Auxiliary safety system

In order to verify that after being implemented, these measures would not lead to new deviations, a restudy and review of those parts should be made. As recommendations for this study do not imply any changes to the process functions, but only addiction of auxiliary safety systems to reduce risk to acceptable levels, there was no need to make a second review of the nodes. 

Safety systems are typically divided into emergency trip/shutdown functions, controlled (slow) shutdown, alarm activation, or startup annunciation of auxiliary equipment such as oil pumps. 

The safety concept considers two nuclear shutdown systems, a set of six reflector rods for reactor scram and power control and a KLAK system of small absorber balls for cold and long-term shutdown. Decay heat removal is made via the heat exchanger, an auxiliary cooling system, and the panel cooling system inside the concrete cavern, or, in case of a failure of these systems, passively by heat transfer via the surface of the reactor vessel. 

As a rule, the main reactor systems are scrutinized more carefully than the auxiliary reactor systems and the order actions are existed for eliminating and mitigating of consequences of main reactor system fails. Therefore the auxiliary reactor system fails may impact on the main reactor systems through places of its contact in significant measure. The influence of auxiliary reactor system fails on main reactor systems and its possible consequences for behavior of the main reactor systems have been analyzed on the basis of the above-mentioned occurrences. Significance of the above-mentioned occurrences for nuclear safety BN-350 NPP have been analyzed too. 

Experimental studies of features of the alloy-structure interaction were carried out by tests of mock-ups, models and real primary components. Fuel subassemblies, control and safety system components, primary pumps, steam generators, sections of main and auxiliary pipelines and their valves have been tested. In the course of tests requirements have been worked out for the temperature conditions of heating up and cooling down modes, as well as for separate structures designed for avoiding damages at multiple condition changes. 

There s a fundamental limitation in traditional gas detection methods. Traditional portable gas detectors provide mobility, but don t communicate. Traditional fixed gas detectors communicate, but aren t portable. The X-zone breaks these limitations 1 combining the mobility of portable systems wHh the communication of forad systems. The result is a level of safety and security unmatched in the industry. This new solution provides flexibility in many applications — from performing confined space entries and area monitoring, to setting up wireless fence lines, to connecting auxiliary safety equipment and transferring alarms to stendtty attendants. With the X-zone, the limits of gas detection are now history. 

Hence, attention focused on passive safety systems and on inherent or intrinsic safety systems. These needed fewer auxiliary systems, they were simpler, with a lower number of parts which could potentially fail, and they did not require as much operator intervention as active systems. 

The active safety systems consist of the safety injection system (SIS), safety depressurization and vent system (SDVS), in-containment refueling water storage system (IRWST), auxiliary feedwater system (AFWS), and containment spray system (CSS). 

A large field for cost savings is also concerned by the design of the nuclear auxiliary systems and not only the safety systems. The solutions used in France and Germany are quite different. The designers knew only their own design and were not able to harmonize themselves. This harmonization process was then made on the Utilities side mainly in comparing each feedback of experience. 

To meet the passive safety requirements of the NGNP, the AHTR uses a reactor vessel auxiliary cooling system (RVACS) similar to that of S-PRISM. It may also use a direct reactor auxiliary cooling system (DRAGS) similar to what was used in the Experimental Breeder Reactor II to supplement the RVACS and reduce the reactor vessel temperature. 

Safety systems and protective equipment are covered in Chapter 16. The present chapter discusses buildings, electrical systems, piping, control systems, auxiliaries such as the equipment used to transport electrolyzers and their components, and the hazards associated with cell rooms. 

Safety injection system Containment spray system Containment hydrogen recombiner system Containment isolation system Auxiliary feedwater system... 

No specific auxiliary feedwater system is included in the design residual heat can be removed by a number of non safety systems ... 

Decay heat removal Primary reactor auxiliary cooling system (PRACS) Steam/water system Active Active N/C operation available Non-safety grade... 

After the Three Mile Island Unit 2 accident the NRC reviewed the auxiliary feedwater system for availability and reliability of components and decay heat removal capability. In particular, the EFW system was scrutinized with regard to the potential for failure under a variety of loss of main feedwater conditions. The safety concern was that a total loss of feedwater, i.e., loss of both main and emergency feedwater, could result in loss of core cooling. The NRC requested operating plants and plants under construction to review both the reliability and the capability of the EFW system to perform its intended safety function i.e., core decay heat removal. The evaluation by the plants was divided into three parts as discussed below. 

The System 80+ Standard Design incorporates pressurizer equipment that is different from current operating plant designs. For example, the Safety Depressurization System (SDS) performs rapid venting and depressurization of the Reactor Coolant System (RCS) when the Auxiliary Spray System is not available (see CESSAR-DC, Section 6.7.1.1 for a description of the SDS). Reliable pressurizer level indication is provided in the Nuplex 80+ Advanced Control Complex consistent with the guidance given in NUREG-0737. 

Much of the nuclear temperature-monitoring> flow-monitoring and other instrumentation which are in the safety clrc ts will ve useful functions for process control. Auxiliary Instrumentation systems which will be provided for N Reactor which a-e not In the safety circuit are described below. 

The two-vessel (catamaran) layout of the floating power unit is used providing separate vessel construction at the shipyard. The reactor plant equipment including auxiliary systems and safety systems will be installed in one vessel the turbo-generator equipment including corresponding systems will be installed in the other vessel. 

A simple scheme of the reactor module and the fact that all modules installed are of the same type make it possible to reduce the number of personnel for operation and maintenance of the modular NPP as compared with the NPP unit comprising one large-power reactor that incorporates many safety systems, such as protection systems, localizing accident systems, and control and auxiliary systems. For example, the safety systems of the AP-1000 reactor [XIX-16] have 184 pumps, 1400 valves, and 40 km of pipelines and cables [XIX-16]. 

Main and auxiliary cooling systems are driven by natural convection. The inherent safety features of the core are enhanced to avoid a core-disruption accident even in anticipated... 

The uncontrolled loss of heat sink also terminates safely if the SGAHRS holds the safety passive function. The reactor vessel auxiliary coolant system (RVACS), which removes heat through the reactor wall to the chimney, serves as a passive system to back up the function of the SGAHRS in accident conditions. 

Main and auxiliary cooling systems of the PBWFR are driven by natural convection. The inherent safety features of the core are enhanced to avoid a core disruption accident even in anticipated transients without scram (ATWSs). Specifically, void reactivity for the case when the core, the axial blanket, and the plenum are totally voided is limited by 3 (design modifications are foreseen to make this effect negative). The bum-up reactivity swing during 15 years of operation without refuelling is minimized down to 1.5% AK/K. 

H2 Procedures used in case of loss of feedwater ARE-ASG (FWFCS (feed water flow control system) -AFWS (Auxiliary feedwater system). Residual heat is extracted using pressurizer relief valves and start up of safety injection and containment spray systems (feed and bleed process). 

In these cases, one can consider that the third barrier is extended from the containment to the operating safety systems and their connected auxiliary systems. 

In order to avoid contamination spreading, the air leaktightness of the rooms containing recirculating systems is controlled and safe (radiation-shielded) storage is provided. The leaktightness between the safety systems and their auxiliary systems is periodically tested. 

Heat removal from PWR plants following reactor trip and a loss of off-site power is accomplished by the operation of several systems, including the secondary system via the steam relief to the atmosphere. The auxiliary (emergency) feedwater system (AFW) functions as a safety system because it is the only source of makeup water to the steam generators for decay heat removal when the main feedwater systems becomes inoperable. 

The capacity of the emergency power supply (DGs) is high enough to supply also such systems as the high pressure boration system (primaiy circuit makeup) and auxiliary feedwater system (for plant startup and shutdown), although these systems are not classified as safety grade. 

Separation between redundant safety systems with their auxiliary supporting features is the basic... 

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