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Title: Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

Abstract

The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remotemore » inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)« less

Authors:
;  [1];  [2]
  1. Commissariat a l'Energie Atomique - CEA (France)
  2. AREVA NP (France)
Publication Date:
Research Org.:
The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)
OSTI Identifier:
20995616
Resource Type:
Conference
Resource Relation:
Conference: 14. international conference on nuclear engineering (ICONE 14), Miami, FL (United States), 17-20 Jul 2006; Other Information: Country of input: France
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ACCIDENTS; CLOSURES; CONSTRUCTION; CONTROL; DESIGN; DRAINAGE; FBR TYPE REACTORS; HEAT EXCHANGERS; IN-SERVICE INSPECTION; METERS; POOL TYPE REACTORS; REACTOR SAFETY; REACTOR VESSELS; SODIUM; STEAM GENERATORS; WELDED JOINTS

Citation Formats

Guidez, J., Martin, L., and Dupraz, R. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor. United States: N. p., 2006. Web.
Guidez, J., Martin, L., & Dupraz, R. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor. United States.
Guidez, J., Martin, L., and Dupraz, R. 2006. "Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor". United States. doi:.
@article{osti_20995616,
title = {Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor},
author = {Guidez, J. and Martin, L. and Dupraz, R.},
abstractNote = {The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2006,
month = 7
}

Conference:
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  • Mandatory safety-related design requirements include the absolute leak-tightness of the heat exchange surfaces separating sodium from water and the prevention of violent sodium-water reactions. After briefly recalling the various leak situations likely, and consequences (wastage and secondary ruptures), the precautions taken during steam generator (SG) designing, manufacturing, and operating phases to minimize leak hazards and satisfy the mechanical design basis requirements adopted for secondary loop components are analyzed. The in-service monitoring systems provided (hydrogen detection, acoustic detection) and the means devised to keep leak consequence within acceptable limits (rupture disks) are described. The actions initiated by signals from these systemsmore » are reviewed (reactor shutdown, SG isolation pressure-relief, cooling and inertization of sodium or water-steam circuits). Plant overhaul procedures, servicing operations on the SG itself (whipping tube localization, damage assessment on adjoining tubes by insertion of an eddy-current probe, removal of damaged tubes) and on the secondary loop after detection and prior to restartup are presented. For each device and situation mentioned, a comparison is drawn, underlining the fundamental differences existing in the different types of SGs and tube materials.« less
  • The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to makemore » sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)« less
  • Sensors installed in special fuel elements for the EBR-II reactor had 30- ft-long leads that would pass from the sodium environment through a sealed bulkhead. A hydrogen-atmosphere, induction-heated brazing furnace was constructed to simultaneously braze 20-26 separate sensor leads at one time. The brazed seals were leak-tight, and the sheath wall has less than 10 percent interaction with the braze alloy. (auth)