PE Nuclear cheat sheet

A PWR primary loop consists of the reactor vessel, reactor coolant pumps, steam generators, and pressurizer; the pressurizer holds pressure with heaters (raise) and spray (lower) to preserve subcooled margin.

In a BWR, recirculation pumps change core flow to control power, and steam flows directly from the vessel to the turbine, making the turbine building a radiologically controlled area during operation.

PWR emergency core cooling has three tiers: high-pressure injection, nitrogen-pressurized accumulators that inject passively on low RCS pressure, and low-pressure injection that later recirculates from the containment sump.

Main steam isolation valves isolate the steam line on a safety injection or high radiation signal to prevent an uncontrolled cooldown that could add positive reactivity.

BWR RCIC and HPCI are steam-turbine-driven makeup systems; RCIC requires DC power for governor and control logic, so loss of DC defeats it.

Steam generator tube integrity is verified by eddy current inspection; primary-to-secondary leakage is bounded by technical specification limits (about 150 gallons per day or a lab-analysis action level).

Control rod drive mechanisms in a PWR are pressure-boundary housings; a rod ejection accident is analyzed as a design-basis reactivity insertion event.

Turbine bypass to the condenser lets reactor power be reduced without a trip; loss of condenser vacuum typically requires a reactor scram because the heat sink is gone.

Reactivity is rho = (k_eff - 1)/k_eff, expressed in percent delta-k/k, dollars (fraction of beta), or pcm (1 pcm = 1e-5).

The effective delayed neutron fraction for a U-235 thermal system is about 0.0065; a reactivity insertion of one dollar (rho = beta) produces prompt criticality.

The six-factor formula gives k_inf = epsilon * p * f * eta, where epsilon is fast fission, p resonance escape, f thermal utilization, and eta the reproduction factor.

LWR reactivity coefficients are negative for fuel temperature (Doppler, prompt) and generally negative for moderator temperature; a positive void coefficient would be a design concern.

Xenon-135 has a thermal absorption cross section near 2.6 million barns; after shutdown it peaks roughly 8-10 hours later because of iodine-135 precursor decay.

Samarium-149 is a stable fission-product poison with no post-shutdown peak; its equilibrium reactivity worth is roughly -0.6 percent delta-k/k.

Subcritical multiplication is tracked with a 1/M plot; the inverse count rate ratio extrapolates to zero at criticality, and each fuel addition step must be sized so the plot stays linear.

Soluble boron provides negative reactivity in PWRs; dilution during startup or refueling must be controlled and boron concentration independently verified.

Criticality safety outside the reactor uses the double contingency principle, with controls on mass, geometry, moderation, and absorption.

Decay heat is about 6.5 percent of full power one second after shutdown, about 1.5 percent at one minute, about 0.4 percent at one hour, and about 0.05 percent at one day.

DNBR is the critical heat flux divided by the local heat flux; PWR technical specifications require DNBR to stay above the design limit (commonly about 1.30) at all allowed power and flow conditions.

BWRs use critical power ratio (CPR) as the dryout-based analogue of DNBR, with a minimum CPR operating limit.

Nucleate boiling heat transfer coefficients are far above single-phase convection, so the clad-to-coolant temperature difference stays small until CHF is approached.

Reynolds number Re = rho * V * D / mu sets the flow regime; Re below about 2300 in a pipe is laminar, and turbulent single-phase coefficients follow Dittus-Boelter type correlations.

Two-phase flow regimes progress through bubbly, slug, churn, annular, and mist flow; the prevailing regime determines which CHF correlation (e.g., W-3, Biasi) applies.

Natural circulation driving head equals the density difference between the hot and cold legs times loop height; it requires a filled loop and an available heat sink.

Choked two-phase flow at a break limits blowdown rate; the Henry-Fauske model is a standard critical flow treatment for LOCA analysis.

Overall plant thermal efficiency for an LWR Rankine cycle is about 33 percent, set mainly by condenser pressure and steam conditions.

Absorbed dose is measured in gray (1 Gy = 1 J/kg); equivalent and effective dose are in sievert, with 1 Sv = 100 rem.

Photon attenuation follows I = I0 * exp(-mu * x); the half-value layer is ln(2)/mu and the tenth-value layer is ln(10)/mu.

Buildup factors correct the exponential attenuation formula for scattered photons and are required for thick or broad-beam shields.

Gamma shielding favors high density and high atomic number (lead, concrete); neutron shielding uses hydrogenous material to moderate followed by boron or cadmium to absorb.

Point-source dose rate falls off as the inverse square of distance, so distance is the cheapest dose-control tool.

US occupational limits in 10 CFR 20.1201 are 50 mSv/yr total effective dose equivalent, 150 mSv to the lens, and 500 mSv to extremities; the public limit is 1 mSv/yr.

Airborne intake is controlled using derived air concentration (DAC) and annual limit on intake (ALI); DAC-hours multiplied by breathing rate gives activity taken in.

10 CFR 20.1101 requires a written ALARA program that weighs economics, state of technology, and social factors against dose reduction.

The front end runs mining, milling to U3O8, conversion to UF6, enrichment, and fabrication of UO2 pellets into zirconium-clad fuel rods.

Separative work units quantify enrichment effort and depend on the feed, product, and tails assays rather than on the mass of uranium processed.

In once-through cycles about 95 percent of the uranium remains in spent fuel; enrichment above 20 percent U-235 defines HEU and is a proliferation concern.

The PUREX process uses tributyl phosphate solvent extraction to separate uranium and plutonium from fission products in irradiated fuel.

Low-level waste classes A, B, and C are regulated under 10 CFR 61; spent fuel and high-level waste repository requirements are in 10 CFR 63.

Spent fuel pools maintain subcriticality by dissolved boron and geometry, remove decay heat through heat exchangers, and provide roughly 10-20 feet of water shielding.

A zirconium cladding fire is the dominant spent fuel pool risk if fuel is uncovered; makeup capability and level instrumentation are key mitigations.

Dry cask storage is passive and air-cooled; fuel must cool in the pool for a minimum licensed period before transfer.

High-level waste immobilization by vitrification into borosilicate glass yields a leach-resistant waste form for geologic disposal.

10 CFR 50 Appendix A General Design Criteria set minimum requirements for plant structures, systems, and components, including the single failure criterion for safety systems.

10 CFR 50.46 LOCA acceptance criteria: peak clad temperature at or below 2200 F, local clad oxidation at or below 17 percent of thickness, hydrogen generation at or below 1 percent of theoretical, coolable geometry, and long-term cooling.

Defense in depth layers accident prevention, mitigation, and containment, and design-basis accidents are analyzed with conservative assumptions including worst single failure and loss of offsite power.

Emergency planning under 10 CFR 50.47 uses a 10-mile plume exposure pathway zone and a 50-mile ingestion pathway zone.

Technical specifications are enforceable license conditions derived from the safety analysis; operating outside a limiting condition for operation triggers the specified required action.

Event reporting: 10 CFR 50.72 requires immediate notification of specified events and 10 CFR 50.73 requires a written report within 30 days.

Post-Fukushima requirements added hardened containment vents, FLEX portable equipment, and seismic/flood re-evaluations beyond the original design basis.

Initial licenses are 40 years under 10 CFR 50, with renewal to 60 years under 10 CFR 54 and subsequent renewal to 80 years.

The IAEA International Nuclear and Radiological Event Scale ranks events from 1 (anomaly) to 7 (major accident).

A limiting condition for operation states the minimum equipment and performance needed for safe operation, along with required actions and completion times.

Refueling core alterations require a licensed senior reactor operator present, verified boron concentration, and compliance with the refueling criticality checklist before any fuel movement.

Inservice testing of pumps and valves follows the ASME OM Code; acceptance criteria include measurable flow, differential pressure, and stroke time.

Emergency diesel generator surveillance verifies start time (typically 10 seconds or less), load acceptance, and continuous rated-load operation for the specified duration.

Containment leak testing comprises Type A integrated rate tests, Type B penetration tests, and Type C isolation valve tests, all at specified pressure and acceptance leakage.

Startup physics testing verifies the critical boron endpoint, isothermal temperature coefficient, rod worth, and incore flux map power distribution.

The maintenance rule in 10 CFR 50.65 requires monitoring the performance of safety-related structures, systems, and components and correcting declining performance.

Pre-outage risk assessment and risk-informed configuration control limit how many redundant safety trains are out of service at once.

Post-maintenance testing demonstrates that a component can still perform its safety function after work is completed.

Austenitic stainless steels 304 and 316 are common for piping; chromium carbide precipitation in weld heat-affected zones causes sensitization and intergranular stress corrosion cracking.

Alloy 690 has largely replaced Alloy 600 for steam generator tubing because of improved resistance to primary water stress corrosion cracking.

Reactor pressure vessel embrittlement is tracked by the nil-ductility transition temperature shift from Charpy tests and evaluated under 10 CFR 50.61 and 50.61a pressurized thermal shock rules.

Zircaloy cladding absorbs little neutron flux but oxidizes exothermically in steam above roughly 1000 C, generating hydrogen and embrittling the clad.

Neutron detection uses proportional counters or fission chambers in the source range, gamma-compensated ion chambers in the intermediate range, and uncompensated ion chambers in the power range.

Reactor protection logic typically uses 2-out-of-4 coincidence and fails safe by de-energizing to trip.

Trip setpoints must account for calibration tolerance, instrument drift, and response time so the analytical safety limit is never exceeded.

Digital instrumentation upgrades must address diversity and defense in depth against common-cause software failure and meet 10 CFR 73.54 cyber security requirements.

Typical process instruments include resistance temperature detectors for coolant temperature, differential-pressure transmitters for flow and level, and strain-gauge transmitters for pressure.