Core stability

*⁵⁾ [Translator’s note] The key regulatory guides authorized by the Nuclear Safety Commission of Japan are downloadable from its homepage < http://www.nsc.go.jp/NSCenglish/>.

thermal conductivities, so that the BWR core has a very good overall dynamic stability, considering the nuclear, thermal and hydraulic characteristics.

3. Zone stability

Zone stability is associated with the neutron flux oscillation properties with phase differences in different zones of the core due mainly to thermal-hydraulic factors combined with nuclear factors. The above-mentioned thermal-hydraulic and nuclear stabilities exclude the risk of zone instability.

4. Plant stability

Pressure increases in a BWR core decrease its void fraction resulting in a positive reactivity feedback. In order to avoid them, a "constant­pressure control policy" is adopted in combination with the functions of the recirculation flow control system and feed water control system, and the reactor response remains stable to any external disturbance.

5. Core stability against xenon oscillation

Changes in xenon density accumulated in the core causes power oscillation. Due to the large negative power coefficients, a BWR core remains stable against the xenon oscillation.

5. Operation and management of the core

i) Refueling

A BWR core is refueled periodically at the end of each operation cycle of about one year (an operation cycle is the period between two consecutive refuelings). When the core loses its capability (reactivity) to generate the rated thermal output at the end of one operation cycle, the number of fresh fuel assemblies required for the next operation cycle is determined. The number and location of irradiated fuel assemblies to be discharged from the core, and the location of fresh reload fuel assemblies to be loaded in the core are determined by considering the actual depletion (burn up) of individual fuel assemblies. The refueled core, also called the reloaded core, must meet the afore-mentioned design requirements (Section 2.3.2) of shutdown margin, thermal-hydraulic design limits and other criteria.

The number of replaced fuel assemblies may vary slightly at each operation cycle, but about one fourth of the fuel assemblies in the core are

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discharged from the core at each cycle. That is, a refueled core has approximately 25% fresh fuel and approximately 75% of partially irradiated fuel. These fresh and reloaded fuel assemblies are arranged in appropriate locations in the core.

2. Shutdown margin test

After the core is refueled in a scheduled outage, the shutdown margin is tested prior to the start of a new operation cycle for confirming that the reactor can be brought to the sub-critical condition at room temperature with the appropriate margin, under the condition that one control rod of the maximum worth is fully stuck outside of the core.

In an ABWR plant two control rods are driven by one control rod drive unit, so the shutdown margin test is performed with two control rods of maximum worth attached to one control rod drive unit stuck outside of the core.

2. Core operation

1. Reactivity control by control rods

The reactor is brought to the critical condition and to the rated power by withdrawing control rods from the core. While in operation at a constant power, the reactivity of the core gradually changes in the course of time; reactivity changes due to consumption and production of fissile nuclides (U-235 and Pu-239) as a result of fuel burn up, reactivity changes due to the burnable poison (gadolinia) burn up, etc. The control rods in the core need to be maneuvered from time to time to compensate for such gradual reactivity changes.

The control rods also adjust the core power distribution. The control rod insertion pattern is determined so that the operation limits (e.g., the minimum critical power ratio and the maximum linear power density) are satisfied.

2. Reactivity control by coolant flow control

Reactor power can partially be controlled by adjusting the coolant flow through the core. The principles of such power control mechanism are discussed below.

If the coolant flow through the core is increased, this enhances the sweeping out of the in-core void. Since the void production rate remains unchanged, the void fraction decreases, and a positive reactivity is added to the core (void coefficient). Consequently, the reactor power

increases until a new equilibrium power level is achieved, when the negative feedback reactivity by the increased void fraction counterbalances the transient excess reactivity initially caused by the decreased void fraction (due to the increased coolant flow). On the contrary, in order to reduce the reactor power level, the coolant flow through the core may be decreased. In the meantime (either to increase or to reduce the reactor power) no control rod maneuvers are needed.

Figure 2.3.22 shows a typical power-flow operating map of a conventional BWR-5. If the control rods are maneuvered (moved in or out) while keeping the speed of the recirculation pumps constant, the reactor power will change along the recirculation pump constant speed line on the graph. On the other hand, the reactor power response to a change in the speed of the recirculation pumps, while the control rod profiles remain unchanged, will be along the flow control line on the graph. The stability limiting line represents the maximum allowable operation condition in the "low flow- high power" region, set for increased stability.

In an ABWR, the stability limit line is eliminated owing to the design measures for increased stability (Figure 2.3.23).

3. Preconditioning interim operating management recommendation (PCIOMR)

Experience has shown that, among various design considerations to prevent fuel cladding failures as already mentioned in the preceding sections, precautions in operation management are equally effective as design considerations to prevent fuel cladding failures due to mechanical interactions between pellets and fuel cladding (Section 2.3.1- (2)-b)-i)). There is a strong correlation between the reactor operation and the cladding failures caused by the mechanical interactions between the pellets and cladding. This operation is called PCIOMR and it is based on the extensive experimental results obtained from the experimental reactor of the General Electric (GE) Company and other sources. The experimental results showed that below a "threshold value" of the linear power density, no cladding failures occurred no matter how fast the power was changed. However, above the

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and other important parameters are monitored by the operation monitoring system (process computers) and other instruments. Thermal output of each fuel assembly, the minimum critical power ratio and other operational parameters are determined by the neutron flux distribution (provided by nuclear instrumentations), the thermal output of the core, the core coolant flow and other parameters. Thermal output of the core is calculated from the plant heat balance data such as feed water flow, their temperatures and reactor pressure. The core flow rate is obtained from the flow rate of the jet pumps. In an ABWR, the core flow is obtained from the pressure difference across the recirculation pumps.

The power distribution of the nuclear reactor is controlled by either the control rod maneuvering or the core flow control, while constantly monitoring the core with instruments.

2. Failed fuel detection

Radioactivity concentration in the coolant is continuously monitored during operation. When a possible leak from a fuel assembly is detected, sipping tests are performed during the periodical inspection outage of the plant. Upon identifying failed fuel assemblies, they are removed from the core and not used in the next operation cycle. Such core management schemes facilitate effective fuel bum up while meeting the safety requirements.

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