Abstract

After the Fukushima DAIICHI accident, new safety requirements were imposed in order to reduce risk of severe accident. One of the principles that have been adopted is the introduction of emergency action levels resulting from the expected consequences. They cover a wide range of component and system malfunctions resulting in emergency, incident, and/or accident conditions. To evaluate those emergency action levels, thermal hydraulic (TH) analyses simulating these malfunction/incident/accident conditions are required. This paper describes the simulation of a real operational incident scenario using a standard thermal hydraulic model of the power plant in the TRACE code that was originally intended for simulation of design basis accidents such as large break coolant accident or loss of flow accident. Special attention was given to the methodology, addressing a long duration of an incident with corrective actions of the operators, and to computational issues leading to model modifications caused by a long duration of the incident along with the necessary conservatisms in the estimated results of the simulated radioactivity release.

References

References
1.
Nuclear Energy Institute (NEI)
,
2011
, “Methodology for Development of Emergency Action Levels,”
Nuclear Energy Institute
,
Washington, DC
, Report No. NEI 99-01, accessed Feb. 3, 2021, https://www.nrc.gov/docs/ML1132/ML113270335.pdf
2.
Nuclear Energy Institute (NEI)
,
2009
, “Methodology for Development of Emergency Action Levels Advanced Passive Light Water Reactors,”
Nuclear Energy Institute
,
Washington, DC
, Report No. NEI 07-01, accessed Feb, 3, 2021, https://www.nrc.gov/docs/ML0920/ML092030210.pdf
3.
SURO
,
2015
, “
Identifikace Vzniku Radiačních Mimořádných Událostí na Jaderných Elektrárnách a Systém Klasifikace Jejich závažnosti—BV III/1-VS
,”
SURO
,
Prague, Czech Republic
.
4.
IAEA
,
2013
, “INES: The International Nuclear and Radiological Event Scale User's Manual,”
IAEA
,
Vienna
, accessed, https://www.iaea.org/publications/10508/ines-the-international-nuclear-and-radiological-event-scale-users-manual
5.
CEZ
, “Temelin Interactive Tour,”
CEZ
, accessed Feb. 3, 2021, http://virtualniprohlidky.cez.cz/cez-temelin-aj/
6.
CEZ
,
2012
, “
Stress Tests of Nuclear Power Plants—Č EZ, a. s. Evaluation of Nuclear Safety and Safety Margins of Temelín NPP
,”
CEZ
,
Prague, Czech Republic
.
7.
Ruščák
,
M.
,
Mazzini
,
G.
,
Kynčl
,
M.
,
Savanyuk
,
S.
,
Hrehor
,
M.
,
Musa
,
A.
, and
Flores y Flores
,
A.
,
2019
, “
VVER 1000 Severe Accident Analyses Using MELCOR Code
,”
ASME J. Nucl. Eng. Radiat. Sci.
,
5
(
3
), p. 030913.10.1115/1.4043378
8.
Mazzini
,
G.
,
Kynčl
,
M.
, and
Ruščák
,
M.
,
2016
, “
Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor
,”
ASME J. Nucl. Eng. Radiat. Sci.
,
2
(
4
), p. 044505.10.1115/1.4034144
9.
U. S. Nuclear Regulatory Commission
,
2008
, “
TRACE V5.840 Theory Manual
,”
U. S. Nuclear Regulatory Commission
,
Washington, DC
, Vol.
1
, p.
696
.
10.
U. S. Nuclear Regulatory Commission
, 2008, “
TRACE V5.840 User's ManuaL Volume 1: Input Specification
,”
U. S. Nuclear Regulatory Commission
,
Washington, DC
.
11.
U. S. Nuclear Regulatory Commission
,
2002
, “
TRACE V5.840 User's Manual Volume 2: Modeling Guidelines
,”
U. S. Nuclear Regulatory Commission
,
Washington, DC
.
12.
SUJB
,
2015
, “
Accident Sequence
,”
SUJB
,
Prague
, Report No. 85/15/2.
13.
CEZ
,
2017
, “
Osnova PředprovoznÍ Bezpečnostní Zprávy (PpBZ 1,2 Revize 1)
,” V00, Díl 15—Bezpecnostní rozbory (Temelin Nuclear Power Plant Safety Analyse Report Revision 1)– zavedení paliva TVSA-T, Kap. 15,” CEZ.
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