RSICC CODE PACKAGE MIS-016

 

1. NAME AND TITLE

 

CCVM Database: CSNI Code Validation Matrix of Thermo-Hydraulic Codes for LWR LOCA and Transients.

 

RESTRICTIONS:

 

RSICC is authorized to distribute CCVM Database for research and education purposes only. Requesters from NEA Data Bank member countries are advised to order CCVM Database from the NEA Data Bank. Non-commercial and non-profit users from other OECD member countries (specifically Canada and the United States) may order CCVM-Database from RSICC.

 

2.  CONTRIBUTORS

 

OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France.

 

3.  HISTORICAL BACKGROUND AND INFORMATION

 

Over the years the NEA Data Bank collected a sizable subset of separate effects test reactor transient and LOCA integral test data (I.T.D.) as defined in the Code Validation Matrix of Document OCDE/GD(97)12. These data with accompanying documentation are available on DVDs. The writing format of the DVD conforms to the standard ISO 9660. Each DVD contains a copy of the INDEX file. It summarizes the complete contents of all DVDs. The reports describing the experiments have been electronically scanned and transformed into PDF files. Each report is stored in a separate subdirectory.


The Integral Experiments are designed to follow the behaviour of a reactor system in various off-normal or accident conditions. The ITD matrix data is suitable for the validation of best estimate thermal-hydraulic computer codes: it consists of phenomenologically well-founded experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of the test predictions.

 

4.  DESCRIPTION OF TEST FACILITIES

 

The BETHSY integral test facility located in the Nuclear Research Center in Grenoble (France) is a scaled down model of a 3 loop 900 eMW FRAMATOME PWR; the overall scaling factor applied to every volume, mass flowrate and power level is close to 1/100, while elevations are 1/1 in order to preserve the gravitational heads.

 

DOEL 2 is a Westinghouse, 2 loop pressurized water reactor (PWR) rated at 392 Mwe (NET) and commissioned in 1975, for which TRACTEBEL was the architect/engineer. This plant located in Belgium in part of the twin concept with DOEL 1, as they share some common engineered safety systems such as the high pressure safety injection system (HPSI).

 

The FIST facility is scaled to a BWR/6-218 standard plant. A full size bundle with electrically heated rods is used to simulate the reactor core. A scaling ratio of 1/624 is applied in the design of the system components. Key features of the FIST facility include:

(a) Full height test vessel and internals;
(b) correctly scaled fluid volume distribution;
(c) simulation of ECCS, S/RV, and ADS;
(d) level trip capability;
(e) heated feedwater supply system, which provides the capability for steady state operation.

 

The FIX-II facility is a volume scaled 1:777 representation of a Swedish BWR with external pumps. The pressure vessel contains a 36 rod full length bundle and a spray condenser at the top to allow steady state operation. The downcomer, bypass channels and guide tube volumes are represented by external piping. The intact loop represents three of the four external reactor loops. The broken loop is constructed such that both guillotine breaks and split breaks may be simulated. The facility is equipped with ADS-simulation, but no ECCS injection are included. The FIX-II loop is also suited to investigate response of pump trips and MSIV closures in internal pump reactors.

 

The Leibstadt Nuclear Power Station is equipped with a direct cycle boiling water reactor belonging to the General Electric BWR product line BWR/6.  The MARK III containment system encloses the nuclear island.  The nuclear system is provided with a 238-inch internal diameter vessel and the core is built up of 648 fuel elements and 84 control rods. Each fuel bundle consists of 62 fuel rods and 2 water rods in an 8 x 8 array.  The rated power of the Leibstadt BWR is 3012 MWt and it is designed for a net power of 942 MWe.

 

The LOBI facility is a 1/700 scale model of a four loop PWR and has two primary loops, the intact loop representing three loops and the broken loop representing one loop of a four-loop PWR.  The reactor pressure vessel model contains an electrically heated rod-bundle with 64 rods and a heated length of 3.9 m.  The nominal heating power is 5.3 MW.  The downcomer is of annular shape.  An upper head simulator is connected to the vessel. Each of the two primary loops contains a pump and a steam generator. The different mass flows in the loops are established by the pump speeds, since the two pumps are identical. Heat is removed from the  steam generators by a secondary system.  ECC water can be supplied from two accumulators, one for each loop.  Cold or hot leg as well as combined injection can be simulated.
     

The LOBI-MOD2 test facility is a high pressure integral system blowdown refill test facility designed, constructed and operated at the Joint Research Centre of the European Communities, Ispra Establishment, Italy. It represents and extensive modification of the original MOD1 configuration and was put into operation after June 1982.

 

The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA.  The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)].  For some components, this factor is not applied; however, it is used as extensively as practical.  In general, components used in LOFT are similar in design to those of a LPWR.  Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA.

 

OTIS is a 1:1686 volume-scaled single loop (one hot log and one cold leg) facility with critical elevations preserved. The single once-through steam generator contains 19 tubes prototypical of a B&W PWR plant. The electrically heated core has a power capability of 180 KW which is representative of 10 per cent scaled power based on a 2584 MWth plant. It contains no primary pump but all other active components are simulated and it operates at full PWR system pressure.

 

The PIPER-ONE facility simulates a General Electric BWR-6 plant. It is characterized by volume and height scaling ratios of 1/2200 and 1/1, respectively. The available core rod electrical power (20% of the nominal value) is sufficient to simulate the nuclear heat decay. No circulation loops are included in the facility, considering their low importance in a small break LOCA and the willingness to achieve the maximum simplicity of the loop operation.

 

PKL-facility simulates the essential primary system components of a typical West German 1300 PWR with regard to their thermohydraulic behaviour. The facility essentially consists of the pressure vessel with the heated bundle, the downcomer simulator, the primary loops with the components steam generator and pump simulator, the injection devices, the break geometry simulator, as well as the separators connected thereto, and the test containment to maintain a back-pressure at the location of break which is expected to be typical for emergency conditions. The number of heater rods and the cross-sections of the testing plant are on a reduced scale 1:134 in comparison with a typical German PWR. The elevations and locations are essentially full scale.

 

ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems:

(a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod;
(b) steam line and feed water line, which are independent open loops;
(c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop;
(d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS.

 

ROSA-IV is a large scale test facility (LSTF) for integral simulation of thermal-hydraulic response of a pressurized water reactor during a small break loss-of-coolant accident or an operational transient. The facility has electric core heating. The overall scaling factor is 1/48. The hot and cold legs were sized to conserve the volume scaling. The four primary loops of the reference PWR are represented by two equal-volume loops. 

 

The Semiscale Mod-1 system used for this test consisted of a pressure vessel with internals, including a 40-rod core with 36 electrically heated rods; an intact loop with steam generator, pump, and pressurizer; a broken loop with simulated steam generator, simulated pump, simulated reflood bypass lines, LOFT counterpart nozzles, and two rupture assemblies; a coolant injection accumulator for the intact loop; high and low pressure suppression tank, header, and heated steam supply system.

 

The Semiscale Mod-2A system used for this test was designed as a smallscale model of the primary system of four loop PWR nuclear generating plant. consists of a pressure v  internals, including a 25-rod core with electrically powered rods and an external downcomer assembly; an intact loop with pressurizer, steam generator, pump, and rupture disc assembly. It is scaled to simulate the three intact loops in a PWR. The broken loop simulates the single loop in which a break is postulated to occur in a PWR. The system also has emergency core cooling for the intact loop from high- and low-pressure injection pumps, a coolant injection accumulator for the intact loop, and a pressure suppression system with steam supply and pressure suppression tank. Twenty-three rods of the 25-rod core were powered equally with a power density of 26.51 kW/m. The total core power for Test S-IB-3 was 1.45 MW.

 

The Semiscale Mod-3B system is scaled to a reference four-loop PWR. The scaling criterion is a modified volume scaling based on the ratio of Semiscale power to the thermal power of the reference plant (Trojan). This scaling produces a scale factor of 1/1705.5. The system consists of a pressure vessel with an electrically-heated, 25-rod PWR core simulator and internals, an external pipe downcomer, and two primary coolant loops. Each loop has an active tube and shell steam generator. The intact loop is scaled to represent three of the four primary loops in a PWR, while the broken loop represents the fourth. Even though S-PL-3 does not incorporate a break, this loop is referred to as the "broken loop". In order to correctly scale the facility and preserve important phenomena, component elevations, dynamic pressure heads, and liquid distributions are maintained as close to the reference PWR values as possible.

 

The SPES (Simulatore PWR per Esperienze di Sicurezza) integral test facility is a three loop scaled-down model of PWR (Westinghouse 3122 type, 3 loops, 2775 MWth core power) designed for thermal-hydraulic safety research program. SPES test program provides experimental data for the development and assessment of system codes used in PWR safety analysis. SPES is an experimental facility which allows a true simulation of a PWR system as close as possible to the characteristic of the Westinghouse 312 type.  The test plant reproduces the primary loops, the most important components and the power channel of the simulated reactor according to significant scaling criteria. The SPES experimental facility, having a 1:427 power-scaling ratio includes a full-length-scale electrical heated power channel and three complete primary loops.  In particular SPES system simulates:

(a) the whole primary circuit;
(b) the secondary system restricted to:
   - steam generator secondary side;
   - main feedwater lines downstream the isolation valves;
   - main steam lines upstream the turbine stop valve.
(c) the most significant auxiliary and emergency systems:
   - charging and letdown systems;
   - safety injection systems (HPIS, LPIS, accumulators, EFWS);
   - steam dump.

 

Two Bundle Loop facility.  Volume scaling ratio = 1:328.  Height ratio = 1:1.  The facilities include the following to simulate a BWR system: pressure vessel, main steam system, feedwater, recirculation, ECCS, ADS, break line, utility system, power supply and instrumentation system.  The pressure vessel has two heated simulated fuel assemblies.

 

The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations.

 

5.  DESCRIPTION OF TESTS

 

BETHSY/6.9C, Loss of residual heat removal system during mid-loop operation

The aim of this test performed in 1992 was to study the accident transient following a failure of the residual heat removal system with the pressurizer and SGI outlet plenum manways open. There was no non condensable gas in the test.  The objectives were:

Simulation of integral plant behaviour under atmospheric conditions (open system) with anticipated loss of Residual Heat Removal System during mid-loop operation
- study physical phenomena under very low system pressure, in particular the behaviour of pressurizer and surge line
- core uncovery and reflooding
- effects of loss of primary coolant and refilling by RHRS
- 'open' calculations requested to allow assessment of code applicability for conditions not anticipated during development of codes (extreme low pressure)

 

BETHSY/9.1B, Cold Leg Break Test

The BETHSY test 9.1b (performed in December 1989) involves 2" cold leg break, combined with the High Pressure Injection System (HPIS) failure. In that case, the state oriented approach requires operators to start an Ultimate Procedure, which consists in fully opening the Steam Generator (SG) atmospheric dumps as soon as they are informed of the unavailability of the HPIS. The presently studied scenario assumes a delayed application of this procedure, which is started only when the core outlet temperature rises significantly higher than the saturation temperature.

 

DOEL2/SGTR, Steam Generator Tube Rupture incident at the DOEL 2, Westinghouse 2 loop PWR

Data measured during the Steam Generator Tube Rupture (SGTR) incident on full scale facility are available.

 

FIST/4DBA1, BWR/4-218 Simulated Double-Ended Large Break Test

4DBA1 simulates conditions in the peak power bundle of BWR/4-218 after a double-ended large break in one of the recirculation suction lines. The BWR/4 break size is much larger than the corresponding BWR/6 break. The scaled break sizes are 1.05 sq.in. in the suction line and 0.1645 sq.in. at the drive line nozzle based on the single bundle FIST scaling. The initial power of 6.09 MW simulates a peak power bundle.

 

FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE

Test 6IB1 investigates system responses to an intermediate break in the recirculation suction line. BWR system licensing evaluations for various size recirculation break LOCA's indicates that a break size of about 0.2 sq.ft., without LPCS operation, is the highest PCT case for the intermediate break LOCA. Test 6IB1 simulates this event.

 

FIST/6MSB1, BWR/6 Main Steamline Double-Ended Break Test

The main steamline break test 6MSB1 simulates a BWR/6 response with a double ended break upstream of the flow limiter in one of the four main steamlines. The effective break area consists of one main steamline flow area and one steam line flow limiter, due to back flow via the bypass header. The turbine bypass opens at about 1 second and the effective break flow area at this time also includes the bypass line area. By about 5.5 seconds, the main steam isolation valve is closed and the break is limited to one main steamline. For simplifying the test operation, and having a bounding simulation, the test was performed with full turbine bypass flow are from 0 to 5.5 seconds and one steamline flow area beyond 5.5 seconds.

 

FIST/6PMC1, BWR/6-128 Isolation Valve (MSIV) Closure without Power Scram

Isolation Valve (MSIV) closure without power scram. HPCS and RCIC are assumed to be functional.The power transient test simulation is based on a transient code calculation for a BWR.

 

FIST/6SB1, BWR/6 Simulated Recirculation Line Break

This test simulates a BWR/6 recirculation line break of 0.05 squ.ft. with a stuck open safety relief valve. In addition, similar to test 6SB2C, HPCS is assumed to be unavailable. Test 6SB1 is performed with a BWR/6 core average power of 4.64 MW. A nominal ADS time delay of 105 seconds is used. ECC water temperature is 120 deg.F. The system pressure does not reach the SRV opening setpoint during the period between MSIV closure and ADS activation.

 

FIST/6SB2C, BWR/6 Recirculation Suction Line Break Test

This small break test simulates a BWR/6 recirculation suction line break of 0.05 sq.ft. with HPCS assumed to be unavailable. Test conditions were: initial bundle power: 5.05 MW; core flow: 42 lb/sec; ADS time delay : 120 seconds; ECC water  temperature: 90 deg.F.

 

FIST/T1QUV, Simulated Failure to Maintain Water Level in BWR/6-218

This test simulates a failure to maintain water level. The reactor immediately scrams, which eventually leads to a MSIV closure. Inventory loss by the decay power boil-off results in bundle uncovery and rod heatup. The FIST test is terminated as temperatures reach 1200 deg.F to protect the bundle heater rods.

 

FIST/T23C, Simulated Failure to Maintain Water Level in BWR/6-218

Test T23C simulates a failure to maintain water level. Inventory loss eventually leads to bundle uncovery and fuel rod heatup. This test investigates system response up to the time when the bundle rod heatup reaches the FIST heater temperature limit of 1200 deg.F.

 

FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients

Pump trip experiment 2032 was a part of test group 2, i.e. the mass flow transient was to simulate the pump coast down with a pump inertia of 11.3 kg-m**2.  The initial power in the 36-rod bundle was 4.44 MW which gave dryout after 1.4 s from the start of the flow transient.  A maximum rod cladding temperature of 457 degrees C was measured. Rewetting was obtained after 7.6 s.

 

FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break

Test 3025 simulates an intermediate size split break in one of the four main recirculation lines. The break area was 31 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle was 3.38 MW, corresponding to the hot channel power of the reactor.

 

FIX-II/3061, BWR FIX-II Pump Trip Experiment 3061, Large Split Break

Test No. 3061 is a simulated large split break. The break area was equal to the scaled down are of one of the four recirculation pipes in the reactor (100 %). The initial power should be equivalent to the average bundle power of the reactor and became here 2.51 MW, i.e. slightly higher.

 

FIX-II/5052, BWR FIX-II Pump Trip Experiment 5052, Guillotine Break Simulation

LOCA test No. 5052 simulated a guillotine break in one of the four main recirculation lines. The total break area, equally divided in the two discharge lines, was 200 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle in the FIX loop was 3.36 MW, corresponding to the hot channel power of the reactor.

 

FIX-II/6261, BWR FIX-II Pump Trip Experiment 626, Transient Dryout Tests

Test 6261 is one of a series of transient dryout tests to simulate events following pressurization transients in a BWR. An internal recirculation pump BWR type is simulated in this test. The test is initiated by positive ramps of pressure and bundle power.

 

LEIBSTADT/STP-2001, BWR/6, Reactor Core Isolation Cooling System Test

The purpose of the test STP-2001 which occurred on 15 October 1984 was to demonstrate the capability of the Reactor Core Isolation Cooling (RCIC) system to maintain the reactor water level above level 1 (-337 cm) following a total loss of feedwater with the HPCS system unavailable.  The test was initiated by tripping the two operating feedwater pumps.  The HPCS was prevented from initiating automatically.

 

LOBI/A1-04R, Loop for Blowdown Investigation, PWR Double-Ended Cold-Leg Break Test

The Test A1-04R was the last test in the power ascension series of the main LOBI experimental programme A1. In A1-04R a double-ended cold-leg break between primary pump and pressure vessel with intact loop cold leg ECC injection was simulated. ECC was injected through accumulator of intact loop into cold leg. The pressurizer was connected to the intact loop. At break initiation, the core power stayed at 5.12 MW (100%) during 3.2 seconds and was then stepwise reduced and reached zero at 50 seconds.

 

LOBI/A1-06, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break

The Test A1-06 is one of a series of three LOBI tests in which the ECC injection mode was studied.  In all experiments, a double-ended 2A break in the cold leg was simulated. The downcomer gap width was 12 mm.  In test A1-06, ECC water was injected from the intact loop accumulator into the hot leg and the cold leg and from the broken loop accumulator into the hot leg only. The pressurizer was connected to the intact loop. At break initiation, the core power stayed at 5.29 MW (100%) during 2.9 seconds and was then stepwise reduced and reached zero at 30 seconds. After blowdown initiation, the intact loop pump speed decreased to 71% after 8.0 seconds and remained constant. The broken loop pump speed diminished to 10.4% after 3.0 seconds and reached zero speed after 27.0 seconds.

 

LOBI/A1-66, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break

The Test A1-66 is one of a series of three LOBI tests in which the ECC injection mode was studied.  In all experiments, a double-ended 2A break in the cold leg was simulated. The downcomer gap width was 12 mm. In test A1-66, ECC water was injected from the intact loop accumulator into the cold leg only. The pressurizer was connected to the intact loop.    After blowdown initiation, the intact loop pump speed decreased to 72% after about 8 seconds and remained constant. The broken loop pump speed diminished to 10% after 3.0 seconds and reached zero speed after about 27 seconds.

 

LOBI/A2-77, Loop for Blowdown Investigation, PWR MOD2 Small Leak Programme Experiment

Test A2-77A represents one of the first experiments of the LOBI-MOD2 Small Leak Programme, which was started in April 1984. It is designed to establish the natural circulation behaviour of the LOBI-MOD2 primary loop under single phase and two phase flow conditions without rupture in the primary loop.

 

LOBI/A2-81, Loop for Blowdown Investigation. PWR MOD2 1% Cold-Leg Break

Test A2-81 simulates a 1% cold leg break in the main coolant pipe of a pressurized water reactor. Cooldown is affected via the secondary loop at a rate of 100 K/h. Emergency cooling water is injected into the primary loop by cold leg high pressure injection (HPIS) representative of two injection pumps; of the remaining two pumps in the reference plant, one is assumed to be in maintenance and the other connected to the broken loop.

 

LOBI/B-R1M, Loop for Blowdown Investigation, PWR Single-Ended Cold-Leg Break Experiment B

The test B-R1M is one of the LOBI experimental program B which was performed in the LOBI test facility.  This test belongs to the interim test program which comprises a total of five large cold leg break tests.  In the B-R1M a single ended cold leg break between the primary pump and the pressure vessel, with ECC water injection into the intact cold loop leg only, was simulated.  The double ended rupture device was used.  The flap on the pump side remained closed and the flap on the vessel side was opened. The isolation valve between the two flaps was maintained in the open position.  In this test the pressurizer was connected to the intact loop.  Test B-R1M represents the smallest large-break test of a series of four which were envisaged to cover the large break spectrum and thus substantiate the influence of break size on blowdown in the case of cold leg break and with intact loop cold leg ECC water injection only.  In these tests the downcomer gap width was 50 mm.

 

LOBI/BT-00, Simulation of a Loss of Feedwater Transient (LOFW)

LOBI tests provide experimental data for checking and improving computer codes. In addition, they demonstrate the performance of reactor safety systems and procedures.  MOD2 is an extension of MOD1 25 large break loss-of-coolant experiments by the addition of components as needed for small break tests, e.g. scaled steam generators, a high pressure injection system (HPIS), secondary auxiliary feedwater injection (AFW), additional instrumentation and complete thermal insulation.  Test BT-00 belongs to the LOBI special transient test program and is part of the B-test series. Within the LOBI special transient test matrix this test is also called "ST-2" (scoping test no.2). Test BT-00 simulates a loss of main feedwater (LOFW) followed by a steam generator dryout and a subsequent cooldown via primary bleed and feed.


Scaling: The power input, the primary circuit coolant mass flow and volume are scaled down from the reactor values by a factor of 712. All the other most relevant quantities such as operating temperature, pressure, lengths and pressure drops along heat transfer surfaces have been scaled 1:1. Also the absolute heights and relative elevations of the individual system components have been kept at reactor values thus preserving the gravitational heads.

 

LOFT/L2-3, Loss of Fluid Test, 2nd NRC L2 Large Break LOCA Experiment

This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m.

 

LOFT/L2-5, Loss of Fluid Test, 3rd NRC L2 Large Break LOCA Experiment

This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coastdown.

 

LOFT/L3-5, Loss of Fluid Test, 5th NRC L3 Small Break LOCA Experiment

This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg noncommunicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980.

 

LOFT/L3-6, Loss of Fluid Test, 6th NRC L3 Small Break LOCA Experiment

This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg noncommunicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980.

 

LOFT/L3-7, Loss of Fluid Test, 7th NRC L3 Small Break LOCA Experiment

This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg noncommunicative-break LOCA was simulated. The experiment was conducted on 20 June 1980.

 

LOFT/L6-7, Loss of Fluid Test, Anticipated Transients with Multiple Failures

This was the seventh in the NRC L6 Series of Anticipated Transients experiments. Rapid secondary side induced cooldown was studied. The experiment was conducted on 31 September 1981.

 

LOFT/L8-2, Severe Core Transient Experiment

Experiment L8-2 is the second experiment in the LOFT Severe Core Transient Experiment Series L8, and was designed to evaluate the effect of primary coolant pump restart on core cooling when the primary coolant system (PCS) is highly voided of liquid. The accumulator and low-pressure injection system were not allowed to inject fluid into the PCS until after core uncovery has occurred and the PCS pumps had been restarted. PCS pump restart did not produce a moderation in the core thermal transient, and Accumulator A flow was unblocked by the operator at a preselected temperature. All plant protective systems were triggered automatically shortly thereafter, followed by a rapid reflood of the PCS.

 

LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures

This was the third of the NRC L9 series of experiments on Anticipated Transients with Multiple Failures. Loss-of-feedwater effects were studied. The experiment was conducted on 7 April 1982.

 

LOFT/LP-02-6, Loss of Fluid Test, 1st OECD Large Break Experiment

The fourth OECD LOFT experiment was conducted on 3 October 1983.  This was the first OECD LOFT large break experiment.  The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs.  This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle.  The experiment was initiated by opening the quick-opening blow-down valves in the broken hot and cold legs.

 

LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment

The seventh OECD LOFT experiment was conducted on 19 December 1984.  It was the first of the two experiments to be performed in the LOFT facility with intentional release of fission products.  Its objectives were to obtain data on fission product release from the fuel-cladding gap into vapor and reflood water and to collect data on transport of these fission products through and out of the reactor coolant system.  The experiment was initiated by a reactor scram with one second delayed opening of the quick-opening blowdown valves.

 

LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel

The eighth OECD LOFT experiment was conducted on 7 March 1985.  It was the second of the two experiments to be performed in the LOFT facility with intentional release of fission products.  Its principal objectives were to determine the fission product release from the fuel during a severe fuel damage scenario and the subsequent transport of these fission products in a predominantly vapor/aerosol environment.  This was the largest severe fuel damage experiment ever conducted, and serves as an important benchmark between smaller scale tests and the TMI-2 accident.

 

LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

The first OECD LOFT experiment was conducted on February 20, 1983.  It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient.  The objective of the experiment was to investigate the performance of primary "feed and bleed" using a "bleed" from the PORV and "feed" from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory.

 

LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development.  The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs).  This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR.  Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coastdown, and United Kingdom minimum safeguard emergency core coolant injection rates.  The nuclear fuel rods were not pressurized.  The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs.

 

LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

The second OECD LOFT experiment was conducted on 23 June 1983.  It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop.  The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip.  The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs.

 

LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

The third OECD LOFT experiment was conducted on 14 July 1983.  It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop.  The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip.  The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs.

 

LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressureinjection System (HPIS)

The sixth OECD LOFT experiment was conducted on 5 March 1984.  It simulated a 1.8-in cold leg break LOCA with no HPIS available.  This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncovery and addressed accumulator injection at low pressure differentials.

 

OTIS/220100, Once-Through Integral Systems, 2-Phase Natural Circulation and Reflux

Two-phase natural circulation and reflux.

 

OTIS/220402, Once-Through Integral Systems, Cold Leg Small Break LOCA

Cold leg small break LOCA.

 

PIPER-1/PO-SB-7, SBLOCA Simulation of Break in BWR-6 Plant at PIPER1 GE BWR Simulator

Test PO-SB-7 simulates a SB-LOCA in a GE BWR-6 plant with the 2.6% break located in the downcomer. Recirculation pump trip, closure of feedwater and of steam lines and scram are assumed to occur at the time of break detection. HPCS and RCIC systems are assumed to be unavailable during the transient, while the actuation of ADS and of low pressure injection systems (LPCI and LPCS) are foreseen, following the signals of low level in the downcomer and of low pressure in the RPV, respectively. The SRV system is assumed to operate in the early part of the transient to limit primary system pressure excursion. The main objectives of the test are:
- the evaluation of the facility behavior in relation to the most significant thermalhydraulic quantities (pressure, temperatures of heating rods, mass and energy overall balances, etc.);
- the demonstration that ADS intervention allows the injection of coolant by low pressure ECCS before the mass depletion in the loop causes important dry-out of heating rods;
- the demonstration of the facility capability to correctly simulate the complex accident scenario in the reference BWR-6, considering its limitations.

 

PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

Test K 9 out of a series PKL-IB was conducted on May 30, 1979 by Kraftwerk Union (KWU) at Erlangen (Germany). The objective of the integral cold leg injection test K 9 (double-ended 200%-break) was to investigate after a LOCA the refill and reflood process within a pressure vessel considering the influence of the simulated primary loops of a pressurized water reactor (PWR). As to the number of rods and the cross-sections of vessel and components the PKL test-facility is scaled down to 1:134 of a typical West Germany PWR. The elevations and locations have been designed in order to reproduce substantially the original extension. Compared to other cold leg injection tests in the same facility (K5. 3a, K5. 4a) test K9 differs in a uniform radial power profile in the core, an increased temperature of injected Emergency Core Cooling (ECC) water (feed water, from 35 to 53 deg.C), an injection mass flow rate highly reduced to 1:3, and a history of it corresponding to that of a typical US-PWR, on the average somewhat higher maximum initial temperatures in the bundle, and a smaller bundle heating power during the initial phase.

 

ROSA-III/912, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test

This is one of the small break LOCA/ECC test series to study the response of a BWR with ECC injection. Run 912 is a 5% split break test at the recirculation pump inlet. The test is initiated with the steam dome pressure at 7.30 MPa, the lower plenum subcooling at 10.8 K, a core inlet flow rate of 16.4 kg/s, and a core heat generation rate of 3.9 MW. The core is quenched after ECCS actuation and at a maximum fuel cladding temperature of 839 K.

 

ROSA-III/916, BWR Rig of Safety Assessment for LOCA, 50% Split Break Test

Run 916 was a 50% split break test at the recirculation pump suction line. A failure of the HPCS diesel generator was assumed. At 190 seconds after the break, a peak cladding temperature of 917 K was reached during the reflooding phase. The full core could be effectively quenched by ECCS.

 

ROSA-III/922, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test

Run 922 is a LOCA test with a 5% split break at the recirculation pump suction. HPCS was assumed to have failed. Initial conditions of the test were as follows: steam dome pressure 7.35 MPa; lower plenum temperature 552.6 K; core inlet flow rate 16.0 kg/s; core heat generation rate 3.96 MW. Core power before break initiation was maintained at 44% of the scaled steady-state power and then reduced along the decay curve that simulates the total heat transfer rate. The test was terminated, when the entire core was quenched.

 

ROSA-III/923, BWR Rig of Safety Assessment for LOCA

Core thermal hydraulics; parallel channel effects and instabilities; void collapse and temperature distribution during pressurization; critical power ratio.

 

ROSA-III/926, BWR Rig of Safety Assessment for LOCA, 20% Double-Ended Break

RUN 926 simulates a 200% double-ended break in the recirculation line with failure of HPCS. Initial conditions of the test were as follows: steam dome pressure 7.37 MPa; lower plenum temperature 553 K; core inlet flow rate 16.3 kg/s; core power before break initiation 3.967 MW. During the transient, the power was changed along the curve which simulates the total heat transfer rate in the core of the reference BWR.

 

ROSA-III/952, BWR Rig of Safety Assessment for LOCA, Reference MSL Break Test

Run 952 is a reference MSL break test performed with a 100% break upstream of the main steam isolation valve and with full ECCS actuation logic. The MSL break is characterized by a system depressurization due to a break flow of high mass quality, which is slower than with a recirculation line break. Continuous flashing of the fluid in the pressure vessel was observed. A slow decrease in the downcomer water level eventually led to actuation of HPCS, but not LPCS or LPCI. About 2/3 of the core was uncovered, but the coolant level recovered quickly following HPCS injection. A peak cladding temperature of 752K was reached.

 

ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was reflooded by LPCS alone.

 

ROSA-III/984, BWR Rig of Safety Assessment for LOCA, 2.8% Split Break Test

Run 984 simulated a 2.8% split break at the recirculation pump suction line in the broken recirculation loop of a reference BWR. The core power is kept constant for 8 seconds into the break and then reduced along the curve that simulates the surface heat flux of a high power nuclear BWR rod. The feedwater supply is shut off between 2 and 4 seconds after the break. The coolant recirculation pumps are tripped to start coasting down at break initiation

 

ROSA-IV/SB-CL-18, Large Scale Test Facility, 5% Cold-Leg Break Test

SB-CL-18 was a 5% cold leg break test. Both the initial steady-state conditions and the test procedures were designed to minimize the effects of scaling compromises. 10 seconds into the break, the turbine throttle valve was closed at a pressurizer pressure of 12.97 MPa. The turbine bypass was inactive, since loss-of-offsite power was assumed concurrently with the scram. The reactor coolant pumps stopped at 265 seconds. The core was uncovered between 120 seconds and 155 seconds into the break.

 

ROSA-IV/SB-CL-27, Large Scale Test Facility, Gravity-Driven Safety Injection

SB-CL-27 simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a PWR. The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred.

 

SEMISCALE/S-06-3(LV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR

Test S-06-2 was conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of- coolant accident in a pressurized water reactor (PWR) system. Initial conditions were 15513 kPa and 563K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density (52.5 kW/m).

 

SEMISCALE/S-06-3(SV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR

Test S-06-2 was conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of- coolant accident in a pressurized water reactor (PWR) system. Initial conditions were 15513 kPa and 563K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density (52.5 kW/m).

 

SEMISCALE/S-IB-3, Semiscale MOD-2A, 21.7% Communication Cold Leg Break LOCA Experiment

This test was a 21.7%, communicative, cold leg break loss-of-coolant experiment with ECC injection. The test was intended to provide reference data for comparison of Semiscale test results to result from LOBI test B-R1M. The test was also intended to provide reference data for evaluation and assessment of reactor safety code capabilities to predict integral blowdown, refill/reflood experiments for intermediate break sizes, and particularly for providing data to extend the code into the reflood regime. Particular emphasis was placed on providing extensive core fluid and heater rod measurements to facilitate this development.

 

SEMISCALE/S-PL-3E-LV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation

The S-PL-3 experiment simulates a loss of offsite power with accompanying failure of the auxiliary feedwater and emergency AC power systems. When primary system pressure reached the safety-relief-valve setpoint, emergency power was assumed to be restored and recovery was initiated by latching open the power-operated relief valve (PORV) and initiating emergency core coolant (ECC) flow to begin the primary feed and bleed recovery.

 

SEMISCALE/S-PL-3E-SV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation

The S-PL-3 experiment simulates a loss of offsite power with accompanying failure of the auxiliary feedwater and emergency AC power systems. When primary system pressure reached the safety-relief-valve setpoint, emergency power was assumed to be restored and recovery was initiated by latching open the power-operated relief valve (PORV) and initiating emergency core coolant (ECC) flow to begin the primary feed and bleed recovery.

 

SEMISCALE/TEST1011, Large break LOCA blowdown starting from isothermal conditions of the primary coolant loop

Scaling Information:
- volume to break area equal to volume to break area of a large PWR
- other parameters not scaled, facility should only yield data for comparison to analyses
  
Parameters offered for Comparison:
- 2 pressures ( upper plenum and upstream break nozzle) ; containment pressure
- 4 differential pressures (hot leg-cold leg, across pump, pressure vessel containment)
- 6 local fluid densities
- 7 flow rates and including break flow
- liquid level
- time to start high pressure injection system (HPIS) and low pressure injection system (LPIS)

 

SEMISCALE/UT-1, Semiscale MOD-2A, 10% Communication Cold Leg Break LOCA Experiment

This test simulates a loss-of-coolant accident resulting from a 10% communicative break in the cold leg of a pressurized water reactor. The break size for this test was 0.228 sq.cm. which is volumetrically scaled to represent a 10% pipe break in a PWR. The Mod-2A system was configured to simulate a PWR with the capability to inject emergency core coolant (ECC) into the vessel upper head. However, for this test, no upper head ECC injection was used. The intact loop accumulator pressure was set at 2.76 MPa as is specified for upper head injection (UHI) plants.

           

SPES/SP-FW-02, Total Loss Feedwater with Emergency Feed Water Delayed in SPES facility

The ISP22 test is a Loss of Main Feed Water in all the three steam generators. The purpose of "LOFW-EFW delayed" experiment is to obtain in the primary system of the facility the thermal-hydraulic conditions similar to the reference plant in the incidental transient of main feedwater total loss, in all the three steam generators, with a delayed intervention of the emergency feedwater system. The most important phenomenologies related to the experiment are the following:
- steam generator heat removal capability in conditions of degraded secondary inventory (due to the main feedwater total loss EFW intervention delayed) and degraded primary inventory (due to mass loss through the pressurizer PORV);
- power channel heat transmissions when the rods are partially uncovered (as a consequence of the primary mass reduction) with or without two phase natural circulation;
- conditions of single phase (steam, liquid) or two phase flow through the pressurizer PORV;
- evaluation of core coolability with EFW delayed intervention;
-EFW intervention effects with void redistribution in the primary circuit and natural circulation establishment (related to residual primary inventory)

 

TBL/311, Two Bundle Loop Facility, Small Break in Recirculation Line

Small break in recirculation line.  The break area and initial power were 3.4% DBA and 4 and 6 MW respectively. Single failure of the HPCS was assumed for the ECCS network and operation of LPCS plus 3LPCI plus ADS was simulated.

 

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system.

 

TLTA/6432, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

64.45 sqcm small-break LOCA; no high pressure cooling system.

 

6.  RESTRICTIONS OR LIMITATIONS

            N/A

 

7.  DATA FORMAT AND COMPUTER

            Data are in binary form.

 

8.  COMPUTER HARDWARE REQUIREMENTS

            N/A

 

10. COMPUTER SOFTWARE REQUIREMENTS

N/A

 

11. REFERENCES

CSNI INTEGRAL TEST FACILITY VALIDATION MATRIX FOR THE ASSESSMENT OF THERMAL-HYDRAULIC CODES FOR LWR LOCA AND TRANSIENTS Prepared by a Writing Group Committee of the PWG 2 "Task Group on Thermal-Hydraulic System Behaviour" - OCDE/GD(97)12, NEA/CSNI/R(96)17 (July 1996).

 

12. CONTENTS OF CODE PACKAGE

The package is transmitted on two DVDs which includes the report cited above along with abstracts, data and reports for the Integral Test Experiments.   

 

13. DATE OF ABSTRACT

May 2012.

 

 

KEYWORDS:        DATA, LOSS-OF-COOLANT ACCIDENT, TRANSIENTS, INTEGRAL EXPERIMENTS DATA, DATABASES, BENCHMARKS