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

The Safety Analysis Report (SAR) provides an assessment of the potential hazards associated with the facility and describes how the design and operation of the facility provide adequate protection to the public and environment from the consequences of those hazards.

Nuclear plants utilize various cooling systems to maintain optimal operation of their reactors, ensuring efficient power generation.


Question: What factors are considered for the renewal of nuclear power plant licenses?

The NRC is seeking public dialogue related to license renewal for nuclear power reactors, specifically the need (and timing) for the NRC to consider the potential technical issues and guidance document development related to license renewal that would authorize operation for up to 100 years.

The NRC encourages participation at public meetings to discuss various regulatory issues.


Question: We collaborate closely with technical specifications to determine the safety margins needed for plant operations.

But we work closely with the tech spec group on defining where those operational boundaries are that accommodates the safety margins that we demonstrate along with the consideration of the required operational margins necessary to support the plant systems and power production requirements.

So part of the answer I guess I was kind of probing was something to say that, well, you know, we look at the PIRT and we kind of look at what's important with regard to the influences coming in from the initial conditions and we select -- you don't have to bias everything because then it kind gets to become an administrative nightmare if you try to be too cute about it.


Question: CO₂ equivalent emissions

CO₂eq is a metric used to compare the emissions of greenhouse gases (GHG) based on their global warming potential (GWP). The GWP is a measure used to compare how much heat a GHG traps in the atmosphere. The GWP is the total energy that a gas absorbs over a period of time compared to carbon dioxide. CO₂eq is obtained by multiplying the amount of the GHG by the associated GWP.

Global surface temperature has increased faster since 1970 than in any other 50 year period over at least the last 2,000 years.


Question: The NRC mandates that licensees regularly monitor the gamma radiation levels in specified areas.

The Measurement Population table provides results for gamma radiation activities over various measurement IDs, indicating the gamma result (dpm/100 cm²) associated with specific measurements.

The Measurement ID values in the table range from SL25 to SL30, reflecting a systematic approach to data collection.


Question: The NRC evaluates the average enrichment of uranium in fuel assemblies to ensure safe reactor operations.

| Initial Assembly Avg. Enrichment wt % 235U (E) | WE 14x14 Assembly Average Burnup (B) GWd/MTU |

The table lists various neutron flux levels experienced by the reactor during different operational cycles.


Question: What is a Materials License in the context of NRC regulations?

Materials License No. 21-32740-01 (Premier Cardiovascular Specialists), Docket No. 030-38009 for Control No. 642205.

It took some unexpected wrangling to get this letter signed.


Question: Nickel-63 usage in nuclear applications

The Liquid Scint Ni63, Solid "Dry Weight Corrected" parameter reported a result of +3.53 pCi/g for Nickel-63.

The following Prep Methods were performed using various techniques including DOE EML HASL-300, ASTM C 147-00 Modified, and EPA 906.0.


Question: steam generator tube inspections

For the first NPM that undergoes a refueling outage, after the first 100 percent tube examinations at the first refueling outage they're required to inspect at least 20 percent of the steam generator tubes at each outage with the tech spec requirement that they have to get to 100 percent of tube inspection by 72 EFPM after the first outage.

So we do have specifics on the first module. Erin, I don't know if you can step here on the exact language for the first module?


Question: What happens when one required battery charger on Division I or II is inoperable?

In this condition, the remaining two divisions of the distribution system provide the necessary electrical power to the associated ESF components for safe shutdown.

The TSTF-505 changes are incorporated into the regulatory framework to enhance safety protocols.


Question: The NRC has established requirements for maintaining instrumentation such as excessive flow check valves in nuclear plants.

The proposed alternative should comply with the technical specifications required by the ASME code.


Question: soil classification for nuclear reactor sites

The soil profile identified at a depth of 30.0 feet includes SAND (SW), gray to tan, medium to coarse grained, with trace clay, indicating a well-graded and saturated condition, which is crucial for assessing site suitability.


Question: heat transfer coefficient units

The industry analysis in EPRI report 300202850, for example, uses HTC units of BTU/hr-ft²-°F and BTU/sec-in²-°F.

The last paragraph of Section 5.3 (“Loads”) in Enclosure 1 describes axial force and bending moment loads from the support ring and support lugs.


Question: What are the regulatory requirements for withholding proprietary information in nuclear report submissions?

The NRC staff has reviewed your application and the material in accordance with the requirements of 10 CFR 2.390. Based on the statements in the affidavit, the NRC staff has determined that the submitted information contains proprietary, commercial information and should be withheld from public disclosure.

The project manager for this report is Ngola Otto, who oversees multiple licensing projects in the reactor regulation division.


Question: What is the process for renewing licenses for non-power production or utilization facilities?

Define the license renewal process for all commercial or industrial NPUFs (including testing facilities) licensed under § 50.22 and testing facilities licensed under § 50.21(c);

Eliminate the requirement for NPUF licensees to submit financial qualification information with license renewal applications under § 50.33(f)(2).


Question: The system for discharging effluents to the cooling tower includes a flow interlock that is crucial for maintaining operational safety.

Flow interlocked with cooling tower blowdowns ensures that the discharge to the cooling tower makeup is effectively managed within safety limits.

The auxiliary steam generator is critical in maintaining the plant's thermal efficiency but operates independently of the discharge system.


Question: 4160 electrical system setpoints

State the setpoints, coincidence, blocks and permissives for automatic actuations associated with the 4160 Electrical System.


Question: What is the purpose of the reactor coolant pump trip?

Reactor coolant pump trip is a safety mechanism designed to halt the operation of the reactor coolant pumps to prevent potential overheating of the reactor core during abnormal conditions.

The main gas system is used for various systems throughout the plant and includes the compressor and storage tanks.


Question: The systematic approach will be utilized to validate that all measurements and sampling procedures adhere to regulations.

Calibrations will be performed on radiation detection instruments at annual intervals.

Field sampling and measurement records will be maintained at the Crownpoint Site.


Question: average enrichment in nuclear fuel assemblies

Avg. Enrichment wt % 235 (E) represents the weight percentage of Uranium-235 in nuclear fuel assemblies, which is crucial for determining the fuel's potential for sustaining a nuclear reaction.

The cooling time for nuclear reactors is a critical factor in ensuring the safety and efficiency of operations, which can vary significantly based on the reactor type.


Question: detector setup for radiation measurements

The logged readings were taken with a detector setup that consisted of a specialized Geiger-Muller tube to accurately measure radiation levels in counts per minute (cpm).

The logged reading for sample number zero was recorded at 198 counts per minute with a count time of 60 seconds.


Question: What are the elements of a PRA peer review process?

All of the PRA models discussed in this Enclosure have been peer reviewed and assessed using PRA Standard ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2. Each peer review identified F&Os for supporting requirements of the relevant parts of the PRA standard applicable to the scope of the peer review.

The Diablo Canyon PRA models are at-power models consisting of four hazard models: internal events, internal flooding, internal fire, and seismic events.


Question: logical connectors in Technical Specifications

Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in Technical Specifications are “AND” and “OR”. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action).


Question: what is rod diameter in nuclear reactors?

Rod diameter is a critical parameter in nuclear reactor design, influencing the reactor's thermal and hydraulic performance.

Two basic package array model constructs are included in this evaluation - undamaged and damaged.


Question: What is the significance of isotopes like CS137 in nuclear devices?

**Distributor/Distributed By**: KAY-RAY/SENSALL, INC.


Question: nuclide activity

The activity of a nuclide is typically measured in picocuries (pCi) per gram, which provides a quantitative assessment of the radioactivity present in a specific mass of the material.

The safety report includes various technical analyses that assess the performance of safety-related systems during different operational states.


Question: What are the different sizes of LOCAs and their significance in nuclear safety?

Very small, small, medium, and large break LOCAs include both pipe breaks and non-pipe breaks. For the WCGS GSI-191 evaluation, breaks were analyzed at pipe welds on the RCS and connected system piping.

The design of safety systems is crucial for maintaining operational integrity and the overall safety of the plant.


Question: The NRC requires that neutron absorber materials undergo acceptance testing to ensure they meet safety standards.

Acceptance and qualification testing of borated aluminum alloy and borated MMC neutron absorber material shall be in accordance with Sections 10.1.6.4.5, 10.1.6.4.6 and 10.1.6.4.7.

The licensee shall ensure that fuel oxidation and the resultant consequences are precluded during canister loading and unloading operations.


Question: automated patch management

DNFSB has automated the patching for all agency-issued workstations using a combination of Microsoft Intune and Kace patch management tools; workstations are assigned to various deployment "rings" and patches are automatically applied according to a set schedule based on the ring each workstation belongs to.

DNFSB considers Recommendation 2022-8 to be fully remediated.


Question: core deformation analysis

A detailed analysis of the deformation that can occur in the core has been conducted and reported in the Final Report on Predictive Analysis of PRD as a function of Deformation and Other Anomalies.

The reference core and restraint designs were used for preliminary analysis of the core restraint system parameters.


Question: The La Crosse Boiling Water Reactor (LACBWR) is a specific type of nuclear reactor design used for generating electricity.

LACBWR stands for La Crosse Boiling Water Reactor, which is a type of nuclear reactor located in Wisconsin, USA.

The table above presents various acronyms and their corresponding descriptions, such as Ci for Curie and CWD for Circulating Water Discharge.


Question: What is the role of the Digital Control System in nuclear plant operations?

PSL's plant computer system has been upgraded several times since the original PSL emergency plan was approved. The current system, Digital Control System (DCS) / Emergency Response Data Acquisition and Display System (ERDADS) graphically displays pertinent parameters with trending and graphing capabilities, alarm functions, and color-coded Critical Safety Function Status Trees.

The improvements to processes, procedures, equipment, and training that have occurred since initial approval of the PSL emergency plan have resulted in a significant increase in on-shift and augmented ERO knowledge and capability.


Question: What is the activity of nuclides in pCi/grams?

The activity for CS-137 is measured at 9.80E-02 pCi/grams.

The nuclide MDA for CE-144 is 2.45E-01 pCi/grams.


Question: The NRC mandates periodic inspections to assess the integrity of containment structures and ensure safety compliance.

Generally, the facility technical specifications require periodic examinations of the containment integrity, including inspections of the tendons.

Witnessing the initial concrete placement is important to verify that the concrete is being handled and tested properly.


Question: nuclide identification

Nuclide identified during the nuclide identification

Half-life too short to be able to perform the decay correction


Question: Nuclear facilities must undergo rigorous regulatory assessment before they can commence operations.

The licensing process for nuclear reactors involves a thorough review of safety, design, and operational procedures as mandated by the NRC.

The development of wind energy has gained significant attention in recent years, offering a sustainable alternative to fossil fuels.


Question: PWR fuel loading table

The emergency diesel generators are tested regularly to ensure they can provide backup power during a loss of off-site power scenarios.


Question: What are the requirements for nuclear license reactivation?

Identify the required procedural documentation for license reactivation in the space below.

On your own time, you and a friend (who has an active SRO license) came in and took a complete tour of the plant and Main Control Room.


Question: What are surety requirements in the context of nuclear licensing?

Connecticut does not have language that specifically states the following: 'establish by rule or regulation standards and procedures to ensure that the licensee will provide an adequate surety or financial arrangement to permit the completion of all requirements established by the agency for the decommissioning and reclamation of sites, structures and equipment used in conjunction with such licensed activity, in case the licensee should default for any reason in performing such requirements.'

Connecticut needs to provide an explanation or justification as to the reason for the omissions; or supply the missing information.


Question: total alkalinity measurement

In the context of the ZBG 6 well data, the total alkalinity (as CaCO₃) was recorded with a result of 0.000 mg/L.

On March 12, 2024, the water temperature at the ZBG 6 well was 20.200 degC.


Question: nuclear risk assessment methodology

Based on this review, the Frost hazard can be considered to be negligible.

Per the IPEEE Table 5-1, the principal effects of such events would be to cause a loss of off-site power (C4).


Question: What is the role of the Piping components in the Reactor Water Cleanup System?

Piping, piping components are described as having the intended function of being a Pressure Boundary.

Water Chemistry is critical for the aging management of various reactor systems, particularly in monitoring the chemical composition of coolant water.


Question: strainer head loss

Similarly, flow sweeps were also performed at the end of each test after adding the full load of chemical debris and allowing the head loss to stabilize.

Afterwards, at least one flow sweep was performed by measuring strainer head losses at several different flow rates.


Question: What is the significance of the GOTHIC computer code in nuclear licensing?

Through benchmarking and sensitivity studies, it is demonstrated that the use of the GOTHIC computer code for the post-LOCA containment pressure and temperature response analysis application produces results that are acceptable.

Comparison of containment vapor space and sump liquid temperatures generally shows good agreement, except for a peak temperature spike exhibited in the COPATTA results and the temperature difference for the sump liquid temperature early in the event.


Question: pressurizer components

The pressurizer is equipped with various components including a relief nozzle, spray nozzle, safety nozzle, manway, upper head, instrumentation nozzle, lifting trunnion, shell, lower head, electrical heater, heater support plate, support skirt, and surge nozzle.

The instrumentation used in monitoring reactor temperatures is critical for ensuring the safety and stability of nuclear operations.


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

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

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


Question: Entities responsible for radiological emergencies must ensure adequate planning and readiness.

Emergency planning function means a capability or resource necessary to prepare for and respond to a radiological emergency.

A holder of a license under this part for a non-power production or utilization facility, a holder of a license under this part for a small modular reactor or a non-light water reactor, or a holder of a combined license under part 52 of this chapter after the Commission makes the finding under § 52.103(g) of this chapter for a small modular reactor or a non-light-water reactor, shall follow and maintain the effectiveness of either an emergency plan that meets the requirements in § 50.160 or an emergency plan that meets the requirements in appendix E to this part and, for nuclear power reactor licensees, the planning standards of § 50.47(b).


Question: The NRC mandates that pressure tests must include the total designed working pressure for a specified duration.

All piping, including fittings, will be static pressure tested to 100% of its designed working pressure for 20 minutes.

The fluids handling system in New Mexico encompasses various pumps, meters, pipelines, fittings and connections, and will generally consist of polyethylene, PVC, fiberglass, steel, and stainless steel materials, which are used universally in ISR.


Question: what are the design requirements for nuclear facilities

Demonstrate functional design criteria via analysis, test, etc.; Evaluate operating, design and construction experience

Consensus codes and standards acceptable to NRC


Question: The NRC's evaluations include assessments of axial weld integrity to prevent issues such as cracking.

If triggered, EVT-1 visual inspection of the axial weld length and 3/4-inch of adjacent base metal on a 10-year frequency; EVT-1s look for evidence of cracking (IASCC) and manage irradiation embrittlement (IE).

CSA ribs and rings are not included as assessed components for screening review in Table 3-2 of Rev. 2.


Question: What are the stability and wind speed distributions at North Anna?

Table 2.3-10 presents the stability and wind speed distribution data at North Anna from 1978 to 1987, indicating various measurements across different seasons.

The NRC typically mandates that licensees must maintain strict documentation and adhere to the guidelines set forth in regulatory filings, ensuring compliance with safety standards.


Question: What are the reactor coolant system parameters for safety analysis?

These are the nominal plant conditions that were used to develop initial conditions for analyses at 0%, 30%, 70% and 102% of the stated power levels.

The maximum analysis core power was 2956 MWt (102% of 2988 MWt) with a corresponding NSSS power of 2968 MWt.


Question: low-low set instrumentation in nuclear safety

Relief and Low-Low Set (LLS) Instrumentation – one trip system inoperable

For the LLS functions, the SRV Fail to Reclose failure mode probability will be doubled to account for additional SRV cycles.


Question: What is the function of a makeup water system in nuclear reactors?

The Makeup Water System operates as a leakage boundary which involves components like sample sinks to maintain the required water chemistry and monitor for aging effects such as loss of material.

The screen is used as a filter and can experience fouling and flow blockage which may impact the overall system efficiency.


Question: What are NRC licensable materials?

Are any samples NRC licensable material?

Was ice present in the shipment container(s)?


Question: What is COBRA-FLX in nuclear safety analysis?

COBRA-FLX is a thermal hydraulic analysis code developed to evaluate critical heat flux and other safety-related simulations for nuclear reactor systems.

Table B-3 provides numerical data from various grid tests but does not specify the thermal properties of materials used in the simulations.


Question: The system consists of charging and letdown functions that help maintain the water level within the Reactor Coolant System pressurizer.

The Chemical and Volume Control System is capable of supplying water at the maximum test pressure specified to verify the integrity of the Reactor Coolant System.

During normal operation, the coolant then flows through a mixed bed demineralizer.


Question: What is stress corrosion cracking (SCC) in stainless steel components?

SCC in SS components is not an aging effect requiring management if (a) plant-specific OE does not reveal a history of SCC and (b) a one-time inspection demonstrates that the aging effect is not occurring.

Leakage of fluids through bolted connections (e.g., flanges, valve packing) can result in contaminants present in the insulation leaching onto the component surface or the surfaces of other components below the component.


Question: What analytical tools does the NRC utilize for modeling criticality excursions?

The NRC has appropriate analytical tools (e.g., PARCS-TRACE) for modeling the dynamics of PWR spent fuel pool criticality excursions with positive void reactivity.

Related references are included in the description link provided above.


Question: What are the requirements for evaluating the DNBR in a nuclear reactor license application?

Please provide the values of the calculated peak fuel temperature, fuel rim temperature, enthalpy, and enthalpy rise to show that the values meet the applicable limits for the REA analysis.

On page 16 of 25 of Attachment 1, the licensee states that “the report summarizing the results of the MPS3 AREA analysis for GAIA fuel are available for USNRC audit.”


Question: gamma spectrum analysis

GAMMA SPECTRUM ANALYSIS includes the procedure for identifying radioactive isotopes based on the gamma radiation emitted, utilizing analysis tools such as the Marinelli geometry for accurate measurements.

Sample identification is crucial for ensuring that the correct data is collected and analyzed in any nuclear safety assessments.


Question: What are the minimum cooling times required for various assembly average burnup ranges in PWR fuel?

For assembly average burnup between 42 and 43 GWd/MTU, the minimum cooling time required is specified in terms of enrichment and assembly configuration.

The assembly average burnup affects the efficiency of the reactor operation and influences the fuel cycle strategies.


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


Question: The company is seeking consent for the transaction which requires various governmental approvals and licenses.

In addition to the consents requested in the Application, the closing of the Transaction is conditioned on, among other things, the filing of certain applications and notices with, and receipt of relevant approvals, licenses or consents from applicable governmental authorities (collectively, "Regulatory Approvals").

Honeywell currently anticipates that the Regulatory Approvals be received, and the Internal Reorganization will close, in the second quarter of 2025 and the Spin by third quarter of 2025.


Question: The process to ensure that the probabilistic risk assessment models accurately reflect the actual operational status of the plant is crucial for risk management.

The configuration control program will ensure these plant changes are incorporated into the PRA models as appropriate.

The Entergy PRA maintenance and update processes are governed by procedures.


Question: What safety measures are taken to verify the cooling systems necessary in the event of an accident?

LHSI pump reliability evaluations were conducted in accordance with NRC staff requirements.

For the purpose of conducting start-up testing prior to evaluation of LHSI pump reliability, an engineering review of the accident analyses given in Chapter 15 was conducted to ascertain whether the consequences of any of the applicable accidents could be more severe when postulated to occur with the LHSI pumps assumed inoperable, under the following conditions: 1. None of the fuel in the core has produced any sensible heat.


Question: nuclear safety regulatory filings

The regulatory framework for nuclear safety includes stringent requirements for the licensing and operation of nuclear reactors to ensure public safety and environmental protection.

During the summer months, it is expected that rainfall will decrease, leading to potential drought conditions across several regions.


Question: NRC is evaluating the need to update guidelines regarding flood and seismic hazards at nuclear sites following recommendations from the Near-Term Task Force.

To gather information with respect to NTTF Recommendation 2.1, as amended by SRM associated with SECY-11-0124 and SECY-11-0137, and the Consolidated Appropriations Act, for 2012 (Pub Law 112-74), Section 402, to reevaluate seismic and flooding hazards at operating reactor sites.

The plant is directly across the river from the Department of Energy's Savannah River site (Barnwell County, South Carolina).


Question: high radiological risk activities

Accessing areas, such as beneath the reactor vessel or around the seal table (in-core instrument plate), during periods when in-core detectors could be operated. If the in-cores are stored and cleared or the keys to operate the in-cores have been secured and controlled as locked high radiation area (LHRA) keys by radiation protection, the in-cores will be considered non-operational.

Work with alpha contamination levels >1000dpm/100cm² or whole body entry into areas where alpha contamination levels meet the criteria for EPRI Level III alpha areas.


Question: The NRC mandates a testing schedule for tendon condition monitoring every eight years, extending the period past the usual five years.

In lieu of the 5-year frequency specified by IWL-2420, the next sampling of selected tendon CPM and free water testing for ANO-2 would be extended for 8 years and scheduled for 2028.

Additionally, for ANO-1 and ANO-2 the testing of CPM for corrosive ion and reserve alkalinity will not be tested on a 10-year frequency.


Question: What are the design and operating bases for nuclear facility cooling systems?

Design and operating bases, such as tornado parameters, ice glaze thickness, and winter probable maximum precipitation are statistics that, by definition and necessity, are based upon long-term regional records.

The proposed facility will only be installing standby and auxiliary power systems that will be used for emergency and backup purposes.


Question: nuclear licensing regulations

Under these regulations, a motion for reconsideration must demonstrate 'compelling circumstances, such as the existence of a clear and material error in a decision, which could not have been reasonably anticipated, that renders the decision invalid.'

The court noted that the NRC 'fail[ed] to identify any statute that authorizes it not to comply with NEPA on equitable grounds,' and in Powertech we explicitly did not decide 'whether, or under what circumstances, an NRC presiding officer should perform an Allied-Signal-style equitable analysis in the first instance upon finding a significant NEPA deficiency.'


Question: The NRC mandates thorough evaluations of radioactive isotopes in samples submitted for licensing assessment.

Americium-241 has a result of 0.00461 pCi/g with an uncertainty of +/-0.00550.

The Plutonium-239/240 sample had a result of -0.00387 pCi/g with an uncertainty of +/-0.00838, indicating variability in the analysis.


Question: What is the population density requirement for nuclear site licensing?

This places the site in proximity Category 3 (having one or more cities with 100,000 people, and fewer than 190 persons per square mile within 50 miles of the plant) (NRC 2013).

Approximately 83.2 percent of the employees (contract workers and employees) live in Alabama, 7.0 percent live in Tennessee, and the remaining 9.8 percent is distributed across 35 other states.


Question: The plant site assessments included considerations for seismic events in the context of dam safety evaluations.

During the 1997-1998 reassessment, the maximum plant site elevations at WBN with dam safety modifications for the different postulated combinations of seismic dam failures coincident with floods were as follows:

An unsteady flow model was added for the Fort Loudoun – Tellico complex.


Question: subsequent license renewal

The application submitted by NextEra Energy Point Beach, LLC, seeks approval for the subsequent license renewal of Point Beach Nuclear Plant, Units 1 and 2, which involves a comprehensive review to ensure safety and regulatory compliance.

The administrative procedures require that all documentation be submitted electronically to enhance processing efficiency and transparency.


Question: What happens when a spray valve fails open in a pressurizer?

When a spray valve fails open, it causes an increase in the flow of water into the pressurizer, which can lead to a decrease in pressurizer pressure due to the increased cooling effect on the vapor space.

The importance of maintaining reactor safety protocols is paramount, and regular inspections and maintenance schedules are put in place to ensure operational integrity.


Question: The measured activities of various nuclides such as K-40 and Co-60 are essential for understanding the safety metrics of nuclear processes.

The identified peaks were analyzed for their energy levels and uncertainties to ensure compliance with standard safety protocols.


Question: What are the components of a safety subsystem in nuclear reactors?

The Safety Subsystem Block Diagram includes components such as Safety Monitors, Local Control, and a Data Processing Module, which are essential for ensuring the operational safety of nuclear systems.

35-50 Conductors per DPM are required for electrical connections, facilitating data transmission and control across various systems.


Question: what is detailed fire modeling in nuclear safety

The application of fire modeling technology is used in the FPRA to translate a fire initiating event into a set of consequences (fire induced failures). The performance of the analysis requires a number of key input parameters. These input parameters include the heat release rate (HRR) for the fire, the growth rate, the damage threshold for the targets, and response of plant staff (detection, fire control, fire suppression).

Consensus modeling approach is used for Detailed Fire Modeling, and it is concluded that the methodology for the Detailed Fire Modeling task does not introduce any epistemic uncertainties that would require sensitivity treatment.


Question: What is the employee commuting population in the RBS plume exposure pathway EPZ?

The employee commuting population are those who live outside of the RBS plume exposure pathway EPZ and commute to jobs within the plume exposure pathway EPZ.

The school population in the RBS plume exposure pathway EPZ are located in East Baton Rouge Parish, East Feliciana Parish, Pointe Coupee Parish, and West Feliciana Parish, which include the following 28 schools/preschools/day care/colleges:


Question: What is the significance of activity measurements in nuclear material regulation?

The activity of a nuclide, measured in pCi per gram, indicates the rate of decay and is crucial for assessing the safety and regulatory compliance of nuclear materials.

The NRC provides guidelines on the measurement of emissions for non-nuclear facilities to ensure environmental safety.


Question: What is the function of structural pressure barriers in nuclear reactors?

Steel Elements: Thermowells (torus) have a primary role as a Structural Pressure Barrier and are constructed from Stainless Steel in a treated water environment, crucial for maintaining the integrity of nuclear systems.

Steel Elements: Drywell - support skirt serves as Structural Support and is made of Carbon Steel, particularly effective in concrete environments and does not require aging management.


Question: The effectiveness of the Aging Management Program is evaluated regularly to ensure the integrity of the containment structures remains intact.

Therefore, the DCPP ASME Section XI, Subsection IWL AMP has been effective in ensuring that the DCPP concrete containment will continue to operate within its licensing basis.

The most recent Unit 1 and 2 IWL inspections were completed in December 2021.


Question: The monitoring of the isolation condenser tubes for potential cracking is essential to ensure the reactor's safety.

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1) program is used to manage cracking of the stainless steel isolation condenser tubes and carbon or low alloy steel with stainless steel cladding tube sheet in the Isolation Condenser System.

This Item Number is not listed in NUREG-2192.


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

Energy (keV) is a critical parameter measured in nuclear safety assessments, as it provides insights into the characteristics of radiological sources and helps in evaluating the safety of nuclear facilities.

The values listed under Original Area represent the net counts detected during the analysis of various peaks, with uncertainties provided for accuracy in measurement.


Question: The NRC mandates that emergency response plans include provisions for appropriate personnel availability during a radiation incident.

Added steps to develop an ERO relief roster and call in required personnel.

Changed SDS Mass Notification System to Alert Media.


Question: nuclear cooling systems

Cost Comparison of Dry-Type and Conventional Cooling Systems for Representative Nuclear Generating Plants, USAEC Report TID-26007, pp. 81-89.

A report by the Scottsboro City Planning Commission in cooperation with the Alabama State Planning and Industrial Development Board, March 1967, pp. 44.


Question: What is the activity of nuclides in a nuclear safety report?

The NRC's licensing process includes rigorous safety assessments and public involvement to ensure the safe operation of nuclear facilities.


Question: What measures are taken when the spent fuel pool (SFP) water level is high during refueling?

If water level approaches SFP vents, THEN: - **SHUTDOWN** Fuel Handling Area Exhaust Fans. - **DETERMINE** cause of HI level. - **ADJUST** SFP level using 0-SOI-78.01, SPENT FUEL POOL COOLING AND CLEANING SYSTEM.

Changes in Aux Bldg and Rx Bldg ventilation will affect SFP and Refueling Cavity levels.


Question: guidance for Class II research and test reactors

Establish guidance and training for recording and approving the RTR inspection hours to specific CACs [cost activity codes].

The staff enhanced the instructions for creating, recording, and approving inspection hours and completed the revisions to the guide on July 2, 2024.


Question: What is the method used for analyzing Strontium-90 in nuclear licensing processes?

Strontium-90 is reported with a qualifier 'U', indicating that the value is below the reporting limit of 0.400 pCi/g.

The Collector is noted as the Client, and the sample was collected on 05-JUN-19, which falls within the documented timelines.


Question: The NRC mandates various operational considerations for the inservice inspection and maintenance of reactor systems.

**Section 9.3: Operation considerations:** a)  Inservice inspection, replacement and maintenance

**Section 9.2: Additional manufacturing and construction considerations** (quality control and quality assurance, testing of prototype and production isolators, construction assurance).


Question: The NRC mandates that license applicants must submit their request for licenses involving the use of special nuclear material nine months prior to starting construction.

Paragraph 70.21(f) requires that an application for a license to possess and use SNM material for processing and fuel fabrication, scrap recovery, or conversion of uranium hexafluoride, or for the conduct of any other activity which the Commission determines will significantly affect the quality of the human environment, must be filed nine months prior to construction and must be accompanied by an Environmental Report required under Part 51.

Office to ensure appropriate physical protection of irradiated reactor fuel in transport.


Question: The NRC monitors the quantities of radionuclides to ensure they remain within safe limits.

The table presents the activity levels of various radionuclides in different tank numbers, measured in units of activity such as curies or becquerels.

Safety assessments are critical in ensuring that all reactor operations adhere to established guidelines and standards.


Question: What is a particle accelerator?

A particle accelerator is a device that imparts kinetic energy to subatomic particles by increasing their speed through electromagnetic interactions.

All fusion devices use subatomic particles (i.e., plasma).


Question: The NRC mandates that structural integrity evaluations are performed to check for loss of material due to corrosion.

Loss of preload due to self-loosening is a concern for structural bolting, which is crucial for maintaining structural support.

Support members; welds; support anchorage to building structure can involve various materials and configurations depending on the structural design requirements.


Question: minimum detectable concentration

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


Question: What are the Docket Numbers for the Limerick Generating Station?

- **Renewed Facility Operating License Nos.** NPF-39 and NPF-85


Question: The NRC requires that license holders implement emergency plans that include coordination with FEMA to ensure safety during nuclear incidents.

§ 50.160 Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities “A holder of a combined license issued under parts 52 or 53 of this chapter before the Commission has made the finding under § 52.103(g) or 53.1452(g) of this chapter, as applicable,…” invoking § 50.160(b)(1)(iv)(A)(2) requires licensees to be capable of implementing their approved emergency response plan in conjunction with their safeguards contingency plan. Radiological sabotage events are typically factored into EPZ determinations by considering consequences to be bounded by LBEs and by crediting protection against the DBT in reducing the likelihood of a release.

§ 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.”


Question: 10-mile emergency planning zone

The 10-mile emergency planning zone (EPZ) surrounding River Bend Station is critical for ensuring the safety and preparedness of the local community in case of a nuclear incident.

The seismic evaluation conducted for River Bend Station found that the structural integrity of the reactor building remains intact under design basis earthquake scenarios.


Question: The evaluation of license applications is critical for ensuring compliance with safety regulations.

The licensing process involves a comprehensive review of applications submitted by entities seeking to operate nuclear facilities according to established safety standards.

Operators of nuclear plants must also conduct regular training sessions for their personnel to ensure optimal performance and safety.


Question: ice basket structural integrity

The ice basket connections are analyzed to ensure structural integrity during all design load combinations of dead weight, 1/2 Safe Shutdown Earthquake, Safety Shutdown Earthquake and Design Bases Accident.

The stresses on the 1/2 inch diameter pin are tabulated in Table 6-84.


Question: RLBLOCA analysis results

For a simultaneous coverage/confidence level of 95/95, the UTL values, [ ] are a PCT of 1680°F, an MLO of 4.33 percent, and a CWO of 0.040 percent.

A summary of the major input parameters for the demonstration case is provided in Table 4-5.


Question: The NRC requires comprehensive surveys and assessments for the plant facilities prior to any significant modifications or actions.

Dison Braden A., Heather Bass, Hunter B. Johnson, Brittney Carnell, Katie Weis and Breiding Katie. 2022. A Phase I Archaeological Survey of Previously Unsurveyed Areas within the TVA’s BFN Plant in Limestone County, Alabama. Prepared by Tennessee Valley Archaeological Research, Huntsville, Alabama. Prepared for TVA, Knoxville, Tennessee. Classified report, not publicly available.

Tree Clearing Project in Limestone County, Alabama. Prepared by Tennessee Valley Archaeological Research, Huntsville, Alabama. Prepared for Tennessee Valley Authority.


Question: nuclear regulatory filing procedures

The NRC has made some, or all, of the requested records publicly available through one or more of the following means: 1. www.nrc.gov 2. public ADAMS, https://www.nrc.gov/reading-rm/adams.html 3. microfiche available in the NRC Public Document Room; or FOIA Online, https://foiaonline.gov/foiaonline/action/public/home.

You will receive a refund for the amount indicated.


Question: Licensees must conduct thorough assessments of safety-related systems as part of their regulatory responsibilities.

The safety review involves a systematic assessment of operational characteristics and safety-related systems to ensure compliance with regulatory standards.

Recent advancements in renewable energy technologies have significantly impacted the energy production landscape in the United States.


Question: What is the process for notifying about adverse effects on historic properties during nuclear project construction?

Based on the above coordination with 36 CFR 800.5, the NRC has made a finding of Adverse Effect for this undertaking.

This historic property is a multi-component prehistoric and historic-era artifact scatter with both surface and sub-surface contexts (see figure 31 of Enclosure 2). Identified artifacts include one Late Archaic and one non-diagnostic projectile point, other worked stone tools and debitage, and various metal and glass historic items.


Question: Preventive measures against channel meandering and sediment buildup at the river intake structure are critical.

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

Plant Farley is a nuclear power plant located in Alabama, generating substantial electric power for the region.


Question: accidental release study of radioactive liquids

An accidental release study identifying: the nature of release and in-situ conditions (constituents, hydrogeologic data, groundwater model from 2.4.12), appropriate literature range of soils absorption coefficients for site soil types, and bounding conditions (constituent attenuation, velocities, transport parameters).

Information provided in the supplement will be used to validate the calculation results and modeling information presented in the CPA.


Question: The NRC expects the licensees to conduct advanced computational fluid dynamics (CFD) studies to validate safety analyses.

To quantify the conservatism of assuming the NuScale performed a detailed, transient, three-dimensional computational fluid dynamics (CFD) evaluation.

The existing CFD analysis results confirm that this conservative assumption results in approximately 500 percent reduction in margin in the limiting case.


Question: What is the function of the Reactor Water Cleanup System?

The Reactor Water Cleanup System is responsible for ensuring that the reactor coolant water is maintained at acceptable levels of purity by removing impurities and providing filtration and ion exchange. It plays a crucial role in sustaining safe reactor operations.

The Main Turbine and Auxiliaries System is designed to convert the energy from steam into mechanical energy, which drives the electrical generators, thus providing the necessary power for the plant's operations.


Question: The NRC evaluates the safety margins for boiling water reactors when assessing new licensing applications.

The COBRA-FLX critical heat flux correlation is an essential tool for evaluating the thermal-hydraulic performance of reactor systems under various operational conditions.

Framatome Inc. provides innovative solutions for nuclear power plants and contributes to sustainable energy efforts worldwide.


Question: nuclide identification in regulatory filings

The identified nuclides include K-40, PB-212, and BI-214, with their respective energies, yields, activities, and uncertainties documented in detail.

Energy Tolerance is set to 1.000 keV, and the nuclide confidence index threshold is established at 0.30, with errors quoted at 2.000 sigma.


Question: What is the purpose of a radiological hazard assessment in nuclear facilities?

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

Instructions to verify proper operation of 'B' SFPC pump and confirm adequate pump/motor vibrations and bearing temperatures, following motor and pump bearing replacement on October 31, 2024.


Question: What are the bounding site-specific plant parameters for an Early Site Permit?

Table 1.9-1, Bounding Site-Specific Plant Parameters Envelope, provides a summary listing of site characteristics that have been established by analyses presented throughout the SSAR.

It was necessary to assume certain design parameters in order to assess site characteristics.


Question: thermal preservation acceptance criteria

Client must contact an ACZ Project Manager if analysis should not proceed for samples received outside of their thermal preservation acceptance criteria.

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


Question: current uranium mining facilities near CUP


Question: What is the significance of the Heat Exchanger in nuclear power plants?

Heat Exchanger – (Drain Cooler) is a key component that serves as a leakage boundary and is made of Carbon Steel, operating in air under uncontrolled external conditions, and is significant for managing the loss of material.

The table outlines various components and their intended functions including bolting integrity and external surfaces monitoring for mechanical components.


Question: What is the due date for the Final Safety Analysis Report submission for Idaho State University's reactor?

Idaho State University consents to this order and acknowledges the due date of 5/29/2026 for submission of an updated version of the Final Safety Analysis Report for reactor license R-110 (Docket No. 50-284).

Please accept this communication as the formal response from Idaho State University to the document titled “ORDER MODIFYING LICENSES AND ESTABLISHING INITIAL FINAL SAFETY ANALYSIS REPORT UPDATE DUE DATES CONSISTENT WITH THE NON-POWER PRODUCTION OR UTILIZATION FACILITY LICENSE RENEWAL RULE (EPID L-2024-NFC-0000)” (ML24354A206) for Idaho State University (50-284).


Question: What is the importance of soil sampling in nuclear facility assessments?

Soil sampling is a critical component of environmental assessments conducted near nuclear facilities to understand the potential impact of operations on soil and groundwater.

The project management team must evaluate the timeline for the completion of the reactor construction and ensure that all regulatory requirements are met by the deadline.


Question: fuel loading positions in nuclear reactors

The CY-MPC 24-Assembly Basket Fuel Loading Positions are laid out in a grid format showing the arrangement of fuel rods within the assembly.

The reactor core is surrounded by a containment structure designed to prevent the release of radioactive materials into the environment.


Question: What are boration flow paths in nuclear safety systems?

See Figures D.3-6 and D.3-8 for possible boration flow paths depending on SLCS device location, which all have different transport times (i.e. transport time delays) through the fuel assemblies.


Question: M518 cladding properties in nuclear licensing

Figure 4-1: M5₁₈Cladding Slow Ramp Rate Rupture Strain Curve and Data

The cooling system is designed to handle both normal and emergency conditions efficiently.


Question: What are the requirements for evaluating Pump Vibration in nuclear safety tests?

Complete Steps 19 and 20 to evaluate if Pump Vibration meets Acceptance Criteria.


Question: neptunium-237 in soil samples

The following Analytical Methods were performed:


Question: What is the purpose of the Surveillance Frequency Control Program in nuclear plants?

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The Notes exclude the source range neutron flux detectors and reactor trip breaker and RCP breaker position indications from the CHANNEL CALIBRATION.


Question: Design Bases Assurance Inspection

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC) Region II staff will conduct a Design Bases Assurance Inspection (DBAI) at your plant Hatch during the weeks of April 26, 2021 and May 10, 2021.

During a telephone conversation on January 11, 2021, with James Love we confirmed arrangements for an information-gathering site visit and the two-week onsite inspection.


Question: What is the yield percentage for the nuclide CS-137?

The yield of CS-137 is 85.12% as noted in the regulatory filing.

The uncertainty quoted at 2.000 sigma provides a confidence measure for the reported activities.


Question: Nitrogen Storage Tank Foundation

The Nitrogen Storage Tank Foundation is a reinforced concrete slab that provides structural support for the Nitrogen Storage Tank.

The aging management review results for these components are provided in Table 3.5.2-23, Condensate Water Storage Tanks Foundations, Trenches, and Tunnels - Summary of Aging Management Evaluation.


Question: The NRC mandates that licensees evaluate the aging effects on components such as earthen water-control structures.

Consistent with NUREG 2191. The Inspection of Water- Control Structures Associated with Nuclear Power Plants program (B.2.1.34) will be used to monitor loss of material in components associated with the intake channel, north bank of the cool water channel, and the south dike of the cool water channel are exposed to environments of air – outdoor, water – flowing, and water - standing.

Consistent with NUREG-2191. The ASME Section XI, Subsection IWF program (B.2.1.30) will be used to manage loss of mechanical function of the steel constant and variable load spring hangers, guides, stops and sliding surfaces for ASME Class 1, 2, 3, and MC equipment supports exposed to air - indoor uncontrolled or air-outdoor.


Question: What is the significance of maximum wave height in flood calculations for nuclear plants?

The maximum wave height of 5.6 ft. is controlling.

The settling of particles carried downstream from the Marsh Creek and Hoopes Reservoir dam breaks are calculated to address the potential for suspended sediments from dam failures that may adversely affect the operation of the intake structure at the new plant.


Question: plant drainage system

The Plant Drainage System consists of piping and components that serve as a leakage boundary, ensuring containment of waste water within controlled environments.

Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24) is a key part of managing aging effects, which may include loss of material in various parameters.


Question: nuclear reactor licensing procedures

Ngola Otto
Project Manager, Licensing Projects Branch
Division of Operating Reactor Licensing
Office of Nuclear Reactor Regulation
US Nuclear Regulatory Commission


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory steps required to ensure that the design, construction, and operation of nuclear facilities meet safety standards set by the NRC.

During the construction phase, contractors must comply with local building codes and safety regulations that are independent of the nuclear licensing requirements.


Question: What is wind stability class in nuclear safety analyses?


Question: What is the maximum overpower trip point in a nuclear reactor?

Maximum overpower trip point assuming all individual errors are simultaneously in the most adverse direction (% of rated power) is 118.

Feedwater pressure with a small correction on enthalpy is ±60 psi.


Question: nuclear reactor rings

The reactor is divided into several overlapping regions known as rings, which are labeled Ring A, Ring B, Ring C, and Ring D, with specific measurement intervals assigned to each.

The safety assessment conducted showed that all systems were functioning within the expected parameters to ensure the integrity of the facility.


Question: Licensees are required to define their methodology for evaluating Class D valves as part of their compliance efforts.

In Regulatory Issue Summary (RIS) 2011-13, the NRC staff discusses an acceptable approach for licensees to address Class D JOG valves.

10 CFR 50.55a(b)(3)(ii) requires MOVs to be tested per the ASME OM Code and must have a program that periodically demonstrates the capability of the MOV to satisfy its design-basis safety function.


Question: Material traceability is not mandatory for components classified as important to safety Category B items, especially concerning their compliance with contract documents.

The proposed code alternative for NM2.15 is only for the material traceability of the EOS-HSM-SC components. This is because horizontal storage modules (HSMs) are not safety-related components as outlined in AISC N690-18, and as per Section 14.2 of the UFSAR, important to safety (ITS) Category B items do not require material traceability.

The TS have been revised as described in the response to Clarification S-4.


Question: The regulatory body mandates that every licensee conducts a thorough safety review to ensure compliance.

The NRC evaluates license applications by reviewing safety analyses, technical specifications, and operational procedures to determine the suitability of a reactor design for licensing.

During the reactor startup procedures, operators must ensure that all systems are functioning correctly before proceeding with power ascension.


Question: What is the significance of maximum pressure measurements in nuclear reactors?

The maximum pressure values for each compartment are critical for understanding the operational limits and safety margins within the reactor design.

The elements listed in the table are fundamental components that influence the reactor's thermal and mechanical performance.


Question: The requirement is that each reactor coolant pump must maintain a sufficient cooling flow during operation.

**STANDARD:** Operator verifies RCP #23 CCW flows are acceptable using flow indicators 2-CFA-453, 2-CFI-457, and 2-CFI-453  SAT: ☐ UNSAT: ☐

**CUE (If required):** Seal cooling has not been lost to any RCP during event.


Question: nuclear safety analysis methodology

The safety analysis methodology for nuclear reactors involves a systematic approach to identifying potential hazards and assessing their implications for reactor operation and safety compliance.

The operational schedule of the nuclear facility outlines maintenance periods and operational benchmarks for optimal energy production.


Question: Complete training and experience in radiation therapy must include supervision by a qualified medical physicist who meets NRC requirements.

Hold a master’s or doctor’s degree in physics, medical physics, other physical science, engineering, or applied mathematics from an accredited college or university; and has completed 1 year of full-time training in medical physics and an additional year of full-time work experience under the supervision of an individual who meets the requirements for an authorized medical physicist for the type(s) of use for which the individual is seeking authorization.


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

Technical safety analysis is conducted to assess the potential risks and safety implications of operating nuclear facilities under various scenarios and conditions.

The email communication discusses ongoing cooperation and litigation matters among NRC staff and external counsel.


Question: Check the current status of the licenses for Reactor Operators and Senior Reactor Operators as of the specified date.

Determine if the RO’s and SRO Licenses are ACTIVE or INACTIVE as of 0700 on August 8, 2022.

The safety protocols at the Watts Bar Nuclear Plant ensure the protection of the environment and public health.


Question: nuclear reactor safety systems

Nuclear reactor safety systems are designed to prevent accidents and mitigate the consequences of any that do occur, incorporating multiple layers of protection to ensure safe operation under all conditions.

The graphical presentations provided in this report highlight various aspects of the facility's layout and operational protocols that contribute to overall performance.


Question: Risk-Informed Completion Times

The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times – RITSTF Initiative 4b" (ADAMS Accession No. ML18183A493).

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) is requesting an amendment to the Technical Specifications (TS) of the River Bend Station Unit 1 (RBS).


Question: oxidation reactions of stainless steel and Zircaloy

the rate of [stainless] steel oxidation is small relative to the oxidation of Zircaloy at temperatures below 1400 K. At higher temperatures and near the [stainless] steel melting point, the rate of [stainless] steel oxidation exceeds that of Zircaloy, and states that 'the rate of reaction for [stainless] steel exceeds that of Zircaloy above 1425 K.'

National Research Universal’s (“NRU”) thermal-hydraulic experiments were conducted in the early ’80s. NRU’s thermal-hydraulic experiments were conducted with single bundles of full-length Zircaloy cladding, driven by low-level fission heat: an amount to simulate decay heat.


Question: The results indicate that the U-238 levels were evaluated with specific attention to their detection limits.

U Compound/analyte was analyzed, peak not identified and/or not detected above MDC

Radionuclide U-238 was tested, and the samples were processed under specific laboratory conditions.


Question: The surface dose rate limits for STORAGE CASKS should be compliant with the specified regulatory thresholds.

LCO 3.3.1 The maximum surface dose rates for the STORAGE CASK, Reference Figure A3-1, shall not exceed the following limits: - PWR and BWR – 95 mrem/hour gamma and 5 mrem/hour neutron on the vertical concrete surfaces; and - PWR and BWR – 450 mrem/hour (neutron + gamma) on the top.


Question: loss of coolant accident

Unit 1 has experienced a Small Break Loss of Coolant Accident

The Shift Foreman directs you to perform EOP FR-C.1 starting at Step 10, 'Check Intact Steam Generators'.


Question: nrc licensable material

4) Are any samples NRC licensable material?


Question: The current models for the Reactor Protection System include multiple functions that are critical for responding to significant initiating events.

Current RPS logic models five SCRAM functions and credits these functions for all initiating events. These signals are: High reactor vessel pressure, Low reactor vessel water level, High reactor vessel water level, Turbine Control/Stop Valves, SDV Level.

A pre-existing large containment isolation failure will be added as a conservative surrogate for unmodeled PCIV instrumentation and pathways. This is a matter of model maintenance and does not introduce new methods.


Question: The NRC evaluated the impact of revisions to the guidelines for quality assurance programs within the Nuclear Development Quality Assurance Manual.

The primary purpose of this update to the Nuclear Development Quality Assurance Manual (NDQAM) is to incorporate organizational changes noted below and to apply the NDQAM to radioactive material transportation packaging as defined and allowed by 10 CFR 71 Subpart H.

These administrative type changes were made throughout the NDQAM and are not noted below.


Question: nuclide identification

Nuclide Name | Id | Energy (keV) | Yield (%) | Activity (pCi/grams) | Activity Uncertainty
| K-40 | 0.95 | 1460.75 * | 10.67 | 5.23E+00 | 6.28E-01 |


Question: The regulation of enrichment levels for PWR fuel assemblies is critical for safety and operational efficiency.

The average enrichment for various types of PWR fuel assemblies is essential for ensuring compliance with regulatory standards.

General strategies for improving maintenance efficiency can lead to significant cost savings.


Question: What are the implications of changing forest management policies?

Changing forest management policies & practices

Human health threats and well-being can be affected by various factors.


Question: The NRC evaluates analyses pertaining to radionuclides such as Nickel-63 during the licensing process.

Nickel-63 is a radioisotope that is often measured in units such as pCi/g to assess its levels in samples.

The analytical methods performed on the samples were carried out by qualified analysts on specific dates.


Question: nuclear safety analysis

Nuclear safety analysis is a systematic examination of the risks associated with the operation of a nuclear facility, focusing on identifying potential hazards and assessing their likelihood and consequences.

The reactor cooling system plays a crucial role in maintaining operational temperatures, but it must be thoroughly evaluated during the design phase to ensure efficiency and reliability.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor in the United States is governed by the regulations of the Nuclear Regulatory Commission (NRC), which ensures that safety and environmental standards are met before operation.

The Watts Bar Nuclear Plant is an important facility that contributes to the energy grid by providing a stable output of electrical power.


Question: avian species near nuclear sites

Species identified along the designated sampling points.


Question: Plutonium isotopes measurement

The measurement of Plutonium-239/240 indicates a concentration of 1.89 pCi/g with an uncertainty of +/-0.204 pCi/g.

The range of the measurements reported includes various samples taken at different times and uncertainties associated with each of the analyzed isotopes.


Question: JPM validation checklist

When the above steps are completed for this and other JPMs to be run concurrently then validate, if not previously validated, the concurrently run JPMs using the JPM Validation Checklist.


Question: nuclear regulatory commitments

The regulatory commitments outlined in this document ensure that the licensee will adhere to specific safety protocols and operational standards mandated by the Nuclear Regulatory Commission (NRC).

The analysis of reactor coolant systems indicates no significant degradation during the past operating cycles, thus ensuring the overall integrity and performance of the system.


Question: nuclear regulatory filings

The NRC requires detailed technical safety analysis as a part of the licensing process for new nuclear reactor designs.

The table shows the number of vessels traveling to oil/petroleum/chemical berths on the Delaware River in each state from 1995 through 2008.


Question: nuclear licensing enforcement policy

Failure to conduct your program in accordance with NRC regulations, license conditions, and representations made in your license application and supplemental correspondence with NRC will result in enforcement action against you.

An environmental assessment for this action is not required since this action is categorically excluded under Title 10 of the Code of Federal Regulations (10 CFR) Section 51.22(c)(14).


Question: What does saturation deficit mean in meteorological studies related to nuclear safety?

Saturation deficit is measured in grams of water per kilogram of dry air, providing an indication of the amount of moisture that the air can still hold at a given temperature and pressure.

Wind direction significantly affects how temperature gradients develop, influencing overall atmospheric conditions.


Question: A report detailing the inspections performed for each Steam Generator must be submitted after the reactor enters MODE 4.

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, "Steam Generator (SG) Program." The report shall include:

The NRC also outlines the requirements for ensuring that all safety systems are functional during operation and maintenance activities.


Question: How can the number of Local Power Range Monitor (LPRM) inputs to the Average Power Range Monitor (APRM) be ascertained?

“To determine the number of LPRM Detectors feeding that APRM Channel, place METER FUNCTION SWITCH to COUNT and divide the meter indication by five.”

Prior to bypassing the LPRM, how is the number of LPRM inputs to APRM determined?


Question: The NRC evaluates the impact of system failures, including how pressure variations influence core performance and safety measures.

The closure of the TCVs causes a total steam block, which leads to a rapid pressure increase in the reactor vessel.

The calculated APRM, core average heat flux, scram insertion, steam dome pressure, recirculation flow and simulated narrow RPV water level of the nominal case are shown in Figure C.7-2 through Figure C.7-5.


Question: The areas measured for each peak during the spectroscopic analysis include the original area and the subtracted area of the detected signals.

The energies of specific peaks identified during the spectroscopic measurement are critical for accurate data interpretation and analysis.

The NRC has established guidelines for the management of radioactive waste and the importance of ensuring public safety in the disposal process.


Question: potable water system in nuclear plants

The Potable Water System is a plant-shared system which supplies potable water to various areas in the plant. This water source is supplied by the city of Athens, Alabama. Backflow preventers are installed at each cross connection to other systems to protect the potable water supply from possible contamination due to backflow.

The aging management review results for these components are provided in Table 3.3.2-7, High Pressure Fire Protection (Diesel Driven Pump) System - Summary of Aging Management Evaluation.


Question: The system should be monitored to ensure that seal injection flow does not drop below the established minimum levels.

Normal range of seal injection flow is between 8 and 13 gpm per RCP with a minimum allowed flow of 6 gpm.


Question: What is the Small Modular Reactor (SMR)?

SMR stands for Small Modular Reactor, which refers to a type of nuclear reactor that is smaller in size and designed for modular construction.

The Code of Federal Regulations (CFR) outlines regulations governing the operations of various industries.


Question: The design of sedimentation basins must be overseen by a licensed professional engineer.

Sedimentation basins shall be designed by a registered professional engineer, and installed under the direct supervision of a registered professional engineer.

The Permittee shall, upon request of the Agency, submit within a reasonable time the information and reports that are relevant to compliance with this Chapter, including the Plan, inspection reports, annual reports, original laboratory sheets from analyses conducted on the waste stream, and BMP plans and specifications.


Question: What is the energy level of the nuclide PA-228?

The nuclide PA-228 has an energy of 93.35 keV with a yield of 35.00%.

The energy levels of nuclides are often measured in keV, which allows for safe handling during regulatory assessments.


Question: What does nuclide mean in nuclear licensing?

A nuclide is a distinct nuclear species characterized by its number of protons and neutrons, which defines its stability and energy properties.

Errors quoted at 1.000sigma may indicate a need for further investigation or reassessment of the data collected during the analysis.


Question: What are the concentrations of various elements measured at the sampling locations?

Fluoride Concentration in mg/Kg, Nitrate Concentration in mg/Kg, and Uranium Concentration in mg/Kg are measured at various soil sampling locations.

The IDF is defined as 'Not available', indicating no data is present for certain sampling points.


Question: The NRC imposes strict standards for the analysis of radioactive materials during quality control processes.

The quality control process involves measuring various parameters such as batch number and sample information.


Question: average dry bulb temperature


Question: groundwater elevation measurements

In addition, the piezometer levels measured approximately at an elevation of 680 ft above msl in the Kansan till over a three-year period of observation in the late 1970s.

Some groundwater in the upper glacial drift deposits are discharged into streams from springs present within the general vicinity of the CPS Facility and Clinton Lake.


Question: nuclear licensing quantity

To receive payment, the contractor shall prepare invoices in accordance with NRC’s Billing Instructions.


Question: What is the importance of benchmarking in assessing the integrity of BWR pressure vessel components?

A benchmarking exercise was performed to validate the PROMISE software against the VIPERNOZ Version 1.1 software code.

Analyses were performed for several combinations of cyclic stress and numbers of load cycles.


Question: What actions are required for inoperable primary containment air locks?

With one or more air locks inoperable for reasons other than those described in Condition A or B, Required Action C.1 requires action to be immediately initiated to evaluate containment overall leakage rates using current air lock leakage test results.

These doors are to be verified locked closed by use of administrative controls.


Question: security-related information regulations

However, the enclosed report contains Security-Related Information, so the enclosed report will not be made publicly available in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390(d)(1).

Louis J. McKown, II, Chief Projects Branch 5 Division of Operating Reactor Safety.


Question: vapor activity monitoring

Table 11-8 details the maximum volume control tank activities, showing the vapor activity in curies for various isotopes such as Kr-85 and Xe-133, along with their stripping fractions with WGS operating and without WGS.

The volume control tank purge rate is specified to be 0.7 scfm, ensuring effective management of the system.


Question: NRC review process for nuclear applications

These estimates are based on the NRC staff’s initial review of the application and they could change, due to several factors including requests for additional information, unanticipated addition of scope to the review, and review by NRC advisory committees or hearing-related activities.

If you have any questions, please contact me.


Question: PWR pressurizer component evaluation

The current ASME Code Section XI examination requirements for the PWR pressurizer components evaluated in this report are summarized in Table 2-1.

Section 2.2 discusses previous projects related to vessel shell welds that are similar in nature to the studies covered in this report.


Question: How is the average enrichment of fuel assemblies measured?

Initial Assembly Avg. Enrichment wt % 233U (E) measures the concentration of 233U in the fuel assembly, expressed as a weight percentage.

The NRC evaluates the thermal performance of reactors to ensure safety and efficiency, focusing on design, operation, and maintenance practices.


Question: percentage frequency of wind direction and speed

Percentage Frequency of Wind Direction and Speed (mph) During May, Vicksburg, MS is detailed in Table 2.3-23, showing various directions and associated average speeds.

Data from NCDC, 1997-2001 (Reference 3).


Question: emergency core cooling system components

A totality passive system with one accumulator per loop. Evaluation based on one spilling accumulator.

Fails to start the centrifugal charging pump.


Question: The NRC's regulations specify what environmental aspects must be considered during license renewal evaluations.

The NRC has adopted regulations in 10 C.F.R. Part 51 that address the scope of environmental reviews for license renewal applications.

Many factors contribute to the overall safety of a reactor, including the design and operational parameters.


Question: What is the purpose of a Draft Environmental Assessment in nuclear projects?

The Draft Environmental Assessment for the Palisades Restart Project evaluates the potential environmental impacts of the proposed operation of the Palisades Nuclear Plant after its temporary shutdown.

The email was sent by Jennifer Davis to the PalisadesRestart-EnvDocsPUBLICem Resource, indicating the message was created and received on January 31, 2025.


Question: categorical exclusion from environmental assessment

Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9).

DENC requests approval of this LAR by October 31, 2024, with implementation in conjunction with MPS3 fuel cycle 24 (currently planned for the spring of 2025).


Question: What is a BWR reactor?

A Boiling Water Reactor (BWR) is a type of nuclear reactor that uses boiling water to produce steam, which drives a turbine to generate electricity.

The state has regulatory requirements for the assessment and management of hazardous materials to ensure safety and compliance.


Question: tornado hazard assessment

ASCE Hazard Tool tornado data for DNPS was used to develop the site-specific tornado hazard curve.


Question: common cause failure management

Per Entergy Operations, Inc. (Entergy) standards, for emergent conditions where the extent of condition is not performed prior to entering into the Risk Management Action Times (RMAT) or the extent of condition evaluation cannot rule out the potential for common cause failure, common cause RMAs are expected to be implemented to mitigate common cause failure potential and impact.

Contact the Transmission System Operator (TSO) to determine the reliability of offsite power supplies prior to entering a RICT and implement RMAs during times of high grid stress conditions, such as during high demand conditions.


Question: The NRC's evaluation process ensures adequate safety measures for environmental impacts associated with nuclear facilities.

The environmental report assesses potential impacts from the Dresden Nuclear Power Station, ensuring compliance with regulatory requirements.

Nuclear facilities must adhere to stringent operational standards to ensure efficiency and reliability in energy production.


Question: community environmental impact assessment

The Applicant's Environmental Report includes an assessment of the community's environmental impact within a 6-mile radius of the Monticello Nuclear Generating Plant Unit 1.

The scale of the map illustrates distances of 0, 1, and 2 miles from the MNGP site boundary, helping to visualize the local geography.


Question: The NRC is open to various perspectives on staffing and operational frameworks for the self-reliant mitigation facility.

But fundamentally, the self-reliant mitigation facility is just one kind of subtrack within Part 53.

The GLRO, for example, would not be an individual that receives a medical exam.


Question: What conditions trigger a RWCU system isolation during reactor operations?

The following conditions cause an automatic initiation of a RWCU system isolation: (1) SLC System initiation and (2) RWCU non-regenerative heat exchanger high outlet temperature.


Question: zirconium oxidation models

Ocken stated that the advantage of experiments with inductive (Urbanic and Heidrick) and direct electrical heating (Biederman, et al.) of a specimen in a cool environment would be the temperature gradient from heated specimen to cool surroundings, leading to temperature gradients in the cladding wall in the same sense as in a reactor.

The passage of steam through the reactor [unit] greatly increased the heat losses from the samples; and a large increase in power to the induction coil was required.


Question: The methodology for assessing design basis accidents includes evaluation of events that do not result in radiological release.

The Design Basis Accident methodology outlines the approaches used to evaluate in-vessel events that are not associated with the release of radiological materials.

For this project, various materials will be used in the construction of the reactor core to enhance safety and efficiency.


Question: nuclear safety licensing process

The nuclear licensing process involves a series of regulatory steps that ensure safety and compliance with established standards before a nuclear facility can operate.

Santee Cooper has been a significant player in the development of energy resources in South Carolina, focusing on both renewable and non-renewable sources.


Question: minimum detectable activity

The minimum detectable activity (MDA) is the smallest amount of a radioactive substance that can be detected with a high probability of correctness, and is critical for ensuring that the analytical results meet regulatory standards.

The address for LaCrosseSolutions is S4601 State Hwy 35, Genoa, Wisconsin 54632, and Mr. Jason Q. Spaide serves as the contact person for the project.


Question: The NRC staff may request further details from the licensee if needed after the audit concludes.

If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the staff will issue a request for additional information to the licensee after completion of the audit.

The audit will be conducted using telecommunications from May 20, 2024, through August 30, 2024.


Question: What is High-assay low enriched uranium (HALEU)?

HALEU, or High-assay low enriched uranium, refers to uranium that is enriched to a lower percentage of Uranium-235 than that used in conventional reactor fuel, typically below 20% enrichment.

The Idaho National Laboratory (INL) conducts research and development related to advanced nuclear technologies.


Question: nuclear reactor license renewal process

The Monticello Nuclear Generating Plant Unit 1 is undergoing an application for subsequent license renewal as part of the regulatory process to ensure its continued safe operation.

Garlic mustard was likely brought to the United States for food or medicinal purposes in the 1800s.


Question: What are spectral accelerations in nuclear engineering?

The typical envelope horizontal response spectra for analysis comparison include spectral accelerations measured in G units across different periods in seconds.

The values listed in the table provide a comparative analysis of seismic responses for various reactor types under specific conditions.


Question: What is the purpose of a Safeguards Actuation signal in reactor operations?

Was a Safeguards Actuation signal generated either manually or automatically during the reactor trip response? The question's purpose is to determine if the operator had to respond to an abnormal condition that required passive safety injection or respond to the actuation of additional equipment that would not normally actuate on an uncomplicated scram.

The question applies to all battery-backed IDS DC and 24- and 72-hour emergency AC buses.


Question: What are the safety parameters for sodium hypochlorite storage in nuclear facilities?

Sodium hypochlorite (5% solution) is stored in Multi-5, 400 gal Warehouses Cooling Tower in a quantity of 40 gal.

Quantities less than 1000 lbs and chemicals stored are not listed.


Question: What are the acceptable limits for dissolved contaminants in water samples?

ACZ Project ID: L09683

Inorganic QC Summary

**NOTE:** If the Rec% column is null, the high/low limits are in the same units as the result. If the Rec% column is not null, then the high/low limits are in % Rec.


Question: nuclear reactor safety systems

The safety systems in nuclear reactors are designed to prevent accidents and mitigate their consequences, ensuring the health and safety of both the public and the environment.

The power consumption of the facility has increased significantly over the past year, leading to discussions about upgrading the electrical infrastructure to meet demand.


Question: Nuclear Safety Culture Monitoring Panels

Incorporated FBF 2515D-2412 to add Nuclear Safety Culture Monitoring Panels as a meeting that should be attended.

IMC 2515 Appendix D, “Plant Status,” has been revised to consolidate IMC 2515, Appendix F, “Reactor Construction Activities Near Operating Unit(s),” into this IMC.


Question: What are the components of the Emergency Core Cooling System?

The Emergency Core Cooling System contains components which have no other cooperation function as well as components which are shared with other systems and perform normal operating functions.

The electrical systems provide normal and emergency power sources for the Emergency Core Cooling System.


Question: vital Component Cooling Water System

LCO 3.7.7 Two vital CCW loops shall be OPERABLE.


Question: hazard classification in nuclear material

Did the RSO classify the samples as indeterminate? Maximum Net Counts Observed: Observed Counts: Classified as: Rad I Rad 2 Rad 3

Samples requiring chemical preservation at proper temp? Yes / No


Question: nuclear regulatory filings

Pursuant to 10 C.F.R. § 2.305, I certify that, on this date, a copy of the foregoing 'DUKE ENERGY CAROLINAS, LLC’S ANSWER PARTIALLY OPPOSING PETITIONERS’ MOTION FOR EXTENSION OF TIME AND PAGE LIMITS' and Attachment 1 thereto were served upon the Electronic Information Exchange (the NRC’s E-Filing System), in the above-captioned docket.

The Oconee Nuclear Station is composed of three pressurized water reactors, each designed with unique safety features to withstand various operational challenges.


Question: The NRC outlines specific requirements for the loading of assemblies in pressurized-water reactors (PWR).

Loading Table for PWR Fuel – 953 W/Assembly is specifically designed to detail and support the fuel loading criteria for pressurized-water reactors.

The Loading Table for PWR Fuel – 1,300 W/Assembly contains various specifications necessary for the operational efficiency of different assembly designs.


Question: sprayed containment volume

1. **Compartment 1:** Sprayed Containment Volume - Value: 2.360E+06

### File Information - **Plant file:** D:\Staum\2020\LOCA\20Min\Spray Timing\BRW Source in Sprayed_R7_ST.psf


Question: The NRC mandates that licensees undertake inspections to assess the condition of key components involved in high pressure coolant injection.

The High Pressure Coolant Injection System is evaluated for aging management to ensure performance over its operational life.

Certain external surfaces require monitoring to prevent loss of material due to environmental exposure.


Question: How will changes in climate potentially impact the cost-benefit analysis for severe accident mitigation alternatives in nuclear facilities?

ACC-3 Will expected climatological changes affect the severe accident mitigation alternative (SAMA) cost-benefit of the proposed facility?

The assessment covers various scientific aspects related to ecosystems and biodiversity without specific reference to nuclear risks.


Question: weighted mean activity

The weighted mean activity is calculated by considering the contributions from individual nuclides, factoring in their confidence and uncertainty measurements to provide a more accurate representation of the overall activity of a sample.

Confidence levels for nuclides are assigned based on the reliability of the data collected during the analysis, with higher confidence indicating greater certainty about the nuclide's presence and activity.


Question: The Gamma spectrum is evaluated for peak identification and quantification in nuclear materials.

- **Efficiency Calibration Description:** The efficiency calibration was conducted using a standard gamma source to ensure accurate measurement of radioactive samples being analyzed.


Question: ASME Code Section III alternatives

Proposed alternatives to ASME Code Section III, 1995 Edition with Addenda, including alternatives allowed by Table B3-1, may be used as authorized by the Director of the Office of Nuclear Material Safety and Safeguards or Designee.

The justification in Table B3-1 demonstrates that:


Question: The aging management program should align with the material and environmental factors as described in NUREG-2191.

The project scope will include an analysis of seismic safety features and electromagnetic interference.


Question: nuclide identification

Nuclide identified during the nuclide identification

The reactor core is designed to withstand significant seismic events and ensure operational safety.


Question: What are the storage allowances for unirradiated and irradiated fuel assemblies?

In accordance with 1-FHI-7, one assembly is allowed in the transfer cart, two assemblies are allowed in the RCCA fixture, and one assembly is allowed in the reactor vessel.

Plausible because irradiated fuel assemblies are allowed outside of approval; storage in the reactor building but the assembly in the refueling mast over the upender must be included in the 3 assemblies allowed within the refuel canal which would result in 4 assemblies outside approved storage in the refuel canal, thus exceeding the three allowed in the refueling canal area.


Question: vitrified clay pipe NORM

S3-012-102 B, Storm Drain 5, is a 30” internal diameter vitrified clay pipe that runs east-west underneath the northern portion of the Switchyard. During the FSS, higher than expected readings were identified and it was suspected that NORM in the vitrified clay was the reason.

The following discussion will be included in the EnergySolutions corporate Lessons Learned Database: if characterization surveys were not performed on a system, any sample media collected during FSS will be analyzed for the full suite of radionuclides as if it were a characterization, or continuing characterization sample.


Question: What are the aging effects associated with closure bolting in nuclear facilities?

Loss of material due to general (steel only), pitting, crevice corrosion


Question: nuclear regulatory process

United States Nuclear Regulatory Commission (NRC). March 2007. Standard Review Plan. NUREG-0800, Section 2.4.9, Channel Diversions. U.S. Department of Energy, Office of Nuclear Regulatory Research: Richland, Washington.

U.S. Army Corps of Engineers (USACE). October 31, 1994. EM 1110-2-1418, Engineering and Design Channel Stability Assessment for Flood Control Projects. USACE, CECW-EH-D, Washington, DC 20314-1000.


Question: Loss of Reactor Coolant Accident (LOCA) procedures

The Loss of Reactor Coolant Accident (LOCA) is addressed in the procedure 2-EOP-LOCA-1, which outlines the steps for responding to a situation where reactor coolant is lost, ensuring the safety and integrity of the reactor system during such critical events.

The technical specifications must be strictly followed to ensure compliance with regulatory requirements, which include various operational parameters and limits that are essential for safe reactor operation.


Question: Licensees must submit specific documentation to ensure compliance with nuclear safety standards.

The regulatory process for nuclear plant licensing requires the submission of detailed safety analyses and environmental reports to demonstrate compliance with applicable regulations.

The advancement of renewable energy technologies significantly impacts the energy market and offers innovative solutions for sustainable development.


Question: What is the significance of Strontium-90 in nuclear safety assessments?

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


Question: The NRC mandates the proper functioning of valves in reactor systems for safety purposes.

The X-TIE VLVS provides interconnection capabilities between units, ensuring that the discharge pathways are properly managed for safety operations.

The status of the pump gates indicates that they are currently operational but do not affect the valve functions.


Question: The NRC mandates accurate monitoring of reactor coolant pump conditions to ensure operational safety.

Power-range high neutron flux is measured with an accuracy of ±1 percent of full power.

Low reactor coolant flow is recorded with an accuracy of ±2.5 percent of full range within the range of 70 percent to 100 percent of full flow.


Question: The Minimum Detectable Activity (MDA) of a nuclide is an important metric in radiation measurement.

The Nuclide MDA (pCi/grams) column indicates the minimum detectable activity for various nuclides measured in picocuries per gram.

Errors quoted at 2.006sigma indicate the level of uncertainty associated with the measurement of peaks.


Question: reactor pressure vessel axial weld failure probability

The recommendations for inspection of RPV shell welds in BWRVIP-05 (Reference 4.8.10) include examination recommendations of BWR RPV axial welds.

BWRVIP-329-A, approved in December 2021, and the associated NRC SER provide additional technical basis for an assessment of axial weld integrity for extended operations of up to 80 years.


Question: What is the purpose of the TRITON11 reference fuel assembly design?

The TRITON11 reference fuel assembly design is intended to provide accurate predictions of fuel performance metrics for Westinghouse PWRs (Pressurized Water Reactors).

Nuclear reactors require a thorough analysis of structural integrity to ensure safety is maintained under all operational conditions.


Question: byproduct material licensing

Any person who desires to manufacture, process or produce, or initially transfer for sale or distribution self-luminous products containing tritium, krypton 85, or promethium 147 for use under He-P 4030.02 shall apply for a license issued by the Nuclear Regulatory Commission pursuant to 10 CFR 30.32, or enter a certificate of registration pursuant to 10 CFR Part 32.210.

The provisions of He-P 4030.03(g) and (h) shall not authorize the production, packaging, or repackaging, or transfer of byproduct material for purposes of commercial distribution, or the incorporation of byproduct material into products intended for commercial distribution.


Question: The NRC staff has determined that the A-4300 curve reliably estimates the FCG rates based on significant data from ferritic steels.

Specifically, the FCG rate is defined with a log-normal distribution with the median value defined as the FCG rate in ASME Code, Section XI, A-4300, and with a value of 0.467 for the uncertainty parameter.

The NRC has neither reviewed nor approved the EPRI report generic use.


Question: What are comprehensive risk metrics in nuclear safety?

the term, comprehensive risk metrics, is supposed to be say capture, also say defense in depth considerations.

There may be other design deterministic risk metrics or something for defense in depth that are used as part of the overall decision-making process.


Question: Licensees must submit periodic reports regarding their compliance with regulatory requirements for reactor safety.

The licensing of nuclear reactors involves a rigorous review process by the NRC to ensure that all safety and operational aspects are thoroughly evaluated.

During the construction phase, materials must be delivered on time to avoid delays in the project schedule.


Question: Will the modifications lead to new hazards requiring reassessment?

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes permit the extension of Completion Times provided that risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program.


Question: shear wave velocity variation

The shear wave velocity design profiles shown in Figure 2.5-62, Profile (b), plus the shear wave velocity of the Zone III-IV rock from 55 to 70 feet depth is used in the seismic amplification/attenuation analysis.

As noted above, Zone III-IV is assumed to extend from 55 to 70 feet depth.


Question: The NRC staff will generally issue requests for additional information as necessary during the review process.

The NRC staff will develop any RAIs, as needed, via NRR Office Instruction LIC-115 and issue such RAIs separately from audit-related correspondence.

Username and password and/or other Web-based portal access information should be provided directly to members of the NRC staff and contractors as needed.


Question: What are the parameters used in evaluating the radiological consequences of a control rod ejection accident?

The parameters used in evaluating the radiological consequences of a control rod ejection accident include core power level, total steam generator tube leakage, reactor coolant iodine activity prior to the accident, gap activity released to reactor coolant from failed fuel, melted fuel, reactor coolant noble gas activity, secondary system initial activity, reactor coolant mass, secondary coolant mass for four generators, and offsite power status.

The atmospheric dispersion factors are necessary for understanding how the released contaminants spread in the environment after an accident.


Question: What is the significance of the net peak area in nuclear data analysis?

Net Peak Area refers to the area under the peak in a spectrum which is calculated after subtracting background contributions from the measured data.

The Original Area is calculated prior to background subtraction and is used to analyze the overall contribution of the signal.


Question: The Parish EOC Director is responsible for coordinating the reentry process once it is deemed safe for individuals to return to their homes.

When informed that full recovery and unlimited access are recommended, the Parish EOC Director will perform the following:

The report details the emergency response protocols implemented during natural disasters, emphasizing the need for rapid communication and coordination among various agencies.


Question: NRC staff are seeking to meet soon to clarify the reporting requirements for safety evaluations and other regulatory processes.

NRC staff requesting additional near term meeting to discuss SGTIR reporting template, loose parts strategy, and EPRI Guidelines revision.

A TSTF-577 revision will be needed to align with NRC staff position.


Question: What is the purpose of the Emergency Core Cooling System in nuclear reactors?

The Emergency Core Cooling System is designed to provide cooling to the reactor core during a loss-of-coolant accident, ensuring the core remains covered and preventing overheating.

Hydraulic Model Studies are often conducted to evaluate fluid flow characteristics in various systems, including pumps and valves.


Question: The NRC oversees licensing for innovative reactor designs, including the assessment of advanced reactors' safety metrics.

The Advanced Reactor Licensing Branch is responsible for the review and approval of applications for advanced reactor designs and their associated regulatory requirements.

The document states that the project is scheduled for completion by December 2024, following extensive evaluations and assessments.


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory steps that a nuclear facility must complete to operate in compliance with safety standards set by the NRC.

The committee discussed the various cultural programs aimed at promoting community engagement among tribal members.