Abstract

Quantification of common cause failure (CCF) parameters and their application in multi-unit PSA are important to the safety and operation of nuclear power plants (NPPs) on the same site. CCF quantification mainly involves the estimation of potential failure of redundant components of systems in a NPP. The components considered in quantification of CCF parameters include motor operated valves, pumps, safety relief valves, air-operated valves, solenoid-operated valves, check valves, diesel generators, batteries, inverters, battery chargers, and circuit breakers. This work presents the results of the CCF parameter quantification using check valves and pumps. The systems considered as case studies for the demonstration of the proposed methodology are auxiliary feedwater system (AFWS) and high-pressure safety injection (HPSI) systems of a pressurized water reactor (PWR). The posterior estimates of alpha factors assuming two different prior distributions (Uniform Dirichlet prior and Jeffreys prior) using the Bayesian method were investigated. This analysis is important due to the fact that prior distributions assumed for alpha factors may affect the shape of posterior distribution and the uncertainty of the mean posterior estimates. For the two different priors investigated in this study, the shape of the posterior distribution is not influenced by the type of prior selected for the analysis. The mean of the posterior distributions was also analyzed at 90% confidence level. These results show that the type of prior selected for Bayesian analysis could have effects on the uncertainty interval (or the confidence interval) of the mean of the posterior estimates. The longer the confidence interval, the better the type of prior selected at a particular confidence level for Bayesian analysis. These results also show that Jeffreys prior is preferred over Uniform Dirichlet prior for Bayesian analysis because it yields longer confidence intervals (or shorter uncertainty interval) at 90% confidence level discussed in this work.

References

1.
Cho
,
J.
,
Han
,
S. H.
,
Kim
,
D.-S.
, and
Lim
,
H.-G.
,
2018
, “
Multi-Unit Level 2 Probabilistic Safety Assessment: Approaches and Their Application to a Six-Unit Nuclear Power Plant site
,”
Nucl. Eng. Technol.
,
50
(
8
), pp.
1234
1245
.10.1016/j.net.2018.04.005
2.
Saputro
,
D. R. S.
,
Widyaningsih
,
P.
,
Handayani
,
F.
, and
Kurdhi
,
N. A.
,
2017
, “
Prior and Posterior Dirichlet Distributions on Bayesian Networks (BNs)
,”
AIP Conf. Proc.
,
1827
, p.
020036
.10.1063/1.4979452
3.
Le Duy
,
T. D.
, and
Vasseur
,
D.
,
2018
, “
A Practical Methodology for Modeling and Estimation of Common Cause Failure Parameters in Multi-Unit Nuclear PSA Model
,”
Reliab. Eng. Syst. Saf.
,
170
, pp.
159
174
.10.1016/j.ress.2017.10.018
4.
Hassija
,
V.
,
Kumar
,
S. C.
, and
Velusamy
,
K.
,
2014
, “
A Pragmatic Approach to Estimate Alpha Factors for Common Cause Failure Analysis
,”
Ann. Nucl. Energy
,
63
, pp.
317
325
.10.1016/j.anucene.2013.07.053
5.
Kang
,
D. D.
,
Hwang
,
M. J.
, and
Han
,
S. H.
,
2011
, “
Estimation of Common Cause Failure Parameters for Essential Service Water System Pump Using the CAFE-PSA
,”
Prog. Nucl. Energy
,
53
(
1
), pp.
24
31
.10.1016/j.pnucene.2010.09.009
6.
Modarres
,
M.
,
Zho
,
T.
, and
Massoud
,
M.
,
2017
, “
Advances in Multi-Unit Nuclear Power Plant Probabilistic Risk Assessment
,”
Reliab. Eng. Syst. Saf.
,
157
, pp.
87
100
.10.1016/j.ress.2016.08.005
7.
Atwood
,
C. L.
,
2013
, “
Consequences of Mapping Data or Parameters in Bayesian Common-Cause Analysis
,”
Reliab. Eng. Syst. Saf.
,
118
, pp.
118
131
.10.1016/j.ress.2013.04.015
8.
Wierman
,
T. E.
,
Rasmuson
,
D. M.
, and
Mosleh
,
A.
,
2007
, “
Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding
,”
U.S. Nuclear Regulatory Commission
,
Rockville, MD
,
Standard Nos. NUREG/CR-6268, INL/EXT-07-12969
.https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6268/
9.
SAS Institute Inc.
,
2009
, “
SAS/STAT® 9.2 User's Guide
,” 2nd ed.,
SAS Institute
,
Cary, NC
.
10.
Mosleh
,
A.
,
Rasmuson
,
D. M.
, and
Marshall
,
F. M.
,
1998
, “
Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment
,”
U.S. Nuclear Regulatory Commission
,
Rockville, MD
,
Report No. NUREG/CR-5485
.https://nrcoe.inl.gov/resultsdb/publicdocs/CCF/NUREGCR-5485_Guidelines%20on%20Modeling%20Common-Cause%20Failures%20in%20PRA.pdf
11.
USNRC,
1975
, “
Reactor Safety Study: An Assessment of Accident Risk in U.S. commercial Nuclear Power Plants
,”
U.S. Nuclear Regulatory Commission
,
Rockville, MD
,
Standard No. NUREG-75/014 (WASH-1400)
, pp.
102
110
.https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr75-014/
12.
Iawa
,
D. D.
,
da Silva
,
Borges
,
D.
,
Guimarães
,
A. C. F. G.
, and
de Lourdes
,
Moreira
,
M.
,
2017
, “
Reliability Study of the Auxiliary Feed-Water System of a Pressurized Water Reactor by Faults Tree and Bayesian Network
,”
International Nuclear Atlantic Conference (INAC 2017)
,
Belo Horizonte, MG, Brazil
, Associação Brasileira de Energia Nuclear (ABEN).
13.
U.S. Nuclear Regulatory Commission
,
2016
, “
CCF Parameter Estimates
,”
U.S. Nuclear Regulatory Commission
, accessed Jan. 6, 2020, https://nrcoe.inl.gov/resultsdb/ParamEstSpar/
14.
Poloski
,
J. P.
,
Knudsen
,
J. K.
,
Atwood
,
C. L.
, and
Galyean
,
W. J.
,
2000
, “
Reliability Study: High-Pressure Safety Injection System, 1987–1997 (DRAFT)
,”
U.S. Nuclear Regulatory Commission
,
Rockville, MD
,
Standard Nos. NUREG/CR-5500, INEEL/EXT-99-00373
, pp.
1987
1997
.https://nrcoe.inl.gov/resultsdb/publicdocs/SystemStudies/nureg-cr-5500-vol-9.pdf
You do not currently have access to this content.