Question: The ASME Boiler and Pressure Vessel Code regulates the design information for these components.

The design information for these components is regulated under the ASME Boiler and Pressure Vessel Code.

This code applies only to manufactured items that meet the relevant standards.


Question: The control room ventilation system includes air supply systems with cooling and heating capabilities for maintaining optimal conditions.

The Control Room, Control Room Area and Switchgear Rooms have 100 percent standby air handling equipment.

Air conditioning maintains the Control Room and Control Room Area at approximately 75°F and the Switchgear Rooms at approximately 85°F.


Question: tornado-generated missile impacts

SSC that are classified as ITS are designed for, or administrative provisions are implemented to protect against impacts from, tornado-generated missiles.

The WCS CISF is located in Andrews County, Texas which has a semi-arid climate with approximately 16 inches of rain per year.


Question: The prevailing wind directions and speeds are crucial for assessing environmental impacts.

Figure 3.1 displays the wind rose for Gallup, New Mexico, illustrating the average annual wind direction and speed conditions from 1976 to 1980 as reported by the U.S. Department of Commerce.

The temperature data collected from various regions were vital in establishing the climate patterns over several decades.


Question: ASME Code stamped vessel regulations

Requires preparation of a Design Specification, Design Report, Overpressure Protection Report, Certification of Construction Report, Data Report, and other administrative controls for an ASME Code stamped vessel.

MPC Enclosure Vessel is designed and will be fabricated in accordance with ASME Code, Section III, Subsection NB to the maximum practical extent, but Code stamping is not required.


Question: assembly average burnup

42 < Assembly Average Burnup ≤ 43 GWd/MTU


Question: The NRC emphasizes the need for a risk-informed approach that balances uncertainty and safety margins during the assessment of PRA accident sequences.

In keeping with NRC policy on risk-informed regulation, the ultimate screening of PRA accident sequences shall be based on the specific application, with appropriate consideration of uncertainties, sensitivities, and maintaining sufficient defense-in-depth and safety margin, such that the user is implementing a “risk-informed” rather than a solely “risk-based” approach.

The documentation should include all pertinent details regarding the facility's operational history and compliance with previous regulatory criteria.


Question: The NRC mandates that all safety evaluations include visual inspections of critical welds to ensure integrity.

The relevant condition for the MGW, MAWs and LAWs is a detected surface breaking crack-like indication.

If the UIA fuel alignment plates screen in for either fatigue on the needed 80-year screening assessment, the fuel alignment plate is identified as a Primary category specific component per Item C10 in Tables 4-2 and 5-2 of MRP-227, Rev. 2.


Question: The percentage of visible plumes that extend beyond 0.25 miles is reported during different seasons...

Data from Calculation ENT0002-CALC-003, Plume Characteristics Of Proposed New Cooling Towers At Grand Gulf, Rev. 2 (Reference 32).


Question: What is the significance of the enrichment percentage of Uranium-235 in nuclear fuel assemblies?

Enrichment wt % 235U (E) is a critical factor in determining the fuel assembly's performance, particularly ranges of enrichment such as 1.3 ≤ E < 1.5 and 1.5 ≤ E < 1.7, which indicate the concentration of the fissile isotope in the fuel.

The data in Table B2-32 is used to standardize the reporting of thermal performance metrics across different types of nuclear reactors.


Question: What are the control room actions to take during an irradiated fuel handling accident?

Operator has performed actions IAW 12-OHP-4022.018-004, Irradiated Fuel Handling Accident in Containment Building – Control Room.

At completion of evolution, announce the JPM is complete.


Question: nuclear fuel assembly burnup

| Avg. Enrichment wt % 235U (E) | 49 < Assembly Average Burnup ≤ 50 GWd/MTU |                         |                         |                         |                         |                         |

Nuclear power plants are rigorously designed to ensure safety during normal operations and potential emergency conditions.


Question: The MRB reviews the technical quality and efficacy of licensing actions related to nuclear materials.

Consideration of the Program’s IMPEP Report. Presentation and Discussion of Findings Associated with Following performance indicators: iv. Technical Quality of Licensing Actions.

The meeting was convened with members discussing various performance indicators related to the materials inspection program.


Question: Risk Informed Completion Time Program

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

The company has implemented new training procedures for all staff to enhance safety protocols.


Question: What are the requirements for diesel generator operability during different MODES?

The DG LOPS actuation function shall be OPERABLE in MODES 1, 2, 3, and 4 because ES Functions are required to be OPERABLE in these MODES.

A Note has been added to the ACTIONS indicating that separate Condition entry is allowed for each Function.


Question: The evaluation of large break loss-of-coolant accidents is critical in maintaining safety protocols.

The Large Break Loss of Coolant Accident (LOCA) analysis evaluates the impact of significant coolant loss on reactor safety systems.

Framatome Inc. provides various technical services related to nuclear energy production and plant operations.


Question: What is the importance of calculating intake-discharge differential temperature in nuclear power plants?

Identifies intake-discharge differential temperature exceeds the limit.

Calculates and records average plant discharge water temperature as 78°F [(76+75+81+80)/4].


Question: The NRC mandates assessments of design stresses for components like jet pump auxiliary wedge assemblies over extended durations.

Five (5) DNPS jet pump (JP) assemblies have had auxiliary wedge assemblies installed to maintain lateral support for the jet pump inlet mixer. The design considered potential aging effects, including loss of preload due to radiation effects based upon a design life of 40 years.

For Unit 2, the maximum fluence at the limiting auxiliary wedge assembly was determined to be 2.80E+18 n/cm² at the end of the SPEO. For Unit 3, the maximum fluence at the limiting auxiliary wedge assembly is 2.78E+18 n/cm² at the end of the SPEO. Each of these projected fluence values are less than the 1.40E+20 n/cm² fluence value assumed in the design basis to assess and qualify the auxiliary wedges.


Question: What are the stages of nuclear licensing for reactor design?

The nuclear licensing process consists of several stages, including pre-application, application submission, and review by the NRC, which ensures that safety standards are met before a reactor can be built and operated.

The design of a reactor must account for various thermal dynamics, which influence how the reactor core achieves criticality during operation.


Question: What regulatory requirements cover the installation of facility components within restricted areas?

The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

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


Question: Reactor Recirculation Control System

Enters and directs actions of QCOA 0202-08, Reactor Recirculation Control System Trouble.

I suspect a clogged cooling line. I have the System Engineer out here investigating the problem as well.


Question: What is the significance of Break Flow Rate in reactor safety assessments?

The Break Flow Rate is an important metric measured in min^-1 that quantifies the flow of coolant in the event of a pipe break, which is critical for sustaining long-term core cooling during post-accident scenarios.

The table outlines various time intervals for assessing performance metrics during reactor operation but does not specifically address flow rates or cooling mechanisms.


Question: What is the intended function of the Auxiliary Feedwater System's pump casing?

Pump Casing (Aux Feedwater Hydrazine Injection Pumps) has the intended function of being a Leakage Boundary (spatial).

The aging management evaluation includes a summary of various components and their functions, but does not specifically define the structural integrity of other reactor systems.


Question: The safety of spent fuel storage involves ensuring that criticality is prevented during storage and that there is adequate radiation shielding.

The design bases of the spent fuel storage system are the following: 1. The prevention of criticality during storage. 2. The prevention of damage to the fuel. 3. Adequate radiation shielding. 4. Protection against radioactivity release. 5. Adequate monitoring of the fuel storage. 6. Ability to withstand design seismic loads.

No degradation of the storage racks is expected; thus no routine inspections are made. Any unusual deterioration would be noted during routine fuel handling.


Question: How do climatological changes impact environmental risks of severe accidents?

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

ACC-3 Will expected climatological changes affect the severe accident mitigation alternative (SAMA) cost-benefit of the proposed facility?


Question: What is the emergency plan classification for a fire at a nuclear plant?

The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. The occurrence of ANY of the following hazardous events: - Seismic event (earthquake) - Internal or external flooding event - High winds or tornado strike - FIRE - EXPLOSION - River reservoir level less than 666 feet as reported by River Operations.


Question: What does S/RO mean in the context of nuclear operations?

The S/RO, or Senior Reactor Operator, is a designation that indicates a qualified individual who has met the regulatory requirements to supervise the operations of a nuclear reactor.

The NRC K/A system number is used to categorize and track the knowledge and abilities required for various job performance measures.


Question: The NRC mandates that operators declare critical components inoperable when redundancy is lost, particularly during offsite power outages.

Declare required feature(s) with no offsite power available inoperable when the redundant required feature(s) are inoperable.

Complete the restoration process within specified timelines to ensure operational readiness.


Question: What is the purpose of the One-Time Inspection program for aging management?

The One-Time Inspection (B.2.1.20) program will be used to manage long-term loss of material of the carbon steel exposed to reactor coolant and treated water in the Reactor Coolant Pressure Boundary System and Reactor Vessel.

Thermal insulation is addressed as a commodity in the Insulation Commodity Group System.


Question: In the Peak Analysis Report, various energy levels in keV are measured corresponding to different peak areas.


Question: What is the purpose of a Safety Analysis Report in nuclear licensing?

A Safety Analysis Report (SAR) provides a detailed assessment of the safety aspects of a proposed nuclear facility, including potential hazards and the measures to mitigate them.

The reactor's cooling system is designed to operate efficiently under a wide range of conditions to maintain optimal temperature levels.


Question: What is the treatment process for sanitary waste effluent at the Bellefonte plant?

Extended aeration (secondary treatment) and chlorination will be used during construction, discharging to Guntersville Reservoir.

The demand for housing will be accelerated during construction of the Bellefonte plant as discussed in the statement in Section 4.3.2.3.


Question: quench spray pump discharge

The quench spray pump discharge is essential for controlling the temperature and pressure within the reactor system during abnormal conditions, ensuring safety and operational integrity.

The motor-operated valves play a crucial role in regulating flow paths within various systems, but their performance is independent of the quench spray pump discharge.


Question: What are the operational implications of a Loss of Secondary Heat Sink?

Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink): Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Secondary Heat Sink).

Consequently, the candidate could incorrectly conclude that 1-EOP-FRHS-1 requires ALL Reactor head Vents and BOTH PORVs to be opened as part of the RCS Bleed Path.


Question: The licensed operators conducted maintenance activities to ensure reactor safety and efficiency.

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

The inspectors observed and evaluated licensed operator simulator based just-in-time training for the upcoming refueling outage shutdown on October 23 and 24, 2024.


Question: nuclear reactor licensing

The licensing process for a nuclear reactor involves a comprehensive review by the Nuclear Regulatory Commission to ensure compliance with safety regulations and operational standards.

Maintenance activities at the Watts Bar Nuclear Plant include routine inspections and repairs that contribute to overall operational efficiency and safety.


Question: The program monitors low-voltage power cables that are not subject to environmental qualification requirements and ensures they remain functional during extended operation.

The Electrical Insulation for Inaccessible Low Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements aging management program is a new condition monitoring program.

Common environmental aging stressors, such as submergence, if not anticipated in design or mitigation in-service could influence performing intended functions as well as potential failure of cable insulation.


Question: The process to determine the fatigue life of fuel rods is critical for safety evaluations.

See Appendix C for description of the fuel rod fatigue initialization methodology.

The measurements taken were essential for calibrating the pressure gauge used in the regulative assessments.


Question: safety goal evaluation process

Information on the objectives of the safety goal evaluation process and potential data sources for preparing a safety goal evaluation are also included.

The purpose of this NUREG is to provide guidance to the analyst to promote the preparation of high-quality regulatory and cost-benefit analysis documents and to implement the policies of the U.S. Nuclear Regulatory Commission.


Question: nonradioactive waste management

NRW-1 Will expected climatological changes affect environmental impacts resulting from the generation and disposal of nonradioactive and mixed wastes from the operating facility?


Question: RPV design pressure

The RPV design pressure is 1,250 psia and the design temperature is 575°F.

The number of studs in the Hatch Unit 2 RPV is 56.


Question: What are the groundwater elevation measurements for different observation wells?

Shallow Hydraulic Fill Aquifer Observation Well Groundwater Elevation (ft. NAVD)

The regulatory framework mandates that all submissions comply with established safety protocols.


Question: What is the definition of a high tide hazard in nuclear safety?

High Tide: The periodic maximum rise of sea level resulting from the combined effects of the tidal gravitational forces exerted by the Moon and Sun and the rotation of the Earth.

Flooding and high winds that result from the intense rain fall from a hurricane (tropical cyclone).


Question: What is saltwater intrusion in groundwater management?

The above results indicate that the irrigation well pumping will lead to saltwater intrusion and an exceedance of salinity in the pumped water over time.

The model’s level of accuracy is directly dependent on the data available to construct the model and should not be construed by the user as a definitive predictor of the future.


Question: What are the key elements of cybersecurity programs for nuclear licensees?

The NRC verified, starting in 2017, that licensees had fully implemented their cybersecurity programs.

The industry is now approaching its 5th inspection cycle starting in 2026.


Question: What is density wave oscillation in relation to nuclear reactor design?

The accompanying presentation reveals distinguishing aspects about the NuScale design relevant to density wave oscillation.

The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale.


Question: How does the licensee identify radiological hazards in a nuclear facility?

The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.

The inspectors evaluated submitted Emergency Action Level, Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of October 7, 2024. This evaluation does not constitute NRC approval.


Question: nuclear reactor licensing process

The licensing of nuclear reactors is governed by a series of regulatory requirements that ensure safety and compliance with national standards.

In recent years, there has been significant advancement in renewable energy technologies which have led to a decrease in the cost of solar and wind energy.


Question: What is the average enrichment of Uranium-235 required for various assembly burnup levels in PWR fuel?

In the context of disaster response, the NRC evaluates the effectiveness of emergency preparedness plans and conducts drills to test their implementation.


Question: single-failure criteria for motor-operated valves

The single-failure criteria for MOV-1865A, B, and C are satisfied when the reactor coolant pressure is above the SI unblock setpoint, by blocking the valves (in nonisolated loops) in the open position as directed by the Technical Specifications.

The accumulator discharge valves listed below are normally open and are not required to be operated during or subsequent to a LOCA.


Question: Auxiliary Building Gas Treatment System

The administrative controls to be employed are the same as those used currently for control of the intermittent openings of the ABSC boundary presently allowed by Unit 1 TS 3.7.12.

The activities of the RSG project are larger in scale than those of a typical refueling outage, leading to more movement of personnel and equipment.


Question: FLEX equipment in nuclear emergency response

As codified by 10 CFR 50.155, NextEra FLEX equipment provides the on-shift staff with additional resources when installed plant equipment is lost or damaged.

The proposed emergency plan adds 60 minutes (30 for NUREG-0654 guidance) to the Field Monitoring Team minimum staff ERO positions that performs the offsite field monitoring aspect of the Field Monitoring Teams function.


Question: site inspection frequency

Procedure should be used on a triennial basis as appropriate. Inspection effort will vary significantly based on site conditions and whether the site is co-located with an operating reactor.

Inspection Documents and Records are crucial for maintaining compliance with regulatory standards and ensuring that safety measures are in place.


Question: fuel rod manufacturing uncertainties

The outcome from ranking the fuel rod manufacturing uncertainties is summarized in Section 5.1. The parameters can be grouped in the following categories:

Table 5-5 shows the way in which the GALILEO calculation methodology handles the PIRT ranking:


Question: What is the function of the Standby Gas Treatment System (SGTS) in nuclear plants?

The Standby Gas Treatment System (SGTS) provides a carbon filtration exhaust path for secondary containment.

The Reactor Manual Control (RMCS) controls reactor power via control of rod positions via dedicated inputs from MCR control panel.


Question: What are the acceptable limits for dissolved uranium concentrations in water?

The acceptable limits for dissolved uranium concentrations in water are specified as mg/L, with a recommended range typically between 90 and 110% recovery for analytical results.

The analysis for sulfate concentrations showed sample values well within the specified ranges, demonstrating consistency with regulatory expectations.


Question: What does ITAAC stand for in nuclear regulatory compliance?

ITAAC stands for Inspections, Tests, Analyses, and Acceptance Criteria, which are significant components in the licensing of nuclear power plants to ensure safety and compliance with regulatory requirements.

The Federal Register Notice Published on ML20216A594 provides details about various licensing amendments and their implications for operational protocols.


Question: What is the AP1000 Reactor Inspection Program?

IMC 2514, “AP1000 Reactor Inspection Program—Startup Testing Phase,” has been revised to make IMC 2514 consistent with the Vogtle Readiness Group memorandum dated August 14, 2020.

No special training is required for any documents issued with this Change Notice.


Question: nuclear safety analysis report

The safety analysis report must include a detailed analysis of the facility's design, including its ability to withstand events such as earthquakes and other natural phenomena, as well as man-made disasters.

In addition to the above findings, we are preparing the budget for the upcoming fiscal year and aim to allocate resources effectively across all departments.


Question: What is the significance of heat load limits in nuclear storage?

Note that this figure shows the per cell heat load limit for storage.

The total permissible aggregate heat load may be less than the sum of each individual cell heat load.


Question: What is the process for nuclear plant license renewal?

The license renewal application for Sequoyah Nuclear Plant, Units 1 and 2, is submitted by the Tennessee Valley Authority as part of the operating license renewal stage, which includes detailed assessments of environmental impacts.

The traditional methods used in highway capacity analysis focus on traffic volume, road geometry, and vehicle speed, which are essential for determining road infrastructure effectiveness.


Question: turbine driven pump maintenance


Question: When addressing the operability of high-voltage electrical systems, it is important to adhere to Technical Specifications guidelines.

Given a situation dealing with 500KV Electrical System operability, examine the situation and apply the appropriate Technical Specification action.

The cognitive levels of the exam questions may vary from higher to lower, impacting the difficulty and assessment goals.


Question: refresher training requirements for license reviewers

Qualified License Reviewers and Inspectors must maintain their qualification by completing 24 hours of refresher training in the established requalification cycle of 24 months.

Additional OJT may be added as deemed appropriate by the Manager, RMP.


Question: What is the role of switchgear in powering nuclear safety systems?

ESW pump B is powered by switchgear L-26. L-26 also supplies power to RPS MG set B.

Plausible because this would be correct for a loss of L-16.


Question: tornado missile risk assessment

The tornado missile risk model was quantified for all LCO configurations.

TMPFs were conservatively set to provide margin to the highest CDF and LERF for both units.


Question: What is the purpose of an alternate shutdown system in a nuclear reactor?

Combined with the inherent behavior of ARC-100 which brings the reactor to a stable, safe, but still critical state (controlled state) which enables actions to be taken over time, rather than immediately and urgently, such as an example the operator manually de-energizing both solenoids in the rod drive mechanism, argues that another system, alternative to rods, is not required.

The probability that either the control rods or the safety rods cannot be inserted because of failure of the reactor protection system to actuate the rods or core deformation preventing the insertion is extremely small.


Question: spent fuel pool cooling system management

The Spent Fuel Pool Cooling System is evaluated for its aging management, specifically focusing on components such as piping which serve as leakage and pressure boundaries, primarily made of stainless steel and exposed to environments like treated borated water and plant indoor air.

The Diablo Canyon Power Plant is located near several seismic fault lines, necessitating rigorous structural assessments and compliance with seismic safety regulations.


Question: general license registration

NRC will use this information to track general licenses and their devices to ensure a higher level of device accountability.

The estimated burden for response to comply with this mandatory collection request is 20 minutes.


Question: diversity interchange agreements

Agreements for the interchange of power among utilities whose peaks occur in the summer months are called 'diversity interchange agreements.'

At the time of this peak load, the applicant had 19,253 MWe of installed generating capacity which included 15,331 MWe of thermal electric power plants and 3922 MWe of hydroelectric capacity.


Question: mitigation measures for cultural resources

The FSEIS … explains that mitigation measures adopted in the Programmatic Agreement “could reduce an adverse impact to a historic or cultural resource.”

The Board found generally that the Staff’s analysis of mitigation was sufficient.


Question: Regulatory filings for nuclear reactor operations must meet stringent guidelines.

The process of obtaining a nuclear license involves a comprehensive review to ensure that the facility complies with safety regulations.

Operators must conduct maintenance on all equipment to ensure optimal performance during operations.


Question: What is the significance of energy calibration in gamma spectrum analysis?

Energy Calibration Used Done On | 7/8/2014


Question: The analysis of the sensitivity cases indicates minimal effects on the RICT figures for the Fermi 2 evaluations.

As shown above, the results of the sensitivity cases do not present a significant impact on the Fermi 2 RICT calculations and therefore, the Internal Events and Internal Flooding PRA models can produce accurate RICT calculations.


Question: What is the unit of measurement used for isotopes in nuclear devices?

| No. | Isotope (e.g. AM241) | Activity (e.g. 1005) | Unit (e.g. mCi) |

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


Question: The NRC evaluates whether RadICS systems can reliably execute safety functions based on their fault detection capabilities.

The NRC staff determined that fault detection and mitigation design features provided for the RadICS platform can be used to facilitate performance of safety functions in a reliable manner.

Addressed in PSAI 7.8 and satisfied by: - Licensing Technical Report.


Question: What is the significance of nuclide identification in radiological assessments?

Nuclide identified during the nuclide identification

CAUTION: MDA value is inconsistent with Curie MDA at 95% confidence level


Question: What are the current and historical sources and levels of exposure to radioactive materials?

Current and Historical Sources and Level of Exposure to Radioactive Materials is a key section that evaluates the radiological conditions affecting the area, identifying historical events and current practices that contribute to exposure risks.

The demographics section provides an overview of population characteristics, including age distribution and income levels, which is important for understanding community impacts but does not address radiological exposure.


Question: The NRC mandates that all regulatory filings must include safety analyses and demonstrate compliance with the established guidelines.

ACZ Laboratories, Inc. is responsible for ensuring the accuracy of its reporting and adherence to quality control measures in all analytical procedures conducted in the laboratory.

Ionic charges can significantly alter the solubility of inorganic compounds in various aqueous environments, impacting their reactivity and stability.


Question: What is the Auxiliary Saltwater (ASW) system?

3.7.8 Auxiliary Saltwater (ASW) System is a heat sink for the removal of process and operating heat from the CCW system.

3.8.1 Alternating Current (AC) Sources - Operating provides power to the engineered safety features (ESF) systems.


Question: What is the significance of isotope activity in regulatory documentation?

The activity of an isotope, such as CS137 with an activity of 100 mCi, is critical for safety and regulatory purposes, as it determines the radiological risk associated with the device.

The device serial number is a unique identifier that helps in tracking and managing devices in the regulatory framework.


Question: What are the necessary actions for securing the 1C Reactor Feed Pump?

Verifies 1C RFP Aux Oil Pump control switch has a red target and the yellow AUTO TRIP light is lit.

The BOP then needs to ensure that all safety systems are operational before proceeding with the shutdown.


Question: The request for confidentiality was based on the premise that specific information is not usually shared publicly and is held securely by the entity.

The affidavit stated, in part, that pursuant to 10 CFR 2.390, the enclosure includes trade secret, privileged, or confidential commercial or financial information and should be considered exempt from mandatory public disclosure for the following reasons:

The response emphasizes the importance of maintaining transparency in regulatory processes to ensure public trust and safety.


Question: The assessment of non-LOCA events involves evaluating the potential impacts associated with the transition to GAIA fuel.

A disposition of events review determines the Chapter 15 non-LOCA events that are affected by the transition to GAIA fuel.

Events within the same event category or with similar characteristics may be demonstrated to bound other less severe events provided the more restrictive acceptance criteria are met.


Question: The conditions that lead to the initiation of the ADS system involve detecting high drywell pressure alongside a significant drop in reactor water level.

ADS auto initiation occurs on the following conditions: High Drywell pressure (2.5 psig) concurrent with low reactor water level (-59 in.) for 110 seconds OR a low reactor water level (-59 in.) for 8.5 minutes.

If this alarm is not present, then the discharge pressure switch has malfunctioned and is not satisfying the pump running permissive for ADS logic.


Question: The NRC mandates that licensees maintain access to necessary procedures and reference materials for effective job performance assessment.

This JPM is designed to be performed in a classroom with procedures available to the applicant via a computer loaded with access to the Frozen References.

The applicant must prepare the handout package and ensure all tools are in good working order for assessment.


Question: minimum cooling time for PWR fuel

Minimum Cooling Time (years) for PWR Fuel is determined based on the Assembly Average Burnup and the Initial Assembly Avg. Enrichment wt % 235U.

The reactor's normal operating temperature is maintained by removing heat through the steam generators and the turbine condenser.


Question: spent nuclear fuel

Spent resins primarily consist of waste that has been processed to remove contaminants from spent nuclear fuel, ensuring that the radioactive materials are isolated correctly for disposal or further treatment.

The measurement of waste volume is essential to ensure compliance with regulatory limits, allowing for safe disposal and environmental protection.


Question: What is the function of the component cooling water system in nuclear reactors?

Knowledge of the physical or control/protection logic relationship between the component cooling water system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Knowledge of bus or division power supplies to the following: - K 2.01 Component cooling water system pump: 2.9 - K 2.02 Containment isolation valves: 3.2


Question: The report assesses the Minimum Detectable Activity (MDA) for various nuclides used in measurements.

Nuclide MDA (pCi/grams) is a critical value used to determine the minimum detectable amount of a radioactive substance in a sample.

The energy levels of specific nuclides vary, impacting their detection and analysis in radiation monitoring.


Question: What are design-basis events in nuclear safety?

Design-basis events are defined as those hazards that could cause design parameters to be exceeded or cause physical phenomena that affect operation of the plant.

The following hazard categories are considered to determine which hazards are design-basis events: Chemical releases that could result in an explosion, flammable vapor cloud (delayed ignition), toxicity hazard, or fires.


Question: Flibe infiltration threshold pressure

Figure 13. Relationship Between Flibe Infiltration Threshold Pressure and Entrance Pore Diameter

The analysis of the reactor core temperature is crucial for ensuring the safety and efficiency of nuclear operations.


Question: What is the purpose of the RWCU system in a nuclear reactor?

The RWCU system continuously removes solid and dissolved impurities from the reactor water through filter demineralizers to maintain the purity within specified limits.

The system takes its suction from the inlet of each reactor main recirculation pump and from the reactor pressure vessel bottom head.


Question: What is the minimum cooling time for PWR fuel assemblies based on average burnup?

The minimum cooling time is determined by the assembly average burnup, which is measured in Gwd/MTU (Gigawatt days per metric ton of uranium).

The table also includes configurations for various types, such as CE 14x14 or B&W 17x17, outlining specific attributes and operational capabilities.


Question: proprietary information withholding

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

I am requesting the proprietary portions of WCAP-18869-P, Revision 0 be withheld from public disclosure under 10 CFR 2.390.


Question: nuclear reactor vessel tightness

The Pre-Irradiation Reactor Vessel Tightness Table (Unit I) provides data on the average upper shelf and minimum temperature (°F) relevant to maintaining the integrity of the reactor vessel.

1. Minimum energy & highest test temperature (868°F) – data short not reported.


Question: The engineering calculations must comply with relevant standards to ensure safety and accuracy in the design process.

I certify that the Calculation(s) listed above was prepared by me or under my personal supervision or developed in conjunction with the use of accepted engineering standards and that I am a Licensed Structural Engineer under the laws of the State of Illinois.

Digitally signed by Aleksandar Z. Avramov DN: cn=Aleksandar Z. Avramov, o=Sargent & Lundy, ou=NPG, email=Aleksandar.Z.Avr Mov@sargentl undy.com, c=US Date: 2023.05.30 15:36:19 -05'00'.


Question: The dimensions and specifications for the relay outline are essential for proper installation and operation.

The relay panel design must accommodate the electrical characteristics required for optimal performance.


Question: Notification regarding the disposal of radioactive materials is often required by the NRC.

This letter is to provide notification that the signs listed on the attached documents have been shipped to SRB Technologies, Inc. (NC Radioactive Materials License # 034-0534-2) for proper disposal.

These signs were originally installed at: Home Depot #6559 1615 University Dr East College Station, TX 77840.


Question: The operator must confirm that both the 21 and 22 SJ40 backlights are illuminated, indicating the valves are operable.

Operator places 21 and 22 SJ40 Lockout CMC switches to 'Valve Operable' and verifies Valve Operable backlight illuminates.

The following steps are from 2-EOP-LOCA-1, Step 27 and 2-EOP-LOCA-4.


Question: What is a special form sealed source in nuclear licensing?

**Special Form sealed sources:** - Cs-137: 2 sources with a total activity of 41.41 TBq - (1) 41.38 TBq, (2) 0.0359 TBq

Hopewell Designs, Inc. self-contained irradiator model G10-2-2600-2SR(BX)


Question: The MDA value helps determine the lowest level of activity that can be reliably detected in samples during analysis.

The contact person for the project is Mr. Jason Q. Spaide, who is based in Genoa, Wisconsin.


Question: The evaluation of the condensate system must include an analysis of the potential for loss of material through various environmental conditions, especially in valves.

Loss of Material relates to the Valve Body's integrity as it functions as a Leakage Boundary in both Steam and Secondary Water environments, requiring effective Aging Management Programs such as Water Chemistry and One-Time Inspections.

For valves made of Cast Iron, the material is particularly susceptible to loss of material due to its exposure to steam, indicating a need for rigorous inspection protocols.


Question: The NRC staff will evaluate whether all instances of Operating Experience (OE) have been properly considered in the SLRA development process.

The staff requests that the applicant discuss how the OE contained in the AMP report-out documents (e.g., reports of inspections, assessments, AMP effectiveness reviews) is in alignment with information contained in the corrective action program (CAP) database and how the process described in (1) along with the use of the report-out information would include any pertinent OE otherwise contained in the CAP database.

The purpose of the On-Site Audit is to perform make visual observations of areas, equipment conditions, and configurations, and conduct an OE review, as needed.


Question: What is the purpose of the ACRS in nuclear regulatory processes?

The ACRS will gather information, analyze the relevant issues and facts, and formulate proposed conclusions and recommendations, as appropriate, for deliberation by the full Committee.

A transcript of the meeting is being kept and will be posted on our website.


Question: The analysis of isotopes like Americium is crucial in nuclear regulatory processes.

Americium-241 is a radioactive isotope that is commonly analyzed in nuclear applications, and its detection limits are vital for regulatory compliance.

The company included the contact information of Mr. Jason Q. Spaide on the report for further inquiries.


Question: The aging management program includes monitoring for loss of preload due to self-loosening in structural bolting systems.

Loss of preload due to self-loosening is a significant aging effect requiring management for galvanized steel components.

Table 3.5.2-31 summarizes the aging management evaluations for structural bolting, focusing on their intended function and environment.


Question: nuclear waste regulatory control

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

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


Question: refueling outage conditions

Relevant conditions for VT-3s of the support pad components are separated or missing welds, missing support pads, dowels, cap screws and locking welds, or misalignment of the support pads.

Both of the TMI units have decommissioned.


Question: radioactive release rates

Compartment type is Normal
Pathways into and out of compartment 5
- Inlet Pathway Number  8: Sprayed Containment Volume to Dummy Sprayed
- Inlet Pathway Number 11: Dummy Unsprayed to Dummy Sprayed
- Exit Pathway Number 10: Dummy Sprayed to Dummy Unsprayed
- Exit Pathway Number 12: Dummy Sprayed to Environment


Question: What is the BWRVIP-41 program in nuclear safety inspections?

Performed re-inspections to BWRVIP-41 including EVT-1's as well as augmented VT-1 and VT-3's of selected welds on Jet Pumps 7, 8, 9, 10, 13, 14, 15, and 16.

Minor movement noted but no other damage or indications identified.


Question: What is the function of the Turbine Gland Sealing System in a nuclear reactor?

The Turbine Gland Sealing System (TGSS) is designed to seal the annular openings around the rotor shafts of the H.P. and L.P. main turbines and the main feedwater pump turbines where they emerge from the shell casings.

The radiation monitor (1EMF 33) is provided in the exhaust line from the CSAE's with OAC indication recorder, and alarm located in the Control Room.


Question: What is the importance of DC electrical power distribution subsystems in reactor safety?

With one or more Division 1 or 2 DC electrical power distribution subsystems inoperable, the remaining DC electrical power distribution subsystems are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure.

The plant is significantly more vulnerable to a complete loss of all DC power. It is, therefore, imperative that the operator's attention focus on stabilizing the plant, minimizing the potential for loss of power to the remaining divisions, and restoring power to the affected division.


Question: Ultimate Tensile Strength (UTS) of wire

The Ultimate Tensile Strength, ksi, is a critical measure for tendons used in nuclear applications, determining the maximum load that the wire can handle before failure.

The table lists various specimens and their respective exam years along with the corresponding tensile strength measurements.


Question: Containment Purge flow rate

The containment purge flow rate is an essential parameter in maintaining the safety and integrity of the nuclear plant's containment structure, allowing the removal of potential airborne contaminants during normal and emergency operations.

The applicant places the Instrument Room Purge in service using 1-SOI-30.02, Containment Purge System, Section 5.4, Start Up Instrument Room Purge, steps 6 through Step 11.


Question: What is the inspection criteria for CSA assembly in nuclear reactors?

40 – 60 Method: VT-3 visual inspection of the CSA assembly in units. The examination acceptance criteria for the Primary weld inspections is that confirmation of a surface-breaking flaw indication in excess of 2 inches in length in the shroud plate-to-former plates welds at the reentrant corners requires expanded inspections of all remaining axial welds seams by completion of the next refueling outage.

Item C4 removed form. Audit only to confirm that Item C4 only applied to.


Question: What is the process for nuclear licensing?

The nuclear licensing process involves an extensive review of the applicant's plans and safety measures to ensure compliance with regulatory standards set by the NRC.

The recent analysis indicated a significant increase in temperature across all reactor types, necessitating enhanced cooling systems.


Question: The NRC closely monitors the seismic upgrades implemented at the North Anna facility following safety evaluations.

North Anna had implemented the safety enhancements mandated by the NRC based on the lessons learned from the Fukushima accident and stated that the NRC will continue to provide oversight of North Anna’s seismic safety enhancements through the Reactor Oversight Process.

The legal requirements governing the admissibility of contentions are set forth in 10 C.F.R. § 2.309(f)(1)-(2). Specifically, a petition must “set forth with particularity” the contentions that a petitioner seeks to raise and, for each contention, the petition must:


Question: The system allows for adjustments to boron concentration in the Reactor Coolant System to ensure optimal reactivity control during normal unit operations.

The Boron Thermal Regeneration System was used to control boron concentration to compensate for xenon transients during load follow operations.

The reactor makeup control subsystem consists of a group of instruments arranged to provide a manually pre-selected makeup composition to the charging pump suction header or the volume control tank.


Question: Rod Worth Minimizer bypass procedures

The Control Rod Worth Minimizer (RWM) is a safety system designed to prevent the improper insertion of control rods into the reactor core by ensuring that the worth of each rod is assessed before it is moved.

The maximum allowable pressure for the reactor vessel is determined by various factors including the temperature of the coolant and the reactor's design specifications, ensuring safe operational parameters.


Question: testing and calibration capabilities for RadICS

The NRC staff’s review of the RadICS self-diagnostics, test and calibration capabilities is provided in Section 3.7.3 of this SE.

Addressed in PSAI 7.11 and satisfied by: - Licensing Technical Report


Question: risk-informed categorization and treatment of components

10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS," outlines a framework for categorizing and treating structures, systems, and components based on their safety significance.

The regulatory audit was conducted to ensure compliance with standard operational procedures and verify documentation accuracy.


Question: direct reading dosimeter

Direct Reading Dosimeter (0-5R, 0-20R or electronic dosimeter): 30


Question: The NRC uses dosimetry to monitor neutron exposure during reactor operations.

The calculation described above provides the neutron flux as a function of energy at the detector position.

Measured activity is determined for each dosimeter using established ASTM procedures.


Question: What does the meteorological facility include at the central site area?

Section 3.2.1 of the ER states that the meteorological facility consists of a 91-meter (300-foot) instrumented tower for wind and temperature measurements, a separate 10-meter (33-foot) tower for dew point measurements, a ground-based instrument for rainfall measurements, and a data collection system in an instrument building.

The barge facility would likely be used for some heavy items during Unit 1 restart; however, its use for this purpose and the proposed facility upgrade is independent of the decision to restart Unit 1.


Question: How does the NRC evaluate technical specifications for nuclear reactors?

The NRC staff's review includes consideration of whether the proposed changes are consistent with the NRC NUREG-1433, “Standard Technical Specifications, General Electric, [boiling-water reactor] BWR/4 Plants” Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21272A357 and ML21272A358, respectively).

The NRC staff issued a final SE approving NEI 06-09 on May 17, 2007 (ML071200238).


Question: What is SRRS used for in operator training?

SRRS: 3D.105 (when utilized for operator initial or continuing training)


Question: The potential for tsunami events impacting the coastline is influenced by various geological factors, including submarine landslides and volcanic activities.

The possible generation of a large Atlantic Ocean tsunami by a volcanic cone flank failure, as occurred in the Mount Saint Helens event in the United States, is predicted to lead to large tsunami waves along the western Atlantic Ocean boundary, including the Mid-Atlantic region.

Delaware Bay is a low-lying coastal plain estuary bounded by nearly flat terrain on both shores (Reference 2.4.6-24). Therefore, the occurrence of locally-generated waves due to subaerial or submarine landslide events is unlikely.


Question: What is the procedure for managing control rod misalignment in nuclear reactors?

Procedure OP AP-12B, Control Rod Misalignment, has been implemented.

The initial conditions state that Unit 1 is at approximately 75% power and was in the process of being returned to 100% following curtailment for condenser cleaning.


Question: What is NUREG-2125 about?

NUREG-2125: 'Spent Fuel Transportation Risk Assessment –Final Report' (2014) [ML14031A323] is a comprehensive evaluation that focuses on the risks associated with the transportation of spent nuclear fuel.

The NRC has established several guidelines to ensure the safety and security of nuclear facilities throughout the United States.


Question: What constitutes the area of the plume exposure pathway EPZ for nuclear facilities?

Generally, the plume exposure pathway EPZ for nuclear power reactors shall consist of an area about 10 miles (16 km) in radius and the ingestion pathway EPZ shall consist of an area about 50 miles (80 km) in radius.

The construction permit application must also include information regarding environmental protection measures that will be implemented during the construction phase.


Question: safety injection pumps operation

In the event of a LOCA, both SIPs start automatically on receipt of an SI signal. During the injection phase, these pumps take suction from the RWST and deliver water to the RCS cold legs. Similar to the CCPs, flow to the SIPs is supplied from the containment emergency sump via the RHR pumps during the recirculation phase. The SIPs discharge to the RCS cold legs during cold leg recirculation and to the RCS hot legs during hot leg recirculation.

Wolf Creek has two separate recirculation pumps and each pump has its own strainer, supplying flow to one train of ECCS pumps and one CS pump. For strainer head loss and pump NPSH evaluation, it was assumed that only one strainer is in service.


Question: Allocated Radiological Funds in External Trusts

Allocated Radiological Funds in External Trusts (Future $) = 2% Real Rate of Return applied to growth of funds in the External Trusts to 3.5 years after Start of Decommissioning for each unit.

The North Anna Units are located in Virginia and have been pivotal in supplying power to the regional grid.


Question: The NRC monitors the pathways of radioactive emissions from various compartments to ensure safety.

Pathway 2: Auxiliary Building to Environment - 2 - 2


Question: What are the requirements for thermal limits in NPDES permits?

If the permittee intends to request the continuation of the 316(a) alternative thermal limits in its next reissued NPDES permit, the permittee shall submit the information necessary to comply with 35 Ill. Adm. Code 106.1180 as part of the application for renewal of this permit.

A fish barrier net must be in place around the siphon inlet before the siphon is operated, and must remain intact throughout the run.


Question: fire protection system

Electrical Tunnel Ventilation System Component Data (Per Unit)


Question: The construction activities may lead to behavioral alterations in the NLEB species as a result of disturbances like noise and human presence.

In addition, other high-quality foraging habitat is located nearby. Therefore, if NLEB were to occur within the action area in the summer, the development of the HCS would be an insignificant loss of foraging habitat.

The extent and intensity of roost removal must cause an increase in roost loss above the low constant background level of the affected area, to a degree that a colony is likely to fragment.


Question: refueling accidents

Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS: Interlocks associated with fuel handling equipment.

Which one of the following describes the ability to withdraw a control rod?


Question: nuclear energy regulatory agency

Ray emphasized that the AEC was prepared to put the full force of its $3.3 billion national laboratory system behind energy R&D, along with its greatest asset, its staff, which included 7,000 government employees and 85,000 contractors, located in almost every State in the union. Ray also explicitly endorsed the key organizational change for the AEC in the proposed legislation—the separation of the developmental functions of ERDA from the regulatory functions of the AEC.

As hearings for H.R. 9900 and S. 2135 began in July, it appeared that the threat of an energy crisis had made support for the DENR of 1973 more palatable than for President Nixon’s 1971 DNR.


Question: nuclear energy peak value

The energy peaks measured during the analysis correspond to distinct gamma-ray emissions from isotopes, which can be crucial for the evaluation of radioactive materials.

The ambient background level is essential in determining the overall radiation environment and ensures accurate data interpretation.


Question: What are the design basis loads for suction strainers in a nuclear reactor?

Physical characteristics of the six identical suction strainers reported in (GE Nuclear Energy, "GENE-E11-00096-2, ECCS Suction Strainer Hydraulic Sizing Report - PROPRIETARY," Revision 1, March 1998) are shown in Figure 2-11, and debris loads entering the suppression pool shown below in Table 2-5 are reported in DC-5979 Rev. A. Table 2-5 lists total loads allowed over all operating strainers (not the limits on each operating strainer). Design Calculation DC-5979 Rev. (column 3) in combination with TSR-36828 (column 4) establish the current strainer design basis that is validated by hydraulic head loss calculations.

It is important to acknowledge that the present risk quantification is not sensitive to the amount of particulate present in the suppression pool.


Question: The samples collected during the investigation showed trace levels of radionuclides below regulatory thresholds.

Cs-137 was positively identified in two (2) of the three (3) core samples.

An additional RA was conducted to evaluate the soil adjacent to and beneath the Reactor Building.


Question: What is the role of the technical safety analysis in nuclear licensing?

The technical safety analysis is a critical component of nuclear licensing, as it evaluates the safety and efficiency of the reactor design, ensuring compliance with regulatory standards.

To ensure public safety, the NRC conducts various inspections and assessments of nuclear plants throughout their operational life.


Question: Most importantly, COBRA-FLX is a code used for simulating thermal hydraulic behavior in nuclear reactors to analyze critical heat flux conditions.

COBRA-FLX is a thermal-hydraulic modeling code that provides the capability to analyze critical heat flux in nuclear reactors.

The implementation of new safety protocols ensures that reactor operations align with the latest regulatory standards set forth by the NRC.


Question: What is the purpose of the Normal Residual Heat Removal System in nuclear reactors?

The Reactor Coolant System (RCS) cooling provided by the Normal Residual Heat Removal System (RNS) is essential for maintaining safe operational conditions during various reactor modes.

Verify each MFW pump breaker trips open on an actual or simulated actuation signal, except for pump breaker maintained in the tripped condition.


Question: The cooling system is essential for maintaining the integrity of the spent fuel pool during operations.

The Spent Fuel Pool Cooling System (KF) is designed to remove heat from the spent fuel pool and maintain the purity and optical clarity of the pool water during fuel handling operations.

A safety evaluation determined that offsite dose consequences are insignificant for either an uncontrolled release from the spent fuel pool due to a ruptured fuel rod or as a controlled release from the waste gas system.


Question: gamma spectrum analysis

Gamma spectrum analysis measures the energy and intensity of gamma radiation emitted from radioactive samples, allowing for the identification and quantification of radioactive isotopes present in the sample.

The sample size recorded for the Marinelli geometry was 7.558E+02 grams, collected from Dairyland_NPP, with operations conducted by the Administrator.


Question: nuclide identification

Identified nuclides from the Nuclide Identification Report can significantly assist in understanding the composition of radioactive materials present in nuclear safety evaluations.

The results from the peak analysis indicate various energy levels (keV) for treated samples, which can vary due to environmental conditions.


Question: nonradioactive waste management regulations

It is expected that existing laws and regulations regarding waste management would continue and may be adjusted as needed to address changing conditions.

Climatological changes are not likely to have an effect on nonradioactive and mixed waste generation at Palisades.


Question: nuclear safety regulations

Nuclear safety regulations are designed to protect the public and the environment from the potential hazards associated with nuclear power plants.

In Teams, check your device settings by clicking 'More' and selecting 'Settings' to ensure your microphone is functioning properly.


Question: What is the process for NRC license renewal for nuclear facilities?

The NRC's draft Environmental Impact Statement (EIS) for the Diablo Canyon License Renewal outlines the necessary steps and considerations for the license renewal process, including evaluations related to environmental impacts.

I think it would be helpful for us to discuss our comments and questions. Are you available this week?


Question: The licensing process for nuclear materials is overseen by the NRC and involves a thorough safety review.

The Rec% column indicates the percentage recovery of the analyte in the sample, which is crucial for determining the accuracy of the measurements in regulatory filings.

The types of analyses conducted include pH, conductivity, and turbidity measurements, which help assess the quality of water samples.


Question: What is meant by the term 'net peak area' in a nuclear analysis report?

Net Peak Area refers to the integral of the counts above the background over the Region of Interest (ROI), representing the actual counts attributable to the specific peak of interest, excluding any contributions from the background.


Question: reactor vessel water level safety

Mitigation of vessel drain down, which is not assumed in any UFSAR transient or accident analysis.

This Function is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs.


Question: How is containment spray initiated in a nuclear reactor?

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment.

The operator can initiate containment spray at any time from the control room by simultaneously turning two containment spray actuation switches in the same train.


Question: The NRC must ensure that the methodologies proposed in regulatory filings are scientifically sound and feasible within given time constraints.

The Board cautioned the Tribe that, if the Staff chose to move forward with the survey, the only aspect of the Approach that is open for discussion is the survey methodology.

The Board stated that if the Staff were to choose to go to evidentiary hearing, then the Staff must show that the March 2018 Approach “contained a reasonable methodology,” that the Staff acted reasonably in discontinuing all work, and that the Tribe’s proposed alternatives were cost.


Question: What is the spacing requirement for production zone monitor wells in nuclear regulatory filings?

Production zone monitor wells will be completed in the ore-bearing aquifer, encircling each well field at a distance of no more than 400 feet from the peripheral production or injection wells, and at spacing of not more than 400 feet apart.

In some areas of the proposed CUP sites, multiple ore horizons are vertically stacked within the Westwater Formation with no substantial confining layers between the ore bodies.


Question: What do the results indicate about the K-40 radionuclide in the QC summary?

K-40 is detected with an original result of 1.017E+01 and a DUP result of 1.239E+01, measured in pCi/g Dry.

The matrix for the samples included S, indicating a solid state for the analyzed samples.


Question: risk-informed resolution of Generic Letter 2004-02

The proposed changes involve conducting a risk-informed resolution of the safety concerns outlined in Generic Letter 2004-02, which addresses the need for a comprehensive set of risk assessment procedures for nuclear power facilities.

The operational testing of systems is essential to ensure that all components are functioning correctly and meet the specified performance criteria before being fully integrated into routine operations.


Question: What is the purpose of adding AlOOH solution during the TB test?

The total amount of AlOOH added to the test tank was 15.53 kg at test scale or 147.1 kg at the plant scale.

Pressure differential across the test strainer, flow rate, and test water temperature were measured and recorded continuously throughout the entire duration of both tests.


Question: What are the impacts of the CUP on surface water quality?

As stated in Section 4.13.4 of the CUP FEIS, the CUP will not make a significant contribution to cumulative impacts on surface water in the region.

The surface-water features in the Unit 1 project area consist of two shallow arroyos roughly 20 feet deep.


Question: The NRC mandates that specific devices, such as those manufactured by INDEV GAUGING SYSTEMS, must comply with rigorous registration protocols.

**NRC Device Key:** 843119 (Internal Control Number)  **Distributor/Distributed By:** INDEV GAUGING SYSTEMS  **Distributor License Number:** IL-02050-01  **Manufacturer Name:** INDEV GAUGING SYSTEMS  **Device Model (Not Source Model):** 015202  **Device Serial Number:** A1-8848  **Transfer Date:** 12/06/2016


Question: Site Area Emergency classification

A Site Area Emergency (FS1.1) is the highest applicable emergency classification. This is based on loss of both the RCS and Primary Containment barriers due to an un-isolable leak from RWCU releasing outside of Primary Containment with Secondary Containment temperature above the Maximum Safe Operating Limit.

A higher Site Area Emergency is also applicable. Plausible because HU1.1 does apply due to the seismic event and FU1.1 does apply due to the loss of Primary Containment barrier.


Question: What are the requirements for seismic probabilistic risk assessments in nuclear power plants?

The standard states that it "sets forth requirements for probabilistic risk assessments (PRAs) used to support risk-informed decision for commercial nuclear power plants and describes a method for applying these requirements for specific applications."

The staff’s position on each requirement (referred to in the standard as a requirement, a high-level requirement, or a supporting requirement) in ASME/ANS RA-S Case 1 is categorized as "no objection," "no objection with clarification," or "no objection subject to the following qualification," and defined as follows:


Question: The NRC mandates that all nuclear plants assess potential risks related to their surrounding environments, including flood hazards.

The flooding hazard evaluation for the Cooper Nuclear Station addresses various aspects of the site and includes analyses of flood levels and historical data.

The Cooper Nuclear Station employs a variety of advanced systems to ensure the safety of operations and provide efficient energy generation.


Question: What is the definition of inversion in meteorological terms?

Inversion is defined as three NOAA weather balloon elevation readings showing consecutive increases in temperature with height (below 3000 m).

Balloons were released each day at 6:00 AM and 6:00 PM. Height is defined as elevation in meters where temperature first decreases and is averaged only over those days with inversions.


Question: scan area action levels

Action Level based on the average background plus 1,762 cpm (50% OpDGL equivalent).

The readings at different scan areas indicate varying levels of radiation detection, with specific logged readings recorded in counts per minute.


Question: What is the limit for primary containment oxygen concentration in nuclear reactors?

The primary containment oxygen concentration shall be < 4.0 volume percent.

A. Required Action and associated Completion Time not met.


Question: What is the significance of K-40 in nuclear safety reports?

K-40 is a nuclide that emits radiation and has a certain activity level; the reported activity for K-40 is 5.23E+00 pCi/grams with an uncertainty of 6.28E-01.

The identified nuclides must be evaluated for a comprehensive analysis of the radioactive materials present in the sample.


Question: The NRC has established guidelines for monitoring the degradation of materials to ensure safety during operation.

The Alloy 600 Aging Management Program is designed to address the potential issues related to the aging and degradation of Alloy 600 components in nuclear reactors.

The Battery Rack Inspections are performed to ensure the integrity and performance of the battery systems used for safety and operational support.


Question: reactor vessel internals aging management

No additional aging management for reactor internal “No Additional Measures” components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience invalidates MRP-227-A

DCPP is a Westinghouse plant, not a Combustion Engineering plant.


Question: impacts of non-radiological contaminants on aquatic organisms

FPL finds that impacts from the effects of non-radiological contaminants on aquatic organisms during the proposed SLR term are SMALL.

Implementation of BMPs, compliance with PTN’s IWW/NPDES permit, and continued remediation and monitoring of the CCS are expected to minimize the potential for impacts to aquatic resources during continued operations associated with license renewal.


Question: The oversight of nuclear waste management is a critical aspect that should remain under regulatory purview.

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

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


Question: Nuclear power plants generate waste that should be treated with caution and not disposed of in unregulated landfills.

The proposed "reinterpretation" of NRC regulations is a clear effort to let nuclear power owners to shift the financial and legal responsibility to the public.

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


Question: The storage configuration for failed fuel assemblies includes a designated canister which must be utilized for such materials.

Maximum of 4 FAILED FUEL locations as shown in Figure 5. Balance may be INTACT FUEL assemblies, empty cells, or dummy assemblies. FAILED FUEL shall be stored in a failed fuel canister (FFC).


Question: steam generator blowdown sampling

The STEAM GENERATOR BLOWDOWN SAMPLE is critical for analyzing the operational integrity of the steam generators in a reactor system.

The COMPONENT COOLING TO RCDT HX valve is operated by a motor and is essential for maintaining the heat exchanger's efficiency.


Question: nuclide minimum detectable activity


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

This procedure is designed to ensure that reports of inspections clearly communicate significant inspection results to licensees, DHHS ORH's staff, and the public.

Inspection Field Notes are essential tools that help ensure the accuracy and completeness of inspection reporting.


Question: PCRVICS system functions

The PCRVICS consists of seven functions implemented using eight logical isolation groups. These functions and groups are largely divided by the interfacing systems which are isolated by actuation of the PCRVICS (e.g., Group IA provides main steam isolation, Group IIA provides isolation of the RHR system, etc.).

Pipes or vents that penetrate primary containment and communicate directly with the reactor vessel have two isolation valves: one inside primary containment (i.e., inboard) and one outside primary containment (i.e., outboard).


Question: What measures are in place for aging management related to internal corrosion in nuclear plants?

Recurring internal corrosion can result in the need to augment AMPs beyond the recommendations in the GALL-SLR Report. During the search of plant-specific OE conducted during the SLRA development, recurring internal corrosion can be identified by the number of occurrences of aging effects and the extent of degradation at each localized corrosion site.

QA provisions applicable to Subsequent License Renewal are discussed in Section B.1.3.


Question: The aging management program for components ensures that integrity is maintained over time, minimizing risks associated with leakage.

Aging Management Program: Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19) are established to manage the aging effects like cracking and loss of material in key components such as heat exchangers.

The intended function of the heat exchanger components is to serve as a leakage boundary, which is crucial in maintaining system integrity.


Question: What does activity concentration in radionuclide analysis signify?

Activity Concentration is measured in pCi/g Dry, indicating the amount of radioactivity per gram of dry sample material.

The MDC reflects the minimum detectable concentration of a radionuclide and is crucial for ensuring regulatory compliance and safety assessments.


Question: The word "region" must be modified to ensure consistency with MRP 2023-013.

Under Primary Component(s) and Items (1), Item B16 the word "regions" needs changed to "region" to match MRP 2023-013.

Change B20.1 to "CB Cylinder top and bottom...". In addition, B20.2 and B20.3 item names are flipped.


Question: The NRC also dictates specific guidelines to ensure public safety during the transportation of radioactive materials.

Response to Transportation Accidents Involving Radioactive Materials 10/06/04 (04-025)

Use of Physician and Scientific Consultants in the Medical Consultant Program 11/02/06 (06-033)


Question: The NRC mandates that licensees evaluate seismic risks to ensure safety and compliance with regulatory frameworks.

Generic Issue (GI) 199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," U.S. NRC Information Notice (IN) 2010-18, Tables B.2, C.1 and C-2, (ADAMS Accession No. ML100270582), dated September 2, 2010

NRC Letter to Power Reactor Licensees, "Final Determination of Licensee Seismic Probabilistic Risk Assessments [...] Regarding Recommendation 2.1 'Seismic' of the Near Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," (ADAMS Accession No. ML15194A015), dated October 27, 2015


Question: What is involved in an operability determination for nuclear systems?

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:


Question: What is the role of the M5 validation database in nuclear licensing?

**Figure 26-2:** M5 Validation Database, (Predicted – Measured) Oxide Thickness vs. Rod Average Burnup and Time

The inspection team conducted a thorough review of the safety systems in the reactor building, ensuring compliance with all regulatory standards.


Question: What inspections are required for Alloy X-750 dowel-to-LGA support pad welds?

40 - 60: VT-3 visual inspections of the accessible surfaces in 100% of the support pad welds on a once every 10-Year inspection basis (longer intervals can be used if approved by the NRC).

The cooling system of the reactor is designed to maintain the temperature within optimal limits to ensure safety and efficiency.


Question: What is the purpose of uncertainty analysis in regulatory assessments?

An uncertainty analysis assesses the range of outcomes, and usually the relative probabilities of different outcomes within the range, produced from a combined propagation of uncertainties in model inputs.

Sensitivity analysis typically does not assess the relative likelihood of different outcomes.


Question: The operation of fluidizing the spent resin is critical for ensuring effective transfer and disposal of the material.

During resin transfer, this nozzle is used to ease the flow of the resin slurry. After resin transfer is complete, the tank is vented to the plant vent and returned to atmospheric pressure.

Spent samples with high chemical concentrations are held up in the chemical drain tank, then sampled.


Question: What is the significance of a Fire Probabilistic Risk Assessment in nuclear safety?

The Fire Probabilistic Risk Assessment (Fire PRA) Summary Document is essential for understanding the potential risks associated with fire hazards in a nuclear facility.

Electronic format is preferred. If electronic media is made available via an internet-based document management system, then document access must allow inspectors to download, save, and print the documents in the NRC's Regional office.


Question: Reactor Vessel Water Level monitoring in safety systems

a. Reactor Vessel Water Level - Low Low Low, (Level 1) | Automatic Initiation: - RX Water Vessel Low Low Low OR D/W Pressure Hi AND CS Time Delay Relay AND Reactor Pressure-Low (Injection Permissive).

b. Drywell Pressure - High | Automatic Initiation: - RX Water Vessel Low Low Low OR D/W Pressure Hi AND CS Time Delay Relay AND Reactor Pressure-Low (Injection Permissive).


Question: What is the purpose of the Emergency Feedwater System in a nuclear reactor?

The Emergency Feedwater System is designed to provide water to the steam generators in the event of a loss of normal feedwater supply, ensuring that the reactor can be cooled and preventing a core meltdown.

The Reactor Protective Instrumentation is essential for monitoring and controlling key safety parameters within the reactor to ensure safe operation and reactor protection.


Question: emergency plan for nuclear facilities

The Emergency Plan for North Anna Power Station includes detailed procedures for responding to a variety of potential incidents, ensuring the safety of the public and the environment.

The facility uses a combination of pressurized water reactors and advanced safety features to maintain operational efficiency.


Question: What is the significance of the Uranium Fuel Cycle in nuclear reactors?

Uranium Fuel Cycle is a series of processes involving the extraction of uranium from the earth, conversion, enrichment, and fabrication into fuel for nuclear reactors, as well as the management of the spent fuel.

The socioeconomics discussion will highlight the cumulative effects of project implementation on local communities and regional economies.


Question: The NRC allows for the use of alternative techniques when volumetric examination coverage cannot reach the standards specified in ASME Codes.

Volumetric examinations, in accordance with ASME Code and Appendix VIII, will be conducted to the maximum extent practical. Based on industry experience, examination coverage of the ASME defined examination volume is expected to be greater than 40 percent.

Section XI single-side Performance Demonstration Initiative UT-2 (PDI UT-2) examinations were performed on the subject weldlet branch connections located in Unit 3. A 45-degree shear wave transducer was used on the 4-inch weldlet side of the welds, as no meaningful ultrasonic examination can be performed from the 14-inch pipe due to the inability of the sound to reach the weld.


Question: leak detection system monitoring requirements

LC 10.5 of SUA-1580 requires daily measurement and documentation of pond freeboard and fluid levels in the leak detection system.

If fluid levels exceed 15 cm [6 in] in the leak detection sumps, a sample will be collected and analyzed for specific conductance and chloride.


Question: The shipments of 11e(2) byproduct material are designed to comply with DOT hazardous materials shipping provisions, ensuring safety during transportation.

11e(2) byproduct material will be transported to a licensed disposal facility (White Mesa) in Blanding, Utah. Current assessments show that the amount of solid 11e(2) byproduct material generated from the CUP will be within that facility’s annual disposal limits.

The transportation impacts associated with shipments of hazardous waste and used oil were not discussed in the CUP FEIS.


Question: Oconee Nuclear Station cultural resources

The Oconee Nuclear Station, with multiple contributing resources, including the reactor buildings, turbine buildings, intake structure, discharge structure, water tower, skimmer wall, and steam generator reinforcement facility.

The South Carolina State Historic Preservation Office provided initial concurrence that these resources meet the criteria for listing in the NRHP on October 7, 2021, and final concurrence of the report and findings on January 4, 2022 (Duke Energy 2022-TN8948, SCDAH 2023-TN9005).


Question: Licensees are required to evaluate the degree of embrittlement of reactor vessels, particularly through surveillance programs monitoring neutron damage and the effects of reactor trips.

Petitioners further contend that the test capsule data provided to the NRC by licensees who are part of the reactor vessel surveillance program where capsule Charpy test data taken from one nuclear reactor vessel is submitted to the NRC as representative of the amount of neutron damage to the RV of another RV is not sufficient to accurately determine the amount of neutron damage (embrittlement) of the latter RV.

First, there is no single test location authorized by the NRC or utilized by licensees where Charpy impact testing is performed. Therefore, the testing performed by one vendor can widely vary in accuracy from another vendor depending on the testing facility's equipment, testing procedures, qualifications of employees conducting the tests, human interpretation of the test results, gravity affects on the testing machine, the vintage of the testing machine, the placement of the test specimen in the testing machine, the machining of the v-notch in the test specimen, etc.


Question: What actions are taken if the mismatch between Tavg and Tref exceeds 5°F?

If Tavg and Tref mismatch can NOT be maintained less than 5°F, THEN: TRIP reactor, AND GO TO 1-E-0, Reactor Trip or Safety Injection.

As Load Dispatcher acknowledge Unit 1 load reduction.


Question: trans-river flow

While the potential for trans-river flow exists, it is likely that such flow would be quickly discharged to the river because flow in the river alluvium is toward the river.

Numerical simulation techniques have been used to further evaluate areas of previously documented trans-river flow on the Georgia side of the Savannah River (Clarke and West, 1998; Cherry 2006).


Question: What requirements are outlined by the NRC for obtaining a nuclear license?

The NRC requires licensees to analyze potential hazards to ensure the safe operation of nuclear facilities.

The quarterly reports detailing operational data are essential for understanding the overall performance of the plant.


Question: What are the criteria for strainer head loss in nuclear reactors?

Strainer head loss does not exceed half of the submergence depth of a partially-submerged strainer.

The results of the screening criteria and determination of a threshold break size are described in Section 2.4.1.


Question: radon mitigation measures

The quitclaim deed for the former mill site indicates that any habitable structures constructed on the property shall employ a radon ventilation system or other radon mitigation measures.

The city of Rifle constructed a waste water treatment facility and associated structures.


Question: The NRC staff's evaluation will take into account the technical aspects related to the licensee’s proposed changes in the licensing application.

NRC’s objectives of the audit are the following: Gain a better understanding of the detailed calculations, analyses, and bases underlying the LAR and confirm the NRC staff’s understanding of the LAR.

The audit team will consist of the following NRC staff from NRR.


Question: What are the aging effects requiring management for heat exchanger components?

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

The Water Chemistry program includes an analysis of the chemical composition of the water used in the reactor systems.


Question: wind direction and speed data

During April, Vicksburg, MS, the percentage frequency of wind direction and speed (mph) is recorded in a table showing various directions with corresponding speed ranges and average speeds.

Data from NCDC, 1997-2001 (Reference 3).


Question: Common Emergency Plan regulatory requirements

The CEP and Site-Specific Annexes were developed consistent with NUREG-0654/FEMA-REP-1 Revision 2, *Criteria for Preparedness and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants*.

Should you have any comments regarding the proposed fleet CEP and associated applicable Site-Specific Annexes, please contact Cynthia Cashwell (Turkey Point) or Peter Pollit (St. Lucie).


Question: The identification of various nuclides is essential for accurate safety assessments in nuclear facilities.

Identified Nuclides include specific radionuclides that are detected during the analysis process to ensure compliance with safety regulations.

The Ambient Background was recorded to understand the baseline measurements necessary for accurate peak analysis.


Question: What are the data quality objectives in nuclear licensing?

Data Quality Objectives (DQOs) are essential in defining the quality of the data required to support regulatory decisions and ensure safety assessments in the context of nuclear operations.

The conclusion of this report summarizes the findings from the remediation activities and outlines the next steps for compliance with regulatory standards.


Question: Americium-241 nuclear analysis

Liquid Scint N63, Solid "Dry Weight Corrected" | HU | | | | | | | | | | | |


Question: corrosion protection medium

Results of corrosion protection medium (CPM) tests to determine absorbed water content, corrosive ion concentrations and neutralization number confirm that acceptance criteria (with one exception) have been met.

Tendon wire yield strength, tensile strength and ductility will not change over time.


Question: Licensees must assess proposed changes, tests, and experiments to determine their impact on the licensing basis, particularly considering the FSAR.

50.59 process relies on adequate technical evaluation of the change “to ensure it is safe and effective.”

Obtain license amendment if CTE requires a TS change, or if CTE meets one or more of the 8 criteria in 50.59(c)(2).


Question: wind speed distribution

The Monthly and Annual Joint Frequency Distributions of Wind Speed and Direction by Atmospheric Stability Class provide statistical analysis of wind patterns relevant to safety evaluations.

Periods of calms are characterized by low wind speeds, where the recorded values indicate a lack of significant atmospheric movement.


Question: Emergency Core Cooling System components

The major mechanical components of the Emergency Core Cooling System are: Cold Leg Injection Accumulators. These accumulators are pressure vessels filled with borated water and pressurized with nitrogen gas. During normal operation each accumulator is isolated from the Reactor Coolant System by two check valves in series.

Each residual heat removal pump is a single stage, vertical position, centrifugal pump. It has an integral motor-pump shaft, driven by an induction motor. The unit has a self-contained mechanical seal cooling system.


Question: latest revisions to emergency response procedures

Latest revisions to this procedure have been approved by the Director of the Office of Homeland Security and Emergency Preparedness.

The document outlines the responsibilities of various agencies during an emergency, ensuring that all personnel are adequately trained.


Question: Onsite Power System

The Onsite Power System consists of all sources of electric power and their associated distribution systems in the generating unit. These sources are the main generator, two emergency diesel generators, the emergency supplement power source, and the batteries.

Figure 8-16 is the station main buses single line; Figure 8-3 is the typical Auxiliary Power Distribution System, Figure 8-35 is the Vital AC-DC Instrumentation and Control System; and Figure 8-34 is the Computer and Auxiliary Control Power System.


Question: What are the arrangements for medical treatment of contaminated personnel at a nuclear facility?

Arrangements for the medical treatment of contaminated injured onsite personnel and those onsite personnel who have received significant radiation exposures and/or significant uptakes of radioactive material are described.

Mishicot Ambulance Service is the primary provider for the transportation of contaminated, injured individuals.


Question: What are the performance acceptance criteria for nuclear fuel rods?

Please confirm that all the performance acceptance criteria and material limits. If any changes are noted, please identify those changes and highlight where in this TR they are justified.

What fabrication specifications will be applied to texture and SPPs for HiFi cladding in production and how do these compare to current generation Zr-2 specifications for the fuel rods?


Question: The COBRA-FLX code employs advanced modeling techniques to assess critical heat flux in reactor systems.

The COBRA-FLX formulation is integral for predicting critical heat flux under various operational conditions in nuclear reactors.

Reactor coolant systems are designed to prevent overheating by removing heat from the core and transferring it to the steam generators.


Question: nuclear liquid scintillation analysis

The following Analytical Methods were performed:


Question: nuclear medicine department transfer of ownership

Control of these assets will be given to Heart Rhythm & Pacemaker Specialist, LLC when Transfer of Ownership is granted by the Nuclear Regulatory Commission.

The name of the practice will be changing to Heart Rhythm & Pacemaker Specialist, LLC. The address remains the same at 25689 Kelly Rd., Ste. D, Roseville, MI, 48066.


Question: What is included in the Lubricating Oil Analysis program for reactors?

Continue the existing DCPP Lubricating Oil Analysis AMP, including the following enhancements: (a) Include periodic sampling and analysis to maintain lubricating and hydraulic oil contaminants, primarily water and particulates, within acceptable limits; (b) Perform sampling for water, particle count, and other parameters to detect evidence of contamination by moisture or excessive corrosion; (c) Include acceptance criteria for lubricating and hydraulic oil associated with the equipment within the scope of the Lubricating Oil Analysis AMP.

The commitment requires implementation of periodic maintenance checks and upgrades for all safety-related systems, ensuring compliance with regulatory guidelines.


Question: What is the significance of radiation protection in nuclear safety?

Radiation protection is a critical component of nuclear safety that focuses on minimizing exposure to ionizing radiation for workers, the general public, and the environment.

The training program includes a comprehensive review of mathematical principles that apply specifically to radioactivity, emphasizing the necessary calculations for effective measurement.


Question: The function of the DG load sequencer is critical during scenarios such as Loss of Offsite Power.

Although all loads that are started on the sequencer are important, the UFSAR states that the DG capabilities AND accident analysis were specifically considered in determining the sequence and timing.

The examinee must have knowledge of the purpose of DG load sequencer during a loss of offsite power.


Question: aging management for reactor vessel internals

Aging Management Review Results for Reactor Vessel Internals indicate that components such as lower girth welds of the core barrel assembly require management due to aging effects like cracking and loss of fracture toughness.

The CRGT assemblies include CRGT guide cards that serve a direct flow function and are made of stainless steel.


Question: definition of damaged fuel in nuclear context

SPENT NUCLEAR FUEL (SNF) assembly that cannot fulfill its fuel-specific or system-related function. SNF is classified as damaged under the following conditions: 1. There is visible deformation of the rods in the SNF assembly.  **Note:** This is not referring to the uniform bowing that occurs in the reactor; this refers to bowing that significantly opens up the lattice spacing. 2. Individual fuel rods are missing from the SNF assembly and the missing rods are not replaced by a solid stainless steel or zirconium dummy rod that displaces a volume equal to, or greater than, the original fuel rod. 3. The SNF assembly has missing, displaced or damaged structural components such that: 3.1. Radiological and/or criticality safety is adversely affected (e.g., significantly changed rod pitch); 3.2. The SNF assembly cannot be handled by normal means (i.e., crane and grapple); 3.3. The PWR SNF assembly contains fuel rods with damaged or missing grids, grid straps, and/or grid springs producing an unsupported length greater than 60 inches.  **Note:** PWR SNF assemblies with the following structural defects meet MAGNASTOR® system-related functional requirements and are, therefore, classified as undamaged: Assemblies with missing or damaged grids, grid straps and/or grid springs resulting in an unsupported fuel rod length not to exceed 60 inches. 4. Any SNF assembly that contains fuel rods for which reactor operating records (or other records or tests) cannot support the conclusion that they do not contain gross breaches.  **Note:** BREACHED SPENT FUEL RODS with minor cladding defects (i.e., pinhole leaks or hairline cracks that will not permit significant release of particulate matter from the spent fuel rod) meet MAGNASTOR® system-related functional requirements and are, therefore, classified as undamaged. 5. FUEL DEBRIS such as ruptured fuel rods, severed rods, loose fuel pellets, containers or structures that are supporting loose PWR fuel assembly parts.

The CONCRETE CASK LID is a thick concrete and steel closure for the CONCRETE CASK. The CONCRETE CASK LID precludes access to the TSC and provides radiation shielding.


Question: emergency operating procedures

In accordance with the guidance of OP1.DC10, Conduct of Operations, how are Foldout Page assignments made during an entry into an Emergency Operating Procedure?

Assignments should be reviewed with the operators as part of an update.


Question: The report includes quantitative measures of radioactive isotopes following established analytical methods.

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

The address of the company is S4601 State Hwy 35.


Question: storage rack criticality analysis

The racks are analyzed using the MCNP-05P Monte Carlo neutron transport program and ENDF/B-VII.0 cross-section library. The methodology used in this analysis is the peak Standard Cold Core Geometry (SCCG) in-core eigenvalue (k∞) criterion.

In this evaluation, in-core k∞ values and exposure dependent, pin-by-pin isotopic specifications are generated using the GE-Hitachi Nuclear Energy Americas LLC (GEH)/GNF lattice physics production code TGBLA06.


Question: Licensees must submit analyses of their emergency response plans to the NRC.

Emergency preparedness is a critical component of nuclear reactor operations, requiring meticulous planning and regular drills.

Under normal operating conditions, a nuclear power plant generates electricity, relying on fission reactions to produce heat.


Question: What are high radiation areas in nuclear facilities?

The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):

1. Workers exiting the containment contaminated area (CA) during the refueling outage.


Question: NuFuels will implement a Cultural Resources Management Plan that outlines the procedures for identifying and protecting historic and cultural resources before any land disturbances.

Prior to engaging in any land disturbing activities, NuFuels will finalize a CRMP to address cultural resources within the project areas pursuant to the NHPA Section 106 review and consultation process.

Mitigation measures for historic and cultural resources are discussed in Section 5.8 of this ER.


Question: What are personnel contamination protection techniques during nuclear cleanup?

Personnel Contamination Protection Techniques Applied During the Three Mile Island 2 Cleanup were critical to ensuring safety and preventing the spread of contamination during the recovery operations.

The cleanup efforts at Three Mile Island Unit 2 involved extensive documentation and analyses to understand the aftermath of the accident.


Question: Licensees must assess the adequacy of any proposed safety systems for compliance purposes.

The reactor's safety systems must be rigorously tested to ensure they meet strict regulatory requirements for operation and emergencies.

Operational performance is often evaluated through various effectiveness assessments for projects unrelated to reactor safety.


Question: What are the records related to safety inspections of nuclear facilities?

A copy of the list of NRC Office of Inspector General investigations CLOSED during Calendar Year 2020.

Appeal the denial of records pertaining to named individual as described in response to NRC-2021-000301.


Question: The height of the wall and its structural aspects must comply with specified dimensions for missile impact assessment.

- Diagram includes various structural elements


Question: aging management evaluations for concrete

Increase in porosity and permeability; cracking; loss of material (spalling, scaling) due to aggressive chemical attack is an applicable aging effect and will be managed by the DCPP ASME Section XI, Subsection IWL AMP (B.2.3.29).

The DCPP ASME Section XI, Subsection IWL AMP (B.2.3.29) will be used to manage the aging effects on other structural materials as outlined in the regulatory framework.


Question: There are no expected modifications in the facilities, equipment, or procedures associated with the licensed program due to the ongoing Transaction.

Describe any changes in the location, facilities, equipment, radiation safety program, use, possession, waste management, or other procedures that relate to the licensed program.

There are currently no changes anticipated in personnel or management having direct operational responsibility for, and control of, licensed activities at the Metropolis Facility because of the Transaction.


Question: During the reactor operation, it is crucial to monitor automatic safety actions to ensure compliance with operational protocols.

- Execute AOP-60:
  - Verify the following automatic actions:
    - SGT B Start per Section G of OP-20
    - RWCU Isolation per Section G of OP-28
    - Isolate Reactor Bldg Vent per Section G of OP-51A

**Role Play**: If dispatched to restore RPS power, report preparations briefs will begin.


Question: What is the significance of ASME Code, Section III in nuclear reactor design?

The vessels of the reactor coolant pressure boundary are designated ASME Code, Section III, 1968, including addenda up to Winter 1968, Class A.

A complete tube-tubesheet complex analysis verified structural integrity for a primary pressure-loss accident plus the design-basis earthquake.