The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.
The PMK-2 projects were initiated in the early 1980’s, parallel to the start-up procedures of the units of the Paks nuclear power plant of VVER-440/213 type. The vendor provided the safety report of the plant, however, without details of analyses, and with short information on the computer codes applied to the analyses. Therefore, to perform independent assessment of the safety of the plant in the country, there was an urgent need for tools of plant system analyses, that is, for thermohydraulic system codes and thermohydraulic system experiments to assess the predictive capabilities of codes. In the first part of the 1980’s, an early version of the thermohydraulic system code RELAP was available through the IAEA. To get results of system tests for VVER-440/213 type NPPs, the design and construction of the PMK-2 facility were initiated, because, at that time, no integral-type facility existed for VVERs that is, PMK-2 (Paks Model Experiment) was the first and the only facility for VVERs. The facility was put into operation in 1985. The PMK-2 was followed by the PACTEL facility for VVER-440 in Finland (1990) and the ISB and PSB facilities for VVER-1000 in Russia (1992 and 1998, resp.).
This report on the PMK-2 projects contains information providing comprehensive descriptions of the first integral-type research programme organized and conducted for VVER-440/213 nuclear power plants, primarily for the Paks nuclear power plant, to study design basis accidents, performing system experiments and getting expertise in code validation primarily through international standard problems, with the participation of experts of 29 countries from the whole world.
Report contains the description of design features of the PMK-2 facility, modelling of the VVER-440/213-specific design solutions, controls, and actions for safety systems. A description and evaluation is given in the 55 experiments performed, which cover an almost complete spectrum of design basis accidents (DBA) corresponding to the most recent version of the safety analysis report (SAR) of the Paks NPP. The OECD-VVER cross reference matrices are developed for the PMK-2 tests for large breaks, small and intermediate leaks, and transients, providing internationally accepted methodology for the identification of major phenomena addressed by the tests. Matrices contain test types occurring in transients and accidents.
Major findings of experiments include results of tests having safety significance, identified by the use of OECD-VVER code validation matrices. In these matrices physical/thermohydraulic phenomena are linked to test types or inversely. The main items of phenomena are as follows: break flow, pressuriser thermohydraulics, heat transfer in SG primary and secondary side, single- and two-phase natural circulation, mixing and condensation during injection from ECCSs, loop seal behaviour in hot leg and clearance in cold leg, core heat transfer including DNB, and dryout. Tests are selected from the cross-reference matrices applied to the PMK-2 tests for large breaks, small and intermediate leaks, and transients. Major findings identified give primarily the basis of the evaluation of the code prediction; that is, if an identified phenomenon is well predicted by the code, the assessment for this phenomenon is successful.
An effective way to increase the confidence in the validity and accuracy of computer codes can be provided by international standard problems (ISP), which were organized and conducted by the OECD/CSNI since 1975, and 20 tests for validation purposes had been performed in about 25 years. The IAEA Standard Problem Exercises were initiated in 1986, and four PMK-2 based exercises were conducted. These activities provided wide frameworks to assess the capabilities of computer codes to represent phenomena occurring in plant transients and accidents and offered a floor to exchange modelling expertise among the best specialists of the field.
Other international frameworks were provided by the EU-PHARE and EU-Framework projects, as well as the US NRC CAMP Program. Altogether 19 PMK-2 tests were applied to study the pressuriser thermohydraulics, large break in the hot leg, primary to secondary leaks, accident analysis methodologies, steam generator heat transfer, and the effectiveness of secondary and primary bleed and feed, as well as accident analysis methodologies.
Methodology and practice applied to the PMK-2 projects include qualitative and quantitative activities. The qualitative assessment is based on the engineering judgement, which means the comparison of the results of a test and computer code calculation, evaluating the results by visual observation. In the quantitative assessment, the comparison is made quantitatively. In the PMK-2 projects, the quantitative assessment is performed by the fast fourier transform based method (FFTBM).
The quantitative assessment is made by the Fast Fourier Transform Based Method (FFTBM) developed at the University of Pisa (Italy). The methodology and an FFTBM code had been implemented at the AEKI. This is a powerful technique for the quantification of the error between measured and calculated parameters of the transient processes, occurring in plants or in scaled-down facilities. The code provides the amplitude and weighted frequency of the deviation of measured and calculated quantities.
The PMK-2 facility is located at the MTA KFKI Atomic Energy Research Institute, Budapest, Hungary. It is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of VVER-440/213 type units of Paks NPP. At the start-up time of Paks NPP, PMK-2 was the first and only integral-type facility for VVERs. Modelling of specific design solutions and modelling aspects of PMK-2 are shortly given below. Details were published in [
A schematic drawing of the primary circuit of the Paks NPP with one main circulating loop is presented in Figure
Schematic drawing of the primary circuit of the VVER-440 plant.
It can be seen in Figure
The steam generators in each loop are large horizontal vessels with a diameter of 3.2 m. The inner diameter of the horizontal heat transfer tubes is 13.2 mm with an average length of 9 m. The total number of tubes is 5536 in 77 rows with an overall height of 1.82 m of the tube bundles resulting in a large water volume in the secondary side of the SG.
The emergency core cooling systems (ECCSs) consist of four safety injection tanks (SITs), three parallel high-pressure injection systems (HPIS), and three parallel low-pressure injection systems (LPIS). The latter is activated at a system pressure of 0.7 MPa. The HPI systems are activated either by low system pressure or low water level in the pressuriser. The original SIT set-point pressure was 5.9 MPa, higher than the secondary pressure of 4.6 MPa. The inlet and outlet temperatures are 540 and 570 K, respectively. The operating pressure is 12.3 MPa, and the core thermal power is 1375 MW.
In most of the transients and accidents—especially in experiments in support of the development of accident management procedures—it is important to start the experiments from the nominal operating parameters of the plant; therefore, the PMK-2 is a full pressure/temperature facility, and the coolant is water. In the transient processes the control and safety signals of the plant are followed. For the design of the PMK-2 the volume scaling criteria were selected which would require the maintenance of the length, elevation, area, volume, and pressure drop relationships. However, it is not possible to maintain the requirements simultaneously; therefore, acceptable compromises are needed.
When the facility was designed in the early 1980s, the range of scaling ratios of integral-type thermohydraulic facilities was, for example, 1 : 48 for ROSA-IV in Japan and 1 : 1705 for SEMISCALE in the USA. Considering the scaling ratio interval and the financial possibilities of the country, a 19-rod core model with 2.5 m heated length was selected which gives the power ratio of 1 : 2070 (39 312/19
The six loops of the plant in the PMK-2 facility are lumped into a single loop, and the break is modelled by orifices of suitable sizes at the hot and cold leg elevations in the upper plenum and downcomer. From the component models the main design and similarity features of the core model, the steam generator model, and the hot leg model are presented.
The cross-section of the 19-rod core model is presented in Figure
Cross section of the core model.
The cross-section of the steam generator (SG) model is shown in Figure
The cross section of the steam generator.
Cold and hot legs are volume scaled, and care was taken to reproduce the correct elevations of the loop seals in both the hot and cold legs. For practical reasons the lengths could not be maintained 1 : 1; therefore, a relatively large tube diameter of 46 mm has been chosen to keep the Froude number which is important to get stratification in two-phase natural circulation. The hot leg with the pressuriser surge line is presented in Figure
Hot leg with pressuriser surge line.
The component models discussed above have about the same vertical position in the loop as the components of the primary circuit in the reference system. The main circulating pump is applied to produce the nominal operating flow conditions and flow coast down following pump trip in the transient process. For the pressuriser the volume scaling, the water to steam volume ratio, and the elevation of the water level are kept; however, the diameter and height ratios cannot be realized. For the hydroaccumulators the volume scaling and elevation are kept with N2 pressure upon the water level as in the plant.
Figure
Scheme of PMK-2 facility.
The nominal parameters of the PMK-2 corresponding to the nominal operating parameters of the reference plant are as follows: primary pressure: 12.3 MPa; coolant temperature at core inlet: 540 K; core power: 664 kW; mass flow rate: 4.5 kg/s; secondary pressure: 4.6 MPa; feedwater temperature: 493 K; steam mass flow rate: 0.36 kg/s; HPIS and LPIS flow: 0.014 and 0.042 kg/s, respectively, for each unit.
Measurement locations are presented in circles as follows: TE: temperature; PR: pressure; LE: (coolant) level; LV: local void. The local void is measured by needle-shaped conductivity probes. The sampling rate of the data acquisition system is 1 s.
The PMK-2 data base, described in detail in [
The experimental results are available in digital form on a CD attached to [
A complete list of tests is given in Tables
7,4% cold leg breaks.
No. of | Project | Test acronym | Dateof | ECCS configuration and/or main objectives |
---|---|---|---|---|
(1) | IAEA-SPE | SPE-1 | 1986 | Minimum ECCS configuration: 0 SIT and 1 HPIS |
(2) | ATKP-2 | SPE-ROV | 1987 | Without ECCS: 0 SIT, 0 HPIS and 0 LPIS, onset of core dryout |
(3) | ATKP-2 | PAV-CM | 1987 | DBA case with ECCS configuration: 3 SITs and 1 HPIS, late SIT actuation |
(4) | IAEA-SPE | SPE-2 | 1987 | DBA case with ECCS configuration: 3 SITs and 1 HPIS, nominal initial power, and delayed steam dump |
(5) | OKKFT G-11 | SP0 | 1988 | DBA case with ECCS configuration: 3 SITs and 1 HPIS, HL/CL connection |
(6) | OMFB 00307/91 | OM1-BF | 1992 | ECCS configuration: 3 SITs, 0 HPIS, 1 LPIS |
(7) | OMFB 00307/91 | OM1-G0 | 1992 | DBA case with ECCS configuration: 3 SITs and 0 HPIS |
(8) | OMFB 00307/91 | OM1-G1 | 1992 | DBA case with ECCS configuration: 3 SITs and 0 HPIS |
(9) | OMFB 00307/91 | OM1-G2 | 1992 | DBA case with ECCS configuration: 3 SITs, 0 HPIS, 1 LPIS |
(10) | IAEA-SPE | SPE-4 | 1993 | DBA case with ECCS configuration: 3 SITs, 0 HPIS, 1 LPIS |
(11) | PHARE SRR3/95 | PHS-BF | 1999 | DBA case with ECCS configuration: 3 SITs, 0 HPIS, 1 LPIS |
(12) | IMPAM-VVER | IMP-21 | 2003 | ECCS configuration: 0 HPIS, 3 SITs, 1 LPIS |
(13) | IMPAM-VVER | IMP-22 | 2004 | ECCS configuration, 0 HPIS, 3 SITs, 1 LPIS |
(14) | IMPAM-VVER | IMP-23 | 2004 | ECCS configuration, 0 HPIS, 3 SITs, 1 LPIS |
(15) | IMPAM-VVER | IMP-32 | 2004 | ECCS configuration: 0 HPIS, 3 SITs, 1 LPIS |
Cold leg breaks of different sizes.
No. of | Project identification | Test acronym | Date of | Break size | ECCS configuration and/or main objectives |
---|---|---|---|---|---|
(1) | OKKFT G-11 | G11-35 | 1989 | 3.5 | Minimum ECCS configuration: 0 SIT, 1 HPIS |
(2) | OKKFT G-11 | CLB-14A | 1990 | 14.8 | Minimum ECCS configuration: 0 SIT, 1 HPIS |
(3) | OKKFT G-11 | CLB-14B | 1990 | 14.8 | DBA case with ECCS configuration: 3 SITs and 1 HPIS |
(4) | OKKFT G-11 | CLB-10A | 1990 | 1.0 | Minimum ECCS configuration: 0 SIT, 1 HPIS |
(5) | AEKI | CLB-10B | 1994 | 1.0 | Minimum ECCS configuration: 0 SIT, 1 HPIS |
(6) | OMFB 0881/95 | OM5-BF | 1995 | 1.0 | Minimum ECCS configuration: 0 SIT, 1 HPIS |
(7) | PHARE SRR3/95 | PHS-05 | 1999 | 0.5 | ECCS configuration: 0 SIT, 3 HPISs |
(8) | OAH-CAMP | OAH-C1 | 1999 | 2.0 | ECCS configuration: 2 SITs, 0 HPIS |
(9) | IMPAM-VVER | IMP-1 | 2003 | 0.5 | ECCS configuration: 3 HPIS, 0 SIT |
(10) | IMPAM-VVER | IMP-31 | 2004 | 30 | ECCS configuration: 0 HPIS, 0 SIT, 1 LPIS |
Hot leg breaks and primary to secondary leaks.
No. of | Project identification | Test acronym | Date of | Break size | ECCS configuration and/or main objectives |
---|---|---|---|---|---|
(1) | OKKFT G-11 | G11-7A | 1989 | 7.4 | ECCS configuration: 0 SIT and 1 HPIS |
(2) | OKKFT G-11 | G11-7B | 1989 | 7.4 | DBA case with ECCS configuration: 3 SITs and 1 HPIS |
(3) | OKKFT G-11 | G11-PS | 1988 | 4.7 | DBA case with ECCS configuration: 3 SITs and 2 HPISs |
(4) | IAEA-SPE | SPE-3 | 1989 | 11.8 | DBA case with ECCS configuration: 3 SITs and 3 HPISs |
(5) | PHARE 4.2.6b | PH4-PS | 1996 | 1.0 | DBA case with minimum ECCS configuration: 2 SITs and 1 HPIS |
(6) | PHARE 4.2.6b | PH4-SLB | 1997 | 32.0 | DBA case with minimum ECCS configuration: 2 SITs, 1 HPIS, 1 LPIS |
(7) | PHARE 2.02 | PH2-PS | 1997 | 4.5 | Maximum ECCS configuration: 4 SITs and 3 HPISs |
(8) | PHARE VVER01 | PHV-11 | 1998 | 4.5 | ECCS configuration: 2 SITs and 2 HPISs |
(9) | PHARE VVER01 | PHV-12 | 1998 | 1.5 | ECCS configuration: 2 SITs and 2 HPISs |
(10) | PHARE VVER01 | PHV-13 | 1998 | 0.7 | ECCS configuration: 1 HPIS and no SITs |
Note: There are three types of breaks:
(i) hot leg break in tests no. 1 and 2,
(ii) leak on the pressuriser in tests no. 5 and 6,
(iii) primary to secondary leaks in tests nos. 3, 4, 7, 8, 9, and 10.
Tests for natural circulation.
No. of | Project identification | Test acronym | Date of | Main objectives |
---|---|---|---|---|
(1) | OKKFT G-11 | G11-TC | 1988 | Study of natural circulation, step by step decrease of primary inventory up to crisis |
(2) | OMFB 00307/91 | OM1-TC | 1993 | One- and two-phase natural circulation, step by step coolant decrease, N2 in the system, crisis |
(3) | PA Rt. | PAV-GFK | 1993 | Study of disturbances: shut-down conditions, N2 in the upper plenum |
(4) | PA Rt. | PAV-FET | 1993 | Study of disturbances: closing of main isolation valve |
(5) | PA Rt. | PAV-GKK | 1993 | Study of disturbances: gas in the SG collectors |
(6) | PA Rt. | PAV-HVM | 1993 | Study of disturbances: cold water injection to the hot leg |
(7) | OMFB, 1044/96 | OM6-GFK | 1998 | Effect of gas injection to the upper plenum |
(8) | OMFB, 1044/96 | OM6-FET | 1998 | Effect of isolation valve closing in the cold leg |
(9) | PHARE SRR3/95 | PHS-TC | 1998 | SG heat transfer study with level decrease, step by step |
(10) | OAH-CAMP | OAH-C2 | 2001 | SG heat transfer study with continuous coolant level decrease in the SG secondary side |
Plant transients and accidents.
No. of tests | Project identification | Test acronym | Date of | Main objectives |
---|---|---|---|---|
(1) | AEKI | LOF-66 | 1986 | Study of system behaviour in total loss of flow, minimum value of DNBR |
(2) | ATKP | ATK-PC | 1987 | MCP rotor seizure, minimum DNBR ratio |
(3) | ATKP | ATK-FW | 1987 | Total loss of feedwater, experimental support to system analysis in Paks NPP |
(4) | OMFB 00307/91 | OM1-FW | 1992 | Total loss of feedwater, effect of secondary bleed with passive secondary feed |
(5) | OMFB 00307/91 | OM1-ST | 1992 | Station blackout, effectiveness of secondary bleed and feed |
(6) | OMFB 00307/91 | OM1-MSH | 1993 | Main steam header break, secondary bleed and feed |
(7) | OMFB 00881/95 | OM5-FW | 1996 | Total loss of feedwater, secondary bleed, and primary bleed and feed |
(8) | OMFB 00881/95 | OM5-ST | 1997 | Station blackout, secondary bleed, and primary bleed and feed |
(9) | PHARE VVER02 | PHV-21 | 1999 | Station blackout with ATWS, density, and boron concentration feedback |
(10) | PHARE VVER02 | PHV-22 | 1999 | Station blackout with ATWS, density, and boron concentration feedback |
The significance of the PMK-2 experiments mainly consists in the creation of a unique, high-quality data base that can be used for code validation, especially for VVER-specific phenomena [
Besides addressing typical phenomena during transients, an important number of PMK-2 experiments support specific cases of plant performance, like operator actions taken in Emergency Operating Procedures and steam generator tube and header ruptures, system behaviour in a LOCA occurring during plant cooldown, and in an ATWS sequence. These data helped to validate codes for VVER-specific design basis accidents or for beyond design conditions.
Tests addressing the Emergency Operating Procedures of the Paks plant supported the following procedure developments: response to inadequate core cooling, response to degraded core cooling, post-LOCA cooldown and depressurisation, and response to loss of secondary heat sink and of loss of all AC power. The results experimentally supported the qualification of procedures and validation of codes.
In the field of primary to secondary leaks the tests are representative for plant response in these VVER-specific accident types and reflect the effectiveness of secondary bleed and feed and of pressure reduction by pressuriser spray and supported the development of effective operator actions to minimize the coolant loss to the atmosphere.
The test investigating the consequences of a LOCA occurring during plant cooldown, when the pressuriser is filled by nitrogen, supplied information on spreading of noncondensable gas along the primary circuit following the cold leg break.
An example on
Quality of measurements strongly depends on the accuracy of measured parameters. In the PMK-2 tests, measurement transducers of temperatures, pressures, differential pressures, levels, local voids, and flow rates are calibrated in accordance with the measurement quality control system. Accuracy estimation includes a calibration constant, absolute maximum error, and standard deviation. In addition, there is a calibration system, and an in-site calibration before each test can be made together with mass and energy balance measurement at operation conditions [
Experiments selected to represent phenomena are as follows: OAH-C1, 2% cold leg break without HPIS, with secondary bleed, PHS-BF, 7.4% cold leg break with secondary side bleed and primary side bleed and feed, PH4-SLB, 32% break in the surge line at the hot leg connection.
Figure
Thermocouple locations in the core model (see Figure
Identification | Number of rods | Level from 0.00 m, m | Levels from core inlet, mm |
---|---|---|---|
TE10 | 10 | 1.044 | 50 |
TE11 | 2 | 1.494 | 500 |
TE12 | 8 | 1.994 | 1000 |
TE13 | 9 | 2.494 | 1500 |
TE14 | 6 | 2.994 | 2000 |
TE15 | 11 | 3.444 | 2450 |
TE16 | 1 | 3.444 | 2450 |
TE17 | 16 | 3.444 | 2450 |
TE18 | 2 | 3.444 | 2450 |
TE19 | 3 | 3.444 | 2450 |
Core model with thermocouple locations for the identification of locations on rods and level (from [
Tables
Cladding temperatures in OAH-C1 test.
Identification | Location rod no./elevation | Temperature excursion | Time | LE11 | Temperature |
TE14 | 6/2.994 | Starts | 1131 | 2.774 | 543.4 |
normal | Max. | 1156 | 2.761 | 628.4 | |
End | 1175 | 2.696 | 549.9 | ||
TE15 | 11/3.444 | Starts | 1118 | 2.968 | 544.4 |
wall | 1. max. | 1180 | 3.232 | 684.0 | |
2. max. | 1263 | 3.210 | 692.7 | ||
3. max. | 1515 | 2.970 | |||
End | 1585 | 2.874 | 546.0 | ||
TE16 | 1/3.444 | Starts | 1119 | 2.947 | 544.7 |
central | 1. max. | 1179 | 3.144 | 697.0 | |
2. max. | 1260 | 3.145 | 702.4 | ||
3. max. | 1515 | 2.970 | |||
End | 1592 | 3.141 | 546.0 | ||
TE17 | 16/3.444 | Starts | 1115 | 3.021 | 544.4 |
corner | 1. max. | 1179 | 3.144 | 680.7 | |
2. max. | 1259 | 3.068 | 681.7 | ||
3. max. | 1514 | 2.886 | |||
End | 1585 | 2.874 | 546.0 | ||
TE18 | 2/3.444 | Starts | 1117 | 2.985 | 544.4 |
normal | 1. max. | 1180 | 3.232 | 701.0 | |
2. max. | 1261 | 3.184 | 708.7 | ||
3. max. | 1515 | 2.970 | |||
End | 1590 | 3.006 | 546.0 | ||
TE19 | 3/3.444 | Starts | 1117 | 2.985 | 544.4 |
normal | 1. max. | 1179 | 3.144 | 711.0 | |
2. max. | 1261 | 3.145 | 712.4 | ||
3. max. | 1514 | 2.886 | |||
end | 1555 | 3.007 | 546.0 |
Cladding temperatures in PHS-05, PHS-BF, and PH4-SLB tests.
Identification | Location rod. no./elevation | PHS-05 | PHS-BF | PH4-SLB | |||
Break size 0.5% | Break size 7.4% | Break size 32% | |||||
—/m | Time | Temp. | Time | Temp. | Time | Temp. | |
TE13 | 9/2.494 | — | — | 479 | 525.4 | 238 | 546.7 |
wall | 717 | 543.4 | 406 | 851.4 | |||
1557 | 470.6 | ||||||
TE14 | 6/2.994 | — | — | 493 | 551.2 | 236 | 553.4 |
normal | 733 | 564.4 | 419 | 927.8 | |||
1578 | 608.2 | ||||||
TE15 | 11/3.444 | 4768 | 603.4 | 525 | 586.7 | 224 | 501.2 |
wall | 749 | 572.7 | 444 | 821.4 | |||
1603 | 685.7 | ||||||
TE16 | 1/3.444 | 4768 | 606.7 | 520 | 591.7 | 223 | 524.0 |
central | 749 | 583.0 | 444 | 943.6 | |||
1603 | 725.2 | ||||||
TE17 | 16/3.444 | 4767 | 598.0 | 512 | 593.4 | 223 | 503.4 |
corner | 748 | 586.4 | 440 | 882.0 | |||
1590 | 667.7 | ||||||
TE18 | 2/3.444 | 4768 | 609.7 | 525 | 598.4 | 224 | 512.0 |
normal | 749 | 584.4 | 444 | 913.0 | |||
1609 | 725.4 | ||||||
TE19 | 3/3.444 | 4768 | 613.0 | 520 | 604.0 | 224 | 517.0 |
normal | 749 | 591.4 | 435 | 916.0 | |||
1600 | 717.0 |
The
Results focusing on dryout occurrence in different regions of the core are presented in Table
Cladding temperatures TE10, TE11, TE12, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in OAH-C1, 2% cold leg break without HPIS, with secondary bleed.
Cladding temperatures TE13, TE14, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in OAH-C1, 2% cold leg break without HPIS, with secondary bleed.
Cladding temperature TE15, TE16, TE17, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in OAH-C1, 2% cold leg break without HPIS, with secondary bleed.
Cladding temperatures TE18, TE19, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in OAH-C1, 2% cold leg break without HPIS, with secondary bleed.
The
Results obtained are shown in Table
Cladding temperatures TE10, TE11, TE12, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PHS-BF, 7.4% cold leg break with secondary bleed and primary side bleed and feed.
Cladding temperatures TE13, TE14, saturation temperature (TS01) and coolant collapsed level in the core model (LE11) in PHS-BF, 7.4% cold leg break with secondary bleed and primary side bleed and feed.
Cladding temperatures TE15, TE16, TE17, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PHS-BF, 7.4% cold leg break with secondary bleed and primary side bleed and feed.
Cladding temperatures TE18, TE19, saturation temperature (TS01) and coolant collapsed level in the core model (LE11) in PHS-BF, 7.4% cold leg break with secondary bleed and primary side bleed and feed.
The
Results obtained are shown in Table
Cladding temperatures TE10, TE11, TE12, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PH4-SLB, 32% break in the surge line at the hot leg connection.
Cladding temperatures TE13, TE14, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PH4-SLB, 32% break in the surge line at the hot leg connection.
Cladding temperatures TE15, TE16, TE17, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PH4-SLB, 32% break in the surge line at the hot leg connection.
Cladding temperatures TE18, TE19, saturation temperature (TS01) and coolant collapsed level in the core model (LE11), in PH4-SLB, 32% break in the surge line at the hot leg connection.
PMK-2 test results have been continuously applied to the validation of different versions of the ATHLET, CATHARE, and RELAP5 codes, since the first IAEA code validation exercise in 1986. In addition, PMK-2 tests figure in the matrices of developmental assessment of internationally recognised computer codes for safety analysis, like ATHLET and RELAP5.
Three small break LOCA and one primary to secondary leak tests served as a basis for VVER-specific Standard Problem Exercises organised under the umbrella of IAEA in the period 1987 to 1995 [
A special small break LOCA test was run to investigate the hot leg loop seal behaviour and the effectiveness of secondary side bleed, with the aim to support code validation activities within the US NRC CAMP programme [
A large number of tests were conducted in EU-PHARE and EU-Framework programmes covering a wide variety of test types including pressuriser surge line break, large primary to secondary leakage, station blackout with ATWS, and tests supporting accident management strategies in VVER-440/213 plants [
Twenty-eight PMK-2 tests were applied to different international code validation exercises, and several versions of ATHLET, CATHARE, and RELAP5 have been assessed by foreign and Hungarian experts. The expertise gained in international cooperations—involving experts of 29 countries participating in PMK-2 projects—helped to better understand VVER system behaviour and reach a high level of modelling of accident sequences. Results of these activities are referenced in [
Since codes applied for safety analysis in Hungary are ATHLET for large break LOCA, RELAP5 for small break LOCA and plant transient analyses, as well as CATHARE, which is used as an independent tool in support of the regulatory authority, validation activity in the country concentrated on these codes [
Methodologies of code validation used in PMK-2 projects include both qualitative and quantitative assessments. The former was almost exclusively applied in the early phase of code validation by integral experiments. It is based on visual observation and engineering judgement of the agreement. Also the quantitative assessment methodology includes a qualitative part, which is a prerequisite to the quantitative part. Quantitative assessment results were obtained by the Fast Fourier Transform Based Method, which had extensively been used also in OECD and IAEA standard problems [
Results of code validation work in the PMK-2 projects are presented by tests selected in a way to represent major test types of the OECD-VVER code validation matrices: different small break LOCAs, pressuriser leak, steam generator header rupture, large break LOCA showing the three typical phases of blowdown, refill and reflood, and station blackout with ATWS. Results are as follows.
The qualitative assessment of RELAP5
Results show that different model approaches applied to the ATHLET, CATHARE, and RELAP5 codes do not influence the quality of predictions in the different types of transients as SBLOCA, LBLOCA, and Transients.
Computer code modelling, development of nodalisations, needs significant users’ expertise and experience gained in simulating plant transients, and vice versa, experiences obtained in the development of computer model for facilities can directly be applied to plant models. Nodalisation may also depend on transient type to be simulated. However, users develop, for example, for a facility a “basic” nodalisation; however, it should be modified in cases when certain phenomena should be described in detail and properly. Methodology for different codes are different. Experts from Germany, France, and USA helped us in developing nodalisation for ATHLET, CATHARE, and RELAP5, respectively. Results of assessment activities in the PMK-2 projects show that similar quality of transient predictions can be achieved by each of the three codes.
The qualitative assessment by engineering judgement refers to the comparison of the results of a test and the results of a computer code calculation, evaluating the results by visual observation. In other words, the qualitative assessment is made by evaluating and ranking the discrepancies between the measured and calculated parameters. In the practice of the qualitative assessment, the comparison by visual observation is limited to the time variations of parameters.
The qualification of qualitative assessment results is as follows. Calculation falls within experimental data uncertainty band, which means that the code calculation is Calculation does not fall within the uncertainty band but shows correct trend. It means that the code calculation is Calculation lies out of the uncertainty band and does not show correct trend, and the reason is unknown. The result of calculation is
The
PMK-2 nodalisation scheme for RELAP5/mod3.2.2 Gamma as applied to the OAH-C1 test.
Results of calculations for the initial conditions and boundary conditions fall into the uncertainty bands of the tests; therefore, the code calculation is qualitatively and quantitatively correct. The same is true for the timing of events, except for the generation of scram. However, it does not affect the overall trend of the time variation of parameters, which are shown in Figures
Measured and calculated primary pressures (PR21) in OAH-C1 test. Calculation by RELAP5/mod3.2.2 Gamma code.
Measured and calculated secondary pressures (PR81) in OAH-C1 test. Calculation by RELAP5/mod3.2.2 Gamma code.
Measured and calculated coolant levels (LE31/LE45) in the hot leg loop seal in OAH-C1 test. Calculation by RELAP5/mod3.2.2 Gamma code.
Measured and calculated cladding temperatures (TE17, TE19) in OAH-C1 test. Calculation by RELAP5/mod3.2.2 Gamma code.
Results of transient calculation are as follows. The overall trend of the time variations of parameters presented in Figures Due to the cold water injection from SITs into the downcomer, direct contact condensation occurs in the downcomer head. The more extensive condensation in the calculation results in higher mass loss through the break: this indicates that the phenomena mixing and condensation are overpredicted. The core heat transfer and especially the timing and value of maximum cladding temperature (1514 s/742 K in the test and 1526 s/726 K in the calculation) are well predicted (Figure The hot leg loop seal behaviour is very well predicted as shown in Figure
The methodology for the qualitative and quantitative assessment of computer code accuracy with applications was presented in several papers [
The procedure for the subdivision of the transient in “phenomenological windows”. For each window: specification of key phenomena, identification of “relevant thermal-hydraulic aspects” (RTA), and selection of parameters characterizing the RTAs are given; qualitative analysis of the results of the measured and calculated time variations of parameters with the subjective judgments are as follows.
In the qualitative phase of assessment the definition of “phenomenological windows” and corresponding RTA’s is highly influenced by the expertise and practice of user.
The quantitative phase of the FFT-based method [
The FFTBM calculates the measurement-prediction discrepancies in the frequency domain. For the calculation of these discrepancies experimental signal (
Normally 14 to 28 variables are selected for accuracy analysis. The most suitable factor for the definition of an acceptability criterion is the total average amplitude
The experiment selected to represent quantitative assessment is the PHS-05 SBLOCA test.
In accordance with the methodology described above, the evaluation of code accuracy needs steps as given below, which are input to the FFTBM code and preliminary, qualitative evaluation of code predictions. Data needed are as follows.
Selected parameters for comparison in PHS-05 test.
Figure no. | Label | Description |
---|---|---|
TE15 | Heater rod surface temperature at core outlet | |
TE63 | Coolant temperature at core inlet | |
TE22 | Coolant temperature in upper plenum | |
TE41 | Coolant temperature at SG inlet | |
TE42 | Coolant temperature at SG outlet | |
PR21 | Pressure in upper plenum | |
PR81 | Pressure in the SG secondary side | |
LE11 | Coolant collapsed level in the reactor model | |
LE71 | Coolant level in the pressuriser | |
LE31 | Coolant level in the hot leg loop seal, reactor side | |
LE45 | Coolant level in the SG hot collector | |
LE46 | Coolant level in the SG cold collector | |
LE52 | Coolant level in the cold leg loop seal, reactor side | |
MA01 | Integrated mass flow through the break |
primary system subcooled: 0 to 350 s, reactor model emptying: 350 to 4720 s, core overheating: 4720 to 5000 s, primary inventory restoration, overfeeding of pressuriser: 5000 to 6997 s.
Table TSE: Time sequence event, SVP: Single-value parameter, IPA: Integrated parameters. exp: experimental values, ATHLET: values calculated by ATHLET Mod2.0A, CATHARE: values calculated by CATHARE2 V1.5, RELAP: values calculated by RELAP5/mod3.3.
RTAs and parameters characterizing RTAs in PHS-05 test for selected time windows: 0 to 350 s, 350 to 4720 s, 4720 to 5000 s, 5000 to 6997 s.
RTAs | Parameters characterizing RTAs | Type | Exp. PHS-05 | ATHLET | CATHARE | RELAP5 | |||
Calc. | Q | Calc. | Q | Calc. | Q | ||||
(1) | |||||||||
Pressuriser emptying | Pressuriser empty | TSE | 350 s | 205 | M | 125 | M | 222 | M |
Primary pressure behaviour | Primary pressure at 8.8 MPa | TSE | 292 s | 284 | M | 184 | M | 321 | M |
Primary pressure at 350 s | SVP | 6.61 MPa | 6.36 | R | 7.42 | R | 7.57 | R | |
Secondary pressure behaviour | Maximum secondary pressure | SVP | 5.34 MPa | 5.34 | R | 5.41 | R | 5.32 | E |
Secondary pressure at 350 s | SVP | 5.32 MPa | 5.33 | R | 5.33 | E | 5.28 | E | |
Pump behaviour | Coast-down begins | TSE | 100 s | E | 100 | E | 102 | E | |
Coast-down ends | TSE | 248 s | R | 236.7 | R | 250 | E | ||
Primary mass inventory | Average break flow | IPA | 0.019 kg/s | M | 0.028 | M | 0.023 | R | |
(2) | |||||||||
Primary pressure behaviour | Primary pressure at 1300 s | SVP | 6.31 MPa | 6.04 | R | 5.42 | R | 6.14 | R |
Primary pressure at 4720 s | SVP | 4.79 MPa | 5.28 | R | 5.34 | R | 4.84 | R | |
Secondary pressure behaviour | Secondary pressure at 4720 s | TSE | 4.79 MPa | 5.07 | R | 5.23 | R | 4.80 | E |
Loop seal behaviour | Cold leg loop seal clearing | TSE | 3182 s | 3203 | E | 2920 | R | 3155 | E |
Hot leg loop seal clearing | TSE | 1272 s | 1192 | R | 1473 | R | 1204 | R | |
Primary mass inventory | Average break flow | IPA | 0.011 kg/s | R | 0.016 | R | 0.011 | E | |
Primary temperature behaviour | Coolant temperature at core outlet at 4720 s | SVP | 534.7 K | 540.5 | R | 541.2 | R | 534.9 | E |
(3) | |||||||||
Dryout occurrence | Onset of dryout | TSE | 4720 s | 4260 | R | 4905 | R | 5135 | R |
Minimum level in reactor model at (s) | SVP | 3.05 m | 3.26 m | R | 1.89 m | R | 2.59 m | R | |
Time of maximum cladding temperature | STS | 4768 s | 4360 | R | 4930 | R | 5160 | E | |
Maximum cladding temperature | SVP | 613 K | 693 | R | 645.9 | R | 593 | R | |
Rewetting | HPIS injection initiated | TSE | 4751 s | R | 4924 | R | 5150 | R | |
End of dryout | TSE | 5000 s | 4380 | R | 5000 | E | 5340 | R | |
(4) | |||||||||
Primary pressure behaviour | Pressure at 5000 s | SVP | 4.61 MPa | 5.37 | R | 4.46 | R | 4.78 | R |
Primary temperature behaviour | Coolant temperature at core outlet at 5000 s | SVP | 532.1 K | 541.3 | E | 530 | E | 534.2 | E |
Coolant temperature at core inlet at 5000 s | SVP | 506.0 K | 511.1 | R | 508.9 | E | 518.7 | R | |
Primary mass inventory | Average break flow | IPA | 0.019 kg/s | R | 0.019 | E | 0.014 | R | |
Total mass through the break | IPA | 94.9 kg | 99.0 | R | 110 | R | 84.8 | R |
Q: E = excellent, R = reasonable, M = minimal, U = unqualified.
By the use of the data presented above, together with the time variations of parameters of the PHS-05 test, calculated by the ATHLET, CATHARE, and RELAP5 codes, the FFTBM code prepares input forms. The nodalisations applied to the system code calculations are as follows: Figure
Nodalisation scheme for PMK-2 facility for the ATHLET 2.0A applied to PHS-05 experiment.
Nodalisation scheme for PMK-2 facility for CATHARE2 V1.5 applied to the PHS-05 experiment.
The time windows, which are shown in Table
Results of quantitative assessment obtained by the FFTBM code are shown in Table
The total values of AA and WF.
Code | ||
---|---|---|
ATHLET Mod2.0-A | 0.25 | 0.03 |
CATHARE2 V1.5 | 0.22 | 0.03 |
RELAP5/mod3.3 | 0.23 | 0.04 |
The qualification is “very good”.
The integral-type facilities have great significance, because they can provide experimental information on transients and accidents anticipated to occur in nuclear power plants. In the 1980s the PMK-2 was the first and the only integral-type facility for VVERs. The PMK-2 was later followed by the PACTEL facility for VVER-440 (Finland 1990) and by the ISB and PSB facilities for VVER-1000 (Russia, 1992 and 1998, resp.).
The PMK-2 facility is a full-pressure thermohydraulic model of the primary and partly the secondary circuit of the Paks nuclear power plant of VVER-440/213 type. The overall scaling ratio is 1 : 2070, the elevation scale is 1 : 1, due to the importance of gravitational forces in natural circulation. The available power is 2 MW that allows establishing nominal conditions of the plant. All the VVER-specific design features are included in the design of the facility. The PMK-2 facility satisfies the criteria of the OECD/CSNI Facility and Test Qualification Matrix.
The first PMK-2 experimental data base covers the list of design basis accidents analysed in the Safety Analysis Report of the Paks NPP, as well as practically all test types described in the OECD-VVER cross-reference matrices. The data base consists of 55 tests which address and help to understand all the important aspects of plant system behaviour in accident conditions. Results are available in digital form at the OECD/NEA data bank.
The significance of the PMK-2 experiments mainly consists in creation of a unique, high-quality data base that can be used for code validation in the three main groups of the OECD-VVER code validation matrices, namely, large breaks, small/intermediate leaks, and plant transients. Phenomena simulated at the level required by the validation matrices are as follows: break flow, pressuriser thermohydraulics and surge line hydraulics, heat transfer in SG primary and secondary sides, single- and two-phase natural circulation, mixing and condensation during injection from ECCSs, loop seal behaviour in hot leg and clearance in cold leg, and core heat transfer in clearing DNB and dryout.
PMK-2 test results had continuously been applied to the validation of different versions of ATHLET, CATHARE, and RELAP5, primarily in international frameworks as: IAEA SPEs, US NRC CAMP programme, EU-PHARE, and EU-Framework programmes. Altogether, 28 PMK-2 tests were applied to different international code validations. During the 20-year period of PMK-2 projects a scientific school has been established with the participation of experts from 29 countries.
Since the codes applied for safety analysis in Hungary are ATHLET for LBLOCA, RELAP5 for SBLOCA, and transients, as well as CATHARE which is used as an independent tool to support regulatory authority, validation activities in the country are concentrated to these codes. Methodologies of code validation in the PMK-2 projects include both qualitative and quantitative assessments. The qualitative assessment is based on visual observation and engineering judgement of the agreement. The quantitative assessment results are obtained by the Fast Fourier Transform Based Method. Last version of the codes as ATHLET MOD2.0A, CATHARE2 V1.5, and RELAP5/mod3.3 have been validated, selecting representative tests.
Expertise gained in the PMK-2 projects was essential for the national reassessment of the safety and for the preparation of the Final Safety Analysis Report of the Paks nuclear power plant on a high level and played an important role in all the modifications performed in the plant, for example, for EOP development, increase of power, handling of primary to secondary leak, and others.
The PMK-2 projects significantly supported the solution of specific problems encountered during the lifetime of the Paks NPP. They corroborated that there is no need to interconnect the hot and cold legs in the operating plant as it had been proposed by Gidropress; they verified the effectiveness of secondary and primary bleed and feed in a group of EOPS like post LOCA cooldown and large break LOCA during cooldown; they continuously supported safety improvement activities in the field of SG tube and SG header ruptures and others.
Hopefully, the PMK-2 data base has significant value for the safety evaluation of all VVER-440/213 type reactors in operation and provides valuable source of scientific information to the nuclear community in the world.
Average amplitude (error function for FFTBM)
Atomic Energy Research Institute
Accident management
Anticipated transient without scram
Bleed and feed
Steam dump valve
Code Assessment and Maintenance Program
Committee for Safety of Nuclear Installation
Data acquisition system
Design basis accident
Density
Departure from nucleate boiling
Departure from nucleate boiling ratio
Differential pressure
Emergency core cooling system
Emergency feed water
European Union
Maximum frequency component of the signal
Sampling frequency
Cut-off frequency
Valve closure in cold leg
Fast Fourier transform
Fast Fourier transform based method
Minimum-maximum frequency
Flow
Gas in upper plenum
Gas in SG collectors
Hydroaccumulator
High pressure injection system
Cold water injection into hot leg
International Atomic Energy Agency
Intermediate break LOCA
Improved accident management
International standard problem
Central Research Institute for Physics
Large break LOCA
Level
Loss of coolant accident
Loss of flow accident
Low-pressure injection system
Local void
Main circulating pump
Main steam header—bleed and feed
Hungarian Academy of Sciences
Nuclear Energy Agency
Nuclear power plant
Number of points in time domain
High-pressure water-cooled loop
National Atomic Energy Authority of Hungary
Organisation for Economic Cooperation and Development
National Middle-Term Safety Research and Development Program
National Committee for Technological Development
Paks Nuclear Power Plant Company
Pump rotor seizure
Paks model experiment
Pressure
Primary to secondary
Relevant thermohydraulic aspect
Safety analysis report
Small break LOCA
Safety injection tank
Standard problem exercise
Station blackout
Transient time duration of signal
Total loss of feed water—bleed and feed
Total loss of feed water
Upper plenum
United States Nuclear Regulatory Commission
Weighted frequency (error function for FFTBM).