Rod drive mechanisms

Nuclear Boiler Assembly. This assembly consists of the equipment and instrumentation necessary to produce, contain, and control the steam required by the turbine-generator. The principal components of the nuclear boiler are (1) reactor vessel and internals—reactor pressure vessel, jet pumps for reactor water circulation, steam separators and dryers, and core support structure (2) reactor water recirculation system—pumps, valves, and piping used in providing and controlling core flow (3) main steam lines—main steam safety and relief valves, piping, and pipe supports from reactor pressure vessel up to and including the isolation valves outside of the primary containment barrier (4) control rod drive system—control rods, control rod drive mechanisms and hydraulic system for insertion and withdrawal of the control rods and (5) nuclear fuel and in-core instrumentation,... 

A cut-away schematic of a BWR equipped with external coolant pumps is shown in Fig. 17. The recirculation system comprises the external piping, pumps, and valves located at the lower region of the vessel. The penetrations through the bottom of the vessel that contain the control rod drive mechanisms that are vital for controlling the reactivity of the core are not shown in Fig. 17. Of particular importance, as far as the integrity of the reactor is concerned, is the control rod drive tubes and related mechanisms, because they are in contact with the coolant and... 

Fig. 17 Cut-away schematic of a BWR vessel and associated piping system for a BWR equipped with external coolant pumps. The underlined labels are indicating regions of IGSCC. The control rod drive mechanisms and penetrations that are located at the bottom of the vessel are not shown.

Nuclear Reactor with a Hole in the Head On March 6,2002, personnel repairing one of the five cracked control rod drive mechanism (CRDM) nozzles at Davis-Besse Nuclear Plant, Oak Harbor, Ohio, discovered extensive damage to the reactor vessel head. The reactor vessel head is a dome-shaped structure made from carbon steel housing the reactor core. The reactor vessel head is placed such that it can be removed when the reactor is shut down to allow spent nuclear fuel to be replaced with fresh fuel. The CRDM nozzles connect motors mounted on a platform above the reactor vessel head to control rods inside the reactor vessel. Reactor operators withdraw control rods from the reactor core to start the operation of the plant and insert the control rods to shut down the operation of the reactor. 

CONTROL SAFETY ROD DRIVE MECHANISM 13. PRIMARY PUMP DRIVE... 

FBTR operating experience has improved the confidence level in the design and operation of core, sodium systems, control rod drive mechanisms, fuel handling machines, steam water system and SG leak detection system. 

For the management of the Control Rod Drive Mechanism System for the reactors WER-1000 operated in Czech Republic and in Ukraine, a computer-based system has been developed in Skoda and is in operation since 1996 on the South - Ukraine NPP and then in Chmelnicka NPP and Temelin NPP. 

Inner tank, control rod drive mechanism support plate,... etc.) ... 

Fertile Particle Predicted Failure Core-Average Release/Blrth for Kr-8Sm Gore-Average Release/Birth for Xe-138 Outer Neutron Control Assembly Inner Neutron Control Assembly Control Assemblies Installed in Vessel Ex-Vessel Detectors and Location Plan View of Reactor Core Vertical Section of Reactor Core Control Rod Drive Mechanism... 

The performance of control rod drive mechanism (CRDM) has been satisfactory with friction force within limits and drop time less than 400 ms. An on-line system to monitor the drop time of control rod (CR) during scram was commissioned. Similarly a system was developed to measure friction force of CR during power operation. The 3 s interlock on CR raise movement, which was introduced before the first criticality was deleted as it was giving rise to large time in raising power and high start up duty demand on CRDM motors. The lower parts of two CRDM were replaced, one due to failure of translation bellows, and another due to failure of gripper bellows. Leaky silicone bellows of one CRDM was replaced in-situ. 

For the in-vessel components, control and safety rod drive mechanism (CSRDM), Diverse safety rod drive mechanism (DSRDM), failed fuel location module (FFLM) and primary sodium pump (PSP), saturation activity was calculated, as these components are expected to be irradiated for 20 years. A cooling time of 2 or 5 years was considered in the case of grid plate components, as these are expected to be handled only for decommissioning purposes. In the case of CSRDM, DSRDM, FFLM and PSP, a cooling time of 2 days has been considered. For the Control and Safety Rod (CSR) and Diverse Safety Rod (DSR), an irradiation time of two years and cooling time of 2 days was considered [4]. 

Control rod drive mechanisms mounted externally to the reactor pressure vessel. 

The CAREM NPP is a light water integrated reactor. The whole high-energy primary system and the absorbers rods drive mechanisms are contained inside a single pressure vessel. 

The CAREM reactor pressure vessel (RPV) contains the core, steam generators, the whole primary coolant and the absorber rods drive mechanisms (figure 1). The RPV diameter is about 3.2 m and the overall length is about 11m. 

The CAREM project involves technological and engineering solutions, as well as several innovative design features that have been properly proved during the design phase. Within CAREM project, the effort was focused mainly on the nuclear island (inside containment and safety systems) where several innovative design solutions require developments. This comprises mainly the reactor core cooling system, the reactor core and fuel assembly, the reactor pressure vessel internals and the hydraulic control rod drive mechanisms. 

The goals of this development were (1) Enhanced safety and reliability, (2) Reduced occupational radiation exposure and radioactive waste, (3) Enhanced operability and maneuverability and (4) Improved economy. Significant improvements were induced by the adoption of the reactor internal pump, fine motion control rod drive mechanism, integral type reinforced concrete containment vessel and digital control system. Cross-section of reactor buildings for 1100MW(e)-class BWR and ABWR are described in Figure 2. 

A first round of the systematic review of the reactor operating eiqierience has been completed for the absorber rod drive mechanisms, main sodium pumps and intermediate heat exchangers (IHXs) which has revealed sensitive points for EFR. Examples are the importance of succeeding with the EFR features which prevent die sodium aerosol problems on the absorber mechanisms, the need to confirm adequate operational flexibility whilst avoiding the check valve at the primary pump outlet and to ensure full benefit is taken of the piston seal experience for the IHX / pool boundary replacing the gas bell originally selected for EFR. 

The performance of reactor systems, sodium systems, control rod drive mechanisms and other safety related systems and auxiliary system were generally satisfactory. The primary and secondary sodium purity was maintained below the plugging temperature of 105 C. The four sodium pumps and Aeir drives are operating well and have logged 83,293 h, 70,615 h, 88,878 h 69,201 h. 

The design of the reactor internals has not been addressed yet, but they likely will be made of graphite or carbon composites to accommodate the high-core outlet temperature required by the NGNP (1000°C). It is possible that carbon-insulated metallic alloy will be used for the core support structure, although this has not been evaluated yet. Control rods will be required to provide for reactor startup, normal operation, and shutdown. The munber and placement of control rods has not been evaluated yet, but the rods will be constructed from carbon composites for the drive shafts and absorber casing and boron carbide or other high-temperature absorber for the neutron absorber. The control rod drive mechanisms will be located above the reactor enclosure head. 

Ageing effects can be detected by checking the performance of a system, structure or component (e.g. drifting of setpoints or deterioration of electronic or mechanical components of valves and valve actuators or control rod drive mechanisms, may cause changes in the performance of a control system). For this reason, the results of the performance test programme, which is dependent on the specific design and operation of the facility, should be examined for evidence of trends which may indicate ageing problems. 

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