Question: What are the requirements for isolating flow paths in nuclear systems?

With both the isolation and control valves inoperable in one flow path, the affected flow path must be restored to OPERABLE status or isolated within a Completion Time of 8 hours. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

The Completion Time of "once per 7 days" is appropriate considering the fact that the devices are operated under administrative controls, valve status indications in the main control room and the probability of their misalignment is low.


Question: The U.S. Nuclear Regulatory Commission evaluates submissions for digital control systems to ensure they maintain safety standards.

The proposed amendment requests would change both the design and technical specifications to permit the use of a new single digital instrumentation and controls (I&C) PPS to replace the previously independent analog instrumentation systems of the reactor protection system, the nuclear steam supply shutdown system (NSSS), emergency core cooling system (ECCS), reactor core isolation cooling system, and end-of-cycle reactivity control pump trip at Limerick.

The NRC relies on comprehensive data from licensees to assess and support their safety-related operational changes.


Question: The NRC evaluates the effectiveness of decay heat removal systems during various operational states.

The LGS Decay Heat Control procedure outlines the strategies to manage decay heat in emergency operations, crucial for maintaining core cooling and safety.

Reference 42 discusses the introduction and references for BWROG Emergency Procedure and Severe Accident Guidelines, providing a foundation for emergency management protocols.


Question: The potential risk posed by waves as a natural hazard is evaluated in the context of external safety assessments.

Waves  | above the main surface of a body of water as a result of the wind blowing over an area of fluid surface. | against waves, including those associated with the combined storm surge hazard evaluated in the FHRR. Based on this review, the Waves hazard is considered to be negligible. There are no configuration-specific considerations for this hazard. This hazard can be excluded from the TSTF-505 program evaluation.

The screening criteria for terrestrial hazards include various factors that impact the structural integrity of the facility, ensuring overall safety during operations.


Question: How do computer safety models predict fuel-cladding temperature escalations?

Computer safety models have failed to accurately predict the onset of the fuel-cladding temperature escalation that occurred in the LOFT LP-FP-2 experiment.

In addition to phenomena which would occur in the design basis accident temperature range: fuel cladding temperatures lower than the 2200°F PCT limit.


Question: radioactive waste handling procedures

As discussed in Section 2.2.6, DNPS has radioactive waste stream handling and shipping procedures.

DNPS operates its air emission sources in compliance with IEPA Title 35 IAC, Subtitle B, Chapter 1, and the CAA.


Question: What is sump energy in nuclear reactor safety analysis?

Sump Energy (Btu) is the internal energy of the water in the sump relative to 32°F.


Question: What are the aging effects of concrete in nuclear facilities?

Increase in porosity and permeability; cracking; loss of material (spalling, scaling) due to aggressive chemical attack.

Masonry walls are critical for structural support and must be carefully monitored for any signs of deterioration.


Question: population growth in the Region of Influence

The Horizon Center site (HCS) is located in Roane County, Tennessee, within the limits of the City of Oak Ridge, which spans portions of both Roane and Anderson counties. Oak Ridge had a population of 31,402 in 2020, while Roane County had a population of 53,404 (USCB, 2020).

The data used in this section are the most current and complete data identified to address each topic. In most cases, those data represent relatively current conditions in the study area (for example, 2015-2019 American Community Survey [ACS] 5-year estimates provided by the U.S. Census Bureau [USCB] and USCB 2020 Decennial Census data), or projected conditions during the expected life of the proposed project.


Question: A licensee is required to monitor the levels of strontium-82 and strontium-85 to ensure they do not exceed regulated limits before patient use.

A licensee that uses a strontium-82/rubidium-82 generator for preparing a rubidium-82 radiopharmaceutical shall, before the first patient use of the day, measure the concentration of radionuclides strontium-82 and strontium-85 to demonstrate compliance with subsection (A).

Except as provided in R12-1-710, a licensee shall require an authorized user of unsealed radioactive material for the uses authorized under Group 200 to be a physician who has: Certified by a medical specialty board whose certification process has been recognized by the NRC or an Agreement State and who meets the requirements in subsection (A)(3).


Question: The NRC emphasizes the importance of evaluating accessible areas when assessing the condition of containment structures for aging effects.

Increase in porosity and permeability due to leaching of calcium hydroxide and carbonation could occur in inaccessible areas of concrete elements of PWR and BWR concrete and steel containments. The GALL Report recommends further evaluation if leaching is observed in accessible areas that impact intended functions.

The DCPP containment is located in a region where weathering conditions are considered “negligible” as shown in ASTM C33, Figure 1. The containment is designed to be exposed to precipitation.


Question: What is the significance of containment accident pressure in the flashing analysis for nuclear reactors?

Flashing would occur if the pressure immediately downstream of the strainer is lower than the water vapor pressure at the sump temperature. The pressure downstream of the strainer can be calculated by adding the strainer submergence (from the top of the strainer) to the containment pressure, before subtracting the strainer head loss. The Wolf Creek analysis showed that, if the containment pressure is assumed to be equal to water vapor pressure at the corresponding sump temperature, a minimum containment accident pressure of 0.97 psi (rounded to 1 psi) needs to be credited to preclude flashing.

No near-field settling was credited in Wolf Creek head loss testing. As described in the Response to 3.f.4, the debris introduction region of the test tank was equipped with hydraulic mixing lines that provided a sufficient level of turbulence to keep debris suspended without disturbing the debris bed.


Question: reduction of loads on the battery

Gen KA # / Rating: 055.EA1.04 - Ability to operate and monitor the following as they apply to a Station Blackout: Reduction of loads on the battery.


Question: Understanding the operational impacts associated with the loss of service air in the compressed air system is crucial for safety regulations.

Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to the compressed air system: (CFR: 41.7 / 45.7)  - **K 5.02** Loss of service air 2.3

Knowledge of compressed air system design feature(s) and/or interlock(s) that provide for the following: (CFR: 41.7)  - **K 4.01** Containment isolation 3.7


Question: What are the responsibilities of a radiation protection technician at nuclear plants?

Specifically, it was alleged that a radiation protection technician (RPT) changed the characterization of dose rates to avoid posting of a locked high radiation area as required.

In July 2019, an OI investigation concerning APINDE was referred to the U.S. Department of Justice—Environmental Crimes Division (DOJ-ECD) and was accepted for prosecution.


Question: What is the function of containment sumps in nuclear reactors?

The containment sumps provide a borated water source to support recirculation of coolant from the containment sumps for residual heat removal, emergency core cooling, containment cooling, and containment atmosphere cleanup during accident conditions.

Safety analysis is crucial during any accident, but mechanisms like emergency core cooling systems must also be tested regularly to ensure functionality.


Question: The NRC assessed the effects of discharging cooling water and determined the impact from scouring would be minimal.

In the 2013 license renewal GEIS, the NRC reviewed the impacts of scouring caused by discharged cooling water and found the impacts would be SMALL.

Natural sediment transport processes could bring fresh sediment into the discharge flow area.


Question: What is the wind speed measurement relevant in nuclear safety assessments?

The NRC requires licensees to maintain clear communication with local authorities.


Question: What is a rod ejection accident in nuclear reactors?

A rod ejection accident occurs when a control rod is inadvertently removed from a reactor while it is operating, potentially leading to an uncontrollable increase in neutron power levels if not managed safely.

The graphs located in Figure 15-185 present data on neutron power levels versus time in seconds.


Question: The analysis of the zirconium-steam reaction is crucial for understanding the safety implications of nuclear reactor design.

Consideration of the zirconium-steam reaction and the ignition and intense firestorm in nuclear reactor fuel rods is well overdue.

broken headers
damaged fuel assemblies
accumulated debris (fuel rod segments, shroud pieces)


Question: The NRC stipulates specific criteria regarding the handling and operations of spent nuclear fuel assemblies.

F. SNF may be loaded uniformly at a maximum heat load of 959 watts/assembly (Figure B2-1). Alternatively, a preferential loading pattern may be applied as described in Table B2-8 and Figure B2-2.

E. Unenriched fuel assemblies and unirradiated (i.e., not inserted in-core) fuel assemblies are not authorized for loading. Low enriched, unenriched, and/or annular fuel pellet axial end blankets are permitted, provided that the nominal length of the blanket is not greater than six (6) inches.


Question: maximum planar average enrichment

Slot A: Undamaged Exxon fuel maximum planar average enrichment 3.71 wt % ^235U.


Question: What is the significance of nuclide MDA in radiation measurement?

The Nuclide MDA (Minimum Detectable Activity) is critical in assessing the sensitivity of radiation detection systems, determining the lowest activity level that can be reliably measured for each radionuclide.

The energy levels of various nuclides were analyzed for their yield percentages and activity, providing important data for understanding the radiation environment.


Question: nuclide minimum detectable activity


Question: What is the safety factor for reactor containment?

Reactor containment is typically assigned a safety factor of 1.5 in regulatory filings, ensuring that structural integrity is maintained under operational conditions.

Factors of Safety at Ground Water Elevations Shown in Reference to Overturning indicate various safety measures in place for below-grade structures.


Question: The NRC mandates that licensees must notify regulatory bodies when there are operational anomalies that affect safety systems.

Unit 2 is operating at 100% power when the following occurs: Annunciator 912-5-C-1, RX BLD 1 LOW DP, alarms. Reactor Building Atmospheric DP gauge is indicating +0.1 in. H₂O vac.

Unit 1 is currently in Mode 4 in preparation for a scheduled refuel outage.


Question: What is the normal source of makeup water to the CCW system?

Normal source is the Transfer Tank, alternate is PWST, CST or FWST.

This is a backup for the CST to supply AFW.


Question: What are the required actions for control rod operability in a nuclear reactor?

A. (continued) A.3 Perform SR 3.1.3.2 for each withdrawn OPERABLE control rod.


Question: The rod block condition is influenced by the status of the source range monitor (SRM) instruments, as they can prevent certain operational actions if they are bypassed.

Loss of high voltage power to SRM B causes an SRM INOP condition to be detected and provides input to the rod block circuitry.


Question: The NRC mandates that modeling of critical heat flux conditions must be conducted during the licensing process.

The COBRA-FLX code provides a unique modeling capability for the prediction of critical heat flux for various reactor types and conditions, which is essential for safety analysis.

Framatome's innovative designs are aimed at improving fuel efficiency and reducing waste generation in advanced reactors.


Question: What is the purpose of an audit in the NRC regulatory process?

The NRC staff acknowledges and will observe appropriate handling and protection of proprietary information made available for the audit.

The following table identifies the NRC audit team members and their respective organization.


Question: Lead Test Assemblies in nuclear licensing

This amendment request proposes to add a License Condition to Appendix D, "Additional Conditions," of the Vogtle Unit 1 and Unit 2 Operating License that authorizes use of four Lead Test Assemblies (LTAs) to be placed in limiting core locations.

In accordance with 10 CFR 50.90, "Application for Amendment of License, Construction Permit or Early Site Permit," Southern Nuclear Operating Company (SNC) requests an amendment to Renewed Facility Operating License (FOL) Nos. NPF-68 and NPF-81 for Vogtle Electric Generating Plant (Vogtle) Units 1 and 2.


Question: What is the process for disposing of tritium waste?

VSS Electric, Inc. has returned 26 Tritium Exit signs to SRB Technologies, Inc. (NC Radioactive Materials License # 034-0534-2) for proper disposal.


Question: fuel rod thermal-mechanical methodology

ANP-10323(P), Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors.

ANP-10339(P), Revision 0, ARITA – ARTEMIS/RELAP Integrated Transient Analysis Methodology.


Question: What is the purpose of baseline water quality assessment in nuclear licensing?

Baseline water quality will be determined by averaging the data collected for each parameter, from each well, for each zone that is to be monitored.

Water quality samples will be obtained and analyzed from the monitor wells described in a above.


Question: What are the requirements for NRC filings regarding environmental evaluations?

In their filings, Organizational Petitioners appear to assert two disparate but related claims in Contention 4. First, Organizational Petitioners assert that certain information in the Combined LR/ER is outdated or stale and that the NRC employs a practice that improperly shifts the burden onto petitioners to fill in that outdated or stale data with more current information.

Thus, like a licensing board in 2004, we are not convinced that this issue is material to the matter pending before the NRC.


Question: What is the unit of measurement for isotope activity in nuclear devices?

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

The Manufacturer name is KAY-RAY/SENSALL, INC.


Question: How does the Chemical and Volume Control System regulate reactivity in the reactor coolant?

The Chemical and Volume Control System regulates the concentration of chemical neutron absorber in the reactor coolant to control reactivity changes resulting from the change in reactor coolant temperature between cold shutdown and hot full-power operation, burnup of fuel and burnable poisons, buildup of fission products in the fuel, and xenon transients.

Drains are sized large enough to facilitate draining of equipment in a reasonable time.


Question: nuclear safety regulations

The purpose of the NRC's safety regulations is to safeguard public health and ensure the safety of nuclear facilities.

Updated guidelines for the construction of residential buildings have been released by the city council.


Question: What are the exemptions under 10 CFR 2.302?

Participants granted an exemption under 10 CFR 2.302(g)(2) must still meet the electronic formatting requirement in 10 CFR 2.302(g)(1), unless the participant also seeks and is granted an exemption from 10 CFR 2.302(g)(1).

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket, which is publicly available at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the presiding officer.


Question: critical heat flux correlation

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation is a pivotal component in assessing heat transfer mechanisms under various operating conditions in nuclear reactors.

The test assembly conditions for non-mixing grid tests are essential for validating the performance and efficiency of advanced reactor designs.


Question: analytical methods for radioactive materials

The following Analytical Methods were performed: DOE EML HASL-300, Am-05-RC Modified, ASTM C 1475-00 Modified, DOE EML HASL-300, Pu-11-RC Modified, DOE RESL-NI, DOE HASL 300, 1.52.3/GA-01-R, EPA 905.0 Modified/DOE RP501 Rev. 1 Modified, EPA 906.0 Modified, EPA EERF C-1 Modified, DOE EML HASL-300, Tc-02-RC Modified, DOE RESL Fe-1, Modified, DOE RESL Ni-1, Modified.

The Certificate of Analysis includes the Company's contact information, address, and project details, which are necessary for identifying the context of the sample analysis and ensuring accurate communication.


Question: The design of the concrete cask is critical for ensuring maximum safety and efficiency during the long-term storage of the TSC.

The concrete cask is the storage overpack for the TSC and provides structural support, shielding, protection from environmental conditions, and natural convection cooling of the TSC during long-term storage.

The American National Standards Institute ANSI N14.6 (1993) and NUREG-0612 govern the TRANSFER CASK design, operation, fabrication, testing, inspection, and maintenance.


Question: What is the activity of isotopes used in nuclear devices?

The activity of the isotopes is measured in millicuries (mCi), which is a unit of radioactivity defined as one thousandth of a curie.

The device model SHLD-1 produced by OHMART/ VEGA CORPORATION is distributed by VEGA Americas, Inc.


Question: What is the procedure for addressing a condenser tube leak in a nuclear reactor?

MALF: CN0094A #21A Condenser Tube Leak Severity: 50

Crew refers to ARP for 2CC2 and OHA G15.


Question: The NRC emphasizes the importance of monitoring the condition of pressure boundary materials to prevent potential failures.

Loss of material due to general (steel only), pitting, crevice corrosion.

The reactor's core design incorporates advanced fuel types that enhance efficiency and safety measures.


Question: The NRC regulates compliance with safety standards, ensuring that all technical specifications are met for the safe operation of nuclear facilities.

The exemption request from 10 CFR 50.55a(h)(2) pertains to the regulatory requirements for the design and installation of certain components important for ensuring safety in nuclear power plants.

The regulatory audit summary includes a detailed overview of distribution to several NRC resource offices, ensuring that all relevant stakeholders are informed.


Question: automatic depressurization safety analysis

D.5.3 Qualification for ATWS analysis involving automatic depressurization.

D.6.1 Initial conditions are essential to establish a baseline for simulation and analysis.


Question: The operational requirements must be adhered to at all times to ensure compliance and safety.

Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval.


Question: What is the purpose of a cumulative effects analysis in nuclear licensing?

The goal of the cumulative effects analysis is to identify potentially significant impacts.

The timeline for the consideration of reasonably foreseeable future actions is the 20-year SLR term.


Question: The NRC outlines the requirements for minimum cool time of spent nuclear fuel assemblies for both PWR and BWR systems.

The cooling time for nuclear reactors is critical for ensuring that operations remain within safe limits.


Question: What are the requirements for pressurizer heater circuit breakers in nuclear reactors?

Each pressurizer heater circuit breaker shall be OPERABLE.

Separate Condition entry is allowed for each pressurizer heater.


Question: activity concentration measurement

Activity concentration exceeds MDC and is gamma identified.

Unit of measurement typically relates to the physical quantity being measured, such as length, mass, or time.


Question: What is the role of the Engineered Safety Feature Actuation System in nuclear reactor safety?

The Engineered Safety Feature Actuation System (ESFAS) portion of the PMS initiates necessary safety systems, based upon the values of selected unit parameters, to protect against violating core design limits and the Reactor Coolant System (RCS) pressure boundary, and to mitigate accidents.

Each division consists of: Field Transmitters and Sensors: operability is addressed by TS 3.3.8, 'ESFAS Instrumentation'.


Question: What is a large break loss of coolant accident?

A large break LOCA has occurred. 1-E-1, Loss of Reactor or Secondary Coolant, has just been resumed after performance of 1-ES-1.3, Transfer to Containment Sump.

Unit 1 is in Mode 5. The examinees are directed to swap RHR from A to B train. A train RHR pump will trip requiring response per 1-AOI-14, Loss of RHR Shutdown Cooling. This is a NEW JPM.


Question: nuclear regulatory process

The nuclear regulatory process involves thorough evaluations by the NRC to ensure that safety, security, and environmental protection standards are met before approval of licenses.

The proposed maintenance schedule outlines the intervals at which equipment should be inspected, ensuring operational efficiency and compliance with standard protocols.


Question: What are the factors influencing the measured breathing rate at various locations in the nuclear facility?

Breathing rate data are collected over time to establish the airflow dynamics in relation to environmental safety protocols.

The convection data shows the flow rate of substances over specified time intervals, illustrating the dispersion patterns in environmental assessments.


Question: release of radioactive liquid waste

21 CVCS MT is being released in accordance with S2.OP-SO.WL-0001 (RELEASE OF RADIOACTIVE LIQUID WASTE FROM 21 CVCS MONITOR TANK)

At 12:00, the crew REQUIRED to stop the release of 21 CVCS MT.


Question: What is the minimum initial assembly average enrichment required for PWR fuel?

Nuclear reactors are complex systems designed to generate electricity using controlled nuclear reactions.


Question: What is the role of limitation and condition (L&C) in nuclear licensing?

The NRC staff reviewed and performed independent calculations to assess Framatome's additional information (ML23137A250) presented at the May 4, 2023, meeting regarding limitation and condition (L&C) 18.

NRC staff will also modify the supporting language in the SE to be consistent with the L&C changes.


Question: minimum initial assembly average enrichment

The minimum initial assembly average enrichment values for the various fuel assembly geometries are provided in the loading tables.

The values presented specify the performance metrics for several types of nuclear reactors, illustrating their efficiency in energy generation.


Question: What is an in-core instrumentation area in nuclear safety?

The in-core instrumentation area is a dead-ended part of the lower Containment compartment.

All ventilation systems are of the recirculation type with the exception of the purge systems and are completely contained within the Containment structure.


Question: lid gasket seal in nuclear canisters

The metal gasket seal between the removable upper head (lid) and the bulkhead on the canister body was found to leak too easily during remote installation of the heads to the bodies using manual long-handled tools.

Various handheld tools were used to facilitate operation.


Question: reactor trip response procedures

Actions to be taken during a reactor trip response include verifying the SG blowdown isolation valves are closed and ensuring the total feed flow is maintained at specific levels until recovery conditions are met.

Verifying the integrity of the containment structure is essential for ensuring the safety of the reactor during normal and emergency operations.


Question: What is the activity measurement for nuclides in terms of pCi/grams?

Activity (pCi/grams) for the nuclides listed includes values such as -2.77E-02 for EU-152 and 1.54E-01 for BI-214.

The energy levels of the nuclides vary widely, with EU-152 having an energy of 121.78 keV and EU-154 at 344.28 keV.


Question: The NRC mandates that each sample must be analyzed for quality control purposes in accordance with established guidelines.

For sample shipping and delivery details, see Sample Receipt & Return form (SRR).


Question: What is a license amendment in nuclear regulatory terms?

A license amendment is a formal request submitted by a licensee to the Nuclear Regulatory Commission (NRC) to change, modify, or update an existing nuclear license under which operations are conducted.

The document outlines the procedures for submitting reports to the NRC, including deadlines and formatting requirements, ensuring that licensees maintain compliance with regulatory standards.


Question: M5 cladding hydrogen model uncertainty

Hydrogen predictions based on the best-estimate (BE) M5 oxidation model were previously compared with measured data in Figure 34-7 of Reference 6. These BE predictions are used to determine a hydrogen model uncertainty that may be used in applications outside of the GALILEO statistical methodology.

Figure 3-8 shows the distribution of M/P values for the M5 BE hydrogen predictions.


Question: nuclear measurements in regulatory filings

| Measurement         | Value           |
|---------------------|-----------------|
| Temperature         | 85.0 °F        |
| Dew Point           | 63.6 °F        |
| Wind Speed          | 10.2 MPH       |
| Wind Direct         | 198 DEG        |
| Diff Temp 250 - 30 FT | -2.0 °F       |
| Precipitation       | 0.1 INCH       |

Byron Station is a nuclear power plant located in Illinois.


Question: risk-informed completion times

By letter dated March 31, 2022, the Tennessee Valley Authority (TVA) requested to amend Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant, Units 1, 2, and 3, respectively, to adopt Technical Specifications Task Force (TSTF) Traveler 505 (TSTF-505), 'Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b,' to permit the use of risk-informed technical specification completion times for certain actions required when limiting conditions for operation are not met.

The U.S. Nuclear Regulatory Commission (NRC) staff has identified the need for a regulatory audit to examine TVA's non-docketed information with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of the licensing or regulatory decision.


Question: The management of waste generated at BFN must comply with state and federal regulations.

Under RCRA standards, BFN is currently classified as a small quantity generator of hazardous wastes with less than 2,200 pounds per month (1,000 kilograms per month) and it is not anticipated that this status will change during the subsequent period of extended operation.

In Section 4.11.1.5 of the 2013 GEIS, the 1996 GEIS was reviewed and it was determined that systems and procedures are in place at nuclear power plants to ensure continued proper handling and disposal of the wastes at all plants (NRC 2013).


Question: The NRC mandates the establishment of radiation exposure control procedures during emergencies.

Radiation Exposure Control is essential to ensure that emergency responders and the public are adequately protected from potential hazards during nuclear incidents.

Public Health & Sanitation plays a crucial role in maintaining community health standards, especially in emergency situations involving hazardous materials.


Question: What is the significance of kinematic viscosity in fluid dynamics?


Question: What is the COBRA-FLX critical heat flux correlation?

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation is detailed in the topical report as a methodology utilized to predict the onset of critical heat flux in nuclear reactor cores under various thermal-hydraulic conditions.

Table B-2 provides an extensive overview of several test assembly conditions used for HTP grid tests, including flow rates, pressures, and coolant temperatures.


Question: What is the strategic approach to ex-plant material harvesting in the nuclear industry?

The purpose of this effort is to develop a more systematic proactive 'strategic' approach to ex-plant material harvesting, rather than the more reactive opportunistic approach to date.

Can you fill in the relevant information for your projects on slides 7-9?


Question: peak differential pressure in pressurizer compartment

The peak differential pressure in the lower pressurizer subcompartment (node two) calculated by THREED is 12.2 psid for a surge-line DER postulated in that node.

A performance-based testing program is conducted to test containment leakage periodically through the operating life of the unit.


Question: Analysis of wind direction frequency patterns can inform safety assessments for nuclear facilities.

The annual wind direction data collected at North Anna reveal that the predominant wind direction during the winter months from 1974 to 1987 was NNW, with a frequency of 25 occurrences.

The unit of measurement used for radiation is typically expressed in sieverts, which quantifies the impact of radiation on human health.


Question: aging management review for nuclear components

To meet 10 CFR 54.21(a)(3), an application must, for each structure and component identified as subject to an aging management review under 10 CFR 54.21(a)(1), demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the period of extended operation.

The staff did not identify the need to include or issue any limitations, conditions, or ALALs in relation to the contents of the MRP-227, Revision 2, as supplemented.


Question: fire risk assessment in nuclear plants

The Fire PRA is built on a peer reviewed internal events model and is subject to cut set reviews with participation of both the development team and utility staff.

From the discussion pertaining to this Task, in general, the Fire PRA development process would have reviewed significant fire initiating events and performed supplemental assessments to address this possible source of uncertainty.


Question: passive margin development

The east coast of North America is characterized by passive margin development throughout the Cenozoic, which involves cooling and subsidence of extended (thinned) Atlantic margin continental crust and a net reestablishment of resubrimation from the eroding Appalachian cordillera, Coastal Plain and offshore depocenters.

Apatite fission track data from southeastern PA indicate roughly 0.6 to 1.2 mi. of post-Cretaceous uplift of the Piedmont relative to the Coastal Plain (Reference 2.5.1-259).


Question: What is bundling in contract requirements?

Bundling occurs when an agency decides to consolidate two or more requirements previously provided or performed under separate, smaller contracts, into a solicitation for a single contract that is likely to be unsuitable for award to a small business due to the diversity, size, or specialized nature of the elements of the performance specified; the aggregate dollar value of the anticipated award; the geographical dispersion of the contract performance sites; or any combination of the factors described.

The Director of the Small Business Program will complete the review within 10 business days and will forward the document to the SBA Procurement Center Representative for review and concurrence.


Question: what is megawatt thermal

MWt stands for megawatt(s) thermal, a unit of power used to describe the thermal output of a nuclear reactor.

National Ambient Air Quality Standards (NAAQS) are established by the EPA to regulate air quality and protect public health.


Question: conventional uranium milling

Marquez-Juan Tafoya is a conventional facility that is currently pending and has estimated reserves of 12.3 million pounds of U₃O₈.

Grants ISR Project is an ISR facility that is currently pending.


Question: What is the significance of the Performance Indicator (PI) in reactor oversight?

The Performance Indicator (PI) is a key component of the Reactor Oversight Process, providing a means to assess various aspects of reactor safety and performance.

The NRC management conducted the meeting to discuss the status of ongoing inspections and regulatory updates related to various facilities.


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

Assembly Average Burnup is a critical parameter for understanding the performance and safety of nuclear fuel, measured in GWd/MTU (gigawatt days per metric ton of uranium).

The use of zirconium alloys in the construction of advanced reactor pressure vessels enhances their resistance to corrosion and high temperatures.


Question: What is neutron fluence in the context of reactor vessel analysis?

Neutron fluence represents the cumulative number of neutrons per unit area that contact the reactor vessel shell and its internal components over a given period of time.

The last capsule withdrawn and tested from Unit 1 was Capsule V at the EOC 11 in 2002.


Question: Commission's policy on public participation in NRC meetings

The Commission's Policy Statement on 'Enhancing Public Participation in NRC Meetings' emphasizes the importance of stakeholder engagement in regulatory processes to ensure transparency and inclusivity.

In regulatory filings, it is critical to adhere to the specific guidelines provided by the Commission for submitting timely and accurate data.


Question: What are the materials used in the construction of safe ends for nuclear reactors?

1. Nozzle                                          | SA-508 Class 2                | 3              |


Question: How is core heat removal managed after a reactor shutdown?

In normal plant operation after a reactor shutdown, the core decay heat is removed by using the DEHC manual or automatic cool down function by manipulation of the turbine bypass valves discharging steam to the main condenser, when the reactor is at high pressure.

When the RHR SDC interlock pressure is clear, the RHR can be used to maintain the reactor at hot shutdown condition or go to cold shutdown conditions.


Question: What are the effects of a malfunction in the flow measurement systems on reactor performance?

Knowledge of the effect that a loss or malfunction of the following will have on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: Flow converter/comparator network: Plant-Specific

The plant is operating at 90% power with the following: - Annunciator 09-5-2-25, Flow Ref Off Norm, alarms.


Question: operating AC electrical power sources

Two offsite circuits and two DGs shall be operable to ensure the availability of vital AC electrical power sources during operation.

At least one AC electrical train, powered by an offsite circuit or an DG, is required to meet accident analyses assumptions.


Question: nuclear device registration

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

The manufacturer name is THERMO MEASURETECH, which is important for understanding the product's lineage.


Question: radiological release procedures

The operator should determine from the briefing a radiological release is not in progress and that no PAR is required.

Candidate reports complete and ready for SEC/SM approval. Completion Time (</= 15 minutes)


Question: Selenium dissolved sample analysis


Question: What does the term 'burst' refer to in the context of nuclear fuel evaluations?

Those evaluations don't tell you what happens to the fuel after the burst. It just tells you that the burst occurred and here's the size of the burst.


Question: The AEC staff states that the aquatic impacts associated with construction of the causeway would be undesirable but acceptable.

The AEC staff also indicates that the peninsula should be preserved as a wildlife region.

The volume of heavy traffic (e.g., delivery trucks and construction equipment) that would use this route if it were the only access to the site would probably cause more rapid deterioration of the already deteriorated structures of possible historic significance in the townsite.


Question: What is the Control Room Emergency Ventilation System (CREVS)?

The action requires restoring the relevant systems to ensure safety and compliance within established timelines.


Question: The NRC stipulates that the management of the heat exchanger components must address material loss due to environmental effects.

Loss of Material is a recognized aging effect requiring management for the Heat Exchanger (Sample Cooler) Shell Side Components, specifically those made of Copper Alloy ($15 Zinc and $8 Aluminum) in Closed-Cycle Cooling Water (Int).

Copper Alloy ($15 Zinc and $8 Aluminum) components exposed to Plant Indoor Air (Ext) do not exhibit any aging effects that require management.


Question: polymer weight loss measurement

Initial Poly Wt. = 66.58 Gm  % Poly Loss = 0.62%

After Thermal Test = 17,067 Gm  Poly Loss = 38 Grams


Question: What is the relationship between exemption requests and licensing actions in nuclear regulatory filings?

The NRC has not provided a clear answer except to say that the exemption request is not a basis for requesting a hearing but may be raised as a contention if it is inextricably intertwined with the licensing action.

Through a previous intervention petition, and a petition for a declaratory order, the Petitioning Organizations have tried to clarify whether the exemption request must be raised as a contention in an intervention petition or if it can be raised in another forum.


Question: What is the role of the North Chilled Water Pump in nuclear plant operations?

4.2.1 Stop 1-PP-82N, North Chilled Water Pump.

Due to Chemistry concerns ESW may be lined up to only one Chiller package at the direction of the Unit Supervisor.


Question: unescorted access authorization

The licensee shall develop, implement and maintain a program, or enhance its existing program, designed to ensure that persons granted unescorted access to the protected area of an ISFSI are trustworthy and reliable and do not constitute an unreasonable risk to the public health and safety or the common defense and security, including a potential to commit radiological sabotage.

Licensees whose ISFSI is collocated with a power reactor may choose to comply with the U.S. Nuclear Regulatory Commission (NRC)-approved reactor access authorization program for the associated reactor as an alternative means to satisfy the provisions of Sections B through G, below.


Question: What is the Phase I SAMA identification process in nuclear licensing?

The Phase I SAMA identification process is described in the ER, where risk insights and potential plant improvements from the PNPP Individual Plant Examination (IPE) and IPEEE were not fully considered.

The NRC staff has concerns with this approach because detailed benefit analysis is not performed on all Phase I SAMA candidates.


Question: nuclear quality assurance plan

All analyses were performed according to ACZ's Quality Assurance Plan.

If the samples are determined to be hazardous, additional charges apply for disposal (typically $11/sample).


Question: minimum safety injection flow requirements

At time 0550 hours, what is the minimum SI flow required from Figure A (attached)?

PSEG CONFIDENTIAL – Possession Requires Specific Permission from Nuclear Training.


Question: groundwater quality impacts

Well installation programs have the potential to impact groundwater quality by mixing water between aquifers through over-penetration or lack of well integrity.

Baseline studies conducted by HRI (HRI 1996) indicate that sufficient available drawdown exists in the Crownpoint area to ensure that the yields of the town of Crownpoint municipal wells will not be affected even in the worst-case drawdown scenario (i.e., if the currently available drawdown were 457 m [1,500 ft], then 433 m [1,420 ft] would still be available).


Question: The report lists the weighted mean activity of various nuclides, expressed in pCi per gram.

The table contains energy lines measured and their corresponding uncertainties, such as the energy line at 1460.75 keV for K-40.


Question: Digital Instrumentation & Controls Licensing Process

DI&C-ISG-06 defines the Digital Instrumentation & Controls Licensing Process Interim Staff Guidance, which outlines the regulatory framework for evaluating digital I&C systems in nuclear facilities.

The Energy Reorganization Act was enacted to restructure the regulatory framework for nuclear energy in the United States.


Question: What is the range for calibrated power measurements in nuclear reactors?

b. Calibrated                          | 4             | 0-120% of full power                        |

Recorded upper                         | 6             | Remaining 2 detectors.


Question: geological formation contact

It is important to recognize that contacts between units both above and below the Vineland are planar and undeformed, ruling out faulting as a potential source for the variability seen in the Vineland/Kirkwood contact.

The physical properties of the soils within the new plant location are discussed in Subsection 2.5.4.7 respective to dynamic response.


Question: The use of safety equipment is crucial when engaging in welding to prevent burns and electric shock.

FOR MAXIMUM SAFETY: - BE CERTIFIED FOR AND WEAR A WELDING HELMET OR FACE SHIELD. - USE PROPER EYE PROTECTION. - AVOID DIRECT EYE CONTACT WITH ARCING. - WEAR PROTECTIVE CLOTHING AND WORK GLOVES WHEN HANDLING.

OSHA M.S.D.S (Material Safety Data Sheet) is available from your manufacturer upon request.


Question: What is the significance of water level measurements in regulatory filings?

Water Level (ft above MSL) is a critical measurement in hydrology that provides essential data for understanding groundwater levels and ensuring safety in nuclear regulatory contexts.

The data table includes various dates for the measurements, highlighting the fluctuation in water levels over time.


Question: The report provides a detailed table of nuclides, including their energy levels and respective activities in terms of pCi/grams.


Question: What does Fission Power Level indicate in a BWR reactor?

Figure D.8-2 illustrates the Fission Power Level for a BWR/3 ATWS PRFO SLCS Case during the initial 600 seconds.

Figure D.8-3 displays the reactor pressures involved in a BWR/3 ATWS PRFO SLCS Case at 2800 seconds.


Question: The licensee's work procedures classified the valve internals based on contamination levels, which should have influenced safety measures during maintenance.

However, the work was performed on a non-alpha II radiological work permit (RWP), which is contrary to licensee procedure. The RWP stated that expected work conditions were less than 20 dpm/100cm2 alpha, and that work should be stopped if greater than expected contamination was encountered.

The significance of this finding is categorized as Green, indicating it does not pose an immediate threat to safety, nor require urgent remedial actions.


Question: Reactor Block Cooling Water System

The Reactor Block Cooling Water System consists of Train 1 of 2, which is designed to ensure effective cooling of the reactor block under operational conditions.

The Outside Air Natural Circulation system is also available to provide natural cooling and enhance airflow in the reactor building.


Question: What is the role of the Pressurizer Relief Tank in a reactor system?

Knowledge of the effect that a loss or malfunction of the Pressurizer Relief Tank/Quench Tank System will have on the following systems or system parameters: Containment.

The reactor coolant system is an essential part of nuclear reactor operations that transfers heat from the reactor core and cools it.


Question: radioactive liquid effluents

The inspectors observed operations, reviewed sample collection and analysis procedures, reviewed radiological effluent monitoring reports and monitoring records, and discussed the radioactive liquid effluent results with licensee staff to verify compliance with aspects of 10 CFR 20.2103 and the license application.

The inspectors reviewed records related to the release of liquid effluents to sanitary sewers to verify compliance with 10 CFR 20.2003 and the license application.


Question: hydrogen-based transient cladding strain limit

The hydrogen-based transient cladding strain limit is a technical specification critical for ensuring the structural integrity of fuel cladding in pressurized water reactors during transient conditions, as outlined in the PWROG-21001-P/NP report.

The regulatory audit plan includes a detailed analysis of compliance with existing safety regulations, assessing both historical performance data and future operational expectations.


Question: chemical and volume control system

The Chemical and Volume Control System is responsible for maintaining the proper chemical environment necessary for the safe operation of the reactor and to control the volume of reactor coolant under normal and emergency conditions.

The aging management program for the valve body involves routine inspections, assessments, and preventive maintenance to ensure that the components function effectively over their intended lifespan.


Question: What are the operational procedures for accumulator outlet valves in a nuclear reactor?

Locally restore power to accumulator outlet valves: 2-IMO-110 (2-EZC-C-5C), 2-IMO-120 (2-EZC-D-1C), 2-IMO-130 (2-EZC-D-1C), 2-IMO-140 (2-EZC-A-5C).

Operator checks RCS subcooling >40°F SAT: [ ]  UNSAT: [ ]


Question: The NRC assesses and approves procedures related to the loss of heat sink during the licensing process.

Units Loss of heat sink procedures will be identical at time of licensing.

Refer to Examination Standard (ES)-4.2, “Developing Written Examinations,” Section B.3, for deviations from the approved written examination outline.


Question: remote inspection concerns

Use of the term “remote inspection” causes concern for some.

Immediate action should be taken to expand telework for inspectors.


Question: nuclear construction permit process

The Terrapower Kemmerer Power Station Unit 1 is currently undergoing a construction permit review, which is essential for officially authorizing the physical construction of the nuclear facility.

The power station will have a significant focus on community engagement and transparency, which are vital to maintaining public trust.


Question: nuclear licensing terms

Nuclear licensing refers to the regulatory process by which a nuclear facility must undergo thorough review and obtain approval from a governing body to ensure safety and compliance with established standards.

The protocols established by the U.S. Department of Transportation are essential for the safe transport of hazardous materials, including radioactive substances, to ensure public safety during transit.


Question: The NRC evaluates safety parameters during licensing processes for integrated pressurized water reactors.

The APRM system is critical in monitoring the reactor power levels and ensuring operational safety during abnormal conditions in a BWR/6 reactor.

Results from the Bypass of one HP Feedwater Heater Test – Core flow indicate fluctuations in water levels that are expected during testing.


Question: What are the thermal limitations defined for nuclear plants during electrical energy emergencies?

The thermal limitations of this permit may be exceeded for a limited period under extreme conditions of electrical energy emergencies or under conditions that are the immediate result of electrical energy emergencies.

System Code Orange (danger) occurs when the electric system is vulnerable to instability due to a single failure, such as a potential transmission fault, loss of a generating unit, or other technical failure.


Question: What are the functions and applications of burnable poisons in a nuclear reactor?

Burnable poisons are materials used in a reactor to absorb neutrons, helping to manage the reactivity of the core throughout its operational life.

The distribution of fuel in the reactor core affects the overall efficiency and performance of the nuclear reactor.


Question: What are the procedures for documenting soil sample collection in radiological surveys?

Sample logs were completed for each boring location interval to include the collection date/time, a photograph, 2”x2” Sodium-Iodide (NaI) gamma measurements with a Ludlum 44-10 probe, large area wipes, and Geiger-Mueller (GM) beta/gamma measurements with a Ludlum 44-9 probe.

Soil samples were collected in accordance with the Pickard Hall Pre-Decommissioning Radiological Work Plan (Work Plan) and the sampling instructions developed for each area.


Question: The knowledge of emergency procedures is crucial for initially assessing operations during emergency situations.

The applicant is required to have knowledge of the bases for prioritizing emergency operating procedures.

A detailed analysis of reactor coolant system failures is essential for maintaining safety and reliability in nuclear operations.


Question: What is the significance of a licensee's performance in nuclear operations?

The performance of a licensee in nuclear operations is evaluated based on specific criteria, ensuring compliance with safety and regulatory standards as established by the Nuclear Regulatory Commission (NRC).

The results of this evaluation indicate a significant improvement in the plant's operational efficiency and safety measures, leading to enhanced public confidence in nuclear energy.


Question: What is the significance of the boric acid concentration in nuclear systems?

Basic system uses 4% boric acid instead of 12% deborating demineralizers.

The system services do not share equipment between units as do North Anna and Surry.


Question: What is the relevance of criticality safety in nuclear power plants?

Criticality safety is a crucial aspect of nuclear facility operations, particularly addressed in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 9.1.1, which outlines the criticality safety of fresh and spent fuel storage and handling.

The document outlines various regulations regarding the structural integrity of nuclear plant components, emphasizing the importance of routine inspections and maintenance to ensure reliable operation.


Question: What is defense in depth in nuclear safety?

The most common defense-in-depth measures apply concepts of redundancy, diversity, independence, and safety margin to ensure systems reliability and availability. The defense-in-depth philosophy applies to the design of a physical protection program, which integrates engineered controls and administrative controls, to provide protection against the DBT for radiological sabotage.

The only mention of diversity is Subpart J § 73.100 Technology-inclusive requirements for physical protection of licensed activities at commercial nuclear plants against radiological sabotage with about 3 pages of detailed requirements (P87119-87122).


Question: What are the requirements for the Fire Protection System when an alarm is triggered?

The alarm should be considered valid if it alarms a second time within 90 seconds of being reset.


Question: Agreement States certification process

Agreement States may also recognize the certification process(es) of specialty boards that meet the requirements in their compatible board recognition requirements.

Letters to specialty boards should include the following language: "In accordance with § 2.390, 'Public inspections, exemptions, requests for withholding,' copies of all communications, including enclosures, will be made available electronically for public inspection in the NRC Public Document Room or from the NRC Agencywide Documents Access & Management System (ADAMS)."


Question: operator licensing examination

The purpose of this letter is to provide U.S. Nuclear Regulatory Commission (NRC) approval of the proposed written examination and operating test under Title 10 of the Code of Federal Regulations (10 CFR) 55.40(b)(4) and to confirm the final arrangements for the upcoming operator licensing examination at Byron Station, Units 1 and 2.

The NRC has approved the subject examinations, and you may administer the written examination during the week of February 17, 2025, in accordance with Revision 12 to NUREG-1021, 'Operator Licensing Examination Standards for Power Reactors.'


Question: action levels in radiation surveys

Action Level based on the average background plus MDCR_surveyor.

All areas identified in the FSS plan were scanned for elevated radiation levels.


Question: What are containment air locks in nuclear safety?

SSCs for the containment air locks can be evaluated by a bounding assessment as permitted by NEI 06-09-A.

The PRA model includes an event that involves a large, pre-existing containment leak; this is bounding for the risk on an inoperable air lock and can be used as a bounding surrogate.


Question: Reactor vessel internals are crucial for maintaining the configuration of the reactor core and facilitating flow distribution.

Reactor Vessel Internals: Top Guide | Structural Support to maintain core configuration and flow distribution | Stainless Steel | Reactor Coolant and Neutron Flux | Loss of material due to pitting, crevice corrosion | Water Chemistry (B.2.1.2) | IV.B1.RP-26 | 3.1-1, 043 | A

The control room operators monitor the reactor's operational parameters to ensure safety and compliance with established protocols.


Question: The reactor maintains offsite power during certain incident conditions, as illustrated in the analysis of neutron power over time.

The reactor's core average neutron power decreases as time progresses after a steamline break while offsite power is maintained, as shown in Figure 15-26.

The data presented outlines the time intervals and corresponding changes in core average neutron power but does not address any maintenance protocols or operational procedures.


Question: What is the significance of the full width half maximum in tracer analysis?

Samples 1204336286 (L1-010-106-AJGS-004-SSDUP), 485191006 (L1-010-106-AJGS-017A-SS), 485191009 (L1-010-106-AJGS-035A-SS) and 485191015 (L1-010-107-AJGS-001C-SS) did not meet the resolution requirement of having a full width half maximum of 100 keV or less for the tracer; however, the tracer yield requirement was met and the tracer peaks are within the tracer region of interest.

The preparation method for this batch involved a dry soil preparation process that follows specific protocols for ensuring the correct analysis of nuclear materials.


Question: Facility Operating License

The NRC is issuing the enclosed Amendment No. 11 to Facility Operating License No. R-126 to reflect that modification.

By order dated February 6, 2025, the U.S. Nuclear Regulatory Commission (NRC) modified Facility Operating License No. R-126 for the University of Utah TRIGA Nuclear Reactor such that it has no fixed term.


Question: The effectiveness of alert systems can be influenced by the technologies and methods used for communication during emergencies.

Many agencies have multiple notification/alerting systems that can deliver both emergency and non-emergency information to their communities.

Considerations for incorporating social media into alerts and warnings before, during, and after emergencies include: Social media outreach is highly dependent on working cellular and data networks that may be impaired or down during and following an emergency.


Question: IEEE Standard Criteria for Safety Systems

In a previous letter addressing implementation of section 507 of the Advance Act, NEI proposed that the NRC eliminate 10 CFR 50.55a(h), which incorporates by reference IEEE Standard Criteria for Safety Systems (IEEE 279-1969, IEEE 279-1971, or IEEE 603-1991).

NEI recommends removing this condition related to leakage checks on mechanical connections. This condition results in no additional repair or replacement action by licensees.


Question: The NRC mandates thorough tests to confirm radiation levels are below regulatory limits.

Identification Energy Tolerance: 1.000 keV


Question: The regulatory process ensures that nuclear facilities operate safely and comply with established standards.

The licensing process for nuclear reactors involves a detailed evaluation of safety systems and operational practices to ensure compliance with regulatory requirements.

The design phase of a reactor typically includes considerations for thermal efficiency and fuel type, which are unrelated to the licensing process.


Question: What are the risk acceptance guidelines for nuclear reactors?

| Region     | Description                                      |
|------------|--------------------------------------------------|
| Region I   | No Changes Allowed                               |
| Region II  | Small Changes
Track Cumulative Impacts       |
| Region III | Very Small Changes
More Flexibility with Respect to Baseline CDF
Track Cumulative Impacts |

The total CDF for internal and external events for WCGS is less than 1.0E-04 yr-1, and the total LERF is less than 1.0E-06 yr-1. As shown in Figure 2 and Figure 3, these values are within the RG 1.174 total risk guidelines (Reference 4).


Question: The NRC's guidelines dictate that responses containing sensitive information must be marked appropriately as Security-Related Information.

If Security-Related Information is necessary to provide an acceptable response, please mark your entire response Security-Related Information in accordance with 10 CFR 2.390(d)(1) and follow the instructions for withholding in 10 CFR 2.390(b)(1).

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and your response if you choose to make one, will be made available electronically for public inspection in the NRC Public Document Room located at NRC Headquarters in Rockville, MD, and from the NRC's document system, the Agencywide Documents Access and Management System (ADAMS), accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html.


Question: The NRC requires monitoring of reservoir conditions to adequately prepare for possible dam failures.

The interim mitigation and monitoring plan was based on the fact that the U.S. Army Corps of Engineers (USACE) Upper Missouri Reservoir System has a very large storage capacity and that it is both unlikely and predictable in advance that the reservoirs will reach the hydrologic or sunny-day reservoir levels over the course of a water year.

Several yearly forecast conditions are published by USACE for their mainstem dams.


Question: What are the requirements for demonstrating the performance of safety systems in nuclear regulatory filings?

The NRC staff determined the RadICS platform test plans are sufficiently comprehensive to demonstrate that a RadICS platform-based safety system will perform its required safety functions in a satisfactory manner.

Application Test planning is therefore a PSAI and should be addressed by PSAI 7.2.


Question: 10 CFR 35.300 written directive

For therapeutic administrations requiring a written directive under 10 CFR 35.300 or 10 CFR 35.1000, we agree that extravasations may cause unintended permanent damage. The regulations call for treating such instances as medical events, which reasonably covers the rationale for the petition.

Dose estimates are largely hand waving and theoretical. We can use a variety of assumptions to guesstimate a dose, but ultimately the skin will be the dosimeter. But results may not be fully appreciated until months later.


Question: puncture drop height requirements

The puncture drop height is based on the requirements of 10 CFR §71.73(c)(3).

The thermal protection design features of the OCA body are critical for preventing damage during transportation.


Question: During the operation of pressurized water reactors, managing the enrichment levels is crucial for regulatory compliance.

The assembly average burnup is reported as a function of the initial assembly enrichment and the time the assembly has spent in the reactor.

The processes for planning and scheduling maintenance activities vary significantly across different types of reactors.


Question: containment performance in BWRX-300

NEDC-33911P provides design requirements, analytical methods, acceptance criteria, and regulatory bases for: Containment and the Passive Containment Cooling System (PCCS) - Containment isolation valves (CIVs) - Analytical methods evaluating containment performance - Acceptance criteria for the BWRX-300 containment performance in accordance with the design requirements specified for the containment, PCCS, and CIVs

The BWRX-300 design emphasizes robust safety features to ensure public health and safety, employing innovative technologies that address potential risks associated with reactor operations.


Question: The NRC emphasizes that Refuel Floor Radiation Monitors must be capable of detecting various types of radiation emissions during fuel handling exercises.

QC(FHP 0110-04, NEW/IRRADIATED FUEL DAMAGE, contains the following CAUTION statement regarding the radioactive release from a damaged fuel bundle: "Refuel Floor Rad Monitors and CAMS may not alarm if the fission product released is the beta emitting isotope Krypton-85." This is because the Refuel Floor and Fuel Pool Radiation monitors are gamma detectors and therefore not sensitive to beta radiation.

Plausible because time after shutdown will reduce the amount of radioactivity reduction. Incorrect because a significant amount of radioactivity is still present.


Question: diesel generators malfunction fire suppression system

DIESEL GENERATORS MALFUNCTION FIRE SUPPRESSION SYSTEM


Question: The NRC mandates the implementation of safety functions for control room systems during reactor operations.

The Control Room Systems are critical for ensuring safety functions, as indicated by the system titles and their corresponding safety functions listed in Form 3.2-2.

The exam level specified for this evaluation is for a Senior Reactor Operator, which dictates the complexity and breadth of the operational procedures being reviewed.


Question: What materials are used in temperature testing for nuclear components?

The table provides a summary of the temperature-adjusted medians for various materials tested under different conditions, including A182 and A600, crucial for understanding material performance in nuclear applications.

The testing status summary indicates that various conditions can significantly affect the efficiency of nuclear reactors, which is a critical factor in their overall safety and operation.


Question: safety injection pump regulations

SAFETY INJECTION PUMP "N" is referenced as a critical system in the operational safety protocols for reactor cooling efficacy and emergency responses.

Operator determines that the AEO must open the following breaker: 2-ABV-D-R2C.


Question: systematic measurement locations

Figure 16-4 – Survey Unit B2-010-103 Interior First Floor Wall Systematic Measurement Locations Map

DESCRIPTION: Administration Building – 1st Floor Wall Systematic Sample Locations


Question: The NRC assesses seismic hazards to ensure the structural integrity of reactor facilities.

Table 3.5-2 lists historical seismic events from 1970 to 2023 that may impact reactor safety evaluations.

The reactor's coolant system operates under high pressure, allowing for efficient heat removal during operation.


Question: nuclear licensing requirements

The licensing process for nuclear facilities encompasses a series of regulatory reviews and approvals to ensure that all safety standards are met before construction and operation.

Reactor coolant pumps are designed to circulate water in the primary loop, helping to maintain the temperature required for efficient operation.


Question: What are the requirements for initial enrichment and burnup for PWR fuel assemblies?

Max. Assembly Avg. Burnup (MWd/MTU) corresponds to the Min. Assembly Avg. Initial Enrichment (wt% 235U) for various cooling times outlined in Table B2-13.

The table indicates that the Minimum Cool Time (yrs) stated is essential for ensuring the longevity and safety of the reactor.


Question: Risk Informed Completion Time Program

If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the Risk Informed Completion Time Program without restoring the inoperable subsystem to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.

If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.


Question: core pressure during blowdown

Core pressure during blowdown is critical for understanding the dynamics of reactor cooling and pressure management during emergency scenarios.

The parameters for measuring time (s) typically include discrete intervals such as 0, 5, 10, and 15 seconds, which are important for data collection but not directly related to core pressure.


Question: What are the administrative controls for hazardous materials at nuclear facilities?

Potentially hazardous chemicals stored on the GGNS site are under administrative controls, including quantities and locations, and are evaluated as necessary to ensure no adverse effect on the safe operation of Unit 1.

Chlorine was also found to be transported on the lower Mississippi River (See [Reference 26] Subsection 2.2.2.2). The most common amount of chlorine in a single shipment was 1100 tons and it was stored in four independent cylindrical tanks.


Question: Probable Maximum Flood (PMF) analysis

The reevaluation of the PMF is in accordance with the guidance in Reference 14.

Three HEC-RAS simulations were performed: a. 21,400 square-mile downstream-centered March storm b. 7,980 square-mile Bulls Gap-centered March storm c. 7,980 square-mile Bulls Gap-centered March storm with Dam failures.


Question: What is the purpose of the Risk-Informed Extended Completion Times in nuclear regulatory procedures?

This document pertains to the adoption of TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times – RITSTF Initiative 4B', which is a regulatory process aimed at improving the management of risk in nuclear plant operations.

The NRC document, 'Risk Assessment of Operational Events, Volume 2 – External Events – Internal Fires – Internal Flooding – Seismic – Other External Events – Frequencies of Seismically-Induced LOOP Events,' focuses on evaluating various external events that could impact nuclear operations.


Question: The aging management review process includes evaluating potential material degradation such as corrosion and cracking.

The One-Time Inspection program (B.2.1.20) will be used to manage loss of material due to pitting and crevice corrosion, cracking due to SCC for aluminum, stainless steel support members; welds; bolted connections; support anchorage to the building structure.

Consistent with NUREG-2191. The One-Time Inspection program (B.2.1.20) will be used to manage loss of material due to pitting and crevice corrosion or cracking due to SCC for components exposed to an environment of air and condensation. See Subsection 3.5.2.2.2.4.


Question: The Minimum Detectable Activity for nuclides can influence regulatory compliance and safety assessments.

The Nuclide MDA (Minimum Detectable Activity) is defined as the lowest level of activity that can be reliably detected in a sample, providing critical information for radiological assessments.

The energy readings for the peaks observed during analysis were measured in keV, which is a unit of measurement commonly used in nuclear physics.


Question: What are the different operational scenarios evaluated for motor-driven pumps during accidents?

Sections 1.4 A, 1.4 B, and 1.4 C are for accident cases involving loss of normal feedwater with loss of offsite power and with different pumps operating.

As shown in the description columns, three different cases were considered.


Question: The response to radiopharmaceuticals and their monitoring highlights the challenges associated with detecting extravasations in treatments compared to chemotherapy.

Monitoring should not be required by regulators such as myself. Monitoring may be a corrective action post-medical event.

Carmine Plott notes, "While no technology exists to prevent extravasations, we rarely have extravasations because we utilize catheters, such as Angiocaths, to establish IV access."


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

A safety analysis is a systematic evaluation of a facility's design and operation to ensure that potential hazards are identified and mitigated, thereby demonstrating compliance with safety regulations.

The latest findings indicate significant advancements in reactor cooling technologies that enhance operational efficiency and reduce fuel consumption.


Question: The licensing process for nuclear facilities involves a thorough review of safety applications and documentation.

The NRC reviews the applications for licenses to operate nuclear facilities to ensure compliance with the regulations and safety standards.

The integration of renewable energy sources is essential for a sustainable future.


Question: What are the groundwater sampling requirements outlined in the NPDES Permit?

We hereby approve of your request to move the upcoming round of groundwater sampling requirements in the current NPDES Permit (SC0001848) from December, January or February to April, May, and June of this year to match the requirements of the Consent Agreement.

I apologize for the delay in responding.


Question: flooding walkdown results

Tennessee Valley Authority, Letter to U.S. Nuclear Regulatory Commission, Fleet Response to NRC Request for Information Pursuant to Title 10 of Code of Federal Regulations 50.54(f) Regarding the Flooding Walkdown Results of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. Enclosure 3, November 27, 2012.

U.S. Department of Commerce, Weather Bureau, Probable Maximum and TVA Precipitation over the Tennessee River Basin above Chattanooga, Hydrometeorological Report No. 41. June 1965.


Question: What are the common species of fish surveyed near the TRISO-X facility?

Fish and Crustacean Species from 2021-2022 Aquatic Surveys of Streams in Proximity to the Horizon Center Site

The NRC requires comprehensive assessments of environmental impacts for new nuclear facilities.


Question: Nuclear facilities must file detailed analyses regarding the impact of their operations on the local and regional environment.

The NRC requires licensees to analyze the environmental impact of their projects as part of regulatory compliance.

The list includes states such as Arkansas and Louisiana, which are part of the Central region.


Question: What is the purpose of a quality assurance process in drug testing for employees?

An SOP to implement a quality assurance process to periodically verify that employees are randomly selected for drug testing without regard to their name, office, or position was developed by SB on June 30, 2005.

The SOP was revised and implemented on January 13, 2006 (Enclosure 1). The SOP implements commitments contained in the Plan.


Question: The requirements for obtaining a nuclear license involve detailed safety analysis and regulatory compliance.

The NRC requires a detailed assessment of safety analysis in order for licensees to meet regulatory standards.

The drawing number 177D4970, Sheet 2, Revision 0, does not contain any proprietary information.


Question: Material choices for the pressurizer components are essential for ensuring structural integrity during operation.

Pressurizer nozzles are typically fabricated from a low-alloy steel forging material such as SA-508, Class 2, while pressurizer vessels are made of a low-alloy steel material such as SA-533, Grade A, Class 2 or SA-533, Grade B, Class 1.

This section discusses the material selections and related properties that will be used in the stress analyses for the PWR pressurizer components selected in Section 4.


Question: emergency plan response time

The proposed emergency plan adds 30 minutes to the augmentation response time to the Lead OSC Supervisor and FIN Supervisor minimum staff ERO position that performs the Supervision of Repair Team Activities function.

This change deviates from the current emergency plan 60-minute ERO response time requirement for the Lead OSC Supervisor, Operations Supervisor and Electrical/I&C OSC Supervisor, and the NUREG-0654 R2 60-minute ERO response time guidance for the Lead OSC Supervisor.


Question: Containment Cooling Service Water system

The safety related function of the valve is to control cooling water flow through the heat exchanger tubes to remove heat from containment.

The station determined the Division I CCSW system was inoperable for 28.5 days; therefore, an Operation or Condition Prohibited by Technical Specifications existed for greater than the Limiting Condition of Operation (LCO) time of 7 days.


Question: What is the purpose of the Fire Protection System in nuclear facilities?

The FPS provides containment isolation for a containment penetration.

Table 3.3.2-12 provides the results of the AMR.


Question: Nuclear facilities must adhere to strict licensing procedures set forth by regulatory bodies.

All applicants for a nuclear power plant license must submit documentation that demonstrates compliance with federal regulations, including safety assessments and environmental considerations.

The temperature of the coolant system is monitored continuously to ensure optimal reactor performance, unrelated to licensing requirements.


Question: generic knowledge and abilities in NRC exams

Revision 0 of the NRC's K/A Catalogs (NUREGs-1122 and -1123) included a list of system-generic K/As at the end of every system and E/APE.

Yes. However, in accordance with Section D.1.b of ES-401 (in NUREG-1021), the specific K/A statements (e.g., K1.03 or A2.11) for the examination outline must be selected in a truly random fashion.


Question: The analysis of common cause failures is critical to ensure the reliability of safety systems in nuclear reactors.

The Defense in Depth and Diversity Common Cause Failure Coping Analysis is a methodology used to assess the potential for multiple failures in safety systems and to ensure that there are adequate defenses in place to cope with such failures.

The Digital Modernization Project aims to upgrade systems and infrastructure to improve operational efficiency and safety at the Limerick Generating Station.


Question: What are the capabilities of a carbon filter in nuclear systems?

Carbon filter capabilities for removal of molecular iodine-131 and methyl iodide-131.

Design specifications for the purge system isolation valves are presented in Table 9-39.


Question: The NRC stipulates that all significant design changes must undergo a rigorous licensing process.

The Integrated Computer Code Analysis (ICCA) is used for evaluating core damage mechanisms and their impact on safety during a nuclear power plant's operational cycle.

Post-test analysis often involves disassembling mechanical components to assess their condition and performance.


Question: population density near nuclear facilities

Population density is a measurement of the number of people living per unit area, commonly expressed in terms of residents per square mile.

The North Anna site is located 10 miles from the nearest town, which features various local amenities and services for residents.


Question: The regulation mandates that high or low limits for sample results must be expressed in the same units as the corresponding measurements if the Rec% column is undefined.

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.

ACZ Project ID indicates the unique identifier assigned to specific projects within the laboratory (L89093).


Question: What is COBRA-FLX in nuclear safety analysis?

COBRA-FLX is a thermal-hydraulic analysis code developed by Framatome for the calculation of critical heat flux and coolant flow behavior in nuclear reactors, which is essential for ensuring safe reactor operation.

Table B-1 presents various test assembly conditions that were utilized for non-mixing grid tests, detailing parameters such as temperature and pressure.


Question: climatological changes severe accident risk

Will expected climatological changes affect average environmental risks of severe accidents due to either changes in severe accident probabilities or associated consequences?


Question: Magothy Formation soil properties

The Magothy Formation was encountered in two of the borings performed for the ESPA investigation (NB-1 and EB-3) and consists primarily of interbedded gray to dark gray, locally mottled, silts and clays, containing trace amounts of lignite and carbonaceous material.

Soils of the Potomac Formation were encountered in borings NB-1 and EB-3, performed for the ESPA subsurface investigation.


Question: nuclear safety regulations

The purpose of the safety analysis is to evaluate potential hazards associated with the operation of the facility and to establish acceptable limits for the radiological doses to workers, the public, and the environment.

Section 17.1 discusses the administrative procedures that need to be followed when filing a report with the Nuclear Regulatory Commission.


Question: License renewal is typically required for the continued operation of nuclear facilities.

TVA has proposed various strategies to enhance operational efficiency.


Question: It is necessary to confirm that the dose rates from the STORAGE CASK loaded with fuel assemblies remain compliant with specified thresholds.

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

The equipment must undergo regular inspections to ensure operational efficiency and safety, addressing any potential mechanical failures.


Question: The NRC mandates that the concentration of soluble boron in the PWR fuel pool must meet specific regulatory criteria for safe operations.

Minimum \(^{10}B\) loading in the neutron absorber material:

The licensee shall ensure that fuel oxidation and the resultant consequences are precluded during canister loading and unloading operations.


Question: NDM acceptability in PRA

The acceptability peer review of an NDM determines whether the NDM is suitable to be used to support a PRA and should be assessed against the NRC-endorsed requirements for NDMs from PWROG-19027-NP, Revision 2, as discussed above.

This section of the RG provides guidance to licensees and applicants on an approach acceptable to the NRC staff to demonstrate the acceptability of the base PRA used, in total or in part, to support an application.


Question: In specific operational scenarios, immediate actions taken by operators can be vital for addressing transients in the plant.

When it is determined that prompt operator action is required, communication will take place indicating action will be taken utilizing the appropriate hardcard.

Upon completion of hardcard actions, or when the transient's effects are minimized, the procedure referenced in the hardcard will be entered and actions taken will be verified using the referenced procedure.


Question: The measurement of the reject flow from the reactor water cleanup system is expressed in terms of flow rate and pressure conditions.

Flow Element 1G33-N011 measures reactor water cleanup reject flow to the condenser.

Flow is rejected to the main condenser only during plant startup.


Question: Category 1 and Category 2 quantities of radioactive material access authorization

Licensees authorized to hold Category 1 or Category 2 quantities of radioactive material must establish access authorization programs in accordance with 10 CFR Part 37, Subpart B.

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, “Implementation Guidance for 10 CFR Part 37, ‘Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material.’”


Question: The licensee shall establish a monitoring network that includes parameters for sampling and frequency.

Monitoring wells installed to comply with the licensee’s Groundwater Discharge Permit issued by the State of South Dakota may be incorporated into this monitoring network.

The Division of Decommissioning, Uranium Recovery, and Waste Programs is responsible for overseeing various aspects of nuclear material safety.


Question: The behavior of Alloy 690 in testing conditions demonstrates no initiation after 4.05 years, signifying its durability.

Data scatter is complicating application of the data to predictions for plant components.


Question: The system's design incorporates both safety valves and relief valves to prevent over-pressurization during critical events such as a loss of heat sink.

The power operated relief valves are also used to provide protection against exceeding 10CFR50 Appendix G limits (including ASME Code Case N-514), as defined by McGuire Tech Spec Figures 3.4.3-1 and 3.4.3-2 during periods of water solid operation.

In the event of a complete loss of heat sink, i.e., no steam flow to the turbine, protection of the Reactor Coolant System against overpressure is afforded by pressurizer and steam generator safety valves along with any of the following reactor trip functions:


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

Half-life too short to be able to perform the decay correction


Question: What is the significance of digital modernization in nuclear licensing?

Digital modernization is an essential process for enhancing the safety and efficiency of reactor operations; it involves the incorporation of advanced digital technologies into control systems and equipment.

The inspections conducted by the NRC are designed to ensure that all units meet the necessary operational standards and regulations.


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

Energy peaks are important in nuclear safety analysis as they indicate specific energies at which detection of radiation occurs, allowing for the identification of isotopes and quantification of radioactive materials.

The original area of the peaks provides a baseline for comparison with the ambient background measurements.


Question: The NRC evaluates potential hazards associated with radioactive materials in the site restoration process.

Strontium-90 is a key isotope assessed due to its radioactive properties and potential impact on safety during the LACBWR Site Restoration Project.

The Dry Soil Prep GL-RAD-A-021 method is employed to ensure accurate analysis of the soil samples gathered from the site.


Question: The regulations stipulate that licensees must ensure organizations involved in emergency response are trained and participate in drills.

Confirm that supporting organizations identified in the Kemmerer Emergency Plan as having an emergency response role, responsibility, or authority, will be requested to participate in scheduled drills and exercises.

Listed in 10 CFR 50.160(b)(1)(iii)(A) through (H) and performance metric and objectives for each of these functions will be developed and used for evaluation.


Question: minimum departure from nucleate boiling ratio (DNBR)

The minimum DNBR is always greater than the safety analysis limit.

This trip circuit is described in detail in chapter 7; however, it is important in this context to note that the average temperature contribution to the circuit is lead-lag compensated to decrease the effect of the thermal capacity of the RCS in response to power increases.


Question: radiological environmental impact assessment

In assessing all the data gathered for 2021 and 2022 and comparing these results with preoperational data, CEG concluded that the operation of DNPS had no adverse radiological impact on the environment.

The REMP results for 2021 and 2022 are presented in Table 3.10-1.


Question: The system continuously observes the primary circuits to identify any leakage of radioactive substances into the cooling water circuit.

The purpose of this monitor is to continuously monitor the nuclear service water loop for activity and to prevent otherwise undetected activity release from the service water system, which represents a potential release path to the environment.

A high alarm signal initiates automatic valve closure on the liquid waste discharge line to prevent further activity release.


Question: The NRC mandates that systems ensuring effective heat removal from the reactor are essential to preventing fuel damage during emergencies.

A system to assure sufficient core cooling during postulated accidents and to remove residual heat following postulated accidents shall be provided. The system safety function shall be to transfer heat from the reactor core during and following postulated accidents such that fuel and reactor internal structure damage that could interfere with continued effective core cooling is prevented.

The use of structural graphite materials in Flibe salt environments as a reflector is considered to represent a new and unique feature not typical of existing licensed light water reactor designs.


Question: natural gas air emissions regulations

Operation of a natural gas plant would result in emissions of criteria pollutants and GHGs released through the heat-recovery steam generator stacks.

A permit from the SCDHEC for air pollutants associated with the operation of the new natural gas alternative would need to be secured.


Question: design or operational changes in nuclear safety

It was determined that no design or operational changes were needed as a result of the downstream evaluations.

Provide a summary of design or operational changes made as a result of downstream evaluations.


Question: The NRC assesses whether licensees have properly evaluated seismic risks...

The NRC requires licensees to analyze seismic and geologic hazards as part of their safety review process.

Drake, A. A. Jr., "Precambrian and Lower Paleozoic Metamorphic and Igneous Rocks – South Mountain and Reading Prong," in C.H. Schultz, ed., The Geology of Pennsylvania, Pennsylvania Geological Survey and Pittsburg Geological Society, p. 37 – 50, 1999.


Question: What is SUNSI in nuclear regulatory processes?

Any contentions in these proceedings that are based upon the information received as a result of the request made for SUNSI must be filed by the requestor no later than 25 days after receipt of (or access to) that information.

If the request for access to SUNSI is denied by the NRC staff after a determination on standing and requisite need, the NRC staff shall immediately notify the requestor in writing, briefly stating the reason or reasons for the denial.


Question: What is the relevance of pCi/g in nuclear regulatory documents?

Co-60 was identified both of the sump samples, at a maximum concentration of 7.16E+00 pCi/g.

The complete gamma spectroscopy reports are presented in Attachment 7.


Question: QPTR determination in reactor safety

2. CALCULATE the QPTR for each operable power detector according to the following equations and RECORD in Attachment 4: A. QPTRₜ = Ratioₜ / Averageₜ ________ B. QPTRᵇ = Ratioᵇ / Averageᵇ ________

1. CALCULATE the average of the relative powers, Averageₜ and Averageᵇ, from each operable power range detector according to the following equations and RECORD in Attachment 4:


Question: freedom of information act process

The Freedom of Information Act (FOIA) gives the public the right to request access to records from any federal agency, including the U.S. Nuclear Regulatory Commission (NRC).

The nuclear reactor operates at a nominal power level of 1000 megawatts, which is essential for its efficiency and safety.


Question: remote inspection procedures

The team recommends that NRR in coordination with NSIR and the regional offices evaluate the current team inspection framework (i.e., one week of in-office or some on-site preparation, one or two weeks on-site with all team members going on-site, and one week of in-office documentation) to determine if and how team inspections might be more efficiently and effectively performed using available technologies.

The issuance of the guidance via the memos mentioned earlier and the March 2020 revision to Appendix E resolved many questions as to how to perform the oversight program during the COVID-19 PHE and supported the immediate needs of the inspectors.


Question: The Aging Management Program specifies procedures related to the oversight of material conditions that affect the structural integrity of components.

The Inspection of Overhead Heavy Load and Light Load (Related to Fuel Handling) Systems (B.2.1.13) program is substituted to manage the aging effect(s) applicable to this component type, material and environment combination.

The TLAA designation in the Aging Management Program column indicates that fatigue of this component is evaluated in Section 4.3.


Question: gamma scintillation detectors

The RE-0017A and B (component water) systems utilize gamma scintillation detectors to measure radiation levels, with a radiation zone of 0.25 mR/h and capable of detecting major isotopes including Co-58, Co-60, I-131, I-133, Cs-134, and Cs-137 in a range of 4.0E-7 to 4.0E-2 µCi/cm³.

The auxiliary building is crucial for housing support systems and ensuring operational continuity of the nuclear facility during maintenance and emergencies.


Question: The NRC reviews the safety features of various mining projects in locations near nuclear facilities.

Conventional oil and gas production areas within 80 km [50 mi] of the CUP are depicted in the report.

While aggregates and precious metals have been produced in McKinley County, there are no pits within 24 km [15 mi] of the CUP.


Question: What is a subsequent license renewal in the context of nuclear plants?

TVA’s proposed subsequent license renewal for the Browns Ferry Nuclear Plant located in Limestone County, Alabama.

Please let me know by April 20, 2022 if you have any questions or comments on the proposed undertaking.


Question: industrial activities near nuclear power plants

There were no significant industrial activities within 5 miles of the plant site. Based on trends of industrial growth and projected population figures, it is not expected that any major industrial expansion will occur in the area.

Virginia State Route 700 provides access to the plant site and State Routes 601 and 652 run parallel with the Lake Anna shoreline and pass about 2.2 miles northeast and 1.5 miles south of the plant site, respectively.


Question: Periodic performance testing is vital for ensuring the reliability of safety systems at nuclear power plants.

The IEEE Standard 338-1971 provides criteria for the periodic testing of protection systems in nuclear power generating stations.

Nuclear power stations utilize various cooling systems to maintain operational temperatures during peak loads.


Question: What is the purpose of the Reactor Core Isolation Cooling (RCIC) System?

**APPLICABILITY:** - MODE 1 - MODES 2 and 3 with reactor steam dome pressure > 150 psig.

**LCO 3.5.3** The RCIC System shall be OPERABLE.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves a series of regulatory steps defined by the Nuclear Regulatory Commission (NRC) to ensure safety, environmental protection, and compliance with federal laws before a reactor can be constructed or operated.

The thermal efficiency of a reactor is an important factor in its overall design, influencing the choice of fuel and the cooling system that will be implemented.


Question: regional transportation network

The existing transportation network in Church Rock includes various classifications of roads such as county roads, state highways, US highways, interstate highways, and Navajo/BIA roads.

The Church Rock Section 8 project area contains several key visual markers that are important for identifying the location within the region.


Question: What is gamma spectrum analysis in nuclear safety assessments?

Gamma spectrum analysis involves the use of a 500 ml Marinelli detector to identify and quantify radioactive samples by examining the energy levels of emitted gamma radiation.

The Dead Time was recorded at 0.29 %, indicating a brief period during which the detector is unable to record new events.


Question: analytical methods for liquid scintillation counting

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

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


Question: The assessment of model parameters such as ADIFF123 is crucial for reactor safety analysis.

Figure 10-1 illustrates the frequency distribution of the ADIFF123 model parameter, which is used in the analysis of reactor safety systems.

Figure 10-2 shows the density of the Zy-4 ACREEPH model parameter, which is unrelated to the ADIFF123 safety analysis context.


Question: The NRC's review process for safety evaluations is crucial for ensuring compliance with licensing requirements.

The Nuclear Regulatory Commission (NRC) oversees the licensing process for nuclear facilities, ensuring that safety evaluations meet federal regulations.

The hearing was conducted via video teleconference due to ongoing health and safety protocols, allowing participants to join remotely.


Question: Licensees are allowed to challenge the immediate effectiveness of an Order if they believe it lacks sufficient grounding in reliable evidence.

Pursuant to 10 C.F.R. § 2.202(c)(2)(i), the Licensee may, in addition to demanding a hearing, at the time the answer is filed or sooner, move the presiding officer to set aside the immediate effectiveness of the Order on the ground that the Order, including the need for immediate effectiveness, is not based on adequate evidence but on mere suspicion, unfounded allegations, or error.

In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section III above shall be final twenty (20) days from the date of this Order without further order or proceedings.


Question: What is EVT-1 in the context of inspections?

Accessible areas on Gussets 4 and 13 were inspected with 55% to 80% coverage obtained using EVT-1. No indications identified. Both Access Hole covers were inspected per draft BWRVIP -180 requirements.

Cracking identified on 0 degree cover. Reference OE 25794.


Question: What is a relief request for underground service water pipe coatings?

The purpose of this teleconference between the U.S. Nuclear Regulatory Commission (NRC) and Duke Energy Progress, LLC (Duke Energy) is to discuss a proposed underground service water pipe coating relief request.

This is a Closed meeting. Contains trade secrets and commercial or financial information (proprietary information).


Question: What are the ASME Code requirements for machining operations in nuclear facilities?

The ASME Code does not specifically address control of machining operations or establish special requirements in this area. It is intended, however, that the manufacturer performing machining operate under a controlled fabrication system which complies with the requirements of ASME, NCA-4000.

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


Question: The operation of small modular reactors entails the withdrawal of significant amounts of make-up water for closed-cycle cooling systems, impacting surface water resources.

For the new nuclear component (SMR) of this alternative, the overall construction and operational impacts on surface water resources would be substantially less than those described in Section 3.5.6.1 for the standalone new nuclear alternative, consisting of the offsite ALWR and the onsite SMR components.

Pipeline construction would have the potential for additional hydrologic impacts. However, water quality impacts would be minimized by the application of BMPs and by compliance with the State NPDES permitting requirements for construction activities and USACE Section 404 permits that would regulate construction of the pipeline in waterways and wetlands.


Question: What is the significance of Americium-241 in nuclear safety assessments?

Americium-241 is a radioactive isotope that can be analyzed for its presence in materials due to its potential health impacts, particularly in assessments related to nuclear safety and environmental monitoring.

The report indicates that the sample was collected on May 17, 2019, and received on July 19, 2019, which reflects the timeline of sample processing.