Question: What structures are important in nuclear facility scoping and screening?

The following structures and commodity groups are addressed in this section: Circulating Water Inlet Tunnel, Component Supports, Cooling Water Structures, Crib Houses, Diesel Generator & HPCI Building, Insulation Commodity Group, Isolation Condenser Pump House, Main Control Room and Auxiliary Electric Equipment Room, Primary Containment, Radwaste Structures, Reactor Building, Stacks, Structural Commodity Group, Switchyard Structures, Turbine Building, Yard Structures.

The total number of components analyzed in this report is significant, ensuring that all potential safety concerns are addressed during the licensing process.


Question: What is the purpose of monitoring noble gas emissions in nuclear facilities?

Operation for noble gas monitoring requires careful measurement and analysis to ensure safety and compliance with regulatory standards.

The auxiliary building is critical for housing support systems necessary for the operation of the nuclear facility.


Question: USGS gages on the Savannah River

#02187252 - USGS Gage on the Savannah River below Hartwell Lake near Hartwell, Georgia

Burtons Dam, GA is part of the NWS Sub-Watersheds in the Savannah River Watershed.


Question: The Turkey Point Nuclear Plant's emergency procedures are tailored to precisely fit its operational context.

This Turkey Point Nuclear Plant (PTN) Emergency Plan Annex supplements the NextEra Common Emergency Plan by providing site-specific information unique to the station.

PTN includes two 802 MWe Westinghouse pressurized water reactors (Units 3 and 4) located on a 3,300-acre site two miles east of Homestead, Florida, United States, next to Biscayne National Park located about 25 miles (40 km) south of Miami, Florida near the southernmost edge of Miami-Dade County.


Question: loss-of-coolant-accident

Loss-of-coolant-accident ducts (both run up the wall of the containment building) were sources of exposure to personnel climbing up to the polar crane.

The decontamination and dose reduction efforts focused on tasks to minimize cumulative worker exposures through the end of the cleanup.


Question: What are the evaluation criteria for RadICS communication methods?

A complete safety RadICS based system design will require further evaluation against this guidance.

Addressed in PSAI 7.10 and satisfied by: - Licensing Technical Report


Question: The NRC mandates a thorough risk evaluation when considering a change to the plant configuration under the Risk Informed Completion Time Program.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

The safety systems must ensure that operations remain within normal parameters to prevent any incidents.


Question: What are the exposure limits for workplace noise according to NIOSH?

The National Institute for Occupational Safety and Health (NIOSH) recommends an exposure limit for workplace noise of 85 dBA for a duration of 8 hours per day (NIOSH 1998).

During construction, wildlife likely will avoid areas where noise-generating activities are ongoing.


Question: non-transportation emissions

Table 13 summarizes the estimated non-transportation and transportation-related emissions from combustion, listing the emissions of THC, NOx, CO, PM10, SO2, and CO2 for various equipment types.

The table above outlines the operational phases of equipment types, highlighting their respective emissions metrics due to different activities.


Question: ITAAC notification types

ITAAC notification types include ICN (Initial Certification Notification), IPCN (Individual Performance Certification Notification), and UIN (Unsolicited Notification) as specified in regulatory filings.

The NRC conducts regular inspections to ensure compliance with the operational requirements set forth in the facility’s license.


Question: maximum water level at CNS

In accordance with the guidelines in ANSI/ANS-2.8-1992 (Reference 2.2-9), the maximum PMF level at the plant site needs to consider the wind setup and wave runup effect from the coincidental occurrence of a 2-year design wind event.

Model results indicated that velocities in that area peaked between 5 and 6 fps.


Question: What are the energy levels and yields for various radionuclides?

CS-137 has an energy of 661.65 keV and a yield of 85.12%.

The table displays different units of measurement for assessing radioactive materials.


Question: The NRC standards require accurate measurements of wind speed and direction to ensure safety in operational environments.

Aerovane Model 120 has a speed range of 2.5-3.0 mph and an accuracy of ±1.5°.

Calibration of instruments must be performed regularly to ensure data integrity and operational efficiency.


Question: nuclear licensing process

The nuclear licensing process consists of multiple steps that ensure the safety and effectiveness of nuclear facilities, including thorough assessments of design and operating plans.

The components used in advanced reactors are carefully selected based on their performance characteristics and ability to withstand extreme conditions.


Question: What does Specification 3.0.5 establish regarding MODE changes?

Specification 3.0.5 establishes limitations on MODE changes when a Limiting Condition for Operation is not met.

For example, if HOT STANDBY is reached in 2 hours, the time allowed to reach HOT SHUTDOWN is the next 11 hours because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours.


Question: The transfer of a licensed nuclear device must be properly documented and communicated to the NRC.

This device was transferred to RAM Services, Inc. (License No. 071-1234-01) on August 20, 2020 for disposal.

The Metropolitan St. Louis Sewer District submitted multiple certified mail letters documenting the nuclear device transfer to the US Nuclear Regulatory Commission but did not receive confirmation of receipt.


Question: What is the purpose of the control room ventilation system in a nuclear plant?

The control room ventilation system is designed to ensure proper air quality and temperature control within the control room environment, thereby maintaining operator performance and safety during normal and emergency conditions.

The mechanical closure components are critical for ensuring that the pressure boundary remains intact, preventing any unintended release of radioactive materials.


Question: The NRC specifies that the developed head of each auxiliary feedwater pump must be verified according to the specified flow test requirements.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program


Question: scaling methodology in reactor testing

The Scaling Methodology for the Kairos Power Testing Program describes how the company evaluates the safety and performance of its reactor designs through systematic testing procedures that apply scaling laws and principles.

The affidavit supporting the request for withholding from public disclosure under 10 CFR 2.390 outlines the legal basis and justification for keeping certain documents confidential, ensuring that sensitive information is not released to the public.


Question: minimum initial assembly average enrichment

Minimum Initial Assembly Avg. Enrichment


Question: What are the assumptions made about the maximum temperature that the PSW concrete can withstand in relation to gamma heating?

It is shown that gamma heating can elevate the temperature within a foot or so inside the biological shield wall concrete.

1. Discuss differences in ventilation induced air flow with that of natural within the cavity (in ft/sec).


Question: Prior regulatory approvals are necessary before constructing new nuclear units to ensure compliance with safety standards.

If Dominion decides to proceed with the development of new units, it would enter into and obtain appropriate regulatory approvals to purchase or lease the ESP site from Virginia Power and ODEC.

The parties would also agree to work cooperatively to control third-party activity that might otherwise present an unacceptable hazard to nuclear operations.


Question: What are the aging management considerations for concrete in nuclear facilities?

Reduction of strength and modulus of elasticity due to elevated temperature (>150°F general; >200°F local) is a significant aging effect for concrete components in nuclear facilities.

Not Applicable. This Item Number is not applicable to the BFN Mark I steel containment. See Subsection 3.5.2.2.1.1.


Question: surveillance frequency requirements for diesel generators

This Surveillance shall be conducted on only one DG at a time.

Verify each day tank contains ≥ 500 gallons of fuel oil.


Question: The scram air header must be vented according to the procedural guidelines established in QCOP 0300-28.

The NSO has directed you to vent the scram air header.

Some operations that are performed from outside of the control room may require multiple steps.


Question: migrant workers statistics

The regulatory structure for nuclear power plants includes multiple layers of oversight to ensure safety and compliance with federal and state laws.


Question: pool-type reactor

The facility is planned to be a pool-type reactor that uses liquid sodium as the coolant instead of water.

The NRC staff is preparing an environmental impact statement (EIS) on the proposed action of whether to issue a CP for the facility, as provided by 10 CFR 51, "Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions," the NRC's environmental protection regulations that implement the National Environmental Policy Act of 1969, as amended (NEPA).


Question: The NRC mandates that operators follow specific procedures when a control rod cannot be driven appropriately.

The SRO directs ATC to enter and take actions per QCOA 0300-02, Inability to Drive a Control Rod: Control Rod Stuck.

Based on the QNE input, the ATC may increase Drive Water Pressure up to 370 psid as indicated on the 1-340-4, DRIVE WTR PRESS gauge at the 901-5 panel.


Question: baseline water quality determination process

The collection of baseline water quality data, and determination of baseline water quality conditions will be very important as the Upper Control Limits (UCL’s) and ground water restoration objectives would be based on this data.

Monitor wells will be installed per the Production Area Hydrologic Test Document, which will be reviewed, and approved by the SERP.


Question: What is the significance of DNBR in reactor safety?

The DNBR (Departure from Nucleate Boiling Ratio) is a crucial parameter in nuclear reactor safety, as it indicates the margin to boiling onset and helps to ensure core cooling under varying operational conditions.

The Core Average Temperature data shows how the temperature fluctuates in the reactor core over time, indicating the thermal response but does not directly relate to the safety margin against boiling.


Question: The test setup for the CTU-2 puncture drop tests is critical for assessing the safety of containment components.

Figure 2.12.1-19 shows the test setup for the CTU-2 puncture drop test, which is intended to evaluate the performance of containment structures under specified conditions.

The design of safety-related systems requires extensive documentation and thorough evaluations to ensure compliance with regulatory standards.


Question: The flow rate of RCIC should be adjusted to maintain desired operational parameters.

Verify RCIC flow increases to 400 gpm. Verifies 400 gpm indicated on RCIC Flow Controller, (FIC 1-1340-1).

Positions MO 1-1301-60 c/s to OPEN and verifies: — OPEN light lit. — CLOSED light out.


Question: The completion of ITAAC involves ensuring that all safety-related systems meet specified requirements.

Completion of ITAAC E.3.9.08.01.02 confirms that the required inspections have been successfully performed and all acceptance criteria have been met.

The Vogtle Electric Generating Plant Unit 3 is designed to operate with a capacity of 1,154 megawatts and incorporates advanced safety features.


Question: What is the stripping fraction for isotopes?

Stripping fractions are defined as the ratio of the number of atoms of a specific isotope removed during a process to the total number of atoms of that isotope initially present.

The NRC reviews applications for new reactor licenses to ensure they meet safety and environmental standards.


Question: What is an NPDES permit for a nuclear power station?

Clinton Power Station, NPDES Permit No. IL0036199, 1998 Summer Temperature Monitoring Data in Salt Creek Below Dam.

The NRC requires regular inspections and assessments of plant operations to ensure compliance with safety regulations.


Question: The NRC evaluates modifications to enhance the safety and reliability of emergency cooling systems in reactors.

Increase the reliability of the low-pressure emergency core cooling system (ECCS) reactor pressure vessel (RPV) low pressure permissive circuitry. Install manual bypass of low pressure permissive.

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


Question: What is the function of the Containment Spray System in nuclear reactors?

The Containment Spray System is shown on Figure 6-194. The spray system consists of two spray pumps and spray heat exchangers in parallel, with associated piping, valves and spray headers in the upper Containment volume.

The realignment procedure for the spray pumps is detained in Table 6-125.


Question: What is the purpose of the Aging Management Review (AMR)?

This chapter provides the results of the AMR for those structures and components identified in Section 2.0 as being subject to AMR.

The major sections of this chapter are: Aging Management of Reactor Vessel, Internals, and Reactor Coolant System (Section 3.1)


Question: general license registration

General License Registration Number: GL-705591-26


Question: NRC guidelines for freezing plant procedures before operator licensing examinations

Guidelines regarding the “freezing” of plant procedures in advance of an initial operator licensing examination have been added to NUREG-1021 as part of Supplement 1 to Revision 9, which was published in October 2007.

Can a plant freeze multiple procedures at different times based on the scope of the procedure change and how its implementation date affects examination development and administration?


Question: What is the significance of unit hydrographs in flood hazard reevaluation?

The unit hydrographs developed for the licensing basis PMF and used in the flood hazard reevaluation are developed from gage data and calibrated to historical events.

The PMF analysis assumed each dam is able to fully open gates and no debris blockage of the spillways occurs.


Question: Conventional nuclear energy systems do not utilize hybrid methodologies in their operation.

Conventional Nuclear does not use a hybrid approach.

The turbine generator converts thermal energy into mechanical energy to drive the generator.


Question: What is the purpose of an Audit Plan in nuclear regulatory submissions?

The Audit Plan provides a framework for evaluating the effectiveness of the licensee's implementation of their safety management system.

Plant Licensing Branch III is responsible for the review and approval of all license amendments and renewals.


Question: In the event of a loss of flow in the component cooling system, operators must identify the key indicators such as pump trip and low flow alarms.

Loss of CCS flow is indicated by alarming conditions including a pump trip, multiple low flow alarms, and low header pressure.

Surge Tank level less than 60% and dropping uncontrolled indicates potential issues that need to be addressed in other systems.


Question: The ASTEC V1.3 analyses for the LOFT LP-FP-2 experiment involved critical evaluations of temperature excursion limits.

The ASTEC V1.3 simulation of the LOFT LP-FP-2 experiment has the onset of the temperature escalation (at the 1.067 m. elevation) occurring at a temperature greater than 1700 K (2600°F); figure 17 also shows that in the experiment the actual onset of the temperature escalation (at the 1.067 m. elevation) occurred at a temperature well below 1500 K (2240°F).

It is noteworthy that, regarding the ASTEC V1.3 simulation of the LOFT LP-FP-2 experiment during reflood, “Recent Advances in ASTEC Validation on Circuit Thermal-Hydraulic and Core Degradation” states: High temperature excursions with extended core degradation and enhanced hydrogen release observed in the test during reflooding were not reproduced by ASTEC due to lack of adequate modeling.


Question: The NRC's evaluation of compliance ensures that licensees maintain safety standards during reactor operations.

The NRC requires licensees to conduct regular safety assessments to ensure compliance with regulatory standards and operational safety for nuclear reactors.

In 2023, the facility upgraded several control systems to enhance operational performance and efficiency.


Question: The NRC mandates that licensees implement systems to evaluate and manage the integrity of pressure boundary components.

Piping, piping components are critical for maintaining the pressure boundary and require a one-time inspection program to evaluate their condition.

The aging management program for insulated piping focuses on monitoring the external surfaces for general and localized corrosion effects.


Question: What are the implications of caution notes in Emergency Operating Procedures?

The Shift Foreman will monitor the Function Restoration Procedures for information only until a specific point in the procedure is reached while performing which of the following procedures?

A. Incorrect. Caution prior to step 1 states: Function Restoration Procedures shall not be implemented prior to completion of step 12.


Question: saturation deficit

Saturation Deficit (gm/kg) is a measurement utilized in various safety considerations regarding the impact of environmental factors on equipment performance.

Total No. Observations indicate the number of data points collected during the study across different wind directions.


Question: The report detailing the activities and findings from the recent refueling outage at St. Lucie Unit 2 includes an Owner's Activity Report, Form OAR-1, submitted after each operational period.

The OAR-1 report is prepared, certified and submitted upon completion of each refueling outage.

Should you have any questions regarding this submittal, please contact Ms. Maribel Valdez, Fleet Licensing Manager, at 561-904-5164.


Question: chemical analysis of selenium


Question: The construction and functioning of the Bellefonte plant is expected to significantly enhance local economies by generating jobs and increasing spending in the area.

Based on past TVA experience, it is anticipated that construction and operation of the Bellefonte plant will promote economic activity in the local area.

These cities similarly face the prospect of an inadequate number of teachers and classrooms for school children of TVA workers.


Question: The activity levels of various nuclides are critical for understanding radioisotope behavior in safety assessments.

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

The confidence value indicates the reliability of the measurement process, with higher values reflecting greater certainty.


Question: What are the characteristics of the stratigraphy in the Crownpoint project area?

The stratigraphy in the Crownpoint project area is presented in Figure 3.3-7. The Recapture Member (termed the Recapture mudstone unit in the Crownpoint project area) is the underlying confining unit and has been confirmed to have a thickness of approximately 75 to 80 m [250 to 260 ft].

Recent soil surveys are more specific and localized in soil classification. While the 1979 TVA soil survey identified three soil units within the Crownpoint project area, the NRCS soil survey depicts nine soil units (Figure 3.3-8).


Question: The enrichment of uranium is critical for determining the performance of nuclear fuel assemblies.

Initial Assembly Avg. Enrichment wt % 235U (E) is a key factor in the loading of nuclear fuel for different assembly types.

The table presents various assembly designs, including CE 14x14 and B&W configurations, along with their specific cooling times.


Question: concentration of radionuclides

The table indicates the concentrations of various radionuclides, including H-3, C-14, and Cs-137, measured in picocuries per gram (pCi/g).

In addition to the radionuclide concentrations, the report discusses various methodologies used in the analysis of nuclear safety assessments.


Question: The integrity of the steam plenum must be maintained to ensure safe reactor operations.

The steam plenum serves as a crucial component in the reactor's primary circuit by providing a space for steam separation and pressure management during normal and accident conditions.

The safety analysis report includes detailed descriptions of the site's seismic analysis and geological evaluations, which are essential for understanding potential natural hazards.


Question: What are the separation criteria for cables entering the Reactor Building?

Separation criteria for cables entering the Reactor Building between elevations 733’ and 767’ are: Cables associated with Engineered Safety Features Train A functions are located between elevation 750’ and 761’. Cables associated with Engineered Safety Features Train B functions are located between elevation 733’ and 750’. Cables associated with reactor protection and Engineered Safety Features instrumentation entering the Reactor Building are separated by routing two channels through penetrations.

Proper installation of the fire stops and seals is assured by following approved manufacturer's installation procedures and techniques. Each installation is visually inspected periodically to verify that its integrity is maintained.


Question: When entering a condition, if required actions are not completed in the specified timeframes, the situation may escalate to the next condition.

The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered.

The Required Action should be pursued without delay and in a controlled manner.


Question: What is a materials engineer in the context of nuclear safety?

Matthew Hiser is a Materials Engineer at the US Nuclear Regulatory Commission, specializing in corrosion and metallurgy related to nuclear safety.

Last passport from the workshop visitors for next week.


Question: What is the significance of the DNBR in nuclear reactor safety?

DNBR, or Departure from Nucleate Boiling Ratio, is a critical safety parameter used to evaluate the heat transfer performance and cooling of a nuclear reactor core.

The Vogtle Electric Generating Plant consists of two units that provide significant electrical generation capabilities through advanced reactor technologies.


Question: The evaporation of noncondensable gases from the main condenser during normal operations is crucial for maintaining efficiency.

The Main Condenser Evacuation System is designed to remove noncondensable gases and inleaking air from the main and feedwater pump turbine condenser shells during plant startup, cooldown and normal operations.

Periodic visual inspections and preventative maintenance are conducted following normal industrial practice.


Question: The NRC mandates that operators verify the status of isolation valves during reactor shutdown procedures.

Verify SG blowdown isolation valves are closed:


Question: SER Coverage testing activity

SER Coverage is a component of testing activity that ensures the reliability and effectiveness of safety-related systems during operational assessments.

The Project Implementation Deliverables section outlines the expected outcomes and timelines for project completion, ensuring that all necessary tasks are addressed.


Question: calibration of GM Survey Meter

S.E. International, Inc. certifies the above-described instrument was calibrated in a known radiation field using a Cs 137 (662keV) beam calibrator.

This certificate may not be reproduced without S.E. International, Inc. TN LICENSEE R-51002-C27.


Question: The report includes the original area and the uncertainties associated with the energy measurements taken during the experiment.

The subtracted area and subtracted uncertainty values provide clarity on the adjustments made to account for ambient background noise in the measurements.

The peaks observed in the spectrum are analyzed using various methodologies to ensure accurate reporting.


Question: The Joseph M. Farley Nuclear Power Plant operates within a licensed framework established by the NRC.

The Farley Plant Site consists of two nuclear reactors: the Unit 1 and Unit 2 reactors, both operating under the license issued by the NRC.

The site is located near the Chattahoochee River, which provides essential cooling water for the reactors.


Question: What are the vascular plant species observed in the HCS study area?

Vascular Plant Species Observed on the HCS Study Area, 2021-2022

The NRC requires that all modifications to existing facilities be reported in a timely manner.


Question: What is the method for calculating the seismic hazard interval annual initiating event frequency?

The seismic hazard interval annual initiating event frequency is calculated (except for the final interval) by subtracting the mean exceedance frequency associated with the g-interval (high) end point from the mean exceedance frequency associated with the g-interval beginning point.

This is common practice in industry SPRAs.


Question: The NRC mandates specific analytical techniques to assess radiological safety in the environment.

The following Analytical Methods were performed: DOE EM. HALS-300, Am-05-RC Modified, ASTM C 1475-00 Modified, DOE EM HALS-300, Pu-11-RC Modified, DOE RESL N-1, among others.

The project detailed in this document is the LACBWR Site Restoration Project, which involves a variety of preparatory and analytical tasks.


Question: What are the results for various isotopes of Americium in the context of solid waste analysis?


Question: minimum distance to project area

The Minimum Distance to the Crownpoint Project Area is measured in meters, indicating the closest proximity that a water right can be located relative to the project.

The average summer flow rate for Well NTUA-2 is recorded at 222 liters per minute, which shows the water output during the peak season.


Question: What is the purpose of the Common Emergency Plan in nuclear facilities?

The Common Emergency Plan is established to ensure that Florida Power & Light Company can effectively respond to any radiological emergency, consistent with NUREG-0654, Revision 2.

The Turkey Point Nuclear Generating Plant includes Units Nos. 3 and 4, which have specific operating licenses that govern their operation and safety protocols.


Question: What is the significance of reactor coolant system pressure criteria in nuclear regulations?

Clarified peak reactor coolant system pressure criterion.

Updated Station Manager title to Site Vice President and Assistant Station Managers titles to Manager-Station O&M and Manager-Station S&L.


Question: What does 'Method Detection Limit' mean in nuclear regulatory context?

The detected analyte was only tentatively identified, and the result is approximate.


Question: What are the oral ingestion ALI values for Americium isotopes?

Americium-241 has an Oral Ingestion ALI of 8E-1 µCi, while Americium-242 has the same value of 8E-1 µCi.

Curium-241 has an inhalation ALI of 3E-1 µCi, showing a significant difference in safety measures for different radionuclides.


Question: ITAAC notification review status

The ITAAC notification review status indicates whether adequate and sufficient information has been provided for the safety review process.

The NRC regulation 10 CFR 50.55a outlines requirements for the design and fabrication of components in nuclear power plants.


Question: The population dose can be evaluated using the population count method as demonstrated in the IRRIDOSE benchmark runs.

Population Dose by Population Count Method integrates exposure data to estimate radiation doses to populations.

The methodology described in the IRRIDOSE benchmark runs allows for the examination of various reactor configurations and their impacts.


Question: What are the criteria for admissible contentions in NRC licensing proceedings?

Admissible contentions explain the issue to be litigated, provide supporting facts or expert opinion and specify sources or documents on which it intends to rely; CLI-21-2, 95 NRC 1 (2022)

NRC's reactor design criteria and standards are intended to ensure the ability to withstand environmental hazards, such as earthquakes and flooding, without loss of capacity to perform their safety functions; LBP-21-2, 95 NRC 49 (2022)


Question: nuclear power plant licensing process

As of the latest report, there are 94 operating nuclear power plants in the United States, which are licensed by the NRC following a rigorous licensing process that ensures compliance with safety and operational standards.

In addition to nuclear power plants, there are also 31 research and test reactors that serve various educational and experimental purposes, licensed under different regulatory frameworks.


Question: What steps are involved in emergency termination during a nuclear incident?

Confirm emergency termination. INITIALS: _______  DATE/TIME: _______/_____

If protective actions had been implemented, verify completion of the actions necessary to allow reentry into affected areas.


Question: The NRC will evaluate the environmental impacts of renewing the Oconee Station operating licenses versus those of alternative energy options.

The environmental impacts of the proposed action (renewing the Oconee Station operating licenses) would be SMALL for all impact categories.

Duke Energy’s 2020 Integrated Resource Plan considered retirement of all coal-fired units to meet its carbon dioxide reduction goals and align with state energy policies and legislation.


Question: safety evaluation report

The Safety Evaluation Report (SER) documents the findings and conclusions of the NRC's review of a license application and outlines the regulatory framework, safety assessments, and compliance evaluations necessary to ensure public health and safety.

The integration of new technologies into the power generation sector requires careful consideration of operational efficiency and reliability in the context of existing infrastructure.


Question: What is clean strainer head loss in nuclear regulatory filings?

This flow sweep data was used to characterize flow regime through the strainer and to adjust the clean strainer head loss (CSHL) to different plant conditions.

Afterwards, at least one flow sweep was performed by measuring strainer head losses at several different flow rates.


Question: The analysis of various isotopes, including Europium-152, provides critical data for ensuring safety and regulatory compliance.


Question: What actions should be taken to manage reactor power levels when parameters indicate rising values?

Unit 2 was at 90% power when the following occurs: Reactor power is 91% and rising.

The reactor pressure is 1000 psig and rising.


Question: secondary side cleaning in steam generators

Advanced scale conditioning agents were applied during this outage in all SGs followed by sludge lancing. A best estimate of deposits removed is a total of 5,000 pounds. Report results of sludge lancing if no chemical cleaning was performed.

For recirculating SGs, report on channel head inspections that were performed on plug locations, divider plate and associated welds, tube-to-tubesheet welds, and cladding.


Question: neutralization number in corrosion testing

Neutralization numbers remained well above the specified acceptance (lower) limits and exhibited no consistent trends over time.

The results of Unit 1 and Unit 2 CPM tests show that concentrations of corrosive ions and absorbed water content have remained well below the specified acceptance (upper) limits and, in addition, exhibit no adverse trends over time.


Question: What is the ECHO report in relation to nuclear facilities?

ECHO (Enforcement and Compliance History Online) Detailed Facility Report, Facility Summary, Duke Energy Oconee Nuclear Station.

The Electric Power Research Institute conducted a Technical Evaluation of the Utility of Intake Approach Velocity as an Indicator of Potential Adverse Environmental Impact under Clean Water Act Section 316(b).


Question: The NRC mandates that assessments of masonry structures be conducted regularly to ensure compliance with aging management standards.

The Masonry Walls aging management program is implemented as an element of the Structures Monitoring program, which is monitored under the Maintenance Rule.

Although not in-scope for License Renewal, these two indications were evaluated and concluded that the identified defect did not compromise the structural integrity of the wall.


Question: What is the purpose of the draft environmental assessment (EA) for the Palisades Nuclear Plant?

The purpose of this letter is to notify your office of the issuance of the draft EA and draft FONSI, including a NHPA Section 106 effects determination for the undertaking.

In addition, HDI distributed Palisades cultural resource management procedures on November 4, 2024.


Question: The NRC mandates that structural fire barriers are essential for preventing the spread of fire within nuclear facilities.

Structural Fire Barrier Walls are intended to function as a fire barrier, utilizing masonry walls to protect against fire hazards in indoor uncontrolled air environments.

The aging management review results also include evaluations under NUREG-2192, focusing on the importance of maintaining regulatory compliance in operational practices.


Question: The NRC stipulates that the SG secondary pressure must stay within safe limits during operational scenarios, taking into account RCS fluid temperatures.

The maximum pressure in the SG secondary side is limited to the saturation pressure at the temperature of the RCS fluid on the SG primary side.

Modification of other biases can generate many cases with similar maximum RCS pressures.


Question: The acceptable temperature level of the Spent Fuel Pool is critical for safety compliance.

Determined that Spent Fuel Pool Temperature is acceptable.

Determined CCW flow rate from F1-197 is 3100 gpm and enters data on Attachment 4.


Question: what is an intact fuel assembly

INTACT FUEL ASSEMBLY is a fuel assembly without DAMAGED FUEL RODS. Connecticut Yankee fuel assemblies with missing fuel rods, or with missing fuel rods replaced with solid filler rods, or with structural damage, are considered INTACT FUEL ASSEMBLIES provided that they have no DAMAGED FUEL RODS. Yankee Class fuel assemblies with missing fuel rods replaced with Zircaloy or stainless steel rods, or with structural damage, are considered INTACT FUEL ASSEMBLIES provided they have no DAMAGED FUEL RODS.

Spent nuclear fuel (SNF) that cannot fulfill its fuel-specific or system-related function.


Question: What is the significance of a subtracted area in nuclear regulatory documents?

The subtracted area refers to the net peak area after subtracting the effects of ambient background, allowing for a more accurate representation of the actual counts attributed to the source being measured.

The energy values and their corresponding uncertainties are crucial for understanding the performance of the detector during measurement.


Question: frazil ice formation in nuclear plants

Frazil ice, fine needle-like structures or thin, flat circular plates of ice, can form on intake structures, thereby causing blockages of the intakes and reducing access to available water (Reference 2.4.7-16). Frazil ice begins to form when the water becomes supercooled (below its

The tidal flow is much greater than the flow required by the intake structure, making the effects of an ice jam upstream and the resulting reduced Delaware River freshwater flow not a critical factor in the design of the new plant intake.


Question: performance requirements for Class I PORVs

If the block valve is associated with a Class I PORV: Restore block valve to OPERABLE status.


Question: grain size measurements

The grain size measurements are related to the distribution of particle sizes within a sample and can be expressed in both U.S. standard sieve openings and millimeters, aiding in the analysis of materials used in construction and safety assessments.

The term 'gravel' can refer to a variety of materials that vary in size and composition, which are frequently employed in different construction and engineering applications.


Question: The protection system must be independent to ensure reliability and redundancy in case of failures.

10 CFR 50, Appendix A, GDC 22 requires protection system independence. The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.

Duke Energy has evaluated the proposed exemption request against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance...


Question: nuclide identification in nuclear regulatory documents


Question: What is the Quality Assurance Program Description in the context of nuclear licensing?

The NRC Step 1 extension letter outlines the closure criteria to the QAP Step 1 review topic as, 'understanding of the scope and application of the Quality Assurance Program to the Aurora.'

Quality assurance, combined with pre-operational testing, startup testing, and the utilization of technical specifications throughout the life of the plant, are all go/no-go gates for safety assurance in the as-built plant.


Question: What is COBRA-FLX and how does it relate to critical heat flux?

COBRA-FLX is a thermal-hydraulic code that provides a correlation for critical heat flux, important for ensuring reactor safety under various operational and accident conditions.

Local conditions from grid tests can vary widely based on the specific test parameters and configurations employed during the simulation.


Question: tritium measurement units

ZBG 6 | 7/16/2024  | AIR TEMPERATURE                |          |         | REG       | 25.000    | degC      | REG         |


Question: dry weight analysis

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

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


Question: The enrichment level of uranium in the assembly is a critical factor in determining its overall safety and performance.

Average Enrichment wt % 235U (E) is a key parameter that affects the cooling time and operational capabilities of the reactor core.

The table provides varying configurations for different reactors including CE, WE, and B&W, but does not explain the significance of uranium enrichment.


Question: Completion Time convention

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

Limiting Conditions for Operation (LCOs) specify the lowest functional capability or performance levels of equipment required for safe operation of the facility.


Question: The data collected from monitoring wells indicates varying levels of fluoride in groundwater.

Monitoring well W-39 is screened within the upper and lower surficial aquifer.

The wells are spaced adequately to monitor the fluctuations in groundwater levels.


Question: What is the ECL Alert in nuclear emergency classification?

An ECL Alert occurs when an automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken in the MCR are not successful in shutting down the reactor.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc.


Question: What are the key components of a nuclear reactor's technical specification?

The technical specifications for a nuclear reactor include items such as limiting safety system settings, channels, and surveillance requirements that ensure safety and operability during various operational states.

Safety management in nuclear plants involves not just compliance with regulations but also engagement with the local community to address any concerns they might have regarding plant operations.


Question: The NRC stipulates that to enhance safety, a Hydrogen Mitigation System must be in place for controlled combustion during an accident scenario.

The licensing requirements relative to the provisions for hydrogen control prescribed in 10CFR 50.44 have evolved from numerous deliberations among the Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and utilities.

Table 6-118 shows the results of the calculation for hydrogen production and accumulation from the following sources: (1) zirconium-water reaction, (2) primary system hydrogen, (3) aluminum and zinc corrosion, and (4) radiolytic decomposition of core and sump solution.


Question: What is the purpose of an OCA in nuclear packaging?

The OCA, or Overpack Container Assembly, is designed to provide a robust protective layer for the contents during transport and storage, ensuring safety and compliance with regulatory requirements.

The GNF NPC Package Assembly is critical for maintaining the integrity of nuclear materials, which involves complex engineering and adherence to various safety standards.


Question: The rules governing inservice inspection of components in nuclear power plants are outlined in the ASME Boiler and Pressure Vessel Code.

The ASME Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2013 Edition, provides comprehensive guidelines for the inspection of critical plant components.

Westinghouse Electric Company LLC provided a document titled Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Revision 3, October 2011.


Question: well-drilling certifications

See the following for a description of well-drilling certifications: SECTION 40-23-320 of SC State House.

Because well drilling records are from several years ago, staff also searched on the following website to determine whether or not the driller was still active: Link.


Question: What is hot standby in nuclear reactor operations?

Hot standby is a stable condition, automatically reached following a unit shutdown. The hot standby condition can be maintained safely for an extended period of time.

A review of the IE instrumentation and control systems was conducted in response to IE Bull. 79-27.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves several key steps, including the submission of an application that outlines safety analysis, environmental considerations, and operational protocols.

During the construction phase, it is vital to conduct regular inspections to ensure that all materials and systems meet the established safety standards.


Question: weighted mean activity of nuclides

The weighted mean activity is calculated as the weighted average of the activities of different nuclides, expressed in picocuries per gram.

The nuclide PB-212 has a confidence level of 0.560 and a weight mean activity of 2.74E-02.


Question: The confinement systems are subject to rigorous testing protocols to ensure their integrity during operations.

These systems are exposed to the test pressure.

The containment leakage monitoring open system is crucial for assessing the performance of safety-related systems during reactor operations.


Question: The NRC staff anticipates the noise generated during operations of the new nuclear section would not exceed existing levels.

The NRC staff concludes that operation-related noise impacts from the new nuclear portion of the combination alternative would be **SMALL**.

Given the distance from shore (10–24 nautical miles), noise from wind turbines would not be audible onshore.


Question: The requirements mandate that licensees must evaluate and monitor the leakage rate of the primary containment for safety compliance.

C. Primary containment air lock inoperable for reasons other than Condition A or B. | 3.6.1.2C.3 | 3.6.1.2C.3 | Added RICT to 3.6.1.2C.3 | TSTF-505 changes are incorporated. However, under certain circumstances, with more than one primary containment airlock inoperable, excessive leakage or a loss of function may occur. Thus, the TS function is.

The NRC provides guidelines on various maintenance protocols for reactor operation, ensuring that safety standards are upheld across all units.


Question: PEP EPZ analysis

This input will establish the specific radionuclide inventory and the quantity released for the events that will be assessed in the PEP EPZ analysis.

Projected doses will be evaluated against three dose-based criteria akin to those in NUREG-0396.


Question: strainer head loss analysis

Strainer head loss analysis was performed for each postulated break in NARWHAL to determine which breaks result in strainer failures due to exceeding head loss based criteria.

Adjustment was made to the perforated plate head loss calculated from the PCI prototype test data.


Question: methods for detecting inadequate reactor core flow

Multiple methods exist for detection of a core blockage condition as manifested by an inadequate RCS inventory or inadequate RCS and core heat removal conditions. The primary methods for detection include CET temperature indication and reactor water level, as monitored by the RVLIS.

Multiple methods are available for Wolf Creek to mitigate an inadequate reactor core flow condition. Upon identification of an inadequate RCS inventory or an inadequate core heat removal condition, the EMGs direct the operators to take actions to restore cooling flow to the RCS including:


Question: To ensure adequate safety measures, it is critical to establish the minimum cooling time required based on enrichment levels.

Minimum cooling time is a critical factor that varies with the average enrichment of uranium in the assembly, specifically in the context of reactor operation and fuel management.

The enriched uranium fuel assemblies are designed with various configurations, including 7x7, 8x8, and 9x9 arrangements.


Question: What is the maximum amount of Sodium-22 that can be possessed under the NRC license?

12.022 millicuries total and no single source to exceed the maximum activity specified in the certificate of registration issued by the U.S. Nuclear Regulatory Commission or an Agreement State.

The licensee is authorized to use these materials for training exercises, instrument operational checks, and calibration and/or reference standards.


Question: The coincidence voting process in the LCL subsystem ensures that RPS and NSSSS/ECCS actuations are handled redundantly, improving safety during trips.

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

The LCL has four PM646A processor modules segmented by function.


Question: The NRC requires that licensees must perform adequate surveys to demonstrate compliance with the radiological criteria for unrestricted use, in accordance with 10 CFR 20.1402.

NRC guidance for conducting ALARA analyses is provided in NUREG 1757, Volume 2, Revision 2. In Section 4.4, “ALARA Evaluation,” of the LTP, the licensee provides its ALARA analysis process and conclusions for the remaining components.

The NS Savannah project is nearing completion and, based on the conservative decommissioning activities, a pre-determined compliance measure for ALARA is not suitable for this vessel.


Question: The measurement and assessment of nuclide yield are critical in understanding radioactive emissions.

EU-152 has a yield of 28.40% at an energy level of 121.78 keV, which is recorded as part of the nuclide identification process.

PA-228 has a half-life of 1.9 years, influencing the timing of safety inspections and assessments.


Question: What are the regulatory requirements for reactor trip instrumentation?

SRO will refer to TS 3.3.1, Reactor Trip Instrumentation, and 3.3.3, Post Accident Monitoring Instrumentation.

The crew will enter 1-AOI-24, RCP Malfunctions During Pump Operations.


Question: elemental iodine release during containment testing

Elemental I-131 after DF of 100 [Ci] is measured to be 3.9488e+05, indicating the amount of iodine present in the containment after the decontamination factor has been applied.

The core power of the reactor is specified as 3658.3 MW, which is critical for understanding the thermal limits of the system.


Question: What is the purpose of the pressurizer in a pressurized water reactor?

The volume of the pressurizer is equal to, or greater than, the minimum volume of steam, water, or total of the two which satisfies all of the following requirements: The combined saturated water volume and steam expansion volume is sufficient to provide the desired pressure response to system volume changes.

The surge line is sized to limit the pressure drop between the Reactor Coolant System and the safety valves with maximum allowable discharge flow from the safety valves.


Question: The NRC mandates an evaluation of potential impacts on the environment during the licensure process.

The licensing process requires a thorough assessment of public health and safety to ensure that nuclear facilities operate in compliance with regulatory standards.

The Braidwood Station has undergone significant technical upgrades to improve its operational efficiency.


Question: The NRC monitors the levels of radiation emitted from nuclear materials, ensuring that all regulatory guidelines are met.

Nickel-63 has a measured value of -1.21 pCi/g, with a qualified uncertainty of +/-1.47 pCi/g as part of the Rad Liquid Scintillation Analysis.

The analytical methods performed include a range of modifications such as DOE EMI HALS-300, Am-05-RC and EPA 905.0 Modified, which are critical for ensuring accurate measurement protocols.


Question: What is the process for obtaining a nuclear license in the United States?

The Nuclear Regulatory Commission (NRC) regulates the use of nuclear materials and oversees the licensing process for nuclear reactors to ensure public health and safety.

The peak daily visitors to Great Smoky Mountains National Park reached 26,537 during a holiday weekend.


Question: The panels designed to blow out have specific vent areas that are critical for pressure relief during operating conditions.

The blowout panels in the subcompartments have vent areas for which an obstruction must be eliminated to provide area for pressure relief.

A sheet-metal panel is considered to blow out only when it is located on the far side of the wire grid door from the break node.


Question: The auxiliary feedwater system has critical roles during operational events, providing essential connections to the steam generators.

The MFWS, from the steam generators to the isolation valves, provides flow paths for the auxiliary feedwater to the steam generators and isolates feedwater flow to the steam generators during a main steam or feedwater line break event or a containment overpressure event.

Two half capacity, high speed turbine driven feedwater pumps are provided with common suction and discharge manifolds.


Question: The calculations for snowmelt rate include wind speeds and dew point temperatures.

The dew point temperatures are used as an input to the energy budget equation to calculate snowmelt rate.

The 100-Year Precipitation estimates for various durations show significant variation in rainfall measurements.


Question: ARITA methodology in nuclear licensing

Licensees using the ARITA methodology must either (1) justify that the comparisons in ANP 10339P for the MB-2 MSLB testing or other relevant comparisons of the ARITA methodology against experimental data accurately or conservatively represent their plants with respect to the [ or (2) otherwise demonstrate that the [ is representative or conservative for each event where it is relevant to the prediction of applicable FOMs. (Reference to Section 3.5.1.2, SEC-6 of Reference 1)

The parameter SEC-2b, (Reference to Section 3.5.1.2, SEC-2b of Reference 1)


Question: license for spent nuclear fuel

a license is hereby issued authorizing the licensee to receive, acquire, and possess the power reactor spent fuel and other radioactive materials associated with spent fuel storage designated below; to use such material for the purpose(s) and at the place(s) designated below; and to deliver or transfer such material to persons authorized to receive it in accordance with the regulations of the applicable Part(s).

Byproduct, Source, and/or Special Nuclear Material


Question: What are the operational requirements for the chilled water system in the Auxiliary Building?

The chilled water system loop which services the safety-related loads in the Auxiliary Building shall be OPERABLE with one of the following configurations: Three OPERABLE chillers and, Two OPERABLE chilled water pumps; Two OPERABLE chillers and, Two OPERABLE chilled water pumps; Three OPERABLE chillers and, Two OPERABLE chilled water pumps from either Unit 1 or Unit 2 (Units Cross-tied).

The Unit 2 Emergency Control Air Compressor (ECAC) is isolated from the chilled water system.


Question: What is the purpose of a proprietary version in regulatory filings?

NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390.

The purpose of this letter is to provide NuScale's response to the NRC's request to docket an audit response.


Question: Duke Energy aims to modify the Oconee Station’s effluent to meet new discharge standards while ensuring compliance with water quality regulations.

In its NPDES permit renewal application, Duke Energy has requested that Oconee Station’s daily maximum discharge temperature be changed to a 7-day average not to exceed 100°F (37.8°C). Duke Energy states that this change would align with the State water quality standard and would not result in any adverse operational impact on the lake’s biological community.

Outfall 002 receives process wastewater and other flows either processed through or entering Oconee Station’s conventional wastewater treatment system.


Question: replacement energy alternatives

To ensure that the replacement energy alternatives are consistent with State or regional energy policies, the NRC reviewed energy-related statutes, regulations, and policies within the Oconee Station region.

The order in which this EIS presents the different alternatives does not imply increasing or decreasing level of impact; nor does the order imply that an energy-planning decisionmaker would be more (or less) likely to select any given alternative.


Question: What are Control Element Assemblies (CEAs) and their inspection methods?

Control Element Assembly (CEA) inspections include VT-3 visual inspections of 100% of the peripheral CEA instrument guide tubes no later than two refueling outages from the beginning of the license renewal period; The VT-3s focus on the CEA supports and the welds connecting the guide tubes to the supports.

For the applicable plant designs, Item C11.1 specifies a minimum 75% population criterion (of inspected components) for VT-3 inspection credit.


Question: What is a relief request in the context of nuclear licensing?

A relief request is a formal submission made by a licensee to the NRC, seeking approval for an alternative approach to a regulatory requirement that would otherwise be applicable during the operation or maintenance of a nuclear facility.

The Palo Verde Nuclear Generating Station is located in Arizona and includes three units, which have been operational since the late 1980s.


Question: The operational release criteria for soil are established through the application of DCGLs for various radionuclides.

The Operational DCGLs are then used as the DCGL for the FSS design of the survey unit (e.g., calculation of surrogate DCGLs, investigation levels). The OpDCGLs for the unrestricted release of soil are provided in Table 4-3.

Instrument DQOs included a verification of the ability of the survey instrument to detect the radiation(s) of interest relative to the Operational DCGL.


Question: specific license for radioactive materials

The Agency shall grant a specific license to manufacture or initially transfer calibration or reference sources that contain americium-241, radium-226, or plutonium for distribution to persons generally licensed under R12-1-306(D) if the applicant satisfies:

Authorization under this Section to produce Positron Emission Tomography (PET) radioactive drugs for noncommercial transfer to medical use licensees in its consortium does not relieve the licensee from complying with applicable FDA, other federal, and state requirements governing radioactive drugs.


Question: The NRC requires thorough documentation and analysis for radioactive materials, including the uncertainty of measurements in pCi/g.

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

The analysis reveals that the soil sample was collected on August 22, 2019, and was received for evaluation on October 21, 2019.


Question: What is COBRA-FLX used for in nuclear simulations?

COBRA-FLX is a computational fluid dynamics code used to analyze thermal-hydraulic behavior in nuclear reactors, providing crucial data on critical heat flux correlations.

Table B-3 presents various local conditions from COBRA-FLX simulations, including parameters such as pressure and temperature, which are essential for understanding the performance of heat exchangers.


Question: The NRC mandates that licensees to adhere to certain reporting standards for analyzed uranium concentration levels.

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

The ACZ Project ID is L89658, which identifies the specific project associated with the submitted data.


Question: What is the purpose of the gas recombiner in the gaseous waste disposal system?

The recombiner system, containing two catalyst beds, reduces the hydrogen content of the gas to be stored, thereby reducing the storage volume as well as the hazard inherent in storing hydrogen.

The steam generator blowdown system vent exhaust is equipped with a vent condenser.


Question: What is the significance of the scaling parameter Π_F in component-level scaling for nuclear reactors?

The scaling parameter from Table 1 used to scale all components as an integrated system is the loop flow resistance number, Π_F.

In KP-FHRs, the very high volumetric heat capacity of the reactor coolant results in low volumetric flow rates, so that Reynolds numbers are commonly in the transition or laminar regimes in heat exchangers.


Question: APRM failure response

APRM 5 recorder indicates 125. HIGH / HI HI lights lit on 901-5 and 37 panels. Channel B Scram Solenoid Lights are OUT. Control Rod H-5 scrams and sticks at position 10.

The control room initiated the emergency shutdown of the reactor due to rising coolant temperatures, which exceeded operational limits.


Question: The NRC mandates that samples undergo detailed evaluations to determine their spectra characteristics.

- **Identification Energy Tolerance**: 1.000 keV


Question: What is the NONROAD2008 emissions model used for in regulatory filings?

EPA originally developed the NONROAD2008 emissions model in 2002 to assist state and local regulatory agencies in developing emission inventories.

The following provides an overview of the methods and the results of the preliminary emissions inventory.


Question: TMI-2 advisory panel effectiveness

All 36 individuals interviewed for NUREG/CR-6252 considered the advisory panel a success, particularly its effectiveness in accomplishing its mission.

The TMI-2 project office acted as a liaison between the NRC and the panel and provided information to the panel on the status of the cleanup.


Question: What are the material specifications for reactor head fabrication?

Head Forging Material (e.g., SA-508 Grade 3, Class 1, SA-788, ASTM E208-91, ASTM A-275) |

General Source and Sequence of Component Fabrication (e.g., Head Forging - JSW, Nozzles - Valinox, Cladding/Buttering & Final Welded Assembly - AREVA) |


Question: nuclear regulatory submissions

Mick Radio Nuclear Instruments, Inc. submitted a series of documents regarding various versions of their SagiNova model, detailing compliance with nuclear regulatory requirements for licensing.

National Institute of Standards and Technology developed SERIES Q, a standard that provides guidelines for measurement accuracy but does not relate to nuclear licensing.


Question: The NRC monitors various river systems that play a critical role in the safety of nuclear facilities.

The Clinch River is essential for the cooling water supply of the Watts Bar Nuclear Plant, ensuring operational safety and efficiency.

Dukes Creek is home to a series of small dams that are primarily used for recreational activities such as fishing and boating.


Question: The Minimum Detectable Activity (MDA) for each nuclide must be reported to ensure accurate measurement and compliance.

**Peak Locate Performed on:** 6/17/2019  7:59:40 AM
**Peak Locate From Channel:** 100
**Peak Locate To Channel:** 4096


Question: What is the ARITA methodology in nuclear licensing?

The ARITA topical report includes a methodology for calculating assembly bow penalties in Reference 1, Section 9.1.3.5.

Framatome’s current rod bow methodology is considered conservative and compliant.


Question: ECCS performance requirements

The intent of 10 CFR 50.46(a)(1) is to ensure ECCS cooling performance design requirements imposed by 10 CFR 50.46 are determined to be met by a rigorous method that provides a high level of confidence in ECCS performance.

Regulatory requirements are largely based on a deterministic framework, and are established for DBAs, such as the LOCA, with specific acceptance criteria that must be satisfied.


Question: The NRC evaluates technical specifications submitted by licensees to ensure compliance with safety standards.

The technical specifications submitted by the licensee must contain information on the design, operation, and safety features of the nuclear facility, including but not limited to instrumentation and control systems.

The contact information for senior executives is crucial for communication and consultation regarding regulatory matters.


Question: The licensing application will include details about setpoint methodology and calculations.

The LAR will address 1.9 “Setpoint Methodology and Calculations” (D.7)

The replacement ECCS CLS does utilize the SOR function.


Question: Licensees must ensure that their digital systems are secure against threats, as detailed in regulatory requirements.

Title 10 of the Code of Federal Regulations (CFR), Part 73, "Physical Protection of Plants and Materials," 10 CFR 73.54, "Protection of Digital Computer and Communication Systems and Networks," requires power reactor licensees to provide assurance that digital computer and communication systems and networks are adequately protected against cyber attacks, up to and including the design basis threat as described in 10 CFR 73.1, "Purpose and scope."

This white paper describes proposed changes to NEI guidance for identifying and protecting Security Critical Digital Assets (CDAs). The changes are intended to improve the efficiency of licensee cyber security programs while maintaining program effectiveness to protect against cyber attacks, up to and including the design basis threat as described in 10 CFR 73.1, "Purpose and scope."


Question: What are the actions related to an inoperable AC electrical power subsystem?

For TS 3.8.7.A, one or more Division I or II AC electrical power distribution subsystem(s) inoperable, RMAs may include:

Walkdown systems prior to beginning the work.


Question: 10CFR Part 55 regulatory requirements

10CFR Part 55 contains regulations governing the licensing of operators of nuclear power plants and the training and qualification of those individuals.

The question history indicates the examination's difficulty level is rated as 3.0, suggesting a medium level of complexity in understanding the content.


Question: What is a critical heat flux correlation?

COBRA-FLX is a critical heat flux correlation used to model thermal hydraulic phenomena in nuclear reactors.

The report outlines various test conditions evaluated during the assembly tests, including temperature, pressure, and hydraulic conditions.


Question: SGBD isolation valves

SGBD isolation valves HV-15216A, HV-15121A normally open, fail closed pneumatic valve isolates blowdown from steam generator No. 1 upon SGDB line break in auxiliary building.

The analysis identifies that the incident response team must undergo annual training to maintain proficiency in emergency procedures.


Question: The construction of small modular reactors is outlined in TVA’s Integrated Resource Plan, emphasizing their role in future energy strategies.

Under TVA’s 2019 Integrated Resource Plan, various scenarios/strategies have been developed that present a mix of energy resources. For purposes of analysis, can any of these strategies be identified as representative? For example, Scenario 6C in the 2019 Integrated Resource Plan would involve the construction and operation of 1,200 of SMRs, 2,500 of natural gas, and the remainder solar and storage.

Traffic on the Tennessee River near BFN includes both commercial and recreational vessels. The river channels and the locks at Guntersville Lock and Dam and Wheeler Dam are more than adequate for handling river traffic.


Question: spent fuel transfer process

Completion of the ISFSI decommissioning process is dependent upon the DOE's ability to remove spent fuel from the site.

To achieve this objective, based upon the oldest fuel receiving the highest priority, DOE would commence maximum rate of transfer of 3,000 metric tons of uranium/year.


Question: The system responsible for dispersing heat from the reactor is crucial for safety monitoring.

Ability to monitor automatic operation of the component cooling water system, including the following: (CFR: 41.7 / 45.5 / 45.13)

Loss of cooling to spent fuel pool cooling system heat exchanger tube leak.


Question: What are the key components of quality assurance in nuclear regulatory practices?

Quality assurance processes are crucial for ensuring the safety and compliance of operations in nuclear regulatory environments, involving systematic monitoring and evaluation of activities against established standards and regulations.

The audit report also includes various recommendations for enhancing operational efficiencies in the waste management sector.


Question: nuclear reactor operator licensing

The regulation 10 CFR Part 55 establishes the requirements for issuing licenses for operators of nuclear reactors, thereby ensuring that only qualified individuals are permitted to operate nuclear facilities safely and competently.

The reactor containment structure is designed to withstand significant external forces and prevent the release of radioactive materials in case of an accident.


Question: What are risk-informed technical specifications in nuclear regulation?

Risk-informed technical specifications are guidelines that provide a framework for using risk assessments in the development and maintenance of technical specifications for nuclear plants, enhancing plant safety and operational efficiency.

The NRC Regulatory Guide 1.200 outlines the approach to ensure the technical adequacy of probabilistic risk assessment results for risk-informed activities.


Question: What are the health impacts of inadvertent liquid releases at nuclear power plants?

However, other radioactive isotopes, such as cesium and strontium, have also been inadvertently released into the groundwater. The types of events include leakage from spent fuel pools, buried piping, and failed pressure relief valves on an effluent discharge line.

Actual usage averaged less than 11.5 million gallons per year from 2016 to 2021 (Section 3.6.3.2). As discussed in Section 3.6.3.2, other groundwater wells are used as needed and their withdrawals are not required to be permitted under a MDNR water appropriation permit.


Question: Office of Nuclear Reactor Regulation responsibilities

Allen Fetter is a senior project manager in the NRC's Office of Nuclear Reactor Regulation.

Environmental project manager at the NRC's Office of Nuclear Material Safety and Safeguards, and my co-project manager over here for section 206 is Allen Fetter.


Question: How does the Component Cooling Water System operate in relation to the Main Control Room?

Multiple ITAAC are performed to demonstrate that the Component Cooling Water System (CCS) provides the nonsafety-related functions of transferring heat from the Normal Residual Heat Removal System (RNS) during shutdown and the Spent Fuel Pool Cooling System (SFS) during all modes of operation to the Service Water System (SWS), that controls exist in the Main Control Room (MCR) to cause the pumps identified in VEGP Combined License (COL) Appendix C Table 2.3.1-1 to perform the listed function, and displays of the parameters identified in COL Appendix C Table 2.3.1-1 can be retrieved in the Main Control Room (MCR).

Displays identified in Table 2.3.1-1 can be retrieved in the MCR.


Question: What is the role of the Core Spray system in nuclear reactors?

Core Spray is a safety system designed to provide cooling water to the reactor core during a loss of coolant accident (LOCA).

The isolation valve operator is responsible for ensuring the proper functioning of valves in various operational states.


Question: EPRI MOV Performance Prediction Methodology

Some licensees that evaluated MOVs using the Electric Power Research Institute (EPRI) MOV Performance Prediction Methodology (PPM) did not address all of the applicable provisions when implementing the EPRI MOV PPM to determine valve operating requirements.

The quick brown fox jumps over the lazy dog during its playful antics in the field.


Question: What role does the Emergency Core Cooling System play in reactor safety?

The plant licensing basis considers long-term core cooling (LTCC) following a loss of coolant accident (LOCA) as identified in 10 CFR 50.46(b)(5). Long-term cooling is supported by the ECCS, which includes the charging, safety injection (SI), and residual heat removal (RHR) systems.

The risk-informed evaluation analyzes the following events: 1. Large, medium, and small LOCAs due to: i. Pipe breaks ii. Failure of non-piping components iii. Water hammer.


Question: What is local intense precipitation, and how is it measured?

Local intense precipitation is typically measured by factors such as duration and intensity, as demonstrated in the tables which show the relationship between different durations and their corresponding intensity values.


Question: The NRC mandates strict regulations for tracking radioactive sources based on their associated risks.

Increased Controls for Risk-Significant Radioactive Sources requires licensees to implement enhanced tracking measures for radioactive materials categorized as risk-significant.

The Medical Use of Byproduct Material - Minor Corrections and Clarifications outlines necessary adjustments to ensure compliance with existing drug approval processes.


Question: What are the estimated air emissions for an NGCC plant?

Emissions for an NGCC Plant Alternative (2,120 MWs) include SO₂ at 210 tons/year, NOₓ at 803 tons/year, CO at 1,854 tons/year, PM₁₀ at 408 tons/year, and CO₂ at 6,798,713 tons/year.

Annual gas consumption is calculated using the formula: Plant size in MWe x heat rate x 1,000 x (1/ heat content) x hours in a year.


Question: What are safety-related functions of reactor vessel internals?

Maintain core assembly geometry within the reactor to ensure core cooling, core reactivity control, and the integrity of the fuel cladding as a radioactive material barrier.

Components that provide interfaces between the major assemblies include the radial keys, clevis inserts, fuel alignment pins, head/vessel alignment pins, upper core plate alignment pins, and holddown spring.


Question: What is the required boron concentration for spent fuel pools?

to increase the required spent fuel pool boron concentration to be ≥ 2000 ppm.

CEG performed a criticality safety evaluation for fuel assembly storage in the Braidwood and Byron Spent Fuel Pool (SFP) storage racks and New Fuel Storage Vaults to support the proposed TS change using a methodology described in Attachments 1 and 8.


Question: What is gamma spectrum analysis in nuclear safety?

Gamma spectrum analysis involves measuring the energy levels of gamma radiation to identify and quantify radioactive isotopes present in a given sample, typically utilizing specialized detectors such as the 500ml Marinelli geometry, which optimizes the detection efficiency for this type of measurement.

The peak analysis was performed at 4:54:50 PM on May 30, 2019, and involved a channel range from 100 to 4096.


Question: minimum assembly average enrichment

| wt % 235U (E)      | 35 ≤ Assembly Average Burnup ≤ 37.5 GWd/MTU | Minimum Cooling Time (years)          |

Emergency planning zones are critical for ensuring the safety of the surrounding populations in the event of a nuclear incident.


Question: aging management for cooling water systems

The following aging management programs manage the aging effects for the Open Cycle Cooling Water System components: Bolting Integrity (B.2.1.10), Buried and Underground Piping and Tanks (B.2.1.27), External Surfaces Monitoring of Mechanical Components (B.2.1.23), Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24), One-Time Inspection (B.2.1.20), Open-Cycle Cooling Water System (B.2.1.11), Selective Leaching (B.2.1.21), Water Chemistry (B.2.1.2).

The following aging effects associated with the Open Cycle Cooling Water System components require management: Cracking, Cracking or Blistering, Flow Blockage, Hardening or Loss of Strength, Long-Term Loss of Material, Loss of Material, Loss of Preload.


Question: The NRC's assessment involves measuring the concentration of radioactive materials in the environment, typically denoted by X/Q factors.

Cumulative Probability of Hourly X/Q Values is crucial for understanding the dispersion of radioactive substances in the environment surrounding a nuclear facility.

Topographic Features 0- to 50-Mile Radius from Plant Site illustrate the geographical landscape that may influence environmental assessments.


Question: H-08 control rod removal impact

The reload safety evaluations for Cycle 28 with the H-08 control rod removed validated all cycle-specific safety analysis limits and determined the FSAR Chapter 15 accident analyses remain bounding with respect to the Cycle 28 safety analysis physics parameters and core thermal-hydraulic parameters with the H-08 control rod removed.

By adjusting the software to remove alarms associated with Control Rod H-08, DRPI would continue to function for all other control rods.


Question: measurement uncertainty recapture

The Measurement Uncertainty Recapture Power Uprate effectively allows for an increase in the licensed thermal power of the reactor after the measurement uncertainty has been recalibrated and recaptured during licensing.

The completion of the environmental review process ensures that all relevant environmental issues have been addressed, promoting both public safety and adherence to regulatory standards.


Question: What is the function of the manual initiation for ECCS pumps?

3.f. Manual Initiation (One pushbutton (PB) for Division II) allows for manual actuation of ECCS pumps, providing a backup mechanism for initiating the safety system in case of automatic system failure.

Failure of RCIC Level 8 trip logic will be used as a conservative surrogate for instrument channels as it is a direct failure of RCIC.


Question: radiation surveillance program

Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-10786, February 1985.

Control of Delta Ferrite in Austenitic Stainless Steel Weldments, WCAP-8324, May, 1974.


Question: Radwaste Building functions

The Radwaste Building is an overall integral part of the Fire Protection Program and essential to the NFPA 805 NSCA Performance Goals.

The aging management review results for these components are provided in Table 3.5.2-10, Turbine Buildings - Summary of Aging Management Evaluation.


Question: The NRC staff has highlighted the need to consider the unacceptable risks associated with the large early release frequency (LERF) in the assessment of safety systems.

Revision 3 of RG 1.174, *An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis* (ML17317A2560), provides risk acceptance guidelines in terms of change-in-risk in combination with total core damage frequency (CDF) and large early release frequency (LERF).

The References in this section incorrectly pointed to Reference 27 and 28, when in fact the intention was to add References 47 and 48 to the list of References and direct the reader to them.


Question: uncertainty parameters in BWR/6

Table C.7-4 provides uncertainty parameters and distributions specifically for the BWR/6 Performance Requirements for Fuel Core (PRFC).

The operation of the backup cooling system is critical for preventing overheating in the event of a reactor failure.


Question: maximum expected earthquake intensity

Therefore, the maximum expected earthquake intensity in the region is estimated to be between VII and VIII (7.5).

Therefore, it is concluded that there are no identifiable active faults in the region of the site.


Question: AST inventory of nuclides


Question: What is the purpose of the Atomic Safety and Licensing Board in regulatory filings?

The Atomic Safety and Licensing Board is responsible for conducting hearings on the safety and licensing of nuclear facilities, including the evaluation of regulatory compliance and the analysis of safety-related issues posed during license applications.

The Federal Power Commission's report focuses on the technical basis for energy conservation and does not discuss nuclear safety or licensing processes.


Question: The NRC assesses the safety and licensing of installations involved in nuclear energy production.

Palisades Request for Reauthorization relates to the operational licenses of nuclear power plants to ensure they meet safety standards.

The mail envelope properties include tracking status details, sender and recipient information, and timestamps.


Question: The requirements for technical specifications are critical to ensure that operations uphold public safety.

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TS) as part of the operating license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public.

Components were excluded based on qualitative or quantitative justification for not being significant contributors to system or functional reliability.


Question: What is the role of risk-informed completion times (RICT) in regulatory assessments?

The use of realistic success criteria is necessary to accurately model the as-built as-operated plant, and to conform to capability Category II of the PRA standard (as required by NEI 06-09 Revision 0-A).

Table E1-1 provides a list of candidate TS Actions to which a RICT may be applied and how the associated SSC is addressed in the PRA.


Question: What are the NRC inspection requirements for nuclear facilities?

As discussed with you at the conclusion of our videoconference, no violations of NRC requirements were identified during the 8/21/2024 inspection.

The attached NRC Form 591M provides the inspection results.


Question: BFN's processes involve stringent methodologies to monitor and mitigate water quality impacts in accordance with environmental regulations.

Compliance with the current NPDES and stormwater regulatory requirements and permit conditions as well as implementation of the SWPPP, BMPs, and the Integrated Pollution Prevention Plan will ensure continued insignificant impact on surface water quality during the subsequent period of extended operation.

Land disturbance is not currently anticipated for continued operations, but could result from routine infrastructure maintenance or renovation activities to maintain and upgrade or replace infrastructure and structures if needed to support BFN operations.


Question: The NRC's aging management program addresses potential degradation in fire protection systems.

Aging Management Program: Fire Water System (B.2.15)

Loss of Material: The aging effect requiring management can occur in various environments.


Question: nuclear regulatory framework

The Nuclear Regulatory Commission (NRC) is responsible for regulating the nation's civilian use of nuclear materials and ensuring the safety of nuclear reactors.

Environmental concerns have been a growing issue in multiple sectors, leading to advancements in renewable energy technology.


Question: What safety-related equipment is located in Fire Areas 28 and 29?

Mechanical piping drawings show location of major pieces of equipment in the area, including Main Steam Isolation Valves, Main Steam Line Piping, and Auxiliary Feedwater Line Piping, all noted as safety-related.

Walls and floors which adjoin other plant areas are sealed with three hour fire rated assemblies.


Question: Common Q Record of Changes

A licensee implementing an application based upon the Common Q platform should perform a review of the current Common Q Record of Changes document to assess the validity of previously derived safety conclusions if changes have been made to the Common Q SPM.

Testing will also address license requirements associated with the LGS Technical Specification, which will include the approved changes for this modification.


Question: What is the purpose of the air ejector off-gas subsystem in nuclear plants?

The condenser air ejector effluent is collected and processed in the air ejector off-gas subsystem.

Each filter assembly is composed of an integral HEPA filter element with a moisture separator all housed in one pressure vessel.


Question: What are the key conditions of NRC Material Licenses?

Amendment No. 49 to your NRC Material License No. 21-10459-01 in accordance with your request.

If there are any errors or questions, please notify the U.S. Nuclear Regulatory Commission, Region III office at (630) 829-9887 so that we can provide appropriate corrections and answers.


Question: The NRC mandates the monitoring of effluent releases to ensure compliance with safety standards.

Basic Pathways for Gaseous and Liquid Radioactive Effluent Releases to the Public are established to outline the regulatory requirements.

The Illinois Wind Resource Map provides an overview of potential wind energy generation across the state.


Question: What is the purpose of the N-19 Core Spray Nozzle in a reactor?

The N-19 Core Spray Nozzle serves as a pressure boundary and is made from carbon or low alloy steel with stainless steel cladding.

Water Chemistry (B.2.1.2) is crucial for managing various aging effects in reactor components.


Question: The surveys conducted ensured that the data collected adhered to the established Data Quality Objectives (DQOs), providing evidence for classification decisions.

The measurement results indicated that all samples were less than a SOF of one (1) when compared to the OpDCGLs.

There were no addendums to the FSS plan.


Question: What is a reactor coolant system?

The reactor coolant system is a vital component in maintaining the safe operational limits of a nuclear reactor, ensuring that adequate heat is removed from the reactor core during operation and in the event of an accident.

The analysis report discusses various safety protocols and measures to ensure the reliable operation of pipelines from the reactor to the fuel storage area.


Question: TRISO fuel fabrication process

This fuel fabrication facility utilizes uranium enriched to less than 20 percent to produce Tri-structural Isotropic (TRISO) based particle fuel products to support a variety of advanced reactors (ARs) being considered and planned by the energy and industrial sectors, the Department of Defense, and the National Aeronautics and Space Administration.

Chapter 3 (Description of Affected Environment) discusses the existing environmental conditions at the HCS, and Chapter 4 (Environmental Impacts) discusses how those conditions are affected, if at all, by the Proposed Action.


Question: aging management evaluation for concrete structures

| Concrete (accessible areas): Interior | Gaseous Release Path, Shelter and Protection, Shielding, Structural Support | Concrete | Air - indoor uncontrolled | Cracking; loss of bond; and loss of material (spalling, scaling) due to corrosion of embedded steel | Structures Monitoring (B.2.1.33) | III.A9.TP-26 | 3.5-1, 066 | A |

| Component Type | Intended Function | Material | Environment | Aging Effect Requiring Management | Aging Management Program | NUREG-2191 Item | NUREG-2192 Table 1 Item | Notes |


Question: nuclide MDA levels

The Line MDA (pCi/M²) and Nuclide MDA (pCi/M²) columns in the Nuclide Data table indicate the minimum detectable activity levels for various nuclides, expressed in picocuries per square meter (pCi/M²).

Detector information includes specific details about the detector such as Detector Name, Sample Geometry, and Nuclide Library used.


Question: The NRC mandates that licensees develop and submit plans for sheltering during a nuclear incident.

Evacuation Message: The message sent to individuals in the vicinity of a nuclear facility to instruct them to leave the area as a precautionary measure.

Reentry protocols define the procedures that must be followed after an evacuation, ensuring the safety and readiness of returning individuals.


Question: What is the significance of nuclide MDA in nuclear regulatory contexts?

Nuclide MDA (Minimum Detectable Activity) values represent the lowest levels of radioactivity that can be detected with a specified level of confidence at a given time and are critical for ensuring compliance with safety and regulatory standards in nuclear facilities.

Emergency preparedness is paramount in ensuring that all personnel understand the potential risks associated with operating a nuclear facility, which includes training protocols and response strategies.


Question: aging management review

As a result, the Structures Monitoring (B.2.1.33) program is expected to adequately manage the cracking due to expansion and reaction with aggregates that could occur in inaccessible reinforced concrete areas of Group 6 structures.

Cracking associated with expansion due to reaction with aggregates has not been observed in DNPS Group 6 concrete structures.


Question: RWCU suction system

These two welds are unique to the RWCU Suction on RR Loop A, which is connected using 4" TEE (includes, FW-12WF1, FW-12WF2, FW-12WF3). Details are below: refer to 6M721-3096-1, AF; Section A4-A4.

This weld is identified as discrepancy between the Weld List and reconciliation spreadsheet.


Question: weighted mean activity of nuclides

The weighted mean activity of a nuclide is calculated using its activity (in pCi/grams) and the associated uncertainty, providing a reliable estimate for regulatory assessments.

Errors quoted at 2.00 sigma indicate the level of statistical confidence in the measurements across the undetermined solutions.


Question: recirculation flow control system

The function of the RFCS is to control reactor power by controlling the reactor core flow rate.

Figure C.5-24 and Figure C.5-25 show comparisons of the simulation results and test data.


Question: gaseous dose rate

Gamma Dose Rate in Air (mrad/year) values are provided for various locations, indicating the level of radiation exposure that individuals may receive from gaseous emissions.

The Total Dose (mrem/year) is measured according to different age groups, giving a clear understanding of how exposure varies throughout life stages.


Question: Personnel must adhere to guidelines for using respirators in areas where containment of radioactive materials is not assured.

Respirators will be used only for operations where it is not feasible to prevent atmospheric contamination by effective engineering controls such as process enclosure or ventilation.

Employees will leave an area where respiratory protection is required at any time for relief from respirator use in the event of equipment malfunction, physical or psychological distress, procedural or communication failure, significant deterioration of operating conditions, or any other condition that may require such relief.


Question: The report includes detailed assessments on atmospheric dispersion with specific reference to seasonal variations in dilution factors.

Tables 2.E-24 through 2.E-27 contain estimates of atmospheric dilution factors for the four seasons, winter through fall.

Tables 2.E-1 through 2.E-7 contain climatological and severe weather data for Columbia, South Carolina.


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

Safety-related systems are those systems whose failure could prevent fulfilling the safety function of the nuclear facility and are therefore subject to rigorous design, testing, and operational protocols.

The reactor pressure vessel is designed to contain the nuclear reaction and provide a barrier against the release of radioactive materials to the environment.


Question: interlocks in safety systems

Knowledge of the interrelations between the (LOCA Outside Containment) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

The operator is instructed to close 8809A, RHR to Cold Legs 1 and 2.


Question: license amendment NRC

The License Amendment, Mail Control No. 644965 Dated January 29, 2025, pertains to the regulatory approval required for modifications to the operational parameters of a licensed facility.

To receive a copy of this document, indicate in the box: - "C" = Copy w/o attach/end - "E" = Copy w/ attach/end - "N" = No copy


Question: The analysis shows that exceeding the Specified Acceptable Fuel Design Limit is an important failure criterion in reactor operations.

Exceeding the DNBR Specified Acceptable Fuel Design Limit (SAFDL) is a failure criterion [ ].

The fuel temperature and clad temperature as a function of time are shown in Figure 4-2.


Question: The analysis of environmental consequences, including mitigation for every resource, is outlined in the regulatory documents.

Following the requirements of NUREG-1748, this section describes the potential environmental impacts by resource for the Proposed Action and No Action Alternative.

The potential impacts described in this section are based on information presented in the CUP FEIS, COP, and ISR GEIS as well as any new information presented in Chapter 3 of this ER.


Question: The NRC has established protocols for review and discussion of regulatory issues related to safety inspections.

The purpose is to discuss the issues related to the NRC staff review of TSTF-577, including TSTF responses to NRC staff request for additional information (RAI) questions.

This is a Category 2 meeting. The public is invited to participate in this meeting by discussing regulatory issues with the Nuclear Regulatory Commission (NRC) at designated points identified on the agenda.


Question: nuclear regulatory commission environmental impacts

The Atomic Energy Act requires the Nuclear Regulatory Commission to promulgate, inspect and enforce standards that provide an adequate level of protection of the public health and safety and the environment.

The Staff also incorrectly asserts that 'Petitioners do not demonstrate that strict application of the rule ‘would not serve the underlying purpose’ of the Category 1 findings in Table B-1' because they 'do not provide any information to demonstrate that there are environmental impacts resulting from the 2011 Mineral earthquake, let alone significant environmental impacts, that would lead to a determination that (a) the environmental impacts for design-basis accidents or (b) the probability-weighted consequences of a severe accident, would be greater than SMALL for this SLR review.'


Question: strontium-90 measurement

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


Question: What is NEPA analysis in the context of nuclear licensing?

For the reasons explained in CLI-22-2, we conclude that the Staff did not conduct an adequate NEPA analysis before issuing Exelon licenses for the subsequent license renewal period for Peach Bottom.

In CLI-20-11, the Commission affirmed the Board’s decision and, with respect to Contention 2A, relied, in large part, on the decision in CLI-20-3.


Question: nondisclosure agreement in nuclear licensing

AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.


Question: radiological hazard assessment

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

The conduct of this dual-site drill successfully demonstrated the ability of the Dominion Energy CERC to continue to effectively function during simultaneous events at multiple nuclear power reactor sites with different technologies and to interface effectively with the State agencies.


Question: surveillance requirements frequency

Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time.

The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment or variables are outside specified limits, or the facility is outside the Applicability of the LCO).


Question: What are the analytical results for dissolved metals in groundwater?

The sample matrix for this analysis consisted of groundwater, which is typically analyzed for various contaminants.


Question: The ICE report must include an executive summary and a description of the technical baseline, detailing how costs are assessed and estimated.

A final draft ICE report will be developed to reflect any changes resulting from the reconciliation meeting.

Reconciliation does not necessarily mean consensus.


Question: minimum detectable concentration

**Client Sample ID:** L3-012-101-RAGS-004-SS


Question: What steps should be taken during a Large Break LOCA?

1-E-1, Loss of Reactor or Secondary Coolant, has just resumed after performance of 1-ES-1.3. The examinee is directed to perform 1-E-1, Loss of Reactor or Secondary Coolant, Step 26, DETERMININE if RHR spray should be placed in service.

Operators responding to a Loss of Auxiliary Air using 0-AOI-10, Loss of Control Air, Section 3.2, Loss of Auxiliary Air, due to an unisolable header rupture.


Question: ECCS flow reduction criteria

(7.a) Checks RCS Subcooled based on core exit T/Cs GREATER THAN 20°F (Subcooled Margin Monitor, YI 31 or Appendix C)

The scenario may be terminated once S/G 1-2 Fault has been isolated (S2CT-3 complete)


Question: design pressure in nuclear systems

Design Pressure (psig) refers to the maximum allowable pressure for equipment, which is critical for ensuring the integrity and safety of the nuclear reactor systems.

The control readout includes various metrics such as conductivity, speed, and vibration, each of which is essential for monitoring the reactor's operational parameters.


Question: The management of nuclear waste and its regulation is a critical issue that has faced continual scrutiny.

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

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


Question: aging management evaluation for reactor buildings

The Aging Management Review Results section outlines the evaluations performed on various components of reactor buildings, including assessments of their intended functions and potential aging effects that require management.

The increase in porosity and permeability of concrete can lead to structural integrity issues if not addressed promptly during aging management evaluations.