Reactor coolant systems components

Pressure retaining parts of components in the reactor coolant system Components of or connected to the primary reactor coolant system that are essential for ensuring the shutdown of the reactor and cooling of the nuclear fuel in relevant operational states and in postulated accident conditions ... 

The totaI heat d i ss i pat i on capab iIi ty of the UHS should include conservative estimates of reactor decay heat, heat stored in reactor coolant system components and structures, and other heat sources removed by the cooling systems. This heat removaI capab i I i ty shouId be avaliable during normal operation and following AOEs. 

The criteria applied in the design of the Reactor Coolant System supports are that the specific function of the supported equipment be achieved during all normal, earthquake, safety valve actuation and Branch Line Pipe Break (BLPB) conditions. (BLPB includes feedwater line breaks and all loss-of-coolant-accident conditions resulting from breaks not eliminated by leak-before-break analysis in piping to branch nozzles of the reactor coolant system.) Specifically, the supports are designed to support and restrain the Reactor Coolant System components under the combined Safe Shutdown Earthquake and Branch Line Pipe Break loadings in accordance with the stress and deflection limits of Section III, ASME Code. 

During pre-operational testing of the Reactor Coolant System, the support displacements will be monitored for concurrence with calculated displacements and/or clearances. Subsequent inspections of supports which are integral with Reactor Coolant System components will be in accordance with Section XI of the ASME Code. 

This group of characteristics specifies the structure around a reactor, designed to protect the reactor from outside intrusion and protect the outside from radiation effects, in case of a malfunction inside the structure. Except for the reactor, the containment usually contains all the reactor coolant system components. They are either designed to withstand the maximum pressure expected after a design basis accident (DBA), in which case they are termed full-pressure containments, or they may have installed systems to reduce the DBA containment... 

APIOOO provides instrumentation and controls to sense accident situations and initiate ESFs for the mitigation. The occurrence of a limiting fault, such as a loss of coolant accident or a secondary system break, requires a reactor trip plus actuation of one or more of die ESFs. This combination of events prevents or mitigates damage to the core and reactor coolant system components, and provides containment integrity. 

The SCOR design is based on a compact reactor vessel that contains all the reactor coolant system components, including the pressurizer, reactor coolant pumps, control rod drive mechanism and dedicated heat exchangers of the passive decay heat removal systems. The single steam generator is located above the reactor vessel. 

The heatup and cooldown of the reactor vessel and the addition of makeup water to the reactor coolant system can cause significant temperature changes and thereby induce sizable thermal stresses. Slow controlled heating and cooling of the reactor system and controlled makeup water addition rates are necessary to minimize cyclic thermal stress, thus decreasing the potential for fatigue failure of reactor system components. 

Phenix is France s experimental fast breeder reactor that is located in Marcoule and is operated by the French Atomic Energy Commission (CEA). The CEA, has decided to renovate the aging plant to extend the life expectancy of the reactor. The program reqnires field inventory and inspection of equipment to support the component life-span evaluation studies. The presence of liquid sodium and the high temperature of the reactor coolant system even at cold shutdown - between 150 and 180°C - make inspection of Fast Breeder Reactors (FBR) a difficult technical challenge. Framatome-ANP has in particular conqileted fom innovative nondestructive tests at Phenix ... 

Reactor Coolant Boundary. The reactor coolant boundary means all those coolant-containing components of nuclear reactors, such as pressure vessels, piping, pumps, valves, and heat exchangers, which are part of the reactor coolant system, or connected to the reactor coolant system, up to... 

Anticipated operational occurrences are off-normal events, usually plant transients, which can be coped with by the plant protection systems and normal plant systems but which could have the potential to damage the reactor if some additional malfunction should happen. Their typical frequency of occurrence may be more than 10 year Some of the anticipated occurrences (PIEs - postulated initiating events) are due to the increase of reactor heat removal (as might occur for an inadvertent opening of a steam relief valve, malfunctions in control systems, etc.). Some are due to the decrease of reactor heat removal (such as for feed-water pumps tripping, loss of condenser vacuum and control systems malfunctions). Some are due to a decrease in reactor coolant system flow rate, as in the case of a trip of one or more coolant pumps. Some are connected with reactivity and power distribution anomalies, such as for an inadvertent control rod withdrawal or unwanted boron dilution due to a malfunction of the volume control system for a PWR. Events entailing the increase or decrease of the reactor coolant inventory may also happen, due to malfunctions of the volume control system or small leaks. Finally, releases of radioactive substances from components may occur. 

REACTOR COOLANT SYSTEM MATERIALS WELD MATERIALS FOR REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS Materla1... 

All components in the Reactor Coolant System (RCS) are designed to withstand the effects of cyclic loads due to RCS temperature and pressure changes. These cyclic loads are introduced by... 

The reactor coolant pump pressure boundary is nondestructively inspected as required by ASME Section III for Class 1 components. The pump casing inspections include complete radiography and liquid penetrant or ultrasonic testing. The pump receives a hydrostatic pressure test in the vendor s shop and with the Reactor Coolant System. Inservice inspection of the pump pressure boundary will be performed during plant life in accordance with ASME Section XI. 

Figure 5.4.14-1 illustrates the Reactor Coolant System support points. A description of the supports for each supported component follows ... 

The structural integrity of the reactor coolant system support components is ensured by quality assurance inspections in accordance with Section III of the ASME Code during fabrication. The non-integral supports are procured by individual equipment specifications which impose appropriate quality assurance requirements commensurate with the respective component s functions. 

Only the systems that are part of the high pressure reactor coolant system are located within the containment. Systems carrying hot pressurized reactor water are not allowed to extend beyond the containment. The reactor water cleanup and the liquid and solid waste handling systems are located in a separate building with concrete walls for separation and shielding of major components. 

Stress analyses were performed to determine the effects of a natural circulation cooldown event (similar to that of the St. Lucie occurrence) on both the St Lucie "class reactor vessel and the System 80 "class reactor vessel. The analyses concluded that should natural circulation cooldown of the reactor coolant system be required and should vessel head voiding subsequently occur, the resulting thermal stresses would not cause any thermal, hydraulic, or fatigue damage to the reactor vessel and its integral components over their design lifetime. 

Unresolved Safety Issue (USI) A-12 in NUREG-0933 (Reference 1), addresses minimizing the susceptibility for lamellar tearing and low fracture toughness of major reactor coolant system (RCS) component supports. 

Major component supports for the reactor coolant system are designed and fabricated in accordance with the ASME Code, Section III, Subsection NF, as described in CESSAR-DC, Section 5.4.14. Thus, "code materials are used in the fabrication of the supports consequently the fracture toughness of these materials is in accordance with code requirements. This issue is, therefore, resolved for the System 80+ Standard Design. 

Concerns have been expressed about damage to primary systems and components as the result of excessive vibration. A major source of vibration for the NSSS is flow-induced vibration (i.e., water flowing through the Reactor Coolant System (RCS)). Flow-induced vibration can lead to damage to the reactor vessel internals and, potentially, interference with control rod movement. 

BRS—boron recycle s) em BTRS—boron thermal regeneration s) em CCW—component coolir water HX—heat exchanger RCS—reactor coolant system RHRS—residual heat removal system RMW—reactor makeup water RWST—refuelii water storj e tank WPS (L)—waste processii system (liquid)... 

Shop fabrication of major components Including the reactor coolant system. 

In these closed test-loops components inserted in the reactor core, the coolant, instrumentation and heat transfer systems were completely separated fi om the main FFTF core, permitting the testing of fuels and materials over a wide range of temperatures in a controlled environment independent of the main reactor coolant system. The open loop test positions and integral components of the reactor core for testing large quantities of candidate fuel pins and assemblies were cooled by the reactor primary coolant system. 

This main group of characteristics provides information on the design and parameters of the reactor, the main components of the reactor coolant systems and important nuclear safety... 

This group of characteristics concerns the components of the reactor coolant system that are designed to circulate coolant through the reactor to remove heat from the core. Based on the type of coolant, they include both liquid pumps and gas circulators (blowers). 

The loss of integrity of the supports of heavy components could further aggravate the plant condition during a design basis accident. A failure of RPV supports would also challenge the integrity of the reactor coolant system and the functionality of safety systems. 

The primary coolant contains dissolved boric acid, which has rather strong corrosive effect on low alloy carbon steel components of the reactor pressure boundary. Reactor coolant that leaks out of the reactor coolant system loses water by evaporation, which results in the formation of highly concentrated boric acid solutions or deposits of boric acid crystals. 

Freeze seals are used to isolate components (such as inboard isolation valves) for maintenance in locations that cannot otherwise be isolated. The seal is created and maintained by applying a cooling agent such as liquid nitrogen to the exterior of the pipe. The cooling agent freezes the water within the pipe section, thus sealing the pipe. When used in the reactor coolant system (RCS) pressure boundary, these freeze seals become a temporary part of the pressure boundary. 

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