric acid to analyze the

development of boric acid concentration.

In parallel the most important

configurations of the degraded core

are defined to be analysed using the

MCNP code to identify the minimum

concentration of boric acid leading to

the recriticality. The evolution of boric

acid concentrations vs. the minimum

value will allow to define conditions

for the applicability or the restriction

of application of demi water for

injecting into the degraded core under

various stages of severe accident


3.3 Control room habitability

This study was performed again

independently for each of Czech NPPs

and the UJV performed these analyses.

The methodology consisted of two

main analytical steps – the first one

covered the integral analyses with the

MELCOR 1.8.6 code for the identification

of fission product distribution

and releases via different leak paths.

The second step included the analysis

of dose rates in the control room based

on the precalculated distribution

of fission products and shielding of

control room walls or window (if


3.4 Corium localization

for Temelín NPP

The issue of the corium localization

is very complex and determines

the solution of others activities, like

a solution of long-term containment

pressure control. Before the

Fukushima Dai-ichi event the analytical

activities were focused only on

the ex-vessel corium cooling (ExVC)

strategy, because of a restriction on an

implementation of any new equipment

for severe accidents. The request

from the NAcP opened a way for

the solution of the in-vessel corium

retention (IVR) strategy as an alternative

one, which applicability has to

be evaluated. The utility opened at the

beginning two preparatory projects,

which enabled to prepare the fisrt

analytical models for the corium

behaviour in the lower head and cooling

conditions of rector pressure

vessel from outside. As the outcomes

from the analytical work identified

some potentials, the complex project

on the IVR solution was initiated in

2015 with expected duration up to

5 years. The project is not focused only

on analytical work, but also on the

experimental confirmation of the

VVER-1000 RPV coolability under the

IVR strategy.

The project is focused on three

types of activities – analytical investigation

with proposing of additional

systems for strategy solutions, designing

of new systems required for the

implementation of strategies, and the

experimental program for the RPV

Operation and New Build

Continuous Process of Safety Enhancement in Operation of Czech VVER Units ı J. Duspiva, E. Hofmann, J. Holy, P. Kral and M. Patrik

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