Safety analysis, fast reactors

Fullwood, R. and R. C. Erdman, 1974, On the Use of Leak Path Analysis in Fault Tree Construction for Fast Reactor Safety, CONF-74040I-P3. 

C and the reaction considered fast at this temperature. Nevertheless, for the safety analysis, an accumulation of 10% is considered realistic. The industrial charge is 4000 kg of final reaction mass in a 4 m3 glass-lined reactor. This vessel is protected against overpressure by a safety valve with a set pressure of 0.3 bar g. The total empty volume of the vessel is 5.5 m3. The vent line has an internal diameter of 50mm and the maximum allowed working pressure of the reactor is 0.3 bar (g). 

The lessons leamt firom the analysis of these incidents and safety reviews have been significant not only in improving FBTR performance but also as an important input for the design of future fast reactors. 

FBTR has been fully commissioned with small core up to a power level of 13.4 MWt and the performance of all the safety related systems has been satisfactory. Large number of modifications was carried out based on experience feed back and analysis of various incidents to improve system performance. Construction, commissioning and operation of FBTR have given considerable amount of experience and confidence, which will help in its smooth and sustained operation at nominal power and will also give useful feedback for the design and commissioning of large fast reactors. 

The predominant method of evaluation in these studies has been fault tree analysis. The Reactor Safety Study also utilized event tree analysis to conveniently document accident sequences and to link the subsystem fault trees into a plant analysis. Failure Mode and Effects Analysis, used extensively in fast reactor safety, is the recommended method for preliminary analysis. All of these methods have application to the analysis of the fuel cycle Including the problems of safe arding special nuclear material. ... 

Jubault, M., et. al.. Fast Neutron Reactor Safety Reliability Analysis of PHENIX Decay Heat Removal Function, Proc. of ENS Conference on Fast Reactor Safety, Seattle, USA, 1979. 

Ricotti, M.E., Kloosterman, J.-L., Lathouwers, D., van derLinden, E., Merk, B., Rohde, U., 2014. Safety Analysis Transient Calculations. EVOL (Evaluation and Viability of Liquid fuel fast reactor system) European FP7 project. Contract number 249696. 

PHWR safety analysis requires a comprehensive set of physical models. Reactor physics analysis may require a transient three dimensional model for the large PHWR cores. The most demanding application is a large LOCA, because of the relatively fast kinetics and the spatial effects associated with flux tilts and shut-off rod (or liquid absorber) insertion. Three dimensional effects are also important in slow loss of reactivity control starting from distorted flux shapes. 

Develop guidelines for safety analysis of the primary coolant pipe-rupture event in liquid metal fast breeder reactor. 

The main pipe rupture accident is the most serious one for the loss of coolant accident of pool type sodium cooled Fast Breeder Reactor (FBR). To simulate this accident, a model is developed based on the OASIS code, which is a French fast reactor system safety analysis code. To abide by the strict accident analysis principles, the main pipe rupture accident is calculated for various position of the pipe. Accident sequence and key parameters, including the fuel cladding temperature of reactor, are obtained for each case. The calculation results show that the fuel cladding temperature is below the safety limitation and the coolant temperature is lower then the saturation temperature of sodium in all cases. 

All these various rules and codes, either for dimensioning or safety analysis, need to be approved by the Safety Authority for each reactor plan. Regarding Fast Reactors, specific rules and codes were set for design, construction and safety analysis, and improved since the demonstrator Phenix time up to the most recent European project (EFR). 

The present study analyzes a postulated break in the primary pump discharge pipe to assure the inherent safety of KALIMER. KALIMER is a pool-type liquid metal sodium cooled fast reactor plant. The main concern of the accident is the amount of subcooling margin reduction, i.e., the degree of increase in the fuel and the coolant temperatures. The stabilization of power associated with reactivity feedback is also an important aspect of the accident. The analysis is performed with the SSC-K code, which was developed on the basis of the SSC-L code for the... 

Analysis of the heat transfer deterioration mechanism by numerical simulation using the k-s turbulence model is in ref. [112]. Transient and accident analysis code for fast reactors, SPRAT-F, and calculation of the Oka-Koshizuka heat transfer correlation for the safety analysis at supercritical pressure are described in ref. [113]. 

Chapter 7 covers the design and analysis of fast reactors. The features of the Super LWR and Super FR are that the plant system configuration does not need to be changed from the thermal reactor to the fast reactor. The analysis of plant control, stability, and safety of the Super FR as well as core design are provided. 

Criticality safety evaluations for handling mixed Pu-U oxide-type fuel elements depend heavily on computational analysis with experimentally validated computer codes and cross-section data. A series of critical experiments has been performed with fast test reactor fuel pins in water at the Battelle-Pacific Northwest Critical Mass Laboratory in support of the Advanced Fuel Recycle... 

---