Question: What is a critical digital asset in the context of nuclear security?

Critical digital assets are defined as those assets whose compromise could adversely impact the safety and security of the nuclear facility.

The Performance-Informed Inspection will enhance the oversight process by focusing on significant areas of concern identified during previous inspections.


Question: The NRC mandates that isotopes be quantified accurately, specifying their activity in mCi.

The manufacturer's device model is essential in identifying its specifications for proper use.


Question: Evaluate the pump's vibration levels against specified criteria to ensure compliance.

Applicant determines that pump vibration is within the Required Action Range for Bearing A.

The NRC continuously monitors operational safety and efficiency through various assessments of plant performance.


Question: The analysis includes monitoring the energy levels identified using energy calibration methods.

**Identification Energy Tolerance:** 1.000 keV


Question: dissolved boron concentration requirements

SR 3.2.1.1 Verify the dissolved boron concentration is met using two independent measurements.

Every 24 hours thereafter while the TSC is in the spent fuel pool or while water is in the TSC.


Question: What is the significance of surveillance reports for post-tensioning systems in nuclear plants?

In this report, surveillances are generally referred to by calendar year and / or time, T, since the SIT date.

The examination schedule proposed in Part 2 of this report is justified if the proposed schedule is supported by an evaluation of each of the 4 post-tensioning system performance categories listed above.


Question: What are the identified nuclides and their properties in nuclear regulatory filings?

The regulatory requirements for maintaining safety protocols in nuclear plants are essential for ensuring the health and safety of the community.


Question: What is the significance of energy lines in nuclear measurement?

Energy lines refer to distinct energy levels at which a specific nuclide emits radiation, and are critical for identifying and quantifying the nuclide's presence and activity in a sample.

The uncertainty in measurement can significantly affect the quality of the data collected during the licensing process, as regulators rely heavily on accurate reporting to ensure compliance.


Question: The NRC evaluates the safety and operational readiness of standby diesel generators through stringent fuel-oil system regulations.

Regulatory Guide 1.137, Revision 1, October 1979, provides guidelines on the qualification and testing of fuel-oil systems for standby diesel generators in nuclear power plants.

Regulatory Guide 1.132, Revision 1, March 1979, focuses on the site investigations necessary for the foundations of nuclear power plants.


Question: The Louisiana Public Alerts, Warnings, and Notifications Plan outlines the necessary training for agencies involved in issuing alerts.

Any member of a PAA whose duties include disseminating public alerts, warnings, and notifications must complete all required FEMA Independent Studies course(s) at the time of renewal.

West Feliciana Parish contracts with its IPAWS-compliant CAP Alert Origination Software Provider, which hosts the IPAWS system.


Question: simultaneous inoperability of control room temperature control system

A self-revealed finding of very low safety significance (Green) and associated non-cited violation of Technical Specification (TS) 5.4.1.a was identified for the licensee’s failure to implement procedure BOP VC-20, VC Chilled Water System to VS Chilled Water System Cross Tie Operation with One Train Inoperable for Maintenance, Revision 4.

After the first occurrence on January 31, 2024, the issue and lessons learned were communicated to the department.


Question: NuFuels plans to assess soil samples from the septic system's leach fields for contamination before concluding the licensing process.

Sludge in the septic tanks will be analyzed for natural uranium, radium-226, and lead-210 prior to removal and prior to requesting license termination.

During the grazing season, NuFuels will collect three separate samples from grazing areas near the processing facility (in the direction of the highest predicted radionuclides). The samples will be analyzed for radium-226 and lead-210.


Question: The NRC mandates that licensees must monitor the progression of various parameters such as the peak cladding temperature (PCT) during accident scenarios.

The high PCT cases in the upper part of Figure 4-2 are mainly influenced by the area of the break. This is demonstrated in Figure 4-3 which shows a general increasing trend in PCT with break size.

The analysis scatter plots for the case set are shown in Figure 4-1 through Figure 4-5.


Question: The energy values measured in kiloelectronvolts (keV) are vital for analyzing the performance of the testing.


Question: cask lid seal concerns

The only cask-related delay in shipment occurred in 1988, when concerns over the O-ring material in the cask resulted in the material being changed to a different grade of elastomer.

The only rail shipment incident occurred in 1987, when the train traveling at slow speed struck a car.


Question: The evaluation process at Turkey Point involves comprehensive safety assessments and regulatory compliance measures.

The Defense-in-Depth (DiD) strategy at Turkey Point Unit 3&4 ensures that multiple layers of safety are in place to protect against potential hazards.

Upgrading the digital systems at Turkey Point will enhance operational efficiency and reliability within the overall grid management framework.


Question: The NRC monitors compliance with safety regulations to ensure public health and safety in nuclear operations.

The Nuclear Regulatory Commission (NRC) has established regulations to govern the licensing of nuclear power plants, ensuring they comply with safety and environmental standards.

In response to recent changes in policy, the NRC updated its procedures to improve efficiency and accountability across all divisions.


Question: What is the purpose of Generator Differential protection in nuclear plants?

Generator Differential protection is a protective function that operates to detect internal faults within the generator by comparing the currents flowing into and out of the generator under normal operating conditions.

In nuclear facilities, maintaining the integrity of containment structures is critical for preventing the release of radioactive materials.


Question: What is the bolting integrity program in nuclear facilities?

The Flow Element Throttle is a critical component in maintaining optimal operational efficiency.


Question: The proposed alternative suggests a different method for managing repair and replacement activities in systems classified under ASME guidelines.

The proposed alternative to use ASME Code Case N-752 focuses on risk-informed categorization and treatment for repair and replacement activities in Class 2 and 3 systems, which is a significant aspect of ensuring safety and compliance.

The audit plan outlines various objectives, including evaluating regulatory compliance and assessing management practices related to system operations.


Question: MISTRAS Nuclear Quality Assurance Program

A comprehensive audit was performed on the MISTRAS Nuclear Quality Assurance Program, including the in-house and field activities affecting quality, conducted during the week of October 21, 2024 (Audit #100-QA-202410).

The attached audit report was completed and sent to MISTRAS on November 20, 2024.


Question: activity of nuclides

The activity of nuclides is measured in picoCuries per gram (pCi/grams), which provides a quantification of the rate of decay of the radioactive material.

The energy levels of the nuclides listed include specific values in keV, which indicate how much energy is released during their decay processes.


Question: demineralizer isotopic decontamination factors

Demineralizer isotopic decontamination factors: Kr and Xe isotopes: 1, Sr and Ba isotopes: 1, Rb-86, Cs-134, and Cs-137: 10, Rb-88, Rb-89, Cs-136, and Cs-138: 1, Other isotopes: 1

Pressurizer volumes (ft³):  Liquid volume: 1080  Vapor volume: 720


Question: What is the activity measurement for radionuclides?

Activity (pCi/grams) is a crucial measurement in assessing the radiological properties of nuclides, indicating the amount of radioactive decay per gram of material.

The yield percentage represents the amount of a specific isotope produced in a nuclear reaction, which is critical for understanding the efficiency of isotopic generation.


Question: risk-informed completion times

The RICT Program provides the necessary administrative controls to permit extension of CTs and thereby delay reactor shutdown or remedial actions if risk is assessed and managed within specified limits and programmatic requirements.

Energy Northwest has evaluated the proposed change to the TS using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration.


Question: Leakage boundary of heat exchangers

Heat Exchanger – (Regen Heat Exchanger) Tube Side Components | Leakage Boundary | Carbon or Low Alloy Steel with Stainless Steel Cladding | Air - Indoor, Uncontrolled (External) | Loss of Material | External Surfaces Monitoring of Mechanical Components (B.2.1.23) | VIII.IP-41 | 3.3.1-080 | A

Heat Exchanger – (Regen Heat Exchanger) Shell Side Components | Leakage Boundary | Stainless Steel | Air - Indoor, Uncontrolled (External) | Cracking | One-Time Inspection (B.2.1.20) | VII.E4.AP-209a | 3.3.1-004 | C


Question: inadvertent criticality during refueling operations

The transient considered is inadvertent criticality due to the complete withdrawal or removal of the highest worth control rod during refueling operations. It is classified as an infrequent incident.


Question: The NRC mandates that each facility's operational limits be identified and justified using accurate analyses.

The licensee is responsible for providing the acceptable values for the limiting conditions for operation, which ensure the safety and integrity of the facility.

Operational procedures must be established to maintain equipment functionality and ensure the workforce is effectively trained.


Question: Reactor Vessel Beltline Components

Figure 4.2.1-1: BFN Unit 1 - Reactor Vessel - Beltline Components

Figure 2-1: BFN Unit 1/2 SBO Alternate AC Source and Recovery Path Boundaries


Question: The strainer design must replicate specific parameters to ensure accurate testing during penetration assessments.

The test strainer for penetration testing was a prototypical strainer module that matched the key design parameters (i.e., all disks dimensions, including perforated plate thickness, hole diameter, pitch, etc.) of the plant strainer.

The effectiveness of the agitation region for each of the penetration tests is shown in Table 3.n.1-1, which documents the quantity of fiber that did not transport to the strainer and was collected from the high agitation or transport regions after the conclusion of each test.


Question: LP3 distribution in regulatory filings

Considering the goodness-of-fit, standard error magnitude, and comparison with observed data, the LP3 distribution was found to be the most suitable.

The effect of this additional river regulation can be observed in the figure with a reduced fit of the distribution with observed data.


Question: The monitoring plan for Sturgeon Impingement is required to comply with conditions set forth in the Environmental Assessment.

The 2021 Spring Sturgeon Impingement monitoring schedule is being submitted in accordance with Term & Condition 11.3 in the 2013 Entergy Indian Point Biological Opinion and Incidental Take Statement, as amended February 9, 2018.

The attached document provides a detailed account of the safety protocols in place for the operation of Unit 3 during its scheduled maintenance.


Question: What is the significance of nuclide identification in regulatory filings?

Nuclide identified during the nuclide identification is crucial for assessing the radiation safety and compliance with regulatory standards.

The energy levels and yields of specific nuclides can vary significantly depending on the conditions under which they are measured.


Question: What determines the National Register eligibility of structures related to nuclear facilities?

Thank you for this additional information. It seems to me that the 1974 cooling towers would not be National Register eligible due to age, the fact they were not part of the original design, and since they have lost their historic context.


Question: The NRC mandates a thorough investigation of damage to internal components of the reactor, which may include areas affected by molten material.

Coating of previously-molten material on bypass region interior surfaces is a critical aspect to evaluate during safety assessments to ensure integrity and functionality.

Hole in baffle plate can lead to issues with reactor performance, but does not directly relate to the assessment of molten material coatings.


Question: The NRC mandates that facilities maintain proper cooling systems to ensure operational efficiency and compliance with environmental standards.

The Cooling Water Structures include station structures as follows: Cooling Lake (includes dikes, associated hot and cold cooling canals), Discharge Canals (includes discharge canal, discharge outfall structure, and emergency spillway), Lake Lift Station and Spillway, Flow Regulating Station (includes diverter gates, control house), Intake Canals, Mechanical Draft Cooling Towers (includes pump intakes), and Ultimate Heat Sink.

The primary function of the emergency core cooling system is to provide cooling to the reactor core during loss-of-coolant accidents.


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory reviews and approvals from the NRC to ensure that the proposed nuclear facility meets safety, environmental, and security requirements.

The station's cooling system is designed to handle increased heat loads during peak operating conditions.


Question: The operational status of the Division I and II electrical power distribution systems is crucial for maintaining safety in nuclear reactors.

LCO 3.8.7 The following Division I and Division II AC and DC electrical power distribution subsystems shall be OPERABLE:

A.1 Restore AC electrical power distribution subsystem(s) to OPERABLE status.


Question: What is the PAD5 Code in nuclear licensing?

The purpose of the meeting was to discuss planned license amendment requests (LARs) to adopt the Performance Analysis and Design Model (PAD5™) Code, the FULL SPECTRUM™ Loss-of-Coolant Accident methodology (FSLOCA), and AXIOM™ Fuel Rod Cladding (with associated exemption request) for the Catawba Nuclear Station, Units 1 and 2, and the McGuire Nuclear Station, Units 1 and 2 facilities.

The NRC staff also provided feedback to the licensee regarding quality aspects and the importance of considering providing information in advance to aid in the NRC staff’s review.


Question: groundwater treatment process

c. Groundwater Treatment      |  X  |                              |

No in-house samples were collected during the month.


Question: dry weight basis analysis

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

Samples were re-prepped to verify the results.


Question: probabilistic risk assessment in nuclear licensing

This includes identification of how the baseline PRA model is modified for use in the configuration risk management program (CRMP) tools, quality requirements applied to the PRA models and CRMP tools, consistency of calculated results from the PRA model and the CRMP tools, and training and qualification programs applicable to personnel responsible for development and use of the CRMP tools.


Question: What are compensatory measures in the context of nuclear licensing?

The enclosure outlines the compensatory measures implemented to ensure safety compliance and operational integrity.

The Docket 030-30429 refers to various licensing aspects including inspections and assessments of safety-related systems.


Question: The NRC mandates that license applicants must provide safety analyses for their proposed reactors.

The regulatory framework requires that each license application for a nuclear facility includes a detailed assessment of safety risks and mitigation strategies.

Adobe Reader allows users to view, print, and annotate PDF documents, ensuring a consistent experience across multiple platforms.


Question: status of abandoned wells

H25 is classified as Temporarily Abandoned or Inactive.

H5 is labeled as Active, indicating it is currently producing.


Question: The analysis confirms that the subcritical condition of undamaged package arrays is maintained as long as the UO2 payload does not exceed 60 kg per ICCA.

Figure 6.15 demonstrates that an undamaged NPC package array of unlimited size (5N = ∞) remains subcritical provided the UO2 equivalent payload is restricted to 60 kg per ICCA.

It is noted that prior evaluations have demonstrated that mixtures with the 60 kg UO2 mass limit and a varying amount of H2O are more reactive than mixtures that contain 60 kg total weight of UO2 and H2O.


Question: Westinghouse has established a framework to classify and protect proprietary information crucial for maintaining competitive advantages.

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information.


Question: What is the purpose of the Turbine Trip - Turbine Stop Valve Closure trip Function?

The Turbine Trip - Turbine Stop Valve Closure trip Function anticipates the loss of heat removal capabilities of the secondary system following a turbine trip.

The LCO requires three channels of Turbine Trip – Low Fluid Oil Pressure to be OPERABLE in MODE 1 above P-9.


Question: The system that helps maintain core temperature during an emergency is critical for reactor safety.

Passive residual heat removal system actuation is a necessary action in response to inadequate core cooling.

The condensate system plays an important role in the overall water management of the plant but is not directly related to core cooling.


Question: The lessons gained during the pandemic will enhance future NRC oversight during emergencies.

NRC inspectors, staff, and management learned a great deal in a short time and the lessons learned and best practices identified should prove highly beneficial for future emergencies that limit or prevent access to nuclear plant sites.

NRC oversight of operating reactors during the COVID-19 PHE was appropriate, mostly as a result of the NRC staff and management’s ability to adapt to telework and the challenges the PHE presented.


Question: COVID-19 impact on nuclear plant work hours

The application provided the following information: A statement that explains how, and for which covered groups at Hatch, the COVID-19 PHE impacts the licensee’s ability to meet the work-hour control requirements of 10 CFR 26.205(d)(1) through (d)(7);

The U.S. Nuclear Regulatory Commission (NRC) has approved the requested exemption from specific requirements of Title 10 of the Code of Federal Regulations (10 CFR) 26, “Fitness for Duty Programs,” Section 26.205, “Work hours,” for Edwin I. Hatch Nuclear Plant, Units Nos. 1 and 2 (Hatch).


Question: What are the regulatory requirements for technical specifications in nuclear plants?

10 CFR 50.36(c) requires the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation, (3) surveillance requirements; (4) design features; and (5) administrative controls. The TS changes described in this amendment request are related to categories (1) and (5). The supporting assessments described herein ensure 10 CFR 50.36 requirements continue to be met through the specification of appropriate safety limits and limiting safety systems settings, and the establishment of administrative controls.

The report summarizing the results of the MPS3 AREA analysis for GAIA fuel are available for USNRC audit.


Question: Framatome Inc. license details


Question: What measures are taken to control personnel contamination in nuclear facilities?

Employees will maintain change rooms, showers, and lockers for clean clothing. An operable and appropriately calibrated alpha survey meter will be made available for employee use at the exit of the change room.

Uranium processing equipment that must be removed for maintenance or repair will be thoroughly decontaminated to prevent the possibility of contamination in the maintenance shop.


Question: What is the total concentration of cyanide measured in the sample?

The cation-anion balance was calculated to be -2.4%.


Question: common Quality Assurance Program

Farley, Hatch, and VEGP 1&2 currently use a common QAP (the QATR); the proposed change will integrate VEGP 3&4’s QAP commitments (from the NDQAM into the fleet’s QATR).

Updates the fleet QAP (QATR) to incorporate improvements in formatting and content.


Question: The presence of warping indicates potential geological instability in certain formations.


Question: What is a float charge in relation to battery systems in nuclear power plants?

Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady-state loads of the associated DC subsystem.

The AOT to bring the unit to MODE 5 is consistent with the time required in RG 1.93.


Question: The samples were assessed based upon their dry weight to ensure compliance with the standards set forth in the analytical procedures.

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

The following samples were analyzed using the above methods and analytical procedure(s).


Question: The analysis of the withdrawal of control rods is essential to understanding the reactor's response during abnormal conditions.

The spurious withdrawal of reactivity control system control rods analysis was included in the PIRT.

The initial steady-state conditions are shown in Table 1.


Question: maximum vent time in emergency operating procedures

The maximum vent time is a critical parameter defined in EOP-FRCI-3, ensuring the safe management of reactor operations during emergencies.

In normal operations, all systems must be regularly checked for compliance with the established safety protocols.


Question: The NRC requires that deviations and failures are documented as Conditions Adverse to Quality and addressed promptly according to regulations.

NRC regulation 10 CFR Part 50, Appendix B, Criterion XVI, states that “measures shall be established to ensure that Conditions Adverse to Quality (CAQ), such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.”

Based on the samples reviewed, the team determined that licensee performance in the use of Operating Experience (OE) and Self-Assessments and Audits adequately supported nuclear safety.


Question: What is the function of Relief and Low-Low Set (LLS) Instrumentation in nuclear safety?

Relief: Prevent vessel overpressure; LLS: Lower Relief Setpoint on three setpoint groups causing 1 (low), 1 (medium), or 4 (high) group LLS SRVs to remain open longer to prevent multiple actuations and postulated pressure loads on containment.

For the relief function, division specific SRV pilot SOVs and supports will be doubled to account for additional SRV cycles.


Question: What is the CIM Technical Report in the context of nuclear licensing?

In July 2010, Revision 2 of the CIM Technical Report, WCAP-17179 (ML102170259) was docketed as part of the AP1000 PMS design.

As stated in MSFIS SER (Reference 20), the NRC considered the low level of complexity of the design as part of their safety evaluation.


Question: nuclear energy peak analysis


Question: What are the essential factors for ensuring the structural integrity of graphite in a nuclear reactor environment?

The qualification approach for structural graphite to be used in a KP-FHR includes three basic elements: Confirmation of unirradiated graphite mechanical and thermal properties, Confirmation that the graphite will maintain structural integrity within a neutron flux; and Confirmation that the grade is compatible with a Flibe coolant environment.

In December 2001, NRC published "Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High Burnup Fuel," which identifies and ranks the phenomena occurring during selected transient and accident scenarios in both pressurized water reactors and boiling water reactors containing high burnup fuel.


Question: re-activation of SRO license

You have been directed to re-activate your license during the 1st quarter to stand the Shift Manager watch on April 4, 2019.

The shift schedule for the month of April has been updated to reflect the new staffing assignments.


Question: significant change approval process

**Recommendation:** Approve the significant change.

Please contact George Odom, Chief, Engineering and Technology Section, Policy and Doctrine Branch, Technological Hazards Division, (202) 802-6989 if you have any questions regarding this memorandum.


Question: What are the safety implications of a gas pipeline rupture near nuclear facilities?

The potential consequences to the McGuire facility from a gas pipeline rupture was considered. Two pipelines, one 36 inch diameter and one 42 inch diameter, located one mile south of the plant, are shown in Figure 2-36. A rupture of the 42 inch diameter pipe was postulated for the accident analysis.

Drinking water is now supplied by Charlotte Mecklenburg Utility Department.


Question: The NRC mandates evaluations of hydrologic conditions that could affect the facility's safety during operations.

Hydrologic failure conditions were assessed using Manning's Roughness Coefficients to determine the impact of land use on water flow and velocity metrics critical for safety analysis.

The site investigation revealed multiple aspects of ecological sensitivity, including the presence of endangered species within the study area which necessitated further environmental impact assessments.


Question: What are the eigenvalue comparisons for fuel assembly lattices?

Single Fuel Pin Cell Eigenvalue Comparisons (CASMO5-Serpent2) [pcm] for Enrichments 5, 8, and 10 wt. % U-235.

Generic PWR Assembly Lattice Pin Fission Rate Comparisons (CASMO5-Serpent2) for 6.3 wt. % U-235 at a Burnup of 0.0 GWd/MTU.


Question: The NRC mandates that licensees conduct evaluations based on the seismic probabilistic risk assessments for their facilities.

Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants is established under 10 CFR 50.69, which requires assessments that account for seismic risks and safety concerns.

The construction of nuclear facilities must comply with rigorous safety standards and regulatory inspections to ensure operational integrity.


Question: emergency preparedness organizational chart

East Feliciana Parish Emergency Preparedness Organizational Chart includes key figures such as the Parish Attorney and members of the police jury, mayors from various towns, and representatives from public safety departments.

The Parish Council on Aging works closely with local health units to provide services to the elderly population in the area.


Question: nuclear regulatory exams

Modified Bank #76 DCPP L061 NRC 10/2016  X


Question: The enhancement of procedures for managing buried piping and tanks can significantly impact long-term operational safety.

Continue the existing DCPP Buried and Underground Piping and Tanks AMP, including enhancements to:  (a) Enhance operating procedures to provide direction to evaluate and close Makeup Water isolation valve MU-0-881 as appropriate in case of a pressure boundary failure further along the flow path or in the event the RWSRs are in use for long-term cooling;  (b) Install impressed current cathodic protection for remaining portions of the buried auxiliary saltwater system discharge and supply piping in contact with soil; (c) Revise DCPP backfill procedure to include the guidance in LR-ISG-2015-01, Appendix B, Section 2.f, including a maximum size that meets ASTM D 448-08 size number 67 (size number 10 for polymeric materials) for backfill that is located within 6 inches of the component.

Enhancements are implemented by 11/02/2024 <br> Initial inspections completed and additional cathodic protection installed by 12/01/2028.


Question: What is the purpose of the Structures Monitoring program in nuclear licensing?

The Structures Monitoring program (B.2.1.33) will manage the aging effect of corrosion for pilings exposed to groundwater/soil or raw water.

The option to perform a fatigue waiver was used in lieu of managing for cracking due to cyclic loading with the 10 CFR Part 50, Appendix J program (B.2.1.31) and ASME Section XI, Subsection IWE program (B.2.1.29) via supplemental surface examinations or leak rate test.


Question: What is the significance of population distribution in nuclear licensing?

Population distribution data is crucial in assessing the potential impact of a nuclear facility on nearby communities, and it informs the development of emergency planning for evacuation and safety protocols.

The table shows the values of population distribution in various directions, measured in a scale of miles.


Question: The NRC has established various guidelines concerning the interactions and responsibilities of licensees in regulatory discussions.

The NRC requires licensees to analyze emergency zones in their safety assessments to ensure that potential impacts on the surrounding population are adequately addressed.

It is important to ensure that all participants are available for detailed discussions, which have included numerous subcommittee meetings and presentations.


Question: The failure of rod position indicators requires thorough investigation, as outlined in the job performance measure.

The task involves responding to a Rod Position Indication failure, which is critical for ensuring the safety and operational reliability of the reactor.

The estimated time to complete the performance evaluation was set to 20 minutes, ensuring efficient assessment of operator skills.


Question: A public hearing may be held if the Director determines there is significant public interest in a permit application or draft permit.

The Director shall hold a public hearing whenever it is found, on the basis of hearing requests, that there exists a significant degree of public interest in a permit application or draft permit.

The Environmental Management Commission meets regularly to review environmental regulations and policies.


Question: The licensing process for nuclear facilities involves thorough safety evaluations by the NRC.

The Nuclear Regulatory Commission (NRC) reviews and approves applications for nuclear reactor licenses based on comprehensive safety assessments and regulatory compliance.

Reactor coolant systems are essential for maintaining safe and stable operational conditions throughout various phases of reactor operation.


Question: What is the purpose of the Control Rod Operability during refueling operations?

Control rods are essential for regulating the fission process in a nuclear reactor, and ensuring their operability during refueling operations is critical for maintaining safety and operational integrity.

The thermal performance of the reactor coolant system directly influences the efficiency of heat exchange, impacting overall plant performance.


Question: Xcel Energy is committed to ensuring that operational changes comply with regulations and minimize environmental impacts during the license renewal period.

Xcel Energy finds that impacts of the effects on terrestrial resources (non-cooling system impacts) for the proposed SPEO are SMALL.

The listed species on or in the vicinity of nuclear power plants also range widely, depending on numerous factors such as the plant location and habitat types present.


Question: weighted mean activity of nuclides

Wt mean Activity (µCi/grams) refers to the weighted mean of the activity concentration of a nuclide expressed in microcuries per gram.

Confidence values are used to indicate the reliability of the nuclide identification and the activity measurements.


Question: What are the space requirements for the Independent Spent Fuel Storage Installations (ISFSIs) at the facility?

The station has two separate ISFSIs. The East ISFSI is comprised of two sections and has space for 10 additional casks. The West ISFSI is one pad and has space for 13 additional casks.

Drawings and/or photos that are highlighted/marked showing the flow paths for releases for both radiological and non-radiological waste paths.


Question: What is the significance of weighted mean activity for nuclides in regulatory filings?

The weighted mean activity is an important parameter for assessing the radiological impact of nuclides, expressed in pCi/grams, as it combines the activity levels with their associated uncertainties to provide a meaningful representation of their radiological characteristics.

Errors quoted at 2.000sigma are standard deviations used to indicate the uncertainty in the measurements and calculations associated with nuclear regulatory assessments.


Question: The Tertiary Aquifer plays a crucial role in understanding how radionuclides move through different geological formations.

The Tertiary Aquifer is a significant geological feature that can affect the transport of radionuclides in the subsurface environment.

Erosion of geological formations can lead to significant changes in surface hydrology and ecological systems.


Question: Licensees must comply with the NRC regulations regarding the approval of safety-related modifications.

The NRC regulations specify that any modifications to safety-related systems must undergo a rigorous approval process to ensure compliance with safety standards.

The report discusses the financial implications of the recent market changes affecting the nuclear industry, outlining potential impacts on investment returns.


Question: What is the high water level trip in nuclear reactors?

Automatic Initiation: Reactor Vessel Water Level (L8) Trip, Feedwater and Main Turbine high water level trip.

1). Automatic Initiation: – APRM/Oscillation Power Range Monitor Upscale Trip, Vessel Water Level (L8) Trip, Turbine TSV Trip.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code provides a detailed methodology for calculating the critical heat flux in nuclear reactors under various operational conditions, ensuring safety during thermal-engineering analyses.

The report includes extensive simulations that analyze thermal-hydraulic performance under a variety of grid configurations, which are essential for validating reactor design features.


Question: What is the purpose of the 125V dc power system in a nuclear station?

The dc power system is rated at 125V to provide a highly reliable source of power for the operation of vital safety and non-safety-related equipment necessary for the proper and safe operation of the station, as well as safe reactor shutdown under all postulated accident conditions.

Batteries are contained in separate missile-protected battery rooms. All rooms are located in the service building, as shown in Reference Drawings 17 and 18.


Question: dam break analysis

The Southern Dam Break Plan includes detailed hydrological modeling and risk assessment related to potential dam failures and their impacts on downstream areas.

During regulatory evaluations, sound engineering principles are applied to ensure that all structures comply with established safety standards.


Question: reactor containment building leakage rate test

Reactor Containment Building Integrated Leakage Rate Test

Notification of Field Work Complete for EC 21-4916-2


Question: ground water requirements for nuclear facility

The expected average consumption of ground water for these uses is approximately 1,310 gpm (Table 1.3-1). The expected maximum consumption of ground water for these uses is approximately 3,570 gpm (Table 1.3-1).

Ground water samples were obtained from test and observation wells on site and from selected private wells within the limits of the well survey.


Question: nuclear power fractional of nominal

The process for submitting a license application requires the applicant to provide detailed information about the intended use of the nuclear material.


Question: What is the significance of nitrate concentration in groundwater monitoring?

The table shows the nitrate concentration in groundwater from various monitoring wells at the Westinghouse Columbia Fuel Fabrication Facility, indicating the extent of nitrate presence in the aquifer system.

Mill Creek flow direction is influenced by the geography of the site, which impacts local hydrology and environmental considerations.


Question: proposed hearing schedule for regulatory filings

TVA respectfully submits the following proposed schedule for consideration by the Board, assuming adjudication before this Board of all four Violations.

TVA understands that the Staff will be filing its own proposed schedule.


Question: What is a Part 21 notification in nuclear regulatory context?

The verification of the PCV did not include a review for Part 21 notifications.

On December 16, 2017, Exelon failed to verify that the PCV met all purchase order requirements as required by SM-AA-102 and accepted the defective part upon receipt of purchase order 637236 at FitzPatrick.


Question: What are the specifics of sample size in gamma spectrum analysis?

The sample size for the analysis is 8.156E+02 grams.

The operator for the analysis was noted as 'Administrator'.


Question: seismic hazard assessment in nuclear power plants

The NRC evaluates the potential effects of natural hazards, including seismic events, on nuclear power plants on an ongoing basis that is separate from the license renewal process.

The natural Hayesville and Cecil soils are somewhat limited for site development involving shallow excavations due to high clay content and unstable excavation walls.


Question: what is a pre-charge sequence in a reactor recirculation system?

Answer Explanation: QCOP 0202-43, REACTOR RECIRCULATION SYSTEM STARTUP, step F.6 Note describes the Pre-Charge Sequence as follows:

1. RECIRC ASD 1(A)/(2)(B) BKR ‘OFF’ and ‘TRIP’ indicating lights are illuminated.
2. PRE-CHARGE STATUS ‘READY’ light is extinguished and ‘IN PROGRESS’ light is illuminated.
3. RECIRC ASD 1(A)/(2)(B) ‘OFF’ and ‘TRIP’ lights are extinguished, and ‘ON’ light is illuminated.
4. PRE-CHARGE STATUS ‘IN PROGRESS’ light is extinguished and ‘COMP’ light is illuminated.

Unit 1 is shutdown. The indications below are observed when the RECIRC ASD 1A control switch is momentarily placed to START, then returned to NORMAL:

- RECIRC ASD ‘OFF’ and ‘TRIP’ indicating lights are illuminated.
- PRE-CHARGE STATUS ‘READY’ light extinguishes and ‘IN PROGRESS’ light illuminates.
- RECIRC ASD ‘OFF’ and ‘TRIP’ lights extinguishes and ‘ON’ light illuminates.
- PRE-CHARGE STATUS ‘IN PROGRESS’ light extinguishes.
- PRE-CHARGE STATUS ‘COMP’ light illuminates and then extinguishes.


Question: What regulations govern the use of sources of radiation?

It shall be unlawful for any person to use, manufacture, produce, distribute, sell, transport, transfer, install, repair, receive, acquire, own or possess any source of radiation unless licensed by or registered with the agency in conformance with rules and regulations, if any, promulgated in accordance with the provisions of this act.

The agency shall have the authority in the event of an emergency to impound or order the impounding of sources of radiation in the possession of any person who is not equipped to observe or fails to observe the provisions of this act or any rules or regulations issued thereunder.


Question: maximum annual average site boundary undepleted decay x/Q value

The maximum annual average site boundary undepleted/2.26-day decay x/Q value for use in determining gaseous pathway doses to the maximally exposed individual.

The maximum annual average resident undepleted/no decay atmospheric dispersion factor (x/Q) value for use in determining gaseous pathway doses to the maximally exposed individual.


Question: procurement documents compliance

b. **Describe** implementation to assure purchased materials, items, equipment, and software services conform to conform to the procurement documents:

c. Are procedural controls adequate and procedure revision current?
   - **YES** [X] **OR NO** [ ] (describe in Findings/Deficiencies current section above)


Question: What is the role of End of Cycle Recirculation Pump Trip (EOC-RPT) instrumentation in nuclear safety?

Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE: 1. Turbine Throttle Valve (TTV) - Closure; and 2. Turbine Governor Valve (TGV) Fast Closure, Trip Oil Pressure - Low.

In accordance with the Risk Informed Completion Time Program.


Question: The NRC mandates that a review of potential tornado missile impacts on spent fuel storage facilities must be conducted during safety assessments.

An evaluation has been performed to determine the maximum damage to spent fuel elements stored in the spent fuel storage pool in the event a tornado missile enters the pool through the North end of the building.

For a discussion of the ventilation and filtration system, see Sections 9.4.2.2 (paragraphs 1 & 2), 9.4.2.3 (paragraph 2), and 5.4.5.


Question: logical connectors in regulatory actions

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector 'OR' and the left justified placement.

The radiological impact of a postulated accident scenario is a key consideration for the evaluation of the overall safety of the reactor design and its adherence to regulatory standards.


Question: The NRC mandates that samples be analyzed for various ions, including potassium, in accordance with established methods.


Question: soil sample analysis requirements

The selection of two (2) soil samples (L1-010-104-FSGS-006-SS and L1-010-104-FJGS-016-SS) met the requirement that a minimum of 10% of the samples collected for the FSS of survey unit L1-010-104 be analyzed for HTD ROC.

Table 5-6 provides a synopsis of the survey design for survey unit L1-010-104.


Question: maximum initial enrichment limits for nuclear fuel

The maximum initial enrichment (wt % 235U) for various assembly types is specified in the loading criteria for spent nuclear fuel (SNF) assemblies.

In order to ensure safe operations, reactors must also conform to environmental protection standards as outlined by the NRC.


Question: average enrichment of 235U in PWR fuel

Initial Assembly Average Enrichment wt % 235U (E) | WE 14x14 Assembly Average Burnup (B) GWd/MTU


Question: environmental resource documentation

We will leverage existing, under review and reviewed, and approved documentation to the extent possible to reduce schedule, costs, and review time for the appropriate environmental resources such as: Meteorology, Noise, Cultural and Historical, Infrastructure, Transportation, Socioeconomic, Etc.

We want to know if there can be consideration of assigning NRC staff who are familiar with previous, recent proximate reviews work on this review.


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX code utilizes the ORFEO-HMP Critical Heat Flux correlation to predict the critical heat flux in the reactor core under various conditions, essential for safety performance assessments.

Table B-1 outlines the test assembly conditions for non-mixing grid tests, detailing parameters such as input flow rates and temperature measurements without reference to critical heat flux.


Question: The energy levels measured for the various peaks indicate the specific interactions taking place within the reactor core.

Peak No. F 1 corresponds to an energy of 238.64 keV, with a net peak area of 21.24 and an area uncertainty of 2.36E+02.

The reactor design must include redundancies to ensure that all safety protocols are met during normal operations.


Question: What are the regulatory challenges for constructing new nuclear plants in Illinois?

In 1987, the Illinois State Legislature passed legislation prohibiting the construction of any new nuclear power plant within Illinois after September 11, 1987 (220 ILCS 5/8-406).

CEG has several fossil fuel-generating plants (natural gas-fired, oil-fired, and dual natural gas- and oil-fired) (CEG 2022e).


Question: The operation of the main feedwater isolation valves must be maintained during specified reactor modes.

Three main feedwater isolation valves (FIVs) shall be OPERABLE in Modes 1, 2, and 3, unless certain conditions are met, such as being closed and deactivated or isolated by a manual valve.

The reactor design includes multiple redundancy systems to enhance overall safety and operate under expected operational conditions.


Question: The evaluation of applications for licenses involves detailed assessments of safety measures and technical specifications to ensure compliance with nuclear regulatory standards.

The NRC's licensing process requires comprehensive safety evaluations and oversight to ensure that plant operations adhere to the established safety regulations and ensure public health protections.

Many aspects of environmental monitoring are considered when assessing the broader impacts of the plant's operations on surrounding ecosystems.


Question: dry weight basis

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 the ANSI/ANS-3.4 disqualifying condition?

Make a clear link to ANSI/ANS-3.4[15.4] conditions/requirements in the submittal.

What are the related medical examination results? (readings, laboratory data, physician observations)


Question: debris generation zone of influence

In the Wolf Creek debris generation analysis, the ZOI for a DEGB is defined as a spherical volume centered at the break in which the jet pressure is higher than the destruction/damage pressure for a certain type of insulation, coatings, or other materials impacted by the break jet.

Identify which debris analyses used approved methodology default values. For debris with ZOIs not defined in the guidance report/SE, or if using other than default values, discuss method(s) used to determine ZOI and the basis for each.


Question: The NRC provides guidance for the assessment of internal flood hazards, stating that pipe breaks outside containment do not generate debris inside containment.

The quantitative risk assessment was performed for LOCAs and SSBIs that occur during full power operation (i.e., Mode 1), which is assumed to be equivalent or bounding compared to the other operating modes.

Consistent with the guidance in NUREG/CR-6850 (Reference 15), internal fire hazards are not assumed to result in pipe breaks.


Question: The NRC can grant exceptions to existing license conditions under special circumstances that do not compromise safety.

The NRC finds that the license condition, for which exemptions are sought, could not be legally met without contravening social distancing requirements and, as such, unnecessarily risking possible COVID-19 infection to persons involved in the performance of the required actions.

An environmental assessment for this action is not required, since this action is categorically excluded under 10 CFR 51.22(c)(14)(v).


Question: What is the heat transfer coefficient at PCT and burst elevations?

The heat transfer coefficients at various times recorded at the cladding surface are significant for understanding thermal aspects of reactor safety, particularly at the PCT elevation of 7.25 ft and burst elevation of 6.25 ft.

The historical data showed fluctuations in the operational capacity, highlighting the importance of consistent maintenance and monitoring of nuclear power stations.


Question: What is peak cladding temperature in nuclear reactor analysis?

The Peak Cladding Temperature is a critical factor in the analysis of Large Break Loss of Coolant Accidents (LOCA), impacting the integrity of the nuclear reactor's fuel cladding.

The table provides a series of mesh point temperatures measured in degrees Fahrenheit alongside corresponding time durations measured in seconds, illustrating numerical results of temperature changes.


Question: containment heat removal systems

Heat exchangers that remove energy from the containment sump are modeled with the available heat exchanger options in GOTHIC. Use of a GOTHIC heat exchanger option dynamically couples the heat exchanger performance to the predicted primary and secondary fluid conditions.

The containment depressurization system consists of the containment QS subsystem and the containment RS subsystem, with casing cooling.


Question: The manual intervention options during pressurizer pressure fluctuations do impact system responses.

When pressurizer pressure increases above the P-11 setpoint, low streamline pressure protection is automatically reinstated and steam pressure protection is removed.

The analyses in Chapter 15 of the steam break accidents and an evaluation of the protection system instrumentation and channel design shows that the Engineered Safety Features Actuation System is effective in preventing or mitigating the effects of a steam break accident.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code employs the ORFEO-HMP correlation to predict the critical heat flux conditions in a reactor, which is essential for ensuring safe operation during thermal transients.

The table B-3 provides local conditions determined from COBRA-FLX simulations that were conducted for various grid tests across different operational scenarios.


Question: FLEX portable equipment failure rates

Two sensitivities have been performed regarding the sources of uncertainty associated with the FLEX modeling in the RICT program.

If procedures for initiating mitigating strategies are not explicit and the associated failure probabilities are not directly analyzed by accepted approaches, technical bases for probability of failure to initiate mitigating strategies should be submitted to NRC for review.


Question: The SLCS provides an alternative method for bringing the reactor to cold subcritical conditions without relying on control rods.

The SLCS is a backup method of shutting down the reactor to cold subcritical conditions by independent means other than the normal method using the control rods.

RG 1.97, Revision 2 is the licensing basis for LGS.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves a series of regulatory steps where applicants must demonstrate their compliance with safety standards set by the NRC, including systematic reviews of design specifications and operational plans.

The coolant system in the reactor is designed to maintain optimal temperatures, ensuring efficient heat removal during both normal and emergency conditions, thus protecting structural integrity.


Question: The distribution of the population living within 50 miles of the nuclear facility is a critical factor in assessing emergency preparedness.

The total resident population between 10 and 50 miles of the PSEG Site is projected to grow, reaching approximately 8,138,635 by 2081.

The table displays various metrics related to reactor safety and performance, including thermal efficiency and operational readiness.


Question: How are the upper and lower limits for nominal trip setpoints in voltage relay systems determined?

The upper limit NTSP for the Division I & II Loss of Voltage Relays’ undervoltage function is as follows: NTSP_U-LO(UV) = 91.5 V

These Voltage Drop (ΔV) values are used in Input 4.5.2 when converting the results from the 4kV to the 120V levels and vice versa.


Question: What are the responsibilities of a Radiation Protection Technician in an emergency response organization?

B. **Radiation Protection Technician**  - Radiological Monitoring Activities  - ERO Radiological Protection  - Emergency Exposure  - Contamination Control Measures

1. **Control Room (CR)**  A. **Shift Manager**  - Organizational Interface and Coordination  - Command and Control  - Facility/Group Management and Supervision  - Contact and Use of External Support Services  - Use of Medical, Fire and Law Enforcement Support  - NRC Notification and Communications  - Event Classification  - ERO Notification  - State and Local Event Notification  - ERF Communications  - Accident Detection and Assessment  - Effluent Release and Dose Assessment  - OSC Team Priorities, Dispatch and Control  - Site Assembly and Accountability  - Site Evacuation.


Question: The construction permit for the TerraPower Kemmerer Power Station Unit 1 is being reviewed for compliance with safety regulations.

The Nuclear Regulatory Commission (NRC) conducts thorough reviews of construction permits to ensure that proposed nuclear facilities meet technical and safety standards before commencement of construction.

The report also indicates that discussions were held about the latest advancements in renewable energy technologies and their impact on the regional power grid.


Question: The NRC evaluates the use of carbon-free energy technologies in combination with demand-side management as a reasonable alternative for energy generation.

The NRC staff considers a combination of carbon-free replacement power generation technologies with demand-side management to also be a reasonable alternative to Oconee Station’s SLR.

Solar photovoltaic power generation uses solar panels to convert solar radiation into usable electricity.


Question: What are the requirements for non-destructive examination of welds in nuclear fabrication?

Progressive PT examination at 2-inch intervals. Final PT again of all welds or repair areas after PWHT (NB 5120) and after hydro IAW NB 5410. PT procedure IAW Section V Article 7 and acceptance criteria IAW NB 5350. PT exam records show acceptable exam results.

Fabrication Process Control Sheets. Non-conformance reports (NCRs). Weld procedures and procedure qualification records. VT and PT examination records.


Question: What is the purpose of the Safety Function Determination Program (SFDP) in nuclear regulatory context?

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.


Question: The criteria for determining the emergency class level rely on systematically established action levels and facility conditions.

A standard emergency classification and action level scheme is established and maintained. The scheme provides detailed EALs for each of the four ECLs in Section IV.C.1 of Appendix E to 10 CFR Part 50.

The PSL EAL scheme is documented in EP-PSL-121, PSL Emergency Action Level Technical Basis Manual.


Question: The NRC publications outline the required analytical procedures for reviewing sample data.

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.

The samples in this SDG were analyzed on an "as received" basis.


Question: What is the significance of UO2 in nuclear fuel performance?

Figure 4-26 presents the calculated versus measured rod diameter increase for UO2 fuel, illustrating important thermal and mechanical properties that affect reactor performance.

The measured rod diameter increase in Zircaloy-4 reflects its structural integrity under irradiation conditions.


Question: storm surge impact on nuclear facilities

Storm surge is a rise in offshore water elevations caused principally by the shear force of winds acting on the water surfaces, typically associated with tropical storms (Reference 2.4-1, Section 3.5). The Cooper Nuclear Station (CNS) is not near any large bodies of water for which storm surge flooding would apply; therefore, the risk to the plant from a storm surge event that could cause flooding at CNS is not expected to be a potential flooding hazard. According to ANSI/ANS-2.8-1992 guidance (Reference 2.4-2), the region of occurrence of a hurricane shall be considered for United States coastline areas and areas within 100 to 200 miles bordering the Gulf of Mexico. Because CNS is located far from any coastline and at a site grade of 903.37 ft NAVD88, hurricanes do not present potential flooding hazards. Therefore, storm surge was screened out as a credible flooding mechanism for CNS.

The Cooper Nuclear Station is designed with multiple safety systems that include redundant safety features to address any potential operational issues during normal and abnormal operations. This ensures that the reactor operates within safe parameters, minimizing the risk of accidents.


Question: Loss of material due to pitting

Loss of material due to pitting, crevice corrosion, MIC is an aging effect requiring management for piping components made from copper alloy in treated water environments.

Piping components made from PVC are used in indoor uncontrolled air environments and show no aging effects requiring management.


Question: contention admissibility standards

When a hearing requestor, such as Petitioners here, seeks admission of a contention submitted as part of a timely intervention petition, that contention must satisfy the six admissibility factors set forth in section 2.309(f)(1).

We thus conclude that each of the Petitioners has established its representational standing to intervene in this proceeding.


Question: What are the effects of pitting and crevice corrosion on stainless steel piping?

Loss of material due to pitting and crevice corrosion

Not applicable. There is no concrete or cementitious material piping and piping components exposed to raw water in the Auxiliary Systems.


Question: The NRC mandates licensees to conduct comprehensive analyses of radioactive materials present at their facilities.

The parameter for Americium-241 is measured in pCi/g, signifying its concentration in samples analyzed during safety assessments.

LaCrosseSolutions is located at S4601 State Hwy 35, Genoa, Wisconsin, and was involved in the LACBWR Site Restoration Project.


Question: ECCS valve testing requirements

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

In accordance with the Surveillance Frequency Control Program.


Question: The analysis of spectrum data involves subtracting background counts from the original area to derive accurate measurements of net peak areas.

The net peak area is calculated by subtracting the ambient background counts from the original area measurements of the peaks recorded in a spectrum analysis.

The energy values in the table represent distinct peaks that are measured in kiloelectronvolts, which correspond to specific radioactive isotopes.


Question: What happens to reactor water level with a failed RCIC pump discharge flow?

With the RCIC pump discharge flow element failed low, the flow input signal to the flow controller lowers. The flow controller senses that flow is below the demanded flow of 300 gpm and further opens the RCIC throttle valve.

A. Incorrect – Reactor water level rises, not lowers. Plausible because indicated flow is low, which would correlate to lowering Reactor water level. Also plausible because this would be correct on the opposite failure.


Question: What is the purpose of the GAIN initiative in the nuclear sector?

The Gateway for Accelerated Innovation in Nuclear (GAIN) initiative was established to provide the nuclear industry with access to the resources and expertise needed to facilitate the faster commercialization of innovative nuclear technologies.

Virtual meetings have become increasingly popular as a means to engage remote participants and discuss critical issues without the need for physical presence.


Question: To prevent deterioration of reactor components, licensees must manage aging effects, including those affecting the reactor vessel's integrity.

Cumulative fatigue damage; cracking due to fatigue, cyclic loading

The structural integrity of the reactor is maintained through regular inspections and adherence to safety protocols.


Question: The NRC mandates an evaluation of changes to assess their impact on reactor structural components and systems.

Identify if change affects materials, environment, loading, degradation, geometry, maintenance or mitigation

Focus on RPV surveillance, PT limits, USE, PTS, LBB, piping materials and supports, chemical control, etc.


Question: nuclide activity in regulatory filings

The activity of a nuclide is expressed in terms of its decay rate, typically measured in picocuries (pCi) per gram of the nuclide.

The temperature rise in the reactor core is critical to maintaining the fuel integrity during operational periods.


Question: ITAAC notification types in regulatory filings

ITAAC Notification Type (ICN, IPCN, UIN) refers to different classifications used in the notification process for Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in nuclear regulatory filings.

The Federal Register is a daily journal of the U.S. government that contains government agency rules, proposed rules, and public notices.


Question: guide lug inserts in reactor design

The staff did not accept Comment #19. Based on the staff's rejection of NEI Comment #1, the staff cannot use MRP-191, Rev. 2 as a basis for citing component-specific aging effects or mechanisms in this ISG. Furthermore, in MRP-227, Rev. 1, Table 3-2, the guide lugs and guide lugs inserts and bolts, are screened in as Existing Program components (i.e., “X” designations) for the mechanisms of fatigue, wear, and radiation-enhanced stress relaxation or creep (ISR/C), even though the EPRI MRP did not specifically reflect fatigue as a cited mechanism in the Existing Program line items for the components in Line Items C13 and C14 in Table 4-8 of the MRP-227, Rev. 1-A.

The applicant could also apply the "RP-320" item as a consistent-GALL item for only some of the referenced guide lug fixture components in the line item if that is consistent with the IPA basis.


Question: BWR Fuel Region Homogenization

The BWR Fuel Region Homogenization Sample Calculation provides detailed information about the area and volume fraction of components such as fuel, void, clad, and interstitial in a boiling water reactor's fuel region.

The Upper Nozzle has a mass of 2.08 kg/assy and a volume of 2.6196E+02 cm³, with a modeled height of 19.0500 cm and a volume fraction of 7.0003E-02.


Question: What process is necessary for nuclear licensing in the United States?

The nuclear licensing process involves several key steps, including the submission of a license application that must demonstrate compliance with safety regulations and technical requirements set forth by the NRC.

Submitting financial documents is essential for the overall success of any project, ensuring that all parties understand their responsibilities and obligations.


Question: What is the role of Bechtel Power Corporation in nuclear power projects?

Bechtel has a history of supporting the nuclear power industry, beginning with the construction in 1950 of the EBR-1 reactor.

MACTEC Engineering and Consulting, Inc. performed geotechnical field investigations and laboratory testing in support of SSAR Section 2.5, Geology, Seismology, and Geotechnical Engineering.


Question: The analysis of uranium isotopes, including Uranium-238 and Uranium-235, is critical for understanding radioactive materials in nuclear licensing processes.

The target analyte was detected in the associated blank.


Question: The NRC requires strict adherence to regulatory processes for the submission of safety analyses and environmental assessments.

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

The city of Genoa, Wisconsin is known for its beautiful parks and recreational facilities, attracting many visitors.


Question: What is the significance of cracking in auxiliary buildings as it relates to seismic PRA?

Seismic F&O 2-1 remains open because the potential for cracking in the auxiliary building has not been fully dispositioned.

Provide a sensitivity study to evaluate the impact of NSSS stiffness on the overall structure and seismic PRA.


Question: What is the requirement for the annual registration of generally licensed devices?

The U.S. Nuclear Regulatory Commission (NRC) requires annual registration of certain devices that are possessed under the general license issued in Section 31.5 of Title 10 U.S. Code of Federal Regulations (10 CFR 31.5).

An invoice for the current amount due will be sent to you under separate cover.


Question: The OHSEP will coordinate with various local agencies and state service organizations to ensure integrated responses are effective.

The Parish Office of Homeland Security and Emergency Preparedness will provide overall direction and control for the Parish items listed in Section 1.0.

The Emergency Operations Center Supplies shall include all necessary materials to effectively support emergency response activities.


Question: The NRC requires a review of the steam generator weld examinations carried out at Arkansas Nuclear One, due to concerns over potential degradation.

The purpose of this meeting was to discuss proposed alternative examination methods for the steam generator welds at Arkansas Nuclear One, Units 1 and 2, in accordance with regulatory requirements.

The financial outcomes of the proposed projects will be evaluated based on their expected returns and risks associated with market fluctuations.


Question: Cultural Resources Report

Pursuant to 10 CFR 2.390, “Public inspections, exemptions, requests for withholding,” Enclosure 2, the Class III Cultural Resources Report, is withheld from public disclosure and may not be made publicly available.

Construction of the transmission and water supply lines will have an adverse effect on 48LN8940 because this area will be graded, disturbed, and modified for the construction and installation of these utilities.


Question: nuclear power plant safety

Ray insisted that the AEC believed “that the care taken in design and operation ensures that the chances of a serious accident happening at a nuclear plant are very small.”

AEC critics were undoubtedly dissatisfied with Ray’s assertions, and they were certainly disappointed when Senator Jackson’s S. 2135 bill to create the DENR lost steam.


Question: radiological emergency response procedures

The purpose of this procedure is to provide guidelines and delineate the responsibilities of the Director of the Office of Homeland Security and Emergency Preparedness (Parish EOC Director) in response to radiological emergencies at River Bend Station.

The Director of the Office of Homeland Security and Emergency Preparedness will ensure warning and notification of persons within the 10 mile EPZ, as necessary.


Question: archaeological site assessment

Determination of Eligibility, Phase III complete


Question: cultural resources management plan

In 1991, a cultural resources environmental assessment and management plan for the Unit 1 project area was completed. The report includes the results of the cultural resource evaluation, which includes a search of archaeological records within the project area. The search identified 37 previously documented archaeological and historic sites. In addition, the report outlines a cultural resource management plan (CRMP) to preserve the cultural antiquities, historical properties, and traditional cultural properties within the Unit 1 project area. The principal objective of the CRMP is to avoid all cultural resources. (Marshall 1991)

Prior to the initiation of the Section 106 process, HRI solicited insight regarding historic preservation and mitigation from the tribal agencies (Navajo Natural Heritage Program and Navajo Nation Historic Preservation Department (NNHPD)) and tribes including the Acoma, Hopi, Laguna, Zuni, and All Indian Pueblo Council.


Question: The reactor must maintain adequate coolant flow when the turbine experiences a failure to avoid operational issues.

Whenever the turbine-generator unit trips at an operating power level above 50 percent power, the reactor also trips. A heat sink is required to remove heat stored in the reactor coolant to prevent actuation of steam generator safety valves for a trip from full power.

If a unit is operating below 67% guaranteed load, loss of one MFP has little or no effect on the unit, since one MFP is capable of delivering sufficient flow to maintain 67% load.


Question: What is the significance of Baseline Core Damage Frequency and Large Early Release Frequency in nuclear safety?

Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) are critical metrics used to evaluate the safety and risk assessments of nuclear power plants.

The report includes a detailed analysis of the operational efficiency of the cooling systems and their impact on overall plant productivity.


Question: Class II Cultural Resources Report

Pursuant to 10 CFR 2.390, “Public inspections, exemptions, requests for withholding,” the Class II Cultural Resources Report is withheld from public disclosure and may not be made publicly available.

If the adverse impacts cannot be sufficiently avoided or minimized, resolution of adverse effects will be considered pursuant to 36 CFR 800.6(b) and an MOA will be prepared consistent with 36 CFR 800.6(c) (see also 36 CFR 800.8(c)(4)).


Question: The NRC evaluates compliance with the National Ambient Air Quality Standards (NAAQS) based on monitored pollutant levels.

Measured Ozone Concentrations and Comparison with NAAQS indicate whether the pollutant levels are in compliance with the established standards.

The table lists various counties and their corresponding monitor numbers without indicating any regulatory requirements.


Question: The licensing process requires a thorough review of safety systems and controls to ensure compliance.

The Nuclear Regulatory Commission (NRC) requires a comprehensive safety analysis to support the licensing of nuclear facilities, which includes assessing safety-related systems and their operational reliability.

The meeting will cover various topics including the regulatory framework, safety assessments, and public engagement strategies.


Question: reactor coolant system boron and shutdown margin

EA 2.01 Reactor coolant system boron and/or shutdown margin (OE related) | 3.6 | 3.9

EA 2.05 Containment radiation | 2.6 | 2.9


Question: What are tolerance limits in nuclear safety analysis?

The tolerance limits are calculated by analyzing the requisite number of cases to achieve at least 95% probability and at least 95% confidence.

The statistical approach described in ANP-10339P-A (Reference 1) is followed to determine acceptance criteria.


Question: What is counting uncertainty in nuclear measurements?

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

The Report Date for this sample is November 14, 2019.


Question: In the event of an unexpected loss of power to essential systems, a manual scram may be initiated by the Unit Supervisor.

Unit 1 is operating at 100% power, when a total loss of 120 VAC ESS Bus occurs.

The NRC exam includes various questions related to reactor operations and emergency procedures.


Question: What is a harvesting project in the context of nuclear licensing?

The harvesting project aims to optimize the management of spent nuclear fuel and other materials for enhanced recovery and sustainability within the energy sector.

The team will present a comprehensive analysis of financial implications and projected cost savings resulting from the new operational strategies.


Question: The NRC provides the ability to appeal any final determinations made in response to requests, and users can also utilize the Agency's Public Liaison for assistance with the FOIA process.

You may appeal this final determination within 90 calendar days of the date of this response.

As modified, any emails from 11/14/2024-11/9/2024 that include 19 specified words/terms (plus 4 additional pejorative terms if any of the other 19 words/terms appeared), which would reflect upon government officials' reactions to the result of the 2024 election.


Question: what are ADS valves in nuclear reactors

The LCO requires five ADS valves to be OPERABLE in order to provide the ADS function. The ECCS analyses are performed with the initial condition of one ADS valve out of service (Ref. 14). Per this analysis, operation of only four ADS valves will provide the required depressurization.

If any Required Action and associated Completion Time of Condition E, F, G, or H is not met, the plant must be brought to a MODE in which overall plant risk is minimized.


Question: groundwater level regulatory impact

Grand Gulf Nuclear Station Potable Water Wells Physical and Chemical Water Analysis, MS Department of Health Environmental Laboratory, data analysis done in 1999 (Work Order No. 021699-006), 2000 (Work Order No. 031500-118), and 2002 (Work Order No. 022502-020).

Handbook of Steel Drainage and Highway Construction Products, 2nd Ed., American Iron and Steel Institute, R. R. Donnelly and Sons Co., 1971.


Question: aquatic resource impacts from cooling systems

During the license renewal term, cooling system intake and discharge would continue to affect aquatic resources.

The implementation of BMPs and adherence to vegetation management protocols would ensure minimal impact on aquatic resources from ROW management and maintenance.


Question: The contact information for individuals in charge of operational components is required.

The document specifies the name, email, and phone number of the Point of Contact (POC) for operating components and field installations, if applicable.

The design and installation of safety valves are critical to maintaining system integrity during operation.


Question: How does PRA contribute to assessing risks in nuclear facility operations?

The role of the PRA in the severe accident and evaluating the impact of things allowed by design changes is crucial for understanding potential risks.


Question: What are the transportation routes leading to the Unit 1 project area?

The primary access to the Unit 1 project area is from Navajo 11. Beginning at I-40 exit 53, the primary access road is reached by the following route: - Drive north on NM 371 for approximately 45 km [28 mi] - Turn left and drive west on Navajo 9 for about 6 km [3.5 mi] - Turn left and drive southwest on Navajo 11 for approximately 2 km [1 mi] - Turn right at the primary access road.

Regional roads and truck accidents were discussed in Section 3.4 of the CUP FEIS. The ISR GEIS (Section 3.5.2) provides general information on regional roads and traffic counts in northwestern New Mexico.


Question: The NRC guidelines mandate specific structural support requirements for piping systems under ASME Class 2 and 3 standards.

Supports for ASME Class 2 and 3 Piping and Components (Constant and variable load spring hangers; rod hangers, clamps, straps, guides, stops, seismic anchors or restraints, support frames; bolting) | Structural Support | Galvanized Bolting | Air - Indoor, Uncontrolled | Loss of Preload | ASME Section XI, Subsection IWF (B.2.1.30) | III.B1.1.TP-229  | 3.5.1-087 | A

Aging Management Programs are essential to ensure the integrity and reliability of components over time.


Question: nuclide identification

Nuclide Library Used: C:\Canberra\Apex\Root\Dairyland_NPP\Library\HOTALB.NLB


Question: The NRC's assessment includes evaluations of components like closure bolting exposed to various environmental conditions.

Loss of material of steel and stainless steel bolting exposed to atmosphere/weather, plant indoor air, or a ventilation atmosphere is managed by the DCPP Bolting Integrity AMP (B.2.3.9), which takes exception to NUREG-1801.

Not applicable. There are no steel, high-strength high-pressure pumps or closure bolting in the Auxiliary Systems.


Question: What is the significance of well completion dates in resource exploration?

The NRC requires periodic inspections of nuclear facilities to ensure operational safety.


Question: nuclear safety system requirements

Nuclear safety systems are designed to ensure the protection of the public and the environment during normal operations and in the event of accidents, and they must comply with stringent regulatory requirements.

The reactor core is composed of enriched uranium dioxide fuel pellets, which are manufactured to high precision standards to ensure optimal performance.


Question: The examination of components used in fire protection systems is crucial for ensuring safety during operations.

Valve Body | Pressure Boundary | Copper Alloy with 15% Zinc or More | Lubricating Oil (Internal) | Loss of Material | One-Time Inspection (B.2.1.20) | VII.G.AP-133 | 3.3.1-099 | A

Gas (Internal) | None | None | VII.J.AP-9 | 3.3.1-114 | A


Question: Each simulator operating test must have the correct event type identified corresponding to the applicant's position during the test.

For the set of scenario columns, fill in the associated event number from Form 3.3-1, “Scenario Outline,” to show the specific event types being used for the applicant while in the assigned crew position for that scenario.

Minimums are subject to the instructions in Section C.2, “License Level Criteria.”


Question: What is the process of background subtraction in radiation analysis?

In the Background Subtract Report, the process involves taking the Original Area of the identified peaks and subtracting Ambient Background to get the Subtracted Area, which reflects the true signal from the background noise.

The Nuclide Identification Report does not provide any identified nuclides, indicating either that no nuclides were detected or that additional analysis is required.


Question: The estimated duration of exposure for industrial workers is specified as 25 years according to regulatory guidelines.

EPA OSWER Directive 9285.6-03 established an exposure duration of 25 years for the industrial receptor.

Assumed to annually spend 1,600 hours indoors and 400 hours outdoors, plus 250 hours (1 hours/day x 250 days) indoors to account for eating lunch on site, early daily arrival, and late daily departure.


Question: thermocouple inspection procedures

3. PLUG TO CONDUIT WELD AT THE FOUR SUPPORT COLUMNS ADJACENT TO THE THERMOCOUPLE COLUMNS.

22. BAFFLE ASSEMBLY SCREEN LOCKING ARRANGEMENTS AT THE TWO TOP AND THE TWO BOTTOM FORMER ELEVATIONS.


Question: The theoretical mixture of UO₂ and H2O is important for understanding maximum burn scenarios.

Fully reflected damaged single package, theoretical UO₂ + H₂O mixture, maximum burn.

Fully reflected damaged single package, theoretical UO₂ + H₂O mixture, maximum burn, flooded package.


Question: Crownpoint Uranium Project regulatory filings

The Crownpoint Uranium Project, which is currently under license SUA-1580, involves the renewal application process that encompasses extensive regulatory review and safety assessments.

The Ponds 5.5 Acres section outlines the geophysical characteristics of the site without addressing the regulatory specifics required for uranium extraction activities.


Question: What is the definition of instrument accuracy in nuclear systems?

Starting with the OM-1994 addendum of the code, the definition of instrument accuracy is clarified to read, “the allowable inaccuracy of an instrument loop based on the square root of the sum of the square of the inaccuracies of each instrument or component in the loop when considered separately. Alternatively, the allowable inaccuracy of the instrument loop may be based on the output for a known input into the instrument loop.”

Environmental effects, process effects, and vibration effects on loop accuracy (see Question 2 of Interpretation 95-07).


Question: What are the functions of containment penetrations in nuclear reactors?

In general, a penetration consists of a sleeve embedded in the concrete wall and welded to the containment structure liner. The pipe, electrical conductor cartridge, duct, or access hatch passes through the embedded sleeve and one or both ends of the resulting annulus are closed off by welded end plates, bolted flanges, or flued heads. The penetrations are designed to maintain the same high degree of leak tight integrity afforded by the containment structure itself.

The equipment hatch is furnished with double O-rings and a bolted dished door. The hatch barrel is embedded in the containment structure wall and welded to the liner. Provision is made for pressurizing the space between the double O-rings of the hatch flange to support leak testing.


Question: regulatory relief for nuclear licensees

As requested by the licensee in letter 20-042 dated May 22, 2020, which included a proprietary enclosure, and letter 20-082 dated December 18, 2020, the regulatory relief below is granted from the license commitments specified in Chapter 4 of the License Application in anticipation of the potential impacts to radiation protection staffing levels from the Coronavirus Disease 2019 public health emergency.

The reactor core design must meet specific thermal and hydraulic performance requirements to ensure that the fuel operates within safe limits during normal and accident conditions.


Question: The NRC defines the conditions for adequate cooling systems in terms of operational requirements during unexpected events.

Tube Plugging Criteria for Safety-Related Room Cooler (Aerofin AL6XN) Cooling Coils outlines the specifications for ensuring that safety-related cooling systems maintain proper function under specified temperature conditions.

The inspection procedure includes various drawings and documents for different engineering changes that are required for system upgrades.


Question: nuclear fuel assembly average burnup

The table lists the cooling times and the corresponding assembly average burnup levels—specifically, the range of average burnup values measured in gigawatt days per metric ton uranium (GWd/MTU) for different types of fuel assemblies.

The reduced pressure in the system is maintained through a series of valves and responsive controls that ensure optimal performance under operational conditions.


Question: What is the impact of debris transport on chemical release calculations in nuclear safety?

All insulation debris is assumed to be in two-phase flow, resulting in an over-prediction of the chemical release calculation.

The quantity of fine debris washed down (transported to the surface and cores) is that do not initiate containment.


Question: Ge-68/Ga-68 generator regulations

An applicant requesting authorization for a Ge-68/Ga-68 generator should commit to the following: Providing instructions and/or training on the manufacturer’s procedures to all individuals involved in Ge-68/Ga-68 generator use, commensurate with the individual’s duties to be performed; Not opening, breaching, or physically modifying the Ge-68/Ga-68 generator in any way; Following the manufacturer’s procedures, including: generator set-up; generator elution; Ge-68 breakthrough testing and frequency when in use; and final disposition; Eluting the generator in accordance with the manufacturer’s stated frequency and procedures to minimize the concentration of Ge-68 in the eluate; Not using an expired generator for preparation of materials that will be administered to patients or human research subjects; Only using a generator that has a clearly marked expiration date; After installation, performing the conditioning procedure following the manufacturer’s instructions properly disposing of the conditioning eluates prior to the first use of eluate for testing or human use; Developing and implementing written procedures for the determination of breakthrough that will detect whether the eluate exceeds the manufacturer’s recommended breakthrough limit; During the course of breakthrough testing, if the eluate exceeds the manufacturer’s breakthrough limits, the eluate will not be distributed or administered to a patient.

Methods that differ from those described here may be deemed acceptable if they provide sufficient basis and information for the NRC staff to conclude that the proposed alternative demonstrates compliance with the appropriate NRC regulations.


Question: Nuclear licensees must comply with regulatory frameworks that govern operational safety and environmental protection.

The regulation of nuclear reactor operations is primarily governed by the Nuclear Regulatory Commission (NRC), which oversees licensing and safety protocols.

The design of the nuclear facility must incorporate advanced cooling systems to ensure operational efficiency.


Question: The Illinois Agency oversees a uranium recovery initiative, which includes licensing and environmental inspections.

The EM Section administers the uranium recovery program.

During the review period, the State finalized regulatory amendments and received NRC correspondence stating 'No Comments' for 2015-3, 2018-2, 2018-3, 2019-1, and 2019-2.


Question: emergency diesel generators

B. (continued) B.2 Declare required feature(s), supported by the inoperable EDGs, inoperable when the redundant required feature(s) are inoperable.

The 72-hour Completion Time is extended to 7 days one time for each of the Division I MDCT fan pedestals (A and C) to allow repair online during Cycle 22, regardless of whether maintenance is completed during the attempt.


Question: welding techniques in nuclear reactor repair

(c) Welding electrodes would be rigidly controlled as to sizes employed and temperature maintenance.

The decision was made to repair Unit 1 Steam Generator Supports at the North Anna site and Unit 2 offsite because of schedule requirements.


Question: nuclear licensing processes


Question: nuclear power plant operations during the license renewal term

2.2.1 Nuclear Power Plant Operations during the Subsequent License Renewal Term


Question: The NRC mandates that licensees assess the integrity of components that have shown leakage, especially when not covered by standard testing protocols.

Operational Leakage is leakage through a flaw in the pressure retaining boundary discovered outside any ASME BPV Code required pressure test.

The proposed amendment to the Commission’s regulations would unnecessarily conflate maintaining ASME Code compliance with operability determinations, which will result in unnecessary exigent requests for Code relief and unnecessary potential plant shutdowns which could result in increased plant risk.


Question: Degraded Voltage Time Delay Function

The 4.16kV Bus Degraded Voltage Time Delay function is comprised of two independent relay functions. The first relay function is the time delay response of the Degraded Voltage relays. The second is the Time Delay Relay function. Taken together, these functions define the duration for the entire 4.16kV Bus Degraded Voltage time delay function prescribed in the Technical Specification.

The lower limit NTSP will be calculated using the lower Allowable Value as follows:


Question: loss-of-coolant accident dose analysis

The loss-of-coolant accident (LOCA) dose consequence analysis, documented in Section 15.4.1.9, assumed an effective elemental iodine removal coefficient (spray lambda) of 10 per hour for the duration of the accident.

In addition to the quench spray headers, there are recirculation spray headers located an average of approximately 85 feet above the operating floor.


Question: What is the role of the Nuclear Service Water System in nuclear plant operations?

The Nuclear Service Water (NSW) System provides assured cooling water for various Auxiliary Building and Reactor Building heat exchangers during all phases of station operation.

In conjunction with the Ultimate Heat Sink, comprised of the Standby Nuclear Service Water Pond (SNSWP), the NSW System is designed to meet design flow rates and heads for normal station operation and also those flow rates and heads required for safe station shutdown under the following conditions while sustaining a single failure:


Question: The calibration factor is derived from the uniformity of the radioactive source spread across the surface of the pipe.

The purpose of the present work was to estimate, by Monte Carlo simulation, the calibration factor assuming that the Cs-137 was uniformly distributed over the entire internal surface of the pipe, instead of in just the 1 ft x 1 m rectangle.

The material definitions are from report PNNL-15870 Rev1.


Question: The Passive Core Cooling System is crucial for maintaining safe reactor operations during core cooling events.

The Passive Core Cooling System provides a safety feature that ensures adequate cooling of the reactor core even under postulated accident conditions.

The design incorporates various components to ensure effective heat transfer and structural integrity under operational conditions.


Question: What are the requirements for reactor licensing under NRC regulations?

The licensing process for reactors involves rigorous safety assessments, compliance evaluations, and public consultations prescribed by the NRC to ensure operational safety and environmental protection.

The document outlines various administrative processes and timelines for submitting responses to regulatory inquiries, maintaining a detailed record of all correspondence and actions taken.


Question: What is a completion time for a required action in nuclear regulatory filings?

Each Required Action has its own Completion Time. Each Completion Time is referenced to the time that Condition B is entered.


Question: What materials are commonly used in containment structures?


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

The CCW System also provides RCP thermal barrier cooling in the event of a loss of charging system RCP seal injection.

At some period of time into the accident scenario, SW flow is re-established to two CCHXs to affect a cooldown of the non-accident unit.


Question: steam moisture trap manifold

Performer locates at least one valve for each Main Steam Line steam moisture trap manifold.

Step B of 2-E-3, Attachment 4, Steamline Isolation (Local), is broken into 3 steps to be evaluated.


Question: What is the significance of a nuclear reactor's license renewal process?

The license renewal process for nuclear reactors involves a thorough review of the facility's safety records, environmental impact, and operational procedures to ensure continued compliance with regulatory standards.

The primary function of safety valves in a reactor system is to prevent overpressure conditions, ensuring the integrity of the reactor vessel and protecting the overall system from potential hazards.


Question: The process for approving the release of gaseous materials involves understanding contamination controls and safety protocols established by the NRC.

(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

The applicant must adhere to the procedural requirements for documenting the start and stop times for all operations related to the waste management system.


Question: nuclear reactor licensing

Nuclear reactor licensing is a regulatory process that ensures reactors are designed, constructed, and operated safely and in compliance with federal regulations and standards.

Mission Analytics Portal (MAP-X) serves as a platform for various external data interactions, enhancing mission planning and analytics.


Question: The NRC has specific protocols for the documentation and approval of licensing amendments.

This refers to your license amendment request dated January 20, 2025. Please find enclosed Amendment No. 7 authorizing the removal of an authorized user, removal of authorized licensed material and the associated area of use, and addition of a new area of storage or use.

You will be periodically inspected by the NRC. Failure to conduct your program in accordance with NRC regulations, license conditions, and representations made in your license application and supplemental correspondence with NRC will result in enforcement action(s) against you.


Question: nuclear licensing process

The nuclear licensing process involves several key steps, including the submission of safety analysis reports, environmental assessments, and compliance with regulatory standards set by the Nuclear Regulatory Commission (NRC).

The drawing number 177D4970, as shown in Sheet 7, represents various components of the reactor design and illustrates the layout and specifications of critical systems.


Question: The mean activity of a nuclide is reported in picocuries per gram, reflecting its radiological significance.

Wt mean Activity (pCi/grams) refers to the average radioactivity per gram for a specific nuclide, which is essential for assessing safety and regulatory compliance.

The table presents details about various nuclides, including their confidence values and associated uncertainties, but does not discuss the impact of radiological assessments on public safety.


Question: what is NRC Form 313

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: The NRC may extend its regulatory authority over fusion devices by categorizing them as utilization facilities, contingent upon specific findings related to atomic energy and public safety.

The NRC Staff Memorandum goes on to note that the AEA requires that, for treatment of a fusion facility as a utilization facility, the 'Commission must find in a rulemaking both that: (1) fusion constitutes ‘atomic energy’ within the meaning of the AEA, and (2) the fusion process is of such quantity as to be ‘of significance to the common defense and security, or in such manner as to affect the health and safety of the public.'

The potential hazard of an activated first wall and dust can be addressed through appropriate material selection so that activation is minimized in the first place, remote maintenance machines to minimize dose to maintenance personnel, and procedures to maintain reasonable waiting periods after operation for maintenance personnel to start work.


Question: The load setpoint for a reactor must be adjusted according to operational guidelines.

Raise the Load Set Setpoint to 110% in accordance with QCGP 3-1 Attachment B step 2.

A Unit 1 DEHC failure resulted in LOAD SET being driven to 85.


Question: The NRC evaluates the potential environmental impacts of the proposed nuclear project before granting a license.

The licensing process for nuclear facilities requires a thorough assessment of the environmental effects to ensure that public health and safety are protected.

The construction plan incorporates advanced seismic designs to withstand potential earthquakes.


Question: The analysis of lightning strike data is crucial for ensuring safety and reliability at nuclear facilities.

Computed Lightning Strike Density at Oak Ridge and Knoxville, Tennessee on the basis of historical data shows 15.1 strikes per square mile at Knoxville Airport for the recorded years.

The reactor core design must comply with the standards set by the NRC to ensure safety during operation.


Question: The NRC oversees the licensing process to ensure compliance with safety and environmental regulations.

The process of nuclear licensing involves multiple steps and a thorough review of each applicant's capability to operate safely and in compliance with regulatory standards.

Various reactor designs have been implemented to address specific energy needs and regulatory requirements.


Question: The NRC performed an assessment of the potential environmental impacts related to the continued operation of the Oconee Nuclear Station beyond its originally established license period.

The FEIS includes the NRC staff’s site-specific evaluation of the environmental impacts of the subsequent license renewal (SLR) for Oconee Station and alternatives to the SLR.

The letter also informed the Tribe about the general process of environmental obligations under various federal statutes.


Question: post-tensioning system performance

The results of the 9 post-tensioning system in-service examinations conducted at Callaway between 1985 and 2021 show that the system is continuing to perform its intended function and that it can be expected to do so until well past the proposed January 2030 deadline for completion of the next surveillance.

Occasional buttonhead loss is normal for BBRV tendons (wires anchored by cold formed buttonheads) and generally occurs during or shortly after tensioning.