Question: nuclear power plant restart activities

The security infrastructure changes including new barrier/wall are part of the proposed restart activities.

Complete facility roofing inspections and repair/replace roofs.


Question: The procedures for notifying about nuclear accidents must be followed in case of emergencies.

EP-MW-114-100-F-01, Rev.K, Nuclear Accident Reporting System (NARS) Form provides a standardized method for reporting nuclear accidents.

This is an Administrative JPM that may be performed in any setting where the necessary procedures and support information can be provided.


Question: The NRC notifies licensees when certain ITAAC have been deleted.

The NRC checklist includes various inspection reports that inform about construction status.


Question: nuclear waste dose limit regulations

The proposed 'dose limit'--really ANY dose limit--is a calculation done by applicant using unverified assumptions to claim the site will be acceptable.

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


Question: The NRC mandates that assessments must be conducted regarding the integrity of pressure boundaries in various reactor components.

Tank (Boric Acid Batching Tanks)     | Pressure Boundary         | Stainless Steel          | Plant Indoor Air (Int) | Loss of Material                  | Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Tank (Boric Acid Holding Tanks)      | Structural Integrity (attached) | Cast Austenitic Stainless Steel | Plant Indoor Air (Ext) | None                               | VII.J.AP-17


Question: The TVA indicates that decommissioning will be performed following the established NRC protocols.

Decommissioning activities would be conducted by TVA in accordance with established NRC guidance.

The construction of additional plant structures that would require decommissioning would not be required to support the subsequent period of extended operation.


Question: The uncertainty associated with the Zr-4 low stress creep calibration database is outlined in previous reports.

The data contained in Figure 5-6 of Reference 2 is taken from the determination database for Zr-4 low stress creep is based on the uncertainty as described in Section 5.4.3.1 of Reference 2.

For the Zr-4 low stress creep model, the test rods used in validation of the model are not used in the.


Question: metal enclosed bus aging management program

The BFN Metal Enclosed Bus aging management program is an existing condition monitoring program and no actions are taken as part of this program to prevent or mitigate aging degradation. The program applies to metal enclosed buses within the scope of subsequent license renewal to identify age-related degradation of electrical insulating material (i.e., thermoplastic organic polymers), metallic, and elastomer components (e.g., gaskets, boots, and sealants).

A Condition Report was initiated to document the cable aging degradation and the cracking of the jacket. A work order was created to correct this aging degradation in the reactor building.


Question: What are the elements involved in predicting the effects on the Standby Gas Treatment System during abnormal conditions?

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High secondary containment ventilation exhaust radiation.

A small steam leak develops in the Reactor Building.


Question: What materials are commonly used in the construction of turbine building components?

Steel Components: structural steel | Structural Support | Galvanized Steel | Air - Indoor, Uncontrolled | None | None | III.B5.TP-8 | 3.5.1-095 | A

The aging management programs are detailed in Section B.2.1.33 of NUREG-2191, which outlines the various strategies essential for maintaining the integrity of safety-related structures.


Question: The activity calculations involve determining the weighted mean of each nuclide's activity based on their respective confidence levels.

Wt mean Activity (pCi/grams) is a critical parameter calculated for each nuclide to ensure accurate reporting in nuclear licensing.

Errors quoted at 2.000 sigma indicate the level of uncertainty in the measurements collected during the assessment.


Question: What are the Energy Group classifications for gamma radiation levels?

The Energy Group (MeV/gamma) classifications are crucial for understanding the source strength at different times after shutdown of a nuclear reactor, detailing the various energy ranges associated with gamma radiation and their corresponding measurements.

The cooling water system is responsible for transferring heat away from the reactor core, ensuring that the temperature remains within safe operational limits during normal operations and emergency situations.


Question: available net positive suction head (NPSH) calculations

System piping size and layout provides adequate NPSH to the containment spray pump during all anticipated operating conditions, in accordance with Regulatory Guide 1.1.

Recirculation sump construction provides straining down to 0.045-inch strainer hole size to prevent entrained particles in excess of that size from entering the containment recirculation sump and containment spray system suction piping.


Question: standby liquid control system

Each BWR must have a standby liquid control system with defined boron injection capabilities.

The BFN SBO coping strategy is to shutdown the blacked-out unit with equipment powered from the 250-V Direct Current (DC) battery system.


Question: regulatory standing requirements

The standing elements the Alliance and the Collective challenge as being extra-statutory and/or *ultra vires*, though, are found in a Commission regulation.

In the Reply, the Alliance and Collective argued that the requirement to include an address and telephone number in the petition simply was a clerical omission and included supplemental affidavits with addresses and telephone numbers for themselves, but not for the members on whose behalf they seek representational standing.


Question: What is the definition of developed low intensity land use in nuclear project areas?

Developed, Low Intensity: Includes areas with a mixture of constructed materials and vegetation. Impervious surfaces account for 20% to 49% percent of total cover. These areas most commonly include single-family housing units.

The project area is approximately 92 km [57 mi] northeast of Gallup (2020 population 21,899) and approximately 90 km [56 mi] northwest of Grants (2020 population 9,163).


Question: Nuclear safety protocols necessitate comprehensive evaluations conducted by licensees during compliance submissions.

The NRC applies a strict regulatory framework that mandates licensees to conduct thorough assessments to ensure the safety of nuclear reactors.

The thermal efficiency of a reactor is a measure of its ability to convert thermal energy from nuclear fission into electrical energy.


Question: The RCS must be verified for uniform boron concentration prior to core loading procedures.

To verify correct and uniform boron concentration, prior to core loading in the RCS and directly connected auxiliary systems.

The alarm signal level will be established for startup testing and operations.


Question: What does ALARA stand for in nuclear safety regulations?

ALARA stands for As Low As Reasonably Achievable, emphasizing the need to keep radiation exposure as low as possible in nuclear safety practices.

The GPS is utilized for accurate location tracking, which is essential for various navigation systems but does not pertain to radiation safety.


Question: What parameters are included in the radiochemistry certification process?

The following parameters are not offered for certification or are not covered by NELAC certificate #ACZ.

The ACZ Project ID is L88959, which refers to specific activities within the Rio Algom Mining Company.


Question: audit discussion topics

The component-specific line items in the tabular audit discussion tables of Enclosures 3, 4, and 5 are color coded as follows in order to convey whether the component-specific topics warrant audit discussions, and if so, the degree of audit discussions that are merited on the component-specific audit topic: 1. **Blue color** highlighted content conveys the component-specific topic does not need to be included in audit discussions. 2. **Yellow color** highlighted content conveys the component-specific topic warrants some limited audit discussion. 3. **Green color** highlighted content conveys the component-specific topic warrants more detailed audit discussions.

The NRC staff requests that EPRI make available the appropriate engineer(s) with knowledge of EPRI MRP I&E guidelines and any appropriate references, to address questions by the NRC staff.


Question: The NRC mandates thorough analysis of gamma spectrum peaks to ensure safety compliance, especially for radioactive material measurements.

Gamma spectrum analysis requires specific calibration to accurately identify and quantify radionuclides present in a sample.

Sample identification is crucial for ensuring proper analysis, and it includes details such as sample type, size, and acquisition dates.


Question: FWHM measurement significance in peak analysis

FWHM (keV) is a measurement that represents the Full Width at Half Maximum of a peak, indicating the width of the peak at its half maximum height, which helps in evaluating the resolution of the energy measurement.

The energy calibration process ensures that the detected energy levels correspond accurately to known standards, which is crucial for maintaining reliability in analysis.


Question: dissolved polonium measurements

Sample concentration is less than 5x LLD; RPD was not used for data validation. Replicate Error Ratio (RER) is less than 2. Precision judged to be in control.


Question: What are the design requirements for the spent fuel pool cooling system?

The existing Spent Fuel Cooling System is designed to maintain the spent fuel pool water temperature within acceptable limits under normal and maximum heat load conditions (Table 9-5).

The system demineralizer and filters are designed to maintain adequate purification to permit unrestricted access to the spent fuel storage area for plant personnel, provide means for purifying transfer canal and refueling pool water during refueling, and provide purification capability for the refueling water storage tank.


Question: The assessment of exemptions under 10 CFR 50.12 is crucial for understanding the licensing process and its impact on operational safety and compliance.

Exemptions pursuant to 10 CFR 50.12 currently in effect for DNPS Unit 2 and Unit 3 were reviewed to determine if they are based upon a TLAA.

SLRA Table 4.1-2 is a summary of the TLAAs identified for the DNPS SLRA.


Question: What is the maximum power level authorized for the facility?

TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.

The Technical Specifications contained in Appendix A as revised through Amendment No. 45, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.


Question: The indicators on the auxiliary shutdown control panel provide critical measurements including pressurizer level and neutron flux.

Pressurizer Level, Pressurizer Pressure, Reactor Coolant Loop D Hot Leg Temperature, and Neutron Flux are among the critical indicators available on the auxiliary shutdown control panel during hot standby.

The auxiliary shutdown control panel controls a variety of pumps, including the Nuclear Service Water Pumps and the Component Cooling Water Pumps, which are essential for maintaining the overall system.


Question: new and significant information in nuclear safety

The NEI 17-04 Revision 1 methodology, which describes a three-stage process for determining whether there is any “new and significant” information relevant to a previous SAMA analysis (NEI 2019b).

Xcel Energy finds that impacts of severe accidents for the proposed SPEO are SMALL.


Question: What is the activity of the isotope CS137?

CS137 has an activity of 1000 mCi as noted in the registration details.

The manufacturer of the device is KAY-RAY/SENSALL, INC.


Question: When a laboratory uses NELAP certification, it signifies compliance with the stringent criteria set by the NELAC standard.

Where the analytical method has been performed under NELAP certification, the analysis has met all of the requirements of the NELAC standard unless otherwise noted in the analytical case narrative.

The equipment used for testing is often calibrated according to specific protocols to ensure accuracy and reliability.


Question: The regulatory framework mandates thorough safety assessments and operational reviews for nuclear facilities.

The NRC requires a comprehensive safety analysis to evaluate the potential risks associated with reactor operations.

The new design incorporates advanced cooling technologies to enhance system efficiency and performance.


Question: What is the function of containment isolation valves in a nuclear reactor?

Phase A containment isolation valves are part of the safety system that isolates various fluid samples from the reactor environment to ensure safety measurements and compliance with regulatory requirements.

The reactor pressure vessel is a crucial component in containing the nuclear reaction and ensuring that the reactor operates safely within designed limits.


Question: What is a nuclide minimum detectable activity (MDA)?

The nuclide MDA (minimum detectable activity) is defined as the lowest activity level of a nuclide that can be reliably detected using a specific measurement technique.


Question: TER funding for regulatory audits

Now that the team has been ramped up with the TER and the technical review of the uncertainties, we’re running through the funds we have in-house.

It looks like there is another $844 to be funded.


Question: technical criteria for nuclear material harvesting

There are 6 "technical criteria" that are scored and combined for a "technical" score.

I’m hoping to use a portion of the meeting this afternoon to discuss and refine this straw man ranking approach.


Question: What is the Risk Achievement Worth (RAW) Methodology?

**Risk Achievement Worth (RAW) Methodology:** Section 4, “BTP 7-19 Position 4 Displays and Controls,” of the D3 Analysis examines the five critical safety functions from SECY-93-0087 and defines the diverse controls to achieve each critical safety function and displays to monitor the performance of these functions from the control room.

The D3 Analysis does not specify what the phrase “RAW of ≥ approximately 2” means or how the RAW values are calculated.


Question: The measurement locations selected for survey unit B2-010-103 were determined using a systematic grid approach with randomization.

The survey unit was split into eight (8) areas, with 14 required compliance measurements each, to accommodate the MARSSIM recommended survey unit size restrictions.

One hundred and twelve (112) total systematic measurements (84 interior and 28 exterior) were included in the survey design for survey unit B2-010-103.


Question: The NRC evaluates the potential for the migration of contaminants in groundwater systems based on hydraulic properties.

The much lower hydraulic conductivities of the Shire member sediments impede the horizontal transport of any contaminants through this aquitard.

Sampling of the water quality in Shire member wells at the New Rifle site indicate that any contamination derived from the site’s former mill operations is limited to the uppermost part of the Shire and has not affected the Molina member.


Question: public comment period in nuclear regulation

The Public Comment Period for the proposed rule package is scheduled from 9/8/23 to 1/22/24, allowing stakeholders to provide feedback.

The Regulatory Basis activity outlines the framework and justification for the proposed regulations in the Commission's review process.


Question: The ability to maintain supply of offsite electrical power to essential safety loads is critical for reactor operation.

From the switchyard three electrically and physically separated circuits providing AC power to each 4.16 kV ESF bus. An offsite circuit consists of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1E ESF buses.

SSCs are modeled consistent with the TS scope and so can be directly included in the ERAT for the RICT program.


Question: What are the requirements for regulatory compliance in nuclear inspections?

In establishing inspection requirements, GEH commits to compliance with 10 CFR Part 50 Appendix B, Criteria X and NQA-1-2015 Edition, Requirement 10.

The type of observation is important for determining the course of action during the inspection process.


Question: What are the fourteen requirements for license amendment requests regarding transmitter calibration intervals using OLM?

In the safety evaluation (SE), the NRC identified fourteen requirements that each licensee must address in any license amendment request (LAR) to extend transmitter calibration intervals using OLM.

The NRC staff is also aware of the setpoint of 6 differential pressure switches serving as Reactor Water Level Narrow Range function have all shifted from their previously calibrated settings by approximately the same amount in the same direction due to a systematic effect resulting from the design of the instrument.


Question: What is the purpose of the ECCS suction strainer in nuclear safety?

The ECCS suction strainers clog at the beginning of suppression pool circulation following successful automatic isolation of the pipe break, with many forms of clean RPV injection sources for high-pressure injection remaining available at Fermi 2.

The dominant condition for non-isolated pipe breaks (LOCA) is when a weld between the RPV and the first isolation valve ruptures.


Question: The RMS components are subject to various environmental conditions that involve indoor air and ventilation.

The materials of construction for the RMS components are: Carbon Steel, Stainless Steel.

Materials of construction for the Sanitary Sewage System are: Carbon Steel (Galvanized), Cast Iron (Gray Cast Iron).


Question: The inspection team assessed various systems for their readiness against adverse weather conditions before winter.

The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of low temperatures and snow advisories of the following systems on December 3, 2024: - Diverse and flexible coping strategies equipment room - Screenhouse - 'B' emergency diesel generator room - Standby auxiliary feedwater annex

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.


Question: The report details the procedures for analyzing radionuclide concentrations, particularly in soil samples.

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

LaCrosseSolutions utilized a methodical approach to ensure the accuracy and reliability of the measurements during the site restoration project.


Question: How do climatological changes affect radiation health impacts?

Climatological changes are not likely to have an effect on radiation health impacts.

Existing radiological regulations will continue to guide radiological monitoring and thresholds.


Question: analyte recovery acceptance criteria

A quality control analyte recovery is outside of specified acceptance criteria

This data report has been prepared and reviewed in accordance with GEL Laboratories LLC standard operating procedures.


Question: What is the purpose of the Lubricating Oil Analysis aging management program?

The Lubricating Oil Analysis aging management program will be consistent with the 10 elements of aging management program XI.M39 Lubricating Oil Analysis, specified in NUREG-2191.

The oil reservoir was flushed and the oil changed out.


Question: nuclide MDA report


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

RFI is focused on areas that maintain the cyber program.

Scheduling of on-site presentations (i.e., drills and exercises).


Question: What changes can be anticipated for control room operators following the implementation of new safety measures?

The effect on the control room operator to me is one of the more important -- I get all of the analysis behind the scenes and stuff like that, but it's not -- that doesn't do a lot for me to operate a plant unless I can see increased operating margins, a few more tech specs to have to comply with, and those sort of things.

I think that we can make a list but stress that it's not complete because we might leave something out.


Question: The aging management program requires consideration of material conditions affecting components like piping and heat exchangers.

Loss of Material is identified as an aging effect requiring management for the Heat Exchanger (Intake Cooling Water) Shell Side Components, which are made of Carbon Steel and operate in a Closed-Cycle Cooling Water environment.

Piping elements that use glass as a material do not exhibit any aging effects requiring management in the Closed-Cycle Cooling Water environment.


Question: GSI-191 debris failure mechanisms

Numerous tests have been performed by the U.S. NRC and industry, as well as regulators and utilities around the world over the last 40 years to resolve issues related to debris and strainer performance. This testing has investigated nearly every aspect of GSI-191 including insulation and coatings destruction from break jets; unqualified coatings failure; blowdown and washdown debris transport; containment pool settling, tumbling, and lift-over-curb debris transport; debris erosion; chemical release, solubility, and precipitation; strainer head loss, vortexing, and penetration; ex-vessel component wear; and in-vessel core blockage and boron precipitation.

Both high and low likelihood equipment configurations were addressed with a bounding evaluation as described in Section 2.4.1.


Question: What is a 500 ml Marinelli sample in gamma spectrum analysis?


Question: What is risk-informed completion time in nuclear reactor licensing?

The first LAR would modify the Grand Gulf technical specifications (TSs) to permit the use of risk-informed completion times (RICTs) in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, "Provide Risk Informed Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b" (TSTF-505) (ML18183A493).

The NRC staff has proposed to conduct a regulatory audit from December 4-6, 2023, to enhance technical understanding of the submitted documentation, with periodic meetings to be scheduled during this period, as needed, before the audit and after the audit prior to finalizing a request for additional information (RAI).


Question: ECCS flow rates in nuclear safety analysis

The Wolf Creek ECCS flow rate per fuel assembly is 52.9 gpm/FA based on the ECCS flow rate used in the in-vessel analysis for the most limiting design basis pump configuration with two ECCS trains in operation at maximum flow rate.

The Wolf Creek is a Westinghouse NSSS plant with an upflow barrel/baffle design.


Question: What is the purpose of ECCS suction strainers in nuclear reactors?

ECCS suction strainers are designed to ensure that Emergency Core Cooling System (ECCS) pumps do not become obstructed by debris during a loss-of-coolant accident, thus maintaining their ability to provide cooling to the reactor core.

Fermi, "Mechanical Engineering Design Verification Review Checklist, MES17004, Rev. 13A," January, 2023. This document outlines the necessary steps and considerations for conducting engineering design verification for nuclear system components.


Question: nuclide identification report

**Sample Title:** B1-010-001-FSFC-D10-GM  **Nuclide Library Used:** C:\GENIE2K\CAMFILES\LACBWR.NLB

**They must ensure compliance with regulatory standards to maintain operational integrity.**


Question: The fluid quality is measured on a scale from 0 to 1 in the system.


Question: What are the potential impacts of herbicide application on monarch butterflies near nuclear facilities?

Continued herbicide application could directly affect monarchs in the action area by injuring or killing individuals exposed to these chemicals.

Impacts on climate change during normal operations at nuclear power plants can result from the release of greenhouse gases from stationary combustion sources, refrigeration systems, electrical transmission and distribution systems, and mobile sources.


Question: radiographic testing requirements for nuclear reactors

The NRC must require licensees to perform a one-time inspection of the continuous circumferential transition cone closure weld on each steam generator (essential 100 percent examination coverage of each weld) employing radiographic testing is required to protect the health and safety of the public during a 60-year or 80-year period extended power operations of PRVs.

Licensee emergency plans are not sufficient to protect the health and safety of the public during a declared General Emergency - due to a Loss of Coolant Accident (LOCA) stemming from a fractured nuclear reactor vessel and resulting reactor core melt down, caused by a damaged RV from the effects of neutron fluence which caused the RV to fracture during a reactor trip and subsequent PTS event.


Question: The NRC mandates that during DBA conditions, at least one train each of containment cooling and containment spray must be operational to keep containment conditions within the specified design limits.

The modeled Containment Cooling System actuation from the containment analysis is based on a response time associated with exceeding the containment High-3 pressure setpoint to achieving full Containment Cooling System air and safety grade cooling water flow.

The effect of an inadvertent containment spray actuation has been analyzed.


Question: The NRC assesses environmental impacts during the licensing process for nuclear facilities.

The NRC evaluates the potential environmental consequences of proposed nuclear activities as part of its licensing process to ensure compliance with regulatory requirements.

In 1993, the Savannah District of the US Army Corps of Engineers published a report on water resources development in Georgia.


Question: The confidence interval for the measurements is set at a level of 95%, which provides a statistical basis for estimating uncertainty.

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

MDC refers to Minimum Detectable Concentration, which indicates the lowest concentration of a substance that can be reliably detected in a sample.


Question: hypothetical accident dilution factors

Conditions for a 0-Min to 60-Min Hypothetical Accident (North Anna, September 16, 1971 – September 15, 1972) | Time Interval | Stability Category | Wind Speed | Wind Conditions | 0-60 min. | F | 1.0 m/sec | Invariant |


Question: What is the purpose of the Radioactive Effluent Control Program in nuclear facilities?

A program shall be established and maintained to implement the requirements of 10 CFR 72.44 (d) or 10 CFR 72.126, as appropriate.

The program shall provide for evaluation and control of the following requirements during the applicable operation: verify that no TRANSFER CASK handling or STORAGE CASK handling using the lifting lugs occurs when the ambient temperature is < 0°F.


Question: Critical Heat Flux Correlation in nuclear reactor safety

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation is a significant parameter used in thermal-hydraulic safety analyses to predict the point at which coolant flow can no longer effectively remove heat from the fuel rods, leading to potential overheating and damage.

Framatome Inc. provides various services to nuclear plants, including maintenance, reactor upgrades, and support for regulatory compliance.


Question: What are Risk-Managed Technical Specifications (RMTS) in nuclear regulation?

Risk-Managed Technical Specifications (RMTS) Guidelines provide a framework for developing technical specifications that incorporate risk insights and risk-informed decision-making processes.

The references listed in the document include a variety of standards and reports related to nuclear safety evaluations.


Question: What is the definition of drain time in nuclear reactors?

A bounding DRAIN TIME may be used in lieu of a calculated value.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).


Question: What is the Auxiliary Feedwater System in nuclear plants?

The Auxiliary Feedwater System is designed to provide cooling water to the steam generators during a loss of normal feedwater supply and to ensure the ability to remove decay heat from the reactor core.

The installation, operation, and maintenance instructions serve as a guide for the operators to ensure the equipment is functioning according to the standards.


Question: What is the Materials Inspection Program in nuclear licensing?

The Materials Inspection Program is designed to ensure that materials used by licensees in their operations adhere to regulatory requirements and safety standards.

The Management Meetings section outlines various types of meetings held to discuss regulatory issues and updates.


Question: The plant's safety report comments on the analysis of large break loss-of-coolant accidents.

The Large Break Loss of Cooling Accident (LOCA) analysis is performed to ensure that the reactor can withstand significant coolant loss without jeopardizing safety.

In 2020, the NRC issued several updates regarding the safety protocols for waste management in nuclear facilities.


Question: What is the definition of tracer peak centroid in nuclear analysis?

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

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


Question: The NRC reviews the aging effects of fire protection systems on auxiliary components to ensure their safety and integrity over time.

Loss of material due to abrasion, flaking, vibration; cracking/ delamination due to chemical reaction; change in material properties due to gamma irradiation exposure; separation

There are no silicate fireproofing/fire barriers exposed to air addressed in screened in portions of Auxiliary Systems.


Question: The Oconee Station site is recognized for its historic properties and cultural significance, including potential archaeological resources.

Historic and cultural resources within the Oconee Station site can include prehistoric era and historic era archaeological sites, historic districts, and buildings; as well as any site, structure, or object that may be considered eligible for listing on the National Register of Historic Places (NRHP).

In the late 1800s, the development of a greater regional railroad infrastructure led to the establishment of several small towns centered around cotton and textile mills.


Question: The NRC specifies that all primary load-bearing structures must meet defined stress limits for both normal and severe service conditions.

All primary load bearing members must satisfy Level A stress limits

Tornado missile load may be reduced or eliminated based on a Probabilistic Risk Assessment for the facility site.


Question: Loss of Coolant Accident (LOCA)

The petition raises the concern that the nuclear power industry does not now have adequate safety margins against the consequences of a Loss of Coolant Accident (LOCA) and unduly risks another core melt accident and potentially large radioactive releases worse than what occurred at TMI.

Two such compelling commemorative presentations on the TMI accident and its consequences can be viewed at http://www.tmia.com/march26.


Question: RHR loop operation verification

SR 3.4.7.1 Verify one RHR loop is in operation.

In accordance with the Surveillance Frequency Control Program


Question: What are the requirements for assessing radiological consequences of design basis accidents?

The PPE approach in RS-002 was used for the evaluation.

The types of facilities that would be used to store radioactive liquids and any associated inventory are unique to each design.


Question: environmental site assessment for reactors

you would want to make sure that you're hiring a competent environmental professional to perform the environmental site assessment that will assess a lot of what Matt was talking about so you can establish the baselines prior to ownership.

I don't know that we can really answer, how would you assess that small nuclear reactor is affecting the environment? I'm not qualified to tell you at this time.


Question: What is the methodology for modeling transient heat transport in reactor systems?

This report presents approximate closure models to complete the scaling analysis for the idealized PHTS. This presents the methodology needed for the scaling and modeling of transient heat transport between fluids and solid structures (both the convective heat transfer from the fluid to the surface of the heat structures and conduction in the heat structures).

The integrated energy balance equation is non-dimensionalized using the following normalized parameter, in addition to the same normalized parameters used previously: \[\theta_{loop} = \frac{T_{loop} - T_{C0}}{T_{H0} - T_{C0}}  \tag{54}\] This yields: \[m_{loop}c_{loop} \frac{1}{\theta_{loop}} \frac{\partial}{\partial t} \left[ T_{C0} + \theta_{loop}(T_{H0} - T_{C0}) \right] = \epsilon_{RX}h_{RX}A_{S,RX}\theta_{reg,RX}(T_{S, RX} - T_{F,RX}) + \epsilon_{HX}h_{HX}A_{S,HX}\theta_{reg,HX}\Delta T_{LMTD,HX}  \tag{55}\]


Question: harvesting of aged materials in nuclear power plants

Harvesting of aged materials from operating and decommissioning nuclear power plants is a process that involves the careful extraction and management of older materials to ensure safety and regulatory compliance in the nuclear industry.

We’re targeting to put together a 20 minute presentation on harvesting that we can discuss with you on Thursday morning.


Question: The staff requested additional information on how the reactive control systems are evaluated during licensing procedures.

PSAR section 5.3.4.3 states: “The RIM Program and assessment of available testing and service data provide an appropriate material surveillance program for the reactor vessel (RV) without the use of material surveillance coupons as described in section 7.1.2.”

Classification for the instrumentation and controls (I&C) systems is not located in this chapter in noted tables.


Question: What does MDC stand for in nuclear analysis reports?

The sample was collected from a soil matrix and was received on 20-SEP-19.


Question: What is the function of the Spent Fuel Pool Cooling and Cleanup System in nuclear reactors?

To demonstrate operation of the SF PCCS components and their associated control circuitry.

Required electrical power supplies and control circuits are energized and operational.


Question: what is estimate uncertainty

Estimate uncertainty represents the inherent accuracy of a cost-benefit estimate; it indicates a function of the level of project definition that is available, the resources used (skill set and knowledge) and time spent to develop the cost estimate and the data (e.g., vendor quotes, catalogue pricing, historical databases) and methodologies used to develop the cost estimate.

Buildings and other structures; their functional systems and equipment; site development features such as landscaping, roads, walks, and parking areas; outside lighting and communications systems; central utility plants; utility supply and distribution systems; and other physical plant features.


Question: The containment structure is crucial for preventing the release of radioactive materials during an accident.

The containment building is designed to protect the reactor core and the public from potential releases of radioactive materials.

This figure is currently unavailable in an electronic format.


Question: The regulatory body mandates that all safety analyses align with established safety margins.

A key aspect of nuclear licensing involves the review and approval of safety analysis reports that demonstrate compliance with safety regulations.

The sample contained various isotopes of Uranium and Thorium, which are essential for understanding the radioactivity levels in the region.


Question: The employment statistics for various sectors show a consistent growth in overall jobs from 2010 to 2020.

Total Employment (Number of Jobs) increased from 29,256 in 2010 to 30,948 in 2020 in Colbert, AL.

The maximum percentage of farm employment in Lauderdale, AL was 3.6% in 2010.


Question: What is the COBRA-FLX correlation in nuclear engineering?

The COBRA-FLX correlation is a critical heat flux correlation developed by Framatome for the assessment of reactor safety under various thermal-hydraulic conditions.

The report also includes comprehensive data on various test assembly configurations and their respective performance metrics during the heat transfer testing.


Question: corrective actions for mechanical failures

The failure to promptly identify and correct a significant condition adverse to quality is a performance deficiency. It is a performance deficiency because it was reasonably within the licensee’s ability to identify and correct the condition prior to additional equipment failures.

The inspectors assessed the significance of the finding using IMC 0609 Appendix A, “The Significance Determination Process (SDP) for Findings At-Power.” The inspectors determined that the finding had very low safety significance (Green) because the finding did not represent an actual loss of the probabilistic risk assessment function provided by these four air conditioning units.


Question: advanced nondestructive evaluations

Advanced nondestructive evaluations (AND) refer to a variety of inspection techniques that assess materials or components without causing damage, ensuring the integrity and safety of nuclear systems.

Increased communication between utilities, regulators, and research is becoming essential for improving safety measures and operational efficiency in the nuclear industry.


Question: What is the role of the Radiological Duty Officer in emergency planning for nuclear plants?

Radiological Duty Officer - A designated member of the FPL General Office Power Resources staff with responsibility for responding to radiological emergencies or incidents on a 24 hour per day basis.

Site wide fire alarms are provided from the unit control rooms using the Plant Paging system which includes fire alarms.


Question: What is included in the Reactor Coolant Pressure Boundary System?

Components of the Reactor Coolant Pressure Boundary System include bolting, flow devices, heat exchangers, piping elements, and valve bodies, which serve functions such as mechanical closure and pressure boundary containment.

The aging management review results for all components will be addressed in the subsequent sections of the document, focusing on the importance of maintaining system integrity.


Question: What is the role of risk-informed assessment in nuclear inspections?

Risk-informed assessments support the determination of inspection requirements based on the risk significance of structures, systems, and components under review as described in the Risk-Informed Baseline Inspection Program.

The NRC conducts regular reviews to ensure that all reactor operating experience is thoroughly documented and analyzed to inform oversight processes.


Question: The regulatory framework overseeing water rights for oil exploration is intricate and requires various submissions to ensure compliance.

Water Right No.  SJ-1361-EXPLOR-159 indicates that the well is classified under the Abandoned status and was used for Exploration activities as authorized by NMOSE.

Mobil Oil has been the operator of multiple wells in the area, indicating a strong presence in oil exploration activities.


Question: What is a Marinelli beaker used for in gamma spectroscopy?


Question: What is the significance of nuclide data in radiation measurement?


Question: NRC outlines the necessary procedures for submitting sensitive data to ensure its protection from the public eye.

In 10 CFR 2.390, NRC specifies the procedures and requirements for persons to submit sensitive information to NRC so that it may be properly protected from disclosure.

Information on procedures and any updates is available at http://www.nrc.gov/reading-rm/sensitive-info.html.


Question: What materials are used for valve bodies in turbine generator systems?

Valve Body      | Leakage Boundary (spatial)      | Copper Alloy (≤15 Zinc and ≤8 Aluminum) | Dry Gas (Int)                | None                                | VIII.I.SP-5                                              | 3.4-1, 054      | A     |

| Valve Body      | Leakage Boundary (spatial)      | Stainless Steel                  | Closed-Cycle Cooling Water (Int) | Loss of Material                     | Closed Treated Water Systems (B.2.3.12)                   | VIII.E.SP-39     | 3.4-1, 026  | B     |


Question: The NRC mandates that all designs must accommodate the expected increases in fuel rod power...

Fuel rod power increases are evaluated to ensure safe operation within designed parameters and to assess potential impacts on the reactor's thermal-hydraulic performance.

The analysis of operational procedures is critical for maintaining efficiency and reliability in power generation systems.


Question: The Turbine Bypass System ensures proper thermal regulation by mitigating excess steam flow to various outlets.

The Turbine Bypass System is designed to provide an artificial Reactor Coolant System load by providing the following steam dump capabilities: 1. 100 percent full load steam dump to atmosphere via main steam safety valves.

Proper operation of the Turbine Gland Sealing System is verified during unit startup.


Question: containment spray system

The Containment Spray System is designed to control the temperature and pressure within the containment building of a nuclear reactor, thereby preventing the release of radioactive materials in the event of an accident.

The piping components in this table are made of stainless steel, which can provide reliable performance in various conditions, ensuring system integrity over time.


Question: nuclide activity in nuclear reports

The table summarizes the nuclide activity and confidence values, revealing important information about the isotopes present in the sample.


Question: The facility is designed to maintain safety during fire incidents by utilizing special systems and equipment, as outlined in regulatory requirements.

This Fire Area complies with 10CFR50 Appendix R, Section III.G.3.

Manual hose stations and portable extinguishers would extinguish fires and normal unit or SSS capability would be available.


Question: The table includes information on the weighted mean activity of various nuclides in pCi per gram.


Question: What are risk-informed completion times in nuclear power operations?

The Risk Informed Completion Times (RICT) Program is detailed in the Technical Specifications, which includes guidelines for assessing the impact of risk on completion times to ensure safe operation.

Renewed Facility Operating License Nos. DPR-19 and DPR-25 are essential for maintaining compliance with regulatory requirements at the Dresden Nuclear Power Station.


Question: What are the requirements for nuclear facility licensing?

For licensing a nuclear facility, the applicant must provide detailed information on the safety measures and operational protocols to ensure compliance with the Nuclear Regulatory Commission (NRC) requirements.

The turbine generator is responsible for converting mechanical energy into electrical energy, playing a crucial role in the overall operation of the power plant.


Question: Emergency Operations Facility (EOF) in nuclear regulation

The Emergency Operations Facility (EOF) provides a dedicated location for support of the site event response activities. The EOF is sized to accommodate ERO responders and NRC, FEMA, and state representatives.

Administrative controls will ensure that remote ERO responders have access to any required data and any equipment necessary to perform their emergency response duties.


Question: What are the requirements for groundwater protection in nuclear licensing?

Groundwater protection and restoration and financial assurance estimates were addressed in LBP-05-17, which mandated changes to uranium standards and required revisions to RAPs for well plugging methods.

Cultural resources were discussed in LBP-05-26, which stated that there was no change regarding their assessment throughout the review process.


Question: The table outlines the average burnup ranges for various assembly types used in Pressurized Water Reactors (PWR).

Minimum Initial Assembly Avg. | 44 < Assembly Average Burnup ≤ 45 GWd/MTU | Minimum Cooling Time (years)

The table includes enrichment percentages for different reactor assemblies such as CE and B&W configurations.


Question: The NRC mandates that every licensed nuclear facility must adhere to specific safety protocols during operation.

The technical specifications for the Surry Power Station include detailed requirements that the facility must comply with to ensure the safety of the reactor operations and the public.

The reactor coolant pump is responsible for circulating the coolant through the reactor core to maintain the proper temperature during operation.


Question: In the event of a severed air line affecting 1-LCV-2-3, what is the expected behavior of the valve?

1-LCV-2-3, DUMPBACK TO CST FROM HW PMP DISCH, will FAIL CLOSED.

The conditions state that Unit 1 is functioning at full power while the air line is compromised.


Question: what is a gas decay tank

The Gas Waste System is switching to fill the gas decay tank selected as standby.

If the operator simulates placing GDT #8/C to the “INLET” position, then provide the following information: RRV-341 green light is LIT; red light is NOT LIT. RRV-381 red light is LIT; green light is NOT LIT.


Question: What is the significance of groundwater elevation measurements in environmental reports?

Table 3.4.1-2 provides a summary of observation well groundwater levels including the TOC elevation, date measured, depth to water, and the resulting groundwater elevation in feet NAVD88.

The NRC requires that all monitoring activities comply with federal guidelines to ensure safety and environmental protection.


Question: The NRC evaluated the licensee’s adherence to safety protocols through a series of supplemental inspections conducted at the facility.

The supplemental inspection plan consists of activities specified in the Inspection Manual Chapter (IMC) 2545, 'Research and Test Reactor Inspection Program,' additional inspection activities necessary to determine that the licensee has implemented adequate corrective actions, and inspection activities to support confirmation that the licensee has met the requirements of the CO.

The U.S. Nuclear Regulatory Commission is dedicated to ensuring the protection of public health and safety, as well as the environment, from potential hazards associated with nuclear reactors.


Question: What is the significance of K-40 in nuclear activity assessments?

K-40 has an energy of 1460.75 keV, a yield of 10.67%, and an activity of 2.98E+00 pCi/grams.


Question: What is the significance of a Loss of Instrument Air in reactor operations?

An Instrument Air leak in the Turbine Building Air Header results in lowering pressures in the 1A and 1B Instrument Air Receivers.

The EOs performing QCOS 5750-09, ECCS Room and DGCGP Cubicle Cooler Monthly Surveillance call in and report the 1B Core Spray Room Cooler flow at 55 gm.


Question: What is the purpose of Metal Storage Overpacks (MSOs) in nuclear storage?

The MSO is the vertical storage module that receives, holds and protects the sealed TSC for storage at the ISFSI. The MSO passively provides the radiation shielding, structural protection, and heat dissipation capabilities for the safe storage of spent fuel in a TSC.

NONFUEL HARDWARE is defined as reactor control components (RCCs), burnable poison absorber assemblies (BPAAs), guide tube plug devices (GTPDs), neutron sources/neutron source assemblies (NSAs), hafnium absorber assemblies (HFRA), instrument tube tie components, guide tube anchors or other similar devices, in-core instrument thimbles, steel rod inserts (used to displace water from lower section of guide tube), and components of these devices such as individual rods.


Question: What are the controls for high radiation areas in nuclear facilities?

♦ RP-AA-460, Controls for High and Very High Radiation Areas

| STEP #  | STEP NO. | STANDARD                                                   | EVAL S/U | COMMENTS                                        |


Question: what is Shutdown Margin (SDM) in nuclear regulations?

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period.


Question: The analysis of samples for radionuclides is integral for understanding radiological safety in nuclear operations.

K-40 has an activity concentration of 1.358E+01 pCi/g Dry and is often analyzed for regulatory compliance in nuclear safety assessments.

Radionuclides such as U-238 are measured in pCi/g Dry and have specific levels of uncertainty associated with their activity concentration.


Question: No Significant Hazards Consideration (NSHC)

the proposed determinations that these amendment requests involve NSHC. Under the Commission’s regulations in 10 CFR 50.92, operation of the facilities in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The Commission is seeking public comments on these proposed determinations. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determinations.


Question: Classification of plant conditions in nuclear safety

Since 1970 the American Nuclear Society classification of plant conditions has been used; this classification divides plant conditions into four categories according to anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows: Condition I: Normal operation and operational transients. Condition II: Faults of moderate frequency. Condition III: Infrequent faults. Condition IV: Limiting faults.

A typical list of Condition I events follows: A. Steady-State and Shutdown Operations 1. Power operation (>5 to 100 percent of rated thermal power).


Question: What are the license renewal boundary drawings in nuclear regulation?

The License Renewal Boundary Drawings include detailed diagrams such as SLR-DRE-M-12; Sheets 1 and 2, and SLR-DRE-M-25, which are critical for understanding the spatial delineations involved in license renewal processes.

The UFSAR References consist of various sections, such as 5.2, 5.4, and 10.3, which outline necessary safety assessments and regulatory requirements for nuclear facilities.


Question: What requirements must be met for a license transfer application in nuclear regulation?

A license transfer application must include the same information concerning the applicant’s financial qualifications to operate the plant as required for an initial application.

The NRC will approve the license transfer if it finds both the proposed transferee qualified to hold the license and the transfer otherwise consistent with applicable law, regulations, and Commission orders.


Question: nominal trip setpoints

The Nominal Trip Setpoints (NTSP) for the components establish a timing threshold, with the specific component having a nominal trip setpoint of 15 seconds.

The table includes the individual component setting tolerances, where the As-Found and As-Left values are noted for various components, detailing their accuracy measurements.


Question: What is the calculation method for equipment operation hours?

The calculation method used to determine the Equipment Operation (hrs/yr) includes the number of vehicles, their capacity for trips, and the duration of operation per trip.

The passenger vehicle is rated at 200 HP and can make up to 100 one-way trips per day.


Question: The NRC evaluates the adequacy of the current analysis related to emergency procedures in the event of a significant incident.

The licensing process for the Palo Verde Units involves detailed examination of Large Break Loss of Coolant Accident (LOCA) scenarios to ensure safety compliance.

Framatome also provides reactor design services for a range of nuclear technologies and is involved in international collaborations.


Question: What is Quadrant Power Tilt Ratio (QPTR) in nuclear safety?

This Instruction provides detailed steps for determination of the Quadrant Power Tilt Ratio (QPTR) and Axial Flux Difference (AFD).

Performance of this instruction satisfies the following Surveillance Requirements (SRs):


Question: The NRC requires that compressive stress characteristics are analyzed to ensure safety during operational phases.

Figure 2.10.1.3 illustrates the relationship between strain and compressive stress for materials with a foam density of 11 lbs./ft³.

The regulatory submission includes detailed operational profiles and timelines essential for compliance reviews.


Question: The program includes regular testing and inspections to ensure the effectiveness of firefighting systems in the plant.

The DCPP Fire Protection AMP is an existing condition monitoring and performance monitoring program comprised of tests and inspections.

Visual inspections are performed through the DCPP External Surfaces Monitoring of Mechanical Components AMP (B.2.3.22) to identify conditions of corrosion and mechanical damage in the CO₂ flow path.


Question: The NRC's proposed changes may result in nuclear reactors being treated as non-radioactive waste materials.

This VLLW proposal could permit, by exemption, entire massive nuclear reactors being dumped or left as if not radioactive.

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


Question: The NRC is assessing the completeness of the construction permit application for the Kemmerer Power Station.

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

The U.S. Nuclear Regulatory Commission (NRC) is soliciting input on aspects of a proposed licensing action in Kemmerer, Lincoln County, Wyoming and providing notification about an upcoming Tribal information workshop.


Question: What are options for harvesting electrical distribution components in nuclear safety research?

Meeting to discuss options for harvesting electrical distribution components to support fire research activities.

When: Tuesday, February 28, 2017 2:00 PM-3:00 PM (UTC-05:00) Eastern Time (US & Canada).


Question: What is the significance of LCO 3.0.2 in nuclear safety regulations?

LCO 3.0.2 establishes the actions that must be implemented when an LCO is not met and: a. An associated Required Action and Completion Time is not met and no other Condition applies; or b. The condition of the unit is not specifically addressed by the associated ACTIONS.

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally.


Question: What is the importance of control system malfunction analysis in nuclear safety?

Explain how “malfunction of any control system” was addressed and describe whether and how a PICS failure analysis was developed. This information is needed to confirm adequacy of the design to perform the functions necessary to ensure TSV safety and its conformance to the design bases and acceptance criteria.

The NRC staff seeks this understanding in order to assist in determining the relationship of HIPS development and TRPS development with respect to software requirements development, in conjunction with open item 7.2.5-07 for life cycle development.


Question: The seismic capacity of the ANO-2 plant includes various evaluations to ensure safety against external events.

The assumed limiting plant seismic capacity used in the ANO-2 seismic penalty calculation has a HCLPF value of 0.30g PGA.

The IPEEE did not identify any containment building vulnerabilities with respect to seismic events.


Question: The NRC stipulates specific thresholds for identifying significant gamma peaks in radiological samples.

The sample was taken from Dairyland_NPP and has a mass of 1.226E+03 grams.


Question: auxiliary feedwater system requirements

E.1 ----------- NOTE ----------
 LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.
 ------------------------------
 Initiate action to restore one AFW train to OPERABLE status.

D. Required Action and associated Completion Time for Condition A, B or C not met.
 OR
 Two AFW trains inoperable.


Question: What is the analytical method for testing radioactive materials?

**Analytical Method:** DOE EML HASL-300, Pu-11-RC Modified

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


Question: What is the Minimum Detectable Concentration (MDC) in nuclear licensing?

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


Question: What is the role of the NRC's Section 106 environmental review?

If you have any questions concerning the NRC's Section 106 environmental review, please contact Staff Archaeologist, Jennifer Davis, by phone at 301-415-3835 or via email at Jennifer.Davis@nrc.gov, or Environmental Project Manager, Mary Richmond, by phone at 301-415-7218 or via email at Mary.Richmond@nrc.gov.

Sincerely, Michelle S. Rome
Chief
Environmental Technical Review Branch 1
Division of Rulemaking, Environmental, and Financial Support
Office of Nuclear Material Safety and Safeguards


Question: The management of loss of coating integrity is critical for maintaining the safety and functionality of various structural components.

Loss of Coating Integrity requires management through a Protective Coating Monitoring and Maintenance program (B.2.3.35).

The atmospheric conditions can significantly influence the durability and performance of carbon steel components.


Question: The analysis of extreme precipitation events, particularly local intense precipitation, is crucial for assessing flood hazards at nuclear sites.

The LIP is a measure of the extreme precipitation (high intensity/short duration) at a given location. The duration of the event and the support area are needed to qualify an extreme precipitation event fully.

A system of barriers and seals prevent water intrusion into safety related structures for events where the water elevation is up to 313.5 ft NGVD-29.


Question: What is the process for collecting subsurface soil samples in nuclear regulatory surveys?

The LTP, Section 5.7.1.1.5.2 states that in Class 1 open land survey units, a subsurface soil sample will be taken at 10% of the systematic surface soil sample locations in the survey unit with the location(s) selected at random.

All designated scan areas were scanned using a Ludlum 2350-1 paired with a Model 44-10 2”x 2” NaI detector operated in the rate-meter mode and using audio response.


Question: Containment spray system failure response

Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Failure of spray pump

Trainer B of RHR Spray has been placed in-service in accordance with 1-E-1, Loss of Reactor or Secondary Coolant.


Question: containment pressure monitoring

Containment Pressure - High 1; Containment Pressure - High 2; Steam Line Pressure - Low; Steam Line Pressure - Negative Rate - High, and SG Water Level - Low Low (Adverse Containment Environment) are critical parameters monitored during reactor operations.

Activities on electrical systems (e.g., AC and DC power) and cooling systems (e.g., essential service water and component cooling water) that support the systems or functions listed in the first three bullets should not be scheduled when a logic train is inoperable for maintenance.


Question: E-Filing system for nuclear regulatory submissions

A filing is considered complete at the time the document is submitted through the NRC’s E-Filing system.

Information about applying for a digital ID certificate is available on the NRC’s public Web site at NRC E-Submittals.


Question: The Parish EOC Director must ensure prompt communication and activation of local emergency resources during radiological incidents at River Bend Station.

The Parish EOC Director will ensure warning and notification of persons within the 10 mile EPZ, as necessary.

The procedures will define the roles of the various local agencies involved in emergency preparedness and response.


Question: What are the estimated doses for whole body and thyroid over a time span of 96 hours?

The table includes values for noble elemental organic aerosol deposition across different compartments.


Question: temporary SW jumper

Approval to use the temporary SW jumper and Intake Canal level probe is requested for a maximum of 35 days during the currently planned 2025 Unit 1 refueling outage.


Question: Hornerstown Formation

The Hornerstown Formation is a well-sorted almost purely glauconitic sand that imparts a distinctive green color to the unit.

The Shark River Formation is a glauconitic sand and mudstone.


Question: The NRC mandates that applicants provide comprehensive details regarding material properties and geometry in their submissions.

**Esteel** = 27.96E   ! Young's modulus (psi)

*Address: 12100 Sunrise Valley Drive, Suite 220, Reston, VA 20191*


Question: What is the average enrichment level for PWR fuel assemblies?

The Minimum Initial Assembly Avg. Enrichment for PWR fuel indicates various levels based on the enrichment factor (E) ranging from 2.1 to E ≥ 4.9 with specific values listed for each type of assembly.

The report also details the operational safety protocols during reactor maintenance, emphasizing the importance of thorough safety checks and procedural adherence.


Question: containment structural integrity examinations

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

Measurement of tendon force applied at the end anchorage.


Question: The specifications for the pressurizer shell and surge nozzle dimensions must meet the criteria established in regulatory documents to ensure safety and compliance.

The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11), Figure 1-2 (Item No. B2.12), Figure 1-3 (Item Nos. B2.21 and B2.22), and Figures 1-4 and 1-5 (Item No. B3.110).

The inside diameter of the pressurizer shell must adhere to the listed geometric parameters to ensure structural integrity.


Question: What are the requirements for nuclear plant licensing?

The NRC requires applicants to submit a detailed safety analysis report that demonstrates the safety of the proposed nuclear facility and its systems.

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


Question: The analysis incorporates evaluations similar to those described in NUREG/CR-6595 for various aspects of the LERF assessment, which relies on conservative estimates.

The Large Early Release Frequency (LERF) analysis uses NUREG/CR-6595 type evaluations for some portions of the LERF evaluation based on conservative assessments.

Most of the notebooks indicate that interviews with knowledgeable plant personnel were conducted to confirm that the systems analysis adequately reflected the as-built, as-operated plant and that plant-specific data was appropriately collected where required.


Question: lamellar tears in reactor materials

Let's look at the third condition where we have potential stresses through-the-thickness and potential lamellar tears present, and see what we can do with that one.

For the most severe shake, a long slender flaw, assuming the Q is for stress over yield stress of about a half, which is about what this turns out to be for that deep buried flaw, the free surface boundary connection is one, and I estimate A to be 1.45 inches.


Question: probable maximum flood definition

The probable maximum flood is an estimated flood that may be expected from the most severe combination of critical meteorologic and hydrologic conditions that are reasonably possible in the region.

The flooding effects on the headwater area of the cooling lake were also evaluated during the dam design and were determined by backwater computations using the USACOE's computer program, "Water Surface Profiles" (CPS, 2002).


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

The Safety Analysis Report (SAR) is a comprehensive document that assesses the safety of a proposed nuclear facility, detailing potential risks and safety measures.

The NRC conducts routine inspections to ensure compliance with regulatory standards and assess the operational performance of nuclear facilities.


Question: probable maximum precipitation methodology

The Probable Maximum Precipitation (PMP) is a critical concept in hydrological studies used to estimate the maximum precipitation that could occur at a specific location, derived from historical data and hydro-meteorological modeling techniques.

The method employed for calculating runoff from rainfall involves using the Snyder Unit Hydrograph, which is a widely recognized technique for estimating the response of a watershed to storm events.


Question: Sealed source leak limit


Question: What are power uprate licensing actions in nuclear plants?

The licensing actions related to power uprate at nuclear facilities involve modifications to the existing operating licenses to allow for an increase in the reactor's thermal power output beyond its originally licensed capacity.

The NRC conducts routine inspections of nuclear facilities to ensure compliance with safety regulations and operational requirements.


Question: What is the purpose of a materials license in the context of nuclear regulatory activities?

A materials license issued by the U.S. Nuclear Regulatory Commission permits the holder to possess and use radioactive materials for specified purposes in accordance with regulatory requirements.

The amendment dated January 8, 2019, included various updates but did not introduce any new licensing requirements or changes to existing protocols.


Question: nuclear reactor capacity

The Vogtle Power Plant in the USA has a total capacity of 2 x 1000 MW, which significantly contributes to the electrical grid.

The Södra Cell VARO in Sweden, manufactured by Doosan Skoda, has a production capacity of 55 MW and was delivered in 2015.


Question: What are the surveillance requirements for containment sumps?

SR 3.6.8.1 Verify, by visual inspection, the containment sumps do not show structural damage, abnormal corrosion, or debris blockage.

The NRC requires licensees to report any modifications to the reactor structure that could affect safety.


Question: What is the weighted mean activity of nuclides in nuclear safety reports?

The weighted mean activity is calculated by combining the activities of different nuclides with their respective uncertainties to provide a more reliable measure of overall radioactivity.


Question: What are the requirements for stormwater flow measurement in nuclear facilities?

All stormwater samples shall be collected from the discharge resulting from a storm event that is greater than 0.1 inches.

The permittee shall complete a review and evaluation of the BMP plan at least once every three years from the date of preparation of the BMP plan.


Question: What is the purpose of the AMP XI.M23 program?

AMP XI.M23, "Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems" will be used to manage cracking, deformation, and loss of material of the carbon steel cranes, hoists, and platforms; and the associated rails, bridges, structural members, and structural components exposed to air - indoor uncontrolled in the Cranes, Hoists, and Refueling Equipment System.

The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24) program will be used to manage loss of material of the carbon steel, cast iron, and galvanized steel accumulators, compressor housings, flame arrestors, heat exchanger components, piping, piping components, silencers/mufflers, tanks, and valve bodies exposed to condensation in the Closed Cycle Cooling Water System.


Question: nuclear fuel performance metrics

Figure 5-2 shows the FGR log (Predicted/Measured) as a function of burnup, illustrating the relationship between the fuel's performance and its exposure to neutron flux over time, providing vital information for predicting fuel integrity and behavior.

The system's capacity for reactor pressure vessel cooling is essential to maintaining operational safety, yet its efficiency can vary significantly based on external environmental conditions.


Question: bypassing nuclear instrument protection systems

D. The BYPASS FUNCTION switches are placed in BYPASS on the failed Nuclear Instrumentation Protection Panel.


Question: The ground motion frequency during seismic events can significantly affect the safety assessments of nuclear power plant structures.

Based on these studies, methodology for the reduction of spectral accelerations in the high frequency range have been developed (Reference 136).

However, these high spectral accelerations are accompanied by small displacements and most sub-systems have adequate energy dissipation capability to accommodate such small displacements without failure.


Question: What is the significance of sump strainer structural analysis in nuclear reactor safety?

Back flushing of the sump strainers is not credited in the PBN analysis; therefore, no structural analysis considering reverse flow was performed.

In Safety Evaluations dated June 6, 2005, the NRC reviewed and accepted analyses demonstrating that a rapidly propagating failure of the large bore RCS piping components at PBN is highly unlikely (leak before break analyses).


Question: What is the purpose of the ASME Code, Section XI in nuclear inspections?

the examinations and tests meet the Inspection Plan as required by the ASME Code, Section XI; and (c) the repair/replacement activities and evaluations supporting the completion of 1R25 conform to the requirements of the ASME Code, Section XI.

The 2007 Edition through 2008 Addenda is applicable to the 4th Inspection Interval.


Question: What is a monobloc guide tube in nuclear reactors?

A monobloc guide tube is designed with a constant outer diameter and features weep holes to facilitate the functioning of control rods in a nuclear reactor.

The heat exchanger operates using water as a medium to transfer heat from the reactor to the turbine system.


Question: What is the purpose of drywell air lock in nuclear safety?

The drywell air lock leakage rate limit does not reflect the ability of the drywell to perform its safety function.

The remaining OPERABLE drywell isolation valve in the flow path is adequate to perform the isolation function.


Question: What are unregistered wells and their implications for groundwater rights?

Unregistered wells can impact water rights and the overall management of groundwater resources within a defined area, as they are not officially documented with regulatory agencies like the New Mexico Office of the State Engineer (NMOSE).

The project area includes various well types, such as domestic and livestock wells, which serve different purposes and can affect the water resource management of the surrounding environment.


Question: In the context of nuclear operations, isolating systems like the RBCLC Supply is critical during maintenance activities.

Isolate RBCLC Supply to the Drywell is an essential task to ensure safety during operations and maintenance activities.

The simulator used during training was equipped with various interfaces that simulate real-world scenarios.


Question: The TSC must be subjected to helium leak testing to ensure confinement integrity before its commissioning.

The TSC shell, bottom plate, all confinement welds, and the COMPOSITE CLOSURE LID shall be fabrication helium leak-tested in accordance with ANSI N14.5 to leak-tight criterion.

The unit weight (i.e., density) of the concrete in the CONCRETE CASK LID can be verified by either test method ASTM C138 or an approved shop fabrication procedure by following the basic equation of p=W/V.


Question: minimum decommissioning fund requirements

NUREG-1307, Revision 18, "Report on Waste Burial Charges: Changes in Decommissioning Waste Disposal Costs at Low-Level Waste Burial Facilities", explains the formula acceptable to the NRC for determining the minimum decommissioning fund requirements for nuclear power reactors, as required by the NRC's regulations.

The report provides adjustment factors, and updates to these values, for the labor, energy, and waste burial components of the minimum formula.


Question: What is the purpose of using a Marinelli beaker in gamma spectrum analysis?

The 500 ml Marinelli is utilized in gamma spectrum analysis to ensure a consistent geometry for the measurement of radioactivity in soil samples.

The Facility located at Dairlyand_NPP is equipped with advanced systems for managing nuclear reactor operations and ensuring safety protocols.


Question: The evaluation of contamination levels in the soil requires adherence to established release criteria as per operational guidelines.

For survey unit L1-010-107, compliance with the unrestricted release criteria was demonstrated through a combination of soil scanning with a Ludlum Model 44-10 gamma detector and the sampling of soil for isotopic analysis.

An FSS Supervisor performed a visual inspection and walk-down of the survey unit on August 6, 2019, prior to performing FSS.


Question: What are the accuracy requirements for instruments under the ASME OM Code?

Accuracy. Instrument accuracy shall be within the limits of Table ISTB-3510-1. If a parameter is determined by analytical methods instead of measurement, then the determination shall meet the parameter accuracy requirement of Table ISTB-3510-1 (e.g., flow rate determination shall be accurate to within ± 2% of actual). For individual analog instruments, the required accuracy is percent of full-scale. For digital instruments, the required accuracy is over the calibrated range. For a combination of instruments, the required accuracy is loop accuracy.

Unit 1 and 2 HPCI pump suction pressure indicators (1) 2E41-PI-R004 exceed this Code allowable range limit.


Question: What does mCi stand for in nuclear licensing?

mCi stands for millicurie, which is a unit of measurement used to express radioactivity in nuclear materials.

The manufacturer of the device is RONAN ENGINEERING COMPANY, which holds the manufacturer license number IND-267-95G.


Question: Are facility licensees required to update their reactor models according to regulations?

The Commission in its statements of consideration during the rule making, emphasized that facility licensee’s would not be required to update their core models in order to comply with the requirements of 55.46.

The fact that a simulator may lead or lag design changes made to the reference plant is accommodated by ANSI/ANS-3.5 (2009, 1998, 1993, and 1985).


Question: The licensing process for nuclear participants requires documentation of compliance with established nuclear regulations.

Nuclear licensing refers to the process by which the U.S. Nuclear Regulatory Commission (NRC) authorizes facilities and activities that involve nuclear materials, ensuring they meet safety, environmental, and security standards.

In addition to safety evaluations, the NRC also conducts public hearings to address concerns about the potential environmental impacts of nuclear power generation.


Question: The NRC evaluates various stability classes based on hourly observations for proper assessment.

Based on Pasquill-Turner calculation of E, F or G from hourly surface observations.

The data collected shows the mean values over several decades for various durations.


Question: How is the transient population assessed for nuclear plant safety?

Table 2.1.3-38 summarizes the transient population annular rings and enclosed population within a 0 to 10 mile radius of the VEGP.

Section 2.1.3-14 displays population by segment for the West-Northwest Sector, which does not directly pertain to transient populations near nuclear facilities.


Question: The NRC guidelines require detailed documentation of operational scenarios affecting safety systems.

NRC:  - Critical Task #1 needs to be more descriptive. Revise as follows:  - With a reactor scram required and the reactor not shutdown, (1a) Inject Boron BEFORE 110°F in the Torus -OR- (1b) Insert Rods (all or all but one) BEFORE exceeding Primary containment design limits [these must be measurable values].

NRC:  - Can the simulator operator tell when the 2nd control rod is selected?
    - If so, revise the Lead Evaluator Note to be a Sim Op Note.
 Response: Fixed.


Question: What are Insurge and Outsurge transients in nuclear plants?

Insurge/Outsurge transients are listed in Tables 5-10 and 5-11 for the Westinghouse/CE and B&W plants, respectively. As shown in Tables 5-10 and 5-11, the evaluated number of Insurge/ Outsurge cycles for a typical B&W plant significantly exceeds those for Westinghouse/CE plants (11,100 versus 5,100 cycles). Furthermore, the maximum temperature difference between the pressurizer fluid and the surge nozzle fluid temperatures is higher for the B&W plants (400°F versus 320/330°F).

The general transients do not vary significantly from one plant design to the other.


Question: The update to the Integrated Safety Analysis (ISA) for the GLE Test Loop is submitted as required by 10 CFR 70.72.

In accordance with 10 CFR 70.72(d)(2) Global Laser Enrichment LLC (GLE) submits an update to the Integrated Safety Analysis (ISA) Summary for the GLE Test Loop.

The GLE Test Loop ISA Summary, Rev.18 which contains Secret Restricted Data is being transmitted in accordance with 10 CFR 95.39, separately from the GNF-A SNM-1097 ISA Summary, Rev. 28, dated January 30, 2025, M250034.


Question: What is Zr-4 M/P in the context of hydrogen measurement?

Figure 3-7 shows the variations of hydrogen levels measured against theoretical levels for Zircaloy-4 M/P sections in a reactor.

The control room monitors various parameters to ensure safe reactor operations, including temperature and pressure readings.


Question: What is the GEXL98 correlation in nuclear engineering?

The GEXL98 correlation uses the same functional form as previous GEXL correlations with different constraints for correlation coefficient parameters based on a hypothetical database created using Framatome’s approved correlation (ACE/ATRIUM 10XM).

The total uncertainty in the GEXL98 correlation's critical power predictions accounts for the fact that there is uncertainty in the correlation's fit to the Framatome, Inc. data provided to GNF-A.


Question: What are Hot Zero Power Tests in reactor licensing?

Hot Zero Power Tests include measurements such as the isothermal temperature coefficient for different bank configurations, reactivity worth of rods, and power distribution measurements related to pseudo-rod-ejection scenarios.

The regulatory authority ensures that all safety systems are tested under various conditions to verify their reliability and functionality.


Question: The Reactor Water Cleanup System is essential for ensuring the purity of reactor water by eliminating corrosive substances and deposition on heat exchangers.

The RWCU System maintains high reactor-water purity to limit chemical and corrosive action, thereby limiting fouling and deposition on heat transfer surfaces.

The component types that require aging management review are indicated in the below Table 2.3.3-21, Emergency Equipment Cooling Water System.


Question: What is a primary containment isolation valve (PCIV) in nuclear reactor operations?

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

The Completion Time of 8 hours for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown.


Question: underground piping inspection requirements

Revise implementing procedures to require that inspections for cracking be conducted on piping with degraded coating to assess the impact of cracks on the pressure boundary function of the component being visually inspected.

The inspection will include excavation of the top surfaces and at least 50 percent of the side surface to visually inspect for cracks in the backfill that could admit groundwater to the external surfaces of the component.


Question: nuclear reactor safety inspections

The U.S. NRC conducts thorough nuclear reactor safety inspections to ensure compliance with regulatory requirements and to identify any potential safety issues that could affect the operation of the facility.

The inspection team also evaluates the training programs of the operational staff to ensure that they are well-prepared for emergency situations.


Question: The evaluation of the survey unit's compliance with release criteria was confirmed through statistical analysis following the DQOs.

The survey unit meets the unrestricted release criteria with adequate power as required by the DQOs.

No anomalies were observed during the performance or analyses of the survey.


Question: What are the applicable regulatory requirements for nuclear licensing?

The applicable regulatory requirements and criteria for nuclear licensing are outlined in regulations set forth by the NRC, which ensure that safety, environmental, and operational standards are met.

The environmental consideration section discusses the impact of the proposed change on local ecosystems and air quality but does not address specific licensing requirements.


Question: dose based license termination regulation

The licensee elaborates that it has adopted an administrative dose standard of 15 mrem/yr and explains the factors influencing this decision.

Section 5.4.5.3 of the LTP discusses the use of area factors and the DCGLEMC for the NS Savannah.


Question: The licensing process for nuclear facilities involves comprehensive assessments and evaluations to ensure compliance with safety regulations.

The nuclear licensing process requires a detailed analysis and submission of safety evaluations to the NRC to assess compliance with regulatory standards.

Nuclear power plants operate by using nuclear fission to generate heat, which is then used to produce steam that drives turbines.


Question: The NRC requires regular monitoring of radionuclide levels in the environment, including assessments of tritium concentrations.

The analysis of the water temperature indicated a measurement of 18.700 degrees Celsius.


Question: The NRC monitors radiation levels using various detector types to ensure safety standards are met.

The Scan Logged Result (cpm) represents the counts per minute detected by the radiation detector during a specific scan.

The graphical representation of the data helps in tracking the history of measurements taken over a period.


Question: The measurement of radioactive isotopes often includes activity (pCi/M²) which represents the concentration of radioactivity over a surface area.

The activity (pCi/M²) indicates the level of radiation emitted by a nuclide per square meter.

Understanding the decay levels of various isotopes can assist in radioactive waste management.


Question: What are the estimated doses of radioactivity from drinking water for infants?

The total dose estimates for adult and teenage populations vary significantly due to the differences in consumption and absorption rates.


Question: What are the steam generator supports in nuclear plants?

We need to discuss the steam generator supports on the North Anna plant units 1 and 2.

We had invited all parties to the OL proceeding for North Anna 1 and 2 to this meeting.


Question: The BOP must verify the operational status of the containment ventilation for Unit 2 following an incident.

**NOTE**: Loss of shutdown power could result in a loss of SFP cooling. Annunciator Window 128-A and 128-B could be indicative of this condition and increased monitoring of SFP level, temperature and radiation levels will be necessary...

- Manually STOP fans AND NOTIFY TSC if any damper NOT CLOSED.


Question: How does the NRC ensure compliance with biocide standards in plant discharges?

The means developed by the applicant for limiting biocide levels to standards can be designed to be effective even though routine verification by direct measurement is not practical.

A further problem is to verify that the standards are actually met.


Question: The assessment of penetration containment isolation valves is crucial in maintaining system integrity during operational anomalies.

For sequences requiring drywell spray, success of vacuum breakers was assumed in the PRA, and the possibility of failure was screened as insignificant.

Columbia is not an outlier in the use of this generic industry accepted data that addresses the operating experience-based probability of containment release pathways being larger than 'small'.


Question: The table summarizes the pathways involving containment isolation faults within the risk assessment of the Salem Unit.

This table lists the containment pathways modeled in the containment isolation fault tree logic of the Salem PRA (Table 5-1 of [21]). Not all containment penetration pathways are summarized in this table (e.g., those pathways that are scrubbed or within a closed loop system would not be a "High" release magnitude and are not listed here nor modeled in the containment isolation fault tree logic of the Salem PRA).

Seismic events are considered for their potential impacts on the performance and safety of reactor systems, but they are not specifically implicated in the pathways detailed in the table.


Question: The likelihood of experiencing an internal fire scenario is evaluated through the CNP Internal Fire PRA, which examines the potential risks involved.

The CNP Internal Fire PRA includes evaluation of risk from internal fire events.


Question: What is the method for determining the QPTR using the operational power detectors?

The QPTR is calculated for each operable power detector according to the following equations and recorded in Attachment 4: A. QPTRₜ = Ratioₜ / Averageₜ B. QPTRᵇ = Ratioᵇ / Averageᵇ.

To ensure accurate power measurements, various operational safety checks must be conducted routinely on the reactor systems.


Question: What is the importance of risk assessment in nuclear safety?

Risk assessment is a fundamental process in nuclear safety that involves the identification, evaluation, and prioritization of risks associated with nuclear operations.

The document outlines a management review process regarding the nuclear project timelines and milestones.


Question: What are the safety systems used in nuclear reactors?

In nuclear reactor design, safety systems are critical for preventing accidents and mitigating consequences in the event of a failure.

Honeywell International produces various models such as 1201/2201 Series and 4201 Series, which are widely used in industrial applications.


Question: Reactor Protection System (RPS) instrumentation requirements

LCO 3.3.1 Four channels of RPS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

**C. Three or more channels inoperable.** C.1 Enter the Condition referenced in Table 3.3.1-1 for the Function.


Question: The NRC specifies that Class 1 piping components must be evaluated for their ability to manage various aging effects.

Piping, piping components: Class 1 greater than or equal to 4" NPS | Pressure Boundary                     | Stainless Steel | Treated water >93°C (≥200°F) (Internal) | Cracking due to SCC, IGSSC                                     | Water Chemistry (B.2.1.2)                  | IV.C1.R-20       | 3.1-1, 097               | A

Aging management must include monitoring environmental conditions to prevent degradation of components.


Question: debris source term management

Procedures are in place to control maintenance activities and evaluate temporary changes that have the potential to affect the debris source term.

At the completion of work activities inside containment during normal operations, the work area is thoroughly cleaned and inspected (including the area below the work activity, if the work was performed on grating), prior to leaving containment.


Question: What is the intended function of the reactor coolant pressure boundary system?

The reactor coolant pressure boundary system is designed to provide a leakage boundary, ensuring that any leakage from the reactor coolant does not affect safety and environmental conditions.

The Dresden Nuclear Power Station, Units 2 and 3, is a facility that has undergone extensive evaluations and inspections to ensure compliance with both safety and operational standards.


Question: What is the significance of the LNT model in dosimetry for nuclear reactors?

The NRC staff notes that TR Section 3.6.6, “Dosimetry,” states “dose is computed with the [linear no threshold (LNT)] model.” Though dose is an input to the LNT model, the LNT model applies to health effects and not dose.

The NRC staff requests that example calculations, or other resources, be made available for audit to assist the NRC staff in understanding how the shine dose conversion factor (SDCF) calculation is being implemented in this EM.


Question: What are the technical issues related to license renewal for nuclear power reactors?

The NRC is seeking public dialogue related to license renewal for nuclear power reactors, specifically the need (and timing) for the NRC to consider the potential technical issues and guidance document development related to license renewal that would authorize operation for up to 100 years, including:

1A - Presentation (no slides) NEI (Earls)


Question: What is the concentration level of radionuclides measured, such as K-40 and U-238?

Activity Concentration is provided in pCi/g Dry for each radionuclide identified in the table, like K-40 with an activity concentration of 1.03E+01.

The units of measurement provided in the table specify pCi/g Dry as the standard unit for reporting activity concentrations.


Question: ultrasonic testing measurements

In response to NRC Generic Letter 87-05, TVA reported in a 1988 follow-up letter the ultrasonic test (UT) measurements for all three BFN units and indicated no reading below nominal 1" plate thickness.

The floor inside the drywell is poured concrete that extends up to elevation 549.92’. A 1” slot around the perimeter of the drywell floor provides for the installation of a joint sealer at the intersection of the concrete floor and drywell shell.


Question: What is the purpose of PWR FLECHT in reactor safety?

PWR FLECHT (Full Length Emergency Cooling Heat Transfer) is a methodology to analyze the thermal responses of reactor fuel rods during emergency situations.

The report discusses various aspects of reactor operations, including maintenance schedules and personnel training.


Question: What are radiation monitoring and exposure controls in nuclear safety?

SDS-RP3-PCD-1014, Radiation Monitoring and Exposure Controls, outlines the measures and instrumentation used to ensure that radiation exposure to personnel and the environment is monitored and controlled effectively.

9.2.1.2 IOEP-2, SONGS ISFSI-Only Emergency Plan EAL Technical Bases Manual provides details on the emergency action level procedures specifically for the San Onofre Nuclear Generation Station.


Question: Reactor Coolant System leakage monitoring

To ensure that the licensee properly monitors for Reactor Coolant System (RCS) pressure boundary leakage or potential unidentified leakage exceeding TS limit, the inspector should routinely determine if the licensee:

Any radiation dose implications associated with repetitive tasks should be reviewed by applicable radiation safety baseline IPs. In the control room or other appropriate locations, review visible portions of radiation monitors or other indications that could provide indication of an apparent uncontrolled release.


Question: uranium oxide fuel assembly specifications

1. Uranium oxide Connecticut Yankee INTACT FUEL ASSEMBLIES listed in Table B2-4 and meeting the following specifications:  a. Cladding Type: Zircaloy or stainless steel as specified in Table B2-4 for the applicable fuel assembly type.

The Connecticut Yankee facility has undergone a series of scheduled maintenance outages to ensure the reliability and efficiency of its operations.


Question: What are the EFPY calculations used in nuclear regulatory filings?

The operating periods used in the growth rate/EFPY calculations and voltage projections are as follows: Cycle 24 - 1.373 EFPY (actual); Cycle 25 - 1.437 EFPY (assumed).

For conservatism, a growth rate curve which bounded the growth rates of the past three cycles was used (Figure 3-5, Figure 3-6 and Figure 3-7).


Question: The NRC mandates a thorough evaluation of safety systems and structures associated with the MAGNASTOR system.

The MAGNASTOR System is designed to safely store spent nuclear fuel and integrates advanced safety features to mitigate risks.

The regulatory filing includes various amendments that address the compliance and operational protocols for storage systems.


Question: reactor containment spray systems

Design Considerations of Reactor Containment Spray Systems, Part III, The Corrosion of Material in Spray Solutions, USAEC Report, ORNL-TM-2412, Part III, Oak Ridge National Laboratory, December 1969.

The Handbook of Hydraulic Resistance provides detailed coefficients for local resistance and friction for various flow scenarios.


Question: What is a spent fuel storage rack analysis?

Spent fuel storage rack analysis description is provided in Section 7.

Convergence checks are described in Section 6.


Question: The digital modernization efforts at Turkey Point Units 3 and 4 aim to enhance regulatory safety measures.

The diversity and defense-in-depth evaluation is critical for ensuring the safety and reliability of digital systems in nuclear power plants.

The main purpose of this document is to outline the operational changes in the Florida Power & Light Co. regarding their energy generation strategies.


Question: what is minimum detectable concentration

The project includes a comprehensive analysis of the site restoration process to ensure compliance with environmental regulations.


Question: resident population distribution

PSEG Site 10 Mile Resident and Transient Population Distribution – 2010

PSEG Site 10 Mile Resident Population Distribution – 2000


Question: What is the impact of SIT pressure on the performance analyses of CE16STD and CE16NGF fuel?

The proposed change to the upper limit of the SR 3.5.1.3 SIT pressure band impacts the ECCS performance analyses for Westinghouse CE16STD and CE16NGF fuel.

The CE16STD large break LOCA AOR limiting PCT and PLO cases have been re-evaluated for an increase in their modeled maximum SIT pressure to 675 psia.


Question: The assessment of the Clinton Power Station’s systems was according to NRC guidelines during the recent inspection.

The integrated inspection report for the Clinton Power Station includes evaluations of the facility's compliance with regulatory requirements and safety standards.

Regular maintenance procedures are essential for ensuring the longevity and reliability of all power systems within the facility.