Nuclear Regulation
NEA/CSNI/R(2013)1
April 2013www.oecd-nea.org

Final Integration Reportof the OECD/NEA ROSA
Project
2005-2009Unclassified

NEA/CSNI/R(2013)1

Organisation de Coopération et de Développement Économiques
Organisation for Economic Co-operation and Development

30-Apr-2013___________________________________________________________________________________________English text only
NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSNEA/CSNI/R(2013)1
Unclassified
Final Integration Report of OECD/NEA ROSA Project
2005-2009

JT03338928
English text only

Complete document available on OLIS in its original format
This document and any map included herein are without prejudice to the status of or sovereignty over any territory, to the delimitation ofinternational frontiers and boundaries and to the name of any territory, city or area.

NEA/CSNI/R(2013)1

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 34 democracies work together to address the economic,
social and environmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help
governments respond to new developments and concerns, such as corporate governance, the information economy and thechallenges of an ageing population. The Organisation provides a setting where governments can compare policy
experiences, seek answers to common problems, identify good practice and work to co-ordinate domestic and international
policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, Chile, the Czech Republic, Denmark,Estonia, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Israel, Italy, Japan, Luxembourg, Mexico, theNetherlands, New Zealand, Norway, Poland, Portugal, the Republic of Korea, the Slovak Republic, Slovenia, Spain,
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This work is published on the responsibility of the OECD Secretary-General.
The opinions expressed and arguments employed herein do not necessarily reflect the official
views of the Organisation or of the governments of its member countries.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consistsof 31 countries: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France, Germany, Greece,
Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, the Netherlands, Norway, Poland, Portugal, the Republic ofKorea, the Russian Federation, the Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom
and the United States. The European Commission also takes part in the work of the Agency.
The mission of the NEA is:– to assist its member countries in maintaining and further developing, through international co-operation, thescientific, technological and legal bases required for a safe, environmentally friendly and economical use of
nuclear energy for peaceful purposes, as well as– to provide authoritative assessments and to forge common understandings on key issues, as input to government
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable
development.
Specific areas of competence of the NEA include the safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear
law and liability, and public information.
The NEA Data Bank provides nuclear data and computer program services for participating countries. In these andrelated tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna, with which it
has a Co-operation Agreement, as well as with other international organisations in the nuclear field.

This document and any map included herein are without prejudice to the status of or sovereignty over any territory, to the delimitation ofinternational frontiers and boundaries and to the name of any territory, city or area.
Corrigenda to OECD publications may be found online at: www.oecd.org/publishing/corrigenda.
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NEA/CSNI/R(2013)1

THE COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS“The Committee on the Safety of Nuclear Installations (CSNI) shall be responsible for theactivities of the Agency that support maintaining and advancing the scientific and technical knowledge
base of the safety of nuclear installations, with the aim of implementing the NEA Strategic Plan for 20112016 and the Joint CSNI/CNRA Strategic Plan and Mandates for 2011-2016 in its field of competence.
The Committee shall constitute a forum for the exchange of technical information and forcollaboration between organisations, which can contribute, from their respective backgrounds in research,development and engineering, to its activities. It shall have regard to the exchange of information
between member countries and safety R&D programmes of various sizes in order to keep all member
countries involved in and abreast of developments in technical safety matters.
The Committee shall review the state of knowledge on important topics of nuclear safety
science and techniques and of safety assessments, and ensure that operating experience is appropriately
accounted for in its activities. It shall initiate and conduct programmes identified by these reviews andassessments in order to overcome discrepancies, develop improvements and reach consensus on technicalissues of common interest. It shall promote the co-ordination of work in different member countries thatserve to maintain and enhance competence in nuclear safety matters, including the establishment of joint
undertakings, and shall assist in the feedback of the results to participating organisations. The Committee
shall ensure that valuable end-products of the technical reviews and analyses are produced and available
to members in a timely manner.
The Committee shall focus primarily on the safety aspects of existing power reactors, othernuclear installations and the construction of new power reactors; it shall also consider the safety
implications of scientific and technical developments of future reactor designs.
The Committee shall organise its own activities. Furthermore, it shall examine any other mattersreferred to it by the Steering Committee. It may sponsor specialist meetings and technical working groups
to further its objectives. In implementing its programme the Committee shall establish co-operative
mechanisms with the Committee on Nuclear Regulatory Activities in order to work with that Committee
on matters of common interest, avoiding unnecessary duplications.
The Committee shall also co-operate with the Committee on Radiation Protection and Public
Health, the Radioactive Waste Management Committee, the Committee for Technical and Economic
Studies on Nuclear Energy Development and the Fuel Cycle and the Nuclear Science Committee on
matters of common interest.”

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FOREWORD

OECD/NEA ROSA Project was performed to resolve issues in thermal-hydraulics analyses relevant tolight water reactor (LWR) safety by using the ROSA/LSTF of JAEA for 4 years from April 2005. In
particular, it focused on the validation of simulation models and methods for various complex phenomenathat may occur during design-basis event (DBE) and beyond-DBE and to increase the level of detail andaccuracy in the analyses of the key phenomena during transients and accidents of interest.
To meet these objectives, the ROSA Project performed 12 LSTF experiments for the following 6 types ofsubject as either separate-effect tests or integral-effect tests.
Test 1

Temperature stratification and coolant mixing during ECCS coolant injection

Test 2

Unstable and destructive phenomena such as water hammer

Test 3Natural circulation under high core power conditions

Test 4Natural circulation with superheated steam

Test 5

Primary cooling through SG secondary depressurization

Test 6

Subjects open to participants

These subjects were defined among the Project participants from 14 countries who share the need tomaintain or improve their technical competence in thermal-hydraulics for nuclear reactor safety
evaluations. System modifications, addition of various types of new instrumentation and/or preparation of
new test sections were performed for the ROSA/LSTF to obtain detailed and specific data necessary toachieve the objectives.
The LSTF experiments generated the data that fulfill current and future needs for the development,
verification and validation of best estimate (BE) and computational fluid dynamics (CFD) codes. The
obtained data as well as technical information and experiences through the post-test analysis by using the
BE and CFD codes have been shared among the participants through the discussions mostly during
program review group (PRG) meetings.
This report briefly summarizes major findings in all the 6 subjects of the ROSA Project with a descriptionof background and objectives for each subject. Appendices describe new measurement instruments andnew test sections furnished for the ROSA Project, and summarize the post-test analyses including answers
to a questionnaire on the analysis effort.

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EXECUTIVE SUMMARY

Background of ROSA ProjectComputer codes are utilized in safety evaluation of light water reactors (LWRs) in order to simulate plant
behavior during design basis event (DBE) and beyond-DBE (BDBE). Complex multi-dimensional singlephase and two-phase flow conditions occur during these events and non-condensable gas is present inmany cases. For this purpose, thermal-hydraulic safety analysis codes, especially best estimate (BE) lossof-coolant accident (LOCA) analyses codes, have been developed with significant efforts to achieve highpredictive capability especially for one-dimensional phenomena such as flows in piping at rather high flow
rate. Such BE codes include APROS, ATHLET, CATHARE, MARS, RELAP5, TRAC and TRACE.
However, there remain needs for experimental work and code development and validation for the complex
flow conditions during DBE and BDBE. Furthermore, the increased use of the BE analysis codes andmethodologies in the safety evaluation of reactors, replacing traditional conservative approaches, requires
the quantification of uncertainties in the simulation models and methods.
Many experimental facilities have contributed to the thermal-hydraulic databases available today including
those for the CSNI code validation matrix (CCVM). However, a large portion of current data is insufficientfor future codes that are to incorporate multi-dimensional simulation capabilities, mainly because thespatial resolution of measurement and the measurement parameters are not enough to assess the simulation
models and methods.
The CSNI set up an international working group SESAR/FAP (Senior Group of Experts on Nuclear Safety
Research Facilities and Programmes) in 1997 to identify how existing expertise in the field of reactor
safety and an appropriate experimental infrastructure could be sustained in the future. One of the maintasks of the SESAR/FAP was to identify test facilities and research programs threatened by closure shortly
and to select facilities and programmes that would be of particular benefit to the member countries, if they
were able to continue in a form of OECD project. Several international co-operative research projects had
been started in the field of thermal hydraulics safety research to implement the recommendation by the
SESAR/FAP: these include the SETH (SESAR Thermal-Hydraulics) Project using two test facilities;
PANDA of PSI, Switzerland and PKL of AREVA, Germany, and PKL Project. Later, the CSNI
established the SESAR/SFEAR (Support Facilities for Existing and Advanced Reactors) in 2005 as afollow-up of the SESAR/FAP to assess the need for and strategy of maintaining key research facilities. In
both working groups, the LSTF (Large Scale Test Facility) in JAEA (former JAERI) was discussed andevaluated as an excellent facility for the thermal-hydraulic safety research. In the early 2000s, JAEA
decided to propose an OECD/NEA project to make the LSTF utilized more effectively and efficiently
among the member countries.
The LSTF was constructed in 1985 for the fourth-phase of ROSA (Rig-of-Safety Assessment) Program in
JAEA following the TMI-2 accident to experimentally investigate thermal-hydraulic phenomena that may
appear during postulated abnormal transients and small-break LOCAs (SBLOCAs). The LSTF is a fullpressure full-height two-loop model of a 4-loop Westinghouse-type PWR with a volumetric scaling of
1/48. More than 180 experiments were performed prior to the OECD/NEA ROSA Project, indicating its
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excellent capabilities to simulate the thermal-hydraulic phenomena under conditions close to those in thereference reactor. One of the important benefits of utilizing the LSTF for the OECD/NEA Project would be
the combination of the facility with the latest instrumentation to measure various types of measurement
parameters with more detailed spatial resolution than ever. In the late 1980s and 1990s, the measurement
instrumentations for the LSTF experiments were designed to be suitable for the verification of onedimensional BE codes. The observed phenomena, however, were often characterized by multi-dimensional
effects, including parallel channels, unstable, multi-phase flows under influences of non-condensable gas.
The measured data from these early experiments sometimes lacked the spatial resolution and parameters towell describe the multidimensional phenomena.
This ROSA Project tries to generate and share the necessary detailed data among the OECD member
countries to fulfill the current and future needs for the development, verification and validation of both the
BE and computational fluid dynamics (CFD) codes, by taking a view of the combined utilization of BE
code with uncertainty evaluation methods such as BEPU (best estimate plus uncertainty).

Objectives of ROSA Project
The ROSA Project is to resolve issues in thermal-hydraulics analyses relevant to LWR safety by using the
LSTF. In particular, it intends to focus on the validation of simulation models and methods for complex
phenomena that may occur during DBE and beyond-DBE of LWRs. The key objective of the ROSA
Project is to provide integral and separate-effect experimental database to validate code predictive
capability and accuracy models. Especially, the LSTF experiments were planned to include phenomena
coupled with multi-dimensional mixing, stratification, parallel flows, oscillatory flows and noncondensable gas flows.
A group to pursue the key objectives was formed among the OECD Member countries who share the needto maintain or improve the technical competence in thermal-hydraulics for nuclear reactor safety (NRS)
evaluations. The experimental program was defined among the participants of the ROSA Project toprovide a valuable and broadly usable database to achieve the objectives.

Work performed in ROSA Project
To meet the objectives, the ROSA Project performed 12 LSTF experiments for the following 6 types ofsubject as either separate-effect test (SET) or integral-effect test (IET).
Test 1 Temperature stratification and coolant mixing during ECCS coolant injection
An IET and an SET were performed to clarify temperature stratification in cold legs and upper downcomerin pressure vessel (PV) under reactor-typical conditions, which is important for the evaluation ofpressurized thermal shock (PTS).
Test 2 Unstable and destructive phenomena such as water hammer
A series of SETs was performed to clarify influences of system pressure on occurrence and intensity ofcondensation-induced water hammer (CIWH) during ECCS injection into cold legs.
Test 3 Natural circulation under high core power conditions
Two IETs were performed to clarify thermal-hydraulic response adverse to core cooling during a SBLOCA
or an operational transient, both with high core power because of failure of reactor scram.
Test 4 Natural circulation with superheated steam

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Two SETs were performed to demonstrate a natural circulation of superheated steam during a severe
accident that may cause containment bypass in case of the rupture of steam generator (SG) U-tubes at
high-pressure sequence.
Test 5 Primary cooling through SG secondary depressurization
Two IETs simulated SBLOCA with a break in a cold leg, focusing on the long-term core cooling at low
system pressures by natural circulation induced by the SG secondary depressurization under influences of
non-condensable gas in SG U-tubes from accumulators.
Test 6 Open subjects
The subjects for this test series were open to the project participants. Candidates were discussed as follows:
1. PV upper-head SBLOCA -- Wall thinning found at the Davis Besse reactor in the US in 2002 raised
a safety issue concerning vessel structural integrity,
2. PV bottom SBLOCA -- A small amount of residue found at around the circumference of two
instrument-tube penetration nozzles at the South Texas Project Unit-1 in the US in 2003 raised asafety issue concerning vessel structural integrity,
3. Large-break LOCA (LBLOCA) focusing on the direct-contact condensation of steam on ECCS
water in cold legs – This phenomenon should influence the containment pressure. A good database
was lacking too.
4. SBLOCA focusing on accident management (AM) measures with symptom-oriented operator
actions.
Three LSTF experiments were dedicated for this test series. The 4th subject above was then decided
being coupled to the 1st and 2nd subjects. Consequently, two IETs (Tests 6-1 and 6-2) simulated the 1st
and 2nd subjects with SG depressurization as an AM measure. One SET (Test 6-3) was performed
corresponding to the 3rd subject. A new test section was prepared to simulate prototypical steam flow
conditions that may appear in cold legs; high flow rate with large superheating at near-atmosphericpressure.
Pre-test analyses were done by JAEA for all the experiments to survey optimum test conditions by using
such computer codes as RELAP5/MOD3.2.1.2, SKETCH-INS/TRAC-PF1 and FLUENT version 6.2 and6.3. The obtained results were distributed and discussed among the Project participants to define detailed
conditions of the LSTF experiments.
System modifications such as addition of new measurement instrumentation and/or preparation of new
test sections were performed to obtain detailed and specific data for each test. The new instrumentations
and/or preparation of new test sections include followings:(1) thermocouple arrays in cold and hot legs, PV downcomer and SG inlet plenum(2) non-condensable gas detectors at SG outlet plenum and PV upper-head(3) gas velocity meters for low-velocity gas flow in hot legs(4) video probes employing bore scopes for visual observation of flows in cold legs(5) nuclear data for the preparatory PWR analyses to simulate LOCA and anticipated transient, bothwithout scram(6) test section for CIWH with fast-response pressure sensors, a pitot tube and a gamma-ray
densitometer(7) test section for direct steam condensation on ECCS water downstream of the break unithorizontally connected to the LSTF cold leg in the loop without pressurizer
All above except (5) were newly prepared for the ROSA Project by JAEA experimental team.

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The obtained LSTF data has been utilized for the post-test analyses for the development, verification andvalidation of the thermal-hydraulic BE safety analysis codes and models as well as CFD codes. The
technical information and experiences obtained through the post-test analysis including methods toimprove the analysis accuracy were shared among the Project participants through the discussions in PRG(Program Review Group) meetings and over e-mail communications.
A questionnaire was performed further with the help from the Project participants who performed Test 61 post-test analysis to clarify their specific efforts in the analysis method to obtain better agreement with
the LSTF data.

Major Results and Significances
The 12 LSTF experiments for 6 subjects provided the following technical findings suitable for the V&V(verification and validation) of safety analysis codes and new issues for the assurance and improvement of
LWR safety.
Test 1 Temperature stratification and coolant mixing during ECCS coolant injection
An SBLOCA experiment with a 1% break at hot leg (Test 1-1, IET) and a steady natural circulation
experiment (Test 1-2, SET) were performed focusing on the temperature stratification in cold legs and PV
downcomer during ECCS injection. In Test 1-1, the primary inventory decreased gradually because of
break flow discharge under influences of the ECCS injection following plant control sequence. The
inventory was stepwise decreased in Test 1-2 where the ECCS injection was done at pre-determined
constant flow rates after steady condition was achieved in each step.
The two experiments revealed that the temperature stratification condition in cold leg depends on theinjection manner of the ECCS and that the liquid level in horizontal leg and flow condition may control
low-temperature coolant flowing down the PV downcomer along either PV inner wall side or core shroud
outer wall side.
Namely, the temperature stratification in cold legs was seen widely in both tests under both single- andtwo-phase natural circulation conditions. The degree of temperature stratification was larger in the intactloop cold-leg (loop with pressurizer), since the shape and the location of the ECCS injection nozzles aredifferent between two loops. It was found that the cold water from the cold leg flowed down in thedowncomer along the PV inner wall under the natural circulation condition when the downcomer liquid
level was higher than the cold leg. The cold water flowed down along the outer wall of the core barrel
cover when the loop flow was almost stagnated or the downcomer liquid level was lower than the cold leg.
Such temperature stratification important for the evaluation of pressurized thermal shock (PTS) was
observed under reactor typical pressure, temperature and flow conditions, and utilized for the validation ofboth CFD and BE codes.
Test 2 Unstable and destructive phenomena such as water hammer
A series of SETs was performed to study onset conditions and consequences of water hammer in a newhorizontal pipe test section attached to the LSTF under wide range of system pressure. It was confirmedthat the Condensation Induced Water Hammer (CIWH) occurs at high ambient pressure up to about 7
MPa.
The LSTF experiments clarified that the water hammer generates greatest pressure pulse peak when thesystem pressure was at around 1 MPa and it may take place up to 7 MPa. The averaged amplitude ofpressure pulses in each CIWH test increased with the system pressure when it was less than 1 MPa, anddecreased as the system pressure increases. A pressure pulse whose amplitude was larger than 10 MPa
formed even at the system pressure of 7 MPa, although the averaged amplitude was less than 1 MPa. The
estimated superficial steam velocity at the inlet of test section was related to the averaged amplitude ofpressure pulses, probably because the steam velocity is proportional to the steam condensation rate. Thisobservation suggested that the large volumetric shrinkage induced by the significant condensation rate8

NEA/CSNI/R(2013)1
produced the large pressure pulse. The water injection rate seemed to affect the dependency of the pressure
pulse amplitude on the system pressure. The effect of the water level at the outlet of test section on theintensity was small. No thermal stratification formed in the liquid phase once the CIWH was induced. The
effective mixing of the condensate with the subcooled water may have resulted in the large condensation
rate.
The onset conditions of and intensity of pressure pulses due to the CIWH was experimentally clarified asabove, which may threaten the integrity of reactor structures even under pressures typical of LWRs.
Test 3 Natural circulation under high core power conditions
Two IETs were performed under an assumption of failure of reactor scram to reveal various complex
thermal-hydraulic phenomena adverse to core cooling under high core power conditions and difficult forcode analyses. The core power for two tests was obtained from PWR analysis by using the SKETCHINS/TRAC-PF1 code. Total failure of high-pressure injection (HPI) system was assumed further in bothtests.
An SBLOCA with 1% cold leg break was simulated with Test 3-1. The natural circulation under high core
power contributed to maintain core cooling effectively for relatively long time until core uncover. Several
phenomena specific to the natural circulation under high core power were observed such as supercriticalflow in hot legs and counter-current flow limiting (CCFL) in SG inlet plenum and U-tubes because ofhigh-velocity steam flow. Coolant accumulation in the SG inlet plenum and U-tubes by the CCFL during
reflux condensation mode caused a small drop in the core liquid level. Steam condensation on coolant
injected from accumulator tank into cold leg induced loop seal clearing in the broken loop.
A loss-of-feedwater (LOFW) transient with delayed auxiliary feedwater actuation was simulated with Test
3-2. Core power was maintained at about 7.4% for more than 7 hours until the automatic power reduction
due to high core temperature, which was caused by a gradual coolant inventory loss due to cycle openingof the PORV. Two-phase natural circulation started in very early stage of the transient. Significant level
oscillation, in a form of slow fill and dump, occurred in all the instrumented SG U-tubes with gradualdecrease in the primary loop flow rate. Very complex reflux cooling phenomena appeared in the SG Utubes in non-uniform manner such that a water column developed rather randomly among the U-tubes witha lower frequency than that of the cycle opening of the RVs and PORV. Temperature excursion appearedin the core in the reflux condensation mode, even when the liquid level was still in the upper plenum,suggesting departure from nucleate boiling (DNB) due to the high core power.
Various and very complicated thermal-hydraulic responses were observed, including inadequate corecooling in the reflux condensation mode, associated with the natural circulation under high core power.
Test 4 Natural circulation with superheated steam
Two SETs were performed, with argon (Test 4-1) and super-heated steam (Test 4-2) as test fluid at thesystem pressure of 1 MPa and 4 MPa respectively, to clarify the onset conditions of superheated steam
natural circulation that may cause bypass sequence during severe accident if SG U-tube(s) rupture underhigh temperature and pressure conditions. The experiment conditions were defined and confirmed throughthe CFD (FLUENT) code analyses and LSTF preparatory experiments. The FLUENT analysis indicatedthat the gas temperature at the hot leg inlet should be higher than that in the SG secondary side by at least
100 K.
The gas temperature at the hot leg inlet in the PV upper plenum was higher than that in the SG secondaryside by about 300 K in Test 4-1 and by 150 K in Test 4-2, which was sufficiently larger than the estimatedtemperature difference. Gas velocity measured in the hot legs was about 0.3 m/s at maximum for bothtests, which was comparable to those estimated with FLUENT code.
Forward high-temperature gas flow from PV to SG and backward low-temperature gas flow from SG to
PV developed in the upper and the lower volumes of hot legs, respectively. In the upper volume, a flat
temperature profile formed at around the leg inlet and the gas temperature decreased toward SG. A flat
temperature profile formed in the lower volume near the SG inlet plenum and the gas temperature
increased toward PV with similar change of profile. This temperature profile could be related to the growth9

NEA/CSNI/R(2013)1of thermal boundary layer and the mixing of gas under the counter-current gas-gas flow conditions.
The natural circulation of superheated steam was confirmed to form in the LSTF primary loops including
counter-current gas-gas flow in hot legs. Further it was found that forward and backward flow existed at
the same time in SG U-tubes and hotter and colder gases mixed in SG inlet plenum. No large qualitative
difference appeared between argon and steam cases.
Test 5 Primary cooling through SG secondary depressurization
Two IETs, both simulating SBLOCA with a 0.5% cold leg break, were performed without (Test 5-1) orwith (Test 5-2) non-condensable gas from accumulator tanks of ECCS to clarify non-uniform parallel SG
U-tube behavior during the SG depressurization as an AM action. In both experiments, SG relief valves(RVs) were fully opened at 10 minutes after the safety injection signal. Total failure of HPI system wasassumed in both tests. To well observe the natural circulation phenomena at low pressures, an enhanced
SG depressurization by fully opening the safety valves (SVs) was performed when the primary pressure
decreased to 2 MPa as well as no actuation of low-pressure injection (LPI) system.
Two-phase natural circulation developed in the primary loops soon after the break valve opening andcontinued until the primary pressure decreased far below 1 MPa with non-uniform flow among the SG Utubes in Test 5-1 without non-condensable gas ingress.
In Test 5-2 with the gas ingress, on the other hand, the primary depressurization rate decreased after noncondensable gas started to enter primary loops and SG U-tubes, suggesting the occurrence of degradationin the condensation heat transfer in the SG U-tubes. Collapsed liquid level decreased gradually in the Utubes after the initiation of non-condensable gas inflow while two-phase natural circulation continued with
non-condensable gas. However, some U-tubes indicated a rapid coolant drain, while single-phase liquid
natural circulation continued in some other tubes, probably because of preferential accumulation of noncondensable gas into such drained U-tubes. Asymmetrical natural circulation between two loops also
appeared in Test 5-2 due probably to different number of forward flow U-tubes under the influences of
non-condensable gas. The natural circulation continued and contributed to maintain core cooling for a longtime even with the gas ingress into SG U-tubes.
In the two LSTF experiments various non-uniform flows in SG U-tubes under the influences of noncondensable gas onto heat transfer during the natural circulation was systematically observed especially at
low system pressures which is important for long-term core cooling.
Test 6 Open subjects
In response to the requests from Project participants, following three experiments (2 IETs and 1 SET) were
performed;
Test 6-1: An SBLOCA with a break at PV upper-head was simulated under assumption of total failure of
HPI system. The break size was equivalent to 1.9% cold leg break to simulate the ejection of one whole
CRDM penetration nozzle. SG secondary depressurization was performed as a symptom oriented AM
measure based on core exit temperature (CET).
The primary pressure started to decrease after the break and became lower than the SG secondary-side
pressure almost simultaneously with the core uncovery, resulting in no reflux flow from the SG. The
relatively lower primary pressure at the onset of SG depressurization after the superheating detection by
CETs made the initiation of coolant injection from accumulator tanks too late to adequately cool the corein time. A significant core temperature excursion then appeared and core power was terminated due to highcore temperature. Slow and late response of CETs relative to the core temperature excursion during coreboil-off was revealed, suggesting a necessity of further study on the CET responses because CETs are
utilized worldwide as one of key parameters to start AM operator action.
The result of this test raised a new safety issue on the reliability of the CET to detect core uncovery. A newtask group was then formed in 2008 under WGAMA (working group on analysis and management ofaccident) of OECD/NEA/CSNI and summarized the state-of-the-art knowledge with recommendations as a
CSNI report “Core Exit Temperature (CET) Effectiveness in Accident Management of Nuclear Power
Reactor” NEA/CSNI/R(2010)9. A paper was presented further for OECD/NEA special session at the10

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NURETH-14 at Toronto in 2011.
This issue is currently under investigation, including a ROSA-PKL counterpart-test with collaboration
between the OECD ROSA2 and PKL2 Projects, because the WGAMA task group identified that multidimensional behavior of superheated steam in the core should be a key and scaling-related when theobtained findings are applied to reactor accidents.
Test 6-2: An SBLOCA with a break at PV bottom was simulated with an asymmetrical SG secondary
depressurization as AM measure under assumption of total failure of HPI system. The break size was
equivalent to 0.1% cold leg break to simulate the ejection of one whole instrumentation nozzle. An
asymmetrical SG depressurization in loop w/o pressurizer (active loop) only was employed, expecting
unbalanced natural circulation between loops and unbalanced transport of non-condensable gas fromaccumulator tanks. The depressurization was initiated at 30 minutes after the safety injection signal,
considering an international common understanding on grace period to start operator action. The SG
depressurization rate was controlled such that the primary depressurization rate becomes to be 55 K/hr.
The SG depressurization was started before the core uncovery, because of very small break and slow
primary depressurization, resulting in continuous core cooling by two-phase natural circulation through theactive loop. After non-condensable gas ingress started from accumulator tanks, however, the primary
depressurization rate greatly decreased because of the degradation of condensation heat transfer in the SG
U-tubes, resulting in a large pressure difference between the primary and SG secondary. The core
uncovery finally started due to gradual boil-off because of the difficulty in the initiation of LPI system.
After the core uncovery started, CETs increased on the inactive loop side, while not on the active loop side
due to condensate fall-back from the SG. The LPI was actuated as the steam generation rate decreased in
the core, and most of the core was quenched. The core boil-off, however, started again because the LPI
was stopped when the primary pressure raised due to the increase in the steam generation rate in the core.
Core power was finally decreased automatically because of too-high core temperature during the second
core boil-off.
A unique database was provided by Test 6-2 including an asymmetric natural circulation core coolingunder the influences of non-condensable gas, which indicated a possibility of insufficient core cooling in along-term transient due to high primary pressure relative to LPI actuation pressure.
Test 6-3: Direct steam condensation on ECCS coolant in cold legs during reflood phase of LBLOCA wassimulated in a test section newly installed downstream of the break unit that is horizontally attached to the
LSTF cold leg. The test section consisted of a simulated cold leg pipe (4.5 m-long, 102.3 mm i.d.), asimulated ECCS injection nozzle, three viewers, a separator, separator downstream pipes for steam andwater. Steam velocity at the inlet of the simulated cold leg pipe ranged from about 20 to 50 m/s being
controlled by the LSTF core power, covering conditions estimated from the PWR LBLOCA analyses. The
inlet steam superheating was as high as possible to best simulate the PWR conditions. Simulated ECCS
water was at temperature of 310 K with a mass flux ranging from about 12 to 207 kg/(m2s) so as topreserve the ratio of ECCS water injection flow rate to the inlet steam flow rate.
Significant steam condensation was observed between the subcooled ECCS coolant and highly superheatedsteam in a short distance from the ECCS injection point. Fluid temperature distribution was non-uniform at
about 50 mm downstream from the ECCS injection point but became to be almost uniform in less thanabout 350 mm. Total steam condensation rate estimated from the difference in the steam flow rates
between the inlet and outlet of test section was in proportion to the simulated ECCS water mass flux until
the complete condensation of steam. Overall condensation rate was linearly proportional to the injected
ECCS coolant.
Image data on liquid droplets under condensation in the high-velocity steam flow was obtained further byhigh-speed video camera for in-depth investigation of the flow conditions under non-equilibrium steam
condensation, necessary for the verification and validation of the models and safety analysis codes.

11

NEA/CSNI/R(2013)1
Conclusions and Recommendations
The OECD/NEA ROSA Project to resolve issues in thermal-hydraulics analyses relevant to LWR safety
was conducted during 4 years since April 2005 among 14 countries by using the LSTF of the ROSA
Program in JAEA. Major objectives of the ROSA Project are as follows:(1) Provide integral and separate-effect experimental database to validate code predictive capability andaccuracy of models for complex phenomena that may occur during DBE and beyond-DBE.
(2) Clarify the predictive capability of codes currently used for thermal-hydraulic safety analyses as well
as of advanced codes presently under development.
The LSTF is a full-pressure full-height two-loop model of an 1100 MWe 4-loop Westinghouse-type PWR
with a volumetric scaling of 1/48, and has been utilized for the investigation of thermal-hydraulic response
during accidents and abnormal transients since 1985, by fully utilizing the capability to demonstrate thekey phenomena.
The ROSA Project investigated various types of thermal-hydraulic responses that may encounter during
accidents and abnormal transients of LWR through 6 subjects with 12 LSTF experiments in either IET or
SET. Care was taken to include such complex phenomena as multi-dimensional mixing, stratification,parallel flows, oscillatory flows and/or those with non-condensable gas. Efforts were taken further toincrease data parameters necessary for better understanding of the phenomena by employing four types of
JAEA-designed new instrumentation as well as two new test sections for specific subjects of CIWH anddirect contact steam condensation during reflood phase of large break LOCA, both with much of thespecially designed instrumentation.
For all the experiments, JAEA performed pre-test analyses to survey optimum test conditions to meet theobjectives of each test and the ROSA Project. The obtained results were distributed and discussed among
the Project participants to define detailed test conditions.
From the 12 LSTF experiments, a unique database was obtained, and was readily shared among the Project
participants to utilize it for the post-test analyses for the verification and validation of the safety analysis
codes and models as well as CFD codes, both under development and improvement. This database may
also be added as appropriate to the CCVM for its next revision. The technical information obtained
through such effort was discussed and shared through the PRG meetings, partly through a questionnaire forthe post-test analysis of PV upper-head break LOCA test (Test 6-1), and over e-mail communications.
Inspired by the result of Test 6-1, a safety issue was newly raised on the effectiveness of CET that is
utilized to start an operator action to cool uncovered core as an accident management measure. A task
group was formed in WGAMA, issued a CSNI report of NEA/CSNI/R(2010)9 and prepared a co-authoredpaper for the NURETH-14 in 2011.
Through the ROSA Project, the LSTF has been recognized useful and suitable for the international
research activities to resolve issues in thermal-hydraulics analyses relevant to LWR safety. Further
utilization of the LSTF would be beneficial to reinforce LWR thermal-hydraulic safety by addressing suchspecific issues as V&V of BE and CFD codes and scaling of thermal-hydraulic phenomena necessary forthe accurate safety evaluation including BEPU through in depth understanding of the complicatedphenomena that may encounter during DBAs and beyond-DBAs.

12

NEA/CSNI/R(2013)1

TABLE OF CONTENTS

1.

BACKGROUND, OBJECTIVES & METHODS ................................................................................. 15
1.1
1.2
1.3

2.

ROSA/LSTF .......................................................................................................................................... 21
2.1
2.2

3.

Background of ROSA Project ........................................................................................................ 15
Objectives of ROSA Project........................................................................................................... 16
Scope of Work of ROSA Project.................................................................................................... 16
Facility Outline and Unique Design Features................................................................................. 21
Design Philosophy and Scaling of Major Components .................................................................. 23

TEST DESCRIPTION AND MAJOR RESULTS ................................................................................. 25
3.1 Test 1 Temperature stratification during ECCS coolant injection.................................................. 25
3.1.1 Background ............................................................................................................................... 25
3.1.2 Objectives ................................................................................................................................. 25
3.1.3 Description of Experiments ...................................................................................................... 25
3.1.4 Experiment Results ................................................................................................................... 26
3.2 Test 2 Unstable and destructive phenomena such as water hammer .............................................. 28
3.2.1 Background ............................................................................................................................... 28
3.2.2 Objectives ................................................................................................................................. 28
3.2.3 Description of Experiments ...................................................................................................... 28
3.2.4 Experiment Results ................................................................................................................... 29
3.3 Test 3 Natural circulation under high core power conditions......................................................... 32
3.3.1 Test 3-1 Small-break LOCA without scram ............................................................................. 32
3.3.1.1
Background ....................................................................................................................... 32
3.3.1.2
Objectives ......................................................................................................................... 32
3.3.1.3
Description of Experiment ................................................................................................ 32
3.3.1.4
Experiment Results ........................................................................................................... 33
3.3.2 Test 3-2 Loss-of-feedwater transient without scram................................................................. 33
3.3.2.1
Background ....................................................................................................................... 33
3.3.2.2
Objectives ......................................................................................................................... 33
3.3.2.3
Description of Experiment ................................................................................................ 33
3.3.2.4
Experiment Results ........................................................................................................... 34
3.4 Test 4 Natural circulation with superheated steam......................................................................... 36
3.4.1 Background ............................................................................................................................... 36
3.4.2 Objectives ................................................................................................................................. 36
3.4.3 Description of Experiments ...................................................................................................... 36
3.4.4 Experiment Results ................................................................................................................... 37
3.5 Test 5 Primary cooling through SG secondary depressurization.................................................... 40
3.5.1 Background ............................................................................................................................... 40
3.5.2 Objectives ................................................................................................................................. 40
3.5.3 Description of Experiments ...................................................................................................... 40
3.5.4 Experiment Results ................................................................................................................... 41
3.6 Test 6 Open subjects....................................................................................................................... 43
3.6.1 Test 6-1 PV upper-head break LOCA with AM measure ......................................................... 43
3.6.1.1
Background ....................................................................................................................... 43
3.6.1.2
Objectives ......................................................................................................................... 43
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NEA/CSNI/R(2013)1
3.6.1.3
Description of Experiment ................................................................................................ 43
3.6.1.4
Experiment Results ........................................................................................................... 44
3.6.2 Test 6-2 PV bottom break LOCA with AM measure ............................................................... 44
3.6.2.1
Background ....................................................................................................................... 44
3.6.2.2
Objectives ......................................................................................................................... 44
3.6.2.3
Description of Experiment ................................................................................................ 45
3.6.2.4
Experiment Results ........................................................................................................... 45
3.6.3 Test 6-3 Direct-contact condensation in cold leg during large-break LOCA ........................... 45
3.6.3.1
Background ....................................................................................................................... 45
3.6.3.2
Objectives ......................................................................................................................... 46
3.6.3.3
Description of Experiment ................................................................................................ 46
3.6.3.4
Experiment Results ........................................................................................................... 46
4.

SUMMARY ........................................................................................................................................... 53

APPENDIX 1

FACILITY MODIFICATION AND NEW INSTRUMENTATION ................................. 57

APPENDIX 2

POST-TEST ANALYSES ................................................................................................. 63

APPENDIX 3

LIST OF EXPERIMENTAL DATA.................................................................................. 91

APPENDIX 4

LIST OF PUBLISHING ................................................................................................... 95

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NEA/CSNI/R(2013)1

1.

1.1

BACKGROUND, OBJECTIVES & METHODS

Background of ROSA ProjectComputer codes are utilized in safety evaluation of light water reactors (LWRs) in order to simulate plant
behavior during design basis event (DBE) and beyond-DBE (BDBE). This involves complex multidimensional single-phase and two-phase flow conditions, which may include non-condensable gas in manycases. For this purpose, thermal-hydraulic safety analysis codes, especially for best estimate (BE) loss-ofcoolant accident (LOCA) analyses codes, have been developed with significant efforts to achieve highpredictive capability especially for one-dimensional phenomena such as flows in piping at rather high flow
rate. Such BE codes include APROS, ATHLET, CATHARE, MARS, RELAP5, TRAC and TRACE.
However, there remain needs for experimental work and code development and validation for the complex
flow conditions during DBE and BDBE. Further, the increased use of the BE analysis codes in the safety
evaluation of reactors, which is replacing traditional conservative approaches, requires the validation andthe quantification of uncertainties in the simulation models and methods.
Many experimental facilities have contributed to the thermal-hydraulic databases available today including
those for the CSNI code validation matrix (CCVM). However, a large portion of current data is insufficientfor future codes that are to incorporate multi-dimensional simulation capabilities, mainly because thespatial resolution of measurement and the measurement parameters are not enough to assess the simulation
models and methods.
Meanwhile, the Committee on the Safety of Nuclear Installations (CSNI) of the Nuclear Energy Agency(NEA) of OECD set up an international working group SESAR/FAP (Senior Group of Experts on Nuclear
Safety Research Facilities and Programmes [1-1]) in 1997 to identify how existing expertise in the field ofreactor safety and an appropriate experimental infrastructure could be sustained in the future. One of themain tasks of the SESAR/FAP was to identify test facilities and research programs threatened by closure
shortly and to select facilities and programmes that would be of particular benefit to the member countries,
if they were able to continue in a form of OECD project. Several international co-operative research
projects had been started in the field of thermal hydraulics safety research to implement therecommendation by the SESAR/FAP, which included the SETH (SESAR Thermal-Hydraulics) Projectusing two test facilities; PANDA of PSI, Switzerland and PKL of AREVA, Germany, and PKL Project.
Later, the CSNI established the SESAR/SFEAR (Support Facilities for Existing and Advanced Reactors[1-2]) in 2005 as a follow-up of the SESAR/FAP to assess the need for and strategy of maintaining key
research facilities. In both working groups, ROSA/LSTF [1-3] in JAEA (former JAERI) was discussed andevaluated as an excellent facility for the thermal-hydraulic safety research. However, the ROSA/LSTF was
not considered for the OECD/NEA Project because of heavy utilization for Japanese domestic safety
research in the late 1990s, for example, to confirm the effectiveness of passive safety installations such as
PCCS (passive containment cooling system) with a horizontal heat exchanger in collaboration withindustry. In the early 2000s, JAEA decided to propose an OECD/NEA project to make the ROSA/LSTF
utilized more effectively and efficiently among the member countries.
The ROSA/LSTF was constructed in 1985 at JAEA (formerly JAERI) following the TMI accident toexperimentally investigate thermal-hydraulic phenomena that may appear during postulated abnormaltransients and small-break LOCAs (SBLOCAs) as described in Chapter 2. More than 180 experiments
15

NEA/CSNI/R(2013)1
were performed prior to the ROSA Project, indicating its excellent capabilities to simulate the thermalhydraulic phenomena under conditions close to those in the reference reactor. One of the important
benefits of utilizing the ROSA/LSTF for the OECD/NEA Project would be the combination of the facility
with the latest instrumentation to measure various types of measurement parameters with more detailed
spatial resolution than ever. In the late 1980s and 1990s, the data obtained from the ROSA/LSTF
experiments were designed to be suitable for the verification of one-dimensional BE codes. The observed
phenomena, however, were often characterized by multi-dimensional effects, including parallel channels,
unstable, multi-phase flows under influences of non-condensable gas. The obtained data in such early days
were sometimes lack in the spatial resolution and parameters to well describe the multidimensional
phenomena.
The proposed ROSA Project tries to generate and share the necessary data among the OECD member
countries to fulfill the current and future need for the development, verification and validation of both the
BE and computational fluid dynamics (CFD) codes, by taking a view of the combined utilization of BE
code with uncertainty evaluation methods.

1.2

Objectives of ROSA Project

The OECD/NEA ROSA Project is to resolve issues in thermal-hydraulics analyses relevant to LWR safety
using the ROSA/LSTF facility of JAEA [1-3]. In particular, it intends to focus on the validation ofsimulation models and methods for complex phenomena that may occur during DBE and beyond-DBE.
The key objective of the ROSA Project is to provide integral and separate-effect experimental database tovalidate code predictive capability and accuracy models. Especially, the ROSA/LSTF experiments wereplanned to include phenomena coupled with multi-dimensional mixing, stratification, parallel flows,
oscillatory flows and non-condensable gas flows.
A group to pursue the key objectives formed among the OECD Member countries who share the need tomaintain or improve the technical competence in thermal-hydraulics for nuclear reactor safety (NRS)
evaluations.
The experimental program was defined to provide a valuable and broadly usable database to achieve theobjectives. Twelve experiments were conducted in six types of subject, which are briefly explained in thefollowing section.
1.3

Scope of Work of ROSA Project

The OECD/NEA ROSA Project consists of the following six types of subject; Test 1 through Test 6,
which are either separate-effect test (SET) or integral-effect test (IET). Twelve LSTF experiments were
performed for the six types of subject as shown in Table 1-1. Table 1-1 contains major subject, testclassification (SET or IET), specific safety significance, suitability of data to use for the validation of BE
and/or CFD code, and additional specific features such as new instrumentation, facility modification and/or
preparatory (pre-test) analyses.
Major significances for each of 6 subjects are as follows:
Test 1 Temperature stratification and coolant mixing during ECCS coolant injection
This test is intended to clarify temperature stratification in cold legs and upper downcomer in pressure
vessel (PV), which is important for the evaluation of pressurized thermal shock (PTS).
Two LSTF experiments were intended to simulate the temperature stratification under the reactortypical conditions; pressure, temperature, temperature difference and flow transient in rather
realistically-shaped horizontal legs with an emergency core cooling system (ECCS) injection port and aconcentric downcomer.
16

NEA/CSNI/R(2013)1
Test 2 Unstable and destructive phenomena such as water hammer
This test is intended to clarify influence of system pressure on occurrence and intensity ofcondensation-induced water hammer (CIWH) during ECCS injection.
A series of LSTF experiments was intended to simulate the CIWH condition under reactor-typicalpressure. Namely, the relevant conditions including temperature difference between steam at saturation
temperature and ECCS water at room temperature were achieved under wide range of the systempressure up to about 7 MPa.
Test 3 Natural circulation under high core power conditions
This test is intended to clarify thermal-hydraulic response adverse to core cooling during a small- break
LOCA and operational transient with high core power because of failure of reactor scram.
Two LSTF experiments were intended to simulate two different IETs with well-simulated core powertransient after the initiation of accident and transient, being coupled with the reactor-typical thermalhydraulic responses; pressure, core power and corresponding steam generation rate and resultant flowtransient including natural circulation with high flow rates of steam and liquid.
Test 4 Natural circulation with superheated steam
This test is intended to experimentally indicate the natural circulation with superheated steam during asevere accident that may cause containment bypass in case of the rupture of steam generator (SG) Utubes at high-pressure sequence.
Two LSTF experiments were intended to simulate the natural circulation including gas-gas countercurrent flows under the reactor-typical conditions; pressure, temperature, temperature difference
between PV upper plenum and SG secondary side, and flow transient in realistically shaped hot leg andSG.
Test 5 Primary cooling through SG secondary depressurization
This test is intended to clarify thermal-hydraulic response in LOCAs, especially for the long-term corecooling at low system pressures by natural circulation induced by the SG secondary depressurization
under influences of non-condensable gas from accumulators.
Two LSTF experiments were intended to simulate IETs through the transient from the initial steadycondition at the nominal operating condition for pressurized water reactor (PWR) to the long-term corecooling at near-atmospheric pressure, with and without the influences of non-condensable gas migration
into SG U-tubes.
Test 6 Open subjects
This test is intended to fulfill the requests from the project participants by means of three LSTF
experiments that respectively simulated two types of small-break LOCA with the break at the top of
upper-head or at the bottom of the PV, and one local phenomena as direct steam condensation on ECCS
coolant in cold legs during the reflood phase in a large-break LOCA of PWR.
Three LSTF experiments were intended to perform two IETs (Tests 6-1 and 6-2) that simulate LOCAs
from the nominal operating condition of PWR with an accident management (AM) measure, and aseries of SETs (Test 6-3) under prototype steam flow conditions; high flow rate and large superheating
at near-atmospheric pressure.
The twelve LSTF experiments above were intended to provide data for the development, validation andverification of computer codes and models, aiming to increase the level of details and accuracy in theanalyses of the key phenomena during transients and accidents of interest.
17

NEA/CSNI/R(2013)1
To meet this objective, system modifications, addition of new instrumentation and/or preparation of new
test sections were performed for the ROSA/LSTF, which are mainly composed of the following;(1) thermocouple arrays in cold legs and downcomer, hot legs and SG inlet plenum(2) non-condensable gas detectors at SG outlet plenum and PV upper-head(3) gas velocity meters in hot legs(4) glass-fiber type video probes for visual observation of flows in cold legs(5) nuclear data for the preparatory PWR analyses to simulate LOCA and anticipated transient, bothwithout scram(6) test section for condensation-induced water hammer (CIWH) with fast-response pressure
sensors, a pitot tube and a gamma-ray densitometer(7) test section for direct steam condensation on ECCS water downstream of break unit for the coldleg (horizontal orientation) in the loop without pressurizer
Measurement instrumentation (2), (3) and (4) were developed by JAEA for the LSTF utilization andapplied to the corresponding ROSA Project experiments. (1), (2), (3), (4), (6) and (7) are presented inAppendix 1.
References[1-1]

OECD/NEA, 2001. Nuclear Safety Research in OECD Countries: Major Facilities andProgrammes at Risk (SESAR/FAP). OECD/NEA, Nuclear Safety, ISBN 92-64-18463-5.

[1-2]

OECD/NEA, 2007. Nuclear Safety Research in OECD Countries: Support Facilities for Existingand Advanced Reactors (SESAR/SFEAR). Nuclear Safety NEA/CSNI/R 6, ISBN 978-92-6499005-0.

[1-3]

The ROSA-V Group, 2003. ROSA-V Large Scale Test Facility (LSTF) System Description for
The Third and Fourth Simulated Fuel Assemblies. JAERI-Tech 2003-037, Japan Atomic Energy
Research Institute.

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NEA/CSNI/R(2013)1
Table 1-1
Test

IET
or
SET1)

Subjects

SET

Temperature stratification and coolant mixing in cold
legs and downcomer during ECCS coolant injection
Temperature stratification under quasi-steadycondition during natural circulation
Temperature stratification under transient during hot
leg small break LOCA (1% cold leg break equivalent)

1
1-1
1-2

ROSA/LSTF Experiments for OECD/NEA ROSA Project

IET

Safety Significances2)

Additional Specific Features
Data(New Instrumentation, Facility Modifications
3)
Utilization
and/or Preparatory Analysis)zzClarify temperature stratification
conditions important for evaluationof pressurized thermal shock (PTS)

BE & CFD
z
z

z
2

SET

3
3-1

IET

3-2

IET

4
4-1

SET

4-2

SET

Unstable and destructive phenomena such as water
hammerNatural circulation under high core power conditions
1% cold leg break LOCA without scramzzNatural circulation of superheated steam

z
z
z
z
z
z

z

Loss-of-feedwater (LOFW) transient without scramwith auxiliary feedwater injection
Natural circulation with superheated steam during highpressure sequence of severe accident
Natural circulation of argon simulating superheatedsteam

Clarify onset conditions of andintensity of pressure pulses due tocondensation-induced water hammer
BE & CFD(CIWH) that may threaten the
integrity of reactor structures underpressures typical of LWRs

Clarify thermal-hydraulic responses
adverse to core cooling during bothaccident and abnormal transient
BE & CFD
under high core power condition due
to failure of reactor scram
Clarify onset conditions ofsuperheated steam natural
circulation that may cause bypass
sequence during severe accident due
to rupture of SG U-tubes
BE & CFD
Measure low-velocity gas flows
during natural circulation, whichwere scarcely done due to difficulty
in measurement

19

z
z
z

z

Array of thermocouples to cold legsfor detailed measurement of
temperature distribution
Video Probe to cold legs for visualobservation of flows
New test section with 67 mm i.d.
and 2050 mm in length for welldefined boundary conditions
Fast response pressure sensors
Array of thermocouples
Pitot tube to measure steam velocity(failed)Gamma-ray densitometer
Pre-test analyses by RELAP5 code
to expect flow conditions
Pre-test analyses on PWR and LSTF
to define core power curve
respectively by SKETCHINS/TRAC-PF1 and
RELAP5/MOD3.2 codes
LSTF preparatory tests to define
break size and core power curve
Pre-test analysis with FLUENT code
to define test conditions
Array of thermocouples to hot legsand SG inlet plenum for detailed
measurement of gas temperature
distribution
JAEA-developed pulsed-wirevelocity meter for measurement ofslow gas flow

NEA/CSNI/R(2013)1
Table 1-1

Test

IET
or
SET1)

5
5-1

IET

5-2

IET

ROSA/LSTF Experiments for OECD/NEA ROSA Project (continued)

Safety Significances2)

Subjects
Primary cooling through SG secondary depressurizationas accident management (AM) measure
0.5% cold leg break LOCA without non-condensable
gas inflow from accumulatorz

0.5% cold leg break LOCA with non-condensable gas
inflow from accumulator tanksz

Clarify thermal-hydraulicresponses, especially for long-termcore cooling by natural circulation at
low system pressures underinfluences of non-condensable gas
Confirm effectiveness of AM
measureAdditional Specific Features
Data(New Instrumentation, Facility Modifications
3)
Utilization
and/or Preparatory Analysis)z
BEz

Open subject (requests from participants)

6

z
6-1

IET

Pressure vessel upper-head break LOCA withsymptom oriented AM measure(1.9% cold leg break equivalent)

6-2

IET

Pressure vessel bottom break LOCA and AM
effectiveness (0.1% cold leg break equivalent)zzClarify consequences of SBLOCA
and effectiveness of AM measure
Investigate responses of core exit
thermocouples (CETs) during coreboil-off
Clarify consequences of SBLOCA
and effectiveness of AM measure

BE & CFDz RELAP5/MOD3.2 code analyses on
PWR and LSTF to define break size
and to study influences and onset
timing of AM action

BE
z

z

6-3

JAEA-developed non-condensable
gas sensor to measure gas
concentration in SG outlet plenum
and PV upper-head
RELAP5/MOD3.2 code analyses on
PWR and LSTF to define break sizeand onset timing of AM action

SET

Steam condensation on ECCS water at reflood phaseduring large-break LOCA (LBLOCA)z

Clarify conditions of steam directcontact condensation on ECCS
coolant during PWR LBLOCA that
may impact on containment pressure
BE & CFD
Obtain steam condensation data
under typical LBLOCA condition
without influences of noncondensable gas, which was scarcedue to difficulty in measurement

1) IET: Integral-effect test, SET: Separate-effect test2) Common to all the tests: Provide LSTF database for verification and development of BE (and/or CFD) codes
3) BE: best estimate codes, CFD: computational fluid dynamics codes

20

z
z
z
z
z

New test section to simulate flow incold leg under well-definedboundary conditions
High-speed video for visualobservation of flows
Array of thermocouples
Pitot tube to measure steam velocity
Gamma-ray densitometer
RELAP5/MOD3.2 code analyses on
PWR and LSTF to define velocity
and temperature of steam at test
section inlet

NEA/CSNI/R(2013)1

2.

2.1

ROSA/LSTF

Facility Outline and Unique Design Features

The LSTF (Large Scale Test Facility) [2-1] shown in Fig. 2-1 is a full-pressure and full-height integral
test facility constructed in 1985 to investigate thermal-hydraulic response during PWR transients andaccidents for the fourth phase of ROSA (Rig-of-Safety Assessment, ROSA-IV) Program at JAEA (Japan
Atomic Energy Agency). The LSTF simulates a 1100 MWe four-loop Westinghouse (W)-type PWR:
Tsuruga Unit-2 reactor of the JAPC (Japan Atomic Power Company).
The Three Mile Island Unit-2 (TMI-2) reactor accident in 1979 led to a reorientation of the LWR safety
research to pursue SBLOCAs and operational/abnormal transients. Of primary importance for the LSTF
was then a good simulation of the natural circulation in the primary loops involving two-phase flows andcounter-current flows. The major design characteristics of the LSTF are summarized in Table 2-1.
Table 2-1 Major Design Characteristics of LSTF to Reference PWR
Items
Primary/Secondary Pressures (MPa)Primary/Secondary Temperatures
Core Height

(K)

(m)

Number of Fuel Rods

LSTF

PWR

PWR/LSTF

16 / 7.4

16 / 6.13

1 / 0.83

598 / 562

598 / 550

1 / 0.98

3.66

3.66

1

1008

50952

50.55

8.14

347

42.6

10

3423(t)

342

1.23

8.8

8.0

0.0488

16.7

342

0.053

0.26

4.91

2

4

2

0.207

0.737

3.56

3.69

6.99

1.89

8.11

8.14

1.0

0.124

2.98

24.0

Number of Tubes in Steam Generator (SG)

141

3382

24.0

Average Length of SG Tubes (m)

20.2

20.2

1

Primary Fluid Volume

V

(m3)

Total Core Power

Q

(MW)

Q / V (MW/m3)
Core Inlet Flow

(ton/s)Pressure Vessel Downcomer Gap(m)

Number of Primary Loops
Hot Leg Inner Diameter
Length L

L/ D
VolumeD(m)(m)(m1/2)

π
4

D2 L

(m3)

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NEA/CSNI/R(2013)1

Fig. 2-1 Large Scale Test Facility (LSTF) of Rig-of-Safety Assessment (ROSA) Program in JAEAThe ROSA/LSTF is the world largest integral test facility for W-type PWR. Multi-dimensional thermalhydraulic response during reactor transient is simulated by using a 10 MW electrically heated full sizecore rod bundle with more than 1000 simulated fuel rods, about 50 mm width annulus downcomer in thepressure vessel, 207 mm inner diameter hot and cold legs, 141 full size U-tubes in each of two steam
generator.
The four primary loops of the reference PWR are represented by two equal-volume loops to best simulate
two-phase flows during reactor accidents and transients mainly by achieving the large-diameter horizontal
legs. The volumetric scaling ratio of the primary loops is 1/48 to those of the reference PWR. The LSTF
can handle wide ranges of primary and secondary pressures from reactor nominal operating pressure toatmospheric pressure.
The LSTF is equipped with all the ECCS (emergency core cooling system) with additional new features
such as ECCS coolant temperature control, temperature stratification in accumulator coolant tank to injectcoolant by flashing of high temperature coolant portion without N2 gas pressurization, though thesefeatures were not used in the OECD/NEA ROSA Project.
Instrumentation around 1900 channels provide detailed information on thermal-hydraulic conditions by
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NEA/CSNI/R(2013)1
measuring such parameters as temperature, liquid level, pressure loss, flow rate, density. Visualobservation of flow is possible using “video-probe”; a periscope that withstands high-temperature
steam/water conditions employing glass-fiber image guide or bore scopes. Non-condensable gas detectors
and detailed temperature measurement was furnished further to the LSTF in the OECD/NEA ROSA
Project.
Since the shakedown test in 1985, the LSTF has provided experimental data including the 5% cold-leg
LOCA simulation for OECD/NEA ISP-26, the Mihama Unit-2 SGTR (steam generator tube rupture)
simulation for the Nuclear Safety Commission of Japan and the AP600 reactor simulation for the
USNRC. Throughout its operational history, the facility has demonstrated excellent experimental
capabilities and has provided unique data for non-equilibrium, non-homogeneous and multi-dimensional
phenomena.

2.2

Design Philosophy and Scaling of Major Components

Design philosophy and scaling of major components of the LSTF are summarized as follows:
Design Pressure covers that in the reference PWR. The rated pressure of LSTF pressure vessel is 17.26
MPa. The experiments can thus be performed under full-pressure conditions to simulate the thermalhydraulic phenomena in the reference PWR via corresponding fluid properties.
Core Power is limited to 10 MWt, which is scaled by 1/48 at core power equal to and less than 14% of
the scaled rated power of the reference PWR.
Elevations are preserved, i.e., one to one correspondence with the reference PWR, considering good
representation of the driving force of natural circulation. As for the horizontal legs, pipe inner diameter(207 mm) is far smaller than that of the reference PWR (737 mm). The top elevation of hot leg innerdiameter is thus set equal to that of the reference PWR. As for the cold legs, elevation of leg horizontal
axis is set equal to that of the reference PWR.
Volumes of major components such as pressure vessel (PV) and pressurizer are scaled by 1/48 to thereference PWR. Since two loops are lumped into one, the volume of SG is twice as much as scaled one.
The scaling factor is about 1/21 to the two-loop PWR such as Mihama Unit-2.
Flow Area is scaled by 1/48 in the pressure vessel and by 1/24 in the primary loops and SGs. The flow
area in the horizontal legs is scaled to conserve the ratio of the length L to the square-root of pipe
diameter D; Error! Objects cannot be created from editing field codes. of the reference PWR to better
simulate the flow regime transitions in the primary loops (Froude number basis). The time scale ofsimulated thermal-hydraulic phenomena is thus one to one to those in the reference PWR.
Fuel Assembly has mostly the same dimensions for the following points as those of PWR 17x17 fuelassembly; diameter, length and pitch of fuel rod, diameter, length and pitch of control rod guide thimble,
and the ratio of number of fuel rods to number of the guide thimbles. This preserves the heat transfer
characteristics of the core. The total number of the simulated fuel rods is scaled by about 1/48. There are1008 heated rods in the current fourth fuel assembly.
Steam Generators are designed to simulate primary-to-secondary thermal interactions during most of
SBLOCAs or operational transients. The rated 10 MWt power sets the maximum steady-state steam andfeedwater flow requirements to 14% of the scaled flow required for the reference PWR operating at fullpower. The secondary pressure is elevated at the initial steady state operation to suppress and control theheat transfer with the scaled heat transfer area of U-tubes and 14% primary loop flow rate. U-tube
bundles have mostly the same dimensions with those of the reference PWR.
Flow Capacities are scaled by 1/48 where practical.
Pressure Loss due to Fluid Flow is designed to be equal to that in the reference PWR for scaled flow
rates.
23

NEA/CSNI/R(2013)1
Break in the reference PWR is simulated by using a break unit that can be connected at 19 locations of
the LSTF. Break size is controlled by orifice or nozzle. The tested maximum size of break has been 10%
cold leg break equivalent for the primary loop.
ECCS is designed with adjustable operational envelopes that cover capacity at the normal envelope andbeyond. Piping and nozzles are provided so that ECCS fluid can be injected into several possible
locations around the primary coolant system other than those typical to the reference PWR for someparameter investigations. All types of ECCS furnished to the reference PWR are provided; high-pressure
injection system (HPIS), accumulator injection system (AIS), low-pressure injection system (LPIS) and
residual heat removal (RHR) system. The LSTF is equipped with a gravity-driven injection system(GDIS) as a kind of passive safety feature, though it was not used in the OECD/NEA ROSA Project.
Process Control System includes logic that considers the following four experimental needs;a)A computer controlled sequence control program drives steady-state and transient events. Constants
vary according to demands for each experiment.

b)

The interlock program incorporates special control features of the reference PWR for facility
equipment trip control.

c)

The component control programs contains logic specific to key individual component control suchas core power decay, pressurizer heater control and pump rotation increase/coastdown.

d)

The facility protection control system governs the equipment safety interlocks such as core overtemperature protection.

Instrumentation are designeda)to simulate the process instrumentation of the reference PWR to control the LSTF system during
SBLOCA and operational transient simulation experiments,b)

to measure thermal-hydraulic phenomena as precisely as possible, andc)to develop new instruments for advanced measurements in need.

As for the point b), three-dimensional fluid behaviors are to be detected by a combination ofthermocouples, water level measurements and gamma-ray densitometers with aid of visual observation by
the video probe at hot and cold legs. As for the point c), non-condensable gas detectors and low-velocity
gas velocity meter were developed in JAEA for the OECD/NEA ROSA Project.
References[2-1]

The ROSA-V Group, 2003. ROSA-V Large Scale Test Facility (LSTF) System Description for
The Third and Fourth Simulated Fuel Assemblies. JAERI-Tech 2003-037, Japan Atomic
Energy Research Institute.

24

NEA/CSNI/R(2013)1

3.

TEST DESCRIPTION AND MAJOR RESULTS

3.1

Test 1 Temperature stratification during ECCS coolant injection

3.1.1

Background

During a small-break loss-of-coolant accident (LOCA) of pressurized water reactor (PWR), theemergency core cooling system (ECCS) is actuated and cold water is injected into cold legs. Injectedwater mixes with hot primary coolant and flows into the pressure vessel through the downcomer. Some of
injected water may flow at the bottom of the cold leg, forming a cold layer less mixed with the upper hotlayer. Insufficient mixing of cold water and hot coolant results in temperature stratification. In case thatliquid level forms in the cold leg, steam condensation on injected water occurs, and local pressure
variations may cause unstable flow oscillations, which in turn promote coolant mixing. Such multidimensional and non-equilibrium flow phenomena shown in Fig. 3.1-1 are of concern for pressurized
thermal shock (PTS) in view of aging and plant life extension. Experiments for thermal hydraulicphenomena pertaining to PTS, are summarized in several project reports [3.1-1~3.1-3].
Computational fluid dynamics (CFD) is a powerful tool for predicting various flow phenomena especiallyin single-phase flows. Some mixing phenomena have been simulated using commercial and in-house
CFD codes relating to boron dilution and PTS scenarios [3.1-4~3.1-7], and two-phase cases have also
been reported [3.1-8]. Applications of CFD codes to analyze such complicated nuclear safety problems as
the temperature stratification phenomena are of interest especially under two-phase flow conditions [3.19]. Although some experimental results have been used for validation of CFD codes [3.1-2, 3.1-3],
experimental data obtained under high-pressure and high-temperature single- and two-phase flow
conditions corresponding to PWR are limited.
3.1.2

Objectives

The objective of Test series 1 was to obtain the three-dimensional temperature distributions in cold legsand downcomer during ECCS coolant injection under PWR conditions for verification of computer codesand simulation models. Two types of experiments were conducted: steady-state and transient tests. In
Test 1-1, the ECCS coolant was injected into the cold legs under steady-state single- and two-phasenatural circulation conditions, and the steady-state mixing phenomena under controlled boundary
conditions were observed. These data would be suitable for an assessment of turbulent mixing andcondensation models. A hot-leg small-break LOCA was simulated in Test 1-2 to obtain the temperature
distribution under transient conditions with actuation of high-pressure and accumulator injection systems.
These data would be suitable for an assessment of simulation codes for various types of mixing
phenomena during LOCAs with system responses under influences of ECCS coolant injection.
3.1.3

Description of Experiments

Test 1-1 (ST-NC-34 in JAEA) was conducted as a separate-effect experiment. The primary andsecondary pressures for the single-phase natural circulation condition were 15.5 and 6.7 MPa,
respectively, with the core power of 2% scaled nominal value and the full primary mass inventory. The
two-phase natural circulation was established at the primary and secondary pressures of about 6.7 MPa
with the same core power but reduced inventories. The ECCS injection rates were 0.3 and 1.0 kg/s for25

NEA/CSNI/R(2013)1
both the cold-legs A and B. The injection duration was 80 s. The injections with the rate of 1.8 kg/s wereadditionally performed for the cold-leg B under the two-phase condition.
Test 1-2 (SB-HL-17 in JAEA) was conducted simulating a 1% hot-leg small-break LOCA. The highpressure and accumulator injection systems were actuated at the specified operational set points. The flowrate of ECCS was a half of the scaled value since a single failure of safety system was assumed.
For Test series 1, 144 thermocouples were newly installed in cold legs and downcomer to obtain thethree-dimensional temperature distribution data. New video probes were also installed in the cold legs,
and the flow phenomena during ECCS coolant injection were observed visually.
3.1.4

Experiment Results

Two types of ECCS injection tests were performed in Test series 1: steady-state natural circulation
condition and transient LOCA condition. The temperature stratification in cold legs was seen widely inboth tests under both single- and two-phase natural circulation conditions. The degree of temperature
stratification was larger in the cold-leg A, since the shape and the location of the ECCS injection nozzles
are different between two loops in the LSTF. It was found that the cold water from the cold leg flowed
down in the downcomer along the inner wall of the pressure vessel under the natural circulation condition
with the downcomer liquid level higher than the cold leg (Fig. 3.1-2(a)(b)), while it flowed down along
the outer wall of the core barrel cover when the loop flow almost stopped (Fig. 3.1-3) or the downcomer
liquid level was lower than the cold leg (Fig. 3.1-2(b)). Besides the temperature variation and distribution
in the cold legs and downcomer, the injection of ECCS coolant with the variation of liquid level wasobserved visually. These experimental data would be useful for evaluation and validation of both the
CFD and BE codes and models: Test 1-1 is for simulation models including turbulent mixing andcondensation [3.1-10, 3.1-11], and Test 1-2 is for transient system analyses with PTS estimation under
actual LOCA conditions.
References[3.1-1]EricksonKirk, M., et al., 2006. Technical Basis for Revision of the Pressurized Thermal Shock(PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report. NUREG-1806.

[3.1-2]

Muehlbauer, P., 2003. Review of experimental database on mixing in primary loop and futureneeds. EVOL-ECORA-D03.

[3.1-3]

Scheuerer, M., 2002. Selection of PTS-Relevant Test Cases. EVOL-ECORA-D05a.

[3.1-4]

Egorov, Y., 2004. Validation of CFD codes with PTS-relevant test cases. EVOL-ECORA-D07.

[3.1-5]

Rohde, U., et al., 2007. Fluid mixing and flow distribution in a primary circuit of a nuclear
pressurized water reactor - Validation of CFD codes. Nucl. Eng. Des. 237, pp.1639-1655.

[3.1-6]

Bieder, U., et al., 2007. Simulation of mixing effects in a VVER-1000 reactor. Nucl. Eng. Des.
237, pp.1718-1728.

[3.1-7]

Toppila, T., 2008. CFD simulation of Fortum PTS experiment. Nucl. Eng. Des. 238, pp. 514521.

[3.1-8]

Coste, P., et al., 2008. Status of a two-phase CFD approach to the PTS issue. Proc. Experiments
and CFD Code Applications to Nuclear Reactor Safety (XCFD4NRS), Grenoble, France.

[3.1-9]Smith, B. L., et al., 2007. Assessment of CFD for Nuclear Reactor Safety Problems.
NEA/CSNI/R 13.

[3.1-10] Farkas, T., Tóth, I., 2008. FLUENT analysis of a ROSA cold leg stratification test. Proc.
Experiments and CFD Code Applications to Nuclear Reactor Safety (XCFD4NRS), Grenoble,
France.
[3.1-11] Watanabe, T., Nakamura, H., 2009. CFD analysis of temperature stratification experiment inOECD/NEA ROSA project. Proc. NURETH-13, Kanazawa, Japan.
26

NEA/CSNI/R(2013)1

Fig. 3.1-1

Fig. 3.1-2

Schematic diagram of temperature stratification phenomena in Test 1

Downcomer temperature in Test 1-1: (a) single phase case (b) two-phase case27

NEA/CSNI/R(2013)1

Fig. 3.1-3

Downcomer temperature in Test 1-2

3.2

Test 2 Unstable and destructive phenomena such as water hammer

3.2.1

Background

Condensation-induced water hammer (CIWH) may occur when highly subcooled water contacts directlywith steam, resulting possibly in failures of the reactor components and structures such as branch piping
due to the generation of large mechanical energy. Thus, data to describe characteristics of CIWH havebeen obtained through many studies. However, those studies focused mainly on CIWH under low
pressure conditions because of large shrinkage of steam volume. Low ambient pressure appears in thecourse of an accident management for the depressurization of the primary system to activate ECCS suchas accumulator and low-pressure injection (LPI) systems. CIWH may also occur in some passive safety
reactors, in which significant depressurization of the primary system is performed to enable gravity water
injection from the in-containment refuelling water storage tank (IRWST).
Though higher ambient pressure is supposed to tend to suppress the onset of CIWH due to higher steam
density, it makes subcooling of the injected ECCS water larger, which is preferable for the onset of
CIWH. That is, under high pressure conditions, there could be a competition between contradictory
effects of pressure and water subcooling. In order to evaluate the influence of CIWH on the integrity ofthe surrounding components and structures, it is crucial to clarify the onset criteria and the intensity of
CIWH under a wide range of pressure and subcooling conditions.
3.2.2

Objectives

The major objectives of Test 2 are to obtain detailed transient thermal-hydraulic data concerning CIWH
in a horizontal pipe through a series of separate-effect tests, and to clarify the onset criteria and theintensity of CIWH. In particular, the effect of system pressure up to approximately 7 MPa in combination
with water subcooling is studied. The obtained data could be suitable for the verification of computercodes and models for the prediction of CIWH.
3.2.3

Description of Experiments

Separate-effect tests (ST-WH-05, 06, 07, 08, 09, 10 and 11 in JAEA) were performed under well definedboundary conditions with a newly furnished horizontal pipe test section that was made of approximately
2.3 m-long stainless steel pipe with inner diameter of 66.9 mm [3.2-1]. The primary parameters weresystem pressure, water subcooling and liquid level (local void fraction) in the test section. One end of the28

NEA/CSNI/R(2013)1
test section was connected to the pressure vessel (PV) downcomer of LSTF. The other end was closed
using a sealing plate. Room temperature water in the LSTF refueling water storage tank (RWST) was
injected to the bottom of the test section near the closed end using the LSTF high pressure injection (HPI)
system. The injected water is discharged to the LSTF downcomer.
Fast response sensors were used for the measurement of CIWH pressure pulses generated and propagatedbelow water level in the test section. Many thermocouples were installed in the test section to estimate theliquid level, temperature stratification in water layer and the passage of water slugs. A gamma-ray
densitometer was also employed to measure the liquid level.
The test procedures are schematically illustrated in Fig. 3.2-1. The liquid level in the LSTF downcomer
was kept higher than the top surface of the test section during the initial state. Water at room temperature
was then supplied to the test section at a specified injection rate. Immediately after this operation, thewater in the PV of LSTF was gradually discharged from the PV bottom. During the test phase, the liquid
level was developed in the test section by the water discharge operation and saturated steam entered into
the test section from the PV. This situation may resemble a two-phase flow encountered in the ECCS
injection line after the injection rate decreased.
3.2.4

Experiment Results

Through a series of separate-effect tests, it was confirmed that CIWH occurred even at high ambientpressure of approximately 7 MPa. The following insights were obtained for the characteristics of CIWH
under a various combination of system pressure and water subcooling.
9

The averaged intensity of pressure pulses in each CIWH test increased with the system pressure
when it was less than 1 MPa, and decreased in the case of higher system pressure as shown in Fig.
3.2-2 (a).
¾

The maximum intensity was much larger than the average intensity. A pressure pulse
whose intensity was larger than 10 MPa formed even at the system pressure of 7 MPa as
plotted in Fig. 3.2-2 (b), although the averaged intensity was less than 1 MPa.

¾

When the reflection wave of a pressure pulse and the original wave were superimposed,
the intensity became almost twice as much as that of the original wave.

¾

The water injection rate seemed to affect the dependency of the pressure pulse intensity on
the system pressure. The effect of the water level at the outlet of the test section on theintensity was small.

¾

It was confirmed that the formation of the intermittent flow such as a slug flow was related
to the onset of CIWH.

¾

No thermal stratification formed in the liquid phase during the period while the CIWH wasinduced. The effective mixing of the condensate with the subcooled water may haveresulted in the large condensation rate.

¾

The steam mass flow rate estimated using mass and energy balances in the test section
monotonically increased with the system pressure up to 4 MPa. The mass flow rate
became almost insensitive to the system pressure when it was larger than 4 MPa.

¾

The estimated superficial steam velocity was related to the averaged intensity of pressure
pulses, probably because the steam velocity is proportional to the steam condensation rate.
This observation suggested that the large volumetric shrinkage induced by the significant
condensation rate produced the large pressure pulse.

References[3.2-1]

Japan Atomic Energy Agency, 2009. Final Data Report of OECD/NEA ROSA Project Test 2(Condensation-Induced Water Hammer Tests : ST-WH-05, 06, 07, 08, 09, 10 and 11 in JAEA).
Japan Atomic Energy Agency, Private Communication.
29

NEA/CSNI/R(2013)1

Downcomer liquid level is set
higher than test section.

Subcooled water is suppliedat specific flow rate.

Water is discharged at ~0.1
kg/s larger than supply rate.

(a) Initial state

Gamma-ray

System pressure
is controlled bycore power.

(b) Test phaseTwo-phase interface distribution becomes
insensitive to downcomer liquid level.
Free fallingdischarge isformed.

Liquid level in downcomer ismuch lower than test section.

(c) Termination stateFig. 3.2-1

Schematic diagram for test procedures

30

Average intensity of pressure pulse (MPa)

NEA/CSNI/R(2013)1

20
0.1kg/s
0.3kg/s
0.9kg/s

15

10

5

0

0

1

2

3

4

5

6

7

8

System pressure (MPa)Maximum intensity of pressure pulse (MPa)(a) Averaged intensity
40
0.1kg/s
0.3kg/s
0.9kg/s

30

20

10

0

0

1

2
3
4
5
6
System pressure (MPa)78(b) Maximum intensity
Fig. 3.2-2

Influence of system pressure on CIWH intensity

31

NEA/CSNI/R(2013)1

3.3

Test 3 Natural circulation under high core power conditions

The high reliability of control rods results in relatively low risk for anticipated transient without scram(ATWS) of PWR. Failure of scram, however, would lead to relatively high core power for a long timeand degradation in core cooling especially when coolant mass inventory decreases. The thermal-hydraulicdata on the natural circulation under high core power, however, is scarce, especially for the data takenfrom integral tests simulating accidents and anticipated transients where the primary coolant inventory
decreases. The data obtained through the ROSA/LSTF experiments would then be valuable to understandthe natural circulation phenomena during the high core power transient and for the validation of the safety
analysis codes.
Based on the motivation above and discussions among the project participants, the following two tests
were performed by using ROSA/LSTF. These tests may include natural circulation under high core power
conditions and several thermal-hydraulic phenomena associated with the natural circulation. The primary
inventory of the both tests decreases, though the mass decrease rate differs much depending on the sizeand location of the opening (break) on the primary pressure boundary.
Test 3-1

Small-break loss-of-coolant accident (LOCA) without scram

Test 3-2

Loss-of-feedwater (LOFW) transient without scram

Background, objectives, description and results are explained for each of the two tests in the following
sub-sections.
3.3.1

Test 3-1 Small-break LOCA without scram

3.3.1.1

Background

Thermal-hydraulic phenomena during small-break LOCA (SBLOCA) without scram may include highpower natural circulation with large flow rate for both the steam and liquid phases in hot legs. The liquidflow is highly accelerated to be supercritical (Froude number (Fr) > 1.0) by high velocity steam at theentrance of hot legs where the liquid velocity becomes higher than the speed of a surface long-wave that
depends on liquid depth [3.3-1, 3.3-2]. The hot leg liquid level then drops largely. The correct predictionof the transition into supercritical flow is important for reactor safety analysis, because the supercriticalflow in the hot leg may affect liquid carryover and liquid accumulation condition at around the inlet
plenum and U-tubes of steam generator (SG). Counter-current flow limiting (CCFL) may take place at
the SG inlet plenum and at the bottom of SG U-tubes that may hold a large amount of coolant, as shownin Fig. 3.3-1. The core cooling condition would then be much degraded when the natural circulation turns
into reflux cooling. Experimental data, however, have been scarcely obtained for such a SBLOCA
transient without scram where the thermal-hydraulic phenomena change as the coolant mass inventory
decreases with time.
3.3.1.2

Objectives

The objectives of Test 3-1 are to clarify thermal-hydraulic response adverse to core cooling under highcore power condition due to failure of reactor scram during the SBLOCA transient. Obtained test data on
the complicated coolant behaviors during natural circulation are suitable for the assessment ofpredictability of computer codes and models for system integral analyses.
3.3.1.3

Description of Experiment

Test 3-1 [3.3-3] was conducted under assumptions of total failure of high pressure injection (HPI) system
and an actuation of auxiliary feedwater as well as loss of off-site power concurrent with the scram signal.
Break size and core power curves were defined through discussions among the project participants basedon the LSTF pre-test analyses with RELAP5 code [3.3-4] and PWR SBLOCA analyses with JAEAdeveloped coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1[3.3-5] as well as LSTF preparatory experiments. Size of the break located at the side of cold leg in the32

NEA/CSNI/R(2013)1
loop without pressurizer (PZR) (loop-B) was 1% that corresponds to 2.8-inch break in the reference
PWR. The core power curve was pre-determined based on the PWR SBLOCA analyses with a detailed
core model, but the portion higher than 10 MW (14%) was cut off.
3.3.1.4

Experiment Results

Two-phase natural circulation took place under high core power, which caused SG relief valves (RVs)
kept opened. Flow in hot legs became supercritical due to high velocity steam and liquid flows during thetwo-phase natural circulation. The hot leg liquid level then became quite low and affected liquidcarryover into the SGs (Fig. 3.3-1). Liquid accumulation in the SG U-tube upflow-side and inlet plenum
happened during reflux condensation mode probably because of CCFL at the inlet of the U-tubes and thebottom of the inlet plenum due to high steam velocity (Fig. 3.3-1). The liquid accumulation caused asmall drop in the core liquid level. It, however, had no significant influences adverse to core cooling inthis test. Steam condensation on accumulator coolant in cold legs induced loop seal clearing in the loopB, but after the initiation of automatic core power decrease due to core temperature excursion during boiloff.
Consequently, the natural circulation under high core power contributed to maintain core cooling
effectively for relatively long time until core uncovery, while several phenomena specific to the natural
circulation under high core power such as supercritical flow and CCFL appeared in hot legs and SG inlet
plenum and U-tubes, respectively.
3.3.2

Test 3-2 Loss-of-feedwater transient without scram

3.3.2.1

Background

Thermal-hydraulic phenomena during LOFW transient without scram include high-power natural
circulation with liquid entrainment in hot leg at the surge line nozzle, and CCFL at the PZR bottom that
may hold a large amount of coolant in the PZR, as shown in Fig. 3.3-2. In the transient following LOFW,PZR power-operated relief valve (PORV) may continue cycle opening, resulting in loss of primarycoolant inventory. The core cooling condition would then be degraded especially after the natural
circulation turns into reflux cooling. A LOFW-induced ATWS experiment performed in the LOFT (Lossof Fluid Test) program revealed that the primary pressure is kept at about 17.2 MPa by cycle opening of
the PZR PORV and safety valve while the primary fluid temperature gradually increases [3.3-6].
Experimental data, however, have been scarcely obtained for such a LOFW transient without scram
where the thermal-hydraulic phenomena change as the coolant mass inventory gradually decreases withtime.
3.3.2.2

Objectives

The objectives of Test 3-2 are to clarify thermal-hydraulic phenomena adverse to core cooling under highcore power condition due to failure of reactor scram during the LOFW transient. Obtained test data on thecomplicated coolant behaviors during natural circulation are suitable for the assessment of predictabilityof computer codes and models for system integral analyses.
3.3.2.3

Description of Experiment

Test 3-2 [3.3-7] was conducted under assumptions of total failure of HPI system and an actuation of
auxiliary feedwater (AFW) as well as loss of off-site power concurrent with the scram signal. Onset
timing of the AFW and the core power curve were defined through discussions among the project
participants based on the LSTF pre-test analyses and PWR LOFW analyses with RELAP5 code. To avoid
injection of cool AFW coolant onto high-temperature structures that may cause significant deformation of
SG structures including U-tubes (bottom) due to thermal shock, AFW was initiated when the SG
secondary-side collapsed liquid level decreased to about 0.5 m. The core power curve was pre-determinedfrom volumetrically-scaled (1/48) power obtained from the PWR LOFW analyses with one-point neutron
kinetics model, and the portion higher than 10 MW (14%) was cut off. The core power curve and AFW
33

NEA/CSNI/R(2013)1
conditions are thus different from those in TMLB’ scenario (station blackout with no recovery of AFW)
[3.3-8].
3.3.2.4

Experiment ResultsCore power was maintained at about 7.4% for more than 7 hours until the automatic power reduction due
to high core temperature. AFW was actuated as planned, providing a continuous primary-to-secondary
heat removal. The primary pressure was maintained at around 16 MPa by cycle opening of PZR PORV
until the core uncovery started. The SG secondary-side pressure was kept at around 8 MPa by frequent
cycle opening of the relief valves till the end of the test. The primary coolant inventory decreased
gradually due to cycle opening of the PORV. Two-phase natural circulation started in very early stage ofthe transient. Liquid level decreased very slowly after the actuation of AFW but remained in the PZR
because of CCFL at the PZR bottom due to high steam discharging rate through the PORV (Fig. 3.3-2).
Significant level oscillation, in the form of a slow fill and dump, occurred in all the instrumented U-tubesof SGs with gradual decrease in the primary loop flow rate with some oscillation (Fig. 3.3-2). A watercolumn developed in the SG U-tubes rather randomly and with a lower frequency than that of the cycle
opening of the RVs and PORV. An intermittent increase in the fuel rod surface temperature preceded coreboil-off and major temperature excursion in the reflux condensation mode, even when the liquid level was
still in the upper plenum. The high core power may have caused departure from nucleate boiling (DNB).
The experiment was terminated when the primary and SG secondary-side pressures reached nearlyequilibrium condition with well-cooled core after the automatic core power decrease started.
Consequently, various and complicated thermal-hydraulic responses appeared associated with the natural
circulation under high core power, including inadequate core cooling in the reflux condensation mode.
References[3.3-1]

Gardner, G. C., 1988. Co-Current Flow of Air and Water from a Reservoir into a Short
Horizontal Pipe. Int. J. Multiphase Flow, 14[4], pp.375-388.

[3.3-2]

Chow, V. T., 1959. Open-Channel Hydraulics. McGraw-Hill Book, New York.

[3.3-3]

Japan Atomic Energy Agency, 2008. Final Data Report of ROSA/LSTF Test 3-1 (High Power
Natural Circulation Experiment SB-CL-38 in JAEA). Japan Atomic Energy Agency, Private
Communication.

[3.3-4]

Carlson, K. E., et al., 1990. RELAP5/MOD3 code manual (draft). NUREG/CR-5535, EGG2596.

[3.3-5]Zimin, V. G., Asaka, H., Anoda, Y., Enomoto, M., 1999. Verification of J-TRAC code with 3D
neutron kinetics model SKETCH-N for PWR rod ejection analysis. Proc. NURETH-9, San
Francisco, CA, USA.

[3.3-6]

Nalezny, C. L., 1983. Summary of Nuclear Regulatory Commission’s LOFT Program
Experiments. NUREG/CR-3214, EGG-2248.

[3.3-7]

Japan Atomic Energy Agency, 2009. Final Data Report of ROSA/LSTF Test 3-2 (High Power
Natural Circulation Experiment TR-LF-13 in JAEA). Japan Atomic Energy Agency, Private
Communication.

[3.3-8]USNRC, 1975. Reactor safety study - An assessment of risks in U. S. commercial nuclear
power plants. WASH-1400 (NUREG-075/14).

34

SG U-tube Upflow-side Liquid Level (m)NEA/CSNI/R(2013)1

SG
PZR
Liquidaccumulationin SG U-tube
due to CCFL

PV

15

10

5

0

0

300

SBLOCA

Hot Leg Liquid Level (m)

No
Scram

Supercriticalflow in hot leg
due to highsteam / liquid
velocities

High-power Natural
Circulation

600

900

1200

1500

1200

1500

Time (s)

0.25

0.2

0.15

0.1

0.05

0

0

300

600

900

Time (s)Fig. 3.3-1

Local phenomena during SBLOCA without scram in Test 3-1

14

PORV

SG

Pressurizer Liquid Level (m)SV

LiquidEntrainment

PZR
Liquidaccumulation
due to CCFL

Significant
leveloscillation in
SG U-tubes

12
10
8
6
4
2
0

PV

LiquidEntrainment

0

2000

4000

6000

12000
14000
16000
1041.41041.6104
8000 10000
1 104 1.2

Time (s)No
Scram
High-power
Natural Circulation

SG U-tube Upflow-side DP (kPa)80

Loss-ofFeedwater

70
60
50
40
30
20
10
14000
16000
18000
22000
8000 10000
1 104 12000
1.2 1041.4
1041.61041.8104 220000
104 2.2
1042.424000
104

Time (s)Fig. 3.3-2

Local phenomena during LOFW transient without scram in Test 3-2

35

NEA/CSNI/R(2013)1

3.4

Test 4 Natural circulation with superheated steam

3.4.1

Background

High pressure sequences of a PWR severe accident could involve natural circulation of highly
superheated steam in the primary cooling system, which acts as a carrier of thermal energy generated in
the degraded reactor core to the boundary structures. The natural circulation is characterized by theformation of a counter-current gas-gas flow in hot leg (HL), forward and backward flows in steam
generator (SG) U-tubes and the mixing of hotter and colder gases in inlet plenum (IP) of SG.
The failure of reactor coolant piping in the primary system is concerned to occur due to high temperaturedegradation of piping materials in such a severe circumstance. Depending upon the flow characteristics, it
is possible that the natural circulation allows high temperature superheated steam to enter SG U-tubes. In
case that the natural circulation of superheated steam results in so-called severe accident-induced steam
generator tube rupture (SGTR) so that containment bypass of radioactive materials occurs. Though theprobability in occurrence of the severe accident-induced SGTR is low, the assessment of risk from thisevent should be made with a reasonably realistic manner since it may cause a serious consequence to the
public and the environment [3.4-1, 3.4-2].
Thermo-fluid dynamics of such a natural circulation flow were investigated through past studies. A seriesof experiments was performed with a 1/7th scale SG model for Westinghouse PWRs [3.4-3]. An
experiment was carried out with LSTF before the OECD-ROSA Project [3.4-4]. In those experiments, thespatial distribution of gas temperature was mainly examined. However, no data was obtained for gas flow
velocity of the natural circulation. The quantification of the flow velocity is anticipated to be ofimportance to characterize the natural circulation and to aid the assessment of thermal loads onto theboundary structures.
3.4.2

Objectives

The present test series is composed of two quasi-steady separate-effect tests, Tests 4-1 and 4-2, to be
performed by using LSTF with newly installed instrumentation for detailed measurement of gas
temperature distribution and flow velocity especially at HLs and IP of SGs. The major objectives of thepresent test series are, considering the limitations of LSTF application to severe accident studies, toclarify important phenomena associated with the natural circulation of superheated steam underconditions as typical as possible and to provide a database for assessment of analytical tools including
CFD codes.
3.4.3

Description of Experiments

Argon was applied as a simulated gas of the primary system in Test 4-1 [3.4-5] at relatively low pressure(1.0 MPa) and superheated steam formed in the core of LSTF was used in Test 4-2 [3.4-6] at higher
pressure (4 MPa). The secondary side of SGs was filled with air and steam in Test 4-1 and Test 4-2,
respectively. The procedures for both tests to satisfy specified test conditions including gas temperaturesdescribed below were deeply examined in advance. Those were confirmed to be appropriate bypreliminary tests, resulting in successful performance of both tests as planned.
The dominant driving force for the natural circulation is density difference of gas, namely temperature
difference of gas, between the inlet and outlet sides of SG U-tubes. Consequently, the most importantcondition is the difference in gas temperature between the primary and secondary systems since thevariation of gas temperature in SG U-tubes is largely influenced by the primary to secondary heattransfer. This condition was defined through a series of pre-test analysis with FLUENT code (version 6.2)
[3.4-7], which indicated that the gas temperature at the inlet of HL should be higher than that in the SG
secondary side by at least 100 K. With this temperature difference as the representative for both tests,
Rayleigh number (Ra) was estimated not to be far from that of a PWR based on MELCOR analysis(~1015) when the gas temperature in the secondary system was assumed at 300 K for Test 4-1 and thesaturation temperature at 4 MPa for Test 4-2.
36

NEA/CSNI/R(2013)1
The newly installed instrumentation includes rakes of thermocouples (TCs) and pulsed-wire velocity
meters (see Appendix 1 for concept and roles of measurement). The rakes of TC and the pulsed-wirevelocity meters were positioned at pressure vessel (PV) and SG sides of each HL. Additional TC rakes
were placed in IP of each SG. The pulsed-wire velocity meter is vertically traversable, and it has astructure of resistance thermometer wires arranged in series in the flow direction. The most upstream wire
is electrically heated and others act as sensors to detect the arrival of the hot gas volume generated by the
upstream heater wire. The gas velocity is evaluated from time difference of detection and distance
between two wires. When natural thermal disturbances exist in a flow, all wires can be applied as sensors.
3.4.4

Experiment Results

The gas temperature in the vicinity of the HL inlet in the upper plenum of PV was higher than that in the
SG secondary side by approximately 300 K in Test 4-1 and 150 K in Test 4-2, which was sufficiently
larger than the specified target mentioned above.
The measurement using the pulsed-wire velocity meters without wire heating showed that the signals
from two wires were correlated with a certain width of time difference, which was considered tocorrespond to the convection of thermal disturbances in a flow. Summarizing all signals, it was identified
that forward flow from PV to SG and backward flow from SG to PV developed in the upper and thelower volumes of HL, respectively. There was also a tendency that the time difference was largest, that is,
the gas velocity was lowest at around the center of HL. Roughly estimated gas velocity wasapproximately 0.3 m/s at maximum for both tests, which was comparable to the results of pre-test
analysis with FLUENT code.
A clear temperature stratification: hotter gas in the upper volume and colder in the lower volume,developed in HL. In the upper volume of HL, a flat temperature profile formed at the vicinity of the inlet
and the temperature decreased toward SG with changing profile to an inclined one. On the other hand, aflat temperature profile formed in the lower volume near the IP of SG and the temperature increased
toward PV with similar change of profile. This feature in gas temperature profile could be related to thegrowth of thermal boundary layer and the mixing of gas under the counter-current gas-gas flow
conditions.
No large spatial distribution of gas temperature was found in the IP of SG. This result implied that gas in
the volume of the IP of SG was well mixed.
The gas temperature in SG U-tubes showed a decreasing profile from the inlet to the outlet in some of six
instrumented U-tubes. The inversed profile or uniform profile in the remaining instrumented U-tubes. The
temperature profile of gas indicated the co-existing of U-tubes with forward flow and those withbackward flow.
As a summary, the findings obtained from the present test series are conceptually illustrated in Fig. 3.4-1.
The natural circulation of superheated steam was confirmed to form in the primary loops of LSTF,
including counter-current gas-gas flow in HL, co-existing of forward and backward flow U-tubes and themixing of hotter and colder gases in IP of SG. There was no large qualitative difference between argon
and steam cases except steam velocity at the top volume of HL. It is anticipated that detailed andquantitative data of the present test series is of sufficient value for the assessment of the predictive
capability of analytical codes including CFD codes.
References[3.4-1]

Bayless, P. D., et al., 1995. Severe Accident Natural Circulation Studies at the INEL.
NUREG/CR-6285, INEL-94/0016.

[3.4-2]

SGTR Severe Accident Working Group, 1998. Risk Assessment of Severe Accident-Induced
Steam Generator Tube Rupture. NUREG-1570.

[3.4-3]

Westinghouse Electric Corporation, 1993. Natural Circulation Experiments for PWR HighPressure Accidents. Research Project 2177-05, Final Report.

[3.4-4]

Hidaka, A., et al., 1999. Evaluation of Steam Generator U-Tube Integrity During PWR Station
37

NEA/CSNI/R(2013)1
Blackout with Secondary System Depressurization. JAERI-Research 99-067, Japan Atomic
Energy Research Institute.
[3.4-5]

Japan Atomic Energy Agency, 2010. Final Data Report of Test 4-1 (Designated JAEA Internal
Run ID of ST-NC-39). Japan Atomic Energy Agency, Private Communication.

[3.4-6]

Japan Atomic Energy Agency, 2010. Final Data Report of Test 4-2 (Designated JAEA Internal
Run ID of ST-NC-40) . Japan Atomic Energy Agency, Private Communication.

[3.4-7]

Fluent Inc., 2005. FLUENT 6.2 User’s Guide.

38

NEA/CSNI/R(2013)1
15
PV Side_Row A
PV Side_Row B
PV Side_Row C
SG Side_Row A
SG Side_Row B
SG Side_Row C

50

Elevation from HL Center (m)Elevation from HL Center (mm)100

58000 seconds
0-50

0.805m from
PV Center

3.086m from
PV Center

-100
400

Test 4-1
HL-A

450
500
550
Fluid Temperature (K)Test 4-1
SG-A

Tube 1_Inlet
Tube 2_Inlet
Tube 3_Inlet
Tube 4_Inlet
Tube 5_Inlet
Tube 6_Inlet
Tube 1_Outlet
Tube 2_Outlet
Tube 3_Outlet
Tube 4_Outlet
Tube 5_Outlet
Tube 6_Outlet
SG_Secondary side
SG_inlet plenum
SG_outlet plenum

10558000 seconds

0
300

600

350
400
Fluid Temperature (K)¾ Temperature
stratification

450¾ Co-existing of U-tubes
with forward flow andthose with backwardflow¾ Gas mixing in IP of SG¾ Formation of counter-current gasgas flow in HL¾ No large qualitative difference
between argon and steam cases
except at the top volume of HL
100

Test 4-1 (Argon)
HL-B

Elevation from Hot Leg Center (mm)Elevation from Hot Leg Center (mm)

100

50

0

MedianPV SideSG Side

-50

-100
-0.8

-0.4

0
Velocity (m/s)

Fig. 3.4-1

0.4

Test 4-2 (Steam)
PV Side of HL-B

50

0

-100
-0.8

0.8

MedianSecond Trial
Fifth Trial

-50

-0.4

0
Velocity (m/s)

Major results obtained from Test 4

39

0.4

0.8

NEA/CSNI/R(2013)1

3.5

Test 5 Primary cooling through SG secondary depressurization

3.5.1

Background

Steam generator (SG) secondary-side depressurization through the relief valves (RVs) is one of major
accident management (AM) measures to cool and depressurize the primary loop via natural circulation
during a small-break loss-of-coolant accident (SBLOCA) of current PWR. The establishment ofprediction methods for natural circulation cooling at near-atmospheric pressure is essential to evaluate theeffectiveness of such SG secondary-side depressurization for long-term core cooling in the current PWR
and of gravity-driven injection in some future reactor design.
The primary cooling through the SG secondary-side depressurization would be influenced by such factors
as coolant mass inventory, break flow, water injection from emergency core cooling system (ECCS) andpresence of non-condensable gas. The LSTF experiments performed before OECD/NEA ROSA Project
revealed that non-uniform flow occurs among SG U-tubes during single-phase liquid and two-phasenatural circulation under influences of system pressure and coolant mass inventory [3.5-1~3.5-3].
The non-uniform flow is characterized by the coexistence of concurrent flow, contributing to heat transferfrom the primary loop to the SG secondary-side, and single-phase reverse flow or stagnant two-phasestratified flow, as shown in Figs. 3.5-1 and 3.5-2. The effective heat transfer area decreases when thenumber of stagnant U-tubes increases especially at low pressures. It is difficult to properly predict suchcomplex phenomena using computer codes. The experimental data is limited on the natural circulation
cooling at low pressures under well-defined boundary conditions. It is thus essential to conductsimulation experiments on SG secondary-side depressurization with detailed observation of local
phenomena such that the number of forward or reverse flow U-tubes may change with the coolant mass
inventory.
3.5.2

Objectives

The objectives of Test 5 are to clarify thermal-hydraulic responses, especially for the long-term corecooling at low pressures by natural circulation under influences of non-condensable gas, and effectivenessof the AM measure. Obtained test data on the non-uniform coolant behaviors under influences of noncondensable gas are suitable for the assessment of predictability of computer codes and models for
system integral analyses.
3.5.3

Description of Experiments

Tests 5-1 [3.5-4] and 5-2 [3.5-5] were conducted with no gas inflow and with gas inflow respectively
from accumulator tanks under the assumptions of an actuation of auxiliary feedwater and total failure ofhigh pressure injection (HPI) system as well as loss of off-site power concurrent with the scram signal.
The size of break at the side of cold leg in the loop without pressurizer (PZR) (loop-B) was 0.5% that
corresponds to 2-inch diameter break in the reference PWR. SG RVs were fully opened 10 minutes after
the safety injection signal. Such boundary conditions as break size, AM action and its onset timing were
defined through discussions among the project participants based on the LSTF pre-test analyses with
RELAP5 code [3.5-6], focusing on the observation of two-phase natural circulation and non-uniform flow
among the SG U-tubes. Further assumptions were made to conduct enhanced SG depressurization by
fully opening the safety valves when the primary pressure decreased to 2 MPa and no actuation of low
pressure injection system, both to well observe the natural circulation phenomena at low pressures.
A JAEA-developed gas measurement device including Zirconia sensor shown in Appendix 1 was newly
furnished in the LSTF to directly measure oxygen concentration in steam by introducing air as noncondensable gas for accumulator pressurization. Non-condensable gas in accumulator tanks enters cold
legs first and migrates to hot legs through pressure vessel (PV) downcomer and hot leg leak simulatinglines that connected the downcomer to the hot legs. Non-condensable gas that entered in SG U-tubes
should have gathered at the SG outlet plena after some U-tubes become empty of liquid. The gas
measurement device was thus connected to the SG outlet plena and to the PV upper-head one by one.
40

NEA/CSNI/R(2013)1

3.5.4

Experiment Results

In Test 5-1, the AFW flow rate was almost a half of the planned flow rate in the loop with PZR, causing
low liquid level in the SG secondary-side. Two-phase natural circulation continued in the loop-B until the
primary pressure decreased to far below 1 MPa with non-uniform flow among the SG U-tubes (Fig. 3.51). The SG U-tube liquid level recovered greatly after the initiation of the enhanced depressurization
especially in the loop with PZR where subcooled coolant column was formed in the U-tubes. Similarlarge level recovery happened in the two long U-tubes of the SG in loop-B. For these U-tubes, thecirculation terminated similar to those of the SG in the loop with PZR (loop-A). The U-tube liquid levelmonotonically decreased with the primary pressure after the initiation of the enhanced depressurization
but in different way in both SGs. In the loop-A, the natural circulation stopped before the initiation of
accumulator coolant injection because of adverse heat transfer around the top portion of the SG U-tubes
which emerged above the two-phase mixture level.
Test 5-2 was conducted with non-condensable gas inflow as a counterpart to Test 5-1. The SG
secondary-side liquid level changed similarly in two loops with the AFW flow rate as planned. SG
secondary-side depressurization was performed as planned but the depressurization rate decreased after
non-condensable gas started to enter primary loops and SG U-tubes, suggesting degradation in thecondensation heat transfer in the SG U-tubes. Collapsed liquid level decreased gradually in most of the
instrumented U-tubes after the initiation of non-condensable gas inflow as two-phase natural circulation
continued with non-condensable gas. However, such tubes with rapid coolant drain probably because ofaccumulation of non-condensable gas and with continuous single-phase liquid natural circulation coexisted in the SG in the loop-B (Fig. 3.5-2). Asymmetrical natural circulation between two loops also
appeared in Test 5-2 due probably to different number of forward flow SG U-tubes under the influencesof non-condensable gas. The natural circulation anyhow continued and contributed to maintain corecooling effectively for a long time. Air concentration in steam was about 40-50% at the SG outlet plena
after the U-tube downflow-side became empty of liquid. A bit larger amount of gas was found to exist in
the SG outlet plenum than the estimation based on the steam pressure and temperature. Oxygen
concentration in steam at the PV upper-head, however, indicates almost zero through the wholemeasurement, implying that no non-condensable gas enters the upper-head.
In summary, various non-uniform flow behaviors were observed in the SG U-tubes in both Test 5
experiments. In Test 5-2 with non-condensable gas ingress especially, three different types of U-tube
behavior coexisted in a SG depending probably on gas accumulation rate as well as asymmetric natural
circulation between two loops. The SG secondary depressurization as an AM action, however, wasconfirmed to be effective and no core heat-up took place through the whole transient in both experiments.
Direct measurement of non-condensable gas in a form of oxygen concentration in steam-air mixture by
the JAEA-developed gas measurement device was successful in Test 5-2 at the SG outlet plena where
non-condensable gas accumulates.

References[3.5-1]

Kukita, Y., et al., 1988. Nonuniform Steam Generator U-tube Flow Distribution during Natural
Circulation Tests in ROSA-IV Large Scale Test Facility. Nucl. Sci. Eng. 99, pp. 289-298.

[3.5-2]

Yonomoto, T., et al., 1997. PWR Small Break Loss-of-Coolant-Accident Experiment at ROSAV/LSTF with a Combination of Secondary-Side Depressurization and Gravity-Driven Safety
Injection. J. Nucl. Sci. Tech., 34[6], pp.571-581.

[3.5-3]

Yonomoto, T., Anoda, Y., 2000. Experimental tests and qualification of analytical methods toaddress thermohydraulic phenomena in advanced water cooled reactors. IAEA-TECDOC-1149,
pp.233-246.

[3.5-4]

Japan Atomic Energy Agency, 2008. Final Data Report of OECD/NEA ROSA Project Test 5-1(Primary Cooling through Steam Generator Secondary-side Depressurization Experiment SBCL-39 in JAEA). Japan Atomic Energy Agency, Private Communication.
41

NEA/CSNI/R(2013)1
Japan Atomic Energy Agency, 2009. Final Data Report of OECD/NEA ROSA Project Test 5-2(Primary Cooling through Steam Generator Secondary-side Depressurization Experiment SBCL-40 in JAEA). Japan Atomic Energy Agency, Private Communication.

[3.5-6]

Carlson, K. E., et al., 1990. RELAP5/MOD3 code manual (draft). NUREG/CR-5535, EGG2596.

SG U-tube Differential Pressure (kPa)[3.5-5]100

ECCS-InjectionSG

80

Non-uniform flow w/o
non-condensable gas
among SG U-tubes

60

Two-Phase
Concurrent
Condensing
Flow

40

20

0

0

3000

6000

9000

1.2 104

1.5 104

Time (s)SG U-tube Differential Pressure (kPa)

Fig. 3.5-1

Hot leg

Cold LegSG U-tube liquid level behaviors without non-condensable gas inflow in Test 5-1

100

ECCS-Injection

Non-condensable Gas Inflow

80

SG

2ø natural
circulation
including noncondensable gas

60

Two-Phase
Concurrent
Condensing
Flow

40

20

0

0

1000

2000

3000

4000

5000

6000

7000

Time (s)Fig. 3.5-2

8000

Hot leg

Cold LegSG U-tube liquid level behaviors with non-condensable gas inflow in Test 5-2

42

NEA/CSNI/R(2013)1

3.6

Test 6 Open subjects

The subjects for this test series are open to be defined through discussions among the project participants.
The candidates based on the discussions were:
1. Pressure vessel (PV) upper-head small-break loss-of-coolant accident (LOCA)
2. PV bottom small-break LOCA
3. Large-break LOCA focusing on the direct-contact condensation in the cold leg and its effects on
the containment pressure
4. Small-break LOCA focusing on accident management (AM) measures with symptom-oriented
operator actions
The fourth was then decided being coupled to the first and second subjects. In summary, three LSTF
experiments were defined for Test 6.
Background, objectives, description and results are explained for each of the three tests in the following
sub-sections.
3.6.1

Test 6-1 PV upper-head break LOCA with AM measure

3.6.1.1

Background

Vessel head wall thinning was found at the Davis Besse reactor in the US in 2002 and raised a safetyissue concerning vessel structural integrity [3.6-1]. Circumferential cracking of penetration nozzle for thecontrol rod drive mechanism (CRDM) may cause a small-break LOCA (SBLOCA) with break at the PV
upper-head, as shown in Fig. 3.6-1. Experimental data, however, had been scarcely obtained for such a
PV upper-head SBLOCA especially with some recovery action such as steam generator (SG) secondaryside depressurization as a symptom-oriented AM action [3.6-2, 3.6-3]. During the core boil-off process, acore temperature excursion may be detected by core exit thermocouples (CETs), which can initiate thesymptom-oriented AM operator action during the PV upper-head SBLOCA transient of interest,
employing a certain criterion defined by industries and regulatory bodies. Detailed study on the thermalhydraulic response associated with the PV upper-head break LOCA through the LSTF experiment wasthen requested from the Project participants in combination with the symptom-oriented AM action.
3.6.1.2

Objectives

The objectives of Test 6-1 are to clarify consequences of SBLOCA due to PV upper-head break with SG
depressurization as a symptom-oriented AM measure based on core exit temperature measured by CETs
during core boil-off, to confirm effectiveness of the symptom-oriented AM measure and to provide LSTF
experimental data suitable for the assessment of predictability of computer codes and models for reactor
safety analyses.
3.6.1.3

Description of Experiment

Test 6-1 [3.6-4] was conducted under assumptions of total failure of high pressure injection (HPI) system
and an actuation of auxiliary feedwater as well as loss of off-site power concurrent with the scram signal.
Break size, AM action and its onset timing were defined through discussions among the project
participants based on the LSTF pre-test analyses and PWR SBLOCA analyses with RELAP5 code [3.65]. The size of break at PV upper-head was equivalent to 1.9% cold leg break to simulate the ejection of
one whole CRDM penetration nozzle. The withdrawal of control rods was not simulated. SG secondaryside depressurization was started as a symptom-oriented AM operator action when the core exit
temperature indicated a certain superheating. Namely, relief valves (RVs) of both SGs were fully opened
immediately after the measured core exit temperature reached 623 K: a criterion for Japanese PWR.

43

NEA/CSNI/R(2013)1
3.6.1.4

Experiment Results

Liquid level in the upper-head was found to control break flow rate as coolant in the upper plenum
entered the upper-head through control rod guide tubes (CRGTs) until the penetration holes at the CRGT
bottom were exposed to steam in the upper plenum (Fig. 3.6-2). The cycle opening of the SG RVs
induced oscillation in the upper-head mixture level. Relatively large size break resulted in a fast primary
depressurization, especially after the break flow turned into single-phase vapor flow. The primarypressure became lower than the SG secondary-side pressure almost simultaneously with the core
uncovery, resulting in no reflux flow from the SG.
It took, however, more than 230 s to initiate the AM action after the core temperature excursion started
because of late and slow response of core exit temperature. The time delay in the detection of steam
superheating by CETs was influenced by such factors as core power distribution, three-dimensional
steam flow in the core due to steam flow towards break through CRGTs, low temperature steam flowfrom the core peripheral region and low-temperature metal structure around core exit [3.6-6]. Significant
temperature distribution resulted in the core and at the core exit as typically shown in Fig. 3.6-3.
The AM action was not effective in the early stage onto the intended primary depressurization to start
accumulator coolant injection, because the primary pressure was lower than the SG secondary-side
pressure at the onset of AM action, as shown in Fig. 3.6-4. Automatic core power decrease procedure toprotect LSTF core was then initiated due to high core cladding temperature during core boil-off. Loopseal clearing was induced when the accumulator injection started because of significant steam
condensation in both cold legs. The experiment was terminated when long-term core cooling was
confirmed by the actuation of low pressure injection (LPI) system.
In summary, the results of Test 6-1 indicated a possibility to have significantly high PCT in the PV
upper-head break LOCA, though some assumptions adverse to core cooling such as total failure of HPI
were made. Slow and late response of CETs during core boil-off was revealed, suggesting a necessity offurther study on the CET responses because CETs are utilized worldwide as an important indicator tostart AM operator action by detecting core temperature excursion during reactor accidents. A possibility
was suggested further that the initiated AM measure becomes ineffective for a certain time duration once
the initiation timing is so delayed, though the AM measure is expected to take effect shortly after the
initiation. A task group on the CET effectiveness in AM was thus formed in the Working Group of
Analysis and Management of Accident (WGAMA) of OECD/NEA/CSNI in 2008 to further discuss the
CET effectiveness.
3.6.2

Test 6-2 PV bottom break LOCA with AM measure

3.6.2.1

BackgroundA small amount of residue including boron around the circumference of two instrument-tube penetration
nozzles of PV lower-head was found at the South Texas Project Unit-1 in the US in 2003 and raised asafety issue concerning vessel structural integrity [3.6-7]. Ejection of two 1.5-inch outer-diameter
instrument-tubes would have caused a SBLOCA at the PV bottom, as shown in Fig. 3.6-1. Liquid-phasebreak is characterized by fast loss of coolant mass with low primary depressurization. The SG secondaryside depressurization as an AM operator action is important to depressurize the primary system tointroduce low-pressure ECCS especially when high-pressure ECCS is in failure. It is then important toconfirm the effectiveness of the AM measure under the influences of non-condensable gas fromaccumulator tanks. Experimental data, however, have been scarcely obtained for such a PV bottom
SBLOCA transient with the AM measure [3.6-8].
3.6.2.2

Objectives

The objectives of Test 6-2 are to clarify consequences of SBLOCA due to PV bottom break with anasymmetrical SG depressurization as an AM measure, to confirm the effectiveness of the AM measure,especially on the natural circulation core cooling and the primary depressurization under influences of
non-condensable gas, and to provide a database suitable for the assessment of predictability of computer
44

NEA/CSNI/R(2013)1
codes and models for system integral analyses.
3.6.2.3

Description of Experiment

Test 6-2 [3.6-9] simulated a SBLOCA with a break at PV bottom; equivalent to 0.1% cold leg break, tosimulate the ejection of one whole instrumentation nozzle. The test was conducted under assumptions of
total failure of HPI system, an actuation of auxiliary feedwater and loss of off-site power concurrent withthe scram signal. Break size, AM action and its onset timing were defined through discussions among the
project participants based on the LSTF pre-test analyses and PWR SBLOCA analyses by using
RELAP5/MOD3.2.1.2 code. Discussion was made especially on the SG depressurization method.
Consequently, asymmetrical depressurization only for the SG in the loop without pressurizer (PZR) was
employed, considering it may cause unbalanced natural circulation between loops and unbalanced
transport of non-condensable gas (nitrogen gas) from accumulator tanks (Fig. 3.6-5), which would be
unique and suitable for the validation of computer codes and models. The AM action was taken 30
minutes after the safety injection signal, considering an internationally common understanding on a grace
period to start operator action. The SG depressurization rate was controlled such that the primary
depressurization rate becomes to be 55 K/hr.
3.6.2.4

Experiment Results

Very small size break resulted in slow primary depressurization with certain influences of SG secondary
conditions. Large pressure difference appeared twice between primary and secondary-side of the SG in
the loop without PZR after the AM action and after accumulator tanks started to discharge noncondensable gas. The large pressure difference in the early timing was caused by adverse heat transferfrom non-depressurized SG to the primary loop where natural circulation stooped soon, while the natural
circulation continued in the loop without PZR after the AM action. When non-condensable gas started toenter primary loops, primary depressurization rate decreased and the large pressure difference resulted,
suggesting the degradation of condensation heat transfer in the SG U-tubes (Fig. 3.6-5).
The core exit temperatures on the side of the loop with PZR increased after the initiation of core uncovery
in the late phase of transient. However, they were kept at the saturation temperature on the side of theloop without PZR due to condensate fall-back from the SG though non-condensable gas weakened steam
condensation in the SG U-tubes. The LPI system was actuated when steam generation rate decreased in
the core, and most of the core was quenched. The fuel rod surface temperature, however, started toincrease again because of the termination of the LPI system when the primary pressure was increased dueto the recovery of steam generation in the core following the core quench. The experiment was terminatedwhen core cooling was confirmed by the second actuation of the LPI system after the initiation of
automatic core power decrease procedure due to high core temperature in the second core boil-off.
In summary, Test 6-2 provided a unique database including asymmetrical natural circulation behavior
because of significant effect of the asymmetric SG secondary-side depressurization under influences of
non-condensable gas. The asymmetric SG depressurization emphasized heat transfer manner between the
primary and SG secondary sides and the natural circulation under the influences of non-condensable gas
in all the three modes: the single-phase liquid, two-phase and reflux condensation modes. The natural
circulation, though asymmetrical, effectively cooled core as long as two-phase mixture covered the core.
The degradation of condensation heat transfer in the SG U-tubes, however, caused a significant delay inthe actuation of LPI as well as obstruction for the LPI actuation resuming.
3.6.3

Test 6-3 Direct-contact condensation in cold leg during large-break LOCA

3.6.3.1

Background

Steam direct-contact condensation may take place on emergency core cooling system (ECCS) water incold legs during reflood phase of a large-break loss-of-coolant accident (LBLOCA) of PWR, as shown in
Fig. 3.6-6. The steam condensation in the cold leg is ignored in the current safety analyses in Japan forthe conservative evaluation of containment pressure response by the overestimation of steam discharge
45

NEA/CSNI/R(2013)1
rate from the break. The steam condensation data are important for the realistic evaluation of thecontainment pressure, especially for the pressure peak that appears during reflood phase of LBLOCA.
Since the LBLOCA is design-basis accident of PWR, many of relevant databases have been obtained bysuch test facilities as LOFT [3.6-10], CCTF, SCTF and UPTF [3.6-11]. Experimental data, however, havebeen scarcely obtained for the steam condensation on ECCS water in cold legs because most of the
LBLOCA experiments are under some influences of non-condensable gas. The boundary conditions were
not well defined in several experiments either. The project participants, therefore, requested to prepare a
new database through the LSTF experiments to validate the associated direct steam condensation models
and the safety analysis codes.
3.6.3.2

Objectives

The objectives of Test 6-3 are to clarify the steam direct-contact condensation on ECCS coolant during
PWR LBLOCA that may impact on containment pressure through the LSTF separate-effect experimentthat properly simulates the flow conditions in the cold leg, and to obtain a database useful for thevalidation and development of safety analysis codes.
3.6.3.3

Description of Experiment

Test 6-3 [3.6-12] was conducted as a separate-effect experiment in a test section newly installed
downstream of the break unit that is horizontally attached to the LSTF cold leg, as will be described inAppendix 1. The test section mainly consists of a simulated cold leg pipe (102.3 mm i.d.), a simulated
ECCS injection nozzle, three viewers, a separator, separator downstream pipes for steam and water. The
three viewers were installed in the downstream of the ECCS injection point to observe visuallycondensing two-phase flow behaviors by high-speed video camera with back-lighting through a light
diffusion plate.
Experimental conditions on primary pressure, inlet steam velocity and superheating as well as mass fluxof ECCS water in the simulated cold leg pipe were defined through discussions among the project
participants based on PWR LBLOCA analyses by using RELAP5/MOD3.2.1.2 code. The primarypressure was near-atmospheric pressure, similar to the calculated condition in the PWR cold leg (698.5
mm i.d.) during reflood phase. Steam velocity at the inlet of the simulated cold leg pipe was controlled by
the LSTF core power. The inlet steam velocity was about 20, 30, 40 and 50 m/s, which mostly covers
conditions obtained by the PWR LBLOCA analysis. The inlet steam superheating was as high as possible
in the LSTF to best simulate the PWR conditions. Simulated ECCS water was injected at temperature of
310 K into the simulated cold leg pipe through a nozzle at angle of 45 degree downward, similar to that inreference PWR. Simulated ECCS water mass flux was ranging from about 12 to 207 kg/(m2s) so as topreserve the ratio of ECCS water injection flow rate to inlet steam flow rate, but less than that to cause
complete condensation of steam as shown in Fig. 3.6-6.
3.6.3.4

Experiment Results

Significant steam condensation took place in a short distance from the simulated ECCS injection point,
and the steam temperature in the test section decreased immediately after the initiation of the ECCS
injection. Fluid temperature distribution was non-uniform at about 50 mm downstream from the ECCS
injection point but became to be almost uniform in less than about 350 mm. Total steam condensation
rate estimated from the difference between the test section inlet and outlet steam flow rates was inproportion to the simulated ECCS water mass flux until the complete condensation of steam as shown in
Fig. 3.6-6. When the simulated ECCS water mass flux is about 148 kg/(m2s); equivalent to full HPI withsingle-failure LPI conditions, for instance, it was found that Inlet steam was completely condensed if inlet
steam mass flux is less than about 195 kg/(m2s).
Clear image data were obtained on droplet behaviors through the viewer at about 0.2 m downstream fromthe ECCS injection point, especially for annular mist flow, as shown in Fig. 3.6-7. The number ofdroplets decreased with increasing distance from the ECCS injection point, and very few droplets
appeared at the viewer at about 2.3 m downstream from the point.
46

NEA/CSNI/R(2013)1
In summary, a steam direct-contact condensation behavior was successfully observed in a new test
section that simulated typical flow conditions that may appear in PWR cold legs during reflood phase of
LBLOCA under well-defined boundary conditions. Steam condensation was so significant that it caused
liquid and steam temperature to become equal in a short distance from the simulated ECCS injection
point. The total steam condensation rate was in proportion to the simulated ECCS water mass flux until
the complete condensation of steam. Clear image data were obtained on droplet behaviors, especially forannular mist flow, which is important for model validation and development.

References[3.6-1]USNRC, 2002. Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure
Boundary Integrity. Bulletin 2002-01, OMB Control No.: 3150-0012.

[3.6-2]

Takeda, T., Asaka, H., Suzuki, M., Nakamura, H., 2005. Thermal-hydraulic Responses during
PWR Pressure Vessel Upper Head Small Break LOCA Based on LSTF Experiment andAnalysis. Proc. ICONE-13, Beijing, China.

[3.6-3]

Takeda, T., Asaka, H., Suzuki, M., Nakamura, H., 2007. RELAP5 Analysis of ROSA/LSTF
Vessel Upper Head Break LOCA Experiment. Proc. NURETH-12, Pittsburg, PA, USA.

[3.6-4]

Japan Atomic Energy Agency, 2006. Final Data Report of OECD/NEA ROSA Project Test 6-1(1.9% Pressure Vessel Upper-head Small Break LOCA Experiment SB-PV-09 in JAEA). Japan
Atomic Energy Agency, Private Communication.

[3.6-5]

Carlson, K. E., et al., 1990. RELAP5/MOD3 code manual (draft). NUREG/CR-5535, EGG2596.

[3.6-6]

Suzuki, M., Takeda. T., Nakamura, H., 2008. Performance of Core Exit Thermocouple for
PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at
OECD/NEA ROSA Project. Proc. ICONE-16, Orlando, FL, USA.

[3.6-7]USNRC, 2003. Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor
Coolant Pressure Boundary Integrity. Bulletin 2003-02, OMB Control No.: 3150-0012.

[3.6-8]

Suzuki, M., Takeda, T., Asaka, H., Nakamura, H., 2006. Effects of Secondary Depressurization
on Core Cooling in PWR Vessel Bottom Small Break LOCA Experiments with HPI Failure and
Gas Inflow. J. Nucl. Sci. Technol, 43(1), pp.55-64.

[3.6-9]Japan Atomic Energy Agency, 2007. Final Data Report of OECD/NEA ROSA Project Test 6-2(0.1% Pressure Vessel Bottom Small Break LOCA Experiment SB-PV-10 in JAEA). Japan
Atomic Energy Agency, Private Communication.

[3.6-10] Bayless, P. D., Divide, J. M., 1982. Experiment Data Report for LOFT Large Break Loss-ofCoolant Experiment L2-5. NUREG/CR-2826, EGG-2210.
[3.6-11] 2D/3D analysis group, 1994. 2D/3D Program Work Summary Report. JAERI 1335. Japan
Atomic Energy Research Institute.
[3.6-12] Japan Atomic Energy Agency, 2010. Final Data Report of ROSA/LSTF Test 6-3 (Steam
Condensation during Large-break LOCA of PWR LB-CL-05 and LB-CL-06 in JAEA). Japan
Atomic Energy Agency, Private Communication.

47

NEA/CSNI/R(2013)1

Coolant flow fromupper plenum toupper-headthrough CRGTs

SBLOCA

Upperhead(CRGT)
Upper
Plenum

SBLOCA

Fig. 3.6-1

Upper
Head

Locations of penetrations at top and bottom of PWR pressure vessel

Upper-head liquid level
influenced break flow rate.

Top BreakLine
Water Level(EL 8.3m)

14

Break Flow Rate

SupportColumn(×10)

Upper Core
Support Plate

Hot Leg

Cold Leg

Downcomer
Upper
PlenumSide Hole(×8/Tube)

Upper Core
Plate
Upper Core
Grid

HeaterElectrodeTop of
Active
Core

10
8
6
4
2
0

Upper-head Collapsed Liquid Level (m)SprayNozzle(×8)

Control Rod
Guide Tube
(×8)

Break Flow Rate (kg/s)

12

0

500

1000

1500

2000

2500

3000

Time (s)2.5Upper-headliquid level

2

1.5

1

0.5

0
0

500

1000

1500

2000

2500

3000

Time (s)Fig. 3.6-2

Coolant flow in pressure vessel upper region at onset of AM action in Test 6-1
48

NEA/CSNI/R(2013)1

Elevation Level at Core (m)

4.0
3.5

P.9 (Top)

3.0

P.8
P.7

2.5

P.6
2.0

3D steam flow
influenced steam
temperatures
along rod bundles
with or w/o CRGT.

P.5 (Middle)

1.5

P.4

1.0

P.3

0.5
-100

Bundle with CRGT
Bundle w/o CRGT
0

Fig. 3.6-3100
200
Steam Superheating (K)

300

400

Core steam temperatures 100 s after initiation of AM action in Test 6-1

R

)
a 16
P
M 14
(

14

12

e
r 10
u
s
s 8
e
r 6
P

Primary Press,
Second. Press.

12

ACC Flow Rate

10

AM Action
8
6
4

4

2

2

0

0

0

500

1000

1500

2000

2500

w
o
l
F
r
o
t
a
l
u
m
u
c
c
A

-2
3000

Time (s)Fig. 3.6-4

Primary and secondary pressures and accumulator flow rate in Test 6-1
49

NEA/CSNI/R(2013)1

SG

Asymmetrical SG
depressurization

PZR

Pressure

AM Action
Primary

12

Second. (loop with PZR)

10

Second. (loop w/o PZR)

8
6

Non-condensable Gas Inflow

4

Core Boil-off

2

PV

0

0

5000

1 104

Primary Loop Flow Rate (kg/s)

Unbalanced
naturalcirculation

1.5 104

2 104

2.5 104

Time (s)50

SBLOCA

Test Section Outlet Steam Flow Rate (kg/s)

14

SG
Pressure (MPa)

Degradationin condens.
heat transfer
due to noncondensable
gas accum.

Fig. 3.6-5

16

Primary loop
flow rate

40

30

Loop with PZRAM Action

Loop w/o PZR

20

10

0

0

5000

1 104

1.5 104

2 104

2.5 104

Time (s)Local phenomena during SBLOCA with asymmetrical depressurization in Test 6-2

0.25

SG

243 kg/m2s (Tsuper; 97-125K)

PZR

195 kg/m2s (Tsuper; 109-126K)

0.2

2

146 kg/m s (Tsuper; 95-104K)

Heat Transferfrom Second.
to Primaryside

97 kg/m2s (Tsuper; 70- 75K)
0.15

PV

0.1
Non-failure HPI &Single-failure LPI
Conditions
ECCS

0.05

0

0

50

100

150

200

250

ECCS Water Mass Flux (kg/(m 2s))

ECCS

Steam
Condensation

LBLOCA

Reflood Phase

Fig. 3.6-6

ECCS water mass flux versus test section outlet steam flow rate (inlet steam mass fluxesof 97, 146, 195 and 243 kg/(m2s)) in Test 6-3

50

NEA/CSNI/R(2013)1

Top
ow
Fl

Flow

Bottom(a) Overall view for flow observation(b) Center view for droplet viewing

Fig. 3.6-7

Image taken through viewer at about 0.2 m downstream from ECCS injection point (inlet
steam velocity of 20 m/s, ECCS water mass flux of 12 kg/(m2s)) in Test 6-3

51

NEA/CSNI/R(2013)1

52

NEA/CSNI/R(2013)1

4.

SUMMARY

OECD/NEA ROSA Project was carried out for 4 years from April 2005 among 14 countries to resolve
issues in thermal-hydraulics analyses relevant to LWR safety using the ROSA/LSTF facility of JAEA.
Major objectives of the ROSA Project are as follows:(1) Provide integral and separate-effect experimental database to validate code predictive capability andaccuracy of models for complex phenomena that may occur during DBE and beyond-DBE.
(2) Clarify the predictive capability of codes currently used for thermal-hydraulic safety analyses as well
as of advanced codes presently under development.
The LSTF (Large Scale Test Facility) of ROSA (Rig of Safety Assessment) Program in JAEA is a fullpressure and full-height two-loop integral test facility constructed in 1985 to investigate thermalhydraulic response during PWR accidents and abnormal transients. The LSTF simulates an 1100 MWe
four-loop Westinghouse (W)-type PWR: Tsuruga Unit-2 reactor of the JAPC (Japan Atomic Power
Company) with a volumetric scaling of 1/48.
Especially, 12 experiments for 6 subjects were planned and performed for the ROSA Project in either
integral-effect test (IET) or separate-effect test (SET) to widely cover thermal-hydraulic responses that
may encounter during reactor accidents and abnormal transients. Care was taken for each of the following
ROSA/LSTF experiments to include phenomena related to multi-dimensional mixing, stratification,parallel flows, oscillatory flows and/or non-condensable gas flows,
Test 1 Temperature stratification and coolant mixing during ECCS coolant injection
Two experiments (IET and SET) relevant for pressurized thermal shock (PTS) under reactor
typical pressure, temperature and flow conditions were performed.
The experiments revealed that thermal stratification condition in horizontal leg depends on ECCS
injection manner and that liquid level in horizontal leg and flow condition may control lowtemperature coolant flowing down along either pressure vessel inner wall side or core shroud
outer wall side.
Test 2 Unstable and destructive phenomena such as water hammer
A series of SETs was performed on onset conditions and consequences of water hammer in ahorizontal pipe under wide range of system pressure.
The experiments clarified that water hammer may take place up to 7 MPa and it generates greatest
pressure pulse peak at around 1 MPa.
Test 3 Natural circulation under high core power conditions
Two IETs were performed under assumption of failure of reactor scram to reveal various complex
thermal-hydraulic phenomena adverse to core cooling under high core power conditions anddifficult for code analyses, which are as follows;
Test 3-1 (1% cold leg break LOCA):Supercritical flow in hot legs under high velocity steam
flow, coolant accumulation in SG inlet plenum and U-tubes by CCFL.
Test 3-2 (LOFW transient): Complex reflux cooling phenomena in parallel SG U-tubes in nonuniform manner.
Test 4 Natural circulation with superheated steam
Two SETs were performed, respectively with argon or steam as test fluid, to clarify consequencesthat may cause containment bypass because of the rupture of SG U-tube(s) during a severe53

NEA/CSNI/R(2013)1
accident. The experimental conditions were carefully defined and confirmed by using CFD codeanalyses and LSTF preparatory experiments.
The experiments provided detailed low-velocity steam flow distribution within hot legs, steam
mixing in SG inlet plenum and parallel non-uniform flow in SG U-tubes.
Test 5 Primary cooling through SG secondary depressurization
Two IETs simulating 0.5% cold leg break LOCA with or without non-condensable gas ingress
were performed to clarify non-uniform parallel SG U-tube behavior during the SG
depressurization as an accident management (AM) action to mitigate reactor accidents byeffective primary cooling.
The experiments systematically observed various non-uniform flows in SG U-tubes underinfluences of non-condensable gas onto heat transfer during the natural circulation especially at
relatively low system pressures that is important for long-term core cooling.
Test 6 Open subjects
Three experiments (2 IETs and 1 SET) were performed to provide necessary information in
response to the requests from Project participants, which are as follows;
Test 6-1 (IET):
PV upper-head break small-break LOCA (SBLOCA) with SG secondary
depressurization as a symptom oriented AM measure based on core exit temperature. This test
raised a safety issue on the reliability of core exit temperature to detect core uncovery. A new task
group was formed under WGAMA (working group on analysis and management of accident) of
OECD/NEA/CSNI in 2008.
Test 6-2 (IET):
PV bottom break SBLOCA with asymmetrical SG secondary depressurization
as an AM measure. Asymmetric natural circulation core cooling was observed under influences of
non-condensable gas, which indicated a possibility of insufficient core cooling in a long-termtransient because of relatively high primary pressure.
Test 6-3 (SET):
Direct steam condensation in ECCS coolant in cold legs during reflood phaseof large-break LOCA (LBLOCA). Significant steam condensation was observed in a shortdistance from ECCS injection point between the subcooled ECCS coolant and highly superheatedsteam. Overall condensation rate was linearly proportional to the injected ECCS coolant. Image
data on liquid droplets under condensation were obtained.
For all the experiments, pre-test analyses were done to survey optimum test conditions to meet the test
objectives. The obtained results were distributed and discussed among the Project participants. The pretest analyses employed such computer codes as RELAP5/MOD3.2.1.2, SKETCH-INS/TRAC-PF1 and
FLUENT version 6.2 and 6.3.
To obtain necessary data during each of ROSA/LSTF experiments, JAEA-designed new
instrumentation were furnished especially to measure temperature distribution at fine spatial resolution
in cold and hot legs, downcomer, SG inlet plenum, non-condensable gas concentration in steam-gas
mixture at PV upper-head and SG outlet plenum, low-velocity gas flow distribution in hot legs and direct
visual observation of flows in all of the four horizontal legs.
New test sections were furnished further especially for each of Test 2 and Test 6-3 to observe local
phenomena in detail with much of the instrumentation specially designed for the purpose.
The obtained data has been readily distributed to and shared among the Project participants and utilized
for the post-test analyses for the validation, verification and development of the safety analysis codes andmodels as well as advanced codes such as CFD codes. The obtained technical information andexperiences, partly through a questionnaire for Test 6-1 post-test analysis, were shared among the
Project participants through the discussions during in PRG (Program Review Group) meetings and overe-mail communications.
In total, the ROSA Project, with the Project participants, successfully planned and performed
ROSA/LSTF experiments and shared the obtained data to conduct the verification, validation and54

NEA/CSNI/R(2013)1
development of the computer codes including future CFD codes.
This Final Integral Report is published in a form of a DVD with the Data Report for all the 12
ROSA/LSTF experiments with the measured data except for some specific data such as fast responsepressure sensor data from Test 2.

55

NEA/CSNI/R(2013)1

56

NEA/CSNI/R(2013)1

APPENDIX 1

FACILITY MODIFICATION AND NEW INSTRUMENTATION

Table A1-1 shows a list of new instrumentation and new test section prepared and utilized in the ROSA
Project.
New instrumentation installed for Test 1 (temperature stratification) consists of thermocouples in the cold
legs and downcomer to measure the detailed temperature distributions and one video probe in each ofcold legs to visually observe flow behaviors below the ECCS nozzle, as shown in Fig. A1-1. The
thermocouples at distance of 746.5 or 1633.5 mm from the PV center are arranged at 20 mm intervals
spot-welded on each of three stainless steel support plates fixed to the inner wall of the cold leg, as shownin Fig. A1-2. Each video probe mainly consists of a sensitive CCD camera, a bore scope and lightsources, as shown in Fig. A1-3. The windows for both bore scope and light are made of sapphire withhigh temperature and pressure resistant.
A new test section for Test 2 (water hammer) was connected to the LSTF downcomer to clarify the onset
criteria and the intensity of condensation-induced water hammer (CIWH) under well-defined boundary
conditions, as shown in Fig. A1-4. The test section is an about 2 m-long and 11 mm-thick stainless steel
horizontal pipe with inner-diameter of 66.9 mm. The test section end is closed by a sealing plate. Water is
injected into the test section through the water supply line at the test section bottom. The test section was
equipped with instruments for fluid and wall temperatures, pressure, pressure fluctuation, fluid density
and so on. Fast response pressure fluctuation sensors were used for the measurement of CIWH pressure
pulses.
New instrumentations have been installed for Test 4 (super-heated natural circulation). The
instrumentations consist of thermocouples in the hot legs and SG inlet plena to clarify the thermal
stratification and mixing of hot and cold gases and JAEA-developed pulsed-wire velocity meters in the
hot legs to measure the velocity distribution in the vertical direction, as shown in Fig. A1-5. The
vertically-traversable pulsed-wire velocity meter has resistance thermometer wires made of platinumrhodium arranged in series in the flow direction, as shown in Fig. A1-6. A wire at the most upstream iselectrically heated while others act as sensors to detect the arrival of the hot gas volume generated by the
upstream heater wire. The gas velocity is evaluated from time difference of detection and distance
between the two wires.
A JAEA-developed gas measurement device was used for Test 5-2 (SG depressurization with noncondensable gas inflow) to directly measure oxygen concentration in steam at the SG outlet plena and PV
upper-head by introducing air as non-condensable gas, as shown in Fig. A1-7. The gas measurement
device includes a Zirconia sensor, a separator, a flow meter and piping. The measurement uncertainty of
Zirconia sensor is ±1.5% of full-scale (25%) for oxygen concentration in steam. Conditions of fluid
sampled at the SG outlet plena is confirmed by draining it from the separator bottom while the glove
valve and valve in the drain line are opened. Small-volumetric flow rate of steam/gas mixture is measured
by using the flow meter in the downstream of Zirconia sensor vessel.
A new test section for Test 6-3 (steam condensation) was furnished in the downstream of the break unithorizontally attached to the LSTF cold leg to simulate a steam direct-contact condensation on ECCS
water in cold legs during reflood phase of PWR LBLOCA under well-defined boundary conditions, as
shown in Fig. A1-8. The test section mainly consists of a simulated cold leg pipe, a simulated ECCS
injection nozzle, three viewers, a separator, separator downstream pipes for steam and water. The
simulated cold leg pipe is an about 4.5 m-long and 6 mm-thick stainless steel horizontal pipe with innerdiameter of 102.3 mm. The three viewers were installed in the downstream of the ECCS injection point toobserve condensing two-phase flow behaviors by using high-speed video camera. The separator separates
57

NEA/CSNI/R(2013)1
steam-water two-phase flow into single-phase steam and water flows at the downstream end of the test
section. The test section was equipped with instruments for flow rate, fluid and wall temperatures,pressure, fluid density and so on. Two vortex-type flow meters were mounted at the inlet of the simulated
cold leg pipe and at the separator downstream steam pipe to measure volumetric flow rates of steam, fromwhich velocities and mass flow rates of steam were estimated.

Table A1-1 List of new instrumentation and new test section
Test No.

Subjects

Contents
Thermocouples in cold legs and downcomer
Video probes in cold legs
New test section with new instrumentation that horizontally
attached to LSTF downcomer
Thermocouples in hot legs and SG inlet plena
JAEA-developed pulsed-wire velocity meters in hot legs

Test 1

Temperature
stratification

Test 2

Water hammer

Test 4

Super-heated natural
circulation

Test 5-2

SG depressurization
with non-condensable
gas inflow

JAEA-developed gas measurement device connected to SG
outlet plena and to PV upper-head

Test 6-3

Steam condensation

New test section with new instrumentation furnished
downstream of break unit horizontally attached to LSTF cold leg

58

NEA/CSNI/R(2013)1

Loop B

Video Probe

CL-B

CL-A

CL and DC (TCs)

Loop A
Fig. A1-1

Fig. A1-2

Locations of new instrumentation for Test 1

Locations of new thermocouples in cold leg in loop A for Test 1
59

NEA/CSNI/R(2013)1

Bore Scope32

Cold Leg
207

Bore scope window

36

Light window

Fig. A1-3

Schematic view of new video probe in cold leg for Test 1

6 T/Cs to measure fluid temperature at pipe upper region

Pitot tube-> Pressure transducer
LSTF
downcomer
nozzleI.D. 67 mm
8 T/Cs in a plane
Water
supplyline
Length ~2 m
8 pressure fluctuation meters
Fig. A1-4

Gamma
densitometer

Geometry of new test section for Test 2
60

NEA/CSNI/R(2013)1

Loop B

SG inlet plenum (TCs)PV side of HL(TCs and pulsed-wire velocity meter)

SG side of HL(TCs and pulsed-wire velocity meter)

Loop A

Fig. A1-5

Locations of new instrumentation for Test 4

Temperature stratification

Thermocouple (TC) rakes
15 TCs
21 TCs

21 TCsTC rakes

Mixing of hot
and cold gases

Flow
Formation of
hot gas volume

L1 L2

Heater wireSensor wire
(Pulsed heating) (Resistance)

SignalPulsed-wire velocity meter Velocity distribution
in vertical direction(Traversable)
Pulse

V1=L1/t1
V2=L2/t2
V12=(L1+L2)/(t1+t2)
t1
t2

Time

Detection of convected hot gas volume from upstream

Fig. A-1-6

Conceptual scheme for newly installed instrumentation in Test 4
61

NEA/CSNI/R(2013)1
SG
VesselAir forPress.
Control

Zirconia
SensorPress.
Control
Valve

Separator
N-8
Nozzle

Glove
Valve

HeaterSG
OutletPlenumZirconia
Sensor
VesselMax. 2kPa
2MPa [gauge]
15.5 Max.
MPa 2MPa

Press.
Transmitter

Drain

Flow

P

F Meter

HeaterExhaustPress.
Control
Valve

Separator

Zirconia
Sensor
VesselGas Flow
Flow
Meter

Fig. A1-7

Overview of gas measurement device connected to SG outlet plenum in Test 5-2

LSTF

PV

F
T

Steam Generator(Loop w/o PZR)

Cold Leg (CL)
Break
Unit

Viewer
L:0.2mECCS Water
P
Break
Orifice

Break
Orifice
Break
Valve

Break
Valve

P: Press., T: Fluid Temp., TW: Wall Temp., L: Level
F: Flow Rate, DE: Fluid Density, PT: Differential Press.

P T

Highspeed
Video

F
T T

45deg

PT

TW

Simulated CL Pipe（i.d.=102.3mm）

T

T

DE

TW

3rd

TW

F
TW

2.3m

T

T
P

F P

Steam Tank

Steam
TW

T

PT

Heater（Line for System Depressurization）

T

DE

2nd

0.2m 0.9m

F

Highspeed
Video

Highspeed
Video

1st

i.d.= 42.6mm

About 1.9m

T

Atmosphere

P L

P TW

About 2.6m

P F
Water
Storage Tank
F
Separator

Fig. A1-8 Schematic view of new test section in downstream of cold leg break unit in Test 6-3

62

NEA/CSNI/R(2013)1

APPENDIX 2

POST-TEST ANALYSES

One of the key objectives of the ROSA Project is to clarify the predictability of computer codes currently
used for thermal-hydraulic safety analyses as well as of advanced codes under development. Many pretest and post-test analyses were performed for the ROSA Project by both the project participants and
JAEA as operating agent to define the boundary conditions of the LSTF experiments and to understandthe test results. Appendix 2 summarizes the major findings obtained mainly through the post-test
analyses.

A2-1 Pre-test Analysis and Post-test Analysis(1) Pre-test analysis
The pre-test analyses are performed to provide information on the expected test results by changing
values of several major parameters to define the test boundary conditions to achieve the objectives of allof the twelve LSTF experiments on the six subjects for the ROSA Project. The obtained results werediscussed at the program review group (PRG) meetings and via e-mails among the project participants.
The pre-test analysis is a kind of blind calculation and provides a good chance to verify the code
predictability. In the ROSA Project, only JAEA performed this analysis as the operating agent partly
because it was considered that the analysis should be completed in relatively short time for the test
preparation. JAEA then mainly used rather old version of the RELAP5 code (MOD 3.2.1.2) for the
LOCA test cases because of long-term experience to use it including several code modifications. Other
than the RELAP5 code, the SKETCH-INS/TRAC-PF1 code developed in JAEA for the threedimensional thermal-hydraulics neutronics coupling in the core was used to define the core powertransient in Test 3-1 (high-power natural circulation) with an assumption of failure of core scram during a
LOCA. FLUENT code; computational fluid dynamics (CFD) code to analyze three-dimensional fluid
flows, was used further to define conditions that cause natural circulation of superheated steam in Test 4.
In the course of the ROSA Project, however, many disagreements were observed between the pre-test
calculation and LSTF test results, though the options and models were selected based on the previous
experiences. One of such disagreements appeared for Test 6-1 (PV upper-head break LOCA) where the
RELAP5 code expected that the single-phase steam discharge from the break, thus, fast primary
depressurization, starts much later than observed in the LSTF test. This happened because of some errors
in the break flow prediction, particularly in the definition of discharge coefficient in conjunction with the
break upstream flow analysis as well as some trial and error effort in the guide tube flow representation.
The faster core uncovery and temperature excursion appeared in the LSTF experiment than expected.
Then, the late and slow response of core exit thermocouples (CETs) to detect steam superheating resulted
late initiation of SG secondary depressurization as an accident management measure. This experience
associated with the pre-test analyses has suggested us the necessity of thorough verification andvalidation (V&V) of computer codes and careful selection and definition of input parameters, options andmodels.
(2) Post-test analysis
The post-test analyses are performed to validate computer codes and models against the obtained (known)
test data, which were performed by many of the project participants. The best estimate (BE) computercodes such as APROS, ATHLET, CATHARE, MARS, TRACE were used for integral effect tests (IETs)
such as LOCA and operational transient simulations. CFD code FLUENT was mainly used for the
separate-effect tests (SETs) such as thermal stratification. Some of BE codes, however, were used forsome SETs such as water hammer. The calculated results were presented at the PRG meetings to share63

NEA/CSNI/R(2013)1
the results among the participants through the discussions on the results. One of the major discussions
was done for Test 6-1 where a certain discrepancy appeared in the pre-test analysis as above. An effort
was done in the ROSA Project further to make a mutual comparison on the analysis methods andencountered problems to extract lessons through the questionnaire on the analysis methods for Test 6-1.

A2-2 Post-test Analysis Results
This section summarizes general points found in the post-test analyses results presented at PRGs by theparticipants. Major outcomes of each result are summarized in Table A2-2-1. The presentation materials
for the PRGs are attached to this report.
The code input for TRAC-PF1 and RELAP5/MOD3.2 code for the ROSA/LSTF with their noding
schematics was given to the Project Participants from JAEA to assist to initiate their code calculations.
Since the Project participants mostly utilize their own computer code, these inputs were utilized as
reference to prepare their own code input. Some Participants utilized their own input that was originallyused for the LSTF at ISP-26 in 1992 with many modifications according to the improvements in thecomputer code afterwards.
A2-2-1 Subjects Found for Models in BE Codes and Their Utilization(1) Break Flow
For all the ROSA Project experiments, orifice was used to simulate the break. Most of the code
calculations were then performed by adjusting the break flow rate by comparing the calculated resultswith the measured data. The adjustment was done for each of single-phase liquid, two-phase and singlephase steam (high quality) discharge flows, depending on the optional input structures for each of codes
and, sometimes, by modifying the codes in the source program. The major parameters were dischargecoefficients and break area in a form of break size and upstream flow conditions such as liquid level inthe upstream pipe (e.g., horizontal leg) and liquid entrainment from a volume to branch pipe that has thebreak at its end. The primary pressure was calculated well once the break flow rate was well adjusted.
This process was necessary to validate the predictive capabilities of each computer code on the points
other than the break flow rate. Further discussions may be necessary on the validation of break flow
model(s) and their proper utilization with a certain discharge coefficient and relevant parameters toascertain the appropriateness of the break flow model utilization in the safety analysis as far as theseparameters are adjusted in each analysis.
(2) Steam Generator U-tubes
Steam generator (SG) of the LSTF has 141 U-tubes with 9 different lengths. Six tubes of them are
instrumented. It is important to simulate well the SG U-tube behavior to simulate well both the mass
inventory and primary-to-secondary heat transfer. Therefore, multiple U-tubes with different length wereused in the code calculations to simulate thermal-hydraulic response in parallel channels well, consideringthat the measured flows indicated wide variation among the U-tubes in many LSTF experiments. The
calculated results well simulated the general trend. However, they did not clearly show the variety amongU-tubes.
Forward-flow tubes, reverse-flow tubes and even stalled tubes co-existed in many cases for the LSTF
experiment. It was difficult to simulate such responses even when 9 U-tubes were used to simultaneously
simulate different size of U-tubes. The natural circulation flow rate was overestimated. The origin(s) of
the difficulties have not been well identified. A single volume representation of SG inlet plenum with
branching to U-tubes may be one of the origins. However, the difference in the tube length may havesmall impact. Since the mixing and/or stratification and phase separation and distribution in the LSTF SG
inlet plenum may differ from those in the reference reactor, care should be taken well when the reactor
cases are dealt with.
(3) Interphase dragProper prediction of interphase drag between gas and liquid phases is indispensable to correctly simulate,for example, mixture level in the core, upper plenum and upper head, and liquid level in horizontal pipes
and liquid accumulation around SG inlet plenum. Mixture level in the core was mostly well predictedonce the mass inventory and distribution of coolant is well predicted. In the hot leg, the loop mass flow
64

NEA/CSNI/R(2013)1
rate and interfacial drag controls the liquid level, thus the condition of criticality for horizontal flow basedon Froude (Fr) number. In Test 3-1 (high-power natural circulation), a few codes failed to properly
simulate the liquid level and flow criticality in the hot legs, probably because of some error in theinterphase drag evaluation, considering that the steam generation rate in the core is well represented bygiving the measured core power history.
(4) Steam condensation
When the ECCS, namely accumulators, injects coolant into cold legs, significant steam condensation
takes place on the subcooled water, temporarily causing fluctuation in the primary pressure, liquid level
drop in the core, loop seal clearing as well as the enhancement of coolant injection from accumulators.
Some difficulties were found to simulate well such responses because the phenomena depend on local
steam-liquid interfacial area and temperature difference between steam and subcooled liquid in multidimensional mixing process. The computer codes simulated the overall phenomena rather well, but alarge uncertainty sometimes appeared in the pressure response, etc.
(5) Influences of non-condensable gas
Steam condensation is significantly deteriorated when some non-condensable gas is mixed in the steam.
The non-condensable gas (Nitrogen gas) enters the primary system after the completion of coolant
injection from accumulators. The influences of non-condensable gas are then important for condensation
heat transfer in the SG U-tubes especially after the SG secondary-side depressurization as one of accident
management measures. Nitrogen gas entered cold legs moves to and accumulates in SG U-tubes and SG
outlet plenum via steam flow towards SGs. Computer codes were found to well calculate the transfer ofgas as a mixture with steam and the accumulation into SG U-tubes. Most of computer codes have somedifficulties in the rigorous prediction of the heat removal through the condensation heat transfer underinfluences of non-condensable gas. Influences of gas ingress into the decreased-number of forward flow
U-tubes are significant because flow reversal and stagnation take place in a large portion of U-tubes.
As a special case, non-condensable gas entered the primary loop under single-phase liquid natural
circulation in Test 6-2 (PV bottom break LOCA) causing temporal and fluctuational increase in the
primary pressure. However, computer codes failed to correctly simulate the phenomena.
(6) Bypass Flow in Pressure VesselThis subject is not related to physical models, but to the proper prediction of the pressure balance acrossthe core shroud in the LSTF. It was pointed-out in several post-test analyses that some adjustment isnecessary for the bypass flow between the upper plenum and downcomer and between hot leg anddowncomer. Since the pressure balance between these two components directly affects the core uncovery
behavior, care should be taken.
A2-2-2 Findings for CFD Codes
Test 1-1 was especially dedicated to the thermal-stratification under ECCS injection into cold legs into
single-phase and two-phase natural circulation and almost stagnant conditions. The thermal-stratification
appeared in cold legs and cylindrical downcomer, but in some different way in two cold legs with two
different types of ECCS injection nozzle. FLUENT code results were reported from three countries withvarious types of turbulent models, basically with k-e model, mostly for steady flow results.
For the single-phase liquid flows, flow near the downcomer, thus downstream side in the cold leg, wasrather well simulated, however, a certain discrepancy persisted for flow near the ECCS nozzle. The LSTF
cold legs have a bend at the downstream or upstream of the ECCS nozzle in both loops, which generates
secondary flow with some mixing effect. Three support columns (o.d. 6 mm) to hold thermocouples at
two locations in the straight pipe portion of each cold leg should have caused further mixing effects.
Anyhow, CFD codes failed to well simulate flows near pipe side-wall (both sides) under influences ofsuch secondary flows and flow mixing.
For the two-phase flows, operating agent reported that a new model on steam condensation at steamwater interface may improve the vertical temperature distribution in liquid flow, thus the temperature
stratification and radial temperature distribution in the downcomer too.

65

NEA/CSNI/R(2013)1

Table A2-2-1
ROSA
Test

OECD/NEA ROSA Project Code Evaluation by Post-test Analyses
Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model

Code

PRG2 2005 November at JAEA, Tokai
1-1

Preparation for Temperature
Stratification Experiment

JAEA
Japan

T. Watanabe

9

FLUENT

Single-phase temperature distribution in cold leg was calculated well using ke model.
9
Noding effect was about 4 K in cold leg for 0.13-0.21 million-node
simulations.
9
Secondary flow due to the curvature was predicted in the cold leg.
9Temperature stratification was seen in most part of the cold leg.
9Temperature stratification was not significant in the downcomer during
single-phase natural circulation.

(pre-test analysis)

PRG3 2006 May at UPV/CSN, Valencia
91-1

Preparation for Temperature
Stratification Experiment(pre-test analysis)

JAEA
Japan

T. Watanabe

FLUENT

Noding effects of pump outlet and downcomer on the cold-leg temperature
distribution discussed.
9
Flow velocity increased at the downcomer inlet.
9
New condensation model was prepared and shown to be necessary for correct
thermal stratification prediction in two-phase case.

6-1

First Results of a Post-Test
Calculation of LSTF Test 61 (SB-PV-09) with the Code
ATHLET(post-test analysis)

GRS
Germany

H. Glaeser,
H. Austregesilo

ATHLET

9

Good accuracy in the break flow rate and the influence of the upper head(UH) liquid level on break flow, but with a little delay in core boil-off start9
Calculated start of core uncovery is strongly dependent on simulation of
coolant flow bypass between downcomer and UH/UP
9
Mixture level tracking model applied to UH
9
Break flow rates calculated with ATHLET specific 1D finite difference
discharge model: discharge coefficient (Cd) = 1 for subcooled and saturated
flow, Cd = 0.9 for steam flow
9Adequate simulation of flow resistances between UP and UH, not only within
CRGTs but also for the downflow between HL nozzles and core upper plate, isessential for a correct prediction of the break flow rates!

66

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)

Code
TRACE

Major results andsuggestions to improve and develop code and/or model
9

6-1

Rosa LSTF upper-head
SBLOCA: Conversion from
TRAC-P to TRACE model.
Steady state comparison(post-test analysis)

UPV
Spain

V. G. Llorence,
S. G. Bermell,
G. V. Martín

TRACE input preparation converting from that for TRAC-P, which lead togood agreement with reference results.
9
Old stgen TRAC-P component seems more flexible than htstr component inorder to model heat transfer.
9This steady state point has been accepted as an initial point of the transient.

6-1
&
6-2

Status Report on theactivities being performed
by AVN in the context of
the OECD ROSA project(post-test analysis)

AVN
Belgium

A.Bucalossi

CATHARE2 9
9

6-1

PV
Upper-head
Break
LOCA
(post-test analysis)

JAEA
Japan

T. Takeda

RELAP5/
MOD3.2

Discharge coefficient 0.61 for two-phase break flow
Nodalization corrected considering coolant paths to CRGT (control rod guidetube) bottom in post-test analysis
9Overpredict break flow rate due to underprediction of break-upstream void
fraction especially during two-phase flow discharge period, suggesting that code
has remaining problems in proper prediction of liquid level behaviors in vessel
upper-head

6-2

PV Bottom Break LOCA
(post-test analysis)

JAEA
Japan

T. Takeda

RELAP5/
MOD3.29

Input deck preparation based on literature.
Vessel Coarse Meshing :¾ 1 mesh (VOLUME) for Upper Head, Upper Plenum, and Lower Plenum.
¾ Fine meshing for Core, Downcomer and Guide tubes
9
Calculated general trends seem reasonable for “Experimental Time-dependentinterventions”-type transients except incorrect vessel level behavior
9
More detailed input deck is necessary to correctly qualify the code on“Signal-dependent interventions”-type transients
9
9

Primary pressure improved by adjusting loop-B SG secondary-side
temperature to meet test data after initiation of AM action
9
Primary loop flow rate improved by increasing interfacial drag by 10 times at
hot leg bend, inlet plenum and upflow-side of SG in loop-B

PRG4 2006 November at JAEA Tokai
6-1

Preliminary post test
calculation of OECD/ROSA
test 6.1 with APROS(post-test analysis)

VTT
Finland

P. Inkinen,
I. Karppinen,
J. Poikolainen

9
9
9
9
9

APROS

67

Very good accuracy for overall trend including core boil-off
New input deck based on RELAP5 model for ISP-26 and of JAEA-provided
Heat loss and upper head initial temperature conditions adjusted in detail
Break model assumptions: break-line modeling, discharge coefficient of 0.86
CRGTs have to be modeled with 2 channels. Center tubes and peripheral
tubes separately.

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model

Code
9
9

6-1

Results of Post-Test
Calculation of LSTF Test 61 (SB-PV-09) with the Code
ATHLET(post-test analysis)

GRS
Germany

H. Austregesilo,
H. Glaeser

ATHLET

More detailed consideration on the previous calculated results
Influence of LSTF downcomer to upper plenum is discussed throughparameter calculation.
9
Possibility of heat loss from pressure vessel suggested as a reason for
temporary large mass in downcomer
9
Video presentation on mass inventory distribution transient

6-1

PV Upper-head Break
LOCA(post-test analysis)

KFKI
Hungary

V. Kerekes
G. ÉZSÖLRELAP5/
MOD3.3

9

Noding for break unit piping with Henry-Fauske model discharge coefficient=0.85 based on parameter analyses against measured break flow rate. However,
100 s sooner in the calculation for the break flow turned from two-phase flow tosingle-phase vapor
9
The processes during single phase steam break flow are mainly influenced bycorrect prediction of the core level, which is strongly correlated with themodeling of the UP-CRGT connection.
9
A little delay in core boil-off start and slower system depressurization

6-1

Case 6.1. Preliminarytransient results(post-test analysis)

UPV
Spain

V. G. Llorence,
S. G. Bermell,
G. V. Martín

TRACE

9

Detailed discussions on obtained results for steady and transient up to 500 s
after the break
9
Break flow overestimated, but the reason and consequences not welldiscussed

6-2

Post-calculation of
ROSA/LSTF Test 6.2 using
TRACE v4.274 Preliminary results(post-test analysis)

PSI
A. Jasiulevicius,
Switzerlan O. Zerkak,dR. Macian-Juan

TRACE

9

First presentation on TRACE (4.274) post-test calculation result -preliminary result
9
TRACE input preparation with 3D vessel component to study asymmetricalbehavior between two primary loops of LSTF
9Good overall prediction but with over prediction in SG pressure and somefluctuation in primary pressure

PRG5 2007 May at NEA Paris
91-2

Temperature Stratification
Experiment(pre-test analysis)

JAEA
Japan

T. Watanabe

RELAP5/
MOD3.2

Three
cases
were
comparedtodefineTest
1-2:(1) 5% cold-leg break without HPI, (2) 1% cold-leg break with HPI, and (3) 5%
hot-leg break without HPI.
9
Cold-leg flow conditions at the ECCS injection timing discussed.

3-1

High
Power
NaturalCirculation(post-test analysis)

JAEA
Japan

T. Takeda

RELAP5/
MOD3.2

9

68

Break flow rate overpredicted due to failure in correct simulation of cold leg
liquid level especially during two-phase flow discharge period, resulting inunderprediction of primary pressure, early significant drop in hot leg liquid level

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or modeland temporal drop in core liquid level with loop seal clearing
9
Liquid accumulation in SG U-tube upflow-side improved by simulating SG
U-tube with fine-mesh model and by applying CCFL model with Wallis-type
correlation to all the SG U-tube junctions

Code

9
9

6-2

First Results of a Post-Test
Calculation of LSTF Test 62 (SB-PV-10) with the Code
ATHLET(post-test analysis)

GRS
Germany

H. Austregesilo,
H. Glaeser

ATHLET

Input modified from that for Test 6-1
Simulation of break line at bottom of pressure vessel with dischargecoefficient Cd = 0.92 for subcooled and saturated flow & Cd = 0.8 for steamflow
9
SG U-tubes divided in 2 groups with different length
9Good simulation of overall trend for the early phase because of goodprediction of sub-critical flow at break for most part of the transient by adequate
simulation of flow resistances at break line, while some discrepancy in cold leg
liquid level under influences of accumulator coolant injection during natural
circulation
9
Some problems to simulate influences of non-condensable gas, causing
oscillation in primary pressure with considerably increasing CPU time
9
Necessary of multiple SG U-tubes modeling pointed out

6-2

Preliminary Post Test
Calculation of
OECD/ROSA Test 6.2 with
APROS
(post-test analysis)

VTT
Finland

P. Inkinen,
I. Karppinen

APROS

9
9

SG U-tubes simulated by 3 groups with different length
Good overall simulation of transient via good simulation of break flow (notthe all the transient discussed)
9
Problems in correct simulation of natural circulation via multiple SG U-tubes
under influences of non-condensable gas at rather low-pressure condition

6-1
&
6-2

ROSA-V tests 6.1 and 6.2
simulations with TRACE(post-test analysis)

PSI
A. Jasiulevicius,
Switzerlan O. Zerkak,dR. Macian-Juan

TRACE

9

Test 6-1 (TRACE v.5.00): Good overall transient simulation, though with aproblem in convergence in break upstream pipe stratified flow and in upper headliquid behavior (Discharge coefficient (Cd) not shown)
9
Test 6-2 (TRACE v.4.274): Good overall transient simulation, though early
SG U-tube drain initiation. TRACE v.5.00 over-predicted break flow rate,causing much early U-tube drain, early termination of natural circulation andhigher primary pressure
9
Necessary of multiple SG U-tubes modeling pointed out vs. one lumped Utube for the reported calculation.

6-2

Case 6.2. Preliminarytransient results(post-test analysis)

TRACE

9
9

UPV
Spain

V. G. Llorence,
S. G. Bermell,
G. V. Martín

69

Comparison of calculated results up to 5,000 s in about 24,000 s transient
SG RV flow rate adjusted during the cycle opening after system isolation(MSIV closure)NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model
9
Good agreement in break flow (single-phase liquid flow)
9
SG U-tubes drain earlier, but horizontal leg emptying greatly delayed –reason(s) are not well understood9
Downcomer level decrease after ECCS injection due to steam condensation
was milder than in LSTF test.

Code

PRG6 2007 November at JAEA Tokai
9

1-1

FLUENT Analysis of Test
1.1
(post-test analysis)

KFKI
Hungary

T. Farkas,
I. TóthFLUENT

Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety
Applications by WGAMA followed; cell skewness, size change between
adjacent cells, aspect ratio, and special care for mesh around pipe connection
9
Gambit mesh generation and steady-state calculations
9
Results from 3 types of turbulent models compared; RSM seems to predict
stratification most closely, while Realizable k-ε strongly underpredicts it. RSM(and to some extent Standard k-ε) tend to overpredict jet dynamics, especially inthe measurement plane „2”. Change from 1st to 2nd order discretization scheme
in Realizable k-ε has little effect.
9
Effect of the CL bend upstream of the injection appeared in all calculated
results as secondary flows.
9
Transient nature of experiment pointed-out; ECCS pipe fluid temperature

3-1

Results from the simulationof LSTF Test 3.1 (SB-CL38) using TRACE v5.0
(post-test analysis)

USNRC
USA

S.O.Marshall

TRACE

9

Good calculation of overall response: pressure, natural circulation flow rate,
supercritical flow in hot leg
9Unexplained temporary increase in primary-side pressure at ~500 s duringhigh-power natural circulation period caused overestimation of break flow rateand earlier core uncovery and natural circulation flow rate, and failure of proper
prediction of super-critical flow in hot legs
9
Problems with accumulator modeling resulted in inaccurate fuel rod and coreexit fluid temperatures and hot-leg level recovery

3-1

Post test calculations of
ROSA LSTF Test 3.1 using
RELAP5/MOD3.3(post-test analysis)

UPC
Spain

V. Martínez,F. Reventós,
C. Pretel,
Ll. Batet

RELAP5/
MOD3.39

Very good calculations in pressure, natural circulation flow rate, SG U-tube
coolant accumulation until the initiation of accumulator injection
9
Detailed adjustment (validation) of pressure difference distribution along the
primary loop and in the pressure vessel, and of downcomer-hot leg (upper
plenum) bypass
9
Significant difference in loop seal behavior around 1000 s and an importantdelay in the loop seal clearing as a result of a wrong accumulator injection.
9
Differences in the core refill and significant fluctuations in the core level70

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model
appeared after accumulator started coolant injection, suggesting that vessel
bypasses and upper head could be a source of uncertainty
9
Low-pressure behavior under influences of non-condensable gas did not agree
well with the data

Code

9

5-1

Primary Cooling throughSG
Secondary-side
Depressurization(post-test analysis)

JAEA
Japan

T. Takeda

RELAP5/
MOD3.2

Simulate 141 SG tubes by 9 parallel flow channels with different lengths,
namely 24 nodes for short to medium tubes and 26 nodes for medium to longtubes
9
Code has remaining problems in proper prediction of primary loop flow rateand SG U-tube liquid level behaviors especially after enhanced depressurization
through fully opening of SG safety valves in both loops.

6-2Preliminary post-test
calculation of OECD/ROSA
Test 6.2 with TRACE5(post-test analysis)

UPV
Spain

S. Gallardo,
G. Verdú,V. Garcia

TRACE

9

Good agreement in general for pressures even after the non-condensable gas
ingress, and primary loop natural circulation flow rate - Note: There was a
problem in N2 gas discharge from accumulator tanks
9
Temporary underprediction of primary pressure between 1000-2000 s as
found in APROS/VTT calculation. There were further discrepancies in primary
mass flow rates, fluid temperatures and PV downcomer liquid level. Necessityof detailed study of upper head behavior is suggested.
9
SG U-tubes divided in 3 groups (short, medium, long)9
Sub-critical flow at break -- Discharge coefficient Cd adjusted to 0.78
9Accumulator valve flow area adjusted

Development of ROSA
model for TRACE 5.0(post-test analysis)

UPM
Spain

G. Jimenez,
C. Queral,
J. Barrera,
A. Exposito,
P. Niesutta,
L. Valle

TRACE

9
9
9

Translation of TRAC-PF1/MOD1 input model into a TRACE input modelPreparation of Animation Mask with SNAP application
Several new points:2-D PZR to avoid thermal stratification, 9 SG U-tubes,
heat loss, break upstream tube
9
Future revisions (adjustments): comparison with RELAP5 input model and
ROSA facility descriptions, Check impact of level tracking option inside 3D
vessel, check & adjust pressure losses in 3D vessel such as bypass flow to upper
plenumSimulation of OECD/ROSA
Test 6.1 and 6.2(post-test analysis)

UPM
Spain

C. Queral,
J. Barrera,
G. Jimenez,
P. Niesutta,
L. Valle,

TRACE

9
Utilization of the above-prepared input model for LSTF
9
Similar nodalization rules to be applied to Almaraz NPP analyses
9
Plant model is validated with a real plant transient
Test 6-1
9
Break flow model with different nodalizations before and after flow
restriction (critical flow point)

6-1
&
6-2

71

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model
9
Adjust bypass flow to upper head and temperature in upper head
9
Late start of core uncovery due probably to bad simulation of coolant
distribution and/or two-phase break discharge flow -- influences of SG U-tube
modeling studied (1 tube & 9 tubes)
9
Simulation with/without core protection system and possibility of PCT higher
than 1200 °C pointed-out
Test 6-2
9
Break flow model with different nodalizations before and after flow
restriction (critical flow point)
9
Primary pressure behavior adjustment by new heat structure & heat loss
9
SG U-tubes simulation with coarse 1-tube model
9
Overestimation of two-phase break flow rate and early core uncovery

Code

A. Ex́ pósito

6-2

Revised Post Test
Calculation of
OECD/ROSA Test 6.2 with
APROS
(post-test analysis)

VTT
Finland

P. Inkinen,
I. Karppinen

9

APROS

Improvement in system nodalization; core region including upper plenum
divided into 3 channels, and heat conduction from upper tie plate to core barrel
modeled
9
Initial 4000 s transient was presented with good overall agreement
9
Temporary underprediction of primary pressure between 1000-2000 s as
found in TRACE/UPC calculation9
Problems in modeling the Upper head and reproduction of liquid level in DC
was noted and probably improper heat loss from the pressure vessel werepointed-out

PRG7 2008 May at GRS Garching
1-1

Stratification test 1.1
Preliminary results of CFD
simulations with Ansys
CFX(post-test analysis)

3-1

OECD/ROSA Test 3.1
TRACE5(post-test analysis)

Tractebel C. Mandy,
engineerin S. Keijersgsuez
Belgium

UPV
Spain

V. Abella,
S. Gallardo,
G. VerdúAnsys CFX9945° injection case selected as more representative for Belgian plants
Wide range simulation including pressure vessel downcomer and cold leg inthe other side9
Shear Stress Turbulence model (Combined 2 equations model)
9Time step 0.05s for transient calculation with 7 internal iterations at
maximum and polynomial function input for ECCS mass flow rate, temperaturesand pressure boundary at downcomer outlet
9First-look analysis results indicated with future planning
9
9

TRACE

72

Overall response was well simulated, including hot leg supercritical flows
Primary pressure decrease rate over-estimated with early accumulator
injection, late core uncovery, probably because of disagreement of break flow

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or modeldue to disagreement in the cold leg liquid level prediction
9
Level drop in downcomer larger than in upper plenum and core
9
Significant accumulator coolant injection with significant steam condensation
and abrupt level recovery in the core, upper plenum and hot legs not well
predicted
9Increase in the break valve friction by factor 2 caused decrease in the breakflow rate and better primary pressure transient, but more delay in core uncovery
initiation
9
Increase in hot leg bypath friction caused better primary pressure, but not sosignificant improvement in the range tested
9
More parameter survey are being planned (ex.) for friction factors in cold legs
and pressure vessel, and CCFL coefficients and modes (location unclear)

Code

9

3-2

High
Power
NaturalCirculation (LOFW w/o
Scram Test 3-2)
(post-test analysis)

JAEA
Japan

T. Takeda

RELAP5/
MOD3.2PZR liquid level behaviors improved by applying liquid entrainment model tosurge-line inlet and CCFL model with Wallis-type correlation to PZR bottom
9
Code has remaining problems in proper prediction of primary loop flow rateand SG U-tube liquid level behaviors with odd oscillation.
¾ Influences of multi-channel model for SG U-tubes are to be investigated bysimultaneously applying CCFL model with Wallis-type correlation to inlet
plenum and all the U-tube junctions of SG

5-1

Preliminary Post-test
Calculation of ROSA/LSTF
Test 5-1 with APROS(post-test analysis)

VTT
Finland

P. Inkinen

APROS

9

Noding completely 1-dimensional (including the core that is divided into 3
regions) with 3 SG U-tube groups
9Good overall prediction, except low-pressure transient especially after SG
depressurization
9
Flow rate in the long U-tube group behaved unexpectedly¾ Flow stopped in the other U-tubes after the RV open as in the experiment¾ After the opening of SVs flow rate decreased in the long U-tube group andincreased in the other U-tube groups
9
Cold leg liquid levels remain at higher level due to lower pressure vessel
liquid level, causing an increase in break flow rate through the end of thetransient

5-2

Primary Cooling throughSG
Secondary-side
Depressurization(Noncondensable Gas Inflow

JAEA
Japan

T. Takeda

RELAP5/
MOD3.2

9

Simulate 141 SG tubes by 9 parallel flow channels with different lengths,
namely 24 nodes for short to medium tubes and 26 nodes for medium to longtubes
9
Code has remaining problems in proper prediction of primary loop flow rate73

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or modeland SG U-tube liquid level behaviors that were observed with different tendency
among instrumented tubes especially after non-condensable gas inflow,
suggesting that uncertainty remains in prediction of steam condensation in SG
U-tubes which typically appears under influences of non-condensable gas.

Code

Test 5-2)(post-test analysis)

9
9

6-1
&
6-2

OECD/ROSA Tests 6.1 and6.2 with TRACE5(post-test analysis)

V. Abella,
S. Gallardo,
G. Verdú,V. Garcia

TRACE

Similar models which differ mainly in the break situation
Transform the model from TRAC-P to TRACE, including¾ Re-design of components (i.e., SG, PV)¾ Sensitivity analyses of physical parameters such as friction coefficients,
discharge coefficients for (break) critical flow model, and for the heat transfer
coefficients.
Test 6-1
9Good overall behavior prediction by using fitting discharge coefficients for
single-phase liquid and two-phase break critical flows. Steam discharge rate wasover-predicted because of lack in the coefficient fitting.
9
Faster depressurization of SGs than in experiment, which made the SG
pressure reached to the primary pressure far earlier. Heater rod surface
temperature excursion rate was influenced by the fast decreasing primarypressure (saturation temperature).
Test 6-2
9Good overall behavior prediction by using a fitting discharge coefficient of
0.78 for break critical flow. Good primary pressure response after noncondensable gas ingress.
9Core uncovery started a little earlier than in experiment
9
Qualitatively good calculation of unbalanced cold leg level.
9
It was suggested that reverse friction coefficient may have a decisive
influence onto the cold leg behavior calculation.

6-1

Post-test Thermal-Hydraulic
PSI
J. Freixa,
analysis of ROSA Test 6.1
Switzerlan A. Manera,using TRACEd
O. Zerkak,(post-test analysis)
J. Dreier

TRACE5

9

UPV
Spain

Single SG U-tube and all bypasses nodalized according to the geometry andthe friction k-factors being adjusted to have the bypasses mass flows during
steady-state match the experimental values
9Good overall response calculated including heater rod temperature thatrecorded PCT, by using default break discharge coefficients. A little delay in thetransition from two-phase flow to single-phase vapor discharge, andoverprediction of steam discharge rate.
¾ Additional (new) discharge coefficient of 0.8 for steam discharge improved74

NEA/CSNI/R(2013)1
ROSA
Test

Title

A TRACE model of the
ROSA/LSTF

Organizati
on

VTT
Finland

Author(s)S. Hillberg

Major results andsuggestions to improve and develop code and/or model
the result.
9
Phenomenological study performed to clarify reason(s) of discrepancy in the
DC and cold leg level behavior during blowdown phase. Additional leakage at
DC-UP by 1% of the core flow at steady state conditions dramatically found to
improve the result.

Code

9
9

TRACE5The model is built with SNAP, input has not been directly edited
As for pressure vessel, for example, 16 axial levels, 4 radial rings, 4
azimuthal sectors
9
A work in progress

PRG8 2008 November at OECD Paris
9
9

1-1

FLUENT Analysis of a
ROSA Cold Leg
Stratification Test
(post-test analysis)

KFKI
Hungary

T. Farkas,
I. TóthFLUENT

Paper for XCFD4NRS in September, 2008
Follow-on of PRG6 results, utilizing almost the same method such as
consideration to the OECD Best Practice Guidelines (BPG)911 cases with variation in 3 turbulence models, standard wall functions, fullbuoyancy effects, SIMPLE pressure-velocity coupling etc.
9Standard k-ε & RSM overestimate buoyancy effect and indicate less intensive
mixing of cold plume, while Realizable k-ε model underestimates buoyancy
effect but with more effective mixing
9
Necessity of more detailed information on the flow field such as velocity
distributions with some turbulence properties pointed-out

1-1

Analysis of Temperature
Stratification
Phenomena
Examination in Cold Legwith FLUENT code
(post-test analysis)

JNES
Japan

Y. Masuhara,
F. Kasahara

FLUENT

9

Analysis of temperature stratification phenomena during single-phase natural
circulation mode with 2 different mesh numbers for each of 2 horizontal legs
and a part of vessel downcomer
9Standard k-εmodel showed a good prediction among compared two types ofk-εmodels9
In loop-B with inclined ECCS nozzle, thermal stratification was not properly
simulated

3-1
&
3-2

OECD/ROSA Test 3.1 and
3.2 Simulation with
TRACE5(post-test analysis)

UPV
Spain

V. Abella,
S. Gallardo,
G. Verdú

TRACE

Test 3-1
9
Revision of the effort for the previous PRG79
Results including liquid levels in core and cold leg improved, but with no
explanation on the change(s) in the analysis method
9
Coolant accumulation in SG U-tubes overestimated at early stage of transient,causing longer time to complete coolant drain

75

NEA/CSNI/R(2013)1
ROSA
Test

Title

Organizati
onAuthor(s)Major results andsuggestions to improve and develop code and/or model
9
Overestimation of break flow at low-pressure transient after completion ofcore reflooding, due probably to temporary overestimation of liquid level in cold
leg-B
Test 3-2
9Overall transient was well simulated, but with overestimation of natural
circulation flow rate in both loops suggesting remaining problems in theprediction of flows in multiple parallel SG U-tubes
9Overestimation of void fraction in the core during two-phase natural
circulation phase, and rather abrupt drop in the liquid levels in pressure vessel
and hot legs at the termination of the two-phase natural circulation

Code

6-1

Plant Applications of ROSA
6.1 Test.
Accident
management
actions in an Upper HeadSBLOCA(post-test analysis)

UPM
Spain

C. Queral,
A. Expósito,
L. Valle,
G. Jiménez,E. Villalba,
S. Beneyto

TRACE

Test 6-1
9
Basically, the same input was used for PRG6 in 2007 Nov.
9
Influences of reflood option and reverse friction factors tested -- combined
utilization may improve the core rod temperature (PCT).
9
Mutual comparison among 5 Participant results and LSTF data -- AEKI is thebest of all, but UPM still overestimated break flow rate at steam discharge
phase, and late core uncovery happened.
9
Revision of input (nodalization etc.) is to be performed.
9
CET response and effectiveness of RVLIS (In-vessel liquid level monitor)
discussed through parametric study on influences of AM depressurization timing
onto PCTPWR analysis
9
Break area adjusted to obtain similar pressure transient than in ROSA 6.1 test
and power scaled with respect to primary volume
9
Influences of available number of accumulators onto PCT studied
parametrically in conjunction with RVLIS condition changes

6-2

Preliminary post-test
calculation of OECD/ROSA
Test 6.2 with TRACE5(post-test analysis)

UPV
Spain

S. Gallardo,
G. Verdú,V. Garcia

TRACE

9

During really transient of 1000-2000 s, underprediction of primary pressure,
overprediction of primary loop mass flow rate and overprediction of downcomer
liquid level appeared.
9
Influences of nitrogen gas inflow were not well predicted in such parameters
as primary pressure, collapsed liquid level in SG U-tube in loop without PZR
and hot leg liquid level.

76

NEA/CSNI/R(2013)1

A2-3 Questionnaire for Test 6-1 Analysis by Project ParticipantsA2-3-1 Background and Objective
Many post-test analyses were performed for Test 6-1 (PV upper-head LOCA with AM measure) by using
such best estimate (BE) LOCA analyses codes as ATHLET, APROS, CATHARE, RELAP5 and TRACE,
and were presented at several PRG meetings of the ROSA Project. The calculated results agreed wellwith the experimental results. Several participants even extrapolated their results into a postulated
condition, as a trial basis, to expect the true peak cladding temperature (PCT) by assuming that the core
power did not trip-off even after the limiting core temperature in the LSTF experiment.
Based on the good agreement of results with the LSTF experiment data, and partly because the five majordifferent computer codes were utilized so far by 4th PRG in 2006, a questionnaire was proposed to draw
practical information mutually of interest among participants. The questionnaire asked some details of themodels and methods used for their post-test analysis, if any. Efforts associated with performing theanalyses such as manpower and time spent to complete the analyses were also asked as well as their
computing circumstances. Replies were compared each other to clarify their practical efforts to obtain thegood results by means of some practical analysis methods including selection and tuning of model(s) andcalculation method(s) that would be applied for the reactor safety analysis (extrapolation from the scaleddown problem).
The questionnaire, however, was not meant to compare the calculated results themselves, considering that
it was possible to incorporate some efforts into the post-test calculations by optimizing the code with thebest accuracy. This attempt is, therefore, different from the ISP (international standard problem) in thisrespect.
The obtained results are summarized in the following.
A2-3-2 Questionnaire
The questionnaire and the mutual comparison of five replies are respectively shown in Attachment A23-1 and A2-3-2. The questionnaire is composed of the following four items;
1. Computer CodeSpecific features of the code and motivations to perform post-test analyses
2. Efforts
Computer and computing time, and manpower characterization
3. Selection Method(s) of Model(s) and Method(s) specific to the CodeStrategy to put priority on specific parameter(s), selection of specific model(s)/method(s),simulation of multi-dimensional phenomena
4. Models and Methods Specific To Test 6-1
Method(s) to use five key models, and strategy to use the models in reactor analyses
A2-3-3 Major Results
For each of 4 items, major results are summarized as follows.
(1) Computer code
The following five computer codes were used; ATHLET Mod 2.1 Cycle A (GRS*), APROS 5.06
(VTT*), CATHARE2v15b mod5.2 (AVN*), TRACE 5.0RC3 (DSE-UPM*) and RELAP5 MOD3.2.1.2*

GRS: Gesellschaft für Anlagen- und Reaktorsicherheit mbH, Germany

77

NEA/CSNI/R(2013)1(JAEA). Three TSOs (technical safety/support organizations) and two universities who support regulatory
body provided answers to this questionnaire by June 28, 2007. The common major motivation to performthe post-test analysis was the validation and verification of their code in preparation for the safety
analysis.
(2) Efforts
The calculation was performed mostly by using a so-called personal computer or its cluster with anoperating system such as Linux and Microsoft Windows XP. The CPU time ratio to the physical time wasranging from about 0.3 to 13, depending on the code and number of meshes.
The total work hours by the personnel who made the calculation (analyst) significantly varied from three
weeks to 1 year depending on their experiences to conduct code calculation. The analyst had at least one
co-worker (and more at several organizations) to discuss with on their analyses.
(3) Selection Method(s) of Model(s) and Method(s) specific to the CodeThree organizations put their priority onto the correct calculation of the break flow rate and liquid levelsin the core, upper head and upper plenum in the pressure vessel. The liquid level in the vessel upper headis important for the correct prediction of break upstream condition. One organization utilized an input
used for ISP-26 (5% cold leg break LOCA test by ROSA/LSTF, 1992) with some modifications
according to the facility modifications since the ISP-26.
The computer codes were used without any modifications except RELAP5 code, to which the break flow
models were incorporated for better calculation of two-phase discharge flow through an orifice. The
current computer codes are considered well improved through many of analysis experiences. However,care was taken to select appropriate models such as VOLUME model to calculate mixture level in thecase of CATHARE code. As for the ATHLET code, 6-equation model (fully separated momentum
equations) or 5-equation model (mixture of momentum equation and drift-flux approach) were applied toappropriate volumes. The break discharge coefficients and bypass flow were adjusted further to havegood agreement with the measured data.
As for the simulation of three-dimensional phenomena, no special measures were taken except the 3-D
noding in the pressure vessel employed by the ATHLET and TRACE codes. The other components were
simulated basically by using zero-dimensional volume components (ATHLET and CATHARE codes)
and one-dimensional components such as pipe.
(4) Models and Methods Specific To Test 6-1
Break Flow
The break flow is the most important parameter for the correct simulation of LOCA because it controls
the mass and energy inventory in the system. Each computer code utilizes it’s own set of critical(choking) flow models to calculate the break flow rate for single-phase (subcooled) liquid flow,(saturated) two-phase flow and single-phase steam flow with a discharge coefficient for each model. Nonequilibrium flow condition may be handled too. In some cases, analysts changed the model(s) for theproper simulation of the flow through orifice break used in the ROSA Test 6-1. When they use default
models, they followed the guideline in the user manual. For all the cases, however, the dischargecoefficients for three flow conditions were respectively adjusted against the measured data. The same setof models and methods including the upstream piping are to be used for the safety analysis. The discharge
VTT: Valtion Teknillinen Tutkimuskeskus, Finland
AVN: Association Vinçotte Nuclear, Belgium
DSE-UPM: Departamento de Sistemas Energeticos - Universidad Politechnica de Madrid, Spain78

NEA/CSNI/R(2013)1
coefficient would then be one of the primary parameters to discuss the influences of uncertainty.
Liquid level in vessel upper head
In ROSA Test 6-1, this and the following two (in total three) parameters are important for the proper
simulation of the break upstream pipe flow conditions. Two methods were used depending on thecomputer codes. The ATHLET and CATHARE codes used a single volume to simulate whole upper
head, being composed of two sub-volumes that respectively represent upper steam phase with liquid
entrainments and lower two-phase mixture phase. The other three codes used vertical stack of single
volumes (pipe). The mixture level is defined from the void fraction. A level tracking model is employedby the TRACE code. Interphase drag was decreased by 1/10 for the proper simulation for the
RELAP5/MOD3.2 code. The calculated results, however, seem to be influenced by the nodalization(number of slices of the vertically stacked volumes) for the latter three codes. For APROS code, flow
resistances were adjusted in upper head as well as in control rod guide tubes, upper plenum anddowncomer. The same model and methods (strategy) are to be used for the reactor safety analyses. Thisparticular case, thus the simulation of mixture level in a large volume such as upper head, would not be somuch influential in the scaling of phenomena.

Vapor/Liquid entrainment in vessel upper head
Every computer codes have an entrainment model for the branch pipe flow. This model was applied to thenozzle that connects to the break upstream pipe, except for the RELAP5 code case. The ATHLET and
CATHARE codes employed further an optional vapor pull-through model considering the high-speed
steam flow. A horizontal pipe was directly connected to the upper head from the break upstream pipe in
the ATHLET code model, while a 45-degree inclined pipe was used to connect the upper head and thebreak upstream horizontal pipe in the LSTF. As for the RELAP5 code, fine mesh representation was
performed to the volume above the top of CRGT. The interphase drag for these fine-mesh volumes wasdecreased to 1/10. The same model(s) and methods are to be used for the reactor safety analyses.
Flows through multiple Control Rod Guide Tubes (CRGTs)
There are eight simulated CRGTs in the LSTF, which connect the upper plenum and upper head. Good
simulation of the flow through the CRGTs is indispensable for the proper prediction of the break
upstream conditions. The parallel channels of CRGTs were simulated in the code calculations by either(two or eight) parallel multi-channels or one lumped flow channel, depending on the core and upper
plenum noding. Eight PIPEs were employed for the TRACE analysis to achieve rigorous simulation by
fully utilizing the 3D VESSEL component. Two parallel pipes were employed for the ATHLET and
APROS analyses by considering the change in the flow conditions depending on the location of CRGTs;
above the high-power or low-power fuel bundles. One lumped flow channel was selected for the
CATHARE and RELAP5 analyses for simplicity. Consequently, It seems that the parallel multi-channel
simulation gave better results in the flow through the CRGTs. Care was taken further in the ATHLET
analyses such that the model option for the drift flux-based interfacial friction in the upper plenum
volume was selected to have one for the pipe geometries rather than for the usual bundle geometries. The
higher interfacial friction was resulted, causing higher resistance for the liquid downflow. In the RELAP5
analysis, the coolant conditions in the upper plenum were controlled by the number of slices of verticallystacked volumes. The same model(s) and methods are to be used for the reactor safety analyses.
Liquid level in the core and rod surface temperature
The interphase drag is defined based on the flow regimes in the core rod bundle for appropriate
representation of vertical void fraction distribution under mixture level in the core. In the ATHLET code,
the code-specific flooding-based drift flux model was used to define the interphase drag. For the
RELAP5/MOD3.2 code, the interphase drag was reduced to 1/10 based on the previous experiences. The
79

NEA/CSNI/R(2013)1
heat transfer in the core was calculated by using the default set of heat transfer correlations for Test 6-1
by all the codes, though the correlations are different from code to code. The same model(s) and methods
are to be used for the reactor safety analyses.
A2-3-4 Summary
Five detailed replies to the questionnaire were compared and summarized to reveal the efforts done by theparticipants to attain the good simulation results for ROSA Project Test 6-1 by using the BE LOCA
codes. Efforts associated with the analyses work such as manpower and time spent to complete theanalyses were included further in the questionnaire as well as the computing circumstances. All theanalysts were found to work for the regulatory body as TSO and university. The motivation to conduct
the post-test analyses was the validation and verification of the BE codes to finally apply it to the reactor
safety analyses.
It appeared that all the BE codes request the analysts to make many selections and decisions against many
thermal hydraulic phenomena simulation, on the noding, models and combination of models andcoefficients, especially when the multi-dimensional phenomena and the multiple parallel flows are
involved. The analyst then may need very long time to build and adjust their input model, depending on
their experiences. They referred user manual and had a cooperation of their colleague(s) to obtain the best
results. This style of work is surely good to minimize the user effect.
The computer codes still have difficulties in the simulation of multi-dimensional phenomena except thosein the pressure vessel for some codes that have a capability of three-dimensional node representation.
The break flow as the most important parameter for the rigorous simulation of LOCA was obtained wellby mainly adjusting the discharge coefficients for break models that are different from code to code.
Efforts were taken to achieve the proper simulation in the break upstream conditions and the mixture
levels in the pressure vessel; core and upper head. Basically, the models and methods used for this posttest analysis are to be used for the reactor safety analyses without changes. It was difficult to fully clarify
the influences of the phenomena scaling within the present questionnaire effort.
Good information was obtained in total to look into the motivation, circumstances and efforts associatedwith the post-test analysis of ROSA Project Test 6-1 by using the BE LOCA codes. The efforts taken forand experiences and information gained through the post-test analyses and the questionnaire would be
beneficial in the preparation of PIRT (Phenomena Identification and Ranking Table) and in the BEPU(Best Estimate plus Uncertainty) analyses for the safety evaluation of current and future LWRs.

80

NEA/CSNI/R(2013)1
Attachment A2-3-1
OECD/NEA ROSA Project
Data/Information Transfer
January 10, 2007
Thermohydraulic Safety Research Group
Nuclear Safety Research Center
JAEA
General questionnaire on the analysis methods for Test 6-1
Various post-test analyses have been presented in PRG3 and PRG4 by the project participants to validate
predictive capabilities of best-estimate (BE) computer codes such as ATHLET, CATHARE, TRACE,APROS and RELAP5/MOD3 against the ROSA/LSTF Test 6-1 data. Each presentation has shown nicecalculated results with accuracy. However, the indicated results seem to rely on their specific models andnodings to simulate thermal-hydraulic phenomena in the test facility based on the long-term experiences
against various subjects. In the presentation, details of such aspects are sometimes not included as a kindof common sense to the relevant researchers/engineers, while not for the other participants. Such a trend
may be implicitly emphasized in post-test analysis. It is thus good to summarize the major aspects ofcomputer codes and associated strategy to select primary models and calculation methods used to analyze
the ROSA/LSTF Test 6-1 as well as motivations to perform the code verification and development.
Following is a list of tentative questionnaires to collect and summarize necessary information to clarify
such details to thoroughly understand the calculated results for mutual comparison and discussions. Every
participating organization of the OECD/NEA ROSA Project is requested to reply to the questionnaires,
even when the code analysis is not performed, by assuming that some analyses are to be performed.

No.

Item

Questionnaire
1. Computer Code

1-1

Name

Name of the code (used or to be used)

1-2

Specific Feature(s)Specific feature(s) (as BE code)

1-3

Motivations(1) Specific objective(s) to develop/improve the code (as BE code)(2) Reason(s) to perform analyses by using the code against ROSA Test 6-1(3) Expected practical subjects for the code to be used for
2. Efforts

2-1

NodingNumber of volumes and junctions

2-2

CPU Time(1) Name of CPU and OS(2) Typical CPU time for physical time(3) CPU time ratio = (Typical CPU Time) / (Physical Time)81

NEA/CSNI/R(2013)1

No.
2-3

Item

Questionnaire

Manpower(1) Number of personnel (calculator) who performed the code calculation(2) Number of personnel who contributed the calculation work through discussions
with calculator(s)(3) Calculator -- a/Regulator, b/Supporting Organization, c/University (i/Professor
or ii/Student), d/Vendor, e/Utility, f/External employee of a/, b/, d/ or e/,g/others (specify)(4) Career (years) of calculator(s) to perform the code analysis of this kind(5) Total work hours for the main calculator to provide the results
3. Selection Method(s) of Model(s) and Method(s) specific to the Code

3-1

Strategy

Key information(s) and/or consideration(s) to put priority on major calculation
parameters such as liquid level in upper plenum

3-2

Selection ofModel(s) andMethod(s)(1) Method(s) to select the model(s) in the code and method(s) to handle the codewith validation method(s) for the selection of the model(s) and/or the method(s).
List all of the relevant model(s) and/or method(s).
(2) Try & error process for the selection and/or elimination of the model(s), thespecific constant(s) and/or assumption(s) for the model(s) and the method(s), if
any.

3-3

3-4

Simulation of multidimensional
phenomena(1) Practical method(s) to simulate 3-dimensional two-phase flow phenomena byusing the code

Statisticaluncertainty
definition method(1) Experiences applicable, for example, for the optimized setup of noding in
combination with liquid level tracking model, if any.

ex.) 3-dimensional temperature distribution around the core during the coretemperature excursion to have an exact prediction of the core exit temperature to
start accident management action.

4. Models and Methods Specific To Test 6-1
4-1

Break flow

(1) Break flow model and specific coefficient with method(s) to define them(2) Strategy to apply them for reactor cases, namely in safety analysis

4-2

Liquid level invessel upper head(1) Method(s) and/or phase separation (inter-phase drag) model(s) to identify thelocation of liquid level (mixture level) in the vessel upper head, in conjunction
with noding modeling(2) Strategy to apply them for reactor cases, namely in safety analysis

4-3

4-4

4-5

Vapor/liquid
entrainment in vessel
upper head(1) Method(s) and/or model(s) to estimate the (two-phase) flow through rather large
volume vessel upper head towards break pipe

Flows throughmultiple Control
Rod Guide Tubes(CRGTs)(1) Method(s) and/or two-phase flow (flow regime definition) model(s) to simulate
multi-channel parallel pipe flows from upper plenum to upper head

Liquid level in thecore and rod surface
temperature(1) Method(s) and/or phase separation (inter-phase drag) model(s) to simulate void
fraction (distribution) in the core that defines two-phase mixture level in theupper plenum and core multi-channel(2) Strategy to apply them for reactor cases, namely in safety analysis(2) Strategy to apply them for reactor cases, namely in safety analysis(2) Boiling heat transfer model(s), definition of dryout, post-dryout heat transfer
model(s) to simulate temperature excursion in the core(3) Strategy to apply them for reactor cases, namely in safety analysis

82

NEA/CSNI/R(2013)1
Attachment A2-3-2
Answers of General questionnaire on the analysis methods for Test 6-1 -- as of June 28, 2007 -No.
1-1
1-2

Answers by GRS(ATHLET)
ATHLET Mod 2.1 Cycle A(version released in July 2006)
The thermal-hydraulic computer code
ATHLET is being developed by the
GRS for best-estimate analyses ofplant transients, design basis andbeyond design basis accidents in lightwater reactors.
The code is composed of several basic
modules: a thermo-fluiddynamic
module, a heat conduction and heattransfer module, a neutron kinetics
module, a general control simulation
module, and a general solver ofdifferential equation systems.
The thermo-fluid-dynamic module
includes a two-fluid formulation, withfully separated balance equations forliquid and vapour. It is complemented
by mass conservation equations for up
to 5 different non-condensable gases
and by a boron tracking model.
Specific models for pumps, valves,
steam separators, mixture level
tracking, critical flow, among others,
are also available within ATHLET.

Answers by VTT(APROS)

Answers by AVN(CATHARE)

Answers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)

APROS 5.06

CATHARE2v15b mod 5.1

TRACE 5.0 RC3

RELAP5/MOD3.2.1.2-

CATHARE2 (Code for Analysis ofThermal-Hydraulics during an Accidentof Reactor and safety Evaluation) has
been developed in Grenoble by the
French Atomic Energy Commission(CEA), Electricité de France (EDF) and
AREVA to perform Best-estimate
calculations of PWR accidents.
Based on a one-dimensional, two fluids,
six equations model with a unique set of
constitutive laws.
1-D AXIAL moduleThis module solves for each mesh thefollowing equations:• 2 energy balance equations(liquid and vapour phase)• 2 momentum balance equations(liquid and vapour phase)• 2 mass balance equations(liquid and vapour phase)• 1 or 2 mass balance equations for
1 or 2 noncondensable gasses
for a total of 6+2 equations per mesh.
It is generally used for piping (i.e. hotlegs, cold legs, core channels, SG Utubes, SG downcomer, SG riser,
feedwater lines, vapour lines, etc).
1-D VOLUME moduleThis module consists of 2 subvolumes(INF and SUP). In each one are solved:• 2 energy balance equations(liquid and vapour phase)• 2 mass balance equations

1) 1D/3D, six/eightequation models, nonequilibrium, nonhomogeneous model.
Also includes noncondensable gas andboron transport.
2) It includes a 3D
VESSEL model.

1) 1D, 2-fluid, non-equilibrium, nonhomogeneous model for 2ø steam-water
mixture that can contain noncondensable gas in steam phase and/or
soluble gas in water phase2) System component models such aspressurizer, accumulator, valve, pump,separator and control system
3) Core power is calculated by pointreactor kinetics models with reactivity
feedback or is given by preprogrammed time-dependent power
table.

--

6 equation thermal
hydraulic model
1 and 3-D neutron
kinetics
integrated containment
model
comprehensivemodeling of control
and automation
systems
graphical userinterface with processcomponent library83

NEA/CSNI/R(2013)1
No.

Answers by GRS(ATHLET)

Answers by VTT(APROS)

Answers by AVN(CATHARE)(liquid and vapour phase)
1 or 2 mass balance equations for
1 or 2 noncondensable gasses
For the total Volume is assigned onepressure equation for a total of 9+2
equations per mesh. For each junction
connected to the VOLUME areassociated two momentum balance
equations (liquid and vapour phase).
(1) Framework of code validation forsafety analysis(2) Code Validation for Vessel
SBLOCAs(3) Safety Analysis for BelgianPower plants

Answers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)(1) Validate the code
capabilities in these
conditions.
(2) We are interested insmall break LOCA
analysis at top andbottom vessel.
(3) We want to simulate
top/bottom vessel
SBLOCA sequenceswith our TRACE
model of Almaraz
Nuclear Power Plant(Westinghouse 3 loop)
and, we need to learn
which are thenodalization, models
and options adequate
in this kind of
sequences.
284 1D cells, 734 3D cells(1) Assess the predictability of
RELAP5/MOD3.2 code and models
for system integral analyses andmodify the models(2) Find out problem(s) in
RELAP5/MOD3.2 code to properly
predict thermal-hydraulic phenomenaduring vessel upper head small break
LOCA(3) Best-estimate plus uncertainty
analyses with RELAP5/MOD3.2 code

•

1-3

2-1

2-2

(1) The calculations have been
performed in the frame of the code
validation.
(2) Post-test calculation of separate
effect tests and integral tests aremain contributors to theenhancement of the validation
status of ATHLET. The selectionof experiments for code validation
takes into account the current
status of code development,
requirements of code users, current
safety issues as well as theavailability and adequacy of new
experimental data. The
OECD/ROSA Test 6-1 has been
recently added to the ATHLET
code validation matrix (Accident
management for a non-degradedcore in PWRs).
80 thermo-fluid-dynamic objects and
56 heat conduction objects, with a
total of 344 control volumes, 287
junctions and 261 heat slabs.
(1) CPU: A Linux Cluster with 2
AMD single core processors (2.4
GHz).
(2) The total CPU time up to the startof LPI injection was 2100 sec(3) An average CPU time ratio of 0.8.

(1) To enhance code
capabilities for safety
analysis andlicensing.
(2) Code validation(3) Safety analysis of
EPR

340 volumes, 342
junctions456 scalar meshes(1) CPU: Intel Pentium 4,
3.00 GHz
OS: Windows XP
Professional(2) 3.75(3) 12000/3200

(1)
(2)
(3)

Intel Pentium 4 3.20 GHz, 4GB
RAM, Linux Fedora Core 1.0
1000s for 13500s
1000s/ 13500s

84(1) vendor_id:
AuthenticAMD, cpu
family: 15, model: 37,
model name : AMD,
Opteron(tm) Processor
252, stepping : 1
cpu MHz : 996.164
cache size : 1024 KB(2)
3.2

About 190 volumes, about 200 junctions,about 180 heat slabs(1) CPU; Intel Itanium 2,
OS; SGI Linux(2) Typical CPU-time; 950 s(3) Typical CPU / physical time ratio;
0.32 (= 950 s / 3000 s)

NEA/CSNI/R(2013)1
No.

Answers by GRS(ATHLET)

Answers by VTT(APROS)

Answers by AVN(CATHARE)

2-3(1) One senior analyst(2) One coworker has contributed tothe analysis through discussions of
the results.
(3)(4) Several years of experience incode development and application
for reactor safety analysis(5) The total manpower is estimatedto 2 persons over 1 month.

(1)
(2)
(3)
(4)
(5)

3-1

No specific strategy was adopted forthis calculation. The ATHLET input
data was strongly based on the data set
used for the calculation of the ROSAIV LSTF small break test SB-CL-18 in
the frame of code validation, taking
into account the modifications
performed in the facility for the
ROSA-V Program.
(1) The selection of methods andmodeling options was largely
based on the recommended values
presented in the code User's
Manual /1/, which in turn reflect
the experience acquired by theextensive code validation. For the
primary circuit of the LSTF facility
the 6-equation model (fully
separated momentum equations)
was applied, except for thepressurizer, control rod guidetubes, upper head and break line,
where the 5-equation model(mixture momentum equation anddrift-flux approach) was used. The
secondary side was simulated withthe 5-equation model.
(2)
Multi-dimensional phenomena can be
simulated approximately with theparallel channel technique (one-Primary pressure andliquid level in the core

Priority on proper break flowrate. The
major investigated parameters werelevels within the vessel (especiallyUpper head, Upper plenum, and core)(1) Standard models used
(2) Break flow dischargecoefficient and flow
resistances indifferent parts of thepressure vessel were
adjusted to fit
experiment.

(1) Standard CATHARE model ofUpper Head and upper plenum(VOLUMEs model)

3-2

3-3

1 person1 personc ii
0600 hours = 4
months(ex. 5 hours a day for
analysis work)

No 3-dimensionalmodelling

(1)
(2)
(3)
(4)
(5)

1 person1 personb-TSO
10 years
3 weeks(2) Break size, Finer Upper Plenumand Upper Head ModelingAnswers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)(1) 1 person(2) three persons of DSEUPM, one full time
analyst and two parttime analyst(3) c ii(4) one(5) one year, because weare performing a newmodel of ROSA
facility.
No specific strategy was
adopted for thiscalculation.

(1) One analyst(2) Four persons of JAEA who contributeto analysis through discussions andone analyst(3) JAEA -- b(4) Analyst who engages in RELAP5code analyses for several years(5) Over 1 month(1) Standard models. Two
dimensionalpressurizer and several
heights in steam
generator U-tubes(2) Adjust head bypass
flowrate

No 3-D model was used

85

(1)

3D Vessel

Proper prediction of break flow rate andliquid levels in upper head, upper plenumand core as essential boundaries of vessel
upper head small break LOCA

(1) A two-phase critical flow model isincorporated by JAEA [1], which
employs the Bernoulli equation(incompressible flow) with a dischargecoefficient (Cd) of 0.61 for singlephase liquid and the maximum
bounding flow theory with a Cd of 0.61
for two-phase flow. A Cd of 0.84 isused for single-phase vapor.
(2) 1) Modify noding for upper head above
top of CRGT and interfacial drag ineach node to properly predict liquid
level in upper head
2) Adjust flow resistances of flow path
between downcomer and upper head tofit downcomer-to-upper head bypass
flow rate to 0.3% of total core flow rateduring initial-steady state.
Multi-dimensional phenomena cannot beanalyzed by RELAP5/MOD3.2 code with
one-dimensional model. As for simulation

NEA/CSNI/R(2013)1
No.

3-4

4-1

Answers by GRS(ATHLET)dimensional, parallel pipes connectedby cross-connection junctions) in
ATHLET.
For Test 6-1, the core region has beenmodeled by two concentric rings,
connected by a cross-connection
object. The inner channel ring includes
the high and middle powered bundles(13 to 20 and 21 to 24). The low
powered bundles are combined in theouter channel ring. The upper plenum
region up to the hot leg nozzles as wellas the control rod guide tubes are
divided accordingly.
In near future it is planned to apply the
GRS methodology for uncertainty andsensitivity analyses for an uncertainty
analysis of the Test 6-1 calculation,
taking into account not only the code
uncertainties (input parameters andmodeling options) but also
experimental uncertainties (e.g. core
bypass flows).
(1) Critical discharge flow rates arecalculated in ATHLET by a onedimensional thermo-dynamic nonequilibrium model, taking into
account the actual geometry of thedischarge flow path /1/. This model
generates tables of critical mass
fluxes for a set of fluid conditionsin the upstream discharge control
volume. For pure steam flow, adischarge coefficient of 0.9 has
been applied, for subcooled andsaturated flow 1.0.
(2) The same strategy is applied forreactor safety analyses.

Answers by VTT(APROS)

Answers by AVN(CATHARE)

Answers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)of multi-dimensional core phenomena, forexample, the use of computer code withthree-dimensional core model such as
TRAC-PF1 is needed.

None(1) Critical flow model,
discharge coefficient
0.86
(2)

(1) The Break is modeled with a TEE
and AXIAL pipe and a Boundary
condition as specified by the user
guidelines. This is the recommended
modeling procedure, particularly forbreaks on horizontal pipes; since ittakes into account the phaseseparation effects which are modeledin the TEE (see qualification reports).
No specific phase separation law wasdeveloped in the TEE for a vertical
pipe. Even without a phase separation
phenomenon, the critical flow
prediction is more accurate whencalculated with 1-D modeling of theflow in the nozzle (see Qualification
reports) than by using a critical flowcorrelation function of upstream
parameters (Break sub-module).

86

No experience inevaluation of uncertainties
in code analysis by using
statistical procedure

No experience in evaluation of
uncertainties in code analysis by using
statistical procedure(1) Default model.
Discharge coefficientsstill not adjusted.

(1) A two-phase critical flow model
employs the Bernoulli equation (theincompressible orifice flow equation)with a discharge coefficient of 0.61 forsingle-phase discharge liquid and the
maximum bounding flow theory with adischarge coefficient of 0.61 for twophase discharge flow. A dischargecoefficient of 0.84 is used for singlephase discharge vapor.
(2) The predictability of the break model
has been assessed through analyses of
LSTF small break LOCA experimentswith break sizes of less than 2.5% [1].
However, it is not sufficient to assess
the predictability of the break model
for cases with larger break sizes.
Reactor safety analysis is conductedbased on knowledge obtained from theNEA/CSNI/R(2013)1
No.

4-2

Answers by GRS(ATHLET)(1) The liquid level in vessel upper
head is determined by the
ATHLET specific mixture level
model /2/. This model is based on
the definition of a nonhomogeneous control volume,
composed by two homogeneous
sub-volumes, one for liquid withbubbles below the mixture level
and the other for vapour withdroplets above the mixture level.
Mass and energy balances are
performed separately for both subvolumes, taking into account massand energy exchanges at theinterface. The upper head isdefined as a branch object.
(2) Similar strategy is applied forreactor safety analyses.

Answers by VTT(APROS)

Answers by AVN(CATHARE)(2) The Same technique is applied to
the reactor case(1) Adjusting flow
(1) The Upper head is modeled using a
resistances in upper
VOLUME component rather than an
AXIAL one.
head, control rod
The volume module represents a
guide tubes, uppercapacity with several connections. It
plenum andis adapted for capacities with large
downcomer, no phasedimensions compared to theseparation applied
diameters of the junctions.
(code standard modelAll thermal-hydraulic quantities areused)assumed uniform in horizontal(2)
planes.
• The velocities inside the volume aresmall compared to the velocities at
the junctions.
• Inertial forces are assumed to be
negligible compared to gravity
forces. Consequently, themomentum equations are simplified
and the pressure field is hydrostatic.
• Phase stratification occurs (voidfraction stratification).
The stratification is represented by a
two-node model with two subvolumes (Ω-,Ω+). The interface
between the sub-volumes has avariable level. In each sub-volume,
enthalpies and void fractions are
assumed to be uniform but not thepressure that has a hydrostatic
gradient. It is assumed that there isliquid in the lower sub-volume
possibly with gas rising towards theinterface. In the upper sub-volume
there is mainly gas, possibly withliquid drops or falling jets.
Scalar variables are calculated inside
the volume in front of each junction(at a distance from the junction equal
to 2% of the volume height forvertical junctions). Flow distribution

87

Answers by DSE-UPM(TRACE)(1) Level tracking andsensitivity to
nodalization

Answers by JAEA
(RELAP5)
future uncertainty evaluationconsidering effects of break model.
(1) The RELAP5/MOD3.2 code
overpredicted break flow rate during
two-phase flow discharge. So, upper
head above top of CRGT was
represented by ten vertical nodes, andinterfacial drag in each node was
reduced to 1/10. Compare thecalculated results before and after thesemodifications.
(2) The same noding model and interfacial
drag are applied to reactor safety
analysis.

NEA/CSNI/R(2013)1
No.

4-3

Answers by GRS(ATHLET)

Answers by VTT(APROS)(1) The ATHLET specific T-junction(1) code standard intermodel /2/ is applied at the junction
phase drag andconnecting the vessel upper headentrainment modeland the break discharge line. Thisused
model simulates multidimensional(2)
flow phenomena of a T-branchingas the superposition of two onedimensional flows: one in the axial
direction of the main flow (upperhead) and one in the axial directionof the branch (break discharge line).
One additional momentum equation
is solved at the branching point.
The phase separation takes into
account liquid entrainment, vapour
pull-through, and it is a function of
the flow regime at the main flow
direction. For Test 6-1 the junction
connecting the upper head to thebreak discharge line was defined ashorizontal.
(2) The T-junction model is also
applied for reactor safety analyses.

Answers by AVN(CATHARE)
between sub-volumes and the phasesorting phenomena are modeled ateach junction, taking into account thetwo-phase jet effect pull-throughprocess.
Mass and energy transfer between thetwo sub-volumes (bubble rise, fall ofdroplets, condensation, evaporation)
are modeled.
For the one-phase liquid case, theupper sub-volume is residual (height=1 cm). Respectively the lower subvolume is residual for one-phase gas
fluid (height = 1 mm).
(2) The same model is applied to theReactor Case
(1) Scalar variables are calculated inside
the volume in front of each junction(at a distance from the junction equal
to 2% of the volume height forvertical junctions). Flow distribution
between sub-volumes and the phasesorting phenomena are modeled ateach junction, taking into account thetwo-phase jet effect pull-throughprocess.
A junction is the connection between
two adjacent modules. A junction iseither upstream or downstream of theelement. This choice is of noimportance in terms of flow and is a
topological control for the element
and the circuit. In an element (except
boundary condition) at least one
upstream and one downstream
junction should be defined; in acircuit, each junction should appear
in two, and only two, different
element definitions, each with adifferent direction key word. For
each element, a junction consists ofan assembly vector node and a scalar88

Answers by DSE-UPM(TRACE)(1) Standard TRACE
entrainment model

Answers by JAEA
(RELAP5)(1) Upper head was modeled by verticalpipe to simulate the corresponding
configuration, while entrainment model
for vertical pipe is not prepared in
RELAP5/MOD3.2 code. The code
overpredicted break flow rate during
two-phase flow discharge. So, upper
head above top of CRGT was
represented by ten vertical nodes, andinterfacial drag in each node was
reduced to 1/10.
(2) The same interfacial drag is applied toreactor safety analysis.

NEA/CSNI/R(2013)1
No.

4-4

4-5

Answers by GRS(ATHLET)The 8 control rod guide tubes in the test
facility are represented by two parallel
pipes in the ATHLET nodalization. The
calculations have shown that an
adequate simulation of the flow
resistance between upper plenum andupper head, not only within the CRGTs
but also for the downflow between hot
leg nozzles and core upper plate, isessential for a correct prediction of thetwo-phase break flow. One main factor
contributing to the good agreement
between calculation and measurement
was the change of the model option forthe calculation of the drift-based
interfacial friction within the upper
plenum volumes: instead of the usual
correlation for bundle geometries,
which results in a higher interfacial
friction and thus in a higher resistancefor liquid downflow, the correlation forpipe geometries was applied.
(1) Interfacial friction coefficients arecalculated in ATHLET as afunction of the drift velocities,
which in turn are determined by
the ATHLET specific flooding
based drift flux model taking into
account the channel geometry andthe flow conditions /2/. The
driftflux model has been validated

Answers by VTT(APROS)(1) CRGTs modeled with
two separate
channels, one forcenter tubes and one
for peripheral tubes(optional CCFL
models for paralleltubes available, but
not used in this case)(2)(1) Flow regime
dependent interfacial
friction correlations(2) Heat transfer model ischosen according flow
conditions. The user
can select fromseveral critical heat
flux correlations,

Answers by AVN(CATHARE)
node that is a copy of the adjacentmodule node.
The two-phase velocities Vl & Vg
are defined at the vector node and thetwo momentum equations are centredon this node. Due to the solution
method, the momentum equations
must be written by one of theadjacent modules.
(2) The same model is applied to theReactor Case
(1) One AXIAL pipe element was used
to simulate the all Guide Tubes ofthe facility. The UH/UP bypass
orifice were also modeled.
(2) The same model is applied to theReactor Case

(1) The flow regime is calculated
automatically by CATHARE in an
AXIAL pipe.
(2) The heat-transfer regime iscalculated automatically by
CATHARE between a wall and a TH element.
(3) The same model is applied to theReactor Case89

Answers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)(1) Every pipe from upper
plenum to upper head issimulated with a PIPE
connected to a 3D
VESSEL.

(1) CRGTs were modeled with one lumped
flow channel. Flow paths between theupper plenum and the CRGT weremodeled to simulate coolant flow at thepenetration holes and at the bottom.
(2) The same noding model for CRGT isapplied to reactor safety analysis.

(1) Upper plenum was represented by fivevertical nodes considering coolant flowat penetration holes at CRGT bottom.
Core was represented by nine verticalnodes according to 9-step chopped
cosine in axial profile of core power,
and interfacial drag in each node was
reduced to 1/10 [2].
(2) Default heat transfer model used inNEA/CSNI/R(2013)1
No.

Answers by GRS(ATHLET)

Answers by VTT(APROS)against numerous separate effecttests, including several tests in the
full-scale UPTF facility. For thecore region the drift correlation forbundle geometries have been
applied.
(2) Heat transfer correlations used in
the core region are the default
ones, recommended in the code
User's Manual.
(3) The same correlations are applied
for PWR safety analyses.

however, the default
model (combination ofBiasi and ZuberGriffith correlations)
was used for test 6-1.
(3) Default models used,
unless any special
reason to use optionalmodels (sensitivity
analyses).

Answers by AVN(CATHARE)

Answers by DSE-UPM(TRACE)

Answers by JAEA
(RELAP5)
RELAP5/MOD3.2 code is chosen. The
default model consists of Chencorrelation for nucleate boiling, ChenSundaram-Ozkaynak correlation fortransition boiling, and Bromley and
Sun-Gonzales-Tien correlation for film
boiling [3].
(3) The same noding model for upper
plenum and core and default heattransfer model are applied to reactor
safety analysis.

References/1/ G. Lerchl, H. Austregesilo
ATHLET Mod 2.1 Cycle A User's Manual
GRS-P-1, Vol. 1, July 2006.
/2/ H. Austregesilo, C. Bals, A. Hora,
G. Lerchl, P. Romstedt
ATHLET Mod 2.1 Cycle A Models and Methods
GRS-P-1, Vol. 4, July 2006.

References[1] Asaka, H., et al., 1990. Results of
0.5% Cold-Leg Small-break LOCA
Experiments at ROSA-IV/LSTF:
Effect of Break Orientation. Exper.
Therm. Fluid Sci. Vol. 3, pp.588-596.
[2] Kumamaru, H., et al., 1999.
RELAP5/MOD3 Code Analyses of
LSTF Experiments on Intentional
Primary-Side Depressurization
Following SBLOCAs with Totally
Failed HPI. Nucl. Technol., Vol. 126,
pp.331-339.
[3] Carlson, K. E., et al., 1990.
RELAP5/MOD3 code manual (draft).
NUREG/CR-5535, EGG-2596.

90

NEA/CSNI/R(2013)1

APPENDIX 3

LIST OF EXPERIMENTAL DATATABLE A3-1 IS A LIST OF LSTF EXPERIMENTAL DATA FOR THE ROSA PROJECT
DISTRIBUTED BY MEANS OF CD-RS AND DVDS.

Table A3-1 List of experimental data using CD-Rs and DVDs
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14

Test No.

CD-R
Test 1-1 (STNC-34)

19
20
21
22
23
24

DVD

CD-R
Test 1-2 (SBHL-17)

Test 2 (STWH-05,
06,07,08,09,1
0 and 11)

15
16
17
18

Media

Test 3-1 (SBCL-38)

DVD

DVD

Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Image data from video probe in cold leg in loop with PZRImage data from video probe in cold leg in loop without PZR
Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Image data from video probe in cold leg in loop with PZRImage data from video probe in cold leg in loop without PZRTest data of ST-WH-05
Test data of ST-WH-06
Test data of ST-WH-07
Test data of ST-WH-08 and 09

CD-R

Test data of ST-WH-10 and 11
Measured test data of all instruments

DVD

Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR

CD-R
Test 3-2 (TRLF-13)

Contents

DVD

Image data from video probe in cold leg in loop without PZR
Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Image data from video probe in cold leg in loop with PZRImage data from video probe in cold leg in loop without PZR

91

NEA/CSNI/R(2013)1

No.
25
26
27
28
29
30
31

Test No.
Test 4-1 (STNC-39)
Test 4-2 (STNC-40)

Test 5-1 (SBCL-39)

32
33
34
35
36
37
38
39

Test 5-2 (SBCL-40)

Test 6-1 (SBPV-09)

40
Test 6-2 (SB41PV-10)
42
43
44 Test 6-3 (LB45 CL-05 and 06)
46

Media

Contents

DVD

Measured test data of all instruments

DVD

Measured test data of all instruments

CD-R

Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Image data from video probe in cold leg in loop with PZRImage data from video probe in cold leg in loop without PZR

DVD

CD-R

Measured test data of all instruments

DVD

Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Image data from video probe in cold leg in loop with PZRImage data from video probe in cold leg in loop without PZR

CD-R
DVD
CD-R
DVD
CD-R
DVD

Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Measured test data of all instruments
Image data from video probe in hot leg in loop with PZRImage data from video probe in hot leg in loop without PZR
Measured test data of all instruments in LB-CL-05
Measured test data of all instruments in LB-CL-06
Image data of high-speed video camera in LB-CL-05
Image data of high-speed video camera in LB-CL-06

Each data CD-R contains followings files other than LSTF data file(s).
(1) A folder BIN that contains lstfread.exe file
This file read data in CD-R, extract data and create data table in MS Excel format.
(2) CDread.bat file
This file activates lstfread. exe in BIN folder and specify CD-R drive name.
For example, current file specifies f drive. Please edit this bat file to change this CD-R drive to be
specified by CDread.bat file.
Please note that the default drive for lstfread.exe is d. You do not need to use CDread.bat file when the
CD-R drive of your computer is d.
(3) TAGLIST.xls file92

NEA/CSNI/R(2013)1

This is a table of all the LSTF data. Please consult this table to seek necessary Function ID and LSTF
System Description Report that indicates the location of sensors.
(4) Readme(HowToUse).doc file

The data reader program “lstfread. exe” converts the data file(s) and creates a csv file of LSTF data by thefollowing procedures.
(1) Prepare a folder named ‘output’.
(2) Copy the folder BIN and CDread.bat file into a work directory the same as the folder ‘output’.
(3) Insert LSTF Data CD-R and confirm RUN ID and Function IDs you want to obtain data. ‘Function
IDs’ are listed in ‘TAGLIST.xls’ file that is included in each of LSTF data CD-R. The contents of‘TAGLIST.xls’ file may change according to the change in the measurement instrumentation(4) Double click CDread.bat to activate it.
(5) Input several parameters requested as follows;(Example)
RUN ID : SP9<== FIX

Number of Data (MAX 10) : 4<== Arbitrary

Function ID :
TE514
PE13
LE3
RC2
Start Time : 0.0<== Arbitrary

End Time : 100.0<== Arbitrary

This Example creates a data table file of Test 6-1 (abbreviated as SP9) in csv format that contains four
data (TE514, PE13, LE3, RC2) for the same measurement time span.

(6) Program starts the preparation of a csv file into the folder ‘output’.

93NEA/CSNI/R(2013)1

94

NEA/CSNI/R(2013)1

APPENDIX 4

LIST OF PUBLISHING

Tables A4-1 and A4-2 are lists of publishing papers and reports respectively based on ROSA Project.
Table A4-1 List of publishing papers
No.

Paper

Distribution Date
F: file1

A. Jasiulevicius et al. of PSI, “SIMULATION OF OECD/NEA TEST 6.2 USINGTRACE,”, ICONE15

F: March 1, 2007

2

3

4

5

M. Suzuki et al., “Effects of three-dimensional steam flow on exit temperaturesduring core boil-off in PWR vessel top break LOCA simulation experiments byusing ROSA/LSTF”, 2007 Fall Meeting of Atomic Energy Society of JapanT. Watanabe,”Simulation of Temperature Stratification during ECCS Water
Injection using FLUENT”, NURETH-12
T. Takeda et al.,”RELAP5 Analysis of ROSA/LSTF Vessel Upper Head Break
LOCA Experiment”, NURETH-12
H. Nakamura et al., “OECD/NEA ROSA Project (1) Current Status of The ROSA
Project”,
M. Kondo et al., “OECD/NEA ROSA Project (2) Condensation-induced water
hammer test employing system pressure as parameter”,
H. Asaka et al., “OECD/NEA ROSA Project (3) High-power Natural Circulation
Experiment and RELAP5 Post-test Analysis”,
Y. Maruyama et al., “OECD/NEA ROSA Project (4) Pre-Test Analysis forSuperheated Steam Natural Circulation Experiments and Verification ofPulsed-Wire Velocity Meter”,
T. Takeda et al., “OECD/NEA ROSA Project (5) Experiment on PWR PV Bottom
Break LOCA and Post-test Analysis”,
All for 2008 Annual Meeting of Atomic Energy Society of Japan
M. Suzuki et al., “Performance of Core Exit Thermocouple for PWR AccidentManagement Action in Vessel Top Break LOCA Simulation Experiment at
OECD/NEA ROSA Project”, ICONE16
H. Nakamura et al., “RELAP5/MOD3 Code Verification through PWR Pressure
Vessel Small Break LOCA Tests in OECD/NEA ROSA Project,”, ICONE16

F: June 14, 2007

F: June 14, 2007

F: Feb. 8, 2008

F: Feb. 8, 2008

6

S. Gallardo et al. of UPV, “Post test calculation of OECD/ROSA Test 6.2,”,
PHYSOR2008

F: Apr. 15, 2008

7

S. Gallardo et al. of UPV, “OECD/ROSA Tests 6.1 and 6.2 with TRACE5,”, CAMP
meeting

F: Apr. 28, 2008

8

T. Takeda et al., “RELAP5 analysis of OECD/NEA ROSA Project experiment
simulating a PWR small break LOCA with high-power natural circulation,”,PHYSOR2008

F: May 8, 200895

NEA/CSNI/R(2013)1

Distribution Date
F: file

No.

Paper

9

T. Watanabe et al., “OECD ROSA Project temperature stratification experimentand CFD analysis”,
T. Takeda et al., “ROSA Project Experiment on Loss-of-feedwater without Scram
and RELAP5 Post-test Analysis”,
Both for 2008 Fall Meeting of Atomic Energy Society of Japan

F: June 24, 2008

10

V. M. Martínez et al. of Technical University of Catalonia, “Code Validation and
Scaling of the ROSA/SLTF Test 3-1 Experiment,”, TOPSAFE2008 meeting

F: July 4, 2008

11

H. Nakamura et al., “Overview of Recent Efforts in ROSA/LSTF Experiments”,
NUTHOS-7

F: Sept. 8, 2008

12

T. Takeda et al., “RELAP5 POST-TEST ANALYSES OF OECD/NEA ROSAPROJECT
EXPERIMENTSON
STEAM
GENERATOR
DEPRESSURIZATION WITH OR WITHOUT NON-CONDENSABLE GASINFLOW”, NTHAS-6

F: Sept. 8, 2008

13

T. Takeda et al., “Plan of ROSA Project Experiment on Steam Condensation during
Large-break LOCA”, 2009 Annual Meeting of Atomic Energy Society of Japan

F: Jan. 23, 200914

T. Farkas and I. Tóth of KFKI, “FLUENT ANALYSIS OF A ROSA TWO-PHASESTRATIFICATION TEST,”, ICONE17

F: Feb. 12, 200915

T. Watanabe and H. Nakamura, “CFD analysis of temperature stratification
experiment in OECD/NEA ROSA project,” NURETH-13--

16

S. Gallardo, V. Abella and G. Verdú of UPV, “FLUENT OECD/NEA ROSA Project
Test 3-2. Simulation with TRACE5,”, CAMP Meeting

F: June 8, 200917

Y. Maruyama et al., “Experiment on Superheated Steam Natural Circulation in
ROSA Project -Characterization of Natural Circulation-,
T. Takeda et al., “LSTF ROSA Project Experiment on Steam Condensation on
ECCS Water during PWR Large-break LOCA”,
Both for 2009 Fall Meeting of Atomic Energy Society of Japan

F: July 22, 200918

J. Freixa and A. Manera, “Analysis of an RPV Upper Head SBLOCA at the ROSA
Facility using TRACE,“ Nucl. Engng. and Des. 240(7) (2010) 1779-1788--

96

NEA/CSNI/R(2013)1
Table A4-2 List of publishing reports
No.
1

2

3

4

5

6

Distribution Date
F: file, P: printed

Report
The ROSA-V Group, “ROSA-V Large Scale Test Facility (LSTF) System
Description for The Third and Fourth Simulated Fuel Assemblies,” JAERI-Tech
2003-037 (March 2003)
Thermohydraulic Safety Research Group, “A Report on RELAP5 Code Pre-test
Analyses to Define LSTF PV-break SBLOCA Experiment Conditions” (Sept. 12,
2005)
Thermohydraulic Safety Research Group, “Brief Description of Computer CodeInputs for ROSA/LSTF Analyses using RELAP5/MOD3 and TRAC-PF1/MOD1
Codes” (June 17, 2005)
Thermohydraulic Safety Research Group, “A Report on RELAP5 Code Pre-test
Analyses to Define ROSA/LSTF Experimental Conditions of PV Bottom LOCA
Experiment” (November 25, 2005)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 6-1 (1.9% Pressure Vessel Upper-head Small Break LOCA Experiment SBPV-09 in JAEA),” (February 16, 2006)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 6-2 (0.1% Pressure Vessel Bottom Small Break LOCA Experiment SB-PV-10
in JAEA),” (March 15, 2006)

P: before PRG1

F: Sept. 12, 2005

F: Oct. 14, 2005

F: Nov. 25, 2005
F: Feb. 17, 2006
P: Oct. 26, 2006
F: March 16, 2006
P: Oct. 26, 2006

7

Thermohydraulic Safety Research Group, “Heat Loss Characteristics of the
ROSA/LSTF System,“ (Aug. 7, 2006)

F: Aug. 16, 2006

8

Thermohydraulic Safety Research Group, “Proposed
Temperature Stratification Experiment,” (Aug. 7, 2006)

F: Aug. 16, 2006

9

Thermohydraulic Safety Research Group, “A Report on Pre-test Analyses andPreparatory Experiments to Define Experimental Conditions of ROSA/LSTF Test
3-1 on High Power Natural Circulation,” (Sept. 22, 2006)

F: Sept. 22, 2006

10

Thermohydraulic Safety Research Group, “Final Data Report of ROSA/LSTF Test
6-1 (1.9% Pressure Vessel Upper-head Small Break LOCA Experiment SB-PV-09
in JAEA),” (Dec. 1, 2006)

F: Dec. 5, 2006
P: ---

11

Thermohydraulic Safety Research Group, “General questionnaire on the analysis
methods for Test 6-1 -- tentative for discussions --,” (Jan. 10, 2007)

F: Jan. 10, 2007

12

Thermohydraulic Safety Research Group, “Proposed Test Conditions andProcedures of Condensation-Induced Water Hammer Experiment,” (Jan. 10, 2007)

F: Jan. 10, 2007

13

14

15
16

Test

ProceduresofThermohydraulic Safety Research Group, “Final Data Report of ROSA/LSTF Test
6-2 (0.1% Pressure Vessel Bottom Small Break LOCA Experiment SB-PV-10 in
JAEA),” (Jan. 17, 2007)
Thermohydraulic Safety Research Group, “Quick-Look Data Report of OECD/NEA
ROSA Project Test 1-1 (ECCS water injection under natural circulation
condition: ST-NC-34 in JAEA),” (Jan. 30, 2007)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 3-1 (High Power Natural Circulation Experiment SB-CL-38 in JAEA),”(March 20, 2007)

F: Jan. 31. 2007
P: ---

Thermohydraulic Safety Research Group, “Proposed Test
Temperature Stratification Experiment,” (March 23, 2007)

F: March 23, 2007

97

Proceduresof

F: Feb.13, 2007
P: May 30, 2008

F: March 20, 2007
P: ---

NEA/CSNI/R(2013)1

Distribution Date
F: file, P: printed

No.

Report17

Thermohydraulic Safety Research Group, “Performance of Core Exit Temperatures
for Accident Management Action in LSTF 1.9% Top Break LOCA Test
(Supplemental Report for Test 6-1),” (May 25, 2007)

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

Thermohydraulic Safety Research Group, “A Report on Pre-test Analyses to DefineExperimental Conditions of ROSA/LSTF Test 5-1 on Primary Cooling throughSteam Generator Secondary-side Depressurization,” (May 24, 2007)
Thermohydraulic Safety Research Group, “Quick-Look Data Report of OECD/NEA
ROSA Project Test 1-2 (1% hot-leg break LOCA experiment with HPI: SB-HL-17
in JAEA)” (Sept. 3, 2007)
Thermohydraulic Safety Research Group, “A Report on Pre-test Analyses to DefineExperimental Conditions of ROSA/LSTF Test 5-2 on Primary Cooling throughSteam Generator Secondary-side Depressurization” (Sept. 5, 2007)
Thermohydraulic Safety Research Group, “Performance of Core Exit Temperatures
for Accident Management Action in LSTF 1.9% Top Break LOCA Test
(Supplemental Report for Test 6-1)” (Sept. 19, 2007)
Thermohydraulic Safety Research Group, “A Report on Pre-test Analyses to DefineExperimental Conditions of ROSA/LSTF Test 3-2 on High Power Natural
Circulation” (Oct. 9, 2007)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 5-1 (Primary Cooling through Steam Generator Secondary-side
Depressurization SB-CL-39 in JAEA),” (Nov. 30, 2007)
Thermohydraulic Safety Research Group, “Final Data Report of ROSA/LSTF Test 31 (High Power Natural Circulation Experiment SB-CL-38 in JAEA),” (Jan. 9,
2008)
Thermohydraulic Safety Research Group, “Final Data Report of OECD/NEA ROSA
Project Test 1-1 (ECCS water injection under natural circulation condition : STNC-34 in JAEA),” (Jan., 2008)
Thermohydraulic Safety Research Group, “Results of Preliminary Test on
Superheated Steam Natural Circulation in Framework of OECD/NEA ROSA
Project,” (Feb. 7, 2008)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 3-2 (High Power Natural Circulation Experiment TR-LF-13 in JAEA),” (Apr.
11, 2008)
Thermohydraulic Safety Research Group, “Quick-look Data Report of ROSA/LSTF
Test 5-2 (Primary Cooling through Steam Generator Secondary-side
Depressurization Experiment SB-CL-40 in JAEA),” (Apr. 11, 2008)
Thermohydraulic Safety Research Group, “Quick-look Data Report of OECD/NEA
ROSA Project Test 2 (Condensation-induced Water Hammer Tests : ST-WH05,06,07,08,09,10 and 11 in JAEA),” (Mar. 14, 2008)
Thermohydraulic Safety Research Group, “Final Data Report of OECD/NEA ROSA
Project Test 1-2 (1% hot-leg break LOCA experiment with HPI: SB-HL-17 in
JAEA),” (July 3, 2008)
Thermohydraulic Safety Research Group, “Final Data Report of ROSA/LSTF Test 51 (Primary Cooling through Steam Generator Secondary-side Depressurization
Experiment SB-CL-39 in JAEA),” (July 3, 2008)
Thermohydraulic Safety Research Group, “Final Data Report of OECD/NEA ROSA
Project Test 3-2 (High Power Natural Circulation Experiment TR-LF-13 in
JAEA),” (Jan. 21, 2009)

98

F: May 25, 2007
P: May 30, 2008
F: May 25, 2007

F: Sept. 4, 2007

F: Sept. 5, 2007

F: Sept. 19, 2007

F: Oct. 9, 2007

F: Nov. 30, 2007

F: Jan. 15, 2008

F: Jan. 20, 2008

F: Feb. 7, 2008

F: Apr. 11, 2008

F: Apr. 11, 2008

F: Apr. 14, 2008

F: July 3, 2008

F: July 3, 2008

F: Jan. 22, 2009

NEA/CSNI/R(2013)1

Distribution Date
F: file, P: printed

No.

Report33

Thermohydraulic Safety Research Group, “Final Data Report of OECD/NEA ROSA
Project Test 5-2 (Primary Cooling through Steam Generator Secondary-side
Depressurization Experiment SB-CL-40 in JAEA),” (Jan. 21, 2009)

34

35

36

37

38

39

40

Thermohydraulic Safety Research Group, “Quick-Look Data Report of Test 4-1(Designated JAEA Internal Run ID of ST-NC-39) Natural Circulation ofSuperheated Steam in Primary Loops during High Pressure Sequences of LWR
Severe Accidents with Argon Gas as Simulated Gas,” (May 28, 2009)
Thermohydraulic Safety Research Group, “Quick-Look Data Report of Test 4-2(Designated JAEA Internal Run ID of ST-NC-40) Natural Circulation ofSuperheated Steam in Primary Loops during High Pressure Sequences of LWR
Severe Accidents,” (June 30, 2009)
Thermohydraulic Safety Research Group, “Quick-Look Data Report of Test 6-3(Steam Condensation Experiment during Large-break LOCA of PWR LB-CL-05
and LB-CL-06 in JAEA),” (Sept. 1, 2009)
Thermohydraulic Safety Research Group, “Final Data Report of OECD/NEA ROSA
Project Test 2 (Condensation-induced Water Hammer Tests : ST-WH05,06,07,08,09,10 and 11 in JAEA),” (Aug. 11, 2009)
Thermohydraulic Safety Research Group, “Final Data Report of Test 4-1(Designated JAEA Internal Run ID of ST-NC-39) Natural Circulation ofSuperheated Steam in Primary Loops during High Pressure Sequences of LWR
Severe Accidents with Argon Gas as Simulated Gas,” (May 6, 2010)
Thermohydraulic Safety Research Group, “Final Data Report of Test 4-2(Designated JAEA Internal Run ID of ST-NC-40) Natural Circulation ofSuperheated Steam in Primary Loops during High Pressure Sequences of LWR
Severe Accidents,” (May 6, 2010)
Thermohydraulic Safety Research Group, “Final Data Report of Test 6-3 (Steam
Condensation during Large-break LOCA of PWR LB-CL-05 and LB-CL-06 in
JAEA),” (June 30, 2010)99

F: Jan. 22, 2009

F: June 2, 2009

F: July 6, 2009

F: Sept. 2, 2009

F: Oct. 10, 2009

F: May 19, 2010

F: May 19, 2010

F: July 1, 2010