Question: technical specifications markup

The Technical Specification Page Markup details the specific requirements, limiting conditions, and operational modes of the nuclear facility as mandated by the regulatory authority.

The attachments include various documents related to the operational performance of the reactor, including inspection reports and maintenance logs.


Question: The report details the net peak area measurements for different energy levels, highlighting the importance of accurately assessing these values in safety analyses.

The table provides a summary of the background subtract report, which outlines the original area and the uncertainties involved in the measurements.


Question: Standby Gas Treatment systems

The Reactor Building is isolated and Reactor Building differential pressure is restored with two trains of Standby Gas Treatment in service.

*30. Ensure open TRAIN B(A) INLET 01-125MOV-14B(A).


Question: What is the significance of background subtraction in nuclear measurement?

Background subtraction is a crucial process in nuclear measurement that involves removing the contribution of ambient radiation to obtain a more accurate representation of the signal from the target material.

The Ambient Background represents the natural radiation present in the environment, which can affect the results of the measurements taken during nuclear experiments.


Question: decommissioning costs of nuclear facilities

New York argues that the PSDAR and DCE are inadequate because HDI allots only one year per unit for reactor-internals and pressure-vessel segmentation.

In its reply brief, New York does not challenge the Applicants’ claim that Entergy, rather than HDI, will pay for disposing of PCB-contaminated material at Indian Point.


Question: The NRC analysis requires detailed meteorological data to assess potential environmental impacts.

The monthly and annual joint frequency distributions of wind speed and direction by atmospheric stability class are crucial for evaluating environmental conditions at nuclear facilities.

The Unit of Measure for radiation exposure is often expressed in millirems or sieverts, which helps to quantify dosages effectively.


Question: What is the site-specific environmental impact statement for nuclear plants?

The Site-Specific Environmental Impact Statement (EIS) for License Renewal of Nuclear Plants evaluates the potential environmental impacts associated with the renewal of the operating licenses for existing nuclear facilities.

During the review process, the staff noted that the proposed changes to the facility were well-researched and presented no significant concerns regarding operational safety.


Question: What is a Curie in radiation measurement?

Ci is an acronym for Curie, which is a unit of radioactivity defined as the quantity of a radioactive material that produces one transformation per second.

The La Crosse Boiling Water Reactor (LACBWR) is a type of nuclear reactor located in Wisconsin and operated from 1970 until 1987.


Question: Does the report include any NRC licensed materials?

Check box if samples include NRC licensed material?

The address for submitting the invoice is 201 C Sante Fe Avenue, Grants, NM, 87020.


Question: The critical heat flux correlation is essential for ensuring the safety and efficiency of nuclear reactors.

COBRA-FLX is a thermal-hydraulic code used to predict critical heat flux conditions in nuclear reactors.

The new safety protocols require extensive training for reactor operators to ensure compliance with regulatory requirements.


Question: What is the energy measurement associated with the nuclide CS-137?

The energy of CS-137 is 661.65 keV, and it has a yield of 85.12%.

The activities of nuclides are recorded in units of picocuries per gram.


Question: The system uses filters with specified efficiency ratings to maintain air quality in the Diesel Building.

Instrumentation for the Diesel Building Ventilation System is provided to control and indicate temperature.

The Diesel Building Ventilation System is composed of the following components:


Question: well drilling records near Savannah River Site

Attachment 2 contains the well drilling records for the area near the Savannah River Site, which are essential for understanding local geology and groundwater resources.

The summary section outlines key findings and conclusions drawn from the performance assessment assumptions.


Question: The methodology used to calculate the SCDF relies on the premise that a single seismic event can trigger failures in one or more safety systems.

The approach taken in the computation of SCDF assumes that the SCDF can be based on the likelihood that a single seismic-induced failure leads to core damage.

The limiting HCLPF approach assumes that a failure of a component with seismic capacity that HCLPF leads directly to core damage (CD).


Question: The impact of population density around a nuclear facility is a critical factor in regulatory assessments.

The population within a ten-mile radius of the facility is a crucial factor in evaluating the potential impact of an accident.

The Savannah River Plant Site Map illustrates the geographic layout and relevant boundaries for various operational areas.


Question: In order to control leakage along the reactor coolant pump shaft, three seals are arranged in series to ensure that leakage to Containment remains minimal.

Leakage along the reactor coolant pump shaft is controlled by three shaft seals arranged in series such that reactor coolant leakage to the Containment is essentially zero.

The anti-reverse rotation device consists of five pawls and a ratchet plate, ensuring that the motor does not rotate in the opposite direction.


Question: What are the training requirements for spent fuel storage inspectors?

Training Requirements and Qualification Journal for Independent Spent Fuel Storage Installation Inspector outlines the necessary qualifications and training mandated for inspectors in this specialized area of nuclear safety.

The Cybersecurity Inspector Technical Proficiency Training and Qualification Journal emphasizes the importance of cybersecurity measures in protecting sensitive information and ensuring the safety of nuclear installations.


Question: What is a Group I isolation in nuclear safety systems?

A high Main Steam Line radiation Group I isolation is triggered to automatically close valves in response to elevated radiation levels in the main steam line.


Question: What is the role of electrical insulation in nuclear plant safety?

Electrical Insulation: Insulate (Electrical)

High Voltage Electrical Insulators: Insulate (Electrical)


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code employs the ORFEO-HMP correlation to accurately calculate critical heat flux in boiling light water reactors under different operating conditions.

Table B-2 outlines various test assembly conditions, detailing temperature and pressure settings for each experiment conducted.


Question: What was the total number of discrete events recorded?

The total number of discrete events recorded was 204, with 67 occurrences in 1973, 67 in 1974, and 70 in 1976.

The data includes the number of inversion occurrences during July at the Grand Gulf site in 1973, 1974, and 1976, categorized by the duration of the occurrences.


Question: fuel failure causes

Fuel failures can be caused by overheating, PCI, hydrating, cladding collapse, bursting, mechanical fracturing, and fretting. When applicable, the fuel rod failure criteria should consider high effects based on irradiated material properties data.

Rod reactivity and insertability must be maintained. This requires that, at a minimum, the following may need to be reviewed: Changes in control rod configuration.


Question: The criteria for screening external hazards often require assessing event damage potential to determine their impact on the plant.

C1. Event damage potential is < events for which plant is designed.

Screening not successful. PRA needs to meet requirements in the ASME/ANS PRA Standard.


Question: units of measurement for isotopes

**NRC Device Key:** 819336  (Internal Control Number)
**Distributor/Distributed By:** NDC Technologies
**Distributor License Number:** 1933-19GL
**Manufacturer Name:** NDC INFRARED ENGINEERING INC
**Device Model (Not Source Model):** 103
**Device Serial Number:** 3108871
**Transfer Date:** 10/07/2011


Question: The quality control measures implemented throughout the analysis of radioactive materials ensure precision in measurements.

The QC (Quality Control) process for radioactive samples includes evaluating results using established acceptance criteria and reporting uncertainties.

The data from Batch 1921313 indicates that Iron-55 samples were analyzed using the liquid scintillation technique on 10/11/19.


Question: The NRC assesses compliance with regulatory standards during routine inspections...

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements.

On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Callaway Plant and discussed the results of this inspection with you and other members of your staff.


Question: What considerations should operators take regarding medication storage during emergencies?

Licensed operators are ultimately responsible for compliance with any conditions stated on their license. If possible, they should consider storing a one-week supply of necessary and critical medications at their work site, having such medications readily available at their homes to take to their work sites on short notice, and making contingency plans to have their medications brought to the site, if needed.

Although the NRC has no formal policy to address this specific situation, individual operators and facility licensees should consider the following regulatory requirements guidance in planning their response to emergency situations that may require operators to remain on-site for extended periods of time.


Question: The active inventory of the SG/DHRS loop reflects the critical parameters essential for safety assessments.

The active loop inventory is discussed in greater detail for Run 1 in Section 5.3.7.7.


Question: The average enrichment of uranium in PWR fuel assemblies can significantly impact their performance and safety measures.

The various fuel assembly designs contribute to different neutron absorption characteristics which must be evaluated for safety analysis.


Question: The void fraction for the hot subchannel at normal operating conditions can reach a maximum of 1.0.


Question: nuclear safety analysis

The safety analysis for the Millstone Power Station was conducted to ensure compliance with the regulatory requirements set by the Nuclear Regulatory Commission, focusing on potential accident scenarios and their mitigations.

The application for the construction of new facilities was submitted by the power plant, detailing the financial projections and the expected timeline for completion.


Question: Main Steam Safety Valves operational requirements

The table provides the relationship between the number of operable Main Steam Safety Valves (MSSVs) per steam generator and the corresponding maximum allowable power expressed as a percentage of the rated thermal power (RTP).

The reactor coolant pump is crucial for maintaining adequate flow and pressure in the reactor system, ensuring effective heat removal during normal operations and under accident conditions.


Question: flood protection features

Structures, systems, and components (i.e., features) relied upon for protection against external floods are designated as "flood protection features" and can be either passive or active. For Plant Farley Units 1 & 2, credited flood protection features are passive and include those credited for mitigating floods from surface and subsurface sources. Farley's credited flood protection features are independent of operating mode and, therefore, changes in operational mode do not affect the flood protection features' ability to protect safety-related SSCs. (Reference 12)

At the time of the CLB analysis, a new spillway was being added at Bartletts Ferry reservoir to increase spillway capacity. That new spillway capacity was not considered in the CLB. The FSAR states that the spillway construction was to be completed in 1985 (Reference 7).


Question: The construction of Fire Area 9-11 includes walls, floors, and ceilings designed to provide significant fire resistance, specifically with 3-hour rated barriers.

Boundary walls separating Fire Area 9-11 from other fire areas are 3-hour rated concrete block or structural steel stud and drywall construction.

Combustible materials and fire protection equipment are listed in Table 3.


Question: Base Case Derived Concentration Guideline Levels

LTP, Section 5.2 states that each radionuclide-specific Base Case DCGL is equivalent to the level of residual radioactivity (above background levels) that could, when considered independently, result in a Total Effective Dose Equivalent of 25 mrem/yr to an Average Member of the Critical Group.

The dose contribution from each ROC is accounted for using the SOF to ensure that the total dose from all ROC does not exceed the dose criterion.


Question: The system requires that nine vacuum breakers located between the suppression chamber and drywell be functional for operation.

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

The safety analysis report must include the site-specific seismic evaluations and the potential impacts on facility operations.


Question: probabilistic risk assessment (PRA) models

Enclosure 3 is not applicable because each probabilistic risk assessment (PRA) model used for the Risk Informed Completion Time (RICT) Program is addressed using a standard endorsed by the Nuclear Regulatory Commission.

The annual maintenance schedule includes inspections and tests to ensure all systems function as designed and meet safety standards.


Question: divisional cable separation

DIVISIONAL CABLE SEPARATION SHOULD MEET THE GUIDELINES OF REGULATORY GUIDE 1.75 "PHYSICAL INDEPENDENCE OF ELECTRICAL SYSTEMS".

SAFETY RELATED COMPUTERS SHOULD BE SEPARATED FROM OTHER AREAS OF THE PLANT BY BARRIERS HAVING A MINIMUM THREE-HOUR FIRE RESISTANT RATING.


Question: The licensing process for nuclear reactors involves rigorous evaluations of safety metrics and compliance with regulations.

The nuclear licensing process ensures that all regulatory requirements are met, addressing safety, environmental impact, and operational efficiency.

The temperatures recorded in the reactor core during normal operation provide crucial data for assessing thermal performance, yet do not encompass the entirety of safety evaluations.


Question: The proposed RSO replacement involves collaboration with a new radiologist group to ensure compliance with safety regulations.

Our current RSO will be retiring this year, and our initial thought was to have him train me to fill his role.

Please don’t hesitate to reach out with further questions.


Question: What is the significance of the containment vessel in nuclear reactors?

Figure 3-91. Containment Vessel, Second Horizontal Mode, f = 25.69 cps (Kalnin's)


Question: What is the role of a containment structure in a nuclear reactor?

The containment structure is designed to prevent the release of radioactive materials into the environment in the event of an accident.

The regulatory process for obtaining a license involves several steps, including the submission of safety analyses and public comment periods.


Question: The regulatory authority mandates that all nuclear power plant operators must conduct regular assessments of safety protocols.

The NRC's regulations require that licensees submit a detailed safety analysis report demonstrating compliance with safety standards.

Nuclear power plants generate electricity through a controlled nuclear reaction that heats water to produce steam.


Question: emergency condenser circulating water system

In the event of a power loss and loss of the circulating water pumps, the nuclear power plant’s emergency condenser circulating water system is automatically initiated and operates as an unassisted siphon system, supplying sufficient water to the condenser for decay heat removal and emergency cooling requirements.

Water flows down the length of the intake canal toward the intake structure.


Question: The report details the Minimum Detectable Activity (MDA) for various nuclides.

Nuclide MDA (pCi/g) values indicate the Minimum Detectable Activity for specified nuclides and are critical for evaluating the performance of the radiation detection system.

The uncertainties in the error quotes are calculated at 1.000 sigma, which helps ensure the reliability of the measurements.


Question: What is included in the PSL Monthly Turtle Report?

The PSL Monthly Turtle Report for January 2025 summarizes data collected on turtle populations in the vicinity of the facility, including nesting activity and environmental conditions.

The report includes various safety procedures implemented at the reactor, such as the response protocols for any potential hazards.


Question: The NRC mandates that all licensees ensure their reactors utilize compliant fuel designs.

The NRC determined that the proposed GAD Rod design meets safety and performance criteria, as outlined in the regulatory framework.

In the assessment of plant operations, unexpected maintenance intervals were identified, leading to adjustments in the overhaul schedule.


Question: Xe-100 safety analysis methodologies

The Xe-100 Licensing Topical Report describes the transient and safety analysis methodologies that ensure the reactor operates safely under varying conditions and scenarios.

X Energy, LLC is actively engaged in research and development to advance the state of reactor technology and improve efficiency in energy production.


Question: What is the function of the Condenser Circulating Water System in a nuclear facility?

The Condenser Circulating Water System is normally used to supply cooling water to the main condenser to remove residual heat from the Reactor Coolant System during the initial cooling period of unit shutdown when main steam is bypassed to the condenser.

Lake Norman water is monitored periodically for consistency of water chemistry and it is the normal common source for the Condenser Circulating Water (RC) and Nuclear Service Water (RN) systems.


Question: baseline inspections of BFBs

To accept the scheduling criteria for performing and completing baseline inspections of the BFBs, the NRC staff confirmed that MRP-227, Revision 2, sets the following updated schedule for performing and completing the baseline UT inspections of the BFBs: 1. For highest ranked Tier 1 BFBs, all baseline inspections of the bolts have been completed.

However, EPRI clarified that the 80 percent parameter placed on the 3-by-10 and 4-by-6 BFB arrays and used as the acceptance criterion for initiating sample-expansion to CBF bolts is independent of the size of the baffle plates or number of BFBs in the plates.


Question: The evaluation focused on Wisconsin's ability to carry out its radiation control program under the Agreement between the NRC and the State of Wisconsin.

The review focused on the radiation control program as it is carried out under Section 274b. (of the Atomic Energy Act of 1954, as amended) Agreement between the NRC and the State of Wisconsin.

The team issued a draft report to Wisconsin on November 5, 2024, for factual comment in ML24295A154.


Question: What are historic and cultural resources at Oconee Station?

Historic and cultural resources within the Oconee Station site can include prehistoric era and historic archaeological sites, historic districts, and buildings; as well as any site, structure, or object that may be considered eligible for listing on the National Register of Historic Places (NRHP).

The current Historic Period in South Carolina began with colonization by the British in 1670 and the establishment of trading posts.


Question: Main Steam Isolation Valve test pressure requirements

The proposed change will revise the Main Steam Isolation Valve (MSIV) test pressure value from ≥40 psig to ≥25 psig in TS Surveillance Requirement (SR) 3.6.1.3.12.

Experience shows that this testing method has been successful, yet this test method creates undue hardships during outages by requiring installation of plugs at the inside steam line nozzle penetrations and expose workers to radiological dose.


Question: What actions are authorized by the Police Jury President during an emergency?

I hereby authorize the Parish EOC Director (or authorized representative) to take the following actions (Check the appropriate box and fill in the blanks):


Question: What is the purpose of the NRC Examination Security Agreement?

I acknowledge that I have acquired specialized knowledge about the NRC licensing examinations scheduled for the week(s) of July 23, 2018. I agree that I will not knowingly divulge any information about these examinations to any persons who have not been authorized by the NRC chief examiner.

Furthermore, I am aware of the physical security measures and requirements (as documented in the facility licensee's procedures) and understand that violation of the conditions of this agreement may result in cancellation of the examinations and/or an enforcement action against me or the facility licensee.


Question: Connecticut should clarify its position regarding the definition of materials produced during the processing of uranium or thorium from ores.

The definition does not include "The tailings or wastes produced by the extraction or concentration of uranium or thorium from any ore processed primarily for its source material content."

Connecticut also needs to delete the later part of the sentence "or, if such regulations should be deemed inappropriate by the Commissioner of Energy and Environmental Protection, on the latest recommendations of the National Committee on Radiation, as published by the United States Department of Commerce, National Bureau of Standards."


Question: The NRC mandates that operators assess and ensure the function of safety-related systems, including back-up power supplies.

In the context of regulatory processes, the NRC emphasizes the importance of maintaining operability of emergency systems, such as the diesel generators used for backup power.

The regulatory framework also includes guidelines on routine inspections and maintenance of equipment, which are crucial for overall safety management in nuclear facilities.


Question: seismic fragility in nuclear plants

The plant level seismic fragility is a curve of the conditional probability of plant damage as a function of seismic hazard input level.

The frequency of the last seismic hazard interval is the exceedance frequency at the beginning point of that interval.


Question: During the internal audit, the NRC reviews applicant documents related to the SLRA for safety evaluation.

Throughout the Aging Management Audit, the staff will review the SLRA, as supplemented and revised, the AMP basis documents and other applicant supporting documents as appropriate and requested, which have been uploaded to the ePortal.

The technical staff will inform the technical branch chief of the need for an In-Office Breakout Session.


Question: NRC adjudicatory E-Filing system

A filing is considered complete at the time the document is submitted through the NRC’s E-Filing system.

Information about applying for a digital ID certificate is available on the NRC’s public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html.


Question: Could you provide the calculated values for pellet thermal relocation as influenced by different burnup amounts?

Sample calculations of pellet thermal relocation displacement (micrometers, µm) as a function of burnup (gigawatt-days per metric ton of uranium, GWd/mtU) generated by GALILEO for power levels of 15, 30, and 45 kilowatt per meter (kW/m) are shown graphically in Figures 3-1, 3-2, and 3-3, respectively. The total pellet relocation as a function of burnup is also shown in these figures.

The assessment of safety systems at nuclear facilities is crucial to ensuring regulatory compliance and protecting public health.


Question: heatup rate correlations

At the elevations where cladding oxidation was significant ([4, 6, and 8 feet]), both the Cathcart-Pawel and the Baker-Just correlations resulted in an over-prediction of the measured heatup rate.

The NRC staff response misrepresents a statement made in the petition.


Question: sump water level calculation after LOCA

The containment sump pool water volume following a LOCA was determined by considering all water sources (i.e., the refueling water storage tank, reactor coolant system (RCS), the safety injection accumulators, and the spray additive tank) and subtracting various holdup volumes.

The heat exchangers, orifices, and spray nozzles were evaluated for the effects of erosive wear for a constant debris concentration over the mission time of 30 days.


Question: The guidelines for evaluating fuel cladding failure include addressing peak radial average enthalpy thresholds and ensuring the cladding pressure differential remains within prescribed limits.

The MPS3-specific AREA analysis for GAIA fuel was performed against the acceptance criteria defined in USNRC Regulatory Guide (RG) 1.236.

The proposed DBLFPBs are outlined in Table 1.


Question: analytical method for detecting Fe55

**Analytical Method:** DOE RESL Fe-1, Modified


Question: The applicant must ensure that safety assessments reflect current environmental conditions around the facility.

The Environmental Report submitted by the applicant must include a comprehensive assessment of the potential environmental impacts associated with the operation of the Dresden Nuclear Power Station.

Historical records indicate that the topography of Illinois has significantly changed since 1855, impacting local land use and development.


Question: The passive autocatalytic recombiner is deemed essential for maintaining safety in design-basis scenarios.

NuScale determined the PAR mitigates design-basis events, making the component safety-related and appropriate for inclusion in the GTS.

The updated guidelines indicate a shift in the methodology for assessing structural integrity during seismic events.


Question: What is the purpose of the Residual Heat Removal System in nuclear reactor safety?

The purpose of the RHR System is to remove decay heat with long-term recirculation capability in post-accident conditions and to provide safety injection.

Each train is provided with a discharge path to both the hot and cold legs.


Question: recirculation spray pump safety features

The outside recirculation spray pump discharge lines are provided with a check valve inside the containment and a normally open, remotely controlled, motor-operated valve outside the containment.

The isolation valve for each LHSI pump suction penetration is a normally closed, remotely controlled, motor-operated valve.


Question: leak-before break analysis

xLPR and leak-before break (LBB) analyses (CIB) - Current Status on MOU Development


Question: The requirements for datasets are specified in ASTM E1921-20, particularly in paragraph 10.3, which stems from the consensus of the ASTM E08 committee regarding necessary dataset sizes.

The PWROG stated that the T0 test data must meet the requirements of ASTM E1921-20. If T0 data was produced in accordance with another version of ASTM E1921 or another test standard (e.g., ASTM E399), the user of the TR must review the data and revise the analysis accordingly.

The PWROG clarified that the second term under the square root sign of Equation 5 of the TR markup is (2 * Equation 12). The NRC staff finds this clarification acceptable because Equation 12 of the TR markup ensures a minimum of 9 degrees Celsius (°C) margin to account for adjustment uncertainty.


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

The COBRA-FLX code employs the ORFEO-HMP critical heat flux correlation to accurately predict the conditions under which critical heat flux occurs in nuclear reactors.

The table displays local conditions from COBRA-FLX simulations specifically for grid tests, offering detailed insights into the performance metrics under various operational parameters.


Question: What happens to the breaker indicating lights when the 6.9kV SD BD 1A-A is inoperable?

The breaker indicating lights for equipment feeding from 6.9kV SD BD 1A-A and 480V SD BDs 1A1-A/1A2-A will be dark. 6.9kV SD BD 1A-A loads are not able to be controlled from the MCR due to the ACBs being electrically inoperable.

Plausible because the loss of any other 125V DC Battery Board will not result in the breaker indicating lights for equipment feeding from 6.9kV SD BD 1A-A and 480V SD BDs 1A1-A/1A2-A being dark.


Question: What is the purpose of regulatory filings in nuclear licensing?

Regulatory filings serve as a means for licensees to demonstrate compliance with safety and environmental standards as mandated by the NRC.

The updated tables offer a detailed representation of the changes made to the safety protocols in different phases.


Question: what is Emergency Core Cooling System (ECCS) shutdown?

In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head).

The parameters assumed in the generic bounding thermal hydraulic analysis for the limiting DBA are based on a combination of limiting parameters for MODE 3, with the accumulators blocked, and parameters for MODE 4.


Question: What is the impact of fuel composition on the reactivity of a damaged package array?

The reactivity of the damaged package array depends on the effectiveness of the fuel in competing with other materials, such as the cadmium, hydrogen, stainless steel or water reflector for absorption of thermal neutrons.

All damaged package array models remain below the accident limit keff, USL = 0.931 for up to 60 kgs UO2 per ICCA.


Question: The NRC outlines the requirements for evaluating environmental impacts in the licensing process for nuclear power plants.

The Generic Environmental Impact Statement for License Renewal of Nuclear Plants assesses the potential environmental effects associated with the renewal of operating licenses for nuclear reactors.

Geothermal energy has been identified as a renewable energy source that can significantly reduce dependence on fossil fuels for power generation.


Question: what is ISR licensing process

Licensing history and extensive litigation have provided the framework for the current ISR licensing process.

Baseline monitoring and regulatory monitoring are essential components in ensuring the safety of nuclear facilities and are conducted regularly.


Question: The proposed alternative seeks to implement different treatment guidelines for repair and replacement tasks in specified systems.

The licensing action proposes to use the alternative requirements of ASME Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section XI, Division 1," for determining the risk-informed categorization and implementing alternative treatment for repair/replacement activities on moderate- and high-energy Class 2 and 3 items instead of ASME Code Section XI, paragraphs IWA-1000, IWA-4000, and IWA-6000 requirements.

The NRC staff and Duke Energy agreed to a regulatory audit to gain a better understanding of the issues noted in the Enclosure and to identify information that may require docketing to support the basis of the NRC staff's licensing decision.


Question: The assessment indicates that watershed changes, particularly due to urbanization, have a minimal effect on runoff over the duration of the project.

Based on this assessment the Tennessee Valley watershed is experiencing a change in impervious area at a rate of about 0.02% per year in a 10 year period.

Additional lands owned by federal agencies, including US Fish and Wildlife Service, US Forestry Service, National Park Service, and the Protected Areas Database, comprise one third of the entire watershed above Guntersville Dam.


Question: The regulation of nuclear facilities necessitates that licensees perform comprehensive evaluations of potential safety hazards.

The NRC oversees the licensing process which requires detailed safety analyses and regulatory compliance for nuclear operations.

Duke Energy Florida, LLC is committed to enhancing the environmental and energy landscape of Florida.


Question: automatic Halon systems

Automatic Halon systems have been provided for the upper cable spreading areas and the QA vault. Both systems are actuated by ionization detectors. The upper cable spreading rooms have a train both of ionization and thermal detectors.


Question: The NRC guidance emphasizes the importance of peer review for changes made to the PRA model, highlighting that it must adhere to specific criteria.

The ASME/ANS RA-Sa-2009 PRA Standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard states that upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this Standard.

If any operator actions related to FLEX equipment are evaluated using approaches that are not consistent with the endorsed ASME/ANS RA-Sa-2009 PRA Standard (e.g., using surrogates), discuss the methodology used to assess human error probabilities for these operator actions.


Question: Best Management Practices in nuclear licensing

The NRC took into account that implementing BMPs required by the National Pollutant Discharge Elimination System (NPDES) permit and construction stormwater pollution prevention plans (SWPPPs) would reduce soil erosion and subsequent impacts on surface water quality.

Ground-disturbing activities and stormwater runoff could result from routine infrastructure, renovation, and maintenance that would be expected during continued operation at PTN.


Question: DTE Electric Company is seeking permission from the NRC to follow BWRVIP Guidelines as a substitute for ASME Code Section XI inspection standards.

The BWRVIP Inspection and Evaluation (I&E) Guidelines have recommended aggressive specific inspections by BWR operators to completely identify material condition issues with BWR components.

These examinations are performed to assess the structural integrity of the reactor vessel interior, its welded attachments and the welded core support structure within the boiling water reactor pressure vessel.


Question: control rod sequence during reactor startup

With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence.

With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence.


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX code employs the ORFEO-HMP critical heat flux correlation, which provides predictive methodologies for assessing thermal-hydraulic performance during operational conditions in nuclear reactors.

The Test assembly conditions for non-mixing grid tests are essential for evaluating the performance of experimental assemblies under various thermal-hydraulic scenarios.


Question: What is shear load in valve testing?

Shear Load refers to the load from the disc when the valve is closed, which is crucial for evaluating the valve's performance and integrity under operational conditions.

Withstand beyond highest in-service loadings applied.


Question: loss of material in concrete structures

Aging Effect Requiring Management: Loss of Material (Spalling, Scaling) and Cracking

Concrete: Below-grade exterior (inaccessible) | Flood Barrier Shelter/Protection Structural Support | Reinforced concrete | Groundwater/Soil | Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling) | Structures Monitoring (B.2.1.33) | III.A3.TP-212 | 3.5.1-065 | A


Question: radial power profiles in nuclear reactors

Figure 5: Comparison of GALILEO and FRAPCON calculated radial power profiles at an enrichment of 3%, Gd2O3 concentration of 8wt% and 60 GWd/MTU burnup

The assessment of seismic performance is essential for ensuring structural integrity during an earthquake.


Question: The design of the BWRX-300 will not allow for sharing of safety-related structures, systems, or components across multiple units.

The capability of the CIVs to perform their design-basis functions is safety significant because it provides assurance that the containment of the BWRX-300 can be safely isolated and prevent radioactive release to the environment that exceeds regulatory requirements.

The NRC staff finds that the BWRX-300 containment design is consistent with the requirements of GDC 5 and is, therefore, acceptable.


Question: The treatment of human errors, particularly post-initiator human errors, is crucial for modeling risk in nuclear safety assessments.

However, HFEs are typically significant contributors to CDF and large early release frequency (LERF).

In general, for procedure-directed actions, the cognitive part of the post-initiator HEP is quantified using the Cause Based Decision Tree Methodology (CBDTM).


Question: nuclear criticality safety

The inspectors interviewed the Emergency Preparedness Manager, Nuclear Criticality Safety Engineering Manager, and a NCS engineer to determine whether the licensee maintained NCS related emergency response capability consistent with the emergency plan and procedures.

The inspectors selected a sample of plant changes based on interviews with licensee staff and the list of plant changes submitted to the NRC on January 6, 2020 (ADAMS Accession Number ML20006F639).


Question: The NRC mandates that licensees perform various tests to assess material properties and ensure structural integrity during their operations.

The following Plates are attached and complete this Appendix:| Plate    | Description                                      | |----------|--------------------------------------------------| | A-1a    | Triaxial Compression Tests on Remolded Samples (Bulk Sample 1) | | A-1b    | Triaxial Compression Tests on Re-molded Samples (Bulk Sample 2) | | A-2     | Triaxial Compression Test Data                   | | A-3     | Rock Compression Test Data                        | | A-4a    | Confined Compression Test Data                    | | A-4b    | Consolidation Test Data                           |

The project involves documentation and detailed analysis of the environmental impact, ensuring that all regulations are strictly followed.


Question: proprietary information in nuclear licensing

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

These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.


Question: The NRC mandates that licensees implement measures to mitigate exposure in high radiation areas.

Controls for high and very high radiation areas are critical to ensuring safety around radiation sources and must be strictly adhered to.

The examination answer key includes several references to administrative procedures and cognitive levels relevant to NRC exams.


Question: maximum internal pressure in nuclear containment systems

The determination of the maximum internal pressure within the ICCAs is based on the ideal gas law. If an ICCA is filled at the minimum normal operating temperature, -40 °F and is allowed to reach the NCT maximum bulk temperature for the payload, 133 °F, the maximum pressure would be the product of the ratio of the absolute gas temperature to -40 °F and atmospheric pressure, 14.7 psia.

The NPC package is designed to ensure safe transport and containment of nuclear materials under various conditions.


Question: evaluation of potential accidents


Question: what is nuclide identification?

Nuclide confidence index threshold = 0.30


Question: What is a limiting condition in safety evaluations?

A limiting condition is defined as the lowest functional capability or performance level of a system or component required to maintain safety.

The primary objective of the project is to ensure that all the necessary safety measures are implemented effectively.


Question: The NRC requires disclosure of devices that are subject to registration, along with their pertinent details.

Our records indicate that you have these devices. Please update the information as necessary.

The Device Model (Not Source Model) must be clearly identified as either new or used to ensure accurate registration.


Question: What are the discharge limitations for NPDES permits?

THE DISCHARGE SHALL HAVE NO SHEEN, AND VISIBLE OIL, FLOATING SOLIDS OR VISIBLE FLOATABLE MATERIALS.


Question: What are the key components of a nuclear safety analysis?

A safety analysis typically includes an examination of the systems, structures, and components that are critical to the safe operation of a nuclear facility, assessing potential hazards and their implications.

Licensing for a new reactor type requires detailed documentation to demonstrate compliance with both safety and environmental standards.


Question: The ventilation system in place is equipped with special features to ensure the air within critical areas remains uncontaminated and safe for operations.

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

The Equipment and Cable Room make-up fans supply 1000 CFM each. The circulation fans supply 55,000 CFM each.


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

The term 'Energy (keV)' refers to the measured energy peak of a radioactive emission, which is crucial for identifying and quantifying isotopes in safety assessments.

Background reports are essential for understanding the environmental conditions surrounding nuclear facilities.


Question: What are the implications of seismic reflection studies for nuclear facility safety assessments?

Seismic reflection studies provide critical insights into subsurface geological structures, which are essential for accurately assessing site safety for nuclear facilities.

The geological studies of North America highlight various landforms, which can vary significantly in composition and structure across different regions.


Question: The NRC mandates that assessments are conducted to ensure the integrity of aluminum alloy components against material loss due to environmental exposure.

This item evaluates loss of material due to pitting and crevice corrosion in aluminum alloy piping, piping components, and tanks exposed to air and condensation environments. There are no aluminum alloy piping, piping components, or tanks exposed to an environment of air-indoor controlled in Auxiliary Systems at DNPS. Plant-specific OE from 2013 to 2023 associated with aluminum alloy components in the Auxiliary Systems has been evaluated to determine if prolonged exposure to air-indoor uncontrolled, air-outdoor, and condensation has resulted in loss of material due to pitting and crevice corrosion. Loss of material has not been identified through this review as an aging effect that has been experienced at DNPS for aluminum alloy in these environments. This provides objective evidence that the external surfaces of these components are not experiencing significant halide (e.g., chloride) exposure that would result in loss of material. Accordingly, DNPS will implement the One-Time Inspection (B.2.1.20) program to demonstrate that the aging effect of loss of material is not occurring in aluminum alloy piping, piping components, and heat exchanger components exposed to air-indoor uncontrolled, air-outdoor, and condensation. Deficiencies will be documented in accordance with 10 CFR Part 50, Appendix B Corrective Action Program. The One-Time Inspection (B.2.1.20) program is described in Appendix B.

This item evaluates loss of material due to pitting, crevice corrosion in aluminum alloy tanks ('Outdoor and Large Atmospheric Metallic Storage Tanks') exposed to air or condensation environments. There are no aluminum alloy tanks exposed to air-indoor controlled or air-indoor uncontrolled environments in the Auxiliary Systems at DNPS. Plant-specific OE from 2013 to 2023 associated with aluminum tanks in the Auxiliary Systems has been evaluated to determine if prolonged exposure to air-outdoor and condensation environments has resulted in loss of material due to pitting or crevice corrosion.


Question: nuclide minimum detectable activity (MDA)

The Nuclide MDA Report provides critical data on the minimum detectable activity (MDA) of various nuclides, including their energy levels, yields, and activities measured in picocuries per gram (pCi/grams).

The Peak Locate performed indicates the specific channels and energy levels at which unidentified peaks were detected, essential for understanding the overall spectrum of analyzed data.


Question: What is the purpose of the Marinelli beaker in radiation measurement?


Question: The DCPP's operational capacity relies on the effective performance of its four-loop PWR, which is critical for safe electricity generation.

Both DCPP Units employ a four-loop pressurized water reactor (PWR) Nuclear Steam Supply System (NSSS) furnished by Westinghouse Electric Corporation.

Any notices, questions, or correspondence in connection with this filing should be directed to:


Question: What is tornado wind fragility analysis in nuclear safety?

The intent of the follow-up review is to ensure that the final fragility evaluation is substantively consistent with the 29 April 2021 draft previously reviewed by SGH, and to review how the fragility is used in the LBDCR to ensure the application is consistent with the underlying inputs, assumptions, and methods used in the fragility evaluation.

Our initial review of the 24 February 2022 draft focused on the following attachments of LBDCR 22-027:


Question: Plutonium-241 regulatory compliance

Rad Alpha Spec | NOM | Sample | Qual | QC | Units | RPD% | REC% | Range | Anslt | Date | Time


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves several steps, including site evaluation, safety analysis, and regulatory reviews by the NRC.

Environmental assessments are conducted to determine the potential impact of projects on wildlife and natural resources.


Question: average burnup in nuclear reactors

"Assembly Average Burnup ≤ 59 GWd/MTU" defines the average burnup of nuclear fuel assemblies, which is critical for assessing their performance and safety in a reactor.

The safety evaluation involves assessing the potential for accidents and the corresponding emergency response strategies to protect public health.


Question: The NRC is evaluating the feasibility of conducting inspections remotely instead of in-person visits.

Use of the term “remote inspection” causes concern for some.

Report relies too heavily on survey results.


Question: remote shutdown systems

Remote Shutdown of Ventilation for Loss of Auxiliary Air 078A4.030 (3.2)

Alignment Upper Containment Monitor to Lower Containment Locally (Unit 2) 073A4.02 (3.6)


Question: safety-related system evaluation

The safety-related system evaluation is critical in ensuring that all systems operate within specified safety margins and comply with regulatory standards.

Regulatory filings must include accurate data and comprehensive documentation to support the overall safety assessment of the proposed modifications.


Question: The NRC evaluates submissions from nuclear facilities to ensure compliance with safety standards.

The NRC requires that all nuclear reactor licensees submit detailed operational plans and safety analyses as part of their regulatory compliance process.

Aerotest Operations Inc. specializes in the operation of research reactors and provides a range of testing services for various scientific applications.


Question: regulatory process for nuclear licensing

The regulatory process for nuclear licensing includes a detailed review of safety analyses and technical specifications to ensure compliance with established safety standards.

The public meeting addressed various topics related to industry developments and operational challenges, but did not specifically cover licensing issues.


Question: What are the safety systems in place at Surry Power Station?

The safety systems at Surry Power Station are designed to prevent and mitigate the consequences of potential accidents, thereby ensuring the safety of the plant and surrounding area.

Virginia Electric and Power Company operates multiple power generation facilities across Virginia, including both fossil fuel and renewable energy sources.


Question: What are the different radiation zones in a nuclear facility?

The radiation zones are classified into six distinct areas: Zone I, Zone II, Zone III, Zone IV, Zone V, and Zone VI, each representing varying levels of radiation exposure.

The safety analysis must include a review of the operational parameters and maintenance schedules for the reactor.


Question: What are common-cause failures in nuclear safety analysis?

The CCF contributions should be modeled so that they can be modified to reflect the condition in which one or more of the components is unavailable.

If the PRA does not model the system for which the TS change is being requested, specialized analyses may be necessary to demonstrate the sufficiency of the proposal.


Question: What are the potential impacts of dam breaches on nuclear safety?

The Flood Hazard Reevaluation Report includes an analysis of dam breaches and failures, which outlines their potential impact on the safety and operation of the Cooper Nuclear Station.

Figure 2.3-4 illustrates the HEC-HMS Plan for Missouri River from Omaha, NE to CNS, providing crucial data for flood management and response strategies.


Question: What is a primary safety system in nuclear reactors?

The primary safety system in nuclear reactors is designed to prevent the overheating of the reactor core and to maintain core cooling under all operational and accident conditions.

The user interface allows for easy navigation between different document sections, enhancing the overall user experience and accessibility.


Question: COBRA-FLX heat flux correlation

COBRA-FLX is an advanced thermal-hydraulic analysis tool utilized for predicting the critical heat flux (CHF) in nuclear reactor systems, which is crucial for assessing safety margins in reactor operations.

The table lists the local conditions from COBRA-FLX simulations, showing various parameters such as pressure, temperature, and flow rate for different grid test scenarios.


Question: nuclear licensing process

The nuclear licensing process is a systematic approach to ensure that nuclear facilities meet safety, environmental, and regulatory requirements before they are allowed to operate.

The reactor containment structure is designed to prevent the release of radioactive materials during an accident.


Question: post-accident sampling drills

Post-accident sampling drills are conducted annually. These drills address capabilities including analysis of liquid and containment atmosphere samples with simulated elevated radiation levels.

Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected.


Question: The NRC collaborates with national laboratories and industry to ensure advanced reactor technologies can be effectively demonstrated.

Support advanced reactor regulatory readiness for demonstrations

Performed gap assessment for demonstration capabilities


Question: According to the NRC, the tornado missile hazard can be evaluated using established criteria that contribute to CDF and LERF values.

Average maintenance configuration estimates for CDF for both units are less than 1E-6/year and LERF was estimated to be much less than 1E-7/year. Therefore, the tornado missile hazard can be screened using Criterion PS4 as defined in Table E4-12.

On March 12, 2012, the NRC issued a request for information associated with NTTF Recommendation 2.1 for Flooding. One of the required responses in this letter directed licensees to submit a Flood Hazard Reevaluation Report (FHRR), including the interim action plan (to address calculated higher flooding hazards relative to the design basis prior to...


Question: security personnel training regulations

The Order requires compliance with the specified security personnel training compensatory measures (CMs).

The Commission requires that it be protected and that access to Safeguards Information be limited in accordance with Title 10 of the Code of Federal Regulations Section 73.21.


Question: What is Environmental Fatigue in relation to reactor components?

NUREG-2191 Section X.M1, Fatigue Monitoring, provides guidance for evaluating the effects of the reactor water environment on the fatigue life of ASME Section III Class 1 components that contact the reactor coolant.

For the SLR application, environmental fatigue screening calculations were performed for all RPV, piping, and other component locations that have an identified CUF value in a BFN CLB stress report or evaluation and are in contact with reactor water.


Question: model uncertainty analysis

To meet the guidance in NUREG-1855 (Reference 12), model uncertainty was addressed for any models, for which no consensus exists. Most of the models used in the PBN risk quantification are consensus models that have been widely used by the industry and accepted by the NRC. However, the following models are not consensus models and were therefore included in the model uncertainty analysis:

The probability of all of the worst-case conditions occurring simultaneously is extremely low. The results of this evaluation show that the parametric uncertainty is low.


Question: The NRC indicated that the actual cladding temperatures in the experiment closely aligned with the predicted values post-multiplier application, suggesting the importance of accurate reaction rate predictions.

In-Vessel Phenomena—CORA reports that in CORA-16, the estimated cladding strain was in the "range of 0.005 to 0.11," at 4200 seconds into the experiment, at locations of the cladding where temperatures were between 1832°F and 2372°F.

Hence, the multiplier also helped the predicted cladding temperatures match the cladding temperatures that occurred in the experiment.


Question: The Supervisor has instructed the operator to execute the procedure for transitioning to Hot Leg Recirculation following the specified steps.

The Unit Supervisor has directed you to proceed with 1BEP ES-1.4, Transfer to Hot Leg Recirculation and to notify him when complete.

UNSAT requires written comments on respective step. * Denotes critical steps: 2, 5, 6 & 10 - 15


Question: What is a Radiation Safety Officer and their responsibilities?

The Radiation Safety Officer (RSO) for this license is Bradley M. Gemlick.

The licensee shall conduct a physical inventory every six months, or at other intervals approved by the U.S. Nuclear Regulatory Commission, to account for all sealed sources and/or devices received and possessed under the license.


Question: functional testing of emergency systems

Functional and operational testing of C High Head FLEX Pump (0FX02PC) following modification of hydraulic cooler, as documented in Work Order (WO) 5394984.

The inspectors evaluated the following maintenance and testing of the alert and notification system: - Annual siren inspection and maintenance records for the period of October 2022 – October 2024.


Question: What is measured in the Rad Alpha Spec Analysis?

The Rad Alpha Spec Analysis includes measurements of various alpha-emitting radionuclides such as Americium-241 and Plutonium-239/240, with results reported in units of pCi/g.

The Rad Gamma Spec Analysis provides data on gamma-emitting isotopes like Cesium-137 and Europium-153, again expressed in pCi/g.


Question: redundancy in load-bearing structures

One characteristic of this frame that is important for further discussion, is the redundancy of this frame.

If certain parts of the frame were for any reason incapable of sustaining all of the load integrity inferred to them what would happen? How would the load be redistributed into the frame.


Question: Reactor Coolant Pump functionality

Reactor Coolant Pump is responsible for circulating coolant through the reactor core to remove heat generated during nuclear fission and maintain safe operating conditions.

Power supply is crucial for the operation of various safety systems in a nuclear plant to ensure uninterrupted monitoring and control.


Question: The response time verification must be performed for functions subject to time limits specified in regulatory guidance.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.


Question: What are the proposed changes for surveillance requirements in reactor licensing?

In attachment 2 to the LAR, ITS table 3.3.1.1-1, the licensee proposed to add calibration surveillance requirement footnotes (similar to TSTF-493, Revision 4, Option A) to Function 3 (Reactor Vessel Steam Dome Pressure – High), Function 4 (Reactor Vessel Steam Dome Pressure – Low), Function 6 (Drywell Pressure – High), and Function 9 (Turbine Control Valve Fast Closure, Trip Oil Pressure – Low).

The footnotes are not applied to these functions in the Hope Creek current TS.


Question: The system involved in chemical treatment to prevent bio-fouling is crucial for maintaining the functionality of associated systems during operations.

The Hypochlorite System, also called the Raw Water Chemical Treatment System, prevents bio-fouling of systems (including the EECW and RHRSW Systems) that use water from Wheeler Reservoir by providing the capability of injecting a biocide into the fluid stream.

The aging management review results for these components are provided in Table 3.3.2-16, Building Heating System - Summary of Aging Management Evaluation.


Question: The analysis of DBA involves examining the behavior of chemical constituents such as sodium, air, and argon gas.

The chemical forms of substance included in the DBA analyses are sodium, air, and argon gas.

The programming language used for SAS4A/SASSYS-1 is ANSI FORTRAN, and the organization of the code follows the FORTRAN convention of the MAIN program with a number of subroutines and functions.


Question: The inspection team requires access to certain key documents like the Original Final Safety Analysis Report Volumes.

Please provide five copies of each CD submitted (i.e., one for each inspector/contractor).

If you have any questions regarding this information, please call the inspection team leader as soon as possible.


Question: What are the specific foldout page requirements during a transition brief in emergency operating procedures?

Specific foldout page parameters and values are required to be repeated back.

The crew is performing a transition brief prior to entry into an Emergency Operating procedure.


Question: What is the annual notification requirement under 10 CFR 50.46?

The annual notification requirement under 10 CFR 50.46 mandates that licensees report on the performance of the emergency core cooling systems and their compliance with established thermal performance criteria.

The safety inspections at nuclear plants are crucial for ensuring operational integrity and protecting public health.


Question: What are the limitations and conditions for TS related to the PRA?

The LAR will provide identification of the TS Limiting Conditions for Operation (LCOs) and action requirements to which the RMTS will apply.

The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent with licensing basis assumptions.


Question: What are MCPR limits for nuclear fuels?

Flow Dependent MCPR Limits MCPRF for GNF2 Fuel demonstrate the limits for different flow rates, indicating the Minimum Critical Power Ratio that must be achieved for safe reactor operation.

Table 4-6 provides an overview of the Cycle Specific SLMCPR (MCPR99.9%) limits for all fuel types, focusing on operational loops rather than fuel specific constraints.


Question: The minimum concentration of boron in the SITs is crucial to prevent dilution from the RCS during a large break LOCA.

The 2300 ppm minimum boron concentration in the SITs assures that the back leakage from the RCS will not dilute the SITs below the minimum boron concentration in the safety analysis.

The LCO establishes the minimum conditions required to ensure that the SITs are available to accomplish their cooling safety function following a LOCA.


Question: What thresholds must be established for ensuring safety in the final rulemaking?

The purpose of these success criteria is to define what functionality needs to be provided in the final Part 53 rule, and will be a guide for the Part 53 development, and future discussions to be held between the NRC staff and industry to ensure that the final rule meets the pre-defined success criteria.

During the NRC's public meeting on November 18, 2020, the NRC discussed draft proposed rule text for Part 53 Subpart B 'Technology-Inclusive Safety Requirements'. We offer the following feedback on the NRC's.


Question: The NRC sets requirements for systems that form the pressure boundary to protect against losses.

Pressure Boundary                             Copper Alloy with less than 15% Zinc | Gas (Internal)                  | None                              | None                              | VII.J.AP-9                                                    | 3.3.1-114         | C

The Heat Exchanger components must undergo regular inspections to assess condition and functionality.


Question: CFD model features in nuclear safety analysis

Significant Features in CFD Model (PBN1) include Loop B Break Location Modeled Mass Source, Strainer Module Mass Sinks, Refueling Canal Drainage, Washdown Spray Sources (all areas in dark blue), Direct Spray Sources (all areas in light blue), and Loop A Break Location Modeled Mass Source.

The strainer module is designed to perform filtration and is an essential component of the reactor's cooling system, ensuring that only clean coolant circulates within the system.


Question: What are valve operating requirements in nuclear safety?

For example, appropriate parameters for calculating valve operating requirements and actuator capability (such as valve friction coefficients, maximum differential pressure conditions, motor torque temperature derating factors, stem friction coefficients, and butterfly valve bearing friction coefficients) need to be adequately addressed.

Some licensees were not properly determining the operating requirements for POVs to perform their safety functions.


Question: What does 10 CFR Part 51, Subpart A, Appendix B, Table B-1 define as 'small' environmental impacts?

SMALL: Environmental effects are not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource.

Decommissioning would at a minimum occur after the expiration of the current license term.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code incorporates a new critical heat flux correlation developed from the ORFEO-HMP thermal-hydraulics methodology, which is designed to improve the modeling accuracy for various reactor conditions.

Table B-3 provides a detailed overview of local conditions derived from COBRA-FLX simulations applicable to grid tests, focusing on pressure, temperature, and flow rates.


Question: background subtraction in radiation measurements

Ambient background refers to the natural radiation present in the environment that can interfere with accurate measurements during spectroscopic analysis.

The FWHM (Full Width at Half Maximum) is a common parameter used to characterize the resolution of peaks in spectroscopic recordings.


Question: average enrichment of U-235 in BWR fuel

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


Question: What should the NRC consider when developing regulations for fusion energy systems?

The NRC’s 10 C.F.R. Part 50, 52, in the federal regulations for large commercial fission systems and the new regulatory approach discussed for advanced fusion systems (i.e. Part 53) address a different suite of risks compared to risks that fusion facilities could create and therefore are not appropriate for fusion systems.

The DOE has created a framework for safe construction and operation of experimental fusion energy devices that has worked well for decades.


Question: What is COBRA-FLX and how does it relate to nuclear safety?

COBRA-FLX is a thermal-hydraulic analysis code developed to predict the behavior of reactor core cooling during various operational and accident scenarios, particularly focusing on critical heat flux conditions.

Framatome Inc. specializes in providing various services and technologies to enhance the performance and safety of nuclear reactors.


Question: The NRC mandates that aging management programs assess the impacts of neutron irradiation on reactor components.

The Neutron Fluence Monitoring (B.3.1.2) program and Reactor Vessel Material Surveillance (B.2.1.19) program will be used to manage loss of fracture toughness of the carbon or low alloy steel with stainless steel cladding reactor vessel beltline shell, nozzle, and weld components exposed to reactor coolant and neutron flux.

The One-Time Inspection (B.2.1.20) program will be used to manage cracking of the stainless steel reactor vessel top head enclosure flange leakage detection line, reactor vessel penetrations, nozzle safe ends and welds exposed to air - indoor uncontrolled in the Reactor Coolant Pressure Boundary System and Reactor Vessel.


Question: The regulatory authority mandates periodic updates to the submitted documentation by licensees.

The NRC requires licensees to submit their licensing actions and any changes made to the technical specifications for review and approval.

The tables listed above indicate various updates that have been deleted to comply with previous regulatory requirements.


Question: The NRC assesses the capability of the HPCS system to actuate and mitigate potential accidents.

3. ECCS Actuation Instrumentation High Pressure Core Spray (HPCS) System

One RVP - Low pressure transmitter supplying two pressure trip units


Question: The NRC technical library stores industry codes and standards that influence its regulatory process.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at—The NRC Technical Library Two White Flint North 11545 Rockville Pike Rockville, MD 20852–2738.

NRC publications in the NUREG series, NRC regulations, and Title 10, "Energy," in the Code of Federal Regulations may also be purchased from one of these two sources:


Question: Risk-Informed Regulatory Framework

The Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors reflects an innovative approach allowing for different regulatory strategies based on the risk associated with the design and operation of advanced nuclear reactors.

Please see attached document for comments.


Question: How to control steam generator levels during a reactor trip?

Place FRV-220 in Manual and control SG #11 Level to maintain it 15% - 67% preventing a Reactor Trip.

Isolate SG 14 by closing: MFW valves, AFW valves – FMO-241, FMO-242, SG Stop valve – MRV-240.


Question: The NRC defines the parameters for fuel density in the context of reactor operations.

Fuel density is a crucial parameter in determining the performance of nuclear fuel and is specified in Section 8.2.

Flooded conditions within the reactor must be carefully modeled in safety assessments to ensure adequate performance.


Question: The minimum detectable activity for a nuclide is a critical parameter in determining the sensitivity of radiological assessments.

Nuclide MDA (pCi/grams) indicates the minimum detectable activity for each nuclide, providing essential information for regulatory compliance in nuclear safety.

The reactor core is a critical component of the nuclear power plant as it houses the fuel assemblies and moderator, playing a key role in sustaining nuclear fission.


Question: What are the acceptable vibration limits for centrifugal charging pumps?

Centrifugal Charging Pump 1A-A Normal Flow Test requires the comparison of pump vibration readings to the Acceptance Criteria, which includes acceptable ranges for outboard and inboard horizontal, vertical, and axial vibrations measured in inches per second.

These readings are evaluated for trending purposes only.


Question: What are the climate change impacts on urban infrastructure systems?

Climate change impacts on urban water, energy supply, transportation, & other essential infrastructure.

Dependence of energy systems on land & water supplies influencing development of these systems & options for reducing greenhouse gas emissions.


Question: What is the purpose of the control room simulator in nuclear facilities?

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: What are baffle-edge bolts in nuclear reactors?

Baffle-edge bolts (All operating Westinghouse units with baffle-edge bolts) | Yes | 40 – 60 Methods and Frequency. VT-3 visual inspections for cracking (IASCC, fatigue) no later than two refueling outages from the subsequent period of extended operation with subsequent inspections on a 10-Year ISI interval basis.

Significant Past Indian Point and Salem operating experience with cracking of Westinghouse design BF bolts has been profile issue for the NRC to the extent that the NRC has had to.


Question: large break LOCA analysis

This report describes and provides results from the RLBLOCA analysis for the Byron/Braidwood Unit 1 with the Framatome GAIA fuel design.

The transient progression for the demonstration case follows that described in Section 3.2.


Question: What happens when the Low NPSH signal for the Main Feed pump is triggered?

The standby Condensate/Condensate Booster Pump starts, any open 1CD152, Condensate Pumps Recirculating Valve closes...

When the reactor reaches a high pressure, the safety valves will automatically open to relieve excess pressure.


Question: The classification of systems, structures, and components in safety-related contexts is differentiated by designations such as Safety-Related, Non-Safety-Related with Special Treatment, and Non-Safety-Related with No Special Treatment.

The Natrium plant uses three safety classification levels: Safety-Related (SR), Non-Safety-Related with Special Treatment (NSRST), and Non-Safety-Related with No Special Treatment (NST). Explanations for each of the three classifications are provided below.

The In-Vessel Transfer Machine (IVTM) moves core assemblies between the core, in-vessel fuel storage racks, and transfer station for removal from the RV.


Question: The design basis for tornado parameters must take into account the specific geographic location of the nuclear power plant.

Design basis tornado parameters have traditionally been based on Regulatory Guide 1.76 (USNRC, 1974a) and WASH-1300 (USNRC, 1974b).

The likelihood of a given point being struck by a tornado can be calculated by using a method developed by H.C.S. Thom (Thom, 1963).


Question: What is the significance of RPV axial weld failure probability evaluations in nuclear reactors?

The recommendations for inspection of RPV shell welds in BWRVIP-05 (Reference 4.8.10) include examination recommendations of BWR RPV axial welds.

Plant-specific RPV dimensions and material chemistry for DNPS Units 2 and 3 were evaluated for the applicability criteria in BWRVIP-329-A Section 5.0.


Question: what is the method for analyzing compounds in nuclear regulatory filings?

U Compound/analyte was analyzed, peak not identified and/or not detected above MDC.


Question: The NRC mandates that keylock switches must be properly rotated to ensure they are secured in the closed position.

Inserts key and rotates keylock switch 1-1RBCA10 counterclockwise to CLOSE.

Place the following keylock switches in CLOSE:


Question: The NRC's guidelines specify testing methods for assessing the oxidation behavior of cladding materials under accident conditions.

Pre-Decisional Regulatory Guide 1.222 “Measuring Breakaway Oxidation Behavior,” 2016, (U.S. NRC ADAMS Accession Number ML16005A135).

Narukawa, Takafumi and Amaya, Masaki, “The Effect of Azimuthal Temperature Distribution on the Ballooning and Rupture Behavior of Zircaloy-4 Cladding Tube under Transient-heating Conditions,” Journal of Nuclear Science and Technology, 53:11, pp. 1758-1765, 2016.


Question: What is the RadICS system in relation to NRC regulatory filings?

The RadICS system is designed to ensure compliance with NRC regulations and safety requirements by providing a standardized platform for instrumentation and control systems in nuclear power plants.

The meeting aims to discuss the general administrative processes that might affect the project's timeline and resource allocation.


Question: Does the question require the assessment of operational conditions to select an appropriate procedure?

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

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


Question: What is the March 2018 Approach in nuclear licensing?

The Staff proposed the approach, and Powertech and the Tribe agreed to it; Powertech’s own petition for review acknowledges that it agreed to the March 2018 Approach.

The question of whether NEPA could be satisfied through an approach other than the March 2018 Approach was not before the Board.


Question: What are the various analytical methods used for radiological analysis?

The following Analytical Methods were performed: DOE EML HALS-300, Am-05-RC Modified, ASTM C 1475-00 Modified, DOE EML HALS-300, Pu-11-RC Modified, DOE EML HALS-300, Pu-1-RC Modified, DOE RESL N-1, DOE HALS 300, 40.5/2.3/6A-01-R, EPA 905.0 Modified/DOE RP501 Rev. 1 Modified, EPA 906.0 Modified, EPA RERF C-1 Modified, DOE EML HALS-300, Tc-02-RC Modified, DOE RESL Fe-1, DOE RESL Ni-1.

The sample is identified as L1-010-105-AJGS-007-SB with Client Sample ID 485191028 and contains parameters such as Nickel-63 with a result of -0.763 pCi/g.


Question: performance objectives in emergency planning

Section 50.160(b)(1)(ii) requires applicants and licensees using § 50.160 to describe in the emergency plan an approach to develop and maintain at the beginning of each calendar quarter a list of performance objectives for that calendar quarter.

Current regulations in appendix E to 10 CFR part 50, section IV.F include a requirement for periodic drills and exercises for nuclear power reactor licensees to be conducted during an eight-year drill and exercise cycle.


Question: MAGNASTOR Confirmatory Order

The MAGNASTOR Confirmatory Order EA-20-054 outlines the requirements and conditions necessary for the continued operation and performance assessment of the MAGNASTOR storage system for spent nuclear fuel.

The safety analysis report must include details about the containment structures and their ability to withstand external hazards without compromising the reactor integrity.


Question: What are burnable poison concentrations in nuclear reactor contexts?

There are no new restrictions or removal of restrictions on BPs based on the approval of PARAGON2.

The numerical benchmarks used the TR covers all these offerings.


Question: What actions are required when two required diesel generators are inoperable at a nuclear facility?

E. Two required DGs inoperable. E.1 Restore one required DG to OPERABLE status.

D. One required offsite circuit inoperable for reasons other than Condition F.


Question: The method for determining flow rates utilizes the conversion of flow values in gpm into the corresponding values of current, denoting the relationships within the system's operation.

By substitution, $$I = \left[(F)*\left( (16 \, mA)^{0.5}\right)[(flow span in gpm)]^2 + 4 \, mA \right]$$

The values of flow rate are critical for understanding how the system responds under different operational conditions, especially regarding the management of input uncertainties.


Question: What is the purpose of inspections in aging management programs?

The aging management program includes inspections such as the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24) to address aging effects like hardening or loss of strength due to polymeric degradation.

The reactor design must ensure sufficient cooling capacity to prevent overheating during operation, thereby maintaining safe performance levels.


Question: surface water quality impacts

Potential impacts to surface water during operation will not be significantly different than those described in Section 4.5.4.1 of the ISR GEIS.

Storm water runoff will be controlled by a SWPPP issued by EPA.


Question: What are the requirements for license renewal environmental reports?

License renewal requirements for environmental reports are specified in Title 10 of the Code of Federal Regulations (10 CFR) Part 51, “Postconstruction environmental reports.”

The NRC staff will also identify environmental resources that must be described and evaluated in the Supplemental Environmental Impact Statement.


Question: The NRC mandates that licensees prepare detailed safety analyses specifically to assess the risk of accidents.

Licensees are required to conduct probabilistic risk assessments in accordance with recognized standards to evaluate potential accidents at nuclear power plants.

The American Society of Mechanical Engineers is a professional organization that develops technical standards in various engineering fields.


Question: What are the requirements for verifying DSC fuel integrity?

SR 3.1.3 Verify that the time limit for completion of DSC transfer is met.

The analysis of reactor core performance must be conducted prior to any vessel modifications.


Question: lead test assemblies in nuclear reactors

Four (4) GAIA LTAs operated in a Westinghouse 3-loop international reactor starting in 2012, successfully completing four (4) 12-month cycles of leaker-free performance.

GAIA lead test assembly (LTA) and reload operating experience is provided in this section.


Question: Cobalt-60 false positive analysis

UI Results are considered a false positive due to high peak-width.

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


Question: What procedures are in place for reacting to an alarm condition?

Alarm Response Procedures (Various) outline the necessary steps that personnel must follow during an alarm condition to ensure safety and compliance with regulations.

The Emergency Plan (ECG) defines the processes for communication and action in case of a significant incident but does not address reactions to alarms specifically.


Question: tube inspection requirements

The location of each tube plugged [or repaired] during the inspection outage (section 5.6.7.c.4 requires the number of tubes plugged or repaired), and the reason each tube was plugged.

This report is only required if sleeving or other repairs are being made. These types of repairs must be approved by the NRC.


Question: What is the activity level of nuclides in pCi/grams?

Activity (pCi/grams) is a measure of the radioactivity of a nuclide, indicating the number of decay events per second in a specific mass of material.

The Energy (keV) value corresponds to the energy of the emitted radiation from the nuclide during decay.


Question: power range monitor malfunction

Crew will enter 1-AOI-4 Nuclear Instrumentation Malfunction, Section 3.4, Power Range Monitor Malfunction, and defeat the failed power range channel.

Crew will enter 1-AOI-20 Malfunction of Pressurizer Level Control System, Section 3.4, Loss of Letdown or Letdown Malfunction.


Question: What information is provided about NRC devices not subject to registration?

The NRC oversees the safe use of nuclear materials to protect public health and safety.


Question: The NRC monitors external surfaces of mechanical components to prevent loss of material and other aging effects.

Water Chemistry (B.2.3.19) and One-Time Inspection (B.2.3.19) are part of the Aging Management Program for the valve body made from carbon steel, which is intended to maintain pressure boundaries.

Demineralized water is used in some systems to improve operational efficiency and extend service life.


Question: The NRC evaluates environmental impacts related to the subsequent license renewal process for nuclear plants.

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

In accordance with Title 36 of the Code of Federal Regulations Section 800.8(c), the NRC staff elected to use the National Environmental Policy Act of 1969, as amended, process to comply with its obligations under Section 106 of the National Historic Preservation Act of 1966, as amended.


Question: turbine emergency shutdown procedures

IF an emergency condition requires the turbine be brought to rest as quickly as possible, THEN perform the following: NOTE: The following step could require closing MSIVs. Ensure all steam loads to main condenser are isolated.

Simultaneously open the following condenser vacuum breaker valves:  • CNDSR A 38MOV-VB-1  • CNDSR B 38MOV-VB-2


Question: NPSH in Containment Spray System

NPSH for the Containment spray pumps is evaluated for both the injection and recirculation modes of operation for the design basis accident.

The design head of the pumps is sufficient to provide adequate flow at limiting conditions.


Question: What are the characteristics of Nukon low-density fiberglass insulation?

A baseline analysis of Nukon includes a size distribution with two categories—60 percent small fines, and 40 percent large pieces per NEI 04-07.

This guidance provides an approach for determining a size distribution for LDFG using the AJIT data, with conservatism added due to the potentially higher level of destruction from a two-phase jet.


Question: What are the requirements for NPDES permits for nuclear power plants?

To operate a nuclear power plant, NRC licensees must comply with the CWA, including associated requirements imposed by EPA or the State, as part of the NPDES permitting system under CWA Section 402.

The EPA authorized the State of South Carolina to assume NPDES program responsibility.


Question: The flow and movement of chemicals in the testing tanks should mirror or be conservative relative to conditions expected in the actual plant.

Transport of chemicals and debris in testing facility tanks should be representative or conservative with regard to the expected flow and transport in the plant-specific conditions.

Wolf Creek is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the Wolf Creek chemical effects analysis. See Figure 3.0.2-1.


Question: What is energy flow in the context of nuclear thermodynamics?

Energy Flow (BTU/sec) refers to the rate at which energy is transferred or converted in a system, measured in British Thermal Units per second.

Average Enthalpy (BTU/lbm) indicates the total energy content per unit mass of fluid, which affects the system's thermal efficiency.


Question: The NRC sets forth the need for, and the methods to establish, temperature monitoring within the EOS-HSM to confirm compliance with safety standards during normal and off-normal conditions.

The user shall establish in the program, measurement locations in the EOS-HSM that are representative of the EOS-HSM thermal performance and directly correlated to the predicted fuel cladding temperatures, air mass flow rates, and NUHOMS® EOS System temperature distributions that would occur with the off-normal and accident blockage conditions, as analyzed in the UFSAR.

If measurements or other evidence indicate that off-normal or accident temperature limits for fuel cladding have been exceeded, verify that canister confinement is maintained and assess analytically the condition of the fuel.


Question: nondestructive examination of welded joints

Various changes were made to Section NN5.5, Nondestructive Examination of Welded Joints, which probably do not warrant further review.

The above revisions do not warrant further review.


Question: What is nuclear licensing in the context of NRC regulations?

Nuclear licensing is a regulatory process by which the Nuclear Regulatory Commission (NRC) grants permission to operate a nuclear facility, ensuring it meets safety and environmental standards.

The Climate Change Impacts in the United States report discusses various environmental changes affecting different regions and sectors, but does not pertain to nuclear licensing processes.


Question: How often are off-hours report-in drills conducted for nuclear facility emergency response?

Off-hours report-in drills are conducted biennially and are unannounced.

Each NextEra site will conduct communications drills once per calendar year.


Question: What is the minimum cooling time for pressurized water reactor (PWR) fuel with an enrichment of 2.9 to 3.1?

The enrichment levels for uranium-235 in PWR fuel assemblies provide critical data that affects reactor performance and safety.


Question: What is the unit of measurement for radioactive isotopes in NRC filings?

Activity is measured in units such as millicuries (mCi).

The manufacturer name listed for the device is Honeywell International, Inc.


Question: boron flow rates during power reduction

Reactivity Briefing Sheet, "Thumb Rules" (page 3), lists boron flows and volumes for different reduction rates.

Steps 1 and 2 may be performed in any order based on plant conditions.


Question: What is CWA Section 401 certification in the context of nuclear facility licensing?

An applicant (in this case, Duke Energy) for a Federal license to conduct activities that may cause a discharge of regulated pollutants into navigable waters of the United States is required by CWA Section 401 to provide the Federal licensing agency (in this case, the NRC) with water quality certification from the certifying authority (in this case, the State of South Carolina).

On November 3, 2021, Duke Energy notified the SCDHEC of a wastewater spill that occurred on November 1, 2021.


Question: Critical Heat Flux correlation in nuclear engineering

The COBRA-FLX critical heat flux correlation provides an assessment of the critical heat flux under various local conditions, which is crucial for ensuring the safety of boiling water reactors.

The results obtained from the grid tests were consistent with previous simulations and demonstrated satisfactory performance across all evaluated parameters.


Question: The RICT Program incorporates probabilistic risk assessment (PRA) as a fundamental part of its analysis and implementation process.

Enclosure 2 provides a discussion of the results of peer reviews and self-assessments conducted for the plant-specific PRA models which support the RICT Program, as required by Regulatory Guide (RG) 1.200, Revision 2, Section 4.2.

Enclosure 5 provides the plant-specific baseline core damage frequency (CDF) and large early release frequency (LERF) to confirm that the potential risk increases allowed under the RICT Program are acceptable.


Question: What is the significance of Energy measurements in nuclear licensing?

Energy (keV) denotes the energy levels observed in the decay of certain isotopes, which is crucial for analyzing the safety and technical compliance of nuclear facilities.

The values shown in the table indicate statistical measurements, boolean, and categorical data necessary for understanding overall plant operations.


Question: What does PSI refer to in pressurized nuclear reactor safety assessments?

PSI-only examinations, as compared to the proposed alternative of PSI + 10 + 20 + 30 + 60, would involve a much more adverse effect on risk-informed principles, particularly since PSI-only examinations would remove any future condition monitoring needed for risk-informed decision making.

The NRC staff calculated a large percent error when using the procedure described in the EPRI report.


Question: The effectiveness of the Structures Monitoring Program is critical to ensuring that age-related degradation of structures is addressed during extended operations.

In 2022, a program effectiveness review was performed for Structures Monitoring Program described in FSAR Section 0.1.33 in support of preparing the BFN Subsequent License Renewal Application. The purpose of the program effectiveness review was to verify that the intent of the existing aging management program activities to identify and correct, as warranted, age-related degradation of structures and structural components within the scope of the License Renewal, is being effectively implemented in the initial license renewal period of extended operation.

Revise implementing procedures to ensure there are no indications of excessive loss of material due to corrosion or wear and no debris or dirt that could restrict or prevent sliding of the surfaces as required by design.


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

Each peak corresponds to a specific energy level and provides insight into the radioactivity of the sample being analyzed.

The facility is equipped with multiple safety systems to prevent any radiation exposure during normal operations.


Question: assembly average burnup limits

The table provides the minimum cooling time in years based on the average enrichment of the fuel assemblies, specifically detailing the average burnup limits for both 59 GWd/MTU and 60 GWd/MTU classifications.

The regulatory review process involves multiple stages where documents are assessed, and feedback is provided through various channels.


Question: The regulatory framework mandates that nuclear facilities must maintain strict compliance with licensing requirements to ensure operational safety.

The licensing process involves comprehensive evaluations of safety systems and operational procedures to guarantee the protection of public health and the environment.

Nuclear power plants generate electricity through the process of nuclear fission, which involves splitting atoms to release energy.


Question: The NRC mandates that licensed individuals perform analysis on contamination or leak test samples.

Analysis of leak test samples and/or contamination shall be performed by persons specifically licensed by the U.S. Nuclear Regulatory Commission or an Agreement State to perform such services.

Each portable nuclear gauge shall have a lock or outer locked container designed to prevent unauthorized or accidental removal of the sealed source from its shielded position.


Question: radiological emergency preparedness

Other systems determined to support components of Radiological Emergency Preparedness Alert and Notification plans may be used as supplemental/additional means of alert and notification, in conjunction with primary and backup ANS, if the responsible Offsite Response Organization (ORO) determines they are capable of providing adequate alert or notification to the public for the full range of hazards in their community.

Messages provided to the stations will include the identity of the affected utility, references to radiological emergency information, such as brochures and websites, for members of the public to use, and a warning to either stay tuned for additional information or additional stations to tune into for more information regarding the emergency.


Question: What are the requirements for using high-strength reinforcement in nuclear plant design?

Consistent with RG 1.142 Rev. 3, high-strength reinforcement (yield strength greater than 60,000 ksi) should not be used in design, unless otherwise justified. If high-strength reinforcement is used, applicants should demonstrate their adequacy for specific use in the design by testing, analysis, or performance evaluation.

The provisions of Section G2.1 assume monotonically increasing loads. If a flexural member is subjected to load reversals causing cyclic yielding over large portions of a web, such as may occur during a major earthquake, special design considerations may apply (Popov, 1980).


Question: Class 1E electrical equipment in nuclear reactors

SR 3.7.20.1 Verify each Class 1E electrical equipment A/C train actuates on an actual or simulated actuation signal.

C. Two Class 1E electrical equipment A/C trains inoperable.


Question: What is the significance of ITAAC notifications in nuclear regulatory processes?

The ITAAC NOTIFICATION provides notification of the status of the ITAAC, including whether it has been satisfied or not.

The NRC construction inspection reports are essential for verifying compliance with construction standards.


Question: What are the allowable occurrences for boron injection in the reactors?


Question: The NRC's review process ensures that licensees maintain adequate emergency preparedness plans.

The acceptance of the license amendment request for the Perry Nuclear Power Plant involves a thorough review of the Emergency Plan as mandated by regulatory requirements.

The email discusses the technical requirements for the maintenance of nuclear reactor components without mentioning any safety plans.


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

The Safety Analysis Report (SAR) provides the necessary information to demonstrate that a facility will be designed and operated in a manner that provides reasonable assurance of the protection of public health and safety.

Licensees must complete the operational readiness assessments before commencing the installation of new systems.


Question: DNPS has three production wells that provide water for various systems at the facility, including cooling and sanitary needs.

The two primary water supply wells are approximately 1,500 feet deep (EGC 2021b). The two wells tap the Cambrian-Ordovician aquifer (AEC 1973).

A summary of monthly surface water withdrawals reported by DNPS between 2018 and 2022 is included as Table 3.6-3b.


Question: The NRC mandates that the external radiation monitoring program must utilize personnel monitoring devices to assess worker safety during operations.

The external radiation monitoring program will ensure the safety of workers and the public. This program is discussed in Section 9.6.1 of the COP. As part of the external radiation monitoring program all personnel will be issued personal monitoring devices such as TLDs for at least the first year of operations. The badges will be analyzed on at least a quarterly basis by a vendor and the results will be reported to the NRC.

The samples will be analyzed for dissolved and suspended natural uranium, total and soluble radium-226, thorium-230, polonium-210, and lead-210.


Question: The proposed modifications do not influence the established safety margins or regulatory limits as specified in the Updated Final Safety Analysis Report.

The proposed change revises the inspection frequencies for SG tube inspections and associated reporting requirements. The proposed change does not change any of the controlling values of parameters used to avoid exceeding regulatory or licensing limits, or any controlling value for a parameter established in the Updated Final Safety Analysis Report or the license.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.


Question: DCPP Internal Coatings Aging Management Program

The new DCPP Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks AMP will be informed and enhanced through the systematic and ongoing review of both plant-specific and industry OE, including research and development, such that the effectiveness of the AMP will be periodically evaluated.

AMP effectiveness will be assessed at least every five years per NEI 14-12.


Question: minimum detectable concentration

The Minimum Detectable Concentration (MDC) is a statistical measure that defines the lowest concentration level that can be reliably quantified in a sample analysis.

The report includes various client samples with specific IDs to ensure accurate tracking and auditing of analytical results.