Question: What is the purpose of a radioactive effluents monitoring program?

The RADIOACTIVE EFFLUENTS MONITORING PROGRAM is designed to ensure compliance with regulatory requirements by monitoring and controlling radioactive effluents released from the facility.

The ADMINISTRATIVE CONTROLS ORGANIZATION section outlines the structure and responsibilities of the regulatory body overseeing safety protocols.


Question: What are typical soil concentrations for radionuclides in residential areas?

Radionuclide H-3 has a residential soil concentration of 228 pCi/g.

Total uranium is measured at 47 mg/kg for residential soil concentration.


Question: metal fatigue in nuclear components

Fatigue is a TLAA; further evaluation is documented in Subsection 3.2.2.2.1.

The One-Time Inspection (B.2.1.20) program will be used to manage loss of material of the stainless steel accumulator, bolting, flexible connection, flow device, piping, piping components, pump casings, rupture disks, and valve bodies exposed to air - indoor uncontrolled and condensation in the Containment Atmospheric Control System, Core Spray System, High Pressure Coolant Injection System, Isolation Condenser System, Low Pressure Coolant Injection System, Primary Containment Isolation System, and Standby Gas Treatment System.


Question: The NRC requires that certain exemption requests must be included for processing license amendments to ensure compliance with regulations.

The staff has cited examples of cases where the exemptions and the license amendments are sort of expressly required.

I just have to say I disagree with it. These amendments are necessary because of other amendments that were issued at the time of the 50.82 certification that eliminated some of these requirements that basically adopted these.


Question: normal operation in nuclear reactors

The normal operation analyses account for [ ]

The results for each criterion (rod internal pressure, cladding corrosion, and cladding hydrogen concentration) are summarized in Table 6-9.


Question: What are the requirements for a nuclear reactor's emergency core cooling system?

The emergency core cooling system (ECCS) is designed to automatically provide cooling water to the reactor core following aLoss of Coolant Accident (LOCA) to prevent overheating and potential meltdown.

The regulatory filings necessitate detailed environmental assessments to ensure that the impact of the nuclear facility on the surrounding area is minimal.


Question: What is the purpose of a reactor recirculation pump variable frequency drive?

The unit was down powered to 72 percent for a loss of a reactor recirculation pump variable frequency drive.

Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, "Light-Water Reactor Inspection Program - Operations Phase."


Question: weighted mean activity

The weighted mean activity of a nuclide is calculated using the average activity concentration of the nuclide in units of picoCuries per gram, alongside the associated uncertainty values.

Errors quoted at 2.000sigma reflect the statistical uncertainty of the measurements in the report.


Question: What are licensing basis events in nuclear regulatory context?

AOOs would be a category of licensing basis event. I'll actually get into that, I think maybe in a couple of slides. I provide the definition of what licensing basis events are.

So in a way it's kind of too bad that it doesn't resolve that, but I think the explanation of what it's doing is pretty clear.


Question: nuclear reactor safety licensing process

The nuclear reactor safety licensing process involves a comprehensive review by the NRC to ensure that facilities meet the required safety standards before being granted a license to operate.

The unemployment rate in Champaign County decreased from 3.30% in 1990 to 2.40% in 2000.


Question: The auxiliary spillway is designed to manage extreme flood events exceeding the levels of the 100 yr flood, ensuring the dam's safety against overtopping.

The auxiliary (emergency) spillway is located east of the dam. The location is chosen on the basis of obtaining a better approach condition. The auxiliary spillway is designed to pass floods more severe than the 100 yr flood and up to and including the PMF. The spillway provides protection to the dam against overtopping. The spillway is an open-cut type with a crest length of 1,200 ft and a crest elevation of 700 ft above msl.

The peak discharge through the auxiliary spillway during the CPS PMF conditions is 102,800 cfs with a corresponding water level in the lake of elevation 708.8 ft above msl.


Question: What is the reporting limit in nuclear regulatory analysis?

The reporting limit (RL) is defined as the lowest concentration of a substance that can be reliably measured and reported by an analytical method.

The counting uncertainty is calculated at the 95% confidence level (1.96-sigma).


Question: Jocassee Dam failure modes

Further, in evaluating Duke’s information regarding these three failure modes using present-day methodologies and regulatory guidance, which included independent confirmatory analyses of the Jocassee Dam breach uncertainties range, the Staff determined that both seismically and overtopping-induced failures of the Jocassee Dam were not “reasonable model[s] of failure” while the sunny-day Jocassee Dam failure mode was deemed “an unlikely, although reasonable, failure mode.”

See Staff 2010/2015 Evaluation Comparison at 3; see also April 2016 Staff FHRR Final Assessment at 3-6.


Question: The NRC evaluates the safety of new reactor designs through stringent technical specifications and thorough reviews.

I am a project manager with the Office of Nuclear Reactor Regulation, Division of New and Renewed Licensees, New Reactor Licensing Branch.


Question: The handling of specific sample types does not require verification at the time they are received for 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 (sulfate) - HCl preserved vial (organics) - Na2S2O3 preserved vial (organics) - HG-1631 (total/dissolved mercury by method 1631).

The ACZ Project ID serves as a unique identifier for tracking purposes.


Question: The extent of the PMP is identified in relation to nearby river basins.

**PMP Storm Extent** is important to define the impact area for flood planning and management activities, specifically against the backdrop of regional river systems.

The Missouri River Basin from Gavins Point to Sioux City is one of the critical areas to monitor during flood assessments.


Question: What is meant by digital modernization in nuclear plants?

Digital modernization enhances the operational performance of Turkey Point Unit 3 and Unit 4 by integrating advanced digital technology into the existing systems.

The cooling systems at Turkey Point have undergone several upgrades to improve efficiency and reduce environmental impact.


Question: What is the purpose of the integrated plant assessment in nuclear license renewal?

The IPA shall identify and list those structures and components subject to an aging management review and demonstrate that the effects of aging (i.e., cracking, loss of material, loss of fracture toughness, dimensional changes, loss of preload) will be adequately managed so that their intended functions will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.29, "Standards of issuance of a renewed license."

The NRC staff reviewed and accepted the technical report on June 17, 2022 (ADAMS Accession No. ML22145A401).


Question: Maximum Allowable Rod Internal Pressure

The maximum allowable rod internal pressure is determined to avoid hydride reorientation, ensuring the structural integrity of the fuel rods under operational conditions.

PWR reactor application examples are critical for understanding different fuel rod designs and their specific operational characteristics.


Question: thermophilic organisms in nuclear safety

Consultation about Thermophilic Organisms. February 19, 2014. ADAMS Accession No. ML15273A445.

Well Water Testing. Retrieved from [IEPA](https://epa.illinois.gov/topics/drinking-water/private-well-users/well-water-testing.html) (accessed February 28, 2023).


Question: What is the regulatory limit for Zircaloy cladding temperatures?

Extrapolation of data from cited experiments indicates that a new regulation that requires minimum allowable core reflood rates in the event of a LOCA is necessary to prevent Zircaloy cladding from exceeding the regulatory limit of 2,200 degrees F under certain conditions.

The NRC disagrees with the comments, because the NRC staff determined that excessive heatup rates were not experienced during the FLHT-1 experiment until temperatures exceeded 2,420 degrees F.


Question: What is the minimum departure from nucleate boiling ratio in reactor safety?

The criterion for reactor protection for the pump coasting loss-of-flow event is that the minimum departure from nucleate boiling (DNB) ratio shall not be less than the applicable critical heat flux correlation limit (e.g., 1.3 for BAW-2 correlation, 1.18 for the BWC correlation, or 1.132 for the BHTP correlation).

No new failure modes are introduced by the proposed technical specification (TS) change.


Question: What is the required action if the Division 1 or 2 125 VDC electrical power subsystem is inoperable?

H. Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Conditions E, F, or G. | H.1 Restore Division 1 or 2 125 VDC electrical power subsystem to OPERABLE status.
 **OR**
 H.2 -------NOTE-------
 Only applicable if the opposite unit is not in MODE 1, 2, or 3.
 Place associated OPERABLE alternate 125 VDC electrical power subsystem in service.

The technical specifications require that systems be maintained to ensure safety margins are not compromised during operation.


Question: The report details the activity levels of various isotopes in terms of pCi per gram, highlighting the regulatory limits for each.


Question: The population data is significant for assessing potential impacts on safety measures in the vicinity of the facility.

The calculation of population within various distances from the reactor is essential for evaluating potential emergency impacts.

The distance measurements are organized by miles and groupings to reflect demographic changes in various areas.


Question: What are the numerical parameters related to reactor safety analysis?

The sums for various numeric parameters included in the analysis are: Sum t 3,004, Sum F 2,688, Sum t**f 4.521, and Sum F**F 4,037.7, indicating the analysis results of integrated factors for safety evaluations.

The NRC evaluates the performance of licensees based on their adherence to safety regulations and operational standards, which ensures consistent regulatory compliance.


Question: The analysis in Figure 3 illustrates how boron dilution affects core reactivity.

As shown in Figure 3, the conservatisms in the boron dilution analysis are such that the

To quantify the conservatism associated with assumed


Question: The instrumentation for the RCIC system must adhere to specific technical specifications as detailed in regulatory guidelines.

Reactor Core Isolation Cooling (RCIC) System Instrumentation - As required by Required Action A.1 and referenced in Tech Spec Table 3.3.5.3-1

Division specific ADS SRV pilot SOVs and supports will be added to the ADS model as conservative surrogates for the LCO condition.


Question: The data shows a decrease in dial readings over time indicating pressure adjustments.

The test results indicate consistent pressure levels during the measurement interval.


Question: The analysis of samples taken from the Pickard Hall site included various measurements of radioactive isotopes, specifically Ra 226 and Ra 228.

Homogenize and select alpha spec aliquot from the samples analyzed for Ra 226/228 Gamma Spec Norm Library U/T Alpha Spec.

The project manager's contact number for inquiries is 865-934-0376.


Question: The analysis of break mass flow rate during various operational conditions is crucial for safety assessments.

BREAK MASS FLOW RATE DURING BLOWDOWN (CD = 0.6, LOW TAVG, MIN SI, COSINE POWER SHAPE, 156-IFBA)

The chart illustrates the variations in temperature across different reactor zones.


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

The Safety Analysis Report (SAR) provides a comprehensive evaluation of the safety aspects of a proposed nuclear facility, including a detailed analysis of potential hazards, system design, and operational procedures.

The cooling system is designed to maintain optimal operating temperatures during full power conditions, but it requires regular maintenance to ensure efficient performance.


Question: What are the inspection requirements for clevis insert bolts in PWRs?

The PWR Owners Group issued OG-21-160 in 2021 to address NEI guidance pertaining to PWR lower radial support clevis insert X-750 bolt inspection requirements.

The fuel alignment pins at this site were constructed from Type 304 stainless steel with malcoated surface treatments, which were considered for potential surface degradation as early as the 1970s.


Question: What is the minimum cooling time required for nuclear fuel assemblies with an average enrichment of 2.5 to 2.7 wt % 235U?

For assemblies with an average enrichment of 2.5 ≤ E < 2.7 wt % 235U, the minimum cooling time required before reprocessing is 4.2 years for CE and 4.8 years for B&W designs.

The NRC requires that all nuclear facility operators maintain accurate records of their operational activities to ensure compliance with safety regulations.


Question: What is included in ACZ's Quality Assurance Plan?

All analyses were performed according to ACZ's Quality Assurance Plan.

If you have any questions or other needs, please contact your Project Manager.


Question: What is the significance of the reactor coolant pressure boundary in nuclear safety?

Reactor coolant pressure boundary | Pressure Boundary  | Stainless Steel | Reactor coolant   | Loss of material due to pitting, crevice corrosion

Piping, piping components, ducting, | Pressure Boundary  | Polymeric     | Treated Water     | Hardening or loss of strength due to polymeric degradation; loss of material due to peeling, delamination, wear; cracking or blistering due to exposure to ultraviolet light, ozone, radiation, or chemical attack; flow blockage due to fouling.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves a comprehensive review of the applicant's proposal, including safety analyses, environmental assessments, and regulatory compliance determinations as outlined in 10 CFR Part 50.

The fuel assembly is designed to enhance neutron moderation and improve the overall thermal efficiency of the reactor core.


Question: Nuclear facility operators must comply with specific licensing requirements and safety regulations.

Licensing is a process by which the NRC grants permission to a facility to operate based on safety evaluations and compliance with regulations.

The storage of spent fuel is a critical aspect of nuclear facility management that requires careful planning and execution.


Question: Licensees must conduct thorough assessments for potential hazards during nuclear operations.

The NRC mandates that all licensees perform rigorous safety evaluations to identify potential risks associated with their nuclear facilities.

The design of the cooling system must accommodate variations in temperature while ensuring efficient heat exchange.


Question: Routine inspections and testing are essential to ensure that the nuclear power plant's systems remain compliant with regulatory standards and operate safely.

Routine maintenance performed on station systems and components is necessary for safe and reliable operation of a nuclear power station.

The water systems at DNPS are discussed in the sections below. The water flow and volumes are presented in Figure 2.2-1.


Question: What is the limit for fiber insulation fines in nuclear sump strainer debris limits?

Fiber Insulation Fines    | 42.70                 | ft³

The available margin for a given debris type is the difference between its debris limit and largest transportable quantity for the range of break sizes considered.


Question: radiological health impacts

Potential Radiological Health Impacts    | Proposed Action   | MILDOS-AREA modeling shows no impacts to the public resulting from the CUP.

Potential Socioeconomic Impacts         | Proposed Action   | Local communities could be positively impacted by the addition of jobs created by the CUP. McKinley County, the State of New Mexico, and the Navajo Nation will be beneficially impacted as a result of taxes on the CUP.


Question: What is the significance of nuclide minimum detectable activity in nuclear safety?

The nuclide minimum detectable activity (MDA) is a measure that indicates the lowest level of activity that can be reliably detected in a sample for a specific nuclide.

The table lists various nuclides along with their energy, yield, and activity, but does not specify any minimum detectable activity.


Question: What must be provided when reporting a device no longer in possession of a licensee?

Provide Information about devices listed in Section 2 or 6, but no longer in your possession.

- License Number of Recipient (if transferred to a specific licensee): LA-5934-01


Question: What is mass flow rate in fluid dynamics?


Question: What is the significance of time to core damage in nuclear safety analysis?

The time to core damage is a critical factor in determining the effectiveness of security measures in a nuclear facility, as it impacts the analyses required for risk mitigation.

Al Csontos, NEI, just wanted to say thank you for the opportunity to present today and provide our feedback.


Question: The cooling water system has not undergone any modifications that would adversely affect aquatic life during the proposed operational term.

DNPS’s continued operation in compliance with its NPDES permits is not expected to adversely affect the aquatic community during the proposed SLR operating term.

During the period of indirect open-cycle operation at DNPS, no effects on the algal community outside the immediate discharge area were detected that could be attributed to DNPS operations.


Question: What are Circulating Water Conduits in the context of nuclear power plant safety?

Revise implementing procedures to add Circulating Water Conduits to the scope of the program.

Revise implementing procedures for monitoring and inspection of water-control concrete structures to also include those parameters as described in ACI 201.1R.


Question: The NRC’s current stance on reporting extravasations disallows the classification of these events as medical events, regardless of the dose equivalent involved.

On May 18, 2020, Lucerno Dynamics, LLC (“Lucerno”) filed a petition for rulemaking with the Nuclear Regulatory Commission (NRC) to amend 10 C.F.R. § 35.2 and 10 C.F.R. § 35.3045 to require the reporting of extravasations that exceed the 0.5 Sv dose equivalent to tissue as medical events.

The University of California, Los Angeles (UCLA) Radiation Safety Committee has reviewed Lucerno’s petition and the relevant literature, and finds that the current NRC policy regarding extravasations established in May 1980 (and reviewed several times during the ensuing years, most recently in 2020) remains sufficient as it stands and does not require additional rulemaking as petitioned by Lucerno.


Question: What are the aging management programs for the Shutdown Cooling System?

The following aging management programs manage the aging effects for the Shutdown Cooling System components: - Bolting Integrity (B.2.1.10) - Closed Treated Water Systems (B.2.1.12) - External Surfaces Monitoring of Mechanical Components (B.2.1.23) - One-Time Inspection (B.2.1.20) - Selective Leaching (B.2.1.21) - TLAA - Water Chemistry (B.2.1.2)

The Shutdown Cooling System components are exposed to the following environments: - Air - Indoor, Uncontrolled - Closed Cycle Cooling Water - Treated Water - Treated Water > 140°F


Question: What is the function of the residual heat removal system in reactor operations?

The residual heat removal system is designed to reduce the temperature of the reactor coolant from 350°F to 140°F over a period of 16 hours.

Plant shutdown encompasses the operations that bring the reactor from hot no-load temperature and pressure to cold shutdown conditions.


Question: What is a Quality Assurance Program in nuclear licensing?

A Quality Assurance Program ensures that all activities related to the design, construction, and operation of nuclear facilities meet specified standards of quality and safety.

The reactor vessel is surrounded by a containment structure that provides a robust barrier to prevent radioactive releases.


Question: What is the purpose of a Control Rod in a nuclear reactor?

Control rods are used to control the fission rate of uranium and plutonium in nuclear reactors. They absorb neutrons and thereby reduce the number of free neutrons available to sustain the nuclear chain reaction.

The temperatures in the reactor core must be maintained within specific limits to ensure safe and efficient operation.


Question: Before confirming contract acceptance, GEH must validate that customer requirements have been clearly defined and that any discrepancies are addressed according to the regulatory requirements.

Proposals and contracts are reviewed to determine customer requirements in accordance with implementing procedures for commercial risk and approval, work planning and scheduling, design inputs and customer technical requirements, and spare and renewal parts. This includes customer’s stated requirements or intended expectations, regulatory requirements, product specification, and any other GEH identified requirements.

Implementing procedures related to new product introduction describe the detailed process for the development and introduction of new products.


Question: nuclear regulatory compliance

Nuclear regulatory compliance is the adherence to the laws and regulations set forth by the Nuclear Regulatory Commission (NRC) to ensure the safe operation of nuclear facilities.

The production of renewable energy can lead to significant reductions in carbon emissions and contribute to climate protection efforts.


Question: J-groove weld qualification requirements

Weld procedures qualified per Section IX and- NB 4330.

Mechanical properties per SB spec Table 2.


Question: What are the environmental resource issues evaluated for the DNPS?

Surface water use conflicts (plants with once-through cooling systems) | 1 | Applicable | 4.5.8

Offsite land use in transmission line ROWs | 1 | Not Applicable | 4.1.3


Question: Charpy impact testing limitations

Petitioners further aver that, while Charpy impact tests are useful in the analysis and prediction of the behaviors of different materials under impact stresses or dynamic loading, such tests cannot directly predict the reaction of a material to real life loading.

The performance of a specimen in a Charpy impact test is, however, influenced by many factors beyond material composition and temperature such as yield strength and ductility, placement, size, and shape of notches, strain rate, and fracture mechanism all affect the performance of a sample.


Question: How is the containment evacuation alarm actuated in nuclear facilities?

The containment evacuation alarm is actuated from the 1PM05J via a switch. The alarm will continue to sound until the switch is returned to its normal location.

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.


Question: What are the requirements for maintaining a nuclear reactor's safety systems?

The maintenance of safety systems is critical for ensuring the continued safe operation of a nuclear reactor, as outlined in NRC Regulation 10 CFR 50.65, which requires licensees to monitor and maintain the performance of safety-related systems.

The financial audit provided insights into the budget allocations for various departments within the organization, detailing the expenditures and projected financial outcomes for the upcoming year.


Question: limiting condition for operation

10 CFR 50.36, "Technical Specifications," paragraph (c)(2)(i) states that limiting conditions for operation are the lowest functional capability or performance level of equipment required for safe operation of the facility, and when a limiting condition for operation is not met, the licensee shall shut down the reactor or follow any remedial actions permitted by the TS until the condition can be met.

General Design Criteria 2, "Design bases for protection against natural phenomena," and General Design Criteria 44, "Cooling water," of Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," provides design considerations for the UHS.


Question: What are the requirements for non-welded fittings according to ASME Section XI?

Case N-878 requires that non-welded fittings be supplied in accordance with the provisions of Section XI, IWA-4142.1(b)(4) or (5), except that the activities will be performed by a Repair/Replacement Organization qualified by the Owner in accordance with 10 CFR 50, Appendix B.

All other ASME Section XI requirements for which relief was not specifically requested and authorized by the NRC Staff will remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.


Question: group work hour controls

Group average work hours for personnel performing the functions identified in B shall be controlled in accordance with the following limits: (a) Normal Plant Conditions: The average number of hours actually worked by personnel performing the functions identified in B, shall not exceed 48 hours per week averaged over consecutive periods not to exceed six (6) weeks. Workers who did not work at least 75 percent of the normally scheduled hours during the averaging period shall not be included when calculating the average.

(c) The licensee shall monitor and control individual work hours to ensure that excessive work hours are not compromising worker alertness and performance.


Question: What does pCi/g dry indicate in nuclear licensing reports?

The activity concentration is reported in pCi/g Dry, which expresses the amount of radioactive material in picocuries per gram of dry sample.

The uncertainty calculations were performed under standard laboratory conditions to ensure the accuracy of the measurements.


Question: river intake structure sediment management

large scale dredging operations to prevent channel meandering due to scouring or substantial sediment build up have not been necessary at the river intake structure.

Figure 4-6: 1953 USGS Dothan, AL Quadrangle Map


Question: The classification of survey unit L1-010-107 as an open land survey unit was guided by MARSSIM standards.

Survey unit L1-010-107, an open land survey unit, has a MARSSIM classification of 1.

The boundary of the survey unit and the location of the soil samples were defined using a Global Positioning System (GPS).


Question: What is the significance of DO concentrations in cooling water discharges?

Overall, the NRC concluded that impacts of plant operation on low DO concentrations and gas supersaturation attributable to cooling water discharges would be SMALL for all plants.

To summarize, discharges to the Illinois River are subject to DNPS’s NPDES permit, and there have been no NOVs related to the NPDES permit with respect to contaminants in surface water from January 2018 through the end of 2023.


Question: AC electrical power system redundancy

However, two factors tend to decrease the severity of this degradation level: a. The configuration of the redundant AC electrical power system that remains available is not susceptible to a single bus or switching failure; and b. The time required to detect and restore an unavailable offsite power source is generally much less than that required to detect and restore an unavailable onsite AC source.

According to Regulatory Guide 1.93 (Ref. 9), with the available offsite AC sources two less than required by the LCO, operation may continue for 24 hours.


Question: The NRC assesses the impact of potential operational issues on the submittal review process, which can affect timelines.

In this case the actual need date is considered to be the date that the ILRT must be performed based on the current 15-year frequency, which is February 2023.

There were no final determinations or regulatory decisions made by the NRC staff during the meeting.


Question: What is the aging management program (AMP) according to NUREG-2191?

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


Question: What is the significance of VTTP codes in nuclear data documentation?

The VTTP codes, such as VTTP-600 and VTTP-616, represent specific technical parameters used in the safety assessment and reporting for nuclear facilities, helping to standardize measurements and ensure regulatory compliance in technical safety evaluations.

The document discusses various units of measurement like cubic meters and kilograms used in different engineering calculations, which highlight the physical properties of the materials used in reactor construction.


Question: nuclide activity measurements

The activity of a nuclide can be measured in picoCuries per gram (pCi/grams), which indicates the rate at which a sample of a radioactive nuclide decays.

Regulatory filings must include detailed descriptions of plant safety systems, including their design, operation, and maintenance protocols.


Question: The proposed emergency plan alters the staffing requirements for the EOF Emergency Director at the Site Area Emergency level by reducing the minimum positions available for Command and Control functions.

The NUREG-0654 R2 technical basis for the ERO staffing guidance states that: For a SAE ECL, or greater, these functions (or a subset of them), typically those associated with protective action recommendations (PARs), should be assigned to an Emergency Director located in the EOF….

Refer to Attachment 1, ERO Staffing Plan Comparison Table, for a side-by-side summary staffing and augmentation comparison between NUREG-0654 R2 Table B-1, the current emergency plan, and the proposed emergency plan.


Question: Strontium-90 detection limits

The following Analytical Methods were performed:  | Method | Description                             | Analyst Comments       |  |--------|-----------------------------------------|------------------------|  | 1      | EPA 905.0 Modified/DOE RP501 Rev. 1 Modified |                        |  | 2      | EPA 906.0 Modified                         |                        |


Question: water-bearing properties of geological formations

Yields in tributary streams yield up to 100 gpm. Mississippi River alluvium may yield up to 500 gpm; source for CGS radial wells in Warren County report yields of 75 to 440 gpm.


Question: The identification of nuclides is critical for ensuring regulatory compliance in nuclear safety assessments.

Nuclide identified during the nuclide identification.

The energy levels of various nuclides can vary significantly, affecting their detection and measurement in safety evaluations.


Question: The NRC conducted simulations to analyze the variability of costs and benefits, using sophisticated tools to incorporate uncertainty into their regulatory assessments.

Monte Carlo simulations involve introducing uncertainty into the analysis by replacing the point estimates of the variables used to estimate base case costs and benefits with probability distributions.

The results of the uncertainty analysis of Alternative 2 net costs using a 7-percent discount rate are provided graphically in Figure A-1.


Question: The process of analyzing the feasibility and effectiveness of possible solutions is an integral part of NRC's regulatory framework.

Identifying and evaluating alternative approaches to resolve problems are key elements in meeting the NRC's regulatory analysis policy.

The draft papers in development or other underlying staff documents supporting the requirements or staff positions.


Question: Class 1E electrical equipment air conditioning system requirements

LCO 3.7.20 Two Class 1E electrical equipment A/C trains shall be OPERABLE.


Question: What is the purpose of the slip clutch in a core dolly system?

In the event of operator error or equipment malfunction the torque generated by core dolly drive mechanism is limited by a slip clutch. The slip clutch is set to prevent damage if the core shroud or any other part of the core dolly comes into contact with an obstruction, such as the core shroud contacting the lead shield doors.

To eliminate the hazard, the previous FIS was modified and an override switch was added to the inside of the cabinet.


Question: What is the purpose of the Flow Accelerated Corrosion program?

In 2022, a program effectiveness review was performed for the Flow Accelerated Corrosion (FAC) Program described in FSAR Section 0.1.14 in support of preparing the BFN Subsequent License Renewal Application.

The Outage Summary reports since 2010 for each of the three units were reviewed.


Question: Limiting Condition for Operation (LCO) requirements

LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.10.

Exceptions to this Specification are stated in the individual Specifications.


Question: What is the peak locate threshold in gamma direct samples?


Question: nuclear power plant safety systems

The safety systems of a nuclear power plant include multiple redundant and diverse systems to ensure that the reactor operates safely under all conditions.

Open and log in to Microsoft Teams. Left click on your initials at the top right corner.


Question: Licensees must submit a safety analysis as part of their nuclear licensing process.

The safety analysis report is a comprehensive document required for the evaluation of the safety of a proposed nuclear facility.

The construction of the facility must comply with local zoning laws and environmental regulations.


Question: What systems require aging management in nuclear facilities?

The Auxiliary Boiler System and the Emergency High Pressure Makeup System are among the key components that necessitate aging management programs to ensure operational safety and reliability.

Cracking due to stress corrosion cracking in aluminum alloys can significantly impact infrastructure but is primarily a material failure issue irrelevant to the licensing processes.


Question: nuclear licensing requirements

Nuclear licensing involves a series of regulatory processes that ensure the design, construction, and operation of a nuclear facility comply with safety and environmental standards set by governing bodies.

The development team conducted an extensive analysis of the software architecture to determine the most efficient design patterns to apply.


Question: guidelines for steam generator maintenance

Before heatup to full power, the bulk water in the steam generators is normally brought into power operation specifications by draining and refilling or by feeding and bleeding. Guidelines for heatup are provided in EPRI, PWR Secondary Water Chemistry Guidelines.

A. Oxygen in the presence of moisture rapidly corrodes carbon steel. These corrosion products may be carried through the feedwater system and form a sludge pile in the steam generator.


Question: What are the specifications for the new fuel vault analysis as outlined in the regulatory documents?

The new fuel vault analysis will be covered by the GESTAR II methodology and is not addressed in NEDC-33938P. See Section 2.2 of LAR RS-22-064 for details.

The analysis NEDC-33938P covers peak reactivity in every spent fuel pool rack cell location.


Question: What are the updates to reactor figures in NRC documents?

Figures 15-131 to 15-136 were deleted as per various updates from the years 1996, 1998, and 2001.

The design of control systems is critical in ensuring the reactor operates within safe limits.


Question: What is the purpose of a radiologically controlled area in nuclear facilities?

A Radiologically Controlled Area (RCA) is designated for the purpose of controlling access and minimizing exposure to radiation in areas where occupational radioactive material is used or stored, ensuring safety for personnel and the public.

The site tour included a walkdown of the chemical treatment ponds, which are critical for managing any residual contaminants before discharge into the nearby waterway.


Question: What is the average enrichment of uranium in a PWR fuel assembly?

The table provides loading information for PWR fuel, detailing the initial assembly average enrichment of wt % 235U across various burnup levels.

The document includes detailed regulatory requirements for nuclear plant operations and maintenance schedules.


Question: Loss of Reactor or Secondary Coolant

1-E-1, Loss of Reactor or Secondary Coolant, is in progress

Unit 1 was at 100% when a Large Break LOCA occurred at 12:01


Question: Monitoring cladding diameter variations over the course of fuel burnup is essential for assessing reactor safety.

Figure 33-3 shows the relationship between measured cladding diameter change and the associated burnup levels.

The reactor core design must consider the thermal characteristics of the fuel as it undergoes operation.


Question: aging management evaluations for safety features

AMP XI.M20, "Open-Cycle Cooling Water System" is the program outlining aging management for components like stainless steel heat exchanger components exposed to raw water, which may experience loss of material due to pitting, crevice corrosion, and microbiologically influenced corrosion (MIC).

Not Applicable. There are no stainless steel or steel heat exchanger tubes exposed to raw water in the Engineered Safety Features systems.


Question: What is an alternative seismic method in nuclear licensing?

The license amendment request proposes to adopt an alternative seismic method for the categorization of structures, systems, and components at the Edwin I. Hatch Nuclear Plant.

The Environmental Report details the expected impacts of the proposed project on local wildlife and habitats, including potential changes in land use and water resources.


Question: The NRC emphasizes the importance of the secondary control rod design, which features an electrical system supported by a pneumatic backup for enhanced safety.

Secondary (Safety) Control Rod - Electrical, backed up by pneumatic

The primary control rod system includes components like the drive housing, electromagnet, and ball screw to ensure precise movement and control.


Question: The U.S. Nuclear Regulatory Commission oversees the licensing of nuclear facilities to ensure compliance with safety standards.

The U.S. Nuclear Regulatory Commission (NRC) is responsible for regulating commercial nuclear power plants and other uses of nuclear materials in the United States.

The Linear Heat Generation Rate (LHGR) is an important parameter in the thermal analysis of nuclear fuel that relates to the amount of heat generated per unit length of fuel rod.


Question: The NRC requires licensees to conduct thorough analyses on various radioactive materials during regulatory assessments.

Strontium-90 is reported with a result of 0.205 pCi/g, which falls under the uncertain bounds of +/-0.154.

The following Analytical Methods were performed: Method 1, EPA 905.0 Modified/DOE R501 Rev. 1 Modified; Method 2, EPA 906.0 Modified.


Question: What is a containment post-tensioning system and why is it inspected?

The containment post-tensioning system is inspected to ensure the integrity and safety of the containment structure during operation.

The technical report was prepared by an experienced engineer with a focus on compliance with standards and regulations.


Question: What are the key requirements for seismic margins assessment in nuclear safety?

However, it should be noted, that consistent with the Commission-endorsed phased PRA Quality Initiative, all risk contributors that cannot be shown as insignificant, should be assessed using a PRA (as defined in regulatory position C.1) to support risk-informed licensing actions.

The staff does not endorse the material in this Part of the standard, and as such, does not have a position (i.e., no objections, no objection with clarification, or no objection with qualification) on any of the material contained in Part 10 of the standard.


Question: What is the temperature reactivity coefficient in nuclear reactors?

The temperature reactivity coefficient is a key measure of intrinsic core safety characteristics for any particular state.

The pre-equilibrium core coefficients are preliminary, and subject to change.


Question: What materials are used in the Main Steam System components?

The materials of construction for the Main Steam System components are:
- Carbon and Low Alloy Steel Bolting
- Carbon Steel
- Stainless Steel

The following aging effects associated with the Main Generator and Auxiliaries System components require management:
- Cracking
- Cumulative Fatigue Damage
- Long-Term Loss of Material
- Loss of Material
- Loss of Preload


Question: What is a Co-60 source in the context of nuclear licensing?

I sent you some information on disposal of the Co-60 source and just making sure you did not need anything else from me.

I hope you were able to take some time off and enjoy the holiday season!


Question: impact of power uprates on reactor pressure vessel

Use guidance for EPUs - Focus on RPV surveillance, PT limits, USE, PTS, LBB, piping materials and supports, chemical control, etc. - Assess and describe performance monitoring

Identify if change affects materials, environment, loading, degradation, geometry, maintenance or mitigation


Question: The analysis of soil samples from the site restoration project involved measuring radiation levels.

The Project aims to facilitate the safe and systematic restoration of the LACBWR site.


Question: boron concentration adjustments during reactor operation

During reactor operation, changes are made in the reactor coolant boron concentration for the following conditions: a. Reactor startup - boron concentration must be decreased from shutdown concentration to achieve criticality. b. Load follow - boron concentration must be either increased or decreased to compensate for the xenon transient following a change in load. c. Fuel burnup - boron concentration must be decreased to compensate for fuel burnup and the associated buildup of fission products in the fuel. d. Cold shutdown - boron concentration must be increased to the cold shutdown concentration.

The reactor makeup control subsystem consists of a group of instruments arranged to provide a manually pre-selected makeup composition to the charging pump suction header or the volume control tank.


Question: What are the requirements for analytical evaluation of pipe wall thinning?

Use of Code Case N-597-3 to mitigate flow accelerated corrosion is authorized subject to the following conditions: The Code Case must be supplemented by the provisions of the EPRI/Nuclear Safety Analysis Center Report (NSAC) report EPRI/NSAC-2021-2, 'Recommendations for an Effective Flow Accelerated Corrosion Program,' issued April 1999 (Ref. 7), for developing the inspection requirements, the method of predicting the rate of wall-thickness loss, and the value of the predicted remaining wall thickness.

Essentially 100 percent (not less than 90 percent) of the examination volume A-B-C-D-E-F-G-H must be inspected.


Question: What is RTS instrumentation in nuclear regulatory filings?

RTS instrumentation is a crucial component that provides real-time monitoring of the reactor's safety parameters, ensuring operational integrity and compliance with regulatory standards.

The list of tables includes detailed descriptions of various systems and instrumentation used at nuclear facilities.


Question: Who is the designated Radiation Safety Officer for this nuclear device?


Question: The restrictions on utilizing sensitized stainless steel aim to enhance corrosion resistance during the various phases of construction and operation.

The use of severely sensitized stainless steel as a pressure boundary material has been prohibited and has been eliminated by either a select choice of material or by programming the method of assembly (see Section 5.4.3.1). This restriction on the use of sensitized stainless steel has been established to provide the primary system with preferential materials suitable for the following: - a. Improved resistance to contaminants during shop fabrication, shipment, construction, and operation.

All the insulation modules are removable, but access to vessel side insulation is limited by the surrounding concrete.


Question: Loss of Normal Feedwater

CLOSE recirc valve for the tripped MFWP:


Question: Zircaloy cladding temperature increase

by the metal-water reaction of zirconium with steam is greater than the heat that can be transferred away from the Zircaloy cladding. This causes the Zircaloy cladding temperature to increase, thereby increasing the diffusion of oxygen into the metal, which in turn increases the rate at which the zirconium-steam oxidation reaction occurs.

To evaluate conformance with the criteria specified in 10 CFR 50.46(b), ECCS cooling performance must be calculated using an acceptable evaluation model for a range of postulated LOCAs of different sizes, locations, and other properties sufficient to


Question: What are the safety limits for reactor core operation?

**2.1.1** The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits shown in Figure 1.

The cycle-specific parameter limits for the specifications listed above are presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.6.b.


Question: What are the identification marking requirements for Type B packages?

For Type B(U), Type B(M) or fissile material package destined for export shipment, mark 'USA' in conjunction with specification marking, or certificate identification; and package identification indicated in U.S. Competent Authority Certificate.

The arrows must either be black or red on white or other suitable contrasting background and commensurate with the size of the package; depicting a rectangular border around the arrows is optional.


Question: What is the GALILEO Fuel Rod Thermal-Mechanical Methodology?

GALILEO Fuel Rod Thermal-Mechanical Methodology is a framework used to assess the behavior of fuel rods under varying thermal and mechanical conditions, specifically designed for Pressurized Water Reactors.

Framatome Inc. provides a range of services and products to the nuclear industry, focusing on enhancing the safety and efficiency of nuclear power operations.


Question: What is the significance of Americium-241 in nuclear licensing?

Americium-241 is measured in picocuries per gram (pCi/g) and is recorded during quality control tests to ensure the safety and regulatory compliance of nuclear materials.

The QC summary indicates uncertainty values for various radioactive materials, which are essential for accurate measurements and evaluations.


Question: What does the HEC-RAS Model indicate in nuclear plant regulatory filings?

HEC-RAS, or the Hydrologic Engineering Center's River Analysis System, is a computer program used for simulating the hydraulics of water flow through natural and constructed channels.

Various rivers and sites are illustrated.


Question: The NRC is involved in the process of evaluating the environmental impacts associated with the subsequent license renewal for the Oconee Nuclear Station.

Oconee Nuclear Station, Units 1, 2, and 3, February 2024 Subsequent License Renewal Public Draft Environmental Impact Statement Meetings-Memo.

Federal Register Notice: Environmental Impact Statements; Notice of Availability.


Question: What should be done when a turbine experiences high vibrations?

As Work Control acknowledge request to initiate Condition Report and initiate work Order for repairs.

When dispatched wait 1 minute and report oil spray is preventing entry into 1A-A EDG Room.


Question: proprietary information in nuclear regulatory filings

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary.


Question: The analysis of transportation impacts related to radiological waste and spent fuel does not indicate any significant changes due to regulatory updates.

The impacts from transportation of radiological waste and spent fuel is not significant.

There will be no changes in the impacts from transportation of radiological waste and spent fuel.


Question: The contour map indicates the piezometric levels throughout the water table aquifer at the site.

The Water Table Aquifer Contour illustrates the varying levels of the piezometric contour, which is crucial for understanding groundwater flow in the vicinity of the plant.

The coordinates for the plant site are defined as 63N and 40E, referencing geological studies conducted in 1987.


Question: What are acceptable methods for adding boron to the reactor coolant system?

Clarified acceptable methods for adding boron to the reactor coolant system.

Made under the provisions of 10 CFR 50.59 except where indicated in brackets.


Question: What methodologies are used for dam breach modeling in nuclear site assessments?

The approach and sequence of steps taken to calculate the maximum WSEL, peak flows and velocities caused by dam failure are as follows: USACE HEC-HMS, version 3.3 (Reference 2.4.4-8) is used to calculate the flows from the dam breach analysis.

There are hundreds of reservoirs within the Delaware River watershed with storage capacity ranging from 15 ac.-ft. to 460,000 ac.-ft.


Question: What are the requirements for testing under NELAC certification?

Except as noted, the test results for the methods and parameters listed on ACZ's current NELAC certificate letter (#ACZ) meet all requirements of NELAC.

All samples and sub-samples associated with this project will be disposed of after December 04, 2024.


Question: nuclear reactor licensing process

The licensing of a nuclear reactor involves a comprehensive examination of safety protocols, regulatory compliance, and operational readiness that must be documented in the submittal for review and approval.

Audit plans are critical in ensuring compliance with regulatory standards and verifying adherence to operational procedures.


Question: source material transfer reporting requirements

For each general license under He-P 4031.03, equivalent agreement state provisions for who greater than 50 grams (0.11 lb) of source material has been transferred in a single calendar quarter, or Nuclear Regulatory Commission 10 CFR 40.22, the name and address of the general licensee to whom source material is distributed; a responsible agent, by name and/or position and phone number, of the general licensee to whom the material was sent; and the type, physical form, and quantity of source material transferred; and

Submit each report by January 31 of each year covering all transfers for the previous calendar year. If no transfers were made to persons generally licensed under He-P 4031.03, equivalent agreement state provisions during the current period, or Nuclear Regulatory Commission 10 CFR 402.2, a report shall be submitted to DHHS/RHS indicating so. If no transfers have been made to general licensees in New Hampshire during the reporting period, this information shall be reported to DHHS/RHS upon request.


Question: The NRC mandates that licensees must conduct analyses of the radiological consequences of accidents.

Licensees must demonstrate compliance with the regulations by providing sufficient data and analysis concerning the safety systems in place and their functional capabilities.

The USDA provides comprehensive statistics on agriculture practices and outputs in South Carolina, detailing crop yields and livestock inventories.


Question: The estimated emissions from typical passenger cars during operation are crucial for environmental assessments.


Question: What requirements must Licensees meet to mitigate beyond-design-basis external events?

Fukushima Dai-ichi leads the staff to conclude that additional requirements must be imposed on Licensees or CP holders to increase the capability of nuclear power plants to mitigate beyond-design-basis external events.

Compliance with NRC requirements plays a critical role in giving the NRC confidence that Licensees or CP holders are maintaining an adequate level of public health and safety and common defense and security.


Question: main steam routing diagram

Figure 3-13. Main Steam Routing Elevation Outside Containment Farthest From Reactor Building Wall

Elevation annotated with: - 20' markers for reference - Various connections and junctions to be noted - Section view indicated


Question: The report lists the Minimum Detectable Activity for various isotopes measured in the investigation.

The Minimum Detectable Activity (MDA) is the smallest amount of a nuclide that can be reliably detected in a sample and is expressed in picocuries per gram (pCi/grams).

Errors quoted at 2.000sigma indicate the confidence levels used to ascertain the integrity of the measurements.


Question: aging management program for concrete

Aging Management Review Results detail that the Structures Monitoring Program (B.2.3.33) addresses various aging effects requiring management for concrete foundations, such as cracking, loss of bond, and increase in porosity.

The outdoor water storage tanks are designed to provide water supply for firefighting and emergency response actions in case of an incident.


Question: risk-informed approach for accident analysis

The proposed change in methodology in this LAR is to use a risk-informed approach to address the effects of accident-generated and transported debris on the containment emergency sumps instead of a deterministic approach.

The dose considerations discussed above demonstrate that compliance would result in substantial personnel exposure due to insulation modifications in containment, which is not commensurate with the expected safety benefit based on the risk evaluation results showing that the risk associated with post-accident debris effects is less than the threshold for Region III in Regulatory Guide (RG) 1.174 (Reference 2).


Question: What does sample receipt verification entail in nuclear regulatory filing?

Sample receipt verification is a process that ensures that all samples listed on the chain of custody received are accounted for and intact, which is critical for maintaining accurate records in regulatory filings.

The shipping container custody seals must be present and intact for all deliveries, as this ensures the authenticity of the materials being received.


Question: What is the relationship between fuel enrichment and cooling time in PWR assemblies?

The loading table outlines the relationship between the initial assembly average enrichment of Uranium-235 (U-235) and the minimum cooling time required for various pressurized water reactor (PWR) vendor designs.

The NRC requires annual inspections of all nuclear facilities to ensure compliance with safety regulations.


Question: The NRC mandates that the amount of tritium released into the environment must be controlled and monitored during reactor operations.

During refueling, the tritium concentration in the containment atmosphere is maintained at a safe level by the containment purge system.

Based on the above, the following conclusions were reached: 1. The tritium levels in plants operating with Zirconium alloy-clad cores are substantially lower than original design predictions.


Question: The Z-port is designed to provide a backup method for component actuation if the primary system encounters a failure.

The Z-port is hardwired to the DPS/RRCS for a diverse means to perform component actuation for certain protection functions in case the PPS fails to do so due to a CCF.

The configuration of existing safety systems must be reviewed regularly to ensure compliance with updated regulatory standards.


Question: The NRC oversees the environmental considerations associated with regulatory activities in nuclear energy.

Jeffrey Rikhoff, NRC has 44 years of combined industry and government experience in NEPA compliance for DOE Defense Programs/National Nuclear Security Administration and Nuclear Energy.

Ted Smith has 39 years of experience, including DOE Power Administration, support of site environmental management programs, and spent fuel management.


Question: What impact do topographic features have on wind direction in nuclear site evaluations?

The topographic features of the ESP site region, in conjunction with the movement of pressure systems and the location of the semi-permanent pressure zones, have a definite influence on the wind direction distribution.

Consequently, low-pressure cells that are generated in the Gulf are frequently forced to move toward the east on the back (west) side of the Blue Ridge Mountains, therefore resulting in a southerly flow of air in the ESP site region instead of a southeast or easterly wind.


Question: What is the role of the New Reactor Licensing Branch in the NRC?

The New Reactor Licensing Branch is responsible for the review of applications for licenses to construct and operate new nuclear reactors, ensuring compliance with safety and environmental regulations.

The financial support provided by the Division of Rulemaking is essential for the ongoing stability of regulatory operations in various sectors.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves several key stages, including the pre-application phase, the application review stage, and the post-licensing activities, which ensure that the reactor meets the required safety and regulatory standards.

The Quad Cities region is known for its scenic views along the Mississippi River, which attract many tourists each year.


Question: What is the COBRA-FLX critical heat flux correlation?

COBRA-FLX is a thermal-hydraulic code developed by Framatome for light water reactors, which includes a critical heat flux correlation that helps in predicting the limits of coolant flow and heat transfer under varying operating conditions.

The report discusses various test assembly configurations used during the HTP grid tests, including their geometrical dimensions and material specifications, but does not address thermal-hydraulic codes.


Question: The NRC is developing a regulatory framework that will support the safe deployment of advanced reactor technologies...

The proposed 10CFR53 outlines a risk-informed, technology-inclusive regulatory framework aimed at advanced reactors, emphasizing safety and performance standards.

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


Question: The NRC requires sufficient automatic controls and indications independent of the PPS to manage event scenarios like MSIV Closure.

For the postulated event of MSIV Closure, concurrent with a postulated CCF of PPS, sufficient automatic control functions, indications that are independent of the PPS, and operator actions, are/will be available to mitigate the event.

The condenser vacuum is assumed to be lost at a rate of approximately 2 inches Hg per second. This event is categorized as an incident of moderate frequency.


Question: human engineering discrepancies

HEDs were initially categorized using a subjective approach that reflected the consensus judgment of the multidisciplinary CRDR team.

The following five categories were designed to be unique so a consensus could be obtained from the review team as to which category each HED should be assigned.


Question: What is the role of a Health Physicist in nuclear licensing?

**Casey Alldredge**  Health Physicist  Materials Licensing Branch  Division of Radiological Safety & Security  Region IV, USNRC

I will give you a call on Monday to discuss.


Question: What is the minimum signal to noise ratio required for SRM operation?

The proposed change to TS SR 3.3.1.2.4 will increase the minimum SRM signal to noise ratio from ≥ 2:1 to ≥ 20:1, when the SRM count rate is ≥ 0.7 cps and less than 3 cps.

The proposed change does not alter the Nuclear Instrumentation system design, create new failure modes, or change any modes of operation.


Question: What are the regulatory guidelines for uranium concentration evaluations?

40 CFR Part 192, Subpart A, §192.02 (c)(3)(ii)(B)(2), specifies that an evaluation of potential hazards from proposed ACLs must consider several factors, including the potential for health risks caused by human exposure to constituents.

While the ponds may not be a current source of recreational fishing, flooding of rivers is commonly observed to result in transferal of fish to nearby surface water bodies.


Question: What is the ultimate heat sink in nuclear power plants?

The power block area, which contains all safety-related structures including the ultimate heat sink (if applicable), is shown relative to the exclusion area boundary, with distance measurements from the power block reference point along the 16 cardinal compass points.

The VCS site is bracketed by the U.S. Geological Survey 7.5 minute quadrangles named Raisin, Bloomington, McFaddin, and Bloomington SW.


Question: NRC K/As referenced in examination questions

2. a. NRC K/As are referenced for all questions.

The sampling process was random and systematic.


Question: What are the proposed changes to 10 CFR Part 51 regulations?

The purpose of this rulemaking is to perform substantial revisions to 10 CFR Part 51 that would align 10 CFR Part 51 with the NRC’s efforts to streamline and update the NRC’s NEPA process as well as incorporate relevant and appropriate conforming changes from CEQ’s July 2020 final rule.

A rulemaking to revise 10 CFR Part 51 regulations would be consistent with ongoing agency efforts to comply with new Federal requirements (e.g., NEIMA, FAST-41) as they relate to the NRC’s environmental reviews, allow flexibility to address advanced reactor designs and new technologies, and codify best practices and lessons learned that could reduce staff and applicant burdens when preparing environmental documents.


Question: potential impacts of lighting during operation

Potential lighting impacts will be minimized by implementing protection and mitigation measures discussed in Section 5.9 of this ER. This will help ensure that the potential visual and scenic resources impacts from lighting will be SMALL during operation.

The No Action Alternative would result in no change to visual and scenic resources in the project areas.


Question: What is the significance of burnup in nuclear fuel?

Average burnup for the power assembly of 25,000 MWd/t or less (this corresponds to a peak local burnup of about 45,000 MWd/t).

For release pressures greater than 1200 psig and water depths less than 23 ft, the iodine decontamination factors will be less than those assumed in this guide and must be calculated on an individual-case basis using parameters in conservatism to those of this guide.


Question: What is the evidentiary hearing process in nuclear regulatory filings?

The evidentiary hearing that is the subject of the instant appeal followed.

Cultural Resources Site Survey Methodology (February 2019 Methodology), which it provided to the Oglala Sioux Tribe for review.


Question: non-power reactor licensing report

A non-power reactor specific technical report will be submitted with the PSAR to implement an ANSI 15.8 QA program, consistent with NUREG-1537.

The non-power reactor will evaluate a maximum hypothetical accident. A formal LMP implementation is not anticipated.


Question: The pages in Appendix A of the Technical Specifications are being updated with revisions that are marked for clarity.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

The table outlines the specific pages that need to be replaced, ensuring that the documents are up to date with the latest information.


Question: regulations for endangered species consultation

If a Federal agency determines, based on the Biological Assessment or biological evaluation, that listed species and/or designated critical habitat may be affected by the proposed project, the agency is required to consult with the Service pursuant to 50 CFR 402.

The MBTA has no provision for allowing take of migratory birds that may be unintentionally killed or injured by otherwise lawful activities.


Question: safety analysis reports in nuclear licensing

Safety analysis reports (SARs) provide information about the design, operation, and safety of nuclear facilities, including evaluations of potential hazards and risks associated with the operation of the reactor.

The environmental impact statement must be submitted for review, outlining the potential effects of the proposed project on local wildlife and ecosystems.


Question: What is the purpose of the reactor bottom head drain in the RWCU system?

The nominal pipe diameter is a critical parameter for ensuring suitable flow rates in the system.


Question: The maximum recorded 24-hour rainfall at the Salem/Hope Creek site is significant for evaluating environmental safety conditions.

Maximum Recorded 24-Hour Rainfall (inches) at the S/HC Site is 10.03.

Periods of record reviewed include various locations such as Dover and Hammonton, each with its extensive historical data.


Question: What is the significance of water discharge measurements in nuclear facility regulations?

Discharge data for Salt Creek at Rowell indicates the changes in water flow dynamics pre-and post-dam construction, which can impact various ecological and engineering assessments in regulatory filings.

Mean monthly runoff data shows that in January, the mean runoff was 0.80 inches, suggesting significant seasonal variability in water availability.


Question: nuclear licensing process

The licensing process for a nuclear facility involves several stages, including site evaluation, system design approval, and operational licensing, all of which must comply with NRC regulations to ensure safety and environmental protection.

The California Department of Health oversees various public health initiatives, including vaccination programs and health education campaigns aimed at reducing the spread of infectious diseases.


Question: What is the purpose of a materials license in nuclear regulation?

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Parts 30, 31, 32, 33, 34, 35, 36, 37, 39, 40, 70 and 71, and in reliance on statements and representations heretofore made by the licensee, a license is hereby issued authorizing the licensee to receive, acquire, possess, and transfer byproduct, source, and special nuclear material designated below; to use such material for the purpose(s) and at the place(s) designated below; to deliver or transfer such material to persons authorized to receive it in accordance with the regulations of the applicable Part(s).

Nuclear Diagnosis, Inc. is responsible for complying with the conditions of their license and must ensure that all authorized uses of the materials are correctly reported.


Question: What are the requirements for dielectric withstand tests in generator circuit breakers?

The generator breaker is qualified for dielectric withstand in accordance with ANSI Standard C37.09.1964, paragraph 09.4.10 including low frequency withstand voltage tests and impulse withstand tests, (rate voltage: 24 kV, basic impulse level: 150 kV) without any compressed air.

Forty successive interruptions at 40 kA r.m.s. symmetrical are performed at minimum rated air pressure and without changing contact fingers.


Question: The NRC requires an analysis of safety systems and their effectiveness in preventing accidents during operation and under transient conditions.

Paragraph 50.34(a)(4) requires that applicants for a construction permit (CP), and subsequently for an operating license (OL) in (b)(4), provide an analysis and evaluation of the design and performance of SSCs of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.

In addition, this analysis must include the determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.


Question: What are the requirements for spent fuel storage cask renewal applications?

10 CFR 72.240, “Conditions for spent fuel storage cask renewal,” provides requirements for applications for the renewal of CoCs for spent fuel storage cask designs, including requirements for TLAA and AMPs.

Regulatory guides are not NRC regulations and compliance with them is not required.


Question: The NRC requested a detailed evaluation of the probabilistic risk assessment as part of the regulatory framework for nuclear safety.

The Risk-Informed Technical Specifications (RITS) initiative aims to provide a structured framework for incorporating risk insights into the technical specifications of nuclear power plants, enhancing operational safety and regulatory compliance.

The IEEE-279-1971 standard outlines essential criteria for the design and evaluation of protection systems in nuclear power generating stations.


Question: Licensees must conduct safety analyses to demonstrate compliance with regulatory requirements.

The NRC requires that licensees submit a safety analysis report detailing potential hazards associated with the facility's operation.

The licensing process involves various administrative procedures that are necessary to ensure compliance with federal standards.


Question: RCS cooldown procedures

The crew will transition to 2-EOP-TRIP-5 if the required RCS Cooldown Rate is greater than a MINIMUM of 50 °F / Hour.

At time 12:30, the crew has just completed the RCS Stabilization Steps of 2-EOP-TRIP-4 and is now preparing to Cooldown and Depressurize the RCS.


Question: The NRC's procedure involves making electronic copies of certain documents available to the public as per regulatory guidelines.

In accordance with 10 CFR 2.390 of the NRC’s “Rules of Practice,” a copy of this letter will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

The licensing process includes comprehensive assessments and documentation to ensure safety and compliance in nuclear facilities.


Question: strontium-90 detection in nuclear analysis

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


Question: What are the criteria for SI termination in nuclear safety systems?

Starting a charging pump with no CCW cooling to RCP thermal barrier


Question: To maintain operational safety, a temporary solution involving an intake canal level probe is required during maintenance activities.

Approval to use the temporary SW jumper and Intake Canal level probe is requested for a maximum of 35 days during the currently planned 2025 Unit 1 refueling outage.

The maintenance activity will also require Intake bays “B” and “D” to be de-watered.


Question: nuclide identification procedures

The Nuclide Identification Report lists various nuclides along with their corresponding confidence levels, energies, yields, and activities in pCi/grams.

The report includes several procedural guidelines for submitting regulatory documents to the NRC, outlining the necessary forms and deadlines for compliance.


Question: What are the NRC's approaches to alternative methods for demonstrating compliance in nuclear safety?

If you get to burst, if you get to significant deformation and dispersal and uncertainty, then it seems to me that's the branch point where you then invoke these alternate -- whether it's Option 2 or Option 4 that we heard about this afternoon.

It's kind of backwards in my mind. Set out some criteria objectives, whether they're functional or there's prescriptive limits or both.


Question: What are the requirements for sump strainer structural qualification in nuclear plants?

The Wolf Creek sump strainer structural qualification analysis evaluated the strainer modules as well as the supporting structures.

The financial forecasting report indicated a significant increase in projected revenue for the upcoming fiscal year.


Question: The NRC evaluates various heat dissipation systems to ensure regulatory compliance for Units.

Initial Screening of Alternative Heat Dissipation Systems - Unit 4 (Base Case & Alternatives 7-9) outlines the methods and rationale used for analyzing the potential heat dissipation systems for effective reactor operations.

Operations-Related Unavoidable Adverse Environmental Impacts detailed the effects of the reactor on the surrounding environment during regular operational activities.


Question: What are the general actions outlined in S2.OP-AB.CR-0002(Q) concerning safety protocols in the control room?

Describe, in general terms, the actions taken in S1/S2.OP-AB.CR-0002(Q) and the bases for the actions.

The examination answer key focuses on cognitive levels and question sources such as New, Modified Bank, and Bank.


Question: The NRC oversees inspections to ensure the safe operation of fuel cycle facilities.

The NRC requires comprehensive inspections of fuel cycle facilities to evaluate operational safety and security measures in accordance with established guidelines.

The 2600 Fuel Cycle Facility Operational Safety and Safeguards Inspection Program outlines procedures for inspections.


Question: What is the role of the Nuclear Safety Project in regulatory comments?

The Nuclear Safety Project aims to ensure that the technology and processes used in nuclear power are safe and effective for public health and the environment, addressing issues raised in regulatory filings and petitions for rulemaking.

The Union of Concerned Scientists regularly submits comments to various agencies to ensure accountability and transparency in environmental monitoring.


Question: What is the intended function of the Primary Water Receiver Tanks in a nuclear plant?

The tank (Primary Water Receiver Tanks) serves as a leakage boundary (spatial).

The tank (RCP Lube Oil Collection) maintains a pressure boundary in the lubricating oil environment.


Question: combined license supporting testing of manufactured reactors

Subpart H, DRAFT Section 53.1480 – Combined license supporting testing of manufactured reactors (ML24344A037)

NRC Staff Response dated February 10, 2023 (ML22341A047)


Question: Nuclear plants must comply with specific regulatory requirements to maintain their operational licenses.

The NRC carefully reviews license applications and renewals to ensure compliance with safety standards.

The power generation sector is facing increased competition from renewable energy sources, which are becoming more efficient.


Question: effect of Raw Cooling Water malfunction on turbine systems

The applicant is required to have knowledge of the effect that a Raw Cooling Water malfunction (turbine building CCW) will have on the Main Turbine Generator System.

Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the Stator Cooling Water System will have on the following: a. Main generator.


Question: NEPA responsibilities in regulatory filings

Accordingly, we find no clear error in the Board's ruling that the Staff has satisfied its NEPA responsibilities, and we deny the Consolidated Intervenors’ petition for review.

American groups themselves. The Staff began its efforts to consult with various affected Tribes in 2011, and a field survey was conducted on the site with three Tribes (although not the Oglala Sioux Tribe) participating.


Question: What is the significance of assembly average enrichment in nuclear fuel?

**Minimum Initial Assembly Avg. Enrichment wt % 235U (E)**

The table provides various enrichment levels for different reactor types, detailing the minimum cooling times associated with specific burnup ranges.


Question: The NRC assesses licensee compliance with external flooding mitigation plans and procedures during inspections.

The inspectors evaluated that flood protection barriers, mitigation plans, procedures, and equipment are consistent with the licensee’s design requirements and risk analysis assumptions for coping with external flooding.

Unit 2 operated at or near 100 percent RTP for the entire inspection period.


Question: What are best management practices (BMPs) in nuclear operations?

The application of BMPs to minimize the area affected, to control fugitive dust, runoff, and erosion from project sites; to reduce the spread of invasive non-native plant species; and to reduce wildlife disturbance in adjacent habitats could greatly reduce the impacts of continued operations and refurbishment activities.

Terrestrial resources are described in Section 3.7.2. No license renewal-related construction activities or changes in operational practices have been identified that would involve disturbing habitats.


Question: The NRC monitors the stability characteristics of different core designs in reactors.

The Axial Stability Index is a critical parameter for the Pressurized Water Reactor Core, particularly with a height of 12 feet, as it influences the reactor's stability during operation.

Table 4-12 displays various metrics related to reactor operation, such as burnup and the corresponding values for Fz and Cₕ.


Question: average burnup in nuclear reactors

46 X Assembly Average Burnup ≤ 47 GWd/MTU


Question: initial assembly average enrichment 235U

Initial Assembly Avg. Enrichment wt % 235U (E)


Question: What criteria must be met for recycling processed liquids in a nuclear plant?

Processed liquids are recycled for reuse within the plant whenever possible, provided that the following criteria are satisfied: - The plant water inventory requires makeup. - The water to be reused satisfies system water quality requirements. - Tritium buildup is less than plant operating limits.

The equipment utilized during liquid waste processing is at the discretion of the operator; therefore, the calculated releases conservatively do not address all possible treatment processes but only the expected process.


Question: surrogate radionuclides

The surrogate ratio for soil is given in Table 5-1.

The level of effort associated with planning a survey is based on the complexity of the survey and nature of the hazards.


Question: What are the standards for summary disposition in NRC proceedings?

Summary disposition is appropriate where there is no remaining material issue of fact. The standards governing summary disposition are set forth at 10 C.F.R. § 2.710(a) and are based upon those the federal courts apply to motions for summary judgment under Rule 56 of the Federal Rules of Civil Procedure.

The expansion of issues for resolution and the continuation of litigation that results from admitting a contention or denying summary disposition does not necessarily have a "pervasive and unusual" effect on the litigation.


Question: What are the requirements for obtaining a nuclear reactor operating license?

The regulatory requirements for obtaining an operating license for a nuclear reactor are detailed in 10 CFR Part 50, which includes provisions for safety analysis, environmental considerations, and operational standards necessary to ensure public health and safety.

The safety culture of a nuclear facility is critical to maintaining operational excellence, and it encompasses the attitudes, values, and behaviors of personnel that prioritize safety across all levels of the organization.


Question: nuclear licensing requirements

Nuclear regulatory licensing requirements involve a series of detailed safety and operational criteria that must be met by applicants seeking to operate a nuclear facility.

The cooling system is essential for maintaining the reactor's temperature within safe limits and preventing overheating during operation.


Question: What are the responsibilities of the Emergency Response Organization in nuclear emergency planning?

The functional responsibilities assigned to the ERO are established and the responsibilities that may not be delegated to other members of the ERO are clearly specified in the emergency plan.

Non-delegable responsibilities include the following: - Event declaration - ORO and NRC Notification - PARs for the general public - Emergency Exposure (Dose limits and KI)


Question: What is the AREA™ methodology in nuclear safety analysis?

The AREA™ methodology is limited to only the GALILEO™ derived thermal-mechanical properties of fuel pins.

The limitations conditions for each of the topical reports used in the REA analysis are addressed in this section.


Question: The net peak area for the energy level of 238.17 keV is indicated as 4.33E+01 in the measurements.

Net Peak Area of the energy level is recorded as 4.33E+01 for the peak at 238.17 keV.

The ambient background measurement for the peak at 238.17 keV is 4.18E+01, with a background uncertainty of 1.86E+01.


Question: What are the test assembly conditions for HTP grid tests?

Table B-2 lists the test assembly conditions for HTP grid tests, including various parameters such as pressure, temperature, and flow rates.

Framatome Inc. provides a wide range of services for nuclear power plants, including design and engineering solutions.


Question: What is the significance of Americium-241 in nuclear analysis?

Americium-241 is a radionuclide that is often analyzed to assess the levels of radioactive contamination in solid matrices, typically expressed in picocuries per gram (pCi/g).

The Certificate of Analysis lists various parameters and their corresponding results for several radioactive isotopes, which include qualifiers, uncertainties, and measurement methods executed during tests.


Question: The operator must ensure proper management of the steam generator system following any significant failure.

Isolate SG #1 IAW 1-E-3 Steam Generator Tube Rupture, to prevent entry into 1-ECA-3.1, SGTR and LOCA – Subcooled Recovery.

Early notification of Radiation Protection and Chemistry could expedite subsequent sampling efforts if needed.


Question: What are the acceptance criteria for testing the Annulus Ventilation System pressure?

Acceptance criteria for the functional tests are as follows: 1. Annulus pressure reaches -1.2 in wg within 22 seconds of a start signal. 2. Annulus pressure reaches -4.2 in wg within 48 seconds of a start signal. 3. The annulus pressure decay time from the beginning of a vacuum decay to -1.2 in wg is ≥278 seconds.

Operational tests were performed prior to initial startup to demonstrate proper functioning of the system.


Question: Westinghouse implements a system to protect sensitive information that is crucial for maintaining a competitive edge in the market.

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information.


Question: The safety analysis must demonstrate that the thermal limits are maintained during operation.

The COBRA-FLX code is used to calculate the critical heat flux, which is a measure of the heat transfer capability in nuclear reactors under different thermal-hydraulic conditions.

The construction schedule for the new reactor was finalized and submitted for review by the project management team.


Question: spurious operations in control systems

Draft Revision 8 of Branch Technical Position (BTP) 7-19 is proposing changes to the guidance regarding the treatment of malfunctions and spurious actuations in NSR control systems.

The NRC staff has determined that spurious operation is a potential outcome of CCF originating from latent design defects.


Question: design life in nuclear licensing

Please clarify how the design life of the plant was determined for the stress and crack growth analyses with respect to the renewed license, which is scheduled to expire in 2044.

During our audit of the validation of the AREVACSC code, the direct comparison with the [[ ]] was unclear.


Question: Sealed sources must undergo testing if they have been used or if they are transferred, ensuring safety compliance.

Sealed sources need not be tested if they are in storage and are not being used. However, when they are removed from storage for use or transferred to another person, and have not been tested within the required leak test interval, they shall be tested before use or transfer.

Records of leak test results shall be kept in units of becquerels (microcuries) and shall be maintained for 3 years.


Question: To assign a LOADING CATEGORY to the CANISTER, the maximum heat load from the fuel assemblies must first be calculated and cross-referenced with the heat load matrix.

Establish LOADING CATEGORY by determining the total heat load for all of the fuel assemblies to be loaded into a CANISTER and then determining the maximum individual fuel assembly heat load for these assemblies.

The CANISTER's design and construction must withstand extreme environmental conditions to ensure the secure storage of nuclear waste.


Question: What information must be provided regarding the radioactive materials possessed by the applicant?

a. Element and isotopes, chemical and/or physical form; and maximum amount which will be possessed at any time.

The Materials Safety Licensing Branch oversees the distribution of exempt products in accordance with the established regulatory procedures.


Question: The NRC makes accommodations for individuals to ensure everyone can participate in meetings.

The NRC provides reasonable accommodation to individuals with disabilities where appropriate.

The Commission's Policy Statement on 'Enhancing Public Participation in NRC Meetings' aims to improve the engagement of the public during regulatory processes.


Question: What is the aging effect requiring management for structural bolting in nuclear facilities?

Loss of material due to general, pitting, crevice corrosion

Increase in porosity and permeability; cracking; loss of material (spalling, scaling) due to aggressive chemical attack


Question: Samples were evaluated based on specific analytical methods, ensuring compliance with procedural acceptance criteria.

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

There are no exceptions, anomalies or deviations from the specified methods.


Question: liquid waste disposal alternatives

The NRC staff evaluates the use of evaporation ponds and surface water discharge, which have historically been used by ISR facilities to manage and dispose of liquid wastes.

After weighing the impacts of the proposed action and comparing the alternatives, the NRC staff determined that the proposed action is the preferred alternative and that the NRC should issue a source materials license for the proposed action.


Question: What is a nuclear reactor licensing process?

The licensing process for a nuclear reactor involves a detailed review of safety, environmental, and operational considerations to ensure compliance with regulatory standards set by the Nuclear Regulatory Commission (NRC).

The recent findings in the environmental study showed a significant impact on the local ecosystem due to increased industrial activities in the surrounding area.


Question: auxiliary feedwater system reliability analysis

An AFW analysis was performed in accordance with Action Items II.E.1.1 and II.E.1.2 of NUREG-0737 using the methodology defined by Appendix III and Annex 1 of Appendix X of NUREG-0611 and NUREG-0635 to determine the system reliability and major contributions to AFW failure under various loss-of-main-feedwater transients.

Inservice inspection and testing is performed as discussed in section 3.9.


Question: What is the purpose of the reactor containment structural integrity test?

To demonstrate the structural integrity of the reactor containment building.

The HEPA filters operate as described in subsection 6.5.1 and section 9.4.


Question: effective regulation in fusion energy

Effective regulation can encourage more private sector investment in current and near-term R&D phases.

Fusion will work in concert with renewables to deeply decarbonize our energy grids.


Question: Pressurizer Safety Valve Operation

Pressurizer Safety Valve Operation                   | 40     | 40

Inadvertent Startup of an Inactive Loop            | 15     | 15


Question: What is the purpose of the Remote Shutdown System in a nuclear facility?

SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown program.

The test verifies the channel responds to measured parameter values with the necessary range and accuracy.


Question: What is fuel qualification in nuclear reactor licensing?

Fuel qualification is the process of evaluating and determining that a specific nuclear fuel can perform its intended functions safely within the operational parameters of the reactor.

The meeting includes a public notice and a list of attendees who discussed various items related to the regulatory process.


Question: critical steps in nuclear regulatory performance assessments

**Critical Step?**  Y  **Standard:**  Attachment 1 completed per the key found on the following page.  This step is critical to ensure accurate information is provided in notifications.

**Performance:**  - [ ] SATISFACTORY  - [ ] UNSATISFACTORY  **Examiner Notes:**  All data CIRCLED below must be accurately marked to pass this JPM.


Question: inservice inspection requirements

This letter provides a copy of Form OAR-1 Owner's Activity Report for inservice inspection activities performed during the previous fuel cycle. This report meets the requirements of the ASME Code Case N-532-5, “Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission.”

This report covers the inservice inspections conducted prior to and during the Unit 2 Fall 2020 refueling outage.


Question: What is the energy range for CE-144 in keV?

CE-144 has an energy of 133.51 keV and a yield of 11.09%.

AM-241 has an energy of 59.54 keV with a yield of 36.30%.


Question: The project areas include significant regions such as the Church Rock Sections and Crownpoint.

The Chaco Watershed is identified with the HUC code 14080106, which falls within the scope of the project areas discussed in this document.

The Navajo Nation is indicated within the map boundaries, showcasing the geographical context of the project.


Question: The area included in the 50-mile emergency planning zone is critical for assessing potential radiological impacts.

The ingestion exposure pathway typically encompasses a 50-mile radius around a nuclear facility to account for potential radiological contamination.

The cooling system design specifications are crucial for maintaining reactor temperatures within safe operational limits.


Question: The analysis methodology for non-loss-of-coolant accidents must consider a variety of factors to ensure safety protocols are met.

The pressurizer is a key component in maintaining the necessary pressure in the primary coolant system of a pressurized water reactor.

The turbine generator converts thermal energy from steam into electrical power, contributing to overall plant efficiency.


Question: electrical conductor insulation

Electrical conductor insulation for inaccessible power, instrumentation, and control cables (e.g., installed in duct bank, buried conduit or direct buried) composed of various organic polymers such as EPR, SR, EPDM, XLPE, butyl rubber, and combined thermoplastic jacket/ insulation shield exposed to an adverse localized environment caused by significant moisture.

Surface cracking, crazing, scuffing, dimensional change (e.g., "ballooning" and "necking"), shrinkage, discoloration, hardening or loss of strength due to elastomer degradation.


Question: The NRC emphasizes the analysis of fire-related issues in reactor designs.

K Not Applicable - This SAMA is based on a plant-specific fire insight. At Dresden, there is not a 480V switchgear room; the 480V busses and panels are located throughout the plant.

SAMA Category Key defines various categories such as A: AC/DC, B: Core Cooling, C: FW/Condensate, and others.


Question: The levels of certain radionuclides are assessed in regulatory filings to ensure public safety.

The unit of measurement known as picocurie per gram (pCi/g) is commonly used in radiation measurements.


Question: Ensure the Emergency Operations Center is fully prepared for any urgent local incidents.

Verify that the Emergency Alert System (EAS) was alerted to standby status by the State EOC.

Review Notification Message Forms and perform time check with State EOC for the purpose of synchronizing clocks in the EOC and on fax machine.


Question: uranium bioassay process

Bioassays will be performed for all workers who are routinely exposed to airborne yellowcake or excessive levels of yellowcake, such as may occur when maintenance work is performed in yellowcake areas.

Results will be obtained within 20 days of the collection and corrected to standard urine specific gravity of 1.02.


Question: overtopping failures in dam safety

Overtop dam break represents failures caused by overtopping of the dam.

Total time for the breach to form from initiation to reaching the maximum size (in hours).


Question: Section 401 Certification requirements

CWA Section 401 [33 U.S.C. 1341(a)(1)] states: “No license or permit shall be granted until the certification required by this section has been obtained or has been waived as provided in the preceding sentence. No license or permit shall be granted if certification has been denied by the State, interstate agency, or the Administrator, as the case may be.” Therefore, the NRC cannot issue its license without a Section 401 Certification or an NRC determination that a waiver has occurred, in accordance with 40 CFR 121.9(c) (TN6178).

Federal and State permits/certifications for activities affecting waterways and wetlands are reviewed by the USACE in consultation with the FWS, the Soil Conservation Service, the EPA, and the delegated State agency.


Question: remote shutdown systems in nuclear plants

Remote Shutdown of Ventilation for Loss of Auxiliary Air 078A4.030 (3.2)

Alignment Upper Containment Monitor to Lower Containment per 0-SOI-90.02


Question: What is the aging mechanism of stainless steel heat exchanger tubes in nuclear applications?

Reduction of heat transfer due to fouling.

Not applicable. This item is only applicable to BWRs.


Question: Licensees must submit a plan detailing decommissioning activities and describing methods to safeguard against radiation hazards during the decommissioning process.

Paragraph 70.38(g)(1) requires that licensees submit a decommissioning plan if required by license condition or the decommissioning procedures and activities have not previously been approved by the NRC and the procedures could increase potential health and safety impacts to workers or the public. Submission of the decommissioning plan is necessary for the NRC to evaluate any health and safety impacts.

When this notification is made, the licensee shall either begin decommissioning or submit within 12 months of notification a decommissioning plan if such a plan is required. This notification informs the Commission that the status of licensed activities has changed and establishes the start date for decommissioning activities that must be completed within specified time periods.


Question: weld inspections for NuScale reactors

Welded connections located between the CNV and CIVs receive 100 percent volumetric inspections on a frequency of once per ISI interval (10 years). The volumetric inspection meets the inspection requirements of BTP 3-4.

The ASME Class 1 assembly configuration connecting the CNV and the CIVs includes a safe-end, CITF, and the CIVs.


Question: What is the significance of the Background Reference Area in regulatory filings?

The Background Reference Area (Boring 16) is a designated site for collecting environmental data relevant to assessing impacts from facility operations, often presented in regulatory filings to demonstrate compliance with safety standards.

Location details for Boring 16 indicate the site was identified as Pickard Hall, with depth measurements ranging from 0-5 ft and 5-9 ft bgs collected in August 2020.


Question: The analysis of heat generation and transfer during postulated accidents is crucial for evaluating containment performance.

The reactor cavity analysis evaluates the effects of various break scenarios on the containment structure's integrity and performance during high-stress conditions.

The detailed nodalization of the steam generator enclosure is essential for understanding the thermal-hydraulic behavior within the primary loop.