Question: The process of obtaining and maintaining a nuclear license involves several technical reviews and safety evaluations.

Regulatory processes for nuclear licensing often require detailed safety analyses and documentation to ensure compliance with safety standards.

The design features of the reactor include advanced control systems and redundant safety measures, which enhance operational reliability.


Question: The reported hours detail the distribution of wind conditions, showing varying speeds observed at different wind directions.

Joint Frequency Distribution of Wind Speed, Wind Direction, and Atmospheric Stability at Clinton Power Station Facility provides a comprehensive overview of meteorological conditions affecting the site, including specific data on wind levels and stability categories.

The Nuclear Regulatory Commission mandates that all facilities maintain a high level of operational efficiency and safety compliance, ensuring that all regulations are met throughout their operational lifetime.


Question: The process of obtaining a nuclear license involves detailed documentation and regulatory review to ensure safety standards are met.

The process for licensing a nuclear reactor encompasses thorough safety evaluations, technical assessments, and regulatory approvals mandated by the NRC.

The development of advanced reactor technologies continues to evolve, focusing on improving efficiency and sustainability in energy production.


Question: The USFWS has identified various species as endangered, highlighting the need for conservation efforts.

The Mexican wolf, Canis lupus baileyi, is listed as an endangered species and is part of the conservation efforts in McKinley County.

The Colorado Pikeminnow, Ptychocheilus lucius, is a fish species that has been studied for its role in the aquatic ecosystem.


Question: What is a Human Reliability Analysis in the context of nuclear safety?

Human Reliability Analysis (HRA) is conducted to evaluate the reliability of human actions within the context of safety and risk assessments for nuclear facilities.

The resolution for issue CF-B1 indicates that there is no objection to the proposed changes without necessitating further action.


Question: The phenomenon of dish filling impacts the fuel rod's free volume and internal pressure predictions during high power operations.

After pellet-cladding contact, fuel creep plays a major role in the axial mechanical interaction between cladding and pellet. During power transients, fuel creep contributes to reduced or increased cladding radial strain depending on the pellet design and conditions.

The dish filling model is calibrated from ceramography on axial cross-sections that allow the observation and measurement of changes in dish geometry after irradiation.


Question: What is the purpose of the Containment Air Release and Addition System?

The Containment Air Release and Addition System is utilized to maintain the Containment pressure between the Technical Specification limits of -0.3 and +0.3 psig.

Pressure decreases in the Containment can be the result of unit shutdown.


Question: NAC-MPC SYSTEM operations and regulatory requirements

The NAC-MPC SYSTEM does not create any radioactive materials or have any radioactive waste treatment systems. Therefore, specific operating procedures for the control of radioactive effluents are not required.

An annual report shall be submitted pursuant to 10 CFR 72.44(d)(3) or 10 CFR 50.36(a).


Question: The meters utilized in the evaporation tanks serve as pressure transducers responsible for monitoring water levels.

The level meters in the tanks are pressure transducers which will be used to monitor and control the water level in each tank.

Confined space regulations in 10 CFR 1910.146 and 10 CFR 1926 and LM confined space requirements will be followed to ensure safe entry.


Question: The topical report provides a rationale for the specific fuel design while meeting regulatory acceptance criteria established for various reactor types.

The GAIA Mechanical Design topical report (ANP-10342, Reference 1) defines the GAIA fuel design and addresses the Acceptance Criteria established in Section 4.2 of the Standard Review Plan (NUREG-0800, Reference 3). ANP-10342 is explicitly NRC-approved for use in all Westinghouse 3-loop and 4-loop 17x17 reactors; therefore, it is applicable to the fuel transition at Byron and Braidwood Units 1 and 2.

Per Section 4.0 of the SER, there are a total of four (4) L&Cs. Based on the assessments in Table 4-5, all the L&Cs are met for the fuel transition at Byron and Braidwood Units 1 and 2.


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) is a critical parameter in assessing the sensitivity of radiological measurements and indicates the lowest concentration of a nuclide that can be reliably detected above background levels.

The peak locate performed on channel 100 indicates various unidentified energy levels present during the assessment of nuclear safety systems.


Question: What is an integrity testing unit in emission factors?

The front end loader is a popular piece of construction equipment, known for its versatility in handling materials.


Question: The Petitioners challenge the effectiveness of the environmental reviews related to natural disasters such as earthquakes on nuclear plant safety.

Specifically, the Petitioners state that these issues include the generic findings regarding the environmental impacts of both design-basis accidents and severe accidents, including accidents caused by or contributed to by earthquakes.

A waiver from Table B-1 is required to litigate these issues in this proceeding.


Question: The activity levels of various nuclides are expressed in pCi/grams, which is essential for ensuring regulatory compliance.

The activity (pCi/grams) of nuclides such as EU-152 and EU-153 is critical for evaluations related to safety and environmental impact.

The energy measurements of specific isotopes like BI-214 are recorded in keV to assess their characteristics.


Question: What is the purpose of the HRA in nuclear safety assessments?

Human Reliability Analysis (HRA) is an important element in the overall safety assessment of a nuclear power plant, helping to evaluate, mitigate, and manage human error risks in the plant's operation and maintenance activities.

The NRC requires that all licensees maintain adequate training programs for their staff to ensure compliance with regulatory standards.


Question: nuclear licensing requirements

The regulatory requirements for nuclear licensing include a comprehensive assessment of safety, environmental impact, and financial qualifications of the applicant.

In the event of a system failure, immediate actions must be taken to ensure the safety of personnel and surrounding areas.


Question: The regulatory framework mandates that various dose locations must be monitored and reported accurately.

The calculated dose values at different locations are essential for ensuring compliance with safety standards.

Locations are determined based on various control parameters that ensure optimal functioning of the reactor.


Question: What does Risk-Informed Decision-Making mean in nuclear regulations?

The term Risk-Informed Decision-Making (RIDM) refers to an approach that incorporates risk assessment into the decision-making process to enhance safety and efficiency in regulatory practices.

The Safety Evaluation Report (SER) is crucial for assessing the overall safety of nuclear facilities and ensuring compliance with established regulations.


Question: The NRC outlines specific geographical coordinates for the reactor containment locations in the technical documents.

11715-FM-1B Machine Location: Reactor Containment, Plan, Elevation 262"-10", Unit 1

12050-FM-1C Machine Location: Reactor Containment, Plan, Elevation 241"-0", Unit 2


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

The COBRA-FLX is designed to provide a more accurate representation of critical heat flux conditions in nuclear reactor conditions based on advanced modeling approaches.

The test assembly conditions are crucial in determining the operational limits of the reactor and ensuring compliance with safety regulations.


Question: What does GSI-191 address in relation to PWR containment sump strainer performance?

GSI-191 identifies the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized water reactors (PWRs) and result in loss of net positive suction head (NPSH) for the ECCS and containment spray system (CSS) pumps, impeding the flow of water from the sump.

Compensatory and mitigative measures have been implemented in response to Bulletin 2003-01 (Reference 4) and GL 2004-02 (Reference 1) to address the potential for sump strainer clogging and related GSI-191 concerns.


Question: population dose reduction in nuclear safety analysis

Considering additional plant improvements since the initial license renewal, this dose reduction factor remains valid for the subsequent licensing term.

According to NEI 17-04, Rev. 1, Section 2.1 (NEI 2019b), population growth is considered new information, but not necessarily significant for the Stage 1 analysis.


Question: Adjust the letdown control valve to maintain the correct pressure in the system.

ADJUST 1-HIC-62-81A, LETDOWN PRESS CONTROL, for desired press, (320 psig at normal letdown temp), and PLACE in AUTO.


Question: What is man-rem in the context of nuclear operations?

The table provides data on the man-rem per unit for nuclear reactors, indicating the radiation exposure experienced by workers during operation.

The NRC requires detailed reporting of all operational anomalies found during routine inspections.


Question: What are the requirements for neutron absorber materials in nuclear reactors?

Minimum \(^{10}B\) loading in the neutron absorber material includes specific effective areal densities for various neutron absorber types, such as borated aluminum alloy and borated MMC, which are required to control criticality within PWR and BWR systems.

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


Question: aging management program for structural materials

Aging Management Program | Structures Monitoring Program

Concrete (Inaccessible Areas): All | Direct Flow Fire Barrier | Concrete | Buried (Structural) (Ext) | Cracking | Structures Monitoring Program


Question: What is the significance of the no-solo restriction for SRO licenses in nuclear operations?

As written, the current no-solo restriction requires another licensed operator to be in view only when the restricted operator actually manipulates a control (i.e., that small subset of apparatus and mechanisms that directly affect reactivity or power) while in the control room or out in the plant; at all other times while performing SRO licensed duties, another SRO would have to be present on site.

When the Operator Licensing Program Office last revised the wording of the “no-solo” license restriction, we tried to minimize the impact that it would have on the facilities and the individuals involved by distinguishing between those activities that require a license per 10 CFR 50.54 (i.e., manipulation of the controls that directly affect reactivity or power) and those that, with the knowledge and consent of a licensed operator or senior operator, can be performed by non-licensed personnel (i.e., apparatus and mechanisms other than controls that may affect reactivity or power).


Question: battery backed IDS bus

During a reactor trip or during the period operators are responding to a reactor trip using reactor trip response procedures, was power lost to any battery backed IDS (Class 1E DC and UPS System) bus (e.g., IDSA-DD-1, IDSA-EA-1, IDSC-EA-3)?

The current question about receiving a safety injection signal (NEI 99-02, Rev. 7, page 21, line 35) should be modified to add a note indicating this question does not apply to the AP1000:


Question: What are the design considerations for containment isolation valves in the BWRX-300?

The capability of CIVs to perform their design-based functions is safety significant because it provides assurance that the containment of the BWRX-300 can be safely isolated and can prevent a radioactive release that exceeds regulatory requirements to the environment.

GEH also stated that a complete description of compliance with SRP Section 3.6.2 and the associated BTP 3-4, using many of the assumptions from ESBWR DCD Section 3.6.1.1 to determine the appropriate protection requirements for protection against dynamic effects, will be provided in future licensing activities.


Question: SCS runoff curve number

The SCS runoff curve number for all subbasins is 98.

The table lists various subbasins along with their area in square miles and lag times in minutes.


Question: safety related equipment power supply

Provide power to safety related equipment.

3 separate and independent diesel generators and associated day tanks, 1 common fuel storage system, 2 fuel transfer pumps, piping, and associated instruments and controls.


Question: What is the review process for the Emergency Plan?

The Emergency Plan and EPIPs will be reviewed on an annual basis and updated as necessary. Updates will take into account needed changes identified by drills and exercises. Management controls will be implemented for evaluation and correction of review findings.

Due to the local relationships and the advancement of technology, the quarterly emergency telephone directory review is being changed to a semi-annual review.


Question: steam generator tube inspections

The purpose of the meeting was for the NRC staff to provide feedback on their review findings regarding traveler TSTF-577, "Revised Frequencies for Steam Generator Tube Inspections" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20160359).


Question: What are the reporting requirements for whole body radiation exposure?

Whole body dose greater than 0.25 Sv (25 rems) is subject to immediate telephone notification and requires a written report within 30 days, as stipulated by 10 CFR 20.2202(a)(1)(i).

Licensees should note that the list is incomplete in that not all potentially applicable requirements have been included.


Question: The tolerances for the Division I and II degraded voltage relays are defined in Input 4.5.2 and can be found in Section 8.4 and Figures 5 through 10.

The existing upper Allowable Value (AV) for the Division I & II voltage setting of the Loss of Voltage relays is 97 volts.

The NRC considers multiple factors in the design of nuclear reactors to ensure both safety and efficiency, focusing on equipment reliability and system redundancy.


Question: The assessment considers potential flood risks from dam failures that are closely located to the plant site.

Reservoirs selected for inclusion in the dam breach analysis are chosen based on volume of water stored and distance from the new plant.

The Delaware River is tidally influenced at the PSEG Site, therefore the 10 percent exceedance high tide is included in the results.


Question: The requested change involves using lead test assemblies to demonstrate how accident-tolerant fuel behaves under operational conditions.

The proposed amendment and exemptions allow for the use of lead test assemblies (LTAs) to demonstrate operating characteristics for accident-tolerant fuel.

SNC requests approval of the proposed amendment and exemptions by July 22, 2023 to support fuel receipt for the fall 2023 Unit 2 refueling outage.


Question: 6.9kV AC distribution system

The 6.9kV AC Distribution System is critical for supplying electrical power to various components and systems within the facility; it connects to multiple switchboards and generators to manage power distribution efficiently.

The generator has a capacity of 27 MVA and serves various units through its defined connections without compromising the operational integrity of the system.


Question: northern long-eared bat habitat loss

In its final rule listing the northern long-eared bat (80 FR 17974-TN4216), the FWS identifies forest conversion and forest modification as two of the most common causes of habitat loss.

Most vehicle activity would occur during daylight hours when bats are less active.


Question: What are the procedures for performing leak tests in radiation safety programs?

Leak tests will be performed at intervals approved by the NRC or an Agreement State and specified in the SSD registration certificate. Leak tests will be performed by an organization licensed by the NRC or an Agreement State to provide leak testing services to other licensees; or using a leak test sample collection kit supplied by an organization licensed by the NRC or an Agreement State to provide leak test kits and/or sample analysis services to other licensees and according to the kit supplier’s instructions. Records of leak test results will be maintained.

We will implement and maintain procedures for routine maintenance of our gauges according to each manufacturer’s or distributor’s written recommendations and instructions.


Question: What are the PM₁₀ emissions estimated for heavy construction activities?

Heavy Construction Operations (AP-42) present PM₁₀ emissions with various activities showing specific emission factors and totals, such as 0.84 tons per year for site preparation and 3.15 tons per year for well field and roads preparation.

The Cementing unit has an emission factor of 6.57 lb/VNM and an estimated PM₁₀ emissions of 0.66 tons per year based on 1,800 operating days, with a maximum distance traveled of 0.5 miles per day.


Question: What is the function of DC power systems in nuclear safety?

The 250 VDC Auxiliary Power System supplies power for backup lighting and the larger DC motor loads such as the turbine emergency bearing oil pump and the generator seal oil backup pump. The 250 VDC System is not essential for a safe shutdown of the reactor.

A system is provided by which all safety related Onsite Power System equipment is identified according to the particular safety channel or train that it is associated with. A color coding method is implemented with four basic colors.


Question: What are the requirements for snubber systems in nuclear plants?

PLANT SYSTEMS BASES SNUBBERS document the requirements and performance criteria for snubber systems used in nuclear plant safety systems.

ELECTRICAL POWER BASES EMERGENCY SWITCHGEAR ROOM COOLING outlines protocols for maintaining optimal temperature in emergency equipment areas.


Question: The performance of the CHANNEL CHECK and other tests is necessary to ensure the reactor trip system measures correctly.

Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1.


Question: The NRC's guidelines indicate that materials in critical components can degrade over time due to environmental factors such as neutron flux.

Material and environmental factors can significantly impact the performance of structural components within reactor systems.


Question: What is the purpose of revising procedures in regulatory licensing?

(Previous version was labeled Rev 12) JPM revised for procedure revision.


Question: risk assessment in nuclear licensing

Risk assessments are conducted to identify potential hazards and evaluate the safety protocols in place for operating nuclear facilities.

The cooling system design relies on advanced computational models to ensure efficiency and accuracy during operations.


Question: probabilistic risk assessment in nuclear licensing

NRC Regulatory Guide 1.174 provides a structured framework for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the licensing basis.

NUMARC 93-01 outlines the importance of monitoring maintenance effectiveness at nuclear power plants to ensure operational reliability and safety.


Question: license amendment requirements

Section 50.92 says that the requirements for initial licenses are -- apply here, apply to license amendments.

You know, we from this table, we attended public meetings. And so, we formulated our


Question: The NRC emphasizes the need for assessing the structural integrity of components in the auxiliary feedwater system.

Pump Casing (Aux Feedwater Hydrazine Injection Pumps) | Structural Integrity (attached) | Stainless Steel | Demineralized Water (Int) | Loss of Material               | Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19)  | VIII.G.SP-87   | 3.4-1, 016   | A     |

The auxiliary feedwater system operates under various environmental conditions that could impact its performance and safety.


Question: The design and analysis of support structures such as ice condensers are critical for maintaining safety during reactor operations.

Figure 6-135. Four Loop Ice Condenser Lower Support Structure General Assembly

Figure 6-147. Test Ice Bed Compaction Versus Ice Bed Height.


Question: What is a high pressure injection system in nuclear reactors?

High pressure injection systems are designed to inject water into the reactor core during an emergency to ensure cooling and to maintain safe operating conditions.

The inspectors verified licensee performance indicators submittals listed below.


Question: What is the maximum core thermal power evaluated for safety in nuclear plants?

The maximum core thermal power evaluated for safety considerations in the safety analysis report (SAR): 2900 MWt.

The methods and parameters used in estimating the source terms in the primary coolant are the same as those given in Regulatory Guide 1.BB.


Question: psychosocial health effects in nuclear facility operations

Although the NRC acknowledges the existence of psychosocial health effects arising from nuclear facility operations, this attribute is unlikely to influence the results of most cost-benefit analyses that it performs.

The value of an injury should include medical costs and the value of lost production (Regulatory Working Group, 1996 [Section 5]).


Question: The operator must respond to alarms regarding the Diesel Generator's lube oil pressure levels and take proper actions if the alarms persist.

Crew should recognize low Lube Oil pressure Annunciator with the Diesel continuing to operate and enter ARI 198-B.

The corrective actions listed must adhere to the established operational protocols and ensure safety compliance during testing.


Question: What is Loss of Excitation in nuclear reactors?

Loss of Excitation is a condition that occurs when the electrical supply to a synchronous generator's rotor is disrupted, leading to a loss of electromagnetic forces necessary for generating power.

Turbine Underfrequency is a protective function that triggers actions when the frequency of the turbine falls below a specified threshold.


Question: What is the impact of loss of dc control power on the Reactor Protection System?

Loss of dc control power will impair the ability of the Reactor Protection System (RPS) to perform its safety function, potentially leading to unsafe reactor conditions if not mitigated through appropriate procedures.

The examination tests cover various aspects of reactor operations and safety protocols, with a focus on maintaining high standards of performance and readiness during operations.


Question: What are the primary nuclides identified in the reactor analysis report?

The reactor cooling system operates continuously to maintain appropriate temperatures within the core, preventing overheating and potential failure.


Question: safety injection pumps

Pump Casing (Safety Injection Pumps)     | Pressure Boundary      | Cast Austenitic Stainless Steel   | Treated Borated Water >60°C (140°F) (Int) | Cracking                        | Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19)         | V.D1.E-12       | 3.2-1        | 020   | A     |

Tank (Accumulator Tanks)                 | Pressure Boundary      | Carbon Steel with Stainless Steel Cladding | Dry Gas (Int)                  | None                              | V.F.EP-22                                                            | 3.2-1           | 063          | C     |


Question: There are no changes needed to the License Renewal Application following these inquiries.

No updates to the LRA are required as a result of these responses.

The committee discussed various safety measures during the meeting.


Question: What is the significance of MDC in nuclear sample testing?

The Minimum Detectable Concentration (MDC) is a critical parameter in the evaluation of nuclear samples, serving as the threshold level for determining whether an analyte can be reliably detected above background levels.

The table lists various Client IDs associated with the samples collected during testing.


Question: What is the purpose of the proposed License Amendment (LAR) in nuclear regulation?

The proposed LAR should demonstrate a sound technical justification and/or undue hardship or unusual difficulty, the requested amendment poses minimal safety risk, acceptable plant-specific containment performance, including a plant-specific risk-informed analysis, and that containment does not have a history of significant degradation issues.

Results from testing performed since approval of the prior LAR will be added.


Question: head forging material specifications

Head Forging Material (e.g., SA-508 Grade 3, Class 1, SA-788, ASTM E208-91, ASTM A-275)

Record Background Information, Review Material Specifications and ASTM Standards


Question: generic fundamentals

The NRC staff compared K 5.30, K 5.31, and K 5.32 on Page 3.4-10 to Section 6 of the report, which contains the reactor theory K/As that are used to develop content for generic fundamentals exams.

Starting with K1.02 the values in our submittal are different through K1.36.


Question: fire protection system

Fire Barriers (Concrete Curbs) | Fire Barrier | Reinforced Concrete | Air - Indoor, Uncontrolled (External) | Cracking | Structures Monitoring (B.2.1.33) | Fire Protection (B.2.1.15)

The aging management program includes regular inspections and monitoring to ensure the structures remain fit for service and comply with safety standards.


Question: The inspector should evaluate the effectiveness of the RVHRP inspections to ensure compliance with safety standards.

The total resources required for RVHRP inspection using IP 71007 is estimated to be 425 direct inspection hours.

The ASME Boiler & Pressure Vessel Code provides guidelines essential for maintaining safety standards in nuclear facility components.


Question: The NRC's regulations mandate that a comprehensive analysis of cost implications must be included in regulatory filings.

The NRC requires licensees to incorporate financial costs and benefits into their evaluation process as part of regulatory compliance.

The NRC's commitment to transparency in the regulatory process is critical for maintaining public trust and ensures stakeholders are informed.


Question: What is the purpose of a groundwater drainage sump pump in nuclear safety?

This system is safety related because it protects a Category I seismic structure. Therefore, the groundwater sump pumps and their associated discharge valves are classified Safety Class 3 denoted in Table 3-4.

The Diesel Generator Crankcase Vacuum System is designed to reduce the concentration of combustible gases in the crankcase.


Question: What are the implications of acetone migration in environmental assessments?

While acetone and MEK can accumulate in relatively high concentrations in areas meeting the criteria just discussed, the authors have never observed its migration beyond 50 meters from the treated area (Mueller, Moreno, Przepiora, Thompson, & Majer, 2011).

At other sites, the similar production of these compounds should rapidly decline with time and attenuate within a short distance downgradient.


Question: The NRC mandates that a comprehensive steam dryer inspection strategy be developed in accordance with prior industry insights to ensure operational safety.

Long term inspections recommended for aging management of the RSDs have been specified by GEH in GE-Hitachi Nuclear Energy Report No. 007N4785 - Revision 0, “Browns Ferry Nuclear Station (BFNS) - Recommendations for Future Inspections - Replacement Steam Dryer,” dated November 2022, which was subsequently submitted to the NRC (Letter from TVA to NRC, “Long Term Steam Dryer Inspection Plan,” dated January 20, 2023) pursuant to Unit 1 Operating License Condition 2.C(18)(i), Unit 2 Operating License Condition 2.C(18)(i), and Unit 3 Operating License Condition 2.C(14)(g) which required that a long term steam dryer inspection plan be submitted based on industry operating experience.

The Browns Ferry Nuclear Station is subject to various operational requirements to maintain efficient performance and oversight of safety protocols within its reactor processes.


Question: The NRC mandates that certain pumps undergo testing to ensure they can operate within specified flow rates during emergency conditions.

Comprehensive pump tests are performed biennially for all pumps in the Inservice Testing Program. Comprehensive pump test flow rates have been established which effectively detect mechanical and hydraulic degradation during subsequent testing.

Pump Periodic Verification Tests (PPVT's) are performed biennially to verify that certain applicable pumps can meet the required (differential or discharge) pressure as applicable, at its highest design basis accident flow rate.


Question: nuclear licensing process

The nuclear licensing process involves a series of steps that must be completed to obtain permission from regulatory bodies to operate a nuclear facility, ensuring compliance with safety and environmental standards.

The primary function of a reactor coolant pump is to circulate coolant within the reactor to maintain safe operating temperatures and prevent overheating.


Question: Framatome considers certain types of information to be proprietary as they can provide a competitive edge.

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.


Question: What is the significance of subtracted area in radiation measurement?

The Subtracted Area is the difference between the Original Area of a peak and the Ambient Background, accounting for any necessary corrections due to background radiation.

The Original Area is calculated before any adjustments for background radiation are applied.


Question: The NRC mandates that the integrity of containment structures must be regularly confirmed through a defined inspection process every five years, as stated in 10CFR50.55a.

Continuing containment structural integrity is verified through regular examinations and tests performed at intervals of 5 years in accordance with the requirements of USNRC Regulation 10CFR50.55a (Reference 6.1) and, by reference therein, ASME Section XI, Subsection IWL (Reference 6.2).

Laboratory analysis of corrosion protection medium samples to determine absorbed water content, concentration of corrosive ions and reserve alkalinity.


Question: inoperable battery in nuclear systems

An inoperable Division I or II DC battery would require entry into Condition B (inoperable subsystem for reasons other than Condition A – see below).

RBS TS do not include a specific Condition for an inoperable battery.


Question: reasonable accommodation in NRC meetings

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

Link to meeting details: https://www.nrc.gov/pmns/mtg?do=details&Code=20200940


Question: What is the purpose of the ASME Section XI, Subsection IWL AMP in nuclear licensing?

Therefore, the DCPP ASME Section XI, Subsection IWL AMP has been effective in ensuring that the DCPP concrete containment will continue to operate within its licensing basis.

The most recent Unit 1 and 2 IWL inspections were completed in December 2021.


Question: Critical Safety Functions monitoring

The crew is monitoring the Critical Safety Functions.

The following parameters are observed: Containment Pressure is 5.8 psig, Power Range NI is 2.1% with an IR SUR of (+) 0.1 DPM, RCS Subcooling is (-) 15 ºF, ALL RCPs are stopped.


Question: What are the characteristics and measurements associated with nuclides?

The NRC requires licensees to submit detailed reports on emergency preparedness and response plans to ensure public safety.


Question: What are the Chemistry Factors for reactor vessel materials?

Initial NTRT °F and fluence values are critical for determining the material properties and performance of the reactor vessel during its operational life, as shown in Table 4.2.2-1 for DCPP Unit 1.

The NRC requires all reactor licensees to submit safety evaluations, which include an assessment of potential hazards and proposed corrective actions.


Question: negative heat transfer coefficients

It is significant that within the first 18.2 seconds of FLECHT run 9573, “negative heat transfer coefficients” were observed at the bundle midplane for 5 thermocouples; i.e., more heat was transferred into the bundle midplane than was removed from that location.

Additionally, the exothermic oxidation reaction of the Zircaloy heated a mixture of steam and hydrogen, and entrained water droplets.


Question: What are the acceptance criteria for emergency core cooling systems in nuclear reactors?

(1) amend its regulations in 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,” to require that the calculated maximum fuel element cladding temperature not exceed a limit based on data from cited experiments, rather than the limit of 2,200 degrees Fahrenheit (F) specified in 10 CFR 50.46(b)(1);

In your petitions, you asserted that data from multi-rod (assembly) severe fuel damage experiments indicate that specific aspects of the NRC’s regulations and associated regulatory guidance pertaining to emergency core cooling system (ECCS) acceptance criteria and evaluation models are not conservative and that additional regulations are needed.


Question: fire risk mitigation in nuclear facilities

Evaluate. This SAMA is considered to limit fuel oil fires as initiating events in the diesel areas, which was estimated for Dresden by setting the initiating event frequencies of the EDG compartments to zero.

Install additional physical barriers to limit dispersion of fuel oil from diesel generator rooms.


Question: What is the definition of Class I structures in nuclear safety?

Class I structures, systems and components are those whose failure could cause significant release of radioactivity or which are vital to a safe shutdown of the plant under normal or accident conditions and to the removal of decay and sensible heat from the reactor.

Utilities are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review.


Question: The evaluation of the inspection results should adhere to the guidelines outlined in ASME standards.

The results of inspection ASME CC 368 should be in accordance with the criteria in the subparagraph 72.1, if the anomalies are present, as described in ASME CC 368.

The staff does not agree with this comment. The staff believes that the differences in the management of the projects are adequately covered in the RG.


Question: What does renewal in entirety mean in the context of nuclear licensing?

Renewal In Entirety means that based on the review of the application, the inspection history, the current license, or a significant change in the applicable regulations, the preparation of a total license revision is warranted. An example is a license that has been amended numerous times since the last renewal, such that the scope of the program has changed.

This procedure defines that licenses are renewed in entirety. Timely and untimely applications for renewal are also discussed.


Question: What are the proposed changes in the licensing amendment application?

The proposed changes have been reviewed and approved by the DNPS Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

Attachment 6 provides the proposed License Condition wording for each unit's renewed facility operating license.


Question: The NRC mandates that measurements must accurately reflect the concentration of radioactive isotopes in environmental samples.

Minimum Detectable Concentration (MDC) is a critical value in radiological assessments that determines the lowest quantity of radioactive material that can be reliably detected in a sample.

The project involved the restoration of the LaCrosse site, ensuring compliance with environmental standards set forth.


Question: What is a tracer peak centroid in nuclear sample analysis?

The tracer peak centroid for sample 485191028 (L1-010-105-AJGS-007-SB) is greater than 50 keV from the expected library energy value for the tracer; however, the tracer yield requirement was met and the tracer peak is within the tracer region of interest.

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, with the following exceptions.


Question: What is the failure probability for isolation valve operation?

Failure Probability (demand) (Note 3) for valve SW-X7D-WI is 3.63E-04 when positioned at 'Open' during a LOCA.

The Main Steam Loop D is comprised of various components that operate under high pressure and temperature conditions.


Question: The NRC evaluates mass and energy release characteristics during cooling system accident scenarios.

Containment backpressure affects mass and energy release rates during the reflood and post-reflood phases.

The alignment of the ND system to the sump will be performed first, then the NI and NV systems follow.


Question: What is a net peak area in nuclear measurements?

Net Peak Area is defined as the area under the peak in a radiation spectrum, which represents the net count of emitted radiation from a specific energy level after accounting for background counts.

The FWHM value indicates the full width of the peak at half its maximum height, which provides insight into the resolution of the measurement system.


Question: nuclear device registration

NRC Device Key: 719665 (Internal Control Number) Distributor/Distributed By: Thermo Process Instruments, L.P Distributor License Number: L03524 Manufacturer name: THERMO MEASURTECH Device Model (Not Source Model): 5197 Device Serial Number: B1483 Transfer Date: 11/15/1981

The radioactive isotope Cs137 has an activity level of 100 mCi.


Question: What are the operational parameters that need to be recorded during the core unloading process?

- Unit 1 is currently in MODE 6 with fuel offload in progress

After identifying the appropriate procedure for the task, the examinee may be given the procedure (as applicable) and told the steps with which to begin.


Question: COBRA-FLX correlation

COBRA-FLX is a thermal-hydraulic analysis code developed to provide a comprehensive methodology for assessing critical heat flux in nuclear reactors, utilizing the ORFEO-HMP correlation for accuracy in different operational conditions.

Figure 7-2 illustrates the relationship between mixture quality (M/P) and pressure, capturing the defining and validation data sets for various thermal-hydraulic scenarios not explicitly related to COBRA-FLX.


Question: The report includes a detailed nuclide MDA (Minimum Detectable Activity) for various isotopes detectability under specific conditions.

The activity is presented in terms of counts per second with no relation to specific nuclides or their detection limits.


Question: SVEA-96 Optima3 transient validation

SVEA-96 Optima3 transient validation for bottom peaked power profile, measured temperature vs minimum CPR simulated.

WCAP-16747-NP Appendices are crucial for understanding the administrative framework of the regulatory submissions.


Question: What are the safety mechanisms in place for the steam lines in the containment?

Each 32-inch main steam line is isolated by an automatically tripped, normally open swing-check valve installed in the direction to prevent flow out of the containment.

The piping inside the containment from the penetration to the component is run so that it is protected from potential missiles generated as a result of an accident.


Question: The NRC monitors the presence of Nickel-63 in waste materials...

The report also contains detailed preparation methods for soil samples used in the analysis.


Question: CTU-4 Free Drop Test

The CTU-4 Free Drop Test No. 8 evaluates the impact of Helicopter Airborne Cartridge (HAC) drops onto the Onsite Container Area (OCA) bottom, providing critical data on structural integrity and safety performance under operational conditions.

The report details various non-proprietary findings related to reactor safety measures and operational protocols, emphasizing the importance of thorough inspections and maintenance of equipment to ensure compliance with national safety standards.


Question: What is PPS architecture in nuclear licensing?

Figure 3.2-2 illustrates the simplified architecture of the Physical Protection System (PPS) that is designed to safeguard the nuclear facility and its contents from unauthorized access and potential threats.

The Westinghouse steam generator design utilizes advanced materials and construction techniques to enhance thermal efficiency and reliability.


Question: Licensees must submit detailed safety analyses to demonstrate adequate protection of public health and safety.

The regulatory framework for nuclear licensing requires comprehensive assessments of safety measures to ensure the protection of the public and the environment.

Many nuclear facilities engage in routine maintenance to ensure their operational efficiency and prolong the lifespan of equipment.


Question: What is the significance of diverse features in nuclear safety systems?

Diverse features from the Plant Protection System (PPS) ensure that different methods are available to maintain safety and security against a range of potential failures.

The operator actions during emergencies are outlined in the Emergency Operating Procedures (EOP) to guide responses effectively under various scenarios.


Question: Probabilistic Risk Assessment

The baseline internal events PRA (including internal flood), fire PRA, and seismic PRA models document assumptions and sources of uncertainty and these were reviewed during the model peer reviews.

The model translation uncertainties evaluation and impact assessment are limited to new uncertainties that could be introduced by application of the CRM tool during RICT Program calculations.


Question: What is a 6-mile radius in relation to nuclear site vicinity?

The 6-mile radius is indicated on the map to show the zone around the Monticello Nuclear Generating Plant (MNGP) relevant for community safety and planning operations.

The various parks and recreational areas listed provide numerous opportunities for outdoor activities and wildlife observation.


Question: TER finalization process

Just a reminder to provide the feedback to PNNL so that they can finalize the TER.


Question: What factors are considered when analyzing population distribution near nuclear sites?

For the year 2020 the estimated population within 50 miles of the site is 1,650,000.

Recent figures on barge and recreational traffic use of the river both upstream at Nickajack Lock and downstream at Guntersville Lock are given below:


Question: CTU-3 Fire Test procedure

CTU-3 Fire Test 15; View ~30 Minutes after Start

The advanced safety features of the reactor design were discussed in detail during the regulatory review.


Question: The amendment request seeks to modify technical specifications to include timeframes informed by risk assessments.

The proposed amendment to the technical specifications includes the adoption of risk-informed completion times, which aims to enhance operational safety and efficiency at Arkansas Nuclear One, Unit 2.

This document also outlines the distribution of audit findings related to the various offices involved in the regulatory approval process.


Question: aging management of engineered safety features

This section provides the results of the aging management review for those components identified in Section 2.3.2, Engineered Safety Features, as being subject to aging management review.

The following tables summarize the results of the aging management review for Engineered Safety Features.


Question: emergency service power system

An EO has closed the ASCO Switch Normal Supply Breaker, supplying the output of the ESS UPS to the ESS ABT.

Unit 1 is in cold shutdown. Electrical Maintenance has just finished performing an inspection of the ESS UPS Inverter.


Question: What are the safety limits for radioactive isotopes in nuclear facilities?

The measured concentrations of various isotopes, such as H-3 and C-14, reveal significant data on release and concentration levels necessary for assessing compliance with safety regulations.

In nuclear power generation, the reactor's efficiency is often measured in megawatts, which represents the electrical output delivered to the grid.


Question: The NRC considers ITAAC notifications essential for confirming that the construction of a facility complies with planned safety protocols.

ITAAC, or Inspections, Tests, Analyses, and Acceptance Criteria, are key components of the NRC's oversight process that help ensure facilities meet safety and regulatory standards during construction.

The Federal Register often publishes proposed changes to regulations and invitations for comment from the public on safety-related issues.


Question: What are the aging management programs for bolting integrity in nuclear facilities?

Aging Management Programs: Bolting Integrity (B.2.1.10)

Pressure Boundary: Aluminum Alloy - Air - Outdoor


Question: What is the unit of measurement for uranium concentration in water quality reports?

Uranium concentration is measured in mg/L.

The pH of water affects the solubility of various compounds and is critical for understanding water chemistry.


Question: The Commission requires that organizations provide the names and addresses of members with standing to ensure proper standing in intervention petitions.

The Appeal Board noted the name and address of the member with standing serves an important and substantive role—it allows the licensing board to engage in the fact finding necessary to determine independently whether a member truly could establish standing.

The recent Supreme Court decisions have influenced the regulatory framework under which the NRC operates, impacting how challenges to its regulations can be addressed.


Question: What is the purpose of component harvesting in nuclear facilities?

Component harvesting involves the systematic collection and reuse of materials and components from decommissioned nuclear facilities to ensure safety and compliance with regulatory standards.

The NRC conducts regular inspections to ensure that all nuclear facilities comply with operational safety standards.


Question: Containment Fan Cooler System

The Containment Fan Cooler System is a critical component of the Containment and Containment Support Systems, designed to manage the temperature within the containment area during both normal and abnormal operating conditions.

The Reactor Shield Ventilation System serves to protect personnel from radiation exposure by maintaining safe levels of radiation within designated areas of the plant.


Question: What is the operational mode of a nuclear plant?

Start RWR Pump A per OP-27, beginning at Step D.1.29.


Question: what is AEA-regulated waste?

AEA-regulated waste includes liquids and solids meeting the definition of 11e.(2) byproduct material as defined by 10 CFR § 40.4: 'The tailings or wastes produced by the extraction or concentration of uranium or thorium from any ore processed primarily for its source material content.'

Disposal methods and potential impacts for the brine generated from RO are discussed in the following section.


Question: regulatory requirements for radium-226

Shall not abandon products containing radium-226. The product, and any radioactive material from the product, may only be disposed of according to Article 4 or by transfer to a person authorized by a specific license to receive the radium-226 in the product or as otherwise approved by the Agency.

The general license in subsection (H) does not authorize the manufacture, assembly, disassembly, repair, or import of products containing radium-226, except that timepieces may be disassembled and repaired.


Question: The NRC mandates an analysis for the various modes of reactor operation, particularly regarding the Reactor Mode Switch and manual initiation of safety systems.

Reactor Mode Switch—Shutdown Position | 1) Manual Initiation
a) Reactor Mode Switch – Shutdown Position
b) Manual SCRAM

1) Automatic Initiation
a) TCV Fast Closure, EHC4 Oil Pressure – Low
b) TSV – Closure
c) Reactor Vessel Steam Dome Pressure - High
d) APRM Neutron Flux – High (Fixed)
e) APRM Neutron Flux – High (Flow Biased)
2) Manual Initiation
a) Manual SCRAM


Question: What is a Mechanical Standby Liquid Control System?

Mechanical Standby Liquid Control System is utilized to provide a means of controlling the reactor coolant inventory in the event of an emergency, ensuring the safe operation of the reactor.

The NRC conducted a routine inspection of the facility to verify compliance with safety regulations and operational guidelines.


Question: The NRC sets specific criteria for the measurement of radioactive isotopes such as Americium and Curium in regulatory filings.

The measurements for Americium-241 are recorded in pCi/g, with a value of 0.00237 U.

The specific QC sample codes for the data log include QC1204246812 and QC1204246811 LCS.


Question: decommissioning funding plan requirements

The Decommissioning Funding Plan must provide reasonable assurance that funds will be available to complete the decommissioning of the facility, as required by 10 CFR 50.75.

The design features of the containment structure must ensure that the reactor is adequately protected from external hazards, such as seismic events or floods.


Question: The NRC assesses various thermal limits including fuel rim temperature and enthalpy to ensure safety during reactor operations.

Minimum margin is defined as the least negative value. A positive value indicates a violation of the limit.

The general timing of the event is provided in Table 4-1.


Question: What is the Cinch River used for in regulatory filings?

The Cinch River is included in the River System Schematic as a significant water body influencing the operational parameters and environmental considerations of the regulatory process for nuclear projects in the area.

The Watts Bar facility operates under the scrutiny of various environmental regulations that ensure the protection of local ecosystems.


Question: emergency preparedness regulations for small modular reactors

The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to include new alternative emergency preparedness requirements for small modular reactors and other new technologies.

This final rule acknowledges technological advancements and other differences from large light-water reactors that are inherent in small modular reactors and other new technologies.


Question: Understanding the function of automatic depressurization during the operational phases of a reactor...

Knowledge of the relationship between automatic depressurization system (ADS) Stage 1–3 actuation and the following systems or components: (CFR: 41.8 / 41.10 / 45.3)

Proper operation of the Engineered safeguards actuation system is crucial for maintaining safety during emergencies.


Question: The usage of logical connectors is essential in Technical Specifications to clarify the relationships among various Required Actions and their conditions.

Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies.

An effective safety culture is critical to ensure that all personnel understand their roles in maintaining operational integrity.


Question: What is the Probable Maximum Flood in hydrologic engineering?

The Probable Maximum Flood (PMF) is defined as the flood that may occur from the most severe combination of meteorological and hydrological conditions that are considered reasonably possible in a particular watershed.

Cooling water canals and reservoirs are designed to facilitate the efficient transfer of heat away from nuclear reactors.


Question: The NRC mandates that projects undergo thorough geological assessments to understand their foundation materials.

Specimen swelled after extrusion from the tube due to the large mica content.

Coloring occurs in streaks, pockets and zones throughout the specimen showing structure of parent rock.


Question: The NRC mandates a clear assessment of cumulative risks associated with external flooding for each unit during the application process.

Section 4.15.2.1 of the SLR ER provides a Cumulative Distribution Function (CDF) estimate of 2.00E-05 per year for external flooding events for each Unit 2 and 3.

Section 3.2.2 of NEI 17-04 states that “an alternate quantification process is required for any SAMAs that reduce the consequences of accidents without reducing the CDF or release category frequencies.”


Question: probabilistic risk assessment in nuclear licensing

NRC Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256) provides a framework for integrating probabilistic risk assessment into licensing decisions for nuclear facilities.

The company submitted its annual financial report to the Securities and Exchange Commission (SEC) outlining its revenues and future business strategies.


Question: What is the significance of long-term onsite storage of spent nuclear fuel?

The EPA stated that long-term onsite storage of spent nuclear fuel may be required for many decades and recommended that the draft SEIS indicate if there will be any changes in the generation of waste, including low-level radioactive waste, mixed low-level radioactive waste, transuranic waste, and hazardous and Toxic Substance Control Act wastes over the life of the program.

Commenters called attention to the beneficial economic impacts associated with the operation of the DNPS. These comments suggest that the proposed continued operation of the DNPS under renewed operating licenses would result in beneficial economic impacts in the affected area from continued property tax revenues that provide financial support for schools, higher education, worker incomes, and philanthropic support to local organizations.


Question: To minimize soil erosion and prevent unnecessary environmental impact, specific measures will be taken during construction activities.

The usual erosion control measures, such as construction of diversion swales, siltation basins, drainage ditches, terracing, and limitation of slope, will be used on access roads as well as in areas where the ground has been disturbed.

Access roads will be held to a minimum and an extreme effort will be made to limit them to tower sites only. Where an access road is necessary, visual impact, as well as soil stability, will be a prime consideration.


Question: The NRC utilizes various references to estimate the breach parameters for earthfill dams based on historical case studies and research.

B is normally 2-4 times hₑ. B can range from 1-5 times hₑ.


Question: The Black Mingo Formation is typically referred to when discussing its hydrologic characteristics, as specific aquifer systems have not yet been mapped.

Specific hydrogeologic units within the Black Mingo Formation have not yet been mapped in this area (Camille Ransom, S.C. Water Resources Commission, personal communic., 1981). Therefore, the Black Mingo "Formation" rather than "aquifer system," is most often used by hydrogeologists in describing the hydrologic properties of this unit.

In the immediate vicinity of the WEC site, wells deeper than about 125 ft would penetrate the Tuscaloosa Formation. Well 30Q-3 reportedly drilled to a depth of 140 ft probably encountered Tuscaloosa sediments.


Question: What is the significance of allegations in nuclear regulatory evaluations?

This table refers to allegations from sources external to the NRC that are still under evaluation.

The plants in the United States are monitored for safety and compliance to ensure public health and environmental protection.


Question: minimum initial assembly average enrichment of 235U

Minimum Initial Assembly Avg. Enrichment is defined by the weight percentage of 235U (E), which specifies the enrichment levels required for different reactor types.

The average temperature at which a reactor operates can vary significantly depending on the design and the cooling system used.


Question: What are severe accident mitigation alternatives in nuclear power plants?

Severe accident mitigation alternatives (SAMAs) are strategies developed by nuclear power plant licensees to minimize the consequences of a severe accident should it occur, thereby enhancing the safety and risk management of the facility.

The Applicant's Environmental Report includes various appendices detailing operational assessments, environmental impacts, and compliance with regulatory requirements, but does not specifically address severe accident scenarios.


Question: The facility's cooling system utilizes water from Wheeler Reservoir to maintain optimal operational temperatures.

The cooling towers, which create the water vapor plumes, operate when river temperatures near one or more of the National Pollutant Discharge Elimination System (NPDES) permit limits and cool the water prior to being released into Wheeler Reservoir.

Although the plant contrasts as an industrial feature in an otherwise rural setting, it is not a dominant feature from most views due to its distance from residences and the presence of wooded areas around BFN.


Question: What are the requirements for performing surveillance on safety systems?

If it is discovered that Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed from the time of discovery up to 24 hours or up to the limit of the specified Frequency, whichever is less.

Entry into a specified Condition in the Applicability of an LCO shall not be made, unless the LCO's Surveillances have been met within their specified Frequency.


Question: The analysis of pressure response includes various compartments such as the reactor cavity and steam generator.

A pressure response analysis is performed on three subcompartments: the reactor cavity, the pressurizer cubicle, and a steam generator cubicle.

The results presented in Tables 6.2-19 through 6.2-21 are for a typical plant, although the conclusions apply to North Anna.


Question: proprietary information

In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act (“FOIA”), 5 U.S.C. §552(b)(4), and the Trade Secrets Act, 18 U.S.C. §1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4).

Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH.


Question: What is the significance of activity measurement in nuclear devices?

The activity of an isotope, expressed in units such as mCi (milliCuries), quantifies the decay rate of the radioactive material and is critical for safety assessments and regulatory compliance involving nuclear devices.

Isotope SR90 has widespread applications in various fields, including geology and environmental monitoring.


Question: bulk electric power systems

The purpose of NERC is to encourage improvement in the coordination of bulk electric power systems at both the national and regional levels.

It has been estimated that the summer peak load for the entire region will increase from 70.0 GWe in 1973 to 154.5 GWe in 1982 for an annual compound growth rate of about 9.2 percent.


Question: Bayesian Inference in trust algorithms

Bayesian Inference is a probabilistic approach that allows systems to update their beliefs about an entity’s trustworthiness based on new evidence. This is done using Bayes' Theorem, which combines prior trust beliefs (based on previous interactions) with new data (such as successful or failed interactions). The result is a posterior probability, which provides an updated measure of trust.

Criteria-based trust algorithms provide more richness and detail that helps show why a user is considered trustworthy or not.


Question: nuclear regulatory safety analysis

Safety analysis is a critical part of the nuclear regulatory process, involving the evaluation of potential hazards and the implementation of appropriate safety measures to mitigate risks associated with nuclear plant operations.

The cooling system in a nuclear reactor is designed to remove heat from the reactor core to ensure safe operating conditions.


Question: peak linear power for nuclear reactors

Peak Linear Power for Prevention of Centerline Melt, kW/ft is defined as >18.0.

These values are typical values based on RCS flow of 388,000 gpm and nominal inlet temperature of 553.1°F.


Question: The NRC assessment requires thorough analysis of radioactive isotopes in waste materials.

The Certificate of Analysis details various parameters including company information, project title, and report date.


Question: What procedures are followed for soil sample collection in nuclear site assessments?

Per the sampling instruction for below slab soil sample collection, at each concrete core location, a 2”x2” NaI gamma measurement was performed and a large area vivo sample of the area was taken after advancing the core 0.5 to 1.0 inches, then a 4-inch diameter core was cut through the basement concrete slab and a final 2”x2” NaI gamma measurement was performed with the detector in the core hole.

Instrument calibration and background data was also recorded on the sample collection log.


Question: What is the purpose of a chemical injection system isolation valve in nuclear plants?

Chemical injection system isolation valve HV-5194 (normally closed, fail closed air-operated valve, non-1E) provides isolation of SG 1 from chemical injection system by remote-manual control from QCP.

Both MDPs fail to operate and the HVs remain in position (normally closed).


Question: tilt measurement in nuclear reactors

Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent power distribution measurement data.

To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the in-core tilt result may be compared against previous tilt values either using the symmetric thimbles as described above or a complete flux map.


Question: What are the key activities involved in aging management for nuclear facilities?

Aging management activities include inspection and monitoring, analysis and evaluation, mitigation and modernization, and repair and replacement to ensure the continued safety and reliability of nuclear systems.

Graphs are essential tools for visualizing data and trends in various fields, aiding in effective communication and understanding of complex information.


Question: What is the purpose of the Chemical and Volume Control System in nuclear reactors?

The Chemical and Volume Control System is designed to provide the following services to the Reactor Coolant System: 1. Maintenance of predetermined water level in the pressurizer, i.e., maintain required water inventory in the Reactor Coolant System, 2. Maintenance of seal-water injection flow to the reactor coolant pumps, 3. Control of water chemistry conditions, activity level, soluble chemical neutron absorber concentration and makeup, and 4. Emergency core cooling (part of the system is shared with the Emergency Core Cooling System).

Adequate operating performance monitoring assures system integrity.


Question: Licensees must submit documentation that demonstrates compliance with safety regulations during all phases of reactor operation.

The regulatory process for nuclear licensing includes the submission of safety analysis reports that detail potential hazards and risk assessments.

Routine maintenance procedures for non-nuclear equipment are essential to ensure the overall functionality of the plant's infrastructure.


Question: fire protection program responsibilities

In accordance with Chapter 3 of NFPA 805 a fire protection plan documented in the Fleet Fire Protection Program Manual (PD-EG-ALL-1500) and the NFPA 805 Fire Protection Program Design Basis Document defines the management policy and program direction and defines the responsibilities of those individuals responsible for the plan’s implementation.

There are two primary communication systems at the McGuire Station: the station Telephone System and the Public Address System.


Question: The instrumentation for the RCIC System must be functional across all specified conditions.

The RCIC System instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.

Separate Condition entry is allowed for each channel.


Question: boric acid corrosion management

The DCPP Boric Acid Corrosion AMP (B.2.3.4) will be used to manage loss of material due to boric acid corrosion of carbon steel miscellaneous structural components, support members, welds, bolted connections, and support anchorage to building structure for piping and component supports exposed to air with potential borated water leakage.

Not applicable. DCPP does not have any piping and component supports exposed to treated borated water.


Question: The NRC analyzed the public feedback to improve the proposed regulation.

The NRC prepared a summary and analysis of public comments received on the 2017 proposed rule and draft regulatory guide, as referenced in the "Availability of Documents" section.

The NRC hosted two public meetings to engage with external stakeholders on the proposed rule and associated draft guidance document during the public comment period.


Question: dry weight analysis

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

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


Question: What is the purpose of the Nuclear Service Water System in a nuclear facility?

The Nuclear Service Water System is designed to withstand a safe shutdown earthquake and to prevent any single failure from curtailing normal station operation or limiting the ability of the engineered safety features to perform their functions.

A flood analysis was performed as required by 10CFR50. Flood analysis criteria were also established in Regulatory Guide 1.59. Exceptions were taken to Regulatory Guide 1.59 and described in Section 3.4.


Question: nuclide activity and measurement

The activity of each nuclide is measured in picocuries per gram (pCi/gram), which quantifies the rate at which radioactive decay occurs.

The yield percentage is an important factor in determining the efficiency of nuclear reactions and is calculated based on the initial fuel mass.


Question: The likelihood of core damage is assessed during safety analyses...

The LPIs perform an important function in assessing core damage during various events by utilizing multiple performance metrics such as the Core Damage Frequency (CDF).

The Safety Evaluation Report summarizes significant findings regarding the environmental impact of nuclear power operations as mandated by regulations.


Question: What is the purpose of the THREED computer code in nuclear safety analysis?

The THREED computer code is used to calculate pressure and temperature transients in various nuclear power plant subcompartments following a

The NRC requires licensees to analyze emergency zones to ensure public safety during a nuclear incident.


Question: What are the key components of a nuclear regulatory filing?

The nuclear regulatory filing process includes documentation of safety evaluations, environmental assessments, and technical justification for proposed actions.

The graphical presentations in the report illustrate various data trends over the past decade.


Question: Strontium-90 analysis in nuclear regulatory filings

The following Analytical Methods were performed: | Method | Description                               | Analyst Comments            |


Question: bentonite seal

After the designated depth of each well was reached, and the PVC screen and casing set, the sand pack and bentonite seal were placed, and then a grout plug was placed from the top of the bentonite seal to the ground surface.

Eight of the observation wells were screened in soil and/or weathered rock and had depths ranging from about 25 to 50 feet.


Question: SIT isolation valve requirements

Verify each required SIT isolation valve is fully open when pressurizer pressure is ≥ 430 psia. In accordance with the Surveillance Frequency Control Program

Verify nitrogen cover pressure in each required SIT is ≥ 250 psia and ≤ 675 psia. In accordance with the Surveillance Frequency Control Program.


Question: What measures are being taken to protect the California Red Legged Frog during site activities?

Some things to potentially consider including in the Protection Measures: A brief training for all onsite staff so they can recognize a CRLF, know when they are most likely to be on roads, and know how to report the sighting.

The potential application of herbicides around Diablo Creek (such as on the perimeters of the north access road, the 230-kV and 500-kV switchyards, and the perimeter between the Raw Water Reservoirs and Diablo Creek).


Question: What are the power-dependent MCRP limits for nuclear reactors?

Limits support operation with any combination of 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

The table lists various conditions for power levels and corresponding MCRP values for different reactor operational states.


Question: radioactive waste management programs

Are there any proposed changes or upgrades to the radioactive waste management programs (low-level radioactive waste, spent fuel, mixed waste) being considered during the license renewal term?

In section 4.11.5.2 of supplement 2, Duke Energy notes that it is a small quantity hazardous waste generator and an infectious waste generator.


Question: What is the significance of dam failure analysis in nuclear safety?

Water control structures for the purpose of these analyses are reservoirs with dams retaining large volumes of water within the Delaware River watershed.

Seismic failure of a dam is modeled as the immediate breaching of the dam and release of water within the reservoir, as opposed to an extreme rain event which is modeled as the breach of a dam structure over a longer specified time period.


Question: NRC Commission Meeting Schedule

The NRC Commission Meeting Schedule can be found on the Internet at: https://www.nrc.gov/public-involve/public-meetings/schedule.html

There are no meetings scheduled for the week of March 1, 2021.


Question: Critical Heat Flux Correlation in nuclear reactors

The COBRA-FLX code includes the ORFEO-HMP correlation for predicting critical heat flux (CHF) conditions in nuclear reactor systems, which is essential for ensuring safe operational limits during high thermal events.

Local conditions from COBRA-FLX simulations indicate variations in pressure and temperature, which significantly affect the thermal hydraulic performance of the reactor core.


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

The energy peaks measured in keV are critical in identifying and quantifying specific radioactive isotopes present in a nuclear material sample, which informs safety assessments and regulatory compliance.

The overall facility design must adhere to stringent safety standards to ensure continued operation and public trust.


Question: The safety analysis report must include an evaluation of potential accident scenarios.

A safety analysis report (SAR) provides a comprehensive overview of the safety analyses performed to demonstrate compliance with regulations and the overall safety of a nuclear facility.

Reactor components must be inspected prior to operational use to ensure they meet design specifications.


Question: The NRC mandates that licensees apply various treatments to eliminate organisms during the reactor decommissioning process.

Disinfection, a process in which organisms are destroyed or inactivated, could be accomplished by several different physiochemical treatments, including thermal energy; ultraviolet, gamma, X-rays, and microwave irradiation; ultrasonic disruption; and chemical agents, with the latter the most common method of disinfecting water and waste waters.

Large amounts of hydrogen peroxide produced large quantities of gas that had to be vented.


Question: The NRC classifies any area where access is limited for safety measures as a Restricted Area.

**Restricted Area** - Any area where access is controlled for the purpose of radiation protection.

The Operational Support Center serves as a staging location for various emergency teams.


Question: What should a Commission paper include regarding transformation initiatives?

The paper should include sufficient information on each recommended transformation initiative to support a Commission decision on whether, why, and how staff should proceed with detailed development and implementation.

The strategy should include a communications plan(s) for each area and a change management plan to foster and sustain (both innovation and) transformation as a characteristic of the NRC’s organizational culture.


Question: What are the human factors engineering design requirements in nuclear plant operations?

The plant design must reflect state-of-the-art human factors engineering principles for safe and reliable performance in all locations that human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions.

The management of normal operations is crucial for maintaining the efficiency and effectiveness of the plant.


Question: What is the significance of minimum initial assembly average enrichment in nuclear fuel?

The Minimum Initial Assembly Avg. Enrichment wt % 235U (E) is critical for determining the performance and safety of nuclear fuel assemblies in pressurized water reactors (PWRs).

The average burnup of nuclear fuel is a measure of how much energy has been produced by the fuel over its lifespan.


Question: radioactive gaseous effluent treatment

On October 31, 2024, the inspectors presented the radioactive gaseous and liquid effluent treatment inspection results to Derek Jones, Plant Manager, and other members of the licensee staff.

The PM task was updated to include banding verification.


Question: saltwater intrusion modeling

The numerical model SEAWAT (Langevin et al., 2008) was used for the saltwater intrusion modeling effort. SEAWAT is a coupled version of MODFLOW and MT3DMS designed to simulate variable density and saturated groundwater flow; it is routinely applied to saltwater intrusion studies.

A copy of these regulatory sections is provided in Attachment 1.


Question: special nuclear material licensing

The proposed Federal action is the issuance of a license to possess and use special nuclear material, under the provisions of 10 CFR 70, that would allow TRISO-X to possess and use special nuclear material in the TRISO-X FFF to produce TRISO-based coated particles and final fuel forms using uranium enriched to less than 20 weight percent 235U.

Nuclear energy is an important source of energy in the U.S. with nearly 20 percent of the nation’s electricity being generated by nuclear reactors.


Question: What are the water quality standards for uranium as per EPA regulations?

Water quality meets EPA and NNEPA primary drinking water standard for uranium.

Water quality characterization was based on samples from six wells (CP-2, CP-3, CP-5, CP-6, CP-7, and CP-8).


Question: What are the design limits for metal primary reactor containment systems?

Regulatory Guide 1.57, Revision 2 discusses the design limits and loading combinations for metal primary reactor containment system components.

Temporary containment opening review guidance is outlined in Appendix B of the document.


Question: total grain-size distributions for tephra deposits

A compilation of total grain-size characteristics for tephra-fall deposits by Costa et al. (2016) included 3 silicic tephra-fall deposits that had measured median diameters and sorting from field data (Table 36).

The eruption Duration was adjusted until the range of column heights, eruption Power, and calculated fall deposit volumes was reasonably consistent with the characteristic VEI range (Table 35).


Question: nuclide identification


Question: What are the projected population densities near nuclear facilities over time?

The table illustrates projected population densities in various sectors around nuclear facilities ranging from 0 to 10 miles for the years 2010, 2013, 2021, 2027, 2037, 2047, 2057, and 2067.

The NRC ensures that licensees comply with the regulations by monitoring their operational practices and safety measures.


Question: containment spray system

An initial test of the containment spray header is made to check the free flow out of the spray nozzles.

These pressure sensors are part of the Containment Pressure Control System which is designed to maintain the containment pressure within design limits.


Question: The safety of drilling operations is paramount, particularly considering the geologic stability of formations being penetrated during the drilling process.

Vertical geologic cross-section depicting formations and their dominant lithology along a northwest-to-southeast transect from Aiken County to Hampton County, South Carolina.

Theories for rock breakage mechanisms for drag bits.


Question: What are the requirements for operating under an NRC license?

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

An environmental assessment for this action was not required, since this action is categorically excluded under 10 CFR 51.22(c)(14).


Question: Subsequent License Renewal Interim Staff Guidance

As discussed in NEI 17-01 "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 for Subsequent License Renewal," the NRC has encouraged applicants for subsequent license renewal to address Subsequent License Renewal - Interim Staff Guidance (SLR-ISG) issues in subsequent license renewal applications.

During the reviews of the first three SLRA's that were based on NUREG 2192, the NRC and applicants identified improvements to the guidance that would assist in preparing and reviewing future SLRA's more effectively and efficiently.


Question: risk-informed inspection programs

we had considered that to be not consistent with risk-informed regulations with the clarification that staff provided that that was meant to be substituted into your risk-informed inspection programs, we will be reconsidering that.

the welds that we've selected this time are based off of our plant-specific risk profiles.


Question: What are the requirements for nuclear license applications?

Nuclear license applications must include a detailed safety analysis that evaluates the potential risks and impacts of the proposed facility on public health and the environment.

During maintenance, the reactor's cooling system is inspected to ensure it operates within the required temperature limits.


Question: The flow rate at the Bellefonte plant is measured in cubic feet per second (cfs), indicating discharge levels during operations.

The Gunterville Dam released a steady flow of 40 cfs on Day 1, declining to 30 cfs on Day 2.


Question: What is the relationship between assembly average burnup and uranium enrichment?

The assembly average burnup is defined as the total energy produced by a nuclear fuel assembly divided by the initial energy content of the fuel, often expressed in GWd/MTU and varies with different initial assembly average enrichment levels of 235U.

Nuclear reactors are designed to convert atomic energy into heat, which can then be used to produce steam for power generation.


Question: standby gas treatment system

Table 3.2.2-7 lists the components of the Standby Gas Treatment System, which includes information on intended function, material, environment, aging effects requiring management, and aging management programs.

The aging management programs for mechanical components emphasize monitoring and inspecting various materials to prevent degradation over time.


Question: What is the response protocol for 1PM01J alarms?

Limit your action(s) to 1PM01J; if actions are needed beyond this panel, inform your evaluator.

Unit 1 is at 90% power with all systems and controls in automatic.


Question: What are the trademarks associated with Framatome?

AFA 3G, AREA, COBRA-FLX, COPERNIC, GAIA, GRIP, HMP, HTP, M5, M5Framatome, Mark-BW, MONOBLOC, ORFEO, and Q12 are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries.

This report outlines the expansion of nuclear safety measures implemented across different reactor types.


Question: What is the significance of differential pressure in reactor systems?

The differential pressure measurement is critical in assessing the flow dynamics within a reactor system, as it indicates the pressure changes across various nodes, which can affect the reactor's operational safety and efficiency.

The recorded times from the data are essential for understanding the response rate of the system under evaluation.


Question: minimum cooling time for nuclear fuel assemblies

The minimum cooling time for nuclear fuel assemblies is determined based on the assembly average enrichment and burnup, which are specified in the loading tables.

In contrast, the operational efficiency of a nuclear reactor is influenced by various factors like fuel type, reactor design, and environmental conditions.


Question: The use of radioactive substances in medical imaging is tightly regulated to ensure safety and effectiveness.

Luckily for my father, or at least I would like to think so today, although I would never know otherwise, the radioactive material was properly injected so that the oncologist made the right health decision for him.

After graduation, I went on to work for a medical device company. The very same company that manufactured the infusion pump used to infuse chemotherapy into him.


Question: What is 0-AOI-29 related to in nuclear safety?

The NUSO will enter 0-AOI-29, Dropped or Damaged Fuel Or Refueling Cavity Seal Failure, which will direct the crew to close 2-ISV-78-600, FUEL TRANSFER TUBE ISOL.

Plausible because checking to ensure CVI has occurred is directed in the same procedure, but another section, for a dropped or damaged irradiated fuel assembly.


Question: average shipment of radwaste

The average annual shipment of radwaste over 40 years is 2,070 cubic feet (58.6 cubic meters).

Applying the RG 4.2 limit of 2.34 cubic meters (82.6 cubic feet) per shipment yields an annual average of approximately 25 shipments, which scales to approximately 75 shipments when normalized to the power level of the reference reactor.


Question: The NRC is reviewing the application for the construction permit of the Kemmerer Power Station Unit 1.

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

The construction permit review will include analysis of the proposed facility's impact on local resources and communities.


Question: What are the Category 1 and Category 2 issues in the NRC License Renewal EIS?

Category 1 covers issues that are known to have no or a small impact on the specified area (as they are generic issues seen at all nuclear plants), whereas Category 2 issues could have moderate or large impacts, and need to be addressed on a site-specific basis.

The NRC staff acknowledges the comment concerning the use of acronyms and the confusion regarding their use.


Question: What are vertical and horizontal lift tubes in nuclear systems?

Vert. L. Tubes represent a critical component in the reactor where the vertical lift tubes transport fluids through the system effectively.

Condensate is often measured in various pressure and temperature conditions to ensure efficiency in the reactor's systems.


Question: maximum permissible limits for radioactive contamination

The level of non-fixed (removable) radioactive contamination on external surfaces of packages offered for transport must be kept as low as reasonable achievable, and shall not exceed the values shown in the following table:

Each vehicle shall be inscribed with the words 'Radioactive Materials Use Only' in letters at least 76 mm (3 in) high in a conspicuous place on both sides of the exterior of the vehicle.


Question: What is the significance of design basis flooding in nuclear power plants?

ANSI/ANS-2.8-1992. Determining Design Basis Flooding at Power Reactor Sites.

Missouri River, Mainstem Reservoir System, River Regulation Manual, Master Manual, U.S. Army Corps of Engineers, Omaha, Nebraska, 1979.


Question: reactor protection system

controls and safety rod actuation by the reactor protection system and diversity of its parts;

If you have any questions or need any additional information, please contact me at riotti@arc-cleantech.com or (732) 890-3602; or Raymond Burski, ARC 100 Licensing Director at rburski@arc-cleantech.com or (504) 909-6436.


Question: The performance of the low pressure coolant injection subsystem is crucial when ensuring the operability of the safety systems during specific operational modes.

Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of eight safety/relief valves shall be OPERABLE.

Each of the listed conditions requires a specific action to ensure compliance with operational standards, including restoration of failed components within designated time frames.


Question: The design of the Nuclear Steam Supply System (NSSS) is a critical factor in the overall efficiency and safety of a nuclear reactor.

The Nuclear Steam Supply System (NSSS) Design is crucial to ensuring the reactor operates within its thermal limits and safety parameters.

The duration for collecting emergency response data is an essential part of safety planning for nuclear plant operations.


Question: COBRA-FLX heat flux correlation

The COBRA-FLX model implements the ORFEO-HMP Critical Heat Flux Correlation to enhance the prediction accuracy of heat transfer in nuclear reactor environments.

The analysis was conducted following the standard procedures outlined in the regulatory framework for nuclear facility operations.


Question: What is the difference between BWR and PWR in terms of safety considerations?

But it should have a way actually for everybody to use. You cannot open the door for one plant and close the door for others. That's just something basically especially BWR. I think we have to realize BWR is different from PWR.

And PWR, yes, there's something like a leak before break and ALS, all of this.


Question: What is an aging management program for nuclear reactor components?

A plant-specific aging management program is to be evaluated because existing programs may not be capable of mitigating or detecting crack initiation and growth due to SCC in the vessel flange leak detection line.

Not applicable. This item is only applicable to BWRs.


Question: What is the purpose of regulatory filings in the nuclear industry?

Regulatory filings are essential for ensuring that nuclear facilities comply with safety standards and operational protocols set forth by the Nuclear Regulatory Commission (NRC).


Question: What is the significance of the Mississippi River alluvium in hydrogeology?

The Mississippi River alluvium is the principal aquifer at the site. With the potential for induced recharge from the river, the alluvial aquifer is capable of meeting plant service water requirements.

Ground water migrates downward through the terrace deposits and accumulates on the terrace clay lenses above the Catahoula or at the terrace/Catahoula contact above the regional water table.


Question: What is the monetary valuation of routine public dose reductions in nuclear facilities?

The steps are as follows: 1. Estimate reductions in public health (routine) risk per facility for implementation (D_{PRI}) and operation (D_{RO}). 2. Convert each reduction in public health (routine) risk per facility from person-rem to dollars via monetary evaluation of health effects.

Because the quantity Z_{PHA} only accounts for the risk of an accident in a representative year, the result is the expected loss over the facility life, discounted to present value.


Question: weighted mean activity calculations

Weight mean Activity is calculated using various factors, including the confidence index and uncertainty associated with each nuclide measurement.

The report includes confidence levels and uncertainties for various nuclides, providing a detailed look at the measurements.


Question: What is the purpose of Required Action C.2 in nuclear safety operations?

Required Action C.2 minimizes the consequences of a loss of decay heat removal event by optimizing conditions for RCS cooling in MODE 6 using IRWST injection. Additionally, the potential for a criticality event is minimized by suspension of positive reactivity additions.

Completion Times are reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems.


Question: flood hazard reevaluation report

The most recent information on External Flooding at Fermi 2 is the Fermi 2 Flood Hazard Reevaluation Report (FHRR) submitted to NRC for review on March 8, 2013.

As outlined in Table 4.0-1 of Reference, all mechanisms that were evaluated were shown to have a hazard elevation less than that of the current design basis.


Question: fire protection acceptance criteria

The specific acceptance criteria will include: No visual indications (outside those allowed by approved penetration seal configurations) of cracking, separation of seals from walls and components, separation of layers of material, or ruptures or punctures of seals; No significant indications of cracking and loss of material of fire barrier walls, ceilings, and floors and in other fire barrier materials; No visual indications of missing parts, holes, and wear; No visual indications of cracks or corrosion of fire damper assemblies; No deficiencies in the functional tests of fire doors; and No indications of excessive loss of material for the visual inspection of the carbon dioxide fire suppression system.

Revise implementing procedures to require, if any sign of degradation is detected (during the inspection of penetration seals) within that sample, the scope of the inspection will be expanded to include additional seals in accordance with the BFN Fire Protection Requirements Manual.


Question: Operators must understand the limits for operation concerning the loss of secondary cooling systems.

K/A is matched because the applicant is required to have knowledge of limiting conditions for operation related to a Loss of Secondary Heat Sink.

The NRC retains authority over the licensing process for commercial nuclear power facilities.


Question: What is the purpose of the Reactor Safety Study?

The initial results of these efforts are expected to be available in 1974. This study is called the Reactor Safety Study and is an effort to develop realistic data on the probabilities and sequences of accidents in water cooled power reactors, in order to improve the quantification of available knowledge related to nuclear reactor accidents probabilities.

The man-rem estimate was based on the projected population within 50 miles of the site for the year 2020.


Question: What is the significance of average peak fuel burnup limits in nuclear reactors?

As discussed in Section 2.2.1 of this report, average peak fuel burn up limit for MNGP during the terms of the extended licenses is not expected to exceed 62 GWd/MTU.

The 2013 GEIS (NRC 2013a) concluded that the environmental impacts from accidents at low power and shutdown conditions were generally comparable to those from accidents at full power, based on a comparison of NUREG/CR-1150 (NRC 1990), which assesses five nuclear plants, and NUREG/CR-6143 (NRC 1995), which specifically analyzes Grand Gulf.


Question: The NRC established that a written examination must effectively evaluate the candidate's ability to operate safely during unplanned scenarios.

For Tier 1 questions, clarified how to test an applicant’s knowledge of how to safely operate the plant during emergency and abnormal conditions.

No longer need to get NRC chief examiner permission to reject/replace a K/A (must document why K/A rejected and why replacement K/A acceptable) on Form 4.1-1 Record of Rejected K/As.


Question: nuclear licensing compliance requirements

The compliance requirements for nuclear licensing are detailed in various regulatory documents, including those issued by the NRC, which outline the necessary procedures and standards that must be met by applicants to ensure safe operation and environmental protection.

The remediation services provided by AECOM encompass a wide range of activities, including soil and groundwater remediation, hazardous waste management, and facility decommissioning.


Question: The safety analysis must account for the cooling times related to various assembly average burnup levels.

### Certificate of Compliance No. 1031 B2-74 Amendment No. 1, Revision No. 3


Question: What is the purpose of the NRC exam for nuclear plant operators?

The NRC's examination process is designed to ensure that operator license applicants possess the necessary knowledge and skills to operate a nuclear power plant safely and effectively.

Watts Bar Nuclear Plant utilizes a variety of systems to manage coolant flow and maintain reactor temperatures.


Question: What are the requirements for evaluating instrumentation and control systems in nuclear facilities?

Instrumentation and control systems must be evaluated to ensure they meet the necessary safety and operational requirements as outlined in regulatory filings.

The research outlines various reactor types and their operational characteristics, with a focus on the implications of design variations.


Question: radiological impact analysis

The radiological impact analysis is a critical component of the licensing process that assesses the potential radiation exposure to the public and the environment resulting from the operation of a nuclear facility.

The analysis was conducted using various compartments to evaluate the distribution and flow of fluids within the system, ensuring that safety regulations are met.


Question: normal power supply to 6.9kV Shutdown Board

CSST C is the normal power supply to 6.9kV Shutdown Bd. 2A-A.

Plausible because CSST A is a 6.9kV power supply to Start Bus A and RCP Start Bus A.


Question: The NRC conducts a safety review to evaluate licensing actions and related inspections at nuclear facilities.

To discuss issues associated with the safety review of licensing actions and/or Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) closures for SNC's Vogtle Electric Generating Plant, Units 3 and 4

This is a Category 1 meeting. The public is invited to observe this meeting and will have one or more opportunities to communicate with the NRC after the business portion of the meeting but before the meeting is adjourned.


Question: emergency power for ECCS

An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety.

Normal ESF operating status and deviations from this status to include the ECCS and associated power sources is controlled by the NAPS TSs.


Question: What is an Incidental Take Report in the context of environmental compliance?

Attached is the Incidental Take Report and Data Collection Form for a deceased Atlantic sturgeon collected on 1/30/2025 from the Salem Generating Station during routine raking of the circulating water intake trash bars.

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