Question: ice condenser isolation valves

K/A is matched because the applicant is required to have the ability to manually operate and monitor Ice Condenser isolation valves in the MCR.

The WBN Desktop Simulator is a crucial tool used for training and evaluating operator performance during normal and emergency conditions.


Question: What are the key components of a proposed audit agenda for nuclear regulatory filings?

The proposed audit agenda includes an entrance briefing with NRC and licensee introductions, a site visit, discussions of audit questions, and a summary of the audit in the exit meeting.

The reactor core consists of fuel assemblies that contain nuclear fuel, which is used to initiate and sustain the nuclear reaction.


Question: structural support evaluation for concrete and steel

Concrete (accessible areas): all Structural Support, Air - outdoor, Cracking due to expansion from reaction with aggregates, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.1.34)

Steel components: all structural steel, Structural Support, Air - outdoor, Loss of material due to corrosion, Structures Monitoring (B.2.1.33)


Question: licensed operator performance evaluation

The inspectors observed and evaluated licensed operator performance in the control room during power reduction and plant shutdown on October 5, 2024.

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:


Question: The regulatory framework mandates that all nuclear facilities adhere to strict licensing guidelines...

The licensing process for nuclear reactors involves an extensive review of safety analyses and technical specifications to ensure compliance with NRC regulations.

Routine inspections of non-nuclear materials are essential for maintaining workplace safety, but they do not directly affect reactor licensing.


Question: neutron irradiation embrittlement

Loss of fracture toughness due to neutron irradiation embrittlement and for CASS, martensitic SS, and PH SS due to thermal aging embrittlement; or changes in dimensions due to void swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear

Not applicable. DCPP is a Westinghouse plant, not a Combustion Engineering plant.


Question: The NRC assesses the impact of natural and human-induced flooding on nuclear facility safety.

Flooding is a significant concern for nuclear facility safety and operations, as it can affect structural integrity and operational capability.

The evaluation of debris loads during a flooding event includes quantifying the potential impact of floating materials on structures.


Question: audit team assignments

The audit team will consist of the following NRC staff from NRR, including contractors: Jason Drake, Plant Licensing Branch 4 (LPL4), Samson Lee, LPL4, Alexander Schwab, PRA Licensing Branch A (APLA), Mihaela Biro, APLA, and others.

To support the schedule established when the NRC staff accepted the LARs for technical review, audit activities will be performed remotely and virtually using Microsoft Teams, teleconference, and any web-based portals or meeting spaces created by the licensee.


Question: What is high pressurizer pressure in nuclear reactors?

High Pressurizer Pressure is a critical parameter that indicates the pressure within the pressurizer system of a nuclear reactor, which is essential for maintaining the proper operational conditions during reactor operation.

The High Under-the-Bioshield Temperature is a measurement that relates to the assessment of thermal conditions affecting structures within the reactor containment area, primarily focusing on the potential for high-energy line breaks.


Question: What does it mean for a well to be temporarily abandoned in regulatory terms?

A well that is classified as Temporarily Abandoned is not in active production, but it has not been permanently plugged and abandoned according to regulatory standards.

The NRC requires all licensees to submit an annual financial report detailing their expenditures and safety upgrades on the facility.


Question: What is the procedure for license expiration in nuclear facilities?

3.1 License Expiration: The license expires on the date specified in the license, unless renewed as provided in the regulations.

The responsibilities of the Manager, Radioactive Materials Program (RMP) include overseeing the implementation of policies and ensuring compliance with safety regulations.


Question: thermal discharge plume effects

The second approach used quantitative hydrothermal modeling to predict thermal conditions under various operating and ambient flow conditions, integrated with metrics of thermal requirements and tolerance limits identified in scientific literature for selected aquatic species representative of the BIC.

For this assessment, a three-dimensional mathematical model, MIKE 21, estimated ambient temperatures under various flow and thermal conditions representative of the ATLs.


Question: nuclear material safety and safeguards

The activities reported here were performed on behalf of the NRC Office of Nuclear Material Safety and Safeguards, Division of Spent Fuel Management.

The authors would like to thank Osvaldo Pensado for his technical review and David Pickett for his programmatic review.


Question: Fermi-2 LOCA weld bin assignments

Table 2-2 summarizes the Fermi-2 LOCA weld bin assignments, which are critical for assessing potential weld integrity during loss of coolant accidents (LOCA).

Table 5-3 lists the isolation-valve weighted failure probabilities, which provides a statistical overview of valve reliability under various operational conditions.


Question: What are the critical steps in nuclear licensing?

The reactor design must comply with all federal and state regulations.


Question: What are the conditions for using the online reference portal for NRC reviews?

The online reference portal will be password-protected, and passwords will be assigned to those directly involved in the review on a need-to-know basis; the online reference portal will be sufficiently secure to prevent NRC staff from printing, saving, or downloading any documents; and conditions of use of the online reference portal will be displayed on the login screen and will require concurrence by each user.

The NRC staff has determined that the audit plan contains proprietary information pursuant to 10 CFR 2.390, “Public inspections, exemptions, request for withholding.”


Question: results of the inspection in a regulatory context

The team presented the results of the inspection to Mr. B. Greene and other members of the NAC and Petersen staff at an exit meeting on December 9, 2021.

The CoC holder acknowledged the results presented and did not identify any of the information discussed as proprietary.


Question: nuclear reactor inspection report

The inspection examined activities conducted under your license as they relate to public health and safety, and to confirm compliance with the Commission’s rules and regulations and with the conditions of your license.

The enclosed report documents the results of this inspection, which were discussed on October 1, 2020, with you and members of your staff.


Question: licensing and regulatory requests

The U.S. Nuclear Regulatory Commission (NRC) staff is conducting an environmental review for a series of licensing and regulatory requests submitted by Holtec Decommissioning International, LLC that seek NRC approvals to reauthorize power operations at the Palisades Nuclear Plant (PNP).

The NRC will hold an informational meeting with Federally recognized Tribal governments on July 16, 2024, from 2:00 p.m. to 4:00 p.m. ET, to provide an overview of the NRC's environmental review.


Question: The NRC mandates that license renewal applications include updated safety analysis reports.

The initial final safety analysis report update due dates are essential for ensuring compliance with regulatory requirements during the license renewal process.

The manual load testing procedures must comply with industry standards to ensure the structural integrity of the facility.


Question: The pressure vessel design must account for various types of stresses, including radial, axial, and hoop stress, each affecting the structural integrity differently.

Figure 7-6 shows various graphical representations of temperature distributions across the heat exchanger. They illustrate the impact of flow rates on thermal efficiency.


Question: The NRC requires regular monitoring of groundwater to assess concentrations of specific contaminants such as molybdenum.

Molybdenum concentrations in well 32-45 KD-R generally continue to trend downward and were below the GPS in the sample collected in October 2024.

Propose removal from monitoring program in a future License amendment.


Question: nuclear reactor design criteria

The design criteria for nuclear reactors establish the technical requirements that structures, systems, and components must meet to ensure safe operation and protect public health and safety.

The maintenance of auxiliary systems is crucial for the overall performance of a power plant, as it ensures that all components function at optimal levels during operation.


Question: The regulatory framework mandates that nuclear licensees must perform thorough safety evaluations prior to operation.

Nuclear power plants must obtain a license from the Nuclear Regulatory Commission (NRC) before they can operate, which includes a detailed safety analysis and compliance with technical safety requirements.

In contrast, renewable energy sources, such as wind and solar, have different regulatory requirements that do not involve the same licensing process as nuclear power.


Question: nuclear regulatory filing process

The nuclear regulatory filing process requires that all documentation related to safety assessments, operational procedures, and compliance measures be submitted for review before any license can be granted.

The reactor's coolant system is designed to maintain optimal operating temperatures and ensure thermal efficiency during normal operations.


Question: What are the Limiting Condition for Operation (LCO) in nuclear technical specifications?

The proposed amendment would modify TS requirements in Section 1.3 and Section 3.0 regarding Limiting Condition for Operation (LCO) and Surveillance Requirement (SR) usage.

The enclosure provides a description and assessment of the proposed changes.


Question: The NRC's licensing process ensures compliance with safety standards for nuclear facilities.

The NRC evaluates license applications to ensure that nuclear facilities operate within strict safety guidelines.

The response to the FOIA request indicates that the information requested may be released entirely after review.


Question: static geotechnical properties of alluvium

Twenty one samples of Upland Complex alluvium were tested for static geotechnical properties. Water content in the alluvium samples varied between 9.9% and 25.5%, with an average of 19.2%. Dry densities ranged between 93.2 and 129.4 pcf, and averaged 106.0 pcf. These densities are typical for medium dense to dense silty and clean sands (Reference 203). PLs for alluvium samples ranged between 0% and 4% and LLs ranged between zero and 21%. This indicates that Upland Complex alluvium is essentially non-plastic.

The fines content of Upland Complex alluvium ranged between 4.7% and 92.9%, with an average of 12.1%, and the clay-size fraction determined by hydrometer analysis ranged between 2.8% and 10.6%. The hydrometer test results show that the fines in the Upland Complex alluvium consist primarily of silt-size material.


Question: The failure of MOV 5106 impacts the operation of the auxiliary feedwater system by preventing it from receiving an open signal.

MOV 5106 fails to open on demand due to a loss of DC electrical power on Train C, which results in no automatic or manual open signal being sent to the valve.

The reactor's cooling system is designed to ensure safe operation under various conditions, while continuously monitoring the temperature and pressure levels.


Question: The pressurizer is designed to maintain the pressure of the reactor coolant system within specified limits during normal and transient conditions.

The pressurizer (PZR) is a component of the reactor coolant system and is used to maintain the pressure in the system, especially during changes in temperature and phase of the coolant.

The diagram shows the relationship between time and the various points of pressure in the system without further context or technical detail.


Question: aging management review

Aging Management Review Results are evaluated to identify the effects of aging on various structural components and systems, ensuring continued safety and reliability throughout the lifecycle of a nuclear facility.

Concrete components are often used in construction due to their high compressive strength and versatility in various environmental conditions.


Question: wind speed frequency distribution

The table presents the monthly and annual joint frequency distributions of wind speed and direction by atmospheric stability class for the VEPCCO North Anna site, specifically at the 150-foot elevation level.

The reactor core is designed with multiple layers of containment to ensure safety and prevent any release of radioactivity during normal operations.


Question: fire protection program

The inspection will verify that plant structures, systems, and components, and/or administrative controls credited in the approved fire protection program can perform their licensing basis function.

The enclosure lists the types of documents that will be needed prior to the information gathering visit.


Question: What is the purpose of the RHR pump in a nuclear reactor?

The RHR (Residual Heat Removal) Pump is utilized in nuclear reactors to facilitate the removal of residual heat from the reactor core during shutdown conditions.

The test pressure direction for valve NI961 is set at a specific level to ensure operational reliability under normal conditions.


Question: What are the investigation levels for Class 2 open land survey units?

For this Class 2 open land survey unit, the “Investigation Levels” for area scanning and soil sample measurement results are those levels specified in the LTP, Table 5-16, and are reproduced below in Table 5-5.

Table 5-6 provides a synopsis of the survey design for survey unit L2-011-101.


Question: Are senior management involved in overseeing the radiation safety measures of the licensee?

A. Is senior licensee management appropriately involved with the radiation protection program and/or RSO oversight?

Summarize problems and/or deficiencies identified during the audit.


Question: The average burnup of WE 14x14 assemblies can vary based on the enrichment levels of U-235.

WE 14x14 Assembly Average Burnup (B) GWd/MTU is a crucial factor that reflects how much energy has been extracted from the fuel per unit mass during its operational lifetime.

The regulations provide guidelines on the maintenance of reactor coolant pumps, emphasizing the need for regular inspections.


Question: Each boring was sampled based on geological data and the final depths were well documented for each location.

The planned maximum depth for each boring location was 15-ft bgs. This depth was based on the MU provided geological data which indicated that perched water resides around 15 ft bgs.

Four discretionary 1-ft soil samples were collected; one each from boring locations PKH-009, 011, 014 and 017 at the 10-11 ft bgs sample section.


Question: SW System operability requirements

If one SSW pump in one SSW subsystem is inoperable, it must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE SSW pumps are adequate to perform the heat removal function. The 30 day Completion Time is reasonable, based on the low probability of an accident occurring during the 30 days that one SSW pump is inoperable, the number of available SSW pumps, and the time required to complete the Required Action.

The function. The 72 hour Completion Time for restoring the SSW subsystem to OPERABLE status was developed taking into account the redundant capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.


Question: What is the procedure for aligning chemical and volume control system makeup pumps in a nuclear reactor?

EK 3.26  Aligning chemical and volume control system makeup pumps to maintain pressurizer level                       2.8

EK 3.11  Blocking streamline/feedwater isolation actuations and/or safeguards actuation below P-11, Pressurizer Pressure below 1,970 psig          3.4


Question: What is the purpose of self-diagnostic functions in nuclear reactor systems?

The LGS PPS application software contains self-diagnostic functions that are carried out within the ITP PM646A processor modules and the MTP. There are many self-diagnostic functions that monitor the system for errors and report these errors up to the ITP to initiate a Division Fault Alarm.

Alarm annunciation by the DCS when it receives a PPS Division Fault Alarm over the AOI data link.


Question: Licensees must prepare various safety analyses in accordance with the regulatory requirements set by the NRC.

Under the regulations of the Nuclear Regulatory Commission (NRC), a licensee must submit a safety analysis report that details the design, operation, and safety of the nuclear facility.

The cooling system must function properly to maintain adequate temperatures in the reactor core during operation.


Question: auxiliary feedwater flow rate

The impact of increasing this auxiliary feedwater flow rate to 970 gpm was subsequently evaluated. That evaluation confirmed that with this increase in auxiliary feedwater flow, the results of the analysis for main steam line break in containment would still be within the acceptance criteria.

Moody contraction coefficient is a dimensionless parameter used to characterize the friction loss in fluid flow through a conduit.


Question: What is the purpose of an Environmental Impact Statement (EIS) in nuclear regulation?

The purpose of the EIS is (1) to ensure that the agency "in reaching its decision, will have available, and will carefully consider, detailed information concerning significant environmental impacts" and (2) to guarantee "that the relevant information will be made available to the larger audience."

This is in keeping with Federal Court’s descriptions of NEPA’s limited requirements.


Question: What is the function of the 125V DC Vital Battery Board III in a nuclear plant?

DISPATCH personnel to restore power USING O-SOI-236.03, 125V DC Vital Battery Board III.

**PLACE** 1-XSW-46-DC, T-D AFW PUMP DC MANUAL TRANSFER SWITCH to EMERGENCY [A2T692].


Question: The ORFEO-HMP critical heat flux correlation has specific applications within defined pressure and mass flux limits.

Table 7-16 provides the range of application for the ORFEO-HMP correlation with the Pressure-Velocity solver, detailing local thermal-hydraulic conditions like pressure and mass flux.

Framatome Inc. engages in the design and supply of components for nuclear power plants globally, providing innovative solutions in the energy sector.


Question: What are the requirements for equipment reliability in safety systems?

The Safety Systems at Cooper Nuclear Station are designed to ensure that critical equipment is capable of performing its safety function under design conditions, with an expected reliability of 99.9%.

The documentation for Cooper Nuclear Station includes details about its operational history and maintenance schedules, which are crucial for assessing overall plant performance.


Question: The NRC mandates that emergency ventilation systems must effectively remove contaminants from the air in controlled environments.

The auxiliary building emergency ventilation system parameters (for each of two trains) include a recirculation flow of 13,950 ft³/min and a discharge flow of 2970 ft³/min, with an elemental iodine filter efficiency of 90%.

The mass of water in the containment sump is measured to be 6.77 x 10^6 lb, which is critical for ensuring the containment integrity during operational scenarios.


Question: The NRC demands rigorous analysis of spectral data during the licensing process.

**Energy Calibration Used On:** 7/8/2014


Question: What is the role of Command and Control during nuclear emergencies?

an analysis or study, that demonstrates there will be no adverse impact on on-site or off-site emergency response if Command and Control augmentation is extended to 90 minutes.

reactor licensees to maintain continuous communications with the NRC when they are requested to do so.


Question: The document discusses the reliability of the Auxiliary Feedwater System (AFWS) and its components in relation to fault tree analysis.

In accordance with NUREG-0611, fault tree analysis was utilized to model and quantify the reliability of the AFWS.

A review of the generic AFWS reliability analysis revealed the generic analysis consisted of: A. Quantifying a point estimate of the probability of failure on demand of the AFWS.


Question: The NRC emphasizes the importance of maintaining multiple layers of safety to prevent system failures.

Defense in depth is a fundamental safety principle that uses multiple layers of protection to prevent accidents and mitigate their consequences.

The document discusses various modernization activities and their impact on operational efficiency without addressing safety principles.


Question: The NRC mandates that licensees conduct thorough evaluations of potential safety issues related to reactor operations.

Clinton Power Station (CPS). Clinton Power Station Updated Safety Analysis Report. Revision 10. 2002.

Bureau of Transportation Statistics. National Transportation Atlas. Airport Locations. 2000.


Question: nuclide activity measurement

The Wt mean Activity (pCi/grams) provides a standardized measure of the radioactivity of a nuclide with a specified uncertainty, allowing for comparison across different isotopes.

The table shows various elements along with their respective chemical properties, but does not provide specific details about any particular nuclide.


Question: The NRC mandates that a timely response is crucial, often requiring licensees to maintain open lines of communication during emergencies.

The NRC may request an open line be maintained. An additional communicator (preferably an RO or SRO) may be assigned to provide continuous updates to the NRC under the following circumstances: If an additional qualified communicator is available AND is not required for access to mitigate the emergency (higher priority activities in the judgment of the IC).

It is important to ENSURE it is accurate and appropriate, and always LAW.


Question: nuclear power licensing process

The licensing process for nuclear power plants involves several key regulatory steps including the application for an operating license, safety assessment, and public hearings.

The performance rating of the applicants was assessed based on their knowledge of various job-related tasks and their ability to respond to emergency scenarios.


Question: nuclide identification

Nuclide identified during the nuclide identification is indicated with a '+' sign.

Energy values can be expressed in keV without reference to their application in nuclear safety or measurement.


Question: The SEB penalties considered during RICT calculations depend on seismic risk assessments, which are crucial for safety evaluations.

If a RICT is being entered during a period when the primary containment is de-inverted, a different SLERF penalty of 8.00E-06/yr (SCDF = SLERF, SCLERP = 1.0) will be applied to address the increased potential for hydrogen deflagration events in the primary containment.

Some primary containment penetrations use MOVs for containment isolation which would require electric power for closure and for an isolation signal.


Question: Licensees must ensure compliance with safety regulations for nuclear facilities.

Nuclear reactor operators must submit periodic safety reports to the NRC as part of their licensing obligations.

The primary purpose of a pressure vessel is to contain liquids or gases at a pressure substantially different from the ambient pressure.


Question: public comment period regulations

Because the public comment period does not close for another six weeks, the staff will not be discussing any public comments received to date, nor can we discuss the staff's thoughts about potential changes to the final rule at this time.

I'm one of the technical leads for the key NRC staff.


Question: radioactivity monitoring locations

Gamma - (TLDs) one between IX columns and sand filters. Radon Progeny - two at the IX and one at the sand filter.

Additional monitoring will be conducted or eliminated at the RSO’s discretion.


Question: What is the purpose of the fuel pit cooling and refueling purification system?

The fuel pit cooling and refueling purification system removes the residual heat from spent fuel stored in the spent-fuel pit and purifies the water in the reactor cavity and spent-fuel pit.

The fire protection system furnishes water and other extinguishing agents with the capability of extinguishing any single or probable combination of simultaneous fires that might occur at the station.


Question: What are the challenges associated with the safe use of advanced reactor fuels?

NMSS/DFM is proactively identifying potential technical challenges and information needs associated with the safe use of AR fuels in the areas of enrichment, fabrication, transport, and storage.

We have experience with advanced technologies with regards to fuel cycle and transportation licensing, and we have the regulatory infrastructure in place.


Question: The requirements for timely notification of emergencies are critical in ensuring public safety during potential incidents.

The Alert and Notification System evaluates the effectiveness of communication channels for disseminating warnings and instructions during emergencies at the River Bend Station.

The Mayor of a town plays a crucial role in local government administration and public services.


Question: The relationship between the shear stress ratio and the number of cycles is significant for assessing material performance under load.


Question: technical justification for regulatory exemptions

The technical justification for the exemption is based on a thorough analysis of the safety implications and operational challenges, demonstrating that the requested exemption meets the criteria of 10 CFR 50.12(a)(1) and is in the public interest.

The facility is equipped with several systems designed for emergency response, ensuring that a robust framework is in place to handle potential crises.


Question: The NRC requires accurate quantification of radioactive isotopes within regulatory submissions.

K-40 is measured as an activity concentration, displayed in picocuries per gram (pCi/g Dry), indicating the level of radioactivity present in a sample.

The report also flags values based on whether the activity concentration exceeds the Minimum Detectable Concentration (MDC), affecting compliance reporting.


Question: What are the hydrologic characteristics affecting the PSEG Site?

The Delaware Estuary is comprised of the Delaware Bay and the tidally influenced portion of the Delaware River.

Average annual precipitation in the Delaware River Basin ranges from 42 inches (in.) for southern NJ in the lower basin, to 50 in. for the Catskill Mountains of southern NY in the upper basin.


Question: The goal is to provide users with tools that enhance their ability to access and submit regulatory information effectively.

Empower stakeholders with data analytical tools, including a secure, flexible, and personalized self-service gateway to retrieve, submit, and/or interact with regulatory related information and assessment tools through a standard web browser.

The system architecture is designed to optimize database performance and ensure data integrity during transactions.


Question: It is critical to ensure that all emergency diesel generators are operational during a tube rupture event to maintain system integrity.

The strips of paper were carefully stacked in the laboratory for analysis and later examination.


Question: ambient background in nuclear licensing

The ambient background level refers to the natural radiation present in the environment, which must be considered when measuring radioactive emissions from nuclear facilities.

The net peak area represents the calculated area under the peak in a spectrum, indicating the quantity of a specific isotope detected.


Question: The NRC oversees the amendments and modifications made by licensees to ensure compliance with safety regulations.

The NRC requires licensees to submit amendments to their licenses for approval as part of the regulatory process in maintaining compliance.

The applicants in the document are located in various cities across Texas and Virginia, indicating a wide geographic distribution of licensees.


Question: Nuclear power reactor operators must prepare exercise scenarios to ensure emergency preparedness testing is effective.

Nuclear power reactor licensees shall submit exercise scenarios under § 50.4 at least 60 days before use in a full participation exercise required by this paragraph 2.a.

In addition, a radiological orientation training program shall be made available to local services personnel; e.g., local emergency services/Civil Defense, local law enforcement personnel, local news media persons.


Question: The document provides data on the precipitation levels measured at the MNGP over a period of many years.

The maximum monthly precipitation recorded was significant in determining the operational conditions of the reactor.


Question: What actions should be taken when the Reactor Protection System fails to signal a reactor trip?

The applicant is required to predict the impacts of the failure of the RPS signal to trip the reactor, and based on those predictions, use the procedure to mitigate the consequences of the event.

Given a set of plant conditions, use FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly DIAGNOSE and implement: Action Steps, RNOs, Notes and Cautions.


Question: minimum detectable concentration in nuclear analysis

No = Peak not identified in gamma spectrum


Question: operable status for UPS inverters

Four Class 1E Vital 120 V UPS inverters shall be OPERABLE.

SR 3.8.7.1 Verify correct inverter voltage and alignment to required AC vital buses.


Question: ARITA final Technical Evaluation Report

The Assessment of Regulatory Issues and Technical Approaches (ARITA) final Technical Evaluation Report (TER) provides a comprehensive review of the technical aspects and regulatory considerations related to the licensing of nuclear reactors, aiming to ensure safety and compliance with NRC standards.

As I work on the documentation, I will consider the implications of the timeline between submittal and contract closeout for our upcoming projects.


Question: cooling system types

DNPS operates a cooling system in three modes: closed cycle, direct open cycle, and indirect open cycle.

Compliance with regulatory requirements and the NPDES permit would ensure that water use conflicts from the continued operation continue to be limited to a minimal impact.


Question: What is the significance of the Unconfined Aquifer System in hydrogeologic studies?

The shallowest aquifer system is the Unconfined Aquifer System which is approximately 30 ft to 50 ft thick in the vicinity of the site.

The hydrogeologic properties of Unit IV have been summarized from a considerable number of Tuscaloosa wells in southeastern Richland County.


Question: low head safety injection

the two low head safety injection (LHSI) pumps. The water is continuously recirculated through the containment to remove heat from the reactor core and containment atmosphere and radioiodine from the containment atmosphere.

The containment sump is a depressed area in the floor to hold water and to provide suction points for the four recirculation spray pumps and, after commencement of the recirculation mode of safety injection.


Question: What is an incomplete SCRAM in nuclear reactor operations?

Incomplete SCRAM is a condition that occurs when the reactor has not fully responded to a reactor shutdown command, potentially resulting in unsafe reactor conditions or failure to achieve critical safety responses.

The ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.


Question: The NRC requires the licensee to submit an Integrated Safety Analysis (ISA) summary for the upcoming year.

This document is a submittal of the Integrated Safety Analysis (ISA) Summary Annual Update for Framatome Inc.'s Richland, Washington Fuel Fabrication Facility; License No. SNM-1227; Docket No. 70-1257.

The attachment lists various electronic files and their designations that are part of the submission.


Question: What is the aging management program for buried and underground piping and tanks in nuclear facilities?

Loss of material of buried steel (with or without coating or wrapping) piping and piping components, as well as the diesel engine generator fuel oil storage tanks, is managed by the DCPP Buried and Underground Piping and Tanks AMP (B.2.3.26), which takes exception to NUREG-1801.

Steel piping and piping components encased in concrete in the MWS, Saltwater and Chlorination System, and Oily Water and Turbine Sump System is included in item 3.3-1, 112.


Question: Early Warning System regulatory approval

The license amendment request (LAR) proposes alternative security measures for the implementation of the Early Warning System in accordance with 10 CFR 73.55(r).

Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) submitted Reference 1 that requested approval of a proposed amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Diablo Canyon Power Plant Units 1 and 2, respectively.


Question: The Division of Public Health oversees various licensure activities to ensure compliance with health standards.

The Nebraska Department of Health and Human Services, through its Division of Public Health, manages the licensing of health facilities to uphold public safety and health standards.

The Deputy Director of the Licensure & Environmental Health Section handles multiple health-related administrative tasks, but not specifically concerning facility licensing.


Question: What are the requirements for maintaining canal level probes in a nuclear facility?

A safety-related, non-missile protected temporary canal level probe will be required to maintain the minimum number of canal level probe channels in accordance with TS 3.7.

An internal flooding walkdown will be performed after installation of the temporary SW jumper.


Question: What is the maximum discharge during a storm event for nuclear safety evaluations?

The maximum elevation recorded was 8 feet.


Question: The expected number of personnel working on the site during construction phases is crucial for planning and safety assessments.

Average Daily Workers during operation is estimated at 75, which is crucial for safety monitoring.

The Decommissioning phase includes specific measures to ensure that all hazardous materials are handled according to regulatory standards.


Question: What is the role of the Medical Support Facilities Communications Systems in emergency scenarios?

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

The State Preparedness Network Radio System serves as a backup communications system to the RBS Internet-based Electronic Notifications and State and Local Hotline.


Question: cumulative dose in nuclear safety reports

The control room data reflects operational metrics that are critical for system performance assessment, but they do not directly relate to the measurement of radioactive I-131.


Question: The NRC assesses the impact of various assumptions regarding the operation of safety systems in their filings.

The flow diversion through SI-857A, Train A RHR to Train A SI Pump Crossover, and SI-857B, Train B RHR to Train B SI Pump Crossover, is not modeled in the fault tree.

This modeling assumption does not impact the risk-informed GSI-191 application.


Question: soil classification tests in nuclear regulatory processes

The scope of the static laboratory program included soil classification, index, shear strength, and consolidation testing.

The safety analysis uses the test data to address geologic hazards and engineering design characteristics.


Question: The NRC stipulates that a minimum of two sets of two Manual Trip Buttons must be in an operable state for compliance.

Two sets of two Manual Trip Buttons shall be operable

Action 6.a: With one CEAC inoperable, restore in 7 days (or take additional actions) (TSTF-505 does not apply a RICT to STS 3.3.3, Required Action A.2)


Question: What is the scoping process in Section 106 consultations for nuclear plants?

The scoping process is a key component of Section 106 consultations, where the NRC initiates engagement with relevant stakeholders to identify the potential effects of a proposed undertaking on historic properties.

The attachments include detailed correspondence and various engineering assessments related to the operation of the Palisades Nuclear Plant.


Question: What are the requirements for analyzing hazardous waste characteristics?

The characteristics of hazardous waste must be identified according to RCRA regulations, including factors such as arsenic and cadmium levels.

The sample ID includes details about the date collected and time collected in military format.


Question: fuel oil chemistry in nuclear reactor plants

The DCPP Fuel Oil Chemistry AMP is an existing mitigative and condition monitoring AMP that manages loss of material in tanks, components, and piping exposed to an environment of diesel fuel oil.

The tanks within the scope of this AMP are listed below: Emergency diesel fuel oil storage tanks, Emergency diesel fuel oil day tanks, Portable diesel electric generator fuel oil tanks, Portable diesel driven fire water pump fuel oil storage tanks, Emergency diesel fuel oil pump head (priming) tanks.


Question: The NRC is focusing on the development of a structured framework for the extraction of aged materials from nuclear facilities as part of their ongoing research.

Given the expected availability of materials from numerous plants and anticipated research needs to better understand aging out to 80 years of operation, the NRC is pursuing a more proactive approach to prioritize the data needs best addressed by harvesting and identify the best sources of materials to address high-priority data needs for regulatory research.

Economic challenges and limited budgets have restricted the resources available to support new research, including harvesting programs.


Question: The safety assessment includes evaluating potential debris transport to the emergency systems following a loss-of-coolant accident.

The ECCS Sump Strainer is designed to prevent debris from entering the Emergency Core Cooling Systems and the reactor core, ensuring long-term core cooling and containment cooling are maintained.

The strainer modules are made of two concentric cylinders formed from perforated plates for straining debris from the water.


Question: Remote Shutdown System operability

Condition A addresses the situation where one or more required Functions of the Remote Shutdown System listed in Table 3.3.4-1 or one or more required ASP controls are inoperable.

Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred.


Question: What is a Technical Support Center in the context of nuclear facilities?

The existing TSC has been selected as the centralized location for the new NSR Controls Platform due to its close proximity to the MCR and the existing PCS.

The proposed new TSC will also provide enhanced power capabilities with a dedicated diesel generator backup with auto transfer on loss of normal power which works in conjunction with an uninterruptible power supply (UPS) to ensure continuity of TSC operations.


Question: large break LOCA

The design temperature for the strainers is 268°F, which is the maximum calculated containment sump water temperature during a large break LOCA.

Finally, guidance was also taken from AWS D1.6, “Structural Welding Code - Stainless Steel” (Reference 38) as it relates to the qualification of stainless steel welds.


Question: The evaluation of the initial conditions of the underground water to assess its safety and regulatory compliance.

Determination of Baseline Ground Water Quality.

Despite the requirements in section 2.309(f)(1), she failed to include an explanation of the basis for her contention; a statement of the facts, opinions, or documents on which the contention is based; an explanation as to how the contention is within the scope of this proceeding; an explanation as to how this issue is material to the findings the NRC must make; or an explanation as to how this shows a genuine dispute with the Combined TR/ER.


Question: In accordance with 10 CFR 50.91, the licensee must notify the State regarding any amendments to their license applications.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using the criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration.

The affidavit sets forth the basis on which GNF's information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4), "Public inspections, exemptions, requests for withholding."


Question: containment enclosure building integrity

Containment Enclosure Building Integrity 18 Month Surveillance

Bus E5 and E6 18 Month Offsite Power Supply Transfer Operability Test


Question: wt mean activity nuclides

The table lists several nuclides along with their corresponding weighted mean activity in picoCuries per gram (pCi/grams).

Comments associated with the nuclides include various identifiers such as '?', 'X', and '@', which provide insights into their analyses.


Question: What is the significance of meteorological evaluations in nuclear project licensing?

A meteorological evaluation, completed as part of the license renewal and provided in Addendum A, demonstrates that wind data from Gallup is representative of the project areas.

Based on the information presented in this section and as summarized in Table 1.1-1 of this ER there is no significant change between the CUP FEIS and this ER in regard to meteorology, climatology and air quality.


Question: The heat transfer coefficient for the pressurizer fluid in the finite element model can be described by HtBF, which is crucial for understanding fluid dynamics within the reactor system.

- HtBF is the HTC for the pressurizer fluid on the pressurizer portions of the finite element model other than the surge nozzle and the pool region at the bottom of the pressurizer.

The cooling system in the reactor is designed to efficiently lower the temperature during normal operation, ensuring stable conditions.


Question: cumulative risk assessment in nuclear licensing

The total average annual change in risk for extended CTs will be compared to the guidance of RG 1.174, Revision 3, Figures 4 and 5, acceptance guidelines for CDF and LERF, respectively.

Also, Revision 3 of RG 1.174 was issued by the NRC in January 2018 and the relevant guidance regarding the implementation and monitoring program remain substantially the same.


Question: What actions are required for operator response in case of an RCS to Thermal Barrier CCS leak?

ENSURE CCS, CVCS, and Locked Valves and Breakers are returned to normal alignment USING the following Instructions as necessary: - 0-SOI-70.01, Component Cooling Water (CCS).

14. WHEN repairs are complete, THEN: a. ENSURE Surge Tank Vents are OPEN: - 1-FCV-70-66 - 2-FCV-70-66.


Question: The NRC evaluates extreme weather conditions during reactor safety assessments.

Only hailstorms with hailstones of size 1.75 inch diameter and larger are included.

Near Smyrna, DE             | April 1, 1993


Question: The NRC assessed the potential effects on populations during the construction of two new ALWR units based on environmental and health factors.

The NRC evaluated the economic impacts from construction of two ALWR units with a total net electrical output capacity of 2,234 MWe at the W.S. Lee Nuclear Station site in Section 4.5 of NUREG-2111 (NRC 2013-TN6435: pp. 4-98 through 4-102).

Potential impacts to minority and low-income populations from operations of a new nuclear alternative would include those common to all replacement power alternatives discussed in Section 3.12.4.


Question: What is the applicability of LCO 3.0.4.b in relation to safety systems?

LCO 3.0.4.b is not applicable to ECCS high pressure safety injection subsystem.

The plant operator must ensure that all systems are functioning correctly during normal operations to maintain safety.


Question: The verification of ITAAC completion is integral to ensuring compliance with the acceptance criteria outlined in regulatory filings.

If the reviewers have verified that the licensee successfully completed the inspections, tests, and analyses prescribed for this ITAAC, and that the acceptance criteria have been met then this determination is based on information available at the time and is subject to the licensee’s ability to maintain the condition that the acceptance criteria are met.

An additional reviewer was assigned to the ITAAC closure review with his or her name entered into the blank at the top of this page based on his or her expertise being required.


Question: What defines the nonfuel hardware in spent nuclear fuel assemblies?

PWR UNDAMAGED SNF ASSEMBLIES may contain NONFUEL HARDWARE. SNF assembly lattices possessing less than the nominal number of undamaged fuel rods specified in Table B2-3 must contain solid filler rods that displace a volume equal to, or greater than, that of the fuel rod that the filler rod replaces.

Fuel assemblies may contain any number of unirradiated (i.e., not inserted in-core) nonfuel solid filler fuel replacement rods.


Question: The NRC expects to charge fees for licensing actions such as issuing new licenses, renewing existing ones, or making amendments to them.

Additional revenue could be gained through licensing actions such as application for a new license, renewal, or amendment.

The overall increase in revenue would be 1.2 million dollars per year (1.3 in current NRC licensing fees minus $100,000 in fees from Connecticut state agencies), while costs would be approximately $500,000 per year (estimated staff costs), resulting in a yearly transfer of approximately $500,000 per year to the General Fund, even granted a potential decrease in licensing fees which would save Connecticut businesses money.


Question: uncontrolled RCCA bank withdrawal

Uncontrolled RCCA Bank Withdrawal from Full Power with Minimum Reactivity Feedback (80 pcm/s Withdrawal Rate)


Question: The analysis of net peak area helps in determining the energy levels of emitted radiation.

The net peak area for F 1 is recorded as 1.39E+02, with a net area uncertainty of 32.52.

The ambient background for peak F 2 is recorded as 4.18E+01 with an uncertainty of 1.68E+01.


Question: minimum cooling time for PWR fuel

The enrichment of uranium-235 in a nuclear reactor is crucial for its operational efficiency and safety.


Question: The systems for monitoring radioactive particulates and gases within the containment structure are crucial for detecting leaks in the reactor coolant system.

Experience has shown that these monitors respond rapidly to reactor coolant system leakage and provide a sensitive indication of such leakage.

The information provided by these instruments can be used to detect increases in containment pressure and temperature that is indicative of a leak from the RCS.


Question: Risk Informed Completion Time Program

A report shall be submitted in accordance with Specification 6.9.3 before a newly developed method is used to calculate a RICT.

The reactor coolant system design is an essential aspect of ensuring operational reliability and safety during normal and emergency conditions.


Question: water chemistry program

Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19) are part of the Aging Management Program to address the loss of material in components like stainless steel flow elements exposed to steam environments.

The Flow Element is intended to act as a throttle within the system, and its intended function is critical for maintaining the operational efficiency of the plant.


Question: The NRC mandates that all three Auxiliary Feedwater (AFW) trains must remain operational during critical plant modes.

Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.


Question: What are the regulatory requirements for exemption requests in nuclear licensing?

The regulatory requirements involved for exemption requests are outlined under 10 CFR 50.12, which provides the criteria under which exemptions may be granted by the NRC.

Environmental considerations play a crucial role in analyzing the impact of installing additional technology at the station.


Question: The area within ten miles of the nuclear facility is known as the emergency planning zone, which is crucial for assessing safety measures.

The 0 - 10 mile area presents the population of the "10-Mile EPZ," which is expanded to include certain communities slightly beyond the 10-mile radius.

The projected population density within the 30+ mile area shows a decrease compared to the areas closer to the facility.


Question: radiological environmental monitoring

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

Ecosystems, Biodiversity, and Ecosystem Services


Question: What is the role of an Environmental and Regulatory Manager in nuclear licensing?

An Environmental and Regulatory Manager is responsible for overseeing compliance with environmental regulations and ensuring that regulatory filings meet safety and technical standards in nuclear operations.

Shawn Donovan holds a business telephone number and an email address to facilitate communication with stakeholders.


Question: The average enrichment of 235U for different assembly designs is crucial for determining fuel performance in pressurized water reactors.

Initial Assembly Avg. Enrichment wt % 235U (E) | 52 < Assembly Average Burnup ≤ 53 GWd/MTU |  |  |  |  |  |  |  |  |  |

From 2.1 ≤ E < 2.3 the assembly designs did not return any values, indicating that the designs have not been specified for this range.


Question: Analysis of gamma spectra requires precise calibration and understanding of energy levels in the samples.

Efficiency Calibration Used Done On: 7/8/2014


Question: Non-Destructive Examination (NDE) procedures

Verify that measures are established and implemented to control Non-Destructive Examination (NDE).

MISTRAS does not perform welding for safety-related purposes.


Question: What is the classification basis for survey unit B2-010-101?

Based on the *La Crosse Boiling Water Reactor Historical Site Assessment* (HSA), survey unit B2-010-101 was identified as a Class 2 structure survey unit.

The characterization surveys for the above grade buildings were conducted between November 8, 2016, and November 18, 2016.


Question: What are initial conditions in the context of AOO analysis?

Initial conditions are the key plant inputs that define the (steady-state) operating conditions prior to a specific transient event.

The high ranked phenomena for AOO analysis are established in Reference 2.


Question: maximax analysis in regulatory decisions

The maximax criterion strategy involves selecting the alternative that maximizes the greatest payoff available, so this approach would be more suitable for a 'risk-seeking' investor, who wants to achieve the best results if the best happens.

The conjunctive step requires an alternative to meet a minimal performance threshold for all criteria.


Question: What information is provided about groundwater extraction and injection practices at the site?

ER 3.5.2.1 states that, “Groundwater elevations at some groundwater wells were strongly correlated with Wheeler Reservoir stage levels while groundwater elevations at other exhibited secondary elevation trends that resembled either groundwater extraction or injection cycles over a period of days.

Please provide measured values, or references to these, for the saturated hydraulic conductivities of the geologic units at the site.


Question: What are the voltage specifications for the rod drive motor generator set?

GENERATOR LINE VOLTS is 260V (247V - 273V), AND GENERATOR LINE AMPS is ≈ 80 DC AMPS.

Direct Maintenance to OBTAIN voltage reading on Running MG Set using a DMM OR equivalent, set to voltage, at back of installed voltmeter.


Question: The NRC stipulates that licensees must adjust load settings based on operational needs, including ramp rates, as specified in operational procedures.

If a ramp rate is requested, as Unit Supervisor, “A rate of 10% is desired.”

In the “Load Setpoint and RAMP” box:
a) Enters 110 for Load Set Setpoint
b) Enters 10 for the RAMP value.
c) Selects OK.


Question: What is the average enrichment level of uranium isotopes in PWR fuel?

Initial Assembly Avg. Enrichment wt % 235U (E) | 54 < Assembly Average Burnup ≤ 55 GWd/MTU Minimum Cooling Time (years)

The NRC requires licensees to submit comprehensive technical specifications for any operational changes.


Question: completion time requirements in nuclear regulatory actions

The Note allows Condition A to be entered separately for each component, and Completion Times to be tracked on a per component basis.

The actions must be completed under controlled conditions to ensure safety compliance in all regulatory scenarios.


Question: What are the maximum allowed concentrations for Co-60 in nuclear safety assessments?

The maximum concentration for Co-60 is reported as 1.49E+01 pCi/g in the sample population data.

The average dose per reactor from Cs-137 is reported as 0.6829.


Question: The NRC's guidelines mandate evaluations for the aging management of critical safety systems such as the reactor core isolation cooling.

Reactor Core Isolation Cooling System - Summary of Aging Management Evaluation for Reactor Core Isolation Cooling System | 3-237

The structures associated with the storage of high-level waste must be designed to withstand seismic events and external flooding.


Question: The layout of the Crow Point facility includes various specialized rooms for storage and processing of materials.

The Crow Point facility features a Processing Room that is equipped with several important chemical substances used in elution processes, including Sodium Carbonate and Hydrogen Peroxide.

The schematic also indicates the presence of cooling towers and bulk chemical storage areas within the facility's design.


Question: The involvement of various companies in the nuclear licensing process is crucial for compliance and safety evaluations.

The NRC accepts submittals from multiple affiliate members engaged in contributing to the nuclear regulatory framework.

The development of new technologies in aerospace continues to push the boundaries of innovation, leading to advancements in flight systems.


Question: The number of Safety Relief Valve actuations is critical to assessing the impact on system integrity during operational periods.

The BFN Tours Integrity Long-Term Program Plant Unique Analysis Report assumed 500 SRV actuations during 40 years of normal operations and the contribution from the postulated worst-case LOCA.

The original Bergen Patterson suppression chamber restraint hydraulic snubbers which were evaluated during the initial license renewal (60 years) were replaced with Lisega hydraulic snubbers not long after the license extension was approved (for all units).


Question: In order to meet regulatory requirements, the Cooling Time for spent fuel assemblies must be considered based on their enrichment and burnup levels.

The Minimum Cooling Time required for assemblies with an average enrichment of 2.1 ≤ E < 2.3 is 4.0 years, as per the loading table for PWR fuel.

The NRC's regulations also include guidelines on waste management and disposal to ensure environmental protection.


Question: What is the purpose of a license amendment in nuclear regulatory context?

The purpose of a license amendment is to allow for changes to the operating parameters or other conditions of the nuclear facility's existing license in order to ensure compliance with safety standards and regulatory requirements.

The document also outlines the timeline for the public release and review processes associated with regulatory filings.


Question: What is the definition of bias in the context of instrumentation?

Bias is described as a systematic or fixed instrument uncertainty that is predictable for a given set of conditions because of the existence of a known direction (positive or negative).

Vendor accuracy is considered an independent and random uncertainty component unless the manufacturer specifically states that a bias or dependent effect also exists.


Question: The calculations for the time required for aerosols to achieve specific decontamination factors are crucial for determining safety protocols.

The time at which aerosol I-131 has achieved a decontamination factor (DF) of 50 and time for elemental I-131 to reach DF of 100 is calculated below for the source term in Attachment D.

The supporting output is found in runs BRW Source in Sprayed_R7_ST.o0 & BRW Source in Unsprayed_R7_ST.o0 for Aerosol and BRW Source in Sprayed_R7_ST_elemental.o0 & BRW Source in Unsprayed_R7_ST_elemental.o0 for Elemental (located later in Attachment H).


Question: What is a Decommissioning Funding Plan (DFP) for ISFSIs?

The purpose of the DFP is to demonstrate the licensee’s financial assurance (i.e., that funds will be available to decommission the ISFSI).

The NRC staff is reviewing the information submitted by Ameren Missouri.


Question: nuclear regulatory filings

The evaluation process for licensing activities is a critical component of nuclear regulation, ensuring that all safety, security, and environmental considerations are adequately addressed.

The upgrades to the reactor control systems included enhancements to the digital technology used across multiple operations to improve efficiency.


Question: What is the unit of measurement for radioactive activity?

Our records indicate that you have these devices.


Question: What is the acute intruder health impact (IHI) dose from drilling into waste tanks?

The acute IHI dose at 500 years from drilling through Tank 13 is 0.14 mSv/yr [14 mrem/yr], which is approximately 9 times higher than the 100-year dose associated with drilling into a 7.6-cm [3-in] diameter transfer line.

Many of the drillers surveyed have experience drilling into hard material such as reinforced concrete and steel, so the sensitivity analysis of drilling into a tank could have instead been the base-case analysis at 100 years after closure.


Question: condensate system design features

Knowledge of condensate system design feature(s) and/or interlock(s) that provide for the following: (CFR: 41.7)

Reactor response to loss of feedwater heating


Question: The NRC's standards for aging management programs include directives regarding material and environmental factors for components used in reactor systems.

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

The internal venturi section of each main steam line flow restrictor is fabricated from centrifugally cast low molybdenum content SA-351 Type CF8 CASS material.


Question: What are the medical disqualifying conditions for nuclear operators?

Section 5.3.7(2) [5.4.7(3)] of ANSI/ANS-3.4-1983 [1996] indicates that a history of disturbance of consciousness within the past five [two] years without a satisfactory medical explanation of the cause (emphasis added) shall be considered a disqualifying condition for operation without restriction.

Applicants who have uncorrected near visual acuity of at least 20/40 in one eye and uncorrected distant visual acuity of at least 20/40 in the other eye do not require a conditional license.


Question: nuclear regulatory licensing

The Nuclear Regulatory Commission (NRC) oversees the licensing and regulation of nuclear reactors, ensuring compliance with safety standards and regulations to protect public health and the environment.

The archaeological sites listed in this inventory are primarily focused on historic artifacts and foundations, providing a glimpse into the daily lives of past communities and their cultural heritage.


Question: The analysis of Charpy upper-shelf energy over the reactor's lifespan is critical in determining safety margins.

Per RG 1.99, Revision 2, the Charpy USE is assumed to decrease as a function of fluence and copper content. The current licensing basis USE calculations were prepared for DCPP Units 1 and 2 reactor vessel beltline values for 40 years. These are TLAAs requiring evaluation using the projected 60-year fluence values.

The surveillance data was plotted against RG 1.99, Revision 2, Figure 2. This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines can then be used instead of the existing lines to determine the Position 2.2 EOLE USE values.


Question: containment flooding procedures

1. **IDENTIFY and ISOLATE unexpected source of water:**   - a. ERCW   - b. CCS   - c. High pressure fire protection   - d. Primary water   - e. DI water   - f. SFP cooling water

2. **CHECK cntmt sump activity and chemistry:**   - a. CHECK RHR suction aligned from RWST.      - IF RHR suction is aligned to the cntmt sump, THEN:        1. NOTIFY Chemistry to sample from RHR system.        2. GO TO Step 3.


Question: what is nuclear licensing process

Nuclear licensing is the regulatory process by which the U.S. Nuclear Regulatory Commission (NRC) grants approval for the construction and operation of nuclear facilities, ensuring they meet safety and environmental standards.

The cooling system in a reactor is designed to remove excess heat during the normal operation of the plant.


Question: What are isothermal tests used for in oxidation correlation development?

For the development of oxidation correlations, limited by oxygen diffusion into the metal, well-characterized isothermal tests are more important than the complex thermal hydraulics suggested by [Robert H. Leyse].

The Baker-Just correlation using the current range of parameter inputs is conservative and adequate to assess Appendix K ECCS performance.


Question: proprietary information criteria

The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA’s research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

In accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) 'Trade secrets and commercial or financial information':


Question: The evaluation of electrical production systems is crucial for maintaining operational reliability.

The reliability of electrical production systems is essential for the safe operation of nuclear facilities, ensuring that all components function as intended during both normal and emergency conditions.

Addressing benefits to the nation, the nuclear industry contributes significantly to energy production and economic growth.


Question: NextEra will maintain arrangements with Offsite Response Organizations to support emergency preparedness and response actions.

NextEra stations will continue to meet the requirements of 10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E, Section IV.F.1.

The hospital’s emergency plan includes cooperation and collaboration of local emergency preparedness officials and an annual full-scale community-based drill.


Question: In the context of nuclear materials, the activity concentration measured in picocuries per gram is crucial for assessing safety and regulatory compliance.


Question: What are the average surface dose rate limits for the YANKEE-MPC concrete cask?

The average surface dose rates of each YANKEE-MPC CONCRETE CASK shall not exceed: - 50 mrem/hour (neutron + gamma) on the side (on the concrete surfaces); - 55 mrem/hour (neutron + gamma) on the top; and, - 200 mrem/hour (neutron + gamma), an average of the measurements at air inlets and outlets.

Prior to or at the beginning of STORAGE OPERATIONS.


Question: The report outlines the minimum detectable activity for various radionuclides.


Question: What are the operational conditions for the Auxiliary Feedwater (AFW) system in a reactor?

At 06:00, LCO 3.7.5 Auxiliary Feedwater (AFW) System, is MET.

At 06:15, ENTRY CONDITIONS to 1-FR-H.1, Loss of Secondary Heat Sink, are NOT MET.


Question: battery charger current limit operation

A faulty charger that is incapable of maintaining established battery terminal float voltage does not provide assurance it can revert to and operate properly in the current limit mode necessary during the recovery period following a battery discharge event that the DC system is designed for.

Condition F, Division 1 or 2 125 VDC battery inoperable as a result of maintenance or testing, represents one division with a loss of ability to completely respond to an event.


Question: What is the confidence index threshold for nuclide identification?

Nuclide confidence index threshold = 0.30


Question: What rights do requesters have under the FOIA regarding agency records?

You have the right to seek assistance from the NRC's FOIA Public Liaison by submitting your inquiry at https://www.nrc.gov/reading-rm/foia/contact-foia.html, or by calling the FOIA Public Liaison at (301) 415-0717.

The NRC has made some, or all, of the requested records publicly available through one or more of the following means: (1) www.nrc.gov; (2) public ADAMS, https://www.nrc.gov/reading-rm/adams.html; (3) microfiche available in the NRC Public Document Room; or the NRC Public Access Link (PAL), at https://foia.nrc-gateway.gov/app/Home.aspx.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code is used to perform thermal-hydraulic analyses and includes the ORFEO-HMP critical heat flux correlation to accurately predict the heat transfer characteristics under various conditions.

The simulations showed that the local conditions varied significantly between the different grid test configurations, impacting the overall results and necessitating a review of the assumptions used in the simulations.


Question: Changes in the disposal pricing structure can affect the overall cost of low-level waste management.

A comparison of the year 2020 Bx factors in Table 2-1 to the corresponding year 2018 Bx factors reported in Revision 17 of NUREG-1307, shows that the values increased for both the Washington site and the South Carolina site. These changes were influenced by increases in the disposal price schedules provided by the operators of those disposal facilities (see Appendix A).


Question: What is the role of summary disposition in nuclear regulatory proceedings?

Summary disposition is not the only option for ending this proceeding. The Board was prepared to proceed to an evidentiary hearing to establish whether further cultural resources information was reasonably obtainable.

The Board stated that the Tribe was bound by the terms it had agreed to in accepting the March 2018 Approach, including the two two-week periods allotted to accomplish the survey.


Question: What is the significance of deposition lambda in nuclear safety assessments?

Deposition Lambda (1 / Hours) values are used to characterize the rate at which elemental organic aerosols deposit onto surfaces in compartment models for evaluating airborne contaminants in nuclear safety assessments.

The total control room doses at time (H) = 2.0000 indicate significant variations across different body organs, highlighting the necessity for continuous monitoring and assessment of radiation exposure in various scenarios.


Question: nuclear licensing process

The nuclear licensing process involves several steps including the submission of applications for construction and operating licenses, which are reviewed by the NRC to ensure compliance with safety and environmental regulations.

The reactor's cooling system is designed to efficiently remove heat from the core without causing any thermal shocks to the structure.


Question: The Nuclear Regulatory Commission oversees the licensing of nuclear facilities to ensure compliance with safety standards.

The NRC is responsible for the regulation of commercial nuclear power plants, nuclear research facilities, and other nuclear applications in the United States.

Oconee County's budget document details the financial allocations and expenditures planned for the fiscal year, providing insight into local governance and community services.


Question: What is the role of Defense in Depth in nuclear safety?

Defense in Depth is a safety philosophy employed in the design and operation of nuclear power plants, which integrates multiple layers of safety systems to protect against potential failures and accidents.

The Westinghouse Non-Proprietary Class 3 documentation provides essential insights into the operational and safety protocols relevant to the Limerick Generating Stations.


Question: What is the purpose of the demineralizer in nuclear plants?

The demineralizer is of the mixed bed type with H+ and OH- type resin which removes corrosion and fission product ionic contaminants from the spent fuel pool water or the Refueling Water System water.

The fuel pool skimmer pump is sized to circulate 150 gpm through the skimmer loop and return the water to the spent fuel pool at four discharge points.


Question: The NRC guidelines indicate that a plant-specific aging management program (AMP) may not be required for specific materials used in divider plate assemblies if certain industry analyses are applied.

For units with divider plate assemblies fabricated of Alloy 600 or Alloy 600 type weld materials, if the industry analyses (EPRI 3002002850) are not bounding for the applicant’s unit, a plant-specific AMP is necessary or a rationale is necessary for why such a program is not needed.

The acceptance criteria for this review are: For units with Alloy 600 SG tubes for which an alternate repair criterion such as C*, F*, H*, or W* has been permanently approved for both the hot- and cold-leg side of the steam generator, the weld is no longer part of the reactor coolant pressure boundary and a plant-specific AMP is not necessary.


Question: What is the significance of enrichment levels in BWR fuel assemblies?

The Average Enrichment w% 235U (E) for BWR fuel assemblies is a crucial factor in determining the performance and safety of the reactor.

The NRC’s regulatory framework emphasizes the importance of thorough documentation for all safety assessments.


Question: ICMF form notification process

Reads the ICMF form to individual  **CUE:** Joe Black repeats back the ICMF and states he has no questions.

Operator SIMULATES the following: 1. Selects the phone labeled “NETS” 2. Depresses pushbuttons to dial 5409  **CUE:** You may pick the phone up.


Question: What are preventative protective actions for food contamination in nuclear events?

Preventative Protective Actions are measures taken to prevent or minimize contamination of food products.

The amount of radioactive material released into the atmosphere, the duration of the release, and weather conditions will determine the severity of contamination of food, crops, land, and near the site of the release.


Question: head loss in reactor cooling systems

In nuclear reactor cooling systems, head loss is a crucial parameter that represents the pressure drop due to friction and other factors in the system components, affecting the flow of coolant.

The timeline for the debris addition indicates categorical events that take place during specific hours, marking significant changes in coolant flow and system behavior.


Question: design basis accident methodology

The design basis accident methodology for in-vessel events without radiological release provides a structured approach to evaluating the potential impacts and consequences of various accident scenarios within the reactor vessel.

The project team submitted several documents for review to facilitate the audit process, ensuring all necessary data and analyses were included.


Question: LLRW generation and disposal in nuclear power

DNPS would continue to manage LLRW on site in accordance with NRC regulations and dispose of LLRW in NRC-licensed treatment and disposal facilities during the proposed SLR operating term.

The expected increase in the volume of spent nuclear fuel from an additional 20 years of operation can be safely accommodated on site during the license renewal term with small environmental impacts through dry or pool storage at all plants.


Question: What is a risk-informed approach to nuclear safety regulations?

The proposed changes to the licensing basis, submitted for NRC approval with Attachment 1, address the additional debris loads at the ECCS strainers for Fermi 2 on the basis that the associated risk is shown to meet the acceptance guidelines in Regulatory Guide (RG) 1.174 and that, in conjunction with the existing licensing basis, adequate safety is demonstrated.

This exemption request is for the purpose of allowing the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated loss of coolant accidents (LOCAs).


Question: A regulatory body oversees the compliance of licensees with safety standards and requirements...

The NRC requires all nuclear power plants to maintain up-to-date certifications to ensure their compliance with safety regulations.

States like Florida and California each have their own laboratory certification numbers for different environmental assessments.


Question: What is an emergency diesel generator?

Emergency diesel generator 1A 36-month operability and engineered safeguards surveillance on November 5, 2024.

The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.


Question: license application process for nuclear facilities

The license application process for nuclear facilities involves submitting detailed technical information and analyses to the NRC, including safety assessments and environmental impact studies.

This document outlines various administrative procedures related to correspondence, such as how to address envelopes and format professional letters.


Question: The monitoring of water levels and chemistry in the spent fuel pool is conducted according to regulatory requirements.

The spent fuel pool water level and chemistry are monitored in accordance with Technical Specifications.

Neither the component nor the material and environment combination is evaluated in NUREG-2191.


Question: What are the requirements for electric power systems in nuclear safety?

Electric power systems shall be provided when required to permit functioning of structures, systems, and components. The safety function for each power system shall be to provide sufficient capacity and capability to ensure that (1) the specified acceptable system radionuclide release design limits for the fusion product barriers are not exceeded as a result of anticipated operational occurrences and (2) safety functions that rely on electric power are maintained in the event of postulated accidents.

The design shall demonstrate that power for important safety-significant functions is provided.


Question: The NRC assesses the economic impact of facilities within the surrounding counties.

Per Capita Income statistics reflect the economic conditions of the counties surrounding the TRISO-X Fuel Fabrication Facility, providing important context for regulatory assessments.

The fuel fabrication facility's operational efficiency is determined by the technical specifications set forth by the NRC.


Question: lamellar tear criteria

6X photograph of sample 5A, showing that lamellar tear is less than 0.2 inch in length, which is subcritical in size for brittle fracture.

The reactor's cooling system must be operational at all times to ensure safety during emergencies.


Question: What is SEMS in the context of nuclear licensing?

For purposes of this paragraph as applied to this license, SEMS means any SEMS that has been referenced in a staff SER, TER, EA, or EIS and supplements and amendments thereof.

The licensee shall maintain records of any changes made pursuant to this condition until license termination.


Question: The energy levels at which specific gamma peaks occur help in the identification of radioactive isotopes.

This peak at 295.02 keV corresponds to a significant energy level identified during the analysis of the emitted gamma rays.

The unit of measurement used for energy in this context is the electronvolt (eV), which is a common unit in particle physics.


Question: The time it takes for the reactor protection system to respond after a monitored parameter exceeds its trip setpoint is crucial for safety.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.


Question: The instrumentation for trip functions related to feedwater levels is critical for reactor safety and operational efficiency.

Feedwater and Main Turbine High Water Level Trip Instrumentation is designed to ensure that the reactor maintains proper water levels while preventing safety risks associated with high water levels in the system.

The instrumentation diversity in the reactor safety systems must meet stringent regulatory standards to ensure reliable operation under various conditions.


Question: nuclear licensing procedures

Nuclear licensing procedures involve a series of regulatory reviews and assessments by the Nuclear Regulatory Commission to ensure that all safety and environmental standards are met before a nuclear facility can operate.

The reactor core design is crucial for ensuring efficient operation and heat management, as it determines the overall power output and safety margins.


Question: seismic hazard interval annual initiating event frequency

The seismic hazard interval annual initiating event frequency is calculated (except for the final interval) by subtracting the mean exceedance frequency associated with the g-interval (high) end point from the mean exceedance frequency associated with the g-interval beginning point.

This is common practice in industry SPRAs.


Question: The licensee must manage their radiation program according to the terms and conditions set forth in their radioactive materials license.

You must conduct your radiation safety program in accordance with the conditions of your license, representations made in your license application, and Florida regulations.

The Florida Department of Health aims to protect and improve the health of all people in Florida through various community efforts.


Question: The NRC documents the Adjusted Reference Temperature for reactor components in accordance with safety standards.

67 EFPY Adjusted Reference Temperature (ART) Values for Beltline Plates and Welds is a critical measure that accounts for material properties under neutron exposure over a specified operational timeframe.

The shell ring welds are subject to rigorous quality assurance processes to ensure compliance with construction and design specifications.


Question: What are the regulations for low-level waste disposal in nuclear power plants?

A new section, 3.6, "Low Activity Waste," is included in this revision (Revision 20), which provides information about this category of LLW and the potential for its disposal at sites other than the four currently operating LLW disposal facilities licensed by the NRC or Agreement States.

Revision 20 to NUREG-1307 assumes that LLW generated during plant operations is disposed of using operating funds.


Question: inspection of internal surfaces in miscellaneous piping and ducting components

The GALL-SLR Report recommends that the GALL-SLR Report AMP XI.M38, "Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components," be evaluated for inclusion of augmented requirements to ensure the adequate management of any recurring aging effect(s).

Plant-specific OE examples should be evaluated to determine if the chosen AMP should be augmented even if the thresholds for significance of aging effect or frequency of occurrence of aging effect have not been exceeded.


Question: What methodology is applied to assess total head loss across the strainer in the context of debris management?

Total head loss across the strainer was determined as the sum of the calculated CSHL (see the Response to 3.f.9) and the measured debris head loss from testing (see the Response to 3.f.4).

Strainer head losses at various sump pool temperatures and plant strainer flow rates are of interest.


Question: dry weight analysis in nuclear soil samples

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

5g used due to the highly radioactive and/or hazardous matrix of samples.


Question: Participants should prepare for the Licensing Practices and Procedures course scheduled in March 2025 at the NRC.

Individuals listed in Enclosure 1 have been accepted for participation in the Licensing Practices and Procedures (G-109) course.

The facility’s telephone number is (423) 855-6500.


Question: ITAAC notifications

ITAAC notifications are critical components of the licensing process, ensuring that construction and operations comply with regulatory requirements and safety standards set forth by the NRC.

The vendor inspection report provides detailed information about the quality assurance and construction activities conducted by the vendors supplying parts and systems for the reactor.


Question: The NRC has specific classifications for wells indicating their operational status.

Wells may be classified as Active, Temporarily Abandoned, or Not Installed depending on their operational status.

The general location of the wells includes areas like north of Broadview Acres and south off-site.


Question: minimum cooling time for nuclear fuel assemblies

The table outlines the minimum cooling time (in years) required based on the initial assembly average enrichment of U-235 (E) for different reactor designs such as CE 14x14 and WE 17x17.

In addition to minimum cooling times, various reactor designs also specify criteria for operational safety measures and emergency response protocols that must be adhered to during fuel management.


Question: What is the process for license renewal in nuclear facilities?

License renewal for nuclear power plants, such as the Browns Ferry Nuclear Plant, involves a regulatory audit concerning the environmental review of the license renewal application as detailed in EPID NUMBER: L-2024-SLE-0000.

In order to ensure appropriate oversight, the NRC conducts various routine inspections and assessments of operating nuclear facilities, including compliance with safety regulations and operational standards.


Question: What is the significance of prestressing forces in nuclear containment structures?

The use of the formulas to predict the future prestressing lost in the containment as the justification for extending the ASME Code's five-year surveillance intervals to ten-year intervals may not be valid.

Several failures have occurred at random times over the past four decades.


Question: 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 water quality of the aquifer within the San Juan Structural Basin is highly variable (fresh to briny). TDS concentrations range from 1,000 mg/L in the shallower portions to over 35,000 mg/L in the deeper portions of the aquifer system.


Question: What information must be included regarding the radioactive material in the application?

RADIOACTIVE MATERIAL: Element and/or mass, chemical and physical form; c. maximum amount which will be possessed at any one time.

SEND APPLICATIONS TO: LICENSING BRANCH DIVISION OF NUCLEAR MATERIALS SAFETY U.S. NUCLEAR REGULATORY COMMISSION, REGION II 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PA 19406-2713.


Question: What is the distribution of wind speed and direction for Stability Class E as recorded between August 1, 1972 and July 31, 1973?

Table 2.E-16 shows the annual average joint frequency distribution of wind speed and direction for Stability Class E for the period from August 1, 1972 to July 31, 1973.

Safety measures for nuclear reactors include regular inspections, maintenance of safety systems, and rigorous training for operators.


Question: 500 ml Marinelli sample type


Question: The components of the Residual Heat Removal (RHR) System are designed to withstand various environmental conditions such as borated water leakage and closed-cycle cooling water.

The materials of construction for the RHR System components are: Carbon Steel, Carbon Steel with Stainless Steel Cladding, Cast Austenitic Stainless Steel, Glass, Nickel Alloys, Stainless Steel.

The CSS components are exposed to the following environments: Borated Water Leakage, Demineralized Water, Dry Gas, Plant indoor Air, Sodium Hydroxide, Treated Borated Water.


Question: What are the analysis inputs for calculating average stresses in reactor pressure vessel studs?

The following inputs are required to calculate the average stresses in the studs due to primary load conditions with one stud out of service: 1. The RPV design pressure is 1,250 psia and the design temperature is 575°F. 2. The RPV inner o-ring radius is 111.0 inches. 3. The RPV stud circle radius is 117.313 inches. 4. The number of studs in the Hatch Unit 1 RPV is 52. 5. The stud shank OD is 6.0 inches, the stud shank ID is 1.0 inches, and the stud shank cross section area is 27.489 in².

The average stresses in the studs due to primary load conditions with one stud out of service were calculated using the methodology outlined in Section 5.1.


Question: How does the facility ensure the safety of personnel from alpha contamination?

An operable and appropriately calibrated alpha survey meter will be made available for employee use at the exit of the change room.

The radioactivity of pipes, drain lines, pumps, or duct work where access can be difficult, will be determined by making measurements at a trap or similar access point.


Question: The NRC evaluates the impact of downstream debris on the functioning of emergency core cooling systems and containment spray systems.

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

The erosive wear rate developed for annealed steel in the WCAP was applied to the equipment, which included pumps, heat exchangers, orifices, and spray nozzles.


Question: The activity concentration of each radionuclide is expressed in picocuries per gram (pCi/g).

The activity concentration was recorded as 1.24E+01 pCi/g for K-40.

The aliquot volume for the analysis was 339.39 g dry.


Question: nuclear operating license amendment

The amendment to the renewed facility operating license is intended to conform to the requirements outlined in Order EA-NRR-2025-0013, which specifies safety and operational protocols.

The committee evaluated various options for the expansion of the research center, considering factors such as funding, staffing, and community engagement.


Question: How is the characteristic time scale for a system defined in reactor safety contexts?

For system response characteristic time scale, the temporal scale is defined using a control volume viewpoint. A system with a nominal control volume has a volumetric rate of a process occurring within it, and the nominal amount of time needed for the process to change in the entire volume is the residence time for the specific process.

By defining a transfer process, transfer area, and a rate at which the transfer process occurs, the entire complex system may be described by characteristic spatial, temporal, and concentration (volumetric or mass) scales of varying magnitudes, depending on the level in the hierarchy.


Question: The U.S. Nuclear Regulatory Commission maintains strict guidelines regarding the protection of sensitive data provided by licensees.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) 'Trade secrets and commercial or financial information.'

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary.


Question: method for rad liquid scintillation analysis

The following Analytical Methods were performed:


Question: What are the requirements for volumetric inspections of welds in nuclear facilities?

Welded connections located between the CNV and CIVs receive 100 percent volumetric inspections on a frequency of once per ISI interval (10 years). The volumetric inspection meets the inspection requirements of BTP 3-4.

The ASME Class 1 assembly configuration connecting the CNV and the CIVs includes a safe-end, CITF, and the CIVs.


Question: What is radioactive source term characterization?

The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.

The inspectors evaluated the calibration and testing of the following radiation detection instruments:


Question: safety class valve designation

The valves in Safety Class 2 are designated with the identification code such as 2SD002B, indicating their function and classification within the safety system.

The P&ID coordination for valve layouts is represented using specific coordinates to ensure accurate integration within the system architecture.


Question: average axial power distribution

Average Axial Power Distribution, Comparison POLCA7/POLCA-T.

A schematic of the POLCA-T primary coolant loop model, steam line connections and feedwater boundary location are shown in Figure C.5-40.


Question: The audit will commence on December 4, 2023, and is expected to conclude by December 6, 2023.

The audit will start at 8:00 a.m. (Eastern Time (ET)) on Monday, December 4, 2023, and conclude on Wednesday, December 6, 2023, at 4:30 p.m. (ET) approximately.

The NRC project manager will coordinate with the licensee to set dates and times to discuss information needs and questions arising from the NRC’s review of the audited items.


Question: What is the role of a NuScale Power Module in nuclear reactor design?

The NuScale Power Module (NPM) is a small modular reactor designed to provide efficient and safe nuclear power generation using advanced safety and economic features.

The Control Building is crucial for the operational management of the nuclear power facility, ensuring that all systems function smoothly, following the established protocols.


Question: The facility must not enter a higher operational mode unless all Limiting Conditions for Operations are satisfied, ensuring safety is maintained during transitions between modes.

Specification 3.0.5 establishes limitations on MODE changes when a Limiting Condition for Operation is not met. It precludes placing the facility in a high MODE of operation when the requirements for a Limiting Condition for Operation are not met and compliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in MODEs were permitted.

Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant before or after a MODE change.


Question: What is the NRC’s obligation under the NHPA regarding tribal consultation?

Under the NHPA’s implementing regulations, the NRC must provide every tribe “a reasonable opportunity to identify its concerns about historic properties, advise on the identification and evaluation of historic properties, including those of traditional religious and cultural importance, articulate its view on the undertaking’s effects on such properties, and participate in the resolution of such adverse effects.”

Even if a party’s involvement is limited, if that limited involvement is by choice, the agency has provided the party with a reasonable opportunity to participate.


Question: nuclear radiation measurement units

Activity (pCi/grams) is a key measurement of the amount of radioactive nuclides present in a sample, providing vital information for safety assessments and regulatory compliance.

The energy output of a reactor is crucial for determining the efficiency of its operation and ensuring safety protocols are met.


Question: chemical accidents in nuclear licensing

Process-related chemicals that will be stored on site may include the following:

Fugitive dust emissions will be generated by vehicles traveling on unpaved roads, while combustion emissions will be generated by equipment used within the project areas and vehicles and trucks traveling to and from the project areas.


Question: The NRC requires a thorough analysis of gamma spectra to ensure accurate monitoring of radioactive materials.

The detector used for the gamma spectrum analysis was a HOTLAB system, which enables detailed characterization of gamma radiation from the samples.

The sample size was measured at 9.753E+02 grams, indicating a substantial quantity of material for analysis.


Question: What is the role of the Plant Protection System (PPS) in nuclear safety?

The major systems that comprise the Digital Modernization Project are the Plant Protection System (PPS) and the Distributed Control System (DCS).

Sections 3 and 9 go into further detail for these systems.


Question: What is operational leakage in the context of nuclear safety regulations?

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

The term, “through-wall,” is an adjective used to describe a condition that extends from one surface to another surface in a component.


Question: What are the categories of decommissioning activities for nuclear reactors?

It is expected that the category, and thus the appropriate direct inspection effort and associated resources for a decommissioning power reactor, will change over time, based on the progress of decommissioning at the facility (i.e., an actively decommissioning facility would be expected to go from Category 1 to 2 and eventually to Category 8 as decommissioning progresses). Categories 4 and 5 are intended for decommissioning single or multi-unit sites in SAFSTOR where all units have permanently shutdown. Categories 6 and 7 are intended for decommissioning units at multi-unit sites where at least one unit is still in operation.

The inspection program also provides appropriate latitude for NRC management to administer, plan, and implement site-specific inspection plans commensurate, in part, with licensee performance, site activities, and safety.


Question: What is the significance of the Potomac Aquifer in nuclear safety analysis?

The Potomac Aquifer is characterized as an extremely heterogeneous fluvial system in the Atlantic Coastal Plain of Delaware, which has implications for water safety and environmental assessments in nuclear regulatory contexts.

Benson, R.H. and R.G. Doyle discuss the development of the United States Middle Atlantic Continental Margin in their work on early Mesozoic rift basins, which focuses on geological formations rather than hydrogeological systems.


Question: What are the knowledge requirements for radiological safety for licensed operators?

Knowledge of radiological safety principles and procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, or alignment of filters.

The NRC requires licensees to maintain detailed records of staff training and qualifications to ensure operational readiness.