Inherent reactivity feedback

The ultimate shutdown system (USS) located at the center of the core is a self-actuated shutdown system. The USS is actuated passively when the temperature of the primary sodium reaches the Curie point. The USS drops neutron absorbers by gravity as a means to bring the reactor to cold critical conditions in the event of a complete failure of the normal scram system and after the inherent reactivity feedbacks have brought the core to a safe, but critical state at an elevated temperature. 

LMRs with oxide-fueled core Models modified and newly developed mto the code so far mclude models for reactivity feedback effects and pool thermal-hydraulics In order to venfy the logic of the models developed, and to assess the effectiveness of the inherent safety features based upon the negative reactivity feedbacks m achieving the safety design objectives of passive safety, a preliminary analysis of UTOP and ULOF/LOHS performance has been attempted... 

Criterion 11 - Reactor inherent protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity. 

The PRISM shutdown systems are backed up by the inherent negative power reactivity feedback of the reactor core. This inherently negative reactivity feedback brings the core to a safe, stable, power state following accidents. 

Incorporation of inherent safety features into the core design, specifically, ensuring a large negative void reactivity feedback ... 

Inherent shutdown system Inherent characteristics based on reactivity feedbacks ... 

The strong coolant temperature-driven reactivity feedback in the fast neutron spectrum core enables autonomous load following whereby the reactor power self-adjusts itself to match heat removal from the primary coolant solely as a consequence of inherent physical phenomena. The system temperatures that are attained following an autonomous power change from the nominal steady state can be optimized through design of the core clamping... 

Reactivity control mechanism - Shutdown rod for reactor start-up and shutdown. - During operation, reactor power autonomously load follows by means of inherent physical processes without the need for any motion of control rods or any operator actions. - System temperatures change corresponding to reactivity feedbacks from fuel Doppler, fuel and cladding axial expansion, core radial expansion, and coolant density effects. - Control rods for possible line reactivity compensation during cycle. - Control rods also provide for diverse and independent shutdown. ... 

The core design provides for an inherent control of power by reactivity feedback within the core, such that acceptable fuel design limits are not exceeded for the defined beyond design basis accidents. 

Rely on inherent and passive safety features including the use of enhanced thermal expansion materials, a fast expansion module, a very low reactivity swing, and other reactivity feedback mechanisms to achieve a high degree of reactor self-control in the ATWS scenarios. 

Criterion 45. Control of the reactor core The reactor core should have prompt inherent nuclear feedback characteristics to compensate for rapid reactivity insertions. 

In some advanced pool type reactors, the consequences of a postulated break of the primary coolant pipe can be mitigated by the inherent safety functions. This is achieved through the reactivity feedbacks induced by the Doppler, as well as sodium, radial and axial expansion reactivity effects. The control rod driveline length increase, and the operation of gas expansion modules (GEM) are also important reactivity feedback mechanisms. 

The inherent safety characteristic against postulated events is the most remarkable superiority of a liquid metal cooled reactor (LMR) to other type of reactors. One of the major threats to the safety of LMR is a loss of flow event accompanied a failure of reactor shutdown systems. This situation is usually referred to as an unprotected loss of flow (ULOF). The inherent safety of the Korean Advanced Liquid Metal Reactor (KALIMER) during the ULOF [I] has been assessed for the situation of all pump trips followed by coastdown. It was assumed that the decay heat is removed by four intermediate heat exchangers (IHXs) and the safety grade system of passive safety decay heat removal system (PSDRS). The results showed that the power was stabilized by the reactivity feedback of the system even though the effect of the gas expansion module (GEM) was not taken into account. 

The present study analyzes a postulated break in the primary pump discharge pipe to assure the inherent safety of KALIMER. KALIMER is a pool-type liquid metal sodium cooled fast reactor plant. The main concern of the accident is the amount of subcooling margin reduction, i.e., the degree of increase in the fuel and the coolant temperatures. The stabilization of power associated with reactivity feedback is also an important aspect of the accident. The analysis is performed with the SSC-K code, which was developed on the basis of the SSC-L code for the... 

The KALIMER design highly emphasizes inherent safety, which maintains the core power reactivity coefficient negative during all modes of plant status and under accidental conditions as well. The reactivity feedback mechanisms consist of Doppler, thermal expansion of the fuel and coolant, thermal bowing of the core, thermal expansion of the core structure and core support structure, and thermal expansion of the control rod driveline. These effects result from either the physics laws, or both the physics laws and core design. 

The ultimate mechanism to ensure public protection from the consequences of postulated ATWS are the inherent negative reactivity feedback when the reactor system temperature increases, and the heat removal function of PSDRS. The analyses of the selected ATWS are conducted to assure the effectiveness of inherent safety features in the KALIMER design. The events considered are Unprotected control rod withdrawal (UTOP), Unprotected loss of heat sink (ULOHS), Unprotected loss of primary flow (ULOF), and combinations of those events. 

The break of one of the four core inlet pipes of KALIMER is analyzed and the inherent safety of KALIMER against a pipe break is evaluated in the present study. The reactivity feedbacks, the power trend and other parameters are predicted to show quite different behaviors depending on whether the GEM becomes effective or not. In the base case, in which the break area is equivalent to the diameter of inlet pipe, the core flow is reduced to about 65% full flow and the GEM level remains above the active core top. In this case, the power is governed by the combined effect of the reactivity feedbacks induced by the Doppler, sodium, radial, axial and control rod driveline behaviors. To overcome the deficiencies in modelling the pipe break with the SSC-K code, the core flow is artificially reduced to about 50% full flow, which results in the operation of GEM. 

The positive reactivity coefficient or negative coolant void reactivity coefficient is necessary for the inherent negative feedback of the Super LWR and Super FR at the loss of coolant accident. The reactor power should decrease automatically at the loss of coolant accident. 

An anticipated transient without scram (ATWS) is defined as an abnormal transient followed by failure of a reactor scram. Since the Super LWR is a simplified LWR, the probability of an ATWS is expected to be on the same order as that of LWRs. An ATWS of the Super LWR is classified as a beyond design basis event (BDBE). A deterministic evaluation of an ATWS is a global requirement because it is a potential safety issue that may lead to core damage under postulated conditions. Also, it is expected that inherent safety characteristics of nuclear reactors, not only reactivity feedback but also reactor system dynamics, can be clearly identified at ATWS conditions due to a scram failure. Therefore, deterministic ATWS analyses are carried out for the Super LWR in Sect. 6.7 as well as the abnormal transients and accidents. 

A positive reactivity ( 0.1) is inserted stepwise by withdrawing the CRs. The feedwater pump speed and the turbine control valve stroke are kept constant. The results are shown in Fig. 7.68 [31]. The reactor power increases about 10% almost stepwise due to the prompt jump and then gradually decreases due to the reactivity feedbacks from the fuel temperature and coolant density. This behavior implies that the Super FR also has inherent self controllability of the reactor power despite the much smaller density reactivity coefficient compared to that of the Super LWR. The main steam temperature increases, which leads to an increase in the main steam and core pressures because the specific volume of the main steam increases. The increase in the core pressure leads to a decrease in the feedwater and core flow rates, which increases the main steam temperature further. As a result, the maximum increase in the main steam temperature is nearly 40°C while that in the Super LWR is only 9°C (see Fig. 4.10). 

The RCSS and NCSS must provide the capability to control heat generation with moveable poisons and to control heat generation with inherent feedback. The moveable poison control function is accomplished both with a primary and a diverse secondary moveable poison control, while control with inherent feedback requires a negative temperature coefficient of reactivity. The NCSS and the RISS within the RS, also perform the function of heat generation control by maintaining the geometry for insertion of moveable poisons into the core. The NCSS monitors the neutron flux. 

EM pumps of both primary units trip with the flow coastdown facilitated by the synchronous motor (SM) system then the fuel and coolant temperatures rise because of the coolant flow decrease. However, the reactor power is decreased via the negative feedbacks resulting from Doppler, fuel expansion, and steel and coolant reactivity coefficients. After the flow coastdown by the SM systems, the primary coolant circulates within the reactor vessel, driven by natural convection, with a flow rate of approximately 20% of the nominal. Then, the inherently decreased power and the convection flow rate are balanced to a steady state. 

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