Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident
-
L. Wu
,
P. Yu
Abstract
Preventing the leakage of radioactive materials is important to nuclear safety. During a station blackout accident in pressurized water reactors, the hot leg creep rupture caused by hot leg countercurrent flow occurs before the reactor pressure vessel failure that caused by lower head rupture. The secondary fission products barrier is lost after hot leg creep rupture. An analysis for this phenomenon was done using the Modular Accident Analysis Program version 4.0.4 code. A station blackout accident for CPR1000 is simulated and the occurrence and influence of hot leg creep rupture phenomenon are analyzed in detail. After that, a sensitivity analysis of the opening of different pressurizer pilot-operated relief valves at five minutes after entering severe accident management guideline (before the hot leg creep rupture occurs) is studied. The results show that reactor pressure vessel failure time can be extended by at least 4 h if at least one pilot-operated relief valve is opened and direct containment heating phenomenon can be eliminated if at least two pilot-operated relief valves are opened.
Abstract
Die Verhinderung des Austretens von radioaktiven Stoffen ist eines der drei Sicherheitsschutzziele beim Betrieb von kerntechnischen Anlagen. Bei einem Station-Blackout-Szenario in Druckwasserreaktoren kommt es vor dem Beginn des Versagens des Reaktordruckbehälters durch den Gegenstrom im heißen Strang zum sogenannten Heißstrang-Kriechbruch. Als Folge wird die zweite Barriere zur Rückhaltung der Spaltprodukte zerstört und die Zeit bis zum Versagen des Reaktordruckbehälters verkürzt. In diesem Beitrag wird eine Analyse zu diesem Phänomen mit dem Code Modular Accident Analysis Program Version 4.0.4 (MAAP4) vorgestellt. Dazu wird ein Stations-Black-Out-Szenario für den CPR1000 modelliert und das Auftreten und der Einfluss des Heißstrang-Kriechbruches im Detail analysiert.
Danach wird eine Sensitivitätsanalyse der Öffnung verschiedener Druckbegrenzungsventile nach fünf Minuten nach Beginn des Betriebs des CPR1000 entsprechend der Richtlinie für das Management von schweren Unfällen (bevor der Kriechbruch des heißen Strangs auftritt) untersucht, entsprechend der Beschreibung der Arbeitsschritte zur Druckentlastung des Reaktorkühlsystems wie sie in der Richtlinie für das Management von schweren Unfällen der Westinghouse Owners Group enthalten sind. Die Ergebnisse zeigen, dass der Heißstrang-Kriechbruch vor dem Beginn des Reaktordruckbehälterversagens auftritt und dass der Beginn dieses Versagens um mindestens 4 Stunden hinausgezögert werden kann, wenn ein vorgesteuertes Überdruckventil geöffnet wird. Das Öffnen von zwei Überdruckventilen kann das Phänomen der direkten Containmenterwärmung verhindern und den Verlust der Integrität der dritten Barriere reduzieren.
Nomenclature
- AC
power Alternating Current power
- APR1400
Advanced Power Reactor 1400
- B-loop
Broken loop
- DCH
Direct Containment Heating
- HLCR
Hot Leg Creep Rupture
- MAAP
Modular Accident Analysis Program
- PORV
Pilot-Operated Relief Valve
- PSRV
Pressurizer Safety Relief Valve
- PWR
Pressurized Water Reactor
- PZR
Pressurizer
- RCS
Reactor Coolant System
- RCP
Reactor Coolant Pump
- RPV
Reactor Pressure Vessel
- SAMG
Severe Accident Management Guideline
- SBO
Station Blackout
- SG
Steam Generator
- SRV
Safety Relief Valve
- TAF
Top of the Active Fuel
- TWCR
Temperature of Water in Core
- U-loop
Unbroken loop
Acknowledgements
The project was supported by the Natural Science Foundation of Fujian Province of China (No. 2020J01038) and Fujian innovation strategy research program (No. 2020R0011).
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© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
Articles in the same Issue
- Frontmatter
- Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident
- Effect of gap design pressure on the LWR fuel rods lifetime
- Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle
- Methodology for analyzing accidents with radioactive material release with code EPZDose
- Development and usage of the digital SAMG system
- Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC
- Corrosion surveillance program for tank, fuel cladding and supporting structure of 30 MW Indonesian RSG GAS research reactor
- Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600
Articles in the same Issue
- Frontmatter
- Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident
- Effect of gap design pressure on the LWR fuel rods lifetime
- Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle
- Methodology for analyzing accidents with radioactive material release with code EPZDose
- Development and usage of the digital SAMG system
- Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC
- Corrosion surveillance program for tank, fuel cladding and supporting structure of 30 MW Indonesian RSG GAS research reactor
- Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600
