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Continue the same-rod-number and same-height comparison from part(a), and assume unchanged pump speed, similar inlet conditions and single-phase forced flow for the normal-operation hydraulic argument.

Core power density increases. The active core volume decreases while total thermal power stays fixed. The fuel's own average volumetric power and the mean fuel linear heat rate do not increase merely from bringing the same rods closer together: fuel volume and rod count are unchanged.

Coolant mass flow rate normally decreases. Tighter passages reduce flow area and hydraulic diameter, increasing the core's hydraulic resistance. The unchanged centrifugal pumps do not enforce constant flow. The pump operating point moves to the intersection of the same pump curve with the steeper system-resistance curve, normally giving smaller flow at greater pump head.^[ref]

The core pressure drop normally increases for the usual pump curve. Its resistance at any given flow is higher. At the new, smaller flow, unchanged external piping and component losses are smaller, whereas a conventional falling pump head-versus-flow curve gives greater available head. The core takes the difference. This conclusion uses the stated pump-curve assumptions; holding flow fixed or holding pump head fixed would be different models. The exact new flow and pressure drop need both system and pump characteristics.

Fuel temperature would normally tend to rise, especially downstream, if the loss of flow increases coolant heating without a compensating improvement in film heat transfer. At fixed inlet state,shows that smaller total mass flow rate raises the bulk coolant temperature rise. At unchanged local fuel linear heat rate, the fuel-to-surface conductive contribution is approximately unchanged if material properties and gap conditions are held fixed. Its boundary temperature and the coolant film drop can nevertheless change.

The local coolant mass flux is $G=\dot{m}/A_f$. Both total flow and flow area decrease, so the sign of the change in $G$, the heat transfer coefficient and the fuel temperature cannot be deduced from total flow alone. Changes in power peaking and inlet conditions add further qualifications to the usual hotter-fuel tendency.

The minimum departure from nucleate boiling ratio would commonly decrease, implying reduced boiling margin, but its sign is conditional. The defining local ratio isThe numerator is the critical heat flux for the local conditions, and the denominator is actual rod-surface heat flux.^[ref] The mean surface heat loading remains approximately unchanged for the same rods and total power. Greater coolant enthalpy rise tends to reduce margin, while changes in $G$, geometry, pressure and local power determine the actual critical heat flux. A geometry-appropriate correlation or experiment is needed for a definite MDNBR change; unchanged hardware does not supply it.

A tighter core is likely to have a more demanding cooling response, but the specification does not determine a unique peak cladding temperature or reflood time in a loss-of-coolant accident. Several effects compete.

At the same pre-accident power and comparable irradiation history, the total decay heat source remains of broadly comparable scale; increasing core power density does not multiply the total decay power by the inverse change in core volume. The core nevertheless has less water within its tighter rod lattice. Its local thermal buffer is smaller, and the same heat must be removed through narrower passages. If the normal-operation fuel temperature is higher, more stored fuel thermal energy must also be removed after shutdown.

Reduced passage size and increased hydraulic resistance can impede emergency-coolant penetration and steam discharge during reflood. Local two-phase flow, entrainment and boiling heat transfer can change, so single-phase pump-curve reasoning is insufficient during blowdown and reflood. Less local water can favor earlier uncovering, but blowdown flow and the much larger inventory elsewhere in the unchanged primary circuit also matter. It would be unjustified to infer a proportionate reduction in the entire primary-circuit inventory.

Loss of water also removes neutron moderation. In the assumed under-moderated reactor, this normally contributes negative reactor reactivity, helping suppress fission power. It does not eliminate decay heat after shutdown. The thermal challenge must therefore be assessed separately from the sign of this feedback.

A quantitative comparison needs break characteristics, initial states, emergency injection performance, bypass paths and a transient two-phase calculation. The defensible conclusion is potentially reduced thermal margin and more difficult reflood, requiring reassessment of the emergency core cooling system, rather than an unconditional prediction that every large-break transient becomes worse in every metric.