Question: The evaluation of core failures due to debris accumulation considers the operational states of all ECCS pumps, particularly identifying scenarios where core damage may occur.

All breaks equal to or larger than the threshold size were considered, which could result in zero damage due to effects of debris.

Risk quantification is based on equipment configurations.


Question: What is the significance of technical specification bases changes in nuclear licensing?

The proposed changes to the technical specifications bases are designed to clarify the basis for various technical specification requirements and ensure ongoing compliance with regulatory standards.

The cooling system of the reactor operates in a closed loop to maintain optimal temperature and efficiency under varying load conditions.


Question: The operator must ensure proper management of steam generator pressure to prevent overpressure conditions that can compromise safety.

Guideline FR-H.2, RESPONSE TO STEAM GENERATOR OVERPRESSURE, provides guidance to address a not satisfied condition (i.e., YELLOW priority) from a loss of steam generator pressure control capability that results from an overpressure condition affecting any steam generator.

The operator is instructed to transfer to FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL, after determining that the SG narrow range level is greater than the upper tap value.


Question: What is the purpose of an isolation valve in nuclear reactor systems?

An isolation valve is used to prevent the flow of fluids within a nuclear reactor system during maintenance or in the event of a failure, ensuring that safety protocols are maintained.

The reactor core design must account for various factors including neutron absorption and thermal conductivity to ensure efficient operation.


Question: What are the design requirements for nuclear facility safety?

(a) • Demonstrate functional design criteria via analysis, test, etc.; • Evaluate operating, design and construction experience

(b) Consensus codes and standards acceptable to NRC


Question: The NRC outlines acceptable levels of various chemical constituents in water samples.

| Ammonium            | mg/L      | 0.01 | WF AVG                  | 10.0                       |

The document details multiple constituents and their corresponding units relevant to water quality analysis.


Question: Cooling time requirements are specifically adjusted based on the average burnup of the fuel assemblies.

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

Amendment No. 12, Revision No. 1 includes a Certificate of Compliance No. 1031.


Question: The NRC mandates that critical heat flux analyses be performed for safety evaluations.

The COBRA-FLX model is used to predict the Critical Heat Flux (CHF) in nuclear reactors, which is essential for ensuring safe operating conditions under various heat transfer scenarios.

Thermal energy systems are integral for many industrial processes, but their configuration can greatly affect operational efficiency.


Question: nuclear power plant licensing process

The nuclear licensing process involves a comprehensive review of safety analyses, environmental impacts, and engineering evaluations to ensure that the facility meets regulatory standards and poses no undue risk to public health and safety.

The reactor will undergo routine inspections to verify compliance with established maintenance protocols and operational procedures.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves multiple steps, including pre-application consultations, the submission of certification and operating license applications, and the subsequent review and approval by the Nuclear Regulatory Commission (NRC).

The reactor's cooling system must maintain optimal temperatures to ensure safe and efficient operations, utilizing a series of pumps and heat exchangers.


Question: nuclear instrumentation malfunctions

Nuclear Instrumentation Malfunctions refer to faults or failures in the nuclear instrumentation systems that can affect plant operations and safety assessments.

The Watts Bar Nuclear Plant has undergone several upgrades and maintenance activities to ensure reliable operation and compliance with regulatory standards.


Question: requirements of the general license in nuclear regulations

C. I am aware of the requirements of the general license, provided in 10 CFR 31.5.

A. All information contained in this registration is true and complete to the best of my knowledge and belief.


Question: price elasticity of electricity

There is much controversy regarding the price elasticity of electricity which measures the sensitivity of electricity demand to changes in its price.

The flow of energy from the TVA system to its customers is shown in Fig. 8.3 for 1960 and 1972.


Question: reactor trip function

Trips all main feedwater pumps which closes the pump discharge valves.


Question: What does initial criticality mean in nuclear licensing?

Initial criticality refers to the state of a nuclear reactor in which the reactor is capable of sustaining a nuclear chain reaction at a steady state under specified operating conditions.

The NRC requires a thorough evaluation of the control systems to ensure that all actions taken during an operational transient are effective in maintaining safety.


Question: The evaluation of seismic performance is crucial for safety assessments during licensing.

Seismic analysis is essential for determining the structural integrity of nuclear facilities in the context of earthquakes.

The control room must be equipped with reliable communication systems to ensure proper interface during operations.


Question: The NRC reviews license renewal applications while considering those aging effects that might not be routinely managed by established regulatory oversight programs.

NRC's review of a license renewal application focuses on those potential detrimental effects of aging that are not routinely addressed by ongoing regulatory oversight programs and the applicant's plans for managing those effects during the period of extended operation; CLI-22-5, 95 NRC 97 (2022)

Motion for reconsideration must demonstrate compelling circumstances, such as existence of a clear and material error in a decision, which could not have been reasonably anticipated, that renders the decision invalid; CLI-21-2, 95 NRC 116 (2022)


Question: What are the environmental impacts assessed for aquatic resources in nuclear plant license renewals?

Effects of cooling water discharge on dissolved oxygen, gas saturation, and eutrophication

Would not adversely affect known historic properties


Question: The NRC mandates evaluations for components like galvanized steel ducting that are used in specific environmental conditions.

Galvanized steel Ducting, piping, and components exposed to Air – indoor, controlled (External)

There are no copper alloy (≤15% Zn and ≤8% Al) piping, piping components, or piping elements exposed to borated water leakage.


Question: The reactor protection system is designed to automatically shut down the reactor under certain conditions.

An automatic reactor scram on a Turbine trip is bypassed when Reactor power is < 38.5% rated thermal power.

Annunciator 902-5 A-12, is actuated by closure of at least 2 Stop valves with reactor power > 40%.


Question: What are the requirements for the Residual Heat Removal (RHR) system during refueling operations?

One RHR loop shall be OPERABLE and in operation.

The required RHR loop may be removed from operation for ≤ 1 hour per 8 hour period, provided no operations are permitted that would cause introduction into the Reactor Coolant System, coolant with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.


Question: Radiation exposure levels are categorized into specific ranges based on counts per minute (cpm).

Color codes are used to indicate the range of radiation exposure levels, where Blue signifies 0 to 10,000 cpm, Green indicates 10,001 to 13,000 cpm, Yellow refers to 13,001 to 16,000 cpm, and Red denotes 16,001 to 19,000 cpm.

The reactor's coolant system is essential for maintaining optimal operating temperatures and preventing overheating.


Question: The NRC mandates that sealed sources must undergo leakage testing at specified intervals or else at a maximum of six months.

Sealed sources, source rods, foil sources, or detector cells containing licensed material shall not be opened or sources removed from source holders or detached from source rods, or foil sources removed from detector cells by the licensee, except as specifically authorized.

Licensed material shall only be used by, or under the supervision of Jeffrey Q. Angel or William Lee Cain.


Question: What are the requirements for traceability in RadICS platform safety systems?

The RadICS platform TR does not address traceability activities associated with application specific logic. Therefore, plant application requirements traceability activities for RadICS platform-based safety systems must be performed during plant application development and thus could not be evaluated in this SE.

Addressed in PSAI 7.2 and satisfied by: [[ - - - - - ]]a,c,e


Question: What is NMGIRD in the context of CHF tests?

Figure A-10 shows the axial geometry of the CHF test CE300, illustrating various heights and distances from EOHL.


Question: What is the fuel clad gap coefficient and its significance in thermal analysis?

The magnitude and time dependence of the heat transfer coefficient between fuel and clad (gap coefficient) have a pronounced influence on the thermal results.

The transient results for the locked rotor accident are shown in figures 15.3.3-1 through 15.3.3-4.


Question: decay constant of nuclides


Question: What is the role of the Parish School Board in a nuclear emergency?

The Parish School Board, under the direction of the Superintendent, has the responsibility for providing school buses and bus drivers for evacuation of students and school personnel from schools within the Parish 10-mile EPZ.

The Pointe Coupee General Hospital is responsible for providing health and medical care services to non-contaminated injured individuals.


Question: During the blowdown phase, the mass flow rate is significantly reduced as pressure drops.

Mass flow rate is defined as the amount of mass passing through a given surface per unit time, measured in pounds mass per second (lbm/s).

The table shows the mass flow rates calculated at different time intervals during the system operation.


Question: The NRC has established specific guidelines to ensure aging management is adequately addressed in the license renewal applications.

The aging management programs outlined in this submission are designed to adequately address the effects of aging on structures and components important to safety over the period of extended operation.

Different methods of reactor design can influence operational efficiency, but they do not directly correlate with aging management requirements.


Question: pressure relief valve functions

The pressure relief valve downstream of the low pressure letdown valve protects the low pressure piping, demineralizers, and filter from overpressure when this section of the system is isolated.

The valve is set to relieve at the design pressure of the heat exchanger.


Question: surface mining regulatory requirements

Surface mining involves the extraction of mineral resources like coal and aggregates from the earth's surface, requiring adherence to strict regulatory guidelines to ensure environmental and safety standards are met.

The distance from facilities to the CUPI can vary significantly, with some located as close as 20 km and others up to 70 km away, indicating the diverse operational scopes of various mining projects.


Question: What are the aging management strategies for ICI thimble tubes in PWRs?

EPRI clarified that the aging management strategy selected for the component type through the end of a 60 – 80-year subsequent license renewal period would involve at least one of the specified aging management strategies accepted in GALL-SLR AMP XI.M16A for managing specific aging effects in the PWR RVI components.

The NRC staff finds the I&E criteria for the “ICI thimble tubes – lower” component type in MRP-227, Revision 2, Table 4-8, Item C20 (as amended inclusive of changes to Item C20 in the EPRI letter MRP 2023-003 (EPRI 2023a)) to be acceptable for implementation based on the following technical assessments.


Question: What is the weight mean activity of nuclides in a regulatory filing?

The weight mean activity of a nuclide is expressed in pCi/grams, which indicates the average radioactivity per gram of the nuclide present in a sample.

The table lists various nuclides and their associated confidence levels, but does not provide information about the safety protocols for handling radioactive materials.


Question: The NRC mandates that seismic evaluations be conducted for structures crucial to nuclear safety.

The seismic design of structures, systems, and components in nuclear power plants must comply with the established criteria to ensure safety during seismic events.

The earthquake event of 1968 produced a 5.5 magnitude tremor that was felt across the mid-Atlantic region, causing concern among the population.


Question: What is the significance of zirconium-steam reaction kinetics in nuclear safety?

In the passage above, Schanz, the author, is referring to zirconium-steam reaction kinetics in 'the high temperature range,' at temperatures far greater than the 10 C.F.R. § 50.46(b)(1) peak cladding temperature ('PCT') limit of 2200°F.

It is significant that in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe ('KfK') GmbH, Karlsruhe, FRG, October 1-4, 1990, problems with SCDAP/RELAP5's modeling of Zircaloy oxidation kinetics, in the 900-1200°C temperature range, were discussed.


Question: What are the criteria for high-energy line breaks (HELBs) according to NRC Branch Technical Position MEB 3-1?

NRC Branch Technical Position MEB 3-1 is the basis for the VEGP criteria for the postulation of high-energy line breaks (HELBs) with the exception of lines that have eliminated postulated breaks based on leak-before-break analysis.

The design of ASME III Code Class 2 and 3 piping systems at VEGP incorporates stress reduction factors for determining the acceptability of the piping design with respect to thermal stresses.


Question: The NRC establishes requirements for the enrichment levels of uranium in nuclear fuel assemblies...

Minimum Cooling Time (years) is needed to ensure proper radioactive decay and safe handling of spent fuel.


Question: The volume of air in the ice condenser is determined based on thermodynamic calculations involving temperature and pressure measurements.

In the calculation of the mass balance for the ice condenser, the compartment is divided into two sub-volumes; one volume representing the flow channels and one volume representing the ice baskets.

The polytropic exponent for these tests is then found from the measured compression pressure and the compression ratio calculated above.


Question: design loads for nuclear structures

SEI/ANSI 7-02 "Snow Loads: A Guide to the Use and Understanding of the Snow Load Provisions of ASCE 7-02" (O'Rourke and Wrenn, 2004).

US Historical Climatology Network (U.S. HCN), Monthly Temperature and Precipitation Data, ORNL/CDIAC-87, NDP-019, Revision 3.


Question: Regulatory commitments are essential in ensuring compliance and safety in operations.

The NRC establishes regulatory requirements to ensure that nuclear facilities operate safely and effectively, thereby protecting public health and the environment.

Technical skills are essential for personal development and workplace effectiveness.


Question: What are the decontamination factors for iodine species in a fuel pool?

The pool decontamination factors for the inorganic and organic species are 400 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the pool). This difference in decontamination factors for inorganic and organic iodine above the fuel pool being composed of 50% inorganic and 50% organic species.

The radioactive material that escapes from the pool to the building is released from the building to the environment over a 2-h time period.


Question: nuclear reactor licensing process

The Division of Operating Reactor Licensing is responsible for reviewing applications for operating licenses and ensuring compliance with regulatory requirements.

The Audit Requests and Audit Questions provide a framework for evaluating the effectiveness of the audit process and ensuring that all necessary information is gathered.


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

K-40, or Potassium-40, is a naturally occurring radioactive isotope with a weight mean activity of 7.09E-01 pCi/grams, which is crucial for understanding the natural background radiation levels in nuclear safety assessments.

The uncertainty in the confidence interval plays a critical role in interpreting experimental results, as it informs the reliability of the measurements and calculations.


Question: reporting of events affecting nuclear power plants

For systems within scope, the inadvertent TS inoperability of a system in a required mode of applicability constitutes an event or condition for which there is no longer reasonable expectation that equipment can fulfill its safety function. Therefore, such events or conditions are reportable as an 'Event or Condition that Could Have Prevented Fulfillment of a Safety Function.'

§50.72(b)(3)(xii) - Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.


Question: personnel dosimeter requirements

Change 'film badges and TLD’s' with 'personnel dosimeter'

ALL PARTS   Clarify/remove all references to x-ray devices whose output is less than 1 MeV.


Question: Consideration of diverse technologies in nuclear safety

The discussion highlights the importance of considering diverse technologies and the associated hazards when assessing the safety of nuclear reactor designs.

Given the various challenges faced in the industry, it is crucial to keep efficient communication lines open to address potential safety concerns.


Question: What is the significance of weighting mean activity in radioactivity measurements?

The weighted mean activity is calculated to provide a more accurate representation of the radioactivity of a sample by taking into account the contributions of various nuclides present in the sample, which allows for better compliance with regulatory standards.

Nuclides can vary significantly in their decay paths and radiation emissions, making it essential for safety assessments to evaluate them individually rather than as a collective group.


Question: What are risk-informed extended completion times in nuclear plant regulations?

The proposed LAR would modify the plant’s Technical Specifications to permit the use of risk-informed completion times in accordance with Technical Specification Task Force (TSTF) Traveler TSTF-505, Revision 2, 'Provide Risk-informed Extended Completion Times - RITSF [Risk-Informed Technical Specification Task Force] Initiative 4b.'

The NRC staff will conduct the audit virtually via Teams using a licensee-established electronic portal available to NRC staff from approximately August 22, 2022, through December 16, 2022, with formal audit meetings to be scheduled during this period as needed.


Question: acceptable bolting pattern analysis methodologies

EPRI MRP needs to be prepared to discuss use of "acceptable bolting pattern analysis methodologies" for BF bolts (e.g. WCAP-15029-P-A). They are being referenced in WCAP-17096, Rev. 3.

supplemental methods that currently remain in effect for Westinghouse-design BF bolts.


Question: What is the function of the GALILEO predictions in nuclear reactor safety analysis?

The GALILEO predictions show good agreement.

Figure 4-42 shows the comparison of measured and predicted rod diameter axial profiles at mid-pellet and inter-pellet planes on ramp PK2/3 from the SuperRamp program.


Question: The assessment of net peak area is critical in determining the performance of the detector during experiments.

Net Peak Area refers to the area under the peak in a spectrum, which is crucial for quantifying the amount of a specific energy being detected.

Ambient background refers to the level of radiation present in the environment that is not associated with the source being measured.


Question: transfer cask usage restrictions

The **TRANSFER CASK** shall not be operated and used when surrounding air temperature is < 0°F. This limit is NOT applicable to the stainless steel MTC or PMTC.

An alternative to crediting site-specific cask sliding for design basis earthquake accelerations up to and greater than 0.37g in the horizontal direction and 0.25g in the vertical direction at the ISFSI pad top surface, the use of the MAGNASORB® system is permitted provided the ISFSI pad has bollards and the cask user validates that the cask does not overturn, g-loads resulting from the cask contacting the bollard is bounded by the cask tip-over accident condition presented in Chapter 3 of the FSAR, and the ISFSI pad and bollards are designed, fabricated and installed such that they are capable of handling the combined loading of the design basis earthquake and any contact between the bollard and cask during the design basis earthquake.


Question: seismic insights determination process

Determination of seismic insights will make use of the FPIE PRA model supplemented by focused seismic walkdowns.

Gather the population of SSCs in the system being categorized and review existing seismic information.


Question: What are the oxidation kinetics models for Zircaloy?

It is significant that "In-Vessel Phenomena—CORA: BWR Core Melt Progression Phenomena Program, Oak Ridge National Laboratory," presented in 1991, explicitly states "[c]ladding oxidation [in the CORA-16 experiment] was not accurately predicted by available correlations."

The predicted and observed cladding thermal response are in excellent agreement until application of the available Zircaloy oxidation kinetics models causes the low-temperature (900-1200°C) [(1652-2192°F)] oxidation to be underpredicted.


Question: emergency depressurization procedures

With a primary system discharging into the reactor building and the discharge cannot be isolated, verify/INITIATE an emergency depressurization when two or more areas of the same parameter (radiation, temperature, or water levels) are above max safe IAW QGA 300 and appropriate RPV blowdown leg within 10 minutes.

The evaluation team must determine if the local reports justify performing an emergency depressurization in such cases, especially where accessibility is limited.


Question: nuclide identification

The Nuclide Identification Report provides essential details, including the nuclide name, its energy in keV, yield percentage, activity in µCi/grams, and the associated activity uncertainty.

Errors quoted at 1.00 sigma Energy Tolerance: 1.000 keV Nuclide confidence index threshold = 0.30 Errors quoted at 1.00 sigma.


Question: radioactive isotopes and their activity

NRC Device Key: 625216 (Internal Control Number)


Question: What is the activity measurement for K-40?

K-40 has an activity of 9.19E+00 pCi/gram.

The yields of the radionuclides vary significantly, with some yielding up to 100%.


Question: monitoring MFWP suction pressure

The reason that the ARI directs MFWP suction press to be monitored is to ensure that it is maintained greater than 250 psid. If unable to recover suction pressure, the crew is required to trip one MFPT and go to 1-AOI-16, Loss of Normal Feedwater.

The second part is plausible because the low suction pressure condition affects both MFPs.


Question: What is a Reactor Protection System (RPS)?

RPS stands for Reactor Protection System, which is designed to automatically monitor and control reactor parameters to ensure safe operations and to prevent conditions that could lead to a release of radioactive materials.

The software verification and validation plan includes procedures for testing software changes before implementation, ensuring that modifications do not introduce new errors.


Question: The operational tests were essential in confirming the functionality of the pumps tasked with providing emergency core cooling.

Routine periodic testing of the Emergency Core Cooling System components and all necessary support systems is performed.

Prototype testing has determined that vortex suppression grating and enclosure solid top plate prevent the strainer modules from ingesting air via vortex.


Question: The review of the license application involves assessing safety measures for reactor operation.

The NRC conducts safety evaluations to determine whether a reactor design meets regulatory requirements for operation.

In addition to the reactor design, the applicant must include comprehensive financial data to support their operational proposal.


Question: Isolation Condenser Weld Flaw analysis

4.7.7 Isolation Condenser Weld Flaw TLAA.

ASME Section III, Class 2 & 3, and ANSI B31.1 Allowable Stress Analyses and associated HELB Analyses.


Question: Monticello Nuclear Generating Plant integrated inspection report

On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Monticello Nuclear Generating Plant. On January 23, 2025, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

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


Question: What is the unit used to measure radioactivity in nuclear devices?

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

Distributor/Distributed By: Thermo Process Instruments, L.P


Question: What is an Audit Plan in the context of nuclear licensing?

The Audit Plan outlines the specific objectives, scope, and criteria for conducting an audit of the licensing process for nuclear reactors, ensuring compliance with regulatory standards.

In the context of reactor safety, the design must account for various potential failure scenarios and operational challenges that could arise during normal and emergency operations.


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

A safety analysis report (SAR) provides a comprehensive assessment of the safety of the proposed facility, including all potential hazards and mitigation strategies.

The reactor coolant system's main function is to remove heat from the core and transfer it to the steam generators.


Question: nuclide activity measurement


Question: What actions are taken if the reactor fails to scram?

Given a failure to scram with Reactor power above 2.5%, the crew will lower Reactor power by one or more of the following methods, in accordance with EOP-3: - Terminating and preventing all RPV injection except SLC, RCIC, and CRD - Tripping Recirculation pump A - Injecting boron

The Reactor power reduction must be initiated within five minutes of the start of the failure to scram.


Question: What is the process for changes to a certificate of compliance in nuclear regulation?

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

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


Question: NEI 14-05 quality assurance guidelines

The proposed amendment to QATR Revision 12 also included a commitment to implement the quality assurance program alternative of the Nuclear Energy Institute (NEI) Technical Report NEI 14-05, Revision 1, "Guidelines for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," dated August 2014 (Reference 9).

The NRC staff found this approach to be acceptable.


Question: Unusual Event emergency classification

When informed that River Bend Station has declared an Unusual Event, the EOC Director will complete the following and document the actions on Attachment 1: A. Obtain the status of the emergency.

The State EOC will assist if you have any questions concerning the emergency.


Question: What are the total estimated disposal costs for spent nuclear fuel in a PWR system?

The total estimated disposal costs for a Pressurized Water Reactor (PWR) are outlined as $196,500,420 in the disposal cost table.

The table lists various components and their corresponding charges but does not detail the operational processes involved in managing reactor safety.


Question: LOCA response procedures

The crew will insert a manual scram on high Drywell pressure due to a break in the 'A' Recirc Loop Suction Line.

Electrical Maintenance has requested the 1A EHC pump secured and removed from service for a motor inspection.


Question: The audit will verify how effectively the technical assessments support the licensee's amended applications according to established regulations.

Understand how the licensee’s proposed program conforms to NRC-endorsed guidance in the Nuclear Energy Institute (NEI) report NEI 06-09, Revision 0-A, "Risk-Informed Technical Specification Initiative 4b, Risk-Managed Technical Specification Guidelines" (ML122860402).

Identify questions and requests that may become formal requests for additional information (RAIs) per NRR Office Instruction LIC-115, "Processing Requests for Additional Information" (ML214114238).


Question: What is the analytical method for measuring Sr90 in solid samples?

**Analytical Method:** EPA 905.0 Modified/DOE RP501 Rev. 1 Modified


Question: standby diesel generators emissions

Table 3.1-6 shows the yearly emissions from standby diesel generators, detailing the environmental impact associated with their operation.

Table 4.1-1 outlines the land use within the ESP site and vicinity, which is unrelated to the emissions of standby diesel generators.


Question: The NRC's procedures allow petitioners to contest findings by providing updated ecological evidence.

Organizational Petitioners do not demonstrate that this issue is “material to the findings the NRC must make to support the action” at issue. See 10 C.F.R. § 2.309(f)(1)(iv).

Nor is this aspect of the contention saved by Organizational Petitioners’ argument that the NRC wrongly requires petitioners, not the applicant, to provide updated information.


Question: aging management program

This GALL-SLR item reflects consideration of leakage from a flanged connection.


Question: What is meant by capable tectonic sources in nuclear site assessments?

Based on previous discussions, there are no capable tectonic sources within 5 miles of the ESP site.

The potential for tectonic deformation at the site is negligible.


Question: OpDCGL remediation criteria

Actions in Table 5-17 are required when the statistical test fails or when an investigation confirms residual radioactivity in excess of 50 percent of the OpDCGL.

An extent of condition review for all survey units was performed (see response to RAI-1a). Three survey units (L1-SUB-TDS, L1-010-101 C, and L1-SUB-CRD) had scanning alarm set points that were much higher than the alarm set points established for FSS performed in other survey units subsequent to these three survey units.


Question: Oconee Station power transmission lines

For license renewal and subsequent license renewal, the NRC evaluates, as part of the proposed action, the continued operation of those Oconee Station power transmission lines that connect to the substation where it feeds electricity into the regional power distribution system.

In addition to cooling and auxiliary water from Lake Keowee, Oconee uses potable water supplied by the city of Seneca public water system and Oconee Joint Regional Sewer Authority for sanitary wastewater.


Question: The NRC must complete a technical assessment of the applications received, which may take up to 180 days for renewals.

Your application for a new NRC license did not include your taxpayer identification number. Please complete and submit NRC Form 531, Request for Taxpayer Identification Number, located at the following link: http://www.nrc.gov/reading-rm/doc-collections/forms/nrc531.pdf.

This is to acknowledge receipt of your: ☑ Letter and/or ☑ Application Dated: 01/24/2025.


Question: The performance of the detector is influenced by the length of the pipe containing radioactive material, with longer lengths resulting in higher measured efficiencies.

This data shows that if the assumed length of pipe that contains residual activity is 1 foot, then the measured efficiency from the NIST-traceable calibration source would be increased by a factor of 1.27 to account for activity distributed around the entire circumference of the pipe.

The results of the MCNP simulation are provided in Table 4 and Figure 3.


Question: What are the safety systems required for nuclear reactor licensing?

Safety systems must be designed to ensure the operational safety of nuclear reactors, including active and passive methods for accident prevention and mitigation.

The efficiency of heat exchangers is critical in maintaining operational performance across various industrial applications.


Question: Cs-137 activity value in nuclear assessments


Question: What is neutron fluence in the context of reactor pressure vessel analyses?

Neutron fluence is the term used to represent the cumulative number of neutrons per square centimeter (flux) that contact the Unit 2 and 3 RPV shell and its internal components over a given period of time.

The current uprated power level of 2957 Megawatts thermal (MWt) is the maximum power level evaluated for the SPEO.


Question: A waiver may be warranted in certain unique circumstances regarding Commission regulations, particularly if strict adherence would fail to achieve intended safety objectives.

Over time, Commission caselaw has defined the nature of such “special circumstances” within the meaning of section 2.335(b), culminating in the Commission’s Millstone decision that identified the following four factors as pertinent to evaluating whether a waiver is appropriate: 1. The rule’s strict application would not serve the purposes for which it was adopted; 2. Special circumstances exist that were not considered, either explicitly or by necessary implication, in the rulemaking proceeding leading to the rule sought to be waived; 3. Those circumstances are unique to the facility rather than common to a large class of facilities; and 4. Waiver of the regulation is necessary to reach a significant safety [or environmental] problem.

The NRC regulatory determination as a rule is generally not subject to adjudicatory challenge, emphasizing the stability and protection offered by established regulations.


Question: visual examination of submerged surfaces

In accordance with ASME Section XI, 2007 Edition through 2008 Addenda, Table IWE-2500-1, Examination Category E-A, Item Number E1.12, Wetted Surfaces of Submerged Areas shall receive a visual (VT-3) examination on 100% of the submerged surfaces during the inspection interval.

During Brunswick Unit 2 second interval inspections (March 2015), there were no (0) discovered submerged surfaces in the sixteen (16) submerged torus bays with substrate 'pit' corrosion exceeding the Owner specified acceptance criteria of 90 mils.


Question: What is drain time in nuclear reactor safety?

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

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


Question: What are the activity levels of specific nuclides?

The activity (pCi/g) for CE-144 is -1.17E+00, for EU-152 it is 3.38E-01, and for EU-154 it is -1.04E-01.

The energy levels of nuclides can vary significantly, with values ranging from hundreds of keV to over a MeV.


Question: offsite power maintenance in nuclear plants

Figure 15-159 illustrates the behavior of offsite power maintained during a steamline break, showing the relationship between pressure and time.

The table above displays the pressure readings at various time intervals after a simulated incident.


Question: The measurement of radiation levels is essential for evaluating safety in nuclear facilities.

Xe-133 is a radioactive isotope of xenon that is produced as a fission product in nuclear reactors and has a half-life of about 5.24 days, making it significant in the context of safety and regulatory assessments.

The total thermal power output of the reactor must be carefully monitored to ensure safe operations and compliance with regulations.


Question: What are the requirements for maintaining safety systems during maintenance activities?

To preserve ATWS mitigation capability, activities that degrade the availability of the auxiliary feedwater system, RCS pressure relief system (pressurizer PORVs and safety valves), AMSAC, or turbine trip should not be scheduled when a logic train is inoperable for maintenance.

The Completion Time of 24 hours (Required Action Q.1) is reasonable considering that in this Condition, the remaining OPERABLE train is adequate to perform the safety function and given the low probability of an event during this interval.


Question: The uncertainties associated with hydrogen model predictions are crucial for ensuring safety in reactor operations.

Hydrogen model uncertainties are developed below for application in downstream safety analysis methodologies.

Figure 3-4 shows the distribution of P-M values for the M5 hydrogen predictions.


Question: spent fuel pool regulations

NOTIFY Unit Supervisor that movement of irradiated fuel assemblies into the Spent Fuel Pool will result in changes to SFP, RHR, and CCS heat load(s), and that 0-SOI-70.01 contains applicable instructions and limitations.

ESTABLISH AND MAINTAIN communications between the following stations DURING FUEL MOVEMENT:


Question: The NRC evaluated the effects of various land disturbing activities on the quality of surface water due to nuclear power plant operations.

The NRC concluded that nuclear power plants’ operation under NPDES permits and the implementation of BMPs would mitigate surface water quality impacts from non-cooling systems to SMALL.

The site does not use nor has plans to access surface water for non-cooling purposes.


Question: The NRC mandates that uncertainty values during the evaluation of results must meet specific confidence levels.

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

The project involves careful planning and coordination between various stakeholders to ensure effective execution.


Question: The NRC assesses the potential effects on copper alloy piping and ensures that any material exposed to specific environmental conditions is evaluated accordingly.

Copper alloy piping and piping components, as well as heat exchanger tubes and components, exposed to dry gas or plant indoor air and ventilation atmosphere with low condensation in the Auxiliary Systems are not subject to aging effects.

There are no copper alloy (≤15% Zn and ≤8% Al) piping, piping components, or piping elements exposed to borated water leakage in the Auxiliary Systems.


Question: What are common weld defects in nuclear reactor components?

6X photograph of sample 13, showing slag entrapment and lack of fusion in weld metal.


Question: licensing basis event analysis

These four TRs represent approaches or methodologies to analyze the safety of the Xe-100 design required by the applicable regulations under Title 10 of the Code of Federal Regulations (10 CFR) Parts 50 and 52.

The audit team will examine supporting documentation provided by X-energy in the on-line portal and hold discussions during virtual meetings.


Question: The NRC oversees the requirements for ensuring that groundwater quality meets established regulatory standards during restoration operations.

The recovery process does not introduce new chemical species to the ground water system but does elevate certain species that are native to the host aquifer.

NuFuels will develop an updated groundwater restoration plan for the entire project that includes a refined restoration schedule.


Question: The evaluation of aging management programs is critical for components subject to significant temperature variations.

A plant-specific aging management program is to be evaluated.

DCPP is located in a "negligible" weathering region as determined by ASTM C33, Figure 1.


Question: The operator must verify that the level control channels are functioning correctly according to the control board indications.

CHECK level control channels NORMAL using control board indications [1-M4] or DCS Operator Display: 1-LI-68-339A 1-LI-68-335A 1-LI-68-320

The operator momentarily places the Control Heaters D to ON where the red light is illuminated.


Question: What is the purpose of the Federal Register in NRC regulatory processes?

On January 17, 2025, the Office of the Federal Register published a notice in the Federal Register correcting a reference in the proposed interpretive rule document that was published on January 17, 2025.

The Office of Nuclear Material Safety and Safeguards oversees the management and security of nuclear materials to ensure public health and safety.


Question: What is the Emergency Feederwater Initiation and Control system?

Two manual initiation switches per actuation train for each of the following EFIC Functions shall be OPERABLE: - a. Steam generator (SG) A Main Steam Line Isolation; - b. SG B Main Steam Line Isolation; and - c. Emergency Feedwater (EFW) Initiation.

The completion time for required actions is determined by the associated condition set forth in the table.


Question: temporary canal level probe

The proposed license amendment request (LAR) would establish the requirements for, and the use of a temporary canal level probe and temporary supply line (jumper) to provide service water (SW) to the component cooling heat exchangers (CCHXs) and facilitate planned maintenance activities.

The staff from the Nuclear Regulatory Commission's (NRC's) Office of Nuclear Reactor Regulation (NRR) has initiated its review of the LAR in accordance with NRR Office Instruction LIC-101, "License Amendment Review Procedures" (ML19248C539).


Question: What are the main safety systems in a nuclear reactor?

The Emergency Core Cooling System (ECCS) is designed to provide cooling to the reactor core in the event of a loss-of-coolant accident, ensuring the core does not overheat and mitigates the risk of a meltdown.

The reactor containment structure is built to prevent the release of radioactive materials into the environment under normal and accident conditions.


Question: What are the key components of an emergency plan for nuclear facilities?

The revised emergency plan implementing procedure outlines the roles and responsibilities of facility personnel during an emergency event, ensuring a coordinated response that complies with regulatory requirements.

The reactor's coolant system is critical for maintaining the core temperature within safe limits during normal operation and in accident scenarios.


Question: nuclear regulatory control

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: The NRC requires licensees to assess whether loss of material from aluminum alloys has occurred due to exposure to various environmental conditions.

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

The Buried and Underground Piping and Tanks program is described in Appendix B.


Question: What is the function of the Emergency High Pressure Makeup System?

The Emergency High Pressure Makeup System is designed to provide a pressure boundary in the event of loss of coolant accidents.

The external surfaces monitoring program evaluates the condition of mechanical components exposed to an uncontrolled indoor air environment.


Question: What are the qualification requirements for structural overlaid wrought austenitic piping welds?

*Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds* outlines the necessary criteria for ensuring the reliability and safety of welded joints in nuclear systems.

The ASME Boiler and Pressure Vessel Code, Section III, provides rules for the construction of nuclear power plant components, detailing safety and quality standards.


Question: What is the minimum detectable activity in nuclear sampling?

MDA: Minimum Detectable Activity refers to the lowest quantity of radioactive material that can be reliably detected by a particular measurement method.

The preparation of the sample involves several steps, including drying, grinding, and homogenization to ensure a uniform distribution of the material.


Question: What are the requirements for monitoring mercury in NPDES permits?

BVPS requests removal of mercury monitoring requirement.

Currently there is not continuous temperature monitoring installed at this location.


Question: The analysis of materials and environments in relation to aging management programs is crucial for ensuring safety and compliance.

The review of 10-years of plant specific operating experience has not identified cracking in copper alloy with 15% zinc or more in air environments due to potential exposure to ammonia or ammonium compounds.

Component is different, but consistent with NUREG-2191 item for material, environment, and aging effect. AMP takes some exceptions to NUREG-2191 AMP.


Question: What is the purpose of a Regulatory Audit Plan in nuclear licensing?

The Regulatory Audit Plan outlines the methodology and procedures that will be used to conduct a regulatory audit associated with the licensing of nuclear reactors, ensuring compliance with safety standards and regulations.

The enclosure contains various administrative details and contact information for the project manager, which do not pertain to the technical aspects of nuclear licensing.


Question: The activity level measured for nuclides is expressed in picocuries per gram, indicating the rate of radioactive decay.

The unit of measurement for the activity of a nuclide is defined in terms of picocuries per gram (pCi/grams), which quantifies the amount of radioactivity present.

The yield of each nuclide is expressed as a percentage, reflecting the efficiency of its production during a nuclear reaction.


Question: Nickel-63 radioactive decay

Nickel-63 is analyzed in the report with a result of -0.325 pCi/g, indicating its presence and concentration level in the sample tested.

The method used for soil preparation was GL-RAD-A-021, which involves techniques to ensure accurate analysis of radiological samples.


Question: The NRC mandates that licensees demonstrate compliance with specific guidelines related to system design and protective actions.

The following clauses were not evaluated because addressing compliance with this guidance is a plant-specific activity that depends on the system design: Clause 5.2, Completion of Protective Action, Clause 5.8, Information Displays, Clause 5.11, Identification, Clause 5.12, Auxiliary Features, Clause 5.13, Multi-Unit Stations, Clause 5.14, Human Factors Considerations. [SER]

Addressed in PSAI 7.11 and satisfied by: Licensing Technical Report.


Question: What analytical techniques are used for evaluating steam generator load in nuclear plants?

The basic analytical technique utilized dynamic results from Westinghouse.

A detailed structural model (Fig. 2) was used to determine load and stresses in the structure.


Question: What is the Nickel Alloy Management Program in nuclear licensing?

The Nickel Alloy Management Program for Nonreactor Vessel Closure Head Penetration Locations is a plant-specific program that manages cracking due to primary water stress corrosion cracking (PWSCC) for nonreactor vessel head nickel alloy component locations.

SNC will submit a program inspection plan for VEGP that includes implementation of applicable NRC bulletins, generic letters, and staff-accepted industry guidance.


Question: What are the potential impacts of the Monticello Nuclear Generating Plant on state-listed species?

Based on the project details provided with the request, the following rare features may be impacted by the proposed project: Black sandshell (Ligumia recta), a state-listed mussel species of special concern, has been documented in the Mississippi River in the vicinity of the proposed project. Mussels may be negatively affected by riverbed disturbance, changes in water flow, and deterioration in water quality including sedimentation/siltation, nutrient loading, and possibly temperature alterations.

We recommend any activities that will impact this structure be done outside of the nesting season (April through July). If the birds exhibit unusual behaviors or other signs of potential distress during operation or maintenance, especially during the nesting season, please contact Erica Hoaglund, (651-259-5772 or erica.hoaglund@state.mn.us).


Question: What is the Atomic Energy Act?

| AEA           | Atomic Energy Act                          |

| RCRA          | Resource Conservation and Recovery Act      |


Question: What are the potential impacts of water use conflicts with terrestrial resources in nuclear plants?

Water use conflicts with terrestrial resources in riparian communities could occur when water that supports these resources is diminished either because of decreased availability due to droughts; increased water demand for agricultural, municipal, or industrial usage; or a combination of such factors.

Continued operations and refurbishment activities could continue to affect onsite terrestrial resources during the license renewal term at all operating nuclear power plants.


Question: The two-dimensional groundwater model for VEGP simulates how water moves through the aquifer post-construction.

A two-dimensional single layer groundwater model was developed to simulate post-construction groundwater flow in the Water Table aquifer at the VEGP site (Appendix 2.4B).

Horizontal hydraulic conductivity values were varied across the model domain based on observed hydrogeologic conditions in order to calibrate the model to recently observed Water Table aquifer groundwater levels.


Question: The NRC assesses the physical protection measures of nuclear facilities to ensure compliance with regulations and enhance public safety.

One hour has been allocated within the resource estimate of this inspection procedure for the inspector(s) to conduct physical protection program status verifications. The purpose of the status verification is to ensure that the implementation of the licensee's physical protection program is maintained in accordance with regulations, license security plans and implementing.

Information that reflects characteristics and attributes of the DBT of radiological sabotage.


Question: groundwater chemistry of aquifers

The summary of the general groundwater chemistry of the Glasford (Illinoian) Sand and Gravel Aquifers includes various constituents, their number of samples, and statistical information such as minimum, maximum, mean, median, and standard deviation.

The table illustrates the quality of groundwater in Illinoian versus Kansan Aquifers, detailing parameters like iron, chloride, sulfate, and hardness measured in mg/L.


Question: What is the engineering behavior of saprolite in relation to nuclear safety evaluations?

The saprolite at the ESP site is comprised of micaceous silty, clayey sand and sandy silt/clay with occasional-to-many relict rock fragments. Depending on the degree of weathering, the saprolite more or less retains the fabric or structure of the parent bedrock.

Evaluation of engineering geology conditions at the ESP site has been performed based on a review of existing site-specific reports, geologic and geotechnical investigations, and geologic literature.


Question: The NRC has determined that the submitted application fulfills all necessary regulatory criteria and can proceed without compromising safety.

The application for amendment by Indiana Michigan Power Company dated April 30, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations set forth in 10 CFR Chapter I;

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;


Question: aging management program

This example provides objective evidence that the existing aging management program activities are being effectively implemented to identify and manage aging effects of in-scope components serviced by raw water.

During a program effectiveness review, it was identified that the first inspections after the period of extended operation had been scheduled to be performed later than required by the program.


Question: What measurements are used to evaluate energy peaks in keV?

The table lists various energy peaks along with their corresponding values in keV, encompassing parameters such as Peak Number, Energy, FWHM, and Net Peak Area.

The nuclear reactor operates continuously within established safety margins and is monitored by several automated systems to ensure compliance with operational standards.


Question: Reactor Coolant System pressure boundary valves

All valves in the Reactor Coolant System which are in contact with the coolant are constructed primarily of stainless steel. Other materials in contact with the coolant, such as for hard surfacing and packing, are special materials.

There are no required tests or inspections performed on the Pressurized Relief Tank.


Question: The categorization of components or functions as HSS or LSS must follow the process outlined in NEI 00-04.

Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201.

The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited.


Question: Monticello Nuclear Generating Plant vicinity

The Monticello Nuclear Generating Plant (MNGP) is situated within a 6-mile radius that includes notable sites such as Oak Savanna Park and Lake Maria State Park, relevant for assessing impact zones during regulatory reviews.

The Mississippi River State Wild and Scenic Recreational District offers various outdoor activities, including hiking and fishing, which attract many visitors each year.


Question: What is the purpose of heat tracing in nuclear facilities?

Electrical heat tracing is required on the recycle evaporator even though redundant electrical heating units are provided in the room in which it is located. Electrical heat tracing is provided on the concentrates piping which leaves the room to insure that the boric acid does not precipitate.

A summary of principal component data is given in Table 9-34 and the code requirements are given in Section 3.2.


Question: The termination of an NRC license involves reviewing documentation related to the controlled substances utilized by the licensee.

In response to your email below, we reviewed your docket file and found the documents listed below that are germane to this licensing action.

Once we get rid of the source, we are going to do the formal reports and documentation.


Question: VCS site location and description

The VCS site is approximately 13.3 miles south of the city of Victoria and located in the southern part of Victoria County, Texas.

Between 5 and 10 miles from the VCS site, there are six industrial facilities that are significant enough to be considered in the hazards screening presented in Section 2.2.


Question: containment spray system

NS13      | CONTAINMENT SPRAY IN                 | M387       | INSIDE         | CHECK      | 1                      | A          | N        |

WL-2      | RCDT HEAT EXCHANGER DISCHARGE        | M375       | INSIDE         | DIAPHRAGM | NORMAL                 | 2          | C        |


Question: What are the requirements for the physical separation of redundant safety-related systems in nuclear plants?

Minimum Separation Requirements for Redundant Transmitter Impulse Lines: The following precautions must be taken in routing the instrument impulse lines for the reactor protection system in order to meet the IEEE single failure criteria. It is necessary that such redundant transmitter impulse lines be physically separated at a minimum distance of eighteen inches (18") in any direction.

The basic objective of impulse line separation is negation of damage to more than one redundant protection or safeguard transmitter impulse line as the result of any one incident.


Question: What is the role of the Senior Resident Inspector in nuclear regulatory processes?

The Senior Resident Inspector is responsible for overseeing the operations and safety practices at the facility, ensuring compliance with NRC regulations and standards.

The General Manager - Performance Improvement and Oversight focuses on enhancing the overall performance of the plant through various strategic initiatives.


Question: What are the activity levels of different nuclides in nuclear safety assessments?

The activity (pCi/grams) of CE-144 is 3.27E+00 and that of EU-155 is 4.66E-02, showcasing the varying levels of radioactivity present in nuclides used for regulatory assessments.

The NRC mandates that all facilities operate within the defined regulatory framework to ensure compliance with safety standards and environmental protection.


Question: The NRC oversees the licensing process for the Turkey Point nuclear facility, ensuring adherence to regulatory standards.

The licensing process for nuclear reactors involves a series of regulatory filings that demonstrate compliance with safety standards and operational protocols.

The discussion regarding the environmental impact assessments is crucial for understanding the broader implications of energy production.


Question: nuclear safety systems

Nuclear safety systems include distinct components, such as reactor core cooling systems and containment structures, designed to prevent the release of radioactive materials in the event of an accident.

The licensing process for nuclear facilities requires extensive documentation and review to ensure all operational protocols are met.


Question: The safety analysis demonstrates how certain nuclides contribute to the overall activity measurements.

?  = nuclide is part of an undetermined solution


Question: Strontium-90 in soil analysis

The results for the parameter Rad Gas Flow Proportional Counting indicate a presence of Strontium-90 at a concentration of 0.0706 pCi/g in the soil sample.

The sample collection was carried out on September 20, 2019, and it was received by the lab on October 21, 2019.


Question: Strontium-90 in nuclear licensing

Rad Gas Flow Proportional Counting | GFPC, Sr90, Solid "Dry Weight Corrected" | Strontium-90 | U | 0.024 | +/-0.0648 | 0.117 | 0.400 | pCi/g | JXC9 | 10/07/19 | 1510 | 1921875

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


Question: quality assurance program in nuclear regulation

The Quality Assurance Program for a nuclear facility is designed to ensure that activities affecting safety are performed in a consistent and rigorous manner, adhering to the regulatory requirements set forth by the NRC.

The reactor design must incorporate systems that can handle extreme temperature variations without failure, ensuring operational efficiency and safety.


Question: operator licensing examination standards

The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 12.

The NRC examiners determined that eleven of the fourteen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.


Question: What mechanisms are involved in control rod reactivity control?

Control rods gravity insertion involves a passive mechanism designed specifically for control rod reactivity control.

For case where turbine trip occurs, a reactor trip signal on turbine trip is anticipated (for power to P8).


Question: minimum cool time for BWR fuel

Sample calculation for the 09b fuel type at 45 GWd/MTU. Cask surface (normal), 1m (accident), and axial (normal and accident) dose rates are significantly below limits at all cool-time meeting assembly heat load limits and are therefore not listed.

In the case of the reactor core design, the dimensions play a crucial role in the overall thermal efficiency and safety performance.


Question: What is the unit of measurement for radioisotope activity?

The unit of measurement for radioisotope activity is mCi, which stands for millicurie.

The manufacturer name for the device listed is TN TECHNOLOGIES, INC.


Question: The NRC mandates precise measurements for detecting environmental contaminants such as Strontium-90 in soil samples.

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


Question: What is an Emergency Action Level (EAL) scheme in nuclear safety?

The NRC staff verified that the numbering, sequencing, formatting, logical progression, and instrumentation and setpoints for this EAL are consistent with the overall EAL scheme development guidance and address the site-specific implementation strategies provided and are, therefore, consistent with a standard EAL scheme, as required by 10 CFR 50.47(b)(4).

The NRC staff determined that the proposed EAL scheme uses objective and observable values, is worded in a manner that addresses human factors engineering and user-friendliness concerns, follows logical progressions for escalating events, and allows for event downgrading and upgrading based upon the potential risk to the public health and safety.


Question: nuclear safety protocols

The regulation of nuclear reactors encompasses a variety of safety protocols to ensure the safe operation and environmental protection standards are maintained during reactor operation and maintenance.

I would just delete that second sub-bullet, I think the first bullet covers the thought we were trying to convey…


Question: How does the NRC describe the need for definitions related to PRA maintenance?

The NRC has developed a glossary of terms in this RG, which includes definitions of some terms that were derived from PWROG-19027-NP, Revision 2, and that were previously defined in ASME/ANS RA-Sa-2009.

The staff neither approves nor disapproves of information contained in the referenced document.


Question: In modeling the dam breach event, the removal of bridges was necessary to allow the flood wave to progress without obstruction at the VEGP site.

Thus, to provide more reasonable results, which allow the flood wave to progress unimpeded downstream (a conservative assumption for modeling the flood elevations at the VEGP site), the bridge structures were removed from the HEC-RAS model.

The normal depth option in HEC-RAS was used to determine the starting water surface elevation at this location.


Question: nuclear regulatory filings

Nuclear regulatory filings must include detailed descriptions of the proposed changes and their potential impact on safety, as outlined in the regulatory guidelines.

The reactor coolant system is responsible for transferring heat away from the reactor core to prevent overheating during operation.


Question: What are the fuel specifications for MAGNASTOR storage?

MAGNASTOR is designed to provide passive dry storage of canistered PWR and BWR fuel. The system requires few operating controls. The principal controls and limits for MAGNASTOR are satisfied by the selection of fuel for storage that meets the Approved Contents presented in this section and in the tables for MAGNASTOR design basis spent fuels.

If any Fuel Specification or Loading Condition of this section is violated, the following actions shall be completed: - The affected fuel assemblies shall be placed in a safe condition. - Within 24 hours, notify the NRC Operations Center. - Within 60 days, submit a special report that describes the cause of the violation and actions taken to restore or demonstrate compliance and prevent reoccurrence.


Question: What is the classification of the wells in this submittal?

Classification  of wells refers to the differentiation based on their characteristics, such as their depth and screen interval, as shown in the well data table.

The ground surface elevation for well W-43 is 138.09 feet, which contributes to understanding the depth and structural integrity of the casing.


Question: What is in-vessel event methodology in nuclear reactor design?

The design basis accident methodology for in-vessel events without radiological release outlines the necessary analyses to be performed on potential events that could occur within the reactor vessel, focusing on the assurance of safety without considering the release of radioactive materials.

The NRC establishes guidelines for conducting routine inspections of facility operations to ensure compliance with safety standards and operational protocols.


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

Gamma measurements provide critical information about radiation levels in the environment, assessed in counts per minute (cpm) to determine the presence and concentration of radioactive materials.

The sampling location was identified as Pickard Hall, and boring number BRA-05 was recorded for the collection of data.


Question: What is the significance of the McGuire Nuclear Station in nuclear licensing?

The McGuire Nuclear Station is an important facility within the nuclear fleet and serves as a reference point for numerous regulatory discussions and safety analyses.

Table 4-3 outlines various updates and changes made in the regulatory framework from 1992, including modifications to operational protocols.


Question: What is the Commission's Policy Statement on enhancing public participation in NRC meetings?

The Commission's Policy Statement on "Enhancing Public Participation in NRC Meetings" aims to encourage and facilitate involvement by the public in the decision-making process of the NRC, ensuring that various stakeholders have the opportunity to express their views and participate effectively.

The ADAMS Accession Number for this document is ML25034A140, which is used for tracking and referencing documents within the NRC's system.


Question: The NRC monitors the levels of specific nuclides and their respective activities in regulatory evaluations.

K-40 has an energy of 1460.81 keV and a yield of 10.67% with a line Minimum Detectable Activity (MDA) of 7.736*003 pCi/M².

The average temperature inside a reactor is crucial for determining the efficiency of the cooling system.


Question: The NRC evaluates components associated with the radiation monitoring system to ensure compliance with safety protocols.

Radiation Monitoring System Components Subject to Aging Management Review

Sanitary Sewage System Components Subject to Aging Management Review


Question: The analysis of mass flux distribution in the core is crucial for assessing safety during large break Loss of Coolant Accidents (LOCAs).

Mass Flux (lbsm/s-ft²) represents the flow of coolant through the reactor core, which is vital for determining the heat transfer capability during accidents.

The table shows the time in seconds for different measurements, including 0, 100, 200, and up to 900 seconds.


Question: What are DFC designated locations in relation to nuclear safety?

DFC designated locations may contain a loaded DFC or a PWR UNDAMAGED SNF ASSEMBLY.

The schematic outlines various components and their placements within the reactor system.


Question: What are the considerations regarding the exclusion of embrittlement effects in regulatory discussions?

The NRC requires that licensees address embrittlement effects, specifically with respect to ΔT < 25°F, in their regulatory filings.

The materials exchange meeting is scheduled for May 22-24, focusing on various industry topics.


Question: The testing procedure includes applying a 1 MHz oscillatory wave with a peak voltage of 2.5kV for validating the relay functionality.

1 MHz oscillatory wave, 2.5kV peak open circuit, 50 and 400 Hz repetition rates, applied for 2 seconds each test, between terminals as shown.

No breakdown, no leakage, relay undamaged and operating properly after test.


Question: The NRC mandates that licensees report transactions concerning nationally tracked sources, detailing all necessary information involved in the process.

Each licensee who manufactures, transfers, receives, disassembles, or disposes of a nationally tracked source shall complete and submit a National Source Tracking Transaction Report as specified in He-P 4021.22(a) through (e) for each type of transaction.

Measurements of fixed contaminant shall not be averaged over more than 1.0 square meter.


Question: carbon filter capabilities

appropriate test the following items: carbon filter capabilities for removal of molecular iodine - 131 and methyl iodide carbon filter iodine collection capability, carbon filter cell leak-tightness integrity, carbon filter flow resistance.

A flow totalizer is also provided for recording the quantity of air released.


Question: Risk-Informed Licensing Amendment Request

The DTE Energy Enrico Fermi Unit 2 Nuclear Generating Station’s Risk-Informed Licensing Amendment Request for closure of debris-related suppression pool strainer performance issues.

This appendix provides a description of how 10 CFR 50 Appendix-B quality assurance requirements, which are specified in DTE Energy Fermi-2 P-20 4701602416, are implemented in the development of this report.


Question: safety-related systems in nuclear licensing

Safety-related systems are those systems that are necessary for the safe operation of a nuclear facility and are designed to mitigate the consequences of postulated accidents.

The new data FGR shows significant advancements in the algorithms used for analyzing reactor performance under varying conditions, enhancing operational efficiency.


Question: stress corrosion cracking in aluminum alloys

SCC is a form of environmentally assisted cracking which is known to occur in high and moderate strength aluminum alloys. The three conditions necessary for SCC to occur in a component are a sustained tensile stress, aggressive environment, and material with a susceptible microstructure.

Deficiencies will be documented in accordance with 10 CFR Part 50, Appendix B Corrective Action Program. The Fire Water System program is described in Appendix B.


Question: nuclear safety licensing process

The nuclear safety licensing process involves a series of steps to ensure that a new reactor design meets established safety standards, which includes comprehensive safety analysis reports and compliance documentation submitted to regulatory bodies.

During the construction of the facility, a variety of materials were chosen based on their availability and cost, including steel and concrete that meet industry standards.


Question: What is an amendment for a nuclear license?

Please find attached the Cover Letter for Amendment to License No. 13-3271901, Michiana Hematology Oncology, P.C. and the Model Delegation of Authority.

Good morning, Martha 😊 Can you please add the attached to ADAMS?


Question: The NRC requires an assessment of the recovery of surrogate tracers such as Americium and Plutonium during liquid scintillation analysis.


Question: What is a non-cited violation in nuclear inspections?

We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Calvert Cliffs Nuclear Power Plant, Units 1 and 2.


Question: What materials are used in MOX fuel for nuclear reactors?

The MOX fuel and Zircaloy-2 cladding alloy are no longer part of the scope in Revision 1 of the methodology document (Reference 2).

The request for additional information on Figures 5-8, 5-11, 5-14 and 5-17 in Revision 0 of the methodology document (Reference 1) is partially obsolete.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves several key steps, including the submission of safety evaluation reports and thorough assessments by regulatory bodies to ensure public safety and environmental protection.

In order to ensure that all operations are conducted safely, personnel undergo rigorous training and certification processes that focus on emergency preparedness and response.


Question: What is the maximum permissible concentration of isotopes in nuclear safety assessments?

Maximum Permissible Concentration (MPC) (c) (uCi/cm³) is specified for isotopes to ensure regulatory compliance and safety.

The total annual release from one unit of Np-239 is calculated to be 2.15E-08 Ci/year.


Question: How do the NRC regulations dictate the management of Ga-68 radiopharmaceuticals?

Accordingly, the requirements of 10 CFR 30.35(e) for developing, funding, updating, and submitting detailed cost estimates for NRC review are unnecessary for the decommissioning of sites licensed for the use of Ge-68/Ga-68 generators.

The NRC staff has determined that these conditions will be sufficient until a more permanent regulatory solution is reached through rulemaking.


Question: What are the licenses required for operating nuclear reactors?

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


Question: The NRC maintains oversight of the operational status of all licensed wells to ensure safety and compliance with regulations.

Wells can be classified as either active or temporarily abandoned, depending on their operational status and regulatory compliance.

The well locations are specifically categorized based on their positioning relative to various landmarks and sites.


Question: What are time-limited aging analyses (TL AAs) in nuclear licensing?

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TL AAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

However, LRA Section 4.3.5 does not clearly describe how the 60-year cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, USAR information and specific system-level knowledge).


Question: loss of material due to corrosion in nuclear reactors

The aging effect of loss of material due to general corrosion does not apply to the carbon steel external surfaces of reactor vessel, nozzle, and safe end components exposed to air - indoor uncontrolled in the Reactor Vessel.

During power operation the insulated reactor vessel, nozzles, and safe end components have an external temperature greater than 212°F and are at a higher temperature than the air - indoor uncontrolled environment.


Question: What is the process for renewing a nuclear materials license?

If you wish to continue your licensed program, you should prepare and submit a renewal application on NRC Form 313 (enclosed), following regulations in 10 CFR and licensing guidance NUREG 1556, “Consolidated Guidance About Materials Licenses.”

You must submit an application for the renewal of your license at least 30 days before the expiration date on the license.


Question: land use impacts in regulatory filings

Depending on the route chosen, land use in the ROW corridor may change from agricultural, forest, wetland, or grassland to industrial use.

Visual impacts during natural gas power plant operations would be similar to those experienced during Oconee Station operation.


Question: The calculations for the flow rates to the environment from the sprayed containment volume are critical in assessing safety measures.

Pathway number 3: Sprayed Containment Volume to Environment includes convection data which impacts the flow rate.

Filtered Inactive Environment to Control Room pathway involves different filter efficiencies but does not relate to containment volumes.


Question: The NRC mandates that licensees ensure the integrity of spent resin storage systems when handling radioactive materials.


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

Minimum Cooling Time (years) for assemblies with an average burnup between 40 and 41 GWd/MTU is specified in loading tables, which must be adhered to for safe operation.

The enrichment of the fuel used in a reactor directly affects the efficiency and performance of the nuclear reactions taking place within it.


Question: Reactor trip events

The inspectors evaluated the following licensee event reports (LERs): 1. LER 05000328/2024-001-00, Unit 2 Reactor Trip due to an Electrical Trouble Turbine Trip (ADAMS Accession No. ML24269A161)  The inspectors did not identify a performance deficiency or violation of NRC requirements.

The inspectors verified no proprietary information was retained or documented in this report.


Question: isotope activity measurement

Distributor License Number: 34-00639-03G


Question: steam generator aging management

Steam Generators (B.2.3.10) and Water Chemistry (B.2.3.2)


Question: What is an RCS soak and its purpose in reactor operations?

The RCS soak is designed to reduce the temperature stresses present in the Reactor Vessel.

The crew is preparing to perform an RCS soak in accordance with 2-EOP-FRTS-1 (RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS).


Question: The safety analysis report must include a detailed evaluation of the plant systems relevant to nuclear safety.

A safety analysis report (SAR) is required to address all potential hazards and assess the adequacy of safety systems at the facility.

In the development of new software, it is critical to follow a structured methodology to ensure consistent and correct implementation.


Question: License applicants must verify compliance with relevant nuclear safety standards and regulations during the application process.

The process of licensing for a nuclear facility includes a comprehensive evaluation of its safety systems and compliance with specific regulatory requirements.

The design of the reactor core is critical to maintaining efficient heat transfer and ensuring optimal fuel utilization.


Question: The NRC stipulates that it is crucial to keep the temperature of the torus below the Heat Capacity Temperature Limit during incident scenarios to ensure primary containment integrity.

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise: - Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, or - Suppression chamber pressure above the Primary Containment Pressure Limit, before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent. It is vital to maintain Torus temperature below this threshold during emergencies to preserve primary containment and continue safe operation of the plant.

It is important to understand the design specifications of the Drywell Spray Initiation Limit curve, which is used to manage temperature and pressure effectively in the containment during various operational conditions.


Question: The NRC calculates the weighted mean activity of each nuclide in a sample.

The table provides detailed information about the energy and yield of various nuclides.


Question: The NRC mandates that facilities conduct safety assessments pertaining to radiological impacts.

NRC regulatory guidelines require that the licensing process includes a thorough evaluation of the reactor's safety systems and their efficacy in preventing accidents.

The operators of the plant reported a significant increase in maintenance activities over the past year, reflecting the commitment to operational excellence.


Question: The NRC is currently reviewing the information related to Licensing and Certification that was submitted on 5/4/23 for further steps.

During our meeting yesterday, we had a path forward for addressing L&C 18 and 19.

Alan Meginnis is the Manager of Product Licensing at Framatome and can be reached at his office or mobile for any inquiries.


Question: What is the significance of P-M diameter change in nuclear rods?

P-M diameter change versus ramp terminal power for ramped rods is reported in Figure 31-3.

The rods are identified by burnup ranges of 15 to 24, 25 to 34, 35 to 44, 45 to 54, and greater than 55 GWd/mtU.


Question: energy peak analysis in nuclear regulatory filings

The background subtracted area is calculated by taking the original area and subtracting the ambient background counts to obtain the net counts.


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

Energy (keV) is a crucial measurement in nuclear safety assessments, as it helps determine the identify of radioisotopes present in a sample based on their characteristic peaks in the spectrum.

Background noise in a reactor can significantly affect the accuracy of measurements but is often measured in terms of counts per minute rather than energy levels.


Question: Groundwater levels are critical for assessing the environmental impact of a nuclear facility, particularly in relation to drilling formations.

Drilling Limestone Formation is an important aspect of geological assessments, but it does not directly relate to environmental monitoring protocols.


Question: critical heat flux correlation

COBRA-FLX is a new methodology developed to evaluate the critical heat flux in a nuclear reactor, which is essential for ensuring safe operation under various thermal conditions.

The importance of maintaining a comprehensive safety culture within all levels of the organization is emphasized in the report, ensuring that everyone is engaged in safety practices.


Question: peak area measurement in nuclear safety

The net peak area is a critical measurement used to analyze radiation detection data, providing insights into the intensity of specific energy peaks.

The ambient background can affect the measurements, necessitating careful calibration and consideration in data collection.


Question: screening for SRO-only questions in nuclear regulations

Does the question require one or more of the following: assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed; knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps; knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures; knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures?

Can the question be answered solely by knowing immediate operator actions?


Question: safety injection systems

SAFETY INJECTION SAMPLE is crucial for ensuring the reactor core remains adequately cooled during an emergency by injecting coolant into the core.

PRESSURIZER SAMPLE refers to a section of the system that manages the pressure within the reactor vessel, contributing to the overall thermal hydraulic stability.


Question: standby gas treatment system impacts

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

The plant is operating at 100% power with the following: - Emergency Diesel Generator (EDG) 102 is out of service for corrective maintenance. - A small steam leak develops in the Reactor Building.


Question: What are the requirements for maintaining secondary containment in nuclear facilities?

Maintaining secondary containment OPERABILITY requires verifying one door in each access opening to secondary containment zones is closed.

When the railroad bay door (No. 101) is closed; all Zone I and II hatches, removable walls, dampers, and one door in each access opening connected to the railroad access bay are closed; or, only Zone I removable walls and/or doors are open to the railroad access shaft.


Question: Uncontrolled RCCA bank withdrawal and its impact analysis

Uncontrolled RCCA Bank Withdrawal from Full Power with Minimum Reactivity Feedback (80 pcm/s Withdrawal Rate)

The core water average temperature ranges from 550 to 620 °F, which is critical for maintaining operational efficiency in a nuclear reactor.


Question: What are the requirements for safety-related systems in nuclear facilities?

The safety-related systems must be designed to satisfy the requirements of applicable regulatory standards to ensure the highest level of safety.

The effective date for the updates discussed in the regulatory filings is 14 October 2000.


Question: The data shows various well numbers alongside their respective elevations measured in feet.

WATER LEVEL MEASUREMENTS are recorded to provide critical data regarding the elevation of various wells.

The information represents a statistical overview concerning the maintenance schedules and operational requirements for nuclear facilities.


Question: nuclear regulatory filings

The licensing process for nuclear facilities is strictly governed by regulations set forth by the Nuclear Regulatory Commission, which ensures that safety standards are met.

The reactor's cooling system is designed to efficiently transfer heat away from the core, preventing overheating during operation.


Question: Reactor Coolant Pressure Boundary System

Reactor Coolant Pressure Boundary System is in scope for License Renewal.

Fire Protection System is essential for ensuring safety during potential accident scenarios.


Question: What is the NRC Form 665s used for in nuclear regulatory documentation?

The NRC Form 665s is used for reporting information related to the decommissioning of nuclear facilities and ensures compliance with regulatory requirements.


Question: The analysis of mass flux and its implications for safety assessments are critical in determining reactor performance.

Fluid mass velocity at PCT and burst elevations (Cₒ = 0.6, low Tₐᵥg, max Sᵢ, cosine power shape, non-IFBA) is a key parameter in the operational assessment of reactor systems.

The time intervals measured in seconds provide critical data points for the evaluation of reactor processes.


Question: What is the purpose of an audit plan in nuclear licensing?

The audit plan outlines the objectives, scope, methodology, and procedures to evaluate the performance and compliance of the licensee with regulatory requirements.

The Senior Project Manager oversees all reactor licensing activities and liaises with the Division of Operating Reactor Licensing.


Question: plutonium-239 analysis

Plutonium-239 is measured as part of the radiochemical analysis for solid samples, providing valuable information for regulatory compliance.

The LaCrosseSolutions facility is located in a region with diverse ecological features that contribute to its unique environmental profile.


Question: The Earth Berm serves to safeguard important safety systems and feature prominently in the structural layout of the facility.

The Earth Berm is a plant feature common to all three units and is intended to provide physical support, shelter, and protection for safety-related SSCs. 10CFR 54.4(a)(1)

The aging management review results for these components are provided in Table 3.5.2-30, Earth Berm - Summary of Aging Management Evaluation.


Question: What is the unit of measurement used for radioactivity in NRC registrations?

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


Question: What does net peak area represent in nuclear measurements?

Net Peak Area is defined as the area of the peak after the background has been subtracted, which is essential for accurate quantification of radioactive isotopes.

The FWHM (Full Width at Half Maximum) is a measure of the width of the peak, expressed in keV, indicating the resolution of the measurement device.


Question: What are the requirements for safety-related piping systems in nuclear facilities?

Piping Systems Relied on for Safety 05/18/05 (05-013) defines the critical importance of ensuring that all piping systems integral to safety are designed, constructed, and maintained according to stringent regulatory guidelines.

The Resident Inspection Program for On-Site Construction Activities at the Mixed-Oxide Fuel Fabrication Facility focuses on the oversight of construction quality and compliance with project specifications, without directly addressing safety-related systems.


Question: During a steam line break, the reactor's coolant temperature and pressure are initially assumed to be at nominal values representative of full power operation.

The DNB analysis is performed using the revised thermal design procedure. Initial core power, and reactor coolant temperature and pressure are assumed to be at their nominal values consistent with steady-state full power operation. Uncertainties in initial conditions are included in the DNBR limit as described in WCAP-11397.

A conservatively large positive moderator density coefficient of 0.5 Δk/ρcm³ (corresponding to a large negative moderator temperature coefficient) and a most-negative Doppler-only power coefficient form the basis for the end-of-life (EOL) maximum reactivity feedback assumption.


Question: The NRC mandates strict procedures for the operation of gas disposal systems to ensure safety and compliance.

The Waste Gas Disposal System is crucial for managing radioactivity release, specifically concerning the safe handling and disposal of radioactive gases.

The training program includes various performance measures designed to enhance operator competency in all phases of reactor operation.


Question: The NRC mandates that aging management programs for monitoring components must include effective inspections and testing to detect any age-related degradation.

The purpose of this section was to verify that the existing aging management program activities, to identify and correct, as warranted, age-related degradation of components monitored by the Compressed Air Program, are being effectively implemented in the initial license renewal period of extended operation.

Control air sampling data has not shown any issues with the sampling process. Occasionally moisture was found in sample lines and removed.


Question: The NRC evaluates the safety of the Palisades power plant based on the latest operational data and regulatory standards.

The Palisades plant seeks reauthorization for power operations, which involves thorough assessments of its adherence to safety regulations.

During the recent meeting, stakeholders discussed various community concerns unrelated to the current licensing process.


Question: The NRC establishes technical requirements for safety systems and components, including performance standards that must be met during operation.

The RHR OUT FROM SUMP system is a safety system designed to provide cooling and maintain cooling water for the reactor in the event of a loss of coolant accident.

The PALS DISCHARGE line is crucial for the safe management of process water used during routine plant operations.


Question: What are the characteristics of Tristructural Isotropic (TRISO) coated particle fuel?

Tristructural Isotropic (TRISO) coated particle fuel is designed with a unique multi-layer coating around uranium-containing fuel particles, which enhances its performance and safety in high-temperature environments.

Liquid metal coolants such as sodium, lead, and lead-bismuth eutectic are utilized in fast reactors to efficiently transfer heat from the reactor core.


Question: gaseous swelling model

The fraction of porosity active in the gaseous swelling model can be distinguished from the remaining initial porosity and rim porosity. The results show that the code reasonably predicts the experimental data.

Figure 4-24 shows the overall comparison of calculated and measured porosity from the different programs available in the validation database.


Question: The system comprises two trains, each able to individually satisfy the design requirements for heat removal and iodine elimination.

The CSS, in conjunction with the containment fan cooler system and the emergency core cooling system, is designed to be capable of removing sufficient heat and subsequent decay heat from the containment atmosphere following the hypothesized LOCA or MSLB to maintain the containment pressure below the containment design pressure.

The containment spray water does not contain substances which would be unstable in the thermal or radiolytic environment of the LOCA or cause extensive corrosive attack on equipment.


Question: control room design criteria

This rulemaking would amend the control room design criteria from the current 5 rem (0.05 Sv) total effective dose equivalent (TEDE) to a revised value of 10 rem (0.10 Sv) TEDE; the value may range up to 25 rem (0.25 Sv) TEDE with consideration of the plant-specific risk profile or risk information.

The amended rule, and subsequent guidance, would align with Commission direction provided in SRM-SECY-98-144 to take a risk-informed, performance-based approach to regulations and guidance.


Question: The NRC stipulates specific criteria for evaluating soil types and their properties during site assessments.

The soil types evaluated in this report include SAND (SW) gray, fine to medium grained, which is further classified based on its density and consistency.

The drilling process had an unexpected refusal at a depth of 32 feet, which indicates a potential challenge in soil sampling.


Question: maximum reactor core power level

Each unit was originally licensed to operate at steady state reactor core power levels not in excess of 2527 megawatts thermal (MwT). An extended power uprate project was undertaken for both units, resulting in maximum core power levels of 2957 MwT being approved by the NRC in 2001.

The Dresden Nuclear Power Station site consists of approximately 2,459 acres along the Illinois, Des Plaines and Kankakee Rivers. The station, associated infrastructure, cooling lake, and most of the site boundary are in Grundy County, Illinois.


Question: The Institute of Nuclear Power Operations OE program involves reporting on operational experiences that are age-related in nature.

The TVA process directs the reporting of plant-specific operating experience on age-related degradation and aging management to the industry through the Institute of Nuclear Power Operations OE program, consistent with the guidance in NEI 14-13, "Use of Industry Operating Experience for Age-Related Degradation and Aging Management Programs."

Each AMP summary in Appendix B of the SLRA contains a discussion of OE relevant to the age. This information was obtained through the review of internal OE captured by the Corrective Action Program, program assessments, program health reports, and through the review of external OE.


Question: percent motor voltage after 2.1 seconds

In conclusion the percent voltage is less than 25% after the 2.1 sec operating time of the first level undervoltage relay which allows a safe delayed transfer.

The relays must not trip because of the depressed voltage occurring during simultaneous motor starting for a LOCA.


Question: What materials are commonly used for structural support in nuclear facilities?

Equipment supports and foundations are typically made of reinforced concrete to provide structural support.

Flood barriers are critical structures that must remain intact during extreme weather conditions to protect vital areas.


Question: The evaluation of k_eff included variations in water density across normal and accident conditions.

By changing the water density (up to the water density at 124 °C and 20% void) and temperature in MCNP, the effect of the moderator (water) density on k_eff was evaluated.

The 4 °C moderator temperature was conservatively used for the base case model (base case fuel and rack).


Question: minimum cooling time for nuclear fuel assemblies

The data shows various enrichment levels for different reactor designs.


Question: What is involved in the decommissioning of nuclear power plants?

Experience with other commercial nuclear plants has demonstrated the feasibility of decommissioning and dismantling such plants sufficiently to restore a site to its former use (NEI 2016).

Continued operation for another 20 years beyond the currently licensed term would not increase the short-term productivity impacts described here.


Question: The pathways for containment volume interactions are critical in understanding reactor safety mechanisms.

Pathway 4: Unsprayed Containment Volume to Environment

The compartments listed show numerical data relevant to various sections of the reactor model.