TY - JOUR
T1 - Natural circulation and stability performance of BWRs (NACUSP)
AU - Aguirre, C
AU - Caruge, D
AU - Castrillo, F
AU - Dominicus, G
AU - Geutjes, AJ
AU - van der Hagen, THJJ
AU - Saldo, V
AU - Hennig, D
AU - Huggenberger, M
AU - Ketelaar, KCJ
AU - Manera, A
AU - Munoz-Cobo, JL
AU - Prasser, H-M
AU - Rohde, U
AU - Royer, E
AU - Yadigaroglu, G
N1 - N05-01
PY - 2005
Y1 - 2005
N2 - From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon.
The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities.
In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed.
Four of the three experimental facilities within the project have yielded a large, unique database.
Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility.
The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood.
Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data.
The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments.
Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map.
State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data.
The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress.
The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.
AB - From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon.
The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities.
In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed.
Four of the three experimental facilities within the project have yielded a large, unique database.
Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility.
The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood.
Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data.
The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments.
Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map.
State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data.
The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress.
The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.
KW - ZX CWTS JFIS < 1.00
M3 - Article
SN - 0029-5493
VL - 235
SP - 401
EP - 409
JO - Nuclear Engineering and Design
JF - Nuclear Engineering and Design
IS - 2-4
ER -