Question: activity of nuclides

The activity of a nuclide, measured in picocuries per gram (pCi/grams), indicates the rate at which a nuclide decays and emits radiation.

The energy of EU-154 at 344.28 keV reflects its decay characteristics and emission profile.


Question: Baseline Security Significance Determination Process

The Baseline Security Significance Determination Process (BSSDP) is a method used to assess the security-related significance of findings identified during inspections.

The next agenda item will involve discussions on the effectiveness of communication strategies used during emergency situations.


Question: What is the expected timeline for receiving the Technical Evaluation Report (TER) from the NRC?

The Technical Evaluation Report (TER) is an essential document that summarizes the NRC staff's findings based on their review of a licensee's submittal and assists in the decision-making process for regulatory actions.

We need to ensure that all submissions are complete and correctly formatted to avoid delays in the review process.


Question: Components categorized under RISC-3 must undergo regular inspection and evaluation to ensure they meet safety standards.

10 CFR 50.69 requires RISC-3 components to have: Inspection and testing, Corrective action, Feedback and Process Adjustment (at least every 2 cycles).

Periodic reviews consider plant modifications, operational or regulatory changes, and updated risk information for decision-making.


Question: The identification and quantification of specific nuclides is critical for accurate radiological assessment.

Nuclide confidence index threshold = 0.30

Activity (pCi/grams) values are measured to determine the concentration of specific isotopes in the sample.


Question: weighted mean activity of nuclides

Weighted Mean Activity is measured in picocuries per gram (pCi/grams) and represents the average radioactivity of a specific nuclide per unit weight, accounting for various isotopes and factors affecting their presence.

Errors quoted at 2.00sigma indicate the level of confidence in the measurements reported for the nuclides within the study.


Question: What is leakage in the context of nuclear reactor operations?

LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

The ESF RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).


Question: What is the COBRA-FLX ORFEO-HMP Critical Heat Flux Correlation?

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation is a mathematical relationship used to predict critical heat flux in nuclear reactors, which is essential for ensuring safe operational limits during reactor operation.

Figure 7-9 depicts the relationship between M/P and equilibrium quality for the ORFEO-HMP correlation, illustrating the combined data set utilized for analysis.


Question: What is a site license for microreactors?

But one way around would be require the manufacturing facility to have a site license. I think that would cover it.

And to Steve's point, this probably is a niche, right, for SMRs that are going to be routinely refueled, whether that would be anything that would be done in the factory versus done at the site since you're planning to do refuelings anyway.


Question: Licensees must comply with regulatory requirements pertaining to their operational safety protocols.

The NRC assesses nuclear license applications for compliance with established safety standards.

The energy market is influenced by various factors including supply chain dynamics and weather conditions.


Question: what is gamma spectrum analysis

Gamma spectrum analysis is performed to determine the energies and intensities of gamma radiation emitted by radioactive samples, which is essential for identifying the isotopes present in the sample.

The sample was taken on 6/13/2019 at 9:49:00 AM, and the procedure utilized was a 500ml Marinelli geometry which optimized the detection efficiency for the analysis.


Question: The NRC mandates that operators must ensure each automatic feedwater valve responds correctly to the actuation signals, whether real or simulated.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

Note: Not required to be performed for the turbine driven AFW pump until 24 hours after ≥ 900 psig in the steam generator.


Question: What is the purpose of an ISFSI in nuclear regulations?

The Independent Spent Fuel Storage Installation (ISFSI) is a facility for the safe storage of spent nuclear fuel and other high-level radioactive waste, ensuring it is managed in accordance with regulatory requirements to protect public health and safety.

The Docket Number refers to the specific identification assigned to a license application or regulatory submission, which helps in tracking the review process and maintaining organized records for public inspection.


Question: The NRC seeks information regarding the technical specifications and processes involved in reactor licensing.

The NRC requires a comprehensive evaluation of the safety systems for nuclear reactors during the licensing process to ensure compliance with regulatory safety standards.

The report details the recommendations for improving communication strategies among regulatory bodies and stakeholders in various industries.


Question: What are the compaction specifications for nuclear facility construction?

When the moisture content and other conditions in any layer are satisfactory, that layer is compacted by the necessary number of passes of approved rollers required to attain not less than 95 percent of the Standard Proctor Density, ASTM D698-66T, Method C.

Logs of core borings in dam foundation are furnished in Figure 6.3 of Attachment 2G-A of Former Appendix 2G.


Question: ASME Code Class 2 components

The SG MS outlet NSWS and SG FW NSWs and NIR sections at Vogtle are ASME Code Class 2 components, whose ISIs are performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable edition and addenda, as required by 10 CFR 50.55a(g).

The component numbers uniquely identify the various elements involved in the examination categories and their respective welds.


Question: The NRC's protocol involves the estimation of various debris types' physical characteristics to assess their impact on transportability and head loss during accidents.

The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

A summary of the material properties of the debris types found within containment are listed in Table 3.c.1-1. See Response to 3.d.3 for the material properties of latent debris.


Question: what is the meaning of RCRA?

RCRA stands for Resource Conservation and Recovery Act, which governs the disposal of solid and hazardous waste.

The State Historic Preservation Officer oversees the preservation of historic sites and coordinates with federal agencies to protect cultural resources.


Question: severe accident analysis

The applicant provided information on a severe accident analysis and provided an update on constructing a meteorological tower for collecting site-specific data.

The NRC staff provided an overview of the process utilized for determining whether an EA or an environmental impact statement (EIS) is required and stated that the evaluation of an EA or EIS for the LME project has yet to be determined and will be evaluated during the acceptance review.


Question: What does the number of indications represent in nuclear safety assessments?

The number of indications refers to the count of detected anomalies or potential issues in the tubing of a steam generator, which are critical for assessing the integrity and safety of the reactor's systems.

The maximum volts recorded in the steam generator are important for understanding the electrical characteristics of the system and ensuring that they operate within safe limits.


Question: The NRC must ensure compliance with water quality regulations and consider the consequences of hydrologic alterations.

Compliance with applicable water quality and water sustainability regulations is essential for the management of water resources in relation to hydrologic alterations.

Global climate changes are projected over this century and beyond with the magnitude of changes after the next few decades dependent primarily on global emissions of heat-trapping gases and the sensitivity of the Earth’s climate to these emissions.


Question: fuel transfer system

The fuel transfer system includes an above-water electric-motor driven transfer car that runs on tracks extending from the refueling canal through the transfer tube and into the spent fuel pool and an upender lifting frame at each end of the transfer tube.

The new fuel elevator consists of a box-shaped elevator assembly with its top end open and sized to house one fuel assembly.


Question: sublethal effects of thermal effluents

Fish and shellfish that are exposed to the thermal effluent of a nuclear power plant may experience stunning, disorientation, or injury. These sublethal effects can subsequently affect an organism’s susceptibility to predation, parasitism, or disease.

The EPA’s 2014 CWA Section 316(b) regulations establish BTA standards for impingement mortality.


Question: Loss-of-Coolant Accident

A Loss-of-Coolant Accident (LOCA) occurs when there is a sudden drop in the reactor coolant inventory, which can lead to core overheating if not mitigated effectively.

The reactor operator reported that the system's voltage levels were within acceptable ranges and did not require any immediate adjustments.


Question: What does 'dry weight' mean in the context of laboratory analysis?

The samples in this SDG were analyzed on a "dry weight" basis.

There are no exceptions, anomalies or deviations from the specified methods.


Question: What is Very Low-Level Waste (VLLW) in nuclear regulations?

Very Low-Level Waste (VLLW) consists of materials that are only lightly contaminated with radioactivity and can include items like protective clothing, tools, or other disposable items used in the operation of a nuclear facility.

The current proposal to reinterpret VLLW rules raises serious concerns about public safety and environmental protection, as it could lead to nuclear waste being disposed of in regular trash or industrial sites.


Question: what is vortexing evaluation in nuclear safety?

Testing was conducted to determine if vortexing is expected to occur. As discussed in the Response to 3.f.3, the vortex tests were performed at both clean strainer and debris-laden conditions.

A flashing failure was recorded for a postulated break if, at any time during sump recirculation, the pressure downstream of the strainer was lower than the vapor pressure at the sump temperature.


Question: What is the significance of nuclide names in regulatory documents?

| Nuclide Name | Energy (keV) | Yield (%) | Activity (µC/grams) | Nuclide MDA (µC/grams) | Line MDA (µC/grams) |

The NRC requires licensees to conduct thorough safety evaluations for the use of radioactive material.


Question: What are the permits required for construction and operation of nuclear facilities?

Permits and approvals required for construction and operation include a range of regulatory approvals such as site permits, environmental assessments, and operational licenses.

Consultations required for construction and operation involve discussions with federal and state agencies, stakeholders, and public interest groups to ensure compliance and safety.


Question: The NRC emphasizes the need for licensees to evaluate potential simultaneous failures in safety systems.

The NRC defines concurrent failures as events where two or more systems fail at the same time, which can significantly impact safety and reliability.

The training for emergency response personnel includes regular drills and exercises to ensure readiness for actual events.


Question: What are the environmental impacts considered in the licensing process for nuclear power stations?

The environmental impacts associated with the operation of the Dresden Nuclear Power Station Units 2 and 3 have been assessed as part of the regulatory process for subsequent license renewal.

The annual report provides detailed data regarding the radioactive effluent released from the Dresden Nuclear Power Station, including statistical analysis and comparison with previous years.


Question: cold leg break transient

The Reactor Coolant System behavior is different for cold leg breaks and the peak Containment pressure occurs later for cold leg breaks.

Three double-ended guillotine pipe break locations are examined; hot leg, cold leg pump suction, and cold leg pump discharge.


Question: What are bedrock aquifers and their importance in water supply?

Bedrock aquifers within the 50 mi radius of the site are shown in Figure 2-4-22. Most of the glacial drift in the study area is underlain by Pennsylvanian bedrock that consists largely of shale and siltstone interbedded with limestone, sandstone, underclay, and coal.

Groundwater base flow averages 0.13 cfs/mi² for years that have below-normal precipitation and 0.58 cfs/mi² for years that have above-normal precipitation.


Question: cultural resources mitigation measures

Mitigation measures to address potential land use impacts during operation and aquifer restoration include: Implementing cultural resources mitigation measures, as described above, for any surface disturbance that occurs during operation or aquifer restoration.

Access roads and roads within the project areas will be removed and reclaimed unless exempted from reclamation by the request of lessees/allottees, in which case the lessees/allottees will assume responsibility for long-term maintenance.


Question: joint systems in excavation stability

Joint Set No. 1, which dips at an angle of about 80 degrees, is one of the most abundant sets observed across the site. Another steeply dipping set, number 3, although observed only in rock outcrops near the site and believed to be widely spaced, could conceivably be found near the proposed excavations.

We believe the shear strength of the relatively fresh rock along the planes of foliation is suitable to support even a vertical cut.


Question: Licensees are required to provide specific information about locations where radioactive materials will be used or stored.

An acceptable location of use or possession specifies street address, city, State, and zip code and does not include a post office box number.

To conduct operations at temporary job sites (i.e., locations where work is conducted for limited periods of time), the address may be stated as “temporary job sites anywhere in the U.S. where the NRC maintains jurisdiction.”


Question: nuclear licensing and regulatory filings

The licensing of nuclear facilities is a complex process that involves the submission of detailed safety analyses, operational procedures, and compliance verification to regulatory bodies, ensuring that the nuclear plant operates within defined safety margins.

Figures C.5-36 and C.5-37 represent data trends and are not included in the document as they do not contribute to the current analysis.


Question: flood barrier construction and materials

Flood Barrier: Intended Function: Flood Barrier; Material: Reinforced concrete; Environment: Air – Indoor, Uncontrolled; Aging Effect Requiring Management: Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling)

Missile Barrier: Intended Function: Missile Barrier; Material: Not specified; Environment: Not specified; Aging Effect Requiring Management: Increase in Porosity and Permeability, Cracking, Loss of Material (Spalling, Scaling)


Question: The NRC assesses whether proposed exemptions have a negligible impact on public safety and environmental concerns.

The proposed exemption must not endanger life, property, or the common defense and security as outlined in regulatory guidelines.

The Technical Assessment provides an overview of the methods used to evaluate the proposed exemption, including various analytical techniques.


Question: What is the purpose of a Final Environmental Impact Statement for nuclear plants?

The NRC staff has completed the environmental impact statement NUREG–1437, Supplement 2, Second renewal, "Site-Specific Environmental Impact Statement (EIS) for License Renewal of Nuclear Plants Regarding Subsequent License Renewal for Oconee Nuclear Station Units 1, 2, and 3, Final Report" (FEIS).

The NRC staff initiated consultation with you regarding this project in August 2021.


Question: wind speed measurements in nuclear safety assessments

The data presented includes a joint frequency distribution of wind speed and direction, categorized by atmospheric stability class, which is critical for evaluating meteorological conditions during safety assessments at the Salem/Hope Creek facility.

The reactor containment structure is designed to withstand various environmental loads, ensuring that safety systems remain functional during extreme conditions.


Question: nuclear licensing requirements

The Nuclear Regulatory Commission (NRC) requires comprehensive documentation and submissions for license applications, including safety analysis reports, environmental reviews, and plans for mitigating potential hazards.

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Question: What are the relevant simulation parameters for TSC calculations?


Question: The process of obtaining a nuclear license involves rigorous assessments to ensure compliance with safety standards.

The licensing process for nuclear reactors includes comprehensive safety evaluations mandated by the Nuclear Regulatory Commission (NRC).

Emergency preparedness regulations require that all facilities establish response plans for unforeseen natural disasters.


Question: What constitutes an SRO-only question in nuclear safety procedures?

Does the question require one or more of the following: assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed?

Can the question be answered solely by knowing immediate operator actions?


Question: What is a control rod scram accumulator in the context of nuclear reactor safety?

Two or more control rod scram accumulators inoperable with reactor steam dome pressure ≥ 900 psig.

Verify all control rods associated with inoperable accumulators are fully inserted.


Question: An ambient noise assessment was performed to evaluate the sound levels around the HCS site.

An ambient noise survey was conducted in January 2022 to characterize the baseline acoustical conditions at the HCS and property boundaries.

The Horizon Center Industrial Park is partially developed, with businesses closest to the HCS consisting of Philotechnics, an industrial facility specializing in radioactive waste repackaging, and the headquarters office of the Community Reuse Organization of East Tennessee (CROET) (Figure 3.7.1-1).


Question: The Commission has determined that the plant can operate safely under the amendments proposed by Indiana Michigan Power Company.

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission’s regulations;

A. The application for amendment by Indiana Michigan Power Company dated April 30, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations set forth in 10 CFR Chapter I;


Question: The NRC mandates that licensees evaluate the clarity of initiating cues to ensure proper operator understanding of a task.

Check each JPM for appropriate administrative topic requirements (COO, EC, Rad, and EP) or safety function requirements and corresponding K/A. Mark in column 1. (ES-301, D.3 and D.4)

For column 6, provide a brief description of any (U)acceptable or (E)nhancement rating from column 5.


Question: impact of battery charger malfunction on DC Electrical Distribution System

K/A is matched because the applicant is required to predict the impacts of a battery charger malfunction on the DC Electrical Distribution System and based on those predictions, use procedures to mitigate the consequences of those abnormal operations.

Given a set of plant conditions, DETERMINE the following: The bases for the required LCO or TR actions.


Question: What is configuration control in nuclear regulatory practice?

Configuration control is a process that ensures changes to the facility or design are thoroughly reviewed and documented, maintaining the integrity and safety of the nuclear plant operations.

The document also describes the roles of various preparers and contributors involved in the approval process, highlighting the importance of proper documentation.


Question: Could you provide information on the examination plans for components governed by the ASME Code during the upcoming outage?

A detailed schedule (including preliminary dates) of: 1.1 Nondestructive examinations planned for ASME Code Class Components performed as part of your ASME Section XI, risk informed (if applicable), and augmented inservice inspection programs during the upcoming outage.

Our inspection dates are subject to change based on your updated schedule of outage activities.


Question: The NRC has guidelines to assess the eruptive hazards within a designated radius of planned construction sites.

Following guidance in Regulatory Guide 4.26, a tectono-magmatic model can be useful in evaluating the large-scale processes that led to the occurrence of Quaternary volcanism at specific locations in the site region.

The bedrock geology is composed of folded, faulted, and deformed sedimentary rocks of Eocene age or older (i.e., >34 Ma) (Dover and McGonigle, 1993; Dover, 1995; M’Gonigle and Dover, 2004; Wittke and Heffern).


Question: How does core stored energy impact containment integrity during a LOCA?

The higher core stored energy impacts the calculated peak containment pressure, temperature, and containment recirculation sumps water temperature following a LOCA.

The pressure increase can be conservatively accounted for by managing the impact of the increased energy deposition into the time period between the end of blowdown and the time of reflow.


Question: surface water usage summary

Table 3.6-5 Surface Water Usage Summary in MGD, 2015 provides detailed data on the water usage in various categories like public supply, domestic self-supplied, industrial self-supplied, and power generation.

The irrigation category reported usage of 0.13 MGD in Wright County and 0.08 MGD in Sherburne County.


Question: Kairos Power is requesting confidentiality for certain proprietary information to prevent competitive disadvantage during the licensing process.

I have personal knowledge of the criteria and procedures utilized by Kairos in designating information as a trade secret, privileged, or as confidential commercial or financial information.

The licensing process for nuclear reactors requires extensive documentation to ensure safety and compliance with regulatory standards.


Question: UO2 fuel type in nuclear reactors


Question: hydrogen purge containment tanks

POST-DBA HYDROGEN PURGE OF CONTAINMENT TANKS (time = 0 is start of purge)

The table shows the total activity levels of various iodine isotopes over designated time intervals.


Question: The NRC needs clarity on the classifications of Actuations and Protective signals during incidents.

Please provide information to clarify which Actuations and Protective signals are primary and which are secondary and describe what the relevant distinctions between primary and secondary functions are.

The licensee's response will address the items listed in Table 4-1, ensuring all necessary details are communicated effectively.


Question: regulatory audit requirements for nuclear licensing

Specifically, 10 CFR 50.34(a)(4) and (b)(4) require that applicants for a construction permit (CP) and an operating license (OL), respectively, provide analysis and evaluation of the design and performance of structures, systems, and components (SSCs) of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.

The sponsor states that information that will be made available during the audit includes V&V documents, code manuals, code base decks, modeling details, analyses, and special methods required by analyses.


Question: What are the surveillance requirements for the HPCI system in nuclear reactors?

For the HPCI System, not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. Vessel injection/spray may be excluded. Verify each ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

In accordance with the Surveillance Frequency Control Program.


Question: What is total body dose in radiation exposure?

Total body dose is the same for various organs, e.g., gastro-intestinal tract, bone, liver, kidney, thyroid, and lung.

The GASPAR code calculates individual doses for various pathways. These doses are totals for the plume, ground, and inhalation pathways.


Question: The Temperature Average (Tavg) required for accurate rod drop time measurements is at least 551°F.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses.

The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality.


Question: weld overlay evaluation

The axial length and end slope of the weld overlay will cover the weld and heat-affected zones on each side of the weld, as well as any stress corrosion cracking susceptible base material adjacent to the weld and provide for load redistribution from the item into the weld overlay and back into the item without violating applicable stress limits of NB-3200.

The effects of any changes in applied loads, as a result of weld shrinkage from the entire overlay, on other items in the piping system (e.g., support loads and clearances, nozzle loads, and changes in system flexibility and weight due to the weld overlay) will be evaluated.


Question: The MAGNASTOR SYSTEM is designed to accommodate various types of PWR and BWR fuel assemblies as well as non-fuel hardware.

UNDAMAGED PWR FUEL ASSEMBLIES, DAMAGED PWR FUEL ASSEMBLIES, PWR FUEL DEBRIS (DAMAGED FUEL), UNDAMAGED BWR FUEL ASSEMBLIES and NONFUEL HARDWARE meeting the limits specified in this section may be stored in the MAGNASTOR SYSTEM.

This section does not cover the specific safety measures required for the handling of radioactive waste.


Question: What is the Ultimate Shutdown System in nuclear reactors?

The primary shutdown system is backed up by an ultimate shutdown system. These control rods use magnetic latches, which can be actuated by either the reactor protection system or automatically when the latch temperature exceeds the magnetic curie point temperature of the latch.

In 2006, GE Hitachi Nuclear Energy (GEH) revitalized the PRISM engineering effort. The revitalized PRISM design differs significantly from the design that was evaluated by the NRC in the 1994 PSER.


Question: What is ASME Code Case N-740-2 in nuclear licensing?

Removal of the 48-hour hold is consistent with the requirements of ASME Code Case N-888-1 and is supported by the white paper that was developed for the proposed change.

The supporting White Paper was previously submitted during the Nine Mile Point, Unit 1 Relief Request ML23089A230.


Question: monitoring smoke containing radioactive materials

SMOKE AND GASES CONTAINING RADIOACTIVE MATERIALS SHOULD BE MONITORED IN THE FIRE AREA TO DETERMINE IF RELEASE TO THE ENVIRONMENT IS WITHIN THE PERMISSIBLE LIMITS OF THE PLANT TECHNICAL SPECIFICATIONS.

Continuous monitoring is provided in appropriate areas throughout the Auxiliary Building to assure safe conditions of radioactivity and to alert operating personnel of any abnormality.


Question: What is the role of the glycol circulating pumps in a nuclear cooling system?

Six glycol circulating pumps are designed to convey the cooled glycol from the ten refrigeration units to the air handling packages (30 per Containment) and to the ice compartment Floor Cooling System of each Containment.

Components requiring periodic maintenance (pumps, heater, control valve) are located in the upper compartment.


Question: nuclear power plant licensing process

The licensing process for nuclear power plants is a regulatory framework established to ensure safe and secure operations of nuclear facilities, including safety assessments, environmental reviews, and public consultations.

The plant’s cooling system is designed to handle peak thermal loads by utilizing a series of pumps and heat exchangers.


Question: The requirement for Fuel Building Exhaust Radiation – Gaseous channels to function properly is crucial during situations involving high radiation levels in the fuel building.

Control Room Ventilation Isolation is also initiated by high radiation in the fuel building detected by Fuel Building Exhaust Radiation – Gaseous channels (GGRE0027 and GGRE0028). The requirements are not repeated in Table 3.7-1. Instead, refer to LCO 3.3.8 for all initiating Functions and requirements.

The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by the unit specific calibration procedures.


Question: What is the process for notifying the MCR in nuclear operations?

Uses three way communication via radio or phone to communicate to the MCR that the local actions of Step 5 of 0-SOI-90.02, Section 8.3.

The performance of the operator is evaluated based on adherence to the established safety protocols and standards as outlined in the regulatory guidelines.


Question: The report highlights the activity levels of various radionuclides in the reactor environment.

Co-60 has an energy of 1173.22 keV and a yield of 100.00%, with a line minimum detectable activity (MDA) of 9.338E+002 pCi/M².

The analysis included extensive data on the operational efficiency of the cooling systems within the facility.


Question: What criteria does the NRC use to evaluate sensitivity thresholds for acceptance margins?

However, the LTR does not define clear criteria to how these determinations are made in accordance with the evaluation model. For figures of merit associated with the non-LOCA topical report, identify the thresholds that determine whether margins to acceptance criteria are sufficient and the thresholds that determine whether figures of merit are insensitive to biases, initial conditions and analysis assumptions.

Modification of initial and boundary conditions in the calculations can generate many cases with similar maximum RCS pressure due to this design feature.


Question: The monitoring of structural integrity is critical in assessing material degradation over time.

Consistent with NUREG-2191. The Structures Monitoring (B.2.1.33) program will be used to manage cracking of reinforced concrete in interior and accessible areas (including drywall floor, reactor pedestal, and reactor shield wall) as well as equipment supports and foundations exposed to air - indoor uncontrolled, air - outdoor, and water - flowing environments in the Cooling Water Structures, Crib Houses, Diesel Generator & HPCI Building, Isolation Condenser Pump House, Main Control Room and Auxiliary Electric Equipment Room, Primary Containment, Radwaste Structures, Reactor Building, Stacks, Switchyard Structures, Turbine Building, and Yard Structures.

The discussion on equipment supports encompasses various structures critical to operational efficiency.


Question: uranium hexafluoride in nuclear fuel production

ConverDyn is one of the largest global providers of uranium hexafluoride (UF6), a crucial component in nuclear fuel production.

NewCo Corp. has not yet been incorporated, but Honeywell expects it to be incorporated as a Delaware corporation.


Question: The NRC reviews proposals to amend technical specifications for the operation of nuclear plants.

The proposed amendment would revise the CNP-2 technical specifications (TSs) to allow a one-time change to permit the current integrated leak rate test interval of 15 years to be extended by approximately 18 months to no later than the plant startup after the fall 2022 refueling outage.

Comments can be submitted electronically or via postal mail, with a deadline for submissions set for February 11, 2021.


Question: The licensee must declare the High Pressure Coolant Injection (HPCI) System inoperable if its initiation capability is lost.

Declare High Pressure Coolant Injection (HPCI) System inoperable.

Only applicable for Functions 1.a, 1.b, 2.a, 2.b, 2.d, and 2.g.


Question: What is the role of debris in the evaluation of ECCS and CSS components?

The objective of the downstream effects, components and systems section is to evaluate the effect of debris carried downstream of the containment sump screen on the function of the ECCS and CSS in terms of potential wear of components and blockage of flow streams.

The verification process involves ensuring that all components meet the regulatory standards set forth by the NRC and that any methodology used is properly documented.


Question: What is a boron recycle system in nuclear reactors?

The boron recycle system is designed to maintain the appropriate concentration of boron in the reactor coolant, which is essential for controlling the reactivity of the core and ensuring safe operation.

The auxiliary building provides the necessary support systems to maintain the overall plant operation efficiency, including water supply and electrical systems.


Question: nuclear reactor licensing process

The process for licensing a nuclear reactor involves multiple phases, including pre-application consultations, submission of the application, and a thorough review by the regulatory authority.

The sampler used for obtaining the soil samples was a Shelby Tube, which allows for minimal disturbance of the soil core.


Question: instrument air compressor package

A 2.03 Instrument air compressor package        | 2.8 | 2.7

Ability to manually operate and monitor in the control room:  (CFR: 41.7 / 45.5 to 45.8)


Question: What is gamma spectrum analysis in nuclear measurements?

Gamma spectrum analysis is performed using a 500ml Marinelli geometry to measure the energy distribution of gamma rays emitted from a radioactive source.

The sample identification includes specific details such as the sample number, type, and the date on which the sample was taken.


Question: cleaning maintenance of nuclear gauges

Any cleaning, maintenance, or repair of the gauge(s) that requires detaching the source or source rod from the gauge shall be performed only by the manufacturer or by other persons specifically licensed by the U.S. Nuclear Regulatory Commission or an Agreement State to perform such services.

The U.S. Nuclear Regulatory Commission's regulations shall govern unless the statements, representations, and procedures in the licensee's application and correspondence are more restrictive than the regulations.


Question: What is Post-Maintenance Testing (PMT) in regulatory filings?

Post-Maintenance Testing (PMT) (IP Section 03.01) involves functional and operational testing of systems following maintenance activities to verify their operability and functionality.

The inspectors evaluated the readiness of the Emergency Preparedness Organization.


Question: The NRC stipulates minimum cooling times based on the average enrichment of fuel assemblies.

The safety analysis must evaluate the potential impacts of various operational scenarios on plant performance.


Question: What are the regulatory processes for relicensing a nuclear plant?

The 2001 letter was for plant relicensing and plant expansion, and focuses on an archaeological site.

We need to know why the towers were determined to be “not eligible” for the NR. Was it because of their ages or did they lack integrity, etc.?


Question: radiation safety procedures

- If personnel dosimetry is provided, do the following:
  - Always wear your assigned dosimeter when using the gauge.
  - Never wear another person’s dosimeter.
  - Never store your dosimeter near a gauge.

- If a gauge becomes damaged, a source becomes dislodged, a source is leaking, the device fails to function properly, or any other nonradiological emergency (e.g., fire or medical emergency) or unusual situation arises, do the following:
  - Stop use of the gauge.
  - Immediately secure the area, and keep people away from the gauge until the situation is resolved.


Question: How is transmitter output current calculated for flow rates in nuclear systems?

The values of 'I' (transmitter output current in mA) that corresponds with the above operating flow rates of interest are determined from above Equation 2.7 as follows:

Flow = 327.24 gpm - Flow Range = 0 to 350 gpm (from Appendix 1)


Question: scram safety definitions

A scram is an emergency shutdown of a nuclear reactor, which is initiated when certain safety parameters are exceeded or conditions warrant an immediate response to prevent an accident.

The dashboard summarizes the current status of electrical systems across multiple years while tracking performance metrics in a linear fashion.


Question: Reactor Protection System instrumentation

Reactor Protection System (RPS) Instrumentation is integral to the safety and automated control of nuclear reactors, ensuring that necessary safety functions can be performed under all operational conditions.


Question: nuclear reactor corrosion management

The Chemical, Corrosion, and Steam Generator Branch (MCCB) is responsible for overseeing the management of corrosion issues in nuclear reactors to ensure safety and compliance with regulatory standards.

The Office of Nuclear Reactor Regulation (NRR) handles various administrative functions, including the review of license renewals and regulatory compliance for operational reactors.


Question: POLCA-T application in boiling water reactor analysis

POLCA-T is a best-estimate computer code for three-dimensional analysis of boiling water reactor transients.

The application methodology for AOO analysis is presented in Reference 2 and Reference 3.


Question: The analysis of local conditions in COBRA-FLX simulations is critical for understanding heat flux behavior.

Local conditions from COBRA-FLX simulations for grid tests are essential for determining critical heat flux characteristics.

The complex designs of modern reactors aim to improve efficiency and safety through advanced technology.


Question: What is the COBRA-FLX model used for in nuclear simulations?

The COBRA-FLX model is used for simulating local conditions in thermal-hydraulic analyses, which include various critical parameters like heat flux correlations under different operating conditions.

Local conditions from COBRA-FLX simulations for grid tests indicate the importance of precise data collection during experimentation.


Question: What are the service conditions for defining allowable stress in nuclear structures?

ASME Section III Service Condition for Definition of Allowable Stress specifies that all primary load bearing members must satisfy Level A stress limits.

Factor of safety against overturning shall be ≥ 1.1, if applicable.


Question: permissible heat loads in nuclear storage

Note that this figure shows the per cell heat load limit for storage.

The permissible aggregate heat load may be less than the sum of each individual cell heat load.


Question: After a reactor trip, the letdown flow should be monitored according to the relevant procedures.

Following the Safety Injection, Letdown Flow __(1)__ be 0 gpm.


Question: minimum detectable concentration

Mr. Jason Q. Spaide is the contact person for the LaCrosseSolutions company.


Question: What is the process for appealing a FOIA response from the NRC?

You may appeal this final determination within 90 calendar days of the date of this response. If you submit an appeal by mail, address it to the FOIA Officer, U.S. Nuclear Regulatory Commission, Mail Stop T-6 A60M, Washington, D.C. 20555-0001.

We are continuing to process your request.


Question: The management of nuclear waste is governed by strict regulatory frameworks that dictate how it should be handled, stored, and disposed of.

The U.S. Nuclear Regulatory Commission (NRC) oversees the safe disposal of nuclear waste, ensuring that licensees comply with regulations for waste management and that environmental protection is prioritized.

The email discusses various regulatory practices related to communication between NRC and external parties, emphasizing the importance of transparency in regulatory processes.


Question: The P-10 interlock is crucial for managing reactor trips by blocking neutron flux signals during power changes.

The LCO requires four channels of Power Range Neutron Flux, P-10 interlock to be OPERABLE in MODE 1 or 2 (two-out-of-four trip logic). The Trip Setpoint is 10% RTP.

The Turbine Impulse Pressure, P-13 interlock is actuated when the pressure in the first stage of the high pressure turbine is greater than 10% of the full power pressure.


Question: Quality Assurance Plan in nuclear license applications

All analyses were performed according to ACZ's Quality Assurance Plan.

This project has been assigned to ACZ's project number, L88949.


Question: The activity of a nuclide is measured in picoCuries per square meter squared (pCi/M²²).

Activity (pCi/M²²) is a critical parameter in assessing the characteristics of a nuclide, reflecting the number of decays per unit time through its yield and energy emission.

Errors quoted at 2.000 sigma indicate the statistical significance of the measured data, helping to determine the reliability of the results.


Question: significant hazards consideration

An evaluation to determine whether or not a significant hazards consideration is involved with the proposed amendment was completed by focusing on the three standards set forth in 10 CFR 50.92(c), “Issuance of amendment,” as discussed below.

Southern Nuclear Operating Company (SNC) is requesting an amendment to Combined License (COL) Nos. NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively.


Question: The nuclear energy sector must adhere to strict regulatory standards when evaluating seismic risks.

The NRC has established requirements for the assessment of site safety and suitability, including the evaluation of potential seismic hazards in order to ensure public safety and the integrity of nuclear facilities.

The USGS monitoring network collects data on groundwater levels and quality across the nation, providing essential information for ecological and environmental research.


Question: What is the purpose of Regulatory Guide 1.174 in risk-informed decisions?

Regulatory Guide 1.174 provides an approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the licensing basis.

The NRC outlines procedures for inspecting and qualifying personnel who perform examinations and tests on nuclear power plants.


Question: station blackout in nuclear regulatory filings

SBO or station blackout is a situation in which a nuclear reactor loses all off-site power and the on-site emergency generators are unavailable or inoperable, potentially leading to a loss of cooling and safety systems.

I would hate to think that someone, okay, well let's just focus on design basis events and we'll throw in a few extra events here just because it's mentioned.


Question: What are the success criteria for core cooling after a containment failure?

Core cooling success following containment failure or venting through non-hard pipe vent paths can be a significant contributor in some PRAs.

There are not significant assumptions or impact on the model. Therefore, the impact of Reactor Building (RB) failure on core cooling is not a source of uncertainty for the ANO-2 base PRA and for applications.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves comprehensive safety reviews, which include assessments of the reactor's design, safety systems, and operational plans to ensure compliance with the regulatory framework established by the commission.

The operational efficiency of a nuclear reactor is vital for electricity production, as it directly influences the cost of energy and supply reliability.


Question: What is the purpose of an Emergency Action Level (EAL) in nuclear facility operations?

The NRC staff has verified that the EAL set is worded in an unambiguous manner that addresses human factors engineering and user-friendliness concerns, is technically complete for each emergency classification level, addresses issues regarding completeness and accuracy raised in Appendix 1 to NUREG-0654, and uses objective and observable values based on site-specific indications.

The licensee chose to modify this EAL set by using a site-specific implementation method that uses a modified numbering format and EAL sequence other than that provided in the generic EAL scheme development guidance.


Question: 10CFR Part 55 compliance

10CFR Part 55 encompasses regulations related to the requirements for licensing operators of production and utilization facilities.

The document also provides guidelines on the cognitive levels expected from applicants during the examination process.


Question: What does FONSI mean in the context of nuclear regulatory processes?

Accordingly, the NRC has determined that a FONSI is appropriate, and an environmental impact statement is not warranted.

The documents identified in the following table are available to interested persons through one or more of the following methods, as indicated.


Question: boric acid solution in nuclear reactors

The concentration of boric acid solution in storage is maintained between 4 and 4.4 percent by weight. Periodic manual sampling and corrective action, if necessary, assures that tanks are maintained.

A temperature sensor provides temperature measurement of each tank's contents. Temperature indication is provided as well as high and low temperature alarms which are indicated on the main control board.


Question: DNPS site boundary and vicinity

The DNPS Site Boundary is delineated in red on the map, providing a clear demarcation of the area under the jurisdiction of the facility.

The scale presented in the figure shows distances of miles, ranging from 0 to 2, which assist in understanding the proximity of various geographic features.


Question: Demineralized Water Makeup System

The Demineralized Water Makeup (DWM) System is a normally operating plant system that is in scope for license renewal. The DWM System is common to both Units 2 and 3 and consists of pumps, storage tanks, demineralizers, and the necessary control and support equipment.

Water is drawn from site wells and placed in the well water storage tank.


Question: The NRC mandates that licensees conduct thorough safety evaluations for their reactors.

The Division of Operating Reactor Licensing is responsible for overseeing the regulatory compliance of nuclear power plants in the United States.

Ngola Otto's email attachment includes detailed project updates and timelines based on recent revisions.


Question: through-wall stress distribution in PWR pressurizer vessels

Through-wall stress distribution at Path P2 for PWR pressurizer vessel bottom head (Westinghouse design)


Question: What is the purpose of a reactor containment structure?

The reactor containment structure is designed to prevent the release of radioactive materials into the environment under accident conditions.

The annual maintenance schedule ensures that all necessary inspections and repairs are conducted in a timely manner to maintain operational efficiency.


Question: byproduct material disposal authority

The authority to establish terms and conditions as the Commission determines necessary to assure that, prior to termination of any license for such byproduct material, or for any activity that results in the production of such material, the licensee shall comply with decontamination, decommissioning, and reclamation standards prescribed by the Commission; and with ownership requirements for such materials and its disposal site;

The authority to permit use of the surface or subsurface estates, or both, of the land transferred to the United States or a State/Commonwealth pursuant to paragraph 2.b. in this section in a manner consistent with the provisions of the Uranium Mill Tailings Radiation Control Act of 1978, provided that the Commission determines that such use would not endanger public health, safety, welfare, or the environment.


Question: What is the significance of energy levels in nuclear measurements?

Energy levels are critical in nuclear measurements as they help identify specific nuclides by their characteristic gamma photon energies emitted during decay processes.

The background uncertainty in measurements can significantly impact the accuracy of the data collected during nuclear safety assessments.


Question: The NRC imposes a safety review to evaluate aging management programs during the extended operation of a nuclear facility.

The NRC's safety review is focused on the applicant's programs for managing the effects of aging during the period of extended operation for specified plant structures, systems, and components that the current licensing basis of the plant may not be sufficient to address.

A hearing request must clearly demonstrate standing, and at least one contention must be identified that meets the admissibility standards set forth in 10 C.F.R. § 2.309(f)(1).


Question: What are severe accident mitigation analyses (SAMAs) in the context of nuclear licensing?

10 C.F.R. § 51.53(c)(3)(ii)(L) excuses Dominion from updating severe accident mitigation analyses (“SAMAs”) prepared in previous licensing and license renewal proceedings.

The probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to groundwater, and societal and economic impacts from severe accidents are small for all plants.


Question: What is the significance of occupational exposure in nuclear safety?

The approximately 14 percent difference in occupational exposure occurred during 2019-2021 (presented in NUREG-0713, Vol 43) as compared to 2018-2020 (presented in NUREG-0713, Vol 42) for the following reasons: (1) there was an outage with additional Steam Generator and RTD work that was not performed during 2018-2020, and (2) a dry fuel campaign was performed in 2021 and was not performed during 2018-2020.

NEPB has documented two reportable inadvertent nonradioactive releases since 2020. A reportable spill occurred in January 2021 when approximately eight gallons of diesel fuel was released due to a hose failure on a pickup truck.


Question: What is the weighted mean activity of nuclides?

The weighted mean activity for each nuclide is calculated to provide an average representation of the activity measured. For example, the weighted mean activity of PB-212 is 2.21E-02 pCi/grams.

Errors quoted at 2.00sigma indicate the level of uncertainty in the measurements taken.


Question: What are the requirements for an applicant's environmental report for nuclear license renewal?

Licensees are required by 10 CFR 51.53(c)(1) to submit with its application a separate document entitled "Applicant’s Environmental Report—Operating License Renewal Stage."

The staff also participated in breakout sessions with applicant personnel to gather information that will likely be used in the supplement to the 2021 draft SEIS.


Question: differential pressure for nuclear safety systems

DP is the pressure load during accident conditions when the strainers are covered with debris. This is the maximum allowable differential pressure that will not result in interaction ratios (IR) greater than 1.0 for the Cold Case.

The allowable compression stress of the stainless steel members is based on the lower allowables from ANSI/AISC N690-1994, "Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities" (Reference 35) as opposed to those provided in the AISC Code.


Question: What are the key components of a reactor protection system?

The reactor protection system includes safety-related components that detect abnormal conditions and automatically initiate safety actions to protect the reactor and prevent accidents.

The cooling system is designed to maintain optimal temperature within the reactor core and ensure efficient heat removal under normal operating conditions.


Question: nuclear licensing process

The nuclear licensing process is a systematic approach mandated by the Nuclear Regulatory Commission (NRC) for the approval and oversight of nuclear facilities, ensuring that all safety, environmental, and operational standards are met before a plant can commence operations.

The cooling system of a reactor must maintain operational integrity during all phases of plant operation, ensuring that heat is effectively removed from the reactor core to prevent overheating.


Question: What does the New Nuclear Initiative aim to achieve in terms of reactor concepts and community engagement?

So we really do want to have more NRC interaction because obviously the more transparency in these discussions that there are, the more you have people being brought along in an iterative conversation.

How do we look at customers we might want to bring into our community?


Question: Level 2 PRA model

The requested information is needed in order for the NRC staff to reach a conclusion on the adequacy of the applicant’s Level 2 PRA model for supporting the SAMA evaluation.

Define each of the release categories in terms of the characteristics (e.g., timing and magnitude of release) that are used to perform the binning and justify that the binning is reasonable.


Question: The equipment must move in both forward and reverse orientations during the scanning process.

Also with Search Unit pointed in opposite direction.

- **Search Unit (1)**
  - Travel (1)
  - Direction:
    <--- Straight Beam Search Unit --->


Question: The assessment of analysis methods ensures compliance with the regulatory standards for sample evaluation.

All sample data provided in this report met the acceptance criteria specified in the analytical methods and procedures for initial calibration, continuing calibration, instrument controls and process controls where applicable.

Samples 485191034 (L1-SUB-PAD-AJGS-005-SB), 485191035 (L1-SUB-PAD-AJGS-003-SB), 485191036 (L1-SUB-PAD-ASGS-013-SB) and 485191039 (L1-SUB-TDS-CJGS-B03-SB) were recounted due to results more negative than the three sigma TPU.


Question: How does grid voltage affect the performance of circulating water pumps?

Since the Circulating Water pumps have synchronous motors, the pump and motor run at synchronous speed regardless of the voltages supplied. This will cause the CW pump to maintain a constant flow even when supply voltage is reduced.

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.


Question: What is the purpose of instrumentation in nuclear reactors?

INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION is designed to detect and measure radiation levels in and around the reactor environment, ensuring safety and compliance with regulatory standards.

The page is intentionally left blank to maintain the formatting of the manual.


Question: The NRC mandates that licensees must provide a detailed analysis of protective measures during emergency situations.

The NRC requires licensees to analyze emergency zones and implement protective actions as necessary to ensure public safety in the event of an incident.

The implementation of response strategies is crucial for effective management of resources and allocation of personnel during routine operations.


Question: post-accident sampling drills

Post-accident sampling drills are conducted annually. These drills address capabilities including analysis of liquid and containment atmosphere samples with simulated elevated radiation levels. This criterion is not applicable if the NPP unit(s) does (do) not have licensing basis requirements for post-accident sampling.

Ingestion pathway and post-plume phase drills are conducted biennially. These drills involve sample plan development, analysis of lab results from samples, assessment of the impact on food and agricultural products, protective decisions for relocation, and food/crop embargos.


Question: isotope concentration

The regulatory filing process for nuclear power plants involves detailed reviews and assessments of safety protocols and operational procedures.


Question: What is the significance of reactor trip in nuclear safety analysis?

However, these functions are not credited in the analysis. No single active failure will prevent operation of the reactor protection system.

The specific results for operation with digital feedwater control system are not presented since the detailed results and conclusions presented for operation with the analog feedwater control system are considered valid for operation with the digital feedwater control system.


Question: What is the impact of a Power Range Monitor (PRM) failure in reactor operations?

NOTE Control rod withdrawal may NOT be possible if a PRM has failed high due to the 103% Rod Withdrawal Stop (C-2) until PRM is defeated. (Annunciator window 83-A).

The event description outlines the sequence of actions an operator takes when verifying control rod positions and checking system levels.


Question: Americium-241 concentration

Americium-241 is measured in the report with a result of 0.0538 pCi/g under the qualifier U, indicating that the measurement has been verified.

The LaCUBR Site Restoration Project aims to restore the site using various methodologies and approaches to ensure compliance with regulations.


Question: uncertainty and setpoint calculation

In order to change from an uncertainty calculation to an uncertainty and setpoint calculation, add the evaluation of the setpoint and Allowable Value, which was previously performed in calculation 19-AN-19.

Entire calculation has been reformatted and reprinted. Revision bars are shown for content changes only, not for format changes or typographical corrections.


Question: The evaluation of compliance with the Technical Specifications is essential for operational safety.

Perform Technical Specification Evaluation and LCO Tracking for Inoperable Turbine Bypass Valves.

All computer access must be monitored by an examiner or proctor to ensure access remains limited to official plant databases and documents (no personal study notes, email, etc.).


Question: What is meant by passive electrical components in nuclear licensing?

Following the identification of the electrical commodities, the criterion of 10 CFR 54.21(a)(1)(i) was applied to identify commodities that perform their functions without moving parts or without a change in configuration or properties (referred to as 'passive' components).

Components and commodities which support or interface with electrical components and commodities (for example, cable trays, conduits, instrument racks, panels and enclosures) are evaluated as structural components.


Question: The NRC evaluates license amendment requests by analyzing relevant technical specifications and regulations.

The proposed LARs (Agencywide Documents and Access Management System (ADAMS) Accession No. ML23159A249 and ML23159A253) requested to modify the Quad Cities technical specifications (TSs) to permit the use of risk-informed completion times (RICTs) in accordance with Technical Specifications Task Force (TSTF)-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b' (ML18183A493).

An audit is conducted to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision.


Question: What accommodations does the NRC provide for individuals with disabilities in meetings?

The NRC provides reasonable accommodation to individuals with disabilities where appropriate.

The policy statement may be found on the NRC website: http://www.nrc.gov/reading-rm/doc-collections/commission/policy/67fr36920.html


Question: What are the conditions evaluated for PWR small-break LOCA scenarios?

Experiments and Void Correlation for PWR Small-Break LOCA Conditions”, Transactions of American Nuclear Society, Vol. 17, Nov. 1973, pp. 369-370.

Heat Transmission, McGraw-Hill 3rd edition, 1954, p. 172.


Question: In comparing the economic viability of nuclear plants, higher burnup fuel cycles can lead to cost reductions over time.

If you compare an 18-month cycle to a 24-month cycle, it goes up. And if you compare a 24-month cycle with current burnup to a 24-month cycle with a higher burnup, costs go down.


Question: The NRC mandates that licensees monitor the degradation of components such as valves over time to ensure system integrity.

Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19)


Question: The outreach efforts must balance public engagement with maintaining the integrity of the NRC's regulatory processes.

OGC is concerned that continuing outreach activities can cause confusion in the minds of the public with respect to the NRC’s decision making process—informal outreach versus formal staff review and adjudication—and create false expectations in regard to the public’s ability to influence agency decisions through informal means.

The Commission determined that a hearing was called for in the public interest; it is not explicitly addressed in the NWPA.


Question: FSAR update requirements

The NPUF LR Rule added new 10 CFR 50.71(e)(4)(ii), which states, in part, that NPUF licensees shall file an FSAR update no more than 5 years from the date of the submittal of the updated FSAR required by 10 CFR 50.71(e)(3)(iv) or 'by order.'

In accordance with 10 CFR 2.202, a licensee must, and any other person adversely affected by this Order may, submit a written answer to this Order under oath or affirmation within 20 days of its publication in the Federal Register.


Question: NRC safety culture policy

The NRC's Safety Culture Policy Statement became effective in June 2011. While a policy statement and not a regulation, it sets forth the agency’s expectations for individuals and organizations to establish and maintain a positive safety culture.

In accordance with 10 CFR 2.390 of the NRC’s “Agency Rules of Practice and Procedure,” a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s document system (ADAMS).


Question: The Technical Specifications, as updated, must be followed precisely by the licensee to ensure safety and compliance.

The Technical Specifications contained in Appendix A, as revised through Amendment No. 252, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.


Question: What is the function of the pressurizer water level channels in reactor safety?

Three pressurizer water level channels are used for high pressurizer water level reactor trip. Isolated signals from these channels are processed with a median select algorithm and then used by the operator's PC for pressurizer water level control.

The basic function of the reactor protection circuits associated with low-low steam generator water level is to preserve the steam generator heat sink for removal of long-term residual heat.


Question: What are the types of systems used for radiological monitoring?

General Radioisotopes Products, Inc. has developed various models for radiological monitoring including the GTC-100, GTC-200, and GTC-300, which are designed to meet the needs of different monitoring applications.

General Electric Medical Systems (GE Healthcare) manufactures a variety of medical imaging devices, one of the most notable being the GE Discovery 610, which is used extensively in healthcare settings.


Question: Nuclear facilities are required to adhere to stringent safety regulations which include detailed technical analyses.

The NRC mandates that all reactor designs undergo a thorough licensing process to ensure that they meet safety and operational standards.

Maintaining proper ventilation in a building is essential for indoor air quality, which involves the circulation of fresh air.


Question: The NRC assesses the measurements and uncertainties of various radiation energies in relation to their environmental impact.

The energy levels and their corresponding uncertainties are critical for determining the radiation characteristics in regulatory assessments.

The tabular data includes various figures representing energy measurements, yet the primary focus remains on ensuring technological advancements.


Question: What is the function of the Process Vent Gas Monitor in nuclear safety?

This channel takes the continuous process vent sample, after it has passed through the particulate filter paper, and draws them through a sealed system to the process vent gas monitor assembly.

The counting rates of limiting isotopes for each channel, except the letdown monitors, are shown in Table 11.4-2.


Question: average enrichment of fuel assemblies

| Initial Assembly       | Avg. Enrichment wt % 235U (E) |       CE        |       WE        |       WE        |       B&W       |       CE        |       WE        |       B&W       |

The maintenance of cooling water systems is crucial for reactor safety and prevents overheating.


Question: The NRC concludes that the findings indicate no substantial harm to environmental conditions.

Accordingly, the NRC has determined that a FONSI is appropriate, and an environmental impact statement is not warranted.

The documents identified in the following table are available to interested persons through one or more of the following methods, as indicated.


Question: The emergency response procedure dictates immediate actions upon detecting a loss of one of the reactor's critical systems.

In accordance with S2.OP.AB.RHR-0001, which ONE of the following completes the statement below?


Question: The measurement of radiation dose at specific locations is crucial for safety assessment in nuclear facilities.


Question: vessel original locking devices

40 – 60 Methods and Conditions: For the VT-3 inspections of the specified item B4 VV original locking device types, the relevant condition looks for evidence of: (1) damage or wear in the U-cover, (2) misalignment of the pressure plate with the jack-screw, or spring retainer, (3) damage to the pressure plate or spring retainer, or (4) jack-screw out of design configuration. Additionally, evidence that the VV is not symmetrical with the mounting ring or that the jack-screw thread extensions from the lower retaining ring threaded flange are unequal.


Question: What are the safety implications of aircraft impact on nuclear plants?

Aircraft Impact is defined as an aircraft (either a portion of (e.g., missile) or the entire aircraft) that collides either directly or indirectly (i.e., skidding impact with one or more structures, systems, or components (SSCs) at or in the plant's analyzed area causing functional failure.

An updated aircraft impact evaluation is documented in UFSAR Section 3.5.6.


Question: What is the significance of uranium supply in nuclear energy generation?

However, because the project has potential to provide a domestic source of uranium that will eventually be used in nuclear reactors to generate electricity, it will have a public benefit.

The following sections describe the benefits and costs of the project for members of the local communities, local governments, and the State of New Mexico.


Question: The NRC assessed the implications of water extraction on the local aquifer's health during the license renewal process.

The NRC recognized that nuclear power plants are not the large-volume groundwater users that would be a leading driver for saltwater intrusion in the plant's locale.

The analysis revealed an improvement in the water quality of the Illinois Waterway over the past decades, although it still faces contamination from various environmental sources.


Question: The measurement of radiation dose at specific locations can significantly affect the safety analysis for nuclear facilities.

The dose at Location 1: EAB3 includes values such as 6.1800E-04 and others, which are crucial for assessing the radiological impact on personnel.

The control room maintains critical operational parameters to ensure the plant runs smoothly and efficiently, independent of dosimetry measurements.


Question: cadmium neutron capture cross-section

The accuracy of the cadmium cross-section behavior as a function of neutron energy has been well known and published since 1955.

Table 6.13 presents the results of the GEMER calculated eigenvalues for 14 of the critical benchmarks and limiting condition cases for the NPC single package and package arrays - with and without cadmium.


Question: What is the role of safety systems in nuclear regulatory compliance?

Safety systems are critical for ensuring the protection of the public and the environment, and they must be adequately designed, implemented, and maintained under regulatory requirements.

The meeting was originally scheduled for a different room, and there was an issue when I attempted to change the location.


Question: What is the significance of nitrate ACL in groundwater monitoring for nuclear facilities?

The table shows nitrate levels measured in well samples over several years.


Question: The evaluation of stress in welds is critical for ensuring structural integrity under operational loads.

For the primary stress evaluation, the original calculated stress is 6.48 ksi in the fillet weld, which is less than the allowable stress value of 13.98 ksi for the Alloy 600 material.

The calculated primary membrane-plus-bending stress intensity is, therefore:

Pt+Pb = (18.31)(3.318/3.182)(0.1227/0.1131)²
= 22.5 ksi < 34.9 ksi = 1.5 Sm


Question: What is the purpose of auditing the probabilistic risk assessment in nuclear licensing?

The NRC staff will audit the PRA methods that the licensee would use to determine the risk impact from which the revised completion times would be obtained, including the licensee’s assessments of internal events (including internal flooding), fire, and seismic PRAs.

The NRC staff will request information and an audit meeting(s) throughout the audit period.


Question: criticality safety evaluation for nuclear transport

Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, NUREG/CR-5661, ORNL/TM-11936, H.R. Dyer, C.V. Parks, April 1997.

MCNP4C / MCNPDATA – A General Monte Carlo N-Particle Transport Code System, CCC-200, Radiation Shielding Information Center, Oak Ridge National Laboratory, July 2000.


Question: What is the role of probabilistic risk assessment in nuclear safety evaluations?

So that's essentially how we had addressed that item. There were no changes to the rule text itself.

Okay. So subpart D, it has the requirements for siting.


Question: mass flow rate in nuclear reactors

CORE INLET AND OUTLET MASS FLOW RATE DURING BLOWDOWN (CD = 0.6, LOW TAVG, MIN SI, 8.5 FT POWER SHAPE, NON-IFBA)

The emergency response plan outlines the procedures to be followed in the event of an unforeseen incident at the facility.


Question: The NRC mandates that licensees submit safety analyses to demonstrate compliance with regulatory requirements.

Licensees must submit detailed safety evaluations as part of the licensing process to ensure compliance with the NRC's safety standards.

Periodic inspections are conducted to verify the operational status of equipment and assess compliance with established guidelines.


Question: What are the potential alternative development sites for the Dresden Nuclear Power Station?

The potential alternative development site indicated in the legend of Figure 7.2-1 includes a yellow area covering 103 acres.

The site boundary is marked in red, which denotes the limit of the Dresden Nuclear Power Station's operational area.


Question: nuclide library

Nuclide Library Used: C:\Canberra\Apex\Root\Dairylend_NPPLibrary\HOTLAB.NLB


Question: license amendment request

By letter dated December 8, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20343A243), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted a license amendment request for the Millstone Power Station, Unit No. 3 (MPS3).

The NRC staff has reviewed your application, as supplemented, and concluded that it does provide technical information in sufficient detail to enable the NRC staff to complete its detailed technical review and make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment.


Question: license renewal application

If you wish to continue your licensed program, you should prepare and submit a renewal application on NRC Form 313 (enclosed), following regulations in 10 CFR and licensing guidance NUREG 1556, 'Consolidated Guidance About Materials Licenses.'

If you do not wish to renew your license, you must dispose of or transfer all licensed radioactive material in your possession in accordance with 10 CFR 30, 40 and 70.


Question: To maintain reactor stability, designers must ensure that any disturbances in power output do not lead to an increase in fluctuations.

The NRC staff conducted its review considering Natrium PDC 12, "Suppression of reactor power oscillations," states: The reactor core; associated structures; and associated coolant, control, and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable system radionuclide release design limits are not possible or can be reliably detected and suppressed.

Though applicable to boiling water reactors (BWRs), the NRC staff considered aspects of chapter 15.9, "Boiling Water Reactor Stability," of NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants," dated March 28, 2007 (ML007550017), relevant to a reactor cooled with single-phase liquid sodium.


Question: The threshold for peak localization during gamma spectrum analysis is set at 3.00.


Question: What is the significance of Minimum Detectable Concentration in nuclear analysis?

MDC: Minimum Detectable Concentration is defined as the smallest concentration of a substance that can be reliably detected but not necessarily quantified, which is crucial in ensuring regulatory compliance and safety evaluations.

The address for LaCrosseSolutions is S4601 State Hwy 35, Genoa, Wisconsin 54632, which provides a location for the company conducting the analysis.


Question: What are the critical steps identified in a Job Performance Measure (JPM)?

Critical steps meet the criteria for critical steps and are identified with an asterisk (*).


Question: nuclear licensing review process

The Nuclear Regulatory Commission (NRC) reviews applications for nuclear licenses through a systematic process that includes technical assessments, public comment opportunities, and considerations of safety and environmental impacts.

The carbon-14 analysis showed a result of -0.010 with an uncertainty of +/-1.77, which indicates the sensitivity and accuracy of the testing method used.


Question: What is the analytical procedure for Pu-238 analysis?

**Analytical Procedure:** GL-RAD-A-011 REV# 27

The samples in this SDG were analyzed on a "dry weight" basis.


Question: What are the key references related to the high temperature oxidation of Zircaloy?

“The High Temperature Oxidation of Zircaloy in Water,” W. A. Bostrum, WAPD-104 (March 1954)

The voting record includes a signed note from each of the four Commissioners. In each case the opening sentence is: “I approve …”}


Question: What is the role of subject matter experts in nuclear environmental audits?

Constellation staff and contractors who are subject matter experts in the disciplines identified on the environmental audit needs list should be available for interviews and tours.

An audit summary report will be issued by the NRC staff within 90 days from the end of the environmental audit.


Question: The NRC stipulates requirements for the processing and management of containment spray water, ensuring that safety measures are properly implemented during operation.

No automatic switchover to the containment sump for containment spray. It is manually aligned to be supplied from RHR.

Not applicable to DCPP, no emergency AC units. Randomly selected AA1.05 (IR 3.3)


Question: What is the population change in the regions surrounding nuclear plants?

The total population in the region of influence increased from 564,115 in 2000 to 696,149 in 2020, reflecting a growth of 21.5%.

The NRC mandates that all licensees must submit a report demonstrating compliance with safety regulations.


Question: What is the RCIC system isolation?

Reactor Core Isolation Cooling (RCIC) is a system designed to provide cooling to the reactor core and is crucial for maintaining safety during transient accidents.

These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.


Question: What actions are required when two offsite circuits are inoperable?

C.1 Declare required feature(s) inoperable when the redundant required feature(s) are inoperable.

E.1 Restore one required DG to OPERABLE status.


Question: What is a Risk Informed Completion Time (RICT) program in nuclear regulations?

Fermi 2 TS 5.6.9 Risk Informed Completion Time (RICT) Program Upgrade Report corresponds to TSTF-591 5.5.17 Risk Informed Completion Time Program and 5.6.6 Risk Informed Completion Time (RICT) Program Upgrade Report respectively.

DTE has evaluated the proposed changes to the TS using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.


Question: What regulations govern the classification and storage of radioactive waste?

The inspectors reviewed documentation and records of facility activities and observed posting and labeling of storage areas and containers to verify compliance with the requirements of 10 CFR 61.55 and the license application.

The inspectors conducted walkdowns and observations and reviewed records of on-site airborne effluent scrubbers and filters to verify compliance with the license application.


Question: What are the proposed changes to the emergency plan at PSL?

The PSL emergency plan is being converted into the proposed NextEra Common Emergency Plan and PSL Emergency Plan Annex, which are based primarily on the NUREG-0654/FEMA-REP-1, Revision 2 guidance.

NextEra will make program wide improvements by standardization of response organizations, duties and responsibilities, processes, procedures, and training.


Question: What is an ISFSI in the context of nuclear licensing?

The BFN site has adequate space onsite to accommodate the construction of an additional ISFSI pad if necessary.

The Alabama Coastal Area Management Program as part of the Coastal Zone Management Program (CZMP) enforces policies on coastal lands and waters in the continuous 10-foot contour seaward in Baldwin and Mobile counties (ADEM 2023).


Question: steam generator tube rupture

Figure 15-230 illustrates the thermal response of a steam generator tube rupture scenario, displaying the coolant temperature over time following such an incident.

The cooling system operates efficiently to maintain reactor temperature within the designated safety limits under normal operating conditions.


Question: What is the significance of MDA in nuclear measurements?

CAUTION: MDA value is inconsistent with Currie MDA at 95% confidence level.


Question: debris transport in nuclear pools

For PBN1, Figure 3.e.1-6 shows that the turbulence (yellow regions) and the velocity (red regions) in the pool (blue regions) generated by the break flow are not high enough to transport the generic small debris present in the pool to the sump strainers during recirculation. Therefore, the transport for small debris blown to lower containment is 0% for PBN1.

The regulatory process requires comprehensive documentation of the safety analysis and all assumptions made during the assessment to ensure compliance with NRC guidelines.


Question: During FLECHT Run 8874, the test bundle experienced unexpected chemical reactions leading to excessive oxidation.

The section of the test bundle from FLECHT Run 8874 that incurred runaway oxidation was critically evaluated to understand the implications on fuel performance.

The reactor's cooling system operates by circulating water through the core to maintain optimal temperature and pressure during operation.


Question: What is the COBRA-FLX critical heat flux correlation?

The COBRA-FLX critical heat flux correlation provides a means for predicting the conditions under which critical heat flux occurs in nuclear reactors, which is essential for ensuring safe and efficient reactor operation.

Table B-3 includes local conditions from COBRA-FLX simulations, detailing various grid test parameters and outcomes.


Question: What is the significance of land use and land cover changes in nuclear regulatory filings?

The NRC evaluates land use and land cover change to assess their potential impacts on environmental resources during the licensing process.

Climate models are often utilized to predict future weather patterns and their potential effects on regional ecosystems.


Question: micrograms per liter

µg/L refers to micrograms per liter, a unit of measurement commonly used to quantify concentrations of contaminants in water and other substances.

The wastewater treatment plant (WWTP) processes sewage and wastewater to remove contaminants before releasing it back into the environment.


Question: The NRC mandates the presentation of electrical lineups related to Loss of Offsite Power and Loss of Coolant Accident conditions.

Electrical lineups for following:
 a. LCO 3.8.9, Condition A & concurrent loss of offsite power (LOOP) & loss-of-coolant accident (LOCA).
 b. LCO 3.8.9, Condition C & concurrent LOOP & LOCA.

High-level discussion of the fault tree(s) and/or event tree(s) for electrical sources associated with Technical Specification (TS) LCOs 3.8.1, Condition D; and 3.8.4, Condition B.


Question: Actions must be taken to manage the power range monitor's status during a failure scenario.


Question: radiological environmental monitoring

Will expected climatological changes affect the level of radiological environmental monitoring for the site?

Will expected climatological changes affect any health impacts from noise associated with plant operations?


Question: What is the role of a pre-decisional enforcement conference in nuclear regulatory processes?

If a PEC is held, it will be open for public observation and the NRC will issue a press release to announce the time and date of the conference.

The final decision will be based on you confirming on the license docket that the corrective actions previously described to the NRC staff have been or are being taken.


Question: The final compression pressure in the upper compartment is influenced by the amount of energy released during blowdown events.

The sensitivity of the upper compartment compression pressure peak versus the amount of energy released is shown in Figure 6-7.

Figure 6-16 gives some insight as to the very large capacity for energy absorption of the ice condenser, as obtained from test results.


Question: The NRC's protocols established for managing unexpected findings of human artifacts are critical in preserving cultural resources.

The 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.

If you have any questions concerning this subject, please contact Mr. Robert Carlson of my staff at (301) 415-8165.


Question: The NRC assesses the impact of neutron flux on the integrity of reactor components.

Loss of fracture toughness due to neutron irradiation embrittlement is a key aging effect for steel reactor vessel components.

Not applicable. DCPP is a Westinghouse plant, not a Babcock & Wilcox plant.


Question: What is the purpose of a radiation survey meter in a radiation safety program?

We will either possess and use, or have access to and use, a radiation survey meter that meets the criteria in the section titled, “Radiation Safety Program—Radiation Monitoring Instruments” in NUREG–1556, Vol. 1, Rev. 2, “Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Portable Gauge Licenses,” in the event of an incident.

Physical inventories will be conducted every 6 months or at other intervals approved by the NRC to account for all sealed sources and devices received and possessed under the license.


Question: The NRC mandates that operators ensure proper function of the RCIC system during specific operational conditions.

The Reactor Core Isolation Cooling (RCIC) system is designed to provide adequate core cooling during a loss of coolant accident by injecting water into the reactor pressure vessel.

The Event Trigger 22 specifies a condition where a parameter must be less than 95 to activate the corresponding command.


Question: Westinghouse is seeking to protect certain proprietary information from public disclosure following the procedures outlined in 10 CFR 2.390.

I am requesting the proprietary portions of WCAP-12306 Revision 4 be withheld from public disclosure under 10 CFR 2.390.

I, Camille Zozula, Manager, Regulatory Compliance and Corporate Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).


Question: What is the role of bolting integrity in nuclear safety?

The last bullet states: “a representative sample of bolt heads and threads is inspected up to a maximum of 25 bolts for each material and environment combination.”

On page 3-603, the last paragraph states: “(2) Cracking due to expansion and reaction with aggregates could occur in inaccessible concrete.”


Question: The NRC mandates that licensees assess potential risks associated with the maintenance of nuclear facilities.

The safety evaluation report provides an extensive analysis of the proposed technical specifications changes for the operating units, ensuring compliance with safety regulations.

The reactor's containment structure is designed to withstand extreme pressure changes that may occur during operational scenarios.


Question: What are the results for the activity of various nuclides?

The activity (pCi/grams) for I-131 shows -3.61E-01, for CS-134 it is -2.02E-02, and for CE-144 it is -4.28E+00.

The energy levels of nuclides like I-131 and CS-137 vary significantly, with I-131 at 284.30 keV and CS-137 at 661.65 keV.


Question: To ensure the effective operation of emergency systems, it is crucial for operators to continuously monitor the status of diesel generators and associated ventilation mechanisms.

(RAI 8a) Revise procedures to direct the operator monitoring a running diesel generator to ensure that the ventilation system is running or take action to open doors or use portable fans.

Replace a supply fan with a different design in service water pump room.


Question: What are the components of the feedwater control system in a nuclear reactor?

Figure 7-35 illustrates the functional diagrams of the DAP & DBP Feedwater Control System, including elements such as Feedwater Flow Control Valves and Feater Pumps.

Quality Assurance requirements apply, and the authority to deviate from these must be documented.


Question: What are the license numbers for the Joseph M. Farley Nuclear Plant?

Mr. G. R. Peterson is the Vice President at the Duke Energy Corporation.


Question: The highest recorded temperature at Fredericksburg was 107°F.

Maximum Temperature refers to the highest temperature recorded in a specific location, with Fredericksburg experiencing a remarkable 107°F.

The fastest observed mile wind speed in Richmond reached 68 mph.


Question: What are the requirements for the Automatic Depressurization System (ADS) initiation?

In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability in both trip systems.

In this situation, the Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities.


Question: What is the significance of wind speed observations in nuclear safety assessments?

The equipment used for monitoring must comply with standards set forth by the regulatory authority to ensure the integrity of the data collected.


Question: nuclear gamma calibrators

The 492 Gamma Calibrators are designed for precise calibration of radiation detection equipment used in nuclear applications.

The GAMMACELL 220 Excel is a high-performance gamma irradiator, primarily used in research and development for sterilization and calibration.


Question: The NRC mandates that licensee programs must adhere to environmental monitoring standards to ensure safety.

The Environmental and Personal Radiological Monitoring and Protection Program SOP outlines the procedures for monitoring radiological exposure and ensuring compliance with regulatory standards.

Management will assess the performance of all staff members on a semi-annual basis to identify training needs and areas for improvement.


Question: What are the requirements for the inspection of the reactor coolant system boundary?

The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined, and accepted by reviewing the following activities in Unit 2 during refueling outage A2R24 from October 7, 2024, to October 16, 2024.

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas: Fire Zone 1.3-2; Unit 2 containment 426' elevation, containment upper area during the week ending October 12, 2024.


Question: SRO justification for operator questions

Does the question require one or more of the following: knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures (when SRO approval required)?


Question: The NRC's evaluation of the FLECHT tests highlights significant deviations in heat transfer coefficients that were overlooked during the data interpretation process.

NRC’s technical evaluation does not say that the 'anomaly,' regarding heat transfer coefficients, was definitely attributed to the data reduction process. NRC’s technical evaluation states that 'some of the anomaly [lower 'measured' heat transfer coefficients] can probably be explained due to a deficiency in the data reduction process.'

The heater rod failures were apparently caused by localized temperatures in excess of 2500°F—i.e., they were not caused by temperatures in the range of 2200 to 2300°F.


Question: What is the role of primary containment isolation instrumentation in nuclear reactors?

The primary containment isolation instrumentation for each function in Table 3.3.6.1-1 shall be OPERABLE.

Separate Condition entry is allowed for each channel.


Question: What is the significance of radioactive decay in safety calculations?


Question: The NRC mandates that all modification of ITAAC must be communicated properly through appropriate notifications.

Currently, the ITAAC Notification Review Status for several submissions has been marked as 'Sufficient Information', indicating that the submissions meet the necessary regulatory requirements for review.

The NRC has established a comprehensive strategy for overseeing the safe operation of nuclear facilities that includes regular inspections and oversight.


Question: valve program plan

The Valve Program Plan is designed to ensure that all safety-related valves are properly maintained and tested according to regulatory requirements to ensure operational safety.

The testing of the system occurs routinely, ensuring that all parameters are within acceptable limits during normal operations.


Question: The licensing process for nuclear facilities involves rigorous scrutiny by regulatory bodies.

Each applicant for a nuclear power plant license must submit a comprehensive set of documents that demonstrate compliance with NRC regulations and safety standards.

The heat exchanger is crucial for maintaining optimal temperature control in the reactor core.


Question: The NRC reviews the impact of thermal aging on material properties for the components within the reactor systems.

Loss of fracture toughness due to thermal aging embrittlement


Question: affected environment near nuclear facilities

Figure 3.1-4 presents the major population centers near the HCS, including Horizon Center Site, Oak Ridge, and Oliver Springs, as well as the Oliver Springs Inc Airport.

The USGS National Hydrology Dataset (NHD) includes various waterbody classifications and demographic information for municipalities.


Question: The NRC distinguishes between extravasation and infiltration in the context of nuclear medicine procedures.

The petitioner fails to make an important distinction in the terminology where “extravasation” is inappropriately considered synonymous with “infiltration”.

The nursing societies have developed national guidance documents for the management of extravasations.


Question: What is the scaling methodology used for in nuclear safety analysis?

This report has been prepared to document the methodology for development of scaling analyses to be used for testing that supports the assessment base of the KP-FHR evaluation model.

The KP-FHR is a U.S. developed Generation IV advanced reactor technology.


Question: exempt distribution license

CK Worldwide, Incorporated has submitted a request for the renewal of their exempt distribution license, which allows for the distribution of certain radioactive materials without the need for a full licensing process.

The table of key personnel includes individuals responsible for overseeing various aspects of the licensing and compliance process, ensuring that the operations meet regulatory standards.


Question: Quality Assurance Program in nuclear licensing

The Quality Assurance Program (QAP) is essential for ensuring that all processes and procedures meet the regulatory requirements for nuclear safety and operational reliability.

The financial projections for the upcoming quarters indicate a steady growth in revenue, driven by an increase in demand for renewable energy sources.


Question: What is the ASME Code Case N-740-2 in nuclear inspections?

Attachment 2: Review of ASME Code Case N-740-2 against the Proposed Alternative Weld Overlay

The regulatory commitments include various measures to ensure the safety and compliance of nuclear operations.


Question: The simulator must be set up correctly before starting the examination.

Go to RUN when the examinee is given the cue sheet.


Question: What is the purpose of the spent resin dewatering subsystem in a nuclear plant?

The spent resin dewatering subsystem receives the resin slurry resulting from backwashing the spent powdered resins from the demineralizer vessels.

The system is installed with a full-flow bypass which ensures full condensate flow and limits pressure drop across the cleanup system.


Question: What is the significance of the Nuclide Minimum Detectable Activity (MDA) in regulatory filings?

Nuclide MDA (pCi/grams) is the minimum detectable activity of a given nuclide, expressed in picocuries per gram.

The energy levels of various nuclides are critical in understanding their behavior in nuclear reactors.


Question: boric acid corrosion program

The NRC staff acknowledged that boric acid corrosion programs are well established performance-based AMPs for managing cracking and potential leaks from primary pressure boundary components in PWRs.

The NRC staff perceived that MRP-232, Revision 2, may be subjecting these nozzles to specific types of NDE inspections, which could potentially conflict with EPRI’s designation of the CB outlet nozzles and SI nozzles as N category, non-managed RVI component types in MRP-227, Revision 2, Table 3-3.