Question: steam generator tube integrity requirements

SG tube integrity shall be maintained AND All SG tubes satisfying the tube plugging or repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

Separate Condition entry is allowed for each SG tube.


Question: The analysis of break sizes is crucial for determining the safety of the nuclear reactor design.

The size of the transition break is important in assessing the safety and operational aspects of the reactor design, specifically during the transition from large-break to small-break loss-of-coolant accidents.

The regulatory framework ensures that all proposals are thoroughly analyzed to protect public health and safety.


Question: What are the safety regulations for coolant systems in nuclear reactors?

Coolant systems are essential for maintaining safe operating temperatures in nuclear reactors and are regulated under specific safety standards to ensure their effectiveness in heat transfer and containment of radioactive materials.

During the teleconference, the Committee discussed various topics, including the importance of workforce training and the impact of technological advancements on reactor design.


Question: What criteria determine the ACTIVE status of a nuclear operator's license?

The applicant determines RO C’s license is ACTIVE.

On 06/15/22, RO C failed the As-Found simulator exam, this would have suspended his LOR status.


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

Energy measurements are crucial in nuclear safety analysis as they provide insights into the behavior of radioactive materials and the effectiveness of safety systems.

Nuclide identification employs library references to determine the existence and concentration of specific isotopes present in a sample.


Question: What is stress corrosion cracking in aluminum piping?


Question: gaseous radwaste system

Gaseous Radwaste System – Summary of Aging Management Evaluation.

The inspection of internal surfaces in miscellaneous piping and ducting components is critical to ensuring structural integrity.


Question: The NRC indicates that no aging management programs are necessary due to the absence of aging effects for the Transmission Connectors.

Because there are no aging effects requiring management, no aging management programs are required for the Transmission Conductors.

The materials of construction for the Transmission Connectors are: - Aluminum, Steel.


Question: The NRC mandates that the risk implications of each application of an extended Completion Time be continuously examined within the framework of the RICT Program.

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

This enclosure provides a description of the process applied to monitor the cumulative risk impact of implementation of the Risk-Informed Completion Time (RICT) Program, specifically the calculation of cumulative risk of extended Completion Times (CTs).


Question: What is a control transient in reactor operations?

The ±10 percent step change in load demand is a control transient which is assumed to be a change in turbine control valve opening that might be occasioned by disturbances in the electrical network to which the station output is tied.

Maximum permissible temperature rates of change, as a function of pressure and temperature.


Question: What measures are required during construction to limit adverse effects on water quality?

For construction considerations, please note the amplification of the text in Section 4.4 which addresses the measures and controls to be utilized during construction to limit adverse effects, including siltation and turbidity.

Due to the lack of adequate technology for the separation of fish from debris, the staff feels this recommendation is infeasible at the present time.


Question: What is the significance of Fission Gas Release in nuclear safety?

Figure 29-16: Fission Gas Release C/M for Ramps, Lower-bound Model shows the modeling aspects of fission gas release, which is critical for understanding reactor behavior during normal and accident conditions.

The analysis conducted in this report includes various parameters of reactor operation, such as temperature and pressure, which are essential for evaluating general safety protocols.


Question: low pressure ECCS injection subsystem

Restore low pressure ECCS injection/spray subsystem(s) to OPERABLE status.

HPCI System inoperable AND Condition A entered.


Question: The cumulative dose received by the thyroid and total effective dose equivalent (TEDE) over time must be carefully monitored during nuclear operations.

Cumulative doses for thyroid and TEDE were calculated over specified time intervals to ensure compliance with safety regulations.

The report details various measurements taken during the operation, including ambient temperature and pressure readings throughout the experiment.


Question: What is the significance of the scale in project site mapping for the Crownpoint Uranium Project?

The scale on the map of the Crownpoint Uranium Project indicates the relative size of the drainage areas, providing critical context for evaluating site drainage management and safety measures.

The project site includes multiple drainage areas such as Drainage Area 1, Drainage Area 2, and Drainage Area 3, which are labeled on the map to assist in location identification.


Question: The effective duration limits for entering MODE 4 after a required shutdown are detailed in the regulatory stipulations.

The time limits of Specification 3.0.3 allow 37 hours for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation.

For example, if MODE 2 is entered in 2 hours, then the time allowed for entering MODE 3 is the next 11 hours, because the total time for entering MODE 3 is not reduced from the allowable limit of 13 hours.


Question: What are the considerations for containment purge during outage activities?

Containment purge may be required to be aligned during certain outage activities that pose a high potential for generation of airborne radioactivity.

Precautions related to two train operation: 1. PER 78024 has documented that annulus delta-P indication, and annulus vacuum, are impacted by higher Unit 1 Shield Bldg Vent flows.


Question: average enrichment of uranium in nuclear fuel assemblies

The Minimum Initial Assembly Avg. Enrichment wt % 235U (E) specifies the required enrichment levels for different assembly types in a pressurized water reactor (PWR) context.

The cooling times for various fuel assemblies vary significantly based on their design and operational history.


Question: tsunami risk assessment

Landslide-generated tsunamis typically cause the greatest levels of inundation on shorelines immediately landward of the slide event.

The tsunami was generated at the Gorringe Bank, approximately 125 miles (201 kilometers [km]) from the Portuguese coast, due to a displacement in the submarine fault.


Question: What is the procedure for shutting down a Reactor Coolant Pump after a seal failure?

RCP #3, seal No. 1 has failed, and per 13003-1, "Reactor Coolant Pump Operation," RCP #3 is required to be immediately shutdown.

All RCP standpipes were last filled over two weeks ago.


Question: What is the purpose of the combined license review procedure in nuclear regulation?

Section 207 of the ADVANCE Act requires the NRC to establish and carry out an expedited procedure for qualifying combined license applications with specified deadlines to meet to the maximum extent practicable, and report delays to the Commission and Congress.

The economic impact of energy policies can significantly affect consumer behavior and market dynamics, influencing how energy resources are allocated across sectors.


Question: What is the significance of dose limits in nuclear waste management?

The proposed "dose limit"--really ANY dose limit--is a calculation done by applicant using unverified assumptions to claim the site will be acceptable.

This proposal is even more bold and expansive than the many past efforts by the nuclear industry and its "regulators" which Congress revoked in the 1992 and the public has repeatedly rejected over the decades of the nuclear age.


Question: What are the requirements for the Ge-68 breakthrough tests in nuclear medicine?

Maintaining a record of the breakthrough tests for at least 3 years. These tests should include the ratio of the measured activity of Ge-68 or Ga-68 corrected for the time of elution, time and date of the elution, time and date of the measurement, and the name of the individual who made the measurement;

Sending a written report to the appropriate NRC Regional Office within 30 days after discovery of a generator that is unable to meet the manufacturer’s stated breakthrough limits of Ge-68 on multiple occasions rendering the generator unusable in human patients and research subjects.


Question: The NRC evaluates material properties for nuclear components using specific standards and regulations.

Axial welds at azimuthal angles of 77°, 197° and 317° are all Electroslag with the same material properties (BAW-2258, Document No. 77-2258-00) but with different fluence values. Thus, the ES weld at 317° which has a higher fluence value is limiting.

The bounding ART in the table is the ESW material with Heat No. 37A334/3496 and ART of 103 °F.


Question: What are the requirements for sealed sources and devices in nuclear licensing?

Applicants must provide the radionuclide and nominal activity for each requested sealed source, the manufacturer’s or distributor’s name, and model number for each device, and the number of gauges for each model. Licensees will be authorized to possess and use only those sealed source and devices specifically approved and registered by the NRC or an Agreement State.

If this individual is not a full-time employee of the licensed entity, his or her position and relationship to the licensee should be specified.


Question: containment monitoring system

Required containment cooler condensate monitoring system inoperable. | C.1 Perform SR 3.4.15.1. | Once per 8 hours

D.1 Restore required containment atmosphere particulate radioactivity monitor to OPERABLE status. | 30 days


Question: The adequacy of Probabilistic Risk Assessment (PRA) models must be supported by sufficient information according to regulatory guidelines.

The NRC has established that for license applications, the technical adequacy of PRA models is crucial, particularly when no PRA standards exist.

Maintaining effective communication with the stakeholders is essential for the successful renovation of the facility.


Question: RCP shutdown procedure

Perform an RCP shutdown as follows: If RCP Oil Lift Pump is NOT available, shutdown of RCP may still proceed. The Oil Lift pump should NOT be started for any RCP that is being stopped due to loss of thermal barrier cooling.

The immediate operator actions of 19000-1, E-0 Reactor Trip or Safety Injection are complete, THEN go to Section 4.2.1 Step 4.d.


Question: East Coast magnetic anomaly

Regionally, the East Coast magnetic anomaly exhibits amplitudes ranging from 300 to 500 nanotesla (nT), and widths from 47 to 80 mi., being narrowest off Cape May, NJ.

Gravity and magnetic modeling has shown that the major contributor to the magnetic low field associated with a basin resulted from the negative induced dipole component from a strongly magnetized mafic body southeast of the basin, and that the effect of the basin field was.


Question: The potential consequences of a large-break loss-of-coolant accident can be mitigated by implementing certain emergency measures.

For large-break LOCAs, the most limiting single failure is the loss of one train of emergency core cooling system (ECCS) injection.

The acceptance criteria for the LOCA described in 10 CFR 50.46(1) are met as follows: - A. The calculated maximum fuel element cladding temperature is below the requirement of 2200 °F.


Question: The validation of Zr-4 material is crucial for ensuring structural integrity in nuclear reactors.

The calibration and validation of Zr-4 are essential to demonstrate that the material meets the necessary safety and quality standards in nuclear applications.

Power histories provide insight into the operational performance of a nuclear reactor over time.


Question: safety injection system operation

At time 12:00: A Reactor Trip and Safety Injection have occurred coincident with a LOOP.

Given: - Unit 2 is at 100% Reactor Power - 21 and 22 CCW Pumps are in service


Question: Nuclear reactor coolant pumps must cease operation within a designated period after losing component cooling supply flow to the oil coolers.

RCPs are required to be stopped within 10 minutes following a loss of CCS flow to the RCP oil coolers.

Plausible because the first part is correct. RCPs are required to be stopped within 10 minutes following a loss of Train A CCS flow. The second part is plausible because the RCPs have also lost cooling flow to the Thermal Barriers, however, this does not require the RCPs to be stopped as long as RCP Seal Injection is in service.


Question: nuclear regulatory licensing process

The licensing process for nuclear facilities involves a comprehensive review of safety, environmental, and operational considerations to ensure compliance with established regulations.

The legal team at Honeywell is responsible for overseeing mergers and acquisitions in various sectors, ensuring that all transactions adhere to corporate policies.


Question: PARAGON2 qualification process

Therefore, the NRC staff finds the use of PARAGON2 as a stand-alone code or part of Westinghouse’s overall fuel licensing process acceptable.

The NRC staff reviewed these trends and found that there is adequate, or in some cases, exceptional agreement.


Question: What is the Primary Element Accuracy in flow measurement?

The Primary Element Accuracy (PEA) for flow element 1G33-N035 is ±0.461%. This value is rounded up to ±0.5% for use in the setpoint calculation.

The upstream distance exceeds the recommended minimum distance of 0.5D in ASME MFC-3M-1989 (Ref. 6.3), Table 7, “Recommended Straight Lengths for ASME Venturi Tubes (for 0.5% Additional Uncertainty).”


Question: baseline conditions assessment

Establishing baseline conditions is crucial to support regulation and environmental impact assessments for nuclear facilities.

The assessment of weak points in the environmental management plan is necessary to ensure sustainable practices.


Question: TVA is seeking NRC endorsement for BWRVIP-321 Revision 1-A to enhance its vessel surveillance program.

As part of this application, TVA also requests NRC approval to implement BWRVIP-321 Revision 1-A, “Boiling Water Reactor Vessel and Internals Project, Plan for Extension of the BWR Integrated Surveillance (ISP) Through the Second License Renewal (SLR),” as required by NRC Safety Evaluation included in BWRVIP-321 Revision 1-A and per 10 CFR Part 50, under NRR IV. H, Paragraph III.B.3.

A non-proprietary version of BFN SLRA Section 4 is provided in Enclosure 1.


Question: Wind direction data is essential for understanding meteorological conditions affecting the reactor environment.

Wind direction: W refers to the observations necessary for assessing how wind patterns may influence reactor safety and operational protocols.

The total number of observations recorded for different wind directions provides a comprehensive overview of meteorological conditions, but does not directly link to safety measurements in nuclear operations.


Question: The assessment of measurement reliability must ensure that results from different methods align appropriately.

Comments/Corrective Actions: The replicate measurement results are in acceptable agreement.

| ID                   | Activity | +20%     | -20%     | ID         | Activity | Acceptable | Y/N|


Question: What are the criteria for withholding proprietary information in nuclear licensing?

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

I am requesting the proprietary portions of LTR-NRC-22-29, Enclosure 2 (Westinghouse Topical Report CENPD-289-P, Supplement 1) be withheld from public disclosure under 10 CFR 2.390.


Question: What is the role of additive manufacturing in nuclear applications?

Additive manufacturing allows for the creation of complex components that may be difficult or impossible to fabricate using traditional manufacturing methods, which can lead to significant reductions in waste and overall manufacturing costs.

The document outlines various management strategies for nuclear facilities, emphasizing the importance of operational efficiency and regulatory compliance.


Question: cooling tower plume effects

The major thrust of the following discussion is aimed at an evaluation of cooling tower plume effects. An assessment of the contribution of moisture to the ambient environment from cooling tower blowdown waste heat discharge is included.

Operation of a new facility at the Grand Gulf site will influence the local climatology. A discussion of the expected extent of this influence is presented in this section.


Question: What are the implications of confidential information in the nuclear industry?

The Annual ISA Summary Update contains commercial information of a confidential nature that is not available in public sources or available to the public.

This information contained in the Annual ISA Summary Update is commercial and confidential because it:


Question: containment loss-of-coolant accident analyses

Containment loss-of-coolant accident analyses, which include the effects of flow through vents with grating, as described in 1, 2, and 3 above, consider only the reduced effective flow areas after subtracting the grating area.

Five nodal configurations of the upper reactor cavity (URC) are considered. These nodal models contain 1, 6, 12, 18, and 36 nodes in the URC.


Question: The analysis of vehicle usage during the decommissioning phase involves estimating the number of trips made by transport vehicles.

During decommissioning, it is estimated that there will be 44 one-way trips per day using passenger vehicles for worker transport.

The equipment types include various makes and models, with different horsepower ratings contributing to distinct operational characteristics.


Question: permanent disability reporting requirements

If an operator develops any permanent physical or mental condition that could adversely affect his or her performance of assigned operator job duties or cause operational errors endangering public health and safety, it must be reported to the NRC within 30 days of learning of the diagnosis (per 10 CFR 55.25 and §55.33(a)(1)).

Section C.3 of ES-605 of NUREG-1021 and Information Notices 04-20, 94-14, and 91-08 (all of which are available via links from the operator licensing web page) provide additional information regarding the staff’s expectations with respect to medical standards for licensed operators.


Question: Common Emergency Plan in nuclear facilities

NextEra Energy Seabrook (NES) is enhancing the emergency preparedness program for Seabrook Station by integrating its site-specific emergency plan into a Common Emergency Plan (CEP) together with the other nuclear plants that are part of the NextEra Energy Nuclear Fleet.

Exeter Hospital is accredited by DNV Healthcare in compliance with 42 CFR 482.15, "Condition of Participation: Emergency preparedness".


Question: nuclear licensing process

The nuclear licensing process involves various stages such as pre-application consultation, formal licensing application, and ongoing compliance and oversight by the Nuclear Regulatory Commission (NRC).

Framatome's trademarks, including GALILEO and M5, signify the company's commitment to innovation in nuclear technology and service excellence.


Question: What does FSWOL stand for in the context of nuclear repairs?

Implementation of the FSWOL alternative described in Attachments 1 and 2 of this request produces an effective repair for future mitigation of SCC in the identified weld and maintains the nozzle geometry to permit future ASME Code, Section XI, Appendix VIII UT examinations as implemented through Code Case N-653-2.

The provisions of this alternative are applicable to the fifth 10-year in-service inspection interval for BSEP, which commenced on May 11, 2018, and will end on May 10, 2028.


Question: The NRC has determined that issues categorized as Category 1 in the GEIS do not require additional evaluations in specific environmental reviews unless there is significant new information.

In the 2013 GEIS, NRC considered Category 1 issues to be those that had been found to result in essentially the same (generic) impact at all nuclear plants (or for some issues, those plants having a specific type of cooling system or other characteristics in common). Under NUREG-1437, Category 1 issues did not need to be re-evaluated in plant-specific environmental reviews absent new and significant information indicating the potential for a change in the level of impact relative to that in the GEIS.

TVA has prepared Table 1.2-1, which indicates the sections in the BFN SLRA ER that respond to each requirement of 10 CFR 51.53(c), to verify conformance with regulatory requirements.


Question: nuclear radiation measurement

**Company:** LaCrosseSolutions
**Address:** S4601 State Hwy 35
**City:** Genoa, Wisconsin 54632
**Contact:** Mr. Jason Q. Spaide
**Project:** LACBWR Site Restoration Project
**Report Date:** August 15, 2019

**Client Sample ID:** L1-010-106-AJGS-035A-SS
**Sample ID:** 485191009
**Matrix:** Solid
**Collect Date:** 13-JUN-19
**Receive Date:** 19-JUL-19
**Collector:** Client


Question: EOP-5 entry requirements

EOP-5 entry is required because East Crescent area water level is above the Max Normal value of 0”. With Annunciator 09-4-2-32 in alarm, the associated pump is required to be running. Additionally, EOP-5 requires operation of the pump.

Plausible because Drywell floor drain sumps automatically isolate on a LOCA signal to contain leakage.


Question: The cooling time required for nuclear fuel assemblies varies based on their initial enrichment and the burnup level achieved after use.

The loading table provides a comprehensive overview of various reactor models including CE, WE, and B&W, highlighting their configurations without focusing on cooling times.


Question: offsite registered water wells

Offsite registered water wells within 2 miles of DNPS center point are monitored for safety and environmental compliance.

The status descriptions of the wells include various formations such as limestone and sandstone, which are essential for understanding the geological context.


Question: The analysis of the identified nuclides can greatly aid in determining the safety measures necessary for nuclear facilities.

Peak Analysis Performed on: 5/29/2019 6:14:38 AM


Question: The cooling water intake at the BFN is designed to comply with Best Technology Available (BTA) standards to minimize fish mortality.

The current NPDES permit (AL 0022080) for BFN requires the facility to comply with BTA standards that limit impingement and entrainment mortality from operation of the cooling water intake.

Based on the information provided above, TVA concludes that the incremental contribution of the continued operation of BFN to cumulative impacts on terrestrial resources is SMALL.


Question: The NRC outlines requirements for analyzing different types of evacuation scenarios based on seasonal and weather conditions.

The evacuation scenarios include specifications for various conditions such as season, day of the week, time of day, and weather, which are critical for emergency preparedness and response.

The table detailing evacuation scenarios includes specific conditions categorized by season, day of the week, time of day, and weather, but does not address the protocol for regular maintenance of safety systems.


Question: The evaluation of potential impacts of NFCS operations on local historic locations has been important for regulatory assessments.

An examination of information acquired from federal and state sources relating to regional historic, scenic, cultural and natural landmarks led to the conclusion that the only adverse effects of NFCS construction and operation on such landmarks within a 5-mile radius of the plant site may have been to subsurface archaeological relics or remains on the plant site itself.

Considering the area encompassed by a 5-mile radius from the NFCS, the number of officially recognized historic sites is relatively small.


Question: emergency power supply requirements

**NUREG-0737**, “Clarification of TMI Action Plan Requirements,” section II.E.3.1, “Emergency Power Supply for Pressurizer Heaters,” provides positions and clarifications to the recommendations in section 2.1.1 of NUREG-0578.

The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.


Question: nuclear reactor design criteria

The design criteria for nuclear reactors must consider factors such as safety margins, structural integrity, and the ability to mitigate the consequences of potential accidents.

Scheduled maintenance activities are essential for ensuring the reliability of power generation facilities, including periodic inspections and routine checks of equipment performance.


Question: The oversight of nuclear facility operations includes comprehensive assessments of license applications and their compliance with safety regulations.

The Nuclear Regulatory Commission (NRC) evaluates license applications to ensure that applicants meet stringent safety and environmental guidelines before operating nuclear facilities.

Nuclear power plants require a significant amount of water for cooling, which is drawn from nearby rivers and lakes.


Question: aging management programs for concrete structures

The following AMPs manage the aging effects for the Intake Structure and Intake Control Building components: Fire Protection (B.2.3.14), RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.3.34), Structures Monitoring (B.2.3.33)

The materials of construction for earthwork and yard structures component types are: Carbon Steel, Concrete, Earthfill (rip-rap, stone), Elastomer, Galvanized Steel.


Question: core inlet and outlet mass flow rate during blowdown

CORE INLET AND OUTLET MASS FLOW RATE DURING BLOWDOWN (CD = 0.4, LOW TAVG, MIN SI, COSINE POWER SHAPE, NON-IFBA)

The data was collected over a time span of 50 seconds, with various conditions applied to measure the reactor's performance.


Question: What is the systematic sample location mapping in a nuclear facility?

**TITLE:** LACBWR Cribhouse Interior - Systematic Sample Locations - (604.2 m²)

The floor measurements include various values such as 2.43, 3.87, and 10.06, which represent the dimensions of different areas.


Question: What are the physical security requirements for Independent Spent Fuel Storage Installations?

PHYSICAL SECURITY REQUIREMENTS FOR INDEPENDENT SPENT FUEL STORAGE INSTALLATIONS (81311) (effective date 10/06/2020)

Please provide the attached information to the inspection team for in-office inspection and review to arrive at our regional office no later than April 26, 2021.


Question: The NRC mandates that applicants submit detailed assessments regarding nuclear operational licensing.

Nuclear regulatory licensing requires a comprehensive submission of safety evaluations to ensure compliance with established standards.

The representatives discussed recent legislative initiatives aimed at enhancing community engagement and transparency in government operations.


Question: The energy levels of various nuclides are crucial for understanding their decay properties and potential detection limits.

Nuclide Name | Energy (keV) | Yield (%) | Line MDA (pCi/M²) | Nuclide MDA Activity (pCi/M²) | Dec. Level (pCi/M²)

The weather conditions can significantly affect the outside radiation measurements, akin to how environmental factors can influence sensor readings.


Question: The licensing process for the MAGNASTOR system involves rigorous safety evaluations to ensure compliance with NRC standards.

The MAGNASTOR system's certification is governed by Certificate of Compliance No. 1031, which outlines the regulatory requirements for its use and operation.

The document also contains information about other storage systems that do not require NRC approval, highlighting various storage technologies.


Question: What are the labeling requirements for radioactive materials under NRC regulations?

Each device meets the criteria in 10 CFR 31.5(c)(13)(i) (revised January 1, 2013, incorporated by reference, and available under R12-1-101. This incorporated material contains no future editions or amendments) and bears a permanent (e.g., embossed, etched, stamped, or engraved) label affixed to the source housing, if separable, or the device if the source housing is not separable, that includes the words, “Caution-Radioactive Material,” and, if practicable, the radiation symbol described in R12-1-428.

The report must include the findings from the quarterly assessment to ensure compliance with all applicable regulations and standards.


Question: What is the nuclide MDA in radiation measurement?

Nuclide MDA (Minimum Detectable Activity) represents the lowest concentration of a nuclide that can be reliably detected in a sample, typically measured in pCi/g.

The Peak Locate From Channel indicates the starting channel number used during the peak analysis process, showing how the data was captured.


Question: Unit 1 has faced a complete loss of alternating current supply, necessitating immediate response to safeguard systems.

Unit 1 has experienced a Loss of All AC power.

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


Question: safety analysis for the MAGNASTOR system

The MAGNASTOR system includes features that are designed to ensure the safe storage of spent nuclear fuel, with robust engineering controls and safety analysis conducted as part of the regulatory evaluation process.

The regulatory process requires thorough documentation and adherence to standards, which is essential for maintaining transparency and compliance within the nuclear industry.


Question: What is the purpose of the Engineered Safety Feature Actuation System (ESFAS) in nuclear plants?

3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

The provisions of Specification 4.0.4 are not applicable to MSIV closure time testing.


Question: Inservice Testing program requirements

The alternative would apply to compliance with American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code of Record, Subsection ISTC, paragraphs ISTC-3630, ISTC-360(f), ISTC-5221(a)(1), and ISTC-5224, for specific IST Program requirements for certain pressure isolation valves (PIVs).

An audit is conducted to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision.


Question: What are the allowable actions for workers within the containment structure during power operations?

Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation area, aligning filters, etc.

The NRC mandates that licensees submit specific reports regarding the performance and conditions of their reactors on a regular basis.


Question: What is the significance of minimum flow area in nuclear systems?

Minimum flow area (ft²) is a critical measurement used to ensure adequate flow rates in nuclear safety systems; it affects the system's ability to manage coolant and other fluids effectively.

The inertia length of a flow path can significantly impact the flow dynamics, determining how quickly disturbances propagate through the system.


Question: what is IPAWS in emergency management

The Integrated Public Alert and Warning System (IPAWS) has been implemented in the 10-mile EPZ for emergency alerting of the general public should an emergency occur at RBS.

Training will be offered to emergency workers on an annual basis. The type of training required will depend on the individuals’ assigned task on the Parish emergency response team.


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

The table provides initial assembly average enrichment levels of wt % 235U (E) for various average burnup (B) GWd/MTU in PWR fuel assemblies.

The NRC ensures that all regulatory processes are adhered to during the submission of documentation by licensees.


Question: The review of the renewal application will be completed within the regulatory timeline set by the NRC.

Renewal actions are normally processed within 180 days.

If you have any questions concerning the processing of this renewal, you may contact the Materials Licensing Branch at 1-800-522-3025, extension 9887 or 1-630-829-9887.


Question: thorium licensing

A. Thorium                                        | A. Metal Alloy Containing Thorium          | A. No Possession Authorized                                                 | A. In accordance with 10 CFR 40.52, the licensee is authorized to distribute pre-packaged tungsten electrodes, containing not more than 4% thorium, to persons exempt from licensing under 10 CFR 40.13(c)(1)(iii), or equivalent regulations of any Agreement State.

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Parts 30, 31, 32, 33, 34, 35, 36, 37, 39, 40 and 71, and in reliance on statements and representations heretofore made by the licensee, a license is hereby issued authorizing the licensee to receive, acquire, possess, and transfer byproduct, source, and special nuclear material designated below.


Question: What is the minimum detectable concentration in nuclear reporting?

The report outlines various technical parameters crucial for assessing the restoration project's safety standards.


Question: Technical Specifications for nuclear reactor operations

Can the question be answered solely by knowing the TS safety limits?

Does the question involve one or more of the following for the TS, TRM, or ODCM:


Question: What is the aging effect in concrete used for nuclear plant structures?

Cracking due to expansion from reaction with aggregates

Fine and coarse aggregates conform to ASTM C33.


Question: How does the presence of certain nuclides in a reactor affect its safety assessments?

CS-137 has a yield of 1.46% and an energy of 475.35 keV, which indicates its significance in nuclear safety evaluations.

The energy level of K-40 is measured at 1460.81 keV and has a yield of 10.67%, contributing to its classification as a stable isotope.


Question: The boundaries of the DNPS site are critical for regulatory compliance and safety assessments.

Several states such as IA, WI, and MI are part of the newly defined locations for nuclear facilities.


Question: What geological formations are significant in the vicinity of the Horizon Center Site?

The geologic formations in the vicinity of the Horizon Center Site include the Kingsport Formation, Lebanon Limestone, Marcellus Dolomite, Mifsudberg Limestone, Pond Springs Formation, and Ridley Limestone.

Rev. 1 of the report was published in January 2025, indicating the most recent updates to the document.


Question: requirements for DC electrical power subsystems

LCO 3.8.4 The Division I and Division II DC electrical power subsystems shall be OPERABLE.


Question: scale information in nuclear regulatory documents


Question: What is the coefficient of compression (Cc) in soil mechanics?

Cc is the Coefficient of Compression, a parameter used to describe the compressibility of soil under load.

The Kirkwood formation is characterized by a specific gravity range of 2.54 to 2.78 across various samples.


Question: thermal emissivity of volcanic tephra

The size of ash particles has a potentially significant effect on thermal emissivity, with finer particle sizes having relatively higher emissivities than coarser particle sizes.

Tephra compositions are mafic (Fuego, Sakurajima), intermediate (Santiaguito, Soufriere Hills Volcano), and silicic (Mono Craters).


Question: To ensure proper functioning, at least two service water loops need to be verified as OPERABLE.

At least two independent service water loops shall be OPERABLE.

Each automatic valve servicing safety related equipment actuates to its correct position on Safeguards Initiation signal.


Question: What is considered proprietary information in nuclear regulatory documents?

Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome Inc. for the control and protection of proprietary and confidential information.

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


Question: What is the unit of measurement used for radioisotope activity?

| Isotope (e.g. AM241) | Activity (e.g. 1005) | Unit (e.g. mCi) |

**Manufacturer name**: TN TECHNOLOGIES, INC.


Question: RISC classifications in nuclear safety

RISC-1, -2, -3 and -4 SSCs: review plant changes, operational experience, and update categorization and treatment every 2 refueling outages.

The cooling system is vital for maintaining the temperature of the reactor core during operation.


Question: The NRC tracks the distribution of devices and requires proper registration of each unit.

The NRC requires licensees to submit specific information about devices subject to registration, including isotopes and their associated activities measured in units such as milliCuries (mCi).

Each device listed must have a unique serial number and is associated with a manufacturer, as noted in the documentation.


Question: What is the purpose of the On-Site Audit in the license renewal process?

The purpose of the On-Site Audit is to make visual observations of areas, equipment conditions, and configurations, and will be performed by an integrated review team.

The staff, with the assistance of the applicant, will conduct a detailed review of scoping and screening methodology for mechanical, electrical, and structural systems and components.


Question: Rad Alpha Spec

Rad Alpha Spec is a measure of alpha radiation, which is a type of radiation emitted by certain radioactive materials, expressed in picocuries per gram (pCi/g).

The QC1204336283 record indicates sample number 485191001 with a quality control status of DUP and various uncertainty levels for measurement.


Question: What are the Post Maintenance Testing requirements after maintenance on a safety-related system?

The Shift Manager has directed you to determine the Post Maintenance Testing requirement(s) to demonstrate LCOAR restoration by completing 1BOL 7.7 Section E.1.

Maintenance has just completed work on valve 1CC9412B, 1B RH CC Outlet Vlv.


Question: nuclear reactor thermal power measurements

The thermal power of a nuclear reactor is the amount of heat generated in the reactor core, typically measured in megawatts (MW). This power level indicates the capacity to produce steam that drives turbines for electricity generation.

The regulations pertaining to the licensing of nuclear facilities are complex and have been developed over decades. Each regulatory body has specific guidelines that need to be adhered to for successful compliance.


Question: Operators must ensure system parameters like temperature and pressure are closely monitored during equipment repairs.

Deviation on the FWRV's is based solely on S/G level, NOT on the output of the 3 element control logic.

Small changes in feedwater flow can lead to increased operational efficiency for the plant.


Question: What is the Toxic Gas impact hazard in nuclear safety assessments?

Based on this review, the Toxic Gas impact hazard can be considered to be negligible.

For shipping impact hazards, according to the IPEEE, due to the physical location of DC Cook buildings and structures, the only danger to the plant from ships/barges are from those that run-aground that collapse the circulating water intake cribs resulting in flow obstruction of all three circulating water system intake lines.


Question: How do deterministic analyses relate to beyond design basis accidents in regulatory frameworks?

We've had Chapter 19 where deterministic, beyond design basis accidents have been formed under the best estimate sense. They do contribute specifically in a defense-in-depth context.

I'm trying to wrap my head around whether we have equivalence with Part 53 in that sense.


Question: The activity levels of various nuclides after treatment are significantly reduced compared to their initial activities.

The expected performance of the Cation Removal Bed is critical for maintaining low levels of contamination in the Chemical and Volume control system.


Question: What is the significance of a Technical Specification in nuclear licensing?

Technical Specifications are written requirements that ensure safe operation of the nuclear facility and establish the conditions under which a reactor may be operated.

The reactor vessel is designed to withstand high pressures and temperatures during normal and accident conditions.


Question: nuclear emergency response plans

For large LWRs, the most notable guidance documents for the development and maintenance of emergency plans are: NUREG-0654/FEMA-REP-1, Revision 1, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” dated November 1980, which provides guidance and evaluation criteria for the development and evaluation of operating power reactors’ and offsite response organizations’ (OROs) radiological emergency response plans.

Additionally, § 50.47 provides EP requirements for nuclear power reactors, including planning standards for onsite and offsite emergency response plans.


Question: Southern Nuclear Operating Company seeks approval to deviate from standard inspection protocols for Class 1 welds during preservice activities.

Pursuant to 10 CFR 50.55a(z)(2), Southern Nuclear Operating Company (SNC), the licensee for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, requests NRC authorization to use an alternative to the requirements of Section XI, IWA-2200 of the ASME Boiler and Pressure Vessel (B&PV) Code, 2007 Edition through 2008 Addenda (code of record) for VEGP Units 3 and 4.

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


Question: unfiltered inleakage to control room

**Filtered Intake Environment to Control Room**

### Pathway 4:
**Unsprayed Containment Volume to Environment**


Question: What is a survey unit in nuclear licensing?

**Survey Unit**: Meets the Acceptance Criteria

The critical value for the survey unit is a statistical measure used to evaluate data consistency.


Question: The NRC staff performed a comparison of the DCGL values proposed by the licensee for surface contamination against the unrestricted release criteria of 10 CFR 20.1402.

The NRC staff finds that the DCGL values proposed by the licensee for the surfaces of the NS Savannah are acceptable for demonstrating compliance with the unrestricted release criteria in 10 CFR 20.1402.

These analyses are not site-specific and therefore are not driven by a limited number of physical parameters selected by the licensee.


Question: What is an aging management program (AMP) in nuclear regulatory context?

None of the components in this section have been evaluated in accordance with the regulatory framework.


Question: What are the operational guidelines for pump performance in a nuclear facility?

Maintain pressure below 100 psig; Ensure flow rates in line with specifications; Monitor temperature and pressure; Regularly inspect pump operation.

The reactor core must be shielded effectively to minimize radiation exposure to workers and the environment.


Question: What is the significance of the 100-year peak flow inundation area in nuclear site assessments?

The 100-year peak flow inundation area is an important consideration in evaluating potential flooding events that may impact the site, ensuring that safety protocols are in place for any nuclear facilities located in these zones.

The Church Rock Section 8 and Section 17 are recognized for their unique geographical characteristics, which include specific contour formations and notable features that contribute to environmental studies.


Question: What is the significance of RCS pressure in nuclear safety regulations?

Performer reads RCS pressure from ONE of the following PAM instruments and determines that pressure is less than 300 psig: RVLIS-ICCM PLASMA DISPLAY on 1-M-4 or 1-M-6, Loop 4 HL PRESS 1-PI-68-70, Loop 3 HL PRESS 1-PI-68-64, Loop 2 HL Press 1-PI-68-63.

Examiners will note performance quality and adherence to standards in determining satisfactory completion of tasks.


Question: The NRC's assessment of the Heat Exchanger tubes involved considerations for aging effects such as cracking.

Heat Exchanger (SI Pump Casing Seal Water) Tubes | Pressure Boundary | Stainless Steel | Treated Borated Water >60°C (>140°F) (Int) | Cracking | Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19)

Piping Elements | Leakage Boundary (spatial) | Glass | Plant Indoor Air (Ext) | None | None


Question: The NRC is working on mitigating adverse impacts related to the construction of Kemmerer Power Station Unit 1 and consulting with local Tribes on these plans.

The NRC staff intends to share a preliminary draft MOA with consulting parties in March 2025.

This historic property is a multi-component prehistoric and historic-era artifact scatter with both surface and sub-surface contexts (see figure 31 of Enclosure 2). Identified artifacts include one Late Archaic and one non-diagnostic projectile point, other worked stone tools and debitage, and various metal and glass historic items.


Question: What is the role of primary protection systems in nuclear safety?

Primary protection systems, such as those identified in the table, are crucial components in ensuring the safety and operational integrity of nuclear facilities, providing the first line of defense against potential hazards.

The equipment listed in the table includes various components such as the Seal Water HX and Component Cooling systems, which serve specific operational functions within the nuclear facility.


Question: public information officer responsibilities

The Public Information Officer (PIO) is responsible for ensuring that accurate and timely information is provided to the public during an emergency situation.

The Office of Homeland Security and Emergency Preparedness is committed to enhancing the resilience of our community against various hazards.


Question: The analysis of background-subtracted peaks is essential for accurate quantification in nuclear measurements.

The background subtract report provides critical information on the original area and subtracted area of nuclear peaks, along with associated uncertainties, essential for understanding the measurement accuracy.

The results illustrate various peaks in the energy spectrum, and their corresponding net peak areas help in understanding the overall signal characteristics.


Question: The figures illustrate the core exit pressure values measured and predicted for the PB2 reactor types.

The measured and calculated core exit pressure for PB2 TT1 and TT2 are shown in Figure C.5-14 and Figure C.5-15.

This report includes detailed analyses of various reactor designs and their compliance with safety standards.


Question: What does the confidence index threshold indicate in nuclide analysis?

Nuclide confidence index threshold = 0.30


Question: The control system relies on multiple channels to ensure that operators can manually initiate emergency actions when needed.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

The required radiation monitors remain OPERABLE if one or both Emergency Exhaust System trains are inoperable or following a Fuel Building Ventilation Isolation Signal (FBVIS).


Question: What is the minimum detectable concentration in nuclear analysis?

**MDC:** Minimum Detectable Concentration

The sample analysis requires careful calibration to ensure accuracy.


Question: groundwater drilling methods

The equipment used for groundwater drilling includes a stainless steel hand auger, which is suitable for shallow depths and can collect soil samples effectively during the drilling process.

The project involves drilling information that includes the date of drilling, depth measurements, and various drill types employed in the field.


Question: online monitoring data retrieval

The first step in performing transmitter drift monitoring is to retrieve the OLM data. OLM data must be retrieved during periods of startup, normal operation, and shutdown.

The AMS CRS software provides functionality for these tasks and will be used to perform data qualification.


Question: The aging management program must adhere to the guidelines established in NUREG-2191 to ensure safety.

The maximum allowable operating temperature for this reactor type is critical in determining performance.


Question: What is radiological access control in nuclear plants?

Radiological access control is now automated using computerized radiation protection work processes. Radiological access control is primarily completed using electronic dosimetry, which is obtained prior to entry into Radiologically Controlled Areas (RCAs).

The automated whole-body monitors are not available, hand-held friskers can be used for personnel contamination monitoring.


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

The weighted mean activity of a nuclide is crucial in determining its safety and regulatory compliance, reflecting the average radioactivity per gram of the substance based on its isotopic composition.

Errors quoted at 2.000sigma indicate the confidence interval for the measurements performed during the analysis.


Question: steam dump load controller

The steam dump load controller would control at 4.7F (minus 3 for rod control) and would be at (9.43% per Degree) 44.3% open.

10CFR55.41: 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: tricolored bats habitat preferences

Tricolored bats primarily roost in trees, often selecting tall, large diameter trees.

Surveys confirm the presence of tricolored bats within the Oak Ridge Reservation, which are more specifically suited to roosts such as: - Live trees - Snags greater than 3 inches in diameter at breast height.


Question: What is the meaning of nuclide Minimum Detectable Activity?

The Minimum Detectable Activity (MDA) is the lowest level of activity that can be detected with a specified level of confidence under specific measurement conditions, reported in picocuries per gram (pCi/grams).


Question: NELAP certification requirements

Where the analytical method has been performed under NELAP certification, the analysis has met all of the requirements of the NELAC standard unless qualified on the Certificate of Analysis.

Please direct any questions to your Project Manager, Edith Kent.


Question: The NRC evaluates technical safety aspects in various renewal applications.

The NRC requires licensees to submit technical specifications that demonstrate compliance with safety regulations.

The sample results indicate that the detected concentration of sulfate was 3320 mg/L, exceeding the PQL.


Question: The NRC mandates that all components must meet specific safety classifications to ensure regulatory compliance.

Passive components in the valve program are classified as Safety Class 2, which is critical for maintaining reactor safety.

The unit dimensions of the valves are listed in the table, indicating sizes ranging from 0.5 to 48 inches.


Question: nuclide MDA report


Question: What is TSTF-505 in the context of nuclear licensing?

The proposed amendment would modify the licensing basis of Monticello to allow for the implementation of the provisions of Technical Specification Task Force (TSTF) Traveler TSTF-505, "Provide Risk-informed Extended Completion Times – RITSTF Initiative 4b" (ADAMS Package Accession No. ML182694041).

The audit plan was provided to the licensee on July 29, 2020 (ADAMS Accession No. ML20211M275). The audit plan included audit questions for the discussions between NSPM and the NRC staff during the week of August 17, 2020.


Question: What is the purpose of the NRC's small business program?

It is the policy of the U.S. Nuclear Regulatory Commission (NRC) to provide the maximum practicable prime and subcontract opportunities to small businesses in support of the agency’s mission and operational requirements.

The Bundling Review Process is designed to evaluate the impact of bundling on small businesses and ensure fair competition.


Question: containment isolation verification

**CHECK Cntmt isolation:** a. Phase A isolation: - Train A GREEN. - Train B GREEN. b. Cntmt vent isolation: - Train A GREEN. - Train B GREEN.

**ACTUATE** Phase A and Cntmt Vent Isolation signal, **OR** Manually **CLOSE** valves and dampers as necessary.


Question: What is the significance of subtracted area in nuclear instrumentation?

The subtracted area represents the net peak area after the ambient background has been removed from the original measurement, providing a more accurate representation of the peak corresponding to the radioactive source.

Errors quoted at 2.00 sigma indicate that the reported uncertainties are derived from a statistical estimation method that provides a measure of confidence in the reported data.


Question: The average calculated concentration of radioactive materials at various distances from the site is crucial for assessing potential exposure levels.

The data shows various miles to radius assessments, which help in evaluating geographic impacts.


Question: What is the purpose of the One-Time Inspection program in managing loss of material in nuclear components?

The One-Time Inspection (B.2.1.20) program will be used to manage loss of material of the aluminum alloy ducting and components, and heat exchanger fins exposed to air – indoor uncontrolled, air – outdoor, condensation in the Control Room Ventilation System, Diesel Generator and Auxiliaries System, Isolation Condenser System, SBO Diesel Generator Ventilation System, and Safety-Related Ventilation System.

Consistent with NUREG-2191 with exceptions. The One-Time Inspection (B.2.1.20) and Water Chemistry (B.2.1.2) program will be used to manage cracking of the carbon or low alloy steel with nickel alloy cladding and stainless steel heat exchanger components, piping, piping casings, and valve bodies exposed to treated water >140°F in the Control Rod Drive System, Nonsafety-Related Ventilation System, Reactor Water Cleanup System, and Shutdown Cooling System.


Question: door dynamic analysis

Using TMD results as input, the door dynamic analysis is performed using the “DOOR” Program. This computer program has been developed to predict door dynamic behavior under accident conditions.

All portions of the door and frame show factors of safety greater than one. The general acceptance criterion is that stresses be within the allowable limits of the AISC-69 Structural Code.


Question: land use impacts during decommissioning

The proposed action does not include a change in onsite land use or additional structures. No additional workers whose housing could offsite land use are proposed.

There are no additional emission sources expected to be added to the PTN site, and licensees are already required to operate within permit requirements.


Question: Licensees must estimate public radiation doses to ensure no individual exceeds set limits during their operations.

When dose rates in an area are high enough that a member of the public could receive a dose in excess of 0.02 mSv [2 mrem] in any one hour or 1 mSv [100 mrem] in a year, licensees must take additional measures to prevent public access to these higher dose rates, such as building enclosures around the gauges.

If, after making a public dose estimate, the conditions used to make the evaluation change (e.g., there are changes to the location of gauges, changes to the type or frequency of gauge use, addition of new gauges, or changes to the occupancy of adjacent areas), then the licensee must perform a new evaluation to ensure that the public dose limits are not exceeded and take corrective action, as needed.


Question: What does it mean for the Reactor Coolant System to operate close to DNB?

Which of the following will cause the RCS to operate closer to DNB?


Question: probabilistic operational assessments

Fully probabilistic OAs required to obtain 72 EFPM

Crack growths estimated from eddy current sizing contain large uncertainties


Question: construction permit application

By letter dated March 28, 2024, TerraPower, LLC... submitted a construction permit (CP) application and corresponding preliminary safety analysis report (PSAR) to the U.S. Nuclear Regulatory Commission (NRC) staff for a reactor facility pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, “Domestic Licensing of Production and Utilization Facilities,” and section 103 of the Atomic Energy Act of 1954, as amended.

The NRC staff plans to audit material which may be needed to make a reasonable assurance finding that the preliminary design protects public health and safety.


Question: What actions are required to vent the VCT at power?

Vent the VCT per BOP CV-34, VENTING THE VCT AT POWER.

Due to VCT pressure, RCP seal leakoff has ...


Question: What is Import/Export Licensing in nuclear regulations?

Import/Export Licensing & Consultations – General

Code   | Description                                                              |


Question: What are the requirements for license renewal in nuclear facilities?

The license renewal process for nuclear facilities requires a comprehensive safety review and an assessment of the environmental impact of continued operation.

Mr. Randall K. Edington serves as the Vice President-Nuclear and CNO at Nebraska Public Power District.


Question: What are the safety systems employed in nuclear reactors?

Safety systems are established to protect the reactor core and prevent the release of radioactive materials in the event of an accident.

In the case of a malfunction, operators must follow the emergency procedures outlined in the operations manual.


Question: loss of condenser vacuum

15.2.5 (Section 3.13) Loss of Condenser Vacuum describes an event where the vacuum in the condenser is lost, potentially affecting the heat exchange process in the reactor cooling system.

The table lists different events in Chapter 15, such as Loss of All Grid Connections and Loss of Feedwater, but these events pertain to different risks and system management protocols.


Question: The license examination outlines are created based on guidelines in NUREG 1021, Revision 11.

The examination is being prepared in accordance with NUREG 1021, Revision 11.

The NRC also oversees the transportation of radioactive materials to ensure safety during transit.


Question: license transfer application regulations

While the Local Petitioners are correct that the exemption request referenced in the application and later submitted to the NRC is subject to an environmental review, they did not challenge any specific part of the environmental analysis HDI provided in support of the exemption request.

We therefore conclude that the Local Petitioners have not submitted an admissible contention.


Question: confidence index for nuclides

Nuclide confidence index threshold = 0.30


Question: high-furnace fluid samples

Measure fluidity, hardness, & chemistry of samples (8-10, includes porcelain)

Provide data supporting conclusions drawn from other methods (NIST- based) processes


Question: The regulatory process requires extensive documentation to ensure that safety-related systems meet the required standards.

The NRC mandates that any proposed changes to the existing license require a thorough safety analysis to ensure compliance with regulatory standards.

The cooling system at Vogtle Units 1 and 2 is designed to operate under high temperatures and pressures while maintaining stability.


Question: What are the target points for measuring surface dose rates in a concrete cask?

Measure dose rates at eight target points (0°, 45°, 90°, 135°, 180°, 225°, 270°, and 315°) on each plane, at center of each inlet and outlet and at a point in between each inlet and outlet.

The inspection includes a thorough review of all mechanical systems used in the reactor containment.


Question: What is the probable maximum flood (PMF) scenario in nuclear safety evaluations?

The peak discharge during the governing PMF event when all dams were assumed to fail is 1,491,618 cfs.

The PMF hydrograph at Three Mile Island is presented in Figure 12.


Question: The NRC has established requirements for conducting Aging Management Reviews on various components of the steam generator system.

Table 2.3.1-4 lists the SG System component types that require an AMR and their associated component intended functions.

Table 3.1.2-4 provides the results of the AMR.


Question: remote work practices for nuclear inspectors

The survey also provided insights as to what respondents consider to be the appropriate amount of telework for resident inspectors.

A perspective was provided that this recommendation should not be to expand but rather to evaluate the use of remote work practices for resident inspectors.


Question: What is an Ultimate Heat Sink in nuclear power plants?

As described in Regulatory Guide (RG) 1.27, Revision 2, 'Ultimate Heat Sink for Nuclear Power Plants,' the predicted response of the UHS temperature to the design basis event is a function of the historical weather including the diurnal variations.

The proposed amendment is to change Technical Specifications (TS) Surveillance Requirement (SR) 3.7.9.2 to allow an Ultimate Heat Sink (UHS) temperature of ≤ 102.8°F until September 30, 2023.


Question: Residual stresses during the manufacturing process can lead to significant issues if they are not properly addressed.

Residual stresses form during the LPBF build process and can lead to warping, cracking, and delamination if not properly managed.

Heterogeneity and anisotropy generally manifest as different properties in the build direction relative to the other two directions due to the nature of the layer-by-layer build process.


Question: What is involved in the severe accident mitigation alternatives audit plan?

The NRC staff plans to conduct an in-office audit related to the severe accident mitigation alternatives, beginning February 13, 2023, by teleconference and online reference portal.

The NRC staff is reviewing the Vistra Operations Company LLC license renewal application for Comanche Peak Nuclear Power Plant, Unit Numbers 1 and 2.


Question: The NRC conducts audits to assess the adequacy of the information provided by licensees and their compliance with regulatory requirements.

The NRC staff will request information throughout the audit period.

The following table identifies the NRC audit team members, including contractors, and their respective focus areas:


Question: environmental impacts assessment for nuclear projects

The environmental report must contain a discussion of alternatives that is sufficiently complete to aid the NRC in developing appropriate alternatives.

These activities would result in environmental impacts that need to be assessed.


Question: What is the purpose of the Emergency Core Cooling System (ECCS) in reactor design?

Specific exemption requests, pertaining to requirements that concern the Emergency Core Cooling System (ECCS) and the containment cooling mode of the ECCS function for core cooling and containment heat removal following a postulated loss of cooling accident (LOCA), are provided as follows:

DTE requests, in Attachment 1, relief from a solely deterministic method in order to enable the use of the risk-informed method to demonstrate acceptable ECCS and containment heat removal performance with regard to the effects of LOCA generated debris.


Question: The NRC also mandates that the maximum concentrations must be evaluated to ensure that they do not exceed the risk-based concentrations.

Determine maximum risk-based concentrations (RBCs) associated with the scenario.

Pond monitoring will continue as part of the long-term monitoring program.


Question: What is the Engineered Safety Feature Actuation System Instrumentation?

3.3.2.1 Engineered Safety Feature Actuation System Instrumentation

As shown in Table 3.3-3. ACTION 20 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable status within 24 hours or, be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1.1; provided the other channel is OPERABLE.


Question: What are the requirements for document control in nuclear licensing?

In establishing procedural controls, GEH commits to compliance with 10 CFR Part 50 Appendix B, Criteria VI, and NQA-1-2015 Edition, Requirement 6.

Changes to documents, other than those defined in implementing procedures as minor changes, are reviewed and approved by the same organizations that performed the original review and approval unless other organizations are specifically designated.


Question: What is the purpose of the RBS Attachment to the Louisiana Peacetime Radiological Response Plan?

The RBS Attachment to the Louisiana Peacetime Radiological Response Plan (LPRRP) is intended as a tool and reference for emergency actions and procedures in response to an emergency at River Bend Station (RBS).

Drills are supervised instruction periods to test, develop and maintain skills in the EOC operations and related areas in the Parish.


Question: What is the purpose of site-specific analysis in nuclear license renewal?

Site-Specific Analysis for MNGP SLR: An analysis of the environmental impact, taking into account information provided in the GEIS and 10 CFR Part 51, Subpart A, Appendix B, as well as current site-specific information.

Chapter 4 generally follows Regulatory Guide 4.2, Supplement 1, Revision 1 (NRC 2013b) regarding content for the license renewal issues identified in 10 CFR Part 51, Subpart A, Appendix B, Table B-1, modified, as described below, consistent with the Commission’s ruling in CLI-22-2.


Question: The license renewal process for nuclear plants requires comprehensive safety evaluations...

The Applicant's Environmental Report presents an assessment of the potential environmental impacts associated with the continued operation of the Monticello Nuclear Generating Plant Unit 1, as part of the subsequent license renewal process.

In 1969, the construction of the Monticello Nuclear Generating Plant was nearing completion, marking a significant milestone in its development.


Question: What is the purpose of the refueling water storage tank in nuclear reactors?

The minimum storage volume in the RWST and the accumulators is sufficient to ensure that, after a Reactor Coolant System break, sufficient water is injected and is available within the Containment to permit recirculation cooling flow to the core, and to meet the net positive suction head requirements of the residual heat removal pumps.

The component design pressures and temperatures are presented in Table 6-123.


Question: inorganic analytical results

The laboratory report includes a comprehensive list of inorganic analytical results for various parameters, such as the concentrations of metals like calcium, magnesium, and uranium, measured in milligrams per liter (mg/L).

Total alkalinity is measured as an indicator of water quality, with values recorded alongside analytical dates and analyst initials.


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX code provides a critical heat flux correlation that is essential for evaluating thermal hydraulic performance in nuclear reactor core designs.

Local conditions from COBRA-FLX simulations for grid tests are evaluated to determine the performance of fuel assemblies under varying operational scenarios.


Question: What are the standard operating procedures for analytical services in nuclear projects?

This original data report has been prepared and reviewed in accordance with GEL's standard operating procedures.

Our policy is to provide high quality, personalized analytical services to enable you to meet your analytical needs on time every time.


Question: The NRC conducts audits to enhance its comprehension of license amendment requests and assess additional information as necessary.

During the review of the LAR, the NRC staff identified items that require further clarification and detailed explanations.

By letter dated September 29, 2023, Constellation Energy Generation, LLC submitted a license amendment request for Braidwood Station.


Question: What is the significance of measurement units in nuclear regulatory filings?

If the Rec% column is null, the high/low limits are in the same units as the result. If the Rec% column is not null, then the high/low limits are in % Rec.

The NRC requires licensees to implement quality control measures to ensure accurate data reporting.


Question: What are the habitat requirements for federally protected species in the context of nuclear site operations?

Section 4.14.6.3 states that existing procedures consider impacts to threatened and endangered species and their habitats as part of operations and maintenance.

Please provide: TVA. 2021b. Clean Water Act Section 316(b) § 122.21(r) Compliance Documentation for the BFN Plant, Limestone County, Alabama. April.


Question: What does proprietary information mean in the context of nuclear licensing?

Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

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


Question: What is the function of a Loss-of-Coolant Analysis in nuclear safety?

Loss-of-Coolant Analysis examines the potential consequences of a loss of coolant accident (LOCA) in a nuclear reactor system, providing essential data to determine safety margins and response strategies.

Long Term Transient Analysis Program for PWR’s (BLKOUT Code) utilizes advanced modeling techniques to simulate various transient scenarios during reactor operations, which aids in understanding system behavior.


Question: detector current comparator

PO places DETECTOR CURRENT COMPARATOR, UPPER SECTION, switch in PRN41 position AND ENSURE the following: ___ CHANNEL DEFEAT lamp illuminates.

PSEG CONFIDENTIAL - Possession Requires Specific Permission from Nuclear Training


Question: The analysis of wind speed and direction is essential for understanding the atmospheric stability conditions at the site.

VEPCO North Anna JFD 35-Foot Level—Period of Record: MAR 1/57-4 - 30/475 Stability Class: G—Elevation: 35 Feet.

The total number of observations recorded for March was 641, with 0 hours of calms registered.


Question: The NRC mandates an evaluation of safety implications during the licensing process of reactor components.

The GNF NPC Package Assembly is detailed under Patent #6,166,391, which provides technical specifications and safety considerations for the packaging of nuclear material.

CTU-1 Post-Test Disassembly; View of OCA Residual Foam Block shows the internal components after testing but does not discuss technical specifications.


Question: The Commission has determined that the proposed changes will not jeopardize public health and safety.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission’s regulations;

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated July 27, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations set forth in 10 CFR Chapter I;


Question: emissions inventory for the CUP

Table 4.6-1. Preliminary Emissions Inventory for the CUP (t/yr) shows the cumulative emissions across various phases including construction, operation, aquifer, decommissioning, and their corresponding total particulate matter.

Decommissioning is a complex phase involving the safe dismantling of the nuclear facility and ensuring that all components are managed in accordance with regulatory frameworks.


Question: Licensees must submit detailed reports that outline the safety measures in place during reactor operation.

The Nuclear Regulatory Commission (NRC) mandates that all licensees perform a comprehensive risk assessment to identify potential hazards and develop effective mitigation strategies, thereby ensuring safe reactor operations.

The project team reviewed the design specifications of various components involved in the soil package construction.


Question: What are the design criteria for flood events at the WCS CISF?

The WCS CISF is not in a floodplain and is above the Probable Maximum Flood elevation and will remain dry in the event of a flood.

For NUHOMS® Systems: Transfer Cask Horizontal side drop or slip down: 80 inches(3)


Question: nuclear licensing process requirements

The nuclear licensing process involves a series of regulatory requirements that must be met to ensure the safety and security of nuclear facilities, including comprehensive safety analyses, assessments of environmental impacts, and evaluations of financial qualifications of prospective operators.

The report outlines several options for improving operational efficiency, including optimizing maintenance schedules and enhancing training programs for staff, which do not relate to nuclear licensing.


Question: The NRC has established guidelines for the risk-informed categorization of systems in nuclear power plants.

10 CFR 50.69 establishes a risk-informed categorization and treatment of structures, systems, and components to enhance safety and efficiency in nuclear power plants.

The document outlines the procedures for submitting petitions and how they will be reviewed by the NRC.


Question: emergency response organization (ERO) staffing requirements

The NUREG-0654 R2 technical basis for the ERO staffing guidance states that: An … RP Supervisor should be staffed within 90-minutes of an SAE ECL, or greater, and is typically staffed in the OSC. The OSC Supervisor can effectively manage the craft [personnel] resources for the additional 30-minutes prior to the [supervisory] respond, as demonstrated through drills and exercises, without compromising the staff’s reasonable assurance finding in accordance with 10 CFR 50.47(a).

The proposed emergency plan does not assign OSC maintenance supervisors for each craft. Specifically, this change reduces the number of OSC craft supervisory minimum staff ERO positions used for the Supervision of Repair Team Activities function from three (3) to one (1).


Question: The CERC is the main hub for coordinating emergency management operations at the plant site, ensuring effective communication and data handling with external agencies.

The CERC is the consolidated emergency operations facility (EOF) for Surry Power Station and North Anna Power Station.

The station communications system is designed to provide redundant means to communicate with all essential areas of the station associated with Surry Units 1 and 2 and to essential locations remote from the station during normal operation and under accident conditions.


Question: How does control rod friction impact the compliance with startup procedures?

In order to eliminate control cell friction and preclude the need for an emergent License Amendment Request, ATRIUM 10 fuel assemblies require discharge from the core or require rechanneling.

The Susquehanna Unit 2 has been able to manage rods that must remain fully inserted during the startup range by reordering the pull sequence, in accordance with BPWS rules, to mitigate the impacts of these out of sequence fully inserted rods.


Question: source term calculation in nuclear safety

For the RADTRAD calculation, a list of 60 core isotopic nuclides and their activities were extracted from the B/B Design Analyses BRW-19-0019-M / BYR19-027 (Ref. 8) for input into the RADTRAD 'NIF' for Westinghouse Fuel and Framatome (GAIA) fuel. RADTRAD uses these activities, in curies per megawatt, then applies nuclide release fractions and a specified core power to calculate a core source term.

Additionally, personnel doses are calculated for the Technical Support Center (TSC) with a control room equivalent occupancy assumption.


Question: What is the importance of secondary containment in nuclear safety?

In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.


Question: What is Magnetic Fusion Energy?

Magnetic Fusion Energy (MFE) has a low density and a high energy confinement time (τ_E).

There are three general approaches to fusion energy.


Question: The diagnostic capabilities of the RadICS platform are crucial for identifying faults and ensuring reliable safety functions.

Hardware and software based diagnostic features of the RadICS platform provide an acceptable method of detecting and reporting computer system faults and failures in a timely manner.

Addressed in PSAIs 7.8 and 7.12 and satisfied by:
  - Licensing Technical Report


Question: Tritium distillation process in regulatory filings

The analytical method for Tritium Distillation is specified as EPA 906.0 Modified, ensuring compliance with standards during evaluation.

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


Question: What is the significance of the Probable Maximum Flood (PMF) in nuclear safety assessments?

The Scenario 1 peak discharge at TMI, obtained for the PMP centered over the centroid of the Susquehanna River subwatershed (upstream of TMI) and center temporal distribution of rainfall hyetograph, is 1,529,654 cfs, with flood arrival time of 120.5 hr.


Question: The NRC mandates that portable gauge licensees must implement at least two independent physical controls to prevent unauthorized removal of gauges.

Title 10 CFR 30.34(i) requires that each portable gauge licensee use a minimum of two independent physical controls that form tangible barriers to secure portable gauges from unauthorized removal, whenever portable gauges are not under the control and constant surveillance of the licensee.

During her inspection in 2019, the inspector toured LGC Global's field office in Highland Park, Michigan, and observed the company's method of securing the portable gauges in the storage room.


Question: NAC-MPC SYSTEM inspection requirements after accidents

A Response Surveillance is required following off-normal, accident or natural phenomena events. The NAC-MPC SYSTEMs in use at an ISFSI shall be inspected within 4 hours after the occurrence of an off-normal, accident or natural phenomena event in the area of the ISFSI.

Transport of the CONCRETE CASK to the ISFSI


Question: What is the purpose of the Gaseous Waste Processing System (GWPS) in nuclear plants?

The GWPS is designed to collect, process, and store gaseous wastes generated by plant operations including anticipated operational occurrences.

In order to control the release of radioactive gases resulting from equipment failure or operator error, the GWPS design has all waste gas decay tanks (WGDTs) (seven tanks for Unit 1, seven tanks for Unit 2, and two shared shutdown tanks) isolated from each other with valves.


Question: The NRC facilitates participation by providing support for individuals needing accommodations during public meetings.

The NRC provides reasonable accommodation to individuals with disabilities where appropriate.

Link to meeting details: https://www.nrc.gov/pmns/mtg?do=details&Code=20210128


Question: What are the actions required for stabilizing PZR level in the reactor coolant system?

REALIGN charging and letdown, AND STABILIZE PZR level: • REFER TO Appendix A, Alignment of Charging and Letdown.

ADJUST seal injection flow between 8 and 13 gpm.


Question: entrainment compliance under CWA Section 316(b) Rule

What is ADEM’s final best technology available determination for entrainment under the CWA Section 316(b) Rule? Please provide documentation of this determination.

Where is the cooling water intake structure discharge/diffusers into Wheeler Reservoir located? Does the discharge location change based on whether the plant is in open mode or helper mode?


Question: Describe the features associated with the various components of the refueling system, such as the manipulator crane and fuel transfer systems.

Outline the interlocks associated with the following Refueling System components: a. Manipulator Crane Bridge and Trolley b. Manipulator Crane Hoist and Gripper c. Fuel Transfer System Conveyor Car d. Fuel Transfer System Upender e. Fuel Handling Crane

The proposed references to be provided are not applicable for this section.


Question: What is a seismic PRA in nuclear regulatory filings?

Duke Energy submitted its seismic PRA (SPRA) on December 21, 2018.

In a letter dated November 17, 2020, regarding the assessment of Oconee Station’s completion of required actions in response to the lessons learned from Fukushima, the NRC staff acknowledged and documented that the actions required by the NRC in orders issued following the accident at the Fukushima Nuclear Power Station had been completed for Oconee Station and that the NRC would continue to provide oversight of Oconee Station’s safety enhancements through the NRC’s reactor oversight process.


Question: The NRC requires licensees to ensure proper calculations for head loss through debris beds in the reactor's cooling system.

These laminar and turbulent fractions were then used to adjust the maximum conventional debris head loss from testing (1.487 psi at 9,300 gpm and 212.7°F, see Table 3.4-7) to plant conditions at 9,100 gpm and 212°F.

In the example calculation below, the variables with the subscript "1" represent testing conditions while those with subscript "2" are for plant conditions of interest.


Question: What are the loading conditions for fuel in the YANKEE-MPC System?

Fuel to be Stored in the YANKEE-MPC System includes specific loading conditions necessary for safe storage and management of nuclear fuel.

The CANISTER HANDLING FACILITY (CHF) is designed to support the handling of spent nuclear fuel with high operational reliability.


Question: What is the role of the High Pressure Core Spray System in reactor safety?

For 3. "High Pressure Core Spray (HPCS) System," the NRC staff noticed that Function 3c. "Reactor Vessel Water Level – High, Level 8" is listed with "Manual Trip" being the only diverse instrumentation available to complete the function.

In addition, RG 1.174, Revision 3, states the licensee should understand that "when the proposed licensing basis change necessitates reliance on programmatic activities as compensatory measures, the licensee should justify that this reliance is not excessive (i.e., not overly reliant)."


Question: What is an Anticipated Transient Without Scram (ATWS) in nuclear reactors?

An Anticipated Transient Without Scram (ATWS) is an anticipated operational occurrence that generates an automatic scram signal, accompanied by a failure of the reactor protection system to automatically shutdown the reactor.

The fire detection and suppression systems at BFN are plant-wide systems that protect a wide variety of plant equipment.


Question: The NRC oversees the renewal of licenses for organizations involved in nuclear activities.

CT    | New Britain  | The Hospital of Central Connecticut         | 06-02388-01  | 3001250  | Region 1       | 10/05/2020    | Amendment           |


Question: What are seismically induced LOCAs and how are they assessed?

Seismically induced LOCA frequencies were determined for WCGS using two approaches: 1. Using representative LOCA fragility parameters presented in EPRI 3002000709 (Reference 19). 2. Using site-specific LOCA fragility parameters calculated per the guidelines provided in NUREG-1903 (Reference 21).

Although some SSBI (e.g., a large break in a main steam line) would be expected to initiate containment spray (CS), the ECCS would not be required for long term core cooling unless there are subsequent failures, such as a stuck open PORV or loss of auxiliary feedwater, that would require feed and bleed cooling.


Question: The NRC mandates that applicants must demonstrate compliance with safety regulations to obtain a license.

The NRC requires licensees to submit safety analysis reports that assess the potential risks associated with operating a nuclear facility.

In 2020, the population in the East-Residential sector reached 102, indicating significant growth over the decade.


Question: What are the conditions for continuing with the startup of Unit 2 given its current state?

Unit 2 is in Mode 3 preparing for a startup.


Question: brine disposal methods

Brine will be generated from RO treatment of the process bleed and, potentially, brine concentration of the aquifer restoration water. Brine will be disposed in lined retention ponds or deep disposal wells as discussed in Section 2.1.2 of this ER.

During decommissioning all contaminated material will be reused for licensed activities, decontaminated below release limits and disposed in an approved landfill or reused, or disposed at a licensed 11e.(2) byproduct material disposal area.


Question: Nuclear regulatory control is crucial for managing the safe disposal of nuclear waste.

Keep nuclear waste under nuclear regulatory control and make the polluters pay to try to isolate it--not release it!

This proposal is even more bold and expansive than the many past efforts by the nuclear industry and its 'regulators' which Congress revoked in the 1992 and the public has repeatedly rejected over the decades of the nuclear age.


Question: technical specification compliance for reactor operators

Includes Technical Specification compliance for an RO.

Instructions: Check the applicants' license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.


Question: What is a Work Breakdown Structure (WBS) in project management?

— the product-oriented grouping of project elements that organizes and defines the total scope of the project; a multilevel framework that organizes and graphically displays elements representing work to be accomplished in logical relationships. Each descending level represents an increasingly detailed definition of a project component. Components may be products or services. The structure and code integrate and relate all project work (technical, schedule, and cost) and are used throughout the life cycle of a project to identify and track specific work scope.

— risks that were not anticipated or foreseen. Unidentified risks might originally be unanticipated because the probability of the event is so small that its occurrence is virtually unimaginable. Alternatively, an unidentified risk might be one that falls into an unanticipated or uncontrolled risk-event category.


Question: nuclear plant licensing process

The nuclear plant licensing process involves a series of regulatory steps to ensure the safe construction and operation of a facility, including thorough safety evaluations and public consultations.

In the event of an emergency, operators must follow specific protocols to ensure the safety and security of the facility, including immediate communication with local authorities.


Question: exemption from force-on-force exercises

Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 - Request for a One-Time Exemption from 10 CFR part 73, appendix B, section VI, subsection C.3.(l)(i) Regarding Annual Force-on-Force Exercises due to COVID-19 Pandemic, dated December 10, 2020

Columbia Generating Station - Exemption from Annual Force-On-Force Exercise Requirements of 10 CFR part 73, appendix B, "General Criteria for Security Personnel," subsection VI.C.3.(l)(1) (EPID L-2020-LLE-0172 [COVID-19]), dated December 17, 2020


Question: The NRC mandates the installation of overpressure protection devices for pressurizer systems.

Blowout panels designed to release at 5.0 psig

Containment is a safety feature employed to prevent the release of radioactive materials.


Question: The NRC mandates thorough evaluations of thermal limits in reactor operations to ensure safety.

The COBRA-FLX thermal-hydraulic analysis tool incorporates a new correlation for critical heat flux, which is essential for maintaining reactor safety under various operating conditions.

Regular assessments of environmental impacts are also part of the regulatory process for nuclear facilities.


Question: What is the significance of uranium isotopes in nuclear analysis?

Uranium-233/234 is reported at 1.43 pCi/g with an uncertainty of +/-0.533, indicating its presence in the sample under evaluation.

The moisture content of the soil sample was found to be 11.9%.


Question: The State IPAWS Administrator is responsible for evaluating and sanctioning applications submitted by local and State Agencies for Code of Government (COG) participation.

The State IPAWS Administrator reviews and approves applications for COGs for all local and State Agencies.

These actions can include but are not limited to evacuation orders including evacuation routes, shelter info, key information, etc.


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

The assembly average burnup is a key parameter in nuclear reactor operation, defined as the total energy produced per unit mass of nuclear fuel, measured in gigawatt days per metric ton of uranium (GWd/MTU).

The technical specifications of reactor components must be adhered to in order to ensure safety and compliance with regulatory standards.


Question: What factors contribute to flood duration parameters in nuclear safety assessments?

Flood duration parameters are critical in assessing the potential impacts of flooding events on nuclear facilities and are determined through detailed hydrological modeling and site-specific analyses.

The NRC emphasizes the importance of regulatory reviews and public consultations to ensure transparency and accountability in the licensing process.


Question: The evaluation of the Adjusted Reference Temperature is crucial for ensuring the integrity of reactor components under operational conditions.

Adjusted Reference Temperature (ART) (°F) is a critical measure used to ensure that reactor materials maintain their integrity under operational conditions, with values of 87.0 °F for the Limiting Plate and 146.4 °F for the Limiting Axial Weld documented in the analyses.

The chemistry factor is a parameter used to assess the impact of material composition on reactor safety and performance.


Question: What are the intervals for volumetric inspections of critical reactor components?

Procedures will be revised to require volumetric inspection of the pressurizer surge line hot leg nozzle every 48 years for management of EAF.

The staff also noted that the piping had been inspected in 1994 with no indications that were unacceptable.


Question: What is the NRC License SUA-1580?

The NRC License SUA-1580 governs the operations of the Crownpoint Uranium Project, ensuring compliance with safety and environmental regulations.

NuFuels, Inc. is committed to sustainable practices and the conservation of natural resources during the execution of its projects.


Question: Accident Initiated Iodine Spike

Accident Initiated Iodine Spike (0.265 µCi/gm steady state)


Question: Américium-241 definition

Américium-241, a synthetic element, is commonly utilized in various nuclear applications and has a specific activity measured in pCi/g, which indicates its radioactivity level in a sample.

The performance of the reactor depends significantly on the design of its cooling systems and the efficiency of heat exchangers.


Question: What are the design parameters for the Nuclear Service Water System?

Nuclear Service Water Flow Requirements (gpm per channel per Unit)

Heat Transfer Rates to Standby Nuclear Service Water Pond. From LOCA Unit Loads and Controlled Shutdown Unit Loads Along with Associated Station Auxiliary Loads Over Thirty Days.


Question: reactivity manipulations in nuclear licensing

Reactivity manipulations performed on the plant-reference simulator for an applicant to meet the experience eligibility requirements may be credited when the simulator, at the time of performance, meets the requirements of 55.46(c)(2)(i) and (ii).

These standards are quite high when applying them to the nuclear and thermal hydraulic characteristics.


Question: The procedure 1-ES-1.1 outlines the necessary steps to safely terminate a Safety Injection, including the directive to stop a CCP first.

1-E-0, Reactor Trip or Safety Injection, Step 4, will direct the crew to actuate Train B of Safety Injection. 1-ES-1.1, SI Termination, will direct a CCP to be stopped first to prevent filling the PZR solid.

Plausible because the Train A SI was inadvertent, and actuating Train B fills the PZR at a faster rate. The second part is plausible because the SIPs have a higher flow rate than the CCPs, but they also have a lower discharge pressure, which is why a CCP is directed to be stopped first.


Question: What are the surveillance requirements for ECCS automatic valves?

Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program.


Question: The recalculation of completion times must occur to adapt to evolving conditions within the plant.

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

When 'Immediately' is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner.


Question: The NRC and Virginia Electric & Power Company evaluated the design of the lower supports for steam generators and reactor coolant pumps in both units.

Discussion between the Nuclear Regulatory Commission and Virginia Electric & Power Company on Steam Generator Lower and Reactor Coolant Pump Supports for North Anna Units No. 1 and 2.

The document includes a corrected copy of the original transcript from April 13, 1976, which is 164 pages long.


Question: Petitioners argue that the Environmental Report provided by Duke does not adequately address significant environmental impacts that could arise from reactor incidents, particularly with regard to the Jocassee Dam.

In their second and third contentions, Petitioners assert that Duke's Environmental Report (ER) is deficient because it fails to discuss purported new and significant environmental impacts associated with reactor accidents at ONS that could be caused by the failure of the nearby Jocassee Dam.

The Secretary of the Commission referred Petitioners’ hearing request to the Chief Administrative Judge, who, on October 4, 2021, assigned the intervention petition to this Licensing Board to rule on standing and contention admissibility/waiver matters and preside at any hearing.


Question: seismic load analysis for nuclear structures

The temporary SW jumper is being rigorously analyzed for stresses arising from load cases consisting of deadweight, thermal, and seismic loads.

Flooding protection and personnel safety will be provided by requiring double isolation for system boundaries that present a significant source of water.


Question: What are the principal aquifers in the Coastal Plain of Richland County?

The principal aquifers include several terrace formations: the Okefenokee, Sunderland, Coharie and the Tuscaloosa Formation.

Geographically, production from shallow holes is possible within any of the terrace formations.


Question: background radiation measurements

Ludlum 2221 with 43-37 probe # (Floor Monitor) measures background radiation across various surfaces, providing counts per minute (cpm). For example, the average for ambient measurements is 1781.4 cpm, while for concrete it is 2782.1 cpm.

In the Background Data Meter #97817, the measurement for cinder block shows a value of 2109.4, which indicates the level of background radiation detected.


Question: The ASME OM Code is essential for ensuring the safety and reliability of pumps in nuclear power plants.

American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code, 2022 Edition.

The pumps listed are all vertical line shaft pumps.