Question: What are the requirements for acceptance criteria for valve assemblies in OMN-12?

Paragraph 4.4.1, “Acceptance Criteria,” of OMN-12 specifies that acceptance criteria must be established for the analysis of test data for pneumatically and hydraulically operated valve assemblies categorized as high safety significant within the scope of the OM Code Case.

Paragraph 5, “Low Safety Significant Valve Assemblies,” of OMN-12 states that the purpose of its provisions is to provide a high degree of assurance that pneumatically and hydraulically operated valve assemblies categorized as low safety significant within the scope of the OM Code Case will perform their intended safety function if called upon.


Question: What is the maximum dose rate in different access zones for nuclear plants?

| Zone | Maximum Dose Rate (mrem/h) | Description                                         |
|------|-----------------------------|-----------------------------------------------------|
| I    | ≤ 0.25                      | Controlled access, unlimited occupancy.             |
| II   | ≤ 2.5                       | Controlled access, limited occupancy.               |
| III  | ≤ 15                        | Controlled access, limited occupancy.               |
| IV   | ≤ 100                       | Controlled access, limited occupancy.               |
| V    | > 100                       | Restricted access, limited occupancy for very short periods. Access controlled as stated in the Technical Specifications.

Controlled access, unlimited occupancy areas: where entry and exit by plant employees and visitors are not under the direct supervision of the plant health physics staff.


Question: What are the technical data needs that are best addressed by ex-plant harvesting?

The technical data needs best addressed by harvesting are essential for understanding materials degradation to ensure safe and extended operation of reactors.

The workshop will feature sessions with time for presentations and open discussion of different aspects of ex-plant materials harvesting.


Question: What is the significance of Alloy 690 in the Steam Generator Program?

The SG material change from Alloy 600 Old (Existing) Steam Generators (OSGs) to Alloy 690 RSGs requires removal from the Unit 2 TS of inspection/repair provisions that are only applicable to the Alloy 600 OSGs.

The purpose of the meeting is to discuss proposed administrative changes to the Steam Generator (SG) Program descriptions contained in WBN Unit 2 TS 3.4.17, 5.7.2.12, and 5.9.9 to remove SG inspection and repair provisions that become invalid with the installation of the Unit 2 RSGs.


Question: What is ADS Stage 4 actuation in nuclear safety systems?

ADS Stage 4 actuation is a critical process involving the activation of safety systems that respond to specific conditions during a reactor's operation, ensuring the protection of the plant and its components.

The proper functioning of the component cooling water system is essential for maintaining the reactor temperature and preventing overheating.


Question: The analysis uses a combination of earlier seismicity data and recent records to estimate earthquake recurrence in the area surrounding the PSEG Site.

The CEUS-SSC (NUREG-2115) was the starting point for probabilistic seismic hazard calculations.

This updated catalog documents additional earthquakes that have occurred after the earthquake compilation for the CEUS-SSC (which included events through December 31, 2008).


Question: The NRC oversees the registration of devices used in radiation applications, including those manufactured by Thermo Measuretech.

NRC Device Key: 750530 (Internal Control Number)


Question: peak centroid in nuclear analysis

The peak centroid is typically defined as the energy value corresponding to the midpoint of the peak's net area, representing the most probable energy of the emitted photons.

The sample title for the nuclear analysis is labeled as B1-010-001-FSFC-D33-GM, indicating a complex identification for tracking and reference.


Question: 10 CFR 72.48 methodology in tip-over analysis

The direct cause of the failure of the 10 CFR 72.48 screening to identify a change in the MAGNASTOR FSAR required methodology in the tip-over analysis was because APS personnel did not ensure that the applicable regulatory requirements and design basis criteria were correctly translated into the tip-over analysis.

A contributing cause of the failure of the 10 CFR 72.48 screening to identify a change in the MAGNASTOR FSAR required methodology in the tip-over analysis was the inadequate rigor during the owner acceptance of linearly scaled generic concrete cask CC5 tip-over analysis and site specific ISFSI pad tip-over analysis for the modification.


Question: The NRC stipulates specific requirements for managing loads under light load conditions for safety assurance.

The Loads Control Program establishes requirements and provides guidance for the management of loads during light load conditions to ensure safety and structural integrity of systems and components.

The temperature measurements during testing must be recorded accurately to ensure compliance with the specified regulatory standards.


Question: The regulatory guidelines necessitate the measurement of radioactivity levels in material samples to ensure safety.

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

The Client Sample ID is L1-SUB-TDS-FJGS-B33-SB, which is used for identifying the specific sample in this analysis.


Question: The detection efficiency was calibrated using energy sources to ensure accurate peak identification during gamma spectrum analysis.

Efficiency Calibration Used Done On: 7/8/2014

The sample size for testing was determined to be 1.392E+03 grams, ensuring a representative measurement for analysis.


Question: What does PRCHK mean in the context of nuclear reactor monitoring?

PRCHK is a detector setup that indicates a particular type of check or measurement used in monitoring processes within a reactor environment.

The NRC requires that all operators understand the importance of maintaining communication with the control room during operations.


Question: refueling process knowledge

Knowledge of the refueling process is essential for ensuring the safe and efficient handling of reactor fuel during operational changes.

Understanding the various equipment control procedures is crucial for maintaining operational efficiency and safety protocols.


Question: What is the purpose of a geologic mapping plan in nuclear site audits?

The Geologic Mapping Condition Audit Plan outlines the necessary characteristics and geological features of the Hermes 1 site, which are crucial for assessing site suitability and safety requirements for nuclear energy projects.

The audit plan includes a list of individuals responsible for reviewing the materials and their respective dates of input.


Question: The NRC requires a detailed explanation of how members' interests may be affected by licensing actions in proceedings.

Petitioner in NRC licensing proceeding must demonstrate standing; CLI-22-1, 95 NRC 1 (2022)

Contention challenging environmental report analysis of solar power as an alternative to license renewal is dismissed; CLI-22-5, 95 NRC 97 (2022)


Question: The NRC specifies the need for a thorough analysis of safety protocols for handling nuclear materials.

Chapter 11 Radiation Protection Program and Waste Management - Added radiation monitoring from Chapter 7.

Chapter 10 Experimental Facilities and Utilization - (no changes – not applicable)


Question: How does the NRC define the use of Technical Specifications in regulatory framework?

The proposed amendment would revise Technical Specifications (TSs) in accordance with Technical Specifications Task Force (TSTF) Travelers TSTF–563, Revision 1, “Revise Instrument Setting Definitions to incorporate the Surveillance Frequency Control Program,” dated May 10, 2017 (ADAMS Accession No. ML17013A469).

Documents submitted in adjudicatory proceedings will appear in the NRC’s electronic hearing docket and will be available to the public at https://adams.nrc.gov/, unless excluded pursuant to an order of the Commission or the presiding officer.


Question: ASME OM Code in-service testing requirements

10 CFR 50.55a (f)(4) states, 'Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the in-service test requirements set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components.'

Individuals use decision-making practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop.


Question: What is the regulatory process for obtaining a nuclear plant license?

In the United States, the Nuclear Regulatory Commission (NRC) is responsible for issuing licenses for the construction and operation of nuclear power plants, which requires a thorough review of safety, environmental, and security concerns.

The Pollution Prevention Act of 1990 focuses on reducing harmful substances from being released into the environment prior to recycling, treatment, or disposal, emphasizing waste reduction and sustainable practices.


Question: The report utilizes a nuclide library for identifying radioactive substances such as isotopes present in samples.

Nuclide Library Used: C:\Canberra\Apex\Root\Dairyland\NPP\Library\HOTLAB.NLB

The FWHM values are critical for understanding the resolution and quality of the spectral data collected during the analysis.


Question: fuel assembly average burnup

For fuel assembly average burnup greater than 45 GWd/MTU, cool time tables have been revised to account for a 5% margin in heat load.

Certificate of Compliance No. 1031 details the regulations governing the operation of licensed nuclear facilities in the United States.


Question: The NRC emphasizes the importance of reviewing structures, systems, and components throughout the safety assessment process.

SSC refers to Structure, System, and Component, key elements in evaluating the safety and functionality of nuclear facilities.

X Energy, LLC is actively developing advanced nuclear reactors that aim to revolutionize energy production.


Question: The NRC mandates that applicants must include detailed assessments of environmental and safety implications in their submissions.

The NRC requires that licensees submit a licensing application that demonstrates compliance with regulatory requirements for the proposed activity, including technical specifications and safety evaluations.

The document provides insights into various tumor scores and their implications for biological health, focusing on metrics of tumor assessment.


Question: NIS calibration data

Calibration based upon 2-SI-92-10. (WO 121879478).

Reactor Engineer Manager approved the document on 03/01/2022.


Question: Class 1E equipment in nuclear safety systems

Standardized Class 1E hardware Modules are critical components designed to meet the stringent safety requirements necessary for nuclear facility operations.

A non-Class 1E set of tools integrated in a software development environment called RPCT serves a different purpose that does not pertain to safety critical systems.


Question: The Containment Cooling System is essential for maintaining the ambient air temperature within the containment structure during normal operation.

The Containment Cooling System, operating in conjunction with the Containment Ventilation system, is designed to limit the ambient containment air temperature during normal unit operation to less than the limit specified in LCO 3.6.5, "Containment Air Temperature." This temperature limitation ensures that the containment temperature does not exceed the initial temperature conditions assumed for the DBAs.

The analysis and evaluation show that under the worst case scenario, the highest calculated containment pressure is 47.8 psig (experienced during a LOCA).


Question: What procedures does the NRC require for evaluating operating experience in the development of aging management programs?

The staff requests that the applicant provides a presentation on the process used to identify and evaluate the pertinent OE that was used to inform the development of the AMPs as discussed in the LRA and basis documentation.

The project team requires continued PNPP document review capability via the ePortal throughout the audit process.


Question: what is a draft environmental assessment in nuclear licensing?

The NRC and DOE LPO staff have completed a draft environmental assessment (EA) and draft finding of no significant impact (FONSI), *Draft Environmental Assessment and Draft Finding of No Significant Impact for the Palisades Nuclear Plant Reauthorization of Power Operations Project* (ADAMS Accession No. ML24353A157).

Comments can be submitted by the following methods: 1. Emailing PalisadesRestartEnvironmental@nrc.gov 2. Going to regulations.gov and searching for Docket ID NRC-2024-0076; or 3. Mailing comments to the Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.


Question: containment isolation valves

Phase A Containment Isolation Valves (Train B) are critical components in ensuring that various systems, such as the cooling water system and gas analyzer flow control, are properly isolated during specific operational scenarios.

The glycol cooled floor return header isolation valve is crucial for maintaining the thermal management of the reactor building's infrastructure.


Question: What are the reactor containment design conditions?

Air Partial Pressure is a function of the service water temperature. Permissible air partial pressure versus service water temperature is specified in the Technical Specifications.

Instrumentation uncertainties for these parameters have been included in the safety analysis.


Question: reporting requirements to the NRC following an unplanned actuation of the reactor protection system

K/A is matched because the applicant is required to have knowledge of the reporting requirements to the NRC following an unplanned actuation of the reactor protection system.

The technical reference document, NPG-SPP-03.5, discusses various procedures for regulatory compliance but does not specifically cover loss of AC power or the reactor protection system.


Question: What is the process for requesting reasonable accommodations in NRC meetings?

Determinations on requests for reasonable accommodation will be made on a case-by-case basis and the agency may choose between which effective accommodations to provide.

If you have difficulty understanding and/or communicating in English, free language assistance or other aids and services may be available, upon request.


Question: coefficient of friction exemption

This exemption is from the requirement to maintain a coefficient of friction on the ISFSI pad surface of at least 0.5.

Maine Yankee originally requested the exemption following the discovery of a winter icing condition at its ISFSI that created an indeterminate coefficient of friction between the vertical concrete casks and the ISFSI pad surface.


Question: What is residual radioactivity in NRC regulations?

Based on your statements and review of close-out survey information, it has been concluded that all licensable radioactive material has been removed from the location of use and residual radioactive material attributable to licensed activities in this location of use does not exceed current NRC criteria.

An environmental assessment for this action was not required because this action is categorically excluded under 10 CFR 51.22(c).


Question: control rod position check

The task involves resetting the Automatic Reinsert Initiator (ARI) and conducting a post-scram reset control rod position check to ensure that the control rods are positioned correctly following a reactor scram event.

The performance evaluation metric used in this measure is based on simulated events and actual performance at the plant location, which determines whether the candidate's performance qualifies as satisfactory or unsatisfactory.


Question: What is online monitoring (OLM) in nuclear safety assessments?

The proposed change uses online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results.

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


Question: The Minimum Detectable Activity (MDA) of a nuclide represents the lowest level of activity that can be distinguished from the background noise.

Nuclide MDA (pCi/grams) represents the Minimum Detectable Activity of the respective nuclide in relation to the sample.

The energy of K-40 is 1460.81 keV, which is useful for assessing environmental radioactivity.


Question: ECCS evaluation model requirements

To evaluate conformance with the criteria specified in 10 CFR 50.46(b), ECCS cooling performance must be calculated using an acceptable evaluation model for a range of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are evaluated.

Comparisons to applicable experimental data must be made and uncertainties must be identified and assessed so that the uncertainty in the calculated results can be estimated.


Question: In the event of a breach of Functional and Operating Limits, immediate corrective actions must be undertaken.

If any Functional and Operating Limit of 2.1 or 2.2 or 2.3 is violated, the following ACTIONS shall be completed: 1. The affected fuel assemblies shall be placed in a safe condition. 2. Within 24 hours, notify the NRC Operations Center. 3. Within 60 days, submit a special report which describes the cause of the violation and the ACTIONS taken to restore compliance and prevent recurrence.

The safety review process involves an assessment of all potential hazards and their mitigations during reactor operations.


Question: radioactive material license

**AM Calibration Radioactive Material License:** MD-31-206-01

**Meter Owner:** Integrated Cancer Care-North


Question: The NRC's framework requires thorough inspections and monitoring for potential issues related to concrete degradation methods.

Consistent with NUREG-1801. Cracking associated with expansion due to reaction with aggregates has not been observed in the DCPP containment. However, the DCPP ASME Section XI, Subsection IWL AMP (B.2.3.29) will continue to inspect and monitor for cracking and indications of ASR-induced or similar degradation. See further evaluation in Section 3.5.2.2.1.8.

Not used. Increase in porosity and permeability and loss of strength due to leaching of calcium hydroxide and carbonation in inaccessible areas of the DCPP containment will be managed by the DCPP ASME Section XI, Subsection IWL AMP (B.2.3.29) under Item 3.5-1, 014. See further evaluation in Section 3.5.2.2.1.9.


Question: TCE concentration measurement

Based upon data collected from January 2019.


Question: What are the methods used for metals analysis in nuclear regulatory documents?

A summary of the metals analysis is provided, including parameters such as Calcium, Iron, and Uranium, along with the EPA methods used for their detection.

The document includes various parameters such as Alkalinity as CaCO3, which is analyzed using specific methods outlined in the EPA regulations.


Question: How are MOVs categorized in nuclear safety regulation?

When applying Appendix III to the ASME OM Code, licensees shall categorize MOVs according to their safety significance using the methodology described in ASME OM Code Case OMN–3, "Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power Plants," subject to the conditions applicable to OMN–3 which are set forth in Regulatory Guide 1.192, or using an MOV risk ranking methodology accepted by the NRC on a plant-specific or industry-wide basis in accordance with the conditions in the applicable safety NRC evaluation.

The NRC requires licensees to maintain clear and accurate records of all inspections conducted on plant equipment to ensure compliance with safety regulations.


Question: The NRC oversight process requires detailed assessments of nuclear fuel assembly performance based on various burnup thresholds.

The initial assembly average enrichment of uranium is crucial for determining the reactor's efficiency.


Question: The NRC defines requirements for the disposal of radioactive waste generated by nuclear facilities and mandates specific processes for reporting and managing these wastes.

Waste oils (petroleum derived or synthetic oils used principally as lubricants, coolants, hydraulic or insulating fluids, or metalworking oils) that have been radioactively contaminated in the course of the operation or maintenance of a nuclear power reactor licensed under part 50 of this chapter may be incinerated on the site where generated provided that the total radioactive effluents from the facility, including the effluents from such incineration, conform to the requirements of appendix I to part 50 of this chapter and the effluent release limits contained in applicable license conditions other than effluent limits specifically related to incineration of waste oil.

The solid residues produced from the incineration process must be handled in accordance with existing waste management protocols outlined in the regulatory framework.


Question: fuel dispersal in nuclear reactors

DG-1434, “Addressing the Consequences of Fuel Dispersal in Light-Water Reactor Loss-of-Coolant Accidents” discusses the potential implications and safety evaluations related to the dispersal of nuclear fuel during emergency scenarios.

The Control Room Design Requirements outlined in 10 CFR 50.67 and GDC-19 focus on ensuring the operational readiness and safety of control room environments in nuclear power plants.


Question: The system includes components such as the reactor vessel, which houses the core and its supporting structures.

The reactor coolant vent system provides the capability to vent the reactor vessel or the pressurizer using only safety-related equipment.

The inlet is at the bottom of the pump; discharge is on the side.


Question: What is the purpose of HVAC systems in nuclear facilities?

The Auxiliary Building HVAC System is designed to support the structural integrity of various components by ensuring an appropriate indoor climate and managing the effects of aging.

The inspection program involves analyzing the vibrations and sound levels in mechanical components to assess their operational status.


Question: What are periodic updates in nuclear regulatory filings?

Periodic update. Updated to reflect changes in office organization.

Revised in response to feedback forms 0737 and 1810. Section 04, written PN is no longer required following a Commissioners’ Assistants Briefing.


Question: The analysis of reactivity in terms of W/F ratio is crucial when assessing package configurations.

Figure 6.14 – NPC damaged package array (k_{eff}) vs. W/F Ratio (17X17 pellet type, triangular pitch, 53 kgs UO₂/ICCA)

The package configuration must also consider the thermal properties of the surrounding materials to ensure efficient heat dissipation.


Question: What are the conditions of NRC license # 21-32248-01?

We hereby agree, as the management of Heart Rhythm & Pacemaker Specialist, LLC, to all of the constraints, conditions, requirements, and commitments identified in the existing Nuclear Regulatory Commission license # 21-32248-01 for Heart to Heart, P.C.

Heart Rhythm & Pacemaker Specialist, LLC focuses on ensuring optimal patient care through advanced cardiac techniques and management of heart rhythm disorders.


Question: The NRC Form 665 is required for various filings and submissions related to nuclear regulatory processes.


Question: missile barrier shelter

Manholes, Handholes & Duct Banks | Missile Barrier Shelter/Protection | Galvanized Steel | Air - Outdoor | Loss of Material | Structures Monitoring (B.2.1.34) | III.A3.TP-302 | 3.5.1-77 | A

Groundwater/Soil | | | | Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling) | Structures Monitoring (B.2.1.34) | III.A3.TP-212 | 3.5.1-65 | A


Question: The analysis of pressure changes at specific points under varying conditions is critical in determining system performance during accidents.

The pressure at Point C, assuming a 3000-GPM flow from the sump, decreases over time as calculated in the table.

In determining the cooling capacity, it is vital to assess the adequacy of the thermal shielding in the reactor's design.


Question: What is the definition of an isotope in nuclear devices?

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


Question: NPDES permits

Impacts of plant effluent discharges, accounting for midcentury projections, are expected to remain similar to prior plant operations and would be managed under current and future renewed NPDES permits.

Accounting for midcentury projections, climatological changes would not result in significant impacts from building activities and continued operations of Palisades.


Question: measures and controls during construction

The measures and controls to limit adverse effects during construction ensure that potential impacts on the environment and public health are minimized.

Water quality monitoring is essential for preventing pollution and ensuring compliance with environmental regulations.


Question: What is the purpose of Technical Specification Surveillance Requirement 4.0.5?

The amendments revised Technical Specification Surveillance Requirement 4.0.5, “Inservice Inspection and Inservice Testing Program,” in accordance with the implementation of a previously approved amendment dated July 31, 2018 (ADAMS Accession No. ML18156162), to adopt 10 CFR 50, “Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.”

The amendments revised the Donald C. Cook Nuclear Plant, Unit 1 and 2, technical specifications to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-569, “Revise Response Time Testing.”


Question: nuclear instrumentation system alignment

The NIS has been aligned in accordance with the Westinghouse Nuclear Instrumentation System Manual and the Precautions, Limitations, and Setpoints for Westinghouse NSSS.

Functions will be calibrated, tested, and verified, utilizing permanently installed controls and adjustment mechanisms.


Question: The program for inspecting heavy load handling systems is being updated to ensure all relevant structural components are regularly monitored for signs of wear or damage.

The existing program is credited and will be enhanced as follows: 1. Revise the implementing procedures to ensure surface conditions are monitored by visual inspection to provide reasonable assurance that loss of material is not occurring due to general corrosion or wear and the bridges, structural members, and structural components do not exhibit deformation or cracking.

Program will be enhanced no later than six months prior to the subsequent period of extended operation.


Question: ambient background in radiological measurements

Ambient Background refers to the background radiation level that is present in the environment and can affect the readings obtained during radiological measurements.

The energy levels in nuclear physics can vary significantly based on the material being analyzed and the techniques used for detection.


Question: What is the function of the Reactor Protection System in nuclear reactors?

The Reactor Protection System automatically initiates reactor trip: 1. Whenever necessary to prevent fuel rod damage for an anticipated operational transient (Condition II). 2. To limit core damage for infrequent faults (Condition III). 3. So that the energy generated in the core is compatible with the design provisions to protect the reactor coolant pressure boundary for limiting fault conditions (Condition IV).

The design modification includes provisions to test the UV trip and the shunt trip device independently.


Question: What are the types of Topical Reports relevant to nuclear reactors?

The following Topical Reports addressed both power and non-power reactors. No departures are anticipated at this time: Regulatory Analysis Topical Report, Fuel Performance Topical Report, Fuel Qualification Topical Report, Graphite Materials Topical Report.

The following Topical Reports did not specifically address a non-power reactor but are considered sufficient. No departures are anticipated: Scaling Methodology Topical Report, Reactor Coolant Topical Report.


Question: transient population effects on housing

A transient population creates a demand for temporary housing and service in the area.

Since 2002, the Census of Agriculture reports the numbers of farms hiring migrant workers.


Question: What is the Tornado Criterion in nuclear licensing?

The proposed LAR demonstrates the FHTSS meets the intent of a Tornado Criterion “T” structure in the SPS UFSAR, for licensing basis tornado winds only, and using a risk-informed approach.

Based upon the above, Dominion Energy Virginia concludes the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 'no significant hazards consideration' is justified.


Question: traveling water screens in nuclear facilities

These are the original structures, which were later replaced with those shown above in Figure 7.

The new structures provide improved safety features and enhance the overall operational efficiency of the plant.


Question: magnetic anomalies in geological studies

Frequency magnetic anomalies imply a well-defined fabric to the magnetic field that in general mimics the Appalachian tectonic fabric for the region.

The interpretation of the magnetic field for exposed portions of the Appalachian orogen in the site region is discussed in detail by Fisher et al.


Question: isotope activity registration

**Distributor License Number**: IL-01010-02


Question: neptunium-237 analysis

Rad Alpha Spec reports on Neptunium-237 with values of 0.866 pCi/g and 0.906 pCi/g within the acceptable range of (75%-125%).

Cobalt-60 samples showed inconsistent results with QC values of U and uncertainties ranging from +/-0.0393 to +/-0.0329.


Question: What is the role of Technical Specifications in nuclear regulation?

The Technical Specifications are a set of requirements that govern the operation and maintenance of nuclear facility systems and equipment to ensure safety and compliance with regulatory standards.

Copies of corrective action documents generated from previous CDBI are necessary for maintaining operational oversight and ensuring the proper implementation of corrective measures.


Question: The inspections will include visual assessments for gray cast iron components exposed to soil conditions and aluminum-bronze components in water environments.

The DCPP Periodic Inspections for Selective Leaching AMP is a new, plant-specific condition monitoring AMP that manages the loss of material due to selective leaching for gray cast iron components subject to a soil environment and aluminum-bronze components in a raw water environment which have experienced selective leaching at DCPP.

The DCPP Periodic Inspection for Selective Leaching AMP will not provide guidance on preventive actions, but techniques like the application of coatings can help mitigate corrosion.


Question: The Loss of Voltage Function ensures reliable power transfer to critical systems.

Therefore, the power supply to the bus is transferred from offsite power to DG power when the voltages on the bus and the offsite power supplies drop below the Loss of Voltage Function Allowable Values (loss of voltage with a short time delay).

The measured output value of the process parameter exceeds the setpoint, the associated device (e.g., undervoltage relay) changes state.


Question: The NRC enforces strict confidentiality regarding the content of examinations for licensing.

I acknowledge that I have acquired specialized knowledge about the examinations to be conducted in conjunction with the U.S. Nuclear Regulatory Commission (NRC) licensing actions.

I also understand that any violations of the NRC regulations could result in enforcement action by the NRC against me or potentially compromise my position as a licensed individual.


Question: regulatory audit summary report

The NRC staff will prepare and issue an audit summary report prior to completing its review of the relief request.

Username and password information should be provided directly to members of the NRC staff as needed.


Question: License applications must include extensive safety evaluations and compliance reports.

The submittal demonstrates that the applicant has fulfilled all regulatory requirements set forth by the NRC, including mandatory safety assessments for the reactor's operational framework.

The design team anticipates that the construction will proceed according to the planned schedule, barring any unexpected weather conditions.


Question: How is the data logging conducted during the survey of the nuclear facility?

The logged readings indicate the operational status of the systems with detailed timestamps and units of measurement.

The survey unit locations are critical for identifying potential areas of concern in the reactor's vicinity.


Question: steam dump system operation

The steam dump system (SDS) and turbine bypass control are critical components that allow for the management of excess steam in the event of a load rejection, ensuring system stability and safety.

The question type is multiple choice, and it is currently active with a status indicating it is not authorized for practice.


Question: The evaluations in the GALL-SLR Report indicate that certain materials may require special attention to manage aging effects effectively.

The GALL-SLR Report recommends the further evaluation of aluminum piping, piping components, and tanks exposed to an air, condensation, or underground environment to determine whether an AMP is needed to manage the aging effect of loss of material due to pitting and crevice corrosion.

Inspections focus on the most susceptible alloys and locations.


Question: What are the automatic actions that occur during a Main Feedwater regulating valve malfunction?

a. Main Steam Isolation valves (MSIV) close  b. Main Feedwater Reg (MFRV) valves close  c. Main Feedwater pumps (MFP) trip

Unit 1 is raising power and currently at 55% power.


Question: What are Risk Informed Completion Times in nuclear regulatory context?

The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493).

On September 2, 2020, the NRC provided a second Request for Additional Information (RAI) (Reference 4) to support their continued review of Reference 1.


Question: What do SAI and MAI stand for in nuclear safety reports?

SAI (Significant Apparent Indication) and MAI (Minor Apparent Indication) are classifications used to evaluate the severity of indications found in safety assessments of the system.

This section lists the voltage readings and locations in the reactor unit without any contextual implication related to regulatory processes or safety indications.


Question: nuclear facility amendment

The licensing process for a nuclear facility includes several steps whereby amendments to existing licenses may be submitted to reflect changes in operations or compliance with regulatory requirements.

The data received shows a variety of hospitals in Indiana seeking amendments to their licenses, indicating ongoing regulatory compliance.


Question: What is a scram in nuclear safety systems?

The term 'SCRAM' refers to a rapid shutdown of a nuclear reactor, typically initiated by the rapid insertion of control rods into the reactor core to halt the nuclear reaction.

The MAINSTEAM HIGH PRESSURE VALVE is designed to control the steam flow from the reactor to the turbine and is crucial for operational efficiency.


Question: What is the PMP temperature profile used in nuclear safety analysis?

The Maximized March PMP Temperature Profile consists of various events that include a 96-hr Pre-PMP, a 120-hr PMP Event, and a 24-hr Post-PMP, with temperature values recorded at various index hours.

The table lists several storms and their corresponding Spas/SPAS-Lite numbers and daily data ranges, showcasing historical meteorological data.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code employs a unique critical heat flux correlation that has been validated against a comprehensive dataset of experimental results, ensuring its accuracy in predicting heat transfer phenomena in nuclear reactor conditions.

The experimental setup included various reactor coolant configurations and their respective thermal-hydraulic performance metrics.


Question: The requirement for license applicants to submit a technical evaluation of safety margins...

Nuclear reactors must demonstrate compliance with safety regulations by conducting thorough safety assessments and risk analyses.

The weather conditions today are favorable for outdoor activities and community events.


Question: What is the DQO process in FSS planning?

FSS planning and design relies on a properly executed Data Quality Objective (DQO) process to ensure, through compliance with explicitly defined inputs and boundaries, that the primary objective of the survey is satisfied.

The baseline condition is technically known as the null hypothesis.


Question: What are the steps involved in canister unloading for nuclear waste?

CANISTER unloading, including reflooding and weld removal or cutting

Transport of the CONCRETE CASK to the ISFSI


Question: containment pressure values

At the containment pressure data table, the values illustrate the pressure at different times: at time 0 seconds, the pressure is 14 psig, decreasing over time to 0 psig at 300 seconds.

The containment system is designed to withstand external and internal pressures during operation and in the event of an accident scenario.


Question: The proper functioning of main steam safety relief valves is crucial for ensuring the reactor can remain in safe shutdown conditions during emergencies.

The main steam safety relief valves (MSSV) are required for decay heat removal and cooldown to safe shutdown conditions.

The inspectors documented a self-revealed, Green finding and associated non-cited violation of Technical Specification 6.8.1.a, for the licensee’s failure to establish, implement, and maintain procedures associated with safety relief valve maintenance.


Question: What is the minimum average burnup for PWR fuel assemblies?

Minimum Initial Assembly Average Burnup < 43 GWd/MTU

The NRC requires licensees to comply with all safety regulations regarding the operation of nuclear reactors.


Question: net peak area in radiation analysis

The Net Peak Area represents the measured area under a peak which is a key parameter in quantifying the radioactive isotopes present during analysis.

The Ambient Background values indicate the level of natural or environmental radiation that may interfere with the measurements.


Question: The preferred method for determining atmospheric stability involves assessing the vertical temperature difference (delta-T).

The vertical temperature difference (delta-T) is the preferred method for determining Pasquill stability classes because it is an effective indicator of worst-case stability conditions (Pasquill class E, F and G) (Regulatory Guide 1.23). Thus, a multi-level tower close to the HCS was preferred.

Surface data were available from the ASOS in Oak Ridge (station KOQT). This station is relatively close, 8.5 mi. (13.7 km) northeast of the HCS. However, this ASOS is unusual in the respect that it is not located in a relatively open location affiliated with an airport or runway.


Question: Public Information Officer responsibilities in radiological emergencies

The Public Information Officer (PIO), assigned duties at the Emergency Operations Center (EOC), and the Parish Spokesperson will be responsible for implementation of this procedure.

A Parish Spokesperson will be assigned to the Joint Information Center (JIC) at the Alert classification.


Question: triton reactor reliability

The TRITON11 reactor is designed with several proven reliability features that enhance its operational effectiveness and safety profile, ensuring compliance with regulatory requirements.

The annual report includes financial data that indicates a significant increase in operational efficiency across various departments.


Question: What is the process for applying for a nuclear reactor operating license?

The Nuclear Regulatory Commission (NRC) requires that licensees submit an application for a reactor operating license, which includes detailed information regarding the design and safety analysis of the reactor.

Environmental noise levels are a critical aspect of public health and safety, influencing regulations and compliance measures.


Question: recycle evaporator package

The recycle evaporator package processes dilute boric acid and produces distillate and approximately 4 weight percent boric acid stripped of hydrogen, nitrogen, and radioactive gases.

Two reactor makeup water storage tanks (one per unit) supply reactor grade makeup water for the station.


Question: fuel loading regulations for manufactured reactors

A manufacturing license may include authorizing the loading of fresh (unirradiated) fuel into a manufactured reactor under Part 70.

Specifies required protections to prevent criticality: At least two independent physical mechanisms in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions.


Question: The NRC's assessment of environmental impacts includes considerations of dam failure probabilities and their implications for reactor safety.

As was noted previously, the Petitioners did present the Board with a dam-failure rate figure higher than the dam-failure rate figure provided in Duke’s ER with the claim that Duke’s failure rate was “wrong.” Hearing Petition at 14-15. But Petitioners’ dam-failure rate has as a primary underpinning a March 2010 NRC Staff-generated generic failure rate for large rockfill dams similar to the Jocassee Dam.

Duke disputes this, arguing that neither its risk analysis nor any other of their concerns can justify a waiver of the appendix B framework given Petitioners’ failure to challenge in any specific way the post-Fukushima Duke flooding analyses and implemented mitigation measures, along with the associated Staff evaluations that ultimately resulted in the closing of the June 2010 Staff CAL.


Question: What actions should be taken if charging flow is high in nuclear reactor operations?

Implements OP AP-17, Loss of Charging, Section B

Checks letdown flow normal (VB2 meter, PPC)


Question: The isotopic inventory at the end of cycle shows the total activity of different nuclides present.

Total at EOC (Ci Assembly) indicates the radioactive decay measurements of various isotopes present in the reactor.

The reactor safety systems are designed to prevent unexpected failures and ensure operational continuity.


Question: What is the significance of solid uranium oxide in nuclear fuel systems?

The chemical and physical form of the fissile material is solid uranium oxide powder in zirconium alloy rods.

Structural analysis confirms that no normal condition of transport will result in a substantially altered package.


Question: What is the licensing process for nuclear reactors?

The licensing process for nuclear reactors involves multiple stages including the submission of an application, safety reviews, and public hearings to ensure compliance with regulatory standards.

In the event of a malfunction, the control room staff are trained to follow established protocols to maintain plant safety and prevent undue hazards.


Question: The program calculates the direct contribution of gamma radiation from the containment structures to external receptors.

The direct gamma radiation contribution from the containment building to an external receptor is calculated by the program CONTAINMENT SHIELD.

The Halon 1301 suppression system provided for the control room underfloor area does not create a hazard for personnel committed to permanent occupancy in the control room.


Question: The NRC suggests using a common cause impact approach to assess seismic risks at Tier 2 sites, which helps in identifying potential seismic insights.

The special seismic risk evaluation process recommended using a Common Cause impact approach in the full power internal events (FPIE) PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations.

The test case information from EPRI 3002017583, developed by other licensees, including Case Study A, Case Study C, and Case Study D, as well as RAI responses and amendments clarify aspects of these case studies and provide additional supporting bases for this application.


Question: In the context of license renewal, it is essential to implement an effective aging management program to address potential degradation of safety systems.

The Aging Management Audit Plan is designed to evaluate the effectiveness of aging management programs in ensuring the continued safe operation of the facility over its extended licensing period.

The distribution list for this communication includes key personnel from various offices, each responsible for different aspects of nuclear regulation and oversight.


Question: defense-in-depth in nuclear safety

Defense-in-depth is a fundamental safety principle in nuclear reactor design, consisting of multiple layers of protection to prevent accidents and mitigate potential consequences.

The evaluation of seismic activity in the region is crucial for ensuring the structural integrity of the plant during potential earthquakes.


Question: The NRC assesses the safety and environmental impacts of uranium recovery operations before granting appropriate licenses.

The NRC's review process includes evaluating the safety assessment submitted by Powertech, which outlines potential health risks and environmental considerations associated with the in-situ recovery of uranium.

The Office of the General Counsel provides legal representation and guidance to the Commission during procedural matters that arise in licensing and other regulatory processes.


Question: What are the maintenance practices for PIV check valves in nuclear reactors?

Each unit has a scoping preventive maintenance (PM) to evaluate and identify PIVs to recommend for inspection and repair (or replace) for an upcoming refueling outage.

The valves are designed for service in a boric acid solution and are comprised of stainless steel materials.


Question: The identification of nuclides and their associated activity levels are critical components in nuclear regulatory assessments.

The nuclide PA-228 has an energy of 89.95 keV and an activity of 1.39E+03 pCi/grams.

The EU-154 nuclide with an energy of 732.30 keV has a yield of 19.70%.


Question: The elevation of various components in a nuclear reactor system is critical for proper functioning and safety protocol compliance.

The NRC evaluates the safety of emergency backup systems during significant operational events.


Question: gamma direct sample identification


Question: The Nuclear Regulatory Commission mandates that licensees submit their licensing documentation and safety assessments for approval before operational changes can be implemented.

The process of nuclear licensing involves submitting detailed plans and analyses for regulatory review to ensure compliance with safety standards set by the NRC.

In addressing issues of waste management, the facility will implement ongoing training programs for staff to ensure proper handling of hazardous materials.


Question: What is involved in the reauthorization process for nuclear power operations?

The NRC conducts an environmental review before reauthorizing power operations at nuclear facilities, assessing the potential environmental impacts and ensuring compliance with regulatory requirements.

Jennifer Davis sent an email regarding the NRC's environmental review processes, ensuring that all stakeholders are informed about any updates and assessments related to plant operations.


Question: What kind of performance monitoring is necessary to ensure that risk-informed changes are effectively managed?

So making sure that we have sufficient performance monitoring in place to make sure that we’re doing risk-informed changes and not risk-based changes is ultimately what I come back to at the end of the day.

How do you set up a monitoring program that has the lowest footprint possible but gets you the most information possible, the most relevant information.


Question: Borated Water Systems Stainless Steel Inspection

The acceptance criteria for the Borated Water Systems Stainless Steel Inspection is no unacceptable loss of material or cracking that could result in a loss of the component intended function(s) as determined by engineering evaluation.

The Bottom Mounted Instrumentation Thimble Tube Inspection Program monitors tube wall degradation of the BMI thimble tubes.


Question: The NRC mandates that every 10 years, multiple pipe-to-soil potential readings be taken to inform future excavation planning.

Alternate pipe-to-soil potential measurement locations should be chosen to provide additional information applicable to selection of future excavation locations. A minimum of six pipe-to-soil potential readings should be acquired during each 10-year period (10 years prior to the SPEO and the first 10 years of the SPEO) and may be acquired anytime during the 10-year period of the scheduled dig (preferably prior to the actual excavation).

Revise implementing procedures to require that new and replacement field applied coating shall meet the guidance of Table 1 of NACE SP0169-2007.


Question: risk-managed technical specifications

Risk-Managed Technical Specifications (RMTS) Guidelines provide a framework for establishing technical specifications that are risk-informed, allowing for more efficient operation of nuclear facilities while maintaining safety.

The ASME Standard ASME/ANS RA-Sa-2009 outlines the requirements for Level 1 probabilistic risk assessments which are essential for identifying the safety and operational viability of nuclear power plants.


Question: What is the method for analyzing dissolved calcium in water samples?

Calcium, dissolved is determined by EPA 200.7.

The chloride concentration was analyzed on 07/15/24 and measured at 38.13 mg/L.


Question: thermal runaway in nuclear fuel

An array of experiments having multirod assemblies of rods with zirconium alloy cladding reveal that thermal runaway begins well below the 2600 to 2700 range. Perhaps the most impressive is LOFT LP-FP-2 where thermal runaway of the fuel bundle was initiated in the 2060 to 2240 degree Fahrenheit range.

On October 27, 2010, the NRC published for public comment a notice of consolidation of petitions for rulemaking.


Question: What are the guidelines for using mulch in seed mixtures for soil management?

Mulch will be used in areas where water retention, soil temperature, or soil crusting are potential problems for seed germination and seedling growth. The mulch will consist of grass hay, straw, or woodchip applied at the rate of approximately 4.5 t/ha (2 ton/acre). It will be anchored mechanically with a mulch tiller, crimper, or if necessary with a chemical compound.

The limiting factor in establishment of plants in the Crownpoint area will be available moisture. However, fertilizer can be applied with proper moisture to effectively establish seeded species.


Question: nuclear device registration

Devices subject to registration include various types of nuclear instruments that must be licensed and listed by the NRC, such as those used for measuring isotopes and their activities.

The distributor license number is a unique identifier used by the NRC to track and manage the distribution of devices.


Question: very low-level waste regulations

I oppose the VLLW plan to reinterpret the rules to let nuclear waste go to places without a nuclear license.

I oppose this pseudo-rule change that would let nuclear waste in my or other communities' local landfill or other properties.


Question: The NRC has issued modifications to the facility operating license for the research reactor at the University of Missouri.

By order dated February 6, 2025 (Agencywide Documents Access and Management System Package Accession No. ML24354A200), the U.S. Nuclear Regulatory Commission (NRC) modified Renewed Facility Operating License No. R-103 for the University of Missouri – Columbia Research Reactor such that it has no fixed term.

If you have any questions, please contact me at (301) 415-4002 or via email at Jessica.Lovett@nrc.gov.


Question: What is the purpose of vibration isolation elements in nuclear structures?

Consistent with NUREG-2191. The Structures Monitoring program (B.2.1.33) will be used to manage reduction or loss of isolation function of the elastomer supports for vibration isolation elements exposed to air- indoor uncontrolled in the Structural Commodities Group: Equipment Supports and Foundations.

Consistent with NUREG-2191 for all component types in the following Structural Commodities Groups: Cable Trays and Supports, Conduit and Conduit Supports, Electrical Panels, Racks, Cabinets, and Other Enclosures, Equipment Supports And Foundations, HVAC Duct Supports, Miscellaneous Steel, Penetrations and Sleeves, Pipe Whip Restraints and Jet Impingement Shields, Piping Supports, and Tube Track.


Question: materials used in the Fuel Handling System

The materials of construction for the FHS components are: Carbon Steel, Nickel Alloys, Stainless Steel.

The materials from which the component types are fabricated, the environments to which they are exposed, the potential aging effects requiring management, and the AMPs used to manage these aging effects are provided for each of the above systems in the following subsections.


Question: CPM sampling and testing

Samples of CPM are collected at each end and analyzed for water content, concentration of corrosive ions and reserve alkalinity.

The remainder of this report consists of the following 7 parts.


Question: What is the ACMUI's role in nuclear regulatory affairs?

The Advisory Committee on the Medical Uses of Isotopes (ACMUI) is a federal advisory committee that provides recommendations to the Nuclear Regulatory Commission (NRC) on issues related to the medical use of radioactive materials.

The committee discussed various topics, including public safety, emissions regulations, and compliance strategies for industrial operators.


Question: What are the inspection requirements for containment buildings according to the ASME Code?

Per the ASME Code, the periodic visual examination and physical testing of containment building concrete are required in accordance with table IWL-2500-1 L-A, as well as physical testing of unbonded post-tensioning systems in accordance with table IWL-2500-1 L-B.

The NRC staff has determined the need for a regulatory audit to be conducted in accordance with the Office of Nuclear Reactor Regulation Office Instruction LIC-111, Revision 1, "Regulatory Audits" (ML19226A274), for the NRC staff to examine the licensee’s non-docketed information with the intent to gain a better understanding of the alternative request.


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

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

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


Question: Licensees must submit detailed evaluations to demonstrate compliance with safety regulations.

The NRC requires that applicants submit a safety analysis report to demonstrate that the proposed design and operation of the facility are in compliance with regulatory requirements.

Many reactors are equipped with multiple safety systems designed to mitigate potential accidents.


Question: acetone and MEK concentrations in groundwater

Following the EHC® and BounTA™ injections, concentrations of metabolic intermediates acetone and MEK increased up to 27 mg/L and 81 mg/L, respectively.

The site-monitoring wells and extraction wells are completed in the shallow gravel unit.


Question: The measurement of Radium isotopes in water encompasses various analytical techniques and precision indicators.

Concentrations that are greater than the LLD are represented by a closed symbol.

Bars indicate precision for Radium 226 and Radium 228 analytical results.


Question: What is the significance of the source term in nuclear safety assessments?

The source term is defined as the estimated amount and composition of radioactive material that could be released during a nuclear incident, which is critical for evaluating the potential impacts on public health and safety.

The various compartments within the facility are integral to understanding how materials and radiation are managed under normal and emergency conditions.


Question: strontium-90 detection methods

Rad Gas Flow Proportional Counting | GFPC, Sr90, Solid "Dry Weight Corrected" | Strontium-90 | U | -0.0135 | +/-0.0736 | 0.141 | 0.400 | pCi/g | JXC9 | 11/06/19 | 1526 | 1933654 | 1

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


Question: NRC License amendment process

This is in reference to Peter Cordeau’s letter dated October 26, 2020, requesting to amend NRC License Numbers 06-06941-01.

Would you please reply to this email to confirm receipt?


Question: What are the requirements for maintaining active reactor licenses?

The ACTIVE/INACTIVE status and time on shift since April 1, 2022 is as follows:

The simulator exam assesses the operational readiness of reactor operators by testing their knowledge and technical skills.


Question: The NRC mandates that licensees must provide a comprehensive environmental report as part of the license renewal application.

The U.S. Nuclear Regulatory Commission (NRC) requires that the license renewal application include an environmental report (ER) that assesses the impacts from continued operation and any refurbishment to be undertaken to enable the continued operation of the unit.

Constellation Energy Generation (CEG) is preparing an application for renewal of the operating license for Dresden Nuclear Power Station Unit 2 and Unit 3 (DNPS) for an additional 20 years.


Question: What is the purpose of the Anticipated Transient Without Scram Recirculation Pump Trip (ATNS-RPT) instrumentation?

Two channels per trip system for each ATNS-RPT instrumentation function listed below shall be OPERABLE: a. Reactor Vessel Water Level - Low Low, Level 2; and b. Reactor Steam Dome Pressure - High.

The NRC mandates that licensees conduct thorough assessments to ensure compliance with safety regulations.


Question: pressurizer temperature monitoring

There are three temperature detectors in the pressurizer, one in the steam phase and two in the water phase. All detectors supply signals to temperature indicators and high-temperature alarms. The steam-phase detector, located near the top of the vessel, is used during start-up to determine water temperature when the pressurizer is completely filled with water. The water-phase detectors, located at an elevation near the center of the heaters, are used during cooldown when the steam-phase detector response is slow because of poor heat transfer.

Between T_hot and T_cold on at least two reactor coolant loops in the event of loss of the reactor coolant pumps and one vital bus. Two hot-leg temperature indications are available at the auxiliary monitoring panel. One of them is installed with a specific separation from the additional temperature indication available in the control room. This separation meets 10 CFR 50 Appendix R Section III.G.2.


Question: Final Status Survey reports

LaCrosseSolutions submitted the Final Status Survey (FSS) Final Reports for Phases 1 through 3, for NRC review, as documented in References 1, 2, and 3.

The revised documents, as well as the IC dose calculations and relevant technical support documents, are provided electronically on the enclosed CD.


Question: What is the maximum allowable exposure under 10 CFR 20?

Maximum allowable under 10 CFR 20: 2.319 × 10⁵ (170 mrem/person)

Exposure totals from the operation of North Anna Units 1 & 2 amount to 12.7 man-rem/yr.


Question: The guidelines require an assessment of dose from radiological sources to ensure safety limits are maintained.

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

At LACBWR, compliance is demonstrated through the summation of dose from five (5) distinct source terms (i.e., basements, soils, buried pipe, above-ground structures, and groundwater) for the end-state.


Question: nuclear licensing terminology

The licensing of nuclear facilities involves a comprehensive review process to ensure that all safety, environmental, and regulatory standards are met before operations can begin.

The fire brigade training procedures include techniques for effective response to various emergencies, ensuring that personnel are prepared for all eventualities.


Question: containment purge system operational procedures

For lower compartment purge, the following sequence should be followed to prevent Upper-to-Lower Containment press imbalance from opening the Ice Condenser Doors: 1. Upper compartment purge should be started first and operated for at least 5 minutes or until containment pressure is 0.2 psid. 2. Stop upper compartment purge, close associated dampers and stop fans. 3. Start lower compartment purge.

Smoke detected in U1 Aux Bldg air intake will SHUT DOWN Containment Purge Sys.


Question: The documentation provides details on background radiation levels surrounding the samples measured in keV.

The Ambient Background for the peak energy of 295.08 keV is 4.18E+01 with an uncertainty of 1.86E+01.

The Peak Centroid for peak M 1 is 477.51 keV, which shows significant variations in net peak area measurements.


Question: The provisions of LCO 3.0.4.b are crucial for safety evaluations during transitions to different operating modes.

There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable diesel generator and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

Improper maintenance of reactor components can lead to extended downtimes and unnecessary costs for the facility.


Question: What is the significance of the Wasatch Formation in aquifer studies?

The Wasatch Formation is described as bedrock (DOE 1999) that defines the base of the alluvial aquifer. The uppermost part of the Wasatch at the New Rifle site is the Shire member, which consists mostly of low-permeability claystone’s and siltstones and is described as an aquitard rather than an aquifer.

Unweathered claystones and siltstones in the Shire member transmit water slowly.


Question: The samples must be analyzed within the defined holding times to ensure data integrity.

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

The analytical methods used for the samples include EPA EERF C-01 Modified and GL-RAD-A-003 REV# 16.


Question: The current regulatory actions for the TN storage systems have been outlined in the recent submittals to the NRC.

Currently TN Americas LLC has the following licensing actions under NRC review or scheduled for submittal by the end of 2020: - An application for Amendment 2 to NUHOMS® EOS System CoC 1042 was submitted to NRC for review and approval on April 18, 2019.

The proposed amendments to various systems ensure that the highest standards of safety and compliance are maintained.


Question: What is an Aging Management Audit in the context of nuclear power plants?

The NRC plans to conduct an Aging Management Audit from June 17, 2024 – January 31, 2025, in accordance with the enclosed regulatory audit plan.

The application for subsequent license renewal of Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (Dresden), Units 2 and 3, was submitted to the U.S. Nuclear Regulatory Commission (NRC).


Question: What is the location of the nearest noise measuring point to the BFN cooling towers?

Section 3.3 of the ER states that the 2012 noise measurements were 'measured at the location of the nearest residence to BFN in the Paradise Shores Community, approximately 1,500 feet from the BFN property boundary.'

Section 3.3 of the ER identifies that in 2020 and 2022 noise levels were measured.


Question: What is the acceptable voltage range for the Degraded Voltage Relay setpoint in nuclear systems?

The Upper Allowable Value – Reset (AVRU) is ≤ 117.71 V and the Lower Allowable Value – Reset (AVRL) is ≥ 116.94 V for the Degraded Voltage Relay setpoint.

The analytical limit values are essential for ensuring the functionality and proper operation of various safety-related systems in a nuclear facility.


Question: command prioritization in RadICS

The design of field device interfaces and the determination of means for command prioritization were not provided in the RadICS TR.

Addressed in PSAI 7.10 and satisfied by:
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Question: The NRC stipulates that training measures must align with the procedures for readiness in reactor operations.

The Job Performance Measure (JPM) is structured to prepare Unit 1 for Ready State Performance (RSP) operations, ensuring that operators are fully equipped to manage reactor safety and functionality.

The review and approval process for the JPM involves several key personnel, each contributing to the validation of the training material and ensuring its compliance with regulatory standards.


Question: The NRC mandates continuous monitoring and inspections to guarantee the integrity of the Core Spray piping during extended operations.

To ensure fatigue usage remains acceptable through the subsequent period of extended operation, ongoing inspections to ensure the Core Spray piping has not become detached will be performed in accordance with the BWR Vessel Internals program (B.2.1.7) and monitoring of fatigue in accordance with the Fatigue Monitoring program (B.3.1.1) will be performed.

The calculated usage of 0.35 at the jet pump diffuser to core shroud support plate is based on original plant design documentation.


Question: The ongoing operations at Oconee Station will not alter socioeconomic conditions as no additional workforce is required during the SLR term.

Duke Energy has no planned activities at Oconee Station beyond continued reactor operations and maintenance.

As discussed in Section 2.1.4, “Radioactive Waste Storage,” of this EIS, Duke Energy stores spent nuclear fuel from Oconee Station in a storage pool and in an onsite ISFSI.


Question: What is the Limited Aging Management Audit Plan for nuclear facilities?

The Limited Aging Management Audit Plan serves as a roadmap for evaluating the aging management of structures, systems, and components at the Oconee Nuclear Station during the subsequent license renewal application review process.

The audit plan details the regulatory framework for managing waste produced at the facility and outlines the routine inspections performed to ensure compliance with environmental standards.


Question: The NRC oversees the licensing of nuclear reactors to ensure compliance with safety regulations.

The Office of Nuclear Reactor Regulation is responsible for the licensing and regulation of commercial nuclear power plants and research reactors.

Contact me for inquiries regarding the audit plan and related processes.


Question: hot spot fluid temperature definition

The hot spot fluid temperature at 11.25 ft 3-in. is a critical parameter that indicates the temperature of the fluid in localized regions, which can significantly affect the safety and operational efficiency of a nuclear reactor.

The measurement of time in seconds is crucial during the operational phase, as it determines the duration for which various processes and reactions occur within the reactor system.


Question: What is meant by a Red Path in nuclear safety procedures?

The Red Path for Core Cooling is the highest condition and must be addressed prior to the other applicable procedures.

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


Question: Peak Ground Acceleration

Peak Ground Acceleration (g) refers to the maximum seismic acceleration experienced at a site due to an earthquake, expressed in terms of gravitational acceleration.

Hazard Interval Occurrence Frequency (yr) indicates the frequency of occurrence of a seismic event within a specific time frame, expressed in years.


Question: gemstone licensees regulations compliance

The staff estimates an average of 9 gemstone licensees complying with the new regulations. The number of hours saved when complying with the new regulations ranges from a low of 36 hours to a high estimate of 44 hours with a mean value of 40 hours.

This results in an averted cost of $35 thousand at a 7-percent discount rate as shown in Table A-13.


Question: The NRC oversees the licensing of nuclear facilities to ensure their safety and regulatory compliance.

The U.S. Nuclear Regulatory Commission (NRC) is responsible for establishing regulatory frameworks to ensure the safety of nuclear reactor facilities.

The attendees of the meeting included representatives from various organizations and public members who contribute to discussions on nuclear safety.


Question: What is the purpose of the Environmental Report in nuclear licensing?

The purpose of the Environmental Report is to assess the potential environmental impacts of the proposed project and to evaluate alternatives to the proposed action.

The meeting's agenda included an introduction by NRC representatives and a presentation by X-energy discussing their developmental plans.


Question: NEPA implementing procedures

The NRC is in compliance with the requirements of the NEPA amendments, which took effect immediately upon their enactment.

Nevertheless, in accordance with the Principles of Good Regulation, the staff has identified opportunities to enhance clarity, efficiency, and balance in NRC's regulatory procedures and documentation.


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

A Safety Analysis Report (SAR) is a comprehensive document that describes the potential hazards associated with a nuclear facility and outlines the measures taken to ensure safety and compliance with regulatory requirements.

The reactor design includes advanced control systems that enhance operational efficiency under various conditions, reflecting the latest technological advancements in the field.


Question: What is Nickel-63 in radiation analysis?

Nickel-63 is a radioisotope commonly measured in radiological analysis, with results typically reported in picocuries per gram (pCi/g).

The report lists various test methods used in the analysis, such as DOE EMI HALS-300 and EPA 905.0 Modified, to ensure compliance with regulatory standards.


Question: N690-18 applies to structural elements that are critical for the safety of nuclear facilities, including aspects such as cable trays and HVAC ducts.

The scope of N690-18 is safety-related steel structures and structural elements in nuclear facilities.

Further information on the scope of steel structures and structural elements covered by N690 is described in SRP Sections 3.8.3 (Concrete and Steel Internal Structures of Steel or Concrete Containments), 3.8.4 (Other Seismic Category I Structures), and 3.8.5 (Foundations).


Question: What are the dose conversion factors for swimming in a nuclear regulation context?

The Dose Conversion Factors for Swimming (mrem/hr per pCi/Liter) are critical for determining exposure levels during recreational activities in water bodies affected by radioactive effluents.

The table 11.6-2 outlines the North Anna Power Station Environmental Radiological Monitoring Program Measuring Equipment used for monitoring radiological components.


Question: Strontium-90 detection methods

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


Question: The NRC requires that licensees conduct flow sweeps to assess the behavior of strainers in various conditions.

Similarly, flow sweeps were also performed at the end of each test after adding the full load of chemical debris and allowing the head loss to stabilize.

Afterwards, at least one flow sweep was performed by measuring strainer head losses at several different flow rates.


Question: What is the function of the Residual Heat Removal system in nuclear reactors?

LCO 3.6.2.3: Two RHR suppression pool cooling subsystems shall be OPERABLE.

The Completion Time that one train of RHR (RHR-A) can be inoperable as specified by Required Action A.1 may be extended beyond the 7-day completion time up to 7 days to support restoration of RHR following pump and motor replacement.


Question: ball valve handle alignment

Performer locates 2-ISIV-90-112M rotates ball valve handle until the handle is parallel with the pipe. *Step is critical to align 2-RE-90-112 to lower containment.*

When the operator indicates that they will take the ball valve handle and move it to be perpendicular with the pipe, indicate that the handle is perpendicular to the pipe.


Question: Considering the implications for siting small modular reactors, the redevelopment of previously used industrial land offers both advantages and challenges.

Redevelopment of a brownfield site might be eligible for government grants and tax credits.

It may be difficult to secure financing due to contamination uncertainties at some sites.


Question: nuclide minimum detectable activity


Question: emergency response organization

The Emergency Plan maintains a response organization with staff to respond and augment the initial response on a continuing basis.

Editorial changes in the document include directions to develop an ERO relief shift roster and clarification of the timing of Follow-Up Notifications.


Question: Control Room Envelope trace gas testing

08WOSR 3.7.10.4-1, Revision 7, “Control Room Envelope (CRE) Trace Gas Test (Train A)”.

ALION Software Error Notification Number 17, dated 11/6/2012 (Attachment N)


Question: The Charpy upper-shelf energy for the reactor vessel must be maintained at a minimum level throughout its operational life.

Appendix G of 10 CFR 50, (Reference 4.8.8, Paragraph IV.A.1.a), states that reactor vessel beltline materials must have Charpy USE of no less than 75 ft-lb initially and must maintain Charpy USE throughout the life of the vessel of no less than 50 ft-lb, unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

The reactor vessel internal components must be analyzed for fluence projections to ensure they meet safety and operational standards.


Question: The NRC does not include credit for the operator's actions in the Emergency Response Organization during the assessment of the plant's risk profile.

Credit for additional resources and capabilities provided by the SAMGs is not taken in the Level 1 core damage sequence analysis.

Assuming a typical human error probability value for scenarios without any assistance from the ERO can lead to a more conservative model assessment.


Question: The reactor operations must adhere to the limitations set forth in the Technical Specifications for safe functioning.

The licensee is authorized to operate the facility at a steady-state power level not in excess of 250 kilowatts (thermal) and to pulse the reactor in accordance with the limitations in the Technical Specifications.

The approved physical security plan, entitled 'Physical Security Plan, Nuclear Reactor Facility, University of California, Irvine,' dated October 2014, consists of documents withheld from public disclosure pursuant to 10 CFR 73.21.


Question: What is the significance of the Risk-Informed Completion Time (RICT) program in nuclear licensing?

The audit will assist the NRC staff with understanding the licensee’s proposed program for implementing RICTs for certain TSs.

The NRC staff will perform the audit to support its evaluation of whether the licensee’s request can be approved per Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit."


Question: The NRC emphasizes the importance of monitoring the integrity of the bolts used in primary reactor components due to aging effects like stress corrosion cracking.

In its response to RAI MRP-227-B&W-1, EPRI MRP clarified that UT inspections applied to the Primary UCB, LCB, or FD bolting types cannot distinguish between cracklike flaw indications induced by an SCC or IASC mechanism from those induced by fatigue or component overload mechanism.

EPRI includes both 40 – 60-year initial license renewal and 60 – 80-year subsequent license renewal versions of these Primary Items in MRP-227, Revision 2, Table 4-1.


Question: What are the procedures for monitoring containment air temperature?

Monitor the containment air temperature at locations - TE-85 or TE-861¹, - TE-87 or TE-82², - TE-89 or TE-903³, - and TE-91 or TE-924⁴.

Establish a fire watch patrol to inspect the containment.


Question: What is the function of the Reactor Protection System (RPS) instrumentation?

Reactor Protection System (RPS) instrumentation - One or more required channels inoperable.

One Neutron Flux – High channel in each RPS trip system.


Question: The NRC sets regulations for measuring radionuclide concentrations in various materials.

The activity concentration of K-40 in the sample was measured at 8.78E+00 pCi/g Dry with an uncertainty of 1.30E+00.

The uncertainty in the measurements for cobalt isotopes must be reported accurately to ensure compliance with safety standards.


Question: What are the costs associated with premature shutdown of a nuclear power plant?

However, if a proposed regulatory requirement is expected to result in a premature shutdown, this cost is shifted to an earlier time with an associated net increase in its present value.

A reduction in the net cost (i.e., cost savings) is algebraically positive; an increase (cost accrual) is negative (i.e., viewed as negative cost savings).


Question: Division III Loss of Voltage

The lower limit NTSP for the Division III Loss of Voltage function is as follows:  NTSP_LL- Div III LOV = 71 V

The determination of the degraded voltage relays undervoltage settings has been moved to IP-E-0032 (Input 4.5.2) and is deleted from this section.


Question: The U.S. Nuclear Regulatory Commission has established penalties for non-compliance to ensure adherence to safety regulations.

This letter acknowledges receipt of your payment of the $7,500 civil penalty proposed by the U.S. Nuclear Regulatory Commission on October 27, 2020.

Your corrective actions will be examined during future inspections.


Question: What is the purpose of the Emergency Core Cooling System in nuclear reactors?

Other systems which operate in conjunction with the Emergency Core Cooling System are as follows: The Component Cooling System cools the residual heat exchangers during the recirculation mode of operation.

The design philosophy with respect to active components in the high head/low head safety injection system is to provide backup equipment so that maintenance is possible during operation without impairment of the safety function of the system.


Question: direct inspection hours for ITAAC

Figure 2 summarizes the direct inspection hours expended for inspections related to ITAAC since the start of construction at Vogtle Units 3 and 4.

The diagram illustrates the growth of renewable energy projects in various sectors.


Question: The Palisades Restart Project involves considerations for assessing environmental effects.

The Area of Potential Effects (APE) for the Palisades Restart Project encompasses the immediate vicinity of the facility and any areas that may be impacted by project activities.

The recipients of this email include multiple stakeholders who are involved in the review process of the project proposals.


Question: What are the surveillance requirements for maximum surface dose rates of storage casks?

SR 3.3.1.1 Verify maximum surface dose rates of STORAGE CASK loaded with a TSC containing fuel assemblies or FBM are within limits.

Technical assessments must include evaluations of the seismic stability of storage facilities and their structural integrity under expected environmental conditions.


Question: The historical evidence suggests that earthquakes reaching intensity levels of VII to VIII have been recorded in the region, indicating significant seismic activity over the years.

Historical records indicate that the maximum earthquake intensity experienced at the site was the Charleston earthquake of August 21, 1886 with an estimated site surface intensity between VI-VII (MM). The maximum earthquake intensity which has occurred within the region is VII to VIII (MM).

There is no evidence of movement along the regional faults since Triassic time or about 180 million years ago. Therefore, it is concluded that there are no identifiable active faults in the region of the site.


Question: What is the impact of gamma radiation on concrete temperature in nuclear reactor bioshields?

NUREG/CR-7171 (ML13325B077) and RIL 2021-07 (ML21238A064) contain equations backed by test results that show the cumulative effect of heating due to irradiation. The heating effect from gamma ray irradiation has been determined to be limited to approximately 1.12° F of temperature change due to gamma irradiation.

In addition, at mid-core height, where peak radiation is expected, thermal concrete expansion might induce additional stress on the steel liner and a finite element model was used to evaluate the radial temperature profile (of the bioshield wall) from heat loads resulting from the reactor vessel temperature under operating conditions.


Question: chemical and volume control system

Table 3.3.2-8: Chemical and Volume Control System – Summary of Aging Management Evaluation outlines various components like heat exchangers, their materials, intended functions, and the aging effects requiring management.

The Heat Exchanger (Excess Letdown) Shell Side Components, for instance, are constructed from carbon steel and are intended to serve as a pressure boundary in both closed-cycle cooling water and plant indoor air environments.


Question: The requirements for the minimum cooling time are influenced by the average enrichment of the fuel assemblies.

Minimum cooling times for fuel assemblies depend on the average enrichment of uranium-235, designated as wt % 235U (E).

The NRC monitors radioactive emissions from nuclear facilities to ensure compliance with health safety standards.


Question: The NRC evaluates the safety and performance of nuclear licensees to ensure compliance with established standards.

The Watts Bar Nuclear Plant is required to maintain a valid operating license issued by the NRC, which includes regular safety evaluations and assessments to ensure compliance with regulatory requirements.

The cooling system at the plant operates using a closed-loop system that minimizes water usage and environmental impact.


Question: What is a radiological emergency response plan?

The Radiological Emergency Response Plan is designed to ensure effective response strategies during radiological emergencies involving nuclear facilities and materials.

The report outlines the recent structural upgrades made to the facility, emphasizing the need for enhanced safety and operational efficiency.


Question: nuclear licensing terminology

Nuclear licensing is the process by which regulatory authorities grant permission for the operation of nuclear reactors, ensuring compliance with safety standards and regulations.

The Division of Safety Systems oversees various aspects of nuclear facility operations, ensuring that safety protocols are followed effectively.


Question: what are surveillance requirements for diesel generators

This Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (11.4 seconds) from the design basis actuation signal (LOCA signal) and operates for ≥ 5 minutes.

The criteria that in ≤ 11.4 seconds the DG achieves ≥ 3750 V and ≥ 58.8 Hz when the DG is started from a standby condition are starting and accelerating design criteria for the DG that are specified to confirm the capability of the DG to recover from a loading transient.


Question: RadICS platform compliance with IEEE guidelines

Based on the processes reviewed and observed during the regulatory audit for RadICS logic identification, the NRC staff determined the RadICS platform complies with the guidance of IEEE Std. 7-4.3.2-2003, Clause 5.11 for its platform logic.

However, assurance that proper hardware and plant application logic configuration is established and maintained is an activity that must be performed during plant application development and implementation.


Question: What are the electrical requirements for the Technical Support Center regarding power sources?

NUREG-0696 [Reference 1] requires that TSC electrical equipment load shall not degrade the capability or reliability of any safety-related power source.

The original TSC location was implemented in Revision 15 of the Emergency Plan [Reference 7]. The TSC power supply design included a 480 VAC motor control center (MCC), MCC-2C2-3, which fed the UPS, distribution panelboards, and all 480 VAC loads required for the TSC.


Question: The analysis of unsprayed containment volumes is crucial for assessing potential leakage rates.

**Unsprayed Containment Volume**:
- 3: 5.0000E+05
- 0
- 0
- 0
- 1


Question: What is the focus of fire protection inspections in nuclear facilities?

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered from October 21, 2024, to November 22, 2024, and the biennial written examination completed on November 22, 2024.


Question: What is included in ITAAC related inspections?

Inspections of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Related Work 08/13/13 (13-017)

The NRC requires licensees to develop a comprehensive security plan that includes physical protection, cyber security, and access control measures.


Question: The risk evaluation must consider the contributions from fire events in determining the overall CDF and LERF.

- the sum of the CDF contribution for all screened fire compartments is <10% of the estimated total CDF for internal fire events.


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

The Safety Analysis Report (SAR) provides a detailed analysis of the safety limits, accident analyses, and design features of the plant to demonstrate compliance with regulatory requirements.

The ASME code outlines the requirements for the design, fabrication, and inspection of pressure vessels used in various industrial applications.


Question: The program implements measures to monitor and mitigate aging effects on components affected by corrosion and elastomer degradation.

The DCPP Open-Cycle Cooling Water System AMP utilizes surveillance and control techniques to manage aging effects caused by elastomer degradation, corrosion, and fouling in components of the ASW System and the CCW heat exchangers.

Coatings of the CCW heat exchanger waterboxes and ASW System piping linings are managed under the DCPP Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks AMP (B.2.3.27).


Question: The NRC mandates that all sample analyses comply with specified standards and procedures.

Ra 226/228 Gamma Spec Norm Library Uf/Tn Alpha Spec

Please do not dispose of samples until requested.


Question: The analysis of gamma spectrum data provides insights into the identification and quantification of radioactive isotopes.

Gamma spectrum analysis is a method used to identify and quantify radioactive isotopes in a sample by measuring the energy and intensity of gamma rays emitted.

The sample taken on 8/8/2019 had a size of 7.919E+02 grams and was analyzed to provide detailed safety assessments.


Question: Diesel Generators operation during LOOP events

The Diesel Generators would be started 2.1 seconds earlier for a LOOP event if the first level relay did not have a minimum operating time of 2.1 seconds.

In conclusion, there is no design requirement to have the Diesel Generators start immediately at 0 volts.


Question: The NRC stipulates that the initial operator licensing process shall adhere to guidelines set forth in NUREG-1021, ensuring thorough evaluations of applicants.

These examination materials have been developed in accordance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors", Revision 11.

These changes improve examination quality and are in compliance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors", Revision 11.


Question: cracking due to reaction with aggregates

If cracking due to expansion and reaction with aggregates was found to be significant, pattern cracking would be expected over most of the surfaces at grade level where the moisture level is higher. This has not been observed.

Since cracking due to expansion and reaction with aggregates occurs at a comparatively slow rate, the required five-year inspection frequency and scope of the Structures Monitoring program (B.2.1.33) are determined to be adequate, and no plant-specific AMP is required.


Question: How is the groundwater travel time estimated for radionuclides?

The groundwater travel time from the VEGP Unit 4 Auxiliary Building to Mallard Pond has been estimated using a two-dimensional groundwater flow model (SSAR Appendix 2.4B) considering the locations of the effluent holdup tanks, and modeled estimates of hydraulic gradients and hydraulic conductivities of the Water Table aquifer and construction backfill material.

An initial screening analysis was performed considering radioactive decay only.


Question: The NRC emphasizes the need for monitoring structural components to manage aging effects effectively.

Concrete Elements: Embedments are used for structural support, using Carbon Steel as the material, which requires monitoring due to potential loss of material over time.

Water - Flowing indicates an increase in porosity and permeability, leading to loss of strength in various materials.


Question: Americium-241 analysis in nuclear licensing

Americium-241 was analyzed with a result of 0.023 pCi/g and an uncertainty of +/-0.0728.

The Company, LaCrosseSolutions, is located at 4601 State Hwy 35 and is involved in various environmental assessment projects.


Question: volumetric examination of reactor vessel welds

Duke Energy proposes not to perform the ASME Code required volumetric examination of the Oconee Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for Fall 2022.

The proposed inspection date is consistent with the latest revised implementation plan, OG-10-238.


Question: The application of the Test 3 model was necessary to ensure conservative outcomes when evaluating debris scenarios.

To verify that the Test 3 penetration model results in the most conservative output, the Test 3 model was applied to Test 1 and Test 2 conditions.

These figures show that use of the Test 3 model results in more fiber penetration than that measured from either of the other tests.


Question: The NRC oversees the licensing process for nuclear reactors to ensure safety and regulatory compliance.

The licensing process for nuclear reactors involves comprehensive reviews to assess safety measures and operational standards, as dictated by the Nuclear Regulatory Commission (NRC).

Many systems are designed to manage operational efficiencies and financial outcomes without directly addressing safety regulations.


Question: The identification of nuclides plays a crucial role in ensuring safety and compliance in nuclear facilities.

| Nuclide Library Used                          |
|------------------------------------------------|
| C:\Canberra\Apex\Root\Dairyland_NPP\Library\HOTLAB.NLB |


Question: The storage and use of gasoline must be closely monitored due to its known hazards including toxicity and the potential for explosion.

Gasoline is stored in quantities of 6000 gallons at the Lower Alloways Creek Township Buildings, with potential hazards including explosion and flammable vapor cloud.

The use of diesel in large quantities, such as 1,000,000 gallons at Hope Creek, presents a separate set of safety considerations, including the risk of explosion.


Question: What are the quality control measures for soil sample analysis in the FSS process?

The implementation of quality control measures as referenced by LTP Chapter 5, Section 5.9 and LaCrosseSolutions LC-QA-PN-001, Final Status Survey Quality Assurance Project Plan (QAPP) (Reference 10) includes the collection of a soil sample for “split sample” analysis on 5% of the soil samples taken in a survey unit with the locations selected at random.

The LTP, Section 5.6.4.4 and Table 5-15, specifies that for Class 1 open land survey units, surface scans will be performed on 100% of the surface area in the survey unit.


Question: The NRC is focused on enhancing the safety and efficiency of advanced nuclear technologies.

The ADVANCE Act aims to accelerate the deployment of advanced nuclear technology to provide clean energy solutions.

The United States Nuclear Regulatory Commission is dedicated to protecting people and the environment.


Question: What are the land requirements for the natural gas combined-cycle alternative at Oconee Station?

The NRC staff assumes the natural gas facility would require 130 ac (53 ha) of land.

The continued use of developed areas would significantly change the impact on terrestrial resources.


Question: The NRC assesses the structural integrity of components by evaluating stress distributions under various loading conditions.

Through-wall stress distribution at Path P5 for PWR pressurizer vessel top head (CE design)

The unit of vertical force is represented as 1000 lbs and is indicated with a dark blue color in the legend.


Question: What is the significance of the integrated inspection report for nuclear facilities?

The Integrated Inspection Report provides a comprehensive summary of the inspection findings and assessments conducted at the Shearon Harris Nuclear Plant, ensuring compliance with NRC regulations and safety standards.

The NRC inspections also assess the environmental impacts of nearby industrial activities unrelated to nuclear power generation.


Question: cold shutdown boron concentration

The following boration will result in a C₈ in the ACTIVE portion of the RCS which will be higher than the calculated cold shutdown C₈.

NOTIFY Chemistry to monitor RCS boron concentration at the following sample points: • RCS hot leg.


Question: pitting and crevice corrosion

Loss of material due to pitting and crevice corrosion could occur for stainless steel piping, piping components, piping elements, and tanks exposed to outdoor air. The possibility of pitting and crevice corrosion also extends to components exposed to air which has recently been introduced into buildings, i.e., components near intake vents. Pitting and crevice corrosion is only known to occur in environments containing sufficient halides (primarily chlorides) and in which condensation or deliquescence is possible.

Not applicable. During the Unit 1 and Unit 2 sixteenth refueling outages, PG&E replaced the DCPP carbon steel CCP casings with stainless steel casings as summarized in item 3.3-1, 005.


Question: What are offsite power source protective SSC’s in nuclear facilities?

Offsite power source protective SSC’s are systems that provide a means to ensure that offsite power supplies remain available to support plant operations in the event of an accident or loss of power situations.

The PBAPS Technical Support Center and its associated structures are designed to provide operational support during nuclear safety events, ensuring that staff can respond effectively.