A severe accident (SA) is defined as an incident involving melting of the nuclear reactor core and the release of fission products (FP) from the fuel and their associated risks. In the SA, the containment may fail, causing the public hazard of fission products released to the environment. This review elaborates the resolved issues of SAs under the condition of a hypothetical SA. SA research that has been performed over the years is briefly described, including various SA scenarios. The SA scenarios involve core melt scenarios from the beginning of core degradation to melt formation and relocation into the lower head and to the containment, the interactions of the molten corium with water and concrete, the behavior of fission products in- and ex-vessel, hydrogen-related phenomena, and all associated risks. The mitigation strategies that have been adopted in existing reactors and advanced light water reactors (ALWR) are also discussed. These mitigation measures can keep the reactor vessel or containment intact and terminate the SA progression. SA analysis codes are then summarized and divided into three categories, namely, systematic analysis codes, mechanism analysis codes, and single-function analysis codes. Next, the unresolved issues of SAs are proposed, including narrow gap cooling, melt chemical interactions, steam explosion loads, molten debris coolability, and iodine chemistry. Further experimental and theoretical research activities should be conducted to resolve these issues; consequently, some recommendations for further research work are also given in the last part of this review. This review aims to add to the knowledge and understanding of SA research in the past few decades and to benefit further research of SAs.
Skip Nav Destination
Article navigation
October 2015
Research Papers
A Review on Analysis of LWR Severe Accident
Y. P. Zhang,
Y. P. Zhang
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
Search for other works by this author on:
S. P. Niu,
S. P. Niu
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
Search for other works by this author on:
L. T. Zhang,
L. T. Zhang
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
Search for other works by this author on:
S. Z. Qiu,
S. Z. Qiu
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
Search for other works by this author on:
G. H. Su,
G. H. Su
1
Department of Nuclear Science and Technology,
e-mail: [email protected]
Xi’an Jiaotong University
, Xi’an 710049
, China
e-mail: [email protected]
1Corresponding author.
Search for other works by this author on:
W. X. Tian
W. X. Tian
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
Search for other works by this author on:
Y. P. Zhang
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
S. P. Niu
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
L. T. Zhang
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
S. Z. Qiu
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
G. H. Su
Department of Nuclear Science and Technology,
e-mail: [email protected]
Xi’an Jiaotong University
, Xi’an 710049
, China
e-mail: [email protected]
W. X. Tian
Department of Nuclear Science and Technology,
Xi’an Jiaotong University
, Xi’an 710049
, China
1Corresponding author.
Manuscript received October 11, 2014; final manuscript received April 8, 2015; published online September 3, 2015. Assoc. Editor: Tomio Okawa.
ASME J of Nuclear Rad Sci. Oct 2015, 1(4): 041018 (20 pages)
Published Online: September 3, 2015
Article history
Received:
October 11, 2014
Revision Received:
April 8, 2015
Accepted:
April 16, 2015
Online:
September 16, 2015
Citation
Zhang, Y. P., Niu, S. P., Zhang, L. T., Qiu, S. Z., Su, G. H., and Tian, W. X. (September 3, 2015). "A Review on Analysis of LWR Severe Accident." ASME. ASME J of Nuclear Rad Sci. October 2015; 1(4): 041018. https://doi.org/10.1115/1.4030364
Download citation file:
Get Email Alerts
Study of TRICO II Reactor Startup and Shutdown Operations Using the OpenMC Calculation Code
ASME J of Nuclear Rad Sci (July 2025)
Adjuster Absorber Rods Return to Service at PLNGS
ASME J of Nuclear Rad Sci (July 2025)
Calculation of Radiation Field and Shutdown Dose Rate for Fusion Reactor Based on cosRMC
ASME J of Nuclear Rad Sci
Related Articles
Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications
ASME J of Nuclear Rad Sci (April,2017)
Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework
ASME J of Nuclear Rad Sci (October,2020)
Pressure Load Estimation During Ex-Vessel Steam Explosion
J. Eng. Gas Turbines Power (May,2009)
PBMR-A Future Failsafe Gas Turbine Nuclear Power Plant?
Mechanical Engineering (August,2011)
Related Proceedings Papers
Related Chapters
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
New Generation Reactors
Energy and Power Generation Handbook: Established and Emerging Technologies