Question: limiting ART values

The limiting ART value for the Unit 1 reactor vessel at 50 EFPY is 146.9°F, which was computed for the Lower to Lower-Intermediate Girth Weld, WF154.

The above ART values (at depth of 1/4T) of the limiting belting locations at 80 years remain below 200°F, which is the Nil-Ductility Transition (RTNDT) limit specified in NRC Regulatory Guide 1.99, Revision 2, Section 3.


Question: What are MSIV isolation signals?

MSIV isolation signals have been overridden as directed by the EOPs.

Another operator has RPV pressure control on the SRVs at 800-1000 psig.


Question: maximum pressure measurements in nuclear reactors

The table outlines the maximum pressure values for various compartments within the reactor, allowing for assessments of the system's safety and performance under specified conditions.

The reactor components are designed to withstand significant thermal and mechanical stresses during operation, ensuring optimal performance and reliability in power generation.


Question: What is the purpose of a Probabilistic Risk Assessment in nuclear regulation?

The Probabilistic Risk Assessment (PRA) Model Update Process is a systematic approach used to evaluate and update the risk profiles of nuclear facilities based on operational data and safety performance indicators.

The cooling system of a reactor is designed to remove heat from the reactor core during normal operation, ensuring thermal equilibrium.


Question: DSI Voltage Distribution

Table 3-5 shows the R2R24 DSI Voltage Distribution for SG-C, detailing the number of indications across various voltage bins at the end of cycle 24 (EOC-24).

The reactor startup sequence is initiated when the systems are verified operational and safety protocols are confirmed to be in place.


Question: license renewal boundary drawings

The License Renewal Boundary Drawings detail the specific areas and dimensions relevant to the license renewal process, ensuring compliance with regulatory requirements set forth by the NRC.

The UFSAR References include critical design methodology sections that outline the architecture of the reactor, providing crucial information for safety assessments.


Question: The radiation monitors at the stack must be calibrated to ensure accurate readings of emissions.

Stack gas radiation monitor 17RM-050A and 17RM-050B indicate downscale.


Question: The NRC requires detailed evaluation of gamma spectrum data for safety assessments.

The gamma spectrum analysis performed on the sample was conducted using a 500 ml Marinelli geometry, which is essential for accurate measurement of radionuclide concentrations.

The sample was taken from the facility on 8/6/2019, shortly after the acquisition process began.


Question: written examination grading checklist

Discontinued use of Rev 11 Form ES-403-1, Written Examination Grading Quality Checklist. These steps integrated into main body of ES-4.4

Clarifications made to list of errors that the NRC staff will not accept when considering examination changes after the exam is administered.


Question: high-level waste licensing process

OPA has written a fact sheet on the high-level waste licensing process that is now posted on the agency’s web site.

The Commission determined that a hearing was called for in the public interest; it is not explicitly addressed in the NWPA.


Question: What is the Framatome fuel rod database?

The Framatome fuel rod database benefits from the merging of the databases associated with COPERNIC, RODEX4 and CARO-E3 fuel thermal-mechanical codes.

The updated regulations will take effect on January 1, 2024, requiring all licensees to submit their annual reports by the end of March.


Question: What are the benefits of the ADVANCE Act in nuclear regulation?

The Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy (ADVANCE) Act (Public Law 118-67) puts OI on a trajectory, alongside the rest of the U.S. Nuclear Regulatory Commission (NRC), the U.S. Government, and the industry, of innovation and change, giving us a once-in-a-generation opportunity to be a vital and key contributor to how energy is produced and distributed safely and securely across the United States and the world.

In reflecting on the past year, investigative deliverables have played a significant role in OI’s success. From October 1, 2023, through September 30, 2024, OI concluded 48 investigations (a 20 percent increase from FY 2023) and 34 assists to staff.


Question: I-131 Curies measurement


Question: LOCA analysis in nuclear safety

Control Room: 1  - Volume: 2.4000E+05  - Occupancy: 0  - Temperature: 1  - Pressure: 0  - Humidity: 0


Question: The NRC requires that aging management programs be in place to evaluate and address material loss in structural bolting due to corrosion.

Consistent with NUREG-1801. The DCPP Structures Monitoring AMP (B.2.3.33) will be used to manage loss of material due to corrosion of steel and galvanized steel structural bolting exposed to an atmosphere/weather environment.

The DCPP RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP (B.2.3.34) will be used to manage loss of material due to corrosion of steel structural bolting and miscellaneous structural components associated with Group 6 structures in a plant indoor air, atmosphere/weather, or submerged environment.


Question: The distance measurements for chemical transport and their relation to safety are crucial in mitigating potential hazards.

Table 2.2-18 presents the distances from various chemical locations to the power block boundary and safety-related buildings, highlighting the proximity risks associated with chemicals like gasoline and propane.

The safe distance for an explosion is determined based on the type and quantity of the material involved, evidenced by various safety regulations.


Question: mechanistic source term analysis

The evaluation of the consequences of all LBEs are supported by mechanistic source terms.

Specific assumptions of the Distributed Control System are provided in the example analyses in this report’s appendices.


Question: What materials are used for spent fuel storage spacers?

Figure 9-20. Spacer for Spent Fuel Storage

The NRC demands rigorous documentation from licensees in their safety analyses to confirm operational sustainability.


Question: The NRC's safety assessments often include evaluations of radionuclides present in waste streams.

The radionuclides measured are an important aspect of environmental monitoring and regulatory compliance in nuclear facilities.

Water quality analysis is critical for ensuring that the drinking water supply meets safety standards.


Question: At DNPS, the groundwater monitoring program is designed to adhere to federal and state regulations, ensuring the protection of water quality.

The NRC considered the issue considering the programs and procedures commonly implemented at nuclear plants, including proper chemical and waste storage and handling, secondary containment and leak detection, use of BMPs and SPCC plans, compliance with federal and state regulations and permits, and groundwater monitoring programs.

Groundwater withdrawals for the three onsite water supply wells are calculated annually.


Question: Residual Heat Removal System aging effects

The Residual Heat Removal System is crucial for maintaining the reactor pressure boundary and is subject to degradation mechanisms such as loss of material and cracking, which require management through programs like Water Chemistry and One-Time Inspections.

Piping elements are evaluated for leakage boundary integrity and may experience issues such as loss of material, but are not subject to the same aging effects as the Residual Heat Removal System.


Question: uranium soil analysis

SW846 3050B/620A Uranium Soil "Dry Weight Corrected"


Question: flow rates in aquifer restoration

The Combined TR/ER states that, in the case of land application for aquifer restoration, “the usage of water from the Madison Limestone or another suitable aquifer will be about 430 to 510 gpm.”

The 500 gpm figure of concern to Ms. Henderson does seem to be established in the Combined TR/ER.


Question: What is an adverse localized environment in nuclear safety?

An adverse localized environment (ALE) is an environment limited to the immediate vicinity of a component that is hostile to the component material, thereby leading to potential aging effects. Electrical insulation used for electrical cables can be subjected to an adverse localized environment.

Air that has been treated to reduce its dew point well below the system operating temperature and treated to control lubricant content, particulate matter, and other corrosive contaminants.


Question: What does the NRC require for the licensing of small and mobile reactors?

A regulatory approach and philosophy that enables safe, scalable deployment of one or more reactor units every week is of vital importance to not only the business model of small and mobile reactor enterprises, but to meet emission goals, satisfy energy demand, and provide baseload generation for increased grid stability.

Alternative sources of energy (e.g., diesel generators and natural gas units) emit greenhouse gases.


Question: assembly average burnup

The assembly average burnup (B) is measured in GWd/MTU, representing the energy produced per metric ton of uranium used in the fuel assembly in a nuclear reactor.

The table outlines various minimum initial assembly average enrichments of 235U in weight percent for different ranges.


Question: What is an integral frame in nuclear reactor design?

An integral frame design in nuclear reactors combines several safety components into a single structural framework that improves stability and minimizes the risk of failure in the event of operational anomalies.

The reactor core is surrounded by a robust containment structure designed to prevent the release of radioactive materials in case of an accident.


Question: All data collected in this report conforms to the requirements set forth in the analytical methods and procedures.

There are no exceptions, anomalies or deviations from the specified methods. All sample data provided in this report meet the acceptance criteria specified in the analytical methods and procedures for initial calibration, continuing calibration, instrument controls and process controls where applicable.

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


Question: The NRC is reviewing environmental factors associated with the reauthorization of the Palisades Nuclear Plant's power operations.

The U.S. Nuclear Regulatory Commission’s (NRC’s) environmental review of the potential reauthorization of power operations at the Palisades Nuclear Plant (Palisades).

The draft environmental assessment (EA) will be discussed at the upcoming Tribal Information Meeting scheduled for February 11, 2025.


Question: reactor vessel integrity

WCAP-18728-NP, Revision 5, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time-Limited Aging Analyses," June 2023

Application for Withholding Proprietary Information from Public Disclosure Pursuant to 10 CFR 2.390


Question: nuclear device registration

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


Question: What is the NRC’s Safety Culture Policy Statement?

The NRC’s Safety Culture Policy Statement became effective in June 2011. While a policy statement and not a regulation, it sets forth the agency’s expectations for individuals and organizations to establish and maintain a positive safety culture.

Failure to conduct your program in accordance with NRC regulations, license conditions, and representations made in your license application and supplemental correspondence with NRC will result in enforcement action against you.


Question: The NRC staff will conduct a virtual environmental audit followed by an on-site visit to evaluate compliance.

An environmental audit will be conducted by NRC staff virtually the week of January 22, 2024, followed by a site-visit the week of January 29, 2024.

An audit summary report will be issued by the NRC staff within 90 days from the end of the environmental audit.


Question: What is the function of the LCL subsystem in nuclear safety systems?

The LCL subsystem consists of the coincidence voting process of the partial RPS and NSSSS/ECCS actuations (i.e., one division) for functional trips and actuations respectively.

The redundant data from each BPL is 'Or-Red' prior to coincidence voting with other BPL safety actuation signals.


Question: site-specific SBT for physical protection

Under the proposed interpretation, licensees and applicants may incorporate law enforcement response into their site physical protection programs using a site-specific SBT.

The proposed interpretation seeks to provide flexibility and regulatory clarity to licensees to consider response from law enforcement as part of the physical protection program.


Question: The NRC requires that each licensee demonstrates compliance with established inspections and tests before commencing operation.

The ITAAC Notification indicates that sufficient information has been provided to ensure compliance with safety and regulatory standards.

The NRC may find it necessary to issue a new letter to address findings from ongoing inspections.


Question: dissolved radionuclides

Lead 210, dissolved is a type of analysis that assesses the concentration of Lead-210 in the water sample, typically reported in units such as picocuries per liter (pCi/L).

Radium 226, dissolved testing involves collecting samples and analyzing them under specified quality control measures to ensure accurate readings, often utilizing standard test methods like EPA 903.1.


Question: What are the gross alpha and gross beta activity levels in water samples?

Gross alpha and gross beta activity levels in the Congaree River for 1972 and 1973 are given in Table 2.8-5. These yearly average gross beta activities range from (5.1 to 5.4) x 10^-9 µCi/ml upstream of the plant discharges and from (4.4 to 5.6) x 10^-9 µCi/ml downstream while the yearly average gross alpha activities range from (0.6 to 1.1) x 10^-9 µCi/ml upstream of the plant discharges and from (0.5 to 2.0) x 10^-9 µCi/ml downstream.

Average and maximum measurements of gross alpha and gross beta activity in water samples from Sunset Lake for the period December 1972 to November 1973 are given in Table 2.8-6.


Question: Zircaloy oxidation correlations

In 2005, regarding the fact that data from isothermal tests are used for the development of Zircaloy-steam oxidation correlations, NRC stated: "For the development of oxidation correlations, limited by oxygen diffusion into the metal, well-characterized isothermal tests are more important than the complex thermal hydraulics suggested by [Robert H. Leyse]. [Robert H. Leyse's] suggested use of complex thermal-hydraulic conditions would be counter-productive in reaction kinetics tests because temperature."

TH-1 test no. 107 and PWR FLECHT run 3724, with lower PCTs at the onset of reflowed of 1154°F and 1187°F, respectively, had cladding temperature increases of 424°F and 427°F, respectively. So at temperatures where the oxidation of Zircaloy does not produce much heat, the results of TH-1 test no. 107 and PWR FLECHT run 3724 are similar.


Question: What are the primary loop piping systems in nuclear reactors?

Only applies to primary loop piping (PLP) systems and reactor coolant pressure boundary (RCPB) components whose failure could result in breaks greater than the TBS.

Scheduled Commission Briefing – March 24, 2011.


Question: fire protection system impairments

List of current fire protection system impairments, including description.

List of all Generic Letter 86-10 evaluations completed since May 2018. Include a short description and/or title of each evaluation.


Question: minimum cooling time for nuclear fuel assemblies


Question: Boiling Water Reactor definition

BWR (Boiling Water Reactor) is a type of nuclear reactor in which water is boiled to produce steam, which drives a turbine that generates electricity.

The American Society of Mechanical Engineers (ASME) develops codes and standards that are instrumental in the operational safety of nuclear facilities.


Question: What are the inspection methods for cylindrical axial welds in nuclear reactors?

If triggered, EVT-1 visual inspection of the axial weld length and 3/4-inch of adjacent base metal on a 10-year frequency; EVT-1s look for evidence of cracking (IASCC) and manage irradiation embrittlement (IE).

CSA ribs and rings are not included as assessed components for screening review in Table 3-2 of Rev. 2.


Question: The guideline states that CCS flow to the thermal barriers should be preserved as long as the CCS temperature stays below 115°F.

CCS flow to the Thermal Barriers is not required to be isolated immediately. There is a Caution in 0-AOI-13 that directs the operators to maintain CCS flow to the thermal barrier should be maintained as long as CCS temperature remains <115°F.

Appendix E aligns High Pressure Fire Protection water to cool CCP 1A-A.


Question: What are the findings related to the voltage distributions for SG-A and SG-B?

The total number of indications recorded for SG-A was 425, while for SG-B it was 558, with average voltages of 0.32 volts and 0.36 volts respectively.

Nuclear reactors require extensive safety protocols to ensure that all operations are conducted within regulatory compliance.


Question: The reactor protection system guarantees the integrity of the nuclear reactor by executing a series of actions when abnormal conditions are detected.

The ARC-100 Reactor Protection System (RPS) inserts the control rods into the core to ensure fuel integrity and to protect the reactor coolant boundary under abnormal operating conditions.

Each rod has two (2) solenoids and both solenoids must be de-energized before the rod can drop.


Question: decommissioning process of a nuclear power plant

The decommissioning process begins when a licensee informs the NRC that it has permanently ceased reactor operations, defueled, and intends to decommission the nuclear plant.

Health impacts to workers from exposure to airborne dust and methane gases.


Question: The analysis of surface water quality includes measuring parameters such as pH levels and dissolved oxygen content.

Surface water quality data indicates pH levels less than 5.0 U, which are critical for assessing environmental conditions.

Waste management practices must comply with state and federal regulations to ensure safety during operations.


Question: What is the aging management evaluation for reactor coolant system components?

Components defined as ASME Section XI components (e.g., reactor coolant pressure boundary components, core support structure components, ASME Class 2 or 3 components, including associated pressure-retaining welds) not managed by other AMR line items in GALL-SLR Chapter IV

The aging management program includes strategies to address various environmental effects on materials used in nuclear reactor components.


Question: radiological exposure during insulation modification

Due to uncertainties in radiation levels, contamination levels, and the required modification scope, it is difficult to predict the total occupational dose associated with insulation removal and/or modifications.

The requested exemption is to allow the use of a risk-informed method to demonstrate that the design and licensing bases for the ECCS are not significantly affected by.


Question: What is the significance of the Detector Current Comparator in nuclear safety systems?

Detector Current Comparator Upper Section switch placed to 'NORMAL'. Step is critical for proper restoration of upper power detector to current comparator.

The performance evaluation measures the competency of the personnel operating the reactor under various conditions.


Question: The projected cumulative dose to the population within a 50-mile radius is calculated to be 2 man-rem annually, which is significantly low compared to the natural background radiation exposure.

Radiological:  Cumulative population dose (50-mile radius):  2 man-rem per year (less than .001 percent of dose due to natural background).

In lieu of tax payments: $6,800,000 per year.


Question: The NRC mandates the evaluation of control room ventilation design inputs to ensure safety standards are met.

B/B Transmittal of Design Information (TODI) No. DIT-BRW-2019-0030/BYR-19-042, “Design Inputs for Accident Dose Analyses – Main Control Room Ventilation” Rev. 0, 01/06/20.

Individual Instruction Guide, N-AN-ENG-RADTRADERROR, Rev. 1.


Question: The analysis of safety systems is crucial to ensure operational reliability in nuclear facilities.

The safety analysis report must include evaluations of systems that are critical to preventing accidents and mitigating their consequences.

The features of modern computer systems often include enhanced processing power and increased storage capabilities.


Question: The control system includes a low fuel oil pressure alarm that indicates when the pressure drops below a certain threshold.

A low fuel oil pressure alarm is located on the diesel control panel and is activated by the following:  a. Diesel Unit 1A - pressure switch 1FDPS5000  b. Diesel Unit 1B - pressure switch 1FDPS5010  c. Diesel Unit 2A - pressure switch 2FDP5010  d. Diesel Unit 2B - pressure switch 2FDP5010  The alarm is activated if the fuel oil pressure falls to 22 psig while the engine is running. The pressure switch turns off the alarm and resets itself when the fuel oil pressure rises to 25 psig. The pressure switch is located at the engine fuel oil intake.

This indication is performed by the following:  a. Diesel Unit 1A - pressure gage 1FDPG5000  b. Diesel Unit 1B - pressure gage 1FDPG5010  c. Diesel Unit 2A - pressure gage 2FDPG5000  d. Diesel Unit 2B - pressure gage 2FDPG5010  This gage is a dual scale gage and indicates the oil pressure upstream and downstream of the engine intake duplex oil filter. Indications are given at the diesel control panel.


Question: The NRC has concluded that the proposed amendments do not raise significant safety concerns under the applicable regulatory standards.

Based on the above, I&M concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.


Question: The NRC enforces regulations that require licensees to maintain compliance with safety analysis reports.

The Nuclear Regulatory Commission (NRC) issues regulatory guidance to ensure that reactor operators comply with safety regulations and licensing requirements.

The weather conditions in Iowa can vary widely depending on the season, impacting local ecosystems and agriculture.


Question: loss of instrument air

The Steam Jet Air Ejector (SJAE) Suction valves are air operated valves that fail closed on a loss of Instrument Air.

During normal operation this would lower Heater Drain levels and Feedwater Heater outlet temperature.


Question: The NRC assesses the adequacy of operability determinations related to various systems in the plant.

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments: Unit 1 motor driven auxiliary feedwater level control valve failure to control from the control room identified on September 29, 2024 (CR 1962284; CR 1962371).

Licensee's periodic evaluation in accordance with 10 CFR 50.65(a)(3) covering the period from October 1, 2019, thru March 31, 2021 (Maintenance Rule Sixteenth Periodic Assessment Report Units 1, 2, and Common).


Question: nuclear licensing process

The licensing process for nuclear reactors involves a series of regulatory submissions to ensure compliance with safety standards and operational protocols as outlined by the NRC.

The performance of wind turbines has shown significant improvement in energy efficiency and output in recent years.


Question: What is the significance of wind speed data in nuclear safety assessments?

Percentage frequency of wind direction and speed (MPH) during May at the GGNS site indicates patterns of wind that can affect the dispersion of airborne contaminants in the event of an incident.

Data from Site Meteorological Tower, 2001 – 2003 (Reference 2).


Question: Steam Generator Tube Rupture protocol

1-E-3, Steam Generator Tube Rupture, Step 8, CONTROL Ruptured S/G level, is in progress.

Which ONE of the following completes the statements in accordance with 1-E-3, Steam Generator Tube Rupture?


Question: residual heat removal system

Residual Heat Removal Service Water System is designed to remove excess heat from the reactor core and maintain the reactor coolant within acceptable temperature limits under both normal and emergency conditions.

Insulated piping, piping components, tanks are classified under pressure boundary and are susceptible to corrosion and cracking under certain environmental conditions.


Question: What is the role of the Nuclear Regulatory Commission in licensing?

The Nuclear Regulatory Commission (NRC) is responsible for regulating nuclear power plants and ensuring their safety, including the licensing process for operators and facilities.

The contact information for the NRC is provided, including the name, address, email, and phone number of Peyton Doub.


Question: The NRC's estimates for costs associated with the disposal of different components suggest substantial variations based on categorization and facility types.

NUREG-1307, Revision 8, introduced an alternative burial cost escalation factor (Bz) that assumed the use of waste vendor services and disposal of Class A LLW at the non-compact disposal facility located in Utah as an alternative to disposal of all decommissioning LLW at a compact-affiliated disposal facility.

These rates assume no volume discounts, which can be substantial.


Question: The report must include a detailed analysis of uncertainties associated with the sample measurements.

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

Batch 2047687 was assessed for analysis quality control without any noted disqualifiers.


Question: The initiative aims to introduce a strategic methodology for managing ex-plant materials instead of relying on opportunistic practices.

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

This effort is just getting underway, but we would like to present the concept at an NRC-industry materials meeting in early June.


Question: The NRC mandates that licensees monitor the levels of radioactive materials in the environment...

Measurement is a crucial aspect in determining the activity of radioactive materials, as shown in the data where various Measurement IDs correspond to specific dpm/100 cm³ values.

The Gamma Result of the measurements indicates the response of the detectors used in the analysis.


Question: The layout includes various significant features relevant to site safety and operations, such as water reservoirs and spent fuel storage areas.

The Independent Spent Fuel Storage Installation is a designated area for securing and managing spent nuclear fuel outside of the reactor building.

The site layout also features recreational areas, such as boat buoys, which are used for navigation and not related to nuclear safety.


Question: license renewal terms in nuclear regulation

the glossary in the 2013 GEIS defines 'license renewal term' as '[t]hat period of time past the original or current license term for which the renewed license is in force'

After the Board admitted the two environmental contentions, the Staff issued the Draft Supplemental Environmental Impact Statement (Draft SEIS) for Turkey Point Units 3 and 4.


Question: What is the role of the core cooling system in a nuclear reactor?

The core cooling system is critical in maintaining the reactor core at a safe operating temperature by removing heat generated during nuclear fission.

The financial implications of adopting new technologies in the energy sector must be thoroughly evaluated to ensure long-term sustainability.


Question: The Nuclear Regulatory Commission oversees the analysis of safety zones around nuclear power plants.

The NRC requires licensees to analyze emergency zones to ensure safety and public health during nuclear operations.

The table provides an overview of hazardous cargo rail accident data, detailing incidents and locations of cargo transportation.


Question: The conditions for new Surveillance Requirements (SRs) following amendment 133 require their first performance to occur at the end of the initial surveillance interval after implementation.

The schedule for the performance of new and revised Surveillance Requirements (SRs) shall be as follows: For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of this amendment. For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

This amendment authorizes the relocation of certain Technical Specification requirements to licensee-controlled documents.


Question: The NRC's approach to examining the economic impacts of regulatory actions is outlined in various guidelines.

SECY-12-0110, “Consideration of Economic Consequences within the U.S. NRC’s Regulatory Framework,” August 14, 2012

The documents identified in the following table are publicly available.


Question: What are the acceptable OM Code Cases for nuclear power plants?

a. Table 1, “Acceptable OM Code Cases,” lists the OM Code Cases that are acceptable to the NRC for implementation in the IST of light-water-cooled nuclear power plants.

RG 1.192, Revision 4, supersedes Revision 3.


Question: What is the purpose of the reactor core isolation cooling (RCIC) system?

Reactor core isolation cooling (RCIC) is a system designed to remove decay heat from the reactor core and provide cooling in the event of a loss of normal cooling capability.

The anticipated without SCRAM (ATWS) system B train after replacement of inverter and ground troubleshooting on October 3, 2024.


Question: What are the key parameters for the inadvertent opening of a pressurizer safety valve?

Table 5-17 presents key parameters for the inadvertent opening of a pressurizer safety valve event.

The reactor was inspected for any signs of corrosion or wear that could impact its operational lifespan.


Question: The system designed to connect field components with safety logic is known as the Component Interface Module.

The CIM is an FPGA-based component that performs the priority module function described in NRC interim staff guidance DI&C-ISG-04, "COMMAND PRIORITIZATION" (ML083310185).

The Double Width Transition Panel (DWTP) connects CIM base plates to the SRNC base, Ovation RNI assembly, and redundant 24 Vdc power feeds.


Question: What is meant by the term 'safety significance' in nuclear regulatory filings?

Safety significance is assessed based on the potential consequences of an event or condition related to nuclear safety and is a critical component in determining the regulatory response.

The reactor core is designed to operate within specified thermal limits to ensure efficient energy production and maintain structural integrity.


Question: What is the ASME OM Code requirement for maintenance on valves?

The ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) is incorporated by reference into 10 CFR 50.55a, “Codes and standards.” ASME OM Code, Subsection ISTC, paragraph ISTC-3310, “Effects of Valve Repair, Replacement, or Maintenance on Reference Values,” requires in part that: When a valve or its control system has been replaced, repaired, or has undergone maintenance that could affect the valve’s performance, a new reference value shall be determined or the previous value reconfigured by an inservice test run before the time it is returned to service or immediately if not removed from service.

Personnel involved: SM [Shift Manager], AOM [Assistant Operations Manager], Ops Director, Engineering Director, IST [Inservice Test] Engineer, AOV [Air-Operated Valve] Engineer, Reg Affairs Manager, Plant Manager, Site Vice President.


Question: The maximum allowed concentration levels for specific isotopes in the circulating water discharge are stipulated in regulatory guidelines.

Maximum Permissible Concentration (c) (µCi/ml) refers to the highest concentration of radioactive isotopes that can be present in the circulating water discharge, as outlined in regulations.

Total C_i (µCi/year) from table 11.2.3-1 provides a summary of annual releases from various isotopes, which include Ba-137m, Ce-144, and H-3.


Question: licensing procedure for nuclear programs

Please provide copy of the Program's licensing procedure(s) or a confirmation that the Program is using NRC’s NUREG-1556.


Question: The NRC stipulates specific requirements for vent positions in nuclear facility designs...


Question: modes of operation in nuclear reactors

Table 3-1 outlines the various modes of operation for reactors, including Power Operation, Startup, Hot Standby, Hot Shutdown, Cold Shutdown, and Refueling, with their respective reactivity conditions and thermal power ratings.

The maintenance schedule for the safety systems must be adhered to ensure operational integrity and compliance with regulatory standards.


Question: The submission outlines the significance of the 100-year peak flow inundation area in evaluating potential impacts of flooding on project sites.

The 100-year peak flow inundation area is a critical consideration for risk assessment and planning in the Crownpoint Uranium Project.

The Crownpoint Uranium Project is located in the Navajo Nation and covers various historical mining sites.


Question: The NRC assesses changes in population that may impact the effectiveness of alert systems.

Review for population changes affecting route alerting

Audit conducted by (Print and Signature):


Question: The NRC mandates a thorough examination of components subject to aging management review for systems such as the spent fuel pool cooling system.

Table 2.3.3-2 lists the components of the spent fuel pool cooling system that are subject to aging management review, specifying their intended functions and roles within the system.

A prime consideration in reactor safety is the assessment of human factors engineering, which aims to minimize operator error and enhance overall system reliability.


Question: The Nuclear Regulatory Commission monitors the cumulative dose exposure from various sources during operations.

Cumulative doses are calculated continuously to monitor the radiation exposure to personnel over time.

In various reactor designs, specific architectural features help ensure operational safety and efficiency.


Question: nuclear licensing process for reactor units

The licensing process for nuclear reactors, including the evaluation of safety, operational limits, and environmental impact, is governed by regulations set forth by the Nuclear Regulatory Commission (NRC).

The financial performance of the Susquehanna Steam Electric Station is a crucial aspect that impacts the energy market dynamics and local economic development.


Question: What is the purpose of a License Amendment Request in nuclear regulatory processes?

A License Amendment Request (LAR) is a request submitted to the NRC by a licensee seeking changes to the conditions of their operating license, which may involve modifications to facility design, operational procedures, or safety measures.

The reactor pressure vessel is a critical component that contains the reactor core, where nuclear fission takes place, ensuring that radioactive materials are kept contained and secure from the environment.


Question: PRA standards endorsed by Regulatory Guide 1.200

The PRA models used to support the RBS Risk-Informed Completion Time Program were developed in accordance with PRA standards endorsed in Regulatory Guide 1.200, Revision 2.

This enclosure is not applicable to the River Bend Station (RBS) submittal.


Question: What is the significance of energy measurements in nuclear safety assessments?

Energy levels, expressed in kiloelectronvolts (keV), are critical for evaluating the radiation emissions from nuclear materials and ensuring safety compliance in regulatory filings.

The table summarizes peak energy values and corresponding uncertainties, indicating the variations in measurements without discussing their implications for safety.


Question: nuclear plant licensing process

The licensing process for the Watts Bar Nuclear Plant involved the review of operational competencies through both an operating test and a written examination administered to the facility’s staff.

The licensed operators completed their training prior to the scheduled oral boards and practical examinations, ensuring their readiness for the assessment.


Question: What is the role of LLS valves in reactor modes?

Therefore, maintaining the LLS valves OPERABLE is not required in MODE 4 or 5.

The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.


Question: What do Technical Specifications (TS) require regarding Limiting Conditions for Operation (LCO)?

The regulation at Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2) requires that TSs include LCOs. Per 10 CFR 50.36(c)(2)(i), LCOs 'are the lowest functional capability or performance levels of equipment required for safe operation of the facility.'

The Monticello TSs use different numbering than the Standard Technical Specifications (STS) on which TSTF-582 was based. Specifically, TS 3.3.5.2, 'Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation,' in the STS was numbered as TS 3.3.5.3 in the Monticello TSs.


Question: nuclear licensing ITAAC process

NRC's ITAAC Index lists notified amendments and statuses related to nuclear construction, including requests for information and deletions as per license amendments.

The NRC conducts periodic inspections to ensure compliance with regulatory and safety standards during the construction of nuclear facilities.


Question: blast furnace slag in nuclear waste management

One intended function of the grouts at the SRS is to impede the release from the tanks of redox-sensitive radioelements such as technetium and plutonium.

Elements that were not likely to be covered in the technical sessions or project reviews were assigned to the QA auditors.


Question: The NRC requires an evaluation of operational experience (OE) to support the development of the license renewal application (SLRA).

The staff requests that the applicant provide a presentation on the process used to identify and evaluate the pertinent OE that was used to inform the development of the AMPs as discussed in the SLRA and basis documentation.

On an as-needed basis, the staff should perform an independent review of age-related OE contained in the CAP database to provide reasonable assurance that the applicant has adequately identified and evaluated pertinent OE in the development of the SLRA.


Question: cladding stress

Cladding Stress is evaluated in BAW-10227 (Reference 4) and involves assessing the criteria, method, and material properties necessary for nuclear safety compliance.

Criteria, Method is discussed in ANP-10342 (Reference 1) in relation to general component stress, which is separate from cladding-specific issues.


Question: What happens when the PRT safety valve remains stuck open?

The PRT is normally aligned to vent to the GW and RCDT system through 1RY469 which fails closed when pressure reaches 6 psi. With the safety valve leaking by, the pressure in the PRT will rise, 1RY469 will close and pressure will reach the rupture discs limit of 100 psi.

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.


Question: The measurements for nuclides require careful attention to their energy levels, yield percentages, and activity ratios.

The table lists various nuclides along with their corresponding energy in keV, yield in percentage, activity in pCi/grams, nuclide minimum detectable activity (MDA) in pCi/grams, and line MDA in pCi/grams.

Safety protocols ensure that all reactor components are inspected and maintained regularly to prevent failures.


Question: What is a technical specification in nuclear licensing?

Technical specifications are documents that provide the safety limits and operating conditions of a nuclear reactor, ensuring that it operates within established safety parameters.

In a thermal power plant, the efficiency of converting thermal energy to mechanical energy is a crucial factor influencing overall plant performance.


Question: What is Strontium-90 and its significance in nuclear safety?

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


Question: HI-STORM UMAX MPC-37 permissible heat load

HI-STORM UMAX MPC-37 Type 1 Permissible Heat Loads for the effective management of spent nuclear fuel are defined in the context of decay heat requirements and long-term storage considerations.

The classification of fuel assembly for MPC-37 in the HI-STORM UMAX System includes specifications on dimensions and arrangement required for proper storage but does not address permissible heat loads.


Question: What does the Rec% column indicate in regulatory filings?

If the Rec% column is null, the high/low limits are in the same units as the result. If the Rec% column is not null, then the high/low limits are in % Rec.

ASTM D516-07/11-16 is a standard test method for determining the sulfate content of water.


Question: PRB soil mass calculations

| Mass of soil in PRB        | 5,003    | U.S. tons  |

| Linear groundwater velocity      | 0.16       | ft/day     |


Question: What is the significance of black abalone in the consultation process for nuclear regulatory submissions?

To be consistent with the U.S. Army Corps of Engineer’s findings in the 2024 dredging permit review and because black abalone have not been documented in the Intake Cove to date, the NRC prefers to hold to this assumption—that the species is not present in the Intake Cove.

We understand that the ESA reinitiation criteria at 50 CFR 402.16(a)(2) would require us to reinitiate consultation if new information in the future indicates that this species is present in the Intake Cove because such information could potentially reveal effects of the action that may affect this species in a manner or to an extent not previously considered.


Question: thermally-induced failure of hot leg tubes

NRC analytical models and research findings continue to show that a TI-SGTR is more probable than predicted by the industry.

WCAP-16341-P (Reference 33) is an analysis specific for Westinghouse plants, and was performed with the knowledge of the results of NUREG-1570 (Reference 26) and NUREG/CR-6595, Revision 1 (Reference 27). This document is used as the primary reference document for modeling pressure- and TI-SGTR.


Question: What is the significance of a recirculation pump trip in reactor operations?

Trip of one recirculation pump motor is assumed. This event is categorized as an incident of moderate frequency.

Core flow stabilizes at new equilibrium conditions.


Question: The operability of each CVS letdown isolation valve is essential in maintaining system integrity during specific operational modes.

Each CVS letdown isolation valve shall be OPERABLE.

CVS letdown flow path may be unisolated intermittently under administrative controls.


Question: The elevation of the ground floor of the reactor building has been set to EL 135 ft.C.D.

The finished grade of the plant has been established at Elevation (EL) 116 ft.C.D. Grade rises abruptly in the area surrounding the reactor building to a grade of EL 135 ft.C.D., with the top of ground floor at EL 135 ft.C.D.

The Susquehanna River drains an area of 27,000 sq. mi. in New York, Pennsylvania, and Maryland.


Question: MILDOS-AREA modeling

As part of the license application MILDOS-AREA modeling was used to calculate potential radiation doses to individuals and populations from releases occurring at the project areas.

Estimated doses at the CUP ISR facilities will not exceed regulatory limits, and overall impacts to public and occupational health and safety will be SMALL.


Question: What is the significance of PB-214 in nuclear safety regulations?

PB-214 is a nuclide that has a weighted mean activity of 5.44E-02 pCi/grams, relevant for its assessment in regulatory filings concerning radioactive materials.

Errors quoted at 2.000sigma highlight the uncertainty limits in measurements, which are crucial for ensuring data accuracy in regulatory assessments.


Question: The reporting limit is crucial for identifying the lowest concentration that can be reliably detected in a sample.

RL: Reporting Limit                 |           |        |             |      |     |       |    |    |              |            |        |

The project involves strict adherence to sampling and analytical methodologies to ensure regulatory compliance.


Question: what is the RadICS platform in nuclear licensing?

Applicants using the RadICS platform are obligated to verify that the requirements of the application are bounded by the established qualification envelope.

Addressed in PSAI 7.4 and satisfied by: - Licensing Technical Report


Question: The NRC endorses thorough testing of components such as safety injection systems to ensure their reliability during normal operating conditions.

The safety injection system is a crucial feature in maintaining reactor safety, designed to deliver coolant to the reactor in case of an emergency.

The leak class for this test is classified as C, indicating acceptable leak rates during the inspection.


Question: soil classification in nuclear regulatory documents

The soil descriptions classify the materials based on granulation and consistency to ensure proper engineering assessments for safety and structural integrity.


Question: What is the purpose of the Wireless Emergency Alerts system at nuclear facilities?

Wireless Emergency Alerts (WEA) will be used to alert the public of an emergency at River Bend Station.

Key emergency personnel in the emergency response organizations and supporting emergency personnel will be contacted via commercial phone or by an alternate mode if necessary.


Question: The NRC mandates the analysis of specific isotopes in soil samples to assess contamination levels.

The subsurface soil samples were analyzed for the full suite of initial ROC, while the surface soil sample was analyzed for Co-60, Nb-94, Cs-137, Eu-152, Eu-154, Eu-155, and Am-241.

The on-site gamma spectroscopy results revealed that Cs-137 was detected at concentrations above MDC in twelve (12) of the continuing characterization samples, with a maximum concentration of 2.70E+00 pCi/g.


Question: NextEra Common Emergency Plan

The proposed NextEra Common Emergency Plan establishes an updated licensing basis for PSL that complies with current NRC regulations in 10 CFR 50.47 (Reference 6.1) and 10 CFR 50 Appendix E (Reference 6.4), and the guidance NUREG-0654/FEMA-REP-1 Revision 2 (Reference 6.8).

The site-specific annexes generally contain unique information such as local agency and support organizations, site-specific agreement lists, etc.


Question: groundwater aquifer impact on nuclear facilities

Describe the regional and local groundwater aquifers, formations, sources and sinks and onsite groundwater use, including present and projected future use. Describe the effects of present and projected groundwater use on gradients and groundwater or piezometric levels beneath the site.

Section 2.5.6 – Include information related to the investigation, engineering design, proposed construction, and performance of all earth, rock, or earth and rock fill embankments used for plant flood protection or for impounding cooling water required for the operation of the nuclear power plant.


Question: Operating procedures that impact safety and compliance must be rigorously enforced in all work activities.

All operating procedures which will affect radiation safety will be reviewed by the SERP.

Supervisory and management personnel will routinely observe their employees at work, and thus will be able to ensure adherence to the written procedures.


Question: The inspection team assessed the quality of inspections regarding health, safety, and security during the oversight process.

The team determined that the inspection plans and reports generated by the inspectors were exceptional. The team found that inspection documents were thorough, complete, consistent, and of acceptable technical quality with health, safety, and security issues properly addressed, and that management promptly reviewed inspection results.

The team evaluated 20 inspections reports and enforcement documentation, and interviewed inspectors involved in materials inspections conducted during the review period.


Question: inorganic analytical results for uranium

Uranium, dissolved has a result of 0.0162 B, with a dilution of 50 as per EPA method 200.8.

Rio Algom Mining Company has a project ID of 36-08 TRA-10032024 and received the sample on 10/05/24.


Question: frazil ice blockage

Frazil ice blockage can occur at intake structures, leading to potential operational challenges for cooling water systems in nuclear facilities.

The hydraulic reference manual provides essential information on the principles of hydrology and the methods used in flood risk management.


Question: alternate POD values in nuclear licensing

The proposed amendment would revise the Watts Bar UFSAR to apply alternate POD values to indications of axial ODSCC at TSPs in the Watts Bar, Unit 2, steam generators for the BOC voltage distribution in support of the Watts Bar, Unit 2, operational assessment until the steam generators are replaced.

The NRC will need to make the findings required by the Atomic Energy Act of 1954, as amended (the Act), and NRC's regulations.


Question: critical mass flow rate correlation

The critical mass flow rate correlation utilized assumes a homogeneous mixture of air, steam and water. Data which show comparisons between measured critical mass flow rates and predictions using the homogeneous critical flow model at low pressures have been analyzed.

Nodalization sensitivity studies were performed before the analysis was begun. The total number of nodes used varied from 6 to 54.


Question: The NRC requires that notifications for ITAAC status modifications are accurately documented in regulatory filings.

The ITAAC notification type can vary, including types such as ICN, IPCN, or UIN, depending on the nature of the notification required by the NRC.

The NRC maintains a checklist for vendor inspections, which includes evaluation criteria that may not directly relate to ITAAC notifications.


Question: Risk Informed Evaluation Process

Risk Informed Evaluation Process (RIEP): New rule does not appear to implement efficiencies and learnings associated with risk informed change programs that most of the fleet has implemented since 2010.

2010 Draft 50.46a rule required substantial implementation burden as compared to the potential benefits utilities would obtain (see ML100260383 & ML10316027).


Question: The regulatory framework requires that systems must demonstrate resilience against potential failures.

The Defense in Depth and Diversity Common strategy ensures that a combination of systems and procedures is employed to cope with failures that may occur in safety-related systems.

The modernization project focuses on enhancing the operational efficiency of existing units through technology upgrades and improved performance metrics.


Question: The reactor's average enrichment of uranium affects its cooling time requirements.

The Assembly Average Burnup is a critical parameter that determines the cooling time required for nuclear fuel, influencing the operational safety and regulatory compliance of the reactor.

The efficiency of the cooling system is vital for maintaining optimal reactor temperatures during operation.


Question: What are nuclides and their significance in nuclear safety?

Nuclide is part of an undetermined solution.


Question: The NRC evaluates the implications of a Loss of Coolant Accident on the reactor's containment integrity.

Loss of Coolant Accident (LOCA) Inside Containment (Main Steam) is a significant safety concern that requires thorough analysis during regulatory reviews.

The Limerick D3 Analysis focuses on various technical aspects of reactor operations and associated regulatory requirements.


Question: What is the COBRA-FLX Critical Heat Flux Correlation?

COBRA-FLX is a computational fluid dynamics tool used to model the thermal hydraulic behavior of nuclear reactors, particularly in predicting the critical heat flux.

The data from the test assembly conditions are essential for evaluating the performance and stability of reactor components under various operational scenarios.


Question: What is the NRC Device Key in regulatory filings?

**NRC Device Key:** 719593 (Internal Control Number)

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


Question: What is the function of second level undervoltage relays in nuclear systems?

The second level undervoltage relays for 4.16 kV Bus 1C1 are also designated as: - 227-1C1-2 (on drawing 6.5.1) - 275S, 276S (on drawings 6.5.10, 6.5.11, 6.5.18)

The relay tolerances are assumed to be 2σ values, particularly for the pickup and dropout settings which consider repeatability at constant temperature and constant control voltage.


Question: The NRC expects that the new nuclear alternative will consist of multiple reactor types, including both Advanced Light-Water Reactors and Small Modular Reactors.

For the new nuclear alternative, the NRC staff assumes that Duke Energy would replace the generating capacity of Oconee Station Units 1, 2, and 3 with a combination of (1) two ALWRs at the proposed W.S. Lee Nuclear Station in Cherokee County, South Carolina, and (2) a single-unit SMR at the Oconee Station site.

Loss of habitat and increased noise generation during construction and operation of the new facilities could cause terrestrial wildlife to move into other habitats in the surrounding landscape, increasing demands on those habitats and competing with other wildlife.


Question: What is the importance of isotopic analysis in reactor coolant systems?

ENSURE Chemistry is performing Isotopic Analysis of the RCS for Dose Equivalent Iodine and Dose Equivalent Xenon.

MONITOR area radiation monitors and recorder 1-RR-90-1 NORMAL.


Question: The NRC specifies various flow rates for safety injection under different reactor coolant system pressures.

Table 3-2 presents high head safety injection flow rates for cold leg pump discharge break spectrum under varying RCS pressures.

The reactor's temperature control system ensures stable operations during varying environmental conditions.


Question: What are the performance criteria for emergency core cooling systems?

Emergency Core Cooling Performance, Westinghouse Electric Corporation, Westinghouse Nuclear Energy Systems Proprietary, 1971.

J. J. Szyszlowski and R. Salvatori, Determination of Design Pipe Breaks for the Westinghouse Reactor Coolant System, WCAP-7503, Revision 1, 1972.


Question: What is the purpose of 10 CFR 50.69 in nuclear safety regulations?

To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation).

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.


Question: Radiopharmaceutical extravasations can affect the diagnostic accuracy in nuclear medicine.

I am much more concerned in ensuring that nuclear medicine technologist fully understand the gravity that the extravasations can have in quality and diagnostic accuracy of patient studies.

And I believe the diagnostic accuracy is much more of a pressing concern than radiopharmaceutical harm.


Question: The licensing process for nuclear facilities requires careful assessment of fuel assembly specifications...

Minimum Cooling Time (years) for various average enrichments of 235U is outlined in the loading table for BWR fuel.

The average enrichment levels must comply with the specified regulatory standards to ensure safety and compliance.


Question: elastomer piping management in nuclear safety

Consistent with NUREG-2191. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24) program will be used to manage flow blockage and loss of material of the elastomer ducting, ducting components, expansion joints and flexible connections exposed to closed cycle cooling water, raw water and condensation in the Control Room Ventilation System, Diesel Generator and Auxiliaries System, Open Cycle Cooling Water System, Safety-Related Ventilation System, SBO Diesel Generator Ventilation System and Station Blackout Diesel Generator System.

Not Used. The loss of material in steel heat exchanger components exposed to condensation is addressed by Item Number 3.3.1-132.


Question: Periodic evaluations are conducted to ensure the longevity and functionality of aging components in a safe nuclear environment.

The DCPP Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements AMP is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE, including research and development, such that the effectiveness of the AMP is periodically evaluated.

Elevated temperatures were noted on a single phase cable connected to two separate feeder breakers, which were subsequently inspected and addressed.


Question: PWR fuel assembly enrichment and burnup

Initial Assembly Avg. Enrichment wt % 235U (E) | WE 14x14 Assembly Average Burnup (B) GWd/MTU

The reactor design must also take into account the containment systems and their efficiency during a potential event.


Question: Exclusion Area Boundary

The Exclusion Area Boundary (EAB) Distance is measured in meters and indicates the perimeter around a nuclear facility where access is restricted to protect the public during potential radiological emergencies.

The data presented shows various measurements for downwind sectors, illustrating the concentration of airborne effluents during specific scenarios.


Question: What are the criteria for triggering a Reactor scram due to vibrations?

A Reactor scram is required if vibrations exceeded 12 mils or if vibrations exceeded 9 mils for more than 15 minutes.

If dispatched to investigate vibrations, wait 2 minutes, then report you can feel vibrations on the Turbine Deck but do not see any obvious cause.


Question: steam generator heat flow calculations

Table 6-21 summarizes the heat transfer correlations used to calculate steam generator heat flow in the SATAN Code.

The primary heat flow analysis also includes various factors that can affect the thermal efficiency of the reactor system.


Question: What is required for the offsite circuit to be operable in nuclear safety systems?

D.1: Restore [required] offsite circuit to OPERABLE status, or

The Fermi 2 TS numbering and wording for this condition are different than in NUREG 1433.


Question: What is the significance of potentiometric maps in environmental assessments for nuclear facilities?

Figure 3.6-7 shows the Shallow Groundwater Zone Potentiometric Map for December 2019, illustrating groundwater flow direction and potentiometric surface contours essential for environmental assessments at the Dresden Nuclear Power Station.

The elevation data table lists the identifier and elevation in feet for various monitoring wells, but it does not provide context about groundwater flow or its implications for environmental safety.


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

The activity of a nuclide, measured in pCi/grams, indicates the amount of radioactivity present and is crucial for assessing safety and compliance in regulatory contexts.

The nuclear reactor core must be designed to incorporate redundancy in cooling systems to ensure operational safety even in the event of equipment failure.


Question: The NRC mandates compliance with the established emergency planning standards to ensure regulatory requirements are fulfilled.

The proposed NAPS Emergency Plan changes continue to assure that regulatory requirements and emergency planning standards associated with emergency response are met.

Dominion Energy Virginia has determined that the proposed change would not revise a requirement with respect to installation or use of a facility or component located within the restricted area, as defined in 10 CFR 20, nor would it change an inspection or surveillance requirement.


Question: What are the human health impacts associated with construction and operation of nuclear facilities?

Overall, the human health impacts associated with the construction and operation of this alternative would be SMALL.

The solar installations and the property occupied by the wind turbines could be taxed at a higher industrial rate than agricultural land, providing a tax benefit.


Question: surface contamination limits

Removable contamination on the exterior surfaces of the TSC shall not exceed: a. 10,000 dpm/100 cm² from beta and gamma sources; and b. 100 dpm/100 cm² from alpha sources.

A.1 Restore TSC removable surface contamination to within limits


Question: What is an RAI in nuclear regulatory submissions?

RAI stands for 'Request for Additional Information,' which is a formal communication from the NRC requesting further details from licensees during the review process of a safety-related document.

According to our revised schedule, the now-anticipated ACRS dates are: ACRS SC: December 12, 2022 ACRS FC: February 2, 2023.


Question: log-normal distributions in nuclear parameters

Normal or log-normal distributions are used for the uncertainties associated with most of the manufacturing parameters (Table 5-5), model parameters (Table 5-8 and Table 5-9), and rod linear power parameters (Appendix B).

The uncertainties for the UO₂ thermal expansion, thermal conductivity, and specific heat were obtained from...


Question: braided weld inspection techniques

The partial penetration welds for the CRDM nozzles to head are inspected by dye penetrant and on completion of welding by Ultrasonic testing methods.

The full-penetration welds in the following areas of the installed irradiated reactor vessel are available for visual and/or nondestructive inspection, as required by ASME Code, Section XI, and described in detail in Section 5.2.5.


Question: Measures for managing hazardous material spills include the constructing of protective barriers and ensuring proper cleanup procedures are followed.

Construct a protective berm completely around the whole area including the working or clean-up area.

Using scoop shovels, load lined barrels.


Question: What is the function of the APRM Inop logic in nuclear safety systems?

The APRM Inop logic is designed to protect against multiple APRMs inoperable. It provides an automatic trip signal if multiple APRMs are inoperable.

The Reactor Vessel Water Level – Low, Level 3 condition requires automatic initiation and manual scram as procedural responses to ensure safety.


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

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


Question: The NRC mandates that licensees implement strict quality control measures for radiological surveys to ensure accuracy and safety.

All designated scan areas were scanned using a Ludlum 2350-1 paired with a Model 44-116 beta/gamma detector operated in the rate-meter mode and using audio response.

A total of fifteen (15) different scan areas (fourteen [14] random and one [1] judgmental), constituting an aerial coverage of 185 m², were scanned using a Ludlum 2350-1 paired with a Ludlum Model 44-116 detector (125 cm² detector area).


Question: The RV Vents and Drains System is designed to maintain the integrity of the reactor coolant pressure boundary.

The RV Vents and Drains System is a part of the reactor coolant pressure boundary (RCPB), which forms a barrier against the release of reactor coolant and radioactive material.

The aging management review results for these components are provided in Table 3.1.2-2, Reactor Vessel Internals - Summary of Aging Management Evaluation.


Question: shielding evaluation

5.1.1 Design Features provide a detailed examination of the shielding design used to protect against radiation exposure, outlining the specific features and materials utilized for effective shielding.

3.2.1 Material Properties discusses the physical and mechanical characteristics of materials used in nuclear reactors but does not address any aspects of radiation shielding.


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

For 40 % Assembly Average Burnup ≤ 41 GWd/MTU, minimum cooling time (years) varies based on enrichment levels of 235U.

The NRC often reviews various aspects of reactor designs during the licensing process, ensuring that all safety protocols are in place.


Question: medical event reporting for radiopharmaceuticals

So if shunting was accounted for prior to the evaluation and shunting occurs and causes that 50 rem dose and 50 percent or more of the dose from the expected administration, then that's like the caveat that I am talking about.

Should a different dose threshold be developed for extravasation or should there be certain caveats like with Yttrium-90 microspheres?


Question: What is the significance of Breaker 1524 in nuclear plant operations?

Performer positions 1-HS-57-62 to TRIP. Breaker 1524 OPENS and Breaker 1122 fails to CLOSE, resulting in the loss of 6.9 KV Unit Board 1C.

Loss of secondary pumps will require the performer to reduce turbine load using 1-AOI-37, Turbine Runback Response.


Question: What is the intended function of N-4 Feedwater Nozzles?

N-4 Feedwater Nozzles are classified as a pressure boundary and made of carbon or low alloy steel with stainless steel cladding.

The aging management programs for reactors are designed to ensure that the structures and components continue to perform their intended functions throughout their service life.


Question: The NRC's new interpretation may allow nuclear materials to be managed without the stringent oversight typically required, raising concerns about public safety.

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

Sending it to hazardous or industrial waste sites is even worse because the radioactivity can react with other chemicals, making leaks to air and water exponentially more harmful than they are already.


Question: What is a Mark I containment system in nuclear reactors?

The Mark I primary containment system for each unit at BFN is a pressure suppression system consisting of a drywell, pressure suppression chamber, vent system, isolation valves, containment cooling system, and other service equipment.

The BFN central site area consists of a reactor building, a turbine building, a service building, a maintenance building, two diesel generator buildings (one for Units 1 and 2 and one for Unit 3), a radioactive waste building, administration buildings, a Diverse and Flexible Coping Strategies equipment storage building, an intake pumping station, a 161-kilovolt (kV) switchyard and a 161-kV export yard.


Question: What is the activity of isotope AM241 in a regulatory document?

AM241 is noted to have an activity of 150 mCi.

The device is distributed by NDC Technologies and has an internal control number of 624168.


Question: nuclear equipment and materials

Specially designed and prepared nuclear equipment, parts and components, materials, software, and technology covered by part 810 of title 10, Code of Federal Regulations (relating to assistance to foreign atomic energy activities);

Interconnection arrangements means arrangements governing the physical connection of two or more networks to allow the use of another's network to hand off traffic where it is ultimately delivered (e.g., connection of a customer of telephone provider A to a customer of telephone company B) or sharing data and other information resources.


Question: The NRC actively oversees the operations of commercial nuclear reactors to ensure safety standards are maintained.

The Reactor Oversight Process is the NRC’s program for overseeing the safe operation of commercial nuclear power reactors.

No findings or violations of more than minor significance were identified.


Question: What is the process for handling RTD failures in reactor systems?

Crew should recognize loop 3 Tc RTD failure and enter 1-AOI-2, Malfunction of Reactor Control System.

Annunciator 92-F EAGLE PROC PROT CH-III RTD FAILURE in alarm.


Question: What is the aging management program for stainless steel piping?

AMP XI.M29, “Outdoor and Large Atmospheric Metallic Storage Tanks,” AMP XI.M32, “One-Time Inspection,” AMP XI.M38, “Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components,” or AMP XI.M42, “Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks”

Consistent with NUREG-2191. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.124) program will be used to manage loss of material of the carbon steel, cast iron, and galvanized steel blower housing, compressor housing, ducting and components, flexible connection, heat exchanger components, piping, piping components, tanks, turbine casings, and valve bodies exposed to condensation in the Containment Atmospheric Control System, High Pressure Coolant Injection System, Isolation Condenser System, Primary Containment Isolation System, and Standby Gas Treatment System.


Question: regulatory audit requirements

The NRC staff will develop any RAIs, as needed, via NRR Office Instruction LIC-115 and issue such RAIs separately from audit-related correspondence.

The eDocs web will be password-protected, and separate passwords will be assigned to the NRC staff who are participating in the audit.


Question: What happens to the coolant temperature during a partial loss of reactor coolant flow?

A partial loss-of-forced-reactor-coolant flow accident can result from a mechanical or electrical failure in an RCP or from a fault in the power supply to the pump or pumps supplied by an RCP bus. If the reactor is at power at the time of the accident, the immediate effect of the loss-of-forced-reactor-coolant flow is a rapid increase in the coolant temperature.

All of the accidents in this section have been analyzed. It has been determined that the most severe radiological consequences will result from the reactor coolant pump shaft seizure accident discussed in subsection 15.3.3. Therefore, doses are reported only for that limiting case.


Question: metal enclosed bus aging management

The Metal Enclosed Bus program (B.2.141) will be used to manage loss of material of the galvanized steel or aluminum metal enclosed bus: external surface of enclosure assemblies exposed to air - outdoor.

No aging management program is required for those applicants who can demonstrate these fuse holders are located in an environment that does not subject them to environmental aging mechanisms and effects due to chemical contamination, corrosion, and oxidation.


Question: analytical limits in nuclear regulatory documents

The Allowable Value contained in the current Technical Specification is the analytical limit from FSAR Chapter 15.

SNC proposes to change the Technical Specification to replace the Analytical Limit with the appropriate Allowable Value.


Question: neptunium-237 analysis

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


Question: voltage decay incidents

The long time for voltage decay (fifteen minutes) was due to relatively long fixed time delay between tap steps.

Several voltage decays were experienced over a period of several hours.


Question: What is the purpose of the Decommissioning Quality Assurance Program?

The Decommissioning Quality Assurance Program (DQAP) is associated with Facility Operating License DPR-40, Docket No. 50-285, and Docket Nos. 72-054 for the ISFSI, and 71-0256 for radioactive material packages.

Update guidance for electronic storage of records.


Question: The report discusses the minimum detectable activity and its importance in regulatory compliance.

The contact person for this project is Mr. Jason Q. Spaide, who handles client communications.


Question: contingency plans in nuclear operations

Contingency plans were formulated to address credible events that could impede the timely completion of the plenum assembly lift and transfer operations. Many subsequent contingency plans were a direct result of lessons learned during reactor vessel head lift operations.

A checkout program for the plenum lift equipment identified several needed modifications. Special equipment was operationally and functionally checked out as part of an integrated system before use in the containment building.


Question: height of CTU-2 testing component


Question: What are the safety measures in place for nuclear power plants to prevent accidents?

Ray insisted that the AEC believed “that the care taken in design and operation ensures that the chances of a serious accident happening at a nuclear plant are very small.”

AEC critics were undoubtedly dissatisfied with Ray’s assertions, and they were certainly disappointed when Senator Jackson’s S. 2135 bill to create the DENR lost steam.


Question: standby liquid control subsystem

I.1 Declare associated standby liquid control subsystem (SLC) inoperable.


Question: The NRC stipulates that prior to construction activities relating to the operation of a nuclear facility, necessary licensing requirements must be fulfilled.

The purpose of 10 CFR 50.10 is to ensure that activities constituting construction of a production or utilization facility on a site on which the facility is to be operated do not begin until a license or authorization is issued by the NRC.

The reactor is now controlled at low power.


Question: RCS unidentified leakage

Conducts activities to identify sources of RCS unidentified leakage. Documents actions taken to identify sources of unidentified RCS leakage in the control room logs or in the corrective action program, as specified in plant administrative procedures. The licensee’s leak identification plan includes actions such as system walkdowns; system surveillance and re-alignment; containment entry (PWR only) and visual inspections for boric acid deposits (PWR only); verification of pumps and valves for possible seal and packing leakages; inspection of pipe flanges and major welds; including instrument lines and connections; and sampling/performing isotopic analysis of atmospheres, filter elements, and sumps.

If the inspector observes significant adverse trends, the inspector should engage licensee and regional management and the appropriate Office of Nuclear Reactor Regulation (NRR) technical branches as outlined in Attachment 1. As applicable, the inspectors should also determine if the licensee enters the appropriate procedure for responding to adverse RCS leakage trends.


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

The unit of measurement for radioactive isotopes is often expressed in millicuries (mCi).

The Internal Control Number is a key identifier for the NRC device, which ensures adequate tracking and registration of devices used in regulated activities.


Question: nuclear license amendment request

In accordance with 10 CFR 50.91, Dominion Energy Virginia is notifying the Commonwealth of Virginia of this LAR by transmitting a copy of this letter to the designated state official.

Should you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.


Question: Surge Tank Vent Valve

The crew initiates S2.OP-AB.CC-0001 (Component Cooling Abnormality) and S2.OP-AB.RAD-0001 (Abnormal Radiation).


Question: The Kankakee River serves as a vital component for managing the cooling water supply at the DNPS facility.

The Kankakee River is the primary source of cooling water for DNPS.

The Des Plaines River is a secondary source of cooling water for DNPS.


Question: What steps are involved in evaluating the adequacy of an evaluation model?

Assess scalability of models, including the preparation of input and calculations to evaluate model fidelity and accuracy in representing physical processes.

1. Specify analysis purpose, transient class and power plant class.


Question: What are the recommendations for steam dryer inspections at Browns Ferry Nuclear Station?

GE-Hitachi provided recommendations for future inspections of the RSDs (GE-Hitachi Nuclear Energy Report No. N077485 - Revision 0, "Browns Ferry Nuclear Station (BFNS) - Recommendations for Future Inspections - Replacement Steam Dryer,” November 2022), which was subsequently submitted to the NRC (Letter from TVA to NRC, “Long Term Steam Dryer Inspection Plan,” dated January 20, 2023) pursuant to Unit 1 Operating License Condition 2.C(18)(i), Unit 2 Operating License Condition 2.C(18)(i), and Unit 3 Operating License Condition 2.C(14)(q) which required that a long term steam dryer inspection plan be submitted based on industry experience.

The resolution standard for these visual examinations are 'best effort VT-1' in accordance with BWRVIP-139-R1-A. Because there are not specific VT-3 resolution requirements per BWRVIP-139-R1-A, the resolution requirements of the recommended VT-3 examinations are the same as those for VT-1.


Question: population distribution in nuclear regulatory filings

The resident and transient population distribution within a 16-kilometer (10-mile) radius of the facility is projected for the year 2040, detailing the population across various directions.

The safety analysis report clearly outlines the methodologies used to evaluate the potential consequences of postulated accidents at the facility.


Question: What is the status of the monarch butterfly in the context of regulatory filings?

The monarch butterfly is a candidate for Federal listing. In 2020, the FWS issued a 12-month finding announcing its intent to prepare a proposed rule to list the monarch as threatened. In 2022, the FWS identified the monarch listing action as a priority because the magnitude of threats is moderate to low; however, those threats are imminent for the eastern and western North American populations.

Researchers identified no suitable habitat for such species in a 2013 botanical field survey associated with the Federal Energy Regulatory Commission’s license renewal of the Keowee-Toxaway Project.


Question: What does the Nuclear Waste Policy Act require for high-level radioactive waste management?

The Nuclear Waste Policy Act provides for the research and development of repositories for the disposal of high-level radioactive waste, spent nuclear fuel, and low-level radioactive waste.

The Endangered Species Act (ESA) was enacted to prevent the further decline of endangered and threatened species and to restore those species and their critical habitats.


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

The RCIC System Instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system.

Diverse RCIC system instrumentation inputs (TS Table 3.3.5.3-1) include multiple sensors for various measurements.


Question: To ensure compliance with aging management regulations, the implementation of specific quality assurance programs is critical.

The implementation of the guidance in MRP-227-A (as supplemented), plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.


Question: The company implements strict heat input controls for welding processes relevant to austenitic stainless steel.

Westinghouse controls the heat input in all austenitic pressure boundary weldments by prohibiting the use of block welding, limiting the maximum interpass temperature of 350°F, and exercising approval rights on all welding procedures.

Of 25 production and qualification weldments tested, representing all major welding processes, and a variety of components, and incorporating base metal thicknesses from 0.10" to 4.0", only portions of 2 were severely sensitized.


Question: strontium-90 regulatory guidelines

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


Question: What is the role of the Electrical Cable Tunnel in reactor safety?

The Electrical Cable Tunnel provides physical support as well as shelter and protection for the electric cables associated with RHRSW (Residual Heat Removal Service Water), EECW (Emergency Equipment Cooling Water), and CCW (Condenser Circulating Water) Systems in addition to the electric fire pumps, all of which are credited to demonstrate compliance with the Fire Protection regulated event.

The aging management review results for these components are provided in Table 3.5.2-31, Residual Heat Removal Service Water Tunnels - Summary of Aging Management Evaluation.


Question: Will the proposed change result in any new kinds of accidents that haven't been previously considered?

The proposed exemption does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed exemption does not affect any safety limits or limiting conditions for operation used to establish safety margin.


Question: What is mechanical peening in the context of welding?

The purposes of mechanical peening were to reduce weld restraining related cracking problems and to reduce dimensional distortion of the weldments during repair.

The advantages of weld joint sequencing are that proper access for welding is provided which promotes generally better welding quality; the reduction of overall welding stresses is thereby gained; and also general dimensional distortion is minimized.


Question: The NRC endorses strategies for maintaining the chemical environment in reactor coolant systems to manage aging effects.

The primary water chemistry program is consistent with the guidelines of EPRI Report 3002005005, "PWR Primary Water Chemistry Guidelines," Revision 7, as endorsed by the NRC in SLR-ISG-2021-02-Mechanical.

The report emphasizes the importance of maintaining comprehensive records and documentation for all maintenance activities performed on the reactor systems to ensure regulatory compliance.


Question: What is boric acid corrosion in nuclear systems?

Loss of material due to boric acid corrosion

Not Applicable. There are no aluminum piping or piping components exposed to raw water in Steam and Power Conversion.


Question: How does the operator assess the radiation levels in a controlled area?

Operator refers to Radiological Survey Map and determines that the work area is in a High Radiation Area (HRA) and Contaminated Area (CA).

Evaluator's Note: RP-AA-300, Attachment 2 provides definitions of survey map symbols and abbreviations.


Question: The NRC necessitates assessments to evaluate the structural integrity of protective barriers in environments that may experience chemical reactions from aggregates.

Concrete (accessible areas): all - Shelter and Protection, Structural Support - Cracking due to expansion from reaction with aggregates - Structures Monitoring (B.2.1.33)

Aging Management Programs are essential for maintaining safe operations as they prevent unexpected failures due to material degradation.


Question: Reactor trip system instrumentation

The Reactor Trip System Instrumentation includes various functional units that monitor and control operational parameters to ensure the safety and functionality of the reactor under different operational conditions.

The table outlines the total number of channels for each functional unit, as well as the channels required to trip, the minimum operable channels, and the applicable modes for operation.


Question: The analysis for ANO-2 indicates that the SSO time has a direct impact on the accumulation of debris in the reactor vessel.

The earliest SSO times for ANO-2 are greater than those assumed in the WCAP-17788 analysis.

The maximum rated thermal power for ANO-2 is less than the analyzed power level in WCAP-17788 for CE plants.


Question: What are the typical units used for reporting radioactive material concentrations in nuclear regulatory filings?

The units of measurement for radioactive material concentrations are typically expressed in picocuries per gram (pCi/g).

The date of the sample collection was recorded as 08/13/19, indicating when the analysis was performed.


Question: The NRC mandates that licensees submit comprehensive reports on operational safety and compliance.

The U.S. Nuclear Regulatory Commission (NRC) oversees the safety and licensing of nuclear reactors through rigorous regulatory processes and detailed evaluations to ensure operational compliance.

AmerenUE operates multiple facilities within Missouri, contributing to the region's energy grid and supporting local economies.


Question: steam generator tubes

SG Tubes: Pressure Boundary, Nickel Alloys, Secondary Water >60°C (>140°F) (Ext), Cracking, Steam Generators (B.2.3.10) and Water Chemistry (B.2.3.2), IV.D1.R-47.

The reactor coolant system is designed to transfer heat from the reactor core to the steam generators, maintaining safe operational temperatures.


Question: nonsafety-related ventilation system components

| Blower Housing (AEER Air Handling Unit) | Leakage Boundary | | Bolting (Closure) | Mechanical Closure | | Drip Pan | Leakage Boundary | | Heat Exchanger – (Decontamination Facility Air Conditioner) Tubes | Leakage Boundary | | Heat Exchanger – (Drywell Coolers) Tubes | Leakage Boundary | | Heat Exchanger – (Heater Steam Units) Tubes | Leakage Boundary | | Heat Exchanger – (Make Up Air Units) Tubes | Leakage Boundary | | Heat Exchanger – (Steam Coil Demineralizer and Duct Steam Coil) Tubes | Leakage Boundary | | Heat Exchanger – (Steam Converter) Shell Side Components | Leakage Boundary | | Heat Exchanger – (Steam Converter) Tube Side Components | Leakage Boundary | | Heat Exchanger – (Steam Heating Condensate Cooler) Shell Side Components | Leakage Boundary | | Heat Exchanger – (Steam Heating Condensate Cooler) Tube Side Components | Leakage Boundary | | Heat Exchanger – (Steam Tunnel Coolers) Tubes | Leakage Boundary | | Heat Exchanger – (Subcoolers) Tubes | Leakage Boundary | | Heat Exchanger – (Turbine Building Inlet Heating Coil) Tubes | Leakage Boundary | | Piping elements | Leakage Boundary | | Piping, piping components | Leakage Boundary | | Pump Casing (AEER A/C Condensate Pump) | Leakage Boundary |

The regulatory body ensures that all safety assessments are conducted in a thorough manner. This includes evaluating the integrity of containment structures and examining emergency evacuation plans regularly to minimize risk in the event of an incident.


Question: What is the purpose of a safety evaluation report in nuclear licensing?

The Safety Evaluation Report (SER) provides an assessment of the safety of a proposed nuclear facility and ensures that the application meets regulatory requirements.

The Environmental Impact Statement (EIS) outlines the potential effects of a proposed project on the surrounding environment.


Question: aging management evaluation of reactor components

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

Concrete (inaccessible areas); exterior above and below-grade; foundation | Flood Barrier, Shelter and Protection, Pressure Boundary, Shielding, Structural Support, Missile Barrier, HELB Shielding


Question: The NRC outlines procedures for measuring radiation levels around storage casks.

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

The TSC mid-plane is defined as approximately 92 inches from the bottom.


Question: weighted mean activity

Wt mean Activity (pCi/grams) represents the weighted mean of the activity concentration of a nuclide in the sample measured, expressed in picoCuries per gram.

Errors quoted at 2.00 sigma indicate that the statistical error metrics are calculated at a confidence level of 95%.


Question: The petitioners assert that the LRA must adequately detail the environment affected and the potential impacts on archaeological and cultural resources.

Organizational Petitioners claim the LRA lacks “an adequate description of either the affected environment or the impacts of the project on archaeological, historical, and traditional cultural resources.”

Petitioners are not required to prove their contentions at the admissibility determination stage and “we do not consider the merits of [petitioners’] arguments” at this stage.


Question: The reactor coolant pump requires an adequate oil pressure as indicated by various pressure switches to prevent malfunction.

Three pressure switches, arranged in a 2/3 coincidence network, are located on the oil lift supply header to prevent the start of the reactor coolant pump if there is inadequate pressure in the system.

In addition to the above pressurizer liquid level signals are transmitted to the Ovation PCS for specific functions as follows:


Question: The DNPS site is designated for industrial use, allowing for various types of developments to support its operation.

The power generating portion of the DNPS site, which includes associated facilities, is zoned as an Industrial District, according to the Grundy County Unified Development Code, 8-1-3-1: Zoning Districts.

Deer hunting is permitted on site with a signed DNPS deer hunt release. Onsite hunting is only permitted with bow and arrow and only during the archery portion of the hunting season.


Question: What does the Inner Piedmont belt consist of in Virginia?

The Western Piedmont belt, referred to as the Inner Piedmont belt in some publications, extends southward from Virginia through North Carolina and into Georgia. It is composed of greenschist facies (low grade) and amphibolite facies (high grade) meta-sedimentary rocks.

The number of plutons and their relation to one another, however, remains uncertain.


Question: Core Makeup Tanks in nuclear power plants

Two redundant CMTs provide sufficient borated water to assure Reactor Coolant System (RCS) reactivity and inventory control for all design basis accidents (DBAs), including both loss of coolant accident (LOCA) events and non-LOCA events.

The outlet line from each CMT is connected to two other vessel injection lines, which provides an injection path for the water supplied by the CMT.


Question: What is the Departure from Nucleate Boiling in reactor safety testing?

Departure from Nucleate Boiling refers to the point at which the coolant in a reactor ceases to efficiently transfer heat, leading to a transition from nucleate boiling to film boiling, which is critical for understanding fuel performance under accident conditions.

The verification test involves rigorous analysis and documentation by licensed professionals to ensure that all components meet the required safety and performance standards.


Question: Licensees must evaluate the safety implications of modifications to their reactor systems.

The regulatory process for nuclear licensing requires applicants to submit detailed safety analyses to ensure compliance with safety standards.

The use of advanced materials in construction can improve the strength and durability of buildings.


Question: aging management evaluations for concrete

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

Cracking due to expansion from reaction with aggregates


Question: Successful completion of training in managing radiation hazards is a critical component for personnel in the nuclear sector.

has successfully completed and passed all the requirements for Radiation Response.

The health department is responsible for ensuring public health and safety through various administrative mechanisms.


Question: What is the significance of the Steam Generator Tube Rupture (SGTR) Accident in nuclear safety analysis?

...the revised analysis of the radiological consequences of a Steam Generator Tube Rupture (SGTR) Accident associated with operation of Unit 2 Replacement Steam Generators (RSGs).

...the increase in the main control room (MCR) thyroid dose for the pre-accident case was a more than minimal increase in the radiological consequences of an accident previously evaluated in the UFSAR.


Question: NRC device registration


Question: What is the Backfit Rule in nuclear licensing?

The Backfit Rule also provides the requirements for consideration of regulatory effects on those items such as procedures that are not covered by manufacturing license finality during the term of that license.

The agency is gaining valuable experience in the review of applications to reauthorize the operation of power reactor facilities that could be applied to the concept of transitioning a micro-reactor between states of being authorized to operate and being not-in-operation.


Question: What are the fuel pin cladding temperature criteria for failure analysis?

The resulting TATNF criteria above were developed after the initial PIRT, however they are consistent with the FOM used within the PIRT.

The hierarchical system decomposition of the Natrium design follows:


Question: impact of low flow rates on component cooling water systems

The applicant is required to have the ability to predict the impacts of low CCW (CCS) flow rates on the Component Cooling Water System and use procedures to mitigate the consequences.

The safety-related systems must be properly monitored to ensure that all operating parameters remain within acceptable limits.


Question: baseline inspections

Yes, the 2025 inspection will be the baseline inspection. The last inspection was performed in 2015. The “Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks,” Aging Management Program (AMP) requires the baseline inspection to be performed within the 10-year period prior to entering the period of extended operation (PEO) which, for Clinton Power Station, will be in 2027.

The next scheduled inspection will be performed in 2025.


Question: What is the purpose of a calibration factor in nuclear measurements?

The correction factor is the ratio of the uniform response to the measured calibration source response.

Measurements were to be made using a Ludlum 44-10 NaI detector placed near the bottom of the inside of the pipe.


Question: nuclear materials licensing

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Parts 30, 31, 32, 33, 34, 35, 36, 37, 39, 40, 70 and 71, and in reliance on statements and representations heretofore made by the licensee, a license is hereby issued authorizing the licensee to receive, acquire, possess, and transfer byproduct, source, and special nuclear material designated below; to use such material for the purpose(s) and at the place(s) designated below; to deliver or transfer such material to persons authorized to receive it in accordance with the regulations of the applicable Part(s).

This license shall be deemed to contain the conditions specified in Section 183 of the Atomic Energy Act of 1954, as amended, and is subject to all applicable rules, regulations, and orders of the Nuclear Regulatory Commission now or hereafter in effect and to any conditions specified below.


Question: licensed material inventory requirements

The licensee shall conduct a physical inventory every 6 months, or at other intervals approved by the U.S. Nuclear Regulatory Commission, to account for all sealed sources and/or devices received and possessed under the license.

Maintenance, repair, cleaning, replacement, and disposal of foils contained in detector cells shall be performed only by the device manufacturer or other persons specifically authorized by the U.S. Nuclear Regulatory Commission or an Agreement State to perform such services.


Question: Containment System for nuclear reactors

The general purpose of the Containment System is to provide a barrier confining potential releases of radioactivity from severe accidents. This is accomplished by maintaining leaktightness within specified bounds. As a design feature, the Containment System is provided primarily for the protection of public health and safety.

Separate emergency power sources are supplied to the redundant active components and separate instrument channels are used to actuate the systems.


Question: leak test analysis

This certifies that leak test analysis was conducted on the sample with the following information.

The results shown below accurately represent the level of removable contamination.


Question: plutonium analysis in nuclear samples