Question: What are the requirements for NELAP certification in analytical methods?

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

Samples 493624002 (L3-012-101-RAGS-004-SS) and 493624006 (L4-OFF-SOL-IJGS-A03-SS SW) were recounted to verify sample results. Recounts are reported.


Question: What is the significance of Technical Specifications (TS) in nuclear regulatory filings?

Each Technical Specification (TS) evaluation must be tied to a separate event and involve entry into one or more TS action statements.

The offsite dose calculation manual cannot be used to meet the minimum number of TS evaluations for SRO applicants.


Question: The NRC mandates that licensees evaluate internal fire risks as part of their safety assessments.

The Salem Internal Fire PRA model addresses risk from internal fires.

There are no major manufacturing or chemical plants within 5 miles of the site.


Question: The NRC emphasizes the need for accurate modeling of zirconium oxidation for regulatory compliance.

The Baker-Just correlation will retain its importance for comparison and licensing purposes (conservative approach). However, it should not be considered for application in best-estimate calculations.

Robert Colmar expressed his criticisms of the housing both in [his written testimony] and Tr. 11399-11419 [emphasis added].


Question: nuclear power reactor licensing process

The licensing process for nuclear power reactors is defined by a series of regulatory submissions and evaluations to ensure safety and compliance with federal regulations.

During the operational phase, the reactor's safety systems must remain functional and ready to respond to any potential incidents or anomalies.


Question: What is 11e.(2) by-product material in the context of nuclear licensing?

Prior to beginning operations, NuFuels will develop and maintain an agreement for the disposal of 11e.(2) by-product material with a facility licensed by the NRC or an Agreement State to accept such material.

NuFuels will maintain and implement a final cultural resources management plan for all mineral operating lease areas and other land affected by licensed activities, pursuant to the National Historic Preservation Act Section 106 review and consultation process.


Question: What is the required amount of boration for control rods in a reactor?

The value of the required boration is 9750 gallons due to 3 control rods not being fully inserted. Boration is required if two or more rods are not fully inserted.

If boration is required, the NUSO is permitted to direct the BOP to refer to EPM-PRP-0, Prompt Response Program, Attachment PR-6A, Prompt Response to RMCS Malfunctions, to initiate the boration.


Question: The dimensions of the nuclear channels are critical for assessing their safety and performance during operation.

| Channel Name  | 83AV  |
|---------------|-------|
| Channel Section | Zone 1 | Zone 2 |
| Dimension      | mm    | in    | mm    | in    |

The compliance date for submitting regulatory documents varies according to the license type and specific requirements established by the NRC.


Question: The ECCS must ensure the reactor core remains cool to prevent overheating after a coolant loss incident in MODE 4.

The function of the ECCS in MODE 4 is to provide core cooling to ensure that the reactor core is protected after a small break loss of coolant accident (SBLOCA).

The pressure within the Reactor Coolant System (RCS) is carefully monitored to prevent any potential breaches that could lead to safety concerns.


Question: loss of instrument air implications

Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to a loss of instrument air: (CFR: 41.5 / 41.7 / 45.7 / 45.8)

Effect on heater drain system, condensate system, and/or main and startup feedwater system (OE related)  |  2.5


Question: valve inspection frequency regulations

In addition to the proposed alternative to conduct inspections of the valve seating surface every three refueling outages via performance of a disk-to-seating surface blue check, and perform repairs as warranted, these additional, programmatic controls further assure cumulative leakage across MOV-856A, B and MOV-864A, B will not exceed the leakage quantities assumed in the safety analyses.

A review of valve performance over the past ten years identified one failure to close on demand for valve MOV-4-864A during the Cycle 32 refueling outage in 2020.


Question: What are the updated aging management criteria for Reactor Vessel Internals Components?

SLR-ISG-2021-01-PVWR: Updated Aging Management Criteria for Reactor Vessel Internals Components for Pressurized Water Reactors, ADAMS Accession No. ML20217L203.

1.7.15 **LR-ISG-2013-01**: Aging Management of Loss of Coating or Lining Integrity for Internal Coatings/Linings on In-Scope Piping, Piping Components, Heat Exchangers, and Tanks, ADAMS Accession No. ML14225A059.


Question: The process of gamma spectrum analysis includes various steps to ensure accurate results.

Gamma spectrum analysis involves assessing the detected gamma rays to identify isotopes and quantify their activity by evaluating the spectrum produced by a detector.

The sample size taken from the Dairyland_NPP facility weighed 7.370E+02 grams, recorded on 5/18/2019.


Question: The activity of the K-40 nuclide is measured in picoCuries per gram.

K-40 has a yield of 10.67% and an activity of 6.25E+00 pCi/grams with an uncertainty of 9.49E-01.

The table includes various nuclides, each with different confidence levels and activity measurements.


Question: What is the purpose of gamma spectrum analysis in nuclear safety?

Gamma spectrum analysis is performed to identify and quantify gamma-emitting radionuclides in a sample, which is critical for assessing potential radiological hazards.

The sample identification includes a detailed description of the sample and the operational parameters, indicating that it was taken from the Dallyland_NPF facility.


Question: The NRC utilizes probabilistic risk assessments to inform decision-making regarding technical specifications.

Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” Revision 2, March 2009, emphasizes the importance of probabilistic risk assessments in evaluating safety measures.

The Diablo Canyon Power Plant is a prominent facility located along the California coastline, known for its contributions to the local energy grid.


Question: What are the regulatory procedures followed by the NRC for ensuring the safety of the nuclear facilities?

The Nuclear Regulatory Commission (NRC) employs a variety of regulatory actions, including licensing, enforcement, and oversight, to maintain safety and compliance at nuclear power plants and facilities.

Nuclear energy has been a subject of debate over its efficacy and safety compared to renewable energy sources, leading to differing opinions on future energy policies.


Question: Plan of Actions and Milestones in nuclear safety

DNFSB’s draft System and Information Integrity (SI) policy contains the following statement: “Flaw remediation shall be incorporated into the organizational configuration management process. Security-relevant software and firmware updates that are unable to be installed within their appropriate remediation window shall be documented in a Plan of Actions and Milestones (POA&M).”

DNFSB updated its Risk Management Framework (RMF) Handbook on September 29, 2022. The RMF Handbook contains the following language: “The assessor and CISO must review the results of the security control assessment to determine the appropriate steps to address the weakness(es). The CISO and designated officials may determine an assessment result identified as “Other than Satisfied” is inconsequential to the security of the system and is a candidate for risk acceptance.


Question: What is Strontium-90 and its significance in nuclear analysis?

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


Question: What are the procedures for welding inspections during pressurizer replacement?

Several procedures are listed in Appendix A to this procedure that contain useful guidance for welding inspections.

In addition, licensee outage activities and actions to reduce shutdown risk should be reviewed in accordance with IP 71111.20, “Refueling and Other Outage Activities.” This is to ensure that licensees consider risk in developing outage schedules; adhere to administrative risk reduction methodologies they develop to control plant configuration; have developed mitigation strategies for losses of key safety functions; and adhere to operating license and technical specification requirements that ensure defense-in-depth.


Question: What requirements does the NRC set for assessments related to seismic and geologic features?

The NRC reviews site-related information, including assessments of geological features and tectonic activity to determine the potential safety implications for a proposed nuclear facility.

The studies conducted for the Dunbarton Triassic Basin focused primarily on its gravitational properties and magnetic anomalies, which are unrelated to nuclear safety assessments.


Question: The AFD is essential for maintaining acceptable power distribution within the nuclear reactor core.

The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core.

The reactor coolant temperature influences several operational parameters but is less critical to the axial power distribution limit determinations.


Question: aging management of piping components

AMR item 3.3.1-064 in Volume 1 of NUREG-2191 manages loss of material and flow blockage of steel and copper alloy piping and piping components exposed to raw water.

Please discuss whether other portions of the hose reel, including hose reel support, will be managed for aging.


Question: What is the purpose of the Reactivity Management Program in nuclear operations?

The Reactivity Management Program ensures that the reactor maintains proper reactivity control to prevent any safety risks during operation and to optimize fuel utilization.

Westinghouse Letter 97TV-G-0066 discusses the company's stance on core tilt and its impacts on operational performance.


Question: The NRC mandates that personnel must conduct thorough assessments of safety protocols for nuclear operations.

The NRC requires all licensees to submit a comprehensive safety analysis report that details the safety systems and their operational reliability.

The Pilgrim Nuclear Power Station is located in Plymouth, Massachusetts, and operates under the license number DPR-35.


Question: What happens when a reactor coolant pump starts at an incorrect temperature?

Figure 15-80 illustrates the startup conditions of an inactive reactor coolant pump at an incorrect temperature, detailing the pressurizer pressure fluctuations observed during the initial seconds of the process.

The time intervals measured during the pump activation provide important data but do not directly pertain to safety concerns regarding reactor operations.


Question: What is a Safety Analysis Report in the context of nuclear regulatory filings?

A Safety Analysis Report (SAR) is a comprehensive document that describes the safety analysis of a nuclear facility, including the identification of potential hazards, the assessment of risks, and the evaluation of safety systems and procedures developed to mitigate those risks.

The financial report presented in this document outlines the expenditures associated with the operational phases of the nuclear reactor and does not touch upon any safety analyses or regulatory requirements.


Question: projected population growth rates

Projected population values are based on the population projection growth trend for the years reported by the IDPH (IDPH 2021; USCB 2020c).

The average annual growth rate for Will County shows variations over the decades, with significant increases expected by 2030.


Question: What key uncertainties could affect the publication of a final rule by the NRC?

The memorandum included a discussion of key uncertainties that could impact publication of the final rule by that date.

On November 2, 2020, staff submitted a Commission memorandum responding to the SRM direction to provide a schedule with milestones and resources to complete the final rule by October 2024 (ADAMS ML20288A251).


Question: boron dilution transport methodology

The results of the boron dilution transport methodology in Section 6.0 demonstrate at least 25 ppm of minimum margin between the core-region boron concentration and the critical boron concentration; or, for cases with less than 25 ppm minimum margin, analysis demonstrates the methodology is conservative with respect to delayed onset of mixing between CNV volumes, due to liquid density differences in the volumes.

In addition, boron precipitation is evaluated and representative results show boron precipitation does not occur for the conditions evaluated for extended passive cooling, thereby demonstrating the core remains in a coolable geometry.


Question: The NRC requires a review of findings and observations (F&Os) to ensure the effectiveness of risk-informed applications that may rely on the PRA model.

The configuration control, or maintenance and upgrade, processes are included as part of PRA peer review, and NRC has access to documentation of these reviews as part of their review of License Amendment Requests.

The NRC reviewed the Nuclear Energy Institute’s (NEI) comment together with the related comments from the PWR Owners Group (See NRC Resolution for comments 17 and 18 in this document).


Question: The required backfill pressure for the CANISTER must adhere to specified limits during operations.

The CANISTER helium backfill pressure shall be 15 (+2, -0) psia. Prior to helium backfill, the CANISTER vacuum pressure shall be ≤ 3 torr.

All fuel assemblies must be removed from the NAC-MPC SYSTEM if the required actions are not completed in the specified time.


Question: The NRC oversees compliance with safety regulations for nuclear materials and facilities.

The Certificate of Compliance for Radioactive Material Packages ensures that transportation of nuclear materials adheres to safety standards set forth by the NRC.

The proposed technical specifications for the WCS Consolidated Interim Storage Facility cover various aspects of storage and management of nuclear materials.


Question: What is the significance of a clad gap release in nuclear reactor safety?

Plausible because a 1% clad gap release to the coolant would cause a significant increase in radiation levels inside the secondary shield, and it may be thought that the monitors would respond to this.

Also, the Charging Pump Room area monitor would be expected to respond.


Question: diverse actuation system

The applicant’s or licensee’s D3 analysis should either (1) demonstrate adequate diversity exists to mitigate plant vulnerabilities without the need for a diverse actuation system, or (2) determine the need for a diverse actuation system to provide adequate mitigation against plant vulnerabilities.

Addressed in PSAI 7.9.3 and satisfied by: - Licensing Technical Report


Question: in situ uranium recovery process

The chief components of an in situ uranium recovery facility include: a. **ISR process**: where a lixiviant stream is continuously recirculated from the recovery plant into injection wells, through the ore bearing strata, and a uranium-rich (pregnant) lixiviant is withdrawn (via production wells) and recirculated to the recovery plant;

Intergranular uranium mineralization makes up only a small portion of the total mass of uranium ore; therefore, after its extraction, the structural integrity of the host aquifer is maintained, and no land subsidence occurs.


Question: minimum detectable concentration

MDC stands for Minimum Detectable Concentration, which is a statistical measure of the lowest concentration of a substance that can be reliably detected in a sample.

The report lists various parameters essential for understanding the qualitative and quantitative aspects of the samples analyzed in the project.


Question: nuclear licensing requirements

Nuclear licensing involves a series of regulatory processes and documentation needed to obtain permission for the construction and operation of nuclear facilities, ensuring compliance with safety and environmental regulations.

The reactor core is designed to optimize fuel utilization while minimizing the generation of radioactive waste.


Question: What is a nuclide and how is it measured?


Question: minimum initial assembly average enrichment

**Minimum Initial Assembly Avg. Enrichment**  **wt % 235U (E)** | **44 < Assembly Average Burnup ≤ 45 GWd/MTU**

**Minimum Cooling Time (years)** — The durations indicated span a range of cooling periods necessary before the fuel assemblies can undergo further processing.


Question: Emergency Action Levels in nuclear regulatory frameworks

The NRC staff considers NEI 99-01, Revision 6, as an acceptable method to develop site-specific EALs that meet the requirements of Section IV.B of Appendix E to 10 CFR Part 50 and planning standard 10 CFR 50.47(b)(4), with the understanding that licensees may need to develop EALs that differ from the guidance document as allowed in R 1.1.

The NRC staff reviewed the application, as supplemented, and verified that the proposed EAL scheme is consistent with the guidance provided in NEI 99-01, Revision 6, to ensure that the proposed EAL scheme meets the requirements of Section IV.B of Appendix E to 10 CFR Part 50 and planning standard 10 CFR 50.47(b)(4).


Question: The NRC mandates that facility operators implement robust safety measures for the management of spent fuel.

The Independent Spent Fuel Storage Installation (ISFSI) at Peach Bottom Nuclear Plant must comply with regulatory requirements specified for the storage and management of radioactive materials.

The inspection process involves a thorough evaluation of the operational metrics and maintenance records of the facility.


Question: What is a Safety Injection System?

The Safety Injection System (SIS) is designed to supply water to the Reactor Coolant System (RCS) during a loss of coolant accident to maintain the core's cooling. It plays a critical role in ensuring the reactor's safety by providing a means of filling the system to prevent overheating.

The overall heat transfer coefficient (U-factor) is a measure used to determine how well a building or construction element conducts heat, which is crucial in energy efficiency evaluations.


Question: quality control in nuclear regulatory filings

Quality control sample(s) analyzed as WG34651, WG34680, WG34856, WG34857.


Question: The NRC mandates that the operators conduct a walkthrough of the surveillance procedures for room coolers after assuming the shift.

After taking watch, CRS reviews completed Room Cooler surveillance procedure (S2.OP-ST.SW-0014) (TS only)

The event list includes various malfunctions and their corresponding actions that the crew must undertake.


Question: average wind speed measurements in nuclear safety reports

The average wind speed recorded at the Dresden Nuclear Power Station from June 1, 2017, to August 31, 2021, is 2.71 m/s.

The total count of hours recorded for wind measurements during this period is 10,992 hrs.


Question: total property tax payments for nuclear facilities

Total Paid by DNPS (USD) ranged from $2,994,443 to $3,507,083 between 2018 and 2022, reflecting the financial contributions of the facility to local tax revenues.

The reactor's cooling system is designed to maintain safe operating temperatures, ensuring the integrity of the fuel and the surrounding environment.


Question: The highest water levels upstream of the canal are significantly influenced by the Keowee Dam, which plays a central role in controlling the flow from the Keowee River Basin.

The Keowee Dam (upstream) is the primary control on the water level upstream of the canal, which allows flow from the Keowee River Basin into the Little River Basin and eventually into the intake canal.

The final maximum water surface levels, determined from this updated computational mesh and the parameters listed earlier are:


Question: nuclear regulatory filing processes

New Question 202.22 was added to highlight and summarize the noteworthy changes made to NRC Form 396 and NRC Form 398 as a result of the October 2017 revisions.

Incorporated formatting changes and hyperlinks throughout to improve ease of use.


Question: The process of subtracting background counts from the total to determine net peak area is crucial for accurate peak analysis.

The net peak area is derived by subtracting the background uncertainties from the original area, ensuring accurate measurement of the count for each peak indicated.

The identified nuclides are used for comparison against the reference library to ascertain their presence in the sample.


Question: What is the meaning of Plasticity Index in geotechnical engineering?

b) PI = Plasticity Index; psf = pounds per square foot; ksf = kips per square foot; ft./sec. = feet per second

a) Laboratory Vs results not representative of in-situ strata, and not used for dynamic profile (Subsections 2.5.4.2.2.2 and 2.5.4.7.1)


Question: The NRC mandates monitoring and calibration of voltage outputs during reactor startup procedures.

Simulates pressing GEN FIELD FLASH pushbutton until Generator voltage stops rising. (GENERATOR LINE VOLTS should indicate ~260V).


Question: MNGP recreational use regulations

While no public visitors were allowed onsite at MNGP for most of 2020 and none in 2021 due to the Covid-19 pandemic, in recent years annual public visitor use of the MNGP conference facility could range as high as an estimated 147 persons in 2016 to an estimated 71 persons in 2019.

The Minnesota portion of the Mississippi River Trail follows the Mississippi River from Itasca State Park to the Iowa border on a combination of roadway shoulders and off-street trails.


Question: What are the requirements for impingement mortality assessments for nuclear power plants?

If the applicant's plant utilizes once-through cooling or cooling pond water intake and discharge systems, the applicant shall provide a copy of current Clean Water Act 316(b) Best Technology Available determinations and, if applicable, a 316(a) variance in accordance with 40 CFR part 125, or equivalent State permits and supporting documentation.

The potential effects of impingement and entrainment during an initial LR or SLR term depend on numerous site-specific factors, including the ecological setting of the plant; the characteristics of the cooling system; and the characteristics of the fish, shellfish, and other aquatic organisms present in the area (e.g., life history, distribution, population trends, management objectives, etc.).


Question: The NRC mandates that licensees demonstrate compliance with the adequate safety measures during reactor operations.

The licensing process involves the submission of detailed documentation, which demonstrates an applicant's ability to operate safely and in accordance with the regulatory framework set forth by the NRC.

The tables in this document provide numerical data related to reactor performance metrics, but do not address licensing or safety protocols.


Question: The NRC mandates the evaluation of potential aging effects on materials within reactor components.

The Heat Exchanger (Boric Acid Vent Condenser) Shell Side Components are subjected to loss of material due to the harsh environment of demineralized water and require detailed aging management programs for monitoring.

The aging management review results contribute to enhancing the overall safety culture within nuclear facilities.


Question: nuclear licensing regulatory process

The Nuclear Regulatory Commission (NRC) evaluates license applications for the construction and operation of nuclear facilities to ensure that they meet safety standards and regulatory requirements before granting permission to proceed.

The cooling system of a nuclear reactor is designed to remove heat from the reactor core and ensure stable operation, utilizing various heat exchange technologies to maintain optimal temperatures.


Question: fuel fragmentation and relocation models

The applicability of fuel fragmentation and relocation models considering the full set of experimental test data presently available and higher fuel burnups being proposed in WCAP-18850-P.

Westinghouse-proposed limitations and conditions applicable to WCAP-18850-P, focused particularly on areas where proposed treatments differ from similar TR methodologies described in WCAP-18446-P-A and WCAP-16996-P-A.


Question: What includes aging management for reactor vessels?

The Aging Management Program for the RV Bottom Mounted Instrument Nozzle focuses on managing various aging effects, including cumulative fatigue damage and loss of material due to boric acid-induced corrosion.

The Diablo Canyon Power Plant has undergone significant modifications to its cooling systems to enhance operational efficiency.


Question: What actions must be taken when the Instrument Air System experiences a leak during operations?

At the direction of the Lead Evaluator, insert an Instrument Air System leak at a 5% severity ramped over 2 minutes: IA02 5:

Reports both Instrument Air Receiver and Scram Air Header pressures are rising.


Question: The ICP-MS methodology employs rigorous analytical processes to ascertain metal concentrations accurately.

Method SW-846 3050B is not a total digestion technique for most samples. It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available.

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


Question: flood-causing event combinations

NUREG/CR-7046, Appendix H (Reference 128) states that the following combinations of flood-causing events provide an adequate design basis for shore and streamside locations:

A query of the NOAA NGDC (Reference 72) produced no recorded tsunami events or observations in the Susquehanna River near PBAPS.


Question: advanced reactor licensing process

The Advanced Reactors Licensing Branch 1 is responsible for overseeing the regulatory process for advanced reactors, ensuring that safety and operational standards are met.

The Office of Nuclear Reactor Regulation works on various aspects of nuclear reactor safety and compliance.


Question: LPBF machine scan strategy

“Scan strategy” refers to all settings used by the LPBF machine (e.g., laser power, laser acceleration, laser beam diameter, acceptable O2 levels) as well as the two-dimensional (2D) path that the laser travels.

A thin layer of metal powder is spread from the powder tank across the surface of the build plate via a recoating mechanism.


Question: The NRC's oversight includes assessing potential impacts on historic properties during mining activities.

the NRC staff finds that the NHPA process is concluded with respect to the Section 24 properties.


Question: What is the significance of preoperational testing for emergency core cooling systems in nuclear reactors?

REGULATORY GUIDE 1.79 (September 1975, Revision 1). PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS This guide is not applicable to Fermi 2, which is a BWR.

The closest transportation line lies about 3.5 miles from the plant.


Question: What is fire risk quantification in nuclear safety?

Fire risk quantification calculates the fire-induced CDF and LERF (or LRF, as applicable) contributions to plant risk and identifies significant contributors to each. In this element, the plant response model is quantified for the set of fire scenarios to produce conditional core damage probability and conditional large early release probability (CLERP) or conditional large release probability (CLRP) values, as applicable.

Circuit failure analysis treats the impact of fire-induced circuit failures upon the plant response.


Question: gamma spectrum analysis

This gamma spectrum analysis was performed using a 500 ml Marinelli geometry detector, with an efficiency calibration done on 7/8/2014 and a peak location threshold set to 3.00 to identify relevant peaks in the energy range.

The sample was taken on 5/29/2019 and the acquisition started on 6/1/2019, with data collected over a live time of 1600.0 seconds and a real time of 1805.1 seconds.


Question: The compliance of the Peach Bottom ECCS CLS Replacement System with the NRC regulations is outlined in the RadICS Platform Topical Report.

The Curtiss-Wright System Development Methodology meets the requirements specified in the RadICS Platform Topical Report and meets all NRC requirements for licensing the Peach Bottom ECCS CLS Replacement System.

The following slides identify those sections of the RadICS Platform Topical Report that specify requirements for applications that utilize the RadICS Platform and define how the System and Application Development Methodology used for the Peach Bottom ECCS CLS Replacement Project complies with those requirements.


Question: The regulatory framework involves evaluation of offsite doses, particularly in situations where the total dose exceeds specified limits such as 1 rem over a month.

Event Sequences with Offsite Doses > 1 rem over the month following event

For plants with event sequence doses > 1 rem over a month beyond the site boundary, population density < 500 ppsm over the radial distance equal to twice the radius at which 1 rem over a month is estimated (or 20 miles)


Question: The minimum detectable activity for isotopes is quantified in pCi/grams, reflecting the sensitivity of the measurement.

Nuclide MDA (pCi/grams) represents the minimum detectable activity for each nuclide listed, ensuring precise quantification in radiological assessments.

Various nuclides exhibit different energy levels measured in keV, which are crucial for analyzing detection capabilities.


Question: The examinations rely on well-defined learning objectives that align with NRC standards for operator licenses.

Senior reactor operator (SRO) questions test at the SRO license level.

The sampling process was random and systematic.


Question: What is the significance of AXIOM™ fuel rod cladding in nuclear safety?

During the meeting, Duke Energy discussed the planned adoption of AXIOM™ fuel rod cladding as part of their license amendment requests, highlighting its potential benefits to the safety and efficiency of fuel assemblies.

The meeting also covered a variety of logistical issues related to the attendance and coordination of future public meetings, including the preferences of various stakeholders.


Question: The adequacy of offsite circuits and diesel generators is critical for maintaining power to the onsite Class 1E Distribution System during outages.

The remaining OPERABLE offsite circuit and DGs are adequate to supply electrical power to both trains of the onsite Class 1E Distribution System.

The 30-day window should permit extensive preplanned preventative maintenance without placing either unit in an action statement of short duration and would allow both units to be operating during such maintenance.


Question: The DNFSB's recommendations on security measures were previously tracked for compliance and response.

Based on this response, all recommendations (1 through 14) are open and resolved.

The Inspector General’s analysis showed trends in operational audits across various government sectors.


Question: What were the findings of the FLECHT run 9573 regarding heat transfer coefficients?

It is significant that in FLECHT run 9573—a test conducted with a Zircaloy bundle—negative heat transfer coefficients were observed at the bundle midplane for 5 of 14 thermocouples and that steam temperatures exceeded 2500°F at the seven-foot steam probe, during a portion of the test.

Additionally, it is rather odd that James S. Moore would claim that “[f]or the hot spot calculation, [negative heat transfer coefficients: calculated as a result of heat transfer from the coolant to the fuel cladding would] never…exist.” And that “[negative heat transfer coefficients: calculated as a result of heat transfer from the coolant to the fuel cladding occurred] well above the hot spot.


Question: What are missile barrier shelters in nuclear facilities?

Manholes, Handholes & Duct Banks are intended as Missile Barrier Shelter/Protection.

Water - Flowing can lead to an increase in porosity and permeability, resulting in a loss of strength.


Question: nuclide identification


Question: evacuation time estimate study

Applicants and licensees need to conduct an evacuation time estimate (ETE) study and maintain the ETE up to date. The ETE is primarily used in the development of protective action strategies and to inform offsite protective action decisionmaking.

Applicants and licensees need to describe in their emergency plans an offsite facility and any backup facilities for coordination of the response with OROs.


Question: What is the purpose of a License Application for a nuclear facility?

The License Application includes a comprehensive description of the facility and its operations, addressing the safety and environmental impacts in accordance with regulatory requirements.

The maintenance schedule for the reactor includes routine checks and alignments to ensure optimal performance throughout its operational life.


Question: RCP seal injection isolation valve

The RCP seal injection isolation valve is classified as a safety-related component and is used to enable or disable the seal injection to the RCP (Reactor Coolant Pump) system as required for operational safety and maintenance activities.

The operational parameters of safety systems are critical to ensuring plant integrity during normal and emergency conditions.


Question: What is the significance of safety classifications in nuclear valve programs?

Safety Class 1 components are those that are vital to the safety of the nuclear reactor as they perform safety-related functions and are required to satisfy the requirements of 10 CFR Part 50.

The valve program plan includes maintaining a detailed inventory of all valves, which is crucial for ensuring operational efficiency.


Question: Minimum initial assembly average enrichment for PWR fuel

Minimum Initial Assembly Avg. Enrichment wt % ^235U (E) ranges from 1.3 ≤ E < 4.9, with corresponding assembly average burnup levels listed in the table.

The assembly average burnup (B) is measured in GWd/MTU and affects reactor performance.


Question: The data shows the distribution of rainfall over several sites during the specified period.

Data represents hourly rainfall in specified areas.

Rainfall measurements are recorded in inches, indicating the amount of precipitation collected.


Question: For questions regarding regulatory permits, you can reach the Permit Section through its manager.

Should you have questions concerning the Permit, please contact Leslie Lowry at 217/782-0610.

Sincerely, Alan Keller, P.E. Manager, Permit Section Division of Water Pollution Control


Question: nuclear quality control

The QC Summary Report for L88709 provides details of the quality control results associated with samples analyzed in the regulatory process.

The samples listed span various units of measurement reflecting different client identifiers.


Question: gamma measurement techniques

Gamma measurements are reported in counts per minute (cpm), typically provided at specific depth intervals over the sampling range to assess radiation levels.

The Sun is a star at the center of the Solar System and is essential for life on Earth.


Question: The NRC requires applicants to assess the potential environmental impacts of postulated accidents in their licensing process.

Duke Energy's 1999 environmental report submitted as part of its initial license renewal application included an assessment of SAMAs for Oconee (Duke Energy 2021-TN8897).

Human health risks would be smaller following nuclear power plant shutdown.


Question: reactor building floor drain sump

The system includes a Reactor Building Floor Drain Sump which is responsible for collecting and managing any water that may accumulate due to spills or leaks in the reactor building.

The operator verified that the Joy Air Compressor is operating efficiently and is providing necessary air supply for system operation.


Question: Regulatory Guide 1.68 outlines the requirements for preoperational and initial start-up test programs for water-cooled power reactors.

The preoperational and start-up programs, as outlined in Tables 14.1-1 and 14.1-2, comply with the intent of Regulatory Guide 1.68, *Preoperational and Initial Start-up Test Programs for Water-Cooled Power Reactors,* dated November 1973, in most cases, and use the same wording as much as possible in order to more clearly address the NRC guide requirements.

The management and direction of the preoperational test program was under the direct control of VEPCO, with the principal responsibility lying with the Supervisor - Engineering Services.


Question: plume exposure pathway emergency planning zone

The TR methodology provides an approach for determining a plume exposure pathway (PEP) emergency planning zone (EPZ) size. This size is based on the area within which public dose, as defined in Title 10 of the Code of Federal Regulations (10 CFR) Section 20.1003, 'Definitions,' is projected to exceed 10 mSv (1 rem) total effective dose equivalent (TEDE) over 96 hours from the release of radioactive materials from the facility, considering accident likelihood and source term, timing of the release sequence, and meteorology.

The draft final rule includes a second EPZ criterion that the plume exposure pathway EPZ is the area where predetermined, prompt protective measures are necessary.


Question: depth of boring wells

Boring BRA-001 has a depth of 10-15 ft below ground surface (bgs).

Pickard Hall was constructed in the early 2000s with a modern architectural design.


Question: The plant's design incorporates features to manage flood risks effectively.

The design basis was reviewed to determine which flood-causing mechanisms are considered in the current design basis.

Emergency electrical power distribution centers manage the power flow to critical systems in the event of a disruption.


Question: what is the allowable value for loss of voltage

**TS AV for Loss of Voltage** ≥ [Nominal Trip Setpoint – (RCA X span voltagepri)] - ≥ [2975 VAC – (0.0032 x 5250 VAC)] - ≥ 2958.2 VAC

Nominal Trip Setpoint is from Technical Specification Surveillance Requirement 3.3.5.2 and is not changed by this request.


Question: What staffing positions are necessary for overseeing repair team activities in the context of emergency response?

Supervision of Repair Team Activities includes critical positions such as Lead OSC Supervisor, Mechanical Supervisor, and Electrical Technician all equipped to manage repair and response efforts efficiently.

Media Information roles such as JIS / JIC Manager and JIS / JIC Coordinator are vital for communicating with stakeholders and maintaining operational transparency.


Question: What is the role of the U.S. Nuclear Regulatory Commission in reactor licensing?

The U.S. Nuclear Regulatory Commission (NRC) is responsible for regulating the nuclear industry, which includes issuing licenses for the construction and operation of nuclear reactors to ensure public safety and environmental protection.

Public Watchdogs petitions for review of the U.S. Nuclear Regulatory Commission.


Question: What is the purpose of the aging management review in nuclear licensing?

This section provides the results of the aging management review for those structures and components identified in Section 2.0 as being subject to aging management review.

In order to take full advantage of NUREG-2191, a comparison between the BFN AMR results and the tables of NUREG-2191 has been performed.


Question: Very Low-Level Waste regulations

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

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


Question: nuclear aging management

| Supports ASME 2 & 3 Constant         | Expansion/Separation Structural Support | Carbon Steel    | Plant Indoor Air (Structural) (Ext)         | Loss of Mechanical Function       | ASME Section XI, Subsection IWF (B.2.3.30)       | III.B1.2.TP-28     | 3.5-1, 057  | B     |

| Supports HVAC Duct                   | Structural Support                  | Carbon Steel    | Plant Indoor Air (Structural) (Ext)         | Loss of Material                 | Structures Monitoring Program (B.2.3.33)         | III.B2.TP-43       | 3.5-1, 092  | A     |


Question: emergency response organization

The amendments revised certain emergency response organization (ERO) positions in the Palo Verde Nuclear Generating Station (Palo Verde) Emergency Plan.

The amendment revised the technical specifications by reducing the reactor control system and secondary side specific activity by 50 percent.


Question: Annulus Ventilation System in nuclear safety

The Annulus Ventilation System is designed to accomplish the following: 1. Produce and maintain a negative pressure in the annulus following a LOCA.

The system functions to discharge sufficient air from the annulus to effect a negative pressure with respect to the containment and the atmosphere in approximately 120 seconds following a LOCA.


Question: What is the significance of the New Madrid Rift in seismic safety assessments?

The New Madrid Rift complex is a significant geological feature that influences seismic safety assessments and regulatory licensing for nearby nuclear facilities due to its potential to generate substantial seismic events.

The closest distance between site and sources is measured from the Bechtel seismic source map, which provides important data for hazard assessments.


Question: The overview of the scan results includes key data for various areas categorized by grid labels.

All areas identified in the FSS plan were scanned for elevated radiation levels.

Complete scan results are provided in Attachment 2.


Question: emergency planning zone analysis

The Emergency Response Organization (ERO) is evaluated to ensure adequate staffing and capability during emergencies, reflecting the current standards set forth in the PSL Emergency Plan.

The site-specific emergency plan incorporates various elements to ensure compliance but does not delve into the details of reactor construction materials or their specifications.


Question: Periodic surveillance testing requirements for plant-specific applications must be established by maintenance activities.

Maintenance activities including periodic surveillance testing will be defined based on plant-specific application requirements.

Addressed in 7.9.1 and satisfied by: Licensing Technical Report.


Question: The process of licensing a nuclear reactor involves detailed reviews of various safety analyses.

The licensing process for a nuclear reactor includes an extensive review of safety systems and operational protocols to ensure compliance with regulatory standards.

Computers and software play a critical role in modern reactor designs, improving efficiency and safety through advanced simulation techniques.


Question: What does counting uncertainty mean in nuclear analysis?

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

The Company name is LaCrosseSolutions and is located at S4601 State Hwy 35, Genoa, Wisconsin 54632.


Question: What is the purpose of a regulatory audit in nuclear licensing?

A regulatory audit is a planned license- or regulation-related activity that includes the examination and evaluation of primarily non-docketed information associated with the LAR.

The proposed LAR would revise the current technical specifications (TSs) to reflect adoption of NUREG-1433, 'Standard Technical Specifications – General Electric BWR/4 Plants,' Revision 5 (ML21272A357).


Question: narrow range RTD time constant

The Narrow Range RTD time constant specifies the response times for various inputs to the Design Temperature (DT), indicating how quickly the system reacts to changes in temperature and pressure.

The reactor trip system instrumentation response times for the manual reactor trip are not applicable.


Question: The NRC stipulates specific criteria regarding the allowed types of fuel and their characteristics for assemblies used in their facilities.

The Connecticut Yankee INTACT FUEL ASSEMBLIES, DAMAGED FUEL ASSEMBLIES, INTACT FUEL RODS, DAMAGED FUEL RODS, or FUEL DEBRIS shall be, or shall be from, assemblies that meet the criteria specified in Table B2-4 for the fuel assembly type or vendor, and meet the following additional specifications:

The average temperature of the reactor coolant system during operation is critical for maintaining safety and efficiency.


Question: What are the common units used for measuring constituent levels in water quality assessments?

The units for measuring various constituents in water quality assessments are typically expressed in micrograms per liter (ug/L).

Silica is a common mineral found in various geological formations and can significantly influence water hardness.


Question: nuclide activity measurements

The activity of a nuclide is commonly expressed in μCi/gram, which indicates the number of microcuries emitted per gram of the nuclide.

The table includes various nuclides and their corresponding energies, yields, and minimum detectable activities (MDA) for regulatory compliance.


Question: nuclide activity measurement


Question: What is the importance of reactor internals acceptance criteria?

Transmittal of WCAP-17096-NP, Revision 3, 'Reactor Internals Acceptance Criteria Methodology and Data Requirements' is crucial for ensuring that reactor internals are evaluated against established safety standards and operational guidelines to maintain safety and efficiency.

Technical Bulletin No. TB-14-5 discusses the degradation of reactor internals lower radial support clevis insert cap screw, highlighting various factors and implications involved in maintenance and inspection procedures.


Question: The methodology outlines the process for sizing the Emergency Planning Zone related to plume exposure for the reactor design.

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.

TerraPower has developed this methodology to provide an acceptable approach for determining the PEP EPZ size when supported by design- and site-specific information for Kemmerer Unit 1.


Question: What are acceptable limits for decay heat in a nuclear reactor?

Decay heat refers to the residual heat generated by the decay of radioactive isotopes in the reactor core after it has been shut down. It is important to manage decay heat to prevent overheating and ensure safety post-shutdown.

Nuclear power plants must comply with a variety of safety regulations to ensure the well-being of the public and the environment during operation.


Question: The strength of a radioactive source can be measured in terms of its specific source strength in MeV per gram per second.

Specific Source Strength (MeV/g·s) is a measurement that quantifies the energy emitted by a radioactive source per unit of mass over time.

Discrete energy measurements can vary significantly between different sources, reflecting the diverse radiative characteristics.


Question: reactor power reduction procedures

Rod Control should remain in automatic for Tavg Control. Reactivity Briefing Sheet, 'Thumb Rules' (page 3), lists boron flows and volumes for different reduction rates.

Condenser backpressure should be maintained less than the limits of Table 1 and Table 2 during load reduction – see next 2 pages.


Question: The NRC stipulates that licensees must ensure that pressure control systems maintain a safe operational environment...

None - The redundant train will operate and provide air flow


Question: tendon end anchorage condition

Tendon end anchorage hardware and adjacent concrete have performed well throughout the life of the plant (through the most recent surveillance in 2021) and show no trends of deteriorating condition.

Corrosion protection medium (CPM) test samples were collected at the ends of each examined tendon during each of the 9 surveillances.


Question: Licensees are required to comply with safety evaluations as part of their application process.

The NRC evaluates submittals to determine whether the proposed activities can be conducted safely and in accordance with regulatory requirements.

The contact information provided includes the agency name, the individual's name, and their address.


Question: What does PCT mean in the context of LOCA analysis?

The PCT of 1680°F occurred in a fresh UO₂ rod.

The UTL results providing 95/95 simultaneous coverage from this evaluation meet the 10 CFR 50.46(b) criteria with a maximum local oxidation of 4.33 percent.


Question: What is the purpose of the operations phase quality assurance program in nuclear plants?

The operations phase quality assurance program for Vogtle Electric Generating Plant (VEGP) is designed to assure the plant's safe and reliable operation and to satisfy the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50.

Quality assurance program requirements formerly contained in VEGP FSAR Section 17.2 are superseded by those contained in the SNC QATR.


Question: What is a Safety System Functional Inspection (SSFI)?

A Safety System Functional Inspection (SSFI) evaluates the effectiveness of safety systems in ensuring proper operational functionality and compliance with regulatory requirements.

The Maintenance Program outlines the procedures for scheduled upkeep of equipment to ensure continuous operation and regulatory compliance.


Question: What is cobalt-60 used for in medical applications?

This amendment adds authorization for Stephanie Lachowicz-Mahoney, M.S., to use cobalt-60 in a gamma stereotactic radiosurgery unit (Elekta Model Esprit only) for calibration, spot checks and training.

You will be periodically inspected by the U.S. NRC. Failure to conduct your program in accordance with U.S. NRC regulations, license conditions, and representations made in your license application and supplemental correspondence with U.S. NRC will result in enforcement action against you.


Question: The NRC will review the request in accordance with established regulatory procedures to determine its appropriateness.

The proposed LAR would extend the Completion Time for Technical Specifications (TS) 3.6.4.3, “Standby Gas Treatment (SGT) System,” Required Action D.1 on a one-time basis to support repair of the SGT common discharge exhaust line.

An audit is conducted to gain understanding, to verify information, and to identify information that will require collecting to support the basis of a licensing or regulatory decision.


Question: uranium dissolved measurement

Uranium, dissolved measurements are regulated under the EPA 200.8 protocol, which specifies testing for uranium concentration in units of mg/L.

The sulfate analysis involved various quality control checks to ensure that the high and low limits on the measured compounds meet the required specifications.


Question: nuclide activity measurements

The activity of a nuclide is typically expressed in picocuries (pCi) per gram, which quantifies the rate of decay of a radioactive substance.

The line MDA (pCi/grams) represents the minimum detectable activity for a given measurement line, which is important in determining the sensitivity of a radiation detection instrument.


Question: The design includes a dual-welded closure system for the transportable storage canister that ensures secure containment of the nuclear materials.

The transportable storage canister (TSC) and the closure lid are dual-certified Type 304/304L stainless steel.

The lower impact limiter is bolted to the cask bottom plate by 16 equally spaced retaining rods and nuts.


Question: The analyses showed a strong correlation between break size and peak cladding temperature, indicating that larger breaks tend to result in higher temperatures during the incident.

In general, plants with high pressure SITs inject early in the transient when the break flow is still high.

These figures illustrate the parameter ranges used in the analysis.


Question: nuclear power plant safety assessment

The Vermont Yankee Nuclear Power Station underwent a comprehensive safety assessment to ensure compliance with regulatory requirements and operational standards.

The Vice President of Engineering at Entergy Nuclear Operations is responsible for overseeing the plant's engineering activities.


Question: What is the COBRA-FLX Critical Heat Flux Correlation?

COBRA-FLX is a thermal-hydraulic code used to calculate local conditions in nuclear reactor cores, particularly for predicting critical heat flux and understanding safety margins during operation.

The reactor must undergo a series of inspections and tests to ensure compliance with safety standards set forth by the regulatory authorities.


Question: action response for radioactive release report

Return at least one instrument to operable status within 30 days or provide an explanation in the next Radioactive Release Report as to why the inoperability was not corrected within 30 days.

Grab samples shall be collected at least once per 12 hours and analyzed within 24 hours of collection or auxiliary samplers are used.


Question: The likelihood of adverse impacts to public health from radiation is considered negligible due to effective modeling techniques used.

MILDOS-AREA modeling shows no impacts to the public resulting from the CUP.

Slight risk of public exposure to non-radiological hazards, including chemicals. Spills and leaks will be mitigated using BMPs.


Question: The data on inversion heights and strengths provides insights into weather patterns at various times throughout the year.

Inversion is defined as three NOAA weather balloon elevation readings showing consecutive increases in temperature with height (below 3000 m).

Balloons were released each day at 6:00 AM and 6:00 PM. Height is defined as elevation in meters where temperature first decreases and is averaged only over those days with inversions.


Question: What is meant by pressurizer water volume in nuclear reactors?

Pressurizer water volume is an important parameter that indicates the amount of water in the pressurizer, which is crucial for maintaining system pressure and regulating coolant flow.

The turbine trip accident analysis requires validation of multiple safety systems to ensure operational continuity during emergencies.


Question: What are the alternative energy sources considered for the Oconee Station replacement?

This section describes three replacement energy alternatives to the proposed action: - new nuclear alternative (Section 2.3.2.1) - natural gas combined-cycle alternative (Section 2.3.2.2) - combination alternative of new nuclear (small modular reactor (SMR)) power, solar photovoltaic, offshore wind power, and demand-side management (Section 2.3.2.3)

In 2021, North Carolina passed additional clean energy legislation that requires the State to reduce electricity-related carbon emissions from electric generating facilities 70 percent by 2030, and to reach carbon neutrality by 2050.


Question: The NRC mandates specific safety evaluations for changes to a facility's licensing process.

The NRC requires that any significant changes to the licensing of nuclear power reactors be accompanied by a comprehensive safety evaluation to ensure compliance with existing regulations.

The organization seeks to enhance public awareness about nuclear energy and its contributions to the energy sector.


Question: Seismic testing was carried out to evaluate the compressional and shear wave velocities in the subsurface layers.

Seismic velocity testing was completed using crosshole techniques at two locations: in the vicinity of boring NB-1 (test holes CH NB-1A, B and C) and in the vicinity of boring NB-8 (test holes CH NB-8A, B and C) (Figure 2.5.4.4-1). Crosshole testing was completed in general compliance with ASTM D 4428 (Reference 2.5.4.4-1).

Crosshole was completed to assist in validating the horizontal compressional (P) and shear (S) wave velocity measurements obtained with the suspension technique.


Question: Westinghouse AP1000 reactor design

The Westinghouse AP1000 is a pressurized water reactor design that employs passive safety features to ensure safe operation and response to potential accidents.

The report also discusses various training programs for operators, emphasizing the importance of hands-on experience and simulation in preparing staff for emergency situations.


Question: What information is available regarding the uncertainties related to the operational limits of PWR transients?

RAI 16 Appendix B in ANP-10323P provides an overview of the application of fuel rod average power and axial power shape uncertainties in the AREVA GALILEO methodology.

Please provide an example analysis for PWR fast and slow transients and how exposure points are selected for these transients.


Question: What is the NRC's requirement regarding the evaluation of steam generator tube-to-tubesheet welds and their susceptibility to PWSCC?

The license has chosen Option 2: Analysis, which requires Callaway to: Perform an analytical evaluation of the SG tube-to-tubesheet welds either determining that the welds are not susceptible to PWSCC or redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function.

Please clarify the ASME Code versions applicable to the analysis for the Callaway SGs. Section 6.2.2.3 of Enclosure 1 of the submittal states that fatigue crack growth is only implemented for ASME Code years 1992 and 1995 with 1996 Addenda in AREVACGC, and it provides a comparison to ASME Code year 2019.


Question: Demonstrate that the materials used in the transportation capsules are capable of withstanding extreme accident conditions.

Provide a calculation that demonstrates the integrity of the Ir-192 source capsule and Se-75 source capsule that is transported inside the QSA Model 1100 would be maintained during and after the fire hypothetical accident conditions (HAC).

The applicant's analysis must cover regulatory requirements in order to proceed with the application review.


Question: land use land cover data

The acreages presented in this table are based on the Multi-Resolution Land Characteristics Consortium land use/land cover data.

This geographic variation creates a small difference between total acreages reported in Table 3.2-1 compared to the MNGP site acreage reported throughout the ER.


Question: What is the minimum detectable concentration in radiation analysis?

Client Sample ID: L1-SUB-TDS-FSGS-B25-SB


Question: The NRC mandates that at least one containment air lock door remains functional during specific operational modes.

Each containment air lock shall be OPERABLE.

Operation may then continue provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.


Question: What is the definition of weighted mean activity in nuclear regulatory filings?

The weighted mean activity represents the average level of radioactivity of a nuclide, adjusted for the uncertainty in the measurements and expressed in terms of picocuries per gram (pCi/grams).

The nuclide is part of an undetermined solution which complicates its identification and application in regulatory analysis.


Question: What is the graded approach to nuclear inspections?

The graded approach to inspection is a process that allocates inspection resources commensurate with the safety significance of the licensed activities being conducted.

The NRC conducts regular training sessions for inspectors to familiarize them with the latest safety protocols.


Question: The Incident Command Post requires reliable communication systems, including various types of radios for effective emergency response.

The EOC also has a number of field deployable communications assets, including: - A cache of VHF two way radios. - A cache of 800 MHz two way encrypted radios. - A field deployable VTAC36 repeater. - Two field deployable patching gateways capable of both 800 MHz and VHF patches.

The backup EOC location is the Manitowoc County Heritage Center, 1701 Michigan Avenue, Manitowoc, WI.


Question: What is the GRIP Bottom Nozzle Connection in nuclear reactor design?

Figure 2-16 shows the GRIP Bottom Nozzle Connection, which is an essential component in the reactor's coolant flow system.

Framatome Inc. specializes in providing innovative solutions for various sectors, including energy and healthcare.


Question: What is the significance of carbon emissions in regulatory filings?

The annual net flux of carbon to the atmosphere from land-use change has been a significant factor in the analysis of environmental impacts related to regulatory frameworks.

Shale gas is considered a major driver of the recent increase in global atmospheric methane emissions, which has implications for climate change strategies.


Question: What is the role of energy efficiency labeling in household appliances?

The importance of energy efficiency labeling of appliances is that it will allow the consumer to select the most energy efficient appliance.

Small savings can be realized by deenergizing transformers whenever possible.


Question: risk-informed regulatory framework

10 CFR Part 53 establishes a risk-informed, technology-inclusive regulatory framework for advanced reactors, allowing for a more flexible approach to safety and compliance.

Attached please find our letter on proposed guidance documents associated with the proposed 20CFR53.


Question: Containment Air Release and Addition System

The Containment Air Release and Addition System is utilized to maintain the Containment pressure between the Technical Specification limits of -3 and +3 psig.

Functional tests are performed before initial operation.


Question: What is a license renewal process in nuclear regulatory filings?

The site-specific environmental impact statement provides information on the subsequent license renewal for the Oconee Nuclear Station, which includes a detailed evaluation of the potential environmental effects of continued operation beyond the current license period.

The official record copy has been filed and distributed to the relevant offices and personnel involved in the approval process.


Question: What is the purpose of the scoping process in environmental reviews for nuclear applications?

The scoping process provided an opportunity for members of the public to propose environmental issues to be addressed in the SEIS and to highlight public concerns and issues.

The NRC staff reviewed each comment submission, including the public scoping meeting transcripts, to identify individual comments.


Question: What is the function of differential flow isolation in nuclear reactor safety systems?

Therefore, the design solution uses a hybrid approach to add a new differential flow isolation at a higher flow rate (i.e., 182.4 gpm allowable value) with the existing 47 second time delay.

The existing components are shown black, with new components shown in red.


Question: NRC license renewal process

Additionally, NSPM does not anticipate any refurbishment as a result of the technical and aging management program information that will be submitted in accordance with the NRC license renewal process.

NSPM does not anticipate the continued operation of MNGP to adversely affect the environment or any cultural or historic resources.


Question: What is the significance of the Sum F in nuclear safety assessments?

Sum F is a crucial metric used in evaluating the performance and safety of various systems in nuclear engineering, providing insights into structural integrity and reliability under operational conditions.

The table lists numerical data and regression parameters related to tendon performance, specifically detailing values such as t, Log T, and various sums indicative of the tendon behavior.


Question: well driller licensing requirements

Rule 10 of the Idaho Well Construction Standards (IDAPA 37.03.09) defines a well driller as “Any person who operates drilling equipment, or who controls or supervises the construction of a well, and is licensed under Section 42-238, Idaho Code.”

Although some would move their drill rig if they could, it is not always possible.


Question: The projected number of fatigue cycles for the high temperature sample lines remains below the 7000 threshold, ensuring the applied stress range factors are still relevant.

Even if all operational Reactor Coolant System transients (i.e., transients 1 through 38) applied to each of these systems, the total number of projected 80-year cycles is less than 7000. Therefore, the stress range reduction factors originally applied for the components within these piping systems remain applicable and these implicit TLAA remain valid through the subsequent period of extended operation.

In addition, various high temperature sample lines were designed in accordance with ANSI B31.1 requirements.


Question: What are the requirements for maintaining records of dose limits for individuals in nuclear facilities?

10 CFR 20.2107 requires licensees to maintain records demonstrating compliance with the dose limits for individual members of the public.

This method, which allows the licensee to take credit for shielding between the gauge and the area in question, begins by measuring radiation levels in the areas, as opposed to using manufacturer-supplied rates at a specified distance from each gauge.


Question: License applicants must provide comprehensive technical safety analyses as part of their submission to the NRC.

All applications for a nuclear license must contain sufficient information to demonstrate that the applicant is technically and financially qualified.

The implications of permanent changes to the facility's design and operation must be rigorously examined to ensure compliance with safety standards.


Question: What activities are involved in groundwater management at mine units?

a. Groundwater Transfer       |   X  | (if yes, complete part 2)

The Mine Unit is in Stability Monitoring


Question: What are the inspection methods for the Lower Grid Assembly ribs?

The LGA lower grid rib sections are elevated to Primary category components in Rev. 2.

For the Revised 40 – 60 item B7 in Table 5-1, MRP-227, Rev. 2: Refer to the analogous column entry statements for the line item B7 on UCB bolts and bolt locking devices, which also applies to the FD bolts and bolt locking devices.


Question: What is the allowable number of trips in nuclear safety analysis?

The allowable number of trips is defined as \( T_{\text{allowable}} = 10^6 / \sum R_{\text{haz}} \), where \( T_{\text{allowable}} \) represents the allowable number of trips (trips/year) and \( 10^6 \) is the total allowable number of hazards per year as specified in RG 1.91.

The adverse wind direction is different for each case because each case uses a specific stability class and wind speed.


Question: What is the purpose of the containment system in nuclear reactors?

K1.02 – Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: Containment isolation/containment integrity. (CFR: 41.2-41.9 / 45.7-45.8)

G2.4.46 – Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43. / 45.12)


Question: nuclear power plant safety inspections

The inspection confirmed that the R.E. Ginna Nuclear Power Plant is operating within the safety requirements set by the NRC, ensuring that all safety protocols were adequately followed during the inspection process.

The detailed technical specifications for the emergency diesel generators indicate that each generator must have a minimum fuel capacity of 5,000 gallons to ensure operational readiness during emergencies.


Question: The application of risk-informed methodologies is essential for evaluating operational safety within the nuclear regulatory framework.

TSTF-505, Revision 2, “Provide Risk-Informed Extended Completion Times – RITSTF Initiative 4b” is a proposal that seeks to integrate a risk-based approach into the regulatory decision-making process surrounding completion times for safety-related actions.

The Columbia Generating Station is located in Richland, Washington, and is a pressurized water reactor that provides a significant amount of electricity to the grid.


Question: nuclear regulatory filings for radiological analysis

The report from LaCrosse Solutions documents sample identification and preparation methods along with analytical procedures used for various radionuclides.


Question: The company Oklo, Inc. is preparing to submit an application for a future license, and their readiness assessment was discussed in a recent meeting.

The purpose of the meeting was to discuss the plan for a readiness assessment for a future license application from Oklo.

The Public Meeting Notice is part of the Agencywide Documents Access and Management System, which provides information about the agency's activities.


Question: The NRC evaluates and monitors the acceptable levels of radiation exposure to ensure public safety.

Source Terms per this calculation 8.8000E+00 Rb-88 1.0000E+00 indicates the specific leakage source term utilized in the analysis.

The reactor has various components that work together to ensure the production of nuclear energy efficiently.


Question: The NRC emphasizes the importance of testing electrical components, especially those containing aluminum, for risk analysis.

Testing the duct in our International High Energy Arc Fault (HEAF) program illustrated the problems with Aluminum HEAFs and is now going through the Generic Issues program.

Please let me know if you require any additional information from me.


Question: The readings recorded during the scans indicate levels of radioactivity in different scan areas.

The investigation required no further samples as all alarms were within the expectations set forth by regulatory standards.


Question: What is the role of numerical convergence in reactor modeling?

PARAGON2 employs spatial and energy acceleration techniques. These techniques utilize rebalancing approaches to update the fluxes at the end of the inner iteration.

The MCNP approach is based on the direct simulation of individual particle transport trajectories.


Question: nuclear licensing safety requirements

The safety requirements for nuclear licensing ensure that the facilities meet stringent standards to protect public health and the environment. These regulations are enforced through comprehensive reviews and inspections.

The amendment process for industrial applications requires a thorough examination of the proposed changes to ensure compliance with current safety regulations.


Question: water level calculations in reservoirs

The backwater profile curve was therefore begun at the dam at Elevation 264.2, and 142,000 cfs was used as the design flood.

The analysis resulted in a highly conservative upper-bound value for the still-water level of 264.2 feet associated with the probable maximum flood.


Question: uranium concentration in soil samples


Question: The NRC evaluates the licensee's analysis of potential flooding scenarios to ensure compliance with safety standards.

The NRC requires licensees to conduct a thorough evaluation of flood hazards at their sites, which includes assessing both internal and external flooding events that could impact safety systems.

The purpose of the Updated Final Safety Analysis Report is to provide a comprehensive overview of the system design and operational characteristics of the nuclear facility.


Question: What are the regulatory requirements for handling tritium in nuclear facilities?

Tritium – addressed through byproduct material regulations and handling procedures

No chain reaction – purely a driven system for direct radiation hazards


Question: The NRC can apply a more general environmental review approach for reactor projects.

Use a generic environmental impact statement (GEIS) approach—which has been effective for reactor license renewals and involves dividing the environmental review into issues that are expected to be generic for new reactor projects and therefore can be examined in a simplified manner and issues that need more in-depth, project-specific evaluations—for new reactor licensing.

Pare down two analytical sections of the EIS—the need for power and alternatives chapters—which are not currently adding much value, especially commensurate with their length.


Question: The maximum average yearly temperature allowed for the MAGNASTOR SYSTEM is set at 76°F, with considerations for high and low temperature extremes.

A temperature of 76°F is the maximum average yearly temperature. The three-day average ambient temperature shall be ≤106°F.

In cases where engineered features (i.e., berms, shield walls) are used to ensure that requirements of 10 CFR 72.104(a) are met, such features are to be considered important to safety and must be evaluated to determine the applicable Quality Assurance Category on a site-specific basis.


Question: auxiliary feedwater system requirements

3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated manual activation switches in the control room and flow paths shall be OPERABLE with:


Question: common cause failure in safety systems

Spurious operations addressed 'within the design basis' include spurious operations that occur as a result of single failures (including cascading effects) or single malfunctions. Consistent with regulatory requirements such as those of GDC 25 or incorporated by reference in 10 CFR 50.55a(h) (IEEE Std 279-1971 or IEEE Std 603-1991), spurious operations as a result of single failures and single malfunctions are expected during the lifetime of the plant and are addressed as part of the design basis.

There is no technical basis for limiting the initiating event to a single safety function.


Question: The onsite release of hazardous materials can include various forms of hazardous substances, including radioactivity.

A release of hazardous material including, but not limited to liquids, combustion products, or radioactivity. Such releases may be concurrent with or induce an explosion or the generation of missiles. In this context, an onsite release of radioactivity is assumed to be associated with low-level radioactive waste.

Per the IPEEE Report, upstream diversion/damming by landslide, ice blockage, or other cause is unlikely. In the unlikely event of upstream diversion or natural damming of the Arkansas River by landslide, ice blockage, or other causes, there would be sufficient storage in Dardanelle Reservoir to permit normal plant shutdown.


Question: The process for handling uranyl nitrate solution at the facility involves its delivery and conversion into a usable form.

Dilute uranyl nitrate (UN) solution enriched to less than 20 weight percent 235U received as feedstock arrives by truck in approved shipping containers licensed by the NRC or DOE. The dilute UN is transferred from the approved shipping containers into a receipt and storage tank located in the TX-1 facility. The received dilute UN solution batch is converted into U3O8 powder to form input batches to the Dissolution process.

Operational activities require a total peak workforce of 1569 workers for TX-1 and TX-2 combined and an annual average of 24 truck deliveries and 89 off-site waste shipments (radiological and non-radiological waste).


Question: What is the confidence level associated with the identified nuclide in radiation detection reports?

K-40 has a confidence level of 0.99, indicating a high certainty in its identification.

The report includes described activities for various nuclides detected within specific ranges.


Question: The NRC regulates the registration of dry casks for storing spent nuclear fuel.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR 72.212(b)(2)), Arizona Public Service Company (APS) is registering the use of dry spent fuel storage casks to store spent fuel under a 10 CFR Part 72 general license at the Palo Verde Nuclear Generating Station (PVNGS) Independent Spent Fuel Storage Installation (ISFSI).

PVNGS Units 1, 2, and 3 have seven owners who are licensees.


Question: What types of bottles are not checked at sample receipt in nuclear material testing?

The preservation of the following bottle types is not checked at sample receipt: Orange (oil and grease), Purple (total cyanide), Pink (dissolved cyanide), Brown (arsenic speciation), Sterile (fecal coliform), EDTA (sulfite), HCl preserved vial (organics), Na2S2O3 preserved vial (organics), and HG-1631 (total/dissolved mercury by method 1631).

ACZ Laboratories, Inc. is located at 2773 Downhill Drive, Steamboat Springs, CO 80487 and can be contacted at (800) 334-5493.


Question: What are the lifecycle emissions of nuclear energy technologies?


Question: The Crownpoint Project Area includes designated controlled regions essential for operational safety.

The controlled area within the Crownpoint Project Area is specifically designated to monitor and manage safety and radiation exposure levels.

The Crownpoint site features various geographical landmarks, including hospitals and other important infrastructure.


Question: reactor operator responsibilities

The reactor operator is given the authority to manipulate controls which directly or indirectly affect core reactivity, including a reactor trip if he deems necessary.

Information concerning these procedures is presented in Sections 11.2.4, 11.3.4, and 11.5.3.


Question: The RCS PTLR must include the established limits for pressure and temperature during various operations and testing phases.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and inservice leakage and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

When a report is required by condition B or F of LCO 3.3.3.1, “Post Accident Monitoring (PAM) Instrumentation,” a report shall be submitted within the following 14 days.


Question: What are the conditions used in the SAFIR simulation for reactor safety analysis?

SAFIR used the following conditions from POLCA-T as input: Fission power, Steam dome pressure, Pressure at TSV/TCV.

The plant control systems and reactor protection system were simulated using SAFIR or boundary conditions.


Question: What is the minimum detectable concentration in nuclear safety analysis?

The sample ID is 485191025 and corresponds to the Client Sample ID L1-010-101-AJGS-001A-SS.


Question: What is the significance of steam line isolation in nuclear reactor safety?

The steam line isolation functions as a safety feature to prevent the release of steam and potential radioactive materials from the reactor, requiring specific channels to be operable for automatic actuation.

The table illustrates the action steps required in the event of a steam line pressure-low situation, requiring one pressure in any two loops.


Question: The influence of climate change on the accessibility and reliability of urban infrastructure is critical for effective adaptation planning.

Dependence of energy systems on land & water supplies influencing development of these systems & options for reducing greenhouse gas emissions.

Public health actions, and action timing, to protect people from some climate change impacts.


Question: nuclide identification

The identified nuclides include K-40 with an energy of 1460.81 keV and a yield of 10.67% with an activity measured at 6.07351E+004 pCi/M².

The energy tolerance for measurements is set at 3.000 keV to ensure accuracy in readings.


Question: reporting requirements for nuclear medicine

In summary, we believe that part 35, title 10 of the Code of Federal Regulations is sufficient and that regulations requiring the reporting of certain Nuclear Medicine injection extravasations as medical events are not required.

would therefore seem that the prime motivation of the petition is to benefit financially and not as Lucerno claims, “to ensure that diagnostic and therapeutic nuclear medicine patients are protected from avoidable irradiation and given access to vital information to understand when and how medical events impact their care”.


Question: requirements for the RadICS platform

Requirements specified as defined in Figure 7-8 of this section of the TR.

The criteria in Table 7.3 are addressed as applicable for the system being developed.


Question: nuclear discharge monitoring requirements

Samples collected to comply with the monitoring requirements specified above shall be collected just prior to discharge and shall represent the discharge during that month. The same, if applicable, shall be reported.

Oil & Grease       |                |             | REPORT


Question: operational flexibilities in nuclear safety analysis

Embedded mechanisms to apply analytical safety margins to gain operational flexibilities

Consideration of "unmitigated consequences" from event scenarios to support Part 53


Question: The NRC outlines specifications for the total count of fuel assemblies and their maximum weight within a canister.

- **Maximum mass U per YANKEE-MPC RECONFIGURED FUEL ASSEMBLY:** 66.33 kg

- **D.** INTACT FUEL ASSEMBLIES shall not contain empty fuel rod positions. A solid Zircaloy or stainless steel rod that would displace an equivalent amount of water as an intact fuel rod or a Zircaloy rod containing a stainless steel or Zircaloy slug shall replace any missing fuel rod.


Question: What is required before making the weld joint on the pressurizer spray line nozzle?

The operating experience has shown that cracking has occurred on the spray line nozzle at the top of the pressurizer. The inspector should verify in the field the fit-up between the spray nozzle and the new pressurizer is within design tolerance before the licensee makes the weld joint.

The inspection should focus on the impact of load handling activities on reactor core or spent fuel and its cooling and plant support systems for the reactor unit and common systems for the other operating unit(s) at the site. Several heavy lifting evolutions are.


Question: What does PMH mean in the context of hurricane analysis?

Thus, the PMH that produces the maximum negative surge has the following characteristics: Central pressure, p0 = 26.65 in Hg. Pressure drop, Δp = 3.5 in Hg. Radius of maximum winds, R = 28 NM. Forward speed, T = 26 kt. Track direction (storm coming from) = 185 degrees. Distance at closest approach = 20 NM (23 mi) east of the mouth of Delaware Bay. Coefficient related to density of air, K = 68.

Negative surges associated with hurricanes Floyd, Gloria, and Charley observed at Lewes are used to determine the relationship between negative surge, maximum sustained winds of each hurricane, and the distance of each storm from Delaware Bay at its point of closest approach.


Question: What are the conditions under which licensees may apply their proposed uncertainty approach?

This SE does not approve the proposed uncertainty approach for parameter licensees may (1) conservatively apply to this parameter, or (2) justify alternative approaches (e.g., via citation of additional representative data) for determining in their implementation submittals.

The processes involved in reactor operation include regular maintenance checks, fuel replacement, and adherence to operational protocols that ensure safety and efficiency.


Question: What are the implications of regulatory changes on nuclear waste management?

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: definition of aging management program


Question: The systems crucial for maintaining safety in nuclear reactors include the Containment Atmospheric Control System and the Primary Containment Isolation System.

The following systems are addressed in this section: Containment Atmospheric Control System (2.3.2.1), Core Spray System (2.3.2.2), High Pressure Coolant Injection System (2.3.2.3), Isolation Condenser System (2.3.2.4), Low Pressure Coolant Injection System (2.3.2.5), Primary Containment Isolation System (2.3.2.6), Standby Gas Treatment System (2.3.2.7)

The results of the analysis indicate that the facility can withstand extreme environmental conditions without significant risk.


Question: decommissioning impact on public health

The decommissioning of a nuclear facility involves the removal of nuclear materials and the restoration of the site to a condition that is safe for the public and the environment, thereby impacting public health and safety during and after the process.

The Peach Bottom Atomic Power Station, Unit 1, has a history of operational efficiency that includes implementing advanced safety protocols and undergoing routine inspections to ensure compliance with regulatory standards.


Question: What is the process for protesting a civil penalty imposed by the NRC?

The licensee may protest the imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice.

If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information.


Question: SBGT Monthly Operability Test

Perform the SBGT Monthly Operability Test with Failure of the Heater to Turn Off After Shutdown

Developed By: Instructor: Print / Sign  |  Date


Question: Hot Leg Recirculation procedure

The Unit Supervisor has directed you to proceed with 1BEP ES-1.4, Transfer to Hot Leg Recirculation and to notify him when complete.

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


Question: What is LCO 3.7.6 Action in nuclear licensing?


Question: ECCS valve control switches

The candidate must interpret the position of ECCS valves during performance of 1-EOP-FRCC-1 (RESPONSE TO INADEQUATE CORE COOLING) and determine what valves must be operated (opened or closed).

With SECs operating in Mode 1 and IAW -EOP-FRCC-1 (RESPONSE TO INADEQUATE CORE COOLING) Step 3, the crew is required to open 1SJ5 and is NOT required to close 11SW122.


Question: The NRC mandates licensees to ensure that safety systems operate effectively through regular performance testing.

The COMPONENT COOLING COMMON PUMP is characterized by its designation as a Safety Class 3 pump with a driver speed rated at GE600 and is tested using the dP method.

Regulatory agencies continually monitor nuclear facilities to ensure compliance with operational standards and safety protocols.


Question: What is required to assess plant conditions in emergency situations?

The question requires an assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed.

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?


Question: What is the significance of the zirconium-water reaction in reactors?

... with regard to the question he just answered, about tests performed by Westinghouse, I am personally aware of some exploratory type tests which were done a long time ago when we first went into the use of Zircaloy cladding in reactors, exploring just exactly the thing you were discussing. That is the temperature range of 1800 to 2000 degrees [Fahrenheit], and the effect of this type of condition on Zircaloy rods to confirm for ourselves that there was no significant metal-water reaction in that range of temperatures in order to confirm for ourselves.

I observed those tests but I was not actually performing the test myself.


Question: nuclear regulatory licensing processes

The NRC licensing process for new reactors is a comprehensive evaluation that ensures all safety and environmental standards are met before construction can begin.

The reactor's coolant system was designed to efficiently transfer heat from the core, maintaining optimal operating temperatures and preventing overheating.


Question: The NRC requires accurate measurement of weighted mean activity for detected nuclides to ensure compliance with regulatory standards.

Wt mean Activity is a measure of the activity concentration of a specific nuclide in a sample, usually expressed in pCi/grams.

The document includes nuclide rejection criteria based on interference analysis, which affects the confidence of identified isotopes.


Question: The geometry of testing critical heat flux was analyzed at specified distances from the effective operating limit.


Question: reactor oversight and safety inspections

The inspection process ensures that the operation of the Sequoyah facility complies with NRC safety regulations and licensing requirements, which are critical in maintaining safety and operational integrity.

The lottery system used for selecting inspectors is designed to minimize bias and ensure a diverse group of experts assess the compliance of the licensee's procedures.


Question: The NRC mandates a comprehensive analysis of the aging effects on structural components, particularly concrete structures that are critical for containment.

Concrete (inaccessible areas): dome, wall, basemat  | Fire Barrier, HELB Shielding, Missile Barrier, Shelter, Protection, Shielding, Structural Pressure Boundary, Structural Support | Concrete  | Atmosphere/Weather (Structural) (Ext) Plant Indoor Air (Structural) (Ext) | Cracking                                              | ASME Section XI, Subsection IWL (B.2.3.29)

Increase in Porosity and Permeability Loss of strength | ASME Section XI, Subsection IWL (B.2.3.29)        | II.A1.CP-102     | 3.5-1, 014  | A, 1


Question: spent fuel assembly storage regulatory requirements

The combination of initial enrichment and burnup of each spent fuel assembly stored in Region 2 or 3 shall be within the Acceptable Domain of Figure 3.7.17-1 or in accordance with Specification 4.3.1.1.

Initiate action to move the noncomplying fuel assembly to Region 1.


Question: What is the requirement for analysis of on-shift personnel in nuclear power plants?

By December 24, 2012, for nuclear power reactor licensees, a detailed analysis demonstrating that on-shift personnel assigned emergency plan implementation functions are not assigned responsibilities that would prevent the timely performance of their assigned functions as specified in the emergency plan.

Identification, by position and function to be performed, of other employees of the licensee with special qualifications for coping with emergency conditions that may arise.


Question: The report lists nuclides identified during analysis and includes details like energy levels and confidence measurements.

| Nuclide Name | Id     | Confidence | Energy (keV) | Yield (%) | Activity (pCi/grams) | Activity Uncertainty |

In the modern nuclear industry, ensuring that safety protocols are followed during reactor operation is crucial for preventing accidents.


Question: Ensure that inspections for the RV head replacement are conducted according to safety protocols and standards.

Conduct RV head removal and replacement inspections in accordance with the inspection plan by performing selective inspections, consistent with the safety significance and inspection resources, of the following areas:

Verify that machining is carried out under a controlled system of operation (travelers, check lists) consistent with the manufacturers overall Quality Assurance (QA) program.


Question: inorganic analytical results

The Inorganic Analytical Results section includes various parameters evaluated through wet chemistry analysis, such as alkalinity, cation-anion balance, and specific ion concentrations like fluoride and cyanide.

The sample was received on 07/13/24, indicating the timeline of the testing process and ensuring compliance with regulatory standards.


Question: gamma spectrum analysis

**Peak Analysis From Channel:** 100
**Peak Analysis To Channel:** 4096


Question: What is the Special Test ABB-NV Database used for in nuclear tests?

The Special Test ABB-NV Database for Tests 70 and 72 collects and analyzes data to ensure the reliability and safety of nuclear systems during testing phases.

The documents also include a variety of appendices that provide detailed descriptions of the operational parameters for different reactor designs.


Question: The fluence methodology employed for calculating neutron fluence in reactor vessels uses the NRC-approved Transware Radiation Analysis Modeling Application (RAMA).

**Transware Radiation Analysis Modeling Application (RAMA)** methodology has been used to develop 80-year fluence projections for the reactor vessel and vessel internal components that are used in evaluating TLAA in Sections A.4.2.1.1, A.4.2.1.2, and A.4.2.2 through A.4.2.16.

By Amendment Nos. 310, 333 and 293 (Units 1, 2 and 3, respectively), dated December 26, 2019, the NRC approved a Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating strategy for BFN Units 1, 2, and 3.


Question: in-service inspection procedures

In-service Inspection - Review of Procedures 01/14/86 (86-03)

RESERVED for Part 52 Initial Fuel Load Witnessing.


Question: What are the requirements for renewing a nuclear license?

The application for the renewal of a nuclear license must be submitted by the licensee at least 30 days prior to the expiration of the current license.

The technical specifications for the reactor's safety systems are outlined in a separate document.


Question: What are the actions required when control rods are not in compliance with BPWS?

B. Nine or more OPERABLE control rods not in compliance with BPWS.¹ | B.1 --------------NOTE---------------- Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1.  Suspend withdrawal of control rods.

The reactor mode switch must be operated manually to ensure proper safety controls are maintained during all phases of operation.


Question: aging management program

AMP is consistent with NUREG-1801 AMP description.


Question: The NRC regulates the possession and use of specific materials linked to nuclear activities, ensuring safety and compliance in various applications.

Special Nuclear Material Plutonium – Unsealed, Less than Critical Mass | Less than 200 grams, total, for biological and chemical testing and instrument calibration.


Question: electrical cable connections environmental qualification requirements

Implement the new DCPP Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements AMP.

Continue the existing DCPP Transmission Conductor and Connections, Switchyard Bus and Connections, and High-Voltage Insulators AMP including enhancements to: (a) Identify transmission and substation components required to support station blackout recovery which are in the scope of license renewal aging management.


Question: What are the properties of structural graphite material used in nuclear applications?

| Property                        | Values*                    |
|---------------------------------|---------------------------|
| Grain size, µm                 | 40µm max, 15µm avg       |
| Bulk density, g/cm³            | 1.75                      |
| Electrical resistivity, µΩm     | 14                        |
| Thermal conductivity, W/mK      | 104.4                     |
| Coefficient of Thermal Expansion, x10⁶/°C | 3.8            |
| Tensile strength, MPa           | 34.3                      |
| Flexural strength, MPa (3 points) | 58.8                     |
| Compressive strength, MPa       | 98.0                      |
| Elastic modulus, GPa            | 10.8                      |
| Ash content, ppm                | [[    ]]                 |
| Boron Content, ppm              | [[    ]]                 |

The regulatory process for the issuance of a nuclear license requires comprehensive technical documents to be submitted for review.


Question: Toxic Substances Control Act

The Toxic Substances Control Act (TSCA) regulates the manufacture, processing, distribution, and use of certain chemicals not regulated by RCRA or other statutes, including asbestos-containing material and polychlorinated biphenyls.

The Magnuson–Stevens Fishery Conservation and Management Act, as amended, governs marine fisheries management in Federal waters.


Question: The NRC demands extensive documentation on the control of radioactive materials involved in regulatory processes.

Plutonium-239/240 is frequently analyzed in quality control processes to ensure compliance with safety standards and regulatory requirements.

The batch of Cesium-137 was tested for radioactivity and showed readings consistent with the expected levels of decay.


Question: The licensing framework requires comprehensive evaluations of system safety and performance.

In charge of the US460 standard design approval application.

I held former positions as a senior reactor officer, I&C manager, worked in safety analysis, reactor engineering, and now work for -- in licensing at NuScale.