Question: heat stress requirements for containment entry

Low Work Rate is defined as sedentary activities involving sitting, standing still, and low physical effort. These include inspections and operation of instruments and powered equipment.

When determining the work rate using Attachment 2, Work Rate Metabolism Guideline, the evaluator should use the most frequent workload activity (>50% of work time) to characterize the overall activity as being Low, Moderate or High exertion. Higher exertion levels result in higher metabolic body heat generation.


Question: What are the success criteria for the Part 53 rule in nuclear regulation?

The purpose of these success criteria is to define what functionality needs to be provided in the final Part 53 rule, and will be a guide for the Part 53 development, and future discussions to be held between the NRC staff and industry to ensure that the final rule meets the pre-defined success criteria.

The NRC discussed draft proposed rule text for Part 53 Subpart B “Technology-Inclusive Safety Requirements” during their public meeting on November 18, 2020.


Question: The measurement of turbidity in water is expressed in NTU (Nephelometric Turbidity Units).

TURBIDITY                         |          |          |           | 0.600      | NTU      | REG         |

Saltstone wells provide crucial laboratory and field data for safety assessments in regulatory processes.


Question: coefficient of reactivity

The coefficient of reactivity is dependent upon core characteristics, such as fuel loading, the dissolved absorber (boron) concentration, and burnable poisons.

The protection system is designed with consideration of the most probable failure modes of the components under various perturbations of the environment and energy sources.


Question: ASME Code Case N-860 inspection requirements

The NRC staff endorses ASME CC N-860 as generally acceptable for use in complying with the requirements in 10 CFR 72.42 and 10 CFR 72.240.

For this alternative, the ISFSI sites should meet all the following criteria:


Question: The cooling system at DNPS is subject to regulatory oversight to ensure minimal impact on local environmental resources.

Therefore, the cooling system at DNPS would continue to have a SMALL impact on terrestrial resources.

Given that there are no recorded bird deaths at DNPS over the past 5 years and with DNPS’s adherence to regulatory and permit requirements, the impacts due to bird collisions with station structures and transmission lines would remain SMALL.


Question: Action levels for contamination differ between restricted and non-restricted areas at the facility.

Action levels: Non-restricted - 200 net dpm; Restricted - 2000 net dpm.

Weekly wipe tests are conducted to monitor background radiation levels across various locations in the facility.


Question: material specifications for auxiliary piping valves

Auxiliary Piping Valves (Class 1) | SA182 Type 304 or 316 or SA351 GR CF8, CF8A or CF8M

The material information contained in this Table is 'Historical' in nature and is intended to be used for reference only.


Question: NELAP certification requirements

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

Samples were reprepared due to high relative percent difference/relative error ratio. The re-analysis is being reported.


Question: ambient background analysis in nuclear regulatory reports

Ambient Background was recorded as 8.36E+01 with an uncertainty of 3.72E+01 during the peak analysis for energy readings in the regulatory report.

Errors quoted at 2.00 sigma indicate the level of uncertainty associated with the peak areas in the nuclear analysis report.


Question: What is the unit of measurement for radioactivity?

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


Question: radiological hazards assessment

The radiological hazards assessment evaluates potential exposures to radiation from plant operations, including normal operations, anticipated operational occurrences, and design basis accidents.

The cooling system of the reactor is designed to maintain optimal temperature and maximize efficiency during operation, ensuring the safety and reliability of the energy output.


Question: The weighted mean activity of a nuclide is calculated in terms of its pCi per gram.


Question: The NRC specifies that devices containing radioactive isotopes must be registered, including details such as activity and measurement units.

The manufacturer TN TECHNOLOGIES, INC. produces various types of equipment for multiple industries, ensuring compliance with safety standards.


Question: minimum containment pressure required to prevent flashing

The minimum containment pressure required to prevent flashing is at least 1 psi lower than the containment pressure from the safety analysis, and this minimum margin is momentary.

The minimum strainer submergence of 1.17 ft was converted to 0.47 psi conservatively, using the density of water at the highest sump pool temperature after the onset of recirculation (268.8°F).


Question: requirements for the renewal of operating licenses

10 CFR 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants"

Regulatory Guide 1.188, Revision 2, "Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses"


Question: The integral frame must be designed to withstand specific loading conditions such as dead weight and dynamic building effects during loss of coolant accident scenarios.

The integral frame is designed to resist various loading conditions such as dead weight, dynamic building effects, and any combination of these during a loss of coolant accident (LOCA).

The pressure measurement ranges are categorized into several thresholds: 0-6 Ksi, 6-24 Ksi, 12-24 Ksi, and greater than 24 Ksi.


Question: The NRC evaluates aging effects associated with pressure boundaries during regulatory inspections.

The Valve Body is a Pressure Boundary component made of Cast Austenitic Stainless Steel and is subject to an aging management program.

Nuclear power plants utilize various types of cooling systems to maintain optimal operational conditions and ensure safety.


Question: What are the safety requirements for hazardous liquid pipelines?

Hazardous Liquid Pipelines must adhere to strict safety standards to prevent leaks and ensure the integrity of operations during transportation of hazardous materials.

Delaware Route 1 is a major road that runs through the state, providing an important transportation link for various types of vehicles.


Question: The enrichment levels for initial reactor assembly are critical for ensuring proper fuel performance.

The reactor cooling systems are designed to remove heat from the core effectively.


Question: The NRC evaluates the effectiveness of licensing actions during inspections.

Technical Quality of Licensing Actions is essential for ensuring that nuclear licensing processes meet regulatory standards and safety requirements.

The areas of responsibility for the IMEP review team members include technical staffing and training, status of materials inspection, and legislation.


Question: What are the requirements for radiation monitoring systems in nuclear facilities?

Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Knowledge of facility protection requirements including fire brigade and portable firefighting equipment usage.


Question: steam generator safety procedures

**5. IDENTIFY** Faulted S/G based on ANY of the following: Any S/G pressure dropping in an uncontrolled manner. OR Any S/G pressure less than 140 psig. OR S/G enclosure temps high: 1. T1002A for 2 and 3, 2. T1003A for 1 and 4. OR Local indication of break in any of the following: Main steam lines. Main feedwater lines. Other secondary piping.

STOP Cntmt spray pumps. AND CLOSE Cntmt spray discharge valves 1-FCV-72-2 and 1-FCV-72-39.


Question: aging effects in nuclear regulatory filings

The following aging effects associated with the Structural Commodities components require management: Cracking, Cracking; loss of material, Cumulative fatigue damage, Hardening or loss of strength; loss of material; cracking or blistering, and others.

The following aging management programs manage the aging effects for the Structural Commodities components: 10 CFR Part 50, Appendix J, ASME Section XI, Subsection IWE, ASME Section XI, Subsection IWF, and others.


Question: What factors need to be considered for environmental impact near nuclear plants?

The report provides demographic information about the Hispanic and Black or African American block groups within a 50-mile radius of the TVA Browns Ferry Nuclear Plant to assess potential environmental impacts and community concerns.

The counties in the figure are essential for identifying community demographics that could be influenced by the operation of the plant.


Question: What is Stokes' Law in relation to sediment particle settling?

Using these assumptions, the particle settling velocity is determined using Stokes’ Law (Reference 2.4.2-2). This law relates particle settling rates to particle diameter, and predicts that larger particles settle faster than smaller particles.

To determine the potential particle sizes that might be carried to the PSEG Site, soil characteristics for the Hoopes Reservoir watershed are determined using soil profiles from New Castle County, DE, where the reservoir is located.


Question: The regulatory limits for radiation levels must be respected during operations.

The Action Level (cpm) is an important threshold that indicates the maximum allowed count rate for radiation detection, defining the point at which regulatory actions may be taken.

The scan results are evaluated for anomalies that could indicate malfunctioning equipment and are documented in detailed reports for assessment.


Question: What are the retention requirements for regulatory filings?

SRRS: 3D.100; There are no retention requirements for this section


Question: flooding hazard reevaluation

The Flooding Hazard Reevaluation Report assesses the potential flooding hazards that could affect the safety and operation of the Vogtle Units 1 and 2, addressing historical data, site conditions, and potential external events.

The report also outlines the procedures and guidelines that the licensee must follow to conduct routine maintenance and inspections of the nuclear reactors.


Question: nuclear material measurement devices

The 8100-NG-T-Kr85 is a specialized instrument designed for measuring the concentration of Kr-85 in nuclear regulatory applications.

Ohmart Corporation's products include the SHLG Series which primarily focuses on liquid level monitoring in various industrial applications.


Question: What is the function of the C-1 control in nuclear safety systems?

C-1 1/2 Neutron flux (intermediate range) above setpoint blocks automatic and manual control rod withdrawal.


Question: The mean activity of radioactive isotopes needs to be assessed for safety calculations.

Wt mean Activity (pCi/grams) describes the average quantity of a nuclide present in a given mass, providing crucial information for regulatory compliance and safety assessments.

Errors quoted at 2.00 sigma indicate the statistical reliability of the measurements, but do not directly relate to the calculation of mean activity.


Question: The deployment of advanced reactors is essential to maintain competitiveness in the nuclear sector.

Without the Proposed Action, the U.S. would lose its competitiveness in the nuclear industry, there would be limited improvement in fuel types for the existing fleet, and AR deployment would be delayed.

The global market for nuclear power could triple by 2050 due to increasing pressures to transition to a carbon-free world.


Question: What is the function of the Auxiliary Feedwater System (AFWS) in a nuclear reactor?

The safety-related function of the AFWS is to maintain water inventory in the steam generators for reactor residual heat removal when the main feedwater system is unavailable.

The turbine-driven pump of mechanical train C is powered by steam tapped from the main steam lines of steam generators 1 and 2.


Question: What is the significance of sample analysis in nuclear licensing?

Sample analysis is crucial for ensuring that the materials used in nuclear facilities meet safety and regulatory standards, allowing for proper assessment of their safety and environmental impacts.

The date collected must be accurately recorded and reflect when the sample was physically taken to maintain a clear chain of custody.


Question: What information is contained in regulatory filings for nuclear safety?

QSA Global, Inc. provides detailed regulatory filings which include essential safety and compliance information necessary for the nuclear licensing process.

The Masterminder 2 is a sophisticated software tool used in various industrial applications to enhance project management and workflow efficiency.


Question: What is the meaning of nuclide activity in nuclear regulations?

The activity of a nuclide is typically expressed in picocuries per gram (pCi/grams) and indicates the amount of radioactivity present in a given mass of the material.

The energy of CE-144 is recorded as 80.12 keV, which is significant for its identification in radiation measurements.


Question: quality assurance program for nuclear repairs

A comprehensive quality assurance program would be required to insure that upon completion the new welds would be acceptable in all respects.

Several meetings were held with our engineers, Stone and Webster, to discuss the repair specification in light of the information available.


Question: site characterization in nuclear regulation

As stated in Section 2.1 of the LTP, the purpose of site characterization is to ensure that the NS Savannah final status survey (FSS) will be conducted in all areas where contamination existed, remains, or has the potential to exist or remain.

The LTP identifies the survey instruments and supporting quality assurance (QA) practices used in the site characterization program.


Question: What is the Gordon aquifer and its significance?

The Gordon aquifer, which ranges in thickness from 18 to 49 m [60 to 160 ft] at SRS, consists of loam and clay and from the Congaree formation and, in some places, sand from the Fourmile Branch and Snapp formations.

The SRS encompasses portions of Aiken, Barnwell, and Allendale Counties in South Carolina.


Question: What are the attributes of the Real-Time Model in nuclear applications?

The Real-Time Model includes attributes that enhance its capability to process and analyze data relevant to nuclear safety and operational efficiency.

The operations team conducted scheduled maintenance on the cooling systems to ensure optimal performance during peak load conditions.


Question: What is meant by the term 'unit of measurement' in nuclear licensing?

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

Manufacturer name: TN TECHNOLOGIES, INC.


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

The Safety Analysis Report (SAR) provides a comprehensive analysis of the potential hazards associated with the licensed nuclear facility, detailing the design and operational features that ensure safety and compliance with regulatory requirements.

Section (b)(5) discusses administrative procedures for submitting additional documentation as requested by the NRC.


Question: Nuclear power plants must adhere to specific design rules to ensure safety and minimize environmental impacts.

The assumption that nuclear plants are safe -- and therefore pose no significant adverse environmental impacts -- because they are designed and built to an array of closely related and interdependent deterministic requirements, i.e., design rules, promulgated by NRC under the Atomic Energy Act.

A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions.


Question: What are the aging effects requiring management for concrete in nuclear power plants?

Increase in Porosity and Permeability Loss of strength

The NRC documents specify that the aging management review includes evaluations of structures and components to ensure their safety and functionality.


Question: What is the average enrichment weight percentage of uranium-235 in nuclear fuel assemblies?

The data presents the average enrichment weight percentage of uranium-235 (E) in various fuel assembly burnup ranges, specifically detailing how they fall within different brackets from 1.3 to 4.9 and above.

The average burnup for a WE 14x14 assembly is shown here in gigawatt days per metric ton of uranium (GWd/MTU), indicating performance metrics during operation.


Question: maximum quantity of U-235 per package

The U-235 limits are determined by enrichment and are not to exceed the limits established below:

The quantity of hydrogenous packing materials is unlimited unless otherwise specified.


Question: The NRC sets forth guidelines for evaluating system performance under emergency conditions, particularly for cooling systems.

The Emergency Core Cooling System (ECCS) is designed to ensure adequate cooling of the reactor core during postulated accidents by providing water to the core immediately following a loss of coolant accident (LOCA).

The containment spray system is responsible for maintaining appropriate pressure levels within the containment vessel to ensure safety during a reactor operation.


Question: A report must include the Minimum Detectable Activity (MDA) for various nuclides, which is crucial for proper regulatory compliance.

The Nuclide MDA (pCi/grams) gives the minimum detectable activity for specific nuclides, crucial for assessing accuracy in radiological measurements.

The table provides information on the energy levels and peak sizes of different peaks identified in the spectrum analysis.


Question: FLECHT run 9573 heat transfer coefficients

Regarding Petitioner’s comments on PRM-50-93/95 dated March 15, 2010 (pages 5-9), concerning FLECHT run 9573 heat transfer coefficients, NRC’s October 2012 DIR states: The comments discuss the negative heat transfer coefficients near the mid-plane elevation in FLECHT run 9573 and that, as pointed out in the data.

In FLECHT run 9573 there were three thermocouples that registered temperatures greater than 2200 degrees F at a time of 18 seconds. These were thermocouples numbered 3D3, 2D2, and 4E3. Each of these three thermocouples was on the interior of the bundle and shielded from the housing by at least one row of heater rods.


Question: What is included in the commercial database for FGR?

The commercial database for FGR (see Table 4-7) is made up of rods with UO2 and gadolinium-bearing fuel irradiated in PWR and BWR reactors.

The reactor's cooling system is designed to maintain operational safety during both normal and emergency conditions.


Question: What are the technology guidelines for virtual NRC meetings?

Technology does not always cooperate, and all participants should prepare their computers, screens, microphone, camera and/or phone well ahead of the meetings.

If the re-occupancy phase allows, external participants may choose to participate from the Commission Hearing Room.


Question: In scenarios where the Licensing Conditions are not satisfied, specific corrective actions must be implemented.

If the LCO (Limiting Condition for Operation) is not met, the required actions must include the verification of established procedures and potential corrective measures as outlined in the regulatory guidelines.

Each action listed must be done in accordance with specified protocols, ensuring that safety and compliance are prioritized.


Question: The NRC mandates that safety-related systems must function effectively under a variety of conditions to ensure operational integrity.

Systems in areas not containing redundant systems or components (outside containment) must have operable spray and/or sprinkler systems to maintain safety.

Power levels must be monitored continuously to ensure compliance with safety regulations and operational efficiency.


Question: What is blowdown mass and energy release in nuclear reactor safety?

Blowdown mass and energy release refers to the measurement of mass and energy exiting through break paths during an accident scenario in nuclear reactors.

The data in Table 6.2-13 presents flow measurements in lbm/sec and energy in Btu/sec for different break paths.


Question: Ultrasonic Examination of reactor components

Ultrasonic Examination (UT) of ASME Class 1, Component ID. 1RPV1-446A, Reactor Vessel Closure Head Meridional Weld

Written Practice Certification Review, ID No. 288663


Question: Nickel-63 regulatory usage

Nickel-63 is a radionuclide commonly encountered in various environmental and industrial applications, known for its beta decay and relatively long half-life of about 100 years.

The Strontium Carrier is tested under the GFPC methodology, yielding a recovery rate of 86.2 within acceptable limits of (25%-125%).


Question: Fire Area definitions

The fire area is defined as a specified region where certain fire safety measures are implemented and assessed to ensure compliance with safety regulations.

The report includes various sections that detail the design specifications and engineering calculations necessary for construction approval.


Question: Americium-241 measurement standards

Americium-241 is measured in units of picocuries per gram (pCi/g), with reported uncertainty values reflecting the precision of these measurements.

The uncertainties for the Curium-243/244 measurements include values such as +/-0.011 and +/-0.0108, indicating the range of possible error around the reported figures.


Question: The generation of hydrogen during reactor operations is primarily influenced by chemical reactions occurring in the reactor coolant.

Includes hydrogen generated by the zirconium-water reaction and dissolved in the primary coolant.

The table contains historical documentation as discussed in section 6.1.7.


Question: The average enrichment of uranium in an initial assembly must meet specific regulatory guidelines.

Minimum Initial Assembly Avg. Enrichment wt % 235U (E) defines the necessary minimum enrichment levels for various burnup ranges in boiling water reactor assemblies.

The data represented in the table reflects various configurations of assembly types and their corresponding cooling times after discharge.


Question: The NRC mandates that licensees follow specific protocols during the defueling process, which includes careful handling of radioactive materials.

The 4-meter-long pieces of highly irradiated stainless steel were highly radioactive with an estimated peak of 3,000 roentgen per hour gamma.

The removal options included the following: Cut the plates into small pieces and load them into fuel canisters or other specially designed containers.


Question: The structures that are reinforced concrete and designed to be seismically qualified are also essential for supporting the auxiliary saltwater system.

The seismically qualified portions of the circulating water conduits and ASW vacuum breaker vaults are reinforced concrete structures founded on compacted fill.

For the purposes of LR and aging management, the breakwaters and earth slope protection structures are evaluated as barriers.


Question: The NRC's regulations define tritium as a byproduct material, which involves specific safety and monitoring requirements.

The NRC currently regulates tritium as a byproduct material, meaning that the NRC’s rules under 10 C.F.R. Part 30 already thoroughly address the risks associated with tritium.

Fission plants create high-level radioactive waste as used nuclear fuel from their power generating process.


Question: What are the criteria for a Method Blank in nuclear analytical procedures?

The blank result (See Below) is greater than the MDC but less than the required detection limit.

Samples were prepped due to high blank activity. The re-analysis is being reported.


Question: technical specifications for emergency core cooling systems

The emergency core cooling system (ECCS) is designed to ensure the long-term cooling of the reactor core in the event of a loss-of-coolant accident (LOCA).

The proposed amendment includes modifications to the administrative controls and technical requirements for the framework of operational safety assessments.


Question: The activity levels of the nuclides must be carefully measured and reported during regulatory filings.

Wt mean Activity (pCi/grams) represents the average radioactivity concentration of the nuclide in terms of picocuries per gram, providing essential data for safety assessments.

Nuclides that were rejected by the interference analysis are designated with an 'X', indicating they do not meet the necessary criteria for inclusion in the safety evaluation.


Question: What are the radiation doses to members of the public from operations?


Question: local PMP rainfall intensity

The maximum PMP rainfall intensity is discussed in Subsection 2.4.2.3.3.2.2.

For the purposes of flood estimation, the Mississippi Basin was divided into drainage areas and subareas.


Question: The license renewal process for nuclear plants involves a detailed environmental assessment.

The License Renewal Environmental Review for the Browns Ferry Nuclear Plant Unit 1, 2, and 3 is an essential process that assesses the environmental impacts of extending the operating licenses of these reactors.

The reactors at the Browns Ferry Nuclear Plant were initially commissioned in the 1970s and have undergone various upgrades over the years.


Question: Routine inspections for nuclear licenses occur without prior notification to the licensee.

Routine inspections should be unannounced.

The inspector has observed the licensee conduct its most risk-significant, authorized activities with no significant findings (for example, the inspector observed the licensee’s radiography crew perform work at a temporary job site and there were no significant findings);


Question: How does the NRC regulate actions taken during a loss of coolant accident?

Verifies Torus level < 27 ft. and directs BOP to start Torus Sprays when Torus pressure exceeds 2.5 psig.

The RHR 26B valve breaker will not reset. I’ll contact EMs to assist in troubleshooting.


Question: What are severe accidents in the context of nuclear safety?

Severe accidents are postulated accidents that are more severe than DBAs because severe accidents can result in substantial damage to the reactor core, with or without serious offsite consequences.

The NRC staff has reviewed the NEI 17-04, Rev. 1, document and endorsed its interim use (NRC 2019a).


Question: Eagle 21 Digital Protection system

The UFSAR refers to an Eagle 21 Digital Protection system. Page 7.2-3 of the UFSAR states the following: "Channel bypass capability exists for Eagle 21 (overpower _T, overtemperature _T and Hi Pressurizer Level) and the Nuclear Instrumentation System (Source Range and Intermediate Range) utilizes channel bypass for calibration/surveillance."

Is the modification scope really limited to analog systems as stated in the analysis or is the digital Eagle 21 being upgraded as well?


Question: The licensee proposes a modification to the testing schedule for SRVs, extending it from 24 months to 48 months under certain conditions.

The licensee proposes an alternative to the certain requirements in ASME OM Code, Division I, Appendix I, subparagraph I-1320(a). Specifically, the proposed alternative would change the test sampling interval, which requires a minimum of 20 percent of the valves from each valve group to be tested, for the applicable SRVs from 24 months to 48 months.

At QCNPs, Units 1 and 2, there is a single Target Rock 3-Stage, Model 74-67F, SRV installed on each unit's Main Steam lines inside the drywell.


Question: extravasation of diagnostic radiopharmaceuticals

In summary, there is no clinical data that supports Lucerno Dynamics' claim that extravasation of diagnostic radiopharmaceuticals is a patient safety issue.

I believe that extravasations are best managed on an institutional level at the discretion of the authorized user and do not require additional NRC regulation.


Question: The NRC requires enhancements to procedures related to the management of aging effects on various materials used in safety-related systems.

The existing program is credited and will be enhanced as follows: 1. Revise implementing procedures to include heat exchangers in the High Pressure Fire Protection (Diesel Driven Pump) System, the Control Air System, and the Control Rod Drive System. 2. Revise implementing procedures to include reevaluation, repair, or replacement of components that do not meet minimum wall thickness requirements. If fouling is identified, the overall effect is evaluated for reduction of heat transfer, flow blockage, loss of material, and (if applicable) chemical treatment effectiveness. For ongoing degradation mechanisms (e.g., MIC), the frequency and extent of wall thickness inspections are increased commensurate with the significance of the degradation.

The implementation schedule for this program will be enhanced no later than six months prior to the subsequent period of extended operation.


Question: The report outlines various radionuclides measured and their respective activities in picocuries per gram (pCi/g) of dry material.

The description of the sample ID includes both a unique identifier and the documentation for its testing process that must be adhered to.


Question: NRC license devices

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

The device model and serial number must be correctly identified to ensure proper tracking.


Question: What are sensitivity studies in the context of nuclear safety?

**Sensitivity studies** are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event.

| PZR heaters                        | Nominal.            |       |


Question: nuclear licensing process

The nuclear licensing process involves several stages, including pre-application, application review, and post-licensing activities, ensuring that applicants meet safety and environmental standards before obtaining approval.

The safety review committee evaluated the performance of the new cooling system, which showed significant improvements in thermal efficiency compared to the previous model.


Question: The NRC has guidelines for the modernization of instrumentation and control systems within nuclear facilities.

This supplement to the regulatory audit plan provides a review framework for the Digital Instrumentation & Control Modernization Project, which is critical for improving safety and reliability in nuclear operations.

The regulatory framework consists of various procedures and guidelines that ensure the safe operation of nuclear reactors in the United States.


Question: Transportable Storage Canister requirements

The TSC shall be dry and helium filled. The following vacuum drying times shall be met as appropriate.

The following table outlines the heat load requirements for various reactor types.


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


Question: What is NPDES Permit No. IL0002224?

Process wastewaters are monitored and discharged to the Illinois and Kankakee Rivers via outfalls in accordance with the DNPS NPDES Permit No. IL0002224, which was issued in 2016, expired in August 2021, and has been administratively extended pending review of the submitted permit renewal application (IEPA 2016b).

Thermal monitoring and reporting requirements are further discussed in Section 3.6.1.2.6.


Question: personal protective equipment in nuclear regulatory processes

How satisfied were you with the NRC's issuance of personal protective equipment (PPE) provided to you for onsite activities?

Please provide your organization
- [ ] Region 1
- [ ] Region 2
- [ ] Region 3
- [ ] Region 4
- [ ] NSIR
- [ ] NRR


Question: Zircaloy cladding oxidation

It is significant that “if the heat removal capability is lost [from the oxidation of Zircaloy cladding materials by steam], it determines a feedback between temperature increase and cladding oxidation;” and that “any failure to remove the heat of the Zircaloy-steam reaction from the fuel cladding can result in an increase in the temperature of the cladding.”

Further, Figure 68 of the same paper depicts that the “start of rapid [Zircaloy] oxidation by H2O [causes an] uncontrolled temperature escalation,” at 1200°C (2192°F), and Figure 13 of the same paper depicts that if the initial heat up rate is 1 K/sec. or greater, a cladding temperature excursion would commence at 1200°C (2192°F), in which the rate of increase would be 10 K/sec. or greater.


Question: The application for license renewal includes an evaluation of how aging management programs will address structural integrity.

In view of the aforementioned ePortal references, it appears the SLRA lacks clarity as to what constitutes the RV support structure, how it comes together, what is the function of the wide flange structural shapes embedded in the PSW, what are the various jurisdictions according to which appropriate AMPs would manage the prevailing aging effects, how the inner and outer parts of “short-columns” are anchored to the PSW, etc.

ePortal DWGs E-511-224, E-411-251, and 583F788 show rods/bolts potentially connecting the inner part of each “short-column” to an underlying wide flange column assembly, while the outer part of the “short-column” (RV exposed side) appears to be held in place (see DWG E-511-224, Detail G) by a baffle assembly.


Question: What is the importance of density in nuclear shielding materials?

The material density of the side walls in the shielding design is 2.3 g/cm³, indicating its effectiveness in attenuating radiation.

The integration values show a convergence of 1.00%, with MDRPN set at 24° and CRPN at 24°, which are crucial for model accuracy.


Question: What is the minimum margin of fuel rim temperature to melt limit in nuclear reactors?

During a prompt transient, the fuel rim temperature can be the limiting temperature during the pulse. For this reason, the fuel rim temperature is considered to for fuel melt in prompt critical cases. For the Byron and Braidwood REA analysis, the minimum margin to the melt limit (value – limit) for the fuel rim occurs at **[ ]**.

From the results in Table 4-6, the minimum enthalpy rise margin is **[ ]**.


Question: The enrichment levels of Uranium-235 in fuel assemblies are critical for reactor performance.

The maximum enrichment level permitted for U-235 in a BWR fuel assembly is specified in the regulatory documentation.

The cooling system of the reactor must be maintained within specified temperature limits to ensure safety and efficiency.


Question: The step-by-step process for computing internodal flow rates and energy inventories in a simulation model is crucial for nuclear safety analysis.

This conceptual flow chart illustrates how three computer codes interact to model the mass and energy releases during a nuclear event, incorporating various nodes for detailed analysis.

Energy conservation principles in nuclear systems are essential for ensuring reactor safety during operational phases.


Question: fuel region box types

1 # inner canister: bottom fuel_region #1 w/gap: body asyy


Question: What is the role of reception centers in emergency evacuation procedures?

Information on the layout and capabilities of reception centers can be found in the Reception Center implementing procedures.

Parish evacuees are directed to the reception center in East Baton Rouge Parish.


Question: The NRC expects licensees to conduct thorough assessments of safety protocols in response to any proposed legislative changes.

The U.S. Nuclear Regulatory Commission (NRC) actively monitors and assesses proposed legislation that may impact regulatory requirements for nuclear facilities.

The meeting was attended by various stakeholders and covered a range of topics including environmental impacts and community relations.


Question: nuclide identification report

Identified nuclides include K-40, CS-137, PB-212, BI-214, PB-214, RA-226, and AC-228, listed with details such as energy (keV), yield (%), activity (pCi/grams), and activity uncertainty.

Errors quoted at 2.000 sigma provide a statistical measure of the uncertainty associated with the nuclide identification process.


Question: The NRC evaluates potential risks for various operational modes, including both at-power and low power/shutdown conditions.

At-Power: Include LOCA initiators - Dominant identified contributor to challenges to ECCS suction strainer.

Seismic induced LOCA events are estimated to be small or negligible contributors to the Fermi-2 LAR.


Question: The NRC evaluates the potential effects of proposed projects on historic properties during license application reviews.

The National Historic Preservation Act Section 106 requires federal agencies to consider the effects of their actions on historic properties.

The proposed site for the facility is located in a region known for its geological stability, which is essential in the context of waste storage.


Question: License applicants must demonstrate compliance with regulatory requirements during technical evaluations.

The application must detail the technical specifications and safety analysis for the reactor design to ensure compliance with NRC regulations.

Nuclear power plants generate electricity through the fission of uranium atoms in a controlled nuclear reaction.


Question: flood hazard reevaluation

The Flood Hazard Reevaluation Report provides a comprehensive analysis of potential flooding impacts on the Cooper Nuclear Station, particularly focusing on identified flood hazards and the methodologies used in assessing flood risk.

The evaluation of dam breaches and failures is critical for understanding broader water management strategies, but does not directly relate to the specific flood hazard concerns addressed in regulatory processes.


Question: The NRC assesses the environmental impact of subsequent license renewals for nuclear facilities.

The NRC implements a rigorous environmental review process for the subsequent license renewal of the Virgil C. Summer Nuclear Station, Unit 1.

The Audit Plan includes various evaluations of compliance with safety standards and practices.


Question: What are the operational parameters for battery cells in nuclear power plants?

Battery cell parameters are critical in ensuring reliable power supply for safety systems during normal and emergency conditions.

The distribution systems must be properly maintained to ensure adequate power delivery for all electrical loads in the plant.


Question: What is the Residual Heat Removal Service Water (RHRSW) system in nuclear facilities?

**DSN005:** Residual Heat Removal Service Water (RHRSW) System

- **DSN018:** Stormwater from Materials and Procurement Complex (MPC) parking lot, the firing range parking lot, the Facilities Maintenance area, the vehicle fuel dispensing area, part of the parking lot southwest of the MPC, and adjacent areas.


Question: What is the role of Primary Containment Isolation Instrumentation in nuclear reactors?

Primary Containment Isolation Instrumentation outlined in Table 3.3.6.1-1 is crucial for the initiation of closure of appropriate primary containment isolation valves when one or more required channels are inoperable.

The NRC sets strict guidelines for the operation of nuclear reactors to ensure they meet safety protocols and regulatory standards.


Question: Strontium-90 analysis in nuclear regulatory filings


Question: seismic risk assessments in nuclear regulation

Incorporation of seismic risk tabs to support RICT Program risk calculations. Conservatism values for the seismic delta CDF and LERF are applicable.

The model, if restructured, will be logically equivalent and produce results comparable to the base PRA logic model.


Question: What are the surveillance requirements for nuclear equipment?

The following SRs are applicable for all equipment required to be OPERABLE: SR 3.5.2.1, SR 3.5.2.3, SR 3.5.2.4, SR 3.5.2.7.


Question: What is a Job Performance Measure in the context of nuclear plant operations?

The Job Performance Measure (JPM) is utilized to evaluate the competency of personnel in performing job-related tasks in a nuclear facility, ensuring that safety and regulatory standards are met.

The Watts Bar Nuclear Plant utilizes various systems to ensure operational efficiency, including advanced cooling mechanisms and automated monitoring equipment.


Question: baseline core damage frequency

The baseline Core Damage Frequency (CDF) serves as a measure of the likelihood of a core damage event in a nuclear reactor, providing a quantitative basis for assessing plant safety.

The reactor containment structure is designed to withstand external pressures and prevent the release of radioactive materials during normal operations.


Question: What are non-minimum augmenting ERO positions in nuclear emergency plans?

Non-minimum augmenting ERO positions are being removed from the emergency plan, but retained in the Emergency Plan Implementing Procedures (EPIPs).

NextEra station and corporate personnel can be called as 'as needed' to support ERO response activities.


Question: The analysis of core deformation indicates that control rods are not likely to bind due to deformation of fuel assemblies.

The maximum deformation (generally inward) of the fuel assemblies has been computed to be between 0.29 and 0.31 cm at BOL and EOL respectively, however the control rods experience lesser deformation, as shown in Table 2-1. Therefore, it is unlikely that the core would deform sufficiently to cause a binding of the control rods.


Question: Makeup Water System aging management evaluation

Aging Management Program: External Surfaces Monitoring of Mechanical Components (B.2.3.22) for the Makeup Water Demineralizer Caustic Day Tanks, which have a Leakage Boundary with carbon steel material.

Concrete (Coated/Lined) is used in the tank for Raw Water Storage Reservoirs to maintain pressure boundary integrity.


Question: What is a Safety Injection System in a nuclear reactor?

The Safety Injection System provides a pressure boundary for components such as piping and pump casings, which are crucial for maintaining reactor safety during emergency conditions.

Piping components, particularly those under Class 1 specifications, are exposed to various environments that can contribute to aging effects requiring management.


Question: What is the purpose of the torus shell in nuclear containment structures?

Steel elements: torus shell serves as a Pressure Boundary, Structural Support, and Water Retaining Boundary.

Steel elements: downcomers are responsible for pressure boundary and structural support in reactors.


Question: What is the weighted mean activity of nuclides?

The table lists various nuclides along with their weighted mean activity measured in picocuries per gram (pCi/grams).

Errors quoted at 2.000sigma indicate the level of uncertainty in the reported measurements, emphasizing the importance of rigorous analysis.


Question: spent fuel storage estimates

The Staff found that safe storage of spent fuel in a geologic repository is technically feasible using currently available technology, with no major breakthroughs in science or technology needed.

financial assurance to be acceptable if it is based on plausible assumptions and forecasts, even if 'the possibility is not insignificant that things will turn out less favorably than expected.'


Question: The NRC's oversight includes reviewing licensee submittals related to regulatory compliance and operational standards.

The ITAAC (Inspections, Tests, Analyses, and Acceptance Criteria) process is essential for verifying that significant components of a nuclear facility conform to the regulations and safety standards.

The NRC checklist consists of various elements aimed at ensuring a comprehensive review of all submitted materials, including reports and notifications.


Question: The transport of debris during an event involves considering both fine debris and its behavior under various conditions.

A conservatively low pool fill-up transport fraction of 15% was used for debris transport into the inactive cavity.

The quantity of fines generated and subsequently transported to the strainers and core is over-predicted.


Question: What is a performance deficiency in nuclear regulatory context?

The licensee's failure to identify the design deficiency during post-maintenance testing was a performance deficiency that was within the licensee's ability to foresee and correct.

The alarm condition prevented the restart of the chiller after the “loss of offsite power” condition.


Question: pneumatic cylinder in nuclear safety systems

A pneumatic cylinder is a vital component in the operation of the rack and pinion drive for safety rods, providing necessary force to position the control rods.

The gear housing protects the internal components of the drive system and ensures smooth operation of the motor assembly.


Question: What is the required operator action when synchronizing the DG1AB system at noon?

Operator closes DG1AB Synchronize Master Close switch at 5 min to 12 o’clock position.

The US acknowledges that the EDG is loading continuously and should be unloaded/tripped.


Question: What is the function of Power Operated Relief Valves (PORVs) in nuclear safety?

2 Power Operated Relief Valves (PORVs) and automatic actuation circuitry | Yes                                   | RCS depressurization, once through core cooling (feed and bleed Automatic pressure relief during inadvertent ECCS actuation) | 1 of 2 PORVs | Feed and Bleed 1 charging pump for injection requires 1 PORV, 1 SI pump for injection requires 2 PORVs. Depressurization 1 PORV required. ATWS Number and combination of SRVs and PORVs varies depending on desired relief capacity.

With both PORVs inoperable due to causes other than excessive seat leakage, within 6 hours either restore at least one PORV to OPERABLE status or close the associated block valves and be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.


Question: LBLOCA debris testing

Strainer submergence was maintained between 14 in and 18.1 in from the top perforated plate of the test strainer. This range of submergence was consistent with the plant strainer submergence for a LBLOCA.

Prior to each conventional and chemical debris addition, the pH of the water inside the test tank was measured. If the measured pH was higher than 8.2, boric acid was added to the test tank to lower the pH to below 8.2.


Question: What is required for a location of use or possession regarding licensed materials?

An acceptable location of use or possession specifies street address, city, State, and zip code and does not include a post office box number.

An NRC license does not relieve a licensee from complying with other applicable Federal, State, or local regulations (e.g., local zoning requirements).


Question: In situations of inadvertent control rod withdrawal, procedures must ensure that operator actions are backed by automatic systems to prevent excessive reactivity changes.

The RBM system initiates a rod block, inhibiting further withdrawal.

For the postulated events of inadvertent control rod withdrawal, concurrent with a postulated CCF of PPS, sufficient controls that are independent of the PPS are available for mitigating these postulated events.


Question: shared sensor configuration in nuclear safety systems

DPS and RRCS share sensors with the PPS in a configuration that maintains diversity and independence between the PPS and DPS/RRCS. Figure 3.5.3-1 is a typical shared sensor configuration for both the DPS and RRCS.

Select valve statuses that are fed back to the ILPs are shared with ILPs in different divisions when the valve status provides an interlock.


Question: nuclear safety regulatory process

Nuclear safety is a crucial aspect of regulatory filings, ensuring that all safety-related systems and protocols are adhered to for the protection of public health and the environment.

Oconee County's 2020 Comprehensive Plan outlines various development initiatives to support community growth and improve quality of life in the region.


Question: Feedwater Controller Failure

Feedwater controller demand is postulated to fail upscale that results in maximum feedwater injection into the RPV, resulting in a rapid water level rise and a trip of the main turbine and the turbine-driven feedpump turbines when water level reaches the Level 8 trip setpoint.

The reactor core is adequately cooled and RCPB integrity is maintained.


Question: What is the significance of Sr-90 in nuclear safety analysis?

Only the HTD ROC Sr-90 was included in the analysis. All analyses met the required MDC.

The total dose to the public from a nuclear facility must be assessed to maintain regulatory compliance.


Question: The goal of the onsite meteorological program is to ensure continuous monitoring of the weather parameters that have a significant impact on the dispersion of radioactive releases.

The applicant has stated that the "objective of the onsite program will be to maintain a continuous instrumentation surveillance of the meteorological parameters involved in the atmospheric dispersion of radioactive effluent releases and to have the data available at any time for assessing the relative concentrations and doses resulting from accidental or routine releases," which includes control room monitoring of appropriate meteorological parameters.

Data recording at the offsite facility is done on magnetic tape with strip charts forming the secondary system. At the temporary onsite facility, strip charts are used for data recording.


Question: The NRC emphasizes the importance of accurately determining the critical heat flux in reactor safety assessments.

The COBRA-FLX code provides a correlation for critical heat flux that is validated against a comprehensive experimental database.

The test assembly conditions for non-mixing grid tests include specific geometries and flow parameters which influence reactor performance.


Question: In scenarios where primary containment cannot be sustained, it is crucial for operators to execute an emergency depressurization of the reactor pressure vessel as stipulated in EOP-4.

Given the inability to maintain Primary Containment conditions inside the Pressure Suppression Pressure, the crew will perform an Emergency RPV Depressurization, in accordance with EOP-4. Emergency RPV Depressurization must be initiated within 15 minutes of Pressure Suppression Pressure being exceeded.

Given a stuck open SRV, the crew will scram the Reactor before Torus water temperature exceeds the Boron Injection Initiation Temperature, in accordance with AOP-36 and EOP-4.


Question: The NRC assesses the ecological impact of construction projects in sensitive environments.

The environmental assessment for the TRISO-X Fuel Fabrication Facility includes evaluations of local flora and fauna, assessing the effects of operations on the surrounding ecosystems.

Tall larkspur was not observed during field surveys but has the potential to occur in Lot 6b of the study area.


Question: What are the methods for uranium analysis in soil?


Question: The stresses on the basket under various design loads must be within specified limits to ensure safety.

The unit of measurement for design loads is expressed in pounds, which is a standard unit for such evaluations.


Question: What is the Emergency Preparedness Significance Determination Process?

0609 App B Emergency Preparedness Significance Determination Process 09/22/15 (15-017)


Question: In the event of a Loss of Control Air, the response involves critical operations that help manage the reactor's cooling system.

To control RCS temperature, the crew is required to LOCALLY THROTTLE RHR flow using 1-HCV-74-36, RHR HX 1A MANUAL BYPASS, AND 1-ISV-74-524, RHR HX 1A INLET ISOLATION.

Unit 1 is in MODE 5 and RHR Train A is in service for shutdown cooling.


Question: Structures Monitoring program

The Structures Monitoring program (B.2.1.33) will be used to manage cracking and loss of material of the concrete elements of Group 2, 3, 8, and 9 structures exposed to air-outdoor.

Consistent with NUREG-2191. At BFN, there are no stand-alone Group 1 or 5 structures, and Group 7 structures are not applicable to BFN.


Question: The lubricating oil system is crucial for ensuring the proper functioning of various critical components during both normal and emergency situations.

The purpose of the LO System is to provide oil for operation of the feedwater pump turbine, main turbine generator, and several large motor-driven pumps that represent both vital auxiliaries and steam and power plant auxiliaries.

The Pump Casing for the Diesel Engine Generator Jacket Water Cooling Pumps serves as a pressure boundary, ensuring system integrity under operational conditions.


Question: The NRC stipulates that at least two qualified circuits must be functional to connect the offsite transmission system with the onsite Class 1E power distribution system.

The following AC electrical sources shall be OPERABLE: Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and three diesel generators (DGs) capable of supplying the onsite Class 1E power distribution subsystem(s); and two supply trains of the diesel fuel oil (DFO) transfer system.

The electrical systems must meet stringent safety criteria to ensure the integrity of reactor operations and safeguard public health.


Question: What is the Louisiana Peacetime Radiological Response Plan (LPRRP)?

The plans that include emergency messaging actions include but may not be limited to: Louisiana Peacetime Radiological Response Plan (LPRRP)

When circumstances arise and there is a need for a public warning, the decision to send a message is ultimately a matter of local objective judgment.


Question: The NRC mandates that operational licensees must implement effective systems to prevent loss of cooling capabilities during shutdown conditions.

D3 Section 3.16 outlines the procedures and requirements necessary to maintain adequate shutdown cooling during loss of cooling events.

The staff recommended that the analysis should include a review of the incident response protocols for other operational areas.


Question: What is a draft Environmental Assessment (EA) in the context of nuclear regulatory processes?

The NRC and DOE LPO staff have completed a draft environmental assessment (EA) and draft finding of no significant impact (FONSI), *Draft Environmental Assessment and Draft Finding of No Significant Impact for the Palisades Nuclear Plant Reauthorization of Power Operations Project* (ADAMS Accession No. ML243534A157).

Pursuant to 36 CFR 800.8(c), the NRC, as the lead Federal agency, invites your office to submit any comments, or concerns on the draft EA and draft FONSI during the 30-day comment period, which ends on March 3, 2025.


Question: scram discharge volume vents

Relief Request RR-V-1 pertains to the scram discharge volume vents and drains, outlining the proposed alternative methods for testing these components as part of the inservice testing program.

The submittal also includes a comprehensive overview of the operational history and achievements of Units 1 and 2, highlighting their contributions to energy production and safety standards.


Question: nuclear licensing process

The licensing process for nuclear reactors involves multiple stages including pre-application meetings, safety reviews, and public comments before final approval is granted by the NRC.

The availability of committee members is crucial for decision-making processes; each individual must ensure they are prepared for upcoming meetings and able to contribute effectively.


Question: The results of the analysis indicate that some samples were received after their designated holding times, which affects their validity.

Samples (See Below) were received and analyzed outside the hold time.

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


Question: Aging management programs are critical for ensuring the integrity of structures and components subjected to aging effects.

AMP XI.S1, “ASME Section XI, Subsection IWE,” and AMP XI.S4, “10 CFR Part 50, Appendix J” will be used to manage cracking of stainless steel and dissimilar metal welds of penetration sleeves and penetration bellows exposed to an air - indoor uncontrolled in the BFN Mark I Steel Containment Structure.

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


Question: The COBRA-FLX codes are utilized to evaluate heat transfer and fluid flow in nuclear reactor cores.

COBRA-FLX is a thermal-hydraulic code used for calculating critical heat flux and assessing heat transfer in nuclear reactor cores.

The proposal includes various enhancements to the reactor design that align with current safety standards and improve efficiency.


Question: nuclear safety analysis

Nuclear safety analysis involves the assessment of the potential risks associated with the operation of nuclear reactors, focusing on the identification of hazards and the evaluation of safety measures to mitigate their impact.

The complexity of the regulatory framework requires multiple layers of oversight to ensure compliance with safety standards and operational guidelines.


Question: The NRC mandates that manufacturers register devices that utilize radioactive isotopes, ensuring they are properly tracked and regulated.

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

Please provide the details of your distributor license and any relevant information regarding your device's specifications.


Question: What is FQ surveillance in nuclear safety?

A similar requirement exists in the Byron and Braidwood TS, Reference 2 and 3 Surveillance Requirement 3.2.1.2; however, FQ(z) is monitored directly instead of FAH.

In order to remain consistent with the current Byron and Braidwood TS and COLR the PDC-A method will be updated to implement FQ surveillance consistent with the Byron and Braidwood TS, (Reference 2 and 3 Surveillance Requirement 3.2.1.2).


Question: What are the limitations for assemblies classified as damaged fuel?

When loading fuel assemblies classified as DAMAGED FUEL, all assemblies in the MPC are limited to 4.0 wt.% U-235.


Question: What is the role of the Emergency Response Organization in nuclear safety?

The Emergency Response Organization is responsible for managing and coordinating the response to nuclear incidents with an emphasis on public safety and regulatory compliance.

The American Nuclear Insurers provides financial protection and risk management solutions for the nuclear industry, ensuring that facilities have the necessary insurance coverage for operational risks.


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

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

The integrity of the inner port cover welds to the closure lid at the vent port and at the drain port shall be verified in accordance with the procedures in Section 9.1.1.


Question: The assessment of uncertainties in fuel characteristics is crucial for evaluating thermal performance.

All these figures exhibit similar extent of data scattering. The uncertainties considered in this response directly impact fuel related characteristics.

Figure 4-26 of Reference 2 shows the predicted versus measured rod diameter increase during ramp on the M5-UO₂ ramp database.


Question: risk-informed technical specifications

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, (ADAMS Accession No. ML023240437), dated November 2002.

The safety evaluation process is critical for assessing the implications of design changes on the overall plant operations.


Question: boron dilution system

The "borate" mode of operation permits the addition of a pre-selected quantity of concentrated boric acid solution at a pre-selected flow rate to the Reactor Coolant System.

If either counter is satisfied before the other has recorded its required total, the pump and valve associated with the counter which has been satisfied terminates flow.


Question: What are the key components of a safety systems assessment in nuclear licensing?

A safety systems assessment evaluates the design and operational characteristics of systems that are essential for ensuring the safety of the nuclear facility under normal and accident conditions.

The reactor must be designed to withstand seismic events and must include a robust containment structure that can prevent the release of radioactive materials in case of an incident.


Question: The assessment ensures that the structural components meet the standards set forth by the ASME Boiler and Pressure Vessel Code.

The original Design Report and previous amendments, in conjunction with this amendment, reaffirm the structural integrity of the components in accordance with the 1968 Edition of Section III of the ASME Boiler and Pressure Vessel Code, with Addenda through Summer 1970.

The calculation results from previous sections include detailed analysis of fatigue considerations under normal operation and faulted conditions.


Question: Strontium-90 analysis in nuclear regulatory documents

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


Question: Regulatory requirements for water withdrawal permits are essential for maintaining compliance with environmental standards.

The NRC requires that all licensees conduct thorough reviews of their water use to ensure adherence to safety and environmental regulations.

The Bridgeton Water Dept. has multiple wells with varying pump rates ranging from 25 to 800 gallons per minute.


Question: The NRC mandates the monitoring of piping systems to ensure the integrity of pressure boundaries in reactors.

Loss of material due to general, pitting, crevice corrosion, MIC is an issue requiring management from the piping components that serve as a pressure boundary.

The aging management program must address various environmental factors that contribute to the degradation of reactor components.


Question: What are the key actions involved in recovery operations and reentry after an incident?

Call the State EOC and inform them that the Recovery Organization is prepared to implement limited or unlimited reentry and/or any assistance from outside sources that may be required.

Ensure that all necessary Parish agencies and emergency personnel are prepared to accomplish the reentry operations when allowed.


Question: What does the Pressurizer Level Control System manage in terms of reactor operations?

The Pressurizer Level Control System manages the de-energizing of pressurizer heaters and responds to high pressurizer level reactor trips and low level letdown isolation.

Unit 1 is operating at full capacity, ensuring that all systems are functioning within the prescribed limits for safety and efficiency.


Question: Americium-241 analysis in nuclear safety


Question: Local Intense Precipitation evaluation

Local Intense Precipitation (LIP) is a critical factor in assessing flood risks for nuclear facilities.

The drainage areas for flood hazard evaluation are organized into zones, with Zone A covering an area of 2.61 acres.


Question: What is the significance of potentiometric surface elevation in groundwater monitoring?

KD Potentiometric Surface Elevations (ft amsl) are crucial in assessing the groundwater conditions and flow dynamics within the monitoring area.

The graphic scale indicates 0, 2,500, and 5,000 feet, which helps in measuring the distance accurately on the map.


Question: What is FWHM in peak analysis of nuclear data?

FWHM (Full Width at Half Maximum) is defined as the width of a peak at half of its maximum height, and it is an important parameter in peak analysis as it indicates the resolution of the detector.

The Peak Analysis report lists several detectors and provides specific energy readings along with their corresponding net peak areas and uncertainty values.


Question: What is non-11e.(2) byproduct material in nuclear regulation?

As determined by NRC Staff on several previous occasions, these waste streams are determined to be non-11e.(2) byproduct material as TENORM and is not subject to off-site disposal requirements as 11e.(2) byproduct material.

Waste generated during the pre-operational phase of the CUP, including but not limited to well field and CCP construction, will include well cuttings and other solid wastes from site construction.


Question: What is the significance of the cultural resources report in nuclear licensing?

The NRC staff invites you to submit any comments that you may have on the enclosed cultural resources report, the NRC staff’s finding of Adverse Effect, and the development of an MOA.


Question: service water piping repair

This regulatory audit is conducted to support the license request for proposed changes to the license and technical specifications to facilitate service water piping repair.

The regulatory filing process requires thorough documentation and adherence to specified guidelines for all submissions.


Question: The configuration of the Non-Mixing GRID features specific axial geometry measurements vital for understanding heat flux behavior.

The NMGIRD, or Non-Mixing GRID, is an essential element in the COBRA-FLX model that influences critical heat flux calculations by defining the axial geometry.

The data table above presents various measurements vital for analyzing the thermal hydraulic performance in reactor systems.


Question: gamma spectrum analysis methodology

Gamma spectrum analysis was performed using a 500 ml Marinelli geometry detector, with specific calibration done on 7/8/2014 to ensure accurate energy identification and peak analysis.

The sample size taken at the Dairyland Nuclear Power Plant was 1.263E+03 grams and it was acquired on 5/20/2019.


Question: Reactor Protection System failure modes

h. Preset: M:RP01BB, Reactor Protection System Manual Scram Failure ‘B’

The plant is operating at approximately 80% power.


Question: What is the maximum short-term rate of withdrawal for demineralized water systems?

The maximum short-term rate of withdrawal from the water source for the demineralized water system is 720 gpm.

The average rate of withdrawal from the water source for the fire protection water system is 675,000 gal/mo.


Question: What is the significance of ITAAC notifications in nuclear regulatory filing?

The ITAAC notification is a formal communication delineating the completion of Inspections, Tests, Analyses, and Acceptance Criteria related to nuclear construction activities and compliance to regulatory requirements.

The NRC regularly reviews construction inspection reports to ensure compliance with federal safety standards.


Question: The NRC's guidance specifies that licensed cladding materials taking into account known embrittlement mechanisms should not face heightened licensing demands.

Licensed alloys that considered known embrittlement mechanisms should not require additional licensing actions.

Effect on risk due to changes at sites are assessed based on RG-1.174.


Question: What is the role of a reactor core in nuclear energy production?

The reactor core is the part of a nuclear reactor where the nuclear reactions occur, containing the fuel assemblies and neutron moderators used to sustain the chain reaction.

The Westinghouse Electric Company provides a variety of services including engineering, procurement, and construction support for nuclear power plants.


Question: What are the requirements for the licensee's program under NRC regulations?

Except as specifically provided otherwise in this license, the licensee shall conduct its program in accordance with the statements, representations, and procedures contained in the documents, including any enclosures, listed below.

The U.S. Nuclear Regulatory Commission's regulations shall govern unless the statements, representations, and procedures in the licensee's application and correspondence impose on the licensee requirements that are more restrictive than or in addition to the regulations.


Question: What is the graphite material qualification program for a fluoride salt-cooled high-temperature reactor?

Kairos’ information contained in Enclosure 1 to this letter contains details of Kairos’ graphite material qualification program for a Kairos Power fluoride salt-cooled high-temperature reactor.

I am making this request for withholding, and executing this affidavit in support thereof, pursuant to the provisions of 10 CFR 2.390(b)(1).


Question: gamma spectrum analysis

Peak Analysis Performed on: 9/14/2017 2:36:38 PM
Peak Analysis From Channel: 100
Peak Analysis To Channel: 4096


Question: The density of helium varies across different types of reactors, specifically PWR and BWR types.

The NRC outlines various safety measures that must be adopted during the construction phase of new nuclear facilities.


Question: What is a Light Water Reactor (LWR)?

**LWR**: Light Water Reactor, a type of thermal reactor that uses ordinary water as both coolant and neutron moderator.

The NRC oversees the safety and licensing of nuclear facilities to ensure compliance with federal regulations.


Question: what are the enrichment limits for BWR fuel assemblies

Assembly Type Enrichment (wt% 235U) Maximum Planar Average


Question: protective action sections

Black circled numbers indicate Protective Action Sections. White lines separate Protective Action Sections.


Question: The NRC expects licensees to demonstrate compliance with a variety of safety and performance criteria to ensure public health and safety.

The COBRA-FLX correlation is used to predict the critical heat flux for various reactor designs under different operating conditions.

Test assembly conditions are crucial for validating the performance of various reactor components and ensuring their operational reliability.


Question: nuclear power reactor licensing process

The licensing process for nuclear power reactors includes multiple stages, such as pre-application consultations, formal application submissions, safety reviews, and public hearings conducted by the Nuclear Regulatory Commission (NRC).

The reactor's cooling system must function efficiently to maintain optimal operating temperatures and ensure safe energy production.


Question: average background radiation levels

The average background radiation level was recorded at 3,365 counts per minute (cpm) across the monitored locations.

The detector type identified in the data table is 44-10, which is used for radiation measurement.


Question: The control room registered an earthquake event confirmed by both seismic monitoring systems at the site.

An earthquake was felt in the Control Room and confirmed with both JAF and Nine Mile Point Unit 2 seismic instrumentation.

The following conditions exist now: - All Control Rods remain full in. - Reactor pressure is 0 psig.


Question: subsequent license renewal

This EIS includes the site-specific evaluation of the environmental impacts of the proposed action, Oconee Station subsequent license renewal (SLR), and alternatives to SLR.

This EIS considers the impacts of all SLR issues applicable to Oconee Station SLR on a site-specific basis.


Question: What are the functions of the MTP in testing the PPS?

The MTP facilitates testing the PPS to help diagnose errors announced by self-diagnostics. Testing includes exercising PPS outputs such as the scram solenoid valves.

A visual alert is generated on the SD when the keyswitch is in the Enable position.


Question: What is the purpose of the Operating Experience Program in nuclear licensing?

The existing program is credited and will be enhanced as follows: 1. An evaluation of operating experience at extended power uprate (EPU) levels will be performed to ensure that operating experience at EPU levels is properly addressed by the aging management programs.

The License Amendment Requests will be submitted to the NRC no later than 12 months prior to entering the subsequent period of extended operation.


Question: Federal Radiological Monitoring and Assessment Center

The FRMAC Operations Plan provides for the coordinated management of Federal technical response activities related to a radiological emergency. Its primary goals include: Assisting the State and the Federal Coordinating Agency with personnel, equipment, and technical resources, as needed; Collecting offsite environmental radiological data; and providing the data and related assessments to involved State agencies and to the Federal Coordinating Agency.

The above localities have a system to call back to the power station and check the message.


Question: CFRP cure rate and CCHX function

Further testing may be utilized to justify a lower percentage cure that precludes adverse impacts to the CCHX function to support restoration as deemed necessary.

Additionally, Page 20 states: In the event the normal SW flow path is required to be restored and is used as the primary flow path to the CCHXs, additional inspection of the CFRP liner would be required to ensure the installation was not adversely impacted.


Question: What is neutron absorber material in nuclear context?

Neutron absorber material is used in nuclear reactors to help control the fission process by capturing free neutrons, often enhancing safety and efficiency.

Limited strategic planning for harvesting has hindered opportunities to improve resource allocation in various projects.


Question: What are the license numbers for the Diablo Canyon Power Plant?

Mr. Peter Wells serves as the Site Vice President for the NextEra Energy Duane Arnold Energy Center located in Palo, IA.


Question: gamma spectrum analysis

Gamma spectrum analysis is performed using a 500ml Marinelli geometry to identify radionuclides present in the sample based on their characteristic energy peaks.

The operator for the analysis was the Administrator who logged the sample acquisition times and specifications for quality control.


Question: What is the purpose of the Fire Water System in nuclear facilities?

Table XI-01 of Volume 2 of NUREG-2191 states, in part, “This program manages loss of material, cracking, and flow blockage due to fouling by conducting periodic visual inspections, tests, and flushes performed in accordance with the 2011 Edition of NFPA 25.”

Given that the language in the procedure is different from the enhancement and basis document, it is unclear what the distance criteria is.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors is a comprehensive procedure that includes safety evaluations, environmental reviews, and public participation, overseen by the NRC.

The data provided in Table 6-125 outlines the various components of the reactor core, including the fuel assemblies and control rods.


Question: The evaluation of Local Intense Precipitation (LIP) includes analyzing areas with limited water flow.

Local Intense Precipitation (LIP) is a key factor in the assessment of flooding potential for plant safety evaluations.

The regulatory filing also includes detailed information about the various types of reactors used in the industry.


Question: What are the permissible uses for yttrium-90 according to NRC regulations?

For medical use as permitted by 10 CFR 35.1000 in a Sirtex Medical Limited brachytherapy afterloader delivery system.

Chemicals can exist in various physical forms, including solids, liquids, and gases.


Question: What is the significance of the vicinity roads in regulatory filings for nuclear facilities?

Figure 3-2: Roads in the Vicinity of the HCS provides a visual representation of the roads surrounding the Horizon Center Site Centerpoint, which is critical for assessing potential transportation impacts related to the proposed projects.

The environmental report discusses the effect of climate variables on local wildlife, emphasizing the unique species found near the facility.


Question: transportation impacts during nuclear plant construction

Transportation impacts during the construction of a replacement nuclear power plant would consist of commuting workers and truck deliveries of equipment and material to the construction site.

Socioeconomic impacts from construction for the new nuclear alternative would include those discussed for all replacement power alternatives in Section 3.10.9.1.


Question: The NRC requires detailed evaluations to substantiate the core operating limits in the context of loss of coolant accidents.

Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits (Proprietary)

James E. Holloway, Vice President - Nuclear Engineering and Fleet Support, affirms the authenticity of the document for Dominion Energy Nuclear Connecticut, Inc.


Question: What are the design bases documents for nuclear components?

Electronic copies of the design bases documents (or system descriptions) for selected components and modifications.

A copy of any internal/external self-assessments and associated corrective action documents generated in preparation for this inspection.


Question: What is the purpose of NRC's Regulatory Issue Summary on security-related sensitive information?

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

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


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code provides a critical heat flux correlation that accounts for various local conditions and flow regimes, optimizing the safety and efficiency of reactor operations during transient scenarios.

The report also discusses the grid tests conducted under controlled conditions to validate the simulation results, focusing on the accuracy of modeling and the reliability of data acquisition systems.


Question: What are climatological normals used for in nuclear site assessment?

Table 3 presents the climatological normals for Knoxville, TN, which are utilized in evaluating environmental conditions for the proposed site.

The NRC requires applicants to demonstrate that their facility complies with all safety regulations regarding structural integrity and equipment reliability.


Question: nuclear regulatory filings

Source Production and Equipment Co. has submitted various specifications such as LA-0612-S-105-S, which are necessary for evaluating applications for nuclear facility licensing.

The weather conditions today are sunny with a light breeze, making it a perfect day for outdoor activities.


Question: The minimum cooling time for a certain assembly burnup is an important factor in nuclear reactor fuel management.

Minimum Cooling Time (years) referenced in the loading table is a critical parameter for determining the suitable conditions for spent fuel storage and management in nuclear reactors.

Nuclear power plants annually undergo thorough inspections to ensure compliance with safety protocols and operational standards.


Question: What is the activity level of CE-144 in pCi/g?

The activity of CE-144 is 1.62E-01 pCi/g.

The energy level of EU-152 is 121.78 keV.


Question: What is the purpose of Fasken New Contention 5?

Fasken New Contention 5 states: ISP’s application fails to adequately, accurately, completely and consistently consider the cumulative impacts of transporting high-level radioactive waste and spent nuclear fuel to and the socioeconomic benefits of the proposed ... project, which precludes a proper analysis under NEPA, and further nullifies ISP’s ability to satisfy NRC’s siting evaluation factors now and anticipated in the future and is in further violation of NRC regulations.

In CLI-20-14, the Commission also considered and denied the first two submissions that Fasken had filed after the Board denied its hearing request.


Question: LPCI AC Input Breaker

Trigger 3, R:ED23, LPCI AC Input Bkr ‘A’ (71-INV-3A), Final=open

The Plant is operating at approximately 5% power during a startup.


Question: How can communities ensure regulatory oversight of very low-level nuclear waste disposal?

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

The whole proposal is a secretive deregulation without public notice or opportunity for individuals or communities to intervene when their landfill or other facility is applying or approved as "specific exempt."


Question: What is the significance of the Pickens-Gilberttown and Southern Arkansas fault zones in seismic studies?

The Pickens-Gilberttown and Southern Arkansas fault zones were incorporated within background seismic source zones across the Gulf Coastal Plain gulfward of the Ouachita Orogenic Belt.

The six EPRI teams developed estimates of maximum earthquake magnitudes for the area encompassing the Pickens-Gilberttown and Southern Arkansas fault zones (Gulf Coastal Plain south of Ouachita Orogenic Belt) that range from mb 4.6 to 7.2 (Mw 4.2 to 7.5).


Question: The NRC specifies the cooling time requirements for nuclear fuel assemblies post-irradiation.

Minimum Cooling Time (years) can vary significantly based on the average enrichment of the fuel assemblies used in a PWR system, with specific guidelines established by regulatory standards.

The reactor pressure vessel is designed to withstand high pressures generated during normal operation, ensuring that safety limits are not exceeded.


Question: How does SNC evaluate suppliers of safety-related items?

Prospective suppliers of safety-related items and services are evaluated to assure that only qualified suppliers are used.

Verification actions include testing, as appropriate, during design, fabrication, and construction activities.


Question: What are the requirements for calculating core damage frequency in nuclear safety assessments?

The estimated core damage frequency (CDF) provides a quantitative measure of the likelihood of a significant nuclear accident and is essential for assessing the safety performance of nuclear facilities.

The installation of new solar panels in residential areas has increased the interest in renewable energy sources among homeowners.


Question: The NRC mandates the submission of ITAAC notifications to ensure compliance during construction phases.

The ITAAC Notification is a requirement imposed by the NRC to maintain oversight during the construction of nuclear facilities.

The construction of a nuclear facility involves various stages, including design, procurement, and assembly.


Question: The NRC discusses the modeling of isotopic composition in reactor fuel pins during operation and depletion processes.

Pin specific isotopic modeling as a function of exposure is performed based on the lattice physics code TGBLA06.

The fuel models considered include 2D geometric modeling of all fuel material, cladding, water rods, and channels.


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

DNBR, or Departure from Nucleate Boiling Ratio, is a critical parameter in assessing the safety of a nuclear reactor core, as it indicates the thermal margin available to prevent the onset of boiling crisis conditions.

The reported temperatures showed a steady decline from 620°F to 590°F over a period of 300 seconds, highlighting the cooling efficiency of the system during operation.


Question: The licensing process for nuclear reactors involves several key regulatory steps.

The NRC must review and approve a Nuclear Power Plant License Application, ensuring compliance with safety regulations and standards before any construction begins.

The regional administrator oversees inspections and compliance but does not involve directly in the licensing processes.


Question: The commission expects licensees to establish comprehensive emergency preparedness programs for their reactors.

The NRC requires licensees to implement emergency preparedness plans that are tailored to the specific characteristics of the reactor design and site.

The report discusses various reactor technologies and their respective benefits to power generation without focusing on emergency preparedness.


Question: maximum enrichment in nuclear fuel assemblies


Question: The NRC evaluates whether the submission meets established ITAAC requirements...

The Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) defined in the combined license specify the necessary conditions to be met for the construction of the nuclear power plant.

The vendor inspections ensure that manufacturing processes align with the design specifications, maintaining quality throughout production.


Question: nuclear licensing requirements for foreign end users

The document outlines the ultimate consignees in foreign countries for non-nuclear end use in the medical, pharmaceutical, chemical, and industrial markets, specifying that licensing is required for these transactions.

The Republic of Korea is known for its advancements in technology, particularly in organic light-emitting diodes production.


Question: What are the fabrication steps for ice baskets used in nuclear safety systems?

The fabrication steps are as follows: 1. The sheet metal is purchased in the hot-rolled and pickled condition. 2. The perforator oils and perforates the material and ships to the basket fabricator. 3. The basket fabricator rolls the perforated metal into a cylindrical shape 12 inches in diameter by 141.57 or 142.35 inches long (for bottom basket or upper basket respectively) and material is degreased. 4. The sides of the rolled cylinder are continuously welded using the gas metal arc process. 5. Following the welding, the cylinder is pickled, washed, fluxed, hot dip galvanized, and dipped in a sodium dichromate bath. 6. The couplings and stiffening ring blanks are cut from sheets or coils or hot rolled, pickled and oiled material. These are formed by a rolling process and are 3 inches high with a roll-formed internal lip and are of a diameter to fit inside the perforated basket. 7. The cruciforms are die-formed from steel strip. 8. Following the forming operations, stiffeners and couplings with cruciforms in place are pickled, washed, fluxed, hot dip galvanized, and dipped in a sodium dichromate bath. Replacement cruciforms are made of stainless steel and will not be subjected to this treatment. 9. The column bottom is fabricated by a procedure similar to item (6) above. The appurtenances are welded in place and the piece is galvanized per item (8) above. 10. The remaining appurtenances are cut to size, machined, welded, where required followed by galvanizing as above, and plated where required. 11. The completed couplings, bottoms, appurtenances, stiffening rings and cylinders are next assembled. The stiffening rings are inserted inside the cylinder until the side is adjacent to the 2.5 or 2.75 inch (for bottom or upper basket, respectively) perforated area in the center of the cylinder and attached by a sheet metal screw and four basket detents. 12. For the column bottom, two U-bolts and nuts and washers fasten the mounting bracket assembly to the plate of the basket end. Swivel Bracket Assemblies may be substituted for the U-bolt assemblies.

Interconnection couplings and stiffening rings are located at the bottom and 6 ft. levels respectively of each basket section. The bottom coupling and stiffening ring are cylindrical in shape and approximately 3 inches high with a rolled internal lip. The lip provides stiffening to the basket. These couplings are attached to the ice basket by sheet metal screws and basket detents.


Question: What does net counts per minute (cpm) measure in nuclear safety assessments?

Net cpm is a measurement used to quantify the rate of radiation detected by sensors, expressed as counts per minute, which is crucial in evaluating the safety and operational performance of nuclear facilities.

The proposed new system aims to improve efficiency and reduce operational costs across all divisions, fostering innovation and streamlining processes.


Question: The cumulative usage factor (CUF) is critical to evaluate the potential for fatigue damage in reactor components.

The program verifies the continued acceptability of existing fatigue analyses through transient cycle counting and calculation of cumulative CUF and CUFen values to demonstrate that they continue to meet the appropriate limits.

The program utilizes the *FatiguePro™* software which is a computerized data acquisition, recording, and tracking program.


Question: emergency diesel generator performance specifications

Southern Nuclear Operating Company (SNC) requests an amendment to the Technical Specifications (TS) for Vogtle Electric Generating Plant (VEGP) Units 1 and 2 to revise emergency diesel generator (EDG or DG) frequency and voltage ranges for TS 3.8.1, AC Sources – Operating, Surveillance Requirements (SR). These changes reflect incorporation of Technical Specification Task Force (TSTF) traveler TSTF-163, Revision 2, "Minimum vs. Steady State Voltage and Frequency" (ML040500733)(Reference 1), as well as changes to the steady state DG operating voltage and frequency ranges consistent with WCAP-17308-NP-A, "Treatment of Diesel Generator (DG) Technical Specification Frequency and Voltage Tolerances," (ML17215A232) (Reference 2) methodology.

The onsite standby power source for each 4.16 kV ESF bus is a dedicated DG. DGs A and B are dedicated to ESF buses A and B, respectively.


Question: What materials are used for fire barriers in nuclear facilities?

Fire Barriers (Penetration Seals and Fire Stops) consist of materials such as Grout, Gypsum, and Silicates, designed to function as fire barriers in indoor, uncontrolled environments.

The Pressure Boundary is typically made from Cast Iron and is found in outdoor, external environments.


Question: What is the maximum differential pressure in nuclear reactor operations?


Question: What is the role of soil analysis in nuclear reactor safety assessments?

The analysis of soil characteristics is essential for ensuring the stability and safety of nuclear reactor foundations, as highlighted in the regulatory requirements set forth by the NRC.

Recent studies have shown that the cooling systems in advanced reactors significantly improve thermal efficiency, thereby enhancing overall reactor performance.


Question: What are the predominant phytoplankton species found near the NFCS discharge?

Phytoplankton collected in the Congaree River in the vicinity of the NFCS discharge in October 1974 were predominately the colonial green algae *Eudorina elegans* (Appendix 2.G, Table 2.G-11). Diatoms, blue-greens and euglenoids were also present. Average cell numbers in the river were approximately 500 cells per milliliter.

Chironomids (*Chironomini* sp), annelid worms (*Limnodrilus udekemianus* and *Naididae* sp) and corbiculoid clams (*Corbicula fluminea*) were also frequently encountered.


Question: low pressure emergency core cooling system

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

The operator will promptly enter T-101 (Reference 14) upon recognition of any of the following entry conditions:


Question: What is the Cathcart-Pawel equation used for in nuclear safety?

Clearly, the Cathcart-Pawel equation is an equation for predicting the oxidation kinetics of Zircaloy specimens in furnaces; it is not adequate for predicting the oxidation kinetics of Zircaloy bundles in a nuclear power plant core in the event of a LOCA.

It seems to me—when the consequences of failure are serious—then the ability of the codes to arrive at a conservative prediction must be verified in experiments of complexity and scale approaching those of the system being calculated.


Question: Petitioners must file their contentions, including expert opinions and supporting references, within a specified timeframe to ensure their participation in the proceedings.

The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact.

The Commission will make a final determination on the issue of no significant hazards consideration.


Question: Utilities must manage their relationships with fuel vendors to meet the unique needs of their respective plants.

Each plant has its own association with a vendor or maybe some do it themselves. Methods are unique.

This is Bob Martin. When it comes to what's limiting, it's always been a bit of a whack-a-mole with methods.


Question: RBCLC temperature control mechanisms

Further control of RBCLC temperature is performed by throttling further open the RBCLC heat exchanger Service Water outlet valves to raise Service Water flow through the heat exchanger.

The only automatic temperature control mechanism, 15TCV-101, is already in the maximum cooling position.


Question: nuclear regulatory certification

Nuclear regulatory certifications are established procedures that ensure compliance with safety and operational standards as defined by regulatory authorities.

The certification process for drinking water quality in states like New Jersey involves regular testing and reporting of water samples to ensure safety standards are met.


Question: maximum TSC transfer time

The time duration from the end of TSC annulus cooling, either by 24 hours in the pool or by the annulus circulating water system, through completion of vacuum drying and helium backfill shall not exceed the following:


Question: what is the Reactor Trip System

The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents. The RTS and related instrumentation are identified in TS Table 3.3.1-1.

Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants. The changes will not impact how the plants operate.


Question: What is a Marinelli sample in nuclear analysis?


Question: The assessments related to qualitative considerations are made by the IDP following the guidance of Section 9.2.

The assessments of the qualitative considerations are agreed upon by the IDP in accordance with Section 9.2.

The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team.


Question: The assessment of the Unit 1 vessel welds indicates they meet the criteria set forth by Equivalent Margin Analysis.

Therefore, Unit 1 vessel welds are bounded by Equivalent Margin Analysis.

The percentage decrease measurements for the capsules show varying degrees of loss in strength, with the 4th capsule exhibiting the largest decrease.


Question: archaeological site inventory

Illinois Inventory of Archaeological Sites within 6 Miles of DNPS includes various site types such as unassigned lithic scatters, artifact scatters, and features that are either eligible or recommended for further testing based on their NRHP status.

The limestone foundation and limestone lined well feature were noted but do not have associated eligibility criteria or recommendations for further action.


Question: water quality data evaluation

Water quality data will be evaluated for reasonableness and agreement with previous analyses by the analyst, and the Environmental Manager in accordance with the procedure outlined in Section 9.12.11.3.

The reviewer will verify that the detection limits are 10% or less than the appropriate values listed within the Tables in 10 CFR 20, Appendix B.


Question: What is the role of the Closed Treated Water Systems aging management program in nuclear safety?

The Closed Treated Water Systems aging management program will be enhanced as follows: 1. Revise implementing procedures to manage the effects of cracking due to stress corrosion cracking in closed treated water systems.

If fouling is identified, the overall effect is evaluated for reduction of heat transfer, flow blockage, and loss of material.


Question: What is the significance of the Design Basis Accident in nuclear safety?

The Design Basis Accident (DBA) represents the range of accidents that a nuclear facility is required to be designed to withstand, ensuring that public health and safety are maintained under credible accident scenarios.

The cooling system operates by circulating water through the reactor core and then passing it through heat exchangers to remove excess heat from the system.


Question: What is the role of monitoring for radiopharmaceutical extravasation in nuclear medicine?

We do not monitor for radiopharmaceutical extravasation. Requiring that for every patient would add significant time to the patient imaging.

These letters were from North Carolina Senator Tillis and Representatives Price, Holding, and Butterfield.