Question: Class 2 fiber specifications

The prepared fiber consisted of predominantly Class 2 fibers as defined in NUREG/CR-6224 (Reference 9, Table B-3) including mostly individual fibers with lesser quantities of fiber shards and small entanglements.

The fiber was then wetted with the pre-heated test water until the base material was saturated and the pressure washer nozzle assembly was submerged.


Question: The NRC Project Manager will manage any modifications to the timeline of the audit with the licensee.

The NRC staff access to the portal may be terminated after completion of the audit at a time agreed between the audit team and Duke Energy.

An audit summary, which will be publicly available, will be prepared within 30 days of the completion of the audit.


Question: nuclide activity in radiation analysis

?  = nuclide is part of an undetermined solution


Question: What are Limiting Conditions for Operation (LCO) in nuclear regulatory context?

LCOs shall be met during specified conditions in the Applicability, except as provided in LCO 3.0.2.

Upon failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5.


Question: What are the regulatory requirements for identifying historical and cultural resources in nuclear project filings?

the NRC’s regulations requiring such identification in the Combined TR/ER to comply with the Commission’s NEPA-implementing regulations.

That omission serves as a basis for the admission of this aspect of Contention 1.


Question: The EAS broadcast continues until the State EOC either instructs to cease or sends a revised EAS message for transmission.

The EAS message is broadcast and repeated at regular intervals until informed by the State EOC to stop or transmits new EAS message to be aired.

RBS will inform all Parishes that in approx. 5 minutes RBS & LDEQ will call the Parish Directors to discuss the PARs.


Question: The NRC emphasizes the importance of understanding changes in materials due to radiation exposure.

Structures exposed to high radiation experience a change in properties due to irradiation effects.

Validate assumptions of aging effects at large scales during the testing of large structural sections.


Question: The NRC's review includes environmental assessments to ensure off-site doses remain unaffected by site boundary changes.

The 10 CFR 50.59 review assessed the impact of the change in the CR3 Controlled Area boundary on off-site dose calculations and effluent releases and concluded that the change:

Duke Energy has evaluated the environmental impacts associated with the proposed release of the subject property and considered those impacts in light of the original FES.


Question: The NRC mandates that licensees develop models to assess strainer fiber penetration under specific plant conditions.

Data gathered from the fiber penetration test was used to develop two curve-fits for quantifying the strainer fiber penetration under plant conditions: a low-fiber curve and a high-fiber curve.

The test water was maintained at approximately 120°F for the duration of the test.


Question: What is the procedure for gamma spectrum analysis in nuclear facilities?


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX heat flux correlation is specifically designed to enhance the predictive capability of critical heat flux for a wide range of operating conditions and geometries, thereby improving safety margins in reactor operations.

The table B-3 outlines local conditions from various COBRA-FLX simulations including parameters such as pressure, quality, and flow; however, it does not discuss the implications of these conditions on safety protocols.


Question: What does gross activity measure in nuclear safety?

Figure 16-3 presents a histogram for gross activity, illustrating the frequency of measurements taken at various upper end values in disintegrations per minute per 100 cubic centimeters (DPM/100CM³).

The table outlines different frequencies corresponding to specific upper end values, but it does not discuss any regulatory implications or safety protocols.


Question: Strontium-90 analysis in soil samples

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


Question: The gamma spectrum does not accurately reflect the source due to calibration issues that were noted in earlier analysis sessions.

**Energy Calibration Used Done On**: 7/8/2014

- **Sample Identification**: L1-SUB-TDS-FSGS-B19C-SB
- **Sample Size**: 9.267E+02 grams
- **Facility**: Dairyland_NPP


Question: What is the significance of the 100-year return period temperature in nuclear safety assessments?

The 100-year return period maximum temperature is not readily available in published documents. It may be possible to calculate a 100-year return maximum temperature using statistical methods and long-term data records; however, given that the temperature provided above was obtained from a period of record that exceeds 100 years (i.e., 1896 to 2000) it is therefore proposed as a 100-year return temperature.

The maximum ambient dry bulb temperature (0 percent exceedance): 117°F (NCDC, 2004)


Question: What are the requirements for qualification documentation in nuclear safety?

Qualification documentation is required to verify that the SIS is qualified to perform its intended functions within the environmental constraints specified.

The RIM Program includes two expert panels: the RIM expert panel (RIMEP) and the monitoring and non-destructive examination expert panel (MANDEEP).


Question: impact assessment of noise during decommissioning

Only the incremental increases in the impacts of termination of plant operations and decommissioning attributable to continued operation during the proposed SLR operating term is within the scope of this issue. No additional impacts related to noise are expected.

**SMALL to MODERATE (construction):** Given the acreage of the solar installations, the potential need for land clearing, and the number of turbines that would need to be installed, noise impacts would range from SMALL to MODERATE and be temporary during construction of each facility.


Question: SAMA maximum benefits

Summary of Aggregate SAMA Maximum Benefits (Sheet 1 of 6)

The reactor experienced an unexpected shutdown due to an electrical fault.


Question: What is the purpose of the RCIC System instrumentation?

The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink and normal coolant makeup flow from the feedwater system is unavailable, such that initiation of the low pressure ECCS pumps does not occur.

The Diverse RCIC system instrumentation inputs include various parameters such as condensate storage tank level and suppression pool water level.


Question: The fixed water sprinkler systems are essential for fire suppression within the nuclear facility, ensuring safety during operations.

The fixed water sprinkler systems are designed to extinguish fires should they occur. Either the normal or SSS shutdown capability would be available.


Question: The Pointe Coupee Parish Emergency Response Procedure includes a list of facilities equipped with special notification devices.

The facilities with tone alert monitors are identified in the Pointe Coupee Parish Emergency Response Procedure for Dispatcher/Communicator.

The overall efficiency of response teams is crucial for minimizing the risks associated with nuclear facilities.


Question: The NRC emphasizes the need for effective groundwater restoration techniques in in-situ recovery operations.

The restoration of ground water at the CUP ISR sites will have the benefit of a previously engineered array of injection and production wells that were initially installed in a configuration to maximize sweep efficiently throughout the uranium ore body, and maximize uranium recovery.

The primary benefit of R.O. treatment will be that a large fraction of the total water extracted will be purified and reinjected resulting in less water consumption and less ground water drawdown in the area.


Question: The NRC's regulations include requirements for performing gamma spectrum analysis as part of safety assessments.

Energy Calibration Used Done On: 7/8/2014

Peak Analysis Performed on: 3/22/2019 6:41:39 AM


Question: The NRC emphasizes the importance of monitoring external surfaces for material loss in mechanical components.

Aging Effect Requiring Management: Loss of Material

Steel piping and components are commonly utilized within nuclear facilities for safely transporting fluids.


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

The safety analysis report is a comprehensive document that evaluates the safety of a nuclear facility, analyzing potential hazards and their impacts to ensure compliance with regulatory requirements.

The reactor pressure vessel must be designed to withstand high temperatures and pressures during normal operation and any postulated accident scenarios.


Question: In assessing the performance of emergency core cooling systems after an accident, it is crucial to ensure that debris does not hinder the effectiveness of these systems.

The licensing basis with regard to effects of debris is determination of a high probability that the ECCS and CSS can perform their design basis functions based on WCGS-specific testing using an NRC-approved methodology.

The risk-informed method used to analyze the impact on long-term core cooling due to LOCA-generated debris is the threshold break approach.


Question: The application of NEPA’s principles necessitates that environmental impact assessments consider the effects on cultural resources, which may include Native American sites that are present but not comprehensively documented.

In promulgating 40 C.F.R. § 1502.22, CEQ explained that the new regulation "requires that analysis of impacts in the face of unavailable information be grounded in the ‘rule of reason.’"

This notice provision does not alter the substantive rights of the signatories to the Programmatic Agreement or of the Tribe.


Question: What is the interaction ratio of strainer components in nuclear safety systems?

As documented below, the interaction ratio of each subcomponent is less than 1, and therefore is acceptable.

The external radial stiffener is designed to provide support under seismic events and has been tested for compliance with safety regulations.


Question: What are the average background radiation levels measured in TLD readings?

| B-9214   | TLD Damaged – not read         |                      |              |                        |          |           |


Question: What is the time required to clear an affected population area at a nuclear site?

Time to Clear Indicated Area of 90 Percent of the Affected Population is a critical metric for emergency response scenarios.

The seasonal impact of weather plays a significant role in determining the overall safety protocols for nuclear facilities.


Question: DSTP values in nuclear safety assessment

The DSTP-100 value is shown to be 0.70259, while DSTP-200, DSTP-300, and onward reflect a systematic increase in measurement significance within the context of nuclear safety assessments.

In order to ensure safety during the operation of the reactor, all staff must undergo training on emergency protocols and procedures for responding to incidents.


Question: The impact of additional debris loads on emergency core cooling effectiveness needs to be analyzed.

Assurance of ECCS strainer performance at Fermi 2 is based on deterministic calculations of debris generation and transport following guidance provided in Regulatory Guide (RG) 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident".

The proposed amendment changes the licensing basis described in the Fermi UFSAR to allow the use of a risk-informed methodology to address potential debris quantities arising from component identification tags/labels and from Min-K® insulation applied in containment penetrations exceeding those currently evaluated in the deterministic licensing methodology.


Question: pressure boundary in piping systems

Piping, piping components are designed as a Pressure Boundary.

The pressure boundary is crucial for preventing leaks in the reactor containment structure.


Question: Project Specific Application Process for nuclear licensing

Development of the Application ED by Curtiss-Wright conforms to the requirements specified in the TR, specifically:

The reactor coolant system is designed to operate at high pressure to ensure that coolant is always present in the reactor core.


Question: What is the definition of specified Frequency in nuclear regulatory context?

Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated Limiting Condition for Operation (LCO).

The NRC recognizes the importance of maintaining a thorough understanding of operational safety protocols to ensure the safe use of nuclear technology.


Question: The average enrichment of uranium in each assembly is critical for determining the burnup rates.

| Initial Assembly Avg. Enrichment wt % 235U (E) | 40 < Assembly Average Burnup ≤ 41 GWd/MTU Minimum Cooling Time (years) | | | | | | | | |


Question: What indicators are available for monitoring hydrogen levels in the control room?

Indication of Hydrogen concentration on PAMS 1, only.

There are indications of containment parameters on VB1, pressure, temperature, but not hydrogen concentration.


Question: nuclear safety system

The safety-related system implemented in nuclear facilities is designed to manage and mitigate potential hazards during various operational states, including normal operation and emergencies.

The logged reading for the survey unit B2103 on 10/16/2019 indicates a count time of 60 seconds and an alarm-1 setting of 1.


Question: volatile iodine partition factor

MC/AE volatile iodine partition factor 0.15

The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.


Question: Loss of Emergency Coolant Recirculation

Per BD-CA-1.1, "Subsequent procedure steps attempt to delay depletion of the RWST volume and thus extend the time the pumps can use the RWST as their suction source."

If the break has NOT been isolated, which of the following identifies the effect that a transition to 1BCA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, has on mitigating the accident?


Question: What changes were approved for the Watts Bar Unit 1 TS?

In support of the replacement of the Watts Bar Unit 1 SGs, similar changes to WBN Unit 1 TS LCO 3.4.7.b and SRs 3.4.5.2, 3.4.6.3, and 3.4.7.2 were approved in Amendment 61 (ML060960075).

Amendment 61 was critical in defining the operational parameters for various safety systems, ensuring compliance with the latest safety regulations.


Question: What are the implications of unresolved technical issues in nuclear licensing?

Therefore, Framatome is requesting that the ACRS Subcommittee meeting be pushed out. We believe that we should agree upon a resolution path for all technical issues that result from the Framatome review of the draft SE as well as those already identified from the late stage L&C’s before rescheduling the ACRS Subcommittee meeting.

As a result, we believe we will need more than the requested 20 Days to perform the review of the Draft SE.


Question: What are the design standards for transporting spent fuel?

NRC establishes regulations for: Package design standards for transportation of spent fuel (10 CFR Part 71)

NRC also inspects and oversees certificate holders, package fabricators, and licensee shippers.


Question: The assessment considered the transient population within the 10-Mile Emergency Planning Zone (EPZ).

The transient population within approximately 10 miles of the site is evaluated and discussed in detail as part of the assessment of the evacuation time estimate (ETE) of the 10-Mile Emergency Planning Zone (EPZ).

The GGNS Unit 1 facility has approximately 750 people who work at the plant site.


Question: What is the significance of access to Restricted Data in nuclear regulation?

Section 145 of the Atomic Energy Act (AEA) of 1954, as amended, requires, before an individual can be granted access to Restricted Data (RD), an investigation and report on the individual’s character, associations and loyalty and a determination that permitting such a person access to RD will not endanger the common defense and security.

The NRC Form 354 will ONLY be requested for individuals requiring Top Secret clearances.


Question: What are the model extents in HEC-RAS for hydraulic studies?

Figure 7-1 Upper HEC-RAS Model Extents shows the defined boundaries for the upper section of the hydraulic model, outlining the area considered for the analysis.

The routine maintenance of the reactor cooling system ensures optimal performance during operation periods.


Question: nuclear licensing inspection results

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

Inspectors reviewed World Association of Nuclear Operators (WANO) report, dated January 2023, during this inspection period.


Question: The determination process for ensuring the integrity of containment structures is critical in maintaining safety standards at nuclear facilities.

The Containment Integrity Significance Determination Process provides a method for evaluating findings related to the integrity of containment structures and their importance in ensuring safety in nuclear operations.

Safety inspections at nuclear reactors are conducted regularly to assess the operational readiness and compliance with established safety standards, ensuring that all systems are functioning as intended.


Question: uncertainty in radiation measurement

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

The Qualifiers in this report are defined as follows: Analyte is a Tracer compound, Result is less than value reported, Result is greater than value reported, BD Results are either below the MDC or tracer recovery is low, FA Failed analysis.


Question: What is a channel functional test in nuclear safety?

A CHANNEL FUNCTIONAL TEST shall be: a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY.

The trip output signals (Neutron Flux Upscale/Simulated Thermal Power - Upscale/APRM Inoperative/OPRM Upscale) from all four APRM channels go to each of the four APRM 2-out-of-4 voters.


Question: The analysis of public health impacts reveals minor effects on surrounding populations.

As indicated in Section 4.13.6 of the CUP FEIS, the CUP will make a minor contribution to cumulative impacts in terms of public and occupational health and safety impacts.

Liquid and solid wastes are divided into two general categories: AEA-regulated waste and non-AEA-regulated waste.


Question: What types of benthic invertebrates are found near nuclear sites?

Benthic invertebrates in the vicinity of the DNPS site include species such as Aquatic Worms, Leeches, Amphipods, and Mayflies.

The NRC requires that licensees adhere to stringent reporting requirements to ensure compliance with safety regulations.


Question: How should the reactor trip breaker indicating lights be interpreted in the context of plant safety?

The reactor trip breaker indicating lights provide crucial information on the status of the reactor trip system during an Anticipated Transient Without Scram (ATWS).

The plant conditions must be assessed in accordance with EP-AA-111 to ensure compliance with safety protocols.


Question: Changes regarding the modification of the main feedwater motor-operated valves are necessary for abnormal situation handling.

Modify corrects inconsistencies on the discussion of automatic closure of the main feedwater MOVs for abnormal events.

Update reflects modification to the condenser tube cleaning (Amertap) systems.


Question: The average enrichment of uranium in a fuel assembly must comply with specific regulations.

| Initial Assembly Avg. Enrichment wt % 35U (E) | 44 < Assembly Average Burnup < 45 GWd/MTU | Minimum Cooling Time (years) |         |         |         |         |         |         |

Reactor design is influenced by factors such as coolant type and fuel management strategies.


Question: nuclear regulatory filings

Nuclear regulatory filings must comply with the standards set forth in the Code of Federal Regulations, which outline the requirements for the licensing process.

The reactor coolant system is designed to maintain a safe and efficient operation, ensuring that heat transfer occurs effectively during normal and emergency conditions.


Question: What are the existing standards for magnetic confinement fusion energy devices?

Existing nuclear codes and standards for construction do not adequately cover the design, manufacturing or construction of the magnetic confinement fusion energy devices (e.g. Tokamak devices) that are currently being considered for future DEMO constructions.

They also do not provide support for the on-going projects, such as ITER and others.


Question: The confidence level indicates the reliability of each nuclide's activity measurements.

The Wt mean Activity (pCi/grams) represents the weighted mean of activity concentrations for the respective nuclides.

Errors quoted at 2.000egrams indicate the level of uncertainty associated with the measurements.


Question: What is the process for obtaining a nuclear plant operating license?

An operating license is issued by the NRC after considering safety, environmental protection, and security aspects of the nuclear plant.

The reactor's cooling system is designed to manage heat generation during normal operations and accidents.


Question: nuclide identification

Nuclear data tables typically include detailed information about various nuclides, including K-40, PB-212, and BI-214, which are critical for identifying their properties such as energy, yield, and activity.

In addition to the radioactive decay data, discussions on the environmental impact of nuclear facilities often arise, emphasizing the need for stringent safety regulations.


Question: What are the steps in the Emergency Alert System (EAS) message flow?

The EAS message is broadcast and repeated at regular intervals until informed by the State EOC to stop or transmits new EAS message to be aired.

As a backup, the State EOC has a dedicated phone line and FAX to the WFMF-FM Control Room.


Question: Are any materials received that require a NRC license for handling?

Are any samples NRC licensable material?

Was ice present in the shipment container(s)?


Question: design review guide for nuclear instrumentation

The DRG is used as a guide for development of Natrium I&C architecture and this ToR


Question: What is the significance of the Wenonah Formation in geological studies?

The Wenonah Formation is of Upper Cretaceous age and consists of marine sediments deposited in a neritic environment during a marine regression.

The Mount Laurel Formation consists of Upper Cretaceous age marine sediments deposited in a near-shore environment during a marine regression.


Question: What are the environmental impacts of nuclear facilities during decommissioning?

The Environmental Impacts Comparison Summary provides a detailed analysis of various impact areas associated with the proposed action compared to the termination of operations and decommissioning of nuclear facilities.

The report outlines the thermal efficiency of natural gas combined cycle (NGCC) power plants, which can influence operational costs and the decision-making process for future energy projects.


Question: The NRC mandates that operators maintain a specific steam generator (S/G) level to ensure proper functionality during emergencies.

- MAINTAIN S/G press (using steam dump or PORV) low enough to ensure condensate flow is maintained.

The monitoring of the Reactor Coolant System (RCS) pressure is critical to ensure safety during operations.


Question: The NRC assesses possible water supply limitations due to less than average rainfall in specific areas, which affects cooling operations at the plant.

A shortage of surface water supplies due to a period of below-average precipitation in a given region, thereby depleting the water supply needed for the various water-cooling functions at the facility.

Strong winds resulting in dynamic loading or missile impacts on SSCs causing functional failure. Hazards that could potentially result in high wind include the following:


Question: gamma measurement in nuclear safety

Gamma Measurement is the measurement of gamma radiation levels, expressed in counts per minute (cpm), taken at specified depth intervals.

The sampling location was Pickard Hall and the boring number designated for this collection is B1(A-1)6.


Question: What is the composition of a fully reflected damaged package array?

Fully reflected damaged 9 × 9 × 2 package array: 60 kgs UO₂ + H₂O mixture, maximum burn (horizontal and vertical views)

The temperature readings from the reactor core indicate a stable operation within the safe limits set by the regulatory framework.


Question: nuclear reactor licensing requirements

The licensing process for nuclear reactors involves a comprehensive review of safety, environmental, and operational factors, overseen by the Nuclear Regulatory Commission (NRC).

In contrast, the framework of our environmental assessment will consider climate change impacts and local biodiversity without direct reference to nuclear operations.


Question: Regulatory filings require detailed assessments of potential impacts on surrounding communities.

The NRC requires licensees to analyze emergency zones to ensure that public health and safety are protected in the event of an incident.

In 2037, the measurements indicate that no incidents have occurred within the analyzed sectors, ensuring a steady assessment of the area.


Question: The objective of this section is to obtain data needed to relate the differential pressure at the reactor coolant elbow tap with the reactor coolant system flowrate.

To establish the optimum continuous spray flowrate.

The spray bypass valves are throttled so that the minimum flow necessary to keep the spray line warm is achieved.


Question: PWR fuel loading patterns

Table B2.2 lists the PWR fuel loading patterns and their corresponding storage locations and heat loads.

The nuclear facility conducted a thorough assessment of its emergency response procedures to ensure public safety during potential incidents.


Question: The analysis of seismic hazard curves is crucial for understanding the potential ground motions that a nuclear facility may experience.

Mean and Fractile Rock Seismic Hazard Curves for 1 HZ Spectral Acceleration.

Comparison of Input and Modeled Velocity Profiles.


Question: What is the role of the REFUEL BRIDGE INTERLOCK BYPASS SWITCH in nuclear reactor operations?

With the REFUEL BRIDGE INTERLOCK BYPASS SWITCH in NORM, the control rod withdrawal blocks generated from the refuel bridge position are active.

Electrical drawing 4E-1411 shows with the Mode Switch in RUN and the refuel bridge moved over the core, Rod Out Block #2 is received.


Question: The system must analyze each blowdown sample to identify the steam generator causing the high radiation levels.

The Steam Generator Blowdown System is designed to operate manually and on a continuous basis as required to maintain acceptable steam generator secondary side water chemistry. The System serves no safety function.

The ability of the Steam Generator Blowdown System to maintain water chemistry within specifications is verified by the following design functions.


Question: What are the conditional tephra-fall hazards associated with volcanic eruptions?

The conditional silicic tephra-fall hazards and associated eruption characteristics for these volcanic fields are summarized in Table 50.

The average silicic tephra-fall deposit of 84.1 cm has a <2.5x10^6/yr probability of exceedance.


Question: The average enrichment of uranium-235 in a boiling water reactor assembly is crucial for maintaining optimal reactor performance during operation.

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

This table lists various enrichment levels for different BWR reactor types, displaying the parameters for fuel assembly configurations.


Question: boron endpoint in nuclear licensing

Boron endpoint is 1322 ppm, with a tolerance of ±24 ppm.

D bank in has a measured value of -4.29 pcm/°F with a tolerance of ±2 pcm/°F.


Question: The NRC mandates that environmental monitoring data be analyzed for various radionuclides...

The following Analytical Methods were performed include EPA 905.0 Modified and EPA 906.0 Modified for assessing sample results.


Question: What are the safety margins used in nuclear reactor licensing?

The safety analysis performed during the licensing process includes evaluations of safety margins that ensure the reactor can safely operate under both normal and abnormal conditions.

The temperature readings from the sensors in the reactor core are critical for continuous monitoring and control of the reactor's thermal performance.


Question: What is the significance of weight mean activity in nuclear licensing?

The weight mean activity (Wt mean Activity) of a nuclide is a crucial parameter used in radiation safety assessments, indicating the average activity of the nuclide per unit mass, typically expressed in microcuries per gram (uCi/grams).

The confidence level for each nuclide is determined through rigorous testing and compliance with established guidelines to ensure the validity of the licenses issued.


Question: emergency operations center

D. Activate the EOC and verify security post established.

A. Obtain the emergency status and report to the EOC.


Question: What does a nitrate acceptable concentration level indicate in regulatory terms?

Location not sampled during the July 2024 event due to insufficient water (recharge).


Question: The complexities involved in the decommissioning process must be effectively communicated to the NRC.

The issues surrounding the decommissioning of the Palisades Nuclear Plant,

Joint Petitioners strongly believe that direct, in-person interaction with the Board and the NRC Staff is of paramount importance to properly address and underscore the gravity of the issues raised in their contentions.


Question: license renewal environmental reporting

Section 51.60(a) of the Commission’s regulations notes that licensees seeking a license renewal (or license amendment to authorize an expanded area of production, as in the Marsland proceeding) are permitted to submit an environmental report 'supplement' that 'may be limited to incorporating by reference, updating or supplementing the information previously submitted to reflect any significant environmental change … .'

At oral argument, but not in their Answers, Powertech and the Staff argued that, as this is a license renewal proceeding, the LRA was not required to include an updated cultural and historic resources section.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves submitting detailed safety analysis reports to the Nuclear Regulatory Commission for review and approval.

The financial statements of the company must be audited at the end of each fiscal year to ensure compliance with accounting standards.


Question: What is the purpose of the TRPS in nuclear safety?

The TRPS, or target solution vessel reactivity protection system, is designed to ensure that the reactor maintains safe operational parameters by controlling reactivity and preventing unintended excursions.

The scope of this audit addresses updates to the SHINE FSAR and associated responses to RAIs on the highly integrated protection system (HIPS), engineered safety feature actuation system (ESFAS) and other systems.


Question: emergency response plans in nuclear regulation

§ 53.855 would require an emergency response plan that demonstrates compliance with the requirements in appendix E to part 50 and § 50.47(b) or § 50.160 but “emergency plans and emergency planning zones are not safety features in the design.”

§ 50.54(dd) stating that a holder of an OL or COL may take reasonable actions that depart from the license in a national security emergency, apparently without any guidance, restrictions or restraints.


Question: Pressurizer pressure channel failure

After the crew enters Mode 1, PZR Pressure Channel 1 will fail high. The crew will respond by placing Master Pressure Controller in Manual to raise PZR pressure to program.

The crew assumes the watch at 4% power, BOL. Control Rods in manual, 21 SGFP in service, 22 SW Pump C/T for strainer motor replacement. Power ascension to 10% in progress IAW S2.OP-IO.ZZ-0003 Hot Standby to Minimum Load.


Question: The regulatory requirements for licensing nuclear reactors necessitate detailed safety analyses and documentation.

The regulation of nuclear reactors includes various processes that ensure safety, security, and environmental protection through oversight by federal and state authorities.

The study focuses on the mechanical properties of laser-deposited materials and their behavior under stress.


Question: chemical effects testing in nuclear plants

Wolf Creek is using the separate chemical effects approach to determine the chemical source term. Alden Research Laboratory, Inc. performed the testing in their test lab in Holden, MA.

The containment sump pool was assumed to be well mixed.


Question: What are unit hydrographs used for in plant site drainage analysis?

The unit hydrographs are used to analyze the drainage basins of the plant site to predict the runoff and timing of water flow.

The analysis includes rainfall rates and other hydrological studies relevant to the site planning.


Question: What is the purpose of a 10-Mile Emergency Planning Zone (EPZ) in nuclear safety assessments?

The transient population within approximately 10 miles of the site is evaluated and discussed in detail as part of the assessment of the evacuation time estimate (ETE) of the 10-Mile Emergency Planning Zone (EPZ).

There is limited commercial fishing within the study area. Most of this occurs on the Mississippi River, the Big Black River and the Bayou Pierre River, with catfish being the most abundant catch.


Question: The NRC mandates that licensees maintain compliance with established safety standards in their operations related to hazardous materials.

Sodium, dissolved content is measured in mg/L as defined by EPA 200.7, ensuring relevant analysis for regulatory compliance.

ACZ Laboratories, Inc. oversees a variety of quality control processes to ensure accurate sample testing and reporting.


Question: The document discusses the circumstances under which an alternate shutdown system might be considered unnecessary.

The use of an alternate shutdown system is deemed unnecessary under existing regulatory guidelines, as current safety protocols adequately address potential failure modes.

Inspecting the integrity of containment structures forms a foundational aspect of maintaining reactor safety.


Question: What is the significance of the rod position sequence in nuclear reactor operations?

The proposed amendment adds a notation to TS 3.1.6 Limiting Condition of Operation (LCO), Condition A, Condition B, and Surveillance Requirement (SR) 3.1.6.1 to allow for operable control rods to comply with the analyzed rod position sequence in lieu of banked position withdrawal sequence (BPWS) through the remainder of Unit 2, Cycle 21.

The proposed amendment modifies TS 3.2.1 Required Actions C.2.2 and D.1 with a footnote to allow for compliance with the analyzed rod position sequence in lieu of only BPWS through the remainder of Unit 2, Cycle 21.


Question: The Critical Heat Flux (CHF) correlation is instrumental in predicting the thermal limits of reactor fuel assemblies.

COBRA-FLX is a code developed to predict Critical Heat Flux (CHF) in heated channels with various geometries, which is essential for ensuring the safety of nuclear reactors.

Framatome Inc. provides a range of services including engineering, design, and manufacturing for the nuclear industry, ensuring compliance with safety regulations and technical standards.


Question: no significant hazards consideration

The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the license amendment request involves a no significant hazards consideration.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice.


Question: nuclear licensing process

The nuclear licensing process involves multiple stages, including pre-application consultations, the submission of a license application, and a thorough review by regulatory authorities before a license can be granted.

The primary objective of reactor design is to ensure efficiency in power generation while maintaining safety standards and environmental protection.


Question: diesel generator loading calculations

The increase in DG loading due to DG voltage and frequency variation is used as input into the existing diesel fuel consumption and storage tank capacity calculation to verify that the increase in fuel oil consumption remains bounded.

Some of these components are fans, chillers, air conditioner units (ACUs) and coolers; the increase in speed will result in an improved performance from the components.


Question: What actions must be taken if more than one block valve is inoperable?

More than one block valve inoperable. Required Actions do not apply when block valve is inoperable solely as a result of complying with Required Actions B.2 or E.3.

Completion Time for Condition G is listed as Immediately, 6 hours, or 12 hours.


Question: Preventive maintenance testing on DC breakers

A finding of very low safety significance (Green) and an associated NCV of TS 5.4.1, “Procedures,” was NRC identified, for the failure to implement component replacement or trip testing as specified in Preventive Maintenance Change Request (PMCR) 83591.

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensee’s performance by conducting an integrated inspection at Braidwood Station, in accordance with the Reactor Oversight Process.


Question: Strontium-90 analysis in nuclear licensing

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


Question: k_eff calculation in nuclear reactor safety analysis

The 55 Kgs Heterogeneous Damaged Array Case with Overlap includes various configurations of reactor rods and their respective performance metrics.


Question: plasticity index in soil samples

The plasticity index is a measure of the plasticity of a soil, defined as the difference between the liquid limit and the plastic limit, and provides insight into the behavior of the soil under varying moisture conditions.

The triaxial CU test measures the strength of a soil sample when subjected to different pressures, which is critical for understanding its stability under load.


Question: The assessment of dam breaches is crucial for understanding flood risks at the site.

Dam breaches can lead to significant flood hazards that need to be carefully evaluated and mitigated during nuclear facility operations.

The Cooper Nuclear Station utilizes advanced cooling systems to ensure efficient reactor operation and maintain safety under various conditions.


Question: What is core damage frequency (CDF) in nuclear safety assessments?

CDF = core damage frequency; IPE = individual plant examination; LR GEIS = Generic Environmental Impact Statement for License Renewal of Nuclear Plants; SAMA = severe accident mitigation alternative; N/A = not applicable.

From Oconee Station’s 1998 ER, the annual person-rem risk result calculated for the 50 mi population was 5 whole person-rem.


Question: The assessment of causal factors and the effectiveness of corrective actions are critical for addressing site issues.

The causal evaluation report(s) and associated documents for the identified issues.

Please assemble the following documents and provide them to the inspector on or before the dates requested.


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

Considers, evaluates, and discloses environmental impacts of continued plant operation and alternatives to license renewal.

Staff review is to determine if the environmental impacts of license renewal are so great that license renewal would be unreasonable.


Question: What is the significance of RTMAX in reactor vessel assessments?

The EOI RTMAX values for all BFN reactor vessel plates and axial welds meet the acceptability criteria for limiting plates and axial welds in BWRVIP-329-A.

Integrated Surveillance Program (ISP) data have been included directly in the evaluation as applicable to the specific vessel per ISP guidance.


Question: What are the results of the groundwater analysis in the Rio Algom Mining Company report?

The results for groundwater analysis included Calcium, dissolved at 414 mg/L, Magnesium, dissolved at 2840 mg/L, and Sodium, dissolved at 353 mg/L.

The Project ID for this report is ACZ Sample ID: L89108-07 and the Date Sampled was 07/18/24 14:40.


Question: What is the significance of radioiodine exposure from dairy products in nuclear regulations?

A potentially significant exposure pathway involves the transfer of released radioiodines to milk, which is then ingested by a young child.

The total integrated annual population exposure to all individuals within 50 miles, by all pathways, is estimated to amount 12.7 man-rem in the year 2000.


Question: open cycle operation

In this mode, water is withdrawn from and discharged directly to the Mississippi River. Open cycle operation is used when river flow exceeds 860 cfs and cooling of the circulating water effluent is not required to keep the discharge canal temperature below permitted limits.

This permit is currently administratively extended. An application for renewal was submitted and received March 29, 2012.


Question: reactor heat removal system

The Residual Heat Removal (RHR) system is designed to remove decay heat from the reactor core and ensure safe conditions within the containment structure during normal and emergency scenarios.

The auxiliary air compressors are critical for supplying compressed air to operate various pneumatic components throughout the plant, ensuring that operational parameters are maintained.


Question: radiological survey procedures

A radiological survey is conducted to assess the radiological conditions of a site or facility, ensuring that all levels of radiation are within acceptable limits as per regulatory standards.

The reviewed document includes data from multiple instruments, but does not identify any particular issues or concerns regarding their operational status.


Question: core damage frequency

Aggregate CDF, H/E, H/I, M/E Reduction: For those SAMAs that were not “pre-screened” from further analysis, these fields provide the percent by which each SAMA reduced the overall plant core damage frequency (CDF), High-Early release frequency (H/E), High-Intermediate release frequency (H/I), and Medium-Early release frequency (M/E).

None of the 33 SAMAs for which quantifications were performed reduced the MB for DNPS by 50 percent or more.


Question: The NRC mandates licensees to regularly inspect fuel oil system components to manage aging effects effectively.

These activities effectively manage the effects of aging by identifying age-related degradation and maintaining potentially harmful contaminants at low concentrations.

Ultrasonic thickness measurements of the Diesel Generator Day Tanks bottom surfaces will not be obtained because the Diesel Generator Day Tanks bottoms are not accessible to perform ultrasonic thickness measurements of the tank bottoms.


Question: What is the significance of MV cables in nuclear safety assessments?

MV cables: ETE-SLR-2023-3345; I&C cables: ETE-SLR-2023-3350; LV cables: ETE-SLR-2023-3351; Section 3.5.2 is for Monitoring and Trending.

It appears to be a typo in this ETE-SLR section with '+-driven'. Should this be 'event-driven'? Also, confirm what triggers the initiation of inspections for water accumulation after event driven occurrences, what parameter(s) drive this initiation of the inspections?


Question: Containment Atmospheric Control System components

The Containment Atmospheric Control System is designed to manage the atmospheric conditions within the containment structure to ensure safe operation and prevent the release of radioactive materials during normal and accident conditions.

The Table 2.3.2-1 summarizes various components and their intended functions within the system, detailing aspects like pressure boundaries and structural integrity.


Question: systematic measurement locations

Figure 16-2 – Survey Unit B2-010-103 Interior First Floor Systematic Measurement Locations Map

Administration Building - 1st Floor Floor Systematic Sample Locations


Question: The NRC mandates regulatory compliance concerning the materials used in sample analysis.

Peak Area Range (in channels): 85 - 8000


Question: The NRC's guidelines mandate that licensees perform biennial evaluations concerning problem identification and resolution procedures.

Enclosure:  Request for Information for Biennial Problem Identification and Resolution Inspection

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


Question: probabilistic risk assessment in nuclear plants

A PRA of each commercial nuclear plant must SHOULD be performed to HELP identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the PRIMARY NUCLEAR safety functions identified in § 53.230.

To identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment.


Question: Strontium isotopes are measured for their activity in soil samples as part of the safety assessment.

The following Analytical Methods were performed:


Question: The NRC has established guidelines to ensure the safety and efficiency of extraction operations in designated project areas.

NuFuels will operate a satellite facility and the CCP within the Crownpoint project area.

The project area encompasses 287 ha [708.8 ac] in Sections 19 and 20 of T17N, R12W and Sections 24 and 25 of T17N, R13W.


Question: What is the significance of nuclide identification in nuclear safety?

The identified nuclides are crucial for assessing the radiological safety and ensuring compliance with regulatory standards.

The Energy Tolerance is set at 1.000 keV which ensures the reliability of measurements.


Question: feedwater flow operation

Critical Task 2: Establish Feedwater flow to SG(s) before RCS bleed and feed is required. RCS bleed and feed is required when any three SG WR levels are less than or equal to 17%.

CHECK secondary heat sink restored: - a. NR level in at least one SG greater than 29% (39% ADV). - b. IF feed flow established to at least one SG: - S/G Wide Range level rising, OR - Incore T/C dropping.


Question: COBRA-FLX Critical Heat Flux Correlation

COBRA-FLX is a thermal-hydraulic computer code developed to model the behavior of nuclear reactors during transient conditions, including the prediction of critical heat flux, which is vital for ensuring safety in reactor operations.

The data presented in Table B-3 outlines various local conditions derived from COBRA-FLX simulations for grid tests and does not pertain to heat flux modeling or safety analysis in the reactor context.


Question: The maximum water surface elevation due to the PMH event is determined from various model simulations that account for changes in antecedent water levels.

The PMH maximum WSEL of 32.1 ft. NAVD using the high-resolution modeling system described in Subsection 2.4.5.6.1 represents the design basis flood for the new plant location.

Tidal current velocities normally range from 2 to 3 ft/sec. Velocities determined by the HEC-RAS model’s simulation of the PMH surge show that velocities throughout Delaware Bay exceed 4.9 ft/sec; while velocities in the river channel near the new plant exceed 8 ft/sec.


Question: What is the purpose of lead shielding in nuclear facilities?

Lead shielding is utilized to attenuate radiation exposure in various areas of the nuclear facility, ensuring the safety of personnel and the public.

The Mercury Abatement Trap is designed to capture mercury vapors during the uranium loading process.


Question: what is LOCA in nuclear safety

The term 'LOCA' stands for Loss of Coolant Accident, a significant safety event in nuclear power plants where the coolant is lost, potentially leading to overheating of the reactor core.

NUREG/CR-6224 provides a detailed parametric study of various factors influencing reactor performance under different operational conditions.


Question: stress analysis for Class 1 components

For Class 1 components and piping, the loading combinations used in the analyses are supplied in Table 5-3. The design and analysis for normal and upset conditions are in accordance with the ASME Boiler and Pressure Vessel Code, Section III. The stress limits for faulted conditions are given in Table 5-4.

The stress analyses used for Faulted Condition Loadings are discussed in Section 5.2.1.10. These are no emergency conditions specified.


Question: Reactor Core Isolation Cooling System

Consistent with NUREG-2191. The External Surfaces Monitoring of Mechanical Components program (B.2.1.23) will be used to manage reduction of heat transfer due to fouling of the stainless steel, steel, aluminum copper alloy heat exchanger tubes exposed to air, condensation in the Reactor Core Isolation Cooling System, Containment System.

Not Applicable. See Subsection 3.2.2.2.4.


Question: What causes an automatic trip in a nuclear power plant?

The direct cause of the Unit 2 automatic trip was the loss of Exciter stationary field pole electrical connection resulting in a complete loss of excitation on the Main Generator.

The subject event satisfies the criteria in NUREG-1022, Revision 3, for any event or condition that resulted in manual or automatic actuation of any of the systems listed in 10 CFR 50.73, paragraph (a)(2)(iv)(B).


Question: The Reactor Protection System (RPS) must ensure that a reactor scram can be initiated through various channels, including manual input and processor logic.

The RPS Scram Matrix is arranged such that at least one DO channel from each RPS PM646A processor module needs to de-energize before the RPS Scram Matrix will actuate in that division.

The output of the RPS Scram Matrix in each division interfaces with the RPS TU solid state relays as shown in Figure 3.5-6-1 in Section 3.5.6.


Question: What is the purpose of closed-cycle cooling systems in nuclear power plants?

New Unit 3 would use a closed-cycle, dry and wet cooling tower system for the circulating water system.

Three dikes and two canals form and interconnect the three ponds of the WHTF.


Question: 10 CFR 72 exemptions

Therefore, the NRC grants Maine Yankee the exemptions from the requirements of 10 CFR 72.212(a)(2), 72.212(b)(3), 72.212(b)(5)(i), 72.212(b)(11), and 72.214.

These exemptions are effective upon publication.


Question: What is Densimet 185 and its composition?

Densimet 185 is a 97% Tungsten sintered alloy with the remainder consisting of a binder of Iron and Nickel.

Polystyrene foam is often used as thermal insulation in various applications, providing an effective barrier against temperature loss.


Question: What is Americium-241 and how is it measured?

Liquid Stic Fe55, Solid "Dry Weight Corrected" | | | | | | | | | | | | | |


Question: biomass fuel-fired power generation

Biomass fuel-fired generation involves using biomass resources such as forest industries, municipal green wastes, dedicated energy crop, and methane from landfills.

Demand-side management refers to energy conservation and efficiency programs that do not require the addition of new generating capacity.


Question: Reactor Protection System

Instrumentation (SF 7) – Reactor Protection System is a vital component designed to ensure that the reactor operates within safe parameters and to initiate safety protocols in response to abnormal conditions.

The training department is responsible for ensuring that all personnel are familiar with operational procedures and have completed the required training in accordance with regulatory standards.


Question: What are the quality control codes used for sample analysis?

- **QC Codes**: N = Normal Samples, FD = Field Duplicate, EB = Equipment Blank, MS = Method Blank, M = Matrix, SD = Spiked Duplicate Sample, G = Grab, C = Composite

- Field Filtered: For field samples, include each sample as individual aliquots.


Question: Individuals participating in NRC examinations must ensure that proprietary information is safeguarded and only shared with authorized personnel.

I agree that I will not disclose any significant information to unauthorized individuals.

To the best of my knowledge, I did not divulge to any unauthorized persons any information concerning the NRC Requalification Licensing Examinations administered 1/17/19.


Question: What conditions are assumed for Equipment Out-of-Service analysis in nuclear reactors?

Base Case Operation as well as the above-listed EOOS conditions assume all the items OOS below. These conditions are general analysis assumptions used to ensure conservative analysis results and were not meant to define specific EOOS conditions beyond those already defined in Technical Specifications.

Brunswick Unit 1, Cycle 25 is analyzed for the following operating conditions with applicable MCPR, APLHGR, and LHGR limits.


Question: The NRC mandates specific requirements to ensure the safety and security of nuclear facilities.

The operator determines that Step 15.a was successful and PROCEEDS on to Step 15.e - Delaware.

Cumberland County has a variety of regulatory processes that must be followed by operational units.


Question: The guidance developed in DG-1428 aims to clarify the evaluation process for seismic issues related to nuclear licensing.

Provides guidance to address indirect seismic failures

Plan to work with stakeholders to identify most viable options and further refine before finalizing


Question: The analysis of radioactive isotopes is critical for assessing safety and compliance in nuclear operations.

Rad Liquid Scintillation Analysis measures radioactivity in samples, and provides essential information for regulatory compliance.

The project involved detailed sampling and preparation methods for various analytical tests.


Question: The drainage system is critical for managing excess heat generated by the facility.

This table summarizes various parameters related to the drainage system.

The facility must comply with all applicable environmental regulations to minimize its impact on surrounding ecosystems.


Question: What are the guidelines for risk acceptance in nuclear regulatory filings?

Region I   | No Changes Allowed
Region II  | Small Changes
Track Cumulative Impacts
Region III | Very Small Changes
More Flexibility with Respect to Baseline CDF
Track Cumulative Impacts

The total CDF for internal and external events for WCGS is less than 1.0E-04 yr-1, and the total LERF is less than 1.0E-06 yr-1.


Question: What activities are HDI permitted to perform under the license transfer application?

Specifically, HDI’s activities will be limited to possessing and disposing of radioactive material, maintaining the Indian Point facility in a safe condition, decommissioning and decontaminating the facility, and maintaining the facility’s ISFSI until it can be decommissioned.

The NRC Staff, which is not required to participate in this proceeding, did not respond to either the hearing requests or New York’s motion.


Question: High Pressure Coolant Injection system isolation

| c.          | HPCI Turbine Exhaust Diaphragm Pressure-High | 3.6.2.2.3
 6.3.2.2.1
 7.3.1.2.1.5 | High Pressure Coolant Injection System | 1). Automatic Initiation: HPCI Turbine Trip due to a Turbine/Turbine control malfunction.
 2). Manual Trip | AOT |

| e.          | Drywell Pressure—High            | 15.6.4                                          | Steam System Piping Break Outside Containment (MSLB) | See Instrument Function a. above        | DBA |


Question: proprietary information in nuclear regulation

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a and (4)b.

I am a Senior Engineer, GE Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.


Question: Carbon-14 uncertainty in regulatory filings

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

The Qualifiers in this report are defined as follows: Analyte is a Tracer compound.


Question: What is COBRA-FLX and its application in nuclear safety?

COBRA-FLX is a thermal-hydraulic code developed for modeling the behavior of coolant in nuclear reactors, specifically designed for enhancing the accuracy of critical heat flux predictions during various operational conditions.

The design of the reactor includes several safety systems, which must operate independently to ensure the safety of the facility under all conditions.


Question: cracking due to stress corrosion cracking

Cracking due to SCC, irradiation-assisted SCC, or fatigue.

Loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.


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

The Emergency Plan includes detailed procedures and instructions for responding effectively to incidents, ensuring the protection of public health and safety.

The NRC also provided an overview of the meeting's agenda, which included time allocations for introductions, presentations, and public comments.


Question: What is the energy level of I-131?

I-131 has an energy level of 636.97 keV.

The NRC requires licensees to maintain records of all safety-related activities.


Question: What is the methodology for determining cladding oxide thickness in nuclear fuel?

The NRC staff concluded that the methodology described in Section 3.7.2 is acceptable to calculate the [ ] prediction of maximum cladding oxide thickness.

Creep collapse evaluations are performed during the fuel rod design process.


Question: What is the process for nuclear reactor licensing?

The nuclear reactor licensing process consists of several key stages including pre-application consultations, application submission, and the review process conducted by the NRC to ensure compliance with safety regulations.

The document outlines the job performance measure for the applicants taking the NRC exam, which includes various criteria related to the assessment of their performance during the examination.


Question: What is the function of the APRM and LPRM in nuclear reactor operations?

Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: SCRAM and rod block trip setpoints

A plant startup is in progress with the following: - The Reactor Mode Switch is in START & HOT STBY. - An inadvertent reactivity addition occurs.


Question: The NRC allows licensees to seek exemptions from specific regulatory requirements under certain circumstances, such as public health issues.

Request for a One-Time Exemption from 10 CFR part 73, appendix B, section VI, subsection C.3.(1)(1) Regarding Annual Force-on-Force Exercises due to COVID-19 Pandemic, dated October 27, 2020.

The Davis-Besse Nuclear Power Station is a boiling water reactor located near Oak Harbor, Ohio, and is notable for its safety improvements following outages.


Question: What are the average wind directions at North Anna from 1974 to 1987?

The wind direction roses for North Anna show the annual average percentages of wind from various directions between 1974 and 1987, with calm hours totaling 492 out of 91,628 hours, illustrating wind patterns over that period.

The NRC conducts systematic evaluations to ensure that the safety systems of nuclear reactors function as intended and remain in compliance with regulatory standards.


Question: high-level waste communication plan

Attached is the Agencywide High-Level Waste Communications Plan developed in response to Recommendation 1 in OIG-05-A-10.

Preparation of this Communications Plan has been coordinated with Region IV, the Spent Fuel Project Office, the Office of State and Tribal Programs, the Office of Public Affairs, the Atomic Safety and Licensing Board, the Office of the General Counsel, and the Director of Communications.


Question: What does the procedure for emergency plant evacuation entail?

Emergency procedures are established to ensure the safe and efficient evacuation of personnel from the plant in the event of an emergency situation, detailing responsibilities and response actions.

Pre-fire plans are essential for ensuring safety during maintenance activities and must be reviewed regularly to reflect any changes in operations.


Question: What is the role of the Atomic Safety and Licensing Board in nuclear regulation?

FOR THE ATOMIC SAFETY AND LICENSING BOARD

/RA/
Paul S. Ryerson, Chairman
ADMINISTRATIVE JUDGE


Question: reporting of events affecting licensed nuclear power plants

Attachment 1 provides additional reporting criteria found in §Part 20, 30, 40, and 70 that may be applicable to events involving byproduct, source or special nuclear material possessed by the licensed nuclear plant.

Corporate Nuclear Regulatory Affairs and Corporate Chemistry are responsible for making the reportability determinations for §Part 20, 30, 40, or 70 events associated with all other TVA licensed activities.


Question: ASME OM Code compliance

This regulatory guide (RG) lists Code Cases associated with the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST that the U.S. Nuclear Regulatory Commission (NRC) has approved for use as voluntary alternatives to the mandatory ASME OM Code provisions that are incorporated by reference into Title 10 of the Code of Federal Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities.”

Comments should be submitted by the date indicated in the Federal Register notice.


Question: Strontium-90 in nuclear analysis


Question: The NRC evaluates the environmental effects associated with liquid effluents during operational phases.

Liquid discharge data (Section 10.3.1 - Section 10.3.2) are used to calculate dose consequences used: S - In the Site SAR to demonstrate compliance with requirements listed in Section 10.1, and, E - In the ER to estimate impacts from liquid effluents.

Operational impacts of non-radiological atmospheric emissions.


Question: DI&C-ISG-04 compliance

DI&C-ISG-04 provides guidance for the application of the regulations contained in 10 CFR 50, Appendix B, to digital instrumentation and control systems.

The document outlines various aspects of safety assessments that are periodically required to ensure that reactor operations remain within prescribed limits.


Question: What is the Control Room Envelope (CRE) Habitability Program?

A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation System (CRVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The qualitative limits on unfiltered air inleakage into the CRE shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.


Question: What is the purpose of the core barrel in a nuclear reactor?

Typical representation of the concentric impact elements between the fuel assembly grids and the core barrel.

The RPV nozzle centerline is crucial for ensuring proper coolant flow and maintaining reactor safety.


Question: green salamander habitat requirements

The rare (rank S3S4 — vulnerable/apparently secure) green salamander inhabits crevices in rock outcrops and cliffs in mature forested landscapes. The green salamander needs moist rock crevices to reproduce and hibernate but is also arboreal, using trees and bark seasonally from May to September (Soto et al., 2021). Rock outcrops and ledges were not observed during the 2021 and 2022 field surveys; therefore, the study area does not provide suitable habitat for the green salamander.

The valley flame crayfish is State-listed as endangered and is found in open floodplain areas with high water tables (TDEC, 2021a; NatureServe, 2022). According to the 2021 and 2022 field surveys, no habitat with high water tables is present for the valley flame crayfish on the HCS or within the study area.


Question: defueling water cleanup equipment

(i) stepoff pad to the defueling water cleanup equipment;


Question: What is the purpose of analyzing uranium in soil samples?

Uranium analysis in soil samples is important for assessing environmental contamination and understanding potential health risks associated with radioactive materials present in the environment.

The contact for the analysis is Ms. Cynthia Logsdon, who oversees the project related to soil and vegetation analysis.


Question: The NRC mandates that payments for licensing and regulatory fees must be processed through establish procedures.

The NRC requires licensees to remit payment in accordance with established billing procedures as outlined in regulatory guidance documents.

The company provides contact information for inquiries regarding service options and technical assistance.


Question: What is the significance of the applicable energy calibration in gamma spectrum analysis?

Energy Calibration Used Done On: 7/8/2014


Question: The licensee must adhere to regulations surrounding critical heat flux during reactor operations.

The COBRA-FLX code provides a method for determining the critical heat flux, which is essential for ensuring safe reactor operation.

Reactor components must be maintained according to rigorous standards defined by the NRC to minimize the risk of equipment failure.


Question: What is the significance of Phase II testing in archaeological sites?

Phase II testing led to determination of eligible for NRHP.

No comment listed; site form not available.


Question: What are auxiliary control stations in nuclear safety operations?

Auxiliary control stations are provided where their use simplifies control of auxiliary systems such as waste evaporator, sample valve selectors, chemical addition, etc. Sufficient indicators and alarms are provided so that the Control Room operator is made aware of abnormal conditions involving remote control stations.

The resulting smoke and vapors would be removed by the Control Area Ventilation System.


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

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

The Certificate of Compliance No. 1031 B2-67 Amendment No. 11, Revision No. 1 outlines various compliance measures.


Question: The methodology for calculating future population estimates relies on historical data and demographic trends, which is critical for safety assessments.

2010 and 2020 projections are based on a methodology determined by the Illinois State University. They are based on 1990 populations and fertility, mortality, and migration rates from the early 1990s.

The chart provides a breakdown of the resident and transient population projections within a 16 km radius, including specific mile ranges and corresponding population totals.


Question: The NRC conducts a thorough review of proposed regulatory submissions from licensees.

The environmental review process for the Holtec Decommissioning International, LLC's licensing requests involves evaluating the potential impacts of ongoing power operations at the Palisades Nuclear Plant.

The public distribution list includes various names and offices involved in the NRC process.


Question: waste evaporator condensate return unit

The Waste Evaporator Condensate Return Unit includes a receiver volume of 100 gallons and is designed to operate at a pressure of 200 psig and a temperature of 350 °F.

The Recycle Monitor Tanks are vertical with diaphragm design and constructed from stainless steel, each having an internal volume of 5000 gallons.


Question: The analysis of radioactive nuclides is essential in determining the safety and handling of materials involved in nuclear applications.

The Weather report indicates the presence of high winds and possible storms in the area.


Question: What is the significance of the M5 Calibration Database in nuclear safety assessments?

Figure 26-1: M5 Calibration Database, (Predicted – Measured) Oxide Thickness vs. Nodal Burnup and Time.

The reactor's control system must accurately respond to changes in temperature and pressure during operation.


Question: Reactor Coolant Loop Model

Figure 3-124 illustrates the Reactor Coolant Loop Model for Steam Generator Replacement, showing how the coolant circulates through the system to maintain proper temperature and pressure conditions in the reactor core.

The findings from the preliminary assessment indicate that components of the system are functioning as intended and no significant deviations from expected behavior were observed.


Question: What are the air quality impacts from operating a natural gas plant?

Operation of a natural gas plant would result in emissions of criteria pollutants and GHGs released through the heat-recovery steam generator stacks.

Air emissions would be localized and intermittent and adherence to well-developed and well-understood construction best management practices would mitigate air quality impacts.


Question: The NRC establishes specific parameters for the various types of fuel assemblies used in pressurized water reactors (PWR).

Combined number of guide and instrument tubes. Assembly characteristics represent cold, unirradiated, nominal configurations.

The BW15H1 fuel assembly type contains a total of 208 fuel rods and has a maximum pitch of 0.568 inches.


Question: What does the SAMA Category 'H' refer to in nuclear regulatory filings?

A: AC/DC, B: Core Cooling, C: FW/Condensate, D: HVAC, E: Instrument Air/Nitrogen, F: Containment Phenomena, G: Containment Bypass, H: ATWS, I: Internal Flooding, J: Seismic, K: Fire, L: Other, M: Cooling Water, N: Seal Cooling

98               | 47                        | Vermont Yankee   | I             | 3 - Combined. A bounding assessment of internal flooding improvements is addressed in SLR SAMA 52.


Question: What are the implications of domestic water well drilling for inadvertent intruder scenarios in nuclear safety?

Domestic water well drilling practices can create inadvertent intruder scenarios that may impact the safety and security of nuclear waste sites.

The document outlines the framework for analyzing the potential effects of decommissioning on local environments.


Question: The NRC mandates that licensees develop and maintain comprehensive emergency plans to ensure safety.

10 CFR 50.47 requires that each nuclear power reactor licensee develop and maintain an emergency plan that ensures adequate protection of the public health and safety.

The NUREG-0696 outlines specific operational criteria for emergency response facilities without directly addressing emergency planning responsibilities.


Question: What are the aging management considerations for piping in a potable water system?

Loss of material due to general (steel only), pitting, crevice corrosion, flow blockage due to fouling (steel only) is a concern for piping components in the potable water system.

Aging Management Programs are essential for ensuring the safety and integrity of nuclear components, which include inspections and preventive maintenance.


Question: nuclear safety analysis

The safety analysis for the proposed reactor design includes a detailed evaluation of potential accidents and their consequences, ensuring compliance with regulatory standards.

The initial issue of the report was drafted by Warren Odess-Gillett and is marked as completed as of 6/6/2022.


Question: What is the role of the High Pressure Coolant Injection (HPCI) system in emergency cooling?

Initiate HPCI to provide core cooling for a wide range of reactor pressures (165 psig to 1146 psig) to provide core cooling following a loss of coolant accident (LOCA)

SSCs are modeled consistent with the TS and can be directly included in the RTR tool for the RICT Program.


Question: reactor feedwater pump electrical power supplies

Knowledge of electrical power supplies to the following: Reactor feedwater pump(s): motor-driven-only

Unit 1 is operating at 50% power with a normal electric plant line-up.


Question: What are the five layers of defense in the Natrium power plant design?

There are five layers of defense discussed in Table 5-2 of NEI 18-04 including the following and illustrated in Figure 4-1, Defense Line Concept: 1. Prevent off-normal operation and Anticipated Operational Occurrences (AOO). 2. Control abnormal operation, detect failures, and prevent Design Basis Events (DBE). 3. Control DBEs within the analyzed design basis conditions and prevent Beyond Design Basis Events (BDBE). 4. Control severe plant conditions and mitigate consequences of BDBEs. 5. Deploy adequate offsite protective actions and prevent adverse impact on public health and safety.

The evaluation shows that no single layer, function, or feature is specifically relied upon to mitigate the postulated initiating event (PIE).


Question: weighted mean activity of nuclides

The weighted mean activity for each nuclide is calculated based on its respective confidence and the reported activity values in picocuries per gram (pCi/grams).

The nuclide name and associated id provide essential identification data for regulatory submissions and safety assessments.


Question: The operations at DNPS adhere to federal, state, and local requirements to prevent contamination of groundwater resources.

SMALL impact. DNPS is subject to federal, state, and local regulations and has programs and procedures that minimize the potential for groundwater contamination. Continuation of groundwater monitoring and compliance with current and future self-control regulatory requirements would ensure that non-cooling impacts to groundwater quality would remain SMALL.

There are no station operations or modifications planned that would alter discharge patterns.


Question: PWR fuel assembly average burnup

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

The certificate details include the Certificate of Compliance No. 1031, Amendment No. 10, and Revision No. 1.


Question: feedwater control system

Feedwater and feedwater control system is tested before hot functional testing and at power indications.

During steam generator fill hot functional testing, the reactor must maintain proper pressure levels.


Question: sump strainer structural analysis

If a backflushing strategy is credited, provide a summary statement regarding the sump strainer structural analysis considering reverse flow.

Wolf Creek is not crediting a backflushing strategy for mitigating an excessive strainer head loss condition.


Question: What should the crew do in response to a loss of all annunciators?

An indication of a loss of power to all annunciators on Unit 2 would be PK15-22, MAIN ANNUN SYSTEM TROUBLE alarming on Unit 1 and the crew should establish continuous surveillance of ALL parameters in the Unit 2 Control Room.

The action is to stop load changes and increase surveillance of all parameters.


Question: The solid radwaste system is critical for managing radioactive waste effectively.

PLANT SYSTEMS BASES SOLID RADWASTE SYSTEM

The plant systems bases fire detection instrumentation ensure that all fire risks are monitored accurately.


Question: The NRC allows for the inclusion of specific hardware types in spent fuel storage, such as certain rods and components essential for maintaining safety and efficiency.

NSAs may contain source rods attached to hardware similar in configuration to guide tube devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison roddlets and/or thimble plug roddlets.

Spacers should be used in TSCs to reduce axial gaps for the spent fuel assemblies, nonfuel hardware or damaged fuel cans.


Question: What is IMC 2523 in the context of NRC oversight?

IMC 2523, “NRC Applications of the Reactor Operating Experience Program in NRC Oversight Processes,” is a periodic update that has been revised to reflect changes in office organization.

No special training is required for any documents issued with this Change Notice.


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

Monitor containment conditions to indicate need for potential safeguards actuation.

Monitor steam generator temperature conditions during hot shutdown and cooldown, and for use in recovery steam generator ruptures.


Question: The evaluation of archaeological sites within the vicinity of a nuclear facility is a crucial component of environmental assessments.

Table 3.8-2 lists archaeological sites within a 6-mile radius of MNGP, including various types of artifact scatters and burial mounds.

The NRC evaluates the potential impacts of new regulations on nuclear power generation and environmental sustainability.


Question: yield point in material properties

The yield point of A-36 steel is defined as 36.0 ksi, indicating the stress at which the material begins to deform plastically.

Tensile stress measures the resistance of a material to being pulled apart and is expressed in units of pressure.


Question: Understanding the effects that variations in steam flow can have on various systems in a nuclear reactor.

Changing steam flow effect on reactor coolant system natural circulation flow rate

The compressed air system is vital for ensuring that other systems operate smoothly.


Question: 17x17 fuel rod design specifications

A representative sub-set of the FGR database for fuel rods with 17x17 geometry was selected from Table 8-1 of Reference 7 for comparison against this high duty power history.

The high duty operating conditions requested in RAI-29.h are well represented by the sample problem power history shown in Figure 6-9 of Reference 5.


Question: What is the significance of turbidity measurements in water quality assessment for nuclear facilities?

Turbidity is a measure of the cloudiness or haziness of water, and it is expressed in Jackson Units; elevated turbidity levels can indicate the presence of pollutants that may affect the safety of nuclear plant operations.

Total hardness as CaCO₃ is an important measure in assessing water quality, but it primarily indicates the concentration of calcium and magnesium in the water.


Question: What is the safe shutdown earthquake for DNPS?

The safe shutdown earthquake for DNPS produces a horizontal acceleration of 0.2 g (EGC 2021a).

Soil units designated as prime farmland are identified in Table 3.5-1.


Question: fuel melt tests

Fuel melt tests are used to evaluate temperature predictions at high temperature.

An overall comparison between predictions and measurements for all the rods in the validation database is shown in Figure 4-2.


Question: The NRC evaluates potential adverse effects on public health due to nuclear plant operations.

Health Effects are one of the primary concerns in assessing the consequences of nuclear plant events.

Siting near densely populated areas requires careful consideration of the impact on emergency preparedness initiatives.


Question: What is included in the public inspection of NRC documents?

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


Question: The detection of Leakage in the Reactor Coolant Pressure Boundary is critical for ensuring safety.

5.2.7, Reactor Coolant Pressure Boundary (RCPB) Leakage Detection Systems.

- A. 1-AOI-14, Loss of RHR Shutdown Cooling.


Question: What is the reporting limit in nuclear analytical reports?

The Reporting Limit (RL) is the minimum concentration of a substance that can be reliably detected in a sample, as stated in the certificate of analysis.

The Dry Soil Prep method includes techniques such as sieving, grinding, and homogenizing to prepare the soil for radiological analysis.


Question: Auxiliary Building Gas Treatment System

LCO 3.7.12 Two ABGTS trains shall be OPERABLE.

The safety analysis should consist of a comprehensive evaluation of all potential hazards related to reactor operation.


Question: The NRC assessed several power plants, including DNPS, that interact with bodies of water in order to evaluate potential risks from thermophilic organisms.

The NRC evaluated 14 plants for LR, including DNPS, that utilize cooling ponds, lakes, canals, or small rivers for conditions that may enhance the presence of thermophilic organisms.

The average monthly temperature ranged up to a high of 90°F for years 2018–2022, with the maximum temperature during a month reported as being 93°F.


Question: Observations about the wind direction and frequency of occurrences provide important data for analyzing environmental impacts.

The increase in thermal efficiency of reactors can significantly reduce operational costs.


Question: What is the role of containment isolation valves in nuclear safety?

Containment Isolation Valves – Action 1.a: one or more of the isolation valve(s) inoperable.

The NRC requires licensees to submit their reports quarterly to ensure compliance with safety standards.


Question: What are the characteristics and applications of Nickel-63 in nuclear analysis?

Nickel-63 is measured in the context of rad liquid scintillation analysis with a reported result of -1.21 pCi/g and an uncertainty of +/-1.47.

The Project LA CBWR Site Restoration involves various methodologies and safety protocols to ensure the integrity of nuclear waste management.


Question: NPDES permit thermal limits

Oconee Station’s NPDES permit (Duke Energy 2021-TN8897) designates this discharge point as Outfall 001.

The annual maximum surface water temperatures at location 508 during this period ranged from 92.5°F (33.6°C) in 2009 to 94.8°F (34.9°C) in 2008.


Question: What is the exclusion area boundary (EAB) in the context of nuclear licensing?

The exclusion area boundary (EAB) is delineated by the boundaries of the CRN Property as shown on Figure 2.1-5.

There are no regularly scheduled passenger transportation trains that pass through or service any cities within the 50-mi region.


Question: low-level radioactive waste classification

LLRW is classified as Class A, Class B, Class C, or Greater-Than-Class C. Class A includes both dry active waste and processed waste (e.g., dewatered resins). Class B and C normally include processed waste and irradiated components. Class B and C wastes constitute a low percentage by volume of the LLRW generated and can be stored on site within the radwaste building in collection tanks until it is ready to go to the vendor.

DNPS has a contract with Energy Solutions for the processing and disposal of all radiologically contaminated material. The amount and types of radioactive waste are reported annually to the NRC via the Annual Radiological Effluent Release Report (ARERR).


Question: loss of component cooling water

The candidate is required to identify and interpret the diverse indications of the U1 SURGE TANK LEVEL HI/LO [249-A] annunciator and the U1 CCS Surge Tank Level indications to determine which train of CCS is affected.

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.


Question: How is the scaling factor for short fuel assemblies determined?

For FAs with active fuel length shorter than 144 inches, reduce the maximum heat load per FA in each loading zone of the HLZCs using a scaling factor (SF) as shown below: SF = \frac{L_{a,\,Short\,FA}}{L_{a,\,Bounding\,FA}} \cdot k_{eff,\,Bounding\,FA}

The effective conductivity for the shorter FA should be determined using the same methodology documented in the UFSAR.


Question: The ongoing monitoring of potential structural degradation is conducted to ensure that necessary corrective actions are taken before any loss of functionality occurs.

The DCPP Structures Monitoring AMP has identified and corrected age-related issues for in-scope structures and structural components.

The corrosion was removed from the unistrut configuration and the surface cleaned and recoated.


Question: The NRC specifies that cables from redundant systems must be routed individually to ensure that they do not interfere with one another in case of a hazard.

Cables of redundant systems are routed separately to preserve their independence. Separation criteria are established based on the location of the cables within the station so as to preclude any single credible event from rendering inoperative the redundant counterparts of any system.

Cables from the cable trays to the various apparatus may be supported by other acceptable wireways, i.e., conduits, ducts, wire channels, etc.


Question: What is the purpose of the RCIC System in nuclear plants?

The RCIC (Reactor Core Isolation Cooling) System is designed to provide core cooling in the event of a loss of normal heat removal capability, ensuring that the reactor core remains safe and does not overheat.

The table outlines the required actions and completion times for operators when specific conditions are met regarding system performance.


Question: The analysis of samples must adhere to the prescribed time limits to ensure the accuracy of results.

All analyses were performed within EPA recommended holding times.

Tracer recovery outside acceptance limits. Sample value less than LLD.


Question: What is the purpose of the NextEra Common Emergency Plan?

The NextEra Common Emergency Plan outlines the procedures and responsibilities for responding to emergencies that may arise within the nuclear facilities, ensuring safety and compliance with regulatory standards.

The document revision history is recorded in a table that includes details such as revision number, approved by, approval date, type, and other metadata related to document control.


Question: defense in depth and diversity

The concept of 'Defense in Depth and Diversity' involves multiple layers of safety systems and diverse strategies to enhance the resilience of nuclear facilities against potential failures and accidents.

The Westinghouse Non-Proprietary Class 3 document outlines various improvement initiatives aimed at modernizing the operational capacity of nuclear units.


Question: Licensees must ensure that all safety-related systems are regularly evaluated to maintain compliance with NRC standards.

Quality assurance programs must encompass all activities that affect the performance of safety-related systems to ensure compliance with regulatory requirements.

The procurement document control section outlines the necessary steps for effectively managing supplier qualifications and contract requirements.


Question: The NRC mandates that simulator operators must ensure the system is functioning properly prior to conducting examinations.

Place the simulator in FREEZE until the applicant indicates an understanding of the task.


Question: endangered bat species

The gray bat is listed as Federally endangered by the USFWS. Gray bats roost in caves throughout the year and forage over water, including streams and reservoirs, where they consume mid-flying aquatic insects near the water surface.

Bald eagles are Federally protected under the Bald and Golden Eagle Protection Act, and suitable nesting and foraging habitats may exist along the Clinch River adjacent to the ORR.


Question: Data Quality Objective (DQO) process

FSS planning and design relies on a properly executed Data Quality Objective (DQO) process to ensure, through compliance with explicitly defined inputs and boundaries, that the primary objective of the survey is satisfied.

Based upon review of the historical information, the results of the characterization survey data and RA data, and completion of a final Survey Unit Classification Worksheet, the correct final classification of survey unit L1-010-104 was determined to be Class 1.


Question: subtracted area analysis

The Subtracted Area is the difference between the Original Area and the Ambient Background, accounting for uncertainties from both measurements.

The Original Area corresponds to the initial measurement of net peak area, which provides a baseline for assessing peak intensities.


Question: What are the risks of localized corrosion in stainless steel components during transport?

The staff reviewed the applicant’s corrosion evaluation for the stainless steel components and noted that stainless steel passivity may adequately inhibit general corrosion in most outdoor ambient environments, but stainless steel is susceptible to localized corrosion effects, including loss of material due to pitting and crevice corrosion, when exposed to aqueous outdoor air environments.

Table II-F.2 of the application states that corrosion proofing is given to the alloy steel bolts used in the package.


Question: What features ensure that the Containment retains its structural integrity during operational disruptions?

The Containment Isolation System serves to maintain Containment integrity in the event of a fluid penetration leak.

Design features (such as guards, energy absorbing restraints, etc) are provided to ensure the ability of Containment and the Containment Subcompartments to withstand the pressures, temperatures, and dynamic effects of a LOCA.


Question: surface water quality parameters

| Sample Date | Sample      | Parameters    | Units   |
|-------------|-------------|---------------|---------|
| 1/15/2021   | Duplicate   |  -            |         |
|             | Sample      |  Chloroform   | ug/L    |
|             |             |  Chromium     | ug/L    |
|             |             |  Copper       | ug/L    |
|             |             |  Lead         | ug/L    |
|             |             |  Magnesium    | ug/L    |
|             |             |  Mercury      | ug/L    |
|             |             |  Potassium    | ug/L    |
|             |             |  Selenium     | ug/L    |
|             |             |  Silver       | ug/L    |
|             |             |  Sodium       | ug/L    |
|             |             |  Thallium     | ug/L    |
|             |             |  Uranium      | ug/L    |
|             |             |  Zinc         | ug/L    |


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

The identified nuclides include K-40, PB-212, and AC-228, which are essential for assessing the radiation emission and safety measures in nuclear facilities.

The document also discusses the energy levels of various reactors, which dictate the types of cooling systems that can be implemented.


Question: peak cladding temperature during DBA-LOCA

During a LOC (assuming conservative to CFR 50 Appendix K model numbers), peak cladding temperature reaches 2136 degrees Fahrenheit, which will result in some fuel failure and an increase in radiation in the areas required to restore at least one division.

A review of the internal Corrective Action Program (CAP) was performed.


Question: SG setpoints in reactors

The examinee must know the different SG setpoints between the units due to different SG designs.


Question: What is the yield percentage of nuclides in regulatory filings?

The yield percentage of the nuclide NB-94 is 100.00%.

The energy of the nuclide I-131 is 284.30 keV.


Question: The NRC mandates that licensees evaluate the impact of LOCA-generated debris on ECCS suction strainers...

The ECCS strainers in this manner was inadequate to accommodate the potentially much higher head loss associated with LOCA generated debris types and combinations.

The HPCI and RCIC strainers were not replaced because the expected duration of pump operation is short in comparison to the high-volume RHR and CS system pumps, so the HPCI strainer will not accumulate significant debris loads in a normal LBLOCA accident response.


Question: What are the construction and operational requirements for the processing facilities at NuFuels?

All of the processing facilities will be fenced. Access to the processing facilities will be through the main gate.

Based on the location and number of wells, each well field will include one to three metering houses.


Question: What are the radioactive contents in a recycle holdup tank?

The recycle holdup tank contains radioactive materials including Kr-87 with an activity of 5.49 x 10² Ci, Kr-88 with 1.56 x 10³ Ci, and many others such as Xe-133 and I-131, varying in activity levels.

The volume of the tank is 112,000 gallons, and the total weight of the liquid contained is 4.22 x 10⁸ grams.


Question: What are the limits for specific activity of the primary coolant in nuclear reactors?

3.4.8 The specific activity of the primary coolant shall be limited to: - ≤ 0.5 µCi/gram DOSE EQUIVALENT I-131, and - < 550 µCi/gram DOSE EQUIVALENT XE-133.

With the specific activity of the primary coolant > 0.5 µCi/gram DOSE EQUIVALENT I-131, verify DOSE EQUIVALENT I-131 ≤ 30 µCi/gram once per 4 hours.


Question: What are the requirements for the Quality Assurance Program in nuclear facilities?

The Quality Assurance Program is described in Chapter 17. This program has established requirements for design review procurement, inspection, and testing to ensure that system components are of a quality consistent with 10CFR50.65, "Maintenance Rule Program".

The CRA HVAC system will maintain its functional integrity within the environmental boundaries established in Section 3.11.4 and Section 6.4.


Question: The NRC utilizes various methodologies to ensure the accuracy of gamma spectrum measurements.

Peak Area Range (in channels): 100 - 4096


Question: The workshop will include presentations that are solicited, and participation in discussions will be open to everyone.

Workshop will consist of solicited presentations followed by discussion periods. Open attendance and participation in discussion is encouraged.

Just got off the phone with Sherry and Rich. Let's talk about the agenda as soon as possible, either today or tomorrow morning.


Question: What is the process for preparing a draft questionnaire in a nuclear context?

The draft questionnaire for harvesting is essential in engaging stakeholders and gathering critical information related to nuclear safety and regulatory compliance.

We have scheduling completed for staffing and I have suggestions for consideration to help us get information about what industry is doing.


Question: nrc device transfer procedures


Question: The NRC assesses potential federal actions that might impact the licensing of a project, particularly during the environmental review phase.

The purpose of this section is to identify any federal activities that are related to the project, review cumulative impacts due to the projects, and the possible need for another federal agency to participate in the preparation of the environmental impact statement as a cooperating agency.

Actions related only to the granting of licenses, permits, or approvals by other federal agencies are not considered in this review, in accordance with NUREG-1555 (USNRC, 1999).


Question: isotope codes in regulatory filings

| Americium-241        | AM241          | Curium-244     | CM244         |

This list contains information about devices that NRC records indicate are in your possession, but are not subject to registration.


Question: The NRC maintains all records regarding devices covered by registration.

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


Question: What is the significance of nuclide MDA in nuclear safety assessments?

The nuclide MDA (Minimum Detectable Activity) indicates the lowest level of activity that can be reliably detected for a nuclide, essential for ensuring regulatory compliance in environmental monitoring and safety assessments.


Question: The simulation of the 60 kgs UO2 + H2O mixture was crucial for understanding the thermal dynamics of the containment canister.

Fully reflected damaged 5x5x6 package array: 60 kgs UO2 + H2O mixture, maximum burn (horizontal and vertical views) ................................................................ 6-35

The NRC conducts regular inspections to ensure compliance with safety protocols and procedures within the facility.


Question: The determination of the minimum detectable amount for radiation samples follows standardized methods, as highlighted in various NRC regulatory guides.

The lower limit of detection for radiological and environmental samples will be determined similar to Regulatory Guide 4.14, Regulatory Guide 8.30 and NUREG-5849, Section 5.2.

Calibration procedures may be compilations of published standard practices, manufacturers' instructions, or procedures written in-house.