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

The COBRA-FLX critical heat flux correlation provides a means to predict local heat transfer characteristics under varying thermal-hydraulic conditions.

Table B-3 presents local conditions derived from COBRA-FLX simulations for grid tests to assess various configurations.


Question: The wiring diagram must include all necessary components and interconnections that define system operation.

The core dolly wiring schematic T3A100E840 provides a detailed representation of the electrical connections and control circuitry relevant to the operational integrity of the system.

It is essential to ensure that all personnel are trained on emergency procedures and evacuation routes in case of an incident.


Question: The steady-state LHGR limits must be adjusted based on core flow and thermal power considerations.

Steady-state LHGRss limits are provided for Framatome Fuel (Table 9). These steady-state LHGRss limits must be modified as follows:

The safety evaluation (Reference 30) concluded that the BEO-III w/ CDA solution is acceptable subject to certain cycle-specific limitations and conditions (Reference 35).


Question: What is a safety analysis report in nuclear licensing?

A Safety Analysis Report (SAR) is a comprehensive document that provides an assessment of the safety of a nuclear facility, detailing various aspects including potential hazards, safety systems, and compliance with regulatory requirements.

The reactor's primary coolant pump is designed to maintain flow during operational conditions and is tested for both efficiency and functionality under various temperature scenarios.


Question: smoothed response spectra

Smoothed response spectra are generally used for design purposes; for example, a standard spectral shape that has been used in the past is presented in RG 1.60.

Describe the population distributions within 50 miles of the site, including any seasonal or transient populations. Specify the low population zone.


Question: The NRC mandates that licensees conduct a peak analysis to assess radioactive isotopes in samples.

Peak Analysis Performed on: 6/3/2019 10:11:43 AM.


Question: The NRC established a dedicated safety review group to ensure the thorough evaluation of safety-related procedures at the facility.

In 1983, the NRC approved a change in the licensee’s organization that created an onsite safety review group (SRG) for independent review and audit.

Expertise from external organizations was needed because many aspects of the cleanup were beyond the expertise of the licensee.


Question: loss of 10400 bus

- Enter TS LCO 3.8.1 Condition A (7 days)


Question: minimum DNBR and reactivity insertion rate at full power

The minimum Departure from Nucleate Boiling Ratio (DNBR) is determined based on the reactivity insertion rate during rod withdrawal at 100 percent power, which must not fall below certain values to ensure safe reactor operation.

The maximum feedback term is critical for ensuring that the reactor remains stable and within operational limits during normal and transient conditions.


Question: How does the NRC facilitate flexibility in meeting acceptance criteria for systems?

The 1426 guidance is the -- we had a brief on that in December also, and it was the risk-informed evaluation process supporting this acceptance criteria for the ECC systems and LWRs.

So, this is just a graphical overview of what we’re going to be talking about for the remainder of the morning.


Question: We are monitoring developments related to the harvesting effort concerning Muhleberg and will keep our communications updated.

Robert Tregoning is the Technical Advisor for Materials at the U.S. Nuclear Regulatory Commission.

Please give my regards to Klaus when you see/speak to him.


Question: What are the training requirements for personnel in nuclear facilities?

The training and qualifications of craft and quality assurance personnel may be reviewed as there have been past problems with the use of inexperienced personnel.

The safety assessment included a thorough evaluation of the reactor's cooling system and its ability to handle emergency scenarios.


Question: The NRC requires comprehensive analysis of chemical constituents in licensee submittals.


Question: The assessment of the physical conditions of the survey unit included a detailed visual inspection before performing the field survey.

An FSS Supervisor performed a visual inspection and walk-down of the survey unit on September 17, 2019, prior to performing FSS.

FSS field activities were projected to take four (4) working days to complete.


Question: The importance of conducting a thorough containment analysis in reactor design cannot be overstated.

The other one I think belongs more under FFRD, which is the role of the best estimate analysis in --

Right, and some of it was just kind of teeing off some of the comments Tom had.


Question: The Automatic Depressurization System (ADS) can be triggered by either of its two divisional trip systems, and both must operate to start this safety feature.

The ADS has two divisional trip systems, and either one can initiate automatic depressurization. Each trip system has two trip logics, both of which must trip to initiate ADS.

Core Spray (CS) is a four-pump, two-loop system that is backed up by the four-loop LPCI mode of the RHR system.


Question: The analysis determined which nuclides should be included in the solution based on specific criteria.

X  = nuclide rejected by the interference analysis


Question: What is the role of LPCI pumps in nuclear safety systems?

For each low pressure ECCS actuation system, there is one push button arranged in a single channel logic.

4 instrument channels arranged in a 1-out-of-2 taken twice logic.


Question: The NRC mandates that applicants must provide safety analyses for the proposed storage solutions...

Figure 6.8b – Fully reflected damaged 5x5x6 package array: 60 kgs UO₂ + H₂O mixture, CTU-2 observed non-uniform burn (horizontal and vertical views)

The units of measurement are based on the density of materials and the specific design requirements for the construction.


Question: What are the considerations for maintaining control room integrity during a severe accident?

The containment is not degraded by a severe accident.

There are programs in which we are testing the control rooms and its leak tightness.


Question: surveillance frequency requirements

Only required to be met during CORE ALTERATIONS. One SRM may be used to satisfy more than one of the following. Verify an OPERABLE SRM detector is located in: a. The fueled region; b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.


Question: What is a cuboid in the context of nuclear engineering?

A cuboid is a rectangular solid with six faces, where opposite faces are equal and parallel, often used in modeling structural components in nuclear plants.

In various reactor types, it is essential to analyze the thermal-hydraulic behavior under postulated accident conditions.


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

The Residual Heat Removal System is designed to remove residual heat from the reactor core following shutdown and maintain the core temperature below the specified limits.

The effective date for the revisions to the system operating instruction is noted as 03-09-2022, reflecting the continuous need for updates and adherence to operational protocols.


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

Energy Calibration is performed to ensure accurate identification of energy levels of detected gamma rays, which is crucial for determining the presence and concentration of radioactive materials in a sample.

The sample size for the analysis was measured to be 9.004E+02 grams, indicating a significant amount of material for the testing process.


Question: The NRC evaluates the environmental impact of steam generator tube ruptures by analyzing mass release scenarios.

Mass flow rate is an important parameter in calculating the atmospheric mass releases resulting from intact steam generator tube failures, as shown in operational data graphs.

The Southern Company's Vogtle Electric Generating Plant is known for its advancements in reactor technology and operational efficiency.


Question: Loss of Coolant Accident (LOCA) regulations

Loss of Coolant Accident (LOCA) Inside Containment (Main Steam) is described in D3 Section 3.34 of the application.

Item No. 44 is still open and pertains to the recirculation line, without further specifics provided in this section.


Question: What are the criteria for scoping safety-related systems in nuclear license renewal?

Safety-related systems and structures are included within the scope of LR in accordance with 10 CFR 54.4(a)(1) scoping criterion. The current definition of safety-related per 10 CFR 54.4(a)(1) is:

The scoping methodology utilized for electrical and I&C systems was similar to the mechanical system-level scoping described above.


Question: The program will evaluate whether aluminum alloy components are affected by pitting and crevice corrosion due to their exposure to specific environments.

This item evaluates loss of material due to pitting and crevice corrosion in aluminum alloy insulated piping and piping components exposed to air and condensation environments.

Barrier coatings which isolate a component from aggressive environments are not credited.


Question: hydrogen water chemistry system

Engineering Report for the Hydrogen/Oxygen Storage Skid Site Location for the Hydrogen Water Chemistry System at Grand Gulf Nuclear Station (Sargent & Lundy Report No. SL-5104 dated 1/16/97).

Mississippi Department of Transportation; Mississippi, August 2002, [http://www.mdot.state.ms.us/ports/default.htm](http://www.mdot.state.ms.us/ports/default.htm).


Question: debris-flow conditions from tephra deposits

The 11.9 cm tephra-fall deposit from a YVF VEI 4–5 pyroclastic eruption does not appear to have the potential for creating debris-flow conditions for potential flood and channel conditions in the NFLMC basin.

A VEI 8 supervolcano eruption from Yellowstone caldera could potentially deposit 54–113 cm of tephra (i.e., Mastin et al., 2014) in the 67 km^3 NFLMC basin.


Question: What is the significance of the Indiana bat in environmental assessments?

The Indiana bat (*Myotis sodalis*) is a federally listed endangered species.

The IESPB also lists state-protected species on the IDNR website.


Question: NRC license conditions

We hereby agree, as the management of Heart Rhythm & Pacemaker Specialist, LLC, to all of the constraints, conditions, requirements, and commitments identified in the existing Nuclear Regulatory Commission license # 21-32248-01 for Heart to Heart, P.C.

Heart Rhythm & Pacemaker Specialist, LLC is committed to providing innovative treatments and advanced technology in cardiac care.


Question: The NRC guidelines suggest that BWR core plate rim holdown bolts must be managed to prevent stress relaxation due to thermal or irradiation effects.

GALL-SLR Report AMP XI.M9 manages loss of preload due to thermal or irradiation- enhanced stress relaxation in BWR core plate rim holdown bolts.

The nickel alloy reactor vessel internals consist of the following; core shroud support assembly, core shroud access hole covers and bolting, jet pump hold-down beams, jet pump repair hardware and core spray repair hardware.


Question: ESAS manual initiation

Condition A applies when one manual initiation channel of one or more ESAS Functions becomes inoperable. Required Action A.1 must be taken to restore the channel to OPERABLE status within the next 72 hours or in accordance with the Risk Informed Completion Time Program.

Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 2).


Question: core damage frequency in nuclear reactors

VSLOCAs have CDFs on the order of 1E-10 (BWRs) and 1E-09 (PWRs) yr⁻¹

Plant risk and safety performance have continued to improve (utilities focusing on maintenance and improvements that positively impact risk, safety, and operability)


Question: carbonate aquifers in the Valley and Ridge province

The primary carbonate aquifers in the Valley and Ridge province are made up of carbonate Mississippian, Cambrian, and Ordovician rocks. The aquifers underlie more than one-half of the Valley and Ridge province in Tennessee and are typically present in valleys and rarely on broad, dissected ridges.

The repeating lithology created by thrust faulting and the location and presence of streams within the Valley and Ridge province of eastern Tennessee, in part, results in localized groundwater movement.


Question: recirculation transport fraction

Recirculation-pool transport fractions were identified for each debris type associated with the location of its initial distribution.

This same analysis was applied for each type of debris at PBN1 and PBN2.


Question: What are the measured values for Neptunium-237 in pCi/g?


Question: What is a tectono-magmatic model in nuclear regulatory evaluations?

Following guidance in Regulatory Guide 4.26, a tectono-magmatic model can be useful in evaluating the large-scale processes that led to the occurrence of Quaternary volcanism at specific locations in the site region.

Volcanism in the ESRP, YVF, and UWRB volcanic fields originates from the thermal and mechanical effects in the upper mantle from the passage of the North American tectonic plate over the Yellowstone hotspot.


Question: The NRC mandates that all licensees ensure compliance with established safety standards during reactor operations.

The licensing process for nuclear reactors involves a comprehensive review of safety analyses, design specifications, and operational protocols to ensure public safety and environmental protection.

The analytical results presented in this appendix will help in evaluating the overall performance and reliability of the systems implemented.


Question: indicated neutron flux

The digital rod position indication system is described in Reference 4.


Question: What are the preventive actions for maintaining bolting integrity in nuclear structures?

Include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.

Revise implementing procedures that inspection of concrete structures, qualifications of inspection and evaluation personnel will be in accordance with ACI 349.3R-02.


Question: The electrical power distribution subsystems must be functional and meet certain operational standards.

Division I, Division II, and Division III AC and DC, and Division I and II AC vital bus electrical power distribution subsystems shall be OPERABLE.

In accordance with the Risk Informed Completion Time Program.


Question: What is the maximum allowable rod internal over-pressure for cladding materials?

Maximum allowable rod internal over-pressure limits are summarized in Table B-1.

A series of results are obtained for different cladding types, at beginning of life BOL and EOL.


Question: The PSEG Site has been authorized to extract significant amounts of groundwater under the permits issued by NJDEP and DRBC.

PSEG has authorization from the New Jersey Department of Environmental Protection (NJDEP) (Reference 2.4.12-22) and Delaware River Basin Commission (DRBC) (Reference 2.4.12-15) for consumptive use of up to 43.2 million gallons of groundwater per month at HCGS and SGS combined.

Groundwater withdrawals in central and southern NJ increased from 1904 to a peak in the mid/late 1970s.


Question: Reactor Vessel Aging Management

Aging Management Program: Water Chemistry (B.2.1.2) is specified for the Reactor Vessel (Shell and Welds) to address loss of material due to pitting, crevice corrosion in the reactor coolant and neutron flux environment.

Cumulative Fatigue Damage is noted as an aging effect requiring management for the Reactor Vessel (Upper Head), with the corresponding program detailed in Section 4.3 'Metal Fatigue'.


Question: The NRC oversees the evaluation of potential flooding impacts on the Sequoyah Nuclear Plant.

Table 9.2-1 shows the PMF Elevations at SQN (TRM 484.5) Resulting from Reevaluation, which is critical for assessing flood threats.

Figure 3-1 illustrates the Sequoyah Nuclear Plant layout, providing an overview of its physical structure and arrangement.


Question: recirculation pump failure analysis

When a recirculation pump ceases operation, it is crucial to evaluate the automatic control actions in the presence of postulated plant protective system common cause failures, which must be diverse from the plant protective system itself.

The reactor vessel head integrity is essential to maintain containment and prevent the escape of radioactive materials.


Question: The motor-operated valves play a critical role in ensuring the low-pressure safety injection system functions properly during various scenarios.

The motor-operated valves are essential components in the low-pressure safety injection system, ensuring prompt response during emergency situations.

Before core loading, various system characteristics are validated to ensure the overall technical safety and operational capacity of the reactor.


Question: The assessment of wind patterns is critical for the operational safety of the facility.

Period of record at GGNS site, 2001 – 2003. Data from Site Meteorological Tower (Reference 2).

The N-NE sector showed a maximum wind persistence of 19 at the Grand Gulf site, compared to 15 at Vicksburg.


Question: How to measure surface dose rates in concrete storage casks?

Measure dose rates at approximate 70-inch diameter at four points approximately on 90-degree axes.

TSC mid-plane – approximately 92 inches from bottom. Measure dose rates at four target points (approximately 0, 90, 180 & 270 degrees) on the mid-plane.


Question: worst-case duct breaks

RC-34-MRCL-A4R1 | Worst Case Duct Breaks | 29" | Small |  7.75 |  8.02 |

The analysis includes various materials such as mineral wool and qualified epoxy for thermal insulation.


Question: The safety systems must ensure accurate measurements of all relevant variables under specific conditions.

The CSS inputs are derived from signals that are direct measures of the desired variables.

The Hydrogen Skimmer System prevents hydrogen pocketing in closed Containment spaces and compartments.


Question: The licensing process for nuclear facilities is governed by specific regulations set forth by the NRC.

The licensing of a nuclear facility involves a comprehensive review process conducted by the NRC to ensure compliance with safety standards and regulations.

The plant's power output varies with changes in reactor temperature and coolant flow rates, impacting overall efficiency.


Question: environmental report for nuclear projects

The Environmental Report describes the project area, which includes details on occupied residences and public buildings in relation to the controlled area as defined by the regulatory framework.

The base map utilizes USGS 1:24,000 quad data and includes drawing coordinates specific to the NM83-WF region.


Question: What are the key components of nuclear licensing safety assessments?

The safety assessment for nuclear licensing must include a thorough evaluation of the potential radiation exposure, operational procedures, and emergency response plans under various operating conditions.

The agreement between parties was finalized following extensive negotiations, ensuring that all terms were clearly understood and accepted.


Question: nuclide identification

The identified nuclides include K-40, CS-137, PB-212, BI-214, and RA-226, with specific measurements of confidence, energy, yield, and activity as recorded in the report.

Errors quoted at 2.000sigma indicate the statistical uncertainty in the measurements, which is crucial for ensuring the reliability of the data presented.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves several key steps, including pre-application consultations, the submission of a safety analysis report, and a thorough review by the Nuclear Regulatory Commission (NRC) to ensure compliance with safety regulations.

The cooling system is designed to remove heat from the reactor and ensure safe operational conditions, utilizing both natural and forced circulation methods.


Question: fluoride concentration in groundwater

Fluoride concentrations in groundwater were recorded with varying levels across multiple monitoring wells, providing crucial data for assessing environmental safety.

The Inferred Bottom of Bluff depicted in the report refers to geological features that are vital in hydrogeological studies but are unrelated to chemical concentrations.


Question: What are the aging effects managed under the PWR Vessel Internals Program?

This program is used to manage the effects of age-related degradation mechanisms that are applicable to the pressurized water reactor (PWR) reactor vessel internal (RVI) components. These aging effects include: 1. Cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cycle loading. 2. Loss of material induced by wear. 3. Loss of fracture toughness due to thermal aging and neutron irradiation embrittlement. 4. Changes in dimensions due to void swelling or distortion. 5. Loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

The staff approved found the augmented updated inspection and evaluation (I&E) guidelines and criteria for PWR RVI components acceptable, as documented in the staff’s safety evaluation of April 25, 2019 (ADAMS Accession No. ML190180001), and approved the use of MRP-227, Revision 1, for PWR-specific design bases in the staff’s letters to the EPRI MRP dated February 19, 2020 and July 7, 2020 (ADAMS Accession Nos. ML2000D6152 and ML201745149).


Question: What is a safety-related system in nuclear licensing?

A safety-related system is a system that is designed to function in case of abnormal operational conditions to ensure the protection of public health and safety.

The reactor's cooling system operates by circulating water through the core to absorb heat, but it must also maintain structural integrity under high pressure.


Question: The loss of preload in components due to thermal or irradiation-enhanced stress relaxation is critical in the management of reactor internals.

Loss of preload due to thermal or irradiation-enhanced stress relaxation

The reactor coolant system plays a crucial role in maintaining safe operation by removing heat from the reactor core.


Question: What is the process for renewing nuclear power plant operating licenses?

The Federal Register 78:37282 discusses the revisions to the environmental review for renewal of nuclear power plant operating licenses, outlining the regulatory framework and requirements that must be satisfied before renewal is granted.

The National Pollutant Discharge Elimination System governs the discharge of pollutants into U.S. waters and is a crucial component of environmental protection policy.


Question: radiological survey

A radiological survey is performed to identify and measure the levels of radiation in specific areas, ensuring compliance with safety regulations.

The instrument data for monitoring equipment is critical for maintaining accurate operational parameters and ensuring reliability in measurements.


Question: LACBWR Circulating Water Discharge Pipe

Assessment of the LACBWR Circulating Water Discharge Pipe Final Status Survey Data for Detection Efficiency and Detector Background

Termination of License: 10 CFR 50.82(a)(6) and 50.82(a)(7)


Question: anticipated research in nuclear licensing

ACRS has an apparent interest in the scope of our anticipatory research, and expressed some concern about a lack of that in DRA.

Your action is to provide the input for your branch following the attached outline and instructions.


Question: The NRC mandates that any changes to the manual materials need to be accurately reflected in the records of the licensees.

The following changes have occurred to the hardcopy or electronic manual assigned to you.

User information, including name, employee number, and address, should be verified regularly.


Question: ACR-700 reactor design

The ACR-700 is designed by Atomic Energy Canada Limited (AECL) and is based on the CANDU 6 design. The ACR-700 is a 1983 MWh light-water-cooled, heavy-water-moderated reactor.

These adjustments, with considerations of the features of the other reactor designs, form the basis of the effluent source terms presented in Table 1.3-7 and Table 1.3-8.


Question: The design basis earthquake motion for the ISFSI pad must account for specific acceleration values during seismic events.

For Yankee-MPC and CY-MPC, the design basis earthquake horizontal acceleration at the top surface of the ISFSI pad is 0.25g in each orthogonal direction and is (0.25g x 0.667 = 0.167g in the vertical direction).

The analyzed flood condition of 15 fps water velocity and a height of 50 feet of water (full submergence of the loaded cask) are not exceeded.


Question: The average fuel enrichment must not exceed certain limits based on burnup for the WE 14x14 assembly design.

The regulatory body monitors the overall performance of nuclear facilities to ensure they meet safety standards.


Question: environmental impacts assessment

Therefore, the comparison presented in this Report is that of the Proposed Action, under which the TRISO-X Fuel Fabrication Facility (TRISO-X FFF) is constructed and would produce TRISO-X fuel while minimizing potential adverse environmental impacts, against the No-Action Alternative, under which the TRISO-X FFF is not constructed.

The weighted scores found in Appendix C of the Site Study of each candidate site for the impact categories listed in Table 2.4-1 were used to assign the foreseeable environmental impacts.


Question: What is the procedure for draining the pressurizer relief tank (PRT) in a nuclear reactor?

The Unit Supervisor directs you to return the PRT level to 85% per BOP RY-4, DRAINING THE PRESSURIZER RELIEF TANK.


Question: What are the aging effects on concrete for nuclear reactor structures?

Increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation.

Not Applicable. This item number only applies to concrete containments and BWR Mark III steel containments.


Question: Understanding how the normal residual heat removal system interacts with other safety systems is crucial for operational integrity.

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)

The engineers discussed the specifications for the containment isolation valves, ensuring full compliance with safety regulations.


Question: What are the important safety protocols regarding the operation of cold leg accumulator isolation valves?

*Underlined portion of this step is critical to minimize the chance of nitrogen injection into the RCS.*

When the first CLA isolation valve is operated this will trigger Event 1 ed06b Loss of 6.9 KV Shutdown Board 1B-B.


Question: The reactor vessel is specifically designed to allow for remote inspection through the use of ultrasonic sensors, ensuring that critical areas can be easily monitored.

The reactor vessel nozzle welds are designed to accommodate remote inspection with ultrasonic sensors. The nozzle area is tapered along the reinforced areas to ensure a smooth transition, and pipe branch locations are...

Sample stations for routine sampling of process fluids are located as shown in the equipment location and general arrangement drawings presented in section 1.2. Proper shielding and ventilation are provided at the local sample stations to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling.


Question: The NRC mandates that surveys must be performed to ensure radiological safety during maintenance activities.

Section 2.3.6 of the LTP describes a MARSSIM based radiological survey performed by RSCS on the NS Savannah exterior hull in September and October of 2019 while the ship was on drydock in Philadelphia, PA.

Five components were identified as Class A waste, and two components were classified as Class B waste.


Question: The distance from the Anna Point facility to the nearest marina is approximately 2.3 miles to the northwest.

Anna Point, located 2.3 miles NNW, features 25 wet slips and 1 ramp.

Lake Anna State Park, situated 4.3 miles NNW, provides a variety of recreational opportunities.


Question: strainer penetration testing

The test water used for fiber penetration testing had a chemical composition prototypical to PBN. The plant condition selected for testing was that of the minimum boron concentration of 0.22 mol/l and the maximum pH of 9.404.

Note that the full strainer surface area was used when applying the penetration test results to determine the total in-vessel fiber load.


Question: What is the significance of 10 C.F.R. § 54.21(b) in license renewal applications?

The Staff raised an issue related to the applicability of 10 C.F.R. § 54.21(b). That regulation requires license renewal 'applicants' to periodically update their applications.

Through NEPA, Congress tasked the NRC with evaluating the environmental impacts of a proposed major federal action before taking that action.


Question: The NRC evaluates inspection findings to assess compliance with regulatory requirements and safety significance.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements.

The results of this inspection are documented in the enclosed report.


Question: gamma spectrum analysis procedure

The gamma spectrum analysis was performed using a 500ml Marinelli geometry and involved measuring the energy and efficiency calibrations done on 7/8/2014.

The sample size was recorded as 9.154E+02 grams and the sample was taken on 8/23/2018 at 9:57:00 AM.


Question: The need to evaluate fast transients is critical when using ¹a,c to assess the performance of fuel rods.

The importance of this phenomenon is high for fast transients if ¹a,c is used as a metric for determining fuel rod failures; see Section D.4.3.

The reactor core design must accommodate various thermal and hydraulic conditions to ensure operational safety.


Question: Completion Times represent the duration allocated for fulfilling a Required Action upon recognizing a situation that necessitates adherence to an ACTIONS Condition.

The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, provided that MAGNASTOR® is in a specified condition stated in the Applicability of the LCO.

The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.


Question: The facility utilizes updated cooling towers to control water discharge temperatures effectively.

To ensure that the NPDES permit limit is not challenged, MNGP replaced its two cooling towers after many years of service with new, upgraded ones with slightly greater cooling capacity.

The electric field created by high-voltage lines can extend from the energized conductors on the lines to other conducting objects, such as the ground, vegetation, buildings, vehicles, and persons if appropriate clearances are not maintained, posing a shock hazard for the public and workers.


Question: What are the population trends in the Region of Influence surrounding the Horizon Center site?

In recent years, the ROI has grown at rates very similar to those of the State. Loudon County has experienced the most rapid population growth of the counties in the ROI, growing 24.2 percent between 2000 and 2010, compared to the other ROI counties that experienced growth rates ranging from 4.4 percent in Roane County to 13.1 percent in Knox County over the same time period.

The data used in this section are the most current and complete data identified to address each topic.


Question: The analysis of the weighted mean activity of various nuclides is crucial for understanding radiation doses in nuclear applications.

Wt mean Activity for nuclide K-40 is measured at 6.00E-01 pCi/grams, indicating its radioactive properties.

Errors quoted at 2.00 sigma indicate the level of uncertainty in the measurements across various assessments.


Question: What does hydraulic conductivity measure in groundwater studies?

Groundwater flow rate per unit hydraulic gradient.

The river surface water elevation and duration for which the low water level conditions exist at the PSEG Site.


Question: The NRC's flood hazard evaluations incorporate field data and other resources to determine risks associated with flooding at nuclear sites.

This flood hazard reevaluation report outlines the methods and data used to assess the potential flooding impacts on the facility.

The photographs included in this section provide a visual context that helps to understand the geographical layout of the surrounding area.


Question: loss of shutdown cooling

Per QCOA 1000-02, LOSS OF SHUTDOWN COOLING, step D.8.b., states if a vacuum is not present in the main condenser, increase reject flow and replace water with the CRD or Condensate systems.

- **Plausible** because tripping recirc pumps removes the pump heat addition into the core.


Question: regulatory submittal for nuclear facilities

The regulatory submittal process for nuclear facilities involves the submission of detailed documentation that demonstrates compliance with safety standards and regulations set forth by the NRC.

The total volume of water used in the cooling system is calculated based on the inches of rainfall recorded over a specific period.


Question: The license amendment includes updates to the existing Facility Operating License, reflecting changes as noted in the enclosed documentation.

Replace the following page of Renewed Facility Operating License No. R-130 with the revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

The University of California-Davis operates various research facilities, which serve to advance education and experiments in the fields of science and technology.


Question: Licensees must evaluate the concentration levels of radioactive materials in residential soils as part of their environmental assessments.

Consultation triggers for residential soil concentration are defined based on levels of radioactive material that may pose risks to the public health and safety.

Many recreational activities are popular in urban areas, providing communities with essential spaces for recreation and socializing.


Question: What is the significance of average daily maximum temperatures in nuclear safety assessments?

Average Daily Maximum refers to the highest temperature recorded on a given day during a specified period, which can impact the operational limits of nuclear reactors.

The temperatures recorded in January show a range with an average daily maximum of -4.4, indicating potential difficulties during the winter months.


Question: laser powder bed fusion

The energy source used in LPBF interacts with the powder bed and melt pool in various ways depending on the total power applied, power distribution, and processing environment.

These parameter control changes are significant technological challenges but offer the possibility of solving AM issues such as intra-component yield and tensile strength variation.


Question: What is the purpose of a Quality Assurance Plan in analytical labs?

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

This project has been assigned to ACZ's project number, L89124.


Question: The Plant Monitoring and Control System is tasked with overseeing the operational parameters of the NIC and ensuring proper interactions during various plant scenarios.

The monitoring and controlling of the NI are accomplished using plant controllers, networks, and Human Machine Interfaces (HMI), which are necessary to operate the plant during normal operations, including commissioning, maintenance, refueling, and accident conditions.

The I&C system does not include a priority module.


Question: aging management program

Aging management reviews are conducted to evaluate the aging effects on structures and components in nuclear facilities, ensuring that any loss of bond, cracking, or material spalling is effectively managed through the Structures Monitoring Program (B.2.3.33).

The auxiliary building serves various critical functions, including acting as a fire barrier, flood barrier, and missile barrier, which are essential for maintaining the safety and security of the reactor operations.


Question: assembly average burnup

The table provides information on the Minimum Cooling Time (years) for different ranges of Assembly Average Burnup for various fuel types, ranging from average enrichment of Uranium-235 (U-235).

Nuclear power plants are designed with multiple safety systems that protect against potential hazards and ensure operational efficiency.


Question: How is the break-even point for regulatory actions determined when dealing with unknown economic values?

A break-even analysis is one alternative that can be used when either risk data or valuation data are lacking.

Shortcomings are inherent in the cost-effectiveness approach.


Question: The NRC restricts access to information classified under various exemptions, including those concerning national security.

The withheld information is specifically exempted from public disclosure by the statute indicated.

The withheld information includes interagency or intragency records that are normally privileged in civil litigation.


Question: device transfer regulations

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


Question: What is the measurement of the energy levels of various nuclides?

The energy levels of nuclides can vary significantly, for example, I-131 has an energy of 636.97 keV.

The NRC focuses on the safety and environmental protection in nuclear facility operations and strives for public transparency.


Question: nuclear power plant decommissioning process

After permanent reactor shutdown, nuclear power plant operators will initiate decommissioning in accordance with 10 CFR 50.82, “Termination of License” (TN249).

The potential environmental impacts of these alternatives are described in Chapter 3.


Question: This evaluation includes an analysis of the environmental impact of the proposed licensing actions for the Palisades Nuclear Plant.

The NRC is conducting an environmental review for a series of licensing and regulatory requests submitted by Holtec Decommissioning International, LLC that, if approved, would collectively support reauthorizing power operations at Palisades Nuclear Plant.

An exemption request was made to allow further development in unoccupied lands adjacent to urban areas, which is not related to the licensing of the Palisades Nuclear Plant.


Question: What is the significance of incorporating approved versions of Technical Reports?

In accordance with the guidance provided on the NRC website, we request that GNF publish an approved version of TR NEDO-11209 Revision 13, within three months of receipt of this letter.

If future changes to the NRC’s regulatory requirements affect the acceptability of this TR, GNF will be expected to revise the TR appropriately or justify its continued applicability for subsequent referencing.


Question: permeability in soil analysis

Overall high plasticity. Low permeability.

Alternating beds of poorly sorted, fine to coarse sands, silt, and clays with quartz gravels scattered throughout.


Question: What are the procedures for core loading in nuclear reactors?

Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

The applicant is required to have knowledge of the process for conducting special or infrequent tests including management responsibilities prior to and during the test.


Question: The NRC specifies requirements for minimum cooling time based on assembly average burnup limits.

### 52 < Assembly Average Burnup ≤ 53 GWd/MTU
#### Minimum Cooling Time (years)

### 53 < Assembly Average Burnup ≤ 54 GWd/MTU
#### Minimum Cooling Time (years)


Question: Devices must be registered with the NRC, which includes updating device information periodically.

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

The manufacturer name should provide details about the products offered and their applications in various fields.


Question: radiological survey requirements

Radiological surveys are conducted to assess the radiological conditions in a given area and ensure compliance with regulatory requirements.

The table displays instrument data including the model, probe serial numbers, and calibration due dates for each instrument.


Question: What are the maximum possession limits for radionuclides?

Application needs to address areas such as: Radionuclides, including maximum possession limits.

Decommissioning (including financial assurance, if required) is an essential process to ensure that all radioactive materials are properly managed after the cessation of operations.


Question: The components of the Nonsafety-Related Ventilation System are subject to various environmental factors that can affect their integrity.

The materials of construction for the Nonsafety-Related Ventilation System components are: - Carbon and Low Alloy Steel Bolting - Carbon Steel - Cast Iron - Copper Alloy with 15% Zinc or More - Copper Alloy with less than 15% Zinc - Galvanized Steel - Glass - Stainless Steel

The Nonsafety-Related Ventilation System components are exposed to the following environments: - Air - Indoor, Uncontrolled - Closed Cycle Cooling Water - Condensation - Raw Water - Treated Water - Treated Water > 140°F - Waste Water


Question: What is the significance of the Minimum Detectable Concentration in nuclear analysis?

MDC (Minimum Detectable Concentration) is a key parameter used to define the lowest concentration of a substance that can be reliably detected in a sample analysis.

The LACBWR Site Restoration Project involves various stages, including site assessment and remediation efforts to ensure safety and environmental compliance.


Question: What is the significance of the seismic risk evaluation in nuclear safety assessments?

A bounding case was developed that eliminated all seismic risk.

The aggregate CDF reduction for the evaluated categories is notable, showing a 9.0% reduction.


Question: What are the temperature expectations for the site based on local meteorology?

Mean monthly temperatures at the site may be expected to range from about 43°F in January to about 80°F in July.

The fastest mile of wind recorded at Chattanooga was 82 mph.


Question: Mountain Run fault zone

Field and aerial reconnaissance performed for this ESP application did not reveal any geologic or geomorphic features indicative of potential Quaternary activity along the Mountain Run fault zone.

Conspicuous bedrock scars in the Piedmont, an area characterized by deep weathering and subdued topography, has led some experts to suggest that the fault has experienced a Late Cenozoic phase of movement.


Question: What are the testing requirements for verifying the capacity of 125 VDC batteries?

According to the LAR, the proposed revised TS SR 3.8.6.6 states: Verify battery capacity is: - ≥ 90% of the manufacturer’s rating for 250 VDC batteries - ≥ 80% of the manufacturer’s rating for 125 VDC batteries when subjected to a performance discharge test or a modified performance discharge test.

Attachment L shows 17 load periods considered for the Cell Sizing Worksheet. Attachment M provides Load Profile for Battery D1. Please explain how the load periods in Attachment L and load profile in Attachment M were developed considering various scenarios shown in question 1.


Question: normalized scaling equation for flow rate

As such, the normalized scaling equation is determined as follows in accordance with Appendix K of Reference 6.1:

K = gainB*(gainA*A + gainC*C)

Per Section 1.0 and Input 4.1, the analytical limit associated with the differential flow setpoint function that is evaluated by this analysis is 235 gpm.


Question: nuclear plant licensing process

Docket Nos. 50-266 & 50-301 License Nos. DPR-24 & DPR-27 reference the licensing and regulatory process specific to Point Beach Nuclear Plant, Units 1 and 2.

Clinton Power Station operates with various fuel types and has undergone several upgrades to enhance its performance and safety.


Question: gamma spectrum analysis

Energy Calibration Used Done On: 7/8/2014

Peak Analysis Performed On: 6/28/2019 6:30:07 AM


Question: uncertainty in radiation measurement

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

The Qualifiers in this report are defined as follows: - ** **  Analyte is a Tracer compound - <    Result is less than value reported - >    Result is greater than value reported - BD   Results are either below the MDC or tracer recovery is low - FA   Failed analysis.


Question: What is the significance of NRC License SUA-1580?

The NRC License SUA-1580, issued to NuFuels, Inc., governs the operations related to the Crownpoint Uranium Project, including regulatory compliance and safety standards.

In January 2025, the Consolidated Operations Plan for the Crownpoint Uranium Project will outline various operational strategies and resource management protocols.


Question: What is the Exclusion Area Boundary in nuclear facilities?

The Exclusion Area Boundary (EAB) defines a designated area around a nuclear facility from which the public is excluded during operational and emergency conditions.

The MNGP Site Boundary encompasses all buildings and structures necessary for the operation and management of the nuclear facility.


Question: How does the NRC determine the validity of gamma spectrum data in safety analysis?

The efficiency calibration for the gamma spectrum analysis was performed on 7/8/2014 to ensure accurate measurements of radioactivity in the sampled material.

The facility Dairyland_NPP is known for its adherence to strict safety protocols and operational excellence.


Question: What is required in the safety assessment for non-power production facilities?

For non-power production or utilization facilities not subject to 10 CFR part 100, the assessment must provide an evaluation of the applicable radiological consequences that demonstrates with reasonable assurance that any individual located in the unrestricted area following the onset of a postulated accident, including consideration of experiments, would not receive a radiation dose in excess of 1 rem (0.01 Sv) TEDE for the duration of the accident.

The site evaluation factors identified in part 100 of this chapter must receive special attention to ensure the acceptability of the site under the required regulations.


Question: high pressure coolant injection system

High Pressure Coolant Injection System is a crucial safety system intended to maintain the core cooling of a reactor during certain accident conditions, utilizing lubricating oil for operation.

The cooling system is designed to ensure that all electronic components are adequately cooled to prevent overheating under standard operational conditions.


Question: What is the function of Strontium-90 in soil analysis?

Rad Gas Flow Proportional Counting involves analyzing soil samples for contaminants, including Strontium-90, which is measured in picocuries per gram (pCi/g).

The Collector noted that the sample was received on October 21, 2019, and was prepared using the GL-RAD-A-021 method.


Question: The licensing process for nuclear reactors involves a comprehensive safety assessment conducted by the NRC.

The NRC reviews and evaluates applications for licenses, including the technical specifications for safety-related systems, to ensure compliance with regulatory standards.

Furthermore, the facility is equipped with advanced cooling systems that enhance reactor efficiency and reduce operational costs.


Question: weathering constant in radionuclide removal

Exponential rate constant to estimate the removal of radionuclides from the surface of vegetation due to weathering. The expected value assumes a weathering half-life of 14 days. The valid range is derived from weathering half-lives of 1 day to 365 days.

Time that edible crops are exposed to contaminated water via irrigation. Crops in South Carolina average a growing time of 70 days.


Question: What is the purpose of DG-1428 in the context of nuclear regulations?

the staff decided to go ahead and revamp that DG and follow the Commission guidance to develop this type of guidance. And that’s DG-1428 which is now titled Plant Specific Applicability of the Transition Break Size.


Question: What are the economic characteristics of the region surrounding the EGC ESP Site?

The principal economic centers in the region include the cities of Bloomington-Normal, Champaign-Urbana, Decatur, and Springfield.

Regional employment trends, by county, are shown in Table 2.5-10.


Question: The disposal of Class A, B, and C low-level radioactive waste must comply with specific regulations.

Low-level radioactive waste (LLRW) is classified as Class A, Class B, or Class C (minor volumes are classified as greater than Class C). Class A includes both dry active waste and processed waste (e.g., dewatered resins). Classes B and C normally include processed waste and irradiated hardware.

Nonradioactive liquid waste at MNGP consists of wastewater from laboratories, showers, and sinks.


Question: What is the probable maximum flood (PMF) in the context of nuclear safety?

The PMF in rivers and streams adjoining SQN is determined by applying the PMP to the drainage basin of these rivers and streams adjoining SQN.

Results of the PMF analysis for the two candidate storms, the 21,400 sq.-mi. storm and the 7,980 sq.-mi. Bulls Gap centered storm, are presented in Table 9.2-1.


Question: administrative job performance measures

For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:

The emergency plan must be effectively communicated to all plant personnel to ensure compliance with regulatory requirements.


Question: What are the allowable radiation dose limits for the public from licensed materials?

The radiation dose received by individual members of the public does not exceed 1 mSv [100 mrem] in 1 calendar year resulting from the licensee’s possession and/or use of licensed materials.

Typical unrestricted areas may include offices, shops, laboratories, a nearby walkway, areas outside buildings, and non-radioactive equipment storage areas.


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

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


Question: What is the process for license termination for non-power production facilities?

(1) A licensee that permanently ceases operations must make application for license termination within 2 years following permanent cessation of operations, and for testing facilities licensed under § 50.21(c) or facilities licensed under § 50.22, in no case later than 1 year prior to expiration of the operating license.

The collection period for any shortfall of funds will be determined, upon application by the licensee, on a case-by-case basis taking into account the specific financial situation of each holder of the following licenses:


Question: What are the updated aging management criteria for reactor vessel internal components?

This SLR-ISG identifies revisions to the guidance for pressurized-water reactor (PWR) vessel internal components in NUREG 2192, “Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants” (SRP-SLR), issued July 2017 (Ref. 2), and in NUREG-2191, Volumes 1 and 2, “Generic Aging Lessons Learned for Subsequent License Renewal (GALL SLR) Report,” issued July 2017 (Ref. 3).

The NRC staff has reviewed three applications to extend plant operations to 80 years (i.e., for Turkey Point Nuclear Generating Units 3 and 4 (Turkey Point); Peach Bottom Atomic Power Station, Units 2 and 3 (Peach Bottom); and Surry Power Station, Units 1 and 2 (Surry)).


Question: What are self-reliant mitigation facilities in nuclear plants?

Lastly, when it comes to self-reliant mitigation facilities, these are facilities where we don't envision that the human would have a credible role in the fulfillment of safety functions.

Again, if fulfillment of safety functions is something that humans have a role in doing at a plant, there's going to be a driver where the onus is going to be on the applicant to demonstrate where that needs to be achieved from.


Question: What is the Process Control Program in nuclear waste management?

The Process Control Program (PCP) is described below. The PCP shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

These procedures provide for preparation and performance of refueling operations. They include procedures to disassemble components, refueling equipment preuse maintenance and checkouts, and methods and limits for performing refueling operations.


Question: The NRC requires licensees to assess and mitigate potential risks from external hazards to ensure the safety of nuclear facilities.

External hazards, such as earthquakes, flooding, and tornadoes, must be evaluated in the context of their potential impact on nuclear facility operations and safety.

The Office of New Reactors oversees pre-licensing activities to ensure that applicants meet regulatory requirements, promoting a transparent and efficient process.


Question: sensitivity analysis in cost estimation

The concept of a sensitivity analysis is discussed in Sections B.4.3 and B.7.4.5 as a subset of the uncertainty analysis.

The techniques available for estimate development are described in Section B.5 and the estimate development process itself is discussed extensively in Section B.7.4.


Question: What is the purpose of an early site permit in nuclear regulatory filings?

Dominion Nuclear North Anna, LLC (“Dominion”) submits this application for an early site permit (ESP) to the U.S. Nuclear Regulatory Commission (NRC) in accordance with the requirements of Title 10 of the Code of Federal Regulations, Part 52, (10 CFR 52) Subpart A, Early Site Permits.

Both Virginia Power and Dominion are direct and indirect wholly-owned subsidiaries, respectively, of Dominion Resources, Inc. (DRI).


Question: What is the purpose of ice condenser RTDs in nuclear reactors?

The ice condenser RTD (Resistance Temperature Detector) location is critical for effectively monitoring the temperatures within the ice condenser system, which plays a vital role in maintaining safety during abnormal operational conditions.

The angles provided in the documentation indicate various reference points, including specific measurements such as 270° and 275°, which help define the physical layout of the reactor components.


Question: What tools and equipment are necessary for the NRC exam at Watts Bar Nuclear Plant?

Tools/Equipment/Procedures Needed:
- Computer
- Frozen References
- JPM 4S Handout Package

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


Question: The aging management program for various valve bodies includes analysis for loss of material and cracking depending on the environmental conditions they experience.

Loss of Material is an aging effect requiring management for the Valve Body made of Carbon Steel in Demineralized Water (Int), as indicated in the Component Cooling Water System – Summary of Aging Management Evaluation.

The Component Type table outlines several valve body types and their intended functions based on material and operating conditions.


Question: what is a postulated LOCA in nuclear safety?

Certain technical specification LCO periods are justified based on NEDO-24708A which states that for postulated LOCAs, one low pressure ECCS (one LPCI loop or one CS loop) and ADS to depressurize is adequate to reflood the vessel and maintain core cooling sufficient to preclude fuel damage.

The environmental monitoring program involves regular sampling of ground water and surface water to ensure compliance with regulatory standards.


Question: Plant Protection System reliability analysis

The PPS Reliability Analysis provides a calculation of the availability for the Plant Protection System (PPS) for LGS.

Both the PPS input instrumentation (e.g., temperature and pressure sensors) and plant actuators (e.g., scram solenoids) are unchanged as a result of the PPS upgrade.


Question: nuclear regulatory filings

Pursuant to 10 C.F.R § 2.305, I hereby certify that copies of the foregoing 'Notice of Withdrawal of Joe Gillespie,' dated January 4, 2021, have been served upon the Electronic Information Exchange (the NRC’s E-Filing System), in the captioned proceeding, this 4th day of January 2021.

The U.S. Nuclear Regulatory Commission provides oversight of commercial nuclear power plants and other uses of nuclear materials, ensuring that they operate safely and in compliance with regulations.


Question: low water level in reactor vessel

Function 4.a/5.a – Reactor Vessel Water Level – Low Low | 6.3 15.6.5 | LOCA | 1) Automatic Initiation
 a) Reactor Vessel Water Level - Low-Low
 b) Drywell Pressure - High
 2) Manual Initiation | DBA

Function 3.e – Suppression Pool Water Level – High | 6.3 15.6.5 | LOCA | 1) Automatic
 a) Suppression Pool Water Level – High
 2) Manual Realignment | DBA


Question: What factors indicate how many containment spray pumps should be activated during a coolant loss scenario?

The interrelationship of CFCUs, containment pressure, and RWST level are monitored to determine how many containment spray pumps are operated with the goal of conserving inventory in the RWST during a loss of emergency coolant recirculation.

The examination aims to evaluate understanding of the basic concepts of reactor safety and operational procedures without delving into complex systems.


Question: What is external flood hazard analysis in the context of nuclear safety?

External flood hazard analysis estimates the frequency of external flood at the site using a site-specific probabilistic hazard analysis that incorporates the available recent site-specific information and uses up-to-date databases.

The technical elements for an other hazards, at-power PRA are the following:


Question: The NRC mandates detailed evaluations for the safety analysis requirements in licensing applications.

The safety analysis must include evaluations of the design basis and beyond design basis events to demonstrate compliance with applicable regulations.

Computers have drastically changed the way data is processed and analyzed across various sectors.


Question: guidelines for virtual participation in NRC meetings

The NRC has provided guidelines for virtual participation in meetings to ensure effective communication and engagement among participants.

All panelists will be briefed on the process and will participate in discussions regarding various safety protocols.


Question: inadvertent releases of radionuclides into groundwater

An applicant shall assess the impact of any documented inadvertent releases of radionuclides into groundwater.

The majority of the inadvertent liquid release events involved tritium, which is a radioactive isotope of hydrogen.


Question: What is the focus of safety-related systems in nuclear facilities?

The NRC emphasizes the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.

The financial assessments for the project were detailed in the budget analysis report.


Question: high temperature reactors

NUREG - “Technical Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’” - Document the staff’s technical evaluation of the 2017 Edition of Section III, Division 5 and associated Code Cases for acceptability and endorsement.

The regulatory guide will update the guidance of RG 1.87.


Question: radionuclide analysis

Table 3-2 provides off-site analysis data for various characterization samples, detailing concentrations of multiple radionuclides including I-3, C-14, and others, measured in pCi/kg.

The technical specifications for the pressure vessel integrity assessment are crucial to ensure the long-term safety and reliability of the reactor system.


Question: The integrity of the reactor coolant system (RCS) is vital to prevent any significant leaks that could affect reactor safety.

The proposed licensing basis change for the use of a risk-informed approach to evaluate the effects of debris does not alter the fuel cladding limits, or previous analysis and testing programs that demonstrate the acceptability of ECCS.

Wolf Creek developed a program plan to manage the risk of Primary Water Stress Corrosion Cracking (PWSCC) degradation in Alloy 600 components and Alloy 82/182 welds.


Question: What is the significance of Cs-137 concentration in nuclear regulatory filings?

The histogram for Cs-137 concentration presents data on the upper end values measured in pCi/g, showing the frequency of occurrences across various concentration levels.

The table outlines various upper end values without indicating any specific safety thresholds or regulatory implications.


Question: When reactor cooling system level drops below 9%, the operator must halt the auxiliary feedwater system and safety injection pumps.

NOTE: If RWST level < 9% then stop CCPs and SI pumps.

Initiates valve closure for the East RHR pump suction.


Question: What are the implications of defeating the N-42 PRM function in reactor operation?

The following annunciators may be affected by defeating N-42 PRM channel: - [67-C] N-42 OVERPOWER ROD STOP BYPASSED. - [82-E] NIS CHANNEL IN TEST. - [83-A] POWER RANGE OVERPOWER ROD WD STOP. - [83-E] POWER RANGE CHANNEL DEVIATION. - [115-C] POWER RANGE FLUX HI. - [115-E] POWER RANGE FLUX RATE HI.

The NRC requires that all modifications to reactor components are evaluated for safety impacts and compliance with existing regulations.


Question: The inspections conducted as part of the Reactor Oversight Process are crucial for ensuring the safety and security of nuclear facilities.

The Reactor Oversight Process includes a comprehensive set of inspections that are designed to evaluate the safety performance of nuclear reactors.

The reactor vessel plays a pivotal role in containing the nuclear reaction and ensuring that heat is efficiently transported to the steam generator.


Question: What is the Magnuson–Stevens Fishery and Conservation Management Act?

Congress enacted the Magnuson–Stevens Fishery and Conservation Management Act (MSA) in 1976 to foster long-term biological and economic sustainability of the Nation’s marine fisheries (TN9966).

Table 3-14 below summarizes the potential for each federally listed species discussed in this section to occur in the action area.


Question: Identify the aquatic organisms listed in the vicinity of the MNGP site.

Table 3.7-1 lists primary producers and zooplankton near the MNGP Site, including species like Gomphonema olivaceum and Keratella cochlearis.

Fish species are often surveyed to assess the ecological health of aquatic systems.


Question: wind speed distribution analysis

Figure 5: Correlation of 1-Yr and Historical Wind Speed Distributions shows the comparison of wind speed data over a one-year period against historical averages, providing insights for future energy generation forecasts.

The introduction section provides a broad overview of the proliferation of nuclear technology and the regulatory framework involved in maintaining safety standards across various reactor types.


Question: What are the requirements for demonstrating compliance with aging management of structures and components in nuclear regulatory filings?

Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

Procedure CL-AMPBD-SLI, “Selective Leaching Inspection Sample Basis Document,” Revision 0, discussed performing metallurgical analysis of components in place of visual examinations and that the sample size could be reduced by a factor of three (i.e., 6.7 percent or a maximum of 9 components) if the entire material and environment population is evaluated using metallurgical analysis in lieu of visual examinations.


Question: What is the NTSP for Division III Loss of Voltage function?

The lower limit NTSP for the Division III Loss of Voltage function is as follows: NTSP_LL- Div III LOV = 71 V.

The Division I & II Degraded Voltage time delay function has an upper and a lower Allowable Value and no Analytical Limit.


Question: The procedure for evaluating soil samples in environmental assessments includes a mandatory analysis of select samples for quality assurance.

The implementation of quality control measures as referenced by LTP, Section 5.9 and LaCrosse Solutions LC-QA-PN-001, Final Status Survey Quality Assurance Project Plan (QAPP) (Reference 11) includes the collection of a soil sample for “split sample” analysis on 5% of the soil samples taken in a survey unit with the locations selected at random.

For this Class 1 excavation survey unit, the “Investigation Levels” for area scanning and soil sample measurement results are those levels specified in LTP, Table 5-16, and are reproduced below in Table 5-5.


Question: The data includes wind speed and direction categorized by atmospheric stability classes over a defined period.

**Stability Class:** A—Elevation: 150 Feet


Question: nuclear power plant candidates


Question: What is a transportable storage canister (TSC)?

The TRANSPORTABLE STORAGE CANISTER (TSC) is the welded container consisting of a basket in a welding composed of a cylindrical shell welded to a baseplate. The TSC includes a closure lid, a shield plate (optional), a closure ring, and redundant port covers at the vent and the drain ports. The closure lid is welded to the TSC shell and the closure ring is welded to the closure lid and the TSC shell. The port covers are welded to the closure lid. The TSC provides the confinement boundary for the radioactive material contained in the TSC cavity.

STORAGE OPERATIONS include all licensed activities that are performed at the ISFSI following placement of a STORAGE CASK containing a loaded TSC at its designated storage location on the storage pad.


Question: Core Heat Flux measurement

Core Heat Flux (fraction of nominal) indicates the heat transfer rate from the nuclear core expressed as a fraction of its design nominal value, with measurements recorded at various time intervals.

The Core Inlet Temperature (Deg.F) is crucial for assessing the thermal performance of the reactor coolant system, with specific values taken at regular time intervals.


Question: The regulatory framework established by the NRC mandates that all nuclear facilities evaluate their safety protocols regularly.

The NRC's regulations stipulate that licensees must conduct thorough safety assessments to ensure that the reactors operate within the established safety limits.

In a typical processing plant, raw materials undergo various transformations before reaching the final product, which then is packaged for distribution to consumers.


Question: What is the role of CVS letdown isolation valves in nuclear safety?

The Actions of TS 3.3.16 Action B do not provide compensatory actions specifically applicable to supporting maintaining RCS inventory in the event of a LOCA, which is the safety function provided by the CVS letdown isolation.

These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room.


Question: What is the impact of a higher trip set point on breaker performance?

They evaluated thermal heating to conductors assuming the control circuit did not function correctly with the higher trip set point.

Design drawings provided sufficient information for the licensee to identify issues during installation, classified as a performance deficiency.


Question: What is a refueling pool trash rack in nuclear reactors?

Figure 3.I.4-1 shows a diagram of the refueling pool trash rack, which is an essential component used to capture and manage debris that may accumulate in the spent fuel pool during refueling operations.

The reactor core is designed to operate under specific thermal and hydraulic conditions to maximize efficiency and safety.


Question: The system must ensure the operation of the RHRSW pumps to maintain proper cooling flow.

Ensure available RHRSW pumps in RHRSW Loop A(B) are running:
  - RHRSW PMP 10P-1A(B)
  - RHRSW PMP 10P-1C(D)

Recognize / report failure of valve(s) to open (148A & 149B)


Question: The NRC will consider request for assistance in communication and other support during meetings.

Determinations on requests for reasonable accommodation will be made on a case-by-case basis and the agency may choose between which effective accommodations to provide.

If you have difficulty understanding and/or communicating in English, free language assistance or other aids and services may be available, upon request.


Question: The NRC must approve any new measurement uncertainty systems implemented with AREVA fuel.

In addition, application of the ARCADIA® code system to a new uncertainty measurement system(s) would require review and approval by the NRC staff prior to implementation.

The benchmarking demonstrates acceptable results, which provide evidence that ARCADIA can successfully model the Byron and Braidwood cores and that this ability can also be extended to include the use of GAIA fuel due to its similarities to the other fuel designs used in the topical report.


Question: The Commission encourages transparency and active engagement from the public during the licensing process.

The NRC’s regulatory framework aims to ensure efficient communication with licensees and stakeholders throughout the licensing process.

The NRC Meeting Details provide an overview of the agenda and participants attending the upcoming discussions.


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

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

The NRC staff will conduct the audit virtually via Teams using a licensee-established electronic portal available to NRC staff from approximately December 16, 2024, through April 25, 2025, with formal audit meetings to be scheduled during this period as needed.


Question: SG heat transfer bias

NuScale will address the impact of SG heat transfer bias on potential SG overfill conditions with the response to audit question A-NonLOCA.LTR-60.

Markups of the affected changes, as described in the response, are provided below.


Question: What are the automatic operations of the Standby Gas Treatment System?

Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Fan start.

The plant is operating at 100% power with the following: | Time (hh:mm) | Condition | |---------------|-----------| | 00:00 | - Annunciator L1-4-3, REACT BLDG VENT RAD MONITOR OFF NORMAL, alarms. <br> - Reactor Building Exhaust radiation monitor 12 indicates downscale. |


Question: steam generator tube leak

Unit 1 has developed a S/G tube leak and Unit shutdown is in progress. Treat all Unit 1 leaks as radioactive.

The Operations Manager will ensure that all safety protocols are followed during the notification process.


Question: What is the significance of the heating rate in the context of zirconium-steam reaction kinetics?

A maximum heating rate of 15 K/sec. indicates that an autocatalytic oxidation reaction commenced.

The recommendation to have “correlations…based on the critical review of experimental data…and their verification against rod bundle experiments” should certainly be applied to zirconium-steam reaction kinetics at temperatures below the 10 C.F.R. § 50.46(b)(1) PCI limit of 2200°F.


Question: unconsolidated-undrained compression test

Unconsolidated-Undrained Compression Test: 93.0

ST-11A has a water content of 27.4% and is classified as SM.


Question: What procedures must be followed when the 2A EDG experiences low lube oil pressure?

For Part 2: At 01:00, the crew is in LCO 3.8.1.1 REQUIRED ACTION d. However, at 04:00, LCO 3.8.1.1 REQUIRED ACTION d is no longer applicable and LCO 3.8.1.1 REQUIRED ACTION b and h are now applicable.

A. Incorrect but plausible. Part 1 is correct. For Part 2, the candidate could incorrectly interpret LCO 3.8.1.1 to allow a MODE change if a risk assessment is performed.


Question: What are the environmental conditions that promote the growth of Legionella bacteria?

**Legionella** optimally grow in stagnant surface waters containing biofilms or slimes that range in temperature from 95°F to 122°F (35°C to 45°C), although the bacteria can persist in waters from 68°F to 122°F (20°C to 50°C) (AWT 2019-TN8518).

Research studies have shown that Salmonella species can survive for several months in water and in aquatic sediments (Moore et al. 2003-TN8515).


Question: NMSS Radiation Safety Manual

Per NMSS Policy and Procedures 7-01, Revision 1, “NMSS Radiation Safety Manual,” dated April 10, 2017 and Management Directive 10.131, “Protection of NRC Employees Against Ionizing Radiation,” dated July 11, 2018, the NMSS radiation safety officer (RSO) should coordinate an audit of the NMSS radiation safety program every two years.

The audit will have a duration of approximately 2 days and will take place during the month of March or April, depending on NRR RSO/ARSO availability.


Question: Tritium measurements in nuclear regulatory filings


Question: What does NEI 10-04 provide guidance on?

NEI 10-04, “Identifying Systems and Assets Subject to the Cyber Security Rule,” Revision 2, provides guidance for determining whether a system and associated digital assets are subject to the requirements of 10 CFR 73.54.

To address the changes described in section 2, “Discussion” of this paper, the following NEI 10-04 Rev. 2 sections require changes:


Question: The NRC's oversight includes evaluating compliance with safety regulations and operating procedures.

The NRC conducted an inspection to ensure compliance with regulatory requirements and assess the safety culture at the facility.

The report further discusses the distribution of inspection findings to various regulatory offices, detailing the roles of individuals involved.


Question: The NRC also outlines procedures for the disposal of low-level radioactive waste.

PWR LLLW Disposition Costs Using a Combination of Non-Compact Disposal Facility and the Washington Disposal Facility (2018 dollars)

Table B-28 provides a summary of the disposal costs for BWR LLLW, detailing the financial aspects without discussing regulatory frameworks.


Question: What is the Versatile Isotope Production Reactor (VIPR)?

Atomic Alchemy, Inc. plans to submit an application for a construction permit (CP), including a preliminary safety analysis report (PSAR), for the Versatile Isotope Production Reactor (VIPR) in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50, “Domestic licensing of production and utilization facilities.”

The audit will allow the NRC staff to identify any required information that is missing from the application, identify technical or regulatory issues that may complicate the acceptance or technical reviews of the application, and become familiar with the content of the application.


Question: What is the role of telework recordkeeping in agency operations?

Develop and implement a checklist for telework recordkeeping to ensure the employee telework files are consistent.

The Notice has been extended to January 2021.


Question: inoperable APRM/OPRM channel

Inoperability of one required APRM/OPRM channel affects both trip systems and is not associated with a specific trip system, as are the APRM 2-Out-Of-4 Voter and other non-APRM/OPRM channels for which Condition B applies.

As noted, Condition B is not applicable to APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f.


Question: reactor coolant pump operability

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components remain capable of performing their intended function: (1) failure of PA032811B1, reactor coolant pump A under-frequency relay during surveillance testing on November 14, 2024

The inspectors observed and evaluated reactor and secondary off normal operations on November 5, 2024.


Question: distance measurement in regulatory filings

Distance measurements in regulatory filings often specify the proximity of various locations to a facility, stated in miles, for safety assessments and emergency planning considerations.

The data table includes locations such as nearest garden and nearest milk cow, which are used to evaluate environmental impacts near the facility.


Question: containment vent failure

Containment vent and failure can cause more than NPSH problems (e.g., harsh environments).

Examples of degraded environments include: (h) harsh environments induced by containment venting, failure of the containment venting ducts, or failure of the containment boundary that may occur prior to the onset of core damage.


Question: initial enrichment limits for nuclear fuel assemblies

Table B2-13: PWR Loading Table – Low Assembly Average Burnup Enrichment Limits provides the Min. Assembly Avg. Initial Enrichment (wt% 235U) for various Maximum Assembly Avg. Burnup values.

The tables also show various heat loads per assembly, denoting the thermal limits that fuel assemblies must adhere to during operation.


Question: nuclear regulatory inspections

PI-IR34-UT-001: Ultrasonic Examination of SG 11 Bottom Head to Cylindrical Ring Weld W-1 09/26/2024


Question: plant data qualification

C.5.1 Plant Data Qualification provides a comprehensive description of the system parameters and performance characteristics necessary for the qualification of plant data, including specific operational challenges and acceptable limits for fission products.

The Peach Bottom 2 Turbine Trip Tests are documented in section C.5.2, which highlight the thermal performance and safety margins during operational upsets.


Question: What are the recommended procedures for implementing PARs during an emergency?

RBS transmits the NMF containing recommended PARs to all Parishes.

The Parish PIOs and State EOC monitor the EAS broadcast to ascertain that the correct EAS message was aired.


Question: 10 CFR 50.69 risk-informed categorization

10 CFR 50.69 provides a framework for the risk-informed categorization of structures, systems, and components (SSCs) in nuclear power plants according to their safety significance, allowing for a more efficient and effective regulation process.

The NEI Technical Report NUMARC 91-06 offers guidelines for industry actions to assess shutdown management, focusing on maintaining safety during refueling and maintenance periods.


Question: control room design criteria for nuclear facilities

The design of the control room shall be such that it affords sufficient accessibility to both the operating and control equipment needed for safe operation, which reflects the criteria set forth in 10 CFR 50.67 and GDC-19.

The reactor's cooling system must operate efficiently to maintain optimal temperatures during both normal and emergency conditions.


Question: The flood stage-discharge frequency relationship for the Mississippi River is calculated using statistical data to predict discharge at various recurrence intervals.

Flood stage-discharge frequency relation for the Mississippi River at Grand Gulf site indicates the stage in feet msl corresponding to various discharge rates across different recurrence intervals.

The data presented in Table 2.4-12 is based on records collected between 1858 and 1973, which is critical for understanding historical hydrological patterns.


Question: What are the necessary conditions for maintaining Reactor water level during a manual scram?

Reactor water level is being controlled in a band of 53° to 95°.

A spurious signal causes all Reactor Recirculation pumps to trip.


Question: The Office of Investigations at the NRC monitors and reports on nuclear regulatory compliance activities.

This report describes Office of Investigations case activities during fiscal year 2024.

Fiscal year 2024 will see an increase in the number of investigations initiated by the NRC.


Question: We welcome feedback and suggestions from stakeholders about ongoing projects.

We have gotten a number of suggestions through here that have impacted the reports that we issued recently, so I'd encourage you, if you have an idea, please share it with us.

There's a lot of work that goes up leading up to being ready to issue those reports.


Question: intermediate air cooling system

The Intermediate Air Cooling (IAC) transfers heat to the atmosphere (the final heat sink) from the Sodium-Air Heat Exchangers (AHXs). Forced flow heat removal via IAC serves as the normal shutdown cooling system for outages.

The reactor core is designed with a negative temperature and power coefficient that is strong enough such that the reactor can accommodate anticipated transients without scram for events such as loss of primary flow, or loss of heat sink.


Question: Please clarify and define the units of temperature degrees Fahrenheit (°F) and degrees Celsius (°C) early in the document.

Please revise text to define °F and °C at first appearance on page 4, line 15 and page 7, line 27 in the SE.

The NRC staff finds the revised text acceptable and revisions have been incorporated.


Question: The NRC assesses the potential traffic impacts of developments associated with new license applications.

We have reviewed the Traffic Impact Study (TIS) and I have discussed a couple of items with the traffic engineer.

At this time, we do not have any issues with the recommendations in the submitted TIS for Triso-X.


Question: What is the role of the contract award document in nuclear regulatory processes?

The contract award document is attached. The initial PoP ended this September, but we put in an extension through Sept. 2018.

Got it, many thanks for the summary report.


Question: What is the VC Chilled Water System and its role in nuclear plant operations?

At 1355, a briefing was held in the MCR to perform the crosstie of the VS chilled water system to the 0A VC chilled water system per procedure BOP VC-20, VC Chilled Water System to VS Chilled Water System Crosstie Operation with One Train Inoperable for Maintenance, Revision 4.

The auxiliary building equipment operator noted the oil level had dropped from a previous rounds reading of 37% to 15%.


Question: An inoperable acoustic monitoring system on relief valves requires operators to quickly respond based on established procedures.

Declares the 3E Relief Valve Acoustic Monitor inoperable and enters TLCO 3.3.b Condition A, 30 days to restore channel to operable status.

The team must ensure that all scheduled maintenance is performed on time to prevent any operational disruptions.


Question: What does steam generator license renewal involve?

Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking.

In December 2015, the NRC published the final 'Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2,' as NUREG-2190, Volume 1 (ML15350A038) and NUREG-2190, Volume 2 (ML15350A041).


Question: nuclear reactor certification

The certification process for a nuclear reactor ensures that it meets all regulatory requirements set forth by the Nuclear Regulatory Commission (NRC) prior to operation.

The various state certifications required for environmental laboratories differ widely in their requirements and procedures.


Question: subsurface soil sample collection

In Class 1 open land survey units, a subsurface soil sample will be taken at 10% of the systematic surface soil sample locations in the survey unit with the location(s) selected at random.

Each soil sample consisted of approximately one (1) liter of soil. The soil media was sifted to remove stones and other debris larger than one (1) centimeter in diameter.


Question: What are the transient populations and vehicles within the 10 mile EPZ for River Bend Station?

The total transient populations and vehicles within the 10 mile Emergency Planning Zone (EPZ) for River Bend Station were documented as 4,389 and 1,736 respectively, divided across various Protective Action Sections (PAS).

The NRC's recommendations for evaluating the safety of nuclear power plants include extensive testing of reactor components and safety systems to ensure adherence to design specifications.


Question: The NRC assesses the consequences of potential breaks in pipe systems using methods applicable to different sizes of breaks.

NUREG-1829 LOCA exceedance frequencies are commonly distributed over candidate weld locations by two different methods: 1. The continuum model where small breaks are assumed to also occur on larger pipes, and 2. The DEGB model where only complete pipe ruptures can occur.

To maintain consistency with PWR risk-informed LARs related to recirculation strainer performance that assign break frequency to individual welds by pipe size, the NRC-reviewed LOCA frequencies from NUREG-1829 are utilized for the Fermi-2 risk informed analysis.


Question: What is the significance of tephra volumes in hazard models for mafic eruptions?

Tephra volumes are the most significant parameter in any tephra-fall hazard model.

Many of the basaltic volcanoes in the site region formed through fissure-type eruptions, which created near-vent accumulations of pyroclastic spatter and scoria.


Question: What is the impact of conditional LOOP probabilities in nuclear safety assessments?

A realistic with slight conservative bias slant on the consequential LOOP probabilities is utilized. Conditional LOOP probabilities can impact important sequences and is considered a potential source of uncertainty for the PRA.

The frequency for consequential LOOP is calculated for an ECCS signal and general plant trips: LOSP-ECCS = 2.4E-02 LOSP-EPRI = 1.0E-03


Question: sensitivity studies in non-LOCA analysis

The sensitivity studies performed on are in addition to those required by TR-0516-49416-P, Revision 4, "Non-Loss-of-Coolant Accident Analysis Methodology." These sensitivity studies are not part of the implementation of the methodology in TR-0516-49416-P.

As described in TR-0516-49416-P, and the associated markups in the original response, it is a characteristic of the design of the NuScale Power Module (NPM) that peak RCS pressure does not challenge acceptance criteria.


Question: The proposed changes to Technical Specifications ensure that the system can maintain safety functions even during operational challenges.

These changes do not adversely impact a safety function assumed in the safety analyses.

In conformance with 10 CFR Part 50, Appendix A, General Design Criteria (GDC) Criterion 20 – Protection system functions, Criterion 21 – Protection system reliability and testability, and Criterion 22 – Protection system independence, the PMS design is


Question: gamma spectrum analysis

Gamma spectrum analysis provides a detailed understanding of the energy distribution of gamma radiation emitted from a sample, which is essential for identifying radioactive isotopes.

The sample size used in the gamma spectrum analysis was 9.294E+02 grams, collected from the Dairyland_NPP facility.


Question: aging management evaluation for concrete

Increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation

Concrete (inaccessible areas): below-grade exterior; foundation


Question: measured temperatures at nuclear sites

Measured Temperatures 100 ft Below the Clinton Lake Dam (1994-2000)


Question: What is radiological survey in the context of nuclear safety?

A radiological survey involves the systematic measurement and evaluation of radiological conditions in a designated area to ascertain the presence of radioactive materials and ensure safety compliance.

The instrument data table lists various instruments used for radiological measurements, including their model numbers and calibration due dates, but does not provide information about their operational context or application.


Question: What is a plant Probabilistic Risk Assessment (PRA)?

The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2.

Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. [###], or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.


Question: What are the intended functions of safety-related systems in nuclear facilities?

Provide physical support, shelter, and protection for safety-related SSCs. 10 CFR 54.4(a)(1)

The component types that require aging management review are indicated in the below Table 2.4-3, Diesel Generator Buildings.


Question: core stored energy

Margin in core stored energy (+20 percent)

The SATAN-V code uses a control volume approach to model the behavior of the Reactor Coolant System resulting from a large break in a main coolant pipe.


Question: What is a Marinelli geometry used for in gamma spectrum analysis?

In gamma spectrum analysis, the 500ml Marinelli geometry is utilized for efficient detection of gamma radiation in liquid samples.

The peak analysis report provides specific details about the channel range and the time at which the analysis was performed.


Question: powder hopper control in LPBF

QC of the powder hopper consists of vertical position control and preventing or minimizing contamination.

Seals in the LPBF machine should also be periodically inspected and replaced to prevent contamination.


Question: aging management review for electrical components

Table 2.4-8 presents the components subject to Aging Management Review, which include various types of concrete, caulking, metal siding, and structural supports critical to the integrity of electrical systems.

Emergency preparedness is essential for ensuring the safety of the public in the event of unforeseen circumstances affecting plant operation.


Question: The evaluation of safety measures is crucial in ensuring compliance with nuclear regulatory standards.

The safety analysis report must include a comprehensive evaluation of potential accidents and their consequences for the reactor type being licensed.

During the reactor startup process, operators need to monitor various parameters to ensure that the reactor remains within operational limits.


Question: What is the Draft Environmental Assessment in the context of nuclear licensing?

The Draft Environmental Assessment for the Palisades Restart Project evaluates potential environmental impacts associated with the proposed project and assesses its compliance with regulatory requirements.

The email from Jennifer Davis indicates that there were several recipients involved in the correspondence regarding the Palisades Restart Project.


Question: audit reporting requirements

ISU should be prepared to provide documents, calculations, and other material, as applicable, supporting the analyses and justifications for the LAR, and the SAR and any procedures supporting the LAR.

Audit activities will be conducted via teleconference and video conference supported activities, as appropriate and efficient to the gathering of information by the NRC staff.


Question: What is the average enrichment of U-235 for BWR fuel assemblies?

Minimum Initial Assembly Avg. Enrichment wt % 235U (E)


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves a series of detailed evaluations and reviews to ensure safety and compliance with regulatory standards, culminating in the issuance of a license by the Nuclear Regulatory Commission (NRC).

Maintaining accurate records is essential for the proper functioning of administrative processes in various organizations.


Question: What are the overall transport fractions for LBLOCA in nuclear reactors?

Table 3.e.6-8 provides the overall transport fractions for LBLOCA with Sprays On in the Steam Generator Compartment, detailing various debris types and their corresponding percentages across different train configurations.

The NRC has established various guidelines to ensure the successful accomplishment of reactor decommissioning projects, emphasizing the importance of safety and environmental protection during the process.


Question: very low-level waste regulations

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

NRC is claiming this reinterpretation is 'voluntary.' How will we ever know which landfills and which states are doing it?


Question: What procedure is followed for starting the 1/2B Control Room HVAC?

Place the 1/2B Control Room HVAC in operation locally per QCOP 5750-09 step F.3. Contact Mechanical Maintenance when ready for the refrigerant check.

The examinee needs to fill in the JPM Start Time when they acknowledge the Initiating Cue.


Question: What is the role of the Hydrogen Monitor in containment atmosphere analysis?

Indication of Hydrogen concentration on VB1, only.

There are indications of containment parameters on VB1, pressure, temperature, but not hydrogen concentration.


Question: The NRC's guidance on marking security-sensitive information ensures that non-public documents are effectively protected.

NRC's Regulatory Issue Summary (RIS) 2005-31 provides criteria to identify security-related sensitive information and guidance for handling and marking of such documents.

The enclosed license document is exempt from public disclosure because its disclosure to unauthorized individuals could present a security vulnerability.


Question: What types of amendments are typically submitted to the NRC?

Amendment requests generally involve changes to facilities, operations, or regulations that affect the licensee's ability to comply with safety and governing legislation.

The Department of Health & Human Services submitted various assurance filings related to medical programs and health initiatives.


Question: The total greenhouse gas emissions from the Dresden Nuclear Power Station include various sources and are measured in metric tons.

GHG calculated emissions are based on the following: Includes all emissions from DNPS, including stationary combustion sources, sulfur hexafluoride, process CO2, hydrofluorocarbon/perfluorocarbon refrigerants, diesel, and gasoline CEG-owned vehicle refueling stations, and ozone-depleting chemical refrigerants.

The workforce commuting calculations are based on the following: Statistical information from the U.S. Census Bureau (USCB) indicates that 5.8 percent of Illinois workers in the transportation, warehouse, and utilities industry carpool to work (USCB 2020d). The number of DNPS employees as of October 2022 was 717.


Question: sump strainer structural analysis

The Sump strainers were qualified using a combination of manual calculations generated in Mathcad, as well as finite element analyses using the GTSTRUDL software and the ANSYS software.

The pipings are free to expand without imposing any thermal loads on the strainers.


Question: What are the expected annual releases of radioactive materials in liquid waste from both units?

Based upon our evaluation, the radioactivity in the liquid effluent from units 1 and 2 exclusive of tritium will be less than 5 curies/year...

It was impossible for us to determine from table 3.3 the calculated annual release of radioactive materials in gaseous effluents as many of the totals shown appear to be grossly incorrect.


Question: The response acceleration spectra are crucial for assessing the seismic performance of nuclear facilities.

The response acceleration spectra represent the response of structures to seismic activity, which is vital for ensuring the safety and integrity of nuclear power plants during earthquakes.

The results of the analysis indicate a need for further investigation into the historical seismic data collected from various regional sites.