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Review and outlook of the integral test facility PKL III corresponding studies

Rückblick und Ausblick zur integralen Versuchsanlage PKL III und ihrer Studien
H. Xu
From the journal Kerntechnik

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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Received: 2021-07-23
Published Online: 2021-12-17

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