Question: The safety systems in nuclear reactors must comply with NRC regulations to ensure effective performance during emergencies.

The NRC requires that nuclear facility designs incorporate various safety systems that are capable of performing their intended functions under postulated accident conditions.

Windows is either a registered trademark or a trademark of Microsoft Corporation in the United States and/or other countries.


Question: critical heat flux correlation

COBRA-FLX is a critical heat flux correlation used to predict the thermal performance of nuclear reactors.

Figure A-45 illustrates the axial geometry utilized in the CHF test K9300.


Question: The Nuclear Regulatory Commission oversees the licensing processes for nuclear facilities through various technical evaluations.

The Turkey Point Units 3 and 4 Application for Renewed Operating Licenses, prepared by Florida Power & Light Company, outlines the necessary documentation and assessments required for the renewal of these licenses.

FDEP's 2021 Conditions of Certification specifies the operational requirements for the Turkey Point Plant's environmental compliance, focusing on emissions and water quality standards unrelated to the licensing process.


Question: The NRC stipulates that effective aging management must be implemented for components to ensure their safety.

Aging Management Programs are critical for evaluating and mitigating the effects of aging on plant systems and components, as identified in NUREG-2191 and NUREG-2192.

The environment for the valve body in this system includes treated water and various gaseous states which can lead to material loss.


Question: risk-informed and performance-based concepts

the NRC’s efforts to establish a modern regulatory framework, including the proposed Part 53 and Subpart H, do not fully reflect risk-informed and performance-based (RIPB) concepts as they are defined in long-standing Commission policy.

To enable the safe and secure civilian use of nuclear energy and technologies, the NRC must break from conventional thinking and practices and deliver an efficient and effective regulatory framework.


Question: nuclear safety analysis

The nuclear safety analysis encompasses a systematic evaluation of potential hazards associated with the operation of nuclear reactors, ensuring that all safety measures are adequate to protect public health and the environment.

The Virginia Earthquake of April 9, 1918, was recorded in the Bulletin of the Seismological Society of America, detailing its impact on various structures and the historical context of seismic events in the region.


Question: This affidavit certifies that the claims made therein are accurate to the signer's knowledge and belief.

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.


Question: The NRC evaluates the accuracy of nuclide confidence indices to ensure safe operations.

Nuclide confidence index threshold = 0.30.


Question: What is the function of the bottom-mounted instrumentation system in reactors?

Bottom-mounted instrumentation system: bottom-mounted instrumentation (BMI) column bodies.

Loss of material due to wear; loss of fracture toughness due to thermal aging embrittlement (CASS only).


Question: deficiency letter in NRC regulatory process

In the deficiency letter, the reviewers should specify what the specific deficiencies are and what additional information is needed to complete the NRC’s review.

Prior to contacting the specialty board, the reviewers should brief the Medical Radiation Safety Team (MRST) Leader to ensure that follow-up is approved.


Question: pressurizer safety valve operation


Question: RCS flow measurement

Is RCS system running? Yes: Use all RCS. No: Step 6: Is SI (Safety Injection) in service? If yes, proceed to step 9. If no, proceed from step 10.

Check the isolation valves. If not isolated -> Exit Procedure.


Question: What is the significance of uranium measurement in water quality assessments?

Uranium measured in mg/L is crucial in evaluating water quality, as it can indicate potential contamination and health risks associated with exposure to radioactive materials.

The Crownpoint Project Area requires extensive monitoring of various constituents to ensure compliance with environmental standards and regulations.


Question: The regulatory framework includes rigorous assessments of nuclear facilities' safety protocols.

These analytical reports serve to support the conclusions drawn by the licensee regarding the adequacy of the proposed safety measures.

GEL Laboratories is tasked with providing sample analysis and technical support for various environmental assessments.


Question: What is the function of a diesel generator starting air control valve?

DIESEL GENERATOR STARTING AIR CONTROL VALVE is responsible for regulating the flow of starting air to the diesel engine to ensure proper initiation of the operation.

The P&ID for the system must clearly indicate the location of active and passive components to ensure compliance with regulatory standards.


Question: temperature differences in BWR control blades during LOCA

The petitioner also states that current BWR components (control blades) would be damaged if the cladding reaches a temperature between 1832°F (1000°C) and 2192°F (1200°C) (page 65 of PRM 50-95). The petitioner’s basis for this statement is based upon the melting reaction between B4C and stainless steel beginning at approximately 1832°F (1000°C) and accelerating above 2192°F (1200°C). LOCA licensing calculations indicate that when the 1832°F (1000°C) cladding temperature is reached, the temperatures in the control blades are at least 392°F (200°C) lower.

NEI discusses the temperature differences between the BWR cruciform control blades and the fuel cladding in the CORA-16 experiment.


Question: What are Emergency Action Levels (EALs) in nuclear safety protocols?

FPL has updated its EAL scheme to the latest NRC endorsed scheme published in NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6. The PTN EALs incorporate the new guidance that has simplified the classification process, including the use of a matrix of EAL initiating conditions that streamlines the process of evaluating EALs against plant conditions.

The overall improvements in technology and information availability over the years have enabled the on-shift staff to assess plant conditions quickly and efficiently, and with less distraction than before. The computational power of modern processors allows for calculation of dose projections that take seconds rather than minutes.


Question: nonroad emission factors

Table 2 lists the nonroad 2008 emission factors for various types of equipment including THC, NOx, CO, PM10, SO2, and CO2 emissions for different machinery used in construction and other nonroad activities.

The hydraulic crane operates by lifting and moving heavy objects using hydraulic power, enabling it to complete tasks efficiently and effectively.


Question: Power-Range High Neutron Flux Trip

Power-Range High Neutron Flux Trip refers to the condition where the neutron flux in the reactor core exceeds a preset high level, prompting automatic safety measures which are critical to maintaining operational safety.

Maintaining consistency in power output requires close monitoring of the reactor's thermal performance to ensure that operational parameters remain within specified limits.


Question: What is involved in a radiological hazard assessment for nuclear facilities?

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

The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:


Question: Thermal Barrier Booster Pump operation

The Unit 1 NUSO has directed you to place Thermal Barrier Booster Pump 1B in service and place Thermal Barrier Booster Pump 1A in standby in accordance with 0-SOI-70.01, Section 8.7, Alternating Unit 1 Thermal Barrier Booster Pumps.


Question: What is riser hole mass flow rate in nuclear reactors?

Figure 18 shows the riser hole mass flow rate and vapor condensation in relation to power output, with all cases utilizing the Clark Void Fraction Model and assuming a total reactor coolant system mass of 15,000 kg.

The NRC emphasizes the importance of regular inspections and maintenance to ensure the safety and integrity of nuclear reactors, though these processes may vary by facility.


Question: The NRC mandates a thorough safety assessment during the licensing of new nuclear reactors.

A safety evaluation shall be performed to ensure that the design and operation of the nuclear reactor meet the regulatory standards set forth by the NRC.

In the inspection process, the personnel must verify compliance with environmental regulations and unrelated safety codes.


Question: The process of pursuing a license from the NRC involves several regulatory requirements...

The Nuclear Regulatory Commission (NRC) requires applicants to submit a comprehensive Technical Safety Review as part of the licensing process.

The heat exchangers in the cooling system are designed to maintain optimal temperatures during reactor operations.


Question: What is a safety system in nuclear reactor regulation?

The Division of Safety Systems is responsible for the oversight of safety-related systems that ensure the safe operation of nuclear reactors.

Please see attached word document that contains a list of items we’d like to discuss with PNNL.


Question: What are the actions prioritized in the event of lost RPV water level indication?

The QGA 100 Level leg override states that, if RPV water level is unknown, (which is the case when all RPV water level instrument reference legs have flashed), then exit this procedure and enter QGA 500-4.

Plausible because this is a QGA 500-4 action. Incorrect because it applies to when the reactor is not shut down. The transition would have been from QGA 101 and not QGA 100.


Question: remotely operated valves associated with piping penetrations

STONE & WEBSTER-PRODUCED REMOTELY OPERATED VALVES ASSOCIATED WITH PIPING PENETRATIONS include various valves such as the RHR Sample Containment Isolation Valve and Hydrogen analyzer suctions, which are critical for system operations.

Piping penetrations and their associated valves play a crucial role in maintaining the integrity of nuclear systems, ensuring that water and air systems operate efficiently without contamination.


Question: The regulations specify the standards for the design of components such as pumps and valves used in nuclear facilities.

Residual heat removal pumps are classified under ASME, a Class II standard, ensuring they meet safety and operational guidelines for nuclear applications.

The shell side of a residual heat exchanger is governed by ASME, Section VIII c, which provides requirements for pressure vessels in general.


Question: regulatory approval to dispose of very low-level waste

the NRC expressly and unequivocally explained in the 2016 Issue Summary that the interpretation it had announced in 1986 was being "supersede[d]" because, in the agency’s view, "it did not provide the correct information regarding regulatory approval to dispose of very low-level waste"

And, as noted supra note 4, the NRC had actually considered numerous Section 20.2002 requests from reactor licensees in Agreement States prior to 2012.


Question: What is the significance of Radionuclides in nuclear regulatory filings?

| Atomic No. | Radionuclide  | Class                   | Col. 1 Oral Ingestion ALCI (µCi) | Col. 2 Inhalation ALCI (µCi) | Col. 3 DAC Air (µCi/ml) | Col. 1 Water (µCi/ml) | Monthly Average Concentration (µCi/ml) |

The NRC periodically reviews the environmental monitoring data collected from nuclear facilities to ensure compliance with regulatory standards.


Question: What does the acronym pCi/g stand for in nuclear regulatory contexts?

The unit of measure for radioactivity in a sample is often given in picocuries per gram (pCi/g), which indicates the radioactive decay rate per gram of the material.

The data for each sample is accompanied by the analysis date and time noted in the report, ensuring traceability and accountability in the testing process.


Question: Reactor Protection System instrumentation requirements

LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

Separate Condition entry is allowed for each channel.


Question: Agencies such as the Environmental Protection Agency are crucial for monitoring environmental health around nuclear facilities.

The National Oceanic and Atmospheric Administration obtains precipitation data through a nationwide network of rainfall gauges.

The U.S. Department of the Interior Geological Survey regularly provides information on geological features across the country.


Question: What is the aging management program for reactor vessel nozzles?

Aging management programs for reactor vessel nozzles include inspections such as One-Time Inspection (B.2.1.20) and assessments of water chemistry (B.2.1.2) to mitigate issues caused by cracking due to SCC and pitting corrosion.

The NRC utilizes a range of methodologies to analyze seismic risks associated with the structural integrity of nuclear facilities during extreme events.


Question: intended functions of nuclear components

The intended functions that the components and structures must fulfill are those functions that are the bases for including them within the scope of subsequent license renewal. A component intended function is defined as a passive component function that must be performed in order for the system or structure to be able to perform the system or structure intended function(s).

Components and commodities which support or interface with electrical components and commodities, for example, cable trays, conduits, instrument racks, panels and enclosures, are evaluated as structural components in Section 2.4.


Question: The non-mixing grid (NMGRID) is a critical component designed to maintain uniform coolant flow across the reactor core.

The distance measurements are critical for determining the layout and organization of reactor components.


Question: What is the role of unconsolidated-undrained compression tests in nuclear facility safety assessments?

The unconsolidated-undrained compression test assesses soil behavior under rapid loading conditions, which is critical for evaluating the stability of structures in nuclear facilities.

The depth and elevation measurements for the soil samples are crucial in determining the geotechnical properties of the site.


Question: The NRC mandates that licensees submit a detailed plan for terminating their licenses, which includes a license termination plan for approval.

In accordance with 10 CFR 50.82(a)(9), “[a]ll power reactor licensees must submit an application for termination of license. The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval.”


Question: fracture toughness and its significance in reactor components

The NRC staff finds the fracture toughness model acceptable for the requested SG nozzles at Vogtle.

The NRC staff observed that the temperature at the NWV and NIR of the FW nozzle remained high, approximately 500 °F or greater, even though the water temperature at the FW nozzle ranged from 32 °F to 430 °F.


Question: The NRC's assessments show that exposure levels can vary significantly depending on the area and depth of the agricultural activities.

In its sensitivity analyses at INL, DOE showed that dose increases by a factor of two when spreading area drops by a factor of two {i.e., from 2,200 to 1,100 m² [0.5 to 0.3 acres]}.

These well diameters are supported by the surveys. For a 15-cm [6-in] residential well, the well diameter is 25 cm [10 in] for the first 12 m [38 ft]; then 20 cm [8 in] down to about 152 m [500 ft]; then 15 cm [6 in] for the remainder.


Question: Primary Containment Control actions

Directs actions of QGA 200, Primary Containment Control when Drywell pressure is above 2.5 psig.

If started, stops injecting Boron by placing keylock switch to OFF.


Question: GAIA fuel assembly design

The GAIA fuel assembly design at Byron and Braidwood Units 1 and 2 is consistent with the dimensional attributes in Reference 1. It incorporates eleven (11) spacer grids, twenty-four (24) guide tubes (GT’s), one (1) instrument tube (IT), and top and bottom nozzles to provide the structural cage for 264 fuel rods.

The cylindrically shaped pellets are sintered to a nominal theoretical density (TD) up to [ ] and a diameter of 0.3225 inches.


Question: What is a topical report in the context of NRC regulatory filings?

PNNL staff to complete assessment of topical report and as-needed supporting documentation (if needed) and compile list of additional RAIs (if needed).

A list of timelines and deadlines for the review from the NRC RPS system paired alongside the milestones from the statement of work. This will be used to help manage workload and progress.


Question: The analysis of potential flooding events is crucial for ensuring the safe operation of the facility during extreme weather scenarios.

Analyzing the effects of dam breaches on flood hazards is essential for evaluating the site-specific risks associated with severe weather events.

The report includes various geotechnical studies and findings from surface water assessments and their implications for environmental safety.


Question: radioactive waste management

Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, & Transportation | Biennial | 0 | 8 | Not Applicable | Not Applicable - completion not required during this inspection cycle

Occupational Dose Assessment | Biennial | 0 | 4 | Not Applicable | Not Applicable - completion not required during this inspection cycle


Question: Control Element Assembly Calculators

The CEACS assess core conditions based on Control Element Assembly (CEA) configurations and input penalty factors to the Core Protection Calculators (CPCs)

With one CEAC inoperable, restore within 7 days


Question: The evaluation of non-piping breaks is a critical component in determining the LOCA frequencies as outlined in regulatory filings.

the LOCA is any sort of rupture of the primary coolant pressure valve. It doesn’t necessarily have to be in a pipe.

But it does need to be consistent with, you know, the code of regulations.


Question: quality assurance program in nuclear licensing

Quality Assurance Program Description NEDO-11209 Revision 9 Mapping to Revision 10 focuses on ensuring that all activities related to nuclear operations meet the required standards for safety, reliability, and compliance with regulatory requirements.

The table of contents was updated to reflect the changes made in the document, ensuring all sections are accurately represented for the reader.


Question: Americium-241 analysis


Question: nuclear license renewal procedures

3.2 Renewal in Entirety: This section outlines the process of renewing a nuclear license in full, ensuring compliance with regulatory requirements and maintaining the safety and operational integrity of the facility.

1.4 Hardcopy Files: Hardcopy files are physical documents that are maintained as part of the official records for the organization.


Question: The HVAC system in the Control Room is crucial for maintaining operational conditions during maintenance.

1/2B Control Room HVAC is to be placed in operation to support maintenance activities on the A Train.

Mechanical Maintenance is in standby to determine if refrigerant needs to be added.


Question: The NRC mandates that licensees perform quality control analyses for isotopes like Uranium-235 and Uranium-238.

The quality control for Uranium-233/234 was conducted with measurements reported in units of pCi/g, indicating the activity of the isotopes per gram of sample.

The liquid scintillation analysis is aimed at measuring the levels of Technetium-99 in various batches, with results expressed in pCi/g.


Question: What is the significance of Chi/Q calculations in nuclear safety?

Chi/Q calculations provide a measure of the concentration of radioactive materials in the air that can be expected at different distances from the source during an accident, expressed in units of seconds per cubic meter (sec/m³).

The Clinton Power Station has a designed output capacity of 1,065 megawatts electric (MWe), making it one of the significant contributors to the grid in the region.


Question: What is the significance of 10 CFR 54.21(b) in the context of nuclear facility licensing?

10 CFR 54.21(b) requires submitting an amendment each year following submittal of the LRA and at least three months before the scheduled completion of the NRC review identifying changes to the current licensing basis (CLB) that materially affect the content of the LRA, including the FSAR supplement.

In the Reference letter, Constellation Energy Generation, LLC (CEG) submitted a License Renewal Application (LRA) for Clinton Power Station, Unit 1 (CPS).


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves multiple stages, including pre-application, application review, and post-licensing activities, to ensure safety and compliance with regulatory standards set by the NRC.

The cooling system in a reactor plays a crucial role in maintaining optimal operating temperatures, thereby enhancing overall efficiency.


Question: What is the purpose of Regulatory Guide 1.174?

REGULATORY GUIDE 1.174 provides general guidance on applications that address changes to the licensing basis.

Examples of application-specific regulatory guidance documents include (1) RG 1.174 and NUREG-0800, Chapter 19.2, which provide general guidance on applications that address changes to the licensing basis.


Question: The evaluation of the impacts of proposed reactor changes is crucial for compliance.

The NRC assesses potential safety-related changes to reactor systems through a rigorous evaluation process that considers regulatory impacts and public safety.

Nuclear power plants generate electricity by using heat produced from nuclear fission, which is a different technology utilizing uranium and thorium as fuel.


Question: Nuclear licensees must submit a safety analysis as part of their regulatory requirements.

The NRC requires a comprehensive safety assessment as part of the nuclear licensing process to ensure safe operation of the reactor.

The ongoing maintenance of the reactor pressure vessel is vital for the longevity of the plant.


Question: average burnup enrichment limits

Table B2-13 PWR Loading Table – Low SNF Assembly Average Burnup Enrichment Limits lists the maximum assembly average burnup and the corresponding minimum assembly average initial enrichment for various heat loads per assembly.

Certification of Compliance encompasses approvals by the regulatory commission to ensure that certain operational and safety standards are met for nuclear facilities.


Question: What is the purpose of the Foreign Soils Permit in nuclear licensing?

The Foreign Soils Permit, identified by reference P330–15–00283 and P330-15–00253, is necessary for the importation of soil samples that may be used in environmental assessments or regulatory submissions.

California's certification for public water systems, identified as 29400, outlines the requirements and standards for maintaining safe drinking water across the state.


Question: construction inspection report

The ITAAC Deleted Per License Amendment 168 indicates that certain requirements are no longer applicable.


Question: What are the ASME Section III requirements for reactor vessel head energy requirements?

Changed Section 03.02.a.2.(a.) from an energy requirement of 75 ft-lbs. for the RV head to agree with ASME Section III, Division 1, Subsection NB-2331 requirements of at least 35 mils lateral expansion and not less than 50ft-lbs absorbed energy.

IP 71007 has been revised to improve effectiveness of this procedure based on feedback from the regions.


Question: The evaluation of reactor component integrity during maintenance requires careful analysis of stress distribution.

The study examines the primary stress on a reactor vessel closure flange when one stud is out of service, ensuring safety during operation.

The analysis identified various configurations for the reactor coolant system that influence thermal efficiency.


Question: What are alternative requirements according to 10 CFR 50.55a?

Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation.


Question: nuclide minimum detectable activity

Nuclide MDA (pCi/grams) is the minimum detectable activity of a particular nuclide in a sample, expressed in picocuries per gram.

Errors quoted at 2.000sigma indicate the statistical confidence level used in reporting the estimated values of the measurements.


Question: What are suppression chamber vacuum breakers in a reactor building?

One line with one or more reactor building- suppression chamber vacuum breakers inoperable for opening.

The structures, systems, and components (SSCs) are not modeled.


Question: What is the Chemistry Factor used in nuclear licensing?

10 CFR 50.61 and 10 CFR Part 50, Appendix G requires the use of ΔRTNDT using the Chemistry Factor (CF) and Fluence Factor (f) for adjustment while the TR methodology specifies the use ASTM E900-15 of To.

The NRC staff does not find this revision acceptable. The bases for which is provided in the revised text in Sections 5.1 and 5.2 of the SE.


Question: uncrewed aerial system incidents in nuclear facilities

At 05:28 CDT, two security officers identified an uncrewed aerial system (UAS) with red and green lights flying above the protected area (PA) near the north perimeter.

The UAS was seen with red and white running lights and traveling towards Route 54 and flying away from the plant.


Question: What is the function of the Engineered Safeguards Actuation System (ESAS) in reactor safety?

The ESAS initiates necessary safety systems, based on the values of selected unit Parameters, to protect against violating core design limits and to mitigate accidents.

Action B.1 to place the channel in trip is delayed in accordance with the RICT.


Question: What does environmental monitoring for licensing entail?

Environmental monitoring for Church Rock, Crownpoint and Unit 1 includes various sample types such as air radon gas, process fluids, and water from different sources, conducted at specified frequencies and analyzed for specific radionuclides.

The grab sample method for monitoring involves collecting a single sample at a specific point in time, which helps in varying the data collected from continuous monitoring methods.


Question: material aging management for nuclear systems

The Fuel Oil Chemistry (B.2.1.18) and One-Time Inspection (B.2.1.20) program will be used to manage loss of material of the aluminum alloy and stainless steel flexible connections, piping, piping components, strainers, and valve bodies exposed to fuel oil in the Diesel Generator and Auxiliaries System and Station Blackout Diesel Generator System.

Consistent with NUREG-2191 with exceptions. The Selective Leaching (B.2.1.21) program will be used to manage loss of material of the cast iron and copper alloy with 15% zinc or more fire hydrants, flow devices, heat exchanger components, ice melt gate, piping, piping components, pump casings, sprinkler heads, tanks, and valve bodies exposed to closed cycle cooling water, raw water, soil, treated water, and waste water in the Closed Cycle Cooling Water System, Compressed Air System, Control Room Ventilation System, Cooling Water Structures, Demineralized Water Makeup System, Diesel Generator and Auxiliaries System, Fire Protection System, Fuel Pool Cooling System, High Pressure Coolant Injection System, Non safety-Related Ventilation System, Open Cycle Cooling Water System, Process Sampling and Radiation Monitoring System, Radwaste System, Safety-Related Ventilation System, Shutdown Cooling System and Station Blackout Diesel Generator System.


Question: Heat Capacity Temperature Limit definition

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise: - Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, or - Suppression chamber pressure above the Primary Containment Pressure Limit, before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent.

It is vital to maintain Torus temperature below this threshold during emergencies to preserve primary containment and continue safe operation of the plant.


Question: The proposed regulatory changes regarding nuclear waste management could lead to an unacceptable situation with unmonitored radioactive materials.

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

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


Question: The analysis of the small break loss-of-coolant accident includes evaluating the pressures in the primary and secondary systems.

Table 4-3 summarizes the pressures recorded in both the primary and secondary systems during the small break loss-of-coolant accident scenario analysis for Byron/Braidwood Unit 1.

The table also indicates various parameters for system performance and operational efficiency, which are critical for operational assessments.


Question: The calculations produced by RADTRAD suggest the release of nuclides under specific conditions in the event of a loss of coolant accident.

The RADTRAD software performs detailed analyses to assess the performance of the Emergency Core Cooling System (ECCS) leakage pathways during postulated accidents, including loss of coolant events.

The plant power levels were recorded at various points in time to monitor performance and assess compliance with safety standards.


Question: boron injection in nuclear reactors

BB8A STEAM GENERATOR BLOWDOWN M304 PISTON T Normal: OPEN 10


Question: Fuel Cladding containment requirements

Fuel Cladding, RCS Pressure Boundary, Containment

The stress limits must not exceed the ASME CC-3400 limits.


Question: What is Nuclide MDA in nuclear safety reporting?

Nuclide MDA (Minimum Detectable Activity) refers to the lowest activity level that can be reliably detected by a measurement system under specified conditions.

The peak locate performance conducted on 8/9/2019 measured various energy levels and sizes of peaks in the detector output.


Question: What is minimum detectable concentration in nuclear analysis?

The report date for the project is October 15, 2019.


Question: What meteorological inputs are used for safety analyses at nuclear sites?

The following meteorological conditions are used as inputs to the model: Pasquill Stability Class G (very stable), with a wind speed of 2.84 miles per hour (mph); ambient temperature of 80.8 degrees Fahrenheit (°F); and an atmospheric pressure of 1 atmosphere (atm). Each of these meteorological conditions represents the 5 percent worst case at the PSEG Site.

The analyzed effects of flammable vapor clouds and vapor cloud explosions from internal and external sources are summarized in Table 2.2-19 and are described in the following subsections relative to the release source.


Question: preliminary reactor operator license application

Preliminary reactor operator and senior reactor operator license applications (OMB control number 3150-0090) and medical certifications (OMB control number 3150-0024) should be submitted at least 30 days before the first examination date.

The burden to the public for these voluntary information collections is estimated to average 2,250 hours per examination or response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collections.


Question: What is the significance of the ARTEMIS™ model in non-LOCA evaluations?

The NRC staff finds ARTEMIS™ acceptably models the best estimate neutronic time-dependent transient responses (e.g., power response to changes in Doppler, moderator, etc.), and that it is an acceptable tool for use in an evaluation model for non-LOCA SRP Chapter 15 events.

For each new release of the ARCADIA code system a change review and validation of the codes is performed.


Question: Southern Nuclear Operating Company is submitting alternative requests and relief requests related to the inservice testing program for specific pumps and valves at their nuclear plant.

Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of the enclosed alternative requests and relief requests for Edwin I. Hatch Nuclear Plant (HNP) Units 1 and 2.

This letter contains no regulatory commitments. If you have questions, please contact Ryan Joyce at 205.992.6468.


Question: nuclear device registration process

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


Question: What steps are required for cultural resource inventories before disturbance in nuclear licensing?

The ISR GEIS (Section 4.5.8) provides steps that should be completed to address historic and cultural resources prior to engaging in land disturbing activities. These steps include: Review existing literature and records to determine presence of known historic and cultural resources. Complete a comprehensive cultural resources inventory that meets the requirements of federal, state, and local agencies. The cultural resources inventory should identify previously documented sites as well as new cultural resource sites, including sites eligible for inclusion on the NRHP.

Potential general impacts to historic and cultural resources are discussed in Section 4.11 of the CUP FEIS and Section 4.5.8 of the ISR GEIS.


Question: The demonstration tests have been criticized for not adequately proving the effectiveness of the systems involved.

An internal investigation of the failure of some of the GE design test apparatus to function properly, concluded: "…the ‘why’ of the situation has come down to the simple fact that we believed GE was doing the job for which they were paid… the GE effort in heater development has been demonstrated to be seriously inadequate."

Based on the careful reading of the memorandum in the light of CNI’s analysis of the tests and of the cross-examination of [Aerojet] and GE witnesses, CNI has concluded that in effect GE tried to approach elements of the test program, and attempted to interpret the results, in ways wholly inconsistent with the technical content of the test program.


Question: What are the activity levels for nuclides in pCi/grams?

| Nuclide Name | Energy (keV) | Yield (%) | Activity (pCi/grams) | Nuclide MDA (pCi/grams) | Line MDA (pCi/grams) |


Question: What is a low population zone in relation to nuclear power plants?

The low population zone (LPZ) for VEGP Units 3 and 4 is the same as the LPZ for the existing VEGP units and consists of the area falling within a 2-mi radius of the midpoint between the VEGP Unit 1 and Unit 2 containment buildings.

Colleges, schools, hospitals, a military base, and the SRS are between 10 and 50 mi from the VEGP site.


Question: It is essential to confirm that the dissolved boron levels are adequate before loading or unloading operations begin.

SR 3.2.1.1 Verify the dissolved boron concentration is met using two independent measurements.

The surveillance requirements are a critical aspect of maintaining operational safety and regulatory compliance.


Question: What factors are considered in the geology of nuclear facilities?

Accurate knowledge of the geologic conditions can provide the ability to evaluate the impact of these holding ponds.

The boundaries between the headward lying mature valley floors and the zones of youthful incisements are usually quite sharp.


Question: Framatome proprietary information procedures

In accordance with Framatome’s policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

The information reveals details of Framatome’s research and development plans and programs or their results.


Question: The regulatory framework mandates that licensees must perform a thorough assessment of safety measures during reactor operation.

Nuclear reactors must undergo a rigorous safety assessment to ensure compliance with federal regulations and operational standards.

The cooling system of a reactor is designed to maintain the appropriate temperature and pressure during normal and abnormal conditions.


Question: What is the energy yield and activity level of CE-144?

CE-144 has an energy of 80.12 keV with a yield of 1.36% and an activity of 4.53E+00 pCi/grams.

The table lists the nuclides along with their energy, yield, and activity, but does not provide any information on external regulatory processes.


Question: The NRC mandates that analytical results must adhere to standardized testing protocols to ensure data accuracy and regulatory compliance.

Cyanide, total has a result reported using EPA Method 335.4, indicating the standardized testing procedures were followed to measure this parameter in the groundwater sample.

The sample matrix includes a variety of components analyzed under different methodologies, providing a comprehensive view of groundwater quality.


Question: The regulations require facilities to use deterministic methods for demonstrating that safety systems can effectively manage design basis accidents.

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

In order to meet a deterministic threshold value for sump debris loads, the debris sources in containment would need to be significantly reduced.


Question: nuclide activity measurements

The activity of a nuclide is measured in picoCuries (pCi) per gram, indicating the amount of radioactive decay occurring over time per unit mass.

The quality of reactor coolant is crucial for maintaining operational safety and efficiency during normal operations.


Question: The licensee must establish procedures to protect the confidentiality of personal information acquired through background checks.

The licensee shall develop, implement, and maintain a system for personnel information management with appropriate procedures for the protection of personal, confidential information.

The application fee (currently $26) is the sum of the user fee charged by the FBI for each fingerprint card or other fingerprint record submitted by the NRC on behalf of a licensee, and an NRC processing fee, which covers administrative costs associated with NRC handling of licensee fingerprint submissions.


Question: What is an alert in nuclear emergency response?

Upon receipt of a classification of Alert by RBS, the Parish dispatcher will notify the OHSEP Director, other key individuals, and other supporting response organizations consistent with this classification and in accordance with established procedures.

The communication center is equipped with various technologies that enhance the efficacy of emergency response efforts and ensure timely coordination among different agencies involved.


Question: Minimum Critical Power Ratio

The NRC staff notes that TS LCO 3.3.4.1, Condition B, is entered when 'One or more Functions with EOC-RPT [end of cycle recirculation pump trip] capability not maintained AND MCPR [Minimum Critical Power Ratio] limit for inoperable EOC-RPT not made applicable'.

Identify in columns 3 and 5 of LAR enclosure 1, table E1-1, all affected SSCs and their purpose more clearly, and in column 6, the divisions affected.


Question: The program includes provisions for assessing the need for calibration based on condition monitoring and drift analysis.

If a high degree of confidence cannot be established that there is no CC failure mechanism that could affect the redundant components, the RICT shall account for the increased possibility of CC failure.

This program provides controls to determine the need for calibration for pressure level, and flow transmitters using condition monitoring based on drift analysis.


Question: What are the measures for controlling nonconforming items in nuclear licensing?

SNC employs measures to identify and control items that do not conform to specified requirements to prevent their inadvertent installation or use.

SNC employs measures for line management to periodically review and document the adequacy of the process, including necessary corrective action.


Question: What is the purpose of AMP XI.M41 in managing HDPE underground piping?

AMP XI.M41, "Buried and Underground Piping and Tanks"

The Outdoor and Large Atmospheric Metallic Storage Tanks (B.2.1.17) program will be used to manage loss of material of the aluminum alloy tanks exposed to treated water in the Demineralized Water Makeup System.


Question: What is net positive suction head (NPSH) in nuclear reactor pump systems?

The containment recirculation spray pumps are capable of meeting NPSH requirements under accident conditions.

Table 6.2-41 discusses the likelihood and consequences of various component malfunctions.


Question: What is the purpose of the Component Cooling System in a nuclear facility?

The Component Cooling System serves as an intermediate system and a second boundary between the Reactor Coolant and Auxiliary Systems and Nuclear Service Water System and assures that any leakage of radioactive fluid from the components being cooled is contained within the station.

Nominal flow rates during various unit operating modes are tabulated in Table 9-22. Typical valve lineups for various system operational modes are shown in Table 9-23.


Question: nitrate/nitrite analysis in groundwater

The conductance of the water sample at 25 degrees Celsius was measured at 12800 uhmos/cm.


Question: What does ALARA mean in nuclear regulations?

ALARA stands for 'As Low As Reasonably Achievable', which is a safety principle aimed at minimizing radiation exposure to workers and the public in the context of nuclear operations.

The Global Positioning System (GPS) is a satellite-based navigation system that provides location and time information in all weather conditions, anywhere on Earth.


Question: NRC Office of Investigations

OI is an independent national investigations program that consists of four regional field offices led by Special Agents in Charge who directly support the regional administrators and report directly to OI headquarters senior staff.

The main body of OI personnel is made up of Federal criminal investigators and special agents (GG-1811s), as well as professional support and technical staff located in both OI field offices and headquarters.


Question: What is the significance of reactor vessel head closure bolts in nuclear safety?

(a) All required reactor vessel head closure bolts fully tensioned. The required number of head closure bolts is at least 51 of 52 bolts.


Question: What should be the procedure for maintaining reactor water level during a scram event?

Maintain Reactor water level within a band of 180-220”. Insert control rods using attachments in AOP-1, Reactor Scram.

A plant startup is in progress with the following: - IRMs indicate mid-scale on Range 7. - APRMs indicate downscale.


Question: air return fans

The air return fans have sufficient head to overcome the divider barrier differential pressure (a maximum of 7.17 psf) resulting from steam flow and fan air flow entering the ice condenser through the lower inlet doors.

In response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," McGuire has the option of starting one air return fan at a containment pressure of 1 psig during certain small break LOCA (SBLOCA) transient events.


Question: What types of materials are used for piping in nonsafety-related ventilation systems?

Piping, piping components are used for leakage boundary, and materials include Carbon Steel and Cast Iron.

Cumulative Fatigue Damage can occur in various components in a reactor due to repeated stress cycles.


Question: The NRC staff inquires about the data underlying the source terms included in the summarized tables, which encompass various loss scenarios.

The NRC staff requests access to documentation of the development of the source terms provided in the summary table (PSAR tables 3.2-1, "Protected Loss of Flow or Loss of Offsite Power" through 3.2-19, "Unprotected Loss of Heat Sink, Full Core and [In-Vessel Storage (IVS)] Fail"). This includes calculations, description of the implementation of the source term methodology, as well as justifications for inputs and assumptions, computer code output, or other relevant documentation.

Foundation bearing capacity and associated settlement assessment report, which should include estimation of rock mass strength and stiffness calculations.


Question: contentions in regulatory filings

In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions that the petitioner seeks to have litigated in the proceeding.

Petitions must be filed no later than 60 days from the date of publication of this notice.


Question: contract performance extension in regulatory filings

The COR is in DORL, so we won’t be responsible for this part. This is just a heads up for your awareness.

I was concerned this was the one John B was COR on for John L, but thanks for the correction.


Question: What is the typical thermal conductivity requirement for nuclear graphite?

Typical nuclear graphite requires thermal conductivity (W/m·K, AG at 25°C): ≥ 100 for extruded and molded graphite, ≥ 90 for isomolded graphite.

Higher strength material has higher modulus, normally lower fracture toughness.


Question: The instrumentation involved in preventing water intrusion into turbines is crucial for maintaining plant safety.

Feedwater and Main Turbine High Water Level Trip Instrumentation<br>One or more feedwater and main turbine high water level trip channel(s) inoperable.

The Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of.


Question: Strontium-90 analysis in soil samples

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


Question: Control Rod Drive System components

The aging management review results for these components are provided in: Table 3.3.2-3, Control Rod Drive System, Summary of Aging Management Evaluation.

Accumulators are crucial in maintaining pressure boundaries within hydraulic control units.


Question: The acceptance review evaluates whether the submitted information meets the necessary regulatory standards for technical specifications revisions.

The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review.

If additional information is needed, the licensee will be advised by separate correspondence.


Question: What is an aging management program in nuclear licensing?

None of the components listed are evaluated in NUREG-2191.


Question: The NRC has established requirements for monitoring boric acid corrosion in pressurized water reactors to ensure safety.

The Boric Acid Corrosion Control Program outlines procedures and guidelines for managing and mitigating the effects of boric acid on reactor components.

The ASME Boiler and Pressure Vessel Code has specific criteria for the fabrication of pressure retaining welds and components.


Question: The NRC mandates that licensees must ensure the operational performance of the CCS systems to facilitate effective cooling during operation.

The CCS to Normal Residual Heat Removal System Heat Exchanger (RNS HX) A Flow Sensor is crucial for monitoring the flow of coolant to ensure safety and operational efficiency.

The flow values for the CCS Pump indicate varying capacities across units, with Unit 3 A displaying a maximum flow of 7662 gpm for Train A HXs.


Question: The acceptable concentration limit for Lead 210 in monitoring well assessments is 62 pCi/L.

Concentrations that are less than the lower level of detection (LLD) are represented by an open symbol.


Question: licensing consultant role in nuclear regulatory processes

Alan Harris, Jensen Hughes, Licensing Consultant, plays a crucial role in providing expert guidance on regulatory filings related to nuclear licensing activities.

The Waterford 3 project team consists of various professionals including project managers and engineering supervisors, each responsible for distinct aspects of the project.


Question: The change introduces a new specification for the containment sump, retaining existing requirements and updating actions for inoperability due to debris.

The proposed change also adds a new specification to the TS for the containment sump. An existing SR on the containment sump is moved to the new specification and a duplicative requirement to perform the SR in TS 3.5.3 is removed. The new specification retains the existing requirements on the containment sump and the actions to be taken when the containment sump is inoperable with the exception of adding new actions to be taken when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits.

The proposed change does not alter the manner in which safety limits are determined or the acceptance criteria associated with a safety limit.


Question: What is the meaning of gamma measurement in nuclear safety assessments?

**Photograph Collected:** Yes
**Remarks:** 8-9, 2" ag. rem.


Question: Refueling operations safety procedures

Refueling Operations must be stopped IMMEDIATELY if communication is lost between the Control Room and Containment.

The reactor core must be inspected regularly to ensure that all safety standards are maintained.


Question: What is the significance of dpm/100 cm² in nuclear measurements?

Gamma Result is measured in disintegrations per minute per 100 square centimeters (dpm/100 cm²), which reflects the level of radioactivity in a given area.

The measurement population is essential for categorizing the various measurements collected during the regulatory assessments.


Question: What are the roles and responsibilities of the Integrated Decision Making Panel in surveillance frequency changes?

Per ER-AA-425-1003, SURVEILLANCE FREQUENCY CONTROL PROGRAM – INTEGRATED DECISIONMAKING PANEL (IDP) ROLES AND RESPONSIBILITY, states the Operations and Engineering are voting members on the Integrated Decision Making Panel (IDP). This panel is responsible for: - Reviewing a proposed Surveillance Test Interval (STI) change for both qualitative considerations and quantitative results - Making recommendations on the way revised surveillance intervals are implemented - Reviewing the cumulative impact of all changes carried out over a period of time - Monitoring the impact of changes on failure rates

The Maintenance Manager is responsible for providing resources for tests/inspections required by the IST program, but not for approval of surveillance frequencies.


Question: What is the purpose of a reactor building containment vessel?

The reactor building containment vessel is designed to provide a physical barrier to the release of radioactive materials into the environment.

The test channel layout is crucial for ensuring that various reactor components can be evaluated under controlled conditions.


Question: The administration will ensure that the exposure to radiation is maintained at the ALARA standard as specified in the regulations.

The Radiation Protection Program will establish administrative controls to limit personnel exposure to As Low As Reasonably Achievable (ALARA) levels in accordance with 10 CFR Part 20 and Part 72.

A radiological environmental monitoring program will be implemented to ensure that the annual dose equivalent to an individual located outside the WCS CISF controlled area does not exceed the annual dose limits specified in 10 CFR 72.104(a).


Question: What is the significance of MDA values in nuclear regulatory filings?

The NRC requires a comprehensive safety assessment for each nuclear reactor design.


Question: What is the purpose of Containment Local Leakrate Preoperational Test in nuclear safety?

A. Objective To determine the leakage rate of the containment penetrations and isolation valves.

The diesel generators function as designed on a simulated loss of offsite power including a simultaneous start of the redundant units, and emergency bus loads are sequenced on the bus.


Question: What is the purpose of NRC device registration?

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

The essential function of the NRC is to ensure that nuclear power plants operate safely and securely.


Question: According to the NRC, components exposed to environmental factors may experience material loss due to corrosion mechanisms.

Loss of material due to selective leaching for copper alloy (>15% Zn or >8% Al) piping, piping components exposed to soil will be managed by Selective Leaching program (B.2.1.21) in the High Pressure Fire Protection (Diesel Driven Pump) System.

Not Applicable. There are no aluminum fire water storage tanks exposed to air, condensation, soil, concrete, raw water, raw water (potable), treated water in screened in portions of Auxiliary Systems.


Question: What are the loading tables for PWR fuel assemblies?

The loading tables for PWR fuel assemblies provide data such as the thermal power level, which includes values like 513 W/Assembly for WE 14x14 Fuel and 1300 W/Assembly for CE 16x16 Fuel.

The NRC maintains rigorous standards for the licensing process of nuclear facilities, ensuring safety and compliance with federal regulations.


Question: licensed material transfer agreement

Confirm that both the transferor and transferee agree to transferring control of the licensed material and activity, and the conditions of transfer, and that the transferee has been made aware of any open inspection items and its responsibility for possible resulting enforcement actions.

The enclosed Exhibit H is a form of letter from NewCo Corp. confirming that it agrees to the direct and indirect transfer of control within the meaning of 10 CFR § 40.46 and implementing guidance.


Question: What is the role of the U.S. Nuclear Regulatory Commission in ensuring public safety?

The U.S. Nuclear Regulatory Commission (NRC) is responsible for regulating the nation's civilian use of nuclear materials to ensure safety in various aspects, including public health, including oversight of reactor safety and radioactive waste management.

A year's subscription consists of 12 softbound issues, 4 indexes, and 2-4 hardbound editions for this publication.


Question: What is the average enrichment for nuclear fuel assemblies?

The cooling time of a nuclear reactor is crucial for safety and operational efficiency.


Question: What is the purpose of the Alternate Low-Pressure Injection/Drywell Spray System?

Additional dedicated Alternate Low-Pressure Injection/Drywell Spray System.

Develop a procedure for alternating operation of low-pressure ECCS pumps for loss of standby service water (SSW).


Question: centerline temperature predictions

The best-estimate character of the results for UO₂ and gadolinium fuel is confirmed by the centerline temperature predictions shown in Section 4.3.2.

The near-perfect agreement demonstrates a lack of bias.


Question: What are Nondestructive Examination procedures in nuclear technology?

Nondestructive Examination Procedure Liquid Penetrant Examination Procedure Review/Work Observation/Record Review of Transportation ISFSI Components.

Calibration of Nuclear Instrumentation Systems 05/12/86 (86-25)


Question: The NRC mandates that licensees assess the potential risks associated with maintenance or testing on safety systems, especially during their inoperability.

For TS 3.5.2, "Emergency Core Cooling Systems (ECCS)," Condition A, one LPI subsystem inoperable, additional RMAs would include: 1. Verify remaining LPI subsystem is available and aligned for standby service.

Minimize activities that could trip the unit.


Question: reasonably foreseeable impacts

we believe the appropriate test for determining whether an action is reasonably foreseeable is whether the other action is in such a sufficiently advanced stage that a person of ordinary prudence would take it into account in reaching a decision.

Ms. Henderson did not file a supplemental brief.


Question: Continuous operation of the Auxiliary Building Secondary Containment is required during the outage of the Replacement Steam Generator.

The proposed technical specification modification will allow for the continuous opening of the Auxiliary Building Secondary Containment Enclosure (ABSCHE) as needed to support the Unit 2 Replacement Steam Generator (RSG) outage.

The environmental assessment includes considerations of possible thermal impacts on local aquatic life and habitat.


Question: The NRC has determined to deny the rulemaking petitions concerning the thermal limits for fuel element cladding.

The Commission has approved publication of the *Federal Register* notice denying two petitions for rulemaking (PRM-50-93 and PRM-50-95) and issuance of the letter informing the petitioner of this action, subject to the enclosed edits.

The meeting will also address various concerns related to safety assessments and community engagement initiatives that are essential for maintaining transparency.


Question: The task involves reducing the speed of the RWR MG Set A due to operational requirements.

The local manual operation involves adjusting the RWR MG Set speed based on specific safety and operational standards.

There are no additional tools or equipment required for this task.


Question: operable channels in plant licensing

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE, restore the inoperable channel to OPERABLE status within 48 hours or in accordance with the Risk Informed Completion Time Program; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.

With the number of OPERABLE 460 volt Degraded Voltage (Functional Unit 7.b) channels one less than the Total Number of Channels or with both 4.16 kv Loss of Voltage (Functional Unit 7.a) channels inoperable on a single bus: a. Immediately declare the affected diesel generator inoperable, and b. Restore the inoperable channel to OPERABLE status within 48 hours or in accordance with the Risk Informed Completion Time Program; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. LCO 3.0.4.a is not applicable when entering HOT SHUTDOWN.


Question: What is the function of a pressurizer in a nuclear plant?

The pressurizer is an essential component of the primary reactor cooling system, maintaining the pressure of the coolant and allowing for the control of coolant temperature and phase.

The Watts Bar Nuclear Plant is equipped with advanced safety systems and protocols to ensure operational integrity during tests.


Question: What is the role of barrier coatings in managing aging effects in nuclear plant components?

The applicant may establish that loss of material due to pitting and crevice corrosion is not an aging effect requiring management by demonstrating that a barrier coating isolates the component from aggressive environments.

There is one stainless steel tank, the Condensate Heat Drain Tank, in the screened-in portion of the Steam and Power Conversion Systems.


Question: What are the requirements for performance-based inspections in radioactive materials programs?

Performance-based inspection (PBI) programs utilize established performance indicators to identify areas needing improvement, enabling a more flexible and effective inspection approach to ensure compliance with applicable regulations.

The Office of Radiological Health engages in various educational programs to raise awareness of radiological safety and to effectively inform the public about radiation-related risks.


Question: The regulatory framework governing reactor designs requires careful consideration of the specific safety and functionality aspects relevant to the proposed technology.

Ultimately, a “plain language” read of the regulations is less legally defensible and aligns less with the Atomic Energy Act than a thorough comprehensive analysis.

A key part of this approach was pioneering with the NRC a new way of thinking for reactor licensing, which Oklo saw as a viable path forward because of the years of positive experience with the NRC staff in pre-application space.


Question: What is the purpose of head loss tests in nuclear safety evaluations?

Head loss tests were performed to measure the head losses caused by conventional debris (fiber and particulate) and chemical precipitate debris generated and transported to the sump strainers following a LOCA.

The testing sequence was performed following a "test for success" strategy with a total of 4 tests, as summarized below.


Question: What is the role of the Radiation Safety Officer in nuclear facilities?

The Radiation Safety Officer (RSO) is responsible for ensuring compliance with federal and state regulations regarding radiation safety, serving as the primary contact for all safety-related issues involving radioactive materials.

The Chief Financial Officer oversees corporate financial activities, including financial planning, risk management, and record-keeping.


Question: The NRC mandates that certain isotopes must be reported along with their activity levels and corresponding units during regulatory filings.

The device's transfer date was recorded as 11/06/2000, which is essential for tracking its history and compliance.


Question: What are Class 1E electric systems in nuclear facilities?

The 125 VDC-120 VAC Vital Instrumentation and Control Power Systems are the only systems classified as Class 1E electric systems and are designed to meet the requirements of IEEE 279-1971, IEEE 308-1971, 10CFR 50 General Design Criteria 17 and 18, and NRC Regulatory Guide 1.6.

Battery Chargers 1EDGA, 1EDGB, 2EDGA, and 2EDGB are supplied power from Motor Control Centers 1EMXF, 1EMXE, 2EMXE, and 2EMXF, respectively.


Question: The analysis of sulfate levels requires careful calibration and method selection to ensure accuracy in reporting.

WG593640 | Sulfate | EPA 300.0 | M1 | Matrix spike recovery was high, the recovery of the associated control sample (LCS or LFB) was acceptable.

WG595064 | Silver, dissolved | EPA 200.7 | MA | Recovery for either the spike or spike duplicate was outside of the acceptance limits; the RPD was within the acceptance limits.


Question: Inspections are performed regularly to ensure the safety of the reactor coolant pressure boundary.

A Condition Report was generated on May 3, 2011, to evaluate operating experience (OE) from Monticello for applicability and utilization at BFN.

On October 15, 2020, during Ultrasonic Test (UT) examination of the Unit 1 safe end to pipe weld on an instrument line connected to pressure vessel nozzle and connecting upstream of a primary containment penetration, a circumferential indication was identified.


Question: What are the total curies of I-131?

The NRC regulates various aspects of nuclear facility operations to ensure safety and compliance.


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

Wt mean Activity (pCi/grams) provides a quantifiable measure of the radioactivity of a nuclide, critical for assessing safety and regulatory compliance.

Errors quoted at 2.00 sigma indicate that the quoted activities have been derived with a certain statistical confidence, which is important in the analysis process.


Question: What is the significance of the 50-mile radius in nuclear regulatory documents?

The 50-mile radius is a common measure used in nuclear regulatory filings to define areas potentially affected by a nuclear facility's operations, including emergency planning and environmental impact assessments.

February 2024 marks the beginning of a new review period for the applicant’s environmental report regarding the Dresden Nuclear Power Station.


Question: What are the implications of reactor coolant system cold overpressure during shutdown?

Knowledge of the relationship between the response to reactor coolant system cold overpressure during shutdown and the following systems or components: (CFR: 41.8 / 41.10 / 45.3)


Question: What are the acceptable range limits for cation-anion balance in water quality data?

Cation-anion balance will be between 0.95 and 1.05.

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: radiological dose assessment team responsibilities

The Dose Assessment Team maintains contact with and transmits instructions to Offsite Monitoring Teams, performs offsite dose assessment calculations, and provides the Radiological Assessment Director with offsite release calculations.

The Operational Support Center Director will be staffed and shall report to the Station Emergency Manager, normally through the EMD.


Question: What are the requirements for a nuclear reactor's safety analysis report?

The safety analysis report (SAR) must provide a comprehensive evaluation of the design and safety features of the reactor, including potential hazards and the safety measures in place to mitigate those hazards.

Nuclear power plants must maintain a specific output level to meet energy demands, ensuring a stable supply for the grid.


Question: The accuracy of instrumentation can significantly impact the evaluation of safety-related systems during a Design Basis Event.

For the purpose of this calculation, RE will be ±10% of calibrated range for trip condition and will include accuracy temperature effects for the trip conditions (ATEn).

Due to Barton 289A and 581A Differential Pressure switches being similarly constructed, it is assumed that drift would be similar.


Question: To track compliance with the License Conditions, each component must be assessed individually when a failure occurs.

The Note allows Condition A to be entered separately for each component, and Completion Times to be tracked on a per component basis.

The immediate completion time is critical when actions must be taken without delay.


Question: Monitoring Well 17-01 KD has recorded nitrate levels consistent with regulatory standards.

The monitoring data from various months shows fluctuations in nitrate concentrations.


Question: The NRC evaluates the potential impacts of environmental conditions on the performance of the plant's systems and structures.

The NRC requires licensees to perform environmental review processes during the license renewal application to assess potential impacts on the environment.

The power station has a variety of systems designed to ensure the safety and efficiency of operation, including backup generators and control systems.


Question: What are QPTR values and why are they important in nuclear regulatory compliance?

Applicant calculates QPTR values by dividing the respective Relative Powers [Step 5] by the Average Relative Powers for each NIS channel detector [Step 6]. The results recorded in appropriate blocks on Attachment 4. Step is critical because the values calculated will be used to determine whether the individual QPTR for each detector is within the Tech Spec limit to ensure compliance.

The tolerance for QPTRs is based on the requirements of Attachment 4.


Question: The analysis of potential accidents is crucial in maintaining reactor safety standards.

The analysis of potential accidents scenarios is a key requirement for maintaining safety in nuclear reactor operations as outlined in regulatory guidelines.

Reactor core cooling systems are designed to ensure heat removal from the core under normal and accident conditions.


Question: What constitutes an environmental assessment under NRC regulations?

An environmental assessment for this action was not required because this action is categorically excluded under Title 10 of the Code of Federal Regulations (CFR) 51.22(c).

You will be periodically inspected by the NRC.


Question: What is Overpower ΔT in nuclear safety?

The Overpower ΔT trip Function ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions.

In MODE 1 or 2, the Overtemperature ΔT trip must be OPERABLE to prevent DNB.


Question: The regulation of nuclear waste is integral to maintaining safety standards and protecting the environment.

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

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


Question: What is the significance of assessing NRELAP5 model responses in nuclear licensing?

The staff confirmed, based on the results of the various assessment sensitivity studies, the applicant’s conclusion that the NRELAP5 model responses are consistent with physics-based expected results and that there are very negligible effects on the event FOMs.

L/C modified to include “unless the method is followed that is described in section 5.2 of the LOCA EM TR that models the riser level instrument setpoint based on mixture level in the riser, using one of the approaches described in detail in section 5.2 (not including the application–specific alternate approach).” [This addition is also reflected in Section 4.5.2.] Additionally, L/C #4 and #9 were combined into one L/C.


Question: The cooling systems for the new units will utilize either closed-cycle designs that minimize water loss or once-through systems that may be affected by water supply conditions.

New Unit 4 would use a closed-cycle cooling system with dry cooling towers in which the exhaust from the plant’s steam turbines would be directed to a surface condenser where the heat of vaporization would be rejected to a closed loop of cooling water.

The ESP site is situated approximately 5 miles upstream from the main dam and adjacent to the existing units.


Question: What is the process of obtaining a nuclear reactor license?

The process of obtaining a nuclear reactor license involves submitting an application to the Nuclear Regulatory Commission (NRC) that demonstrates compliance with safety, environmental, and security regulations.

In recent years, advancements in renewable energy technologies have significantly impacted the energy market, leading to increased investment in solar and wind power alternatives.


Question: A spent nuclear fuel assembly that exhibits visible deformation of its rods is considered damaged.

SPENT NUCLEAR FUEL (SNF) assembly that cannot fulfill its fuel-specific or system-related function. SNF is classified as damaged under the following conditions: 1. There is visible deformation of the rods in the SNF assembly.

The CONCRETE CASK LID is a thick concrete and steel closure for the CONCRETE CASK. The CONCRETE CASK LID precludes access to the TSC and provides radiation shielding.


Question: What are the performance criteria for NRC examinations?

**Performance Rating:**  SAT ____  UNSAT ____


Question: What do KEFF and SIGMA represent in nuclear reactor calculations?

KEFF is the effective neutron multiplication factor which indicates whether a nuclear chain reaction is self-sustaining, and SIGMA represents the cross-section of a specific material used in reactor calculations, indicating the probability of neutron interactions with that material.

The radioactive decay of a material is characterized by its half-life, which indicates the time it takes for half of the material to decay.


Question: The process of groundwater restoration aims to achieve regulatory compliance with established standards for water quality during operational activities.

The primary goals of the aquifer restoration program are to: Restore the groundwater consistent with 10 CFR Part 40, Appendix A, Criterion 5(B)(5). Complete aquifer restoration concurrently with ISR uranium extraction in accordance with 10 CFR § 40.42.

The management of liquid effluents requires careful consideration to ensure compliance with regulatory standards during extraction processes.


Question: What is included in the draft environmental assessment for nuclear plant operations?

As part of this review, 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. ML24353A157).

The NRC is the lead Federal agency for the environmental review under the National Environmental Policy Act of 1969, as amended, with the U.S. Department of Energy (DOE) Loan Programs Office (LPO) serving as a cooperating agency.


Question: The weighted mean activity for the nuclide is calculated based on the average activity per gram, taking into account uncertainties.

The weighted mean activity of K-40 is 4.30E+00 pCi/grams, which reflects its radioactivity measured per gram of the material.

The confidence level for the nuclide PB-212 is 0.523, indicating a moderate degree of certainty in the measurement results.


Question: What is the function of the Source Range Monitor system in nuclear reactors?

The Source Range Monitor (SRM) system is crucial for monitoring the neutron flux in the reactor, especially during startup and low-power operation, to ensure safe control of the reactor manually.

A Reactor startup is in progress with the following: All SRMs are fully inserted and indicate between 200 and 300 cps. All IRMs indicate downscale on Range 2.


Question: COBRA-FLX critical heat flux correlation

COBRA-FLX is a computer code developed for the prediction of critical heat flux (CHF) in light water reactor thermal-hydraulic systems, and it has been validated against a wide range of experimental data to ensure its accuracy.

The reactor coolant pump is an essential component that helps circulate the water through the reactor core, ensuring efficient heat exchange and maintaining appropriate temperature levels.


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

A safety-related system is defined as a system, structure, or component that is relied upon to ensure the safe operation of a nuclear facility under normal and accident conditions.

Nuclear power plants convert nuclear energy into electricity, providing a stable power source for thousands of homes.


Question: NRC regulations include safety measures for radiation protection in nuclear facilities.

Authority and responsibilities for the radiation protection are defined in regulation 35.24, which lays out the standards necessary to ensure the safety of personnel and the public.

Licensees must submit a plan that outlines their strategy for handling emergencies, including training and resource allocation.


Question: clevis insert bolts in reactor vessels

Alignment and interfacing components: clevis inserts, clevis insert bolts are used for Structural Support in the reactor coolant and neutron flux environment, and are designed to withstand aging effects such as change in dimension, cracking, loss of material, and loss of preload as part of the aging management program.

The reactor coolant system is a crucial aspect of maintaining the safe operation of the nuclear reactor, ensuring that heat is removed effectively from the reactor core.


Question: What are the ECCS response times in nuclear safety analysis?

This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.


Question: nuclear material safety

The NRC’s Office of Nuclear Material Safety and Safeguards oversees the regulation of materials used in the nuclear industry to ensure safety and compliance with federal laws.

The NRC encourages public involvement and provides several avenues for stakeholders to participate in the regulatory process.


Question: Activities impacting quality are conducted under specific controlled conditions as defined in relevant procedures and guidelines.

Planning and performance of activities affecting quality are accomplished under suitably controlled conditions as delineated in appropriate implementing procedures and instructions.

This section describes the GE Hitachi Nuclear Energy (GEH) quality assurance program, as documented through this Quality Assurance Program Description (QAPD) and its implementing policies, procedures and instructions, and identifies the associated requirements for planning, implementing, monitoring, and maintaining the program.


Question: The regulation of heavy metals in the water supply is crucial for safety, particularly concerning elements like lead and cadmium.

The mean pH level recorded in the project area was 8.75, indicating a slightly alkaline water quality.


Question: nuclear reactor license applications

The LaSalle County Station, operating under License Numbers NPF-11 and NPF-18, has undergone a stringent review process to ensure regulatory compliance and operational safety.

The inspection was conducted by a team of qualified inspectors who reviewed various aspects of facility operations during the defined inspection period.


Question: The NRC emphasizes the need for adequate monitoring of groundwater levels to understand flow dynamics at the site.

The groundwater discharges to the Colorado River along the southern site boundary. As the water level in the Colorado River rises above the water level in the adjacent aquifer during each spring, the groundwater reverses its flow direction near the river begins to recharge the aquifer, which may create an immobile or stagnation zone in the alluvial aquifer along the river.

NRC finds DOE's response acceptable and that NRC's concurrence on the GCAP will be based on the revised text in the GCAP; however, DOE must be clear that removal of the downgradient wells will be substantial change in the GCAP, a change which would require concurrence by NRC.


Question: What is the methodology for determining the Plume Exposure Pathway Emergency Planning Zone?

This Topical Report describes TerraPower's methodology to be used to determine the Plume Exposure Pathway (PEP) Emergency Planning Zone (EPZ) for the Natrium™ reactor, a TerraPower and GE Hitachi technology.

The purpose of submitting this Topical Report is to provide information to the U.S. Nuclear Regulatory Commission (NRC) to facilitate efficient and timely review of the TerraPower PEP EPZ sizing methodology.


Question: impact of AC distribution system malfunction

Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following: Major system loads

Since the 23SPT transfer relay remains energized, the operator may incorrectly determine that the 24 SPT and 2A EDG transfer scheme is defeated causing the 2A 4KV Bus to remain de-energized.


Question: To ensure adequate cooling during an emergency, operators must prepare to align safety injection systems effectively.

Manually Align one train of Safety Injection to establish HIGH Head Injection Flow.

Check BIT Flow (E-0, Attachment A Step 2) - 1-IFI-51 - 1-IFI-52 - 1-IFI-53 - 1-IFI-54.


Question: In step F.6, the control switch for the Automatic Standby Diesel is turned to START, and subsequently returned to NORMAL.

IF breaker does NOT close within 3 seconds of completing the pre-charge sequence, THEN the sequence must be repeated.

Plausible because the RECURC ASD 1A BKR ‘OFF’ and ‘TRIP’ indicating lights illuminate.


Question: The technical assessment of the pumps must demonstrate their ability to meet required hydraulic performance standards.

For pumps meeting all the above criteria, no further hydraulic evaluation was required.

The evaluation concluded that the hydraulic performance of the pumps would not be affected by the sump debris.


Question: emergency procedure priorities

In accordance with QGA 300, Temperature/Radiation Leg, if a primary system is discharging into the Reactor Building and cannot be isolated, then before any temperature, radiation, or water level reaches the max safe value per Table R, T, W, scram the reactor and enter QGA 100. This takes priority over other QGA, QCOP or QCAN required actions.

Plausible because it is a required action in QCAN 901(2)-4 C-14, HIGH RCIC TURBINE EXHAUST DISCHARGE PRESSURE. Incorrect because the valve is located high up on the wall between the RCIC room and the Torus requiring a scaffold to reach and the QGA required action has a higher priority.


Question: What is the Anticipated Transient Without Scram Mitigation System?

The Anticipated Transient Without Scram Mitigation System (ATWS) is designed to ensure the safe operation of a nuclear reactor during anticipated transient scenarios where normal safety systems may not function as intended.

The diesel driven fire pump surveillance includes regular checks of the pump's operational readiness and function to ensure reliable fire protection for the facility.


Question: The state has implemented a regulatory framework to ensure the safe handling of radioactive materials in various programs.

The Radioactive Materials Program provides oversight and regulation of the use of radioactive materials within the state, ensuring safety and compliance.

The office also offers various health services including vaccinations and disease prevention programs, aimed at improving overall public health.


Question: What is the purpose of a technical specifications document in nuclear licensing?

Technical specifications describe the specific design and operational characteristics of a nuclear facility, ensuring compliance with safety standards and regulatory requirements.

The safety evaluation report is a comprehensive document that details the various safety analyses conducted and their findings regarding the nuclear facility's operations.


Question: IT capabilities for nuclear inspections

Continue to educate staff and licensees about the capabilities and benefits of available IT tools.

Consider revisions to inspection procedures or guidance to indicate activities and inspection requirements can be performed either fully remotely, partially remotely, or onsite.


Question: What is radiological emergency response training?

Radiological emergency response training is provided to those who may be called on to assist in an emergency.

Initial and continuing training is conducted to ensure ERO personnel are properly qualified to their specific position.


Question: what is extravasation in nuclear medicine

So, I'm just trying to highlight that staff was initiating, or evaluating, extravasation.

MS. LOPAS: And, Ralph, we are very aware of all those comments. We have access to all of them. We've been coordinating with the petition group, so yes.


Question: The NRC mandates that all devices containing isotopes must be registered appropriately, including details like activity and units of measurement.

The device serial number assigned to this equipment is B8390, which helps in tracking and managing regulatory compliance.


Question: License applications must include comprehensive details on safety analysis regarding reactor operation and design.

The safety analysis report must demonstrate that the reactor can operate under a variety of conditions without leading to undue risk to public health and safety.

The plant's maintenance schedule is designed to ensure that all systems are regularly checked and updated according to the latest industry standards.


Question: In quantitative analyses, it is crucial to establish upper tolerance limits to maintain conservative estimates.

To ensure that the calculations are conservative, a one-sided tolerance limit of β is used: β* = β - t(α,v) · se(β)

The resulting linear regression model, which includes the independent variable 1/FᴀH, yields R² values larger than 99%.


Question: The assessment of radioactive isotopes such as Niobium-94 is crucial for ensuring accurate measurements in safety evaluations.

The QC number for the sample batch is QC1204335681, which provides a reference for quality control processes.


Question: nuclear reactor safety systems

Nuclear reactor safety systems are critical for preventing accidents and mitigating their consequences, ensuring that operations comply with established safety standards.

The thermal efficiency of a power generation system largely depends on the operating conditions and the design of the turbines, which convert thermal energy into mechanical energy.


Question: The evidentiary hearing process is crucial for resolving disputes in nuclear licensing.

The parties are now pursuing an evidentiary hearing that should complete this proceeding in the coming months.

Extensive analysis of technical safety topics is necessary for ensuring compliance with nuclear regulations.


Question: energy spectral analysis

Energy (keV) refers to the kinetic energy of photons in the X-ray or gamma-ray spectra, measured in kilo-electronvolts (keV).

The FWHM (Full Width at Half Maximum) represents the resolution of the spectral data, indicating how well different energy peaks can be distinguished from one another.


Question: What is the significance of duration in nuclear regulatory reporting?

The table lists the duration in hours alongside the number of occurrences from May 1968 to October 1971, indicating how often certain durations were observed during the reporting period.

The reactor core temperature must be maintained below the maximum allowable limit to ensure safe operation and prevent overheating.


Question: The Board indicated that there is a requirement for Powertech to address any leaks by monitoring excursions and implementing corrective measures.

In addition to the license condition imposed by the Board, License Condition 11.5 requires Powertech to monitor for excursions and take corrective action—

The Tribe argues that the Board committed legal error in holding that, while “small faults and joints may be present in the project area, their presence does not support Intervenors’ assertions [regarding the impacts of the faults and joints.]”


Question: The NRC has evaluated how cooling towers affect nearby plants and concluded that any possible impacts are minimal.

In the 2013 license renewal GEIS, the NRC reviewed the effects of cooling tower operation on surrounding vegetation. Terrestrial habitats in the vicinity of nuclear power plant cooling towers have been exposed to deposition of cooling tower drift particulates (including salt), deposition of water droplets on vegetation from drift, structural damage from freezing vapor plumes, and increased humidity. Generally, deposition rates from these cooling towers have been below those that are known to result in measurable adverse impacts on plants, and no deposition effects on agricultural crops or plant communities have been observed at most of the nuclear power plants.

Figure 3.7-2 shows the National Wetlands Inventory mapped wetlands within the MNGP site. There are no wetlands along the south bank where the discharge is located. The flow and width of the Mississippi River would minimize the potential for impact to wetlands on the north bank.


Question: What is the NRC Safety Culture Policy Statement?

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

Enclosed is Amendment No. 31 to your NRC Material License No. 21-18525-01 authorizing the new area of use in accordance with your request.


Question: design basis assurance inspection

The Design Basis Assurance Inspection program is intended to evaluate the design basis of facility systems, structures, and components in relation to the applicable regulatory requirements and to ensure that the plant is capable of operating within its design and licensing bases.

The inspection team reviewed the plant's operational procedures and discussed the various training programs in place for the staff, ensuring that all personnel are adequately qualified for their roles.


Question: The traffic counts near the plant's entrances show significant variations over the years.

Total Average Annual Daily Traffic Counts on County Road 75 NE indicate that, near the NW of the MNGP plant entrance, traffic numbers increased from 1,050 in 2000 to 3,650 in 2008 before falling to 1,600 in 2016.

At the Monticello Nuclear Generating Plant, various safety protocols are in place to ensure operational integrity and compliance with federal regulations.


Question: The NRC's analysis suggests that heat transfer models can be formed using data primarily derived from stainless steel cladding.

The TRACE simulations…demonstrate that it is possible to develop heat transfer models based on data obtained primarily from stainless steel rods and conservatively simulate FLECHT run 9573.

There are also serious problems with the fact that NRC compared the results of its TRACE simulations of FLECHT run 9573 to the average value of different thermocouple measurements taken at each elevation and not to the maximum values of the cladding temperatures measured at each elevation.


Question: What are the requirements for operating a digital rod control system?

Ability to operate and/or monitor the following as they apply to rod control system malfunctions: (CFR: 41.5 / 41.7 / 45.5 to 45.8) - **AA 1.01** Digital rod control system 3.3

Maintaining power margin greater than zero during dropped rod recovery (OE related) 3.6


Question: degraded voltage relay setpoint

The Degraded Voltage Relay is set with specific voltage setpoints that determine its reset and dropout criteria, which include upper and lower analytical limits and allowable values based upon the voltage levels monitored.

Primary Voltage is calculated using the formula: (Relay voltage + Wire Voltage Drop dV) * RCF * Nominal VT Ratio of 35, where dV equals 0.010 and RCF equals 0.9987.


Question: The NRC recognizes the importance of features related to ground instability, such as liquefaction, in evaluating geological safety concerns.

Liquefaction-related features have been identified locally within the Saline River source zone in Ashley County and Desha County, Arkansas (Reference 152).

The chronology of deformation for each of the sites is summarized in Table 2.5-4.


Question: What are the requirements of 10 CFR 50.55a regarding valve position indication?

Compliance with 10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication, is required upon licensee adoption of the 2012 Edition of the ASME OM Code as their code of record for the 120-month IST interval.

10 CFR 50.55a(b)(3)(xi) nor NUREG-1482 require concurrent testing and ISTC-3700 clearly states that the observations of valve indicating lights and the supplemental position indication need not be concurrent.


Question: what is a nuclide

Nuclide is part of an undetermined solution

Peak Locate Performed on: 3/29/2019 10:37:41 AM


Question: What are seismic design loads in nuclear safety?

The above studies lend substantial support to the conclusion that the basis for seismic design load calculations is in fact correct.

The effect of sublimation as well as ice melt during and after DBA on the seismic loads has been evaluated.


Question: nuclear licensing process

The nuclear licensing process is a detailed regulatory framework established by the NRC to ensure that nuclear facilities operate safely and comply with federal safety regulations.

Tomorrow, the industry representatives will provide insights into their operations and address any questions posed by the committee.


Question: What is the role of the Reactor Vessel in nuclear safety?

The Reactor Vessel is designed to contain the reactor core, the reactor internals, and the reactor core coolant-moderator and to serve as a high-integrity barrier against leakage of radioactive materials to the drywell and the release of fission products to the environment.

The nuclear safety function of the Reactor Vessel to maintain reactor core assembly geometry for any event to ensure core cooling, core reactivity control, and the integrity of the fuel cladding as a radioactive material barrier.


Question: The recovery percentages for various surrogates were measured in accordance with acceptable limits.

The acceptable limits for recovery percentage are defined as (15%-125%) for various tracers used in the analysis.

The following analytical methods were performed according to the established guidelines, ensuring compliance with industry standards and safety protocols.


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

The Safety Analysis Report (SAR) is a comprehensive document that provides detailed information about the safety analyses performed to demonstrate that the facility can be operated safely and that all possible accidents have been considered and adequately addressed.

Regular inspections of the reactor are essential for ensuring operational efficiency and longevity.


Question: The NRC mandates that all debris characteristics must be analyzed during the test conditions to ensure safety and efficacy.

Test debris quantities were determined based on plant debris loads and the ratio of the test strainer area to the net surface area of one plant strainer.

For pulverized acrylic, silica flour and paint chips, the conversion from mass to volume was done using the densities presented earlier in this section.


Question: functional reliability and testability of protection systems

The protection system shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

The reactor core consists of fuel elements arranged in a specific layout that optimizes thermal efficiency and neutron economy, ensuring that the reactor operates within safe and controlled parameters.


Question: What is the purpose of a nuclear reactor's containment structure?

The containment structure is designed to prevent the release of radioactive materials in the event of an accident by providing a robust physical barrier around the reactor core.

The document discusses various operating systems compatible with the latest software, including Windows, Mac, and Linux, each offering unique features for user interface design.


Question: The measurement locations for the isotopic carbon source monitoring system are critically laid out across the reactor building's basement.

The current operational procedures must align with the established safety protocols and guidelines set forth by the NRC.


Question: operability determination processes

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:


Question: What does the NRC require for the environmental review of license renewals?

The NRC staff has initiated a supplemental environmental review.

To the extent possible, the NRC staff requests the information identified in the Environmental Audit and Information Needs list (Enclosure 2) be made available on the Peach Bottom online reference portal prior to the audit.


Question: nuclide activity measurements

The activity of a nuclide is measured in picocuries per gram (pCi/grams), which quantifies the amount of radioactive decay occurring per unit mass of the material.

The energy levels of the nuclides, expressed in keV, are crucial for understanding their behaviors in various safety contexts, including shielding and detection.


Question: What is the significance of calm wind measurements in meteorological assessments for nuclear facilities?

Bellinger compared winds from the Oak Ridge ASOS with winds from nearby Tower W during January 2005 – February 2011, accounting for this definition of calm winds employed by the ASOS.

A local environmental firm (Holian Environmental, LLC) conducts independent audits of the tower every six months.