Abstract
This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.
Abstract
Dieser Beitrag gibt einen Überblick über die wichtige integrale Testanlage (ITF) PKL (Primärkreislauf), die auf der Grundlage eines 4-Kreislauf-Druckwasserreaktors (DWR) mit einer Leistung von 1 300 MWe konzipiert wurde, und konzentriert sich insbesondere auf zwei Aspekte: (1) die Tests in jeder Entwicklungsphase der Anlage bis 2020, die einen typischen Mikrokosmos der nuklearen Sicherheitsforschung darstellt; (2) die Simulation der Tests der PKL-Anlage mit Hilfe von thermohydraulischen Systemcodes (STH), insbesondere RELAP5, TRACE und ATHLET. Die Ergebnisse aus der Literatur zeigen, dass alle diese Codes die Unfallszenarien in der PKL-Anlage bis zu einem gewissen Grad reproduzieren und die komplexen Phänomene sowohl im Reaktordruckbehälter (RDB) als auch in den Schleifen gut simulieren können, mit Ausnahme einiger lokaler Phänomene (z.B. Spitzenhüllrohrtemperatur (PCT)). Darüber hinaus werden in diesem Papier einige Vorschläge für weitere PKL-Tests vorgestellt. Besonders hervorgehoben werden die Sensitivitätsstudien der Anfangsbedingungen (ICs) und der Randbedingungen (BCs), die Teststudien im Zusammenhang mit den Richtlinien zur umfassenden Schadensbegrenzung (EDMGs) und den FLEX-Strategien, den ATWS Transienten, dem detaillierten Kernschnittmodell, der Kombination mit anderen ITF- oder separaten Effekten-Testanlagen (SET) und den Tests an fortgeschrittenen Konzeptionsreaktoren.
Nomenclature
- ACC
Accumulator
- AM
Accident Management
- ATWS
Anticipated Transients Without Scram
- BC
Boundary Condition
- BDBA
beyond design basic accident
- CCFL
Counter Current Flow Limitation
- CFD
Computational Fluid Dynamics
- CL
Cold Leg
- DBA
Design Basic Accident
- EDMG
Extensive Damage Mitigation Guidelines
- ELAP
Extended loss of alternating current power
- GRS
Gesellschaft für Anlagen- und Reaktorsicherheit (in German)
- HL
Hot Leg
- HPSI
High-pressure safety injection
- IBLOCA
Intermediate break loss-of-coolant accident
- IC
Initial Condition
- ITF
Integral Test Facility
- KWU
Kraftwerk Union (in German)
- LBLOCA
Large Break loss-of-coolant accident
- LOCA
Loss of Coolant Accident
- LOFT
Loss-of-Fluid Test
- LPSI
Low-pressure safety injection
- LSTF
Large Scale Test Facility
- MSLB
Main Steam Line Break
- NC
Natural Circulation
- NCG
Non-Condensation Gas
- NEA
Nuclear Energy Agency
- NPP
Nuclear Power Plant
- OECD
Organization for Economic Co-operation and Development
- PCT
Peak Cladding Temperature
- PKL
Primary Loop (in German abbreviations)
- PSA
Probabilistic Safety Assessment
- PRZ
Pressurizer
- PWR
Pressurized Water Reactor
- RCS
Reactor Core Coolant System
- RHRS
Residual Heat Removal System
- ROCOM
Rossendorf Coolant Mixing Model
- RPV
Reactor Pressurized Vessel
- SBLOCA
Small Break loss-of-coolant accident
- SBO
Station Blackout
- SET
Separate Effects Test
- SG
Steam Generator
- STH
system thermal-hydraulic
- TH
thermal-hydraulic
- TMI
Three-mile Island
- UPTF
Upper Plenum Test Facility
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Artikel in diesem Heft
- Frontmatter
- ANSYS-CFX simulation of the SRBTL test loop core with nanofluid coolant
- Evaluation of human factor engineering influence in nuclear safety using probabilistic safety assessment techniques
- Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm
- Application research on neutron-gamma discrimination based on BC501A liquid scintillator
- Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3
- Study on specific heat capacity and thermal conductivity of uranium nitride
- Review and outlook of the integral test facility PKL III corresponding studies
- Effects of some level density models and γ-ray strength functions on production cross-section calculations of 16,18O and 24,26Mg radioisotopes
- An application research for near-surface repository of strontium-90 sorption kinetic model on mudrocks
- Calendar of events
Artikel in diesem Heft
- Frontmatter
- ANSYS-CFX simulation of the SRBTL test loop core with nanofluid coolant
- Evaluation of human factor engineering influence in nuclear safety using probabilistic safety assessment techniques
- Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm
- Application research on neutron-gamma discrimination based on BC501A liquid scintillator
- Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3
- Study on specific heat capacity and thermal conductivity of uranium nitride
- Review and outlook of the integral test facility PKL III corresponding studies
- Effects of some level density models and γ-ray strength functions on production cross-section calculations of 16,18O and 24,26Mg radioisotopes
- An application research for near-surface repository of strontium-90 sorption kinetic model on mudrocks
- Calendar of events