Question: The process of applying for and obtaining a nuclear license involves thorough regulatory review.

Nuclear power plants must obtain a license from the Nuclear Regulatory Commission (NRC) before they can operate, ensuring that they comply with safety and environmental regulations.

The quarterly maintenance check is essential for ensuring that all systems in the facility are operating correctly.


Question: What is the purpose of the Component Cooling Water System in nuclear plants?

The Component Cooling Water System is designed to perform a pressure boundary function and is essential for maintaining system integrity in nuclear plants.

The strainer element serves to filter the closed-cycle cooling water, ensuring that any potentially harmful debris is caught before it can impact system performance.


Question: radiopharmaceutical extravasation monitoring

I agree with the previous comments by the nuclear medicine physician and technologist statements that there are very few extravasations and hence very little potential for untoward consequences from extravasation.

The need is to monitor all quality processes and ensure compliance with safety standards, which is crucial for maintaining operational integrity.


Question: voltage-based alternate repair criteria

The voltage-based alternate repair criteria are designed to address the integrity of steam generator tubing that is affected by outside diameter stress corrosion cracking (ODSCC).

U.S. NRC letter to Energy Harbor Nuclear Corp., "BEAVER VALLEY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 305 AND 195 TO MODIFY PRIMARY AND SECONDARY COOLANT ACTIVITY TECHNICAL SPECIFICATIONS (EPID L-2019-LLA-0223)," September 2020.


Question: What is the significance of energy peaks in radiation measurement?

Energy peaks in radiation measurement indicate the specific energies at which gamma rays are detected, with each peak corresponding to a particular radionuclide or interaction process.

The ambient background for the measurements was noted to be 4.18E+01 counts, which contributes to the overall data uncertainty.


Question: What are the aging management programs for closure bolting in nuclear facilities?

Closure bolting is used for mechanical closure and structural support, it is typically made of high-strength steel, stainless steel, or other types, and is evaluated for aging effects such as cracking due to Stress Corrosion Cracking (SCC) and loss of preload due to thermal effects, which is managed under the Bolting Integrity program.

Heat exchanger components are categorized under pressure boundaries and made of materials such as stainless steel and are subject to loss of material due to pitting and crevice corrosion, requiring a one-time inspection.


Question: What is the reporting limit for nuclear sample analysis?

The reporting limit (RL) for the analysis indicates the minimum concentration of a substance that can be reliably detected by the analytical method used in the test.

Counting Uncertainty is calculated at the 95% confidence level (1.96-sigma).


Question: The design of the casing cooling pumps ensures they are capable of handling significant flow rates, which is critical for maintaining safety.

The casing cooling pumps are seismically designed, QA Category I, single-stage, centrifugal pumps. The design capacity of the Unit 1 casing cooling pumps is 3000 gpm and the design capacity of the Unit 2 casing cooling pumps is 1000 gpm.

All spray headers have a combination of spray nozzle types that are oriented to obtain a wide distribution of varying size spray droplets.


Question: What measures are required to ensure a nuclear plant operates within acceptable limits during load following?

At the level of regulation, what we would essentially mandate is that one of three measures has to be in place to keep the plant from departing from acceptable operating regime during load following.


Question: reactor coolant system

The reactor coolant system is designed to remove heat from the reactor core and transfer it to a secondary system for power generation or to heat exchangers for cooling.

The emergency response team reviewed the evacuation plans and ensured that all personnel were familiar with the emergency exits and assembly points.


Question: NRC stakeholders believe the agency has shifted its expectations regarding the resolution of Key Technical Issues (KTIs) over time.

Many stakeholders believe that NRC is changing what it expects of DOE regarding the resolution of the KTIs.

Some stakeholders find that the agency has since learned to engage the public and keep them informed.


Question: The NRC has established guidelines for regulatory compliance related to safety inspections and audits.

The Agreement State Program allows the NRC to delegate some regulatory responsibilities to state governments, provided they meet certain safety standards.

The NRC allocates resources to improve its technical capabilities by investing in research and development initiatives.


Question: What are the consequences of varying the pressure within the VCT on the leakoff from the RCP seals?

A2.05: Effects of VCT pressure on RCP seal leakoff flows

The NRC emphasizes the importance of understanding reactor operations and their safety implications over time.


Question: What is the purpose of reactor building isolation valves?

This periodic verification is necessary to ensure that the reactor building penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur.

With two reactor building isolation valves in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 1 hour.


Question: What is GALILEO Fuel Rod Thermal-Mechanical Methodology?

In Reference 1 the NRC accepted the Topical Report ANP-10323P, Revision 1, "GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors" for referencing in license applications and requested that Framatome publish approved versions of the report.

There are no commitments contained within this letter or its enclosures.


Question: The NRC requires that any alternative to the Inservice Testing Program must include a detailed technical evaluation about the mechanical systems involved.

By letter dated March 15, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22074A315), as supplemented by a letter dated June 28, 2022 (ML22179A357), Tennessee Valley Authority (the licensee) submitted a request for an alternative to the Inservice Testing (IST) Program requirements for Sequoyah Nuclear Plant, Units 1 and 2, respectively.

The ERW attached drawing shows 8” and 10” swing check valves (page 3 of Attachment 1) for the Salem plant. Clarify that this drawing is also applicable for TVA 8” and 10” valves.


Question: nuclear fuel cycle

The nuclear fuel cycle encompasses several steps, including the mining of uranium, enrichment of uranium, use in nuclear reactors, and ultimately the management of spent fuel.

The safety of a nuclear reactor is significantly dependent on the design of its containment structures, which prevent the release of radioactive materials in the event of an accident.


Question: Degraded Voltage Time Delay Relay

For the Division III Degraded Voltage Time Delay Relay, the following relationship has to be maintained: NTSP₍ᴠ₀ʳ₊ᴅᴵₕ₁ₗₙₙₐₕ₁₌₍₍₌₦ₐ₌₌ₑ₌₌₌testₒ₉₊₩ = NTSP₍ᴠ₀ʳ₊ᴅᴵₕ₁ₗₙₙₐₕ₁ ₐ₊₈ₛ₊₇₏₊ₐ₉₊₏₊ₛₜₓₓ₏ₙ₌₌₌₌ ₦₌₌ₚ≡₍₁⁷ ₌ₓ₟₍₍₌ₜₑ≤

The As-Left and As-Found Tolerances for the Degraded Voltage Time Delay Functions were not calculated in input Calculations IP-E-0035 (Input 4.5.5) and IP-E-0036 (Input 4.5.6), and therefore, will be calculated here. The time delay functions' tolerances will.


Question: What are the criteria for SI termination in nuclear reactor operations?

Checks if SI Termination Criteria is MET: • RCS Subcooling >40°F – NO • Secondary Heat Sink (SG >13% or AFW Flow >240x103) • RCS Pressure rising or stable – NO Lowering • Pressurizer Level >20%

Maintains SG narrow range levels 27% - 50%. Directs chemistry to initiate post-accident sampling.


Question: Licensees must submit detailed reports that include safety analysis and system design for review.

The safety analysis required for reactor licensing applications must demonstrate conformance with specified regulatory criteria and standards.

Regular maintenance and inspections are critical for the long-term operation of the facility to ensure reliability and service continuity.


Question: The assessment of surface dose rates must include multiple target points on the cask.

Measure dose rates at eight target points (0, 45, 90, 135, 180, 225, 270, and 315 degrees) on each plane, at center of each inlet and outlet, and at a point in between each inlet and outlet.

Each cask features a significant amount of concrete that provides structural integrity and helps in radiation shielding.


Question: defueling tools in nuclear safety

Experience revealed that tool inventory and control were essential to efficient operations. Spare parts should be provided for all tools that risked breakdown during normal defueling activities and that could cause a significant delay in defueling activities.

Various fluids were tested for use in the defueling hydraulic system. After initiating defueling, a problem was encountered with the


Question: What is the purpose of SSC categorization in nuclear power plants?

The categorization of Structures, Systems, and Components (SSCs) is essential for ensuring that safety-related features in nuclear power plants are identified and managed according to their significance to safety, as outlined in the Regulatory Guide 1.201.

The NRC also assesses the environmental impacts of proposed actions as part of the licensing process in order to ensure compliance with the National Environmental Policy Act.


Question: Liquid Process Radiation Monitor

**DESCRIPTION:** Liquid Process Radiation Monitor Channel/Recorder Check

**REQUIREMENTS:** ODCM SR 2.1.2.c.2, Table 2.1-2


Question: The critical heat flux correlation is validated using specific test data described in the documentation.

The HTP CHF test data is described in Table B-2. This is used exclusively for validation of the correlation with no contribution to the defining data set (see Table 6-4).

Table B-2 includes various safety parameters that are critical for ensuring reactor integrity during operation.


Question: aging management evaluation of nuclear components

Aging Management Program: Structures Monitoring Program (B.2.3.33)

Liner Spent Fuel Pool: Structural Pressure Boundary | Stainless Steel | Submerged (Structural) (Ext) | Cracking Loss of material


Question: The Turkey Point Units are subject to a variety of federal, state, and local regulations that oversee their environmental impacts and operational procedures.

The comprehensive regulatory controls and permits in place and PTN’s compliance with them, guided by their internal procedures, would mitigate impacts to aquatic resources from PTN’s continued operations during the SLR operating term.

The Turkey Point barge basin was developed in association with the original Units 1 and 2 and is used for transport of material and large components to the PTN site.


Question: What tools are available for stakeholders to interact with regulatory information?

Empower stakeholders with data analytical tools, including a secure, flexible, and personalized self-service gateway to retrieve, submit, and/or interact with regulatory related information and assessment tools through a standard web browser.

The reactor core is designed to sustain a controlled nuclear reaction while ensuring the containment of radioactive materials.


Question: The NRC mandates that all license applicants submit comprehensive data regarding their safety and operational protocols.

The NRC requires that all licensees submit various reports related to safety analysis for their nuclear facilities.

The comprehensive review of reactor designs involves evaluating their thermal efficiency and power output at varying operational levels.


Question: The permittee must ensure that all necessary reports, including the Discharge Monitoring Reports, are signed by the responsible official or duly authorized representative.

All reports and forms required to be submitted by this permit, the AWPCA and the Department's Rules, shall be electronically signed (or, if allowed by the Department, traditionally signed) by a “responsible official” of the permittee as defined in ADEM Administrative Code Rule 335-6-5-.14 or a “duly authorized representative” of such official as defined in ADEM Administrative Code Rule 335-6-5-.14 and shall bear the following certification:

If the E2 Reporting System is down on the 28th day of the month in which the DMR is due or is down for an extended period of time, as determined by the Department, when a DMR is required to be submitted, the permittee may submit the data in an alternate manner and format acceptable to the Department.


Question: demineralizer system operation

The demineralizers are capable of operation in continuous or intermittent service. The system is operated to maintain the water quality requirements of the nuclear steam supply system supplier. In addition, the ion concentrations, particularly sodium and chloride, are maintained within the effluent requirements given in paragraph 10.4.6.2.4.4.

Table 10.4.6-2 lists the concentrations of corrosion products and other impurities in the influent condensate used as a design basis for two operating conditions, startup or design maximum and expected operation.


Question: What are the Docket Numbers relevant to a nuclear reactor licensing application?

If you have any questions, please contact me by telephone at 301-415-2048 or by e-mail to Justin.Poole@nrc.gov.


Question: What are the implications of allowing Very Low-Level Waste to be disposed of without a nuclear license?

The proposed "reinterpretation" of NRC regulations is a clear effort to let nuclear power owners to shift legal and financial responsibility to the public.

Sending it to hazardous or industrial sites is even worse because the radioactivity can react with other chemicals, making leaks to air and water exponentially more harmful than they are already.


Question: What are the requirements for providing medical evaluations to substantiate an applicant's fitness for operation?

“Supporting medical evidence” consists of the findings, laboratory data, examination results, diagnoses and treatment plans (such as prescribed medications, use of therapeutic devices and planned monitoring) that support a determination of whether or not an individual meets the physical condition and general health requirements to be licensed as an operator.

Commonly reported conditions involving medications include diabetes and thyroid disease.


Question: The NRC stipulates that two automatic actuation logic channels for engineered safety features must be operational to ensure safety.

Functional Units 1.e, 2.d, 3.d, 4.d, 5.e, 6.d, and 8.e – Two ESF Automatic Actuation Logic channels shall be operable

The PRA model only includes credit for a single under voltage relay per vital bus.


Question: What are the population siting criteria for advanced nuclear plants?

They did screen out those that didn't meet current population siting criteria.

This just elaborates again on that report that Christine previewed.


Question: What is the function of the Control Rod Drive System (CRDS) in a nuclear reactor?

Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following: Boration and dilution.

Since rod control speed can vary from 8 SPM to 72 SPM with Rod Control in automatic, all distractors are plausible.


Question: start-up physics tests in nuclear reactors

Hot Zero Power Tests include reactivity computer checkout and measurements of boron endpoints at various control bank configurations such as ARO and D banks.

Power Ascension Tests consist of several measures including a 30% power flux map and pseudo-dropped-rod tests at 50% power.


Question: Class IV VRM area definition

The Crownpoint project area is within a Class IV VRM area.

The site is approximately 5 km [3 mi] east of the border of the Navajo Nation.


Question: What is the operation of the Containment Spray system during E-1.3?

The Containment Spray pumps do not trip on RWST level; they are stopped by the operator at 4% RWST level. The system is aligned to RHR for spray purposes.

The crew will align Containment Spray from a train of RHR.


Question: What is the process for submitting license amendment requests to the NRC?

January 21, 2021 – Pre-submittal teleconference with NRC


Question: financial surety in nuclear licensing

The HMC 2023 surety was last adjusted on February 12, 2024, in the amount of $99,453,303.

This was based on a 3.6 percent increase in the consumer price index for all urban consumers (CPI-U) from the end of March 2023 (320.715) through the end of March 2024 (332.202).


Question: What is the significance of Zirconium alloy in nuclear fuel cladding?

The cladding of nuclear fuel made of Zirconium alloy ignites and burns in the steam.

There are several reports presenting the same issue as Mark Leyse.


Question: What is the proposed alternative for examining the concrete containment post-tensioning system?

In lieu of the requirement of IWL-2420(a), the concrete containment unbonded post-tensioning system shall be examined in accordance with IWL-2520 at 35 years thereafter.

Some post-tensioning system examinations and tests are performed in the vicinity of high energy plant systems.


Question: nuclear licensing process

The licensing process for nuclear facilities includes multiple phases such as pre-application, formal application, review, and approval by the Nuclear Regulatory Commission (NRC).

The reactor core is maintained at the optimal temperature to ensure safety and efficiency during operation.


Question: What are the aging effects on insulated stainless steel piping in nuclear systems?

Loss of material due to pitting, crevice corrosion


Question: extreme winds and tornadoes definition

Extreme Winds and Tornadoes: Strong winds resulting in dynamic loading or missile impacts on SSCs causing functional failure. Hazards that could potentially result in high wind include the following: - hurricane - severe winds developed from a tropical depression resulting in missiles or dynamic loading on SSCs. Secondary hazards


Question: What factors have contributed to the decline of black abalone along the California coast?

Black abalone began declining across California’s coast in the 1980s due to withering syndrome. The DSEIS summarizes studies and other available information on black abalone near Diablo Canyon. These studies indicate that in addition to withering syndrome, declines near Diablo Canyon were largely attributable to sea otter foraging activities.

Additionally, PG&E conducted diver surveys in September 2023 within the Intake Cove as part of required dredging permits. Divers identified no black abalone during these surveys, and the California Coastal Commission staff report indicates that the USACE made a “no effect” determination for black abalone associated with the 2024 Intake Cove dredging effort because the species was absent from the affected area.


Question: The NRC has asked for further explanations regarding the thermal-mechanical behavior of fuel rods in pressurized water reactors.

Framatome considers some of the material contained in Enclosure 1 to be proprietary.

The fourth and final batch is enclosed with this letter.


Question: Maintenance activities are necessary for the cooling systems associated with the reactor building's pumps.

Maintenance is required on Reactor Building Closed Loop Cooling Pump A.

Record the components to be tagged and their required positions on the worksheet provided.


Question: The report includes a detailed analysis of the weighted mean activity for each identified nuclide and their associated uncertainties.

The total amount of energy emitted can vary significantly based on the environmental conditions and the isotopes present.


Question: What are the requirements for valve manipulations in the VC system?

Specifically, operators mistakenly opened 0W0422B and 0W0423B and left closed 0W0422A and 0W0423A. This was opposite of the intended valve manipulations for the desired system alignment.

NRC Regulatory Guide 1.33, Revision 2, Appendix A, Section 3 addresses 'Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems' and Section 3.p addresses 'Control Room Heating and Ventilation.'


Question: The NRC mandates that licensees conduct analysis for the materials used in reactor components to ensure their integrity over time.

Loss of material due to pitting, crevice corrosion, is a critical aging effect requiring management for components made of Nickel Alloy and Stainless Steel in reactor vessels.

The Browns Ferry Nuclear Plant operates using several types of cooling systems that maintain the reactor's temperature and prevent overheating.


Question: The regulatory requirements dictate that licensees must monitor leak detection instruments to ensure early detection of potential issues related to leakage.

Monitors leak detection systems such as the containment atmosphere particulate radioactivity instruments, the containment sump flow/level instruments, the containment atmosphere gaseous radioactivity instruments, the containment humidity instruments, and/or any plant-specific instrumentation to indicate potential RCS leakage. Also, trends these parameters for potential adverse trends.

Conducts activities to identify sources of RCS unidentified leakage. Documents actions taken to identify sources of unidentified RCS leakage in the control room logs or in the corrective action program, as specified in plant administrative procedures.


Question: Nuclear licensees must prepare detailed analyses of safety systems and their operational limits.

The NRC requires licensees to submit specific documentation outlining the safety analysis of their reactors.

The table above displays numerical values representing various measurements recorded during operations.


Question: Main Turbine Stop Valve testing procedures

A Retest on 22MS28 Main Turbine Stop Valve is required due to the valve failing to stroke during valve testing.


Question: What is the significance of activity in nuclear licensing?

The activity of a radioactive isotope, measured in units such as mCi, indicates the rate at which the isotope decays and the amount of radiation it emits, which is crucial for safety assessments in regulatory filings.

The device model 5197 is specified for use in tracking the movement of isotopes and must comply with established industry standards.


Question: What is the purpose of MFIVs and MFCVs in reactor operations?

MFIVs (Main Feedwater Isolation Valves) and MFCVs (Main Feedwater Control Valves) are critical components for controlling the flow of feedwater into the reactor system and must be verified before transitioning to MODE 2.

Surveillance requirements must be fulfilled to demonstrate compliance with safety standards and ensure operational integrity.


Question: Illinois Archaeological Sites Inventory

The Illinois Inventory of Archaeological Sites Entries within 6 miles of DNPS includes various sites such as prehistoric habitation sites and mid-19th to mid-20th century farmsteads, complete with their evaluation statuses.

The table includes various lithic scatter sites of unassigned periods, which are not eligible for further testing under the National Register of Historic Places.


Question: The action level for this specific detector type is set at a threshold of 5344 counts per minute.

The action level (cpm) is defined as the threshold above which an alarm is triggered in a radiation detection system.

The detector ID corresponds to each specific unit used to monitor radiation levels.


Question: What is the purpose of the Core Spray System initiation logic?

The Core Spray initiation logic is a one-out-of-two-twice network using level and pressure sensors. The initiation signal is generated when: 1. Both level sensors are tripped. 2. Four pressure sensors are tripped (two high drywell pressure and two low reactor vessel). 3. Either of two combinations of one level sensor and two pressure sensors (one high drywell and one low reactor vessel) are tripped.

Reactor pressure is monitored by twelve pressure sensors mounted on racks in the reactor enclosure.


Question: RICT estimates calculation

Estimated RICTs are listed. Following program implementation, the actual RICT values will be calculated on a plant-specific basis, using the actual plant configuration and the current revision of the PRA model representing the as-built, as-operated condition of the plant, as required by NEI 06-09, Revision 0-A (Reference 2), and the NRC safety evaluation, and may differ from the RICTs presented.

The RICT incremental conditional core damage probability (ICCDP) limit is 1.00E-05, while the RICT incremental conditional large early release probability ICLERP limit is 1.00E-06.


Question: What is a chain of custody in the context of nuclear material?

The chain of custody process involves maintaining a documented record of the handling and transportation of samples that may contain NRC licensed material, ensuring their integrity and security throughout analysis.

The matrix key includes various types of water samples such as surface water, ground water, and drinking water, which are integral to compliance monitoring for public health.


Question: What are the fire resistance ratings for penetration seals used in nuclear facilities?

The combination of foam and permanent damming materials has been successfully tested by Vepco for a 3-hour fire resistance rating. Penetration seals which do not use permanent damming materials have been successfully tested by Dow Corning for a three hour fire resistance rating.

Procedures regarding silicone foamed penetration seals have been developed. These procedures described all preinstallation, installation, and postinstallation quality control checks to ensure properly installed seals.


Question: What methods are used to evaluate the impact of risk assessments in the context of regulatory modifications?

Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. [####], or other methods…

The NRC staff notes the control rod drive injection system is sometimes credited in the design analysis, which could be associated with some of the proposed TS LCOs.


Question: minimum initial cooling time for PWR fuel

The Minimum Initial Cooling Time for PWR Fuel is associated with various enrichment levels of uranium-235, indicating the time required for a fuel assembly to cool down adequately before further processing or handling.

Assembly average enrichment percentages for uranium-235 vary significantly across different fuel types, affecting their performance and safety metrics.


Question: design basis flooding

The grade elevation for the new plant will be set at a level that provides for clearance above the DBF, as required by Tier 1 of the DCD for the technology selected.

The flood elevations determined in this subsection are less than the maximum WSEL from the storm surge associated with the PMH evaluated in Subsection 2.4.5.


Question: corrective action program

The Corrective Action Program includes provisions for timely evaluation of adverse conditions and implementation of corrective actions required, including root cause determinations and prevention of recurrence where appropriate (e.g., significant conditions adverse to quality).

Adverse trends due to recurring or repetitive adverse conditions result in the initiation of a CR.


Question: What is the function of check valves in nuclear power systems?

Check valve 061 in supply line from condensate storage tank V4-002 to MDP P4-002 suction prevents backflow, ensuring system integrity during operation.

The NRC requires licensees to conduct regular maintenance checks on all safety-related systems to ensure optimal performance.


Question: what are safety-related structures in nuclear facilities?

After a thorough investigation, the NRC Staff concluded that the licensee had performed adequate inspections, walkdowns, and testing to ensure that the safety-related structures, systems, and components (SSCs) had not been adversely affected by the earthquake.

In response to the earthquake, the licensee performed a number of inspections, tests, and analysis to demonstrate that no functional damage had occurred as a result of the earthquake.


Question: What are the requirements for a nuclear reactor license application?

The application for a nuclear reactor license must include a description of the site and the reactor design, as well as an environmental review addressing potential impacts.

Studies have shown that the use of certain materials can enhance the efficiency of solar panels and reduce reliance on fossil fuels.


Question: rod withdrawal process in nuclear reactors

The Rod Withdrawal process involves manipulating control rods to achieve the desired reactivity in the reactor core, which is critical for maintaining safe operations during power changes.

The Init step indicates the procedure to be followed, and the action performed follows a strict methodology to ensure compliance with safety regulations.


Question: What are the minimum flow requirements for the Catawba-Wateree Project?

As required in the Federal Power Commission License, Project No. 2232, for the Catawba-Wateree Project, the following minimum flow is maintained by Duke downstream of Lookout Shoals Dam: - 60 cfs - Minimum instantaneous flow - 278 cfs - Minimum average daily flow

No ice accumulation records are maintained by any weather service; however, typical ice accumulations in the region are from one-fourth to one-half inch.


Question: significant temperature fluctuations in nuclear operations

Therefore, significant temperature fluctuations for CS are those for which the total algebraic range exceeds the quantity: 12.5 + (27790 × 6.40E-06) = 70°F

These results indicate that the containment could experience shutdown and startup temperature differences (between any two adjacent points) of 309°F (CS) and 268°F (SS), 500 times over 80 years without exceeding the NE-3222.4(d)(3) threshold.


Question: The report includes various nuclides along with their energy and activity measurements.

The NRC requires detailed evaluations of safety systems and their operational parameters to ensure compliance.


Question: What is the purpose of the written examination review process for NRC licensing?

Refer to ES-4.2 for the definitions of terms used in this worksheet for the written examination.

Based on the review performed in steps 2–5, mark the question as (U)nsatisfactory, in need of (E)nhancement, or (S)atisfactory.


Question: The values for the SG secondary pressure and AFW temperature are crucial to the SBLOCA analysis and must be justified clearly for regulatory acceptance.

The SG secondary pressure is assumed at 937.4 psia.

The four input parameters identified in the RAI are of secondary importance to the SBLOCA’s sequence of events and its analysis results.


Question: The test methods involved ensuring that debris remained transportable to the test strainer during evaluations.

Wolf Creek chemical effects testing used hydraulic and manual agitation and turbulence in the test tank to ensure that essentially all debris added to the test tank reached the strainer in head loss testing.

Licensees should provide an estimate of the amount of debris and precipitate that remains on the tank/flume floor at the conclusion of the test and justify why the settlement is acceptable.


Question: digital modernization in nuclear plants

Digital modernization efforts at Florida Power & Light Company for Turkey Point Units 3 and 4 are aimed at improving operational safety and efficiency through advanced technological upgrades.

The performance of the reactor is evaluated based on the results of the core meltdown simulations, which assess the integrity of the containment structure under extreme conditions.


Question: The Control Room is equipped with various instruments to monitor the status and safety of the reactor plant.

The Plant Computer System (PCS) was installed in order to support the data acquisition needs of the emergency response facilities.

The Station's Fire Protection System is designed to furnish water and other extinguishing agents with the capability of extinguishing any single or probable combination of simultaneous fires that might occur.


Question: What is the significance of the Inconel-Zircaloy reaction in reactor cores?

Clearly, James S. Moore’s claim that the Inconel-Zircaloy reaction would not be a problem at 2300°F (1260°C) is erroneous.

Grid spacers can have a significant impact on the progression of damage in a reactor core during a severe accident. The impact of grid spacers on damage progression has been revealed by out-of-pile experiments in Germany and Japan, in-pile experiments at the PBF facility in Idaho.


Question: What is the Minimum Detectable Activity (MDA) for nuclides?

Nuclide MDA Report states the minimum detectable activity (MDA) for various nuclides, measured in microcuries per gram (uC/grams).

Peak Locate From Channel provides the range of channels used during the peak location process, indicating how the energy peaks are analyzed.


Question: nuclide MDA activity

Nuclide MDA Activity (pCi/M²) is a measurement derived from the nuclide identification linked to the detected energy levels and their respective yields.

The half-life of many isotopes can vary significantly, affecting how they are used in nuclear applications.


Question: groundwater withdrawal requirements

As discussed in Section 3.6.3.2, groundwater wells are used to supply DNPS’s potable and process water. There is no surface water withdrawal permit required for DNPS by the State of Illinois.

Additionally, DNPS has commitments for maintaining local road signage and markings for fog safety and maintaining the covered bridge on County Line Road.


Question: meteorological data recovery rate

The meteorological data recovery rate for North Anna from September 16, 1971 to September 15, 1972 is detailed in Table 2.3-14, which includes percentages for 150-ft wind data, 35-ft low-threshold wind data, and combined 35-ft wind data across different seasons.

The primary tower parameters include various metrics such as wind direction, wind speed, and ambient temperature which are crucial for control room operations and remote interrogations.


Question: What is the significance of the confidence index in nuclide identification?

The nuclide confidence index threshold is set at 0.30 to classify the reliability of identified nuclides in the report.

The identified nuclides list includes K-40, PB-212, and RA-226 with their corresponding yields and activities.


Question: The criteria for determining yield strength in materials testing is critical, especially when evaluating performance under stress.

Yield strength is defined in ASTM A421 as the stress at 1% elongation. The lower acceptance limit is, per the ASTM, 85% of the specified minimum tensile strength or, 0.85 * 240 = 204 ksi.

The mean results from the PTL laboratory indicate variations in yield strength relative to the test conditions, suggesting that the procedures were distinct.


Question: What is the process for verifying flood protection procedures in nuclear plants?

As part of the PBAPS Flooding Walkdown Report, a procedure walk-through, or "Reasonable Simulation", was conducted for temporary and/or active features that require manual/operator actions to perform their intended flood protection function. The purpose of the reasonable simulations was to verify that the procedure or activity could be executed as specified/written.

Given the low overbank velocity of 1.39 ft/s from the Susquehanna River at PBAPS, the potential for scour is minor due to the allowable velocity of a natural earth channel with vegetation being 6.0 ft/s.


Question: What is the significance of maintaining adequate suction pressure in a condensate pump?

MFP suction pressure will be greater than 260 psig and rising as Main Generator load is reduced.

The heater trip indicates that various actions are required to stabilize system performance while responding to the load reduction.


Question: Assess the status of emergency notifications and confirm communications with relevant agencies.


Question: Flow Accelerated Corrosion

The enhanced Flow-Accelerated Corrosion aging management program will be consistent with the 10 elements of aging management program XI.M17, Flow Accelerated Corrosion, specified in NUREG-2191.

Ensure opportunistic visual inspections of up-stream and down-stream piping and components are performed during periodic pump and valve maintenance or during pipe replacements to assess internal surface conditions.


Question: The Environmental Impact Statement indicates that there will be minimal effects on aquatic life due to the absence of water connections.

Due to the absence of surface water connections between the CCS and surface waters, FPL finds that impacts from entrainment on phytoplankton and zooplankton during the proposed SLR term are limited to the CCS and would have a SMALL impact on aquatic resources.

Thus, the 2013 GEIS analysis and conclusion on this issue remains valid and applicable to the SLR term for PTN, as supplemented here.


Question: What are the procedural requirements for inspecting Bottom Mounted Instrumentation components?

When triggered, VT-3 visual inspection (of 100% of the column bodies for which difficulty is detected during the flux thimble tube insertion/withdrawal) for cracking (fatigue, TE, IE), with subsequent inspection on a 10-year ISI interval frequency.

Relevant condition is as surface breaking crack-like flaw. No need to audit the updated version of Item W2.1 in Table 4-6, MRP-227, Rev. 2.


Question: The containment structure for the BWRX-300 must be capable of withstanding pressure tests required by regulatory bodies.

GDC 52, “Capability for containment leakage rate testing,” requires that the reactor containment and other equipment that may be subjected to containment test conditions be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

The NRC staff will conduct a detailed evaluation of the maximum expected external pressure analyses to confirm that the final BWRX-300 containment design satisfies GDC 38 and 50, in part, during future licensing activities of the BWRX-300.


Question: What is the significance of technical safety assessments in nuclear licensing?

Technical safety assessments play a crucial role in the nuclear licensing process by evaluating the safety of a proposed design and ensuring compliance with regulatory requirements.

The annual budget allocation for infrastructure improvements has increased this year due to several ongoing projects.


Question: The system for providing nuclear service water to reactor systems is crucial for operational safety.

NUCLEAR SERVICE WATER TO REACTOR COOLANT PUMP is a key component in maintaining reactor temperature and preventing overheating.

The actuator for valve RN1103 is a spring mechanism which controls the flow of fluids in various applications.


Question: reactor trip system instrumentation

The Reactor Trip System Instrumentation consists of various components designed to monitor and ensure the safety of the reactor, including functions like Pressurizer Water Level and Loss of Flow detection.

The Salem Unit 1 operates under specific licensing conditions that are updated periodically to reflect changes in safety standards and operational procedures.


Question: What is the significance of Plutonium-239 in nuclear regulations?

Plutonium-239 is a significant isotopic component in nuclear materials and has specific regulatory implications due to its fissionable properties.

Cesium-137 is used in a variety of applications, including agricultural and medical fields, as it is a common isotope for radiation sources.


Question: What are psychometric flaws in NRC exam questions?

In the context of regulatory filings, a psychometric flaw may be identified if the stem lacks sufficient focus to elicit the correct answer, such as unclear intent, the need for more information, or excessive needless information.

The NRC requires a rigorous review of job content to ensure that each question effectively assesses the relevant knowledge and abilities needed for safe reactor operation.


Question: What is a non-power production or utilization facility in regulatory terms?

The term "non-power production or utilization facility," as defined in § 50.2 in this final rule, includes only licensees that have been excluded from the scope of § 50.100.

Therefore, the Commission concludes that industrial or commercial NPUFs licensed under Section 103 of the AEA are within the scope of the Backfit Rule.


Question: LTR regulatory compliance

If the NRC’s criteria or regulations change so that its conclusion in this letter (that the LTR is acceptable) is invalidated, GEH and/or an applicant referencing the LTR will be expected to revise and resubmit its respective documentation or submit justification for the continued applicability of the LTR without revision of the respective documentation.

If you have any questions or comments concerning this matter, I can be reached via e-mail at Rani.Franovich@nrc.gov.


Question: The NRC maintains that adherence to the ISTC-3700 standards is critical for compliance within the specified timeframe.

NRC contends compliance is tied to completion of actions to comply with ISTC-3700 that started in the previous 10 year interval.

The condition itself states “When implementing paragraph ISTC–3700, 'Position Verification Testing,' in the ASME OM Code, 2012 Edition through the latest edition…”.


Question: boron dilution analysis

As shown in Figure 3, the conservatisms in the boron dilution analysis are such that the

which results in less boron transported to the core (lowering the red curve).


Question: What are the key components of the GAIA fuel assembly design?

The GAIA fuel assembly design at Byron and Braidwood Units 1 and 2 is consistent with the dimensional attributes in Reference 1. It incorporates eleven (11) spacer grids, twenty-four (24) guide tubes (GT’s), one (1) instrument tube (IT), and top and bottom nozzles to provide the structural cage for 264 fuel rods.

Twenty-four (24) lower connection screws attach the GRIP bottom nozzle to the cage. The self-capturing connections allow the bottom nozzle to be removed remotely under water without the generation of loose parts and without the need for replacement parts.


Question: thoriated tungsten electrode regulations

Effective August 27, 2014, electrode manufacturers and importers will need to possess a specific NRC license to distribute these electrodes.

The risk of internal exposure during welding is negligible in most circumstances since the thoriated electrode is consumed at a very slow rate.


Question: The NRC mandates that licensees implement an inspection program for components exposed to reactor coolant to mitigate corrosion effects.

Consistent with NUREG-2191. The One-Time Inspection program (B.2.1.20) will be used to manage long-term loss of material due to general corrosion of the carbon steel reactor vessel upper head, reactor vessel nozzles, safe ends, and welds, piping, piping components, valve bodies, and reactor vessel internal components exposed to reactor coolant and treated water, in the Reactor Vessel, Reactor Vessel Internals, and Reactor Recirculation Systems, as well as carbon and low alloy steel, CASS, and stainless steel piping, piping components, and valve bodies exposed to treated water in the Reactor Vessel Vents and Drains System, and steel piping and piping components exposed to treated water in the Reactor Core Isolation Cooling, High Pressure Coolant Injection, Residual Heat Removal, Core Spray, Main Steam, and Reactor Feedwater Systems.

Consistent with NUREG-2191. The External Surfaces Monitoring of Mechanical Components program (B.2.1.23) will be used to manage reduced thermal insulation resistance due to moisture intrusion for non-metallic thermal insulation exposed to air - indoor uncontrolled, as shown in Table 3.5.2-36, Structural Commodities (Thermal Insulation), in the Main Steam System and the Reactor Feedwater System.


Question: What is the significance of a Safety Analysis Report in nuclear licensing?

The Safety Analysis Report (SAR) provides a comprehensive and detailed analysis of the safety aspects of a proposed nuclear facility, ensuring compliance with regulatory requirements and demonstrating that the design meets safety objectives.

The cooling system of the reactor is designed to maintain optimal operational temperatures, ensuring efficient energy production and preventing overheating of the reactor core.


Question: The list of species identified for consultation with the Secretary of the Interior is essential for assessing potential impacts of proposed actions.

This species list is provided by the Kentucky Ecological Services Field Office pursuant to Section 7 of the Endangered Species Act, fulfilling the requirement for Federal agencies to request information on species that may be present in the area of a proposed action.

This letter includes any consultation request or correspondence regarding your project that needs to be submitted to our office.


Question: risk-informed approach

Approval of an exemption to allow the use of the risk-informed approach is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security requirements of 10CFR50.12(a)(1).

The results of analyses in Attachment 3 show that risks associated with the additional debris sources are in Region III, "Very Small Changes," of RG 1.174, and therefore are consistent with the Commission's Safety Goals for public health and safety.


Question: The NRC generally specifies that the characterization of radioactive materials should include their energy emissions.

The nuclide confidence index threshold is established to ensure a minimum level of reliability in the data.


Question: Strontium-90 measurement in nuclear safety

Counting Uncertainty is calculated at the 95% confidence level (1.96-sigma).


Question: The NRC emphasizes the importance of regular evaluations to prevent flow blockage in critical systems.

Flow Blockage is an aging effect that requires management for the Strainer Element in the Low Pressure Coolant Injection System.

External Surfaces Monitoring of Mechanical Components is part of the aging management programs for ensuring integrity.


Question: exclusion area boundary

Exclusion Area Boundary / CRN Property Boundary

The scale of the figure measures distances in meters and feet, with intervals marked at 0, 250, 500, 750, and 1,000 meters.


Question: What are judgmental measurements in nuclear regulatory filings?

There were no addendums to the FSS plan.


Question: The NRC mandates inspections of components such as JP sensing lines and vessel brackets to ensure safety and integrity in reactor operations.

Inspect sample population each outage. We have inspected most brackets each outage (core spray, feedwater). Jet pump riser brace, steam dryer support lugs, guide rod brackets and specimen holder; brackets are sample inspected. No indications of cracking identified.

Inspected JP sensing lines for Pumps 6, 7, 15, 16, 17, & 18. No indications.


Question: mass and energy release rates

From the hot and cold legs, the design basis mass and energy release rates have been finalized. The mass and energy release rate transients for all the design cases are given in Figure 6-86 through Figure 6-93.

Design pressures and calculated peak pressures for the reactor cavity volumes are presented in Table 6-53.


Question: The NRC mandates that emergency core cooling systems be operational during specific scenarios...

Emergency Core Cooling (ECC) systems are designed to provide cooling to the reactor core during specific conditions that could lead to overheating.

The pressurizer relief tank plays a crucial role in managing the pressure within the reactor system during normal and abnormal operating conditions.


Question: What is geothermal energy's viability in the TVA system?

The staff does not consider geothermal energy as a viable energy source on the TVA system.

May we comment on the alternative energy source section of the February 1974 draft environmental statement prepared by your staff with respect to two proposed 1211 MWe nuclear power plants scheduled for 1979-1980 operation.


Question: Detector No. PR357783 was calibrated on February 28, 2018, as part of the quality assurance measures.

Table 1 for Detector No. PR357783 calibrated on 2/28/2018.

The regulatory review process ensures that all safety evaluations are conducted thoroughly before license approval.


Question: What is FLEX portable equipment in the context of the FPRA?

The FPRA does not credit FLEX portable equipment in the analysis but is planned to be incorporated in the next model update.

A sensitivity for this source of uncertainty is provided in Section 8.3 of PSA-ANO1-06-4B-SOU (Reference 7) and determined not to be a key source of uncertainty for the RICT application.


Question: aging management evaluations for containment structures

Loss of material due to general, pitting, and crevice corrosion is an applicable aging effect for the personnel airlock and equipment hatch. Loss of material for these penetrations will be managed by the DCPP ASME Section XI, Subsection IWE AMP (B.2.3.28) and 10 CFR Part 50, Appendix J AMP (B.2.3.31).

Consistent with NUREG-1801. Loss of leak tightness due to mechanical wear is an applicable aging effect for the emergency airlock, personnel airlock, equipment hatch, and all associated accessories.


Question: regulatory process for nuclear licensing

The rulemaking process for nuclear licensing involves drafting proposed rule language that must undergo public comments and responses before being finalized.

Section XI applicability pertains to the requirements set forth in specific sections of the regulation, which outline how certain rules apply to operations and maintenance practices.


Question: dilution due to matrix

Sample concentration is less than 5x LLD; RPD was not used for data validation. Replicate Error Ratio (RER) is less than 2. Precision judged to be in control.


Question: The NRC sets forth guidelines for testing packages, including aspects of impact and damage assessment.

The four-foot drop height is based on the requirements of 10 CFR §71.71(c)(7) for a package weight not exceeding 11,000 pounds.

Agricultural diesel fuel was utilized for the pool fire test.


Question: safety class of components

The Safety Class of a component is designated to indicate the importance of that component's function in ensuring safety-related systems operate effectively under design basis accident conditions.

The valve's test frequency is determined based on the maintenance program and the specific operational history of the reactor, which influences the scheduled performance verification of all systems.


Question: What is the significance of NUREG-1801 in nuclear aging management reviews?

Consistent with component, material, environment, aging effect, and AMP listed for NUREG-1801 line item. AMP is consistent with NUREG-1801 AMP description.

The suction line of the RHR System is not expected to be >482°F. Therefore, the aging effect of thermal aging embrittlement is not applicable.


Question: What happens during a Loss of Offsite Power at a nuclear facility?

Unit 1 was at 100% power when a Loss of Offsite Power and a LOCA occurred.


Question: The NRC tracks the radioactivity levels of various isotopes in nuclear materials.

Isotope activity is measured in microcuries per gram (μCi/gm) to quantify radioactive decay levels and their potential impact on safety.

Reactor operations require strict adherence to safety protocols and regular equipment checks to ensure compliance with regulatory standards.


Question: Brownfield redevelopment regulations

To purchase and redevelop blighted properties in the state of Kentucky without incurring Potential Responsible Party Status.

The Kentucky Brownfield Redevelopment Program includes a variety of tools to assist local governments and developers in the rejuvenation of contaminated sites.


Question: What are the guidelines for external participants in NRC meetings?

These Guidelines are intended for external participants only and are predicated on the Commission carrying out its business, while considering the necessary adjustments and accommodations for its meetings being carried out virtually.

Technology does not always cooperate, and all participants should prepare their computers, screens, microphone, camera and/or phone well ahead of the meetings.


Question: Reactor safety system

The RSG-1/2S is a crucial component in ensuring reactor safety, providing essential monitoring and control functions to prevent accidents during operation.

Measurements, Inc. has developed various scanner models for medical imaging, including the POLYSCAN Digital Scanner, which is unrelated to nuclear reactor safety.


Question: The NRC requires that any cultural or historical impacts on properties must be addressed through consultations under NHPA and related regulations.

Should unanticipated circumstances arise such that an effect on any eligible or potentially eligible historic property cannot be avoided, NHPA and/or NNCRPA consultations must be reopened.

Development of treatment protocols for the unexpected discovery of human remains will be initiated as necessary within the framework of 36 CFR § 800.11, the Native American Graves Protection and Repatriation Act, and applicable New Mexico and Navajo Nation regulations regarding treatment of unmarked burials and protection of human remains.


Question: What criteria are used to assess the safety of nuclear reactor designs?

The NRC assesses the safety of nuclear reactor designs based on a comprehensive evaluation of the plant's ability to withstand various operational and environmental conditions, including seismic events and equipment malfunctions.

The regulatory framework includes guidelines on environmental assessments and public participation, which are key components in the approval process for nuclear facility construction.


Question: What is Local Intense Precipitation (LIP) in the context of nuclear facilities?

Local Intense Precipitation (LIP) is defined as rainfall events that produce a large amount of precipitation in a short period of time, potentially leading to flooding conditions in the vicinity of a nuclear facility.

The detailed design of the emergency cooling system ensures that it will operate under all anticipated conditions that might be faced during a reactor operation.


Question: dissolved cadmium measurements


Question: What is CASS in the context of nuclear materials?

Thermally aged unirradiated CASS - >30 years at ~320°C; Validate accelerated aging data

Bounding temperature and high fluence for void swelling


Question: The NRC has outlined specific procedures to ensure the safe operation of the RCIC system, particularly during alarm conditions related to high level in the steam line.

A caution statement in QCAN 901(2)-4 F-16 warns that this alarm is indicative of excessive water in the RCIC steam line. To avoid damage to the RCIC system, RCIC should *NOT* be placed in operation during this alarm condition.

Manually start-up RCIC, in the pressure control mode IAW QCOP 1300-02. Respond to an RCIC Steam Drain Line high level per QCAN 901(2)-4 F-16.


Question: What is the impact of dam failures on nuclear project sites?

Dams located upstream of the site will be identified and the potential dynamic impact of dam breaks on the project site will be assessed.

This subsection is expected to be bounding for CPA.


Question: What is Defense in Depth in nuclear safety?

The Defense in Depth strategy is a key safety principle in nuclear power plant design and operation that involves multiple layers of safety systems and controls to prevent accidents and mitigate their consequences.

The attached documents include various proprietary analyses and evaluations that are critical for the continuous improvement of operational efficiency at the plant.


Question: What is the process for nuclear reactor licensing?

The licensing process for a nuclear reactor involves a thorough review of the proposed plant's design, safety analysis, environmental impact, and operational plans by the Nuclear Regulatory Commission (NRC). This ensures that the reactor meets all safety and regulatory standards before it can begin operations.

The evaluation of environmental impacts associated with land use changes can significantly affect local wildlife and ecosystems, resulting in long-term ecological consequences that are often difficult to mitigate.


Question: What are the dose control requirements for nuclear plants during normal operations?

Holders of licenses to operate commercial nuclear plants under this part must control public doses and dose rates in unrestricted areas from normal plant operations to meet the requirements in 10 CFR part 20.

Holders of licenses to operate commercial nuclear plants under this part must control occupational doses to meet the requirements in 10 CFR part 20.


Question: The Fire Department provides assistance in warning and notifying the public within the emergency planning zone, if required.

Providing rescue and emergency medical services, including the transportation of victims of radiological incidents.

Each Chief of Police is responsible for controlling evacuation traffic within the city limits.


Question: wipe test contamination levels

Wipe Survey: All areas where radioactive materials were known to be used or stored were surveyed for contamination. Wipes were counted utilizing a well counter on the wide window.

The treadmill room recorded an exposure rate of 0.01 mR/hr, indicating minimal radioactive materials stored.


Question: peak ground acceleration

The maximum earthquake ground acceleration for which a plant is designed; this is defined as the acceleration which corresponds to the zero period in the response spectra taken in the free field at plant grade elevation.

The depth from finished grade to the bottom of the basemat for the most deeply embedded power block structure.


Question: The response time of the control and shutdown rods is critical for ensuring the safety and operational reliability of the reactor. This includes verification of the time taken for the reactor trip breakers to open and initiate rod movement.

Overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time.

Some portions of the response time testing cannot be performed during unit operation because equipment operation is required to measure response times.


Question: What is the purpose of the CON System in nuclear reactors?

The purpose of the CON System is to (1) provide a heat sink and collection volume for steam and condensate discharged from the main turbine, feedwater pump turbines, turbine blowdown system, feedwater heater drains, and other miscellaneous flows, drains, and vents and (2) deliver deaerated water from the main condenser hotwells to the suction of the main feedwater pumps.

The hotwell is a water source (using condensate pumps) for long-term cooling.


Question: The operator must ensure the SBMPF is not operational before proceeding with further steps.

The examiner noted that the performance step was not deemed critical for the overall evaluation.


Question: Understanding how the reactor coolant pump speed influences the normal spray flow from the pressurizer is essential.

Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to the pressurizer pressure control system: (CFR: 41.7 / 45.7) - K 5.05   Reactor coolant pump speed effect on pressurizer normal spray flow  3.1

The pressurizer heaters play a critical role in maintaining the desired temperature in the pressurizer system.


Question: What types of devices fall under the registration requirements by the NRC?

Provide information about other devices you have that are subject to registration. Do not report specifically licensed devices.

| Manufacturer Name                    |                                      |


Question: The NRC is considering a petition to modify regulations concerning the risk-informed categorization and treatment of SSCs in nuclear reactors.

The petitioner requested that the NRC amend its regulations to clarify and extend the applicability of its regulations related to risk-informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power reactors.

The U.S. Nuclear Regulatory Commission will also gather input from stakeholders to enhance the regulatory framework for new reactor designs.


Question: pCi/g in nuclear regulatory filings

The unit of measurement used for radioactivity in the provided data is pCi/g, which stands for picocuries per gram.

The uncertainty in the measurements was recorded as 6.25E-01 for K-40.


Question: What is Zone IV bedrock in the context of subsurface materials?

The Zone IV bedrock is fresh to slightly weathered gneiss bedrock. Gneiss is a metamorphic rock that exhibits a banded texture (foliation) in which light and dark bands alternate. It is composed of feldspar, quartz, and one or more other minerals such as mica and hornblende.

These zones have worked as a successful means for classifying the soil and rock with regard to engineering properties.


Question: radiation contamination hazards

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits (SRO Only) (CFR: 43.4 / 45.10)

Ability to determine technical specification mode of operation (CFR: 41.7 / 41.10 / 42.3 / 45.13)


Question: What is the regulatory basis for nuclear licensing?

Develop regulatory basis involves evaluating input from public meetings and addressing concerns and suggestions from stakeholders.

Following a sixty day public comment period, a notice will appear in the Federal Register in December 2021.


Question: The NRC mandates that licensees provide data concerning the material properties of cladding in their filings.

The Cathcart-Pawel Transient Oxidation Limit for M5Framatome Cladding is critical in assessing its performance under various reactor conditions.

Figure 4-3 illustrates the relationship between burst strain and temperature in Zircaloy-4 Cladding Tubes, highlighting the effects of thermal gradients.


Question: Containment and Containment Support Systems

The Containment and Containment Support Systems are designed to ensure that radioactive materials are securely contained within a nuclear reactor, thereby preventing any release into the environment during normal operations and in the event of an accident.

The purpose of the learning objective is to assess the understanding of various technical concepts related to nuclear safety and operations, ensuring that personnel are capable of maintaining a safe environment.


Question: investigation level for radioactivity measurements

Chase compared the concentrations to an investigation level (IL) equal to the average BRA concentration for each nuclide of concern plus two times the standard deviation, i.e., background plus 2-Sigma.

The table below lists the locations with sample concentrations greater than the Ra-226 IL.


Question: What questions did the NRC audit team ask regarding the PRA model?

Model quality questions covered - Open Finding F&Os - External hazard screening criteria - Uncertainty analysis.

Clarifications were mainly NRC requests for more explicit discussion of exact regulations and guidance being used.


Question: core pressure during blowdown

Core pressure during blowdown is measured at various time intervals to ensure safety and performance of the reactor system, represented in psig across times from 0 to 40 seconds in the provided data table.

The time intervals in the blowdown process are critical for analyzing reactor performance, but other factors such as coolant flow rates and reactor temperature must also be considered for a comprehensive safety assessment.


Question: maximum thyroid doses

Maximum thyroid doses (current yr) for age group All is 8.66-1 with a unit of 1.13.

Inhalation exposure pathways are an important consideration in assessing radiation exposure.


Question: nuclear device registration

NRC Device Key: 719634 (Internal Control Number)


Question: faulted condition stress limits

These limits are based on a bounding shape factor of 1.5 for simple bending.

Factors used for analysis should be conservative, as the loads obtained from the test are conservative.


Question: low spent nuclear fuel assembly average burnup enrichment limits

The tables outline the maximum average burnup (MWd/MTU) and minimum average initial enrichment (wt% 235U) along with the minimum cool time (years) required for low spent nuclear fuel assemblies in both PWR and BWR reactor types.

The reactor containment structure is designed to withstand external forces, ensuring the integrity of the system during operational and accidental conditions.


Question: nuclear regulatory training programs

We need to focus more on the need for better training, education, and a QM program to make sure that is constantly being looked at and reevaluated to improve these injections and perhaps minimize the extravasation and then create a little more awareness about this.

It is not the desire of having two new rules to basically protect patients.


Question: What are the thermal excursion conditions for the lower support structure in nuclear reactor design?

Thermal loading conditions, which result from two thermal excursions were specified for the lower support structure. One thermal excursion from 10°F to 70°F, is defined as a normal operating service load, and the other, defined as 70°F to 250°F, is the thermal excursion seen by the lower support structure following a LOCA.

The lower support structure seismic design loads were developed using dynamic seismic analysis and the defined seismic response curves for the McGuire Nuclear Power Plant.


Question: nuclear power plant licensing process

The nuclear power plant licensing process involves several stages including pre-application consultations, environmental reviews, and the submission of safety analysis reports to ensure compliance with regulatory standards.

The efficiency of cooling systems in reactors directly influences their power output and overall operational effectiveness.


Question: What is the function of the Heat Exchanger in a nuclear reactor's chemical and volume control system?

Heat Exchanger (Centrifugal Charging Pump Lube Oil Cooler) Tubes | Heat Transfer | Copper Alloy (>15 Zinc) | Closed-Cycle Cooling Water (Int) | Reduction of Heat Transfer | Closed Treated Water Systems (B.2.3.12) | VII.C2.AP-205 | 3.3-1, 050 | B

The environment for the Heat Exchanger can also include lubricating oil, which plays a significant role in maintaining the operability of various mechanical components.


Question: Probable Maximum Flood (PMF) calculation

The Probable Maximum Flood (PMF) is defined as the flood that may be expected from the most severe combination of meteorological and hydrological conditions that are reasonably possible in the geographic area of interest, often calculated using historical data and empirical models.

The drainage area for the site is critical in determining the water management strategies necessary for operations and maintenance of the facilities.


Question: What are the reporting requirements for diesel generator failures in nuclear facilities?

If an individual diesel generator (DG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that DG in that time period shall be reported within 30 days. Reports on DG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.7.2.12, Steam Generator (SG) Program.


Question: What are the safety measures associated with Davis-Besse Nuclear Power Station?

The Davis-Besse Nuclear Power Station is equipped with various safety systems, including the Emergency Core Cooling System (ECCS), which is designed to operate in the event of a loss-of-coolant accident, ensuring the reactor core remains adequately cooled.

The construction of the Davis-Besse Nuclear Power Station began in 1970, with its first commercial operation commencing in 1977, contributing to the energy supply needs of the region.


Question: The NRC mandates that licensees must assess and prepare for potential emergencies by implementing effective action levels.

South Texas Project, Units 1 and 2, Docket Nos. STN 50-498; STN 50-499; Response to Request for Additional Information - South Texas Project (STP), Units 1 and 2 License Amendment Request for Emergency Action Level Scheme Change (TACs MD4195 and MF4196)

The audit conducted on 03/03/2020 assessed the quality of emergency preparedness at the South Texas Project.


Question: topographic quadrangles in environmental reviews

The USGS Hillsborough 7.5-minute topographic quadrangles provide detailed representation of the project location, including features such as contour lines, bodies of water, and roads, which are crucial for environmental assessments.

The project footprint is crucial for understanding the overall scope of the regulatory submission and locating key features within the proposed site.


Question: The potential for accidental liquid releases to impact groundwater sources is evaluated in this study.

This section provides a conservative analysis of a postulated, accidental liquid release of effluents to the groundwater at the VEGP site.

There are two effluent holdup tanks, each with a capacity of 28,000 gal., for each AP1000 unit.


Question: design heat load cooling water system

Clinton Lake was designed to provide cooling water to the CPS and remove the design heat load from the circulating water before the water circulates back into the plant.

A rank-order method was used to analyze low-flow frequency for the Rowell gauging station under postdam conditions.


Question: nuclide MDA report


Question: The NRC mandates operators to manage the reactor coolant system pressure effectively during emergencies.

Safety Injection (SI) was initiated successfully on both trains.

The crew has just completed the RCS cooldown to the target temperature and the steam dumps are in Auto in MS Mode.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves multiple stages, including pre-application, application review, and post-grant activities, to ensure compliance with safety regulations and operational standards set by the NRC.

The design and construction of the reactor must also consider factors such as site characteristics, natural hazards, and suitability for intended operations to address potential safety concerns.


Question: NRC audit process

The NRC's licensing project manager informed licensee staff by telephone on February 29, 2024, that the NRC staff had completed its audit.

There were no open items resulting from the audit.


Question: environmental impact statement requirements

The National Environmental Policy Act, however, obligates every federal agency to prepare an adequate environmental impact statement before taking any major action, which includes issuing a uranium mining license.

Certainly it would be preferable for the FEIS to contain all relevant information and the record of decision to be complete and adequate before the license is issued.


Question: emergency core cooling system

Two Emergency Core Cooling System (ECCS) trains shall be operable

One or more reactor building air locks inoperable for reasons other than Condition A or B, restore within 24 hours


Question: The NRC defines the minimum detectable concentration as the smallest amount of a substance that can reliably be identified in a sample.

The results were collected from a batch method that used specific detection limits to ensure accuracy.


Question: What does the term 'defective area in weld metal' refer to in nuclear component assessments?

10X photograph of area of sample 4B, showing that the defective area is entirely in weld metal.

Photo #39 is used to document the condition of the weld without detailing specific defects.


Question: nuclear safety analysis reports

Nuclear safety analysis reports are essential documents that evaluate the potential risks associated with nuclear reactor operations and provide necessary safety measures to ensure public health and environmental protection.

The software used in calculating the thermal efficiency of power plants is often based on complex algorithms that require extensive data inputs and can simulate various operating conditions.


Question: The NRC assesses the integrity of components exposed to reactor coolant to prevent cracking.

Cracking of steel with stainless steel cladding, stainless steel, and nickel alloy PZR components, as well as nickel alloy RV core support lugs, exposed to reactor coolant is managed by the DCPP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD AMP (B.2.3.1) and the DCPP Water Chemistry AMP (B.2.3.2).

The NRC evaluates various designs to ensure that aging management programs are effective and comply with safety standards.


Question: nuclear regulatory process EMDAP

EMDAP (Evaluation Model Development and Application Process) is a structured method for the development and application of evaluation models that assess the safety and performance of nuclear facility systems and components, ensuring compliance with regulatory requirements.

The development of the assessment base requires a thorough understanding of the modeling capabilities, algorithms employed, and the necessary data to support the evaluation process.


Question: What is the range for pressurizer pressure in nuclear reactors?

The range for pressurizer pressure (psia) is 600 to 2400.


Question: friction factor in nuclear engineering

A typical closure relationship is the use of a friction factor to approximate the irreversible pressure losses that occur as fluid moves through a pipe—where the magnitude of the friction factor is a function of the roughness of the pipe wall and whether the flow is laminar, in transition from laminar to turbulent, or turbulent.

The governing equations, numerics, and integrated performance of the EM must be examined to ensure proper evaluation of safety measures in nuclear reactors.


Question: The NRC stipulates requirements for analyzing the decay constants of various isotopes for safety evaluations.

The decay constant is a parameter that signifies the rate at which a particular isotope will decay over time.

The data shows that L-132 has a decay constant of 5.5E-07 sec.


Question: What are the acceptance criteria for emergency core cooling systems in nuclear reactors?

Commission Decision on Rulemaking for Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors states: The principal changes to the AEC’s regulations were to lower the maximum allowed fuel cladding temperature, in the event of a loss-of-coolant accident (“LOCA”), from 2300°F to 2200°F, and to add a 17% local cladding oxidation limit.

The primary participants included the Commission Regulatory Staff, four reactor manufactures, a consolidated group of electric utility companies, and the Consolidated National Intervenors (“CNI”), a group of about 60 organizations and individuals.


Question: What are the effluent limitations for thermal discharges in nuclear plants?

In accordance with the Federal Water Pollution Control Act, this permit may be re-opened to insert a more restrictive thermal limit or the requirement to conduct a 316(a) study if it has been shown that the thermal component(s) of the surface water discharges affect the safety and propagation of a balanced, indigenous population of shellfish, fish, and wildlife in and on Mississippi River.

The permit application shall include analytical data for monitoring stations SD001, SD003, and SD004 for parameters in accordance with Minnesota Rules 7001.1060 and 7001.1061 and instructions of the applicable EPA Permit Application Form 2C.


Question: potential ecological impacts during aquifer restoration

Potential ecological impacts during aquifer restoration will be SMALL, since surface disturbing activities will be limited. Potential impacts associated with aquifer restoration activities will include vegetation and habitat alteration due to the response and cleanup of potential spills, noxious weeds, wildlife mortality from vehicle collisions and displacement due to noise, dust, and human/mechanical presence. Potential impacts are expected to be less than during operation due to a smaller workforce and decreased traffic.

The No Action Alternative will result in no change to existing ecological conditions at the project areas or in the region. Land disturbance will be avoided and the area will continue to provide low to moderate quality vegetation communities and wildlife habitat typical of the region.


Question: The NRC requires adequate provisions for safe shutdown systems to minimize risk during emergencies.

the equipment needed for safe shutdown is a potential means of improving the reliability of operator actions in MCR abandonment scenarios.

alternate shutdown panel were enhanced to include at least one complete division of safe shutdown equipment controls.


Question: What does the document state about the requirements for Management Reviews?

100-QC-004 defines requirements for Management Review inputs, outputs, and associated records.

The audit checklist reviewed locations across various states for their management performance metrics and compliance.


Question: The enrichment level of uranium in a fuel assembly must meet specific regulatory criteria to ensure safety and efficiency.

Minimum Initial Assembly Avg. Enrichment wt % 235U (E)


Question: The activity measured in pCi per grams indicates the concentration of radioisotopes present in a sample.

Activity (pCi/grams) is a measure of the concentration of a specific nuclide in a sample, indicating how much radioactive decay is expected from that material.

The peak size is determined by the count rate over a specified interval, reflecting the strength of the signal detected.


Question: reactor coolant iodine activity

C. Reactor coolant iodine activity prior to accident: An assumed preaccident iodine spike, which has resulted in the DE of 60 µCi/g of I-131 in the reactor coolant.

I. Secondary coolant mass, 4 generators (g): 1.9 x 10^8.


Question: What properties of molten salts are crucial for MSR licensing?

Consequently, the molten salts’ fluid thermophysical properties (density, viscosity, thermal conductivity, and heat capacity) will change and influence the thermal hydraulic behavior and, as a result, the accident progression of the reactor.

Currently, the available molten salt properties data in literature are insufficient for the majority of licensing activities due to the large gaps across the desired operating temperature ranges and poorly understood changes in chemical compositions.


Question: What are the emergency core cooling system flowpaths and their operation?

The applicant is required to have the ability to monitor and manually operate the transfer of ECCS flowpaths.

The NRC emphasizes the importance of emergency preparedness and response measures for nuclear facilities.


Question: The NRC must assess and determine whether the proposed license amendment will meet regulatory requirements outlined in the Atomic Energy Act of 1954.

The proposed amendment would revise the Watts Bar UFSAR to apply alternate POD values to indications of axial ODSCC at TSPs in the Watts Bar, Unit 2, steam generators for the BOC voltage distribution in support of the Watts Bar, Unit 2, operational assessment until the steam generators are replaced.

Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.


Question: nuclear plant environmental impacts

Therefore, TVA finds that the impacts to community services and education from continued operation of BFN during the subsequent period of extended operation are SMALL, which is consistent with the NRC’s conclusion in the 2013 GEIS (Table 4.8-1).

In the 2013 GEIS, the NRC found that transportation impacts associated license renewals are dependent on changes in the workforce, local road network capacity, traffic patterns, and


Question: nuclide energy lines in regulatory filings

@  = nuclide contains energy lines not used in Weighted Mean Activity


Question: The regulated heights for PCT and Burst elevations are critical for understanding fluid dynamics in reactor safety evaluations.

**PCT Elevation** = 7.25 ft
- **Burst Elevation** = 5.75 ft

Fluid quality measurements are taken at various time intervals to assess system performance.


Question: What is the impact of decommissioning on land use for nuclear plants?

The proposed action does not include a change in onsite land use or additional structures. No additional workers whose housing could offsite land use are proposed. Therefore, the land needed for and impacted by decommissioning activities would not be changed by the additional 20 operating years.

There are no additional emission sources expected to be added to the PTN site, and licensees are already required to operate within permit requirements.


Question: What is the process for submitting regulatory filings to the NRC?

The NRC requires licensees to submit a plethora of regulatory documents that include safety evaluations, environmental reports, and financial qualifications as part of their licensing process.

Show device settings include options to adjust video quality, change audio input, and view participant settings.


Question: off-site analysis for radionuclide samples

Table 4-1 provides information on dose significant radionuclides and mixture for soil, which is critical for understanding off-site analysis of contamination levels.

Table 5-2 outlines action levels for Survey Unit L2-011-104, detailing the thresholds that must be met to confidently determine the need for further investigation.


Question: What are stress multipliers for nuclear nozzles?

The NRC staff noted that the multiplier of 2.2 adequately addresses the variation of up to 2.2 times the hoop pressure stress discussed above for the Westinghouse SG MS nozzle.

The NRC has neither reviewed nor approved the EPRI report for generic use.


Question: regulatory process for RAI discussions

The idea of having RAI discussion was first suggested during a meeting with Framatome in Q3/Q4 last year, and Framatome indicated at the time they were in favor of them in order to help resolve NRC concerns and reduce the number of potential limitations and conditions.

I won’t commit to an end date during the drop-in, but I will say that Ngola will reach out to them the first full week of March with a proposed schedule based on their feedback on ending RAI discussions before May or a fast turnaround on draft submittals.


Question: What is a SWPPP in the context of environmental mitigation for nuclear projects?

Prior to construction, NuFuels will prepare and submit a SWPPP to address mitigation measures for erosion and sedimentation.

Erosion and sedimentation will be mitigated by constructing roads and pipelines away from arroyos where possible.


Question: Exclusion Area Boundary

The Exclusion Area Boundary (EAB) Distance is measured in meters and varies depending on the downwind sector, which is defined by cardinal directions. The distance in meters for the EAB is consistently recorded as 975.

The period of record spans from May 1972 to April 1977, during which various environmental data was collected and analyzed to ensure compliance with regulatory standards.


Question: socioeconomic impacts of nuclear projects

Potential socioeconomic impacts related to the proposed project area are discussed in Section 4.10 of this ER and will primarily be positive. The CUP will result in increased revenues for the State of New Mexico, McKinley County, the Navajo Nation, and local businesses. Project implementation also will result in increased employment in the area. In addition to the additional tax base afforded by the project, NuFuels also will implement mitigation measures to enhance the potential for positive socioeconomic impacts.

The reactor design incorporates a passive safety system that operates without the need for external power sources, ensuring safety under all conditions.


Question: What are the mechanisms of fiber penetration through strainers in debris testing?

Fiber can penetrate through the strainer by two different mechanisms: prompt penetration and shedding. Prompt penetration occurs when fiber reaching the strainer travels through the strainer immediately. Shedding occurs when fiber that already accumulated on the strainer migrates through the bed and ultimately travels through the strainer. Both mechanisms were considered during the Wolf Creek fiber penetration testing.

After the introduction of each batch, transportation of fiber from the debris hopper into the tank and to the strainer was verified. Any non-transported fiber was collected, and later dried and quantified. The total non-transported fiber for the entire test amounted to less than 1 gram (vs. total tested fiber load of >13000 grams). Therefore, no fiber settling was credited.


Question: nuclear regulatory process

There are no retention requirements for this section


Question: What are the ring distance measurements for nuclear facilities?

The NRC defines various ring distance measurements as follows: A (0 - 16 kilometers), B (16 - 40 kilometers), C (40 - 60 kilometers), and D (60 - 80 kilometers).

The population data indicates significant residential densities in regions such as SE with 574,975 and N with 342,452.


Question: What are the failure probabilities associated with the isolation valves for the main steam system?

The failure probability for the isolation valve operator for the main steam system train (A) is 3.63E-04 under Above TAF Saturated Steam conditions.

The nominal pipe diameter measurement is used to categorize the welds and isolation valves associated with different systems in the reactor.


Question: Dr. Jason Himmel has undergone comprehensive training for TheraSphere operations and radiation safety protocols.

I believe Dr. Jason Himmel has satisfactorily completed training and work experience in accordance with the U.S. Nuclear Regulatory Commission’s Microsphere Brachytherapy Sources and Devices Guidance (Rev 10.2, April 20, 2021) sections A and B to become an AU for TheraSphere and that this physician is able to independently fulfill the radiation safety-related duties as an AU for TheraSphere.

The clinical use training included three (3) hands-on patient cases for the medical use of TheraSphere under the supervision of an AU for TheraSphere on the following dates: 10/27/2024; 11/10/2024; 11/27/2024.


Question: nuclear sampling system

To demonstrate the capability of the nuclear sampling system, liquid and gaseous, to collect representative samples from the RCS and auxiliary system process streams.

Construction acceptance testing is complete.


Question: The regulation of nuclear safety is of utmost importance in maintaining public trust and protecting the environment.

Nuclear safety regulations encompass a series of guidelines and requirements that licensees must follow to ensure the safe operation of nuclear facilities and the protection of public health and the environment.

The publication highlights various historical agreements and decisions made during the early years of the Atomic Energy Commission, focusing on administrative structures and policy formulation.


Question: The NRC mandates that licensees provide accurate thermal emissivity values for materials when calculating convective heat transfer coefficients.

The coefficients so obtained are small and are about what one would expect from the mechanisms of natural convection and radiation to steam.

The values of the calculated convective heat transfer coefficients depend to some extent upon the value used for the thermal emissivity of stainless steel, since the convective heat transfer is obtained after subtracting the radiative heat transfer from the total.


Question: U.S. Nuclear Regulatory Commission definition

Connecticut needs to delete the statement indicated above, and correct the definition of 'Commission' to read, 'U.S. Nuclear Regulatory Commission.'

Connecticut states that the definitions are not included in the proposed legislation but will be adopted in regulation. However, this definition was omitted from the list.


Question: The NRC plans to engage with licensee personnel during the virtual audit to clarify any questions regarding the reviewed items.

The audit will be led by staff from the NRR, Division of Operating Reactor Licensing (DORL).

The audit activities will be performed using Microsoft Teams and other virtual meeting spaces established by the licensee.


Question: The NRC evaluates aging management programs for spent fuel pool cooling systems to ensure their effectiveness in maintaining safety standards.

The aging management program for Heat Exchanger (Spent Fuel Pit) Tubes includes monitoring the reduction of heat transfer due to aging effects that may require management.

The New and Spent Fuel Storage Racks are designed to provide structural support for the storage of both new and spent fuel within the plant.


Question: RICT calculations

Examples of calculated RICTs are provided in Table E1-2 for each individual condition to which the RICT applies (assuming no other SSCs modeled in the PRA are unavailable). Following TSTF-505 implementation, the actual RICT values will be calculated using the actual plant configuration and the current revision of the PRA models representing the as-built, as-operated condition of the plant, as required by NEI 06-09, Revision 0-A and the NRC safety evaluation, and may differ from the RICTs presented in Table E1-2.

The RICT Incremental Conditional Core Damage Probability (ICCDP) limit is 1.00E-06, while the RICT ICLERP limit is 1.00E-06.


Question: 6.9kV shutdown boards

Loads designed for BO auto sequence start will require performance of Section 3.3 to allow loads to be restarted when board is energized.

Shift Manager agrees board can be energized.


Question: What are invasive species and their impact on ecosystems?

Invasive species are nonnative organisms whose introduction causes or is likely to cause economic or environmental harm or harm to human, animal, or plant health (81 FR 88609-TN8375).

The Keowee-Toxaway State Park lies at the north end of Lake Keowee. It was established through a partnership between Duke Energy and the State of South Carolina and includes 1,000 ac (400 ha) open to camping, fishing, boating, and other recreational amenities.


Question: The NRC's radiation safety assessments often involve analyzing data collected over time, including background radiation measurements.

Env. Background File: C:\Canberra\Apex\Root\Dairyland_NPP\Data\0000001364.CNF

The results indicate that the peak areas for energy levels from 76.14 keV to 1461.75 keV show variations that need to be explored further.


Question: The NRC mandates that the system be ready for the Main Turbine Valve testing procedures under the specified conditions.

This JPM requires no malfunctions. The initial conditions support the required power level to perform Main Turbine Valve testing procedure.


Question: The NRC mandates that each licensee must effectively manage scenarios such as a Small Break LOCA to ensure core cooling adequacy.

Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling. (CFR: 43.5 / 45.13)

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 / 45.3)


Question: The applicant is responsible for calculating acceptable dose limits based on their submissions.

The proposed "dose limit"—really ANY dose limit—is a calculation done by the applicant using unverified assumptions to claim the site will be acceptable.

This proposal is even more bold and expansive than the many past efforts by the nuclear industry and its "regulators" which Congress revoked in the 1992 and the public has repeatedly rejected over the decades of the nuclear age.


Question: What happens to the Safety Injection Pump suction flowpath during a Large Break LOCA?

At 12:10, the SAFETY INJECTION Pump SUCTION flowpath (2) ___ AUTOMATICALLY TRANSFER to the RHR DISCHARGE.

Which ONE of the following completes the statements in accordance with 1-ES-1.3, Transfer to Containment Sump?


Question: The analysis performed to ensure that NPSH margins are adequate reflects the methodologies outlined in regulatory guidance.

The void fractions shown in Table 3.f.14-3 were used to adjust NPSH required (NPSHr) per the guidance provided in RG 1.82 Appendix A when calculating the NPSH margin in the Response to 3.g.16.

The approach to the degasification analysis and correction to the pump NPSHr described above is consistent with the method used by Salem Nuclear Generating Station in their GL 2004-02 response.


Question: The reactor mode switch is critically important for initiating control actions within the reactor system.

The reactor mode switch provides input to the PPS and therefore functions initiated by this switch appear to rely on PPS functionality.

The analysis in Section 3.23, "Rod Withdrawal Error," discusses various operational parameters that are essential for ensuring reactor safety.


Question: What is the importance of chemical spray during design basis event testing?

The chemical spray or demineralized water spray during design basis event (DBE) testing needs to be conservatively injected after the peak of the environmental profiles (temperature, pressure).

Information presented regarding aging may be better suited to be with the aging details presently in Clause 7.4.1.9.3 (Age Conditioning).


Question: What is meant by 'activity' in the context of isotopes?

Activity is a measure of the decay rate of a radioactive isotope and is usually expressed in units like millicuries (mCi).

The distributor license number is 29-30342-02G, and it is associated with Metorex Incorporated.


Question: What is the significance of Tritium in nuclear safety assessments?

Tritium is a radioactive isotope of hydrogen that is produced in nuclear reactors and is often measured in regulatory filings related to nuclear safety and environmental monitoring.

The pH level of a sample is an important parameter in analyzing water quality, which can affect the behavior of various contaminants.


Question: What is the significance of the Rosemont shear zone in geological assessments for nuclear sites?

The Rosemont shear zone is mapped as a northeast striking fault, interpreted variably as a dextral shear zone or high-angle thrust fault (References 2.5.3-32 and 2.5.3-36).

Several faults have been mapped within the site vicinity that offset basement or overlying units, but are presumably not active.


Question: The eruption of calderas can generate extensive pyroclastic flows that vary in volume and impact.

Although caldera-forming eruptions have not been observed, it is generally agreed (e.g., Wilson, 2008) that such large-volume pyroclastic flows originate from multiple vents located primarily along the caldera margins.

The Yellowstone Volcanic Field (YVF) is the largest Quaternary volcanic system in the site region, having undergone three major caldera-forming eruptions (VEI7–8) since 2.1 Ma that produced more than 3,700 km³ of magma as pyroclastic deposits (Christiansen, 2001).


Question: The NRC evaluates the potential for critical heat flux in nuclear reactors to ensure safety standards are met.

COBRA-FLX is a critical heat flux correlation tool used for evaluating thermal-hydraulic behavior in nuclear reactor systems.

The guidelines outline the necessary documentation for various aspects of nuclear power plant operations, including maintenance schedules and inspection protocols.


Question: What is the significance of Strontium-90 in regulatory analyses?

Strontium-90 is a radioactive isotope that is commonly analyzed in environmental samples, as its presence can indicate contamination and thus has regulatory implications in sectors such as nuclear safety and environmental monitoring.

The project includes various methods and descriptions for sample preparation and analytical procedures to ensure accurate data collection and reporting.


Question: What are the considerations in developing a cost estimate for regulatory actions?

When assigned the task of developing a cost-benefit estimate, the cost analyst should gather general project information, including the following: project background, project scope, pertinent contract or subcontract information, if applicable, estimate purpose, classification, how the estimate will be used, and techniques anticipated, project schedule.

Numerous sources are available from trade associations and governmental organizations.