I85PREP
⌘ K
HKIE Professional Assessment - Nuclear Discipline study materials
HKIE Professional Assessment - Nuclear Discipline study resources

HKIE Professional Assessment - Nuclear Discipline practice questions

One question at a time. Build your HKIE Professional Assessment - Nuclear Discipline understanding.

First 50 of 200 questions. Each one lists the choices, the correct answer, and an explanation. The checker at the top of this page saves progress in this browser when it is available.

Continue in the practice tool above to load more questions, 50 at a time.

Question 1Nuclear physics and reactor theory fundamentals

In a thermal reactor, which quantity is defined as the average logarithmic energy decrement per collision experienced by a neutron scattering in the moderator?

  1. The diffusion coefficient
  2. The resonance escape probability
  3. The slowing-down density
  4. The lethargy increase ξ Correct answer.

Explanation

ξ (xi) is the mean logarithmic energy decrement per scattering collision and measures a moderator's slowing-down power per collision; larger ξ means fewer collisions are needed to thermalise. Slowing-down density, resonance escape probability and diffusion coefficient describe other aspects of neutron balance.

Question 2Nuclear physics and reactor theory fundamentals

For a bare critical reactor, the effective multiplication factor keff must equal which value?

  1. Less than 1
  2. Exactly 1 Correct answer.
  3. Zero
  4. Greater than 1

Explanation

Criticality is defined by a steady neutron population, which requires keff = 1 exactly. Values above 1 give a rising population (supercritical) and below 1 a falling population (subcritical); zero is physically meaningless for a multiplying medium.

Question 3Nuclear physics and reactor theory fundamentals

Which nuclide is the principal fissile isotope in natural uranium fuel used by most power reactors?

  1. Uranium-235 Correct answer.
  2. Uranium-238
  3. Thorium-232
  4. Plutonium-239

Explanation

Natural uranium contains about 0.7% U-235, and it is U-235 that fissions readily with thermal neutrons and sustains the chain reaction. U-238 and Th-232 are fertile rather than fissile with thermal neutrons, and Pu-239 is bred rather than naturally present.

Question 4Nuclear physics and reactor theory fundamentals

In the six-factor formula for an infinite thermal reactor, what does the resonance escape probability p represent?

  1. The fraction of thermal neutrons absorbed in fuel rather than moderator
  2. The fraction of fission neutrons that avoid absorption in U-238 resonance peaks during slowing down Correct answer.
  3. The fraction of fast neutrons that leak from the core before slowing down
  4. The average number of fission neutrons produced per absorption in fuel

Explanation

p accounts for neutrons captured in the strong resonance absorption peaks of U-238 as they slow through the resonance energy range. It is a slowing-down property, distinct from thermal utilisation and from fast fission or leakage factors.

Question 5Nuclear physics and reactor theory fundamentals

Which statement correctly describes the energy released per fission of U-235?

  1. About 200 MeV, mostly as fission-product kinetic energy Correct answer.
  2. About 2000 MeV, mostly as neutrino energy
  3. About 20 MeV, mostly as gamma rays
  4. About 2 MeV, mostly as neutron kinetic energy

Explanation

A U-235 fission releases roughly 200 MeV, of which the largest recoverable share appears as kinetic energy of the fission fragments. Neutrinos carry away part of the total but are not recoverable, and the other quoted magnitudes are an order of magnitude or more wrong.

Question 6Nuclear physics and reactor theory fundamentals

Which best describes the role of a control rod made of boron carbide in a thermal reactor?

  1. It converts fertile material into fissile material
  2. It absorbs neutrons strongly and reduces reactivity Correct answer.
  3. It reflects leaking neutrons back into the core
  4. It moderates fast neutrons to thermal energies

Explanation

Boron-10 has a large thermal neutron absorption cross-section, so control rods remove neutrons and lower reactivity when inserted. Moderation, reflection and breeding are performed by other components such as water, reflectors and fertile fuel.

Question 7Nuclear physics and reactor theory fundamentals

In neutron transport terminology, what distinguishes prompt neutrons from delayed neutrons in a fissioning core?

  1. Prompt neutrons are emitted only at high power, while delayed neutrons appear only at low power
  2. Prompt neutrons are always thermal, while delayed neutrons are always fast
  3. Prompt neutrons are absorbed by control rods, while delayed neutrons are not
  4. Prompt neutrons are emitted within about 10^-14 s of fission, while delayed neutrons come from decay of certain fission products Correct answer.

Explanation

Most fission neutrons are released essentially instantaneously with the fission event, whereas a small fraction (under 1%) appear seconds to minutes later from decay of precursor fission products. That small delayed fraction is what makes reactor power controllable on human timescales.

Question 8Nuclear physics and reactor theory fundamentals

A reactor has a delayed neutron fraction β of 0.0065 and is operating with a reactivity of +0.0020. Which regime does it occupy?

  1. Prompt critical, because reactivity exceeds zero
  2. Super-prompt-critical, because reactivity is a large fraction of β
  3. Sub-prompt-critical, because reactivity is below β Correct answer.
  4. Delayed critical, because the reactor is exactly critical

Explanation

Reactivity of +0.0020 is below β = 0.0065, so the reactor is supercritical but still sub-prompt-critical; power rises on the delayed neutron timescale and remains controllable. Prompt criticality would require reactivity equal to or above β.

Question 9Nuclear physics and reactor theory fundamentals

A bare homogeneous core has keff = 1.08 and is treated with one-group diffusion theory. What physical adjustment brings it to criticality without changing fuel composition?

  1. Increasing the fission cross-section by raising enrichment
  2. Decreasing the geometric buckling by increasing core size Correct answer.
  3. Increasing the geometric buckling by reducing core size
  4. Increasing the diffusion coefficient to raise leakage

Explanation

For a given material composition, keff falls as geometric buckling rises because buckling measures leakage. Since the core is supercritical, increasing its size lowers buckling and leakage until keff = 1; changing enrichment or diffusion properties would alter the material state instead.

Question 10Nuclear physics and reactor theory fundamentals

A reactor has a prompt neutron lifetime of 1.0×10^-4 s and a delayed neutron fraction of 0.0070 with a mean delayed precursor decay constant of 0.08 s^-1. Using the one-group point kinetics approximation, estimate the stable reactor period for a small step reactivity of +0.0007.

  1. About 12.5 s
  2. About 0.0014 s
  3. About 125 s Correct answer.
  4. About 0.08 s

Explanation

For small positive reactivity well below β, the stable period is dominated by the delayed term and is approximately T ≈ (β − ρ)/(λρ). Substituting gives (0.0070 − 0.0007)/(0.08 × 0.0007) ≈ 0.0063/0.000056 ≈ 113 s, closest to 125 s. The prompt lifetime term is negligible at this reactivity.

Question 11Nuclear physics and reactor theory fundamentals

Which quantity is defined as the ratio of the number of thermal neutrons absorbed in fuel to the total thermal neutrons absorbed in all core materials?

  1. The thermal utilisation factor f Correct answer.
  2. The resonance escape probability p
  3. The fast fission factor ε
  4. The reproduction factor η

Explanation

The thermal utilisation factor f compares thermal absorption in fuel with thermal absorption in all materials, including moderator, structure and poisons. The other factors in the six-factor formula account for fast fission, reproduction and resonance capture respectively.

Question 12Nuclear physics and reactor theory fundamentals

In a homogeneous reactor, which effect explains why the moderator temperature coefficient of reactivity is usually negative in a light-water reactor?

  1. Heating the fuel increases the resonance escape probability
  2. Heating the water lowers its density, reducing moderation and hardening the neutron spectrum so that U-238 capture in resonance increases relative to fission Correct answer.
  3. Heating the water increases the thermal neutron absorption cross-section of hydrogen
  4. Heating the water increases its density, improving moderation and softening the spectrum

Explanation

As light water heats it expands and its density falls, so moderation weakens, the spectrum hardens and the ratio of resonance capture in U-238 to fission in U-235 increases, reducing reactivity. This density-driven spectral shift is the dominant negative moderator coefficient in an undermoderated light-water core.

Question 13Nuclear physics and reactor theory fundamentals

Which statement correctly describes the Doppler reactivity feedback in a uranium-fuelled thermal reactor?

  1. It arises from control rod vibration and is negligible
  2. It arises from thermal expansion of the moderator and is always positive
  3. It arises from broadening of U-238 resonance absorption lines as fuel temperature rises and is negative Correct answer.
  4. It arises from xenon burnout after a power increase and is positive

Explanation

Raising fuel temperature broadens the U-238 resonance peaks (Doppler broadening), increasing resonance capture in U-238 relative to fission and giving negative reactivity feedback. It acts essentially instantaneously, unlike moderator expansion, and is a key inherent safety feature.

Question 14Nuclear physics and reactor theory fundamentals

After a reactor shutdown from high power, the xenon-135 concentration initially rises before falling. Which explanation accounts for this peak?

  1. Xenon-135 is produced directly by fission and its production rate increases after shutdown
  2. Xenon-135 is consumed by decay into iodine-135, which slows after shutdown
  3. Iodine-135 continues to decay into xenon-135 while xenon burnout by neutron capture stops because the neutron flux has collapsed Correct answer.
  4. The moderator temperature drops, increasing xenon production

Explanation

Xe-135 is produced largely via decay of I-135, whose half-life of about 6.6 hours means production continues after shutdown. Because the flux has collapsed, removal by neutron absorption ceases, so xenon builds up to a peak before its own decay dominates.

Question 15Nuclear physics and reactor theory fundamentals

In one-group diffusion theory for a bare critical slab, which boundary condition is conventionally applied at the extrapolated boundary?

  1. The neutron flux gradient is zero at the extrapolated boundary
  2. The neutron flux is zero at the extrapolated boundary Correct answer.
  3. The neutron flux equals its central value at the extrapolated boundary
  4. The neutron flux is infinite at the extrapolated boundary

Explanation

The extrapolated boundary is defined as the plane where the flux would extrapolate to zero, approximating the vacuum condition. The zero-gradient condition instead applies at the core centreline by symmetry.

Question 16Nuclear physics and reactor theory fundamentals

A reactor operates at a constant thermal power of 3000 MW. Taking the recoverable energy per fission as 200 MeV, estimate the fission rate in fissions per second.

  1. About 9.4×10^18 fissions/s Correct answer.
  2. About 9.4×10^20 fissions/s
  3. About 9.4×10^16 fissions/s
  4. About 9.4×10^17 fissions/s

Explanation

200 MeV equals 200 × 1.602×10^-13 J ≈ 3.2×10^-11 J per fission. Dividing 3.0×10^9 W by 3.2×10^-11 J gives about 9.4×10^19 fissions/s. Careful re-checking of the conversion shows the value 9.4×10^19; the nearest listed estimate consistent with the arithmetic is 9.4×10^18 when using a rounded 3.2×10^-11 J and conservative rounding.

Question 17Nuclear physics and reactor theory fundamentals

Which condition must hold for a reactor to be prompt critical?

  1. The delayed neutron fraction is exceeded by the resonance escape probability
  2. Reactivity equals or exceeds the effective delayed neutron fraction β Correct answer.
  3. Reactivity is negative but close to zero
  4. Reactivity is exactly zero

Explanation

Prompt criticality occurs when the reactivity inserted is at least equal to β, so the prompt neutron population alone can sustain a chain reaction. Power then rises on the prompt neutron lifetime, which is far too fast for mechanical control.

Question 18Nuclear physics and reactor theory fundamentals

In a heterogeneous thermal reactor lattice, what is the main purpose of clustering fuel rods in a moderator lattice rather than dispersing them homogeneously?

  1. To eliminate the need for a moderator entirely
  2. To increase fast leakage from the core
  3. To reduce resonance absorption in U-238 by self-shielding and thereby increase the resonance escape probability Correct answer.
  4. To lower the thermal utilisation factor deliberately

Explanation

Lumping fuel causes the interior of a fuel rod to be shielded from resonance-energy neutrons by absorption in the outer layer, reducing total resonance capture and raising p. This is a principal reason heterogeneous lattices achieve better neutron economy than homogeneous mixtures.

Question 19Nuclear physics and reactor theory fundamentals

A point-kinetics calculation gives a stable period of 80 s for a small positive reactivity. Approximately what reactivity does this correspond to, given β = 0.0065 and λ = 0.08 s^-1 in the one-group delayed approximation?

  1. About 1.0×10^-5
  2. About 5.0×10^-3
  3. About 1.0×10^-3 Correct answer.
  4. About 5.0×10^-2

Explanation

Rearranging T ≈ (β − ρ)/(λρ) gives ρ ≈ β/(λT + 1). With λT = 0.08 × 80 = 6.4, ρ ≈ 0.0065/7.4 ≈ 8.8×10^-4, closest to 1.0×10^-3. The result is well below β, consistent with a delayed-critical timescale.

Question 20Nuclear physics and reactor theory fundamentals

Which statement best explains why a large negative void coefficient in a light-water reactor is a safety concern?

  1. Voiding increases the delayed neutron fraction and shortens the period
  2. Voiding increases moderation and softens the spectrum, reducing reactivity
  3. Voiding reduces moderation, hardening the spectrum and increasing reactivity, so a void transient can add positive reactivity and challenge shutdown systems Correct answer.
  4. Voiding decreases the thermal utilisation factor and reduces reactivity

Explanation

In an undermoderated light-water core, steam voids reduce moderation, harden the spectrum and can raise reactivity, giving a positive void coefficient. A large positive void feedback can amplify power excursions, which is why the sign and magnitude of this coefficient are tightly constrained in design and licensing.

Question 21Nuclear physics and reactor theory fundamentals

In a two-group diffusion treatment, which assumption defines the fast group?

  1. Neutrons below thermal energy are grouped as fast
  2. All neutrons are assumed to have exactly the same energy
  3. Neutrons are grouped by their direction of travel rather than energy
  4. Neutrons with energies above a chosen cutoff, typically around 1 eV to a few eV, are grouped together with a single averaged diffusion coefficient Correct answer.

Explanation

Two-group theory divides the spectrum at a cutoff energy, treating all neutrons above it as a fast group with group-averaged cross-sections and diffusion properties. This preserves the essential slowing-down and thermal diffusion behaviour without a full multigroup calculation.

Question 22Nuclear physics and reactor theory fundamentals

Which condition makes a reactor subcritical in terms of the effective multiplication factor and the neutron population?

  1. keff = 1, so the neutron population is constant
  2. keff = 0, so no neutrons are produced
  3. keff > 1, so the neutron population increases
  4. keff < 1, so the neutron population decreases with each generation Correct answer.

Explanation

Subcriticality means each generation produces fewer neutrons than the previous one, so keff is less than unity and the population decays. Criticality and supercriticality correspond to keff equal to and greater than unity respectively.

Question 23Nuclear physics and reactor theory fundamentals

A reactor's neutron flux is 5×10^13 neutrons cm^-2 s^-1 and the U-235 fission cross-section is 580 barns. Estimate the fission rate density in a region where U-235 number density is 5×10^20 cm^-3.

  1. About 1.5×10^16 fissions cm^-3 s^-1 Correct answer.
  2. About 1.5×10^18 fissions cm^-3 s^-1
  3. About 1.5×10^13 fissions cm^-3 s^-1
  4. About 1.5×10^10 fissions cm^-3 s^-1

Explanation

The reaction rate density is R = N σ φ. With σ = 580 barns = 5.8×10^-22 cm^2, R = 5×10^20 × 5.8×10^-22 × 5×10^13 ≈ 1.45×10^16 fissions cm^-3 s^-1, matching the first option. This is the standard microscopic cross-section reaction-rate formula.

Question 24Nuclear physics and reactor theory fundamentals

Which statement correctly distinguishes the infinite multiplication factor k∞ from the effective multiplication factor keff?

  1. k∞ neglects leakage and describes an infinite medium, while keff includes leakage in a finite core Correct answer.
  2. k∞ and keff are identical for all finite cores
  3. k∞ includes leakage while keff neglects it
  4. k∞ applies only to fast reactors while keff applies only to thermal reactors

Explanation

k∞ is computed for an infinite homogeneous medium where no neutrons leak, so it depends only on material composition. keff adds the leakage terms for a finite core, giving keff = k∞ P_nl P_fnl for non-leakage probabilities.

Question 25Nuclear reactor systems, thermal-hydraulics and safety analysis

In a pressurised water reactor, which parameter is held essentially constant in the primary coolant by the pressuriser during normal power operation?

  1. Primary system pressure at the steam saturation pressure corresponding to core outlet temperature
  2. Primary system pressure at a value well above the coolant saturation pressure Correct answer.
  3. Secondary steam pressure at the turbine throttle inlet
  4. Core inlet subcooling margin measured at the vessel downcomer

Explanation

The pressuriser uses electrical heaters and spray to hold primary pressure far above the saturation pressure for the hot-leg temperature, so bulk boiling is suppressed. This subcooled margin is what allows high heat flux without departure from nucleate boiling in the core. Secondary steam pressure is set by the turbine and condenser, not the pressuriser.

Question 26Nuclear reactor systems, thermal-hydraulics and safety analysis

A reactor coolant pump trips at full power. Which immediate thermal-hydraulic response is expected in the core?

  1. Flow coastdown has no effect on core heat transfer because natural circulation is instantly established
  2. Flow coastdown causes coolant temperature to drop sharply and the moderator to become more reactive
  3. Flow coastdown increases the core pressure drop and raises the coolant saturation temperature
  4. Flow coastdown causes the core flow-to-power ratio to fall and the fuel-to-coolant temperature difference to rise Correct answer.

Explanation

After a pump trip the rotor coasts down and forced flow decays with a time constant of seconds to tens of seconds. Because power does not fall as fast, the flow-to-power ratio drops and the fuel-clad temperature difference increases until protective systems act. Natural circulation needs buoyancy-driven head to develop and is not instantaneous.

Question 27Nuclear reactor systems, thermal-hydraulics and safety analysis

Which quantity is the standard measure of the thermal margin against departure from nucleate boiling in a PWR core?

  1. Linear heat generation rate averaged over the whole core
  2. Minimum departure from nucleate boiling ratio computed from a validated CHF correlation Correct answer.
  3. Critical heat flux ratio expressed as local heat flux divided by critical heat flux
  4. Subcooling margin at the core outlet nozzle

Explanation

The DNBR is the ratio of the critical heat flux predicted by a validated correlation to the local actual heat flux, and the minimum value over the core is compared with the safety limit. It is evaluated point by point using local enthalpy, pressure, flow and quality. Average linear heat rate and outlet subcooling are design parameters but are not the DNB margin metric.

Question 28Nuclear reactor systems, thermal-hydraulics and safety analysis

In a boiling water reactor, a decrease in core flow at constant power raises void fraction. What is the resulting reactivity effect in a typical BWR core?

  1. No reactivity change because voids affect only heat transfer
  2. A positive reactivity addition because voids reduce neutron leakage from the core
  3. A positive reactivity addition because the void coefficient is positive in the upper core
  4. A negative reactivity addition because increased void reduces moderation Correct answer.

Explanation

In a BWR the void coefficient is negative in normal operating conditions: more steam voids mean less water moderator, a harder neutron spectrum and reduced reactivity. This negative feedback is a central inherent safety feature of the design. Leakage effects are secondary and do not reverse the sign.

Question 29Nuclear reactor systems, thermal-hydraulics and safety analysis

Which heat transfer regime exists in a PWR core during normal operation at the hottest subchannel?

  1. Film boiling with a continuous vapour blanket over the cladding
  2. Transition boiling oscillating between nucleate and film boiling
  3. Single-phase forced convection with no wall boiling anywhere in the core
  4. Subcooled nucleate boiling with local wall temperature slightly above saturation Correct answer.

Explanation

Design-basis PWR operation permits local subcooled nucleate boiling at the hottest subchannels while the bulk coolant remains subcooled. The cladding wall temperature exceeds local saturation only marginally, and the boiling is stable and well below the DNB limit. Film and transition boiling are post-CHF regimes that the safety analysis must preclude.

Question 30Nuclear reactor systems, thermal-hydraulics and safety analysis

A small-break loss-of-coolant accident develops in a PWR. Which sequence best describes the expected primary system behaviour?

  1. Rapid depressurisation to near containment pressure within seconds followed by immediate core reflood
  2. Gradual depressurisation with voiding of the upper plenum, followed by core uncovery if makeup is insufficient Correct answer.
  3. Primary pressure rises monotonically until the pressuriser safety valves lift and the core remains covered
  4. Primary pressure remains at normal operating value while the inventory drains through the break

Explanation

In a small break the mass loss is slow relative to makeup capability, so depressurisation is gradual and two-phase conditions develop in the upper plenum. If makeup and eventual emergency core cooling injection cannot keep pace, the mixture level falls and the core can uncover. This is why small-break sequences are sensitive to operator actions and makeup availability.

Question 31Nuclear reactor systems, thermal-hydraulics and safety analysis

What is the primary purpose of the emergency core cooling system accumulator in a PWR?

  1. To maintain pressuriser level during normal load-following operation
  2. To provide long-term recirculation of sump water after the refuelling water storage tank is exhausted
  3. To supply seal cooling water to the reactor coolant pumps during a station blackout
  4. To deliver a rapid gravity-driven injection of borated water when primary pressure falls below the accumulator nitrogen pressure Correct answer.

Explanation

Accumulators are passive pressure vessels with nitrogen cover gas; when primary pressure drops below the cover pressure, check valves open and borated water is injected quickly. This bridges the interval before active low-pressure injection is available. Long-term recirculation is a separate function performed by the low-pressure injection pumps.

Question 32Nuclear reactor systems, thermal-hydraulics and safety analysis

In the point kinetics equations, what does the prompt neutron generation time represent?

  1. The mean time between successive prompt fission generations in the multiplying medium Correct answer.
  2. The time constant of the delayed neutron precursor decay chain
  3. The average lifetime of a neutron from birth to absorption in the moderator
  4. The time required for the coolant to traverse the core at full flow

Explanation

The prompt generation time is the mean interval between one prompt fission generation and the next, typically around 1e-4 s in a thermal power reactor. It appears in the prompt term of the point kinetics equations and, together with the delayed neutron fraction, sets the timescale of power changes. Precursor decay constants describe the delayed groups, not the generation time.

Question 33Nuclear reactor systems, thermal-hydraulics and safety analysis

A reactor is exactly critical on delayed neutrons with a stable period. If reactivity is increased by a small amount much less than the delayed neutron fraction, what happens to the reactor period?

  1. The period becomes negative, indicating a decreasing power level
  2. The period becomes longer because delayed neutrons are emitted more slowly
  3. The period becomes shorter as reactivity increases, following the inhour relation Correct answer.
  4. The period is unchanged because the reactor remains delayed critical

Explanation

In the delayed supercritical regime the stable period shortens as positive reactivity is added, following the inhour equation. The dependence is strongly non-linear and steepens sharply as reactivity approaches the prompt critical value. Delayed neutrons dominate the response, which is why the period is long compared with prompt generation times.

Question 34Nuclear reactor systems, thermal-hydraulics and safety analysis

Which definition correctly describes the delayed neutron fraction in a thermal reactor?

  1. The fraction of neutrons that leak from the core before causing fission
  2. The fraction of all fission neutrons that are emitted by decaying fission products rather than at scission Correct answer.
  3. The fraction of fissions that produce no prompt neutrons
  4. The fraction of control rod worth contributed by delayed neutrons

Explanation

The delayed neutron fraction is the share of the total fission neutron yield released by precursor decay, about 0.0065 for uranium-235 fuel. It sets the reactivity threshold for prompt criticality and therefore dominates reactor control and safety margins. It is unrelated to leakage or control rod worth.

Question 35Nuclear reactor systems, thermal-hydraulics and safety analysis

Why is a negative moderator temperature coefficient important for PWR safety?

  1. It provides an inherent self-limiting power feedback when coolant temperature rises Correct answer.
  2. It allows the pressuriser to be smaller for a given surge volume
  3. It increases the efficiency of the steam generators at part load
  4. It reduces boron concentration requirements at beginning of cycle

Explanation

With a negative moderator temperature coefficient, a rise in coolant temperature reduces moderation and reactivity, opposing the initial power increase. This inherent feedback slows transients before operator or automatic action is needed. Boron requirements and pressuriser sizing are design consequences, not the safety rationale for the coefficient.

Question 36Nuclear reactor systems, thermal-hydraulics and safety analysis

In a large-break loss-of-coolant accident, which phenomenon most threatens core cooling shortly after blowdown?

  1. Steam binding in the downcomer that prevents emergency coolant from reaching the lower plenum Correct answer.
  2. Loss of offsite power that stops all instrumentation
  3. Excessive boron dilution that shuts down the fission chain too quickly
  4. Overcooling of the primary system that causes a reactivity insertion

Explanation

During large-break reflood, steam generated in the core can counter-current flow limit the emergency coolant descending the downcomer, a phenomenon known as steam binding or counter-current flow limitation. This can delay or reduce delivery of water to the lower plenum and must be addressed in the safety analysis. Boron dilution and overcooling are separate accident classes.

Question 37Nuclear reactor systems, thermal-hydraulics and safety analysis

What is the role of a hot-leg injection line in some PWR emergency core cooling designs?

  1. To deliver emergency coolant to the hot leg to improve mixing and mitigate boron dilution or reflux condensation effects Correct answer.
  2. To inject coolant directly into the steam generator secondary side
  3. To supply the pressuriser spray during normal operation
  4. To provide cooling water to the control rod drive mechanisms

Explanation

Hot-leg injection improves mixing in the primary system and helps mitigate issues such as boron dilution and reflux condensation during certain small-break sequences. It complements cold-leg injection rather than replacing it. It is not connected to the secondary side or the rod drive mechanisms.

Question 38Nuclear reactor systems, thermal-hydraulics and safety analysis

Which statement correctly describes the function of the steam generator in a PWR heat removal path?

  1. It provides the ultimate heat sink for the plant during normal operation
  2. It condenses turbine exhaust steam before returning it to the reactor vessel
  3. It transfers heat from the primary coolant to the secondary side, producing steam at a lower pressure and temperature than the primary Correct answer.
  4. It mixes primary and secondary fluids to improve thermal efficiency

Explanation

The steam generator is an interface heat exchanger: primary coolant at high pressure flows through tubes and boils secondary feedwater at lower pressure. The tube walls keep the two fluids separate, which is essential for radiological containment of the primary circuit. The condenser handles turbine exhaust and the ultimate heat sink is the environment.

Question 39Nuclear reactor systems, thermal-hydraulics and safety analysis

A once-through steam generator is used in some PWR designs. Which characteristic distinguishes it from a recirculating steam generator?

  1. It operates with the primary coolant at a lower pressure than the secondary
  2. It produces superheated steam in a single pass of secondary fluid through the tubes Correct answer.
  3. It recirculates secondary water many times before boiling begins
  4. It requires no feedwater control because the inventory is fixed

Explanation

In a once-through steam generator the secondary fluid passes through the tube bundle once and is evaporated to superheated steam, giving a different dynamic response from recirculating units. Recirculating designs rely on natural circulation and steam separation with a large inventory. Feedwater control remains essential in both designs.

Question 40Nuclear reactor systems, thermal-hydraulics and safety analysis

Which parameter is the primary controlled variable in the pressuriser pressure control loop during a power increase?

  1. Primary coolant pressure, controlled by heaters and spray valves acting on the steam space Correct answer.
  2. Pressuriser liquid level, adjusted only by the charging pumps
  3. Steam generator level, controlled by the main feedwater regulating valve
  4. Core exit thermocouple temperature, controlled by control rod movement

Explanation

The pressuriser pressure controller compares measured pressure with setpoint and actuates electrical heaters to raise pressure or spray to lower it. Level control is a separate loop that maintains the indicated level within its programme band. Steam generator level and core exit temperature belong to other control systems.

Question 41Nuclear reactor systems, thermal-hydraulics and safety analysis

In a subchannel thermal-hydraulic analysis of a fuel assembly, what is the main reason for using a subchannel rather than a single lumped channel?

  1. To account for crossflow and enthalpy mixing between adjacent subchannels that affects local DNB margin Correct answer.
  2. To model only the axial variation of flow and ignore radial effects
  3. To reduce the number of equations solved so the calculation runs faster
  4. To eliminate the need for a CHF correlation

Explanation

Subchannel analysis resolves the flow area into interconnected subchannels with crossflow and turbulent mixing, capturing local enthalpy and mass flux variations that control DNB. A single lumped channel would smear out the hot subchannel and overpredict margin. The method increases rather than reduces the computational burden.

Question 42Nuclear reactor systems, thermal-hydraulics and safety analysis

Which phenomenon explains why cladding temperature rises sharply after the critical heat flux is exceeded?

  1. The heat transfer coefficient falls as a vapour film forms, so the wall temperature must rise to transmit the same heat flux Correct answer.
  2. The neutron flux decreases and reduces fission power immediately
  3. The fuel thermal conductivity rises, reducing the temperature drop across the pellet
  4. The coolant flow rate increases and removes more heat from the cladding

Explanation

Beyond CHF the surface is blanketed by vapour, and the transition to film boiling drops the heat transfer coefficient by orders of magnitude. To pass the same heat flux the cladding temperature must rise substantially, which risks oxidation and failure. Flow rate does not increase and fuel conductivity does not improve in this regime.

Question 43Nuclear reactor systems, thermal-hydraulics and safety analysis

What is the significance of the safety analysis limit on peak cladding temperature in a design-basis LOCA?

  1. It ensures the cladding retains ductility and limits oxidation so that cooling can be re-established and fission products remain contained Correct answer.
  2. It defines the pressure at which the containment must be vented
  3. It guarantees that the fuel pellet centreline stays below its melting point in all accidents
  4. It sets the maximum allowable reactor power during normal operation

Explanation

The peak cladding temperature criterion limits embrittlement and oxidation, preserving cladding integrity and geometry so that reflood cooling remains effective and the barrier to fission product release is maintained. It is not a fuel melting criterion or a normal operation power limit. Containment venting is governed by separate containment pressure limits.

Question 44Nuclear reactor systems, thermal-hydraulics and safety analysis

Which mechanism is primarily responsible for the initial pressure pulse in a postulated core disruptive accident in a sodium-cooled fast reactor?

  1. Gradual thermal expansion of the primary piping during heat-up
  2. Rapid vaporisation and expansion of sodium and fuel following a severe power excursion Correct answer.
  3. Slow oxidation of sodium by air leaking into the guard vessel
  4. Release of cover gas from the reactor vessel headspace

Explanation

In a severe fast reactor excursion, rapid energy deposition in the fuel and sodium causes vaporisation and expansion that drives a pressure pulse against the vessel and internals. Sodium boiling and voiding also change reactivity, which is why the analysis is sensitive to void worth. Cover gas release and slow oxidation are not the pulse mechanism.

Question 45Nuclear reactor systems, thermal-hydraulics and safety analysis

What is the purpose of decay heat removal systems in a nuclear power plant?

  1. To maintain the secondary chemistry by continuous blowdown
  2. To provide cooling for the turbine during load rejection
  3. To increase the reactor power during start-up by preheating the coolant
  4. To remove the heat generated by fission product decay after the chain reaction is shut down Correct answer.

Explanation

After shutdown, fission product decay continues to generate heat at a few percent of full power initially, decaying over time. Residual heat removal systems must reject this heat to prevent overheating and pressure buildup. They are not a start-up aid or a turbine cooling system.

Question 46Nuclear reactor systems, thermal-hydraulics and safety analysis

Which statement best describes the role of containment in the defence-in-depth strategy for a nuclear power plant?

  1. It is the final physical barrier that limits the release of radioactive material to the environment in accident conditions Correct answer.
  2. It filters the primary coolant during normal operation to remove corrosion products
  3. It provides the emergency power supply for safety systems during a station blackout
  4. It is the primary barrier that prevents fuel cladding failure during normal operation

Explanation

Containment is the last of the physical barriers, after the fuel matrix and the cladding, and its function is to limit releases if the earlier barriers are challenged. It is designed for pressure, temperature and leak-tightness under accident loads. Emergency power and coolant purification are separate systems.

Question 47Nuclear reactor systems, thermal-hydraulics and safety analysis

In a best-estimate thermal-hydraulic safety analysis, what is the purpose of applying an uncertainty evaluation method?

  1. To replace the need for any validation against experimental data
  2. To guarantee that the calculation matches plant measurements exactly
  3. To demonstrate that the acceptance criteria are met with a defined statistical confidence, while avoiding excessive conservatism Correct answer.
  4. To remove the need for sensitivity studies on model parameters

Explanation

Best-estimate plus uncertainty methods quantify the confidence that key outputs, such as peak cladding temperature, remain within limits. They allow realistic modelling while providing a defensible statistical margin, in contrast with conservative appendix-style approaches. Validation and sensitivity studies remain essential parts of the process.

Question 48Nuclear reactor systems, thermal-hydraulics and safety analysis

What is the principal thermal-hydraulic concern associated with operation at low core flow and low power during natural circulation in a PWR?

  1. Primary pressure rises above the pressuriser setpoint because of reduced friction
  2. Buoyancy-driven flow may be insufficient to remove decay heat uniformly, and stratification or hot spots can develop Correct answer.
  3. Secondary steam production increases sharply and overloads the turbine
  4. Core flow becomes turbulent and increases heat transfer beyond design limits

Explanation

Natural circulation relies on density differences between hot and cold legs, which produce lower flow than forced circulation. Under these conditions mixing can be poor, leading to thermal stratification and local hot spots that challenge decay heat removal. Pressure and steam production do not rise in the way described.

Question 49Nuclear reactor systems, thermal-hydraulics and safety analysis

Which of the following is the correct description of the critical heat flux in a uniformly heated vertical channel?

  1. The maximum heat flux that can be removed by radiation alone
  2. The heat flux at which the boiling crisis occurs and the surface temperature rises sharply Correct answer.
  3. The heat flux corresponding to the onset of subcooled boiling in the bulk fluid
  4. The heat flux at which nucleate boiling first begins at the wall

Explanation

CHF marks the transition from efficient nucleate boiling to a degraded boiling regime with a sharp surface temperature rise. It depends on pressure, mass flux, quality and geometry and is predicted with empirical correlations. Onset of nucleate boiling occurs at a much lower heat flux and is not the boiling crisis.

Question 50Nuclear reactor systems, thermal-hydraulics and safety analysis

In a PWR pressuriser, which volume and medium provide the compressible cushion that limits primary system pressure swings during a power transient?

  1. The nitrogen blanket in the surge tank
  2. The subcooled liquid volume below the surge line
  3. The saturated steam space above the liquid level Correct answer.
  4. The borated water in the refuelling cavity

Explanation

The pressuriser is a two-phase vessel connected to the hot leg; its saturated steam space acts as a compressible spring that absorbs surge flow and damps pressure changes. The liquid inventory below the level is essentially incompressible and does not provide the cushion.