Question: The NRC has outlined the need for engagement and monitoring of industry research related to safety protocols.

Forwarding draft IAD UNR – includes 3 tasks: void swelling, engagement / monitoring industry research, IASSC CGR disposition curves at ASME Code.

My biggest disappointment is that I didn’t see anything on harvesting at all. It would be good if they at least indicated that we should pursue opportunities as they present themselves or something similar.


Question: once-through cooling system

License renewal applicant with a permitted once-through cooling system is not required to further evaluate the environmental impacts of that system or potential alternatives; CLI-22-5, 95 NRC 97, 103 (2022).

Discretionary intervention is allowed only where at least one petitioner has established standing and identified an admissible contention, such that a hearing will be held; CLI-21-2, 95 NRC 1, 1 (2022).


Question: What is the significance of thorium in radiochemistry?

Thorium 230, dissolved, is a parameter that is analyzed in radiochemistry but is not offered for certification or covered by NELAC certificate #ACZ.

The ACZ Project ID for this documentation is L89090 and is associated with the Rio Algom Mining Company.


Question: What is primary-plus-secondary stress evaluation in nuclear engineering?

For the primary-plus-secondary stress evaluation, the reported maximum stress intensity is 30.35 ksi, with an allowable stress intensity of 48.9 ksi.

Based upon the above evaluations, the use of a 1.30” ID heater sleeve is acceptable relative to the criteria contained within the ASME Code.


Question: Explain the impact of gas binding on natural circulation within a reactor.

Gas binding can significantly impede natural circulation within a reactor by trapping gas bubbles in the coolant channels, reducing the effective flow of the coolant and potentially leading to overheating.

The axial temperature and enthalpy profiles in a reactor coolant channel are influenced by various factors including the inlet temperature and heat generation rate.


Question: The NELAC standard outlines various requirements for laboratory accreditation.

The NELAC standard sets forth comprehensive guidelines that laboratories must follow to obtain and maintain accreditation.

Effective analysis of laboratory results depends on the precision of the instruments used in testing.


Question: weighted mean activity

The weighted mean activity is a statistical measure used to estimate the average radioactivity of a set of nuclides in a sample, weighted by their respective uncertainties.

Errors quoted at 2.0000 grams indicate that the measurements may have inherent inaccuracies, influencing the reported values.


Question: What is the minimum initial assembly average enrichment required for low burnup PWR fuel?

The Low Burnup PWR Fuel Loading Table specifies several parameters including maximum assembly average burnup and minimum assembly average initial enrichment, all of which are crucial for ensuring safe reactor operation.


Question: structural bolting aging management

The study also assessed various environmental impacts on structural integrity.


Question: auxiliary feedwater system functions

The auxiliary feedwater system is actuated by the low-low steam generator water level signal.

Following the reactor and turbine trip from full load, the water level in the steam generators will fall due to the reduction of steam generator void fraction, and because steam flow through the safety valves continues to dissipate the stored and generated heat.


Question: calibrated ion chamber current

Calibrated ion chamber current has 4 channels available with a range of 0-120% of full power and an indicated accuracy of ± 2% of full span.

The average flux of the top and bottom ion chamber varies between 0 to 120% of full power, with an accuracy of ±3% for indication, ± 2% for recording.


Question: What is the RICT program in nuclear regulatory processes?

A qualitative assessment, utilizing the experience and judgment of the PRA engineer, is performed considering the potential change in basic event importance measures for each application. This assessment utilizes the experience and judgment of the PRA engineer to determine if there are any issues that are individually negligible but could collectively impact the RICT program.

PRA model of record updates will typically be performed once every two fuel cycles but may be completed sooner depending on the changes or issues documented in the MCR database and at the discretion of management (NEI 06-09-A, Section 2.3.4, Item 7.1, and Section 2.3.5, Item 9.1).


Question: What is the significance of 10CFR50 Appendix R regarding fire safety in nuclear plants?

This Fire Area complies with 10CFR50 Appendix R, Section III.G.3.

Manual hose stations and portable extinguishers would extinguish fires and normal unit or SSS capability would be available.


Question: What are the requirements for notifying the NRC about license termination?

By 10 CFR 30.36(d) and/or license condition, notify NRC, promptly, in writing, and request termination of the license: a. When you decide to terminate all activities involving materials authorized under the license whether at the entire site or any separate building or outdoor area;

You will receive a reminder notice approximately 90 days before the expiration date.


Question: Water levels at various monitoring points are essential for understanding groundwater conditions.

The Church Rock Project involves various wells that monitor geological and hydrological conditions.


Question: anticipated transients without scram

Anticipated Transients Without Scram (ATWS) refers to events that are expected to occur during the normal operation of a reactor, where the reactor does not automatically shut down as per its designed safety systems.

The process of licensing a nuclear facility involves extensive reviews and assessments by the Nuclear Regulatory Commission (NRC) to ensure compliance with safety regulations.


Question: water chemistry analysis in nuclear licensing

The water chemistry results include parameters such as alkalinity as CaCO3, bicarbonate as CaCO3, and conductivity @25C among others, which are essential in understanding the quality of groundwater in nuclear projects.

The package included the Project Information section detailing the Company, Project ID, Sample ID, Date Sampled, and Date Received, although this information does not pertain to the water chemistry analysis.


Question: What are significance levels used by the NRC in environmental impact assessments?

The NRC assigns a significance level of SMALL, MODERATE, or LARGE for most site-specific issues.

Oconee Station is located on the shore of Lake Keowee in eastern Oconee County, South Carolina, approximately 8 mi (13 km) northeast of the city of Seneca, South Carolina.


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

The intent of this EAL is to ensure that an emergency classification level is declared upon a loss of vital direct current (DC) power that compromises the ability to monitor and control operable safety systems.

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


Question: The report must include a discussion on the condition monitoring of steam generator tubing over time.

Discuss trending of tube degradation over the inspection interval and provide comparison of the prior operational assessment degradation projections to the as-found condition.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.


Question: shutdown cooling system components

Heat Exchanger is designed for Heat Transfer and is made of Copper Alloy with less than 15% Zinc, operating in Closed Cycle Cooling Water (External) and Treated Water (Internal), with aging effects including Reduction of Heat Transfer that requires management through Closed Treated Water Systems (B.2.1.12) and a One-Time Inspection (B.2.1.20).

Piping components can experience cumulative fatigue damage, which is managed through TLAA and one-time inspections as noted in various aging management programs.


Question: The NRC mandates licensees to conduct regular testing and inspections of fire protection systems to ensure proper functionality and safety standards are met.

The existing program is credited and will be enhanced as follows: Revise implementing procedures to ensure that flushes are in accordance with NFPA 25, 2011 Edition, section 7.3.2.1, to mitigate or prevent fouling, which can cause flow blockage or loss of material, by clearing corrosion products and sediment.

Program enhancements, associated with inspections/tests required to be implemented prior to the subsequent period of extended operation, will be implemented no later than six months prior to the subsequent period of extended operation.


Question: What is the latent heat of vaporization in nuclear engineering?

--latent heat of vaporization      | 2.3        |

Describe effects of pressure and temperature on density or specific volume of a gas | 2.4        |


Question: The results of gamma spectroscopy analysis indicated the presence of cesium-137 in all samples collected for review.

Gamma spectroscopy results revealed that Cs-137 was positively identified in all six (6) of the investigational samples, at a maximum concentration of 1.57E+00 pCi/g.

There was not an acceptable agreement between standard and duplicate results using Cs-137 in the comparison.


Question: What is the significance of assembly average burnup in nuclear fuel?

42 < Assembly Average Burnup < 43 GWd/MTU

The NRC requires a comprehensive review of all safety-related systems during the licensing process.


Question: quality assurance program in nuclear regulation

The Quality Assurance Program ensures that all activities related to safety, reliability, and performance of safety systems are thoroughly evaluated and documented to meet regulatory requirements.

The implementation of various safety measures includes regular training for personnel and the establishment of feedback mechanisms to enhance operational effectiveness.


Question: What is a passive containment cooling system in nuclear safety?

**EA 1.09** Passive containment cooling system 3.7

The component cooling water system is essential for maintaining optimal temperatures in reactor operations.


Question: What is the purpose of radiation shielding in nuclear facilities?

Materials required to meet radiation shielding objectives are primarily concrete, water, and steel.

Seals, gaskets (including elastomers), compressible joints, and moisture barriers within the scope of license renewal include those items that perform license renewal intended functions such as shelter and protection, structural pressure barrier, and HELB/MELB shielding for structures and components that are within the scope of license renewal.


Question: buried piping inspection program

The following examples of operating experience provide objective evidence that the Buried Piping and Tanks Inspection Program will be effective in assuring that intended functions are maintained consistent with the current licensing basis for the subsequent period of extended operation.

Revise implementing procedures to state that cracks in cementitious backfill that could admit groundwater to the surface of the component are not acceptable.


Question: What is a homogenous matrix in the context of nuclear sampling?

*Homogenous Matrix* Sample 485191030 (L1-SUB-SRX-AJFC-009-CV) was non-homogenous matrix.

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


Question: IPAWS-WEA alert system

IPAWS-WEA is used as the primary alerting method to alert residents and the transient population within the RBS plume exposure pathway EPZ. IPAWS-WEA broadcasts alerts and warnings from cell towers to any WEA-enabled mobile device within a geographically targeted area, alerting and notifying individuals of an emergency.

The RBS License contains no special license conditions relating to the type of Alert and Notification System.


Question: What is the TLAA related to neutron embrittlement?

Certain aspects of neutron irradiation embrittlement are TLAA as defined in 10 CFR 54.3. TLAA are required to be evaluated in accordance with 10 CFR 54.21(c)(1).

As discussed in item 3.1-1, O14, loss of fracture toughness due to neutron irradiation embrittlement could occur in the RV shells.


Question: What is weighed mean activity in nuclear regulatory filings?

Weighted mean activity is a measure used to quantify the average activity of a nuclide per gram, accounting for the respective uncertainties and contributions from different isotope compositions.

The table lists various nuclides along with their corresponding confidence levels and comments, which are essential for assessing the overall radiological safety of a facility.


Question: What are the specifications for measuring activity concentration of radionuclides?

The activity concentration of radionuclides is measured in pCi/g Dry, indicating the amount of radioactive material per gram of dry weight.

The equipment used for measuring the activity concentration includes various detectors and calibration standards to ensure accuracy.


Question: What is the function of the Standby Gas Treatment System in a nuclear plant?

The Standby Gas Treatment (SGT) system is designed to remove radioactive particulate and iodine from the containment atmosphere during an accident scenario.

The Main Turbine Pressure Regulation System is crucial for maintaining the proper pressure within the turbine to ensure efficient steam flow and energy conversion.


Question: NEPA environmental impacts

Duke has violated NEPA and 10 C.F.R. §51.53(c)(3)(iv) by failing to address “new and significant information regarding the environmental impacts of license renewal of which [Duke] is aware.”

The Commission should review 10 C.F.R. §51.53(c)(3) and the categorical exclusions of 10 C.F.R. Part 51, [Subpart] A, Appendix B, and require Duke to address those impacts in a complete environmental impact analysis, as set forth in 10 C.F.R. §51.45.


Question: emergency termination definition

Emergency termination means that RBS has the plant in a safe and stable condition. The Parish will then start planning for recovery and reentry into areas that may have been affected.

Ensure the Sheriff Department/law enforcement resources are stationed to assist in the evacuation of people and to provide security and control access to the evacuated area.


Question: What are suppression chamber-to-drywell vacuum breakers?

Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening.

Twelve suppression chamber-to-drywell vacuum breakers shall be closed.


Question: Risk Informed Completion Time Program

The Risk Informed Completion Time Program requires recalculation of the RICT to reflect changing plant conditions.

The 7-day Completion Time may be applied as discussed in Example 1.3-2.


Question: What is the purpose of a nuclear licensing application?

A nuclear licensing application is required to obtain permission from the regulatory body to construct or operate a nuclear facility, ensuring compliance with safety standards and regulations.

The energy efficiency of the new reactor design has shown promising results during preliminary testing phases.


Question: The NRC mandates that licensees must evaluate critical heat flux under various operating conditions.

The COBRA-FLX code employs the ORFEO-HMP correlation to predict the critical heat flux for a range of thermal-hydraulic conditions in nuclear reactors.

Table B-2 outlines the specific conditions under which various test assemblies were evaluated, detailing parameters such as temperature and pressure.


Question: The process of managing extraction steam and auxiliary systems is crucial when transitioning reactor power levels during operations.

IF #3 Extraction Steam in service to Auxiliary Steam Header, THEN PERFORM one of the following (NA option NOT used): - PLACE Auxiliary Boiler in service per 0-SOI-12.01. - PLACE U2 #3 Extraction Steam in service to the Auxiliary Steam Header per 0-SOI-44.01. - REMOVE Building Heat System from service using 0-SOI-44.01.

CHECK Permissive 70-C, P-8 LO PWR-FLOW TRIP BLOCKED, LIT.


Question: What is BOP VC-20 in nuclear operations?

Contrary to the above, on July 7, 2024, the licensee failed to implement step F.2.f.6 and F.2.f.7 of BOP VC-20.

The inspectors verified no proprietary information was retained or documented in this report.


Question: What happens to the DNBR when one RCCA is fully inserted?

For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value.

Following reactor trip, normal shutdown procedures are followed.


Question: spent nuclear fuel

Spent nuclear fuel (SNF) is the used fuel that has been irradiated in a nuclear reactor and is no longer able to effectively sustain a nuclear reaction.

Stormwater Pollution Prevention Plan outlines the measures that should be taken to prevent pollutants from entering storm drainage systems.


Question: post accident sampling procedures

A contingency plan, controlled by normal Chemistry procedures, has been developed for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere.

In order to support the data acquisitions need of the emergency response facilities, the PCS has been installed.


Question: Operational Derived Concentration Guideline Levels

The dose contribution from soil in survey unit L1-010-106 is 0.2603 mrem/yr TEDE, based on the average concentration of the ROC in samples used for non-parametric statistical sampling.

Survey unit L1-010-106 is acceptable for unrestricted release.


Question: What is the function of flow devices in a reactor recirculation system?

Flow Device (Class 1) is designed to function as a Pressure Boundary and Throttle in the Reactor Recirculation System.

The subsequent license renewal boundary drawings include various document identifiers ranging from 1-47E817-1-SLR to 3-47E844-2-SLR.


Question: The guidance for assessing digital safety systems must include considerations for common cause failures.

Final Draft Revision 8 of Standard Review Plan Branch Technical Position 7-19, “Guidance for Evaluation of Defense-in-Depth and Diversity to Address Common Cause Failure Due to Latent Defects in Digital Safety Systems,” dated November 23, 2020, ADAMS Accession No. ML20328A157

Documentation of Receipt of Applicable Official NRC Notices to the Advisory Committee on Reactor Safeguards for November 2020, dated November 19, 2020, ADAMS Accession No. ML20317A122.


Question: Equivalent Margin Analysis in nuclear reactor safety

Therefore, vessel plate is bounded by Equivalent Margin Analysis.

Comparison of Limiting % Decrease Value to Limit:


Question: What is the significance of the steam-zirconium reaction in nuclear accidents?

Therefore the fraction of released fission products is significantly less than was anticipated from the fuel melting or a so-called “steam-explosion” scenario. Also, the fiery steam-zirconium reaction results in a much higher than anticipated (from any other scenarios) rate of Hydrogen production, which in turn requires a review of containment designs.

In all three of the compared cases, it took several hours of ill-fated actions or in-actions of the operators to cause the ignition condition.


Question: What is the double contingency principle in nuclear safety?

Criticality accident sequences: Use the double contingency principle (DCP) with the criteria presented in the three bullets shown in NUREG-1520 p. 5-A-7 and 5-A-8 (i.e., DCP +3). Use -4 (risk-indexed highly unlikely) for risk indexing.

The NRC staff and TRISO-X agree in principle that the options identified above would resolve the definitions of unlikely and highly unlikely.


Question: The NRC mandates that licensees comply with specific analytical methods to ensure the accurate reporting of radioactive substance levels.

The purpose of the LACBWR Site Restoration Project is to restore the site to meet current regulatory requirements.


Question: nuclide minimum detectable activity MDA

The nuclide minimum detectable activity (MDA) is defined as the smallest activity level of a specific nuclide that can be reliably determined by the measurement process.

The energy of K-40 is measured at 1460.75 keV and has a yield of 10.67%.


Question: What are Function Restoration Procedures in nuclear reactor safety?

The Shift Foreman will monitor the Function Restoration Procedures for information only until a specific point is reached while performing: 1. E-1.3, Transfer to Cold Leg Recirculation 2. ECA-0.2, Loss of All AC Power With SI Required 3. ECA-2.1, Uncontrolled Depressurization of All Steam Generators

The examination outline includes guidelines on knowledge of the operational implications of EOP warnings, cautions, and notes.


Question: defense in depth in nuclear safety

Defense in depth is a key safety principle applied in the design and operation of nuclear facilities, which ensures multiple layers of protection against accidents and unauthorized access to nuclear materials.

The contributors of this report include senior specialists and coordinators responsible for operations at Limerick, ensuring thorough reviews and assessments are conducted.


Question: What documentation is necessary for ensuring compliance with the operational standards of nuclear units?

Framatome Inc., Doc. No. 51-9234739-001, Florida Power and Light Turkey Point Nuclear Units 3&4 Digital Modernization Cyber Security Plan, dated July 6, 2021


Question: elastomer degradation in nuclear safety systems

Consistent with NUREG-2191. The External Surfaces Monitoring of Mechanical Components (B.2.1.23) program will be used to manage hardening and loss of strength of the elastomer ducting and components exposed to air - indoor uncontrolled in the Standby Gas Treatment System.

Consistent with NUREG-2191. The External Surfaces Monitoring of Mechanical Components (B.2.1.23) program will be used to manage loss of material of the carbon or low alloy steel with stainless steel cladding, carbon steel and cast iron accumulator, blower housing, compressor housing, ducting and components, flexible connection, gearbox, heat exchanger components, hoses, piping, components, pump casings, tanks, turbine casings, and valve bodies exposed to air - indoor uncontrolled and air - outdoor in the Containment Atmospheric Control System, Core Spray System, High Pressure Coolant Injection System, Isolation Condenser System, Low Pressure Coolant Injection System, Primary Containment Isolation System, and Standby Gas Treatment System.


Question: The NRC assesses the sufficiency of information in licensing applications.

Sufficient Information is a status indicating that the NRC has enough data to proceed with the regulatory review process.

ITAAC Deleted Per License Amendment 85 signifies that the item has been removed from the list of inspection requirements.


Question: What is an AMP in the context of nuclear aging management?

Aging Management Programs (AMPs) are established to monitor the aging effects of components based on their material, environment, and operational conditions as specified in NUREG-1801.

The DCPP Water Chemistry AMP and the DCPP ASME Section XI Inservice Inspection Subsections IWB, IWC and IWD AMP will be used to manage this aging effect for CASS components.


Question: aging management review results

The DCPP Structures Monitoring AMP (B.2.3.33) will be used to manage loss of material due to corrosion of steel (Non-ASME) support members, welds, and other structural components that are exposed to plant indoor air or atmosphere/weather environment.

The DCPP Fire Protection AMP (B.2.3.14) will manage loss of material for cable trays credited for fire protection.


Question: What is the significance of Nickel-63 in nuclear analytical methods?

The NRC mandates that all licensees must conduct thorough assessments to ensure compliance with safety standards.


Question: The Medical Support Facilities Communications Systems are designed to ensure reliable communication between ambulances and hospitals.

A coordinated communication link for hospitals and ambulances is provided by the Medical Support Facilities Communications Systems.

Fixed, mobile, and hand-held radios operating on law enforcement and fire/rescue frequencies provide for communication means between the Parish EOC and representatives from the following organizations:


Question: control of heavy loads at nuclear power plants

RG 1.244, Revision 0, “Control of Heavy Loads at Nuclear Facilities” (ML21006A346) provides guidelines for managing the risks associated with lifting heavy loads in nuclear power facilities, emphasizing safety and compliance with regulatory standards.

The revision history for IP 50003 outlines the changes made to the inspection procedure, including updates to the inspection criteria and compliance deadlines.


Question: The NRC oversees the assessment and management of radioactive waste at recovery facilities.

Radioactive Waste Management and Transportation at Uranium Recovery and 11e.(2) Byproduct Material Facilities 10/08/21 (21-034)

Emergency Preparedness and Fire Protection at Uranium Recovery and 11e.(2)


Question: gross beta activity in airborne particulates

Annual average gross beta activity in airborne particulates for 1973 was measured at 3.5, 1.6, and 3.5 for various locations.

The data collected at 6.5 miles NE of the plant includes records from 1/73 to 6/73 with 24 samples analyzed.


Question: What are the requirements for nuclear reactor licensing?

The licensing process for nuclear reactors involves a comprehensive evaluation of safety, operational, and environmental factors as mandated by regulatory bodies such as the Nuclear Regulatory Commission (NRC).

In the declaration, the applicant must provide data on the projected economic impact and market analysis to demonstrate the viability of the project.


Question: The review of the safety analysis report is essential for ensuring compliance with regulatory standards.

The safety analysis report must include a detailed assessment of potential accidents and their consequences to comply with NRC regulations.

The communication between team members is vital for maintaining an efficient workflow.


Question: What is the purpose of closing the containment isolation valve during a nuclear emergency?

Closing the isolation valves is essential to safety since failure to do results in radiation release continuing beyond containment.

Given the major event in this scenario (SBLOCA), the open path between containment and the Auxiliary building represents the loss of a second fission product barrier.


Question: failure probability of main steam isolation valve

Failure Probability (demand) (Note 3) represents the likelihood of the valve failing to operate as intended during a specific demand event, which is essential in determining the reliability and safety of the main steam isolation system.

Inboard Valve Number (Note 2) is associated with the identification of valves within the system and does not directly relate to the performance evaluation of the main steam isolation valves.


Question: What is the purpose of inspections according to NRC standards?

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted.

The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems, including the 345 kV and 138 kV switchyard from impending severe weather with expected temperatures below freezing on December 2 and 9, 2024.


Question: Pressurized Thermal Shock regulations

The Pressurized Thermal Shock (PTS) Rule was approved by the U.S. Nuclear Regulatory Commission on June 20, 1985, and appeared in the Federal Register on July 23, 1985. The PTS Rule outlines regulations to address the potential for PTS events on reactor vessels in nuclear power plants.

The stated administrative controls and the use of redundant dual setpoint relief valves will not require operator action until 10 minutes after a transient is in progress.


Question: The Engineered Safeguards Actuation System is essential for automatic responses in emergency scenarios.

Safety Function 2, Reactor Coolant System Inventory Control, Engineered Safeguards Actuation System, added knowledge K 4.36, Containment vacuum relief actuation.

1.10.1, revised the definitions of emergency and abnormal evolutions to be consistent with the definitions provided in the other NRC K/A catalogs.


Question: The NRC assesses potential discrepancies in evaluation models, particularly regarding thermal and fluid dynamics.

There was one evaluation model change or error correction made during the reporting period.

The EXEM BWR-2000 methodology has remained stable without any changes or errors during this evaluation period.


Question: What is the significance of matrix solid swelling in nuclear fuel modeling?

The matrix solid swelling model is classified as an independent model, typically based on the swelling rate coefficient, and is described in Section 7.4.1 of Reference 5.

Fuel diameter and length changes result from the combination of several interrelated models including but not limited to matrix solid swelling, short- and long-term densification, relocation, gaseous swelling, and fuel creep.


Question: Monitoring of the Containment Spray System involves using instrumentation to measure critical parameters such as pressure and flow during operation.

This section describes the instrumentation employed for the monitoring and actuation of the Containment Spray System. This instrumentation includes that which is used during normal and post-accident operation for measurement of temperature, pressure, and flow.

The air test lines, for checking that spray nozzles are not obstructed, connect downstream of the isolation valve.


Question: What are the necessary considerations for scheduling a briefing with stakeholders?

Usually, briefings like this cannot be completed in less than an hour.

I have attached a draft one-pager that could be used to brief Brian and Chris on the harvesting efforts in the context of their questions regarding the ANL travel.


Question: The potentiometric surface of the Black Mingo Aquifer indicates the levels of groundwater flow and pressure in relation to the surrounding geographical features.

The Mill Creek flows through the identified zones, influencing the surrounding hydrology and ecology.


Question: What is the purpose of the Emergency Core Cooling System in a reactor?

The Emergency Core Cooling System (ECCS) is designed to provide cooling to the reactor core under accident conditions, ensuring that the core remains covered with coolant to prevent overheating and potential core damage.

The Code of Federal Regulations (CFR) provides guidelines that govern nuclear power plant operations and safety regulations across the United States.


Question: The analysis indicates that there are no recent seismic activities related to the faults mapped within the Chopawamsic Terrane.

Based on the absence of geomorphic expression, absence of seismicity, and absence of offset of Quaternary surficial deposits it is concluded that the Sturgeon Creek fault is not a capable tectonic source.

Three other bedrock faults are mapped within the Chopawamsic terrane within 5 miles of the ESP site.


Question: What is the purpose of the ADVANCE Act in relation to nuclear energy?

The ADVANCE Act (Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024) aims to promote the development and deployment of advanced nuclear technologies to enhance clean energy production.

The United States Nuclear Regulatory Commission is responsible for ensuring safe and secure nuclear energy operations across the country.


Question: What are the energy levels and yields for specific nuclides?

The notification process for license modifications often involves submitting detailed documentation for approval to ensure compliance with regulatory standards.


Question: What is the purpose of a Technical Specifications (TS) document in nuclear regulation?

The Technical Specifications (TS) document outlines the safety limits, limiting conditions for operation, and other key operational parameters that must be adhered to by the nuclear facility to ensure safe operation and compliance with regulatory requirements.

The analysis of the system used to monitor environmental conditions includes parameters like temperature, humidity, and pressure, which can significantly influence the operational efficiency of the reactor.


Question: What are the units used for measuring radioactivity in nuclear licensing?

The measurements reported are given in pCi/g Dry, which stands for picocuries per gram of dry material, indicating the radioactivity per unit weight of the dry sample.

The uncertainty of the measurement is calculated as 1.69E+00, providing an estimate of the variation or error in the measurement results.


Question: What is the purpose of an affidavit in nuclear regulatory submissions?

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.


Question: The Service Air System plays a crucial role in providing necessary pressurized air to various systems throughout the plant, including support for safety-related functions.

The Service Air System provides pressurized air to: - Hose connections throughout the plant and yard - Miscellaneous equipment in the SLC System, Ametrap Condenser Tube Cleaning System (subsystem of the CCW System), Condensate Demineralizer Air Surge System, and the Radwaste System

The aging management review results for these components are provided in Table 3.3.2-11, Control Air System - Summary of Aging Management Evaluation.


Question: The report requires a confidence index threshold for nuclide identification.

Nuclide confidence index threshold = 0.30


Question: The radiation sources in the chemical and volume control system are primarily determined by the radionuclides present in the reactor coolant.

Radiation sources in the CVCS consist of radionuclides carried in the reactor coolant.

The maximum blowdown rate to the blowdown demineralizers is 1.2 percent of the main steam rate.


Question: What is the initial RCS flow rate in nuclear reactor safety analysis?

Initial RCS flow rate is biased to the low condition.


Question: The NRC utilizes guidance documents to ensure that licensing processes for advanced reactor applications are efficiently reviewed.

This REP was developed with consideration of the Nuclear Energy Institute (NEI) draft “Industry Guideline for Development of a Regulatory Engagement Plan” and Appendix A to the NRC’s DANU-ISG-202-01, “Interim Staff Guidance, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications – Roadmap,” March 2024.

The EN REP is not intended to duplicate that effort.


Question: Assessment of aging effects is vital for maintaining the integrity of concrete structures such as those in PWR and BWR designs.

Increase in porosity and permeability due to leaching of calcium hydroxide and carbonation could occur in inaccessible areas of concrete elements of PWR and BWR concrete steel containments. Further evaluation is recommended to determine the need for a plant-specific AMP or plant-specific enhancements to ASME Code Section XI, Subsection IWL and/or Structures Monitoring AMPs, essential to managing these aging effects if leaching is observed in accessible areas that impact intended functions. Acceptance criteria are described in BTP RLSB-1 (Appendix A.1 of this SRP-SLR).

The Primary Containment structure is completely enclosed and sheltered within the air - indoor environment of the Reactor Building and mounted on support structures that transmit loads to the concrete foundation of the Reactor Building as shown in UFSAR Figures 3.8-1, 6.2-1, and 6.2-2.


Question: 100-year return interval precipitation

The 100-year return interval 48-hour duration liquid equivalent precipitation is available for the HCS from NOAA, 2020a.

A network of 10 meteorological towers and two sound detection and ranging instruments (SODARs) have been installed around ORNL and the Y-12 National Security Complex (Y-12) as part of the environmental monitoring program supporting the Oak Ridge Reservation (ORR) (ORR, 2020; Chapter 6).


Question: What is the significance of the eastern prairie fringed orchid in the context of environmental assessments for nuclear power projects?

The eastern prairie fringed orchid (*Platanthera leucophaea*) is listed as endangered at the state level and threatened at the federal level.

The lakeside daisy (*Hymenoxys herbacea*) is listed as an endangered species at the state level and threatened at the federal level.


Question: The evaluation of the Emergency Core Cooling System (ECCS) components is critical for ensuring proper core cooling during accident conditions.

Table 6-123 summarizes the Emergency Core Cooling System Component Parameters which are essential for maintaining core cooling during an emergency.

Table 6-108 outlines the analysis performed on malfunctioning annulus ventilation systems, assessing the potential impacts on plant operations.


Question: material licensing requirements

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Parts 30, 31, 32, 33, 34, 35, 36, 37, 39, 40, 70 and 71, and in reliance on statements and representations heretofore made by the licensee, a license is hereby issued authorizing the licensee to receive, acquire, possess, and transfer byproduct, source, and special nuclear material designated below; to use such material for the purpose(s) and at the place(s) designated below; to deliver or transfer such material to persons authorized to receive it in accordance with the regulations of the applicable Part(s).

The license shall be deemed to contain the conditions specified below.


Question: nuclear activity measurement

The complexity of radiation monitoring systems requires comprehensive training and adherence to strict regulatory standards to ensure effective assessments.


Question: What are the key features of FLEX equipment in PRA models?

FLEX portable equipment is credited in the ANO-2 FPIE PRA (which includes IF) and FPRA.

The addition of bounding impacts for seismic events has no impact on baseline PRA or RTR model.


Question: fuel cladding temperature limit

Verify that the time to complete the transfer of the TSC from the TRANSFER CASK to the CONCRETE CASK and from a CONCRETE CASK to another CONCRETE CASK assures that the fuel cladding temperature limit of 400°C is not exceeded.

A program shall be established to implement FSAR requirements for ISFSI operations.


Question: The documentation specifies that measurements of radioactive isotopes such as U-235 must be reported for regulatory compliance.

U-235 is a radioactive isotope that is crucial for understanding nuclear reactions and is often evaluated in regulatory filings.

The sample analysis includes various forms of uncertainty to ensure the accuracy and reliability of the measurements.


Question: What is the significance of nuclide activity in monitoring radiation levels?

The Wt mean Activity (pCi/M²) of each nuclide listed in the report indicates the amount of radiation emitted per square meter, which is crucial for determining the contamination levels in the environment.

Errors quoted at 2.000 sigma indicate the level of uncertainty in the measurements reported, which is important for establishing the reliability of the data.


Question: The NRC has initiated a review process for the application concerning the subsequent renewal of licenses for the Oconee Nuclear Station.

This appendix contains a chronological listing of correspondence between the U.S. Nuclear Regulatory Commission (NRC) and external parties as part of the agency’s environmental review of the Oconee Nuclear Station Units 1, 2, and 3 (Oconee Station) subsequent license renewal application.

For scoping comments, see Appendix A of this environmental impact statement and the NRC’s “Scoping Summary Report” (Agencywide Documents Access and Management System [ADAMS] Accession No. ML21357A089; NRC 2022-TN8905) and “Second Summary Report” (ML23304A138; NRC 2024-TN9478).


Question: The calculation of Risk Informed Completion Times must consider various hazard groups such as internal flooding risks.

A RICT calculation must include the following hazard groups: internal flood and internal events PRA model, internal fire PRA model, and seismic penalty factor. Changes to these means assessing the hazard groups require prior to NRC approval.

A report shall be submitted in accordance with Specification 5.6.9 before a newly developed method is used to calculate a RICT.


Question: aqueous hydrogen fluoride production

The DOE-Portsmouth produces 70% aqueous hydrogen fluoride but not anhydrous fluoride.

Natural UF6 has the same chemical hazard as depleted UF6.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor includes a comprehensive review of the facility's design, safety analysis, and operational procedures to ensure compliance with regulatory standards.

Framatome specializes in providing nuclear technology and services, including the development of advanced reactor solutions.


Question: Zircaloy cladding oxidation rates

Therefore, the NRC concludes that the petitioner did not provide sufficient information to support the revision of RG 1.157 and 10 CFR Part 50, Appendix K, to require that the rates of energy release, hydrogen generation, and Zircaloy cladding oxidation from the metal-water reaction of zirconium with steam considered in evaluation models used to calculate ECCS cooling performance be calculated based on data from cited experiments, instead of using the Cathcart-Pawel or Baker-Just equations.

Isothermal tests provide this necessary temperature control. However, previous studies have applied the derived correlations to transients that include complex thermal-hydraulic conditions to verify that the proposed phenomena embodied in the correlations are limiting.


Question: The NRC has approved guidance for evaluating the safety of internal components within a BWR reactor.

As part of the BWRVIP initiative, the BWR reactor internals and attachments were subjected to a safety assessment to identify those components that provide a safety function and to determine if long-term actions were necessary to ensure continued safe operation.

Inspection services, by an Authorized Inspection Agency, will be applied to the proposed alternative actions of this alternative request.


Question: nuclear licensing process for technical safety

There are no additional markups to TR-0516-49416-P as a result of this supplemental response.


Question: The NRC staff agrees with Framatome's evaluation that the swelling rate of pellets and their densification are not interconnected and vary significantly across different levels of burnup.

Framatome responded that the densification and solid swelling phenomena are not directly related and are predominant over different burnup ranges.

The densification is primarily impacted by the [        ], therefore if bounding values of [        ] are input, then this should be acceptable.


Question: average enrichment for PWR fuel assemblies

The average enrichment of fuel assemblies in this table is categorized by different ranges of average burnup, specifically focused on the wt % of 235U for each type of assembly configuration listed.

The cooling system is designed to ensure that the reactor core remains adequately cooled during both normal and emergency conditions, preventing overheating and maintaining safety.


Question: What is the purpose of a risk-informed approach in nuclear safety regulation?

The proposed risk-informed approach meets the key principles in RG 1.174 (Reference 2) in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in a very small increase in risk, and is monitored using performance measurement strategies.

No new operator actions are implemented that could affect occupational radiation exposure. No physical modifications or changes to operating requirements are proposed for the facility, including any SSCs relied upon to mitigate the consequences of a LOCA.


Question: aging management evaluation for concrete structures

Increase in porosity and permeability; cracking; loss of material (spalling, scaling) due to aggressive chemical attack


Question: What does NELAC certification ensure in laboratory analyses?

Except as noted, the test results for the methods and parameters listed on ACZ’s current NELAC certificate letter (#ACZ) meet all requirements of NELAC.

If the samples are determined to be hazardous, additional charges apply for disposal (typically $11/sample).


Question: An understanding of the classifications of emergency conditions is vital for ensuring proper responses during nuclear incidents.

Knowledge of the emergency action level thresholds and classifications.

The evaluation settings for the training include a classroom environment where various scenarios may be practiced.


Question: What is the significance of axial flux difference in nuclear reactors?

The AFD is sensitive to many core related parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration.

Safety analyses are critical for ensuring the integrity of the reactor under various operational scenarios.


Question: The NRC mandates that licensees provide detailed safety analysis reports for the construction and operation of nuclear facilities.

The safety analysis report must assess the potential impacts of nuclear facility operations on the surrounding environment and public health, ensuring compliance with regulatory standards.

The geological survey indicated that the soil composition in the area consists mostly of clay and sand, which could affect the stability of future constructions.


Question: nuclear licensing telework policies

The Telework Directive and Operating Procedures have been updated to reflect the changes identified by the OIG. However, due to the COVID-19 pandemic, the Board approved Notice N-125.1 to provide flexibility on the use of telework during the pandemic.

The proposed action meets the intent of the recommendation. This recommendation will be closed when OIG verifies that DNFSB has developed and implemented a checklist for telework recordkeeping to ensure the employee telework files are consistent.


Question: RCS boron concentration

Chemistry reports the RCS boron concentration has dropped more than 50 ppm and is LESS than the limit specified in the COLR.


Question: What is the source of power for the Train A Glycol Return Containment Valve?

The power supply to Train A Glycol Return Containment Isolation Valve, 1-FCV-61-193, is from a 125V DC Vital Battery.

The four options provided each describe a different battery board, but do not specify how they relate to safety protocols.


Question: primary confinement isolation in nuclear safety

1) Main Steam Tunnel Differential Temperature – High


Question: What is the Risk Achievement Worth (RAW) methodology in nuclear safety assessments?

**Risk Achievement Worth (RAW) Methodology:** Section 4, “BTP 7-19 Position 4 Displays and Controls,” of the D3 Analysis examines the five critical safety functions from SECY-93-0087 and defines the diverse controls to achieve each critical safety function and displays to monitor the performance of these functions from the control room.

The D3 Analysis does not specify what the phrase “RAW of ≥ approximately 2” means or how the RAW values are calculated.


Question: What is the significance of Technical Specification TSTF-505 in nuclear licensing?

Technical Specification TSTF-505 provides a pathway for licensees to propose changes to the regulatory framework for the maintenance of certain safety-related systems.

The quarterly review of operational data allows for a comprehensive assessment of plant performance and highlights the importance of adhering to scheduled maintenance activities.


Question: The NRC's regulatory framework mandates compliance with safety standards for spent nuclear fuel storage systems.

This certificate is issued in accordance with 10 CFR 72.238, certifying that the storage design and contents described below meet the applicable safety standards set forth in 10 CFR Part 72, Subpart L, and on the basis of the final safety analysis report (FSAR) of the cask design.

The concrete cask body is a reinforced concrete structure with a carbon steel inner liner.


Question: The NRC ensures that nuclear plants are designed to endure severe environmental conditions without compromising safety.

of the Surry Updated Final Safety Analysis Report (UFSAR) describes conformance with Atomic Energy Commission (AEC) Criterion 2 which corresponds to 10 CFR, Part 50, Appendix A, GDC 2, “Environmental and dynamic effects design bases,” which states that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.

Gain a better understanding of the detailed calculations, analyses, and bases underlying the LAR and confirm the NRC staff’s understanding of the LAR.


Question: general license regulations for nuclear devices

C. I am aware of the requirements of the general license, provided in 10 CFR 31.5.

A. All Information contained in this registration is true and complete to the best of my knowledge and belief.


Question: What are the assessments and analyses performed for hurricane impacts on nuclear safety?

The NRC's assessment of hurricane impacts includes detailed analyses of historical storm tracks, tide data, and meteorological criteria for hurricanes, which inform the site safety analysis and emergency response planning for nuclear facilities.

PSEG's Site Safety Analysis Report discusses several regulatory processes and requirements that nuclear facilities must meet for licensing, including quality assurance protocols and risk assessments, but does not specifically address hurricane impacts.


Question: What is thermal aging embrittlement in cast austenitic stainless steel?

Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) is a phenomenon that occurs in austenitic stainless steel components, particularly those subjected to high temperatures over extended periods, leading to a loss of ductility and an increase in brittleness.

Periodic Inspections for Selective Leaching are essential to ensure the structural integrity of piping systems, involving planned examinations to assess material degradation due to environmental factors.


Question: What are the compliance requirements for the RadICS Platform in nuclear licensing?

The Curtiss-Wright System Development Methodology meets the requirements specified in the RadICS Platform Topical Report and meets all NRC requirements for licensing the Peach Bottom ECCS CLS Replacement System.

The following slides identify those sections of the RadICS Platform Topical Report that specify requirements for applications that utilize the RadICS Platform and define how the System and Application Development Methodology used for the Peach Bottom ECCS CLS Replacement Project complies with those requirements.


Question: What are the radioactivity levels of radionuclides in soil samples?


Question: The NRC's regulatory framework mandates thorough evaluations during licensing processes.

The NRC conducts detailed safety evaluations as part of the licensing process before granting operating licenses to nuclear power plants.

The name Alfred Krall has been associated with various regulatory activities within the NRC.


Question: free drop test methodology

Figure 2.12.1-3 illustrates the setup for CTU-1's Free Drop Test No. 1, depicting the method used to assess the performance of safety equipment.

The reactor coolant system operates under high pressure to ensure that the core is adequately cooled during normal and emergency conditions.


Question: The communication systems installed in medical facilities ensure reliable links between hospitals and ambulances.

A coordinated communication link for hospitals and ambulances is provided by the Medical Support Facilities Communications Systems.

Fixed, mobile and hand-held radios operating on law enforcement and fire/rescue frequencies provide for communication means between the Parish EOC and representatives from the following organizations:


Question: Species diversity and habitat utilization research was performed to assess the fauna within the specified site area.

A mammalian survey was also performed during the spring of 1973 to determine species diversity and habitat utilization.

The third most preferred habitat area appeared to be cultivated and old fields.


Question: What are the requirements for the operational performance criteria in a walkthrough test?

The operating test shall not duplicate items from the applicants' audit test(s) (see Section D.1.a.).

The associated simulator operating tests (scenario sets) have been reviewed in accordance with Form ES-301-4, and a copy is attached.


Question: The NRC has highlighted the importance of understanding stress corrosion cracking (SCC) in materials, particularly in relation to maintaining reactor component integrity.

SCC refers to stress corrosion crack initiation and growth of susceptible materials under roughly constant stress operating conditions due to the corrosive environment.

While fatigue can be a concern and lead to component failure, it is generally addressed conservatively through design standards and has not generally led to many safety-significant failures or flaws.


Question: What do the site boundaries represent in a nuclear facility?

The Site/Exclusion Area Boundary is represented by a dashed line on the map, indicating the boundary within which certain regulatory requirements are applicable.

Lake Michigan is a prominent geographical feature influencing the environmental aspects of the Palisades Site.


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) represents the lowest activity level that can be reliably detected for a specific nuclide in a given sample, expressed in pCi/grams.

The Peak Locate From Channel indicates the range of channels wherein the peak was detected during the analysis process.


Question: The NRC assesses the environmental impacts of operating cooling systems at nuclear plants to protect terrestrial resources.

For initial license renewals, the NRC codified its conclusion that doses to terrestrial organisms from continued operations and refurbishment associated with license renewal are expected to be well below exposure guidelines developed to protect these organisms.

The plant releases small quantities of radioactive materials in a gaseous form and does not make routine releases of radioactive liquids.


Question: The NRC will conduct an overview of the proposed undertaking during the visit.

The purpose of the site visit is to conduct an overview of the proposed undertaking, focusing on the operational aspects of the Palisades Nuclear Plant.

Holtec has reserved parking spots for the NRC and Tribal governments.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves multiple stages, including pre-application consultations, the submission of safety analysis reports, and public hearings, ensuring compliance with regulatory standards set by the Nuclear Regulatory Commission (NRC).

Plant operators must consistently monitor the weather conditions to ensure safe operations and manage thermal efficiency.


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX software is essential for analyzing critical heat flux in nuclear reactors and is an advancement of the original COBRA code designed to provide improved thermal-hydraulic modeling capabilities.

The results from the grid tests indicated a significant improvement in the efficiency of the cooling system utilized in the reactor core.


Question: The wind-wave activity calculation involves analyzing the effective fetch for the site.

DRAWING SHOWING FETCH AND RADIALS USED TO DETERMINE EFFECTIVE FETCH FOR WIND-WAVE ACTIVITY CALCULATION AT INTAKE SCREENWELL AND SITE.


Question: The regulation of radiopharmaceuticals is critical for ensuring patient safety during medical procedures.

The NRC regulates the use of radiopharmaceuticals to ensure their safe application in medical treatments and diagnostic procedures.

The findings indicate a substantial increase in the number of radiological treatments administered to patients over the five-year period.


Question: The equations by Cathcart-Pawel and Baker-Just fail to properly account for the hydrogen production observed in specific experimental tests under critical conditions.

The NRC staff further concluded that use of the Baker-Just equation results in sufficiently conservative calculations of the metal-water reaction rate that are appropriate for conservative ECCS evaluation models.

The NRC neither agrees nor disagrees with the comments. The cited experiments were performed to better understand reactor behavior under severe conditions.


Question: The NRC has determined that sufficient justification was not provided to modify regulations concerning core reflow rates.

The NRC is denying PRM-50-93 and PRM-50-95. The petitioner did not present sufficient new information or arguments to support the requested changes.

The NRC reaffirms that its existing regulations continue to provide reasonable assurance of adequate protection of public health and safety.


Question: What is the function of the AACDG in a nuclear facility?

The AACDG (commonly referred to as the station blackout diesel generator) can also be started and connected, from the Control Room, to power either 4.16 kV switchgear. The AACDG is a non-seismic backup power source with a higher rating than the safety-related DGs.

Control Room Chillers are crucial for maintaining the temperature of electronic equipment within the control area, ensuring their optimal operation and reliability.


Question: The NRC oversees the storage and organization of various technical documents crucial for reactor operations, such as the New Production Reactor documentation.

**New Production Reactor (NPR)** documents at INL Storage (125 boxes-Idaho Falls). Working with Red Ink to scan and organize files for Export/Classification Reviews. Effort is underway.

**Clinch River Breeder Reactor (CRBR) Project** documents. Contract initiated in Feb 2020. Scanning on hold (COVID-19). Iron Mountain will proceed as soon as possible (235 boxes + 75 reels of microfilm).


Question: What is the purpose of monitoring MFWP suction pressure in nuclear operations?

The reason that the ARI directs MFWP suction press to be monitored is to ensure that it is maintained greater than 250 psid. If unable to recover suction pressure, the crew is required to trip one MFPT and go to 1-AOI-16, Loss of Normal Feedwater.

Plausible because the first part is correct. The reason that the ARI directs MFWP suction press to be monitored is to ensure that it is maintained greater than 250 psid. The second part is plausible because the low suction pressure condition affects both MFPs.


Question: Lead 210 regulation and measurement

The testing team employed a series of standardized methods to ensure that all data collected are both reliable and accurate, following established protocols for data analysis.


Question: nuclear decay products

The table provides a comprehensive overview of various nuclides along with their decay products and fractions, pertinent to understanding nuclear decay processes.

The reactor's cooling system includes three separate loops designed to maintain optimal temperature and pressure during reactor operations.


Question: Dominion has to provide an evaluation of any new evidence that may influence the decision concerning the renewal of the reactor licenses.

As required by 10 C.F.R. § 51.53(c)(iv), which requires an applicant to include in its ER a review of new and significant information, Dominion reviewed the North Anna PRA “to determine if there is any new and significant information regarding the SAMA analyses” that would affect the NRC’s decision to renew the licenses for an additional 20-year term.

In its first step of its qualitative screening process, the ER states that Dominion collected 283 industry SAMAs for evaluation.


Question: What is the purpose of the Masonry Walls aging management program?

The BFN Masonry Walls aging management program is an existing condition monitoring program and includes all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4.

Steel bracing, reinforcing and supports that are part of the masonry wall design are managed by the Structures Monitoring program (A.2.1.33).


Question: The NRC assesses the structural integrity of safety-related systems to ensure compliance with safety standards.

The spent-fuel pool is an essential safety-related structure that must maintain its integrity to prevent the release of radioactive materials.

Emergency preparedness drills are critical for ensuring that personnel can effectively respond to unexpected situations.


Question: What is the role of aerosol removal in nuclear containment systems?

Sprays: Aerosol Removal Data provides the removal coefficients for aerosols in the sprayed containment volume, indicating how effective the removal process is over time.

Recruitment: 10 (%) indicates the percentage of receptors utilized in the deposition study, but does not relate to aerosol removal efficiency or containment processes.


Question: Quality Assurance Program changes in nuclear regulation

SNC is proposing changes to its Quality Assurance Program (QAP).

The purpose of the meeting is to discuss the primary changes to the QAP as well as SNC’s plan for submitting information to the NRC.


Question: gamma spectrum analysis

Gamma spectrum analysis was performed on the sample taken on 8/9/2019 1:48:00 PM using a 500 ml Marinelli geometry.

The sample size recorded was 6.981E+02 grams and the facility conducting the sampling was Daryland_NPP.


Question: The NRC conducts assessments to ensure compliance with safety protocols for radioactive materials used in reactor facilities.

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

The following Prep Methods were performed:


Question: seismic hazard evaluation

For the ESP seismic hazard evaluation, the 1989 EPRI PSHA was reproduced using the 1989 seismic sources and 1989 ground motion models.

These ground motion models were used without any correction for soil conditions, because, for purposes of the 1989 EPRI study, the North Anna site was considered a rock site.


Question: ECCS injection subsystem requirements

Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.


Question: The NRC recognizes that the development of operating procedures must take plant-specific activities into account during the application process.

Operating procedures and retirement assessment activities are considered by the NRC staff to be plant specific activities. The NRC staff agrees that these activities can be performed as part of the application development. It is therefore acceptable for these activities to be addressed during application development and implementation.

Addressed in PSAI 7.2 and satisfied by: - Licensing Topical Report - Operating procedures and retirement assessment will be performed by Constellation.


Question: What is the significance of regulatory communications in nuclear safety?

The NRC requires timely and effective communications regarding safety-related regulatory matters to ensure that all licensees understand their obligations under the law.

Hi Ken, I’ve compiled a list of the different topics where we would like to touch base with the PNNL team for additional discussion.


Question: What are the requirements for a motion for reconsideration under NRC regulations?

Under these regulations, a motion for reconsideration must demonstrate 'compelling circumstances, such as the existence of a clear and material error in a decision, which could not have been reasonably anticipated, that renders the decision invalid.'

The court noted that the NRC 'fail[ed] to identify any statute that authorizes it not to comply with NEPA on equitable grounds,' and in Powertech we explicitly did not decide 'whether, or under what circumstances, an NRC presiding officer should perform an Allied-Signal-style equitable analysis in the first instance upon finding a significant NEPA deficiency.'


Question: Reactor coolant system temperature must be maintained within specific thresholds to ensure safety during operation.

The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits shown in Figure 1.

The cycle-specific parameter limits for the specifications listed above are presented in the following subsections.


Question: What is the concentration of Nickel-63 measured in the nuclear sample?

Nickel-63 was measured at a concentration of 0.774 pCi/g in the sample QC1204338395.

Technetium-99 was reported to have a concentration of -0.357 pCi/g in the sample QC1204338401.


Question: The NRC staff will engage with experts to discuss environmental information needs related to the audit.

The supplemental environmental information needs of the U.S. Nuclear Regulatory Commission (NRC) staff for the Peach Bottom Nuclear Plant, Units 2 and 3 subsequent license renewal (SLR) application have been described in three categories: virtual tours, meetings, and information needs and document requests.

The online portal will be password-protected, and separate passwords will be assigned to the NRC staff who are participating in the audit.


Question: fiber penetration testing in nuclear strainer systems

The test strainer for penetration testing was comprised of two prototypical strainer stacks that matched the key design parameters of the plant strainer stacks.

The upstream portion for debris introduction and mixing, and the pit region where the test strainer was installed, as shown in Figure 3.4-1.


Question: The licensing process requires thorough evaluation to ensure compliance with safety standards.

The Terrapower Kemmerer Power Station Unit 1 is under construction and is undergoing a detailed review for its construction permit.

The document distribution includes various members from different departments, such as NMSS and NRR.


Question: What are the key components of a Safety Analysis Report for nuclear facilities?

The Salem Generating Station Updated Final Safety Analysis Report outlines the various systems designed to ensure safety, including reactor core safety, containment structures, and emergency response protocols.

The atmospheric conditions in the Eastern United States can vary greatly, affecting everything from precipitation patterns to temperature fluctuations throughout the year.


Question: The analysis of new fuel racks requires adherence to several regulatory benchmarks and guidelines.

New fuel rack analysis description is detailed in NEI 12-16, RG 1.240, using MCNP 6.2.

Spent fuel storage rack analysis description is not applicable.


Question: Details about the implementation procedures for reception centers are available in the official documentation.

Information on the layout and capabilities of reception centers can be found in the Reception Center implementing procedures.

Parish evacuees are directed to the reception center in East Baton Rouge Parish.


Question: What is the significance of I-131 in nuclear safety assessments?

I-131 is a radioactive isotope used in various medical applications, but it is also significant in nuclear safety assessments due to its health risks when released.

Containment structures are designed to prevent the release of radioactive materials during normal operations and accidents.


Question: The ASHPLUME2 model uses input parameters to simulate ash plume dispersion from volcanic events.

Parameters are input into the ASHPLUME2 code through the ASHPLUM.IN file.

The average speed during certain measurements was recorded at various atmospheric pressure levels.


Question: The aging management programs are evaluated against the standards set forth in NUREG-2191, ensuring component integrity over time.

The TLAA designation in the Aging Management Programs column indicates that cumulative fatigue damage due to fatigue for this component is evaluated in Section 4.3.


Question: dry weight basis

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

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, with the following exceptions.


Question: RadICS platform design capabilities

The NRC staff determined the RadICS platform design is generally capable of supporting timely recognition, location, replacement, repair, and adjustment of malfunctioning equipment.

Addressed in PSAI 7.11 and satisfied by: - Licensing Technical Report


Question: What are the measurements associated with radioactive isotopes in terms of energy and activity?


Question: The NRC requests an evaluation of the heat transport systems used in advanced reactors.

The Primary Heat Transport System (PHTS) is a crucial component in which heat is transferred from the reactor core to the heat exchangers in the nuclear power plant system.

The regulatory framework requires licensees to submit periodic financial reports detailing their operational expenditures and revenues.


Question: Reactor vessel head replacement inspections

Conduct RV head post-installation verification and testing inspections in accordance with the inspection plan.

Comprehensive NRC inspection of RV replacement activities involves coordination of inspections to review engineering, safety assessment, welding, NDE, quality assurance, radiation protection, security, and testing.


Question: What does TSS refer to in the context of nuclear reactor safety analysis?

The TSS (Temperature Safety System) is analyzed through results such as average liquid temperatures and exit pressures to ensure safety during reactor operations.

The NRC has set comprehensive regulations that govern the operational procedures for nuclear facilities.


Question: What are license renewal boundary drawings in nuclear regulations?

The License Renewal Boundary Drawings include various sheets such as SLR-DRE-M-12, Sheet 2 and SLR-DRE-M-21, which provide detailed layouts and specifications for the license renewal process of nuclear facilities.

The UFSAR references various sections like 1.2.4.4.2 and 9.3.1.6 that outline fundamental design and safety parameters critical to the operational integrity of the facility.


Question: What is gamma spectrum analysis in nuclear safety assessments?

Gamma spectrum analysis is a critical process used to identify and quantify radioactive isotopes in samples, ensuring accurate monitoring of safety and compliance in nuclear facilities.

The sample was identified as L1-010-106-AJGS-029-SS and taken on 6/10/2019 at Dairyland_NPP, reflecting the operational timeline for regulatory filings.


Question: import/export license applications

The audit found that OIP approved license applications in a timely manner in compliance with 10 CFR 110 and effectively coordinated with other Federal agencies.

However, license processing costs are under-recovered and the internal coordination relating to the proper review and approval of import/export license applications requires improvement to correct documentation weaknesses.


Question: The calculation includes doses measured at specific locations throughout the facility.

Dose Location 1 is defined in terms of its coordinates, measurements, and dose values represented in scientific notation.

Simulation parameters for the calculations were established to ensure accuracy and reliability in the risk assessment.


Question: The assessment of uncertainty in probabilistic risk assessments is crucial for evaluating nuclear safety.

The treatment of parameter and model uncertainty is essential for developing accurate probabilistic risk assessments (PRAs) for nuclear facilities.

The report outlines various methodologies and standards regarding nuclear power plants and their operational efficiencies.


Question: The components of the primary loop in a nuclear reactor include critical elements such as the reactor vessel and the coolant pump.

The General Arrangement of Primary Loop Equipment features a three-loop layout that includes essential components like the Reactor Coolant Pump, Reactor Vessel, and Steam Generator.

The scale provided for the diagram is measured in feet, ensuring accurate representation of the physical dimensions of reactor components.


Question: What is the function of the Auxiliary Component Cooling Water system in a nuclear facility?

Auxiliary Component Cooling Water is designed to provide cooling to auxiliary systems during operation, and it is classified as a process system that is offline, with a mechanical code of ASME III-3.

The Turbine Building Drain stops flow from turbine and control building drains system to the waste water retention basin.


Question: What is the significance of G/Gmax curves in seismic analysis for nuclear plants?

G/Gmax and Material Damping Curves for Soil Layer 1 provide important insights into the soil's response during seismic events, which is critical for ensuring the safety and stability of nuclear structures.

Site Stratigraphic Column illustrates the various geological layers present at the site and their respective attributes, which impacts overall site characterization but does not specifically address seismic response.


Question: What temperature range is suitable for the growth of Legionella bacteria?

Legionella is susceptible to temperatures from 95°F to 122°F (35°C to 50°C); can persist in stagnant water.

Infections occur when the amoeba enters the brain through the nose while swimming in warm lakes, rivers, or hot springs.


Question: What is the yield percentage of K-40?

The energy tolerance for measurements is set to 1.000 keV.


Question: ground motion seismic analysis

The type of ground motion is categorized into various levels based on the observed seismic activity, which include strong shaking associated with local seismic sources that could affect infrastructure.

Probability assessments typically range from low to high based on the historical seismicity, with measurements taken from geological surveys across the region.


Question: The report indicates that the analysis of Sr-90 was conducted on the selected samples.

Sr-90 was not detected on the off-site analysis of samples L2-011-101-FSGS-005-SS and L2-011-101-FSGS-013-SS.

Gamma spectroscopy results revealed that no ROC were positively identified any judgmental sample.


Question: What is the purpose of a check valve in nuclear safety systems?

Check valves are utilized in nuclear systems to prevent backflow and ensure that fluid only moves in one direction, enhancing the safety and operational reliability of the system.

The test pressure for check valves is critical to ensure that they function properly under normal operating conditions.


Question: What is the role of containment structures in nuclear reactors?

Containment structures are essential to prevent the release of radioactive materials in the event of an accident and are designed to withstand extreme pressures and temperatures.

The heat exchangers in the reactor coolant system play a critical role in maintaining optimal temperature during normal operations and disasters.


Question: What are the key safety systems in a nuclear reactor?

Safety systems at a nuclear power plant are designed to prevent accidents and mitigate consequences in case of a potential incident, ensuring the safety of the public and the environment.

The design of the reactor vessel is crucial for maintaining high pressure and temperature, which are necessary for efficient nuclear fission.


Question: high pressure coolant injection system operation

Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: Turbine operation: BWR-2,3,4

A plant transient results in the following:

| Time (min) | Condition                              |
|------------|----------------------------------------|
| 0          | Reactor water level is 126.5” and lowering. |
| 5          | Reactor water level is 222.5” and rising.   |
| 10         | Reactor water level is 222.5” and lowering. |
| 15         | Reactor water level is 126.5” and lowering. |


Question: The assembly average burnup must be considered for various enrichment levels.

| Minimum Initial Assembly Avg. Enrichment wt % 235U (E) | Assembly Average Burnup (B) GWd/MTU |

The reactor coolant system is designed to ensure effective heat removal during normal operations.


Question: How is the containment evacuation alarm activated in the facility?

The containment evacuation alarm is actuated from the 1PM05J via a switch. The alarm will continue to sound until the switch is returned to its normal location.

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


Question: The NRC mandates that operators assess the structural integrity of bolting to prevent material loss due to environmental conditions.

The Structures Monitoring Program (B.2.3.33) is designed to manage the aging effects such as loss of material in structural bolting due to atmospheric and weather-related influences.

None of the components listed in the aging management evaluation require immediate repair or replacement.


Question: What are design-basis accidents in nuclear plants?

Design-basis accidents are defined as events that could occur as a result of postulated accidents and are considered in the safety analysis to ensure that the plant can withstand certain conditions without releasing significant quantities of radioactive material.

Microbiological hazards to plant workers can arise from exposure to bacteria or viruses present in the workplace environment, making it essential to implement health and safety measures.


Question: HYSPLIT to MACCS conversion

The HYSPLIT model outputs concentrations of pollutants, which can be formatted as input for the MACCS system to assess dose, health effects, and risk.

Release location and aerosol properties are critical factors that influence the dispersion of contaminants during a nuclear incident.


Question: The analysis of peaks within the gamma spectrum is critical for identifying isotopes present in a sample.

Peak Area Range (in channels): 100 - 4096


Question: radioactive material license

Pursuant to the Radiation Control Act, 1963, and 180 NAC 3 of the Nebraska regulations, 'Control of Radiation', and in reliance on statements and representations heretofore made by the licensee designated below, a license is hereby issued authorizing such licensee to transfer, receive, possess and use the radioactive material(s) designated below; and to use such radioactive materials for the purpose(s) and at the place(s) designated below.

The Associate RSO(s) for this license is/are: Randy Safety.


Question: What is the significance of the activity of nuclides in radiation safety assessments?

The table indicates the weight mean activity of the nuclide PB-214 is 4.53E-02 pCi/grams, which is crucial for assessing its radiological impact in safety evaluations.

Errors quoted at 2.00 sigma indicate the statistical accuracy of the measurements, which helps in interpreting the reliability of the data presented.


Question: electric power distribution subsystem

The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, and 3 to ensure that: a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOO or abnormal transients; and b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained, in the event of a postulated DBA.

The overall reliability is reduced, however, because a single failure in the remaining power distribution subsystems could result in the minimum required ESF functions not being supported.


Question: The NRC stipulates that systems essential for safe operation must be supported by structures that comply with relevant safety regulations.

Provides physical support, shelter, and protection for systems, structures, and components relied upon in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for Fire Protection (10 CFR 50.48). 10 CFR 54.4(a)(3)

Provides physical support, shelter, and protection for nonsafety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of function(s) identified for 10 CFR 54.4(a)(1). 10 CFR 54.4(a)(2)


Question: What is the purpose of the Technical Specifications for a nuclear reactor?

Technical Specifications are part of the operating license for a nuclear reactor and contain the requirements necessary to operate the facility safely.

Licensees must submit periodic reports detailing their compliance with the financial obligations set forth in the initial application.


Question: SAI indicators reporting

The SAI (Safety Assessment Indicator) is a critical measurement used in evaluating the safety performance of specific components within the reactor system, providing insights into operational integrity and safety compliance.

TOTAL is a classification in the data table, indicating the aggregate values of given measurements across various locations of the reactor, but does not pertain to the assessment of safety indicators.


Question: What is the significance of pCi/g in nuclear measurements?


Question: How does the harvesting of components relate to the renewal of licenses for nuclear power plants?

Please describe your interest in harvesting components that were aged in representative light water reactor (LWR) environments in nuclear power plants (NPPs).

What should NRC be doing differently to get ready for SLR? Would you like NRC to have another public meeting on SLR/harvesting? Other suggestions?


Question: The NRC licensing process involves verifying that the reactor power level stays within specified limits.

VERIFY Reactor power level less than or equal to 48%.


Question: What is the NARWHAL software used for in nuclear licensing?

The use of the NARWHAL software in place of the WCAP base model spreadsheet. The NARWHAL software follows the WCAP-16530-NP-A methodology for chemical product generation.

Licensees should list the type (e.g., AIOOH) and amount of predicted plant-specific precipitates.


Question: control rod position measurement

Control Rod Position: Each group is indicated during rod motion. These signals are used in conjunction with the measured position signals (4b) to detect deviation of any individual rod from the demanded position. A deviation will actuate an alarm and annunciator.

The highest of the four \(T_{avg}\) channels into the auctioneer will be passed to the recorder.


Question: What is the significance of discharge measurements in nuclear regulatory filings?

The discharge measurements recorded in cubic feet per second (cfs) represent the rate at which water is discharged from a nuclear facility, which is critical for regulatory compliance and environmental monitoring.

The NRC mandates regular inspections and assessments of the reactor core to ensure safe operations, which are vital for preventing any potential safety hazards.


Question: significant hazards consideration

If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing.

A State, local governmental body, Federally recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1).


Question: What is the activity of CS-137 according to the nuclide identification report?

CS-137 has an activity of 1.21E+00 pCi/grams in the identified nuclides section.

The interference corrected report provides weighted mean activities and uncertainties for various nuclides.


Question: aging management evaluation for fire protection systems

Fire Water System (B.2.15) involves management programs that address aging effects requiring management such as loss of material in components, particularly in carbon steel piping used in raw water.

Periodic Inspections for Selective Leaching (B.2.3.41) are crucial for maintaining the integrity of buried and underground piping and tanks.


Question: The NRC oversees the safety oversight process during the construction of new reactors, ensuring compliance with established regulations.

The Construction Reactor Oversight Process provides a framework for NRC oversight activities during the construction of new nuclear reactors, ensuring safety and compliance with regulatory requirements.

The inspection process for vendor inspections focuses on evaluating the quality and safety of materials used in reactor construction.


Question: The NRC issued an ITAAC notification to address the deletion of components related to the construction of nuclear facilities.

ITAAC Deleted Per License Amendment 113 reflects the NRC's decision to remove specific inspection requirements from the licensing process, marking a significant regulatory adjustment.

The inspection reports include detailed analyses of construction processes and materials used in the facility, ensuring compliance with established safety standards.


Question: The NRC develops regulations to guide the decommissioning process for nuclear facilities, ensuring that materials can be safely harvested and reused.

Forward-looking lessons learned from past harvesting programs include pitfalls to avoid and strategies to improve likelihood of success.

Short 5-10 minute presentations followed by open discussion will focus on the sources of materials, including decommissioning plants.


Question: What is the ALARA principle in nuclear materials safety?

Maintaining Effluents from Materials Facilities As Low As Reasonably Achievable (ALARA) is a principle aimed at ensuring that radiation exposure is kept to a minimum while still allowing for effective operations.

Decommissioning Inspection Procedure for Materials Licensees provides the guidelines for assessing the status and compliance of materials licensees during the decommissioning phase.


Question: The NRC mandates that inspections during the pressurizer replacement process must include evaluations of heavy lifting procedures and rigging preparations.

Conduct pressurizer removal and replacement inspections in accordance with the inspection plan by performing selective inspections, consistent with the safety significance and inspection resources, of the following areas: a. Review the following welding and non-destructive examination (NDE) activities: 1. Where applicable, special procedures for welding and NDE. 2. Training and qualifications for personnel performing welding and NDE. 3. NDE including radiography results and work packages for selected welds. 4. Completion of any pre-service NDE requirements, where required.

Training of personnel is crucial for ensuring safety in all operations, but particularly in high-risk areas such as assembly line production and material handling.


Question: mean activity of nuclides

Wt mean Activity (pCi/grams) represents the weighted mean activity of the nuclide, providing a measure of the radioactive decay rate per gram of the nuclide material.

Confidence levels for the analysis can vary, and it is essential to consider the overall context of the measurements being reported.


Question: Wetlands can be found along the borders of various water bodies, and they significantly contribute to ecological functions such as erosion control and water quality improvement.

As defined in 33 CFR 328.3, wetlands are those areas that are inundated or saturated by surface or ground water at a frequency and duration sufficient to support a prevalence of vegetation typically adapted for life in saturated soil conditions.

However, detailed mapping of EFPC on the FIRM ends upstream of the HCS at the Highway 95 crossing of EFPC.


Question: What is a nuclide MDA in nuclear measurements?

The Nuclide MDA (Minimum Detectable Activity) is a critical parameter used to assess the lowest level of radioactivity that can be reliably detected in a sample, typically expressed in pCi/grams.

Peak Locate From Channel: 100 indicates that the identification of the peak energy was made using channel 100 in the spectroscopic analysis.


Question: The NRC certifies that the packaging and contents comply with the safety standards for transporting radioactive materials as defined in Title 10, CFR Part 71.

This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards set forth in Title 10, Code of Federal Regulations, Part 71, “Packaging and Transportation of Radioactive Material.”

This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or other applicable regulatory agencies, including the government of any country through or into which the package will be transported.


Question: The accuracy of the primary flow element was determined to be ±1.3% for use in calculations.

Therefore, the Primary Element Accuracy (PEA) for flow element 1G33-N040 is 0.727% plus 0.5% = ±1.227%. This value is rounded up to ±1.3% for use in the setpoint calculation.

The distance downstream of the flow element to check valve 1G33-F051 is 2'-3". This is equivalent to 7 pipe diameters.


Question: The ACRS advises the Commission on technical safety assessments and licensing issues.

The ACRS is a group of experienced technical experts, independent of the NRC staff, that reports directly to and advises the Commission on safety issues related to the licensing and operation of nuclear power plants, along with issues of health physics and radiation protection.

Halnon has expertise in nuclear plant operations as well as quality standards, security, maintenance and engineering processes.


Question: peak energy measurements in nuclear safety

The weather patterns affected our scheduling of maintenance operations throughout the year.


Question: What is the purpose of safety injection samples in nuclear reactors?

Safety injection samples are taken from various points within the reactor coolant system to monitor the integrity and safety of the system during operation.

The valve test type note indicates the normal operation conditions under which the leak class is determined.


Question: Discuss the allowable limits on radionuclide releases within the Natrium design framework.

Natrium PDC 10 utilizes the concept of specified acceptable system radionuclide release design limits (SARRDLs) instead of the SFR-DC 10 concept of specified acceptable fuel design limits (SAFDLs).

Audit meetings will take place in a virtual format, using Microsoft Teams or another similar platform.


Question: The process for obtaining a nuclear operating license includes various stages of review and public comment.

The licensing process for commercial nuclear power plants involves detailed safety evaluations and assessments by the NRC.

The reactor's turbine system was designed to maximize efficiency and reduce waste during operation.


Question: The licensing process requires thorough evaluation of reactor performance under various operating conditions.

Ability to evaluate the following parameters and/or conditions as they apply to a degraded grid: (CFR: 41.7 / 43.5 / 45.6)

The system was extensively tested to ensure reliability and safety across multiple operational scenarios.


Question: LCO 3.0.4.b requirements for system availability

LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.


Question: The NRC has specific guidelines for managing the integrity of piping components in safety systems that are exposed to various environmental conditions.

The External Surfaces Monitoring of Mechanical Components program (B.2.1.23) will be used to manage loss of material of elastomer during components exposed to air - indoor uncontrolled in the Standby Gas Treatment System.

Not Applicable. There are no aluminum piping, piping components, or tanks exposed to air with borated water leakage in ESF Systems or RCIC System.


Question: What are NPDES permit requirements for nuclear plants?

As part of the current NPDES permit requirements, MNGP is required to conduct annual fish seining studies to observe the relative abundance and species composition of the small fish community in the vicinity of MNGP.

The most recent electrofishing studies were conducted from 2016–2021 and summarized in biennial reports.


Question: Xcel Energy conducted a thorough review of current meteorological data for the Monticello Nuclear Generating Plant site, confirming prior assessments.

For the proposed SLR of MNGP, Xcel Energy completed a review of the most recent meteorological information available from public sources and from MNGP monitoring to confirm the conclusions of those previous reviews remain valid.

The terrain is generally flat with rolling hills.


Question: gamma measurement procedures in nuclear safety


Question: fluid-structure interaction in seismic design

In the SDAA, the reactor pool is modeled with FLUID elements of ANSYS and using the fluid-structure interaction (FSI) technology.

33 questions resolved in audit, resulting in the following details and updates added to the Final Safety Analysis Report (FSAR): modal analysis, double building model dimensions, and pool sloshing.


Question: What is the significance of a COL application in nuclear licensing?

A Combined License (COL) application is a regulatory submission to the NRC that seeks authorization to construct and operate a nuclear facility based on a design that has received approval from the NRC.

The reactor core needs to be evaluated for thermal-hydraulic performance under various operational conditions, which is essential for ensuring safety and efficiency in power generation.


Question: The fuel assembly must experience adequate cooling time to ensure safety regulations are met.

The table shows various ratings associated with the assemblies used in commercial nuclear reactors.


Question: alternatives to ASME Code

Proposed alternatives to ASME Code, Section III, 2001 Edition with Addenda through 2003, other than the alternatives listed in Table 2.1-2 of the FSAR, may be used when authorized by the Director of the Office of Nuclear Material Safety and Safeguards or designee. The request for such alternatives should demonstrate that:

The American National Standards Institute ANSI N14.6 (1993) and NUREG-0612 govern the TRANSFER CASK design, operation, fabrication, testing, inspection, and maintenance.


Question: What is the function of the in-containment refueling water storage tank during an emergency?

In-containment refueling water storage tank actuation 4.1

The automatic depressurization system is used to maintain safety during abnormal conditions.


Question: What are the safety protocols for reactor design approval?

The safety evaluation process for reactor design approval includes a comprehensive review of the design’s adherence to applicable safety standards, regulatory requirements, and the identification of potential hazards.

The committee also discussed the importance of public engagement in the rulemaking process, emphasizing transparency and the need for effective communication with stakeholders.


Question: What are the requirements for control room simulators in nuclear facility licensing?

The inspectors evaluated the adequacy of the facility licensee’s control room simulator in modeling the actual plant, and for meeting the requirements contained in 10 CFR 55.46.

The inspectors evaluated the effectiveness of remedial training conducted by the licensee, and reviewed the adequacy of re-examinations for licensed operators who did not pass a required requalification examination.


Question: NPDES permit renewal status

ER Section 3.5.1.3 states that TVA has applied for and anticipates renewal of their NPDES permit (No. AL0022080) in 2023.

Please provide a knowledgeable person to discuss the condenser circulating water system and the two modes of operation (open and helper modes), including the consumptive water use while operating in the two modes.


Question: What are circular flood sources in nuclear facilities?

Circular Flood Sources Types refer to specific configurations used in nuclear facilities to ensure safety and manage flood scenarios effectively.

Alpha-Omega Services, Inc. provides various models of equipment for different applications in the nuclear sector.


Question: battery thermal analysis in transport packages

Demonstrate that the batteries, which are new components to the package, do not affect package temperatures as reported in section 3.3.1 of the application and are bounding during normal transport conditions (NCT).

Clarify and correct as necessary the weld details for the support saddle (Item 5) to the housing base (Item 6) shown on the SAR drawing IR100ST-B, Sheet 3, Revision 0.