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COMMISSION THE EUROPEAN COMMUNITIES

Directorate-General for Research, Science and Education

Xll/0/3

AND TECHNOLOGY

European Community Water Reactor

Safety Research Projects

VOLUME

bECEMBER 1983

DOCUMENT EXCLUS/VEMENT INTERNE

r

COMMISSION THE EUROPEAN COMMUNITIES

Directorate-General for Research, Science and Education

Xll/0/3

NUCLEAR SCIENCE AND TECHNOLOGY

VOLUME

DECEMBER 1983

DOCUMENT EXCLUS/VEMENT INTERNE

INTRODUCTION

This is the eigth compilation of Community research formats to be produced by the Commission. The following guidelines were used in compilation of the index which is in two volumes of :

1) the first page relevant to each project is always on the right hand side

of the document when opened;

2) all pages have a number, even if blank;

3) within each class (chapter) the formats are assembled in the following

order of country :

Belgium

Fed. Rep. of

Germany

Denmark

France

Ireland

Italy

JRC Ispra

Luxembourg

Netherlands

United Kingdom

4) Updated formats will be inserted in the relevant replacement position.

When additional pages have to be inserted they will be numbered with the proceeding page number plus an oblique and an extra number (for example page 53/1 will be inserted following page 53).

5) Formats for new projects will normally be inserted following the last

format of the relevant country within that class (chapter).

6) If a project is entered under more than one class (chapter), the full

format is given only once in the most important position.

XII/1090/83-EN

- I -

VOLUMES I AND II -WATER REACTORS

TABLE OF

1. BLOWDOWN AND EMERGENCY

A model for analysis of a BWR

under LOCA conditions

Evaluation of PWR coolant accident :

coordinated general program of code development

Development of an advanced code

for the study of the loss of coolant accident in a pressurized water reaction (CATHARE)

Evaluation -verification of RELAP and

TRAC codes. SEMISCALE, PKL,

calculations

Evaluation verification EDITH project

(small break)

New separate effects test in support to

CATHARE code

Two phase plant simulator

BETHSY system loop

Core rescue system (SSN)

Project FIX. II, Blow-down heat transfer

during LOCA

1.1. Phenomena prior to ECCS initiation

(no entries)

Pre and Post test analysis of the BWR

Series experiments of research project

RS 168/150.396

Country Page No.

Denmark 1

France 5

France 15

France 23

France 35

France 37

France 41

France 43

Italy 45

Sweden 49

FRG 51

-II -

Vent-valve assembly, functional reliability

and pressure rel,ief performance during the first blowdown-phase

Experiments

on the relief in a model reactor pressure vessel with internals at BWR conditions

Design, precomputation and evaluation of the

HDR blow-down experiments on dynamical loa

dings and deformations of reactor-pressure vessel internals

Experimental data acquisition and processing

of the dynamic behavior of the pressure vessel test internals in the HDR blow-down experiments

Development and

verification of coupled fluid-structure dynamics codes for analysis of dynamic stresses and deformations of reactor vessel internals during LOCA Laboratory experiments for validation and enhancement of fluid-structure dynamics codes relevant to initial phase of LOCA

Mechanical response of the core barrel

clamping and the control rod guide tubes during blow-down

Development

of condensation-and slip models of two-phase-flow computer codes

Mechanical

effects of a LOCA on P.W.R. internals

Critical and unbounded flow characteris

tics

Blow-down experiments by PIPER apparatus

with internal structures

Analysis of thermal and hydraulic transients

following a LOCA in L.W. reactors and in associated containment systems r

Country Page No.

FRG 55

FRG 59

FRG 63

FRG 67

FRG 69

FRG 73

FRG 75

FRG 77

France 79

Italy 81

Italy 83

Italy 85

-Ill -

Development of two phase flow and safety

instrumentation

NKA/SAK-3 reactor safety -small break

LOCA analysis

NKA/SAK-5 reactor safety heat transfer

correlations

Development and operation of rod-shaped

ultrasonic probes for liquid-level detec tion within single-and two-phase media under blow-down conditions

Transition boiling heat transfer in forced

convection

Development of a two-phase pump model for

PWR main coolant pumps

Velocity ·-measurements of two-phase flows (water/steam) by correlation of temperature tracer

Loop blow-down investigations (LOBI) pro

ject : influence of PWR primary loops on blow-down

Treatment of small

leaks in PWR's with methods of transient analysis

Analysis of integral and separate effects

tests (LOFT, SEMISCALE, IV)

Joint test rig for tests and calibration

of different methods of two-phase mass flow measurements

Experimental

facility for non-steady state two-phase flow

Development

of radionuclide methods for measuring two-phase mass flows

LOCA thermohydraulics experimental study of

critical flow and two-phase flow : MOBY and SUPER projet

Country Page No.

Belgium 87

Denmark 89

Denmark 91

FRG 95

FRG 99

FRG 101

FRG 103

FRG 107

FRG 113

FRG 117

FRG 121

FRG 125

FRG 127

France 131

-IV -

LOCA thermohydraulics experimental study

of blow-down in a loss of coolant accident (LOCA) of a P.W.R. : OMEGA project

LOCA thermohydraulics reflood studies

PERICLES project

LOCA thermohydraulics : steam generator

behavior in accident conditions : PATRICIA

I and II project

Evaluation -verification : LOBI project

Modeling of the thermal-hydraulics in a degraded

core transients of BWR's for system control faults

Analysis of thermal and hydraulic tran

sients following a LOCA in L.W. reactors and in associated containment systems

Blow-down experiments by

PIPER apparatus

with internal structures

LOCA analysis in LWR : blow-down Jet

Forces

Analysis of thermohydraulic transients in

LWR following LOCA

on two-phase critical flow in con nections with LOCA in light water reactors

Finalized research activity in the field of

thermohydraulic transients following LOCAS for the ENEL unified project PWR reactor

Thermohydraulic behavior of a pressurizer

during accident conditions

Thermohydraulics of steam generators under

accident conditions

Influence on DNB of subchannel obstructions

and rod bowings in BWR bundles with nonuniform heat flux profiles

Country Page No.

France 137

France

141

France 145

France 149

France 159

Italy 163

Italy 165

Italy 167

Italy 169

Italy 173

Italy 177

Italy 179

Italy 181

Italy 183

Italy 185

- V -

Thermohydraulic behavior of a pressurizer

during accident conditions

Steady-state post-dryout experiments in

4x4 BWR geometry

Post-dryout experiments in tubular geometry

Blow-down experiments by PIPER apparatus

with internal structures

Analysis of thermal and hydraulic transients

following a LOCA in L.W. reactors and in associated containment systems -II as international standard problem 15

Thermal-hydraulic accident analysis

LOCA uncertainty margins (GSD4.2)

Water cooled reactor depressurization

studies : CSNI standard problem calculation for ECCS

Post dryout heat transfer at low quality

and low pressure

Theoretical modelling of two-phase flow in

complex geometries

Blow-down of a steam drum

Thermohydraulic safety studies multipin

cluster rig

Effects of fuel pin ballooning on reflood

heat transfer

LOCA uncertainty margins (GSD4.2>

Analysis of thermal and hydraulic transients

following a LOCA in L.W. reactors and in asso-

Country

Italy Italy Italy Italy Italy

Sweden

U.K. U.K. U.K. U.K. U.K. U.K. U.K. ciated containment systems Italy

Page No.

187
189
191
'JI93 195
197
199
201
203
205
207
209
211
213
215
-VI -

Analysis of thermohydraulic transients in

LWR following LOCA

uncertainty margins (GSD4.2)

1.2. Performance of ECCS

Fluid dynamic effects in the upper end box

region of the fuel bundle during refill and reflooding

Reflood tests with regard to the primary loops

(PKL)

Test phases IB-IE and II

and start-up of the UPTF

Supplementary supplies and services for

the construction and start-up of the UPTF within the 2D/3D projectquotesdbs_dbs26.pdfusesText_32

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