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Question 1Nuclear safety: neutron balance, heat removal, radiation and responsibility

According to Article 5 of the Nuclear Safety Law of the People's Republic of China (《中华人民共和国核安全法》), which entity bears the primary and comprehensive legal responsibility for the nuclear safety of a nuclear facility?

  1. The nuclear facility operating organization (核设施营运单位) Correct answer.
  2. The National Nuclear Safety Administration (国家核安全局)
  3. The engineering design institute that designed the nuclear power plant
  4. The regional nuclear and radiation safety inspection station

Explanation

Article 5 of the PRC Nuclear Safety Law establishes the foundational statutory principle that the operating organization of a nuclear facility bears comprehensive and ultimate legal responsibility for nuclear safety (营运单位对核安全负全面责任). Regulatory oversight by the National Nuclear Safety Administration does not relieve or diminish the operating organization's legal safety obligations.

Question 2Nuclear safety: neutron balance, heat removal, radiation and responsibility

Nuclear Safety Culture (核安全文化) is a foundational concept in nuclear governance. Which of the following best reflects the core personal attribute expected of every nuclear professional at the individual level?

  1. Prioritizing electricity production quota targets over procedural pauses when minor discrepancies occur.
  2. Attempting to troubleshoot and bypass safety interlocks independently without logging documentation.
  3. A questioning attitude, rigorous and prudent approach, and open communication (质疑的态度、严谨审慎的工作方法和开放的沟通). Correct answer.
  4. Concealing minor operational anomalies to avoid regulatory scrutiny and maintain site reputation.

Explanation

According to IAEA INSAG-4, HAF safety codes, and Chinese nuclear safety guidelines, the individual level of Nuclear Safety Culture requires a questioning attitude (never assuming everything is normal), a rigorous and prudent working method (adhering strictly to validated procedures), and open, non-punitive communication reporting errors and near-misses.

Question 3Nuclear safety: neutron balance, heat removal, radiation and responsibility

In nuclear reactor physics, what are the four physical factors that comprise the Four-Factor Formula for the infinite multiplication factor (k_inf = epsilon * p * f * eta)?

  1. Elastic scattering factor, inelastic scattering factor, capture probability, and leakage probability
  2. Fast fission factor (epsilon), resonance escape probability (p), thermal utilization factor (f), and thermal reproduction factor (eta) Correct answer.
  3. Thermal conductivity, heat transfer coefficient, mass flux, and void fraction
  4. Core power peaking factor, axial offset, linear heat rate, and burnup factor

Explanation

The classical Four-Factor Formula describes the infinite multiplication factor (k_inf) in an infinite medium without leakage: epsilon is the fast fission factor, p is the resonance escape probability, f is the thermal utilization factor, and eta is the thermal reproduction factor (number of fission neutrons produced per thermal neutron absorbed in the fuel).

Question 4Nuclear safety: neutron balance, heat removal, radiation and responsibility

How does the Six-Factor Formula relate the effective multiplication factor (k_eff) of a finite reactor core to the infinite multiplication factor (k_inf)?

  1. k_eff = k_inf / (P_FNL * P_TNL)
  2. k_eff = k_inf + P_FNL - P_TNL
  3. k_eff = k_inf * P_FNL * P_TNL (where P_FNL and P_TNL are the fast and thermal non-leakage probabilities) Correct answer.
  4. k_eff = k_inf * (1 - beta_eff)

Explanation

In a finite reactor core, neutrons can leak out of the system during moderation and diffusion. The effective multiplication factor is given by the Six-Factor Formula: k_eff = k_inf * P_FNL * P_TNL = (epsilon * p * f * eta) * P_FNL * P_TNL, where P_FNL is the fast non-leakage probability and P_TNL is the thermal non-leakage probability.

Question 5Nuclear safety: neutron balance, heat removal, radiation and responsibility

A nuclear reactor operating in a subcritical state has an effective multiplication factor k1 = 0.9950. Control rods are withdrawn, increasing the multiplication factor to a supercritical value of k2 = 1.0020. What is the total reactivity change (Delta rho) inserted into the core?

  1. +7.00 pcm
  2. +70.2 pcm
  3. +7021 pcm
  4. +702.1 pcm (approximately +7.02 * 10^-3 Delta k/k) Correct answer.

Explanation

Reactivity is defined as rho = (k - 1) / k. Initial reactivity rho1 = (0.9950 - 1) / 0.9950 = -0.0050251. Final reactivity rho2 = (1.0020 - 1) / 1.0020 = +0.0019960. The reactivity change is Delta rho = rho2 - rho1 = 0.0019960 - (-0.0050251) = +0.0070211 Delta k/k. Since 1 pcm = 10^-5 Delta k/k, Delta rho = 0.0070211 * 10^5 = +702.1 pcm.

Question 6Nuclear safety: neutron balance, heat removal, radiation and responsibility

Why are delayed neutrons (缓发中子) essential for the safe and controllable operation of a commercial nuclear power reactor?

  1. Delayed neutrons lengthen the average neutron generation lifetime from ~10^-4 seconds to ~0.1 seconds, allowing mechanical control rod systems to govern reactor power safely. Correct answer.
  2. Delayed neutrons have much higher energy than prompt neutrons, which prevents fuel melting.
  3. Delayed neutrons eliminate the buildup of xenon-135 and samarium-149.
  4. Delayed neutrons cause the reactor to operate without needing any moderator.

Explanation

Prompt neutrons are born within ~10^-14 seconds of fission, which would lead to reactor periods of milliseconds if the core were prompt critical. Delayed neutrons, emitted following the radioactive beta-decay of precursor fission products (with half-lives up to 55 seconds), extend the effective mean generation lifetime to ~0.08-0.1 seconds, providing human operators and mechanical control systems ample time to manage reactivity safely.

Question 7Nuclear safety: neutron balance, heat removal, radiation and responsibility

A critical reactor experiencing a small positive step reactivity insertion rho = 0.0015 (where rho < beta_eff). If the effective delayed neutron fraction is beta_eff = 0.0065 and the one-group average precursor decay constant is lambda = 0.08 s^-1, what is the approximate stable asymptotic reactor period (T)?

  1. 4.17 seconds
  2. 41.7 seconds Correct answer.
  3. 417 seconds
  4. 0.42 seconds

Explanation

For small positive reactivity insertions well below prompt critical (rho < beta_eff), the stable reactor period is given by the asymptotic inhour approximation: T = (beta_eff - rho) / (lambda * rho). Substituting values: T = (0.0065 - 0.0015) / (0.08 * 0.0015) = 0.0050 / 0.00012 = 41.67 seconds.

Question 8Nuclear safety: neutron balance, heat removal, radiation and responsibility

What physical mechanism is responsible for the Fuel Doppler Temperature Coefficient (多普勒温度系数) in a low-enriched uranium oxide thermal reactor core?

  1. Thermal expansion of coolant water pushing fuel rods further apart.
  2. Decreased neutron absorption in boron control rods at higher temperatures.
  3. Doppler broadening of the resonance capture cross-section peaks of Uranium-238 due to increased thermal agitation of fuel nuclei, increasing neutron capture and introducing prompt negative reactivity. Correct answer.
  4. Electromagnetic repulsion between delayed neutrons and cladding atoms.

Explanation

As fuel temperature rises, the thermal agitation of Uranium-238 nuclei increases, broadening the width of its sharp resonance capture peaks (e.g., at 6.67 eV). This Doppler broadening reduces self-shielding, allowing more epithermal neutrons to be absorbed non-fissionably in U-238, which immediately reduces resonance escape probability (p) and provides an instantaneous, inherent negative reactivity feedback.

Question 9Nuclear safety: neutron balance, heat removal, radiation and responsibility

Why does a high concentration of soluble chemical shim (boric acid, H3BO3) in the primary coolant of a Pressurized Water Reactor (PWR) at Beginning of Cycle (BOC) make the Moderator Temperature Coefficient (MTC) less negative (or potentially positive if uncontrolled)?

  1. Boron atoms chemically react with zircaloy cladding to produce prompt neutrons.
  2. Boric acid freezes at elevated temperatures, blocking core cooling channels.
  3. Boric acid increases the fast fission factor of Uranium-235.
  4. As coolant heats and expands, water density decreases, removing dissolved boron (a strong neutron poison) from the core along with water, which adds positive reactivity that counteracts the negative effect of under-moderation. Correct answer.

Explanation

When coolant water expands with rising temperature, its density decreases. In a borated core, this expansion expels not only water (moderator) but also the dissolved boron (neutron absorber) from the active core volume. The loss of boron poison inserts positive reactivity, which offsets the negative reactivity caused by reduced neutron thermalization, making the overall MTC less negative. Safety limits strictly cap dissolved boron (often utilizing burnable poisons) to keep MTC non-positive at power.

Question 10Nuclear safety: neutron balance, heat removal, radiation and responsibility

Why do commercial Light Water Reactor (LWR) safety standards require the reactor core to be designed with a negative Moderator Void Coefficient (空泡反应性系数)?

  1. To ensure that any steam void formation reduces moderation and automatically drives reactor power down, preventing runaway power excursions. Correct answer.
  2. To ensure that boiling instantly increases reactivity to maintain turbine speed.
  3. To prevent radioactive tritium gas from dissolving into steam.
  4. To eliminate the need for primary coolant circulation pumps.

Explanation

A negative void coefficient of reactivity ensures that if steam bubbles (voids) form due to excessive power or localized overheating, the reduced moderator density decreases thermalization in the under-moderated core lattice, thereby inserting negative reactivity and inherently reducing reactor power without requiring operator intervention.

Question 11Nuclear safety: neutron balance, heat removal, radiation and responsibility

Following a sudden reactor shutdown from 100% steady-state power, what causes the Xenon-135 concentration to increase to a peak maximum (known as the 'Xenon Pit' or 'Iodine Pit') approximately 8 to 11 hours post-trip?

  1. Control rods begin emitting massive amounts of xenon gas upon full insertion.
  2. Iodine-135 (half-life 6.57 hours) continues to decay into Xenon-135, while the primary mechanism for Xenon-135 removal during operation-neutron absorption (burnout)-ceases due to the collapse of neutron flux. Correct answer.
  3. Xenon-135 is produced by cosmic ray activation of containment structural steel.
  4. Primary coolant cooldown condenses gaseous xenon into the fuel pellets.

Explanation

During high-power operation, a large inventory of Iodine-135 (T_1/2 = 6.57 h) accumulates. Upon shutdown, neutron flux drops to near-zero, halting the rapid destruction of Xenon-135 (sigma_a = 2.6 * 10^6 b) via neutron capture (Xe-135 + n -> Xe-136). Because I-135 continues to decay into Xe-135 faster than Xe-135 decays away (T_1/2 = 9.14 h), Xenon-135 concentration rises to a peak 8–11 hours post-trip before gradually decaying over several days.

Question 12Nuclear safety: neutron balance, heat removal, radiation and responsibility

How does the post-shutdown behavior of Samarium-149 (钐-149) poison differ fundamentally from that of Xenon-135?

  1. Samarium-149 decays away within 2 hours, whereas Xenon-135 remains forever.
  2. Samarium-149 adds positive reactivity at high temperatures.
  3. Samarium-149 is a stable (non-radioactive) isotope; thus, following shutdown, its concentration builds up from Promethium-149 decay and reaches a permanent, non-decaying plateau until the reactor is restarted. Correct answer.
  4. Samarium-149 only exists in fast breeder reactors.

Explanation

Unlike Xenon-135 (which is radioactive and decays with a 9.14-hour half-life), Samarium-149 is completely stable. After shutdown, accumulated Promethium-149 (T_1/2 = 53.1 h) beta-decays into Sm-149. Because Sm-149 does not undergo radioactive decay and neutron flux is absent to burn it out, its negative reactivity reaches a constant, non-decaying maximum that remains until restart.

Question 13Nuclear safety: neutron balance, heat removal, radiation and responsibility

In reactor physics, what is the geometric profile of the differential control rod worth (微分棒位价值) curve along the axial height of a homogeneous reactor core?

  1. A linear horizontal line of constant value from core bottom to top
  2. Maximum differential worth at the very top and bottom tips of the core
  3. An exponential curve that drops to zero at the core center
  4. A bell-shaped (cosine-squared) curve with maximum differential worth near the axial center of the core where neutron flux is highest Correct answer.

Explanation

Differential control rod worth (d_rho / dz) is proportional to the square of the unperturbed thermal neutron flux (phi^2(z)). In a cylindrical core with an axial cosine flux distribution, differential worth is lowest at the core top and bottom boundaries (where flux approaches zero) and peaks sharply at the axial mid-plane where neutron flux and importance are highest.

Question 14Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the physical definition and safety significance of the Nuclear Enthalpy Rise Hot Channel Factor (F_DeltaH)?

  1. The ratio of the maximum integrated fuel rod linear power (enthalpy rise along the hottest coolant channel) to the core average fuel rod enthalpy rise. Correct answer.
  2. The ratio of the total core thermal power to the generator electrical output.
  3. The ratio of secondary steam pressure to primary system operating pressure.
  4. The ratio of control rod insertion speed to gravity drop speed.

Explanation

F_DeltaH (nuclear enthalpy rise hot channel factor) is defined as the ratio of the maximum integrated rod power in the core (the maximum enthalpy rise along the limiting coolant flow channel) to the average rod power. It is a key 2D radial peaking metric used in thermal-hydraulic design to verify that Departure from Nucleate Boiling (DNB) limits are satisfied across the core.

Question 15Nuclear safety: neutron balance, heat removal, radiation and responsibility

In a Pressurized Water Reactor thermal-hydraulic safety analysis, the Critical Heat Flux predicted by the qualified CHF correlation at the hot spot is q''_CHF = 2.60 MW/m^2. If the actual calculated local operating heat flux is q''_actual = 1.85 MW/m^2, what is the Departure from Nucleate Boiling Ratio (DNBR), and does it meet a minimum design safety limit of 1.25?

  1. DNBR = 0.71; fails the safety limit
  2. DNBR = 1.41; satisfies the safety limit with margin Correct answer.
  3. DNBR = 4.45; satisfies the safety limit
  4. DNBR = 0.95; fails the safety limit

Explanation

The Departure from Nucleate Boiling Ratio is defined as DNBR = q''_CHF / q''_actual. Substituting the given values: DNBR = 2.60 / 1.85 = 1.4054 = 1.41. Because 1.41 >= 1.25, the core hot spot operates safely above the regulatory DNB design basis limit, ensuring that a vapor film will not blanket the cladding surface.

Question 16Nuclear safety: neutron balance, heat removal, radiation and responsibility

What physical phenomenon occurs at the Departure from Nucleate Boiling (DNB) point on a fuel rod cladding surface in a PWR?

  1. Liquid water suddenly freezes onto the fuel rod cladding.
  2. The fuel cladding instantly dissolves into the primary coolant.
  3. Individual nucleate steam bubbles coalesce into a continuous, insulating blanket vapor film (film boiling), causing a sharp drop in heat transfer coefficient and a rapid escalation in cladding surface temperature. Correct answer.
  4. Neutron absorption increases so fast that fission stops immediately.

Explanation

In subcooled and nucleate boiling, heat transfer is highly efficient as bubbles detach rapidly. When heat flux exceeds the Critical Heat Flux (CHF), bubbles coalesce into an insulating vapor film (Departure from Nucleate Boiling). Because steam has a thermal conductivity much lower than liquid water, the convective heat transfer coefficient plunges, causing a rapid, potentially damaging spike in cladding temperature.

Question 17Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the primary technical reason for imposing strict regulatory limits on the Linear Heat Generation Rate (LHGR, kW/m or W/cm) of fuel rods during normal operation and Anticipated Operational Occurrences (AOOs)?

  1. To prevent turbine generator blades from over-speeding.
  2. To prevent electrical short-circuits in excore neutron detector cables.
  3. To ensure that water in the secondary steam generator does not turn into steam.
  4. To keep the maximum centerline temperature of the UO2 ceramic fuel pellet below its melting point (~2800 °C) and prevent excessive fission gas release and cladding stress. Correct answer.

Explanation

Imposing limits on Linear Heat Generation Rate (LHGR) ensures that peak centerline fuel temperatures remain below the melting point of uranium dioxide fuel pellets (~2800 °C for unburned UO2, decreasing slightly with burnup). Preventing fuel melting prevents volumetric expansion, severe fission gas release, and excessive mechanical pellet-cladding strain.

Question 18Nuclear safety: neutron balance, heat removal, radiation and responsibility

A PWR reactor core produces 2900 MWth of total thermal power across 157 fuel assemblies. Each assembly contains 264 active fuel rods of active fuel length 3.66 m. If the overall nuclear heat flux hot channel factor is F_Q = 2.25, what is the peak linear power density at the core hot spot in kW/m?

  1. 43.0 kW/m (approximately 430 W/cm) Correct answer.
  2. 19.1 kW/m
  3. 32.5 kW/m
  4. 86.0 kW/m

Explanation

Total number of fuel rods in core = 157 * 264 = 41,448 rods. Total active fuel length = 41,448 * 3.66 m = 151,700 m. Core average linear heat generation rate = 2,900,000 kW / 151,700 m = 19.12 kW/m. The peak linear heat generation rate is obtained by multiplying the core average by the hot channel factor F_Q: q'_peak = 19.12 * 2.25 = 43.02 kW/m (equivalent to 430 W/cm).

Question 19Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the chemical reaction mechanism and danger of the Zirconium-Water Reaction (锆水反应) during a high-temperature LOCA transient?

  1. Zr + O2 -> ZrO2 (endothermic), which absorbs all available oxygen from containment air.
  2. Zirconium dissolves into pure water to form strong nitric acid that corrodes safety injection pipes.
  3. Zr + 2H2O -> ZrO2 + 2H2 + Heat (exothermic), which generates large quantities of explosive hydrogen gas and releases extra chemical heat that exacerbates core overheating. Correct answer.
  4. Zirconium absorbs water molecules into its crystal lattice without releasing any gases or heat.

Explanation

At elevated temperatures (> 1000 °C), zirconium alloy cladding reacts exothermically with steam: Zr + 2H2O -> ZrO2 + 2H2 + Heat. This reaction produces brittle zirconium dioxide, releases explosive hydrogen gas into containment, and adds substantial exothermic chemical energy that can trigger an autocatalytic thermal runaway.

Question 20Nuclear safety: neutron balance, heat removal, radiation and responsibility

What metallurgical degradation mechanism occurs in the steel wall of a Reactor Pressure Vessel (RPV) due to long-term exposure to fast neutron fluence (E > 1.0 MeV)?

  1. Complete liquefaction of low-alloy steel into molten iron at room temperature.
  2. Spontaneous radioactive decay of all iron atoms into gaseous argon.
  3. Severe volumetric expansion causing the vessel wall to double in thickness.
  4. Irradiation embrittlement (辐射脆化), characterized by an upward shift in the Ductile-to-Brittle Transition Temperature (DBTT / RT_NDT) and a decrease in Charpy Upper Shelf Energy (USE). Correct answer.

Explanation

Fast neutron bombardment (E > 1.0 MeV) creates displacement cascades (point defects and dislocation loops) and promotes copper/nickel-rich nanocluster precipitation in low-alloy RPV steel (e.g., SA-508 / 16MND5). This irradiation embrittlement shifts the Ductile-to-Brittle Transition Temperature (RT_NDT) to higher operating temperatures and lowers the Charpy Upper Shelf Energy (USE), reducing fracture toughness.

Question 21Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is Pressurized Thermal Shock (PTS, 承压热冲击) in a Pressurized Water Reactor, and why does it pose a severe risk to RPV structural integrity?

  1. A transient event combining rapid cooldown of the RPV inner wall (thermal shock from cold ECCS water injection) with high or repressurized primary system pressure, creating high tensile stresses in the neutron-embrittled vessel wall that could propagate existing flaws. Correct answer.
  2. An electrical transient where high-voltage switchgear discharges lightning directly into the pressurizer.
  3. A sudden drop in steam generator secondary pressure caused by closing the main steam isolation valves.
  4. A mechanical vibration caused by coolant flow resonance in the core barrel.

Explanation

Pressurized Thermal Shock (PTS) occurs during severe overcooling transients (such as small-break LOCA or main steam line break) where cold emergency core cooling water contacts the hot inner RPV wall while the system remains at high pressure or repressurizes. The steep thermal gradient induces high tensile thermal stresses at the inner wall, which, combined with internal pressure and irradiation-embrittled steel, risks unstable brittle crack propagation through the vessel wall.

Question 22Nuclear safety: neutron balance, heat removal, radiation and responsibility

In boiling two-phase flow systems, what characterizes Density Wave Oscillations (DWO, 密度波振荡)?

  1. Static cavitation in centrifugal pump impellers caused by low suction head.
  2. A dynamic, low-frequency two-phase flow instability caused by delayed feedback between inlet flow perturbations, boiling boundary displacement, two-phase density changes, and channel pressure drops. Correct answer.
  3. A chemical corrosion wave moving through secondary water treatment resins.
  4. A purely ultrasonic acoustic wave generated by safety relief valves opening.

Explanation

Density Wave Oscillations (DWO) are the most common dynamic two-phase flow instability in boiling channels (e.g., BWR cores and steam generators). An inlet velocity perturbation alters the boiling boundary and two-phase mixture density; the delayed propagation of this density wave creates a phase lag in two-phase pressure drop that can feed back into the inlet flow, sustaining self-excited flow oscillations.

Question 23Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is Axial Flux Offset (AO or Delta I) in a PWR core, and why must it be controlled within narrow target bands during power operation?

  1. The ratio of turbine electrical megawatts to generator reactive volt-amperes.
  2. The physical tilt of the reactor building foundation relative to true vertical.
  3. The normalized difference between power generated in the top half of the core and bottom half of the core (Delta I = (P_top - P_bottom) / P_total), controlled to prevent axial xenon oscillations and localized fuel peaking. Correct answer.
  4. The phase lag between electrical current and voltage in primary coolant pump motors.

Explanation

Axial Offset (AO or Delta I) measures the axial core power distribution imbalance: Delta I = (P_top - P_bottom) / P_total * 100%. Maintaining Delta I within strict technical specification limits (using control bank maneuvering and boration/dilution) prevents destabilizing axial xenon oscillations, limits peak linear heat rates, and ensures adequate shutdown margin.

Question 24Nuclear safety: neutron balance, heat removal, radiation and responsibility

A 3000 MWth PWR operates at full power for one full year before scramming. According to the Way-Wigner decay heat approximation (P(t)/P0 = 0.066 * t^-0.2 for long operation), what is the estimated core decay heat power remaining at t = 10^4 seconds (~2.78 hours) post-trip?

  1. 3.14 MWth (0.10% of nominal power)
  2. 300 MWth (10.0% of nominal power)
  3. 0.31 MWth (0.01% of nominal power)
  4. 31.4 MWth (approximately 1.05% of nominal power) Correct answer.

Explanation

Using the Way-Wigner decay heat formula: P(t)/P0 = 0.066 * (10^4)^-0.2. Since (10^4)^-0.2 = (10)^-0.8 = 0.15849, the fractional decay power is P(10^4)/P0 = 0.066 * 0.15849 = 0.01046 (1.046%). For a 3000 MWth reactor, the decay heat generation is 3000 MWth * 0.01046 = 31.4 MWth.

Question 25Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is Pellet-Cladding Interaction (PCI, 弹丸-包壳相互作用) stress corrosion cracking in nuclear fuel rods?

  1. A failure mechanism where rapid power ramps cause thermal expansion of UO2 pellets against the inner cladding wall, which, in the presence of corrosive volatile fission products (principally iodine), induces stress corrosion crack initiation and propagation through the zircaloy cladding. Correct answer.
  2. Corrosion of external control rod drive housings by ambient humid air.
  3. Chemical dissolution of ceramic pellets by boric acid inside the fuel matrix.
  4. Erosion of the outer cladding surface by high-velocity primary coolant flow.

Explanation

Pellet-Cladding Interaction (PCI) is a stress corrosion cracking (SCC) phenomenon. During rapid local power increases, the fuel pellet expands faster than the cladding, exerting high localized tensile hoop stresses on the inner cladding surface. In the presence of chemically aggressive volatile fission products (especially iodine and cadmium), microscopic stress corrosion cracks initiate and propagate, risking cladding breach.

Question 26Nuclear safety: neutron balance, heat removal, radiation and responsibility

In thermal reactor physics, what is the most probable energy (E_mp) of thermalized neutrons in thermal equilibrium with a moderator at room temperature (T = 293.15 K / 20 °C)?

  1. 1.0 MeV
  2. 0.0253 eV (corresponding to a most probable velocity of 2200 m/s) Correct answer.
  3. 2.0 MeV
  4. 100 keV

Explanation

In thermal equilibrium at room temperature (T0 = 293.15 K), neutrons follow a Maxwell-Boltzmann energy distribution. The most probable energy is E_mp = k_B * T0 = (8.617 * 10^-5 eV/K) * 293.15 K = 0.02526 eV = 0.0253 eV, corresponding to the standard thermal neutron velocity of v0 = 2200 m/s.

Question 27Nuclear safety: neutron balance, heat removal, radiation and responsibility

According to HAF 102 and the IAEA Defense-in-Depth framework, what is the specific objective of Level 3 of Defense-in-Depth (纵深防御第三道防线)?

  1. Prevention of abnormal operation and failures through conservative design and quality assurance.
  2. Control of abnormal operation and detection of failures to prevent anticipated operational occurrences from escalating into accident conditions.
  3. Control of accidents within the design basis (控制设计基准事故), ensuring core cooling and containment integrity via Engineered Safety Features (ESFs). Correct answer.
  4. Mitigation of off-site radiological consequences through public evacuation and food ingestion controls.

Explanation

Level 3 of Defense-in-Depth focuses on controlling Design Basis Accidents (DBAs). Even if postulated initiating events occur despite Levels 1 and 2, engineered safety features (ESFs such as ECCS, containment spray, and auxiliary feedwater) actuate to bring the plant to a safe shutdown state, maintain core cooling, and keep radioactive releases within regulatory limits.

Question 28Nuclear safety: neutron balance, heat removal, radiation and responsibility

What are the Three Physical Barriers (三道物理屏障) engineered into a commercial Pressurized Water Reactor to prevent the release of radioactive fission products to the environment?

  1. Control rods, steam generator tubes, and the turbine condenser
  2. Demineralizer bed, feed pump, and cooling tower
  3. Biological shield, switchyard transformer, and emergency diesel generator
  4. Fuel pellet matrix & cladding (燃料包壳), Reactor Coolant Pressure Boundary (一回路压力边界), and Containment Structure (安全壳) Correct answer.

Explanation

In commercial PWRs, the three physical barriers are: (1) First Barrier: ceramic fuel matrix and zircaloy cladding containing fission products at the source; (2) Second Barrier: the Reactor Coolant Pressure Boundary (RCPB, including the reactor vessel, piping, and pressurizer); (3) Third Barrier: the prestressed concrete containment building with a leak-tight steel liner.

Question 29Nuclear safety: neutron balance, heat removal, radiation and responsibility

How is the Single Failure Criterion (单一故障准则) defined in nuclear safety design under HAF 102?

  1. A safety system must be able to perform its required safety function despite the occurrence of any single random failure in any component of the system, in combination with the initiating event. Correct answer.
  2. A nuclear power plant is only allowed to experience one operational trip per calendar year.
  3. No single operator is permitted to enter the main control room alone.
  4. All redundant safety trains must be connected to a single common electric breaker.

Explanation

Under HAF 102, the Single Failure Criterion (SFC) requires that a safety group/system must be capable of fulfilling its safety mission in the event of any single random failure occurring in any component, alongside the postulated initiating event and consequential failures, typically achieved via N+1 or N+2 redundant safety trains.

Question 30Nuclear safety: neutron balance, heat removal, radiation and responsibility

Which engineering design principle is specifically implemented to prevent Common Cause Failures (CCF, 共同原因故障) that could simultaneously disable all redundant trains of a safety system?

  1. Using identical software, components, and manufacturers for all trains to maximize uniformity
  2. Diversity (多样性) in operating principles/manufacturers, and physical separation (物理隔离) using spatial distance and fire/missile barrier walls Correct answer.
  3. Eliminating all backup battery systems
  4. Connecting all safety cables in a single shared conduit

Explanation

Common Cause Failures (CCF) occur when multiple redundant components fail due to a shared cause (e.g., common manufacturing defect, maintenance error, fire, flood, or software bug). Nuclear safety mitigates CCF through diversity (different physical operating principles, diverse equipment suppliers, diverse software logic) and physical separation (fire barriers, flood barriers, and independent cable routing).

Question 31Nuclear safety: neutron balance, heat removal, radiation and responsibility

A safety injection system requires water flow to the reactor vessel. The system fails if both pump A and pump B fail (an AND gate), OR if the common discharge check valve C fails closed (an OR gate). The independent failure probabilities are P(A) = 0.01, P(B) = 0.01, and P(C) = 0.002. What is the total system failure probability using the rare-event approximation?

  1. P_sys = 0.00010 (1.0 * 10^-4)
  2. P_sys = 0.02200 (2.2 * 10^-2)
  3. P_sys = 0.00210 (2.1 * 10^-3) Correct answer.
  4. P_sys = 0.00002 (2.0 * 10^-5)

Explanation

In Fault Tree Analysis (FTA), the minimal cut sets are {A, B} and {C}. The failure probability of the parallel pump branch (AND gate) is P(A) * P(B) = 0.01 * 0.01 = 0.0001. This branch is in series with valve C (OR gate). Using the rare-event sum approximation: P_sys = P(A and B) + P(C) = 0.0001 + 0.002 = 0.0021 (2.1 * 10^-3).

Question 32Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the operational function and staging sequence of the Emergency Core Cooling System (ECCS) during a large-break Loss-of-Coolant Accident (LBLOCA) in a standard PWR?

  1. Low-Pressure Injection only after 48 hours with zero water delivery during the first hour.
  2. Draining all water out of containment to prevent electrical shorts.
  3. Injecting compressed steam directly into the fuel rods.
  4. High-Pressure Safety Injection (HPSI) and passive Accumulator discharge for rapid core reflood during initial depressurization, followed by Low-Pressure Safety Injection (LPSI) and eventual switchover to containment sump recirculation for long-term core cooling. Correct answer.

Explanation

During a large-break LOCA, the RCS depressurizes rapidly. Passive nitrogen-pressurized accumulators inject large volumes of borated water once RCS pressure drops below ~4.5 MPa to rapidly reflood the core, while active HPSI and LPSI pumps deliver continuous borated water. Once the Refueling Water Storage Tank (RWST) empties, LPSI aligns to the containment sump to provide long-term decay heat recirculation cooling.

Question 33Nuclear safety: neutron balance, heat removal, radiation and responsibility

How do passive ECCS Accumulators (安全注射安注箱 / 蓄能器) operate automatically without requiring AC electrical power or active instrumentation trip signals?

  1. They are pressurized with nitrogen gas (~4.0-5.5 MPa); when RCS pressure drops below accumulator pressure during a LOCA, the pressure difference forces open mechanical check valves, passively discharging borated water into the cold legs. Correct answer.
  2. They rely on manual valve operation by control room operators using handwheels.
  3. They use diesel-driven booster pumps that start on high radiation signals.
  4. They boil water using electrical heaters to force steam into the core.

Explanation

Accumulators are passive pressure vessels filled with borated water under a high-pressure nitrogen cover gas (~4.0-5.5 MPa). During normal operation at 15.5 MPa, two series check valves are held tightly closed by primary system pressure. When a break causes RCS pressure to drop below accumulator pressure, the differential pressure immediately pushes the check valves open, discharging large volumes of water passively.

Question 34Nuclear safety: neutron balance, heat removal, radiation and responsibility

What are the dual primary safety functions of the Containment Spray System (CSS, 安全壳喷淋系统) during a high-energy pipe break inside containment?

  1. Pumping out seawater to flood the turbine building and cooling main transformer oil.
  2. Rapid steam condensation to reduce containment pressure/temperature, and chemical scrubbing (using alkaline additives like NaOH or Na3PO4) to wash out airborne radioactive fission products (principally elemental iodine). Correct answer.
  3. Heating the containment air to dry out electrical instrumentation.
  4. Extinguishing external forest fires outside the plant perimeter fence.

Explanation

The Containment Spray System serves two critical safety functions: (1) Thermal-hydraulic pressure suppression: cold water droplets condense steam, reducing containment pressure and temperature below design limits; (2) Radiological fission product scrubbing: chemical additives (such as sodium hydroxide or trisodium phosphate) maintain an alkaline pH (~7.5-9.0) that absorbs airborne radioiodine gas and fixes it into non-volatile ionic forms in the sump water.

Question 35Nuclear safety: neutron balance, heat removal, radiation and responsibility

In PWR containment isolation logic, what is the operational distinction between Phase A Isolation and Phase B Isolation?

  1. Phase A opens all valves, while Phase B closes all doors.
  2. Phase A applies only during refueling outages, while Phase B applies during civil construction.
  3. Phase A isolation (actuated on Safety Injection) closes all non-essential process lines penetrating containment, while Phase B isolation (actuated on High-High Containment Pressure) closes all remaining process penetrations including cooling water lines to RCPs, while leaving only safety injection and containment spray operational. Correct answer.
  4. Phase A is operated manually by local handwheels, while Phase B operates automatically.

Explanation

Phase A containment isolation is triggered automatically by a Safety Injection (SI) signal, closing non-essential process lines (e.g. sampling, letdown, drains) to establish baseline containment boundary integrity. Phase B isolation is triggered by High-High containment pressure (indicating a severe high-energy line break), closing remaining penetrations (such as component cooling water lines to reactor coolant pumps) while maintaining essential ECCS and containment spray lines.

Question 36Nuclear safety: neutron balance, heat removal, radiation and responsibility

How does China's indigenous Generation III+ Hualong One (华龙一号, HPR1000) safety design differ in architecture from the Westinghouse AP1000 design?

  1. Hualong One has zero active safety pumps and relies 100% on manual bucket water refilling.
  2. AP1000 uses heavy water as moderator, while Hualong One uses molten sodium.
  3. Hualong One does not utilize a containment building.
  4. Hualong One employs a comprehensive 'Active plus Passive' (能动与非能动相结合) safety philosophy with 3 independent active electrical safety trains combined with passive secondary/containment cooling, whereas AP1000 relies primarily on fully passive safety systems (PRHR, CMT, IRWST, PCS). Correct answer.

Explanation

Hualong One (HPR1000) features an 'Active + Passive' safety philosophy: 3 fully independent active safety trains (3 * 100% redundancy for ECCS, EFWS, and emergency diesels) augmented by passive systems (Passive Secondary Residual Heat Removal PRS, Passive Containment Heat Removal PCS, and In-Vessel Retention IVR). AP1000 is an evolutionary passive design that replaces active safety injection pumps with gravity-driven CMTs, PRHR, and natural circulation containment cooling.

Question 37Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the driving physical mechanism for the Passive Residual Heat Removal Heat Exchanger (PRHR HX) in an AP1000 reactor during a loss of normal AC power?

  1. Natural thermal circulation (自然循环) driven by the fluid density difference between the hot primary coolant entering the top of the PRHR C-tube heat exchanger and the colder, denser coolant exiting the bottom and returning to the cold leg, discharging decay heat to the In-containment Refueling Water Storage Tank (IRWST). Correct answer.
  2. High-pressure compressed air pistons driving mechanical water pumps.
  3. Emergency diesel generators turning magnetic coupling turbines.
  4. Electric immersion heaters boiling water inside the core.

Explanation

The AP1000 PRHR heat exchanger is immersed in the large, elevated In-containment Refueling Water Storage Tank (IRWST). When actuation valves fail open, natural circulation starts automatically: hot coolant from the reactor hot leg rises by buoyancy into the PRHR tubes, cools by transferring heat to the surrounding IRWST water pool, increases in density, and flows by gravity back into the reactor cold leg.

Question 38Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the severe accident management objective of In-Vessel Retention (IVR, 堆内熔融物滞留) combined with External Reactor Vessel Cooling (ERVC) in Generation III+ reactors like Hualong One and AP1000?

  1. To accelerate the melting and breach of the reactor pressure vessel lower head to vent steam into the basement.
  2. To submerge and flood the reactor cavity with water, cooling the exterior lower head of the RPV to ensure the molten core debris (corium) is safely retained inside the vessel, preventing vessel breach and ex-vessel corium-concrete interactions. Correct answer.
  3. To pump molten corium directly into the secondary steam generator tubes.
  4. To generate large amounts of hydrogen to burn off oxygen in the containment.

Explanation

In-Vessel Retention via External Reactor Vessel Cooling (IVR-ERVC) is a key severe accident mitigation strategy. By flooding the reactor cavity with water from the IRWST or passive water pools, the outer surface of the lower RPV head is submerged. Nucleate boiling on the vessel exterior extracts sufficient heat to keep the outer steel wall solid, retaining the molten corium inside the RPV and preventing vessel failure and molten core-concrete interactions (MCCI).

Question 39Nuclear safety: neutron balance, heat removal, radiation and responsibility

How do Passive Autocatalytic Recombiners (PARs, 非能动自催化复合器) function inside containment to mitigate hydrogen explosion hazards during a severe accident?

  1. They ignite high-voltage electrical spark plugs to trigger controlled deflagrations.
  2. They pump hydrogen gas through compressors into outdoor storage tanks.
  3. They utilize platinum or palladium catalysts on internal metal plates that spontaneously catalyze the exothermic recombination reaction of hydrogen with atmospheric oxygen (2H2 + O2 -> 2H2O) without external power, setting up natural convective airflow that depletes hydrogen concentration below flammability limits. Correct answer.
  4. They absorb hydrogen by dissolving containment air into sulfuric acid.

Explanation

PARs are passive safety devices containing noble metal (platinum/palladium) catalysts. When hydrogen gas enters the open bottom of the PAR housing, it catalytically reacts with oxygen at room temperature or above without requiring electrical power: 2H2 + O2 -> 2H2O + Heat. The reaction heat warms the gas, generating a natural chimney convection flow that continuously draws hydrogen in and depletes it below detonable thresholds (< 4-10% by volume).

Question 40Nuclear safety: neutron balance, heat removal, radiation and responsibility

In nuclear safety systems design, which of the following examples best illustrates the Fail-Safe Principle (故障安全原则)?

  1. A valve that locks in its current position when electrical power fails.
  2. An alarm system that turns completely silent whenever a sensor cable breaks.
  3. A pump that speeds up to maximum power when its cooling water fails.
  4. Control rod drive mechanisms designed such that loss of electrical power de-energizes the holding coils, causing control rods to drop into the core by gravity. Correct answer.

Explanation

The Fail-Safe Principle mandates that upon loss of electrical power, pneumatic pressure, or control signal, components automatically transition to the orientation that places the plant in the safest condition. Control rod holding coils de-energizing on power loss to allow gravity insertion (scram) is the classic nuclear engineering application of fail-safe design.

Question 41Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the primary objective of the Equipment Qualification (EQ, 设备鉴定) program for safety-class electrical and mechanical components in nuclear power plants?

  1. To verify and document through type testing, analysis, or operating experience that safety-related equipment will reliably perform its safety functions under harsh environmental conditions (high radiation, LOCA temperature/pressure transients, chemical spray, and seismic vibrations) throughout its qualified life. Correct answer.
  2. To ensure components are purchased at the lowest possible market auction price.
  3. To verify that equipment colors match architectural interior decoration guidelines.
  4. To guarantee that components never require maintenance for 100 years.

Explanation

Equipment Qualification (EQ) is a rigorous statutory process ensuring that safety-related electrical, instrumentation, and mechanical equipment is capable of performing its intended safety functions when exposed to the harsh environmental conditions (peak temperatures, high pressures, steam, chemical spray, cumulative radiation doses, and seismic ground motions) characteristic of postulated Design Basis Accidents.

Question 42Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the specific scope and risk metric evaluated in a Level 2 Probabilistic Safety Assessment (Level 2 PSA / 二级PSA)?

  1. Quantifying initiating event frequencies and core damage frequency (CDF).
  2. Evaluating economic loss to local real estate and insurance markets.
  3. Modeling physical phenomena inside containment during severe accidents, containment failure modes, and determining the Large Early Release Frequency (LERF) and radioactive source terms to the environment. Correct answer.
  4. Calculating the financial return on investment of safety equipment upgrades.

Explanation

Level 1 PSA assesses accident sequences up to core damage to calculate Core Damage Frequency (CDF). Level 2 PSA takes the core damage plant states from Level 1, evaluates severe accident progression and containment response (e.g., steam explosions, hydrogen detonations, direct containment heating, basemat melt-through, containment bypass), and calculates the Large Early Release Frequency (LERF) and source term magnitudes.

Question 43Nuclear safety: neutron balance, heat removal, radiation and responsibility

According to Chinese nuclear safety regulatory standards for newly constructed Generation III/III+ commercial nuclear power plants, what are the quantitative safety goals for Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)?

  1. CDF < 10^-2 /reactor-year and LERF < 10^-3 /reactor-year
  2. CDF = 0 (absolute zero risk required by mathematical law)
  3. CDF < 10^-1 /reactor-year and LERF < 10^-2 /reactor-year
  4. CDF < 10^-5 /reactor-year and LERF < 10^-6 /reactor-year (with practical elimination of large radioactive releases) Correct answer.

Explanation

In accordance with HAF 102 and NNSA technical requirements for Generation III/III+ nuclear power plants (such as Hualong One), the quantitative probabilistic safety goals establish that the Core Damage Frequency must be less than 10^-5 per reactor-year and the Large Early Release Frequency must be less than 10^-6 per reactor-year, practically eliminating the need for large-scale off-site emergency evacuation.

Question 44Nuclear safety: neutron balance, heat removal, radiation and responsibility

In the beta-factor Common Cause Failure (CCF) model used in nuclear PSA, what does the parameter beta represent?

  1. The fraction of the total component failure rate that is attributable to common cause events affecting multiple redundant components simultaneously (lambda_CCF = beta * lambda_total). Correct answer.
  2. The delayed neutron fraction emitted by Uranium-235.
  3. The beta radiation dose rate inside the reactor containment.
  4. The ratio of pump discharge pressure to suction pressure.

Explanation

In reliability engineering and PSA, the beta-factor model partitions a component's total failure rate (lambda) into independent failures (lambda_ind = (1-beta)*lambda) and common cause failures (lambda_CCF = beta * lambda), where beta represents the conditional probability that a component failure is shared concurrently by all components in the common cause component group.

Question 45Nuclear safety: neutron balance, heat removal, radiation and responsibility

Why is a Steam Generator Tube Rupture (SGTR, 蒸汽发生器传热管破裂) event categorized as a primary-to-secondary containment bypass accident risk in a PWR?

  1. Because steam generator tubes are located entirely outside the plant boundary fence.
  2. Because high-pressure radioactive primary coolant leaks across ruptured tubes into the lower-pressure secondary circuit, allowing fission products to bypass the containment barrier and escape directly to the atmosphere via secondary safety relief valves or condenser steam air ejectors. Correct answer.
  3. Because the steam generator causes the reactor core to instantly freeze.
  4. Because SGTR accidents only occur when the reactor building containment dome is removed.

Explanation

A Steam Generator Tube Rupture (SGTR) breaches both the reactor coolant pressure boundary (1st/2nd barrier interface) and containment isolation integrity. Because the secondary steam circuit extends outside containment to the turbine building and atmospheric relief valves (ADV/MSSV), primary coolant inventory and volatile fission products can discharge directly into the environment, bypassing the containment boundary.

Question 46Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is an Interfacing Systems LOCA (ISLOCA, 系统间失水事故) in a nuclear power plant?

  1. A leak occurring between two computers in the administrative office.
  2. A water leak inside the plant cafeteria dishwashing system.
  3. An accident where high-pressure primary coolant breaches isolation boundaries and overpressurizes a connected low-pressure auxiliary system (such as RHR or CVCS) located outside the containment building, causing auxiliary piping rupture and an unisolated primary coolant leak outside containment. Correct answer.
  4. An intentional transfer of primary coolant into secondary feedwater storage tanks during normal power operation.

Explanation

An Interfacing Systems LOCA (ISLOCA) occurs when isolation valves separating the high-pressure RCS (~15.5 MPa) from low-pressure piping systems (e.g., Residual Heat Removal / RHR designed for ~3-4 MPa) fail. The resulting overpressurization bursts the low-pressure piping outside containment, discharging primary coolant inventory directly into auxiliary buildings and bypassing containment.

Question 47Nuclear safety: neutron balance, heat removal, radiation and responsibility

What constitutes a Station Blackout (SBO, 全厂断电) event at a nuclear power plant?

  1. The complete and simultaneous loss of all external offsite AC electrical power concurrent with the total failure of all on-site emergency diesel generators, leaving only Class 1E DC station batteries and diverse alternate AC power sources available. Correct answer.
  2. A planned maintenance shutdown of plant administrative office lighting.
  3. A failure of the main generator that triggers a routine automatic transfer to the offsite grid.
  4. A localized trip of a single 400V lighting distribution panel in the warehouse.

Explanation

Station Blackout (SBO) is the simultaneous loss of all offsite AC power (grid loss) coupled with the failure of all main onsite emergency AC power sources (Emergency Diesel Generators). During an SBO, the plant must rely on vital Class 1E DC batteries, turbine-driven auxiliary feedwater pumps, passive cooling systems, and dedicated Alternate AC (AAC) diesels to prevent core uncovery throughout the coping duration.

Question 48Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is an Anticipated Transient Without Scram (ATWS, 未停堆的预计运行事件), and which engineering system provides diverse mitigation against it?

  1. A routine turbine trip where control rods insert normally within 2 seconds.
  2. An anticipated operational occurrence (such as loss of feedwater or turbine trip) where the Reactor Protection System (RPS) fails to insert control rods, mitigated by an independent Diverse Actuation System (DAS) that trips the turbine, initiates auxiliary feedwater, and activates diverse emergency boration. Correct answer.
  3. An emergency drill conducted without sounding sirens.
  4. A scheduled fuel reload performed while the reactor remains at 100% full power.

Explanation

An Anticipated Transient Without Scram (ATWS) occurs when an anticipated transient demands a reactor scram, but a common-cause failure in the primary Reactor Protection System (RPS) prevents control rod insertion. Nuclear safety regulations mandate an independent, diverse mitigation system (Diverse Actuation System / Diverse Protection System) that automatically trips the turbine, starts auxiliary feedwater, and initiates emergency boration.

Question 49Nuclear safety: neutron balance, heat removal, radiation and responsibility

What is the primary objective of a Containment Integrated Leak Rate Test (ILRT / Type A Test) conducted on a nuclear reactor containment building?

  1. To test whether the containment building can float on water in the event of a tsunami.
  2. To clean the external concrete surfaces using high-pressure chemical solvents.
  3. To verify that the total integrated leakage rate of the entire containment structure and all penetrations at the calculated peak accident pressure (P_a) does not exceed the statutory maximum allowable leakage rate (L_a). Correct answer.
  4. To test the acoustic reverberation time inside the dome for public speaking.

Explanation

An Integrated Leak Rate Test (ILRT / Type A test) measures the overall integrated leakage rate of the containment pressure boundary by pressurizing the sealed containment to its peak calculated design accident pressure (P_a, typically ~0.4-0.5 MPa) and verifying over a 24-hour test period that the leakage rate remains strictly within statutory limits (typically L_a <= 0.1-0.2% of containment air mass per 24 hours).

Question 50Nuclear safety: neutron balance, heat removal, radiation and responsibility

What are the Three Fundamental Principles of Radiation Protection established by the ICRP and enacted in China's Basic Standards for Protection against Ionizing Radiation (GB 18871-2002)?

  1. Confidentiality, Speed, and Profitability
  2. Maximum dilution, Open release, and Natural attenuation
  3. Chemical neutralization, Thermal combustion, and Magnetic containment
  4. Justification of Practices (正当性), Optimization of Protection / ALARA (防护的最优化), and Individual Dose Limitation (剂量约束与限值) Correct answer.

Explanation

GB 18871-2002 incorporates the three core international radiation protection principles: (1) Justification (实践的正当性: no practice involving radiation shall be adopted unless it produces sufficient net benefit); (2) Optimization / ALARA (最优化: doses must be kept As Low As Reasonably Achievable, economic and social factors taken into account); (3) Dose Limitation (剂量限值: individual doses must not exceed statutory regulatory limits).