Question: The NRC requires a comprehensive assessment of the potential environmental effects associated with the proposed actions, especially during the license renewal process.

The environmental report must include an analysis that considers the environmental effects of the proposed action and alternatives available for reducing or avoiding adverse environmental effects. [10 CFR 51.45(c)]

The information submitted should not be confined to information supporting the proposed action but should also include adverse information. [10 CFR 51.45(e)]


Question: gamma spectrum analysis

Energy Calibration Used Done On: 7/8/2014


Question: What is a major processor in nuclear licensing?

Major processor means a user processing, handling, or manufacturing radioactive material exceeding Type A quantities as unsealed sources or material or exceeding four times Type B quantities as sealed sources but does not include nuclear medicine programs, universities, industrial radiographers, or small industrial programs.

Natural radioactivity is the process by which unstable atomic nuclei lose energy by emitting radiation.


Question: The review process for new nuclear reactor designs involves rigorous safety assessments.

The NRC's licensing process for nuclear reactors ensures that all designs comply with safety standards and regulations.

The preview feature allows users to see how their changes will look before finalizing them.


Question: The NUSO must refer to Technical Specifications when dealing with the Rod Position Indicator (RPI) issues.

The NUSO is required to enter Technical Specifications due to, only LCO 3.1.8, Rod Position Indication, not being met.

Plausible because the applicant could incorrectly determine that the indications provided indicate that an RCCA has dropped.


Question: SLR application review

Duke Energy will submit a high-quality SLR application to support an 18-month Staff review.

The RNP SLR ER will be consistent with the 2024 revised NRC guidance to leverage new efficiency opportunities where applicable.


Question: What is the purpose of a regulatory audit in nuclear licensing?

A regulatory audit summary will be provided within 90 days of the completion of the audit.

The regulatory audit will be held via Teams.


Question: nuclide identification report


Question: What are the licensing requirements for nuclear reactors?

The Nuclear Regulatory Commission (NRC) establishes requirements for licensing nuclear reactors, which include safety assessments, environmental reviews, and operational guidelines.

The temperature is measured in degrees Celsius and is crucial for understanding thermodynamic properties in various engineering applications.


Question: control rod position indication

The RPI System indicates within 12 steps of the group step counter demand position when LCO 3.1.5, "Rod Group Alignment Limits," met.

A deviation of less than the allowable limit, given in LCO 3.1.5, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits).


Question: satellite auxiliary power systems

Both Unit 1 and Unit 2 SAT disconnect links are INSTALLED.

If notified as US, acknowledge failure of acceptance criteria and direct candidate to complete the surveillance.


Question: interim qualification in nuclear regulatory processes

A candidate who has not completed all of the requirements for final certification in his or her qualification program may obtain interim qualification to independently perform his or her specific work activities in the discipline for which prescribed training has been completed.

The NRC offers fundamental health physics courses that should first be considered.


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

The COBRA-FLX correlation is a significant modeling tool used to predict the critical heat flux in nuclear reactors, specifically designed for use in reactor thermal-hydraulic analysis.

The document includes various technical specifications for reactor assembly configurations to ensure compliance with safety regulations and operational efficiency.


Question: What are the aging management programs for steel piping exposed to condensation in nuclear systems?

The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24) program will be used to manage loss of material of the carbon steel piping, piping components, tanks, and valve bodies exposed to condensation in the Condensate System, Main Generator and Auxiliaries System, Main Steam System, and Main Turbine and Auxiliaries System.

Consistent with NUREG-2191. The Lubricating Oil Analysis (B.2.1.25) and One-Time Inspection (B.2.1.20) program will be used to manage loss of material of the carbon steel and cast iron blower housing, heat exchanger components, piping, piping components, pump casings, tanks, and valve bodies exposed to lubricating oil in the Feedwater System, Main Generator and Auxiliaries System, and Main Turbine and Auxiliaries System.


Question: reactive inspection process

The August 10, 2020, event at Duane Arnold was potentially risk significant and warranted consideration for chartering a reactive inspection.

The operating experience group is aware of this event and will consider the specifics for further learnings and dissemination.


Question: What are the functions of the Emergency Response Facility Control System (ERFCS)?

ERFCS provided plant monitoring, data acquisition, and critical plant data in the form of real-time status displays. ERFCS monitors were located in the MCR, TSC, LEOF, and CERC. Signal inputs for most components monitored by the ERFCS were taken from instrument terminals in the MCR.

Revision 49 of the Emergency Plan [Reference 14] reflected implementation of a design change which replaced ERFCS with the PCS.


Question: The NRC defines the criteria for evaluating potential reactivity initiated accidents during reactor operation.


Question: The system measures coolant temperature using resistance temperature detectors positioned at critical points in the reactor's coolant loops.

Instrumentation necessary to monitor the reactor vessel level and to detect conditions of inadequate core cooling is provided by the reactor vessel level indication system which is described in Section 7.9.2.2.

Steam dump control, using T_avg control, will accommodate excess power generation.


Question: What is the significance of Verification and Validation (V&V) in nuclear licensing?

The NRC staff noted that, in NextEra letter dated 10/05/2023, the response to RAI Sections 1.1 and 1.3 indicates that Task Support Verification (TSV) and Design Verification (DV) information will be made available via two separate sets of results summary reports (RSRs), with the first set being submitted in May 2023 and the second set being submitted in January 2024.

Please confirm—when the execution of PV2 testing will occur.


Question: The MAGNASTOR® system has been approved for general use by holders of specific nuclear reactor licenses.

The MAGNASTOR® system, which is authorized by this certificate, is hereby approved for general use by holders of 10 CFR Part 50 and Part 52 licenses for nuclear reactors at reactor sites under the general license issued pursuant to 10 CFR 72.210, subject to the conditions specified by 10 CFR 72.212, and the attached Appendix A and Appendix B.

The holder of this certificate who desires to make changes to the certificate, which includes Appendix A (Technical Specifications and Design Features) and Appendix B (Approved Contents), shall submit an application for amendment of the certificate.


Question: The Internal Events PRA model was developed following the guidelines provided in NUREG-1855 to enhance the understanding of uncertainty sources.

The Internal Events baseline PRA model uncertainty report was developed, based on the guidance in NUREG-1855 and EPRI 1016737.

Fermi 2 has FLEX portable equipment but this equipment is not credited in the PRA models supporting the TSTF-505 application.


Question: RCS primary temperatures

Figure 5-232 shows the code-to-data comparison of the RCS primary temperatures at two locations: at the SG primary side's inlet (upper plenum) and outlet (mid dowcomer).

Figure 5-233 shows the code-to-data comparison of the PZR level. Figure 5-234 shows the code-to-data comparison of the RPV level.


Question: The report provides an overview of key findings related to the future of materials harvesting in nuclear applications.

In particular, please review how your presentation and contribution to the discussion is characterized, in case you feel it should be clarified in any way.

Comments, edits, and suggestions on any other parts of the report are welcome.


Question: The NRC will issue a draft Supplemental Environmental Impact Statement (SEIS) for public review as part of its licensing process.

The NRC staff plans to issue a draft SEIS for public comment in June 2025.

During the scoping process, the NRC staff did not identify any Federal, State, local, or Tribal agencies as cooperating agencies for the SEIS.


Question: waste gas vent valve

The WASTE GAS VENT VALVE is designated as 2WG41 and is listed with a position of ≤100%.

The GDT STOP TO PLNT V is listed with a position of CLOSED.


Question: What is Minimum Detectable Activity in nuclear measurements?

The City of Genoa is located along the Mississippi River, providing a scenic backdrop for recreational activities.


Question: A rapid increase in feedwater injection can lead to an unexpected reactor shutdown as safety protocols react to elevated water levels.

Feedwater controller demand is postulated to fail upscale that results in maximum feedwater injection into the RPV, resulting in a rapid water level rise and a trip of the main turbine and the turbine-driven feedpump turbines when water level reaches the Level 8 trip setpoint.

This event is considered to be an incident of moderate frequency.


Question: proprietary information classified under 10 CFR

The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) 'Trade secrets and commercial or financial information.'


Question: What are the actions to take when an ECCS subsystem is inoperable?

For Technical Specification (TS) 3.5.2, Action a – "With one ECCS subsystem inoperable," RMAs could include: Actions to minimize the magnitude of the risk increase - Verify system alignment of low pressure ECCS. Defer planned maintenance or testing that affects the reliability of those safety systems that provide a defense-in-depth.

The following sections provide example RMAs.


Question: Off-Gas System aging effects management

The following aging effects associated with the Off-Gas System components require management: Cracking, Cumulative fatigue damage, Loss of Material, Loss of preload.

The materials of construction for the Off-Gas System components are: Copper Alloy, High-strength steel, Stainless steel, Steel.


Question: maximum surface dose rate for STORAGE CASK

The maximum surface dose rates for the STORAGE CASK, shall not exceed the following limits: PWR and BWR – 120 mrem/hour gamma and 5 mrem/hour neutron on the vertical surfaces; and PWR and BWR – 900 mrem/hour (neutron + gamma) on the top.

Separate Condition entry is allowed for each MAGNASTOR® SYSTEM.


Question: heat exchange coefficient calculation

The heat exchange coefficient is the net rate at which heat is lost or gained by a body of water, defined by the following equation (Ref 1): K = 15.7 + (0.26 + B) f(w), BTU/ft²/Day/F, where: - K = Heat Exchanger Coefficient, - B = Vapor Pressure slope, - f(w) = Wind Speed function.

If the temperature of a layer is less than the temperature of the layer below it, then the layers are mixed since a cooler, more dense surface layer would be unstable.


Question: present value calculation for accident risk

To calculate the present value for the public health (accident) and offsite property attributes, when the monetary value or cost C₀ can occur with a frequency f, NUREG/CR-2723 provides the following formula: $$ PV = \int_{t_l}^{t_f} C_0 e^{-rt} dt = \frac{C_0 [e^{-rt_l} - e^{-rt_f}]}{r} $$ where - \( t_l \) = time of onset of accident risk  - \( t_f \) = time of end of accident risk

For an annual discount rate of 7 percent (r = 0.07 per year), the present value of avoided risk (monetized) becomes: $$ PV = \frac{(10,000) \text{ yr}^{-1} }{0.07} \left[ e^{-0.07(0)} - e^{-0.07(25)} \right] = \$118,000 \text{ per facility} $$


Question: engineered safety features in nuclear reactors

One of the control room system JPMs must be an engineered safety feature.

Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function.


Question: The analysis of the small break loss-of-coolant accident considers the mass flow rates in the steam generator system.

The small break loss-of-coolant accident (LOCA) analysis is critical for evaluating the performance of the reactor's cooling system during an emergency scenario.

Key measurements such as temperature and pressure are essential for monitoring the reactor's operational safety and efficiency.


Question: stress multiplier for leakage probability

Based on this multiplier, an increase in the applied stress can be accommodated; under typical inspection scenarios, as much as an 80% increase in stress can be accommodated compared to a possible variation of 20% based on the varying R/t ratio.

These stress multipliers account for the effect of geometrical variations (in terms of R/t ratios) discussed in Section 4 of this report.


Question: The NRC specifies that there must be two independent functional chilled water loops operational at all times.

LCO 3.7.12 Two independent essential services chilled water loops shall be OPERABLE.

The completion time for restoring the ECW train must adhere to the established Risk Informed Completion Time Program.


Question: control room notification procedures

Ensure Nine Mile Point 1 Control Room (due to effect on 115 KV line loading) has been notified that 10100 Bus will be transferred from reserve to normal service.

• Ensure AUTO MANUAL T-3 LTC CONTROL switch is in MAN.


Question: The NRC requires licensees to maintain criteria for flow similarities when conducting experimental tests.

Similitude Criteria to Preserve Flow Characteristics Between Prototype and Scaled Experiment in Twisted Elliptical Tube Tests outline the necessary dimensions and characteristics that must match between models to ensure accurate testing and results.

Regulatory filings must include detailed information about changes in plant design and equipment maintenance schedules, ensuring compliance with all applicable laws.


Question: Software Life Cycle Process

The Common Q SPM only includes the Software Life Cycle Process Planning Documentation as outlined in SRP BTP 7-14, Section B.2.1.

The following table provides the cross reference between the documents listed in BTP 7-14 Sections B.2.2 and B.2.3 and the name of the Westinghouse LGS PPS corresponding document.


Question: What are the criteria for fuel rod bow in reactors?

The criteria for fuel rod bow have been met by calculating DNBR and LHGR burnup thresholds and penalties. The thresholds and penalties will be verified on a cycle-specific basis.

Local areas with excess water-to-uranium ratios can develop and can increase the local power peaking.


Question: The NRC mandates that licensees conduct extensive safety evaluations for operational reactors.

The ACL program involves a comprehensive evaluation of analytical results and field measurements to ensure compliance with safety standards.

Annual financial audits are critical for the stability of utility companies, impacting their operational budget and infrastructure investments.


Question: The NRC mandates that any hazardous materials be clearly identified during the sample collection process.

RCRA metals include substances such as mercury, which is categorized as flammable and ignitable, and cadmium, which is listed as polyhalogenated biphenyls.

Samples can only be analyzed under certain conditions and require specific handling to avoid contamination.


Question: What is the Final Safety Analysis Report?

The Final Safety Analysis Report (FSAR) is a document required by the NRC that provides a comprehensive analysis of the safety of a nuclear facility, including details on its design, operation, and safety measures implemented to protect public health and safety.

Fire Probabilistic Risk Assessment (FPRA) is used to evaluate fire risks associated with nuclear facilities, focusing on the potential for ignition and its impact on safety.


Question: The heat generated during the Zircaloy-steam reaction can significantly affect cladding temperatures during reactor shutdowns, as evidenced by the TH-1 test no. 130.

In TH-1 test no. 130, the reactor shutdown when the PCT was approximately 1850°F; and after the reactor shutdown, cladding temperatures kept increasing because of the heat generated from the Zircaloy-steam reaction (of course, there would have also been a small amount of actual decay heat) and the peak measured cladding temperature was 2040°F.

Analyses using the Baker-Just and Cathcart-Pawel correlations would also most likely under-predict the amount of heat that Zircaloy oxidation generated in TH-1 test no. 130, which I discussed in the last PRB meeting.


Question: minimum detectable concentration

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


Question: The facility is intended to safely store spent nuclear fuel and greater-than-class-C waste until they can be moved to another location for disposal.

The WCS CISF provides interim storage for SNF and GTCC waste loaded in canisterized systems until retrieval of the canisters for transport to a repository or other site.

The total storage capacity for Phase 1 is limited to 5,000 metric tons of heavy metal (MTHM) for SNF and 510,000 pounds of GTCC waste.


Question: What is the unit of measurement used for radioactive isotopes?

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


Question: The hydrology of the area indicates that the surficial aquifer will direct any discharge towards local water bodies, specifically Mill Creek.

The piezometric surface of these aquifers also indicates flow in a southeastward direction.

As a result of these facts, accidental discharge of industrial wastes into the surficial aquifer underlying the NFCS holding ponds would be confined to the aquifer and would not migrate to the lower aquifer as long as present casings in the deep holes remain effective to migration of fluid (static level of surface aquifer about 130 feet above MSL—static level of lower aquifer 122 feet above sea level).


Question: nuclear power plant reactor types

There are several types of nuclear reactors, including Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs), and Advanced Gas-cooled Reactors (AGRs), each with distinct operational characteristics and safety features.

The adoption of renewable energy sources has gained momentum over the past decades due to increasing concerns about climate change and environmental impact.


Question: maximum magnitude in seismic studies

Subsequent hazard studies have used maximum magnitude (Mmax) values that are within the range of maximum magnitudes used by the six EPRI models.

The largest historical earthquake to occur in the CVSZ was the body-wave magnitude (mb) 5.0 Goochland County event on December 23, 1875 (Reference 80).


Question: What is the process for notifying the NRC about a change of control for a materials use license?

The letter includes the information required (following the guidance provided in NUREG 1556 Vol.15, Rev 1 Section 5 and Appendix E) to determine if this organizational change is an actual transfer of control of our NRC Materials Use License.

The Hospital’s legal team has been working diligently to ensure that all of our licenses and registrations are revised to reflect the upcoming organizational change.


Question: What is the purpose of the Standard Review Plan for Non-Power Reactors?

The Standard Review Plan for Non-Power Reactors provides guidance for NRC staff on the review of non-power reactor licensing applications and is designed to ensure a consistent and thorough evaluation of safety and regulatory compliance standards.

The proposed rule outlines the necessary steps for draft regulatory and backfit analysis, detailing how such analyses should be conducted to assess the implications of regulatory changes.


Question: What is ASME Section XI, Subsection IWE in the context of nuclear aging management?

The ASME Section XI, Subsection IWE program (B.1.219) will be used to manage loss of material of the steel elements in the downcomers and torus shell ring girders exposed to air - indoor uncontrolled or treated water in the BFN Mark I steel Containment Structure.

Fatigue is a TLAA; further evaluation is documented in Subsection 3.5.2.2.1.5.


Question: What is the eutectic composition in sodium tetraborate solutions?

Finally, when the system temperature is -0.42°C (+31.24°F), the remaining liquid freezes to a solid with a boron concentration of 2220 ppm. The composition of this solid is known as the eutectic composition.

The driving force for diffusion between crystals of sodium borate through the pure ice matrix is a concentration gradient.


Question: What areas are defined as essential fish habitats in relation to nuclear plant locations?

No essential fish habitat (EFH) areas located within the RNP vicinity

Eleven state-listed species occur within Chesterfield, Darlington, and Lee counties.


Question: safe distances for explosion events

The Table 2.2-18 provides data on safe distances for explosion events related to various chemicals, indicating the necessary distance to safety-related buildings and the power block boundary.

The analysis of the various chemical distances shows significant variation with respect to the type of chemical stored and the geographical features of the site.


Question: What are the expected noise impacts from the operation of two ALWR units?

The NRC staff evaluated the noise impacts from operations of two ALWR units with a total net electrical output capacity of 2,234 MWe at the W.S. Lee Nuclear Station site in Section 5.8.2 of NUREG-2111 (pp. 5-69 through 5-70). In that analysis, the staff considered noise levels from draft cooling towers, pumps, loudspeakers, and transformers. The staff concluded in NUREG-2111 (NRC 2013-TN6435) that noise impacts from operation of two ALWR units would be minor. The NRC staff incorporates the analysis in Section 5.8.2 of NUREG-2111 (pp. 5-69 through 5-70) here by reference. Therefore, the noise impacts from operation of the ALWR portion at the W.S. Lee Nuclear Station site would be SMALL.

Noise sources during construction of a new nuclear alternative would include those discussed for all replacement power alternatives in Section 3.3.6.2 of this EIS.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors includes multiple stages, such as the application for construction permits, operating licenses, and the review of safety evaluations by the NRC.

The coolant system is designed to maintain optimal temperatures and pressures within the reactor core during normal operations.


Question: total mass flow in nuclear safety analysis

The total mass flow (lb) for various scenarios is presented, showing the mass released over time for ruptured and intact steam generators.

The table also includes values for intact steam generators, indicating their performance under different operational conditions.


Question: The NRC mandates that calculations for average relative power values must be accurately performed to ensure proper regulatory compliance.

The Standard requires that the applicant calculates the average relative power values from data collected from NOB Sheets A-2 for the respective NIS Power Range Detectors and records them in appropriate blocks on Attachment 4.

The examiner notes highlight that an acceptable range is ± 0.0002 to account for rounding errors.


Question: safety factors against overturning and sliding

Minimum Factors of Safety Against Overturning and Sliding at Design Ground-Water Elevations for Design-Basis Earthquake


Question: nuclear power plant noise impact assessment

The NRC staff concludes that the overall noise impacts associated with construction of the combination alternative would be **SMALL to MODERATE**.

Vessel-traffic-related noise would be intermittent and decrease as the distance from shore increases.


Question: What is the activity level of CS-137?

CS-137 has an activity of 85.12 uCi/grams.

The energy emitted by I-131 is measured at 636.97 keV.


Question: The regulatory framework mandates assessments for water resources adjacent to nuclear facilities.

The Savannah River is a critical hydrological feature impacting the environmental considerations for nuclear licensing.

Daniels Branch is a tributary that serves as a key waterway for local ecosystems.


Question: What are the high-fluence IAD effects in nuclear materials?

High-fluence (up to 50 dpa) IAD effects with representative LWR exposure conditions.

The initial discussions for this research began in 2006, and harvesting of materials took place in 2013.


Question: What is a partial length fuel rod in BWR reactors?

Partial length fuel rods are utilized in BWR designs to optimize fuel usage and achieve better core management.

The location sketches provide critical information about the fuel rod arrangement in the reactor core.


Question: byproduct material regulations

Medical Use of Byproduct Material Parts 20, 32 and 35

Requirements for Certain Generally Licensed Industrial Devices


Question: What is the purpose of Technical Specification Surveillance Requirement 3.8.6.6?

Surveillance Requirement 3.8.6.6 ensures that the equipment performance is regularly monitored and maintained according to the specifications set forth in the Monticello Technical Specifications.

The Monticello Nuclear Generating Plant is located in Monticello, Minnesota, and has been operational since 1971.


Question: The departure from nucleate boiling ratio quantifies the thermal margin available to prevent boiling crisis scenarios during reactor operation.

A calculated DNBR value greater than this design limit ensures that there is, at a minimum, a 95% probability with 95% confidence (95/95) that a departure from nucleate boiling will not occur.

Framatome has developed and continues to use numerous CHF correlations for its fuel assemblies.


Question: CISCC risk factors for dry storage canisters

A combination of a susceptible material, sufficient applied or residual tensile stresses, and an environment where chloride ions are present is necessary to induce CISCC.

To date, examinations of welded austenitic stainless steel canisters have found no evidence of CISCC or localized corrosion, such as pitting or crevice corrosion, at any site.


Question: defense-in-depth categorization review

inadequate/failure to perform defense-in-depth categorization review for multiple components.

Three Green NCVs and one licensee identified NCV:


Question: What is the relationship between cladding strain and oxidation in BWR experiments?

The 1997 ORNL paper discusses all of the BWR CORA experiments—CORA-16, -17, -18, -28, -31, and -33—and states that 'concurrent cladding strain and oxidation in the β Zircalo y phase regime [which commences above approximately 1922°F] must be considered in the experimental analysis' of the BWR CORA experiments.

NRC's conclusion is unsubstantiated, as the information presented in this section indicates.


Question: The presence of nuclides and their respective activities are important in assessing radiation safety.

Wt mean Activity (pCi/grams) indicates the average radioactive decay rate of a nuclide per gram of material.

Errors quoted at 2.00sigma signify the range of statistical uncertainty in the measurements.


Question: emergency preparedness for small modular reactors

Regulatory Analysis for the Final Rule: Emergency Preparedness for Small Modular Reactors and Other New Technologies outlines the necessary criteria to ensure that emergency preparedness measures are adequate for the unique operational characteristics of small modular reactors.

The Environmental Assessment for the Final Rule—Emergency Preparedness for Small Modular Reactors and Other New Technologies discusses the ecological impacts of implementing new safety protocols.


Question: What is the purpose of a decommissioning funding plan (DFP) in nuclear regulation?

The exemption would relieve a licensee from the requirement for a DFP, for the possession and use of Germanium-68 (Ge-68)/Gallium-68 (Ga-68) generators, when certain conditions are met.

Enclosed is a revision to the July 29, 2016, memorandum that clarifies the technical basis for the subject exemption authorization.


Question: What is an aging management program in nuclear regulation?

the first block in their program process for 1829 was aging management programs.

What do those boxes constitute? In that way, there are differences into how we would respond based upon what the expectations are from staff.


Question: What is the purpose of other hazard analysis in nuclear safety?

Other hazard analysis establishes the frequency of occurrence of different intensities of the hazard being analyzed and uses a site-specific probabilistic evaluation that is based on recent available data and site-specific information.

The study of nuclear physics encompasses various phenomena, including radioactive decay and neutron interactions, which are fundamental to understanding the behavior of nuclear materials.


Question: uranium ISR recovery methods

Uranium ISR methods have been improved over the years to increase efficiencies and minimize costs. The primary recovery method for uranium is now ISR rather than conventional surface or underground mining. ISR has lower operating costs and also reduces exposure of radioactive materials to the atmosphere.

It is not possible to precisely quantify all the economic benefits and costs of the project for any one community because many of the benefits, such as tax revenues, depend upon prevailing market prices, which are subject to change due to unpredictable economic and political factors.


Question: What is the significance of altered thermal stratification of lakes in nuclear power operations?

Altered thermal stratification of lakes can impact aquatic life and water quality due to changes in temperature profiles resulting from turbine discharge of cooling water.

The NRC evaluates the impacts of air quality on public health and the environment, including emissions from nuclear plants that operate under specific regulatory requirements.


Question: The NRC is responsible for overseeing the review of licensing requests related to the operations at the Palisades Nuclear Plant.

In a letter dated July 1, 2024, the U.S. Nuclear Regulatory Commission (NRC) notified your office that it is conducting an environmental review for a series of licensing and regulatory requests submitted by Holtec Decommissioning International, LLC (HDI) that, if approved, would collectively support reauthorizing power operations at the Palisades Nuclear Plant (Palisades).

To keep consulting parties informed and in accordance with the NRC’s Principles of Good Regulation, the NRC and DOE LPO staff have performed various activities.


Question: What is the NIS Calibration for reactor power range detectors?

NIS Calibration (i.e., Conversion) Factors are used to determine the relationship between current measurements and the reference power for detectors in a nuclear power plant.

The BOTTOM (B) DETECTOR parameters include R_b = 0.03181 V/µA and b_b = -1.763 µA/∆Φ, which are critical for understanding the calibration requirements.


Question: 11e.(2) byproduct material transportation

Figure 3.2-4. Anticipated 11e.(2) Byproduct Material Transportation Route

Figure 3.1-5. Surface and Mineral Ownership within the Unit 1 Project Area


Question: nuclear reactor safety systems

The safety systems in a nuclear reactor are designed to prevent accidents and mitigate their consequences by ensuring the reactor operates within safe limits under all conditions.

The regulatory body requires detailed plans for the construction of new facilities and their compliance with safety standards.


Question: What does potentiometric surface mean in groundwater studies?

The potentiometric surface elevation contour indicates the level to which water would rise in tightly cased wells, reflecting the pressure in the aquifer.

The Mississippi River has an elevation of 905.67 feet, providing important hydrological context for regional water management.


Question: The NRC distinguishes between devices that require registration and those that do not...

NRC regulations state that certain devices are not subject to registration if they meet specific criteria for use and safety.

The serial number of the device is essential for tracking its usage and compliance within the registry.


Question: automatic load sequencer

[One required] [automatic load sequencer] inoperable. | 3.8.1.F.1

DC Sources - Operating | 3.8.4 | DNPS TS is organized to take advantage of the availability of the opposite unit's 125 VDC subsystem for operability of certain shared safety-related equipment.


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

The COBRA-FLX is a computational tool that utilizes the ORFEO-HMP critical heat flux correlation to analyze heat transfer during reactor operations, particularly under conditions that affect thermal hydraulic safety.

The table includes a variety of test assembly conditions such as pressure, temperature, and flow rate, which are essential for the evaluation of reactor performance in non-mixing grid tests.


Question: 10 CFR Part 53 regulation overview

10 CFR Part 53 establishes a risk-informed, technology-inclusive regulatory framework for commercial nuclear plants, allowing for more flexible approaches to licensing based on the specific technologies and risks involved.

The staff introduction focused on the qualifications and roles of the members in the Nuclear Regulatory Commission, emphasizing their extensive experience in nuclear safety and regulatory practices.


Question: The NRC's current guidelines permit up to 5.0 weight percent uranium-235 in nuclear fuel, but advancements in technology may lead to the need for exceptions for higher enrichment levels.

Current licensing framework allows for the use of ≤ 5.0 weight percent uranium-235; however, technology developments may require numerous exemptions to utilize fuel enriched above 5.0 weight percent.

Rulemaking would provide for a generically applicable standard informed by public input, providing consistent and transparent communication, rather than individual licensing requests as discussed in SECY-21-0109, “Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors.”


Question: What are the properties of PB-214 regarding its activity?

The weight mean activity of PB-214 is 2.22E-01 pCi/grams with an uncertainty of 4.52E-02.

X = nuclide rejected by the interference analysis.


Question: licensing-basis events

Analyses of licensing-basis events other than design-basis accidents (DBAs) must be performed for LBEs as required by § 53.450.

It is difficult for this commentator to follow the complete logic and the criteria basis so needs some explanation.


Question: industry initiatives in regulatory actions

The presence of industry initiatives is potentially very important in the estimation of costs and benefits and, as such, its treatment in the regulatory analysis should be explicitly considered.

Issues related to adequate protection of public health and safety are deemed the responsibility of the NRC and thus cannot be addressed through industry initiatives.


Question: maximum water surface elevation

Ice Jam Breach Maximum Water Surface Elevation (ft. NAVD) is defined as the highest water level reached during an ice jam event, measured in feet relative to the North American Vertical Datum.

The coincident wave runup measures the increase in water level due to waves acting on the shoreline, which can vary greatly depending on local conditions and coastal configuration.


Question: nuclear security baseline inspection

On December 1, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Palisades Nuclear Plant and discussed the results of this inspection with you and other members of your staff.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, “Public Inspections, Exemptions, Requests for Withholding.”


Question: What are the requirements for submitting technical documentation in nuclear licensing?

The NRC requires that all licensees submit technical documentation that adheres to the standards set forth in 10 CFR Part 50, which includes requirements for safety analysis, environmental assessments, and emergency procedures.

The dimensions of the items listed above are critical for inventory management, ensuring that all vendors are stocked adequately for operational needs.


Question: The locations of the monitor wells are critical for understanding subsurface conditions.

The monitor wells in the Terrace Unit are installed to track groundwater levels and assess the aquifer's response to environmental conditions.

The map details include a scale of 1 inch for every 300 feet, providing a visual aid for understanding distances.


Question: The average enrichment of uranium in a PWR fuel assembly is critical for defining its performance metrics.

Initial Assembly Avg. Enrichment wt % 235U (E) values are essential in determining the fuel cycle behavior and performance of the reactor.

Reactor coolant systems play a crucial role in maintaining the operational temperatures of the reactor core.


Question: site-specific analysis for nuclear power plants

As part of the Site-Specific Analysis, Xcel Energy considered relevant new information on geology and soils since initial license renewal.

Nuclear power plant cooling systems may compete with other users relying on surface water resources, including downstream municipal, agricultural, or industrial users.


Question: The RICT Program implements strategies to address uncertainties identified in the Probabilistic Risk Assessment (PRA) modeling.

The purpose of the DNPS FLEX equipment is to provide independent means capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink resulting from a beyond design basis external event (BDBEE) by providing adequate capability to maintain or restore core cooling, containment cooling, and Spent Fuel Pool (SFP) cooling capabilities to all units on site.

The approach taken is, therefore, to review these documents to identify the items which may be directly relevant to the RICT Program calculations, to perform sensitivity analyses where appropriate, to discuss the results and to provide dispositions for the RICT Program.


Question: The regulatory framework mandates comprehensive assessments of nuclear facility risks...

The safety analysis report must detail the potential accidents and their consequences, including radiological assessments and mitigation strategies.

The construction of the reactor requires compliance with seismic performance criteria and well-defined engineering standards.


Question: probabilistic risk assessment in nuclear regulation

In August 1995, the NRC adopted the following policy statements regarding the expanded use of PRA: PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices.

Regulatory Guide 1.174 provides general guidance for use by licensees in the implementation of risk-informed analyses and explains that a risk-informed approach can provide valuable insights and guidance for use in interactions between the NRC and licensees.


Question: subsequent license renewal

This application seeks a subsequent license renewal for Dresden Nuclear Power Station Units 2 and 3, addressing the environmental impacts associated with extended operations beyond the original licensing period.

The report includes an extensive overview of onsite construction activities, detailing the initial site preparation efforts and the installation of facility infrastructure.


Question: Measurement units for area in reactor safety assessments


Question: nuclear reactor construction inspection program

The Construction Inspection Program: Inspections of Construction and Operational Programs is designed to ensure that all construction activities meet regulatory requirements and that reactors are built and operated safely.

The Low-Level Waste Program includes management and disposal of lower activity waste materials generated from nuclear power operations.


Question: transportation of radioactive waste

Fasken does not explain—or even address—why the Board should reach a different result than we reached with respect to a virtually identical claim in this very proceeding, and which the Commission affirmed in CLI-2014-14.

However, Fasken does not explain why, under NEPA or the NRC's environmental regulations, the DEIS must provide a more thorough analysis of hypothetical future shipping routes.


Question: nuclear licensing affidavit

Framatome submitted an application for withholding and affidavit regarding the proprietary information contained in the documents because it pertains to safeguarding the competitive position of the company.

The Millstone Power Station Unit 3 operates under multiple regulatory regimes to ensure compliance and maintain high safety standards.


Question: What is a bump-up factor in nuclear safety testing?

Licensees should provide the details and the technical basis that show why the bump-up factor from the particular debris bed in the test is appropriate for application to other debris beds.

Wolf Creek is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the Wolf Creek chemical effects analysis. See Figure 3.0.2-1.


Question: The NRC specifies that the assembly's average enrichment must meet certain percentage standards for safe operation.

Minimum Initial Assembly Avg. Enrichment wt % 235U (E) indicates the required concentration of Uranium-235 in the fuel assemblies used in BWR reactors.

The technical specifications for maintaining cooling systems focus on various environmental impacts and do not directly relate to nuclear fuel enrichment.


Question: proprietary information policies

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

The attached documents are bracketed and marked to indicate the bases for withholding.


Question: What are the prerequisites for categorization in nuclear licensing?

The categorization prerequisites include a systematic identification and evaluation of the technical requirements that apply to varying levels of safety significance, enabling informed decisions during the regulatory process.

The NRC evaluates applications based on a comprehensive assessment of all technical specifications and operational practices, ensuring that safety is maintained throughout the lifecycle of the nuclear facility.


Question: ASME Class 3 components supports

Supports for ASME Class 3 Piping and Components (Support members; welds; bolted connections; support anchorage to building structure) | Structural Support | Stainless Steel Bolting | Air - Indoor, Uncontrolled | Loss of Preload | ASME Section XI, Subsection IWF (B.2.1.30) | III.B1.1.TP-229 | 3.5.1-087 | A


Question: To comply with the industry’s requests, the recommendations involve monitoring strategies for the various aging mechanisms affecting PWR internals.

To support the industry in preparing PWR internals aging management programs to satisfy license renewal commitments.

The team developed screening criteria with susceptibility levels for the eight postulated aging mechanisms relevant to PWR internals and their effects and updated those screening criteria based on the latest research and the need to extend operation for SLR.


Question: loss of ADS initiation capability

1 hour from discovery of loss of ADS initiation capability in both trip systems

Not applicable when trip capability is not maintained.


Question: impact on transportation during nuclear construction

The TVA proposes to use two access roads to the site during most of the construction program, one via an existing road south of the plant site and a new permanent access road from U. S. 72 north of the site.

Using the percentages discussed above, estimates were prepared by the staff for selected employment levels (1000 men and 2000 men) to provide typical figures for the Jackson County area regarding housing demand, school-age children, and total population influx.


Question: The NRC requires licensees to provide longitudinal and short transverse CVN data for A-36 steel used in supports, ensuring adequate fracture toughness.

The four original samples of material were charpy V-notch tested at 80°F and 125°F in the short transverse direction (through thickness direction) to supplement the information obtained by the longitudinal direction testing that was previously provided to the NRC.

An analysis of the scrap material available from both Unit 1 and Unit 2 structures has shown the only material that will satisfy the criteria stated above are the ASTM A-36 W14 x 426 lb beams from pieces 21 through 26 of the Unit 1 steam generator structures.


Question: The NRC staff's review will assess whether the environmental report fulfills the regulatory requirements outlined in 10 CFR Part 51.

The objectives of the NRC staff's readiness assessment are to review the draft application for a multi-site service provider license for remediating abandoned uranium mine waste using high-pressure slurry ablation (HPSA) technology.

The readiness assessment will take place virtually through an online reference portal.


Question: The NRC guidelines stipulate that licensees must conduct inspections to assess the integrity of the piping systems due to potential aging mechanisms.

Piping, piping components | Leakage Boundary | Carbon Steel | Steam (Internal) | Loss of Material | Water Chemistry (B.2.1.2) | VIII.B2.SP-160 | 3.4.1-014 | B

The table outlines various components of the main steam system, including details on material and intended function.


Question: What is the purpose of the Licensing Practices and Procedures (G-109) Course?

The NRC provides funding to the Agreement States for training and the associated travel.

Students selected to attend the March 31-April 3, 2025, Licensing Practices and Procedures (G-109) Course in Chattanooga, Tennessee are provided in Enclosure 1.


Question: detection limit in soil analysis

Westinghouse Electric Company, LLC is located in Columbia, South Carolina.


Question: aging effects in nuclear systems

The following aging effects associated with the Hypochlorite System components require management: Cracking, Hardening or loss of strength, Long-term loss of material, Loss of material, Loss of preload.

The materials of construction for the Hypochlorite System components are: High-strength steel, Nickel Alloy, Polymeric, Stainless Steel, Steel.


Question: In the event of operational disturbances, it is critical to verify the functionality of control air systems, particularly in maintaining compressor operations.

CHECK 480V power available to at least two C&S&S compressors.

ENSURE Emergency Purge Shutoff valve NOT closed as indicated on Dryer Panel.


Question: What is the procedure for retesting a Main Turbine Stop Valve at nuclear power plants?

Perform Retest on a Main Turbine Stop Valve IAW S2.OP-PT.TRB-0003

The NRC requires licensees to maintain accurate records of personnel training and qualifications.


Question: What is the role of AC electrical power distribution systems in nuclear facilities?

The onsite Class 1E AC electrical power distribution system for each unit is, for the most part, divided into redundant and independent AC electrical power distribution subsystems (Divisions 1 and 2).

Charging current then reduces exponentially during the remainder of the recharge cycle.


Question: What are the accessible and inaccessible portions of canister confinement boundaries?

| Accessible Portions                              | Inaccessible Portions                      |
|--------------------------------------------------|-------------------------------------------|
| Standardized Advanced NUHOMS® System Canisters   | Inner Bottom Cover Plate (IBCP)          |
| NUHOMS® 24PT1                                   | IBCP to Shell weld                        |
| - Shell                                         | Siphon and Vent block (S&VB)             |
| - Shell long seam welds                          | S&VB Cover Plates                         |
| - Shell circumferential welds, if present       | Inner Top Cover Plate (ITCP)             |
|                                                  | ITCP to shell weld                        |
|                                                  | S&VB Cover to S&VB welds                 |
|                                                  | S&VB to Shell weld                        |

The NRC requires extensive documentation for the operational procedures of nuclear facilities.


Question: Nuclear Regulatory Commission licensing process

Authorization for possession and use of byproduct materials at temporary job sites under exclusive federal jurisdiction shall be obtained from the Nuclear Regulatory Commission by either: 1. Filing a Nuclear Regulatory Commission Form-241 in accordance with 10 CFR 150.20(b); 2. Applying for a specific Nuclear Regulatory Commission license.

Before byproduct material is used by a specific licensee at a temporary job site in another state, authorization shall be obtained from that state if it is an agreement state, or from the Nuclear Regulatory Commission for any non-agreement state, either by filing for reciprocity or applying for and obtaining a specific license.


Question: What is the significance of archaeological site inventories near nuclear plants?

Table 3.8-3 Illinois Inventory of Archaeological Sites Entries within 6-miles of DNPS (Sheet 2 of 28) lists various archaeological sites and their statuses relevant to the regulatory considerations for the plant.

The Illinois archaeological sites are categorized primarily by their trinomials and locations, influencing local historical preservation efforts.


Question: The licensing process for nuclear materials requires thorough documentation and regulatory compliance.

LDS-XXX (XXX = NUCIDE MASS NUMBER) refers to a specific method used for categorizing isotopes based on their mass numbers, which is critical in safety and regulatory assessments.

Eckert & Ziegler Isotope Products has developed various series of isotopic products, each tailored for specific applications in medical and industrial fields.


Question: specific conductance in nuclear regulatory filings

SPECIFIC CONDUCTANCE           | 1.570    | 3.362       | U         | 24.000   | us/cm  | REG         |


Question: What are the regulatory issues for nuclear facilities at brownfield sites?

Today’s meeting will discuss topics related to Section 206 of the ADVANCE Act, 'Regulatory issues for nuclear facilities at brownfield sites.'

This meeting was noticed on the NRC’s Public Meeting Schedule page on December 30, 2024.


Question: ARITA TER regulatory process

ARITA TER stands for Automated Risk Informed Technical Assessment for Licensing that is utilized in the regulatory assessments of nuclear reactor operations.

The meeting is scheduled for Tuesday at 2 PM to discuss the latest safety protocols for all operating facilities.


Question: What are the affiliated states in the Southeastern region for nuclear licensing?

The Southeastern region includes Alabama, Florida, Georgia, Mississippi, Tennessee, and Virginia as affiliated states.

The Midwest region consists of Indiana, Iowa, Minnesota, Missouri, Ohio, and Wisconsin.


Question: What is the significance of 10CFR50 Appendix R in relation to fire safety at nuclear facilities?

This Area complies with 10CFR50 Appendix R, Section III.G.2.a.

Automatic deluge systems are installed on each of the station transformers located south of the Turbine Buildings.


Question: What are Low Head Safety Injection (LHSI) requirements in nuclear safety?

The proposed change adds operability requirements to TS 3.7.E and revises TS Table 3.7-6, “Accident Monitoring Instrumentation,” and TS Table 4.1-2, “Accident Monitoring Instrumentation Surveillance Requirements,” to add Low Head Safety Injection (LHSI) flow indication due to reclassification of the Regulatory Guide (RG) 1.97, Revision 3, variable from a Type D Category 2 variable to a Type A Category 1 variable.

The proposed change does not alter the condition or performance of equipment or systems used in accident mitigation or assumed in any accident analysis, and the safety analysis acceptance criteria are not impacted by this change.


Question: What is the purpose of the Rapid Borate Stop Valve in Reactor Control?

Locates and discusses opening 1CV175, Rapid Borate Stop Valve by disengaging clutch and rotating handwheel (counterclockwise).

CLOSE the Air Supply Isolation Valve, 12CV160 A/S, AIR SUPPLY FOR 12CV160.


Question: critical heat flux correlation

The COBRA-FLX code is employed to calculate the critical heat flux under various conditions which is essential for ensuring safe reactor operation.

Table B-3 provides a summary of local conditions derived from COBRA-FLX simulations, detailing the grid tests conducted during the assessment.


Question: What is the procedure for analyzing groundwater samples in nuclear regulatory contexts?

ACZ Laboratories, Inc. (ACZ) received 1 groundwater sample from Rio Algom Mining Company on December 20, 2024. The sample was received in good condition. Upon receipt, the sample custodian removed the sample from the cooler, inspected the contents, and logged the sample into ACZ's computerized Laboratory Information Management System (LIMS).

The sample was assigned ACZ LIMS project number L92173. The custodian verified the sample information entered into the computer against the chain of custody (COC) forms and sample bottle labels.


Question: nuclear detector setup

The detector setup included a model 44-116, with a serial number of 375009, used for measuring radiation levels at specific survey unit grids.

Logging the readings took place every minute, documenting the logged readings along with corresponding count times.


Question: What is the significance of nuclide activity measurements in nuclear regulatory filings?

The table provides detailed measurements for various nuclides, including their activity in pCi/grams, which is critical for understanding their radiological implications in nuclear safety assessments.

The reactor coolant system plays a significant role in maintaining the temperature and pressure of the reactor during operations.


Question: Uranium concentration measurements


Question: manual actuations in accident scenarios

The LGS design basis assumes manual actuations are not required for 10 minutes after a postulated accident (PA) commences (e.g., LGS UFSAR Chapter 7.5.2.4.1, Initial Accident Event).

The digital modernization project aims to improve the reliability and performance of the reactor systems.


Question: What is the identification energy tolerance for gamma direct samples?

**Identification Energy Tolerance:** 3.000 keV


Question: Framatome applies specific criteria to identify proprietary information, ensuring these details are protected.

The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

My name is Gayle Elliott. I am Deputy Director, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.


Question: The NRC evaluates the integrity of components involved in mechanical closures regularly to ensure safety and operational reliability.

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

Heat exchanger components            | Pressure Boundary                    | Copper alloy     | Raw water             | Loss of material due to general, pitting, crevice corrosion, MIC; flow blockage due to fouling | Open-Cycle Cooling Water System (B.2.11) | VII.C1.AP-179   | 3.3-1, 038             | A     |


Question: LCO 3.0.3 requirements

The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.0.3.

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


Question: What is the significance of turbine-generated missiles in nuclear safety assessments?

Per UFSAR Section 14.1.13.1, the turbine-generators are safe and reliable and the chance of a failure during normal operation, which could endanger the reactor and associated Seismic Class I nuclear systems, is extremely small.

Not applicable to the site because of location (no active or dormant volcanoes located near plant site). Based on this review, the Volcanic Activity impact hazard can be considered to be negligible.


Question: The NRC mandates the detailed reporting of all hazardous materials in certain regulatory contexts.

Cyanide, total analysis must be conducted in compliance with defined quality control standards to ensure safety and regulatory adherence.

The Iron, dissolved sample results must include accurate measurements of concentration levels to ensure compliance with specified limits.


Question: spent fuel storage racks

The spent fuel storage racks are designed to maintain the stored spent fuel in a spatial geometry that precludes the possibility of criticality.

Anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest reactivity.


Question: The assessment of design-basis accidents is crucial for ensuring safety in nuclear reactors.

NuScale Power has submitted a safety analysis report detailing the evaluation of potential design-basis accidents for its small modular reactor.

The office of NuScale Power in Corvallis, Oregon, is responsible for overseeing company operations and communications.


Question: The samples analyzed using the specified methods were evaluated for compliance with the resolution standards.

Samples 1204332689 (L1-010-106-AJGS-004-SSDUP), 485191006 (L1-010-106-AJGS-017A-SS), 485191009 (L1-010-106-AJGS-035A-SS) and 485191015 (L1-010-107-AJGS-001C-SS) did not meet the resolution requirement of having a full width half maximum of 100 keV or less for the tracer; however, the tracer yield requirement was met and the tracer peaks are within the tracer region of interest.

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


Question: The NRC aims to support the ongoing functionality of present-day nuclear power stations in the U.S.

Enable continued operation of existing U.S. nuclear reactors.


Question: centrifugal charging pump calibration

CHECK required M&T E is within its current calibration cycle as evidenced by an affixed calibration sticker:

The CCP speed increaser may be subject to erratic vibration readings induced by resonance, particularly when using the magnetic probe holder.


Question: What is gamma spectrum analysis in the context of nuclear safety?

Gamma spectrum analysis is conducted using a 500 ml Marinelli geometry detector to assess the radioactivity of samples collected from a nuclear facility.

The sample identification indicates that the sample size is 8.248E+02 grams and was taken from Dairyland_NPP.


Question: nuclear safety culture monitoring process

Nuclear Safety Culture Monitoring Process

Snapshot Self-Assessment: 50.59 71111/17T


Question: What is the energy and yield of CS-134?

CS-134 has an energy of 795.84 keV and a yield of 85.40%.

Emergency response plans must be regularly updated and tested to ensure effectiveness.


Question: What are critical tasks related to isolating a faulted steam generator?

The scenario may be terminated once Critical Task SZCT-3, Isolate the faulted S/G is complete, which corresponds with completion of EOP E-2, Appendix HH.

The crew diagnoses that the high reading on the radiation monitor is an instrument failure based on comparisons with other monitors, the rate of failure, etc.


Question: nuclear steam supply sampling system

The Nuclear Steam Supply Sampling System is designed to manage and sample the integrity of nuclear steam supply systems across various aging effects requiring management, particularly focusing on components such as heat exchangers made from carbon steel under different environmental conditions.

The Closed Treated Water Systems are crucial for maintaining the cooling parameters within the reactor system, ensuring that operational efficiency is preserved throughout the lifespan of the plant.


Question: The NRC establishes guidelines for assessing the impact of radiation on reactor materials.

The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50, Appendix H.

Heatup and cooldown requirements for the reactor vessel material are discussed in Section 5.2.4.


Question: The calculations demonstrate that the interaction ratios for components are crucial for ensuring safety under seismic loads during both Operating Basis Earthquake and Safe Shutdown Earthquake conditions.

The results of the calculation show that the interaction ratios for the strainer assembly components are below 1.0 for both Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) loads, and the strainers meet the acceptance criteria for all applicable loadings.

The interaction ratios for the components in the models are provided in Table 3.k.2-1. The Interaction Ratio is nominally 'Actual' divided by 'Allowable.'


Question: The regulatory requirements for structures, systems, and components used in non-light-water reactors involve a careful assessment of their safety significance and performance-based interactions.

As described in Regulatory Guide 1.233 “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors: “The selection of licensing-basis events (LBEs); classification and special treatments of structures, systems, and components (SSCs); and assessment of defense in depth (DID) are fundamental to the safe design of non-LWRs.

The NRC requires comprehensive assessments to ensure the safety and reliability of all reactor designs regardless of their classification.


Question: What are the licensing provisions for an Independent Spent Fuel Storage Installation?

24P-ISFSI licensed per general license provisions of 10CFR72.

Natural Gas Combined Cycle (NGCC) with Carbon Capture and Storage (CCS) system.


Question: cladding surface temperature in nuclear reactors

Figure 2-4 illustrates the relationship between cladding surface temperature and time during unit loading and unloading operations.

The schedule for the upcoming regulatory inspection was reviewed and approved by the management team.


Question: The Minimum Detectable Activity (MDA) for nuclides is determined by factors including energy and yield of radiation emitted.


Question: What are the activities involving the assessment of nuclear criticality safety?

88015 Nuclear Criticality Safety 10/20/20 (20-052)


Question: What is Shutdown Cooling in nuclear reactors?

Shutdown Cooling is lined up on the "A" loop of RHR.

When the NSO attempts to start up Shutdown Cooling, the MO 1-1001-29A valve breaker trips open and will NOT reset.


Question: What is the graded approach to emergency planning in nuclear regulations?

The NRC also considered a graded approach to EP that would take into account the magnitude of any credible hazard involved, the particular characteristics and status of a facility, and the balance between radiological and non-radiological hazards.

The NRC’s performance-based framework, inspection and enforcement program, and design-specific review process provide reasonable assurance that protective measures can and will be taken in the event of an emergency at an SMR or ONT facility.


Question: The NRC expects that multiple PCTE entities will apply for licensing under the revised regulations.

The staff estimates an average of 9 PCTE entities will apply for licenses under the new regulations.

For cost analysis considerations, the staff estimates for NRC rulemaking implementation costs of ($780 thousand) at a 7-percent discount rate as shown in Table A-8.


Question: What isotopes and their activities must be reported in the registration of devices?

| Isotope (e.g. AM241) | Activity (e.g. 100) | Unit (e.g. mCi) |

**Initial Transferor License Number (if known)**  ___________________________


Question: What are Unknown Location (UNL) components in FPRA analysis?

Components in this category are referred to as Unknown Location (UNL) components because specific cables were not identified for the components.

To understand the impact on masking, a RICT specific sensitivity is provided in PSA-ANO2-06-4B-SOU (Reference 7), Section 8.1.


Question: flood hazard reevaluation

The Flood Hazard Reevaluation Report for the Fort Calhoun Station evaluates potential hydrologic failure velocities to ensure regulatory compliance with safety standards.

The figures included in this report depict various metrics related to system operations and maintenance schedules for the Fort Calhoun Station.


Question: PWR fuel assembly average burnup

For fuel assembly average burnup greater than 45 GWd/MTU, cool time tables have been revised to account for a 5% margin in heat load.

Certificate of Compliance No. 1031
Amendment No. 14


Question: control switch operation for 1RY8000B

• CLOSE 1RY8000B by placing control switch to CLOSE


Question: What are ASME Code rulemakings in the context of nuclear licensing?

The American Society of Mechanical Engineers (ASME) develops codes and standards that govern the design, materials, fabrication, and inspection of nuclear reactor components to ensure they perform safely under various conditions.

The references section includes various documents and sources deemed relevant for the regulatory filings but does not discuss specific technical safety systems.


Question: The report discusses the uncertainties associated with the activity of various nuclides.


Question: The instrumentation responsible for detecting a loss of power must be able to sense undervoltage conditions effectively.

Loss of Power (LOP) Instrumentation provides undervoltage sensing capability of 4.16 kV Emergency Bus.

The PRA modeling is not fully developed.


Question: Cycle 24 voltage growth

Figure 3-12: Cycle 24 Voltage Growth vs. BOC Voltage


Question: The evaluation shows how potential dam failures could impact nuclear safety at the facility.

The flood hazard reevaluation report includes analysis of potential dam breaches and the resulting impact on safety systems at the Cooper Nuclear Station.

The document details the design specifications for the reactor cooling system, emphasizing its importance in maintaining operational efficiency.


Question: nuclear material transaction report

LEUN Nuclear Material Transaction Report


Question: What is the significance of hydrogeology in nuclear licensing?

The 1980 Davis & Floyd and LETCo reports provide a fairly comprehensive assessment of the hydrogeology of the Terrace Aquifer System in the area between the main plant and Sunset Lake.

The most pertinent of the references consulted are summarized below.


Question: average enrichment of nuclear fuel

Minimum Initial Assembly Avg. Enrichment wt % ≤ 35U (E) indicates the average enrichment of nuclear fuel assemblies required for operation.

Certificate of Compliance No. 1031 B2-130 Amendment No. 5, Revision No. 2 outlines regulatory approvals necessary for reactor design.


Question: PZR level control

PZR Level - GREATER THAN 17%
- Perform the following:
  1. Verify letdown isolation valves closed:
     - 2-ORV-111
     - 2-ORV-112
  2. Verify PZR heaters are off.
  3. Verify excess letdown isolation valves closed:
     - 2-ORV-113
     - 2-ORV-114

If narrow range level in any SG continues to rise, THEN stop feed to that SG.


Question: What are Items Relied On For Safety (IROFS) in nuclear regulation?

Items Relied On For Safety (IROFS) are defined and implemented to maintain each identified controlled parameter within its specified Limiting Condition of Operation, thereby ensuring the subcriticality of a system or process.

The availability and reliability of the IROFS established for criticality safety are assured through management measures which are described in Chapter 11.


Question: removable contamination survey

Monthly wipe tests are carried out by the RSC or RSO in all areas where radioactivity is stored or handled in unsealed form.

Each authorized user has a monitor available to him/her in his/her department to perform the surveys.


Question: What happens when two recirculation pumps trip at the Limerick Generating Station?

Trip of two recirculation pump motors is assumed. The vessel water level due to rapid flow coast-down is expected to reach the high level trip, thereby shutting down the main turbine and feed pump turbines and scramming the reactor. This event is categorized as an incident of moderate frequency.


Question: The operator must monitor the operation of the Emergency Gas Treatment System (EGTS) to ensure that the fans are running and the dampers are open.

CLOSE 2-FCV-67-143, CCS HX B OUTLET ERCW FLOW CNTL BYP.


Question: Reactor Protection System

Reactor Protection System is pivotal for ensuring the safe operation of the reactor by monitoring and controlling the reactor's parameters, thus preventing potential safety hazards.

The communications system plays a crucial role in facilitating the interaction between control room operators and the field personnel, ensuring effective operational management.


Question: aging management programs for nuclear systems

Aging Management Programs are designed to address the long-term loss of material, which is a key factor that requires management to ensure the continued safe operation of components like the valve body in the core spray system.

The table provides a detailed overview of component types and their intended functions within the nuclear power facility.


Question: What is COBRA-FLX and its significance in critical heat flux analysis?

COBRA-FLX is a thermal hydraulic code used to calculate critical heat flux (CHF) in the boiling water reactor (BWR) systems, which is essential for ensuring the safety and efficiency of reactor operation.

The assembly conditions tested in the non-mixing grid experiments include various parameters like pressure, temperature, and flow rates that have significant implications on the overall performance of the reactor.


Question: Reactor Thermal Power measurement

Corrected Venturi Reactor Thermal Power may be used to verify that Reactor power is within the licensed core power limits for up to 60 hours with the LEFM NOT available.

A pre-test briefing has been conducted with the Shift Manager (SM), Unit Supervisor (US), or Work Control Center SRO (WCC-SRO) in accordance with PMP-4030-OB-001, Pre-Job Briefs and Post-Jobs Reviews.


Question: loss of material due to pitting and crevice corrosion

Loss of material due to pitting and crevice corrosion could occur in aluminum piping, piping components, and tanks exposed to air, condensation, underground, raw water, or waste water environment for a sufficient duration of time.

The timing of the one-time or periodic inspections is consistent with that recommended in the AMP selected by the applicant during the development of the SLRA.


Question: The NRC mandates that licensed facilities adhere to specific standards for testing the containment systems.

Specification C-1003(Q), *Specification for In-service Inspection of the Containment Building Post-Tensioning System & Exterior Concrete Shell*, Revision 16.

Ameren Non-Conformance Report NCR 89N353-004.


Question: What are the design loads considered for seismic events in nuclear facilities?

The lattice frames and support columns are designed to withstand dead loads, live loads, seismic loads including impact and accident loads and remain within the allowable limits established in Section 6.2.2.16.

Structural loads are not transmitted through the lattice frames and columns to the containment structure.


Question: What is a nuclear reactor's core thermal power?

The core thermal power of a nuclear reactor refers to the total thermal energy produced during the fission process, measured in megawatts (MW).

The emergency response plan outlines steps to address potential safety issues that may arise during normal operations.


Question: seismic sources in nuclear regulation

The seismic sources listed include various locations with associated maximum magnitudes (Mmax) and distances, which are pertinent for assessing seismic safety in the context of nuclear regulatory evaluations.

In many situations, the operational efficiency of a reactor is influenced by the quality of its mechanical systems and maintenance procedures.


Question: What are the impacts of climate change on energy systems?

Dependence of energy systems on land & water supplies influencing development of these systems & options for reducing greenhouse gas emissions.

Climate change impacts on urban water, energy supply, transportation, & other essential infrastructure.


Question: What is the purpose of a QA program in nuclear regulation?

The Radics LLC 10 CFR Part 50, Appendix B based QA policy is established in the RadICS QAPD (QAPD-001), which was reviewed by the NRC staff during the regulatory audit in Toronto.

Assuring supplier quality during application development is the responsibility of the licensee.


Question: liquid radwaste exposure

Maximum Individual Whole-Body and Body-Organ Exposure Due to Liquid Radwaste in the North Anna Reservoir.

Process Radiation Monitoring System Counting Rates of Relevant Isotopes (Cpm/µCi/cc)


Question: nuclear diesel generator reliability

Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," July 2, 1984.

Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Implementation of Topical Report Nuclear Energy Institute NEI 06-09, “Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines,” Revision 0-A (CAC Nos. ME9555 and ME9556).


Question: NRC power reactor licensing environmental reviews

Environmental reviews are a critical component of the NRC power reactor licensing process, ensuring that potential environmental impacts are thoroughly assessed before project approval.

The North Anna 3 project is reviewed for compliance with agency standards, evaluating various factors such as construction timelines and contractor evaluations.


Question: What is the purpose of an IPS in a nuclear facility?

The IPS is part of the original plant and is essential for plant operations, providing critical functions such as water intake and debris management.

The project would involve minimal ground disturbance for installation of monorail supports and the concrete dumping container pad in a highly disturbed area that has been dedicated to plant operations since the plant was constructed.


Question: probabilistic hazard analysis

Other Hazard Analysis establishes the frequency of occurrence of different intensities of the hazard being analyzed and uses a site-specific probabilistic evaluation that is based on recent available data and site-specific information.

The other hazard plant response analysis uses a model that includes all important initiating events and other important failures caused by the effects of the hazard that can lead to core damage or large early release (or large release).


Question: FMFIA compliance and material weakness

The term "internal controls," as envisioned by the FMFIA, is synonymous with "management controls" and encompasses program and administrative areas, as well as the accounting and financial management areas.

The auditor shall provide an opinion on the fair presentation of the Principal Statements in accordance with generally accepted accounting principles.


Question: Control systems in nuclear power plants

Generic Letter 89-19, “Request for Action Related to Resolution of Unresolved Safety Issue A-47 Safety Implication of Control Systems in LWR Nuclear Power Plants, Pursuant to 10 CFR 50.54(f),” Nuclear Regulatory Agency, September 20, 1989

The Limerick Generating Station utilizes advanced systems for managing operational safety and adherence to regulatory standards.


Question: Initial License Examination

Indiana Michigan Power Company, the licensee for D.C. Cook Nuclear Plant (CNP) Units 1 and 2, is providing as enclosures to this letter the Initial License Examination based on 4 Senior Reactor Operator (SRO) (U), 6 SRO (I), and 6 Reactor Operator (RO) candidates.

The examination, currently scheduled August 2020, is being prepared in accordance with NUREG 1021, Revision 11.


Question: The analysis of soil samples for radioactive content is crucial in various environmental assessments.

Rad Gas Flow Proportional Counting is one of the methods used to measure radioactivity in samples.

The dry soil preparation method involves drying samples before further analysis, but does not assess radioactivity directly.


Question: What is the purpose of the CUF monitoring in reactor component fatigue analysis?

Therefore, the CUFs were projected for these limiting locations in lieu of projecting their transient cycles.

Operating experience at VEGP and similar units has demonstrated that the analyzed numbers of design basis transients are, in general, conservative for a 40-year life.


Question: The evaluation of matrix spike recovery should indicate the reliability of the analytical method used.

Matrix spike recovery was low, the recovery of the associated control sample (LCS or LFB) was acceptable.

Sample prep or analysis performed past holding time. See case narrative.


Question: The NRC mandates that a comprehensive Nuclear Quality Assurance Plan must be developed by the licensee.

The Nuclear Quality Assurance Plan (NQAP) is a documented program developed by the licensee that ensures safety and regulatory requirements are met during the construction and operation of the nuclear facility.

The Open-Cycle Cooling Water (OCCW) system is utilized for cooling purposes and is integral to maintaining safe operational temperatures in the plant.


Question: What are the different probability maps used to assess seismic hazards in nuclear projects?

Peak Acceleration (%g) with 10% Probability of Exceedance in 50 Years.

The NRC requires licensees to follow stringent protocols during the testing phase.


Question: emission factors for industrial incinerators

1. AP-42 Chapter 2.1, Refuse Combustion, Table 2.1-12 does not list HAP emission factors for the Industrial/Commercial Multiple Chamber Combuster source category. Therefore, the emission factors for Mass Burn/Modular Excess Air Combustors (Tables 2.1-2 and 2.1-7) are shown above. As these emission factors are for mass burning of municipal solid waste, they are believed to be conservative (i.e., they should overestimate emissions from the small incinerator operated at Westinghouse).

Therefore, the highly controlled SD/FF (spray dryer/fabric filter) and DSI/FF (duct sorbent injection/fabric filter) emission factors have been selected from Tables 2.1-2 and 2.1-7, respectively.


Question: What is the significance of safety related power sources during emergency conditions in a nuclear facility?

Incorrect because under accident conditions, a safety related power source is preferred over a non-safety related power source.

The reports of smoke in the 1/2 EDG room and voltage/frequency oscillations suggest an imminent failure of the 1/2 EDG.


Question: What analytical methods are used for gamma Ni59 in solid samples?

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


Question: What is the significance of the SAMA Category in nuclear safety evaluations?

"SAMA Category" Key: A: AC/DC, B: Core Cooling, C: FW/Condensate, D: HVAC, E: Instrument Air/Nitrogen, F: Containment Phenomena, G: Containment Bypass, H: ATWS, I: Internal Flooding, J: Seismic, K: Fire, L: Other, M: Cooling Water, N: Seal Cooling

The SLR SAMA Number provides a unique identifier for each evaluated mitigation alternative, ensuring that they can be systematically referenced throughout regulatory filings and discussions.


Question: If you have concerns about the findings in the inspection report, you must submit your comments within 30 days to the NRC.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Virgil C. Summer.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, “Public Inspections, Exemptions, Requests for Withholding.”


Question: radioactive waste management inspection

IP 84850: Radioactive Waste Management Inspection, “Standards for Protection Against Radiation,” and 10 CFR Part 61, “Licensing Requirements for Land Disposal of Radioactive Waste”

Many of these inspection guidance documents are applicable to programs outside the reactor decommissioning inspection process.


Question: What are the temperatures and pressures reported in Unit 2 containment during the emergency procedures?

The following conditions exist in Unit 2 containment: Containment temperature is 140 F, Containment hydrogen concentration is 1.8%, RCS pressure is 1200 psig.

Unit 2 has experienced a small break LOCA.


Question: What is the purpose of a containment structure in a nuclear reactor?

The containment structure is designed to prevent the escape of radioactive materials to the environment in the event of a severe accident, thereby protecting public health and safety.

During the commissioning phase, the licensing board reviewed the operational parameters and relevant safety evaluations, ensuring compliance with regulatory standards.


Question: What are the requirements for a Level 1 and Level 2 PRA under 10 CFR 50?

10 CFR 50.71(h)(1) requires that each holder of a COL under Subpart C “Combined Licenses,” of 10 CFR Part 52 shall develop a Level 1 and a Level 2 PRA and that the PRA must cover those initiating events and modes for which NRC-endorsed consensus standards on PRA exist 1 year prior to the scheduled date for initial loading of fuel.

10 CFR 52 governs the issuance of early site permits, standard DCs, COLs, standard design approvals, and manufacturing licenses for nuclear power facilities pursuant to the Atomic Energy Act of 1954, as amended, and Title II of the Energy Reorganization Act of 1974.


Question: critical heat flux in nuclear reactors

Critical Heat Flux is the condition at which the transition from a single-phase to a two-phase heat transfer occurs, significantly affecting the cooling of the reactor core.

Instrumentation to Limit Maximum Power Output ensures that the nuclear reactor operates within specified thermal limits and enhances safety by preventing overheating.


Question: significant hazards consideration

The proposed no significant hazards consideration determination states that the amendment would not significantly increase the probability of any accident previously evaluated.

The distribution list includes multiple resources and individuals, such as the NRR/DORL/LPL3/PM and RidsNrrDssStsb Resource, who are involved in the regulatory processes.


Question: nuclide energy levels

The listed nuclides and their corresponding energy levels in keV provide crucial data for understanding their behavior in nuclear processes and regulatory assessments.

The uncertainties presented in the table highlight the importance of precise measurements in nuclear safety evaluations.


Question: The NRC mandates that specific quantities of boric acid must be added based on the adjustments in power levels during reactor operations.

During the operator's action sequence, the charging must be verified to ensure that the flow rate maintains operational safety.


Question: What is the required cooling time for assemblies with average burnup between 40 and 41 GWd/MTU?

Minimum Cooling Time (years) is specified for different ranges of average enrichment of assemblies, such as for average burnup levels of 40 < Assembly Average Burnup ≤ 41 GWd/MTU.

The results indicate that various factors affect the stability of the reactor under different operational conditions, including the reactor type and operational history.


Question: The process of maintaining the integrity of samples throughout the analytic procedures is crucial for accurate results.

Chain of custody procedures are essential to ensure that samples collected for analysis remain uncontaminated and are properly tracked from the point of collection to testing.

The samples collected were documented with IDs and associated timestamps for following up on the collection process.


Question: The NRC requires licensees to identify and mitigate potential impacts on local wildlife during plant operations.

Gray wolves are categorized as state endangered species and are found in habitats such as forests, prairies, and tundras.

The table lists various species and their corresponding legal status to ensure proper conservation measures are followed.


Question: What are the critical meteorological parameters for NRC licensing in nuclear projects?

For purposes of NRC licensing, the most critical meteorological parameters are wind speed, wind direction, and atmospheric stability (which can be derived from wind speed and variability in wind direction).

The wind speed distributions from the two locations are also comparable.


Question: What is the importance of chain of custody in nuclear regulatory filings?

The chain of custody ensures that samples are securely collected, handled, and transported, maintaining their integrity for accurate analysis.

The sample collection time zone was marked as Central, and additional notes reference shipping and delivery details.


Question: The NRC mandates that licensees adhere to specific criteria to ensure the safe operation of advanced reactors.

The Natrium Advanced Reactor is designed to meet the Principal Design Criteria that ensure safety and compliance with regulatory standards.

The correspondence outlines the reviewers and stakeholders involved in the licensing process, making sure all parties are informed and engaged.


Question: The NRC requires licensees to calculate the P/T limits using data validated by RG 1.99, Revision 2.

The NRC staff performed independent calculations of the RTNDT values using the regulatory positions and methods in RG 1.99, Revision 2, and verified that the 1/4T RTNDT values reported in Table 7-2 of WCAP-18455-NP, Revision 1, and the 3/4T RTNDT values reported in Table 7-3 of WCAP-18456-NP, Revision 1, are valid for the P/T limits calculations.

The NRC staff confirmed that the data are credible using the credibility criteria defined in RG 1.99, Revision 2.


Question: The RICT Program involves an analytical process that ensures effective management of risk in technical specifications.

This level of training includes the following attributes: Specific training on the revised TS, record keeping requirements, case studies, and hands-on experience with the CRM tool for calculating RMAT and RICT.

These individuals need a broad understanding of the purpose, concepts, and limitations of the RICT Program.


Question: The yield of radioactive isotopes can vary significantly, influencing their application in different contexts.

The yield of CE-144 is 11.09%, which is an important factor when considering its use in safety and regulatory contexts.

The energy levels of various nuclides are measured in kiloelectronvolts (keV), which indicates the energy emitted during radioactive decay.


Question: eccs throttle valves

The position of throttle valves in the flow path is necessary for proper ECCS performance. These valves are necessary to restrict flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow.

This SR requires verification that each ECCS train containment sump inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion.


Question: setpoint adjustments in nuclear reactors

In the "Confirm Setpoint" box: Selects OK AND In the "Load Set" box: Verifies both "Load Ref Cmd" and "Load Reference" indicate 110% after approximately 2 minutes.

The examinee should inform you that the task is complete.


Question: What is the significance of nuclide confidence in nuclear analysis?

The nuclide confidence index threshold = 0.30 indicates the minimum confidence required for identifying a nuclide in nuclear analysis.


Question: The acceptance criteria for the NPC package are defined based on calculational uncertainty and the effective multiplication factor.

If the calculational bias \( \beta = k_c - 1 \), the bias is negative if \( k_c < 1 \), and positive if \( k_c > 1 \). Thus, the acceptance criteria may be rewritten as,
\[ 1.00 + \beta - \Delta k_u \geq k_{eff} + 2\sigma + 0.05 \]

Validation of GEMER consists of performing calculation of benchmark experiments including the area of applicable to the uranium oxides.


Question: Building 277 has been designated for specific survey activities and reference sample collection.

ANAD0200419Ni63 Building 277 interior survey locations background reference samples; M88 ACADA units never stored or used in this building

The telephone number for inquiries is 586 825 4503, and the fax number is 586 939 4140.


Question: What is risk-informed regulation in nuclear licensing?

The NRC is increasingly using risk-informed approaches to develop regulations as opposed to the prescriptive and deterministic regulations developed for the legacy reactors.

Lengthy timescales, rising costs, uncertain regulatory outcomes, and pressing competition from natural gas-fired power plants have combined to discourage investment in new nuclear fission stations in the United States.


Question: Radium isotopes analysis in soil samples

Ra 226/228 Gamma Spec Norm Library U/T Alpha Spec

Please do not dispose of samples until requested.


Question: The assessment of seismic hazards at the Natrium™ site is essential for ensuring the safety of the reactor.

An analysis was conducted to evaluate potential volcanic hazards at the proposed Natrium™ site near Kemmerer, Wyoming.

The map legend includes the distribution of various physiographic provinces, which is essential for understanding the geological characteristics of the area.


Question: AFW suction valve failure modes

AFW suction valve HV-5113, normally closed, de-operated MOV, train C also has specific failure modes, for example, it may fail closed or fail to open upon command with condensate storage tank V4-001 depleted.

8E1. None - Same as item 8A1, except only pump P4-003 available, but this is sufficient for safe shutdown.


Question: nuclear device transfer

Please see the enclosed General Licensee Registration forms documenting that the nuclear device listed under GL-705439-26 is no longer in possession of the Metropolitan St. Louis Sewer District.

In addition to the certified mail attempts, multiple phone calls were made to contact the NRC via phone but an automated "office is closed" message was received each time.


Question: radionuclide release impact assessment

SMALL impact. Water from plant uses continues to be processed and monitored in compliance with licensing and permitting resulting in SMALL impacts and do not warrant additional mitigation measures.

No impact. Issue is not applicable because MNGP uses a once-through cooling system and two MDCTs, but does not utilize cooling ponds.


Question: The NRC mandates that licensees perform surveillance to ensure the proper functioning of electrical power distribution systems.

SR 3.8.9.1 Verify correct breaker alignments to required AC, DC, and 120 VAC bus electrical power distribution subsystems.


Question: What are Manning's n roughness coefficients used for in hydraulic modeling?

Manning's roughness coefficients for the streambed are selected from the HEC-RAS Hydraulic Reference Manual (Reference 93).

Creating a calibrated HEC-HMS model for estimating surface runoff requires the refinement of an initial set of parameters used to run the model and produce preliminary results.


Question: Probabilistic Risk Assessment is a crucial aspect of the safety analysis for nuclear reactors.

Probabilistic Risk Assessment (PRA) examines the risks associated with potential events and their consequences on the operation and safety of nuclear power reactors.

The reactor's cooling system consists of multiple redundant components that ensure its efficiency under varying conditions.


Question: The NRC regulates the appeal process allowing for challenges to specific agency actions.

Pursuant to 10 C.F.R. § 2.311(c), the only agency action subject to appeal is adjudicatory memorandum and order LBP-25-01.

The storage of spent nuclear fuel requires stringent safety measures and long-term management strategies.


Question: normal residual heat removal system

Knowledge of the physical or control/protection logic relationship between the normal residual heat removal 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: (CFR: 41.7)


Question: The auxiliary building's ventilation system utilizes multiple components including a filtration unit and an energy valve to maintain safe operational conditions.

The auxiliary building ventilation system includes components such as intake, exhaust, filters, and ducts that are critical for maintaining air quality in safety-related areas.

The safety valve is a crucial device that helps prevent excess pressure from building up within the system.


Question: What is the communication protocol used for notifying the MCR in nuclear operations?

Uses three way communication via radio or phone to communicate to the MCR that the local actions of Step 5 of 0-SOI-90.02, Section 8.3.

When the performer calls the MCR to report completion of local operator actions of Step 5 of 0-SOI-90.02, Section 8.3, acknowledge report.


Question: ERO augmentation time requirements

To adequately support the elimination or extension of the two 60-minute responders, the licensee should show that two on-shift positions are identified to fill the 60-minute responder’s role to 'Notify licensee, State, local and Federal personnel [and] maintain communication.'

This technology allows for a single communicator to perform as the NRC communicator and have collateral duties throughout the additional 30 minutes until relieved by the TSC ENS Communicator.


Question: What are the actions required when a Division 1 or 2 125V DC battery charger is inoperable?

For TS 3.8.4.A, one required Division 1 or 2 125V DC battery charger inoperable, RMAs may include: Actions to increase risk awareness and control. Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established.

For preplanned RICT entry, an equipment outage schedule may be used that identifies and plans all needed resources and provides logic ties between critical activities.


Question: LOCA limits in nuclear safety

This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA \( F_q(Z) \) limits.

Reducing THERMAL POWER by \( \geq 1\% \) RTP for each \( 1\% \) by which \( F^C_q(Z) \) exceeds its limit, maintains an acceptable absolute power density.


Question: The report provides details on how to subtract the background counts from the measured peak areas in order to obtain the net values.

The net peak area is calculated by subtracting the ambient background from the original area measurements.

The FWHM (full width at half maximum) values indicate the resolution of the detected peaks, which varies depending on the energy of the radiation.