Question: nuclide activity measurements

The activity of each nuclide is expressed in picocuries per gram (pCi/grams), providing a clear indication of the radioactive decay rate for that nuclide, which is essential for assessing safety and compliance.

The report includes a comprehensive summary of the regulatory framework governing the licensing process for new nuclear reactors, detailing the steps required for submittals and evaluations.


Question: What is the significance of 10 CFR 50.54 in nuclear emergency planning?

Title 10 CFR 50.54(q)(3) states, that the licensee may make changes to its emergency plan without NRC approval only if the licensee performs and retains an analysis demonstrating that changes do not reduce the effectiveness of the plan.

Enforcement: The ROP’s significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRC’s ability to regulate using traditional enforcement to adequately deter non-compliance.


Question: What is the role of the PHOENIX-P/ANC Nuclear Design System in reactor core qualification?

Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, dated June 1988 (ADAMS Accession No. ML2015M337)

Fuel Rod Bow Evaluation, dated July 1979 (ADAMS Accession No. ML19208D506 and ML20079K315)


Question: What is wind pressure distribution in reactor buildings?

Wind pressure distribution on the reactor building is critical for evaluating the safety and structural integrity of the facility under wind loading conditions.

Elevation markers indicate specific heights at which pressure calculations are made, with EL. 750' and EL. 740' being two such measures used in the analysis.


Question: reactor coolant pump startup procedures

Figure 15-80 illustrates the startup of an inactive reactor coolant pump at an incorrect temperature, showing the detailed pressure readings over time during the procedure.

The reactor pressure vessel is a crucial component in maintaining the reactor's safety and efficiency, as it contains the reactor core and provides a barrier against radiation.


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

Nuclide MDA (pCi/grams) is the minimum activity of a nuclide that can be detected in a sample, expressed in picocuries per gram.


Question: nuclear licensing exemption

The exemption request evaluation for the nuclear facility outlines the specific criteria under which exemptions may be granted by the NRC, including detailed assessments of safety and operational impacts.

The license number DPR-72 indicates the regulatory status of a specific facility, but does not provide any information about the exemption request process.


Question: What is the procedure for starting the Unit 1 Diesel Generator Cooling Water Pump?

Dispatches EO to start U1 Diesel Generator Cooling Water Pump.

Acknowledges and reports annunciators 901-3 H-2, AREA HI TEMP STEAM LEAK DETECTION.


Question: Review the measures taken to manage the risks associated with heavy equipment in nuclear facilities.

In accordance with recommendations provided in NUREG-0612, Section 5.1, review how the licensee addressed the potential for accidental dropping of the steam generator inside the reactor containment building.

The American Society of Mechanical Engineers (ASME) B30.2 outlines safety standards for overhead and gantry cranes.


Question: wind speed measurement in nuclear regulations


Question: The NRC's Maintenance Rule outlines evaluations that must be conducted to assess the effects of operational changes on system performance.

The existing Maintenance Rule (MR) monitoring programs under 10 CFR 50.65(a)(1) and (a)(2) provide for evaluation and disposition of unavailability impacts which may be incurred from implementation of the RICT Program.

Plant modifications to reduce risk impact of future planned maintenance configurations.


Question: LPCI Coupling inspection process

Bolts 1-24 (of 48). No indication of cracking or damage.


Question: Relative Percent Difference in radiochemistry analysis

The Relative Percent Difference (RPD) obtained from the sample duplicate (DUP) is evaluated against the acceptance criteria when the sample is greater than five times (5X) the current required detection limit (RL). In cases where either the sample or duplicate value is less than 5X the RL, a control limit of +/- the RL is used to evaluate the DUP result.

For PS, PSD, and SDILT results, the values listed are the measured amounts, not final concentrations.


Question: plenum gas temperature

The plenum gas temperature in GALILEO is calculated using a simple equation given below.

Audit calculations were performed by the NRC staff using FRAPCON, and the plenum gas temperature was found to be the same or higher (more conservative) in GALILEO than in FRAPCON.


Question: What are the reasons behind the NRC's policy against allowing other operators to observe an operating test?

The bases for this policy include the desire to minimize undue stress on the operators (or applicants) that are being evaluated and the need to minimize crowding in the simulator (for the examinees, NRC examiners, facility evaluators, operations and training representatives, and simulator operators that have to be there).

There are instances where the crew being examined may want another operator to observe.


Question: What is the function of thermal regeneration demineralizers in a Reactor Coolant System?

The function of the thermal regeneration demineralizers is to store the total amount of boron that must be removed from the Reactor Coolant System to accomplish the required dilution during a load cycle in order to compensate for xenon buildup resulting from a decreased power level.

The letdown reheat heat exchanger is used only during boron release operations and it is then used to heat the process stream.


Question: National Inventory of Dams in nuclear assessments

The USACE maintains the National Inventory of Dams (NID), which provides characteristics for each dam (location, height, and volume).

Sunny day project failures are postulated to occur due to non-hydrologic and non-seismic causes as required by Reference 19.


Question: Browns Ferry Nuclear Plant inspection results

The Browns Ferry Nuclear Plant was subject to an inspection from October 1, 2024, to December 31, 2024, by resident inspectors and senior operations engineers.

The inspection reports were submitted to the agency for review and approval in accordance with standard operational procedures.


Question: The boron concentration limits for nuclear fuel assemblies are critical for maintaining the safety of the reactor.

Specified soluble boron concentrations are independent of whether an assembly contains a nonfuel insert.

Amendment No. 12, Revision No. 1 was issued to update the licensing parameters required for reactor operation.


Question: What regulations govern the procedures for performing maintenance on safety-related equipment in nuclear facilities?

Regulatory Guide 1.33, Appendix A, Section 9, ‘Procedures for Performing Maintenance,’ specifically addresses the need to have appropriate procedures for preventive maintenance that can affect the performance of safety-related equipment.

The inspectors concluded that the failure to replace or trip test MCCBs in accordance with PMCR 83591 constituted a performance deficiency that warranted further evaluation.


Question: The results of the analysis conducted on the samples provided to the laboratory on the specified date are encapsulated in this report.

All analyses were performed according to ACZ’s Quality Assurance Plan. The enclosed results relate only to the samples received under L88961.

ACZ retains analytical raw data reports for ten years.


Question: The analysis of groundwater samples can reveal essential information about dissolved metals concentrations.

Molybdenum, dissolved was analyzed using EPA Method 200.8, resulting in a concentration of 0.0634 mg/L in the sampled groundwater.

The sample was collected on 11/12/24 and received for analysis on 11/15/24, indicating a standard procedure for handling laboratory samples.


Question: What does the TDEC define as a man-made or natural watercourse?

TDEC defines a WWC as “man-made or natural watercourses, including natural watercourses that have been modified by channelization: that flow only in direct response to precipitation runoff in their immediate locality; whose channels are at all times above the ground water table; that are not suitable for drinking water supplies; and in which hydrological and biological analyses indicate that, under normal weather conditions, due to naturally occurring ephemeral or low flow there is not sufficient water to support fish, or multiple populations of obligate lotic aquatic organisms whose life cycle includes an aquatic phase of at least two months [TDEC Rule 0400-40-03-.05(9)].”

NWI features mapped within the 3-mi. (4.8-km) vicinity include lacustrine, riverine, palustrine scrub-shrub, palustrine forested, and open water wetlands as summarized in Table 3.4.4-1 (USFWS, 2022).


Question: equivalent boron content

ASTM D7219-19 provides that high purity nuclear graphite for a gas-cooled reactor to contain an equivalent boron content (EBC) equal to or less than 2 ppm as a purity specification.

The number of specimens per data set is determined by the sample cutting plan and the need to support a Weibull analysis.


Question: nuclear reactor safety systems

Nuclear reactor safety systems are designed to prevent or mitigate the consequences of accidents, ensuring the health and safety of the public and the environment.

The outreach strategy involved engaging various stakeholders to gather their inputs and concerns regarding the proposed project.


Question: Class 103 license for reactors

EN expects that each Xe-100 unit will have its own Class 103 license issued through review of a common application.

The submitted FSAR will be amended periodically during the construction phase and NRC staff review in response to design changes and NRC requests.


Question: The Inservice Testing Program for the Braidwood Nuclear Power Station is critical for ensuring the reliability of systems important to safety and operations.

The Inservice Testing Program for Braidwood Nuclear Power Station is designed to ensure that safety-related systems perform their intended functions during their operational life.

Commercial service dates for Unit 1 and Unit 2 are essential for maintaining regulatory compliance and operational readiness.


Question: The RFPC System aids in controlling the pH levels of the recirculation fluid to ensure proper absorption of fission products during a DBA.

The Recirculation Fluid pH Control System (RFPC) is a subsystem of the Containment Spray System that assists in reducing the iodine fission product inventory in the containment atmosphere resulting from a Design Basis Accident (DBA).

The containment spray system functions to cool and maintain pressure within the containment structure, particularly during emergency scenarios.


Question: What are the criteria for bounding PWR fuel assembly loading?

Assembly characteristics represent cold, unirradiated, nominal configurations.

The maximum active length of the BW15H2 assembly is 0.4988 inches.


Question: What is the significance of PMF in dam safety evaluations?

The reassessment of the PMF involved evaluation of three candidate flood events: 1. March 7,980-square-mile 2. March 21,400-square-mile 3. March 12,030-square-mile

The reassessment addressed the effects of these dam safety modifications on maximum flood levels at WBN and on warning time available for safe plant shutdown.


Question: What is the Dakota-Glen Canyon aquifer system?

The Dakota-Glen Canyon aquifer system comprises four permeable zones including, in descending order, the Dakota Sandstone aquifer, the Morrison Formation aquifer, the Entrada Sandstone aquifer, and the Glen Canyon Sandstone aquifer.

The Uinta-Animas aquifer within the San Juan Structural Basin comprises the San Jose, Animas, and Nacimiento formations and the Ojo Alamo Sandstone.


Question: What is an IGM Grid in the context of nuclear reactor design?

The IGM Grid is a critical component in the mechanical design of nuclear reactors, serving to support and stabilize fuel assemblies within the reactor core.

The licensing process for nuclear facilities involves several stages, including preliminary consultations and submission of safety-related documents to regulatory authorities.


Question: What is the holding time in nuclear sample analysis?

Samples (See Below) were received and analyzed outside the hold time.

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


Question: The report contains a detailed list of identified radioactive substances and their characteristics.


Question: What are the environmental impacts of the Browns Ferry Nuclear Plant license renewal?

Table 8.1-1 summarizes environmental impacts of the Proposed Action (SLR) and the reasonable alternatives, for comparison purposes based on the 10 topical environmental resource areas identified in the Generic Environmental Impact Statement (GEIS) (NRC 2013) but consistent with the topical organization of Chapter 4 of this document.

TVA concludes the environmental impacts of the continued operation of BFN, providing approximately 3,900 megawatts electric (MWe) of base-load power generation through 2056, would be smaller overall than impacts associated with any of the other reasonable alternatives analyzed.


Question: The abrupt changes in temperature caused by thermal effluent can lead to adverse conditions such as cold shock for aquatic organisms.

Cold shock can occur when organisms acclimated to the elevated temperatures of a thermal plume are abruptly exposed to temperature decreases when thermal effluent stops.

Stimulation of nuisance organisms has been satisfactorily mitigated at the single nuclear power plant with a once-through cooling system where it was previously a problem.


Question: What is the purpose of a safety analysis report in nuclear licensing?

A Safety Analysis Report (SAR) is a comprehensive document that describes the design and performance analysis of a nuclear facility, demonstrating that the facility will operate safely and in compliance with regulatory requirements.

The cooling system of a reactor is designed to maintain the core temperature within safe limits during normal operation.


Question: What are the RCRA metals and their implications in nuclear safety?

RCRA Metals include arsenic, mercury, barium, selenium, cadmium, silver, chromium, and lead, which are of high importance in regulatory compliance and environmental safety.

Chain of custody involves documenting the collection, transfer, and analysis of samples to ensure their integrity throughout the testing process.


Question: The structures monitoring program evaluates the impact of aging effects on concrete components in nuclear facilities.

Increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation

The material used for construction needs to withstand various environmental conditions.


Question: What is the role of the U.S. Nuclear Regulatory Commission in nuclear licensing?

The U.S. Nuclear Regulatory Commission is responsible for ensuring the safe use of radioactive materials for beneficial civilian purposes while protecting people and the environment.

The Order of the Secretary states that the Time to File has been extended for certain submissions but does not affect other ongoing adjudications.


Question: ESW pump 46P-2B operation

The CRS has directed you to supply cooling water to EDGs A and C from ESW pump 46P-2B per OP-22 Section G.21.

The candidate determines where to obtain a controlled copy of OP-22.


Question: What is the role of unqualified coatings in reactor safety assessment?

The unqualified coatings on the reactor vessel head above the permanent cavity seal ring were assumed not to become debris for the postulated breaks.

The generated quantities of the unqualified coatings are shown in Table 3.h.5-2.


Question: Americium-241 measurement in nuclear analysis

Rad Gas Flow Proportional Counter|           |         |                  |       |      |         |      |      |              |       |        |        |


Question: What are the redundancy requirements for emergency core cooling systems?

Redundancy of Flow Paths and Components for Long-Term Emergency Core Cooling is vital to ensure that if one system fails, another can take over and maintain core cooling.

In the event of a loss of coolant accident, proper procedures must be followed to stabilize the reactor and prevent core damage.


Question: What are the zirconium-steam reaction correlations used in NRC safety models?

The 2011 IAEA report does opine that the two zirconium-steam reaction correlations—the Baker-Just and Cathcart-Pawel correlations—that computer safety models use for NRC’s legally-binding simulations of LOCAs are reliable for intact fuel cladding.

For example, at one location on the central fuel bundle (at the 42-inch elevation) when cladding temperatures had reached just below 2200°F, the fuel cladding heat up rate had increased to approximately 21.4°F per second; at the same location, between approximately 2200°F and 2780°F, the average heat up rate was approximately 36.3°F per second.


Question: The keylock switch 1-1RBCB10 must be turned to the CLOSE position to ensure safety.

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

CRITICAL STEP: The secondary coolant system operates to manage the reactor temperature effectively.


Question: Loss of Auxiliary Air response procedures

Control Room operators are responding to a Loss of Auxiliary Air using 0-AOI-10, Loss of Control Air, Section 3.2, Loss of Auxiliary Air, due to an unisolable air header rupture.

Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task, return the cue sheet I provided you.


Question: The NRC mandates that personnel involved in nuclear operations must undergo appropriate training and certification to ensure compliance.

NDT Level II, certified 12/13/2021, expires 11/22/2024, Vision exam expires 10/19/2024 (current for project reviewed).

The checklist includes names of personnel along with their job titles and completed training status.


Question: What is the flow rate measured in lbm/sec for Break Path No. 1?

The flow rate for Break Path No. 1 is measured in lbm/sec, which is a unit that denotes the mass flow rate of a substance in pounds mass per second.

The values listed in the table show various energy measures in Btu/sec, which represent the amount of energy transferred or converted per second.


Question: What are the seven cost elements of the decommissioning estimate?

There are seven cost elements of the decommissioning estimated in relation to the fixed price services contract.

The appropriations for FY2025 will be included in the CY2024 report.


Question: Co-60 contamination levels

- **Contamination, less than:** .0002 Microcuries


Question: The NRC mandates that licensees must conduct an assessment of possible emergency conditions and their respective impact zones.

The NRC requires licensees to analyze emergency zones to ensure adequate planning for potential incidents.

The system's architecture is designed to address Common Cause Failures (CCF) and enhance the overall safety mechanisms in place.


Question: ISU is expected to provide key materials and justifications for the License Amendment Request (LAR) during the audit process.

ISU should be prepared to provide documents, calculations, and other material, as applicable, supporting the analyses and justifications for the LAR, and the SAR and any procedures supporting the LAR.

The NRC staff performing this audit will be: Xiaosong Yin (Project Manager, and Technical Reviewer), Duane Hardesty (Senior Project Manager, and Technical Reviewer), Joshua Borromeo (Branch Chief).


Question: The NRC collects and evaluates information on groundwater levels and well performance.

The total monthly groundwater extraction from the wells amounted to 4.2117, showing varied usage patterns across the months.

The data was collected over a year to assess the impact on local ecosystems.


Question: FCS Open Land Area definition

The FCS Open Land Area SUs are specific units that denote regions maintained in a natural state to protect ecological integrity while allowing controlled activities.

The report also details various measures taken to ensure the safety of operational personnel during maintenance procedures.


Question: What temperatures are associated with core reflood initiation during a loss-of-coolant accident?

Depending on the plant design, core reflood starts at cladding temperatures of between 1300°F (or less) and 1600°F; these are significantly lower than in FLECHT Run 9573 and at flooding rates substantially above the 1.1 inch/second of this test.

In PRM-50-93, Petitioner discusses the NRU reactor thermal-hydraulic experiments on several pages (pages 14-20, 24, 73-74, 75) and Appendix D lists data from the 28 tests conducted in Thermal-Hydraulic Experiment 1, yet NEI has not commented on the NRU reactor thermal-hydraulic experiments in NEI’s comments.


Question: annual summary of ambient conditions

Table 3-5: Annual Summary of Dry Bulb Temperature, Relative Humidity, Wind Speed and Precipitation from Tower for 2019 provides a detailed overview of environmental conditions including temperature, humidity, and precipitation over the year.

The integration of TRISO fuel into the reactor design presents several advantages, including enhanced safety, greater efficiency, and reduced waste generation.


Question: The samples in this SDG were analyzed on a 'dry weight' basis.

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

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


Question: What does nuclear licensing involve in regulatory processes?

Nuclear licensing is a regulatory process that involves obtaining permission from the Nuclear Regulatory Commission (NRC) to operate a nuclear facility, ensuring compliance with safety standards and regulations.

The use of high-quality fuel sources is essential for maintaining optimal reactor performance and efficiency.


Question: What is the Minimum Detectable Concentration (MDC) in nuclear safety assessments?

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

The Project involves several phases, including land clearing, soil remediation, and final site grading.


Question: What is fuel fragmentation in nuclear reactors?

Fuel fragmentation, relocation, and dispersal can occur during a loss of coolant accident (LOCA) and must be assessed for fuel performance and safety evaluation.

The latest advancements in digital instrumentation have improved monitoring capabilities across various reactor components and systems.


Question: Loss of All AC power

Unit 1 has experienced a Loss of All AC power.

At 1000 DC Bus 112 125V Batteries have a suspected ground with a discharge rate as shown on the meter:


Question: Minimum Detectable Activity in nuclear analysis

MDA: Minimum Detectable Activity is the smallest amount of radioactivity that can be reliably detected by the measurement system.

The Certificate of Analysis provides a comprehensive overview of the sample parameters and the associated qualities.


Question: environmental monitoring for radioactive materials

Environmental Monitoring & Air Sampling for Radioactive Labs is crucial in ensuring safety and compliance with regulatory standards.

The course on Medical Uses of Radiation focuses on the applications and safety measures in medical settings.


Question: What is the significance of common cause failure (CCF) in software development processes for nuclear regulatory compliance?

This document focuses on systematic failures due to a latent defect in software, and an approach to providing reasonable assurance through a quality software development process that the common cause systematic failure of an application is adequately addressed.

General Comment on Section 5 titled, 'NRC Regulatory Framework Versus Implementation Level Activities to Address Software CCF'.


Question: What is the significance of CETCs in determining core cooling conditions?

The first block of 1BST-2, CORE COOLING, checks if CETCs are less than or greater than 1200°F. If greater than 1200°F then an inadequate core cooling condition has been reached.

Screening for SRO-only linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.


Question: The NRC mandates specific procedures for ensuring the functionality of safety systems like the HPCI throughout various operational scenarios.

Emergency core cooling systems, such as the High Pressure Coolant Injection (HPCI), are critical safety systems required for maintaining reactor safety during abnormal operation.

In the scenario setup, the operator must verify the control switches for the 1C RFP are in PTL, as indicated by the hang INFO cards.


Question: What are the stages in the design and implementation of nuclear reactor systems?

The design and implementation of nuclear reactor systems typically encompass several key stages, including conceptual design, detailed design, licensing, construction, and operational phases.

Safety analysis reports are crucial for understanding the operational limits and performance measures of a reactor under varying conditions.


Question: The NRC reviews the packing fraction of materials used in reactor cores, such as B₄C.

Therefore in the evaluation of the reactivity of the B₄C ball system, each ball is modeled as one of 38.5 mm in diameter filled with B₄C with 92% of theoretical density.

All blue - Volume = 32,269 mm³ - B₄C (92% TD) volume = 29,880 mm³ - HT9 volume = 2389 mm³.


Question: steam generator tube inspection requirements

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, “Steam Generator (SG) Program.”

The safety committee reviewed the annual report and made recommendations for the upcoming inspections.


Question: What is the significance of Cs-137 in soil data analysis?

The percentage of the total of Cs-137 to the other ROCs went up in both cases when compared to the Cs-137 value in Table 41 of RS-TD-31319-001, Revision 5 of 0.829.

The data set was also examined to verify that the IC dose was insignificant when compared to the five main ROCs.


Question: What is the purpose of liquid scintillation analysis in nuclear measurement?

Rad Liquid Scintillation Analysis is performed to measure the activity of radioactive isotopes in liquid samples, providing critical data for safety assessments and regulatory compliance.

The Project is focused on the restoration of the LACBWR site, which includes various technical challenges that need to be addressed prior to the final cleanup.


Question: ASME Code requirements for nuclear safety

Adherence to and reconciliation of ASME Code requirements: The ASME Code does not specifically address control of machining operations or establish special requirements in this area. It is intended, however, that the manufacturer performing machining operate under a controlled fabrication system which complies with the requirements of ASME, NCA-4000. Non-code machine shops should operate under a system which meets the requirements of 10 CFR 50 Appendix B, 'Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,' for their scope of activities.

The inspection should focus on the impact of load handling activities on reactor core and spent fuel and its cooling and plant support systems for the reactor unit and common systems for the other operating unit(s) at the site.


Question: mitigation measures for ecological impacts

Mitigation of potential vegetation impacts will consist of avoidance of juniper and pinyon pines and temporary and permanent revegetation of disturbed areas.

Given these factors and the limited use of the project areas by most vertebrate species of concern, impacts to those species are expected to be minimal.


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

The action level is the threshold for radiation measurements that indicates a need for further investigation and possible corrective actions.

The facility operates several containment structures designed to prevent the release of radioactive materials into the environment.


Question: EOF Emergency Director staffing requirements

The proposed emergency plan does not assign an EOF Emergency Director minimum staff ERO position at the Site Area Emergency classification level for the Command and Control function.

The ERO Key Function Analysis compares and evaluates the current emergency plan on-shift and minimum staff ERO positions against those in the proposed emergency plan and in NUREG-0654 R2 guidance.


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

Energy levels indicate the specific points at which nucleons can exist in an atom, allowing for analysis of emitted radiation during nuclear reactions.

Ambient background measurements are essential for accurate data interpretation in various experimental setups.


Question: The document details revised proposals for using specific code cases to address piping materials.

Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping

If you have any questions, please contact Tom Loomis (610) 765-5510.


Question: What is the significance of NUREG-2191 in aging management programs?

Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP.

Plant-specific aging management programs often address unique conditions and requirements that may not be covered by generic guidelines.


Question: nuclear licensing and regulatory filings

Nuclear licensing is the process by which a facility must receive approval from the regulatory authority to operate, including assessments of safety and compliance with environmental standards.

The primary objective of the operation phase is to maintain efficiency and reliability across all systems, ensuring economic viability and continued service to customers.


Question: What are the differences between design success criteria and PRA success criteria?

Differences in success criteria typically arise due to the requirement in the PRA standard to make PRAs realistic rather than bounding, whereas design basis criteria are necessarily conservative and bounding.

Examples of calculated RICT are provided in Table E1-2 for each individual condition to which the RICT applies (assuming no other SSCs modeled in the PRA are unavailable).


Question: What is Control Bank Worth in nuclear operations?

To determine Control Rod Worth, the candidate may use BCB Unit 2 Figure 2(c) or interpolated from BCB Unit 2 Table 1-7, for 6350.7 EFPH, page 3 of 5 to obtain the (-)116.3 pcm value.

The NRC requires compliance with the annual report submission deadlines to ensure timely regulatory oversight.


Question: What are the requirements for testing and inspections of carbon filters in nuclear facilities?

The manufacturer was required to verify by appropriate tests the following: a. Carbon filter capabilities for removal of molecular iodine - 131 and methyl iodide b. Carbon filter iodine collection capability c. Carbon filter cell leak-tightness integrity d. Carbon filter flow resistance.

Analysis has shown that if both outside air intake locations are closed, assuming a 1/32" crack around all duct openings, cable openings, and door openings due to aging of the sealant, a 1/8" waterage pressure differential would cause a 950 CFM leak rate.


Question: laceration observations in nuclear safety reports

A close-up view of the laceration observed to the caudal peduncle (01/29/2025).

A left, dorsolateral view of the head showing the characteristic elongated rostrum (01/29/2025).


Question: The analysis must ensure that the core's thermal and hydraulic properties are accounted for following a steam pipe failure.

The analysis of the steam pipe rupture has been performed to determine: A. The core heat flux and RCS temperature and pressure resulting from the cooldown following the steam line break. The LOFTRAN code has been used.

To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the limiting conditions of the cases analyzed.


Question: The NRC evaluated the effects of license renewal on local ecosystems and determined the overall impact would be minimal.

After reviewing the site-specific environmental issues in this EIS and conducting an impacts determination consistent with Commission direction in CLI-22-02 and CLI-22-03, the NRC staff concluded that issuing subsequent renewed licenses for the Oconee Station would have SMALL environmental impacts.

In Chapter 3 of this EIS, the NRC considered the following alternatives to renewing the Oconee Station operating licenses: - no-action - new nuclear (advanced light-water reactor and a small modular reactor) - natural gas-fired combined-cycle - combination – small modular reactor, solar photovoltaic, offshore wind, and demand-side management.


Question: What is the relationship between nuclear power plants and socioeconomic impacts?

Socioeconomic impact analyses of resources (e.g., housing) affected by changes in regional population are based on employment trends at nuclear power plants.

As a result, a site-specific evaluation of applicable Category 1 issues is required to be included in SLR ERs.


Question: What is the function of the heat exchanger components in a nuclear reactor?

Heat exchanger components are intended for heat transfer and are typically made of stainless steel, operating in treated water at temperatures exceeding 60°C (approximately 140°F).

Reactor coolant pressure boundary components are crucial for maintaining the integrity of the reactor pressure boundary and are designed to handle the reactor coolant environment.


Question: What is ECCS leakage and its impact on dose calculation in nuclear safety analyses?

The ECCS leakage path RADTRAD file using Westinghouse fuel was rerun, for Revision 8, using the correct model reducing the dose from the Rev7 value of 1.43 rem to 0.33 rem.

The dose results from the GAIA source term are more limiting than Westinghouse fuel.


Question: If the battery cells are not operating at the required voltage, what is the implication for Unit 2?

Grounds have been detected in a few cells of the 2A 125 VDC Battery, and 3 connected battery cells have a Float Voltage of 0 Volts.

Unit 2 is currently operating at full power and has no issues with reactor performance.


Question: Time Critical Operator Action in nuclear safety

According to the FSAR, with the single failure of the TDAFW considered, it is assumed that 390 gpm total AFW flow will be delivered to two of the intact SGs at 1 minute after the trip and an additional 195 gpm of AFW flow will be delivered to the third intact SG at 10 minutes after the trip. The AFW flow initiated at 1 minute after the trip is delivered from the MDAFWP that is aligned to two intact SGs. All flow from the other MDAFWP aligned to both the third intact SG and the faulted SG is initially assumed to spill out the break. Subsequently, a time critical operator action (TCOA) is taken within 10 minutes to isolate the faulted SG and direct AFW flow from this MDAFWP to the third intact SG.

A main feedwater line break occurs on Unit 1.


Question: What are the scoping criteria for mechanical systems in SLR analysis?

Each of these systems and structures were evaluated against the scoping criteria in 10 CFR 54.4(a)(1), (a)(2), and (a)(3), to determine if the system or structure performs or supports a safety-related intended function, if system or structure failure could prevent the satisfactory accomplishment of a safety-related function, or if the system or structure performs functions that demonstrate compliance with the requirement of one of the SLR regulated events.

For the purposes of system level scoping, a bounding approach is used for plant electrical and instrumentation and control (I&C) systems.


Question: ASME Section XI examination requirements

All nondestructive examinations of the required examination surface or volume shall be conducted to the maximum extent practical. When performing VT-1, surface, radiographic, or ultrasonic examination on a component with defined surface or volume, essentially 100% of the required...

Approval is requested by April 13, 2021 to support completion of preservice inspection activities for Unit 3.


Question: downcomer boiling effects

Downcomer boiling phenomena in a conventional pressurized water reactor has an important effect on the transient behavior of a postulated large-break LOCA..., because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling and finally result in a failure of the nuclear fuel rods.

It is significant that reflood rates could be affected by downcomer boiling and refl0od downcomer bypass.


Question: What is the significance of Ra-226 and Ra-228 in nuclear safety monitoring?

Precision bar values may exceed the scale.


Question: What is a Voluntary Party certification in relation to environmental applications?

Other than being an owner or operator of the Site, it is not a “liable party” of the Site, as defined by Tennessee Code Annotated section 68-212-202(a)(4)(B-E);

It is financially viable to meet the obligations of a Brownfield Voluntary Agreement (BVA).


Question: nuclear licensing process

The nuclear licensing process involves a thorough review of safety and environmental assessments to ensure compliance with regulatory standards before a reactor can operate.

The final report is critical for ensuring that all previous findings are documented correctly and submitted to the appropriate regulatory bodies.


Question: What are the key subsurface units discussed in the NRC audit submittal?

Specifically, revised PSAR Figure 2.5-3 shows approximately 20 ft of clay fill underlain by a thin lens of alluvial clay that thickens towards the center of the profile and a thick layer of residuum clay to the northwest (B) section of the profile that thins towards the center.

Additional audit team members may be added as needed.


Question: The NRC conducts regular evaluations to ensure compliance with safety regulations at nuclear facilities.

The 2515 Light-Water Reactor Inspection Program is designed to assess the operational safety and regulatory compliance of light-water reactors during their operational phase.

Nuclear plants often undergo extensive environmental reviews to evaluate their effects on local ecosystems.


Question: uranium fuel cycle

The uranium fuel cycle encompasses all stages of uranium production, from mining and milling to fuel fabrication, reactor operation, and ultimately the disposal of spent fuel.

The section on socioeconomics analyzes the potential impacts of the Monticello Nuclear Generating Plant on local economies and job markets.


Question: Thyroid TEDE calculation

- Natural deposition - Powers' Model, Compartment 1
  - Deposition Lambda (1 / Hours)
    - Noble Elemental Organic Aerosol
      - 1.0000E+00  2.0000E+00  6.0000E+00


Question: What is the purpose of Eddy Current Examination in steam generator inspections?

The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations: Eddy Current Examination.

The inspectors observed and evaluated licensed operator performance in the control room during reactor shutdown for the refueling outage on November 1, 2024.


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

Nuclide confidence index threshold = 0.30


Question: licensing and regulatory requests for nuclear plants

the U.S. Nuclear Regulatory Commission (NRC) notified your office that it is conducting an environmental review for a series of licensing and regulatory requests submitted by Holtec Decommissioning International, LLC (HDI) that, if approved, would collectively support reauthorizing power operations at the Palisades Nuclear Plant

To keep consulting parties informed and in accordance with the NRC’s Principles of Good Regulation, the NRC and DOE LPO staff have performed a series of outreach activities including conducting virtual Tribal information meetings and site visits.


Question: The NRC conducts evaluations of detected contaminants to assess compliance with safety standards.

Lead 210, a radioactive isotope, was analyzed using the EICHROM OTW01 method in various sample types to establish its concentration and compliance with regulatory limits.

ACZ Laboratories, Inc. is located in Steamboat Springs, Colorado, and offers a range of analytical services.


Question: What is the primary containment isolation instrumentation for nuclear reactors?

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

Separate Condition entry is allowed for each channel.


Question: National Ambient Air Quality Standards compliance

Compliance with the National Ambient Air Quality Standards (NAQS) is indicated by Y (Yes) and N (No).

Measured Concentrations of Particulate Matter with NAQS and Comparison


Question: What is the average enrichment requirement for PWR fuel assemblies?

Initial Assembly Avg. Enrichment wt % 235U (E) - 1.3 ≤ E < 1.5

The NRC requires that the safety systems of a nuclear power reactor must be designed to monitor various operational parameters.


Question: analytical method for sample analysis

Sample Analysis Requested: Analytical method requested (e.g. 25069, 608/8770/4) and number of containers provided for each code (e.g. 25069 - 30, 608/8770/4 - 1).

For Lab Receiving Use Only: - Custody Seal Intact? [ ] Yes [X] No - Cooler Temp: ______ °C - Sample Collection Time Zone: - [ ] Eastern - [ ] Pacific - [X] Central - [ ] Mountain - [ ] Other: __________


Question: Risk Informed Completion Time

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, “Risk-Managed Technical Specifications (RMTS) Guidelines.”

The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities."


Question: What are the nuclear qualifications for a pressure transmitter?

The device measures the pressure and temperature effectively within a wide range of operating conditions.


Question: What are RISC-2 SSCs in nuclear regulatory context?

RISC-2 SSCs are identified and receive additional review and consideration given their risk significance.

For one plant, ventilation supporting equipment needed to support beyond design basis functions so additional analysis was performed to validate equipment would not be challenged.


Question: In order to maintain compliance with safety regulations, operational assessments are conducted to evaluate the safety margins of the plant's components.

The Integrated Inspection Report for Fermi Power Plant, Unit 2 includes evaluations of safety systems to ensure that they meet the required regulatory standards during operations.

The distribution list for the report contains various stakeholders, including NRC staff and other involved parties.


Question: What are the key competencies required for operators at nuclear power plants?

The competencies required for operators at nuclear power plants include interpreting and diagnosing events and conditions, complying with and using procedures, and demonstrating supervisory ability among others.

The North Anna Power Station is a crucial facility located in Virginia, operating under specific regulations to ensure safety and compliance.


Question: What is the Auxiliary Feedwater Pump reported action upon high stator temperature?

Identifies Stator temperature high (~ 350°F); notes rapid temperature rise (PPC).

Observes PK09-17, Motor Driven Aux FW PP alarming, and informs the Shift Foreman.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves multiple steps including the submission of safety and environmental reviews by the applicant, comprehensive assessments by regulatory bodies, and public comment periods to ensure adherence to safety standards.

The cooling system is designed to maintain the reactor temperature within specified limits while the plant is operational, ensuring efficient energy production.


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

COBRA-FLX is a thermal-hydraulic code used in the nuclear industry to predict critical heat flux (CHF) conditions in reactor coolant systems.

The analysis of the reactor’s behavior during various operational scenarios is crucial for ensuring operational efficiency and effectiveness.


Question: What is the purpose of the LCO 3.3.1 Conditions in a nuclear regulatory filing?

LCO 3.3.1 Conditions A, W, and X entered.

Loop 3 Tc RTD fails High. Requires entry into 1-AOI-2, Malfunction of Reactor Control System, Section 3.2, Instrument Failure.


Question: The NRC assesses the reliability of the safety measures implemented in the reactors.

The numerical data presented in this table reflects various parameters analyzed during structural evaluation of the tendons.

The total sum of measurements indicates a significant variance in structural integrity across different test scenarios.


Question: Measurements of groundwater levels indicate fluctuations over time, reflecting changes in environmental conditions.

Water Level Below Land Surface (ft) values are recorded on various dates, indicating the depth of water in relation to the ground surface.

The NRC uses various methods to ensure the safety and reliability of nuclear reactors during operation.


Question: The system utilizes boundary and initial conditions from the test apparatus to ensure accurate modeling.

The model is fed with boundary conditions and initial conditions from [1𝑏𝑐 as input conditions.

The equipment was calibrated to ensure that it meets performance specifications for reliability and efficiency.


Question: maximum rate of water withdrawal

| Southern Cotton Oil Co. |                              |                               |          |                   |          |           |


Question: high radiation area controls

♦ RP-AA-463, High Radiation Area Key Controls

**SYSTEM:** Generic Administrative – Radiation Control


Question: cladding corrosion limits

The cladding limits on corrosion thickness and hydrogen content are proposed in Section 3.1.1 for each of the cladding materials included in GALILEO application methodology.

Parameter uncertainty values in Table 5-9, a [ ] factor is used to calculate the one-sided 95th percentile values.


Question: The term used to describe the comprehensive document that details safety analysis is the Updated Final Safety Analysis Report.

UFSAR | Updated Final Safety Analysis Report

All other product and corporate names used in this document may be trademarks or registered trademarks of other companies, and are used only for explanation and to the owners’ benefit, without intent to infringe.


Question: environmental impacts of severe accidents

As relevant here, in the 2013 GEIS the NRC reached a generic determination that the environmental impacts of severe accidents would be “not significant,” or “small.”

The Staff performed a site-specific analysis of severe accident mitigation design alternatives (SAMDA) in a NEPA document for North Anna at the renewed operating license.


Question: nuclear reactor licensing process

The nuclear reactor licensing process includes a comprehensive evaluation of the reactor design, safety systems, and operational procedures to ensure compliance with regulatory standards set by the Nuclear Regulatory Commission (NRC).

The thermal performance of the reactor is crucial for maintaining optimal operational efficiency and ensuring that fuel is utilized effectively throughout the reactor core.


Question: expected dose calculation for accessing hatch

Applicant determines the following: EXPECTED DOSE to open hatch to access 1-FCV-63-72 (8 minutes/60 minutes/hour) x 4 rem/hr = 533 mrem. Step is critical to calculate expected dose.

The nuclear power plant's emergency preparedness program includes regular drills and training exercises to ensure staff readiness.


Question: Individuals seeking to maintain their certification as radiographers must submit renewal applications to the appropriate agency, detailed in section E.39a.

Applications for examination to renew a certification ID card shall be filed in accordance with E.39a.

All reciprocal recognition of licenses and registrations by the Agency will be granted in accordance with Part C of these regulations.


Question: How do we include load and residual stresses in our calculations?

We assume due to residual stresses, plus load stresses, that one and a half times the yield strain is what we need.

In fact, that calculation would follow a curve like this and, if I come down here, I can get a value of delta or delta Y which is correct, according to this finite element analysis.


Question: What is the role of the Reactor Engineer in the NAPS Emergency Plan?

The current NAPS Emergency Plan maintains the Revision 18 requirement for the Reactor Engineer position augmented at 60 minutes from an Alert or higher classification. The Reactor Engineer is responsible for supporting the technical team in the TSC in analyzing reactor physics, hydraulic, and thermodynamic problems and development of solutions.

Revision 18 of the NAPS Emergency Plan identified maintenance activities as being performed by personnel available on a 16-hour per day, 7-day per week basis, with coverage during remaining hours being provided as an ancillary duty of other on-shift personnel.


Question: nuclide activity measurement

The table provides the weighted mean activity of various nuclides, expressed in picocuries per gram (pCi/grams), along with their confidence levels and uncertainty measures.

The interference corrected report lists various confidence levels but does not detail how these affect safety evaluations or regulatory compliance.


Question: The geological materials identified in the boring logs include variations of sand and clay.

SAND (SW): gray, medium to coarse grained, trace mica, well graded, saturated


Question: What is the significance of the RCS/SG pressure mismatch alarm during plant operations?

PK06-06, RCS/SG PRESSURE MISMATCH alarms and clears several times in a minute and is now in alarm.

In accordance with OP1.DC10, Conduct of Operations, what action should be taken by the Shift Foreman?


Question: aging management programs for reactor systems

The following aging management programs manage the aging effects for the Reactor Recirculation System components: - ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1) - Water Chemistry (B.2.1.2) - BWR Stress Corrosion Cracking (B.2.1.5) - Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) (B.2.1.8) - Flow-Accelerated Corrosion (B.2.1.9) - Bolting Integrity (B.2.1.8) - Open-Cycle Cooling Water System (B.2.1.11) - Closed Treated Water Systems (B.2.1.12) - One-Time Inspection (B.2.1.20) - Selective Leaching (B.2.1.21) - ASME Code Class 1 Small-Bore Piping (B.2.1.22) - External Surfaces Monitoring of Mechanical Components (B.2.1.23) - Lubricating Oil Analysis (B.2.1.25)

The materials of construction for the Fuel Assemblies components are: - Not Applicable (Fuel Assemblies are short-lived components)


Question: isotope activity unit


Question: What is the intended function of a Spray Nozzle in fire protection systems?

Spray Nozzle	Spray	Copper Alloy (>15% Zinc)	Atmosphere/Weather (Ext)	Loss of Material	Fire Water System (B.2.3.15)

RCP Oil Collection Reservoir	Pressure Boundary	Carbon Steel (Galvanized)	Plant Indoor Air (Ext)	None	None


Question: nuclear regulatory licensing process

The licensing process for nuclear reactors involves a series of safety evaluations conducted by the NRC to ensure compliance with regulatory requirements before a reactor can begin operations.

The NRC holds regular public meetings to discuss various topics related to nuclear safety and regulation, providing a platform for stakeholder engagement and input.


Question: The NRC is assessing various methods for monitoring cable conditions over time.

Evaluate the effectiveness of commonly used cable condition monitoring methods

Completion of the LOCA Testing and final report will be by December 2021


Question: The operational guidelines stipulate that the 125-Vdc electrical power distribution must ensure independence and redundancy to maintain safety during potential failures.

Condition C represents one Class 1E DC electrical power subsystem and associated ESF equipment with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation.

If one of the required DC electrical power subsystems is inoperable for reasons other than Condition A or B (e.g., inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition.


Question: What actions can reduce thermal stratification in a reactor?

Which of the following actions will reduce thermal stratification without the possibility of exceeding RPV metal heat-up rate limits?

Unit 1 was in Mode 3 with the RHR Loop A of Shutdown Cooling (SDC) in service and ONLY the 1B Reactor Recirc pump running when the following occurred:


Question: The security controls in conjunction with Zero Trust Architecture ensure that no unauthorized access is granted to sensitive systems.

All controls should be incorporated into a part of the Zero Trust Reference Architecture, and no access control policy in the policy store should conflict with the controls implemented by the CSP.

Table 1 maps the security controls identified in section 2.2 to applicable ZTA components identified in section 2.1.


Question: What is the significance of 125V DC Vital Battery Bd III in a nuclear facility?

Loss of 125V DC Vital Battery Bd III can significantly impact the availability of major equipment necessary for safe reactor operation.

Generator Aux. Panel 2-L-39 is important for controlling various auxiliary functions within the nuclear facility.


Question: Licensees must provide detailed assessments of their reactor safety systems to comply with NRC regulations.

The licensee's request for amendment included evaluations demonstrating that the systems will perform their intended safety functions under design basis conditions.

The maintenance schedule for the reactor includes inspections and routine check-ups to ensure proper functioning of all systems.


Question: What is the preventive maintenance deferral process in nuclear regulation?

Preventive Maintenance Deferral Process refers to the method of identifying and managing risk-significant components and systems, ranked by risk worth; if the list uses system designators, provide a list of the associated equipment/system names.

Electronic copies of the Updated Final Safety Analysis Report (or equivalent), technical specifications, and technical specification bases are crucial for understanding plant safety and operational parameters.


Question: The NRC outlines the calculations necessary to determine the Probable Maximum Flood (PMF) for specific drainage basins.

The PMF in rivers and streams adjoining SQN is determined by applying the PMP to the drainage basin of these rivers and streams adjoining SQN.

Results of the PMF analysis for the two candidate storms, the 21,400 sq.-mi. storm and the 7,980 sq.-mi. Bulls Gap centered storm, are presented in Table 9.2-1.


Question: The saturation margin is critical to maintaining system safety under lowered pressure conditions.

The saturation margin will be decreased to approximately 20°F.

The list of Station Drawings below is provided for information only.


Question: What is Uranium Dioxide?

| UO2     | Uranium Dioxide                        |

M5 is a registered trademark of Framatome Inc. or its affiliates, in the USA or other countries.


Question: weighted mean activity of nuclides

The weighted mean activity of a nuclide is calculated by considering the activity of the nuclide per gram, allowing for a more accurate representation when comparing different isotopes.

The uncertainty in measurement can often arise from various factors, including the calibration of instruments and environmental conditions during data collection.


Question: The NRC provides criteria to evaluate design attributes that mitigate common cause failure in digital safety systems.

The BTP provides guidance for evaluating any diversity and defense-in-depth means credited to address vulnerabilities to CCF caused by latent defects in system hardware, software or software-based logic, as well as, the effects of any unmitigated CCF outcomes on plant safety.

This BTP is not applicable to proposed modifications performed under the change process in 10 CFR 50.59, "Changes, tests and experiments." Review criteria for single random failures and cascading failures from shared resources (i.e., not due to latent design defects in DI&C Structures, Systems and Components (SSCs)) are not covered in this BTP.


Question: What is the significance of scale in safety analysis reports for nuclear sites?

The scale of the topographic map is crucial as it provides a reference for distances and dimensions which are essential for safety analysis and regulatory compliance at the Clinch River Nuclear Site.

The topographic map includes details about various elevations and landmarks that are vital for environmental assessments.


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

The project requires various sampling methods to ensure accurate analysis.


Question: The surface derived concentration guideline levels (DCGLs) were established to ensure safety in areas where radioactive materials may have been present.

Based on the long-lived radionuclides of concern, the licensee selected the following surface derived concentration guideline levels (DCGLs), in units of disintegrations per minute per 100 square centimeters (dpm/100 cm²):

Using the guidance provided in NUREG-1575, revision 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), the licensee divided the various rooms and areas into two or three survey units, depending on location.


Question: Before an area can be released for unrestricted use, it must meet specific radiation standards regarding radionuclide concentrations.

Before release of an area to unrestricted use, NuFuels will provide information to the NRC verifying that radionuclide concentrations meet applicable radiation standards.

The preliminary schedule for ISR related activities and restoration has been discussed in Section 1 of this COP.


Question: A nuclear facility must obtain proper licensing to ensure compliance with safety regulations.

The licensing process for nuclear reactors involves a series of comprehensive evaluations by the NRC to ensure that the facility operates within established safety standards.

The analytical methods employed in this laboratory report are designed to determine the concentrations of various chemical elements in the samples.


Question: aluminum solubility correlation

An aluminum solubility correlation was used to determine a maximum precipitate formation temperature, which effectively delays the onset of aluminum precipitation.

Licensees proceeding from block 7 to diamond 10 in the Figure 1 flow chart should justify any deviations from the WCAP base model spreadsheet (i.e., any plant-specific refinements) and describe how any exceptions to the base model spreadsheet affected the amount of chemical precipitate predicted.


Question: What are the PCT and Burst Elevations in nuclear reactor safety analysis?

- **PCT Elevation** = 9.0 ft
- **Burst Elevation** = 8.25 ft

- **Time (s)**:
  - Ranges from 0 to 300 seconds.


Question: The review of the safety analysis report is critical for obtaining a nuclear license.

The Nuclear Regulatory Commission (NRC) reviews submitted safety analysis reports to ensure compliance with regulatory requirements before issuing licenses.

Radiation monitoring systems are essential for detecting and measuring ionizing radiation levels in and around a facility.


Question: thermal aging effects in stainless steel

The cast austenitic stainless steel used in the RCP casing is subject to time-dependent thermal aging during service. Thermal aging causes an elevation in the yield strength of the material, a decrease in fracture toughness, and a decrease in the crack initiation energy integral.

In the stability analyses, cracks are postulated at various locations in the pump casing. These postulated cracks would be subject to the various conditions the pump casing experiences.


Question: What are fixed air sampling locations in nuclear monitoring?

The State of Louisiana maintains two fixed air sampling locations around the River Bend Station.

Louisiana Department of Environmental Quality maintains the pre-designated sampling locations found in OP-4 in case of an emergency.


Question: containment leakage rate testing

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

The SFDP identifies where a loss of safety function exists.


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) report provides critical information on the minimum activity levels that can be reliably detected for specific nuclides, expressed in units of pCi/grams.

The unidentified peaks were measured with energy values of 351.87 keV and 583.31 keV, indicating the presence of certain isotopes during the analysis.


Question: The NRC mandates that licensees evaluate the projected values of the RTPTS whenever there are significant changes in anticipated neutron fluence levels.

10 CFR 50.61(b)(1) provides rules for protection against PTS events for PWRs. Licensees are required to perform an updated assessment of the projected values of the RTPTS whenever there is a significant change in projected values of RTPTS or upon request for a change in expiration date for operation of the facility.

These components are less than 1.0 x 10¹⁷ n/cm²; therefore, these components do not need to consider irradiation embrittlement and are not considered further.


Question: The NRC has set specific modeling guidelines that must be adhered to for plant-specific applications.

Plant-specific licensing actions referencing EMF 2103, Revision 3, analyses should include a statement summarizing the extent to which the guidelines were followed, and justification for any departures.

The EM is approved based on models that are specific to AREVA proprietary M5® fuel cladding. The application of the model to other cladding types has not been reviewed.


Question: The NRC standards mandate that detector efficiency must be calibrated to ensure accurate gamma spectrum readings.

Efficiency Calibration Used Done On: 7/8/2014


Question: What is meant by NRC Device Key in the context of nuclear regulations?

The NRC Device Key is a unique identifier assigned by the NRC to devices that may be subject to regulatory oversight, ensuring proper tracking and registration.

The manufacturer must provide documentation regarding the operational procedures for the safe handling of the device in accordance with industry standards.


Question: What is a nuclide in nuclear safety reports?

**Peak Locate Performed on:** 5/27/2019 11:21:22 AM


Question: This Draft Standard for Trial Use serves as a framework for constructing devices that utilize fusion energy technology.

This Draft Standard is not an approved consensus document and is issued as a Draft Standard of ASME for trial use.

Following the 2-year research and comment period, this Draft Standard, along with comments received, will be submitted to a consensus committee for the Project Team.


Question: What is the role of a safety analysis report in nuclear licensing?

A Safety Analysis Report (SAR) demonstrates that the design of the plant and its operation will not adversely affect the public health and safety.

The meeting will be held to review the findings of the latest seismic study conducted at the facility.


Question: The Spent Fuel Pool Cooling and Cleanup System is designed to effectively manage the temperature and clarity of the fuel pool water.

The Spent Fuel Pool Cooling/Cleanup System removes residual heat from the fuel assemblies and maintains the fuel pool water within specified temperature limits. It minimizes corrosion product buildup, controls water clarity in the fuel pool water so that the fuel assemblies can be efficiently handled underwater and minimizes fission product concentration in the fuel pool water.

The aging management review results for these components are provided in Table 3.3.2-26, Radwaste System - Summary of Aging Management Evaluation.


Question: What are the inspection procedures for non-power reactors?

The NRC requires extensive training programs for reactor operators to ensure safe and efficient plant operations.


Question: The NRC defines the conditions under which isotopes like TH-230 must be assessed for regulatory compliance.

TH-230 is a radionuclide that is monitored for its potential radiological impact and its regulatory requirements under NRC guidelines.

Analyzing the results from various samples can help determine the overall quality and reliability of the testing process.


Question: aging management review

The Aging Management Program for the Hatch Cover within the containment building involves monitoring the loss of material due to exposure to structural elements within the atmospheric environment, as specified in the Structures Monitoring Program (B.2.3.33).

Insulation materials such as Calcium Silicate are used primarily for thermal insulation but may also experience reduced thermal insulation resistance and moisture intrusion, necessitating monitoring by the External Surfaces Monitoring of Mechanical Components (B.2.3.22).


Question: PWR fuel assembly burnup values

The table outlines various operational parameters for reactor systems, including coolant flow rates and pressure settings.


Question: emergency core cooling system activation

Figure 3.1.3-2 depicts the divisional arrangement of the ECCS functions.

The arrangement of different regulatory bodies often impacts the approval timeline for nuclear projects.


Question: What is the meaning of Nuclide in the context of nuclear safety?


Question: The NRC evaluates license amendments that incorporate risk-informed approaches to safety.

The use of risk-informed completion times, as outlined in TSTF-505, Revision 2, provides a methodology for assessing the safety implications associated with extended completion times.

The reactor utilizes a pressurized water system to ensure effective heat removal from the fuel assembly.


Question: The NRC mandates the use of manual initiation pushbuttons in the ADS trip system to ensure safety protocols are met.

4.h. Manual Initiation (Two pushbuttons (PB) per ADS trip system)

Comments indicate that failure of Train A ADS SOVs to open is viewed as an equivalent surrogate for the analysis.


Question: The allowable concentration levels for Nickel and Uranium are specified as 6.8 mg/L and 1.6 mg/L, respectively.

**Nickel ACL** = 6.8 mg/L
**Uranium (Natural) ACL** = 1.6 mg/L


Question: What are the benthic invertebrates found near the MNGP site?

This table lists the benthic invertebrates in the Mississippi River in the vicinity of the MNGP site, including various classes and orders such as Class Oligochaeta and Order Ephemeroptera.

The NRC requires comprehensive environmental assessments for any proposed nuclear facility to evaluate potential ecological impacts.


Question: top nozzle design in nuclear reactors

The GAIA fuel assembly design incorporates Framatome’s standard top nozzle design. The top nozzle consists of two (2) high strength stainless steel bi-block frames, welded together to form a box-like structure. The upper structure interfaces with the reactor internals and the core components. The lower structure and grille flow-hole pattern is designed to balance low pressure drop and strength requirements.

The locking mechanism is designed to rotate 90° in either direction to lock or unlock, and provides a positive indication when rotation is complete.


Question: The licensing process for nuclear reactors includes comprehensive safety evaluations and public review phases.

The NRC's review process of license applications includes a detailed examination of technical specifications and safety features of nuclear reactors.

Regular maintenance of industrial equipment is essential for ensuring operational efficiency and reducing unexpected downtime.


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

The weighted mean activity (in picocuries per gram) is a crucial metric for each nuclide, reflecting its concentration and potential radiological impact, which is integral to regulatory assessments.

The confidence values for the specified nuclides indicate the reliability of the measurements, with higher confidence reflecting more precise assessment and validation of the data collected.


Question: What is a nuclide MDA in radiation detection?

Nuclide MDA (Minimum Detectable Activity) refers to the lowest level of radioactivity that can be reliably detected by a measurement system, typically expressed in picocuries per gram (pCi/grams).

The energy levels of gamma-ray emissions for various isotopes are crucial in radiation safety, although they do not directly measure radioactivity.


Question: water quality measurements in nuclear facilities

Westwater Canyon Aquifer Water Quality in the Unit 1 Project Area shows various constituents measured in mg/L such as Arsenic, Barium, and Uranium, among others.

The table also highlights how contaminants can vary significantly based on seasonal changes, which is crucial for continuous monitoring.


Question: What is the significance of the ACTIVE license status for nuclear reactor operators?

The ACTIVE/INACTIVE status and time on shift since April 1, 2022 is as follows for each of the ROs:

All perform off-shift assignments at the plant.


Question: COVID-19 regulatory exemptions for nuclear facilities

The NRC is providing compiled tables of exemptions using a single Federal Register notice for COVID-19 related exemptions instead of issuing individual Federal Register notices for each exemption.

These licensees have committed to certain security measures to ensure response readiness and for their security personnel to maintain performance capability.


Question: The NRC outlines that certain procedures related to radiopharmaceuticals must be reported under strict regulatory frameworks to ensure safety.

However, if even a tiny fraction of the millions of injections of diagnostic radiopharmaceuticals were to be classified as medical events, the time and resources that would have to be devoted to reporting them under the burdensome requirements of 10 CFR 35.3045(d) – (g) would be enormous and would contribute little to the health and safety of North Carolina residents.

Approximately 75 million radiopharmaceutical administrations are performed annually in the United States, many of them in North Carolina.


Question: biological assessment requirements

50 CFR 402.12 outlines the contents of a Biological Assessment, which includes results of on-site inspections, expert views, literature review, effects analysis on species and habitat, and consideration of alternative actions.

The NRC staff concurs with BA/BE, consultation letter, and NRC sends consultation package to the Service(s).


Question: The licensing process for nuclear reactors includes the assessment of safety measures and operational protocols.

The Nuclear Regulatory Commission (NRC) is responsible for ensuring that all aspects of nuclear facility operations are in compliance with regulatory requirements through rigorous licensing procedures.

The development of advanced materials has significantly influenced the efficiency of renewable energy systems, offering new possibilities for sustainable power generation.


Question: How is radioactive liquid waste treated at nuclear power plants?

The liquid radioactive waste treatment system consists of process equipment and instrumentation necessary to collect, process, monitor, and dispose of radioactive liquid wastes from the plant.

The reactor coolant-water purity will be maintained by means of (a) the makeup and purification, and (b) the chemical and boron recovery.


Question: The NRC sets specific limits for various water quality constituents in the context of environmental monitoring.


Question: The NRC mandates that licensees ensure all safety-related systems have adequate physical support and protection to comply with regulatory requirements.

Provides physical support, shelter, and protection for safety-related systems, structures, and components. 10 CFR 54.4(a)(1)

UFSAR References include specific sections that address the requirements for various safety-related systems and their evaluations.


Question: What procedures should operators follow when there are high seal leakoff flow alarms in RCP systems?

In accordance with AR PK05-01, what action should be taken by the Shift Foreman?


Question: What is the significance of low-income census block groups in regulatory filings for nuclear plants?

Total Number of Block Groups of Population Within 50-Mile Radius is 5,518.

The safety analysis must include all potential hazards and malfunction scenarios.


Question: nuclear reactor product code

All products manufactured by Gamma Industries, Inc. under the product code NR-0300-S-864-S belong to the T Series.

Gamma Industries, Inc. produces a variety of models, including those distinguished as R Series, such as Model R-5-A and R-12-G, that offer different specifications.


Question: The AREA methodology can only be applied to pressurized water reactor scenarios, particularly for control rod ejection accidents.

The AREA™ methodology is limited to the evaluation of a control rod ejection accident in a PWR.

The plants discussed in this report are equipped with advanced safety systems to mitigate potential risks during operation.


Question: The NRC emphasizes the importance of safety evaluations in the renewal of reactor licenses, particularly addressing environmental impacts.

The environmental review process for the subsequent license renewal of H.B. Robinson Unit 2 will assess environmental impacts and ensure compliance with regulatory requirements.

During the meeting, various stakeholders discussed strategies for improving operational efficiency and reducing downtime in maintenance operations.


Question: The NRC mandates the collection of samples when elevated radiation levels are noted during scans.

A total of two (2) alarms were verified during scanning, which prompted the collection of one (1) investigational soil sample and one (1) subsurface soil sample.

The implementation of survey specific QC measures included the collection of four (4) samples (L2-011-104-FQGS-003-SS, L2-011-104-FQGS-004-SS, L2-011-104-FSGS-003-SS SPLIT, and L2-011-104-FSGS-009-SS SPLIT) for split and duplicate sample analysis.


Question: What is the purpose of analyzing samples on a dry weight basis?

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

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


Question: isolation valves actuation requirements

This SR demonstrates that the CVS letdown isolation valves actuate to the isolation position in response to an actual or simulated actuation signal.

The Frequency of 24 months is based on the need to perform this surveillance during periods in which the plant is shutdown to prevent any upset of plant operation.


Question: What are the requirements for surveillance frequencies in nuclear operations?

Each time the example activity is to be performed, the Surveillance must be performed within 12 hours prior to starting the activity.

The logical connector “AND” indicates that both Frequency requirements must be met.


Question: minimum allowable core reflood rates

Extrapolation of data from cited experiments indicates that a new regulation that requires minimum allowable core reflood rates in the event of a LOCA is necessary to prevent Zircaloy cladding from exceeding the regulatory limit of 2,200 degrees F under certain conditions.

In the FLHT-1 test, the test conductors were unable to prevent a temperature excursion and runaway oxidation by increasing the coolant flow rate when peak cladding temperatures reached approximately 2,200 degrees F.


Question: What does the Pasquill-Turner calculation indicate about inversion occurrences?

Based on Pasquill-Turner calculation of E, F or G from hourly surface observations.

The Total row summarizes all occurrences by year, providing an overview of inversion events from 1955 to 1964.


Question: The measurements of Tritium samples indicate various radiochemical qualities expressed in pCi/g.

The Analyst is a Tracer compound which affects the interpretation of the results.


Question: The identification of various nuclides is crucial in nuclear material analysis.

Identified nuclides include K-40, PB-212, and BI-214, which are essential for understanding the radiological characteristics of materials.

The compliance with safety regulations is mandatory for all nuclear facilities to ensure operational integrity and public health.


Question: Reliability and Integrity Management in nuclear reactors

ASME Section XI has historically provided rules for in-service inspection (ISI) and the replacement and repair of components during the operating life of light water reactors.

Nuclear Power Plant component life management requires a combination of analysis, inspection, testing, and monitoring activities.


Question: aging management evaluations

Consistent with NUREG-1801. The DCPP Structures Monitoring AMP (B.2.3.33) will be used to manage cracking and distortion due to settlement.


Question: transient population projections

Transient population is projected to 40 years beyond the 2027 commencement of operation date for the last unit.

The largest of these events is the Smokin’ in the Water Celebration, which occurs over a two-day period during the July 4th holiday and contributes approximately 12,000 peak daily visitors.


Question: The procedures established for responding to emergencies must align with regulatory standards.

Emergency response procedures are essential for ensuring that safety protocols are followed during unforeseen situations.

The administrative process for filing regulatory documents involves several key steps and requires thorough documentation.


Question: Defense in Depth and Diversity principles in nuclear safety

The Defense in Depth and Diversity principles involve multiple layers of protection and the use of diverse systems to enhance safety and reliability in the case of failures or faults within a nuclear system.

The Westinghouse Non-Proprietary Class 3 documentation provides essential information regarding industry standards and practices, ensuring compliance with regulatory requirements.


Question: What are the requirements for evaluating coating damage in buried piping for nuclear facilities?

Revise implementing procedures to require that where damage to the coating has been evaluated as significant and the damage was caused by nonconforming backfill, an extent of condition evaluation will be conducted to determine the extent of degraded backfill in the vicinity of the observed damage.  **Program Element Affected: Element 7 - Corrective Actions**

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


Question: Standard Penetration Test method

Drive sampling by the Standard Penetration Test (SPT) method was conducted with automatic trip hammers, generally at 2.5-ft., 5.0-ft., and 10-ft. intervals.

Mud rotary wash boring equipment was used to advance boreholes pursuant to ASTM D 5783 (Reference 2.5.4-3-3), and to collect disturbed samples pursuant to ASTM D 1586 (Reference 2.5.4-3-7).


Question: sulfur dioxide standards in nuclear regulation

This figure illustrates the nonattainment areas for sulfur dioxide (SO₂) based on the 2010 and 1971 standards, highlighting regions within a 62-mile radius of the Dresden Nuclear Power Station Units 2 and 3.

The document outlines the overall operational capacity of the station and its contribution to regional energy production.


Question: What are the loading procedures for an OP-RMSC package?

This section delineates the procedures for loading the RMSC payload into the OP-RMSC package.

Remove the eight (8) hex bolts/washers and the closure lid from the OP-RMSC package body.


Question: low-low alarm in nuclear reactors

Actuation of the low-low alarm requires the operator to initiate emergency boration procedures unless the condition is due to a Unit runback.

Any unexpected change in the position of the control bank under automatic control, or a change in coolant temperature under manual control, provides a direct and immediate indication of a change in the reactivity status of the reactor.


Question: Americium-241 measurement


Question: Cs-137 and Co-60 characterization survey

Statistical Quantities for Cs-137 and Co-60 from the Characterization Survey indicate the mean concentration in surface soil samples for Co-60 is 0.045 pCi/g and for Cs-137 is 0.049 pCi/g.

The samples from asphalt exhibited a maximum concentration of Co-60 at 0.062 pCi/g and a minimum concentration of Cs-137 at 0.037 pCi/g.


Question: what is High Pressure Coolant Injection

The inspectors evaluated the following licensee event reports (LERs): 1. LER 05000333/2020-003-00, High Pressure Coolant Injection Inoperable Due to Oil Leak (ADAMS Accession No. ML2016161A405).

Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, "Light-Water Reactor Inspection Program - Operations Phase."


Question: The NRC requires that all licensees maintain comprehensive records of their safety systems and regulatory compliance efforts.

The NRC mandates that licensees submit detailed safety analysis reports to ensure compliance with all regulatory requirements.

Each department of the Army has specific procedures for equipment maintenance and operational readiness.


Question: The proposed update to the CMS5 Topical Report aims to adjust for higher burnup and enrichment levels of PWR fuels, which is essential for the ongoing improvement of reactor efficiency.

Supplement 1 will extend the range of applicability of the CMS5 Generic PWR Topical Report to higher enrichment, higher burnup pressurized water reactor (PWR) fuels.

This letter also outlines NEI's ongoing efforts to enhance communication and collaboration with NRC licensees to ensure compliance with regulatory requirements.


Question: What is the role of a Notary Public in regulatory filings?

Subscribed and sworn before me this 18th day of September, 2024.

Mr. George Carver, being duly sworn, deposes and says:


Question: The Xe-100 reactor must achieve criticality under specific conditions before entering steady-state operation.

It is expected that the fuel temperature coefficient of reactivity will differ in a fresh core or pre-equilibrium core compared to normal operating conditions due to different isotopic makeup of the fuel.

The total number of pebbles in the core is measured.


Question: What are the requirements for fail-safe testing of valves in nuclear facilities?

Subsection ISTC-3560 states that "Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuator power in accordance with the exercising frequency of ISTC-3510."

The fail-safe test is generally performed at the same frequency as the stroke time exercise test.


Question: gamma spectrum analysis

**Identification Energy Tolerance:** 1.000 keV

**Detection System:** Multiple detectors were evaluated to ensure accurate measurements across various frequencies.


Question: Emergency Action Level scheme

The amendment approves an acceptable EAL scheme that is required for operation of the facility.


Question: The NRC's review process for license amendment requests is closely linked to previously established license conditions.

As we stated in our brief, the exemption request and the license amendment requests are inextricably intertwined because the staff's position is that the exemption request must be approved in order for the staff to make the necessary findings to issue those amendments that would authorize power operations at Palisades since prohibition would remove -- the prohibition needs to be removed to allow power operations through the exemption request.

Okay. Now you have indicated that you consider the exemption request to be inextricably intertwined with the four license amendment requests.


Question: nuclear medicine injection extravasations

since we can help clinicians identify, characterize, and reduce the frequency of nuclear medicine injection extravasations, why would we not do that?

It is unbelievable to think we still have no tests to assure that what everyone thinks are routine PET scans actually work.


Question: Understanding how a constant enthalpy expansion through a valve influences downstream temperature is crucial for operational safety in pressurizer systems.

Leak by of 2PR1 is a constant enthalpy expansion process through a valve. Using steam stables and with PZR pressure at 415 psig (430 psia), the hg is approximately 1205 BTU / lbm. Using the Mollier diagram, the throttling process will go from 415 psig to 0 psig with hg (1205 BTU / lbm) remaining constant. This will correspond to a tailpipe temperature of approximately 330 °F.

Since the Mollier Diagram is very complex to read and interpret then any temperature between 212 and 450 °F is plausible. 212 °F is the boiling point for water / steam at 0 psig and 450 °F is boiling point for water / steam at 515 psig.


Question: remote inspection practices

Expand use of remote work practices for resident inspectors.

Revise ROP program documents to include best practices and guidance developed since the COVID-19 PHE began.


Question: The regulatory framework establishes requirements for the safe operation and maintenance of nuclear facilities.

Regulatory frameworks for nuclear power plants detail the necessary safety measures and compliance standards required for safe operations.

Documentation is essential for tracking the progress of the construction and commissioning of the nuclear facility.


Question: annual fees for reactor licenses

Each person holding an operating license for a power reactor, testing facility, or research reactor; each person holding a combined license under part 52 of this chapter after the Commission has made the finding under 10 CFR 52.103(g); each person holding a part 50 or part 52 power reactor license that is in decommissioning or possession only status, except those that have no spent fuel onsite; and each person holding a part 72 license who does not hold a part 50 or part 52 license and provides notification in accordance with 10 CFR 72.80(g), shall pay the annual fee for each license held during the Federal fiscal year in which the fee is due.

The FY 2019 annual fees for licensees authorized to operate a research reactor or testing facility licensed under 10 CFR part 50, unless the reactor is exempted from fees under § 171.11(b), are as follows:


Question: signal number converter


Question: DNBR values in safety analysis

Table 5-18 summarizes the state values for the Departure from Nucleate Boiling Ratio (DNBR) during the inadvertent opening of a pressurizer safety valve event.

The design of the reactor coolant system is critical for maintaining the overall safety and efficiency of the power plant.


Question: The NRC monitors the performance of monitoring wells using various metrics including specific conductance and total dissolved solids.

Monitoring Well 304-3, for example, reported a specific conductance of 8,247.3 µS/cm and a total dissolved solids measurement of 5,860 mg/L during its sample date.

The facility's discharge permit requires compliance with parts of the Clean Water Act.


Question: What is streamflow and how is it measured?

Streamflow (cfs) refers to the volume of water flowing through a specific point in a river or stream over time, typically measured in cubic feet per second (cfs).

H = Water surface elevation in El. ft msl


Question: Licensees are required to ensure accurate indication of valve operation through additional means such as flow meters during testing processes.

Mandatory Condition 12 for licensees intending to adopt the 2012 edition of the ASME OM Code reads as follows: “[xi] OM condition: Valve Position Indication. When implementing paragraph ISTC–3700, 'Position Verification Testing,' in the ASME OM Code, 2012 Edition through the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section, licensees shall verify that valve operation is accurately indicated by supplementing valve position indications with other indications, such as flow meters or other suitable instrumentation to provide assurance of proper obturator position for valves with remote position indication within the scope of Subsection ISTC including its mandatory appendices and their verification methods and frequencies.”

The information contained in these tables identify those valves that are required to be tested to the requirements of ASME OM Code, the test methods and frequency of testing, the associated relief requests, and other applicable information.


Question: What is the role of the heat exchanger in nuclear steam supply systems?

Heat Exchanger (Pressure Liquid Sample Cooler) Tube Side Components | Leakage Boundary (spatial) | Stainless Steel | Borated Water Leakage (Ext) | None | VII.J.AP-18

The aging management program must consider the effects of various environments on material degradation over time.


Question: Aging Management Audit

Throughout the Aging Management Audit, the staff will review the LRA, 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 staff requests for communication with the applicant will require approval of their respective technical branch chief and the NLRP branch chief.


Question: The NRC mandates that seismic analysis must be conducted for seismic hazard assessments at nuclear sites.

Seismic studies are essential for evaluating potential earthquake impacts on the structural integrity of nuclear power facilities.

Public Service Enterprise Group (PSEG) operates multiple generating stations, including Salem and Hope Creek, which contribute to the regional energy supply.


Question: What are the health risks associated with thorated tungsten electrodes?

Thorated tungsten electrodes contain thorium, a radioactive material that can pose health and environmental risks at elevated exposure levels.

Choose thorium-free tungsten electrodes such as those containing cerium, lanthanum, yttrium, or zirconium.


Question: The NRC specifies limits for reporting detectable levels of various substances in nuclear safety assessments.

RL: Reporting Limit indicates the lowest concentration of a substance that can be reliably detected in the analysis of samples collected from nuclear facilities.

The project aims to perform site restoration activities at the LaCrosse site, ensuring compliance with all relevant environmental regulations.


Question: What is the estimated PM10 emissions from vehicles on unpaved roads?

Operation Phase Total Estimated Fugitive PM10 Emissions: 17.38 tons/yr

The integrity testing unit typically operates under varying conditions to ensure optimal performance and reliability.


Question: What is a facility safety program (FSP) in nuclear regulation?

Establish a facility safety program (FSP) using a risk-informed, performance-based process to proactively identify new or revised hazards and performance issues.

Consider measures to mitigate or eliminate the resulting risks.


Question: What is the purpose of the Job Performance Measure (JPM) in nuclear training?

The Job Performance Measure (JPM) is designed to assess the capabilities and performance of operators in executing their roles effectively, specifically during training scenarios such as a SGTR event.

The NRC emphasizes the importance of regular maintenance checks for all safety-related systems to ensure their reliability during operations.


Question: What is the total surface area considered for debris generation analyses in reactor containment?

However, for conservatism, a total surface area of 200 ft² was assumed in the PBN1 and PBN2 debris generation analyses.

Labels, tags, stickers, placards and other miscellaneous or foreign materials were evaluated via walkdown.


Question: What is the role of the NRC in regulatory audits for nuclear facilities?

The U.S. Nuclear Regulatory Commission (NRC) conducts audits to assess compliance with safety and operational standards at nuclear facilities.

Virtual tours and meetings are scheduled to ensure effective communication among all parties involved in the project.


Question: What are advanced small modular reactors (SMRs)?

Advanced Small Modular Reactors (SMRs) are a type of nuclear reactor that are designed to be smaller than traditional reactors and can be constructed in factories and then transported to locations for installation.

The estimated use of water in the United States in 2015 provides an overview of water consumption across various sectors, indicating trends and usage patterns.


Question: What role does the Fire Department play in responding to radiological incidents?

Assisting in radiological monitoring and decontamination of emergency personnel and vehicles, if necessary.

The Police Departments in Port Allen, Brusly and Addis under the direction of the Chiefs of Police, will be responsible for generally maintaining law and order in the community, including the security of key facilities.


Question: The ongoing evaluation of operational experience is crucial for maintaining safety standards in nuclear facilities.

Ongoing review of operating experience is addressed in Appendix A, section A.1.6 and Appendix B, section B.1.4.

The applicant states: (a) why the program’s examination methods will be sufficient to detect the recurring aging effect before affecting the ability of a component to perform its intended function, (b) the basis for the adequacy of augmented or lack of augmented inspections, (c) what parameters will be trended as well as the decision points where increased inspections would be implemented (e.g., the extent of degradation at individual corrosion sites, the rate of degradation change), (d) how inspections of components that are not easily accessed (i.e., buried, underground) will be conducted, and (e) how leaks in any involved buried or underground components will be identified.


Question: What is the significance of corrosion data trending in nuclear plant inspections?

The program includes trending of corrosion data and the decision point where increased inspections would be implemented.

Components susceptible to RIC that are not easily accessed (i.e., buried) are included in the evaluation of inspection locations using risk insights based on susceptibility to aging effects and consequences of failure.


Question: The NRC mandates that licensees must provide documentation verifying that their systems and components meet safety requirements.

The NRC requires licensees to submit detailed technical specifications and safety analyses for nuclear facilities to ensure compliance with regulatory standards.

Krohne-Anteus, Inc. provides a variety of measuring instruments for industrial applications, specializing in flow, level, and pressure measurement.


Question: Category 1 findings in nuclear licensing

Category 1 issues, which are listed in 10 C.F.R. Part 51, Subpart A, Appendix B, Table B-1, include the environmental impacts of design basis accidents and severe accidents, including accidents caused by or contributed to by earthquakes.

By its title, Table B-1 constitutes a 'Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants.' It is a summary, not a prescription for applying Table 1 findings.


Question: What is an ITAAC and why is it important in nuclear licensing?

An Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) is a regulatory requirement published by the NRC that describes the necessary verification processes to ensure that a construction project complies with the regulatory standards and the conditions of the license.

The NRC conducts regular inspections to ensure that licensees follow the prescribed operating procedures to maintain safety and compliance.