Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading
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R. Pabarcius
Abstract
Starting from early to mid 1970s the first generation of channel-type graphite-moderated boiling water reactors socalled RBMK types were designed and went into operation. Nowadays, all designed RBMKs represent three generations of reactors having significant differences with respect to their safety design features. The Ignalina NPP with two RBMK-1500 units being a second generation type were originally the most powerful reactors in the world. The change of important neutron-physical characteristics, being specific safety features for RBMK type reactors during RBMK-1500 core upgrading processes are analyzed until final shutdown and their impact on safety during reactor lifetime is discussed.
Kurzfassung
Anfang bis Mitte der 1970er Jahre wurde die erste Generation von graphitmoderierten Siedewasserreaktoren vom Typ RBMK entwickelt und in Betrieb genommen. Heute repräsentieren alle entwickelten RBMKs drei Generationen von Reaktortypen mit signifikanten Unterschieden in Bezug auf ihre Sicherheitsaspekte. Das KKW Ignalina mit zwei RBMK-1500 Einheiten der zweiten Generation gehörte ursprünglich zu den leistungsfähigsten Reaktoren der Welt. Die Änderung wichtiger Neutronen-physikalischer Eigenschaften, als spezifische Sicherheitsmerkmale für Reaktoren vom RBMK Typ werden analysiert während der Erhöhung der Lebensdauer des Reaktorkerns bis zur endgültigen Abschaltung und ihr Einfluss auf die Sicherheit während der Lebensdauer des Reaktors wird diskutiert.
References
1 Almenas, K.; Kaliatka, A.; Uspuras, E.: Ignalina RBMK-1500. A Source Book. Extended and Updated Version, Lithuanian Energy Institute, Kaunas, Lithuania, 1998Suche in Google Scholar
2 Pabarcius, R.; Tonkunas, A.; Bubelis, E.; Clemente, M.: Uncertainty and sensitivity analysis of void reactivity and power reactivity coefficients in an RBKM-1500 reactor core. Kerntechnik70 (2005) 114–119Suche in Google Scholar
3 Pabarcius, R.; Tonkunas, A.; Listopadskis, N.: Determination of RBMK-1500 reactor passport characteristics. Kerntechnik73 (2008) 18–25Suche in Google Scholar
4 Robotko, A. V.: Implementation of measurements to reduce the void reactivity coefficient of the Ignalina NPP, presented at the Conference on “Void reactivity coefficient and maintenance of RBMK-1500 safety”, Ignalina NPP, Visaginas November 23–27, 1992Suche in Google Scholar
5 Technical safety justification of Ignalina NPP with RBMK-1500 reactors, Tech. Rep. D040-1110, RDIPE, Moscow, Russia, 1989Suche in Google Scholar
6 In-depth safety assessment of Ignalina Nuclear Power Plant, Tech. Rep., Ignalina NPP, Visaginas, Lithuania, 1996Suche in Google Scholar
7 Review of the Ignalina Nuclear Power Plant safety analysis report, summary, Tech. Rep. 55, RISKAUDIT, Paris, France, 1997Suche in Google Scholar
8 Uspuras, E.: State of the Art of the Ignalina RBMK-1500 Safety. Science and Technology of Nuclear Installations, 2010, Article ID 102078, 11 pages10.1155/2010/102078Suche in Google Scholar
9 Passport of Ignalina NPP Unit 2 No. 2004/2, PTOed-1234-2B8Suche in Google Scholar
10 Jurkevicius, A.; Tarasov, A.; Krivoshein, G.: RBMK-1500 Reactor Core Upgrading at Ignalina NPP. Uranium-Erbium Fuel and New Design Control Rods Implementation. Proceedings of ICAPP'03, Cordoba, Spain, May 4–7, 2003Suche in Google Scholar
11 Verification of the STEPAN code for 3D neutron-thermohydraulic RBMK calculation. Report RRC KI, No. 33-09/9, 1995Suche in Google Scholar
12 QUABOX//CUBBOX – HYCA manual, GRS, 1993Suche in Google Scholar
13 Sorokin, N. M.; Gabaraev, B. A.; Cherkashov, Yu. M.: Safe operation and life extension of RBMK plants. Nuclear Engineering and Design236 (2006) 1648–165610.1016/j.nucengdes.2006.04.006Suche in Google Scholar
© 2011, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading
- Strategy, main stages and progress of the Ignalina Nuclear Power Plant decommissioning
- Environmental safety aspects of the new solid radioactive waste management and storage facility at the Ignalina Nuclear Power Plant
- Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment
- Burn up extension in a PBMR-400 full core using weapon grade plutonium fuel mixed with thorium
- A study on the damage of potential first wall materials in a nuclear fusion reactor using plutonium bearing salt
- An analytical benchmark of MYRRHA ADS in cylindrical geometry
- Dosimetric characteristics of three new design 125I brachytherapy sources
- Determination of 89Zr production parameters via different reactions using ALICE and TALYS codes
- Novel dose calculation and characterization of 32P intravascular brachytherapy stent source
- Cyclotron production of 85Sr by proton irradiation of natRb
- Lessons learnt from PSA for new and advanced reactors in Russia
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading
- Strategy, main stages and progress of the Ignalina Nuclear Power Plant decommissioning
- Environmental safety aspects of the new solid radioactive waste management and storage facility at the Ignalina Nuclear Power Plant
- Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment
- Burn up extension in a PBMR-400 full core using weapon grade plutonium fuel mixed with thorium
- A study on the damage of potential first wall materials in a nuclear fusion reactor using plutonium bearing salt
- An analytical benchmark of MYRRHA ADS in cylindrical geometry
- Dosimetric characteristics of three new design 125I brachytherapy sources
- Determination of 89Zr production parameters via different reactions using ALICE and TALYS codes
- Novel dose calculation and characterization of 32P intravascular brachytherapy stent source
- Cyclotron production of 85Sr by proton irradiation of natRb
- Lessons learnt from PSA for new and advanced reactors in Russia
