Question: TRISO fuel fabrication process

The TRISO-X Fuel Fabrication Facility is designed to produce TRISO fuel, which consists of small fuel particles coated with multiple layers of ceramic materials that provide excellent resistance to high temperatures and harsh conditions.

The figure illustrates the typical stratigraphy within the Valley and Ridge Province, highlighting various formations and their characteristics, such as the Fort Payne Formation and its classification as a strong unit.


Question: The analysis of the high pressure core spray (HPCS) system reveals that it will actuate several times over its operational period.

The process piping in penetration 1MC-33 directs flow from the high pressure core spray (HPCS) system to containment when the pumps are tested during refueling outages.

For the calculation associated with penetration 1MC-65, the “Backwash” transient which was assumed to occur 2,450 times in 40 years was revised to 110,000 occurrences over 60 years.


Question: flood hazard reevaluation

The unit hydrographs developed for the licensing basis PMF and used in the flood hazard reevaluation are developed from gage data and calibrated to historical events.

Spillway gate blockage due to maintenance issues or debris has been evaluated and determined to not represent a significant hazard for SQN.


Question: What is the methodology for sizing the Plume Exposure Pathway Emergency Planning Zone?

The purpose of this section is to discuss the regulatory basis that supports the PEP EPZ sizing methodology.

Also provided are examples where the NRC has considered and approved reduced EPZ sizes.


Question: The diversity of fish species in Lake Keowee is monitored to assess environmental health and ecological balance.

Since creation of the lake, Duke Energy, as well as state agencies and other organizations, have monitored Lake Keowee’s fish populations.

This species prefers warmer waters and has a lower lethal temperature limit of approximately 41 to 45°F (5 to 7°C) (Parsons and Kimsey 1954-TN920).


Question: What are the operator actions necessary during a rod withdrawal error in a nuclear reactor?

In the event of a rod withdrawal error, operators must follow specific procedures detailed in the EOP to ensure reactor safety, which includes immediate actions to stabilize reactor conditions and prevent an escalation of the incident.

The cooling system plays a crucial role in maintaining the reactor's operational temperature and preventing overheating, ensuring the safety and integrity of the reactor components.


Question: nuclear licensing requirements

The licensing process for nuclear reactors involves rigorous safety assessments, environmental reviews, and public consultations to ensure compliance with regulatory standards set forth by the NRC.

The primary functions of series circuits include providing a path for electric current and maintaining a constant current flow in all parts of the circuit.


Question: The NRC has established procedures for the submission of regulatory documents by licensees.

The enclosed Order has been forwarded to the Office of the Federal Register for publication.

Enclosures include several documents that provide detailed information relevant to the regulatory framework.


Question: An audio assessment will be conducted to detect any potential issues with valve performance.

An audio test will be performed on check valves to assist in determining valve degradation.

The centrifugal charging pump operates effectively within its designed flow parameters during normal testing conditions.


Question: The fuel assemblies' average enrichment must remain within specified limits to comply with regulatory standards.

| wt % 235U (E) |      48 < Assembly Average Burnup ≤ 49 GWd/MTU      |                                      |                             |


Question: The SGTS is designed to ensure that any released radioactive materials are treated effectively before emissions occur.

The SGTS maintains a negative pressure within the secondary containment and filters the exhaust air to reduce radioactive contamination in gases prior to their release via the chimney.

The License Renewal Boundary Drawings provide essential references for the moderation and stability of reactor components and their compliance with operational protocols.


Question: What is the significance of the cpm unit in nuclear safety monitoring?

The term 'units (cpm)' refers to counts per minute, a measure used in radiological monitoring to indicate the level of radiation detected by the instrument.

The survey check was conducted for a total of 60 minutes to ensure data accuracy during the logging process.


Question: holding time in nuclear sampling

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

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


Question: nuclide identification

IDENTIFIED NUCLIDES
Nuclide Name       Id      Energy   Yield(%)   Activity          Activity Uncertainty
                   Confidence  (keV)               (pCi/grams)

Peak Analysis Performed on 5/9/2019 7:20:40 AM


Question: The NRC mandates that systems monitor effluents to ensure public health safety.

Demonstrates that the operations will be conducted so that all airborne effluent releases are as low as is reasonably achievable.

For land applications (i) concentrations of radioactive contaminants in soils to show that levels of radium and other nuclides in the soil will not exceed the standard in 10 CFR part 40, Appendix A (ii) impacts to ground-water and surface-water quality (iii) impacts on land use, particularly crops and vegetation (iv) exposures and health risks that may be associated with radioactive constituents reaching the food chain.


Question: The NRC's amendment to the facility license mandates updates to the Emergency Plan following a review of Florida Power & Light Company's submission.

Accordingly, by Amendment No. 294, Subsequent Renewed Facility Operating License No. DPR-31 is hereby amended to authorize revision of Turkey Point's Emergency Plan as set forth in Florida Power & Light Company's application dated December 6, 2019, as supplemented by letters dated June 8 and November 5, 2020, and evaluated in the NRC staff's safety evaluation dated January 27, 2021, and to reflect the issuance of this amendment by hereby amending paragraph 3B.

This license amendment is effective as of its date of issuance and shall be implemented within 184 days.


Question: The NRC monitors environmental constituents that may exceed safety standards.

The Dakota Sandstone (Unit 1) has been found to have uranium exceeding Primary MCLs.

The Westwater Canyon (Crownpoint) aquifer has a listed pH level with respect to secondary standards.


Question: What are the conditions for entering Required Action based on system limits?

When a system is determined not to meet the LCO, Condition A is entered. If the system is not restored within 7 days, Condition B is also entered, and the Completion Time clocks for Required Actions B.1 and B.2 start.

The Completion Time for required actions can vary significantly depending on the complexity of the issue being addressed.


Question: What are the requirements for operating examinations at UUTR reactor?

Additionally, Supplement section 2. states: Andrew Allison and Edward Goodell intend to perform their annual operating exams after successful completion of all restoration maintenance and prior to the requested 1 expiration date of December 31st, 2025.

It is the staff’s understanding that RRR will be used for special refresher training and the annual operating exams will be completed after maintenance and restart of UUTR reactor.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves a systematic review by the NRC to ensure compliance with safety standards and regulations before any construction or operation can commence.

Several models of nuclear reactors have been developed over the years, including pressurized water reactors and boiling water reactors, each with unique design characteristics.


Question: high pressure coolant injection system

Pressure Boundary                           | Copper Alloy with 15% Zinc or More | Lubricating Oil (External)   | Loss of Material                 | Lubricating Oil Analysis (B.2.1.25)                  | V.D2.EP-76       | 3.2.1-050               | C     |

Heat Exchanger – (Lube Oil Cooler) Tube Side Components | Pressure Boundary     | Carbon Steel                       | Air - Indoor, Uncontrolled (External) | Loss of Material                      | External Surfaces Monitoring of Mechanical Components (B.2.23) | V.E-44           | 3.2.1-040               | A     |


Question: What are the rules for construction of fusion energy devices?

Division 4 Fusion Energy Devices (FED) issued in November 2018 a Draft Standard for Trial Use of proposed code rules entitled “Rules for Construction of Fusion Energy Devices ASME FE.1-2018.

The Draft Standard is not an approved consensus standard.


Question: radioactive isotope information

The teams conducted simulations to test the performance of the reactor under various scenarios.


Question: What are the repair capabilities addressed in the NRC staff determination for the RadICS platform?

The NRC staff determined the RadICS platform design is generally capable of supporting timely recognition, location, replacement, repair, and adjustment of malfunctioning equipment.

Some aspects of a system repair capabilities must be determined during application development and therefore compliance with this position should be confirmed during plant application development.


Question: Atrium BWR Enhanced Accident Model

The EMF-93-177, Rev. 1, Supplement 3, Atrium BWR Enhanced Accident Model topical report details the safety analysis regarding the behavior of the reactor in postulated accident scenarios, providing insights into the likelihood and consequences of potential events.

The completed U.S. Nuclear Regulatory Commission Form 897 is an important document for ensuring compliance with regulatory requirements during the licensing process.


Question: What is the critical rainfall amount defined for one-hour storm events under specific hydrometeorological conditions?

The 1-hour PMP event distribution was developed using Hydrometeorological Report 52 (HMR 52) (Reference 14). Per NUREG/CR-7046, the UEP event is defined as a 1-hour/1-square-mile PMP event.

The levee feature was used to represent the Kontek Vehicle Barriers in the FLO-2D model to account for the vehicle barrier's effects on local drainage patterns.


Question: Field applied coatings at BFN are essential for ensuring the integrity of the piping systems during repair and maintenance activities.

During BFN construction, field applied coatings were applied to steel pipe not mill coated, fittings of mill coated pipe and joints of mill coated pipe by wrapping them with coal tar protective coating in tape form.

Soil corrosiveness of the environment in which the buried pipe exists along with site operating experience and BPWORKS™ High Risk Ranking are utilized to develop direct inspection prioritization to assess effectiveness of the coating system.


Question: nickel alloys in nuclear applications

The material used for the SG Primary Nozzle to Safe End Welds is Nickel Alloys.

Boric Acid Corrosion (B.2.3.4) is a significant aging management program.


Question: CANISTER vessel hydrostatic pressure testing

The CANISTER closure lid-to-shell weld (MPC-LACBWR only) is performed in the field following fuel loading. The CANISTER is then hydrostatic pressure tested as defined in Appendix 9.A and described in Appendix 8.A.

No overpressure protection is provided. The function of the CANISTER is to confine radioactive contents under normal, off-normal, and accident conditions of storage.


Question: What is included in the SAMA analysis regarding external events?

Applicants for license renewal are required by 10 CFR 51.53(c)(3)(ii)(L) to consider SAMAs if not previously considered in an environmental impact assessment, related supplement, or environmental assessment for the plant.

In response to NRC information requests to address near-term task force recommendations following the accident at the Fukushima Daiichi Nuclear Power Plant, and in response to Generic Issue (GI)-199, new seismic hazard curves have been developed for each nuclear power plant site.


Question: What is a reactor building-to-suppression chamber vacuum breaker?

Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE.

In accordance with the Risk Informed Completion Time Program.


Question: The NRC must evaluate how medical events, particularly those involving radiopharmaceuticals, are reported and processed by licensees today.

So a medical event is not a process improvement mechanism. The other thing about medical event reporting of extravasations, this would be really unprecedented in terms of how extravasations are reported.

Nowhere else are non-radioactivity extravasations reported to any state or federal agency.


Question: Integrated Safety Analysis definitions

The purpose of the call was to discuss the options for resolution of the definitions of unlikely and highly unlikely.

The U.S. Nuclear Regulatory Commission (NRC) staff conducted a closed Microsoft Teams meeting with TRISO-X, LLC (TRISO-X) on December 16, 2024, from 3:00 p.m. until 5:00 p.m. to align on the Integrated Safety Analysis (ISA) update from the December 3-4, 2024 Audit Visit.


Question: nuclide identification report

The identified nuclides include those that are cataloged in the nuclide library used for the analysis, which provides essential data for understanding radionuclide compositions in safety evaluations.

The fitted singlet has a well-defined peak area and the uncertainties are quoted at 2.000 sigma, indicating a high level of accuracy in the measurements.


Question: nuclide MDA report

| Peak No | Energy (keV) | Peak Size (CPS) | Peak CPS (%) | Uncertainty | Peak Type | Tolerance Nuclide |


Question: The high pressure coolant injection system is critical for emergency situations in nuclear reactors.

The proposed amendment revises the allowable value for CNS Technical Specifications (TS) 3.3.5.1, "ECCS Instrumentation," Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Function 3.f, "High Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)," from ≥ 490 gallons per minute (gpm) to ≥ 523 gpm.

Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed Facility Operating License No. DPR-46 for Cooper Nuclear Station (CNS).


Question: The NRC mandates that licensees adhere to quality assurance standards including those specified in 10 CFR Part 50, Criterion XVII.

In GEH QATR QAPD Section 17.8, "Commitment," GEH commits to compliance with Appendix B to 10 CFR Part 50, Criterion XVII, NQA-1-2008 and the NQA-1a-2009 Addenda, Requirement 17, GL 88-18, "Plant Records and Storage on Optical Disk," and Regulatory Information Summary (RIS) 2008-18, "Guidance on Managing Quality Assurance Records in Electronic Media."

In Part IV, "REGULATORY COMMITMENTS," GEH commits to the following regulations and guidance documents in Part I and II:


Question: radioactive materials licensing

5. RADIOACTIVE MATERIAL - Element and its atomic number, chemical and physical form, and maximum amount.

The applicant understands that all statements and representations made in this application are binding upon the applicant.


Question: What is the minimum detectable concentration for radioanalytical measurements?

The sample type analyzed was pCi/g dry, indicating a specific activity concentration measurement.


Question: The analysis of surface soil samples revealed the presence of radionuclides like Cs-137 and Co-60, indicating potential contamination.

For surface soil samples, Cs-137 was detected at concentrations above Minimum Detectable Concentration (MDC) in eight (8) of the samples, at a maximum concentration of 1.39E-01 pCi/g.

In total, sixteen (16) surface soil samples, nineteen (19) subsurface soil samples, and nine (9) asphalt samples were collected in survey units L2-011-101 and L3-012-101.


Question: What is an evapotranspiration cover system in nuclear site management?

HMC is proposing to modify the 1995 NRC-approved large tailings pile (LTP) cover design as an evapotranspiration (ET) cover system.

Excavation of the borrow areas would not exceed a depth of 3 m (10 ft) below ground surface.


Question: class 1E electrical distribution systems

3.8.9 Distribution Systems - Operating | Class 1E AC, DC, and 120 VAC vital bus electrical power distribution subsystems | Yes | (1) Provide necessary power to the ESF systems

Inoperability of the pressurizer heaters will be conservatively bounded by assuming an increase in the frequency of a reactor trip initiating event by a factor of 10; this reflects the adverse impact on pressure control due to inoperable pressurizer heaters.


Question: What are the initial conditions for a plant startup?

A plant startup is in progress. Another operator is stationed to adjust 71T-4 load tap changer, if required. Nine Mile Point Unit 1 Control Room and Power Control have been notified of all impending bus transfers and tap changer operations. All prerequisites have been met for the bus transfer. Both RPS buses are powered from their respective MG sets. Main Generator load is approximately 80-85 MWe.

Transfer Bus 10100 from Reserve to Normal per OP-46A Section D.20.


Question: The analysis of potential accidents is crucial for ensuring safety in nuclear reactors.

The ARITA methodology helps in assessing the risk of non-loss of coolant accidents (non-LOCA) while considering various safety parameters.

The turbine system efficiency is often evaluated during peak load conditions to optimize power generation.


Question: What considerations are made for the operational configuration of the AFWS during safety evaluations?

The AFWS can only be configured in modes consistent with the preliminary Technical Specifications and the preliminary plant procedures.

In order to properly model the AFWS, certain assumptions had to be made; in order to be consistent with the NRC's generic evaluation of AFWSs, the scope of the analysis had to be altered accordingly.


Question: structural support aging management

Loss of material due to pitting and crevice corrosion, cracking due to SCC

One-Time Inspection (B.2.1.20) is a crucial component of the aging management program.


Question: pitting corrosion in pressure boundary components

Loss of material due to general, pitting, crevice corrosion.

Loss of material due to selective leaching.


Question: The NRC evaluates the complexity associated with design changes, often categorizing them by the number of CDA submissions required.

Critical systems often face high complexity levels, requiring significant evaluation due to a large number of CDAs and internetworking with advanced operating systems.

Component-level replacements are typically less complicated and may include straightforward tasks like minor adjustments to existing systems.


Question: What are the specific requirements for steam generator tube inspections in nuclear plants?

The inspections of the steam generator tubes are performed to ensure that they meet the regulatory safety standards set forth by the NRC, including evaluations for wear, corrosion, and structural integrity during the operational lifecycle of the nuclear plant.

During the conference call, various stakeholders discussed the future operational plans for Unit 2, highlighting the importance of regular maintenance and support systems to optimize performance.


Question: Chi/Q calculations for nuclear accident analysis

Chi/Q is expressed in sec/m³ and refers to the ratio of concentration of radioactive material to the rate of release in a specific downwind sector.

The table indicates distances in meters for the Low Population Zone (LPZ) surrounding the Clinton Power Station, which are crucial for assessment in various environmental analyses.


Question: The amendment for the Edwin I. Hatch Nuclear Plant, Unit No. 2 has been submitted, ensuring adherence to the Atomic Energy Act and regulatory standards established by the NRC.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 11, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations as set forth in 10 CFR Chapter I;

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission.


Question: What is the purpose of the ADVANCE Act of 2024 in nuclear oversight?

Section 507 of the Act requires the NRC to develop and submit to the appropriate committees of Congress a report that identifies specific improvements to the nuclear reactor and materials oversight and inspection programs carried out pursuant to the Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.) that the Commission may implement to maximize the efficiency of such programs through, where appropriate, the use of risk-informed, performance-based procedures, expanded incorporation of information technologies, and staff training.

The meeting notice and agenda are located at Agencywide Documents Access and Management System (ADAMS) Accession No. ML25014A022.


Question: The NRC mandates that operators must assess conditions related to the failure of safety systems such as level transmitters.

SG #1 Level Transmitter 1-LT-3-38 Fails Low

Failure Of Automatic MFW Pump Control or Instrument is addressed in section 3.6.


Question: The activity of a nuclide is measured in picocuries per gram (pCi/grams)...

Confidence values are derived based on the established criteria for nuclear licensing.


Question: safety-related systems in nuclear reactors

The current data set includes information on various safety-related systems critical to maintaining the integrity and operational safety of the reactor environment.

The design specifications for the control room layout ensure that operators have direct access to vital information and controls necessary for efficient operations.


Question: What does SLRA Table 4.3.1-1 indicate about the projected cycles for test-condition transients?

Specifically, SLRA Table 4.3.1-1 describes the current cumulative cycles (up to 12/31/2019) and 80-year projected cycles for the test-condition transients (i.e., “turbine roll test,” “primary side hydrostatic test,” and “primary side leakage test” transients).

The staff needs to clarify whether some of the tests will be conducted during the normal operation or other (non-test) transient events such that there is no need to separately count the test transients.


Question: The Maintenance Rule (MR) monitoring programs under 10 CFR 50.65 ensure the evaluation of the unavailability impacts stemming from the RICT initiative.

The Maintenance Rule (MR) monitoring programs under 10 CFR 50.65 provide for evaluation and disposition of unavailability impacts that may be incurred from implementation of the RICT program.

In addition to impacting cumulative risk, the implementation of the RICT program may potentially impact the unavailability of SSCs.


Question: regulatory filings for nuclear safety analysis

Nuclear regulatory filings include detailed safety analyses that demonstrate compliance with safety regulations and standards.

The reactor pressure vessel is designed to withstand high pressures during operation and is a critical component of the overall reactor design.


Question: What are the health services provided by the Nebraska Department of Health and Human Services?

The Nebraska Department of Health and Human Services offers a variety of health services including health promotion, health disparities, and lifespan health services.

Sara Morgan oversees the Health Promotion & Prevention Section, which focuses on improving public health and addressing health disparities among populations.


Question: The NRC establishes the necessary qualifications and responsibilities for Reactor Operators and Senior Reactor Operators according to 10 CFR Part 55.

NUREG-2103, “Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000,” provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The catalog is designed to ensure equitable and consistent examinations.

This catalog is organized into six major sections. K/As are grouped according to the major section to which they pertain.


Question: What is the process for transferring nuclear devices according to NRC requirements?

Transferred to another general licensee (Complete Parts 2 and 3)

Provide information about devices listed in Section 2 or 6, but no longer in your possession.


Question: What are the appendices included in the NRC Phase 1 guidance revisions?

The following Appendices were developed during Phase 1: Appendix A, “Qualitative Factors Assessment Tools;” Appendix B, “Cost Estimating and Best Practices;” Appendix C, “The Treatment of Uncertainty;” Appendix D, “Guidance on Regulatory Analysis Related to American Society of Mechanical Engineers (ASME) Code Rules;” and Appendix E, “Special Circumstances and Relationship to Other Procedural Requirements.”

The NRC staff published the draft revision, in the Federal Register (82 FR 18163, April 17, 2017) for a 60-day public comment period.


Question: equipment tripped in load shedding

Equipment tripped whenever load shedding is initiated includes various high-pressure and low-pressure heater drain pumps for both units.

Unit 1 Start-up Feed Pump 1-FW-P-1A Running indicates the operational status of feed pumps during startup but does not relate to load shedding.


Question: What does lamellar tearing refer to in welding?

This photo shows that lamellar tearing is continuous with the weld cracking - not 'fortuitously ... in the vicinity of a large weld defect.'

Whether one caused the other cannot be determined by examination.


Question: What precision is specified for the wind direction measurement?

The wind direction, wind speed, sigma theta, and 197-33 ft. delta-temperature (0.01° C precision) sensors added at the 197 ft. level.


Question: wind energy feasibility for baseload power

As is the case with other renewable energy sources, the feasibility of wind energy providing baseload power depends on the location (relative to electricity users), value, accessibility, and constancy of the resource.

Biomass resources used for biomass fuel-fired power generation include agricultural residues, animal manure, wood wastes from forestry and industry, residues from food and paper.


Question: What is the analytical method used for Tritium Distillation in nuclear testing?

- **Analytical Method:** EPA 906.0 Modified

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


Question: What is the procedure for appealing a denial of unescorted access by the NRC Facilities Security Branch Chief?

When the NRC Facilities Security Branch Chief makes a determination to deny or revoke unescorted access for an individual nominated as a reviewing official, the individual shall be afforded an opportunity to appeal this determination to the Director, Division of Facilities and Security. The determination must be appealed within 20 days of receipt of the written notice of the determination from the NRC Facilities Security Branch Chief.

Before the NRC Facilities Security Branch Chief makes a determination to deny or revoke unescorted access to the ISFSI for an individual nominated as a reviewing official, the individual shall be afforded an opportunity to submit information relevant to his or her trustworthiness and reliability.


Question: The State Emergency Operations Center coordinates assistance during emergencies with local and federal agencies.

Determine and issue, through the State EOC, the appropriate EAS message.

Inform all emergency response personnel that full recovery and unlimited reentry has begun.


Question: Columbia Fuel Fabrication Facility inspection report

The Columbia Fuel Fabrication Facility is licensed under License Number SNM-1107 and operates within the regulatory framework established by the U.S. Nuclear Regulatory Commission to ensure compliance with safety and operational standards.

The inspection report indicates that the facility has undergone multiple evaluations by inspectors, focusing on the adherence to regulatory requirements and operational protocols.


Question: What are the requirements for CANISTER transfer in nuclear operations?

The CANISTER shall be transferred from the TRANSFER CASK to a CONCRETE CASK, or to a transport cask, or returned to an appropriate location for UNLOADING OPERATIONS.

Separate Condition entry is allowed for each NAC-MPC SYSTEM.


Question: What is a Liquid Release Form in nuclear regulatory filings?

Updated Liquid Release Form for BCP 400-TWX01 revision 69.

This JPM previously administered as JPM SA-d on 2016 NRC Exam.


Question: what is a rem in radiation measurement?

Rem (Roentgen Equivalent Man) is a unit of radiation dose that relates exposure to the biological effects of the exposure (absorbed exposure or dose). A unit related to the rem is the millirem (mrem). 1 mrem = 1/1000 rem.

An Operational Support Center (OSC) is an assembly area that serves as the staging location for Damage Control Teams, the Fire Brigade, the First Aid Team, and the Search and Rescue Team.


Question: What does the flood analysis entail for nuclear facilities?

The flood analysis section provides a framework for assessing the Probable Maximum Flood (PMF), which is critical for ensuring that the facility can withstand extreme flood conditions and maintain safety operations.

The engineering properties of foundation materials play a significant role in determining the construction viability of structures in seismic zones.


Question: What is the purpose of the HPCI system in nuclear reactors?

The HPCI system, or High Pressure Coolant Injection system, is designed to provide emergency core cooling and maintain reactor safety during accidents that could lead to overheating of the reactor core.

The Control Center Emergency Air Division I Radiation Monitor is critical for monitoring radiation levels and ensuring the safety of personnel during operational activities.


Question: What does the term best estimate signify in the context of regulatory filings?

The applicant will come in with different definitions of what their best estimate evaluation might be or how they're going to apply conservatisms or no in various aspects of the analysis.

I think what industry presented today on their understanding of what best estimate today aligns very closely with what we think other than possibly the sampling over the break size -- entire break size range based on frequency.


Question: The evaluation of the adequacy of safety systems is a critical component of the regulatory process.

A safety system must be designed to mitigate potential accidents and to ensure that the reactor operates within approved safety margins.

The inspection team consisted of several experienced professionals who conducted thorough reviews of the plant's operations.


Question: The NRC mandates that valve operators adhere to specific actuation signals for safety compliance.

The actuation signal for the RHR PUMP TO COLD LEG Valve No. NI174 is MANUAL/ REM, indicating the operator's control over the valve's operation.

The isolation time for the RHR OUT FROM SUMP Valve No. NI184B is designated as OPEN/ CLOSED, reflecting its operational flexibility.


Question: What is a small-break LOCA in nuclear reactors?

Ability to determine and/or interpret the following as they apply to a Small-Break LOCA: Existence of head bubble.

The NRC requires licensees to analyze emergency zones for potential impacts during operations.


Question: The NRC guidelines specify that accidental releases from the recycle holdup tank must be thoroughly modeled and their consequences evaluated.

This accident is defined as an unexpected and uncontrolled postulated rupture of the recycle holdup tank (RHT).

The postulated fuel handling accident has been analyzed for three cases: 1. A fuel handling accident outside the containment in the fuel handling building; 2. A fuel handling accident inside the reactor containment building with the containment airlocks and equipment.


Question: The NRC mandates that licensees perform aging management for materials in systems that are subject to wear and environmental effects.

Loss of material due to general (copper alloy only), pitting, crevice corrosion, MIC; flow blockage due to fouling

Not Applicable. There are no fiberglass or HDPE piping or piping components exposed to raw water in Auxiliary Systems.


Question: The submitted data includes activity measurements for various radionuclides involved in licensing assessments.

The activity (pCi/grams) of the nuclide provides essential information in safety evaluations and regulatory compliance.

The NRC outlines procedures for managing environmental aspects related to nuclear activities, focusing on both terrestrial and aquatic ecosystems.


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

All licensees are required to submit a Safety Analysis Report that outlines the safety measures implemented at their facilities.

Participants in the virtual meetings will be expected to adhere to the scheduled agenda and contribute to discussions.


Question: What is the role of CFCUs in maintaining containment pressure during a large break LOCA?

The minimum equipment assumed to be in operation to maintain containment pressure below design pressure is one train of Containment Spray and two CFCUs.

Correct. The minimum is 2 CFCUs and a train of containment spray.


Question: What is the calculation for final charging/seal flow in a centrifugal charging pump?

Final Charging / Seal Flow = 11.55 x \(\sqrt{\Delta p @ Test Gauge 1 - FT - 62 - 93C}\)

CHECK the differential pressures recorded in Steps 6.2[10] and 6.2[11] are not within the lower 1/3 of the range of the associated test gauge.


Question: product numbers for nuclear materials

The vendor provided various product numbers such as NR-0460-S-922-S and NR-0462-D-801-B for identification of nuclear components.

The Minnesota Mining and Manufacturing Company has multiple product designs and configurations available for purchase.


Question: What are the different activities of nuclides in reactor coolant?

Normal Plant Operation Source Terms (based on ANSI N237) provide data on Reactor Coolant Activity for various nuclides under specific conditions.

Group I - Noble Gases includes nuclides such as Kr-83, Kr-85, Kr-87, and continues to list Kr-88, Kr-89, Xe-131m, Xe-133m, with distinct parameters and activities.


Question: nuclide activity measurement

The activity of a nuclide is expressed in picocuries (pCi) per gram, and it represents the rate of decay or disintegration of the radioactive atoms in a given sample.

The yield percentage indicates the proportion of the substance that is produced during a specific process, often used in chemical reactions.


Question: What is the role of the FLEX system in nuclear safety?

The post-initator operator actions identified for the FLEX system are listed in Table E9-5. These HFEs were identified through procedure and FLEX strategy reviews and refined during the system modeling process.

The NRC conducts regular inspections to ensure compliance with safety regulations and operational guidelines.


Question: What are the probabilities associated with tephra-fall eruptions from UWRB?

The probability of a tephra-fall eruption from a future eruption from the UWRB volcanic field is:

5.7x10^{-5}/yr × 1 × 0.028 = 1.6x10^{-6}/yr

Eruptions in the UWRB will have the potential to develop buoyant tephra plumes from VEI2–3 events.


Question: POV test acceptance criteria

For example, POV test acceptance criteria must be properly translated from design calculations to test procedures.

The results of the performance evaluation must be documented and reviewed to ensure compliance with established standards and regulations.


Question: What are invasive species and their regulations in Illinois?

Although the IDNR does not maintain a list of general invasive plant and animal species, the agency, along with the Illinois Department of Agriculture, maintains invasive species lists that are regulated through three regulations: the Illinois Exotic Weed Act, Illinois Injurious Species Rule, and the Illinois Noxious Weed Law.

The Midwin National Tallgrass Prairie is immediately east of the Des Plaines Conservation Area and is crossed by a short segment of the Goodings Grove transmission corridor.


Question: fuel delivery and storage plans

The purpose of the meeting is for USO to discuss with the NRC staff how they will meet the requirements for plans for the delivery and storage of fuel.

This covers closed meeting sessions requested pursuant to 10 CFR 2.390(d)(1), which relates to correspondence and reports to or from the NRC which contain information or records concerning a licensee's or applicant's physical protection, classified matter protection, or material control and accounting program for special nuclear material not otherwise designated as Safeguards Information or classified as National Security Information or Restricted Data.


Question: Primary Containment Air Lock in nuclear safety regulations

**Primary Containment Air Lock** is defined under Tech Spec 3.6.1.2, and changes related to TSTF-505 are incorporated; however, under certain circumstances, with more than one primary containment air lock inoperable, excessive leakage or a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or there is a loss of function.

The DNPS TS do not contain this TS Condition; therefore, TSTF-505 changes are not incorporated.


Question: What is nuclear licensing and its significance in reactor safety?

Nuclear licensing is a regulatory process that ensures nuclear installations comply with safety standards and meet the necessary requirements for operation and management.

The procurement process requires vendors to provide documentation in order to validate their supply chain and manufacturing capabilities.


Question: What is the definition of breached spent fuel rod?

BREACHED SPENT FUEL ROD: Spent fuel with cladding defects that permit the release of gas from the interior of the fuel rod. A fuel rod breach may be a minor defect (i.e., hairline crack or pinhole), allowing the rod to be classified as undamaged, or be a gross breach requiring a damaged fuel classification.

The NRC requires that all safety systems be tested and verified regularly to ensure compliance with operational standards.


Question: What is TS 3.5 in nuclear regulatory requirements?

TS 3.5 requires that the technical specifications undergo rigorous review to ensure they meet the standards for safety and operational requirements for the reactor.

The NRC maintains a comprehensive database of all records related to nuclear facility operations, which includes past inspections and future assessments.


Question: What are the MAPLHGR limits for Framatome Fuel?

Table 14 outlines the Average Planar Exposure and the corresponding ATRIUM 11 MAPLHGR limits, indicating values such as 0.0 GWd/MTU equating to 12.0 kW/ft, and progressing to 69.0 GWd/MTU with a limit of 7.2 kW/ft.

Nuclear reactors require stringent safety systems to ensure operational reliability; these systems include emergency core cooling, containment systems, and multiple redundancy in safety controls.


Question: What are the requirements of the general license 10 CFR 31.5?

C. I am aware of the requirements of the general license, provided in 10 CFR 31.5.

WARNING: FALSE STATEMENTS MAY BE SUBJECT TO CIVIL AND/OR CRIMINAL PENALTIES. NRC REGULATIONS REQUIRE THAT SUBMISSIONS TO THE NRC BE COMPLETE AND ACCURATE IN ALL MATERIAL ASPECTS.


Question: subsequent renewed operating licenses

Duke Energy. 2021. Letter from S.M. Snider, Vice President, Oconee Nuclear Station, to NRC Document Control Desk, dated June 7, 2021, regarding 'Duke Energy Carolinas, LLC (Duke Energy) Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, 50-287 Renewed License Numbers DPR-38, DPR-47, DPR-55 Application for Subsequent Renewed Operating Licenses.'

The Federal Insecticide, Fungicide, and Rodenticide Act, as amended, provides a framework for regulating pesticide use and promoting environmental safety.


Question: What are the characteristic time ratios important for quasi-steady natural circulation observation?

Based on the bottom-up scaling analysis presented here, \( \Pi_{Pr} \) and \( \Pi_{Gr} \) are the two relevant characteristic time ratios for the heat exchanger.

As mentioned in Section 3.1, the experimental objectives for a scaled test resulting from this scaling effort would be to cover normal operation as well as both phases of the natural circulation scenario in a single facility.


Question: passive containment cooling system

Ability to predict and/or monitor changes in parameters associated with operation of the passive containment cooling system, including the following: (CFR: 41.7 / 45.5)


Question: What are the nearby transportation routes to the Grand Gulf Nuclear Station?

The GGNS site area is accessible by both river and road. U.S. Highway 61 and State Highway 18 connect Port Gibson (15 miles southeast of the site) with Natchez, Jackson, and Vicksburg.

The nearest military facility was England Air Force Base in Alexandria, Louisiana, approximately 100 miles to the southwest; however, it was officially closed in 1993.


Question: What measures were taken to enhance the access and maintenance of the reactor head?

So dropping the pool level allowed us to add those access ports that really improved the maintainability of the design and some of the required sequences.

The containment effectively is a big condenser, and so it can really draw the pressure.


Question: Updates to the documents include corrections for the bolting integrity program and modifications to the cooler tube configuration.

The LRA markup for Table 3.3.1, Item 3.3.1-64 in Supplement 2, Change #8 is being replaced to include the changes made via Supplement 2, Change #3 which deleted reference to the Bolting Integrity program.

Accordingly, LRA Table 3.3.1 and Table 3.3.2-15 are revised.


Question: The NRC specifies design flow rates for safety systems in critical applications.

| Type                     | Vertical Centrifugal |


Question: What is meant by the threshold head loss acceptance criteria for debris in nuclear licensing?

When determining the threshold head loss acceptance criteria, a break is considered a failure if the quality of any debris type for this break exceeds the tested quantity.

The quantity of fines generated and subsequently transported to the containment floor would result in agglomeration.


Question: Following the upgrade of the probabilistic risk assessment models, a report must be drafted detailing the newly developed methods and their assessment.

A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The annual safety review consists of comprehensive inspections, evaluations, and assessments to ensure the safe operation of the nuclear facility.


Question: automatic depressurization system

E.1 Declare Automatic Depressurization System (ADS) valves inoperable.

Not applicable when trip capability is not maintained.


Question: What are the measured lead concentrations in air quality monitoring?

Measured Lead Concentrations and Comparison with NAAQS

Air Quality Monitor Location includes various sites across the urban area.


Question: The regulatory process requires thorough inspections of bolting mechanisms such as those in the core barrel assembly.

The visual VT-3 inspections for the specified locking device types look for evidence of broken, separated, or missing bolt locking devices or welds.

No need to audit the 40 – 80 criteria in Item B11 of Table 5-1, MRP-227, Rev. 2.


Question: The analysis of Americium isotopes is critical for understanding the contamination levels in the environment.

Americium-241 was analyzed and found to have a result of -0.00746 pCi/g with an uncertainty of +/-0.02534.

Iron-55 showed a result of -0.689 pCi/g with an uncertainty of +/-6.54, indicating its presence in the waste stream.


Question: What are the plans for the engagement with PNNL regarding the safety evaluation and ACRS interactions?

The Division of Operating Reactor Licensing oversees the licensing process for nuclear reactors, ensuring that safety and regulatory standards are met.

The review process for contracts can be complex, often requiring coordination between multiple departments.


Question: analog input module failure modes

Analog input failures are complicated by the overlaying of failure modes attributable to analog input module error condition and the individual analog input channel errors.

The FMEA credits the AC160 window watchdog timer WWDT that is described in the Common Q Topical Report.


Question: The evaluations conducted show that the flaw depth will not exceed allowable limits for the plant's operational life.

Since the analyses indicate that the allowable flaw depth will not be reached for the remaining plant life, the flaw evaluations of the pressurizer safety and spray nozzle structural weld overlays will remain valid beyond the 2045 end date of the PEO, in accordance with 10 CFR 54.21(c)(1)(i).

During Unit 1 Refueling Outage 19 (1R19) in 2015, while performing an ISI, DCPP identified a weld flaw indication located in ASME Code Class 1 pressurizer spray line pipe weld WIB-378.


Question: critical heat flux correlation

COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation provides a framework for analyzing the critical heat flux conditions in nuclear reactors, crucial for maintaining safe operation under various thermal hydraulic scenarios.

Figure A-21 Axial geometry of CHF test CE580 illustrates the distance measurements from the EOH, which are essential for precise engineering assessments, but do not relate specifically to critical heat flux correlation.


Question: What are the certification requirements for drinking water in Louisiana?


Question: The overhead cranes will be designated to ensure they are not left in the proximity of sensitive structures during seismic activities.

A seismic PRA was performed to respond to the IPEEE – Seismic program at SPS. The quantification concluded that no severe accident vulnerabilities exist at SPS from a potential seismic event.

In addition to the contingency action plan discussed in Section 3.3, the following project constraints and compensatory measures will be implemented:


Question: What is the purpose of additional cool time required for nuclear fuel?

Table 5.8-49 outlines the additional cool time required for 23 kW damaged PWR fuel to ensure safety and regulatory compliance.

Table 5.8-38 compares gamma sources for CEA primary absorber materials, focusing on radiation shielding properties.


Question: The analysis of storm surge at the proposed site combines predicted values with concurrent wind conditions and high tide estimates.

In order to satisfy the combined events criteria specified in Section 9.2.2 of ANSI/ANS-2.8-1992, (Reference 2.4.5-1) storm surge at the new plant is evaluated combining probable maximum surge and seiche with wind wave activity concurrent with the 10 percent exceedance high tide, and effects of hurricane-associated precipitation.

Water levels further increase due to wind blowing directly over the bay.


Question: What are the steps to take during a Power Range Monitor failure?

1. **PLACE** control rods in MANUAL. 2. **CHECK** rod motion STOPPED.  (p) **TRIP** Rx.  **GO TO** 1-E-0, Reactor Trip or Safety Injection. 3. **CHECK** S/G levels NORMAL.  **CHECK** S/G Level controllers in AUTO AND controlling S/G levels.

The NRC requires licensees to perform routine maintenance to ensure all systems are operational and compliant with safety standards.


Question: horizontal SSE spectrum

Figure 2.5-54A shows four horizontal ground spectra—the mean 5 x 10⁻⁵ return period RG 1.165 high- and low-frequency scaled spectra (from Figure 2.5-51), the performance-based spectrum (from Figure 2.5-53) and the selected hard rock SSE spectrum (previously shown in Figure 2.5-48), which is the envelope of the other three spectra.

Table 2.5-26 shows the mean 10⁻⁴ and mean 10⁻⁵ ground motion amplitudes, the calculated values of A_R, the calculated values of the SF, and the resulting performance-based spectral amplitudes A(f). These results were calculated using the same source and ground motion assumptions as for the reference probability approach.


Question: The minimum detectable activity for a specific nuclide is crucial for determining regulatory compliance in nuclear safety assessments.

Nuclide MDA (pCi/grams) represents the minimum detectable activity of a specific nuclide in the sample, indicating the lowest level of activity that can be reliably detected.

Errors quoted at 2.000sigma reflect the uncertainty surrounding the measurements obtained during the study.


Question: The measurement of gamma radiation in various depths during the sampling process is crucial for assessing background radiation levels.

The gamma measurement recorded at a depth of 10-11 feet was 7.9K cpm, indicating significant radiation activity in that interval.

During the sampling, photographs were collected to visually document the sampling location and surroundings.


Question: nuclear waste regulatory control

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: What is the role of Chain of Custody (COC) in nuclear sample receipt?

Chain of custody documents included with shipment are required to ensure proper handling and tracking of samples in nuclear regulatory processes.

The sample containers must be identifiable, and the secondary thermometer device serial number should be recorded if applicable.


Question: What is the flow rate for circulating water at Monticello Nuclear Plant?

The facility includes various components such as the discharge canal and retention pond, which play critical roles in the overall environmental management of the site.


Question: NRC enforcement actions

Failure to conduct your program in accordance with NRC regulations, license conditions, and representations made in your license application and supplemental correspondence with NRC will result in enforcement action against you.

The NRC’s Safety Culture Policy Statement became effective in June 2011.


Question: maximum surge for PMH scenarios

Each PMH evaluated in the above table exhibited a central pressure, P₀ = 26.65 inches of mercury; pressure drop, Δp = 3.5 inches of mercury; and track direction from 138 degrees (moving northwest). The tide is specified as static at the 10% exceedance high tide at the mouth of the Delaware Bay.

The radius of maximum winds, R, is critical for estimating potential impacts from severe weather events, particularly in coastal regions where storm surges can be exacerbated by local geography.


Question: What are the effects of dredging on aquatic organisms near nuclear power plants?

In the 2013 LR GEIS (NRC 2013-TN2654) analyzed the effects of dredging on aquatic organisms as a new issue and concluded that the effects of this issue would be SMALL during the initial license renewal term for all nuclear power plants.

Dredging entails excavating a layer of sediment from the affected areas and transporting that sediment to onshore or offshore areas for disposal.


Question: What is a categorical exclusion in nuclear licensing?

The submittals are either not covered by the criteria for CATEX or, in the case of the license transfer request, do not fall within the factual basis underlining the corresponding CATEX.

It's the environmental report that they submitted that was of continued relevance.


Question: thermal aging embrittlement of CASS

The DCPP Thermal Aging Embrittlement of Cast Austenitic Stainless Steel AMP is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE, including research and development, such that the effectiveness of the program is periodically evaluated.

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


Question: Halon fire suppression system

Halon/carbon dioxide fire suppression system piping, piping components | Pressure Boundary | Steel | Air - indoor uncontrolled | Loss of material due to general, pitting, crevice corrosion | Fire Protection (B.2.1.15) | VII.G.AP-150 | 3.3-1, 058 | A

Closure bolting | Mechanical Closure | High-strength steel | Air | Cracking due to SCC; cyclic loading | Bolting Integrity (B.2.1.10) | VII.A-04 | 3.3-1, 010 | A


Question: NRC regulations and information required for an Early Site Permit (ESP) application

NRC regulations (Reference 1) require that an ESP application contain certain corporate information about the applicant.

Dominion Nuclear Connecticut, Inc., also an indirect, wholly-owned subsidiary of DRI, is the licensed operator of the Millstone Power Station.


Question: The NRC mandates that the letdown flow path must be carefully adjusted to ensure operational safety during the licensing process.

Step is critical to align letdown flow path.


Question: What does the backfit assessment process involve?

The backfit assessment process is designed to evaluate whether a proposed modification to a regulatory requirement in light of new information is necessary to maintain safety and regulatory effectiveness.

The current scope of rule language is a clarification only.


Question: The program is designed to ensure that instrumentation circuit cables and connections remain functional despite potential aging effects from environmental conditions.

The DCPP Insulation Material for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements AMP is a new AMP. This AMP will provide reasonable assurance that the intended functions of insulation material for electrical cables and connections that are not subject to the environmental qualification requirements of 10 CFR 50.49 and are exposed to ALEs caused by temperature, radiation, or moisture are maintained consistent with the CLB through the PEO.

The safety protocols implemented during emergency situations are evaluated through regular drills and training exercises for all staff.


Question: What are the requirements of the Protected Equipment Program in nuclear facilities?

During their review, the inspectors noted a trend involving the implementation of the Protected Equipment Program. Specific CAP entries reviewed by the inspectors included:

The inspectors performed a detailed review of several issue reports (ARs) related to non-cited violations (NCVs) and observations made during the 2024 Comprehensive Engineering Team Inspection (CETI). These included, but were not limited to:


Question: The NRC mandates extensive analysis of geological formations surrounding nuclear facilities to ensure they are safe and sustainable.

Clay, medium gray (N5), very plastic, no bedding or laminae (massive), very low permeability.

Shale, medium dark gray (N4), hard, blocky with conchoidal fracture, dry, lustrous sheen. Very low permeability.


Question: closed treated water systems

The DCPP Closed Treated Water Systems AMP includes condition and corrosion monitoring activities, which provide for opportunistic and periodic inspections of system components and corrosion coupons.

The jacket cooling water chemistry has remained stable at DCPP for 39 years.


Question: peak fuel-cladding temperature limit

NRC has Overlooked Specific Data Cited by Petitioner from Experiments in which Runaway Oxidation Commenced at Temperatures Lower than the 10 C.F.R. § 50.46(b)(1) 2200°F Peak Fuel-Cladding Temperature Limit

NRC’s TRACE Simulations of FLECHT Run 9573 Are Invalid because They Did Not Simulate the Section of the Test Bundle that Incurred Runaway Oxidation


Question: What are external hazard PRA plant models?

The external hazard PRA plant model shall include wind-caused significant initiating events and other failures that are significant contributors that can … shall include wind-caused significant initiating events and other failures that are significant contributors that can …

Table 9-2.3-1 describes the necessary modifications to the HLR statements and their implications for regulatory compliance.


Question: drilling through reinforced concrete

Three out of the six drillers surveyed had experience drilling through reinforced concrete and steel: a. One driller had experience unexpectedly drilling through an old foundation and hitting one small piece of rebar. b. Another driller had experience drilling through steel well casing. When the casing shifted or a weld broke, he ended up drilling through 0.64 cm [0.25 in] thick casing (a total of 1.27 cm [0.5 in] of steel). c. Another driller hit reinforced concrete when drilling near an old bridge. He could not move the rig and used a hammer to drill through two, 20-cm [8-in] layers.

Cuttings are left in the vicinity of the well near the drill rig. The landscaper will level out the property, including the cuttings, afterwards. Sometimes the cuttings are used on roadways. One driller who uses air rotary said that he will create a mound about 3-m [10-ft] square and 0.3 m [1 ft] high around the drill head. He typically drills residential wells at 1- to 5-acre sites, and the owner will move the cuttings afterwards.


Question: nuclear safety briefing

10:15am – 10:30am: Safety Briefing & Security Badging


Question: What are Stress Intensity Factors (SIFs) used for in reactor safety assessments?

Stress Intensity Factors (SIFs) for Nozzle Corner Cracks are calculated to determine the structural integrity of components in nuclear vessels under stress conditions.

The Journal of Engineering Material and Technology published research on corruption in engineering practices that failed to meet industry standards.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves several key steps, including the submission of a Construction Permit Application (CPA) and an Operating License Application (OLA), both of which must be reviewed and approved by the NRC to ensure compliance with safety regulations.

The maintenance of security systems is crucial for safeguarding sensitive information and ensuring that all internal and external threats are appropriately managed.


Question: What is the importance of hydrostratigraphy in safety analysis for nuclear sites?

Hydrostratigraphy of the New Jersey Coastal Plain: Sequences and Facies Predict Continuity of Aquifers and Confining Units.

Gravity Anomaly Map of North America, Decade of North American Geology (DNAG), Geological Society of America, scale 1:500,000,000, 5 plates, 1987.


Question: The NRC's analysis of risk indicates a low level of population dose risk due to severe accidents at Oconee Station, particularly in comparison to the 1996 LR GEIS.

The Oconee Station SAMA analysis performed during initial license renewal used a full-scope, Level 3 PRA with analysis of both the internal and external events.

In general, the population dose risk measures calculated for Oconee Station show relatively low risk of environmental impacts compared to other nuclear power plants (NRC 2023-TN7802).


Question: The model includes factors that affect uncertainty assessment.

The model uncertainty determination also considers the effects of various components and parameters.

Developed models in different fields often rely on empirical data to validate predictions.


Question: The neutron fluence calculations for the reactor vessel components are essential for evaluating material degradation during extended operation.

The Unit 1, Unit 2, and Unit 3 reactor vessel beltline component fluence analyses have been satisfactorily projected through the subsequent period of extended operation.

The tables include data for various components and their responses under different operational stresses.


Question: setpoint methodology

1.9 Setpoint Methodology and Calculations: N/A – setpoint methodology and calculations are not changing.

1.8 (Changes to) Technical Specifications: Not included. Technical Specification Changes is part of the LAR. LTR Appendix A provides the safety case for eliminating certain PPS surveillance requirements in accordance with WCAP-18461-P-A (Reference 54).


Question: What are the requirements for filing a petition to intervene in a nuclear regulatory action?

As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) the name, address, and telephone number of the petitioner; (2) the nature of the petitioner’s right to be made a party to the proceeding; (3) the nature and extent of the petitioner’s property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner’s interest.

The NRC’s regulations are accessible electronically from the NRC Library on the NRC’s Web site at https://www.nrc.gov/reading-rm/doc-collections/cr/.


Question: What does 10 CFR 51.45(b)(3) require for environmental reports?

Per 10 CFR 51.45(b)(3): The environmental report must contain a discussion of alternatives that is sufficiently complete to aid the NRC in developing appropriate alternatives.

LRA Table 8.6-1 is titled "Comparison of Predicted Environmental Impacts" but does not provide the potential impacts for each phase in each project area (CBP and MEA) and therefore does not provide a complete comparison of predicted impacts.


Question: What is the process for submitting a construction permit application to the NRC?

The NRC staff is reviewing a construction permit (CP) application submitted by TerraPower, LLC (TerraPower), on behalf of US SFR Owner, LLC, a wholly owned subsidiary of TerraPower, for the proposed Kemmerer Power Station Unit 1.

An environmental report was included as part of the application (ML240884027). These documents may be obtained online in the NRC’s ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html.


Question: License holders must submit documentation to demonstrate compliance with regulatory standards for reactor safety and operational integrity.

The licensing process involves a comprehensive review where the Nuclear Regulatory Commission evaluates the applicant's safety measures, operational protocols, and emergency preparedness plans.

Thermal performance testing is typically conducted to assess the efficiency of the heat exchangers in the system.


Question: What are the anticipated environmental impacts of the uranium ISR methods?

Due to the benign nature of uranium ISR methods and the lack of unique environmental resources in the area, the potential environmental impacts of the CUP will be minor.

The affected (baseline) environment for the Proposed Action is described in Chapter 3, potential environmental impacts are described in Chapter 4, mitigation measures are described in Chapter 5, the monitoring program is described in Chapter 6, and Chapter 7 presents a discussion of the benefits and costs of the project.


Question: What is the purpose of the NRC's Agencywide Documents Access and Management System (ADAMS)?

You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html.

For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.


Question: The NRC mandates that components like heat exchangers must maintain their structural integrity over time.

Heat Exchanger (Boric Acid Distillate Cooler) Shell Side Components | Structural Integrity (attached) | Stainless Steel | Closed-Cycle Cooling Water (Int) | Loss of Material | Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) | VII.C2.A-52    | 3.3-1, 049 | E, 1  |

Heat Exchanger (Boric Acid Distillate Cooler) Shell Side Components | Leakage Boundary (spatial)   | Stainless Steel | Plant Indoor Air (Ext) | None                              | VII.J.AP-17                                          | 3.3-1, 120     | C         |


Question: Wenonah Formation geology

The Wenonah Formation typically consists of very dark gray to greenish-black, sandy clay with trace shell fragments and trace to little glauconite, and locally consists of clayey and silty, fine-grained to medium-grained sand with trace to little glauconite.

The Mount Laurel Formation consists of Upper Cretaceous age marine sediments deposited in a near-shore environment during a marine regression.