Abstract
The economic simplified boiling water reactor (ESBWR) is a boiling water nuclear reactor of generation III+. The U.S. Nuclear Regulatory Commission (NRC) approved the ESBWR design as the world's best light-water nuclear reactor in 2014. It has the lowest core damage frequency (industry standard indicator of safety) of any Generation III or III+ reactor. It can cool automatically for more than seven days without using electricity or human intervention. During the operation, the ESBWR is designed to produce electricity while emitting almost no greenhouse gases. The energy generated by an ESBWR will prevent the emission of approximately 7.5 million metric tons of CO2 per year compared to standard electricity production on the U.S. grid. The analysis present in this paper aimed to characterize the thermal-hydraulic simulations of full-scale ESBWR design. The analysis presented will help in recognizing the improvement needed in the reactor design and its passive safety systems. The analysis is performed for normal steady-state and postulated design basis accident scenarios (break in one of the reactor main-steam lines inside the contentment). The simulation results obtained by the code REALP/SCDAPSIM/MOD3.4 are compared with the TRACG04 and MELCOR2.1 code results to determine the code predictability and accuracy under accident conditions of the newly proposed design of the ESBWR nuclear reactor. A parametric study is also carried out with different in-containment break sizes to determine code results' sensitive nature due to natural circulation flow phenomena, which are highly unstable. The analysis shows that the code has produced reasonably accurate results under the tested system thermal-hydraulic conditions compared with the TRACG04 and MELCOR2.1 code simulation results. This agreement between the three codes simulation results demonstrated that the publicly available RELAP5 code model has adequate features to simulate the ESBWR thermal-hydraulics. It has been also demonstrated that for the postulated accident conditions, the design of passive safety systems is capable to capture the accident progression without any active power.
Issue Section:
Thermal Hydraulics and CFD
Keywords:
thermal-hydraulics,
generation III+ nuclear reactors,
ESBWR,
nuclear reactor design evaluation,
safety analysis,
RELAP/SCDAPSIM/MOD3.4,
natural circulation,
passive safety system characterization,
advance reactors
Topics:
Accidents,
Boiling water reactors,
Containment,
Design,
Pressure,
Safety,
Steam,
Water,
Heat,
Economics ,
Simulation,
Flow (Dynamics),
Simulation results,
Temperature
References
1.
International Atomic Energy Agency,
2005
, Chernobyl's Legacy: Health, Environmental and Socio-Economic Impacts and Recommendations to the Governments of Belarus, the Russian Federation and Ukraine the Chernobyl Forum (INIS-XA–798),
D.
Kinly
III
, ed.,
International Atomic Energy Agency (IAEA)
,
Vienna (Austria)
, p. 55
, accessed Jan. 20, 2019, https://inis.iaea.org/search/search.aspx?orig_q=RN:360932632.
Ball
,
R.
, and
Simpson
, 1994
, “
Research Report on Centre for Environmental & Risk Management
,”
Canadian Nuclear Association
,
Alberta, Canada
, p. 25
, accessed Jan. 20, 2019, https://www.iaea.org/sites/default/files/cns_8th_national_report_-_final_canada.pdf3.
International Atomic Energy Agency
, 2016
, “
Report: Sustainable Development and Nuclear Power
,” IAEA, Vienna, p. 116
, accessed Jan. 20, 2019, https://www.iaea.org/publications/11084/nuclear-power-and-sustainable-development4.
Hirschberg
,
S.
,
Burgherr
,
P.
,
Spiekerman
,
G.
, and
Dones
,
R.
, 2004
, “
Severe Accidents in the Energy Sector: Comparative Perspective
,”
111
(1–3
), pp. 57
–65
.5.
Levenson
,
M.
, and
Rahn
,
F.
, 1981
, “
Realistic Estimates of the Consequences of Nuclear Accidents
,” Nucl. Technol.
,
53
(2
), pp. 99
–110
.10.13182/NT81-A326146.
OECD
, 2010
, “
Comparing Nuclear Accident Risks With Those From Other Energy Sources
,” Nuclear Development, Nuclear Energy Agency, Paris, France, p. 47
pages, accessed Jan. 20, 2019, https://oecd-nea.org/jcms/pl_14538/comparing-nuclear-accident-risks-with-those-from-other-energy-sources7.
Allison
,
C. M.
, and
Hohorst
,
J. K.
, 2010
, “
Role of RELAP/SCDAPSIM in Nuclear Safety
,” Sci. Technol. Nucl. Installations
,
2010
, pp. 1
–46
.10.1155/2010/4256588.
Woo
,
K. S.
, and
Yoo
,
Y. J.
, 2008
, “
A Data Report on PUMA-E Large Break LOCA Main Steam Line Break without One ICS Test
,” Purdue University, School of Nuclear Engineering, West Lafayette, IN, Report No. PU/NE-07-21, p. 97
.9.
Ishii
,
M.
,
Hibiki
,
T.
,
Lim
,
J.
,
Lee
,
D. Y.
,
Rassame
,
S.
,
Cheng
,
L.
,
Choi
,
S. W.
,
Yang
,
J.
, and
Woo
,
K. S.
, April 2008
, “
A Data Report on PUMA-e Small Break LOCA Bottom Drain Line Break Test with One ICS Unit Disabled
,” Purdue University, School of Nuclear Engineering, West Lafayette, IN, Report No. PU/NE-08-06, p. 101
.10.
Ishii
,
M.
,
Hibiki
,
T.
,
Lim
,
J.
,
Lee
,
D. Y.
,
Rassame
,
S.
,
Choi
,
S. W.
, and
Yang
,
J.
, 2008
, “
A Data Report on PUMA-E Large Break LOCA Feed Water Line Break Test with One ICS Unit Disabled
,” Purdue University, School of Nuclear Engineering, West Lafayette, IN, Report No. PU/NE-08-28, p. 97
.11.
Ishii
,
M.
,
Hibiki
,
T.
,
Lim
,
J.
,
Lee
,
D. Y.
,
Rassame
,
S.
,
Choi
,
S. W.
, and
Yang
,
J.
, September 2008
, “
A Data Report on PUMA-E Small Break LOCA GDCS Drain Line Break Test with One ICS Unit Disabled (GDLB-4500-BDBA-3 and -7)
,” Purdue University, School of Nuclear Engineering, West Lafayette, IN, Report No. PU/NE-08-24, p. 67
.12.
Rassame
,
S.
,
Hibiki
,
T.
, and
Ishii
,
M.
, 2017
, “
ESBWR Passive Safety System Performance Under Loss of Coolant Accidents
,” Prog. Nucl. Energy
,
96
, pp. 1
–17
.10.1016/j.pnucene.2016.12.00513.
Gamble
,
R. E.
, 1997
, “
Pressure Scaling Analysis of the SBWR and Integral Test Facilities Post Test Review
,” Proceedings of ICONE5 Fifth International Conference on Nuclear Engineering
, May 26–30, Nice, France, ICONE5-2073, p. 17
.14.
Boyack
,
B. E.
, 1989
, “
Quantifying Reactor Safety Margins: Application of Code Scaling, Applicability, and Uncertainty (CSAU) Evaluation Methodology to a Large Break, Loss-of-Coolant Accident
,” Nuclear Regulatory Commission, Washington, DC, Report No. NUREG/CR-5249, p. 45
.15.
Ferng
,
Y. M.
,
Pei
,
B. S.
, and
Ding
,
T. J.
, 1995
, “
Investigation of Scale-Up Capability of RELAP5/MOD3 and Scaling Distortion in the IIST Facility Via Natural Circulation Experiments
,” Nucl. Technol.
,
109
(3
), pp. 398
–348
.10.13182/NT95-A3508816.
GE-Hitachi Nuclear Energy
, 2007
, “
ESBWR Design Control Document Tier 2, Chapter 1
,” GE-Hitachi Nuclear Energy, Wilmington, NC, Report No. 26A6642AH, p. 203
.17.
GE-Hitachi Nuclear Energy
, 2010
, “
ESBWR Design Control Document Tier 2, Chapter 6: Engineering Safety Features
,” GE-Hitachi Nuclear Energy, Wilmington, NC, Report No. 26A6642AT Rev. 09, p. 197
.18.
GE-Hitachi Nuclear Energy
, 2011
, “
The ESBWR General Plant Description
,” GE-Hitachi Nuclear Energy, Wilmington, NC, Report No. 26A6642AW Rev. 00, p. 201
.19.
GE-Hitachi
Nuclear Energy, 2005
, “
Technical Report on ESBWR Design Control Document Tier 2 Chapter 7: Instrumentation and Control Systems
,” GE-Hitachi Nuclear Energy
, Wilmington, NC, Report No. 26A6642AW, p. 267
.20.
Jack
,
T.
, 2010
, “
MELCOR DBA Containment Audit Calculations for the ESBWR Plant (Final)
,”
Sandia National Laboratories
, Albuquerque, NM, p. 86
.21.
Karve
,
A. A.
, and
Fawcett
,
R. M.
, 2011
, “
A Minimum Hot Excess Core Design Strategy for ESBWR
,” Prog. Nucl. Energy
,
53
(6
), pp. 571
–582
.10.1016/j.pnucene.2011.01.00822.
Cheung
,
Y. K.
,
Shiralkar
,
B. S.
, and
Marquino
,
W.
, 2005
, “
Performance Analyses of ESBWR ECCS and Containment Systems
,” Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP 2005)
, Seoul, South Korea, May 3, Paper No. 5485, p. 13
.23.
Lahey
,
R. T.
, Jr.
, and
Moody
,
F. J.
, 1979
, The Thermal‐Hydraulics of a Boiling Water Nuclear Reactor
,
American Nuclear Society
, LaGrange Park, IL, p. 12
.24.
Yadigaroglu
,
G.
, 1978
, “
Two-Phase Flow Instabilities and Propagation Phenomena, in Von Karman Inst. for Fluid Dyn, Two-Phase Flows in
,” Nucl. Reactors
,
2
(1
), pp. 353
–403
.https://link.springer.com/chapter/10.1007/978-94-009-2790-2_1125.
Fukuda
,
K.
, and
Kobori
,
T.
, 1979
, “
Classification of Two‐Phase Flow Instability by Density Wave Oscillation Model
,” J. Nucl. Sci. Technol.
,
16
(2
), pp. 95
–108
.10.1080/18811248.1979.973087826.
Nayak
,
A. K.
, and
Vijayan
,
P. K.
, 2008
, “
Flow Instabilities in Boiling Two-Phase Natural Circulation Systems: A Review
,” Sci. Technol. Nuclear Instal.
,
2008
(1
), pp. 1
–121
.10.1155/2008/57319227.
Rohatgi
,
U. S.
, and
Duffey
,
R. B.
, 1998
, “
Stability, DNB, and CHF in Natural Circulation Two-Phase Flow
,” Int. Commun. Heat Mass Transfer
,
25
(2
), pp. 161
–174
.10.1016/S0735-1933(98)00003-728.
Boure
,
J. A.
,
Bergles
,
A. E.
, and
Tong
,
L. S.
, 1973
, “
Review of Two-Phase Flow Instability
,” Nucl. Eng. Des.
,
25
(2
), pp. 165
–192
.10.1016/0029-5493(73)90043-529.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2020
, “
Non-Hyperbolicity of Conservation Equations of RELAP5 Two Fluid Model in Nuclear Reactor Safety Results—Part I: Investigation and Eigenvalue Analysis
,” ASME J. Nucl. Rad. Sci.
,
3
(1
), p. 014503
.10.1115/1.404716130.
Saraswat
,
S. P.
,
Munshi
,
P.
, and
Allison
,
C.
, 2020
, “
Characteristics and Linear Stability Analysis of RELAP5 Two-Fluid Model for Two-Component, Two-Phase Flow
,” Ann. Nucl. Energy
,
151
(1
), pp. 101
–121
.https://doi.org/10.1016/j.anucene.2020.10794831.
Saraswat
,
S. P.
,
Munshi
,
P.
, and
Allison
,
C.
, 2020
, “
Linear Stability Analysis of RELAP5 Two-Fluid Model in Nuclear Reactor Safety Results
,” Ann. Nuclear Energy
,
149
, pp. 107720
–117
.10.1016/j.anucene.2020.10772032.
Bhadouria
,
V. S.
,
Akhter
,
Z.
,
Akhtar
,
M. J.
, and
Munshi
,
P.
, 2021
, “
Automated Microwave Monitoring of Hidden Objects for Strategic and Security Applications
,” J. Electromagn. Waves Appl.
,
151
(1
), pp. 112
–119
. 10.1080/09205071.2021.195340433.
Mishra
,
G.
,
Mandariya
,
A. K.
,
Tripathi
,
S. N.
,
Joshi
,
M.
,
Khan
,
A.
, and
Sapra
,
B. K.
, 2019
, “
Hygroscopic Growth of CsI and CsOH Particles in Context of Nuclear Reactor Accident Research
,” J. Aerosol Sci.
,
132
, pp. 60
–69
.10.1016/j.jaerosci.2019.03.00834.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Alison
,
C.
, and
Khanna
,
A.
, 2017
, “
Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0
,” ASME J. Nucl. Rad. Sci.
,
3
(4
), p. 041009
.10.1115/1.403718835.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2016
, “
Thermal Hydraulic and Safety Assessment of First Wall Helium Cooling System of a Generalized Test Blanket System in ITER Using RELAP5 Code
,” ASME J. Nucl. Rad. Sci.
,
3
(1
), p. 014503
. 10.1115/1.403468036.
Saraswat
,
S. P.
,
Munsh
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2018
, “
Thermal Hydraulic and Safety Assessment of LLCB Test Blanket System in ITER Using Modified RELAP/SCDAPSIM/MOD4.0 Code
,” ASME J. Nucl. Eng. Radiat. Sci.
,
4
(2
), pp. 021001
–021010
.10.1115/1.403882337.
Saraswat
,
S. P.
,
Ray
,
D.
,
Munshi
,
P.
, and
Allison
,
C.
, 2019
, “
Analysis of Loss of Heat Sink for Iter Divertor Cooling System (new Tungsten divertor design) Using Modified RELAP/SCDAPSIM/MOD 4.0
,” ASME J. Nucl. Rad Sci.
,
5
(4
), p. 042202.38.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2020
, “
In-Vessel Loss of Coolant Accident Assessment of a Generalized LLCB TBS in ITER using RELAP/SCDAPSIM Code (No. 2757).
EasyChair, Mumbai, India, accessed Feb. 3, 2021, https://easychair.org/publications/preprint/d2VW39.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2020
, Thermal Hydraulic Analysis of in-Vessel Loss of Coolant Accident and Loss of Flow Accident of First Wall Helium Cooling System of Generalized LLCB TBS in ITER Using Modified RELAP/SCDAPSIM MOD4. 0 Code (No. 2657). EasyChair, Mumbai, India, accessed Feb. 3, 2021, https://easychair.org/publications/preprint/RZL640.
Ray
,
D.
,
Saraswat
,
S. P.
,
Kumar
,
M.
,
Singh
,
O. P.
, and
Munshi
,
P.
, 2021
, “
Build Up and Characterization of Ultraslow Nuclear Burn-Up Wave in Epithermal Neutron Multiplying Medium
,” ASME J. Nucl. Eng. Radiat. Sci.
,10.1115/1.404972741.
Ray
,
D.
,
Kumar
,
M.
,
Bhadouria
,
V. K.
,
Saraswat
,
S. P.
, and
Munshi
,
P.
, 2020
, “
A Study of Transverse Buckling Effect on the Characteristics of Burnup Wave in a Diffusive Media
,” Int. Youth Nucl. Congr.
,
1
(1
), pp. 8
–13
.42.
Bhadouria
,
V. S.
,
Ray
,
D.
,
Saraswat
,
S. P.
,
Akhter
,
M. J.
, and
Munshi
,
P.
, 2020
, “
Microwave Technology for Personal Monitoring as a Nuclear Security Application
,” Int. Youth Nucl. Congr., 1(1), pp. 14
–17
.43.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2018
, “
Thermal-Hydraulic Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/MOD 4.0
,” International Conference on Fluid Mechanics and Fluid Power (FMFP)
, Dec. 10–12,
IIT Bombay
, Mumbai, India, p. 4
.44.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2018
, “
Investigation of Non-Hyperbolicity of Conservation Equations of RELAP/SCADAPSIM in Nuclear Reactor Safety Results by Method of Characteristics
,” Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety
,
IIT Bombay
, Mumbai, India, Dec. 15–20, p. 4
.45.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2018
, “
Investigation of Non-Hyperbolicity of Conservation Equations of Thermal-Hydraulic Code RELAP5 for Two Component and Two Phase
,” International Conference on Fluid Mechanics and Fluid Power (FMFP)
,
IIT Bombay
, Mumbai, India, Dec. 10–12, p. 4
.46.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2020
, “
Thermal-Hydraulics of Loss of Heat Sink Accident of Indian Test Blanket System in ITER
,” International Youth Nuclear Congress (IYNC)
, Mar. 10–12, Sydney, pp.1–4. https://www.researchgate.net/publication/341480149_Thermalhydraulics_of_loss_of_heat_sink_accident_of_Indian_test_blanket_system_in_ITER47.
Saraswat
,
S. P.
,
Munshi
,
P.
,
Khanna
,
A.
, and
Allison
,
C.
, 2020
, “
Non-Hyperbolicity of RELAP5 Two Fluid Model Under Loss of Flow Accident for a Boiling Water Reactor
,” International Youth Nuclear Congress (IYNC)
, Mar. 10–12, Sydney, pp. 5–8.https://www.researchgate.net/publication/341480139_Nonhyperbolicity_of_RELAP5_two_fluid_model_under_loss_of_flow_accident_for_a_boiling_water_reactor48.
Boucher
,
T. J.
,
di Marzo
,
M.
, and
Shotkin
,
L. M.
, 1992
, “
Scaling Issues for a Thermal-Hydraulic Integral Test Facility
,” Proceedings of the U.S. Nuclear Regulatory Commission, Nineteenth Water Reactor Safety Information Meeting
, Bethesda, MD, Oct. 28–30, pp. 367
–387
.49.
Ishii
,
M.
,
Hibiki
,
T.
,
Cheng
,
L.
,
Choi
,
S. W.
,
Yang
,
J.
,
Lim
,
J.
,
Lee
,
D. Y.
,
Rassame
,
S.
,
Woo
,
K. S.
, and
Yoo
,
Y. J.
, 2007
, A Data Report on PUMA-E Large Break LOCA Main Steam Line Break Test With One ICS Unit Disabled
,
Purdue University
,
West Lafayette, IN
, p. 125
.50.
Ishii
,
M.
,
Hibiki
,
T.
,
Lim
,
J.
,
Lee
,
D. Y.
,
Rassame
,
S.
,
Choi
,
S. W.
,
Yang
,
J.
,
Cheng
,
L.
, and
Woo
,
K. S.
, 2008
, A Data Report on PUMA-E Small Break LOCA Bottom Drain Line Break Test With One ICS Unit Disabled
,
Purdue University
,
West Lafayette, IN
, Report No. PU/NE-08-06, p. 235
.51.
Vierow
,
K. M.
, and
Schrock
,
V. E.
, 1992
, “
Condensation in a Natural Circulation Loop With Noncondensable Gas Present, Part I Heat Transfer
,” Japan-US Seminar on Two-Phase Flow Dynamics
, Berkeley, CA, p. 58
.52.
Boyer
,
B. D.
,
Parlatan
,
Y.
,
Slovik
,
G. C.
, and
Rohatgi
,
U. S.
, 2006
, “
An Assessment of RELAP5 M0D3.1.1 Condensation Heat Transfer Modeling With GIRAFFE Heat Transfer Tests
,”
Brookhaven National Laboratory,
New York, p. 213
.53.
Yoo
,
J. M.
,
Kang
,
J. H.
,
Yun
,
B. J.
,
Hong
,
S. W.
, and
Jeong
,
J. J.
, 2018
, “
Improvement of the MELCOR Condensation Heat Transfer Model for the Thermal-Hydraulic Analysis of a PWR Containment
,” Prog. Nucl. Energy
,
104
, pp. 172
–182
.10.1016/j.pnucene.2017.09.012Copyright © 2022 by ASME
You do not currently have access to this content.
