Abstract
An integral analysis methodology for BWR ATWS has been developed. The method covers main scopes about ATWS events, including radiological consequence, primary system integrity, fuel integrity, containment integrity, and long-term shutdown and cooling capability. The primary techniques about this methodology were described herein. The methodology has been applied on Kuosheng nuclear power station to show the applicability. Under this framework, some suggestions were proposed for further development of this methodology. This methodology can give a way to evaluate safety of BWR plants confidently. Further, it can be a tool for developing emergency procedures about severe accidents, or exploring strategies and checking their effectiveness.
Abstract
Es wurde eine integrale Analysemethodik zur Berechnung eines ATWS Ereignisses in einem Siedewasserreaktor entwickelt. Die Methode deckt die Hauptbereiche von ATWS-Ereignissen ab, einschließlich der radiologischen Folgen, der primären Systemintegrität, der Brennstoffintegrität, der Containmentintegrität und der langfristigen Abschalt- und Kühlfähigkeit. Die wesentlichen Techniken dieser Methodik werden in diesem Beitrag zusammengefasst.
Die Anwendbarkeit dieser Methodik wird durch die Beschreibung der Berechnung für ein ATWS am Kernkraftwerk Kuosheng gezeigt. In diesem Rahmen werden einige Vorschläge für die weitere Entwicklung dieser Methodik gemacht. Diese Methodik kann als Werkzeug zur zuverlässigen Bewertung der Sicherheit von SWR-Anlagen dienen. Außerdem kann sie bei der Entwicklung von Notfallprozeduren für schwere Unfälle oder der Erforschung von Strategien und deren Überprüfung eingesetzt werden.
Nomenclature
- AOO
anticipated operation occurrence
- ARI
alternate rod injection
- ATWS
anticipated transient without scram
- BOC
beginning of cycle
- BWR
boiling water reactor
- CHF
critical heat flux
- CPR
critical power ratio
- DBA
design basis accident
- DR
decay ratio
- EAB
exclusion area boundary
- ECPR
experimental critical power ratio
- EOC
end of cycle
- GE
general electric company
- INER
institute of nuclear energy research
- LPPF
local power peaking factor
- LPZ
low population zone
- MOC
middle of cycle
- MSIVC
main steam isolation valve closure
- MUR
measurement uncertainty recapture
- NSSS
Nuclear Steam Supply System
- OCPR
operating critical power ratio
- PBTT
peach bottom turbine trip
- PCT
peak cladding temperature
probability density function
- PREGO
pressure regulator failure open
- RCPB
reactor coolant pressure boundary
- RCIC
rector core isolation cooling
- RHR
residual heat removal
- RPS
reactor protection system
- RPT
recirculating pump trip
- RPV
reactor pressure vessel
- RRCS
redundant reactivity control system
- SGTS
standby gas treatment system
- SLCS
standby liquid control system
- SLMCPR
safety limit minimum critical power ratio
- SPU
stretch power uprate
- TPC
Taiwan power company
- TT
turbine trip
- TTWB
turbine trip with bypass
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© 2021 Walter de Gruyter GmbH, Berlin/Boston
Articles in the same Issue
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations
Articles in the same Issue
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations