# International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)

Authors: N. Aksan
Journal: Science and Technology of Nuclear Installations
 (2008)

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## Abstract

Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of
principal working group no.2 on coolant system behaviour (PWG2) and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium) were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects. ISPs are providing unique material and benefits for some safety-related issues.


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## Body

## 1. Introduction

Large transient thermal-hydraulic system codes are widely used to
perform safety and licensing analyses of nuclear power plants and also used in
the design of advanced reactors. Evaluation of the capabilities and the
performance of these codes can be accomplished by comparing the code
predictions with measured experimental data obtained on different test
facilities. In this respect, parallel to other national and international
programmes, OECD Nuclear Energy Agency (OECD/NEA) Committee on the Safety of
Nuclear Installations (CSNI) has promoted, over the last thirty years some
fourty eight international standard problems (ISPs) [1, 2]. The first international
standard problem (ISP) was organized in 1975 on the famous “Edwards blowdown
pipe” experiment. These ISPs were performed in different fields as
in-vessel thermal-hydraulic behaviour, fuel behaviour under accident
conditions, fission product release and transport, core/concrete interactions,
hydrogen distribution and mixing, and containment thermal-hydraulics. Roughly, 60% of these
ISPs concerned the thermal-hydraulic behaviour.The main goal of ISP exercises is to
increase confidence in the validity and the use of the different
tools that are used in assessing the safety of nuclear installations. These
tools may vary to some extent in different
countries and are extremely complex. Therefore, the ISPs were considered as an effective way to get a common understanding
and judgment about the code/user capabilities on an international basis.
Indeed, in an ISP the predictions of
different computer codes with respect to a given physical problem may be compared with the results of an experiment
or/and among each other.While the developmental
assessment still belongs to the organisation developing the codes, ISP exercises can be considered as a
complementary activity, assessing the codes through the analysis of experts different from the code developers and
covering much wider ranges, specifically in terms of thermal-hydraulics scenarios and value of
parameters.The objectives of the ISP may be summarized as(i)
to contribute to better understanding of postulated events,(ii)
to compare and evaluate the capability of codes (mainly best
estimate codes),(iii)
to suggest improvements to the code developers,(iv)
to improve the ability of code users,(v)
to address the so called scaling effect. Standard problems are performed as “open” or “blind” (double blind)
problems. In an “open” problem, all participants know the
results of the experiment in detail before performing their calculations.
In a “blind” exercise, the results are locked until the code
users submit the calculation results for comparisons. A so called “double blind” exercise
consists of a “blind” one for which no other experimental
data related to the test facility has been published or made available to the
ISP participants before submission of
results. For blind exercises the participants are keenly encouraged to
run post test calculations when the experimental results are released. Those
post test calculations are
sensitivity studies, where various options and/or models are tested in order to
see how they affect the results, also to better understand the reasons
for eventual discrepancies resulting from
comparing “blind” results and experimental data.As mentioned in [3], both integral and separate effect experiments may be
considered for ISP exercise. Also best-estimate codes are
preferably used. The reader will also find in the same reference a
complete description of the organisation of an ISP exercise.A global review and synthesis on the
contribution that small break LOCA ISPs have made to address nuclear reactor
safety issues was initiated by the principal
working group no. 2 (PWG2) in September 1993. Further to this request of the PWG2, an action has been put, during
the thirteenth meeting of the Task Group on Thermal-Hydraulic System Behaviour (TG-THSB), to carry out this review and synthesis
work on previous small break LOCA ISPs. As a result of this synthesis
work, a short overview report was written on this subject [4] by a group of
experts in the TG-THSB. In order to limit
the effort, five ISPs were selected for this evaluation, but not strictly based on small break LOCA scenarios; ISPs in which
similar phenomenon to small break LOCA was observed are also considered
(i)
ISP 18:
LOBI Mod2 1% small break LOCA [5];(ii)
ISP  20: Doel 2 steam
generator tube rupture event [6];(iii)
ISP  22: SPES-simulating
loss of feedwater transient in Italian PWR [7, 8];(iv)
ISP 26: ROSA-IV LSTF 5% cold leg small break LOCA experiment
[9];(v)
ISP 27:
BETHSY 0.5% small break LOCA with loss of high-pressure injection [10]. The ISPs 18, 22, and 27 were “blind” exercises, while the ISPs 20 and 26 were “open” ones. The
ISP 18 is the “oldest” ISP retained
in this review and synthesis work, since such an ISP may
be considered as a milestone in the transition process between the first
generation codes (i.e., RELAP4) and the new generation
of advanced computer codes (e.g., TRAC, RELAP5, ATHLET, CATHARE). It is to be noted that there
were small break LOCA ISP exercises previous to ISP-18, for example, LOFT and semiscale
small break LOCA tests, but they were not considered in this review process due
to advancement of the codes relative to the application of the first generation
codes in these ISPs. Moreover, at that time some of these new
codes were in their development phase. In addition, one may consider
that, since 1985, the objectives of ISP were slightly changed due to the reason
that all codes passed their developmental phase.While the ISP 22 initiating event is not a small break, it has been
considered in this evaluation since specific phenomena observed during the
experiment are similar to those observed during small
break accident. Moreover, it might give the opportunity to fill the gap between
BETHSY and LOBI test facilities for scaling purposes.ISP 20 has been retained in this evaluation as far as scaling effect has to be addressed. Indeed, the ISP 20 is the unique
exercise based on a transient occurring in a full-scale two-loop PWR nuclear
plant.Other internationally conducted research programmes in this
same area have been completed in the time period here considered,
including ISPs, for example, ISP 25 and ISP 33. Examples are the OECD-LOFT project
or LOBI experiments analyzed by a CEC devoted task group. However, resources limitations
and willingness to keep some homogeneity for the discussed transients (i.e.,
ISP 25 is based on a separate effects test, ISP 33 addressed
the behaviour of WWER plants; LOFT is a nuclear facility scaled
down with criteria different from those of LOBI, SPES, BETHSY, and LSTF; in
addition most of the LOFT, LOBI, and LSTF data were not openly
available to the whole OECD community)
supported the conclusion to restrict the investigation range,
though recognizing the fundamental contributions given by the above mentioned
programmes in this same area.The outcome from each considered ISP and in particular the evaluation of the
comparisons between
measured and predicted system behaviours are described in detail in the “final
comparison reports,” from [5] to [10], and therefore will not be repeated here. Identically, this
synthesis work will not deal with
the “user effects” that has been separately addressed and analyzed in
detail in [11].In this paper, some of the aspects addressed in [4] will be summarized in order to provide an overview on the lessons
learned from the small break LOCA ISPs. Section 2 will give an overview on the
development of small break LOCA issue. Main phenomena and relevance of small
break LOCA to reactor safety in a PWR are shortly described in Section 3. A
short overview of ISPs and expected technical findings are dealt within
Sections 4 and 5. After a presentation of the involved facilities and plant and
a description of the different selected
tests (Section 6); Section 7 deals with relevant ISP statistics. Section 8
presents the “lessons learned” from the selected ISP activities with some conclusions and
recommendations. This also constitutes the main objective of the presented activity.


## 2. Origin of Small Break LOCA Issue  (System Thermal-Hydraulics  before and after TMI-2)

In early 1970s, former US
Atomic Energy Commission convened a public hearing to explore the safety
question in relation to the effectiveness of systems to mitigate the
consequences of a loss of coolant accident in a nuclear reactor, in case it
happens. Ultimately, after extensive public hearings, in 1974, the interim
regulations were modified to provide a set of specific requirements for
computer codes for ECCS analyses in and a new section, 10 CFR 50.46 [12, Appendix K], requiring ECCS
meet established standards. This included a definition that LOCAs are
hypothetical accidents that would result from the loss of reactor coolant, at a
rate in excess of the capability of the reactor coolant makeup system, from
breaks in pipes in the reactor coolant pressure boundary up to and including
break equivalent in size to the double-ended rupture of the largest pipe in the
reactor coolant system. The safety criteria prescribed in 10 CFR 50.46 are
applicable to both large and small break LOCAs. That is to say the limits on
peak cladding temperature, cladding oxidation, and hydrogen generation must not
be exceeded in a design basis accident. Calculations of ECCS performance using
the conservative prescriptions of [12, Appendix K] resulted in the large break LOCA generally being the
most limiting accident. At the time, there was a major safety research programme
to support code development for large break LOCA and also some limited work on
small break LOCA.The March 1979 accident at the
Three Mile Island Unit 2 (TMI-2) reactor led to an extensive reorientation of
light water reactor safety research programmes and also regulatory changes. The
TMI-2 accident was a small break LOCA, an event given significantly less
attention because of the major emphasize on the large break LOCA at the time.
Consequent to TMI-2, small break LOCA and plant operational transients received
major attention. The experimental simulation of the natural circulation
phenomena in the primary loops, including those in the two-phase stratified and
counter-current flow regimes, is of primary importance to the thermal-hydraulic
response of a nuclear power plant during such transients. Since these phenomena
are significantly dependent on facility scale and geometry, large-scale tests
for a primary system geometry representative of operational nuclear power
plants are required. Either operational facilities were modified to carry out
small break LOCA experiments or there were new facilities designed and
constructed (see Section4). It is to be noted that unlike the large break
LOCA, the sequence of events following a small break LOCA can evolve in a
variety of ways. Operator actions, reactor design, ECCS set points, break size,
and location will have a bearing how the small break LOCA scenario unfolds.
Therefore, in order to predict the integral system behaviour during a small
break LOCA, a best-estimate code must have sufficient modelling capabilities
to take these factors into account. These codes are also needed to be assessed
against integral system tests. After having been successfully assessed against
data from a large number of scaled test facilities, best-estimate codes become
the ultimate repository of all previous thermal-hydraulic safety research. ISP
activities are a part of this process (see Section 4).


## 3. Small Break LOCA in a PWR with Relevance  to Nuclear Reactor Safety and  Main Phenomena

The major characteristic difference between a
small break and a large break LOCA is in the rates of coolant discharge and
pressure variations with time. In general, small break LOCAs are characterized
by an extended period (this can be tens of minutes to several hours at the
lower end of the break spectrum) after the occurrence of the break, during
which the primary system remains at a relatively high pressure and the core
remains covered. As soon as the pumps are tripped, either automatically or
manually, gravity-controlled phase separation occurs and gravitational forces
dominate the flow and distribution of coolant inside the primary system. The
subsequent sequence of events, whether or not the core uncovers and is
recovered or reflooded, depends not only on the location, shape, and size of
the break, but also on the overall behaviour of the primary and secondary
systems. This behaviour is strongly influenced by both automatic and operator
initiated mitigation measures. In general, the reactor system response to a
small break is slower compared to events after a large break. This allows more
time, and different possibilities, for operator interventions. Another
principal difference is the domination of gravity effects in small breaks
versus inertial effects in the large breaks.It is to be noted that there is no unique path
of development of events following a small break LOCA in PWRs. The scenarios
may change drastically by many factors such as the reactor design (e.g., U-tube
or once-through steam generators, such as TMI-2), the break size, the core
bypass size (allowing some fraction of the inlet cold leg flow directly into
the core upper structure without passing through the core), and most
importantly, by different operator interactions. As an example, the primary
circulation pumps may be shut down early in a small break LOCA transient or
they may be allowed to run and circulate the coolant through the core for a
long time. These alternative actions can make a large difference in the nature
of discharge flow, early heat removal from the core, and the liquid inventory
in the system after one hour or so in the transient. Another important
possibility of different interactions is through the steam generators. The
secondary side of steam generators can be isolated (no feed water flow) or they
can be used for a controlled heat removal. It is also possible to cool the
reactor through the so-called “feed and bleed” process (on the primary side).
Either of these actions will have a major effect on the course of the
transient. It is not the intent in this section to provide a catalogue of all
possible scenarios following small break LOCA accidents. But it is important to
note that an adequate set of modelling capabilities for any of the plausible
scenarios will be equally adequate for all other relevant scenarios. This is because
the phenomena and processes are the same but their interactions and timing of
various developments change in different operations. Therefore, in order to
predict the integral system behaviour during a small break LOCA, a
best-estimate code must have sufficient modelling capabilities to take these
factors into account.During a PWR small break LOCA, there is the
potential for three distinct core heat ups. The first heat up is caused by loop
seal formation and the manometric core liquid level depression. Naturally
occurring events including loop seal clearing and break uncovery mitigate this
heat up. The second heat up occurs following the core quench caused by loop
seal clearing and is caused by a simple core boiloff. During this period the
primary pressure is decreasing to the accumulator set point and the steam
produced by the core boiloff leaves the system via the break. Any heat ups that
occur during this period are mitigated by the reflood from the accumulator
water. The third possible heat up can occur following depletion of the
accumulator tanks and before LPIS injection begins. One drawback to the reflood
process accompanying the accumulator injection is a decrease in the ongoing
depressurisation process such that another possible heat up occurs before the
LPIS primary pressure set points are reached and long-term cooling is provided.
Various factors affect the magnitudes of the three potential core heat ups.
Some examples are break size, break direction and location, availability of
HPIS, and the degree of upper head to downcomer bypass flow. Although the
magnitudes of the core heat ups may vary, ECCS performance must be such that
the criteria, for example, 10 CFR 50.46 [12] is not exceeded.The interested readers can obtain further
details on small break LOCA in [13].


## 4. A Short Overview of ISPs and Technical  Domains Covered by Them

A compilation of all ISPs performed between 1975
and 1997 can be found with a brief description of each ISP in [1] and an
extended list of ISPs (from 1975 to 2007) is also provided in Table 1.Table 1
List of CSNI international standard problems (ISPs) [2].
No.Completion dateTitle11975Standard problem 1-Edwards pipe blowdown test21975Analysis of semiscale blowdown test 11, LB LOCA31977CSNI standard problem 3; comparison of LOCA analysis codes, CISE, blowdown41978United states  standard problem 6 and international standard
problem 4: comparison of the standard problem calculations with measured
experimental data for semiscale test S-02-6, SB LOCA51979United states  standard problem 7 and international standard
problem 5: final comparison report on LOFT test L1-4, LB LOCA61978ISP-6: calculations comparison report-determination of water level
and phase separation effects during the initial blowdown phase71979comparison report on OECD-CSNI LOCA standard problem no. 7: analysis
of a reflooding experiment, ERSEC81979Semiscale MOD1 test S-06-03 (LOFT counterpart test), LB LOCA91981LOFT test L3-1 preliminary comparison report, SB LOCA101981comparison report on OECD-CSNI LOCA standard problem no. 10:
“refill and reflood experiment in a simulated PWR primary system (PKL)111984LOFT L3-5 and L3-6 comparison
reports, SB LOCA121982ROSA-III 5% small break
test, Run 912, BWR-SB LOCA131983international standard problem 13 (LOFT experiment L2-5) preliminary
comparison report, LB LOCA141985behaviour of a fuel bundle simulator during a specified heatup and
flooding period (REBEKA experiment) (results of posttest analyses)151983LOCA experiment at FIX-II facility, BWR161985rupture of a steam line within the hdr containment leading to an early
two-phase flow: results of posttest analyses: final comparison report171984Marviken BWR standard
problem181987LOBI-MOD2 small break LOCA experiment
A2-81191987behaviour of a fuel rod bundle during a large break LOCA transient
with a two-peaks temperature history (PHEBUS Experiment): final comparison
report201988Doel 2 steam generator tube rupture event: final report211989PIPER-ONE experiment PO-SB-7: simulation of small and intermediate
break LOCA for BWRs221990SPES-loss of feedwater transient in Italian PWR. final comparison
report (1990) and evaluation of posttest analyses (1992).231989Rupture of a large diameter pipe in the HDR containment241989SURC-4-core-concrete interaction test251991ACHILLES-N2 injection from accumulators and faster (best estimate)
reflood rates261992ROSA-IV LSTF-cold-leg small-break LOCA experiment271992BETHSY-small break LOCA with Loss of HP injection281992PHEBUS SFD B9+-experiment on the degradation of a PWR type core291993HDR experiment E11.2-hydrogen distribution inside the HDR
containment under severe accident conditions: final comparison report301992BETA II core-concrete interaction experiment (Test V5.1): comparison
report311993CORA-13 experiment on severe fuel damage32—FLHT-6 experiment, cancelled331994PACTEL-VVER-440 natural circulation stepwise coolant inventory
reduction341994falcon experiments FAL-ISP-1 and FAL-ISP-2, fission product transport351994NUPEC hydrogen mixing and distribution test M-7-1: final comparison report361996CORA-VVER severe fuel damage experiment (test W2)371996VANAM M3-a multi compartment aerosol depletion test with hygroscopic
aerosol material-comparison report381997loss of the residual heat removal system during mid-loop operation (BETHSY)391997fuel coolant interaction and quenching (FARO)401999STORM test SR11-aerosol deposition and resuspension in the primary
circuit411999RTF experiment on iodine behaviour in containment under severe
accident conditions422003PANDA tests (six different phases) related to passive safety systems
for advanced light water reactors432001UMCP boron dilution test442002Four open and one blind KAEVER aerosol depletion tests with three
differently soluble materials and uniform thermal-hydraulic conditions with
slight volume condensation452003QUENCH-06, fuel rod bundle behaviour up to and during reflood/quench
(severe core damage)462004PHEBUS in reactor experiment (FP-1) on the degradation, fission
product release, circuit and containment behaviour following overheating of
an irradiated fuel rod bundle472005Based on experiments performed in the TOSQAN, MISTRA and ThAI
facilities for containment thermal-hydraulics (e.g., gas distribution,
natural convection, heat and mass distribution…)482005Containment capacity (integrity and ageing of components and
structures). 1 : 4 scale model of a prestressed concrete containment vessel
(PCCV) of a nuclear power plant (SANDIA II mock-up)containment analysis standard problems (CASPs)CASP-11980Comparison report on OECD-CSNI containment standard problem no. 1:
“steamline rupture within a chain of compartments” (Battelle
Institute test D15)CASP-21982Comparison report on OECD-CSNI containment standard problem no. 2: “water line rupture in a branched compartment chain” (Battelle Institute test
D16)CASP-31983Final comparison report for containment standard problem exercise 3
(australian lucas heights blowdown/containment rig, small-scale
two-compartments basic containment experiment)The very first ISPs from 1975 to roughly 1980 focused on LOCA
thermal-hydraulics as it was one of the main concerns of that time. We find
there ISPs based on separate effects tests (Edwards blowdown pipe, CISE blowdown
test, Battelle blowdown test, tube reflooding test ERSEC) and ISPs based on the
two only available system experiments for PWRs at that time, that is, SEMISCALE
and LOFT.After Three Mile Island (TMI-2) accident, ISPs
started to move from the large breaks to the small breaks. They included ISPs
on LOFT L3 small break LOCA series tests for PWRs, ROSA III, and FIX II tests
for BWRs. Some large break tests were still selected: PKL reflooding test, as
reflooding was considered as a remaining issue; LOFT L2-5, as it was a
significant “concluding” nuclear test for large breaks.During this period (beginning 80s), two ISPs
were initiated in a new domain for ISPs at that time which was the domain of
thermo-mechanical fuel behaviour during LOCA. These were ISPs on REBEKA test
(nonnuclear) and on PHEBUS LOCA test (nuclear).In parallel to the ISPs dealing with the primary
circuit, ISPs (in a first step called CASPs) were organized in the beginning of
the 80s on containment experiments either system experiments (BATTELLE Model
Containment) or very small scale experiment (AAEC-Australia). These ISPs
covered large break situations. They were followed in the mid 80s by ISPs on
HDR containment tests (large break in PWR) and Marviken test (BWR).During the second half of the 80s and during the
beginning of the 90s, the ISPs related to thermal-hydraulics were characterized
by a full and coherent series based on the experiments which were decided and
built after TMI in order to well study small break and transient situations including
operator actions. They included ISPs on LOBI-mod2, SPES, ROSA IV, BETHSY
facilities for PWRs (lessons learned from these ISPs are provided in [4],
summary of which is included in this paper), and PIPER-ONE facility for BWRs.
Besides this series, one ISP investigating the effect of noncondensable gases
on reflood was performed (ACHILLES), and the first and only one ISP based on
real plant was organized in 1988 on the DOEL 2 steam generator tube rupture
event.End of the 80s, the interest of ISPs moved
clearly to the severe accident area. ISPs on core degradation were held based
on CORA (nonnuclear) and PHEBUS SFD (nuclear). Core concrete interaction was
investigated with two ISPs (SURC4 and BETA2). Containment questions and
especially hydrogen problems were the subject of two ISPs based on HDR and one
ISP based on NUPEC test. In addition, an ISP was also organized on FALCON
facility to investigate fission product behaviour with simulants.One of the extensions of domain covered by ISPs
is constituted by the move towards VVER related problems with PACTEL ISP
(thermal-hydraulics) and CORA VVER ISP (Core degradation).In continuation of ISPs on thermal-hydraulics
and severe accident, shut down states are investigated with an ISP on BETHSY
and steam explosions with an ISP on FARO. STORM and RTF experiments provided
data for aerosol behaviour in primary circuit and iodine behaviour in
containment under severe accident conditions. UMCP facility was used to assess boron
dilution models.Recent ISPs are PANDA test with six different
phases related to passive safety systems for advanced light water reactors;
QUENCH-06 and PHEBUS FP-1 tests for severe core degradation; and TOSQAN, MISTRA,
and ThAI facilities for containment thermal-hydraulics.This overview shows the extraordinary large
range of technical domains, which have been covered by ISPs. These domains
reflect of course the successive changes in the area of concern for nuclear
reactor safety research. This demonstrates also that the concept of ISP initiated
in the thermal-hydraulic area and extended to several other technical areas, is
certainly very productive and useful. We will, in the next sections, analyse in
general and also for a specific subject of small break LOCA what are the
outcomes and the benefits produced by this activity and how it may explain its
success.


## 5. The Expected Technical Findings  from ISP Activity

The basic material of the technical findings
from ISP activity is made of the several predictions obtained with several
codes by several code users of a given physical experiment. From these material
different cross-comparisons can be made which we will now review.(i) The first class of comparisons is the comparisons between code predictions and
experimental results. Such comparisons are evidently contributing to the code
assessment. However, some particularities to this contribution should be
emphasized.(a)
This assessment belongs of course to the “independent” assessment.
Considering the generally very large number and very large variety of participants
to ISPs, the “independent” character is certainly one of the most accentuated that we can
afford. For those who are thinking that the independence of assessment is a very
important feature, the results of ISPs are unique.(b)
The number of code calculations in the comparison between
code predictions and
experimental results is certainly the largest that we can imagine on a single
test. Almost no individual can do such work at least because of financial
limitations. Besides this number of calculations, there are numerous
differences in the physical models used in the different codes. The comparisons
with experimental results are then very instructive on the effect of these
models differences on the capabilities to predict the experiment. Often all codes
available in OECD countries (and sometimes in the world) are represented during
the ISP execution. A complete international view is then obtained on the status
of the predictive capabilities of the phenomena studied in the ISP.(c)
It is clear that the large amount of work produced by the participants and by the
organizing country requires that no mistake should be done in the process. As a
consequence, the experimental test must be first very carefully selected.
Therefore, it is very often one of the best and one of the most significant
tests of the experimental programme to which it belongs. The organisation of
the ISP requires also that all necessary information be transmitted to the
participants in a very comprehensive way. Consequently, the organizing country
must do a very high control of test results and of documentation. This last
requirement led particularly the OECD/NEA working groups to define standards
for test documentation. These standards are summarized in the CSNI report no.
17 [3] and have shown to be quite general and useful, in particular, as they
have been used in several other areas than ISP. As the need arises, certain
revisions are introduced into this report. Finally, the efforts made on the
test selection, on the test control and on the test documentation provide most
often a technical quality of very high level to the ISPs activities.(d)
The high-level grade of documentation obtained by following the prescribed standards and the
strict selection of the tests based on their physical and safety significance
make the ISPs tests very good candidates for inclusion in validation matrices.
ISPs tests may often be considered as international reference tests. Their
already wide distribution and their consequent availability is also a favouring
factor for such choices.(ii) The second class of comparisons is constituted by the comparisons between different codes.
It is the common experience of analysts that understanding and analysing the
code responses is a very difficult exercise. Indications are most often
required in order to give directions for the analyst in its search of
understanding the physical models pertinence. A first group of indications is
given by the analysis of the discrepancies between calculations and
experimental results, which has been discussed above. A second group relates to
the discrepancies between the results of different codes. This last group is
often very valuable because the differences of models between the codes can be
quite easily identified. Consequently, the analyst can focus immediately on the
concerned physical models and evaluate their relative capabilities in reference
with the experimental data. By the wide variety of codes used, ISPs give good
opportunities for doing extensive analysis of this kind.(iii) The last category of comparisons, which ISPs allow, is the comparison of the results
obtained with the same code by different users. The major differences between
the calculations with the same code can be mainly attributed to the users of
the code and this effect has been called the “user effect.” Indeed
this effect is a major finding of ISPs activity. It has been discovered very
early by running the very first ISPs on thermal-hydraulics. The development of
thermal-hydraulic advanced codes was expected to decrease this effect, but the
last thermal-hydraulic ISPs have shown that there was still a significant
“user effect” with these advanced codes. Detailed studies of this
effect have been made on different ISPs and especially on ISP 26 [11]. In
addition to the identification of the user effect, ISPs have contributed
largely to its understanding. ISPs are really providing data, which are
absolutely unique on this crucial subject. Even though some suggested ways to
reduce the user effect have been proposed, it remains that we are quite far
from controlling it. This user effect has also appeared as a generic question
and not only in the thermal-hydraulics area where it has been discovered. In
particular the several ISPs, which have been recently performed in the severe
accidents area, have shown the importance of such an effect.In the coming sections, specific analysis and further discussions will be provided on selected
small break LOCA and transient ISPs.


## 6. Outline of Involved Facilities and  Tests for SB-LOCA ISPs

### 6.1. Facilities and Plant Hardware

In this section, information is given concerning some hardware
features that are relevant for the considered ISP tests. Figure1 shows the sketch of LOBI, SPES, BETHSY,
and LSTF facilities and of the Doel plant.Figure 1
Sketch of the facilities considered for the experimental data base evaluation.The relative elevations of important system components like core, steam generators
U-tubes, loop seals can be seen; the number of loops constituting the
system is reported too. The most important design parameters of the
considered facilities and of the plant are given in Table2. All the considered facilities can
operate at the reference plant nominal pressure for both primary and secondary loops. The height scaling ratio is
equal to one in all cases, so the gravity heads are properly simulated. The maximum allowed power is equal to
the reference reactor value multiplied by volume scaling ratio only in the cases of LOBI and SPES. In other cases, a decay
power value is allowed, ranging around 10% of the nominal value. This
scaling limitation prevents, among the other things, the possibility to have simultaneously rightly scaled temperatures
and flowrates in nominal conditions. In these facilities, the choice is
generally made to preserve hot leg fluid temperature during steady state operation, before any transient; alternatively, it
is possible to preserve the cold leg fluid temperature and nominal flowrate (hot leg temperature not preserved); as a
consequence of the former choice, secondary side fluid temperature and pressures must be higher than the
reference plant nominal values (a real plant at hot standby conditions, 10% of nominal power, exhibits the same behaviour, roughly 70 bar at secondary side);
still, primary pumps have not the maximum allowable flowrate and head properly scaled, although in the case of BETHSY, primary pumps have full
flowrate capacity and preserve the head in single phase flow conditions. The different
criteria utilized for the pressurizer result from Figure 1, as well for
defining the minimum elevation of the loop seal. In the facilities (SPES, BETHSY, and LSTF), the L/√D scaling is adopted for the design of hot and cold legs
piping also preserving the volume scaling [14].Table 2
Relevant hardware characteristics of considered PWR simulators and Doel-2 nuclear plant.
QuantitySPESLOBI/Mod2BETHSYLSTF/ROSA IVDOEL-21Reference reactorW-PWRKWU-PWRFRA-PWRW-PWRW-PWRReference reactor power (MWt)277539002700342311872Mximum power (MWt)/% of nominal power9.0/1385.4/1003.0/1010.0/141187.0/1003ReportedKv1/4271/7121/1001/481/14No. of rods97644281064216595Operating pressure of primary loop (MPa)15.15.815.5115.515.56Operating pressure  of
secondary loop (MPa)6.16.546.917.3/7.45.887Primary loop volume (m3)0.6220.6432.888.3168.58No. of U-tubes for
each steam generators13/13/138/2434/34/34141/1413260/32609Internal diameter of U-tubes (mm)15.419.619.719.619.610L/D ratio of Hot leg57.273.1/119.13817.810.6411Total head (“a” in Figure1) mmax16.0816.7218.3418.414.7min15.9116.4716.8716.913.212Linear rod power at 5%
overall power(*)(Kv/m)1.27(°)1.080.860.911.1213ActualKv(+)1/6401/6191/1441/481/2.5(*)The % value is the power related to the reported Kv.(+) Related to LSTF.(°)To compensate heat losses.Nevertheless, the position of the hot leg axis with respect to the top of the active fuel may
be not the same as in the reference nuclear power plant; in BETHSY, this position is preserved with respect to the
reference reactor, as well the bottom line of the cold leg elevation to the
bottom of active fuel, this leads to different elevations for hot
and cold leg axes. For all the multiloop facilities, each primary (and secondary)
circuit is equal to the other; thus nearly symmetrical thermal-hydraulic conditions occur in the various loops. An exception
is represented by LOBI, where one loop (intact) simulates three loops of the reference reactor and the other simulates a
single (broken) loop. Hardware parameters like pump geometrical configuration, presence, and characteristics
of bypass flow paths (mostly in the vessel) can play an important role in the considered test scenarios.


### 6.2. Outline of the Experimental Scenarios

The experiments A2-81, SP-FW-02, SB-CL-18, 9.1b, and the SGTR transient,
respectively from LOBI, SPES, LSTF, and BETHSY facilities and Doel
plant (Figure1 and Table 2), were submitted by the facility owner organisations to the CSNI and
were discussed and approved at working group and principal working group levels.
The list of host organisations (i.e., proposing the exercise, writing the final reports, and chairing the workshops) for each ISP, is given
in Table 3. The procedures outlined in [3] for assignments of ISPs have been
generally followed.Table 3
List of host organisations for small break LOCA related ISPs.
ISPHost organisation18JRC
(Ispra)20TRACTEBEL
(Brussels)22ENEA (Rome)26JAERI (Tokai Mura)27CEA/CENG (Grenoble)The main characteristics of the mentioned tests are reported in Table4. The main phenomena occurring during SB-LOCAs are listed in Table 5
[15], making use of a phenomena matrix developed in state-of-the-art report
(SOAR) on emergency core cooling thermal-hydraulics [15]. In the same table, a qualitative evaluation of the
capabilities of facilities is provided, according to three judgment levels. For completeness and in order to
give an example of the possible use of this table, in the last two columns, an overall evaluation of
the Relap5/Mod2 and CATHARE codes in addition to their performances is reported, considering each of the
phenomena listed and the pre- and posttest calculations [15].Table 4
Main characteristics of the considered transient.
ISPFacility/plantTestTypeSecondary sideEmergency systemsRecoveryEnd of testident.significant conditionsin primary side(*)procedure(s)ISP 18LOBIA2-81SBLOCAAb = 1%Amax in cold legImposed 100K/hrHPIS in cold legSecondary
system feed and bleed4500ISP 20DOELSGTR 1979SGTRAb = 0.5(+)% of AmaxEFW and steam
Relief valves activeHPIS in cold leg: pressurizer
sprays and heaters—3000ISP 22SPESSP-FW-02LOFW-loss of feed waterBoildown of secondary side
and EFW active in one loopPressurizer PORV and
heatersEFW in one loop8000ISP 26LSTFSB-CL-18SBLOCAAb = 5% of AmaxSteam relief valves activeAccumulators and LPIS in
cold legRHR actuation1000ISP 27BETHSY9.1.bSBLOCAAb = 0.4% of AmaxEFW and pressure control
activeAccumulators
and LPIS in cold legDepressurizarion of secondary side8000(*)scram is assumed in
all cases following a low pressure signal.(+)rough evaluationTable 5
Suitability of tests facilities, judgment of the experiments, and (example of) evaluation of RELAP5/Mod2 and
CATHARE code capabilities as from [13].
FacilitiesExperimentsPhenomenaSPESLOBI/Mod2BETHSYLSTFISP 18ISP 20ISP 22ISP 26ISP 27RELAP5/Mod2 CodeCATHARE V 1.3Performance(°)Code PerformanceNatural circulation in one-phase flow, primary side++++o−−−−++Natural
circulation in two-phase flow, primary side+++++−o−o++Reflux condenser
mode and CCFLoo++o−+++ooAsymmetric loop
behaviour+o+o−o+o+ooLeak flowo+++−−o++o+Phase separation
without mixture level formation++++−−−−o×+Mixture level and
entrainment in steam generator secondary sideooo+oo+ooooMixture level
and entrainment in the coreoo+−−−+++ooStratification
in horizontal pipesoo++o−−++o+Emergency core
cooling mixing and condensationooo+o−−++−oLoop seal clearance++++−−−++++Pool formation
in upper plenum/CCFLo−oo−−−oo−oCore wide void
and flow distribution−−−o−−−+o−×Heat transfer
in covered core++++o−+++++Heat transfer in
partially uncovered coreoo++−−+++ooHeat transfer
in steam generator primary side+++++−+++++Heat transfer in
steam generator secondary sideo+++o−+o+++Pressurizer
thermal-hydraulicsoooo−o+oo++Surge line
hydraulics (CCFL, chocking)oooo−−+−ooOne and
two-phase pump behaviouro+++−−o−o++Structural heat
and heat lossesooooo−−oo++Noncondensable
gas effect on leak flow−+oo−−−−−××Phase separation in T-junctionsoo++o−−++−+Thermal-hydraulic
nuclear feedback−−−−−−−−−××Boron mixing and
transport−−−−−−−−××Separator
behaviour−−−−−−−−−××(°) The best performance of the
code is considered due to number of submissions.Note: the following symbols are
used in this table; for test facility versus
phenomenon: + suitable for code assessment, o limited suitability, − not suitable, for phenomenon versus experiments: + experimentally well
defined, o occurring but not
well characterised, − not occurring or
not measured, for phenomenon versus code calculation: + well
predicted, o qualitatively
predicted, −
not predicted,× not applicable.The significant trends of variables
with reference to the selected tests are shown in Figures2 through 7, while
details of the experiments are given below.Figure 2
ISP-18 (LOBI): experimental trends of primary and secondary side pressures and broken loop cold leg density.ISP 18: The test in LOBI simulated a 1% cold leg break with HPIS intervention (Figure 2). From a
phenomenological point of view, the whole transient can be divided into three main phases:(i)
the forced circulation period,(ii)
the two-phase natural circulation period,(iii)
the reflux condensation period. During the first phase, after the opening of the break device, the primary system
pressure decreases down to 13.2 MPa within 32 seconds, triggering both SG isolation and core power
decay. Simultaneously, secondary system cooling is activated causing an upper
limit to the increase in secondary pressure.
At 45 seconds pumps coast down begins and at 74 seconds HPIS starts to inject
water into the primary system. At 121
seconds pump coast down completion ends the forced circulation phase, and two-phase natural circulation is established in the
loops. As voiding proceeds, natural circulation stops and heat exchange with
the secondary system is accomplished by reflux condensation occurring in the steam generator U-tubes.An important feature of the test is the liquid mass distribution
inside the primary loop which is affected by the bypass flow paths in the
vessel and by heat transfer across steam generators mainly during natural
circulation and reflux condensation periods. Since HPIS is sufficient to avoid
core uncovery, no dry out is measured
during the test.ISP 20: The considered transient in Doel
plant is the steam generator tube rupture (SGTR) accident (with a longitudinal crack
of 7 cm long located in the ascending leg of the U-bend of one of the U-tubes) occurred in Doel plant in 1979 and
constituted the first (and, so far, the unique) standard problem related
to a plant system (Figure 3). At the moment
when the event occurred, the reactor was subcritical with all control rods down and the pressurizer heaters on. In
the secondary side, the steam lines were both isolated by the MSIV and
no condenser vacuum was available. The main feed water pumps were not
operational and water level in both SGs was manually
controlled by means of a letdown system. The auxiliary feed water pumps were not
running. The plant conditions remained well below the
safety margins during the whole transient.Figure 3
ISP-20 (Doel-2): registered data trends of primary side pressure and pressurizer level.The condensation induced by the pressurizer spray and in the secondary side of
steam generators at the liquid-steam interface is the relevant phenomena to be
predicted by codes. However, quite large uncertainties
characterize the trends of the main quantities as well as the time of
actuation of the main systems, typically reflecting the features and
capabilities of plant instrumentation and recording systems.ISP 22: The test in the SPES facility consists of a loss of feed
water with delayed actuation of emergency feed water in one of
the three loops of the facility. The transient evolves through 5 phases (Figures
4 and 5) from the following.(i)
The accident beginning to scram: due to the loss of feed
water, the downcomer level drops quickly in each steam generator. As
the low level set point is achieved, the scram occurs, causing the core
power to shutoff and the main steam isolation valves to close.(ii)
Scram to pressurizer PORV opening: after scram a quick
depressurization occurs in primary side as a consequence of temperature
decrease. The steam generators U-tubes then dry out, the primary
temperature rises continuously, causing primary system pressurization up to the
pressurizer PORV opening.(iii)
Pressurizer PORV opening to pumps trip: while the
primary temperature is rising continuously and is approaching the
saturation value, the pumps are switched off when the fluid subcooling
at the inlet reaches the set point value.(iv)
Pumps trip to emergency feed water activation: due to the
progressive voiding of the primary side, a core heat up occurs and the emergency feed
water activation signal in one of the steam
generators is generated by the high rod surface temperature set point.(v)
Emergency feed water activation to the end of the
transient: emergency feed water activation causes a quick
repressurization in the affected steam generator and reestablishes heat
transfer between the primary and the secondary sides, with a consequent big
decrease of primary temperature and pressure. The secondary
level in the affected steam generator increases steadily until the
initial value is restored. The following main features of the test can be pointed out.(i)
The pressure control of the primary system by the pressurizer PORV cycling and the consequent mass depletion cause
rod surface temperature excursion roughly two hours after the transient beginning.(ii)
The actuation of emergency feed water in one loop leads to primary system depressurization,
pressurizer draining, core quench, and brings the facility to safe shut down
conditions, allowing the possibility of accumulators actuation.Figure 4
ISP-22 (SPES): experimental trends
of pressurizer pressure and level.Figure 5
ISP-22 (SPES): experimental trends of steam generator pressure and level.ISP 26: The experiment in the LSTF test facility is originated by a 5% break in the cold
leg of the loop without pressurizer, the HPIS is not available
(Figure 6). Following the break opening the primary pressure went down and scram occurred at 9 seconds.
The core was temporarily uncovered, at first time, between about 120 and 155 seconds after break
opening. The reason for this was a core level depression amplified by a manometric effect caused by
condensation at the top of U-tubes and consequent liquid holdup in the ascending and descending legs of
U-tubes. At about 140 seconds, loop seal clearing occurred and caused a temporary core temperature recovery.
After loop seal clearing, the break flow changed from low quality to high quality two-phase flow and the
depressurization of primary loop was accelerated. By about 180 seconds after
the break, the primary loop pressure decreased below steam generator secondary side pressure. Thereafter,
the steam generator no longer served as heat sink and the energy removal from the primary system
occurred through the discharge of coolant from the break. It is noted that loop
seal clearing occurred before the reversal in primary and secondary pressures. The core was uncovered again after
about 420 seconds due to vessel inventory boiloff; the heater rods in the upper part of the core showed
superheating up to about 80 K. The core was covered with two-phase
mixture again after about 540 seconds by the accumulator water injection. The
peak cladding temperature in the test was approximately 740 K, observed during the temporary core
uncovery just before the loop seal clearing.Figure 6
ISP-26 (ROSA-IV): experimental trends of primary and secondary side pressures and rod surface temperature.Figure 7
ISP 27 (BETHSY): experimental trend of primary side pressure.The occurrence of two dry out and quench conditions constitutes the main peculiarity of this transient.
The mass distribution in the loop and the heat transfer with secondary side constitute further challenging
phenomena for code assessment.ISP 27: The test
in BETHSY facility is an SBLOCA with the break (roughly 0.5%)
located in the cold leg of the loop
with the pressurizer (Figure 7); HPIS
is not available. Three different phases can be recognized during the transient:(i)
subcooled blowdown;(ii)
mass depletion in primary side;(iii)
ultimate procedure.Subcooled Blowdown
Following the break opening the primary pressure falls down and scram occurs when the pressure reaches 13.1 MPa. safety injection signal (SI) occurs at 11.9 MPa. Following SI signal,
turbine bypass occurs and main feed water is off. Before SI, secondary side pressure is controlled through the spray
condenser and remains constant at 6.91 MPa; when turbine bypass occurs the
pressure threshold becomes 7.03 MPa. Auxiliary feed water injection starts 30 seconds after SI signal, and pump coast down
initiates 300 seconds after the same signal. During this phase, the pressurizer and surge line empty leading to the
relatively fast depressurization of the primary side; in the same period owing to the diminution of the
heat transfer from primary to secondary side, the mass flowrate in the secondary side starts to
decrease.Mass Depletion
The second phase is characterized by mass depletion and almost constant
pressure and temperature in primary loop (saturation values). Oscillations in
break flowrate in the first period
of phase 2 testify of little voiding of the cold leg of the broken loop. Later
on, with pumps at rest, once the upper
head to downcomer bypass steam flows to the broken cold leg, mostly steam flows at the break (stratified conditions
with liquid level upstream the break lower than the elevation of the exit
nozzle axis). Loop seal clearing is recognized
to appear in only one of the two intact loops and stops with the
occurrence of the first core uncovery. Secondary side conditions (mostly
levels) remain constants in this period. At
the end of this phase, a second core uncovery occurs, which causes the trip for the predefined ultimate procedure when the
core maximum clad temperature reaches 723 K.The Ultimate Procedure
This phase of the test consists in fully
opening the dump valves in secondary side due to accumulators and LPIS actuation; three different parts can be
distinguished during the last phase of the transient (A, B, and C, resp.). In the part A, starting with the
ultimate procedure initiation and ending with accumulators isolation, intense condensation in the U-tubes induces
liquid fall back to the core, which is cooled from the top, then accumulator
injection allows the clad temperature to turn around and the core to be rewetted. Part B is related to the period
from the accumulators’ isolation up to LPIS actuation. A continuous mass depletion of primary side without ECC
injection characterizes this phase. No dry out situation occurs in this period
during which the primary pressure decreases
down to achieving the set point for LPIS actuation. Very early during part C, LPIS
flowrate becomes larger than break flowrate leading to recover the primary
coolant system. In this period filling up
the primary loop occurs causing, among other things, direct contact condensation between the cold liquid injected by
LPIS and the steam present in the primary loop.


## 6.1. Facilities and Plant Hardware

In this section, information is given concerning some hardware
features that are relevant for the considered ISP tests. Figure1 shows the sketch of LOBI, SPES, BETHSY,
and LSTF facilities and of the Doel plant.Figure 1
Sketch of the facilities considered for the experimental data base evaluation.The relative elevations of important system components like core, steam generators
U-tubes, loop seals can be seen; the number of loops constituting the
system is reported too. The most important design parameters of the
considered facilities and of the plant are given in Table2. All the considered facilities can
operate at the reference plant nominal pressure for both primary and secondary loops. The height scaling ratio is
equal to one in all cases, so the gravity heads are properly simulated. The maximum allowed power is equal to
the reference reactor value multiplied by volume scaling ratio only in the cases of LOBI and SPES. In other cases, a decay
power value is allowed, ranging around 10% of the nominal value. This
scaling limitation prevents, among the other things, the possibility to have simultaneously rightly scaled temperatures
and flowrates in nominal conditions. In these facilities, the choice is
generally made to preserve hot leg fluid temperature during steady state operation, before any transient; alternatively, it
is possible to preserve the cold leg fluid temperature and nominal flowrate (hot leg temperature not preserved); as a
consequence of the former choice, secondary side fluid temperature and pressures must be higher than the
reference plant nominal values (a real plant at hot standby conditions, 10% of nominal power, exhibits the same behaviour, roughly 70 bar at secondary side);
still, primary pumps have not the maximum allowable flowrate and head properly scaled, although in the case of BETHSY, primary pumps have full
flowrate capacity and preserve the head in single phase flow conditions. The different
criteria utilized for the pressurizer result from Figure 1, as well for
defining the minimum elevation of the loop seal. In the facilities (SPES, BETHSY, and LSTF), the L/√D scaling is adopted for the design of hot and cold legs
piping also preserving the volume scaling [14].Table 2
Relevant hardware characteristics of considered PWR simulators and Doel-2 nuclear plant.
QuantitySPESLOBI/Mod2BETHSYLSTF/ROSA IVDOEL-21Reference reactorW-PWRKWU-PWRFRA-PWRW-PWRW-PWRReference reactor power (MWt)277539002700342311872Mximum power (MWt)/% of nominal power9.0/1385.4/1003.0/1010.0/141187.0/1003ReportedKv1/4271/7121/1001/481/14No. of rods97644281064216595Operating pressure of primary loop (MPa)15.15.815.5115.515.56Operating pressure  of
secondary loop (MPa)6.16.546.917.3/7.45.887Primary loop volume (m3)0.6220.6432.888.3168.58No. of U-tubes for
each steam generators13/13/138/2434/34/34141/1413260/32609Internal diameter of U-tubes (mm)15.419.619.719.619.610L/D ratio of Hot leg57.273.1/119.13817.810.6411Total head (“a” in Figure1) mmax16.0816.7218.3418.414.7min15.9116.4716.8716.913.212Linear rod power at 5%
overall power(*)(Kv/m)1.27(°)1.080.860.911.1213ActualKv(+)1/6401/6191/1441/481/2.5(*)The % value is the power related to the reported Kv.(+) Related to LSTF.(°)To compensate heat losses.Nevertheless, the position of the hot leg axis with respect to the top of the active fuel may
be not the same as in the reference nuclear power plant; in BETHSY, this position is preserved with respect to the
reference reactor, as well the bottom line of the cold leg elevation to the
bottom of active fuel, this leads to different elevations for hot
and cold leg axes. For all the multiloop facilities, each primary (and secondary)
circuit is equal to the other; thus nearly symmetrical thermal-hydraulic conditions occur in the various loops. An exception
is represented by LOBI, where one loop (intact) simulates three loops of the reference reactor and the other simulates a
single (broken) loop. Hardware parameters like pump geometrical configuration, presence, and characteristics
of bypass flow paths (mostly in the vessel) can play an important role in the considered test scenarios.


## 6.2. Outline of the Experimental Scenarios

The experiments A2-81, SP-FW-02, SB-CL-18, 9.1b, and the SGTR transient,
respectively from LOBI, SPES, LSTF, and BETHSY facilities and Doel
plant (Figure1 and Table 2), were submitted by the facility owner organisations to the CSNI and
were discussed and approved at working group and principal working group levels.
The list of host organisations (i.e., proposing the exercise, writing the final reports, and chairing the workshops) for each ISP, is given
in Table 3. The procedures outlined in [3] for assignments of ISPs have been
generally followed.Table 3
List of host organisations for small break LOCA related ISPs.
ISPHost organisation18JRC
(Ispra)20TRACTEBEL
(Brussels)22ENEA (Rome)26JAERI (Tokai Mura)27CEA/CENG (Grenoble)The main characteristics of the mentioned tests are reported in Table4. The main phenomena occurring during SB-LOCAs are listed in Table 5
[15], making use of a phenomena matrix developed in state-of-the-art report
(SOAR) on emergency core cooling thermal-hydraulics [15]. In the same table, a qualitative evaluation of the
capabilities of facilities is provided, according to three judgment levels. For completeness and in order to
give an example of the possible use of this table, in the last two columns, an overall evaluation of
the Relap5/Mod2 and CATHARE codes in addition to their performances is reported, considering each of the
phenomena listed and the pre- and posttest calculations [15].Table 4
Main characteristics of the considered transient.
ISPFacility/plantTestTypeSecondary sideEmergency systemsRecoveryEnd of testident.significant conditionsin primary side(*)procedure(s)ISP 18LOBIA2-81SBLOCAAb = 1%Amax in cold legImposed 100K/hrHPIS in cold legSecondary
system feed and bleed4500ISP 20DOELSGTR 1979SGTRAb = 0.5(+)% of AmaxEFW and steam
Relief valves activeHPIS in cold leg: pressurizer
sprays and heaters—3000ISP 22SPESSP-FW-02LOFW-loss of feed waterBoildown of secondary side
and EFW active in one loopPressurizer PORV and
heatersEFW in one loop8000ISP 26LSTFSB-CL-18SBLOCAAb = 5% of AmaxSteam relief valves activeAccumulators and LPIS in
cold legRHR actuation1000ISP 27BETHSY9.1.bSBLOCAAb = 0.4% of AmaxEFW and pressure control
activeAccumulators
and LPIS in cold legDepressurizarion of secondary side8000(*)scram is assumed in
all cases following a low pressure signal.(+)rough evaluationTable 5
Suitability of tests facilities, judgment of the experiments, and (example of) evaluation of RELAP5/Mod2 and
CATHARE code capabilities as from [13].
FacilitiesExperimentsPhenomenaSPESLOBI/Mod2BETHSYLSTFISP 18ISP 20ISP 22ISP 26ISP 27RELAP5/Mod2 CodeCATHARE V 1.3Performance(°)Code PerformanceNatural circulation in one-phase flow, primary side++++o−−−−++Natural
circulation in two-phase flow, primary side+++++−o−o++Reflux condenser
mode and CCFLoo++o−+++ooAsymmetric loop
behaviour+o+o−o+o+ooLeak flowo+++−−o++o+Phase separation
without mixture level formation++++−−−−o×+Mixture level and
entrainment in steam generator secondary sideooo+oo+ooooMixture level
and entrainment in the coreoo+−−−+++ooStratification
in horizontal pipesoo++o−−++o+Emergency core
cooling mixing and condensationooo+o−−++−oLoop seal clearance++++−−−++++Pool formation
in upper plenum/CCFLo−oo−−−oo−oCore wide void
and flow distribution−−−o−−−+o−×Heat transfer
in covered core++++o−+++++Heat transfer in
partially uncovered coreoo++−−+++ooHeat transfer
in steam generator primary side+++++−+++++Heat transfer in
steam generator secondary sideo+++o−+o+++Pressurizer
thermal-hydraulicsoooo−o+oo++Surge line
hydraulics (CCFL, chocking)oooo−−+−ooOne and
two-phase pump behaviouro+++−−o−o++Structural heat
and heat lossesooooo−−oo++Noncondensable
gas effect on leak flow−+oo−−−−−××Phase separation in T-junctionsoo++o−−++−+Thermal-hydraulic
nuclear feedback−−−−−−−−−××Boron mixing and
transport−−−−−−−−××Separator
behaviour−−−−−−−−−××(°) The best performance of the
code is considered due to number of submissions.Note: the following symbols are
used in this table; for test facility versus
phenomenon: + suitable for code assessment, o limited suitability, − not suitable, for phenomenon versus experiments: + experimentally well
defined, o occurring but not
well characterised, − not occurring or
not measured, for phenomenon versus code calculation: + well
predicted, o qualitatively
predicted, −
not predicted,× not applicable.The significant trends of variables
with reference to the selected tests are shown in Figures2 through 7, while
details of the experiments are given below.Figure 2
ISP-18 (LOBI): experimental trends of primary and secondary side pressures and broken loop cold leg density.ISP 18: The test in LOBI simulated a 1% cold leg break with HPIS intervention (Figure 2). From a
phenomenological point of view, the whole transient can be divided into three main phases:(i)
the forced circulation period,(ii)
the two-phase natural circulation period,(iii)
the reflux condensation period. During the first phase, after the opening of the break device, the primary system
pressure decreases down to 13.2 MPa within 32 seconds, triggering both SG isolation and core power
decay. Simultaneously, secondary system cooling is activated causing an upper
limit to the increase in secondary pressure.
At 45 seconds pumps coast down begins and at 74 seconds HPIS starts to inject
water into the primary system. At 121
seconds pump coast down completion ends the forced circulation phase, and two-phase natural circulation is established in the
loops. As voiding proceeds, natural circulation stops and heat exchange with
the secondary system is accomplished by reflux condensation occurring in the steam generator U-tubes.An important feature of the test is the liquid mass distribution
inside the primary loop which is affected by the bypass flow paths in the
vessel and by heat transfer across steam generators mainly during natural
circulation and reflux condensation periods. Since HPIS is sufficient to avoid
core uncovery, no dry out is measured
during the test.ISP 20: The considered transient in Doel
plant is the steam generator tube rupture (SGTR) accident (with a longitudinal crack
of 7 cm long located in the ascending leg of the U-bend of one of the U-tubes) occurred in Doel plant in 1979 and
constituted the first (and, so far, the unique) standard problem related
to a plant system (Figure 3). At the moment
when the event occurred, the reactor was subcritical with all control rods down and the pressurizer heaters on. In
the secondary side, the steam lines were both isolated by the MSIV and
no condenser vacuum was available. The main feed water pumps were not
operational and water level in both SGs was manually
controlled by means of a letdown system. The auxiliary feed water pumps were not
running. The plant conditions remained well below the
safety margins during the whole transient.Figure 3
ISP-20 (Doel-2): registered data trends of primary side pressure and pressurizer level.The condensation induced by the pressurizer spray and in the secondary side of
steam generators at the liquid-steam interface is the relevant phenomena to be
predicted by codes. However, quite large uncertainties
characterize the trends of the main quantities as well as the time of
actuation of the main systems, typically reflecting the features and
capabilities of plant instrumentation and recording systems.ISP 22: The test in the SPES facility consists of a loss of feed
water with delayed actuation of emergency feed water in one of
the three loops of the facility. The transient evolves through 5 phases (Figures
4 and 5) from the following.(i)
The accident beginning to scram: due to the loss of feed
water, the downcomer level drops quickly in each steam generator. As
the low level set point is achieved, the scram occurs, causing the core
power to shutoff and the main steam isolation valves to close.(ii)
Scram to pressurizer PORV opening: after scram a quick
depressurization occurs in primary side as a consequence of temperature
decrease. The steam generators U-tubes then dry out, the primary
temperature rises continuously, causing primary system pressurization up to the
pressurizer PORV opening.(iii)
Pressurizer PORV opening to pumps trip: while the
primary temperature is rising continuously and is approaching the
saturation value, the pumps are switched off when the fluid subcooling
at the inlet reaches the set point value.(iv)
Pumps trip to emergency feed water activation: due to the
progressive voiding of the primary side, a core heat up occurs and the emergency feed
water activation signal in one of the steam
generators is generated by the high rod surface temperature set point.(v)
Emergency feed water activation to the end of the
transient: emergency feed water activation causes a quick
repressurization in the affected steam generator and reestablishes heat
transfer between the primary and the secondary sides, with a consequent big
decrease of primary temperature and pressure. The secondary
level in the affected steam generator increases steadily until the
initial value is restored. The following main features of the test can be pointed out.(i)
The pressure control of the primary system by the pressurizer PORV cycling and the consequent mass depletion cause
rod surface temperature excursion roughly two hours after the transient beginning.(ii)
The actuation of emergency feed water in one loop leads to primary system depressurization,
pressurizer draining, core quench, and brings the facility to safe shut down
conditions, allowing the possibility of accumulators actuation.Figure 4
ISP-22 (SPES): experimental trends
of pressurizer pressure and level.Figure 5
ISP-22 (SPES): experimental trends of steam generator pressure and level.ISP 26: The experiment in the LSTF test facility is originated by a 5% break in the cold
leg of the loop without pressurizer, the HPIS is not available
(Figure 6). Following the break opening the primary pressure went down and scram occurred at 9 seconds.
The core was temporarily uncovered, at first time, between about 120 and 155 seconds after break
opening. The reason for this was a core level depression amplified by a manometric effect caused by
condensation at the top of U-tubes and consequent liquid holdup in the ascending and descending legs of
U-tubes. At about 140 seconds, loop seal clearing occurred and caused a temporary core temperature recovery.
After loop seal clearing, the break flow changed from low quality to high quality two-phase flow and the
depressurization of primary loop was accelerated. By about 180 seconds after
the break, the primary loop pressure decreased below steam generator secondary side pressure. Thereafter,
the steam generator no longer served as heat sink and the energy removal from the primary system
occurred through the discharge of coolant from the break. It is noted that loop
seal clearing occurred before the reversal in primary and secondary pressures. The core was uncovered again after
about 420 seconds due to vessel inventory boiloff; the heater rods in the upper part of the core showed
superheating up to about 80 K. The core was covered with two-phase
mixture again after about 540 seconds by the accumulator water injection. The
peak cladding temperature in the test was approximately 740 K, observed during the temporary core
uncovery just before the loop seal clearing.Figure 6
ISP-26 (ROSA-IV): experimental trends of primary and secondary side pressures and rod surface temperature.Figure 7
ISP 27 (BETHSY): experimental trend of primary side pressure.The occurrence of two dry out and quench conditions constitutes the main peculiarity of this transient.
The mass distribution in the loop and the heat transfer with secondary side constitute further challenging
phenomena for code assessment.ISP 27: The test
in BETHSY facility is an SBLOCA with the break (roughly 0.5%)
located in the cold leg of the loop
with the pressurizer (Figure 7); HPIS
is not available. Three different phases can be recognized during the transient:(i)
subcooled blowdown;(ii)
mass depletion in primary side;(iii)
ultimate procedure.Subcooled Blowdown
Following the break opening the primary pressure falls down and scram occurs when the pressure reaches 13.1 MPa. safety injection signal (SI) occurs at 11.9 MPa. Following SI signal,
turbine bypass occurs and main feed water is off. Before SI, secondary side pressure is controlled through the spray
condenser and remains constant at 6.91 MPa; when turbine bypass occurs the
pressure threshold becomes 7.03 MPa. Auxiliary feed water injection starts 30 seconds after SI signal, and pump coast down
initiates 300 seconds after the same signal. During this phase, the pressurizer and surge line empty leading to the
relatively fast depressurization of the primary side; in the same period owing to the diminution of the
heat transfer from primary to secondary side, the mass flowrate in the secondary side starts to
decrease.Mass Depletion
The second phase is characterized by mass depletion and almost constant
pressure and temperature in primary loop (saturation values). Oscillations in
break flowrate in the first period
of phase 2 testify of little voiding of the cold leg of the broken loop. Later
on, with pumps at rest, once the upper
head to downcomer bypass steam flows to the broken cold leg, mostly steam flows at the break (stratified conditions
with liquid level upstream the break lower than the elevation of the exit
nozzle axis). Loop seal clearing is recognized
to appear in only one of the two intact loops and stops with the
occurrence of the first core uncovery. Secondary side conditions (mostly
levels) remain constants in this period. At
the end of this phase, a second core uncovery occurs, which causes the trip for the predefined ultimate procedure when the
core maximum clad temperature reaches 723 K.The Ultimate Procedure
This phase of the test consists in fully
opening the dump valves in secondary side due to accumulators and LPIS actuation; three different parts can be
distinguished during the last phase of the transient (A, B, and C, resp.). In the part A, starting with the
ultimate procedure initiation and ending with accumulators isolation, intense condensation in the U-tubes induces
liquid fall back to the core, which is cooled from the top, then accumulator
injection allows the clad temperature to turn around and the core to be rewetted. Part B is related to the period
from the accumulators’ isolation up to LPIS actuation. A continuous mass depletion of primary side without ECC
injection characterizes this phase. No dry out situation occurs in this period
during which the primary pressure decreases
down to achieving the set point for LPIS actuation. Very early during part C, LPIS
flowrate becomes larger than break flowrate leading to recover the primary
coolant system. In this period filling up
the primary loop occurs causing, among other things, direct contact condensation between the cold liquid injected by
LPIS and the steam present in the primary loop.


## 7. The Results of Some Statistical Analysis  for Small Break LOCA ISPs

In the framework of the ISP activity evaluation, interesting information may come
from the statistical analysis considering the number of participants to the ISP, including countries
and organisations, as well as the adopted thermal-hydraulic system codes. The
main goals of the effort are to get an overview of the interest towards the ISP
activity from the international scientific community, and to derive information about the engagement
by different organisations in the use of large thermal-hydraulic system codes.A wide database is available for making statistical evaluations; this is included
in the ISP reports approved and issued by CSNI and in the individual
ISP participants written contributions normally distributed (among
participants) at the ISP workshops. A comprehensive analysis would require
establishing homogeneous indices for interpreting the data, for example,(i)
computers have strongly evolved lowering the needed calculation time in the period
1985–1995 (in some cases, the calculation
time increases just because transients take longer times);(ii)
codes having sophisticated capabilities of noding a specific zone of nuclear power plant (i.e.,
volume component in CATHARE) may need less overall number of node for having
the same detail of plant description;(iii)
once an acceptable convergence is reached from a numeric point of view,
the increase in number of time steps might not lead to any benefit; calculation time may be reduced by
the progress in physical modelling reducing the interaction number and meshing size. However, a number of quantities could be used to characterize the results of an extended
statistical analysis, for example, [16]. Following a
discussion among the participating working group members, it was found that
most of the data (e.g., numbers of used meshes or nodes) averaged on the
number of participants could be misinterpreted or even misleading
considering the present situation. This is originated by the reason outlined above, specifically, including the
different levels of qualification of the scientists directly involved in the calculation and even the different
purposes for organisations in participating in an ISP. As an example, it was found that the
consideration of the number of input deck nodes for the different participants should not give a
reasonable index of the “quality” of user nodalization itself.The lack, in the ISP documents of an exhaustive description of calculation resources, prevented the
possibility to use the time needed for the calculation of ISP exercises, as a
parameter eventually identifying a “code speed” index.Keeping in mind the above, the following quantities were selected for the present analysis:(i)
kind and number of participants to the ISP,(ii)
thermal-hydraulic codes used for the ISP calculation. In relation to the first item, it seemed interesting to correlate the participants with the different
ISPs and with the adopted codes used, considering the total number of
participations to the ISPs for each participant.The second item gives an idea of the differences in the use of each code. It must
be emphasized that the results of the analysis might not be indicative
of the actual number of users for each code. More detailed
information in this context should be gathered by specific collaborative
programmes like Club des Utilisateurs du
CATHARE (CUC), Code Assessment and Maintenance Program (CAMP) or specific
“institutionalized” series of conferences like Relap5 International
Conference.Specific parameters to characterize the two items identified above, which were retained suitable for evaluating the
overall impact of ISP activity in the scientific community are(1)
number of participants to the specific ISP,(2)
participants per ISP,(3)
number of countries per ISP,(4)
participants per code per ISP,(5)
codes used per ISP. ISP phases (e.g., pre- and posttest) are considered in Tables 6,  7, 8 and the information related to
items (1) to (5) are given in these tables. As already mentioned, further information on statistical evaluation,
considering a large number of parameters, can be found in [4, 16].
It is to be noted that there are six types of organisations who participated in
the small break LOCA ISP exercises. These are covering a wide range of organisations:
research centres, universities, licensing authorities, industry, utility, and others
(e.g., engineering companies).Table 6
Participants per code per ISP.
ISP 18ISP 18ISP 20ISP 22ISP 22ISP 26ISP 27ISP 27pretestposttestposttestpretestposttestposttestpretestposttestRELAP4/Mod621—32—11RELAP5/Modl10111————RELAP5/Mod24454171011RELAP5/Mod2.5—————211RELAP5/Mod3——————76CATHARE 111——————CATHARE 2——112136TRAC/PF131—4—11—DRUFAN-M23———————ATHLET  1.0—————111SMABRE——11—3——NOTRUMP1——1—211FRACAS1————1——Dinamyka——————1—Tech-m4—————11—Moot———1————ATHENA1———————SATAN-M—————1——Table 7
Countries, Participants, and Codes used per ISP.
ISPTypeNo. of participantsNo. of countriesNo. of codesISP 18Pretest27118ISP 18Posttest665ISP 20Posttest754ISP 22Pretest17148ISP 22Posttest443ISP 26Posttest17149ISP 27Pretest23179ISP 27Posttest17147(*) Note: the numbers reported in the number of
participants column do not coincide with the sum of the numbers in Table6
because either a single
participant might have submitted more than one calculation officially
recognized, or a group of participants took part together in the ISP submitting a single calculation.Table 8
Calculations per code groups per ISP.
CodesISP I8ISP 18ISP 20ISP 22ISP 22ISP 26ISP 27ISP 27prepostpostprepostpostprepostRELAP family
of codes16668391919CATHARE family
of codes11112136TRAC family of
codes31—4—11—Others6—13—932Total2688165202727Detailed statistical data and analysis are included in [4]; in this paper, a few conclusions drawn from
the analysis of the statistical data are given as follows.(i)
A large number of codes have been used in the different ISPs. It is
possible to see a predominant use of RELAP family of codes specifically from most of universities and research
centres.(ii)
A number of participants still use first generation (e.g., RELAP4) or proprietary codes (NOTRUMP).(iii)
The number of participants increased after ISP 20 essentially due to the fact that since the time of ISP
22, the ISP activity was open for the non-OECD countries. The positive effect
was to allow Eastern countries to get information about Western countries
safety methodologies. A “negative” impact of this was
the increment of the scientists participating to the ISP for the first time,
making more difficult to get objective conclusions from the discussions about
the ISP itself.(iv)
The use of well established or “frozen” versions of codes allows the verification of the degree of assessment of the concerned code version against a
full transient.(v)
Fourty six organisations took part in the small break LOCA ISP activities; very few organisations took part to
more than five of the considered ISP cases.(vi)
Of the above organisations, almost 82% belong to the research/university side (specifically, 54% research institutes and 28% universities).


## 8. Some Lessons Learned from the Small  Break Loca ISP Activity

The contents of this section are based on the answers received to a questionnaire
[4] that was sent to fourteen members of TG-THSB who were involved in
the analysis of most of the small break LOCA ISPs, from the
conclusions included in each of the ISP final report (CSNI reports, [5] to [10]), and from the
discussions of a working group, which took place during the meeting in Pisa University in 1995.As mentioned in Section7, eighteen different codes were used by the participants for these
ISPs. It is not the purpose here to produce a detailed analysis of
calculational performances, code by code, and ISP by ISP; but in a more synthetic approach, to derive the main outcomes
from the five ISPs, specifically taking into account the following four items identified in the questionnaire:(i)
code deficiencies and capabilities,(ii)
progress in the code capabilities,(iii)
possibility of scaling,(iv)
other comments. It should be mentioned that from ISP 18 to ISP 27, more and more physical
phenomena were involved in the transients which were dealt
within the ISP exercises, such as core-uncovery and heatup, pressurizer
discharge, secondary side voiding and filling, low pressure two phase
flows as well as interacting operator actions. The involvement of various
phenomena during an ISP exercise must be considered as challenging for the codes, and as
well as code users. Furthermore, increasing overall complexity and longer time
durations of the transients to be
calculated, can be noted during the process of going from the earlier to the
latest considered small break LOCA ISPs.


### 8.1. Code Deficiencies and Capabilities

The code user is clearly the best judge of the performance of his own calculations.
The invested resources, the depth of the quality assurance used when setting up
the nodalization, and the possibility to interact with
the experimentalists play a major role in the quality of the results, this can
only be known to the user. So, in order to
get a general, but not in depth evaluation of submitted results, two steps were
considered as follows:(a)
list of relevant thermal-hydraulic phenomena in each test, making reference to the
list in Table5, also looking at the facilities suitability;(b)
identification of phenomena which were not well predicted
by the majority of submitted calculations. The quality of experimental data also had a role in selecting code
deficiencies. A list of generic code
deficiencies, which were identified, is provided in Table 9. As code
deficiency, it was meant a situation where either the phenomenon is not
predicted to occur in the calculation, or the phenomenon was predicted but at a given time the quantity ¦Yc–YE¦/¦YE¦ was larger than 0.20 (see also
[9]). In this case, Y is a relevant thermal-hydraulic
quantity representing the assigned phenomenon and the deviation of calculated
from experimental quantity.Table 9
General code deficiencies for the considered ISPs.
ISP no.Identification no. of DeficiencyCode deficiencies18(1)Break flow(2)Stratification in cold and hot leg(3)Mass distribution in primary side(4)Mixing in the downcomer20(5)Steam condensation(6)Level simulation in secondary side22(6)Unability to predict mass inventory in secondary side(7)Heat transfer between EFW and hot SG walls(8)Pressurizer behaviour including PORV leak(9)Coolant mass distribution in primary circuit(1)Break flow(2)Stratification in cold and hot leg26(10)CCFL in SG plena(3–11)Core level depression(12)Core uncovery and heat-up27(1)Break flow(2)Stratification in cold and hot leg(3)Mass distribution in primary side(4)Mixing in the downcomer(12)Core uncovery and heat-up(13)Low pressure periodIt can be seen from Table9 that thirteen main code deficiencies have been found, some
of those being common to different ISPs. A comprehensive and systematic qualitative or quantitative code calculation
accuracy evaluation is well beyond the scope of the present paper. In this
respect, some example results are provided in [8, 9, 17] in relation to ISPs22,
26, and 27, respectively. Slightly different criteria are adopted for achieving either a qualitative judgment (e.g.,
good, average, and poor) or a quantitative evaluation (e.g., quantification of the
accuracy through the fast fourier transform- (FFT-) based method). For this type of evaluations, the
interested researcher could refer directly to the mentioned documents. Additional notes on selected
items are provided below.Let us first deal with the break flowrate problem (item 1) in Table9 appearing in
all ISPs, but not in ISP 20 and 22; many participants have experienced
wrong predictions of this parameter among the ISPs, leading to deviation
(sometimes large) from the actual transient. Although a very accurate
prediction of this quantity is not requested for safety studies, where a stated
range of break flowrate may be and is generally used, the capability of codes to reasonably predict
two-phase critical flowrates versus leak
geometry and upstream conditions becomes significant when the efficiency of operator actions (use of discharge devices, e.g.) has to be investigated. For the considered ISPs, various levels of agreement on the break flowrate
predictions were observed, and these results were often correlated with the
resources invested in this part of the work and the user's experience in this
field. It appears however that some break models are still having difficulty to calculate for the
whole range of break upstream conditions. In this
area, an example of complex interaction between code nodal inadequacies, user assumptions, interpretation of data provided by
experimentalists is given in [18] by using the RELAP5 code. This
sensitivity study about break discharge coefficients, performed during the ISP 27 posttest analysis, showed the large
influence of this parameter upon the time scale shifting appearing in
blind calculations. Even though, these coefficients had been previously adjusted by using the separate effect test
experimental data provided by the ISP host organisation. This mentioned study
pointed out and also emphasized the need for code assessment procedures to verify the overall agreement on
integral test transients.However, in general, break flow can be largely influenced by the upstream flow
conditions, which are strongly related to the mass distribution in
the entire system and to the overall system behaviour.
Therefore, just “tuning” the break flowrate might introduce a
compensation of errors and, as well as, it might result in complete
wrong conclusions. This also results in excluding to provide the ISP participants
with the measured break flow. For complicated geometries (such as
valves), geometry effects on break flow are even more important. The critical
flow performance of the valves must be characterized and supplied as input to the code.Another key parameter in these considered ISPs
is the coolant mass distribution in the primary circuit (item 3 in Table9, relevant to ISPs 18, 22, and 27), which is strongly
related to the two-phase structure and flow regimes. Interfacial shear stresses, counter-current flow limitations,
transitions between flow regimes are directly related to the coolant mass distribution. The need for a better
prediction of this distribution prompted the development of second-generation (“advanced”)
two-phase thermal-hydraulic codes.
These codes proved their ability to qualitatively predict the physical
phenomena involved during the different transients, such as stratified flows in
horizontal pipes, loop seal clearing, interfacial transport in core, and steam
generator U-tubes. Nevertheless, some weaknesses revealed during the first of
the considered ISPs and, concerning void distribution in vertical or horizontal components, still
appeared unresolved in ISP 27 (see Table 9).Additional specific comments are connected with the thermal coupling between fluid
and structures, both in primary and secondary
sides. This is a consequence of both the scaling ratio of the facilities
involved, and of the operating procedures applied; this has been
a subject of discussion during most of the ISP related
workshops. Inaccuracies due to different reasons in accounting for
the fluid structure and thermal coupling, that is, lack of
suitable noding and inadequate consideration of heat losses, may have
a role in various calculation discrepancies. In every case, codes have
demonstrated their ability to qualitatively describe these phenomena
(fluid-structures heat transfer), provided that a sufficient
amount of care and work had been spent to correctly define the geometry and
thermal boundary conditions.In ISPs 26 and 27 discrepancies remain in predicting core heatup, though fluid distribution is
predicted adequately. Similarly “hot wall delay” effect
in steam generators downcomer is not satisfactorily calculated in ISP 22. These examples
raised questions about the relevant heat transfer models in the considered
conditions.At last, some specific aspects specifics for one or two ISPs, such as secondary side level
prediction (ISPs 20 and 22), and low pressure refilling of the
primary coolant system (ISP 27), highlighted model weaknesses
in these fields for most of the codes.From the point of view of the code capabilities, it must be indicated that
experienced users are able to get the relevant phenomena
even in the case when complex scenarios are involved. Such a qualitative
judgment has been supported by quantitative evaluations, that is, quantification of accuracy considering
experimental and calculated trends, in the cases of ISP 22 and ISP 27 (see also below).However, looking generally to a single ISP, a wide range of results is achieved even considering
the use of same code versions. This emphasizes the role of the user in setting
up the nodalization and also in interpreting
the initial and boundary conditions supplied by the experimentalists. In
conclusion, in an ISP framework, owing to
different reasons (see also below) the user effect may overshadow the reasons for code deficiencies, thus preventing the possibility to
identify code capabilities


### 8.2. Identification of Progress in Code Capabilities

Firstly, it must be emphasized
that one of the reasons why progress is difficult to measure, is that it is difficult to isolate phenomena in an
integral test. Owing to this fact, it is also
difficult to judge even making reference to each single code, since there is also no clear feedback between the
ISP activity and the code developers, as already mentioned. In fact, ISPs have
been proved more useful to provide information on the capabilities of the thermal-hydraulic codes, especially when
posttest calculations or parametric studies were conducted, than to identify the deficiencies or failures. In
this case, returning to the use of more analytical work or separate effect tests is however necessary to modify or extend the
individual physical models; this step has allowed some progress in code capabilities. The direct contact condensation, or
stratification and phase separation
models in horizontal pipes constitute an example of this.Progress was also observed in
using parallel channel simulation in attempting to better represent 2D or 3D behaviours with the codes used,
which are basically one dimensional. One of the most important progresses has
been obtained in the area of users guidelines. Thanks to the large number of participants, often using the same
code versions, with different nodalizations and option choices, the ISP pre-
and posttest calculations, formed a wide “database” for the so called
“user effect.”The small break
LOCA ISPs provided a useful information basis, not only for experienced code
users to increase their capability from one ISP to the other, but also for new code
users to improve their know-how by exchanging ideas and meeting more experienced people in the frame of ISPs.


### 8.3. Possibility of Scaling

Although the considered five ISPs address the problem of scaling, either because the
plant transient is expected to be very similar to that observed in
the facilities which are properly scaled, or because of the
different scales of the facilities addressing the same thermal-hydraulic
phenomenon, or because a plant transient is considered (ISP 20), the commonly reached
conclusion is that small break ISPs alone are
not sufficient to check code accuracy in this field. The counterpart tests
performed making reference to the same scenario in terms of boundary and
initial conditions, on different facilities, are much more valuable for this task [17, 19, 20].However, it is considered interesting to bring to the attention hereafter the results of a common
evaluation, which was made in preparing CSNI report on “lessons learned from
OECD/CSNI ISP on small break LOCA” [4].Two items are identified to judge the possibility of using the small break LOCA ISP exercises in scaling activities.(A)
Realism of involved physical phenomena as far as plant is concerned.(B)
Possibility to assess the code in different scaled facilities in relation to
the same scenario (evaluation whether the small break LOCA ISP scenario
can be found in different scaled facilities). The analysis of each small break LOCA ISP related to the above two items gives the following results.
(i)
ISP 18, item (A): test scenario expected to be similar in the plant.(ii)
ISP 18, item (B): limited suitability because the test scenario not available in other facilities.(iii)
ISP 20, item (A): this is a plant scenario.(iv)
ISP 20, item (B): the same scenario has been considered in one of the LOBI experiment.(v)
ISP 22, item (A): qualitatively, phenomena expected to be the same as in the plant, but timing is different.(vi)
ISP 22, item (B): test suitable for scaling because the same experiment was repeated in different facilities.(vii)
ISP 26, item (A): plant scenario expected to be the same (local phenomena might be different).(viii)
ISP 26, item (B): test suitable for scaling because the counterpart test activity deals with similar scenario.(ix)
ISP 27, item (A): plant overall scenario expected to be the same.(x)
ISP 27, item (B): difficult to assess the code scaling capabilities, because the similar
test scenario is not available from other facilities. As a result of the above, ISP 22 and ISP 26 related experiments appears to be the most suitable for studying scaling. Even though it
is a plant, ISP 20 mostly suffers of limitations due to inadequacy of the database obtained from the
plant, both in relation to plant hardware and data recording, as already mentioned.


### 8.4. Other Comments

An additional outcome from the small break LOCA ISP activity in the second half of
90s appeared is linked to the area of works about quantitative accuracy evaluation of
codes. The results of the calculations for ISP 22 and ISP 27 have been used to check
some of these methods and proved very
useful for this purpose [16, 21].Another lesson from these small break LOCA ISPs concerns the experience gained by the
code users in performing calculations on various facilities and
transients, improving their understanding of the code
capabilities and weaknesses. Opening this activity to Eastern countries (since
ISP 22) was thus a unique opportunity specifically for small countries to have
access to relevant experimental data, and to improve
their know-how in relation to the use of codes and nuclear reactor safety.A further lesson from small break LOCA ISPs concerns the identification and characterization of user
effects [11]. Different code users utilizing the same code version and getting the
same available information from experimentalists (ISP host organisation)
produce quite different results especially in “blind” standard problems, but as
well as in “open” standard problems. ISP 25 (not included in the present study)
and ISP 26 (here considered) were used as
basis for the influence of the user on the results of calculations (see [11]).
Among the various out comings, it was found that, potentially, user effects can
be very important and may overshadow code
deficiencies or capabilities (same conclusion as in Section 8.1).


## 8.1. Code Deficiencies and Capabilities

The code user is clearly the best judge of the performance of his own calculations.
The invested resources, the depth of the quality assurance used when setting up
the nodalization, and the possibility to interact with
the experimentalists play a major role in the quality of the results, this can
only be known to the user. So, in order to
get a general, but not in depth evaluation of submitted results, two steps were
considered as follows:(a)
list of relevant thermal-hydraulic phenomena in each test, making reference to the
list in Table5, also looking at the facilities suitability;(b)
identification of phenomena which were not well predicted
by the majority of submitted calculations. The quality of experimental data also had a role in selecting code
deficiencies. A list of generic code
deficiencies, which were identified, is provided in Table 9. As code
deficiency, it was meant a situation where either the phenomenon is not
predicted to occur in the calculation, or the phenomenon was predicted but at a given time the quantity ¦Yc–YE¦/¦YE¦ was larger than 0.20 (see also
[9]). In this case, Y is a relevant thermal-hydraulic
quantity representing the assigned phenomenon and the deviation of calculated
from experimental quantity.Table 9
General code deficiencies for the considered ISPs.
ISP no.Identification no. of DeficiencyCode deficiencies18(1)Break flow(2)Stratification in cold and hot leg(3)Mass distribution in primary side(4)Mixing in the downcomer20(5)Steam condensation(6)Level simulation in secondary side22(6)Unability to predict mass inventory in secondary side(7)Heat transfer between EFW and hot SG walls(8)Pressurizer behaviour including PORV leak(9)Coolant mass distribution in primary circuit(1)Break flow(2)Stratification in cold and hot leg26(10)CCFL in SG plena(3–11)Core level depression(12)Core uncovery and heat-up27(1)Break flow(2)Stratification in cold and hot leg(3)Mass distribution in primary side(4)Mixing in the downcomer(12)Core uncovery and heat-up(13)Low pressure periodIt can be seen from Table9 that thirteen main code deficiencies have been found, some
of those being common to different ISPs. A comprehensive and systematic qualitative or quantitative code calculation
accuracy evaluation is well beyond the scope of the present paper. In this
respect, some example results are provided in [8, 9, 17] in relation to ISPs22,
26, and 27, respectively. Slightly different criteria are adopted for achieving either a qualitative judgment (e.g.,
good, average, and poor) or a quantitative evaluation (e.g., quantification of the
accuracy through the fast fourier transform- (FFT-) based method). For this type of evaluations, the
interested researcher could refer directly to the mentioned documents. Additional notes on selected
items are provided below.Let us first deal with the break flowrate problem (item 1) in Table9 appearing in
all ISPs, but not in ISP 20 and 22; many participants have experienced
wrong predictions of this parameter among the ISPs, leading to deviation
(sometimes large) from the actual transient. Although a very accurate
prediction of this quantity is not requested for safety studies, where a stated
range of break flowrate may be and is generally used, the capability of codes to reasonably predict
two-phase critical flowrates versus leak
geometry and upstream conditions becomes significant when the efficiency of operator actions (use of discharge devices, e.g.) has to be investigated. For the considered ISPs, various levels of agreement on the break flowrate
predictions were observed, and these results were often correlated with the
resources invested in this part of the work and the user's experience in this
field. It appears however that some break models are still having difficulty to calculate for the
whole range of break upstream conditions. In this
area, an example of complex interaction between code nodal inadequacies, user assumptions, interpretation of data provided by
experimentalists is given in [18] by using the RELAP5 code. This
sensitivity study about break discharge coefficients, performed during the ISP 27 posttest analysis, showed the large
influence of this parameter upon the time scale shifting appearing in
blind calculations. Even though, these coefficients had been previously adjusted by using the separate effect test
experimental data provided by the ISP host organisation. This mentioned study
pointed out and also emphasized the need for code assessment procedures to verify the overall agreement on
integral test transients.However, in general, break flow can be largely influenced by the upstream flow
conditions, which are strongly related to the mass distribution in
the entire system and to the overall system behaviour.
Therefore, just “tuning” the break flowrate might introduce a
compensation of errors and, as well as, it might result in complete
wrong conclusions. This also results in excluding to provide the ISP participants
with the measured break flow. For complicated geometries (such as
valves), geometry effects on break flow are even more important. The critical
flow performance of the valves must be characterized and supplied as input to the code.Another key parameter in these considered ISPs
is the coolant mass distribution in the primary circuit (item 3 in Table9, relevant to ISPs 18, 22, and 27), which is strongly
related to the two-phase structure and flow regimes. Interfacial shear stresses, counter-current flow limitations,
transitions between flow regimes are directly related to the coolant mass distribution. The need for a better
prediction of this distribution prompted the development of second-generation (“advanced”)
two-phase thermal-hydraulic codes.
These codes proved their ability to qualitatively predict the physical
phenomena involved during the different transients, such as stratified flows in
horizontal pipes, loop seal clearing, interfacial transport in core, and steam
generator U-tubes. Nevertheless, some weaknesses revealed during the first of
the considered ISPs and, concerning void distribution in vertical or horizontal components, still
appeared unresolved in ISP 27 (see Table 9).Additional specific comments are connected with the thermal coupling between fluid
and structures, both in primary and secondary
sides. This is a consequence of both the scaling ratio of the facilities
involved, and of the operating procedures applied; this has been
a subject of discussion during most of the ISP related
workshops. Inaccuracies due to different reasons in accounting for
the fluid structure and thermal coupling, that is, lack of
suitable noding and inadequate consideration of heat losses, may have
a role in various calculation discrepancies. In every case, codes have
demonstrated their ability to qualitatively describe these phenomena
(fluid-structures heat transfer), provided that a sufficient
amount of care and work had been spent to correctly define the geometry and
thermal boundary conditions.In ISPs 26 and 27 discrepancies remain in predicting core heatup, though fluid distribution is
predicted adequately. Similarly “hot wall delay” effect
in steam generators downcomer is not satisfactorily calculated in ISP 22. These examples
raised questions about the relevant heat transfer models in the considered
conditions.At last, some specific aspects specifics for one or two ISPs, such as secondary side level
prediction (ISPs 20 and 22), and low pressure refilling of the
primary coolant system (ISP 27), highlighted model weaknesses
in these fields for most of the codes.From the point of view of the code capabilities, it must be indicated that
experienced users are able to get the relevant phenomena
even in the case when complex scenarios are involved. Such a qualitative
judgment has been supported by quantitative evaluations, that is, quantification of accuracy considering
experimental and calculated trends, in the cases of ISP 22 and ISP 27 (see also below).However, looking generally to a single ISP, a wide range of results is achieved even considering
the use of same code versions. This emphasizes the role of the user in setting
up the nodalization and also in interpreting
the initial and boundary conditions supplied by the experimentalists. In
conclusion, in an ISP framework, owing to
different reasons (see also below) the user effect may overshadow the reasons for code deficiencies, thus preventing the possibility to
identify code capabilities


## 8.2. Identification of Progress in Code Capabilities

Firstly, it must be emphasized
that one of the reasons why progress is difficult to measure, is that it is difficult to isolate phenomena in an
integral test. Owing to this fact, it is also
difficult to judge even making reference to each single code, since there is also no clear feedback between the
ISP activity and the code developers, as already mentioned. In fact, ISPs have
been proved more useful to provide information on the capabilities of the thermal-hydraulic codes, especially when
posttest calculations or parametric studies were conducted, than to identify the deficiencies or failures. In
this case, returning to the use of more analytical work or separate effect tests is however necessary to modify or extend the
individual physical models; this step has allowed some progress in code capabilities. The direct contact condensation, or
stratification and phase separation
models in horizontal pipes constitute an example of this.Progress was also observed in
using parallel channel simulation in attempting to better represent 2D or 3D behaviours with the codes used,
which are basically one dimensional. One of the most important progresses has
been obtained in the area of users guidelines. Thanks to the large number of participants, often using the same
code versions, with different nodalizations and option choices, the ISP pre-
and posttest calculations, formed a wide “database” for the so called
“user effect.”The small break
LOCA ISPs provided a useful information basis, not only for experienced code
users to increase their capability from one ISP to the other, but also for new code
users to improve their know-how by exchanging ideas and meeting more experienced people in the frame of ISPs.


## 8.3. Possibility of Scaling

Although the considered five ISPs address the problem of scaling, either because the
plant transient is expected to be very similar to that observed in
the facilities which are properly scaled, or because of the
different scales of the facilities addressing the same thermal-hydraulic
phenomenon, or because a plant transient is considered (ISP 20), the commonly reached
conclusion is that small break ISPs alone are
not sufficient to check code accuracy in this field. The counterpart tests
performed making reference to the same scenario in terms of boundary and
initial conditions, on different facilities, are much more valuable for this task [17, 19, 20].However, it is considered interesting to bring to the attention hereafter the results of a common
evaluation, which was made in preparing CSNI report on “lessons learned from
OECD/CSNI ISP on small break LOCA” [4].Two items are identified to judge the possibility of using the small break LOCA ISP exercises in scaling activities.(A)
Realism of involved physical phenomena as far as plant is concerned.(B)
Possibility to assess the code in different scaled facilities in relation to
the same scenario (evaluation whether the small break LOCA ISP scenario
can be found in different scaled facilities). The analysis of each small break LOCA ISP related to the above two items gives the following results.
(i)
ISP 18, item (A): test scenario expected to be similar in the plant.(ii)
ISP 18, item (B): limited suitability because the test scenario not available in other facilities.(iii)
ISP 20, item (A): this is a plant scenario.(iv)
ISP 20, item (B): the same scenario has been considered in one of the LOBI experiment.(v)
ISP 22, item (A): qualitatively, phenomena expected to be the same as in the plant, but timing is different.(vi)
ISP 22, item (B): test suitable for scaling because the same experiment was repeated in different facilities.(vii)
ISP 26, item (A): plant scenario expected to be the same (local phenomena might be different).(viii)
ISP 26, item (B): test suitable for scaling because the counterpart test activity deals with similar scenario.(ix)
ISP 27, item (A): plant overall scenario expected to be the same.(x)
ISP 27, item (B): difficult to assess the code scaling capabilities, because the similar
test scenario is not available from other facilities. As a result of the above, ISP 22 and ISP 26 related experiments appears to be the most suitable for studying scaling. Even though it
is a plant, ISP 20 mostly suffers of limitations due to inadequacy of the database obtained from the
plant, both in relation to plant hardware and data recording, as already mentioned.


## 8.4. Other Comments

An additional outcome from the small break LOCA ISP activity in the second half of
90s appeared is linked to the area of works about quantitative accuracy evaluation of
codes. The results of the calculations for ISP 22 and ISP 27 have been used to check
some of these methods and proved very
useful for this purpose [16, 21].Another lesson from these small break LOCA ISPs concerns the experience gained by the
code users in performing calculations on various facilities and
transients, improving their understanding of the code
capabilities and weaknesses. Opening this activity to Eastern countries (since
ISP 22) was thus a unique opportunity specifically for small countries to have
access to relevant experimental data, and to improve
their know-how in relation to the use of codes and nuclear reactor safety.A further lesson from small break LOCA ISPs concerns the identification and characterization of user
effects [11]. Different code users utilizing the same code version and getting the
same available information from experimentalists (ISP host organisation)
produce quite different results especially in “blind” standard problems, but as
well as in “open” standard problems. ISP 25 (not included in the present study)
and ISP 26 (here considered) were used as
basis for the influence of the user on the results of calculations (see [11]).
Among the various out comings, it was found that, potentially, user effects can
be very important and may overshadow code
deficiencies or capabilities (same conclusion as in Section 8.1).


## 9. Conclusions

The ISPs are part of an important ongoing programme promoted by OECD/CSNI during
the last thirty years and gave, among the other things, the possibility to disseminate
the safety culture and to homogenize the knowledge of scientists from different countries of the world, in a relevant area of the nuclear technology. In
addition, the ISP activity gives a real challenge to all participants to analyze an experiment in detail in
the frame of an international activity and compare the own calculation
results with other results (and the data). Furthermore it is a big challenge to all codes, which are used for
comparing with the other codes.The present work focuses on a limited part of the entire programme, making
reference to five ISPs that deal with phenomenon typical of small break LOCAs in PWRs. Four different
facilities based on experiments and an actual plant transient are involved. The considered set of
standard problems represent an answer in the system thermal-hydraulic area to the concerns raised
by the TMI-2 accident and have been proposed in a period when advanced codes have been made available; definitely, the
discussed ISPs and the advanced codes might be considered as complementary elements for ensuring reliability in
safety evaluations in the area of long lasting transients (as opposed to short transients like large break LOCA)
potentially affected by operator actions.In the frame of the presented activity, the involved experimental facilities and
the reference tests have been characterized adopting the list of twenty
two phenomena proposed when setting up the CSNI code validation matrix
for integral test facilities. This led to establishing qualitative similarities
among the different transient scenarios and demonstrated that the latest small
break LOCA ISPs, which were performed in the largest scale facilities, cover much broader
ranges of phenomena relevant to nuclear reactor thermal-hydraulics.Whatever is the kind of ISP, “blind,” “open,” “double blind,” the quality of a
calculation, that is, the degree of agreement between code results
and experimental data, depends upon several factors ranging from
capabilities of code physical models, to user experience, to nodalization
details and qualification, to the quality of the information
supplied by the experimentalists, integration of this information into the
input of the codes. So, as already mentioned, finalized conclusions
regarding the submitted calculations cannot be drawn without the direct contributions
of the code users and the experimentalists; on the other hand, this is the subject of the comparison reports issued by OECD
as a summary of each ISP, they are listed here as references.Considering the above, the conclusions reached are of a quite general nature and involve aspects
that are common to the different ISPs, as well as to small break LOCA related ISPs.It was noted that large numbers of countries (more than 20) and organisations (more than 50) took part at least in one small break
LOCA ISP: these essentially include all countries using nuclear power to generate electricity
(one exception strictly connected with political reasons can be observed).
However, only few organisations participated in all the considered ISPs and
many organisations took part in one ISP only. Furthermore, in the recent years the number of
code users increased and among these
users, there were less experienced ones; this must be considered carefully when deriving conclusions from the ISP activities. Assuming that the advanced
codes were available to most of the participants since the time of the
ISP 18 (first of the considered ISP), this together with the statistical evaluations done in the
frame of Section7 and [4], lead to the following conclusions.(a)
The objectives in the participation to the ISP changed over the time, being
mostly connected with code development at the beginning and mostly
focused toward user training in the latest ISP; this might not be true
for codes that did not reach an adequate maturity at the beginning of
the considered time frame.(b)
Notwithstanding the large effort necessary to organize or even to participate
in an ISP, the cumulative experience gained by a single organisation
or by a single group of scientists inside one organisation is
generally not transferable or at least has not been transferred. This is especially
true in a nonnegligible number of cases where the participant organisation
or the group of scientists dissolved and did not leave any track of
the acquired experience. This concerns code developers,
experimentalists, and code users, and may be considered as a problem common to
the whole area of system thermal-hydraulics.(c)
The ISPs got more demanding with the time. There was a significant progress in the code
capabilities; for example, the ISP 27 (BETHSY) could be calculated only with very
large difficulties (or in some cases could not be calculated at all) at the time period when
the ISP 18 (LOBI) was performed. A list of thirteen deficiencies coming from the considered ISPs and common to most of the utilized
codes has been identified as in Section 8.1. This is not an exhaustive
list, but underlines one positive result of ISP exercises. However, it must be observed that very
slow or almost no progress has been done in the identified areas in the past decade.An additional aspect that should be brought to the attention is that the ISPs are
not part of a general finalized code assessment programme that,
historically, has been the objective of cooperations
like International Code Assessment Program of USNRC (ICAP), Code Assessment and
Maintenance Program of USNRC, follow up to ICAP (CAMP), Club
des Utilisateurs du CATHARE (CUC), and so forth or of nationally funded researches.
In most of the cases, this prevented a direct improvement of codes based on the results
of ISPs (see also below), although code deficiencies detected in the frame of ISPs, owing to the relevance of the ISPs themselves, were always brought to the attention
of code developers.Furthermore, inadequacy or lack of direct feedback from the results of
ISPs to code model improvements is in some cases the consequence of the need to
fix time frames and deadlines; this may prevent the achievement of “optimized” results with an assigned code version. For some particular codes, too frequent releases
of different code versions also put obstacles as far as that feedback is concerned. The use of ISPs as exercise for
proving or achieving some user qualification, also contributed to the above conclusion.Although a detailed evaluation/judgment of each ISP activity is not the purpose of the effort done in the present
framework, it seemed worthwhile to add few specific conclusions applicable to single ISPs.(i) A large mismatch may exist between the huge effort from the host organisation and the participants as a whole on one side, and the
final result of the exercise.(ii) Incomplete or even misleading information supplied by the host
organisation in some cases testify of the complexities of the general code assessment problem and could hinder to
facilitate the achievement of meaningful conclusions.(iii) In some cases, participants underestimated the effort
necessary to set up suitable nodalization including correct
consideration of initial and boundary conditions; this constitutes an
additional reason preventing more satisfactory conclusions of the activities.(iv) Especially, as a consequence of the above, quite vague
formulations can be found in the general conclusions of the ISP reports.(v) A large range of results obtained
by participants using the same code version gives interesting information
about uncertainty in selection of input parameters and uncertainties of code
models as well as experimental data errors (see [11]).


### 9.1. Recommendations

General recommendations coming from the performed activity can be summarized as follows, covering different
aspects connected with small break LOCA ISPs.(i) The participation into ISP activities of non-OECD countries should be
continuously encouraged;
especially small countries not having the capabilities for wide national research
programmes, can get
substantial benefits from ISPs.(ii) Notwithstanding obvious drawbacks (e.g., lack of suitable instrumentation, inaccuracy of data base,
etc.) a future ISP based on an actual plant transient, if any, is highly
recommended.(iii) A better characterization of the experiments of
ISPs, also in view of a qualitative evaluation of code performance, could be based on the 67 phenomena identified for the
CSNI separate effects tests code
validation matrix made available in mid 90 s [22, 23], future ISPs should directly consider this.(iv) The interaction between ISP host/proposing organisation and CSNI working
groups has been quite satisfactory as far
as the test selection is concerned, but could be improved especially in relation to the evaluation of the results and for
defining the impact of these in the thermal-hydraulic
and nuclear safety areas.(v) The inadequacy of a direct feedback (indirect feedback may exist) between ISPs
results and code developers has already been stressed. However, indirect
feedback exists, as ISPs revealed the
important role played by physical phenomena such as phase separation at the junctions,
stratification in horizontal components (ISP 18), or secondary side heat
transfer (ISP 27). Then, valuable information
for improving the code model must be the result of independent confirmatory analyses performed utilizing
data from separate effects tests facilities (SETF), for example, a code
inadequacy possibly identified when performing the analysis of one ISP in an integral test facility should be confirmed
and characterized by calculations based on SETF experiments. In this sense, SETF-based ISPs are also strongly
recommended.(vi) The list of code deficiencies given in the Section8.1 could be
used as basis for planning future ISPs
in separate effects tests facilities together with phenomena relevant in 2D/3D
geometrical configurations. Clearly, codes should also be improved
as far as possible, when a model inadequacy is found.(vii) “Blind” types of ISPs should be preferred to “Open” types,
especially when a posttest (“Open”) phase of the ISP can be planned and
reliable data can be supplied to the participants since the
beginning. This gives a better opportunity to evaluate the user effect and
better represents the overall situation that is faced when performing
plant related calculations.(viii) The experience acquired so far, the database available from different national and international programmes and the cost of an ISP,
suggests not to propose additional ISPs in the frame of small break LOCAs;
transients evolving at low pressure, scenarios involving complex accident management procedures or of
specific interest for the new generation reactors are not part of this recommendation.(ix) Some of the discussed ISPs have been utilized as sample basis for
addressing the problems of user effects and quantification of the accuracy of calculation
results. However, some specific
efforts should be devoted from future ISP host organisations, possibly in cooperation with CSNI, in the
areas of user effects, user qualification, and quantification of the accuracy. It could even be
standard part of the ISP activity.(x) In relation to user effect, in a long-term view, a part of the problem
can be solved by improved codes, which remove the need for the user to make ad hoc
assumptions in order to compensate for code limitations or complete lack of modelling; an
example of this is modelling pressure drop at geometric discontinuities.(xi) In connection with the above, when applicable, the
problem of evaluating the uncertainty by system thermal-hydraulic codes when
predicting scenarios relevant to nuclear power plants could
be addressed in the frame of activities similar to the ISPs.Finally, considering the effort expended in the preparation of ISPs, it would be
very useful if this information was
catalogued and stored so that it could be easily accessed for future posttest analyses.


## 9.1. Recommendations

General recommendations coming from the performed activity can be summarized as follows, covering different
aspects connected with small break LOCA ISPs.(i) The participation into ISP activities of non-OECD countries should be
continuously encouraged;
especially small countries not having the capabilities for wide national research
programmes, can get
substantial benefits from ISPs.(ii) Notwithstanding obvious drawbacks (e.g., lack of suitable instrumentation, inaccuracy of data base,
etc.) a future ISP based on an actual plant transient, if any, is highly
recommended.(iii) A better characterization of the experiments of
ISPs, also in view of a qualitative evaluation of code performance, could be based on the 67 phenomena identified for the
CSNI separate effects tests code
validation matrix made available in mid 90 s [22, 23], future ISPs should directly consider this.(iv) The interaction between ISP host/proposing organisation and CSNI working
groups has been quite satisfactory as far
as the test selection is concerned, but could be improved especially in relation to the evaluation of the results and for
defining the impact of these in the thermal-hydraulic
and nuclear safety areas.(v) The inadequacy of a direct feedback (indirect feedback may exist) between ISPs
results and code developers has already been stressed. However, indirect
feedback exists, as ISPs revealed the
important role played by physical phenomena such as phase separation at the junctions,
stratification in horizontal components (ISP 18), or secondary side heat
transfer (ISP 27). Then, valuable information
for improving the code model must be the result of independent confirmatory analyses performed utilizing
data from separate effects tests facilities (SETF), for example, a code
inadequacy possibly identified when performing the analysis of one ISP in an integral test facility should be confirmed
and characterized by calculations based on SETF experiments. In this sense, SETF-based ISPs are also strongly
recommended.(vi) The list of code deficiencies given in the Section8.1 could be
used as basis for planning future ISPs
in separate effects tests facilities together with phenomena relevant in 2D/3D
geometrical configurations. Clearly, codes should also be improved
as far as possible, when a model inadequacy is found.(vii) “Blind” types of ISPs should be preferred to “Open” types,
especially when a posttest (“Open”) phase of the ISP can be planned and
reliable data can be supplied to the participants since the
beginning. This gives a better opportunity to evaluate the user effect and
better represents the overall situation that is faced when performing
plant related calculations.(viii) The experience acquired so far, the database available from different national and international programmes and the cost of an ISP,
suggests not to propose additional ISPs in the frame of small break LOCAs;
transients evolving at low pressure, scenarios involving complex accident management procedures or of
specific interest for the new generation reactors are not part of this recommendation.(ix) Some of the discussed ISPs have been utilized as sample basis for
addressing the problems of user effects and quantification of the accuracy of calculation
results. However, some specific
efforts should be devoted from future ISP host organisations, possibly in cooperation with CSNI, in the
areas of user effects, user qualification, and quantification of the accuracy. It could even be
standard part of the ISP activity.(x) In relation to user effect, in a long-term view, a part of the problem
can be solved by improved codes, which remove the need for the user to make ad hoc
assumptions in order to compensate for code limitations or complete lack of modelling; an
example of this is modelling pressure drop at geometric discontinuities.(xi) In connection with the above, when applicable, the
problem of evaluating the uncertainty by system thermal-hydraulic codes when
predicting scenarios relevant to nuclear power plants could
be addressed in the frame of activities similar to the ISPs.Finally, considering the effort expended in the preparation of ISPs, it would be
very useful if this information was
catalogued and stored so that it could be easily accessed for future posttest analyses.



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*Source: 814572-2008-04-21.xml*