Safety analysis reactor protection system

Chapter 15 lists the assumptions used in the turbine trip e analysis. These assumptions are chosen so that they tend to maximize the required pressure relieving capacity of the primary and secondary valves. The analysis demonstrates that sufficient relieving capacity has been provided so that when acting in conjunction with the reactor protective system the safety valves will prevent the pressure from exceeding 110% of the design pressure. 

Operability requirements should be stated for the reactor protection system and for instrumentation and logic for other safety systems, together with limits on response times, instrument drift and accuracy, where appropriate. Interlocks teqnired by the safety analysis report should be identified and relevant operability reqnire-ments should be stated. 

The results of the analysis show that the overtemperature AT reactor protection system signal provides adequate protection against the reactor coolant system depressurisation events. The calculated DNBR remains above the design limit. The long-term plant response due to a stuck-open ADS valve or pressuriser safety valve, which cannot be isolated, is bounded by the small-break LOCA analysis. 

The relief capacities of the pressuriser safety valve is determined from the postulated overpressure transient conditions in conjunction with the action of the reactor protection system. An overpressure protection report is prepared according to Article NB-7300 of Section III of the ASME code. Reference 6.2 describes the analytical model used in the analysis ofthe overpressure protection system and the basis for its validity. 

A.809. All elements of the reactor power regulating system shall be described (design criteria and reliability analysis). All interfaces between the power regulating system and the reactor protection system should be identified and analysed to confirm that they do not lead to degradation of safety. 

The configuration related to the CEFR reactor protection system is also introduced in this simulation. The various rupture positions and all the transient parameters of the primary loop, as well as the accident sequence are obtained with the help of the system safety analysis code OASIS. This code is based on the original French code OASIS, modified to simulate the CEFR. 

Safety variables, safety settings, trip signals, and self-check results of the reactor protection system are provided to the Supervision and Control System (SCS), through a one way, electrically isolated interface. This enables the efficient use of reactor instruments, avoiding duplication, while improving presentation by the use of SCS visual display units, and enabling enhanced recording/analysis of safety events. 

The safety analysis should establish the design capabilities and protection system set points to ensure that the fundamental safety functions are always maintained. The design basis events are the basis for the design of the reactivity control systems, the reactor coolant system, the engineered safety features (for example, the emergency core cooling system, the containment system and containment protection... 

The safety/relief valves are dual-function valves discharging directly to the pressure suppression pool. The safety function provides protection against overpressure of the reactor primary system. The relief function provides power-actuated valve opening to depressurize the reactor primary system. The valves are sized to accommodate the most severe of the following two pressurization transient cases determined by analysis ... 

For research reactors, consideration is given to automatic actions to attain a safe state in the case of an external event when these actions are compatible with the speed of development of the external events. The facility should have protection capabihties in all operating modes and conditions. The systems in charge of this are considered safety related and consequently categorized for external events. In particular, operational limits and conditions of a seismic scram system including surveillance tests and intervals are based on the safety analysis for seismic events. Reference [5] provides information on automatic seismic trip systems for NPPs and other facihties. 

A. 1609. The settings of all protection system functions that are used in the safety analysis shall be listed. Typical protection system functions are reactor trip, isolation valve closures and backup cooling. 

A. 1701. This chapter of the SAR shall contain the operational limits and conditions (OLCs) important to safe reactor operation which have been derived from the safety analysis. The OLCs represent an envelope of parameters, developed by the operating organization, which will protect the reactor, the staff, the general public and the environment from undue exposure if they are not exceeded. Therefore, it is essential that the OLCs are understood by the responsible operating personnel. The OLCs include safety limits, safety system settings, limiting conditions for safe operation, and surveillance and administrative requirements. Additional information is contained in paras 601-608 of Safety Series No. 35-S2. 

Within process industries characterized by large production units and high levels of automation, risk and accident analysis is focused on the avoidance of low-probability events entailing serious consequences for the plant and its environment. Safety analysis is based here on causal or probabilistic models of the accidental chain of events that can serve to identify deficiencies in the design of the plant and its protective system as well as to predict the level of risk involved in an operation. Methods developed are fault tree analysis, MORT (Johnson 1975) and INRS (Leplat Rasmussen 1984). A detailed analysis of the actual, individual incident or failure is performed to identify these possible weak spots in the plant and its operation. It is a common experience that human acts play an important role in such industrial mishaps so, especially after the reactor incident at Three Miles Island in 1979, much effort has been spent on developing suitable predictive tools for the... 

Qualitative reliability analysis are used to identify possible ways in which a system can fail. The calculation can result in all combinations of components and human failures that lead to safety (protection) system failure, which prevents the safety system to shut-down the reactor upon request. For this analysis a top-down logic model, known as a Master Logic Diagram (MLD), similar to a fault-tree, could advantageously be used with a top event of a safety system failure upon request. The results of this analysis can be used to prove the fulfillment of the important single failure design criteria. 

ANS 56.3 ANS 54.6 ANS 56.8 ANS 56.9 Overpressure protection of low-pressure system connected to the reactor coolant pressure boundary Pressure/temperature transient analysis for LWR containments Reactor containment leakage testing requirements Environmental envelopes to be considered in safety-related equipment... 

SSCs important to safety shall be designed and located, subject to compliance with other safety requirements, so as to minimize the effects of fires and explosions. A fire hazard analysis and an explosion hazard analysis shall be carried out for the research reactor facility to determine the necessary ratings of the fire barriers and means of passive protection and physical separation against fires and explosions. The design shall include provisions to prevent or limit the formation of explosive atmospheres. Fire detection systems and fire fighting systems of the necessary capability shall be provided. 

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