Control Rod Withdrawal

The reactor protective interface provides signals for protective actions. Examples of protective actions include control rod withdrawal interlocks and startup rate reactor trips. 

The safety demonstration tests in the HTTR are conducted to demonstrate an inherent safety feature, that is an excellent feature in Shutdown of the HTGRs, as well as to obtain the core and plant transient data for validation of safety analsis codes and for establishment of safety design and evaluation technologies of the HTGRs. The safety demonstration tests consist of Reactivity insertion test - control rod withdrawal test and Coolant flow reduction test as shown in Figure 6. In the control rod withdrawal test, a central pair of control rods is withdrawn and a reactivity insertion event is simulated. In the gas-circulators trip test, primary coolant flow rate is reduced to 67% and 33% of rated flow rate by running down one and two out of three gas-circulators at the Primary Pressurized Water Cooler without a reactor scram, respectively. 

Pressurized Conduction With Control Rod Withdrawal (SRDC-3, -4)... 

Control rod withdrawal occurs when rotation of a selected drive motor is commanded by the rod control instrumentation. Control rod insertion can occur either by actuation of the drive motor or as a gravity powered movement following receipt of a reactor trip signal. 

AOO-3 is a control rod withdrawal followed by reactor trip and cooling on the HTS. It results in a slower cooldown rate in the core support graphite structure than AOO-1 (C). No structural or heat removal consequences have been identified for this event. 

DBE-2. which is an HTS transient without control rod trip, and DBE-3. which is a control rod withdrawal without HTS cooling, do not result in rapid temperature transients, heatup or structural loading of graphite reactor internals components. 

DBE-4. which is a control rod withdrawal with Reactor Cavity Cooling System (RCCS) cooling, results in internals graphite temperatures essentially the same as AOO-2. Structural support and heat path functions are not affected. 

It has now been established that, under the circumstances prevailing at the time of the accident with the control rods withdrawn much too far from the core, inserting the control rods would initially have added reactivity to the reactor instead of reducing it. This would have occurred whether the control rods were manually or automatically inserted and could have triggered the power rise or made an existing power rise worse. The designers have now imposed control rod withdrawal limits which would prevent this situation from arising. 

Reactor cooling pumps were in full operation, four of which were necessary to provide an electrical load during the test. This meant that the pmnps were operating close to cavitation with an unstable characteristic, and the reactor channels were overcooled and contained mainly saturated water with a small quantity of steam near the outlet, requiring further control rod withdrawal. 

All 8 circulating pumps brought into operation. Four of these provided the electrical load for the turbo generator. Increased water content in core reduced reactivity, necessitating more control rod withdrawal. " Pumps close to cavitation as cool feedwater quantity low. " Instability in steam pressure and drum levels—water level and steam pressure reactor trips disconnected. " Difficulty in maintaining core output... 

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. 

Prolonged control rod withdrawal at full and low power BDBA... 

Computer codes must be qualified. The adequacy of the nodalization should be demonstrated as recommended, for instance, in Ref. [15], Integrated computer codes are preferable. For instance, analyses of reactivity initiated accidents with spontaneous control rod withdrawal should employ a three dimensional (3-D) neutronic code with a built-in multichannel thermohydraulics code in order to determine the distortion of the neutron field and the redistribution of thermohydraulic parameters in the group of fuel channels affected by the distorted power density distribution. 

In order to determine the performance of the ALPES system with respect to the detection of an inadvertent control-rod withdrawal incident, control-rod insertion tests were performed at SUPERPHENIX in December 1996, at power level of 50 and 80 % nominal power. 

Neutronic calculations have shown that the consequences of such a test are symmetrical to those of a control-rod withdrawal. 

Local fault (Control-Rod Withdrawal Accident scenario)... 

Corrosion of a large proportion of the control rod material before the SNF has substantially corroded away. This could have the same effect on core reactivity as control rod withdrawal during a reactor start up, and cause criticality. 

Reactivity and power distribution anomalies inadvertent control rod withdrawal boron dilution due to a malfunction in the volume control system (for a PWR) wrong positioning of a fuel assembly. 

Reactivity and power distribution anomalies uncontrolled control rod withdrawal control rod ejection boron dilution due to the startup of an inactive loop (for a PWR). 

Interlocks are used on the intermediate range neutron monitors to ensure that all units are operating properly and on the proper range. Control rod withdrawal is blocked if the ratio of reactor power to recirculation flow exceeds a predefermined value. 

An electronically positioned mechanical rod stop system (RSS) prevents unprotected control rod withdrawal and excessive reactivity insertion. Components in the RSS include a redundant controller, a rod stop drive selector, and a limited capacity power supply which controls power to the rod stop adjustment drive motor for each control rod. 

The AOT may reasonably be expected to occur sometime in the service life of a plant. These events lead to no significant releases of radioactivity. The partial list of AOT include RPV pressure increase, RPV wafer (moderator) temperature decrease, control rod withdrawal, RPV coolant inventory decrease, reactor core coolant flow decrease, reactor core coolant flow increase, core coolanf femperature increase, and excess of coolant inventory. 

This may be an incredible accident for the STAR-LM, depending upon the provision and usage of control rods. Traditionally, reactivity insertion accidents result from unanticipated control rod withdrawal. The STAR-LM may operate without reactivity compensation due to fine adjustment of control rods. In such a case, control rods would not be partially inserted in the core such that inadvertent rod withdrawal could not realistically occur. In the past, this scenario has been analyzed for the STAR-LM by postulating a prescribed reactivity insertion at a prescribed ramp rate. 

The transient overpower (TOP) due to a control rod withdrawal, the loss of primary flow (LOT) and the loss of heat sink (LOHS) due to a loss of the heat removal capability of the secondary system are commonly postulated as accident scenarios for power reactors. Even though the loss of external power is commonly superposed on these events, this does not lead to any serious problem if the reactor is safely tripped. Severe accidents, where the failure of a scram system is superposed on the abovementioned accidents, are surveyed below, for the LSPR. The analytical methods employed are described in [XXV-7]. 

The FUJI MSR has very low excess reactivity, and even in the case of a malicious action of control rod withdrawal, the reactor would have no prompt criticality accidents with a release of radioactivity to the environment. 

The principal accidental sources of positive reactivity are 1) core flooding, 2) cold water insertion during equilibrium reactor operation and 3) excessive control rod withdrawal rates. There are two types of failures or errors which must be considered in any discussion on withdrawal of control rods. The two failiires or errors are 1) increased rod withdrawal speeds due to Inadvertent admission of air or gas into the hydraulic systems, and 2) steeper ramps resulting from maximum strength rods being involved in the rod withdrawal. 

Figure 7.6.2.3 (See Figure A-k of HW-71 8 V0L2) illustrates the case In which the excursion is assumed to occur at equilibrium level. The assumed reactivity ramp rate in this case is 0.0 per cent per second which corresponds to control rod withdrawal in excess of the simultaneous movement of the maximum number of contTOl rods together with gas in each hydraulic system so that the rods come out faster than normal. A balanced-flux is assumed, however. The power variation in Figure 7 6.2.3 is that for a nominal central tube operating initially at kw.

International practice considers the analysis of ATWS for a variety of initiating events such as loss of feedwater, loss of load, turbine trip, loss of condenser vacuum, loss of off-site power, closure of main steamline isolation valves, uncontrolled boron dilution, inadvertent control rod withdrawal, etc. ATWS analyses are performed in general by using best-estimate tools to determine the preventive (e.g. a diverse scram system) or mitigative measures (e.g. initiation of turbine trip and emergency feedwater supply) which need to be implemented for strengthening plants defence in depth. 

Subcritical multiplication techniques are used to measure the approach to criticality by uranium fuel addition, changes in core geometry, and control rod withdrawal. The basic concept is to measure the effect on neutron multiplication by positive reactivity additions. 

If the neutron flux diminishes during each succeeding generation when K is less than 1,0, why is it that the count rate increases in your reactor after each increment of control rod withdrawal, while subcritical ... 

The subcritical count rate has increased by a factor of 20 since initial control rod withdrawal began, what is the effective multiplication factor (Kg f) for this reactor with the rods partially withdrawn ... 

---