Question: The components of the pipeway structure are subject to multiple environmental conditions including weather exposure and concrete encasement.

The materials of construction for the radwaste storage facilities component types are: - Carbon Steel - Concrete

The following aging effects require management: - Cracking - Distortion - Increase in porosity and permeability - Loss of bond - Loss of material - Loss of preload - Loss of strength


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

COBRA-FLX is a thermal-hydraulic code developed for the analysis of nuclear reactor cores, which incorporates the ORFEO-HMP critical heat flux correlation to improve safety assessments for core thermal performance.

The test assembly conditions for HTP grid tests are documented in Table B-2, detailing parameters such as pressure, temperature, and flow rates, which contribute to the overall testing methodology.


Question: What is an environmental impact statement in the context of nuclear plant licensing?

The Generic Environmental Impact Statement for License Renewal of Nuclear Plants provides an assessment of the potential environmental effects associated with the renewal of operating licenses for nuclear power plants.

The Browns Ferry Nuclear Plant Units 1, 2, and 3 Environmental Statement includes a detailed review of the anticipated impacts on land use and local economies.


Question: The NRC evaluates natural hazard risks such as flooding from storm surges during the safety assessment of nuclear facilities.

Flooding that results from an abnormal rise in sea level due to atmospheric pressure changes and strong wind generally accompanied by an intense storm. Secondary hazards resulting from a storm surge include, but are not necessarily limited to, high tide, river and stream overflow, and waves.

Dynamic forces on structures' foundations due to the expansion (swelling) and contraction (shrinking) of soil resulting from changes in the soil moisture content.


Question: What are the holding times for sample analysis in nuclear-related testing?

All analyses were performed within EPA recommended holding times except for parameters flagged with "H" flags (H2, HE).

The custodian verified the sample information entered into the computer against the chain of custody (COC) forms and sample bottle labels.


Question: 10 CFR 50.55a and nuclear licensing alternatives

10 CFR 50.55a(y) “Definitions” was incorporated into the rule during the most recent change to 10 CFR 50.55a [89 FR 58039]. 10 CFR 50.55a(z) “Alternatives to codes and standards requirements” provides a licensee to seek “Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation.”

Leak tight or within the limits required to maintain operability as required by plant quality assurance program and procedures. Deletion of this condition would provide an efficiency gain by removing these redundant and unnecessary regulatory requirements.


Question: What is the role of closure bolting in mechanical systems?

Closure bolting provides mechanical closure in systems, ensuring that components are securely fastened to prevent leaks and maintain system integrity.

Heat exchanger components include various materials, such as stainless steel and gray cast iron, which are essential for maintaining system performance under varying environmental conditions.


Question: nuclear plant event notification form

The Shift Manager directs you as the Plant Communicator to complete EMD-32a, Nuclear Plant Event Notification form for approval.

With indications of multiple rod drops, the crew tripped the reactor and entered 2-OHP-4023-E-0, Reactor trip or Safety Injection.


Question: nuclear reactor safety systems

Nuclear reactor safety systems are critical components designed to prevent accidents and mitigate the consequences of any potential accidents that may occur during nuclear operations.

The effectiveness of renewable energy sources has gained significant attention in recent years, as they provide a cleaner alternative to fossil fuels and contribute to efforts against climate change.


Question: The NRC allows Licensees to submit requests for modifications to inspection protocols related to the reactor pressure vessel.

The Order imposes interim inspection requirements for reactor pressure vessel (RPV) heads and associated penetration nozzles.

In anticipation of numerous requests for relaxation of the Order for inspections of specific penetration nozzles, the Order specified that the NRC staff would evaluate requests for relaxation of inspection requirements for specific penetration nozzles using its procedure for proposed alternatives to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code pursuant to 10 C.F.R. §50.55a, “Codes and standards.”


Question: What factors influence atmospheric transport and dispersion of gaseous effluents from nuclear reactors?

This guide provides criteria for characterizing atmospheric transport and diffusion conditions for evaluating the consequences of routine releases.

The types of facilities that would be used to store radioactive liquids and any associated inventory are unique to each design.


Question: Class 1E Power System compliance

The PPS complies with IEEE 603-1991 Clause 8.1 because it is using the existing LGS Class 1E Power System that meets its licensing basis for an electrical power source.

Section 3.5 identifies all the external PPS interfaces and there is no direct interface between the PPS and the plant HVAC.


Question: Electrical Insulation aging management program

The BFN Electrical Insulation for Inaccessible Low Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements aging management program is a new condition monitoring program.

Periodic actions will be taken to prevent in-scope power cables from being exposed to significant moisture.


Question: What is the method used for analyzing Strontium-90 in solid samples?

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


Question: What is the procedure for handling confirmed excursions in nuclear licensing?

In the event of a confirmed excursion per LC 12.1 of SUA-1580, NuFuels will notify the EPA or NMED and NRC by phone within 24 hours and by letter within 7 days.

Each retention pond will be equipped with a leak detection system and inspected on a daily basis for leakage.


Question: The minimum cooling period for assemblies after they reach a certain burnup is crucial in the licensing process.

Minimum Cooling Time (years)                                     | 58 < Assembly Average Burnup ≤ 59 GWd/MTU

The table illustrates various configurations and their associated initial assembly average enrichment percentages.


Question: AC vital bus distribution subsystem requirements

With one Division I or II AC vital bus distribution subsystem inoperable, the remaining OPERABLE AC vital bus is capable of supporting the minimum safety functions necessary to shut down and maintain the unit in the safe shutdown condition.

It is, therefore, imperative that the operator's attention focus on stabilizing the plant, minimizing the potential for loss of power to the remaining vital bus, and restoring power to the affected vital buses.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves several key stages including pre-application discussions, submission of a license application, and reviews by the Nuclear Regulatory Commission to ensure compliance with safety standards.

The analysis of seismic data involves utilizing various sensors and instruments to measure ground motion, which can be crucial for understanding the geological characteristics of a site.


Question: The COBRA-FLX tool is essential for analyzing critical heat flux under various local conditions.

The COBRA-FLX code employs a critical heat flux correlation to ensure proper safety analysis during reactor operations.

The data collected during the simulations are crucial for determining grid performance in various configurations.


Question: What are the pressure values for the reactor coolant system during cold leg operations?

The RCS Cold Leg Pressure (psia) values are significant for understanding the parameters affecting reactor safety.

Under normal operations, the reactor experiences various temperature fluctuations that can affect the overall efficiency of power generation.


Question: The NRC conducts assessments of materials related to nuclear safety to ensure any degradation risks are adequately managed.

The Corrosion and Metallurgy Branch focuses on the assessment and management of materials issues, including corrosion, that may affect the safety of nuclear reactors.

The meeting was scheduled to discuss the plans for the upcoming harvesting workshop for various research topics.


Question: What are the technical specifications for the MAGNASTOR system?

The MAGNASTOR system is designed with specific technical features that ensure its compliance with safety regulations, including the ability to store spent nuclear fuel in a secure and environmentally safe manner.

Nuclear power plants must adhere to strict operational protocols to maintain efficiency and safety during routine operations and emergencies.


Question: What is the recommended minimum cooling time for fuel assemblies with an average enrichment of 2.9 to 3.1 % 235U?

The table specifies minimum cooling times for various assembly average enrichments, such as for 2.9 ≤ E < 3.1, the minimum cooling times vary across different reactor designs, with CE showing 14.2 years for 14x14 configuration.

In reactor operation, monitoring parameters such as temperature and pressure is crucial for ensuring safety and efficiency.


Question: The transportation of the yellowcake product involves specific procedures and routes to ensure safety and compliance during shipments.

Following drying and packaging of the yellowcake product, the product will be sold to utilities.

The layout of the well field is shown in Figure 1.4-8. It is in the floor of the valley and will not be materially affected by the nearby escarpments.


Question: The Tribe asserts that the Board's refusal to vacate the license does not comply with federal regulations regarding NEPA.

Given the Board’s decision, the Tribe argues that NEPA and the NHPA prohibit the Board from leaving the license in place and asserts that 'the proper remedy is that employed by federal courts up[on] a finding of a violation of NEPA: to vacate the decision and remand back to the agency for further proceedings necessary to achieve compliance.'

In a separate context, the health and safety of the community must be considered in conjunction with any project that affects their environment and livelihoods.


Question: The NRC's oversight includes evaluating the geography of project areas involved in nuclear activities...

The Unit 1 Project Area is designated for regulatory review in the context of nuclear safety and licensing.

New Mexico and Arizona are known for their diverse landscapes and rich cultural heritage.


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

A containment system is designed to prevent the release of radioactive materials into the environment during an accident by providing a physical barrier that protects the reactor core and the spent fuel.

The operational efficiency of a nuclear power plant depends on continuous monitoring and maintenance of its auxiliary systems, which support the primary reactor functions and contribute to overall safety.


Question: fuel performance database


Question: What is the significance of the REMP program in relation to environmental safety at MNGP?

Furthermore, MNGP implements a REMP program that includes sampling fish populations.

The latest available U.S. census Hmong population data, in geographic areas smaller than counties, was the 2015 census tract estimates.


Question: nuclear operating corporation authority

Michael T. Boyce, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

Notary Public
Expiration Date: 1-11-2027


Question: The NRC mandates that Quality Assurance programs must remain uncompromised during any change evaluation as per regulatory requirements.

The 10 CFR 50.54(a) / 10 CFR 50.55(f) screening/evaluation confirmed that the individual and cumulative changes made during this update to the NDQAM did not result in a reduction in Quality Assurance program commitments.

Changes were made to the NDQAM, Part I, Section 2.4 and Part II, Sections 17.1 and 17.2 to apply the NDQAM to radioactive material transportation packaging as defined and allowed by 10 CFR 71 Subpart H.


Question: What are continuous soil core samples used for in geophysical testing?

Continuous soil core samples were obtained at 20 of the 32 observation well locations using rotosonic drilling methods to expedite the drilling process.

Thirty-two observation wells were installed in 16 locations during exploration at the PSEG Site.


Question: The emergency diesel generator is activated when specific voltage conditions are met to ensure safety during operation.

The emergency diesel generator starts when a safety injection signal is received, at approximately 74% voltage on the bus for 2 seconds, or a 90% degraded voltage level exists for 56 seconds.

The 480V emergency switchgear buses H and J are arranged as shown in Reference Drawing 5.


Question: The components of the Air Conditioning System are critical for maintaining safety during adverse plant operations.

The Air Conditioning system is needed to meet NSPC Vital Auxiliaries Performance Criteria.

In the event of a real or spurious accident signal during a fire event, components within the Air Conditioning System may load shed to meet the NSPC Vital Auxiliaries Performance Criteria.


Question: What is the function of the Containment Purge Exhaust in a nuclear reactor?

Containment Purge Exhaust facilitates the removal of potentially contaminated air from the containment area, ensuring that the internal atmosphere remains controlled and safe during normal operation and in emergency situations.

Type 'C' Sealed indicates a specific condition of local leak rate which might affect the assessment of the overall containment integrity.


Question: What is the significance of FWHM in peak analysis for nuclear safety reports?

F 1 has a FWHM (Full Width at Half Maximum) of 3.11 keV, which is an important parameter in determining the resolution of a peak in spectral analysis.

The Ambient Background for F 7 is measured at 1.88E+01 and exhibits a corresponding uncertainty in background measurement.


Question: What is geological evidence of liquefaction in the Saline River source zone?

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

Geological field investigations in the vicinity of Saline River were conducted to evaluate faults initially identified in road-cuts (Reference 151).


Question: The NRC mandates that all necessary details in submissions to the agency must be accurate and complete in every significant aspect.

I am aware of the requirements of the general license, provided in 10 CFR 31.5.

A physical inventory of the devices subject to registration has been completed, and the device information on this form has been checked against the device labeling.


Question: Hatch Unit 2 operation evaluation

Hatch Unit 2 Operation with Two Studs Out of Service Evaluation

Check again for flange contact links which came out of compression


Question: The operability of Service Water System loops is critical during operation modes.

The completion time for restoring required actions must not exceed 72 hours.


Question: What is the process for nuclear reactor licensing?

The nuclear reactor licensing process involves several stages, including pre-application consultation, formal application submission, and the review of technical and safety information by the NRC.

The testing of equipment ensures that all components meet operational standards and are free of defects.


Question: What is the significance of normality assumptions in modeling manufacturing parameters?

The assumption of normality for the manufacturing parameters (Table 5-5), steady-state model parameters (Table 5-6), transient model parameters (Table 5-7), and power distribution uncertainties (Appendix A) from Reference 2 are justified in the following discussion.

The steady-state model parameters were processed with the Shapiro-Wilk test.


Question: What is the role of NUSO/SM in declaring an Appendix R Fire?

The NUSO/SM is required to declare an APPENDIX R Fire.

Which ONE of the following completes the statements in accordance with 0-AOI-30.1, Plant Fires, and EPIP-1, Emergency Plan Classification Logic?


Question: The assessment also includes an analysis of the core shroud's ability to endure thermal shock during low-pressure coolant injection events.

Neutron irradiation embrittlement may affect the ability of core shroud to withstand a low-pressure coolant injection thermal shock transient.

The maximum stress intensity factor calculated for the N16 nozzle transient is 83.99 ksi-in² at 1/4T.


Question: The sensitivity analyses conducted for the Non-Loss-of-Coolant Accident Analysis Methodology are performed beyond what is strictly required by the guidelines.

The sensitivity studies performed on [placeholder] are in addition to those required by TR-0516-49416-P, Revision 4, "Non-Loss-of-Coolant Accident Analysis Methodology." These sensitivity studies are not part of the implementation of the methodology in TR-0516-49416-P.

The NRC recommends that applicants incorporate feedback from the staff into their regulatory submissions to ensure compliance with safety standards.


Question: What is the concentration of H-3 in µCi/ml?

H-3 has a concentration of 8.5E-10 µCi/ml.

The isotope Rb-89 has a release rate of 4.7E-05 Ci/yr.


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

Enter the mailing address where correspondence regarding your device(s) should be sent.


Question: The operational status and performance of the Nuclear Service Water System are monitored using specialized instrumentation, informing the plant operators about any potential leaks.

All system components are hydrostatically tested and full operational tests are performed before station operation. During normal operation, periodic tests and inspections are performed per the requirements of the plant Technical Specifications, Inservice Inspection Program, and Inservice Testing Program to ensure operability of the Nuclear Service Water System.

A flood analysis was performed as required by 10CFR50. Flood analysis criteria were also established in Regulatory Guide 1.59. Exceptions were taken to Regulatory Guide 1.59 and described in Section 3.4.


Question: What are the requirements for a Nuclear Power Plant under a Regulatory Licensing Framework?

The NRC applies a rigorous licensing process that requires licensees to submit comprehensive safety analysis reports, technical specifications, and environmental impact statements to ensure compliance with federal regulations.

The cooling system of the reactor is designed to remove heat from the core while maintaining the integrity of the fuel during operation.


Question: At least one of the tasks within the training framework must focus on handling situations categorized as emergencies or abnormal conditions.

At least one of the tasks conducted in the plant shall evaluate the applicant’s ability to implement actions required during an emergency or abnormal condition.

Four to six of the tasks for RO and SRO-I applicants shall require execution of alternative paths within the facility licensee’s operating procedures.


Question: RWCU Differential Flow Setpoint Analysis

Calculation IP-32-0132, Rev. 0, “RWCU Differential Flow Setpoint Analysis for Detecting Large Leaks”.

Design Specification 22A3735, Revision 5, “Leak Detection System”.


Question: What is the objective of the thermal and hydraulic design safety review?

In accordance with SRP Section 4.4, the objectives of the thermal and hydraulic designs safety review are to provide reasonable assurance that there should be a 95-percent probability at the 95-percent confidence level that a hot fuel rod in the reactor core will not experience a departure from nucleate boiling (DNB) or a transition condition during normal operation or anticipated operational occurrences (AOOs).

The U.S. Nuclear Regulatory Commission (NRC) staff plans to conduct a virtual audit to facilitate the TR review.


Question: The NRC mandates that proper procedures must be in place for the operation of standby pumps during refueling operations.

4.2 Residual heat removal flow. (√ instruments used).

SAT:  ☐  UNSAT:  ☐  Standard (CS): Identifies requirements are NOT met.


Question: The NRC does not have the authority to consider local regulations during its licensing processes for nuclear facilities.

The Commission has instructed its licensing boards that a proceeding before them is not the appropriate forum in which to consider the impacts of a local regulation or whether a regulatory waiver or a non-NRC permit may be required for a project to move forward.

Moreover, beyond citing the APA generally, Organizational Petitioners have not argued that the alleged local ordinance binds the NRC or this Board.


Question: TH-230 testing results

TH-230 is a radioactive isotope that can be analyzed to determine its quantity in various samples, with results indicating if it was detected or not during the evaluation process.

The samples listed in the table range from L88716-1 to L88716-8, each associated with a unique client ID, but do not provide detailed information regarding the specific analytes being tested.


Question: seismic LOCA frequency

As described in Section 2.3.4, the large break seismic LOCA frequency was calculated to be 6.2E-07 yr⁻¹ based on representative fragility parameters from EPRI 3002000709 (Reference 19).

Completeness uncertainty was qualitatively determined to be low.


Question: What are defueling canisters used for in nuclear facilities?

The defueling canisters were designed to accept and confine core debris ranging in size from particles (known as fines) of about 0.5 micron in diameter up to partial-length fuel assemblies of full cross section.

Designers began work on the defueling canister in 1980 without knowing the physical details of the fuel debris.


Question: What is the process used for analyzing peak areas in gamma spectrum data?

The Peak Analysis report details that the analysis for peak identification was performed on 5/29/2019, focusing on channels ranging from 100 to 4096.

The sample taken at Dallyland_NPP on 5/17/2019 was 8.768E+02 grams, collected in a 500 ml Marinelli container.


Question: emergency termination procedures

If protective actions had been implemented, verify completion of the actions necessary to allow reentry into affected areas.

Confirm IPAWS Activation and Authorize EAS, as necessary (Use Attachment 2).


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

Energy Calibration Used Done On: 7/8/2014

The Sample Taken On: 6/4/2019 2:42:00 PM


Question: The NRC issues updates and notifications when certain ITAACs are completed or modified, keeping stakeholders informed about safety matters.

The NRC will publish a Federal Register notice when an Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) is found to have sufficient information, marking a milestone in the regulatory oversight process.

The construction process requires a substantial amount of material resources, which includes steel and concrete.


Question: maximum initial enrichment of uranium

| Absorber 0.0225 *10 g/cm²* | Absorber 0.02 *10 g/cm²* |


Question: The NRC will ensure that notifications about lost property are streamlined into a unified submission process.

Consolidate the notification of stolen NRC property to one NRC form.

The proposed actions meet the intent of this recommendation.


Question: What is the verification process for the RadICS platform in nuclear licensing?

Applicants using the RadICS platform are obligated to verify that the requirements of the application are bounded by the established qualification envelope.

Addressed in PSAI 7.4 and satisfied by: Licensing Technical Report.


Question: wind speed measurements

Regulatory filings often require comprehensive documentation, including safety analysis reports and operational assessments, to ensure compliance with safety standards.


Question: total core damage frequency (CDF)

The license amendment request (LAR) provide the plant-specific total core damage frequency (CDF) and total large early release frequency (LERF) to confirm applicability of the limits of Regulatory Guide (RG) 1.174, Revision 3.

Tables E5-1 and E5-2 list the CDF and LERF point estimate values that resulted from a quantification of the baseline internal events (including internal flooding) and fire Probabilistic Risk Assessment (PRA) models.


Question: What is the purpose of the nonconformance or corrective action program in nuclear procurement?

Verify that deficiencies identified/reported by customers, to the supplier, (e.g., receipt inspection rejections, source verification rejections, return material authorizations, site nonconformances, etc.) are adequately evaluated and entered into the supplier's nonconformance or corrective action program, as applicable.

Adequacy of corrective actions taken, as a result of the issues identified during the last NUPIC audit, will be provided by audit team members assigned to the checklist sections which identified the previous issues.


Question: nuclear regulatory inspection results

On December 31, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at McGuire Nuclear Station.

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


Question: The activity of nuclides is a crucial measurement in regulatory filings pertaining to nuclear safety.


Question: What is volcanic hazard characterization in the context of nuclear reactor licensing?

The TR describes TerraPower’s volcanic hazard characterization for the design and operation of the proposed Natrium reactor at the Kemmerer Power Station Unit 1 site.

The NRC staff will summarize its observations in an audit report to be provided to TerraPower, as discussed below.


Question: The NRC conducts inspections to ensure compliance with radiation safety protocols and license conditions.

The inspection was an examination of the activities conducted under your license as they relate to radiation safety and to compliance with the Nuclear Regulatory Commission (NRC) rules and regulations and the conditions of your license.

During this inspection, certain of your activities, as described below and/or attached, were assessed at Severity Level IV, in accordance with the NRC Enforcement Policy.


Question: nuclear regulatory compliance

Nuclear regulatory compliance involves adhering to the guidelines and standards set forth by the Nuclear Regulatory Commission (NRC) to ensure the safe operation of nuclear facilities.

The cooling system in a nuclear reactor must maintain a stable temperature to prevent overheating, which can lead to system failures.


Question: nuclear safety system

The Nuclear Safety System is designed to ensure that the reactor operates within safe parameters and to protect the public and environment in case of any operational anomalies.

The implementation of new technologies in the field of nuclear power generation is subject to extensive research and development.


Question: NRC Form 361 Reactor Plant Event Notification

6.2.7 COMPLETE NRC Form 361, Reactor Plant Event Notification Worksheet.

Verbal Notifications will be performed primarily via commercial telephone. Cellular and/or satellite phones may be used as the back-up means of communication.


Question: The review process for nuclear licenses includes evaluations of various safety systems and regulatory compliance.

The Nuclear Regulatory Commission (NRC) assesses all licensing applications to ensure that the proposed activities meet safety and environmental protection standards.

The names listed as cc via email include various representatives from the Department of Energy, highlighting the involvement of multiple stakeholders.


Question: The analysis of background noise is crucial for accurate peak area calculations in nuclear measurements.

An analysis of peaks from the data indicated specific energies at which significant counts were detected, vital for understanding the source of the radiation.

Nuclear power plants require regular maintenance to ensure operational safety and efficiency in energy production.


Question: HSS categorization process in nuclear licensing

SSC Categorization
      - HSS
        - Section 10.1
        - All Components in Pathway HSS

- Confirm:
  - Section 5
  - Section 6
  - Section 8 (in PRA)
  - Section 9


Question: what are extravasations in nuclear medicine

For those that routinely extravasate, process improvements will be made, and then reporting burden will be minimal for them as well.

The nuclear medicine community has been against regulatory reporting since the late 1970s.


Question: The safety of nearby residents is a primary concern when considering operational changes at nuclear plants.

It is critical that you support the Mark Leyse petition (PRM-50-93) and proceed to rescind the "power uprate" procedures of the Nuclear Industry, Nation wide.

My son and his family live next to Diablo Canyon Plant at Avila Bay, CA, which is built on an earthquake fault.


Question: What are the responsibilities of a Vice President Engineering in a nuclear operating corporation?

Michael T. Boyce, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

SUBSCRIBED and sworn to before me this 11 day of September, 2024.


Question: The NRC mandates the evaluation of physical inspections of structural components to ensure their integrity.

The inspection results showed that no indication of cracking or damage was found during the evaluation of Bolts 25-48 and the Center 1/3rd of Shroud Head/Separators.

The report also discusses the installation of new control systems and their expected impact on reactor efficiency.


Question: chemical injection method

Licensees should provide the one-hour settled volume (e.g., 80 ml of 100 ml solution remained cloudy) for precipitate prepared with the same sequence as with the plant-specific, in-situ chemical injection.

The direct chemical injection method was not used in head loss testing for Wolf Creek. See Figure 3.0.2-1.


Question: The regulatory amendment process requires licensees to submit detailed changes for review.

Replace the following page of Renewed Facility Operating License No. R-110 with the revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

The AGN-201M Research Reactor is a pool-type reactor designed for educational and research purposes.


Question: What is the significance of activity measurement in nuclear devices?

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


Question: Isolation Condenser System components

The Isolation Condenser System includes various components such as bolting, flexible connections, heat exchangers, piping elements, and tanks, which are subject to aging management review to ensure their intended functions are maintained over time.

The tanks for storing diesel oil, such as the Diesel Oil Day Tank, play a critical role in maintaining pressure boundaries within the system but do not directly contribute to the isolation condenser's heat transfer capabilities.


Question: What is K IC in the context of fracture toughness?

K IC (ksi/in) is a critical stress intensity factor used to characterize the fracture toughness of a material; it indicates the ability of a material to resist crack propagation under stress.

The Curve 1 and Curve 2 equations represent energy absorption values for various conditions, highlighting the relationship between stress and material behavior.


Question: The NRC identifies Salem as a site where unique seismic insights related to seismic correlated failures must be taken into account when categorizing safety systems.

The NRC concluded that the methodology used by PSEG in determining the GMRS was acceptable and that the GMRS determined by PSEG adequately characterizes the reevaluated hazard for the Salem site.

The Salem SSE and GMRS curves from Reference [26] are shown in Figure A4-11.


Question: The NRC mandates licensees to follow specific regulatory requirements to ensure safety and compliance.

An environmental assessment for this action was not required, since this action is categorically excluded under 10 CFR 51.22(c)(14).

You will be periodically inspected by the NRC.


Question: What is the ARITA Technical Review in nuclear regulatory contexts?

The ARITA Technical Review (ATR) is a comprehensive assessment by the NRC to evaluate the safety and regulatory compliance of advanced reactor designs during their licensing process.

The Microsoft Teams meeting link provides an easy way for participants to join the session remotely, enabling flexible attendance options.


Question: What is a steam generator and its role in nuclear reactors?

A steam generator is a heat exchanger that is used to convert water into steam in a nuclear reactor system, playing a critical role in the generation of electricity by powering the turbines.

The NRC conducts regular assessments to ensure compliance with safety regulations and evaluates the performance of the regulatory framework.


Question: BWR steam dryer replacement

Replacement Steam Dryers (RSD) designed and manufactured by GE-Hitachi Nuclear Energy with a design life of 40 years were installed in Unit 3 during Unit 3 Refueling Outage 18 (U3R18) in the Spring of 2018, in Unit 1 during Unit 1 Refueling Outage 12 (U1R12) in the Fall of 2018, and in Unit 2 during Unit 2 Refueling Outage 20 (U2R20) in the Spring of 2019 in preparation for Extended Power Uprate implementation during each unit's subsequent operating cycle.

Inspections and evaluations are performed in accordance with BWRVIP-41-R4-A and BWRVIP-138-R1-A. The repair design criteria in BWRVIP-51-A would be used in preparing a repair plan for jet pump components.


Question: The regulatory framework necessitates that design of the canister handling facility must consider multiple environmental load combinations.

All primary load bearing members must satisfy Level A stress limits.

Tornado wind load for the CHF site is categorized under Level D conditions.


Question: What are the implications of inadequate core cooling in nuclear reactors?

Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to inadequate core cooling: Loss of coolant accident, Loss of secondary heat sink, and Loss of passive residual heat removal system.

The contents of the reactor building shall be examined and, as necessary, changed to ensure that all structures, systems, and components meet the standards for safety and operational performance.


Question: What is the impact of RCP seal LOCAs on reactor safety?

The assumed timing and magnitude of RCP seal LOCAs giving a loss of seal cooling can have a substantial influence on the risk profile.

SBO events are important contributors to baseline core damage frequency (CDF) at nearly every U.S. nuclear power plant.


Question: What is the significance of ICCA spacing in damaged packages?

For these cases, the nominal 11.75” (29.8450 cm) center-to-center ICCA spacing is uniformly reduced by 1/8” (0.3175 cm) increments to 11.25” (28.575 cm) to quantify the effect (if any) on ICCA spacing within the damaged package.

The following 2D images are provided to clarify the damaged package array model constructs and associated sensitivity studies.


Question: The NRC emphasizes the importance of modeling assumptions in dose criterion calculations, particularly regarding acceptable public exposure risk.

Clarity is needed on the modeling assumptions, which heavily influence dose criterion calculations, including consideration with respect to the realistic exposure risk to the public that would be acceptable to the NRC.

For sites where the EPZ coincides with the site boundary, the EAB, LPZ, PDD & PCD could all be set to the site boundary.


Question: safety analysis methodology for Xe-100 reactor

The examples of Xe-100 safety analysis methodology in this report’s appendices were focused on a full power nominal operating state.

The steady-state mass and heat balance are developed by Xe-100 Design and reported in the Xe-100 Heat and Mass Balance Report, pending design finalization.


Question: The NRC has established guidelines for managing aging effects that can lead to equipment failures within reactors.

Dissimilar metal welds (DMWs) exposed to an uncontrolled air environment. Cracking due to stress corrosion cracking (SCC) is an applicable aging effect for these materials when exposed to temperatures in excess of 140°F.

Connection of stainless steel components to the steel drywell or torus involve DMWs.


Question: control room dose regulations

MEMBER KIRCHNER: Ron, I have one. For the people working on control room dose, I really commend you to -- first as Dennis pointed out, a larger summary of his excellent from our December presentation.

As far as the meeting -- full committee meeting, staff floated this true best estimate wording and if they want to, that might be a good thing for them to expand on and be a little more specific, if they can.


Question: The audits of supplier quality assurance programs commence every three years and may be adjusted based on changes in procurement scope.

Supplier quality assurance programs are audited on a triennial basis starting with the year that the first audit occurs, and if the procurement scope significantly changes from the originally audited scope.

Continued use of suppliers that have exceeded the time period for audits or surveys due to exigent conditions is allowed, if the following conditions are met:


Question: What tools and guidance are necessary for NRC staff to ensure safety in regulations?

We need to step back. Instead of looking at all these legal theories, look at what we are doing. Does it make sense? It can make sense if it's done right. But if we don't give the NRC staff the tools and the guidance that they need, how do we know it's going to be done right?

CHAIR KRAUSE: All right. Thank you all for your presentations today. Your responses to our questions will help us in making our decisions on the hearing requests. I'm also grateful for your understanding of our change in plans due to the snow. We did receive snow last night and this morning.


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

The ITAAC Notification (ICN) is a communication to the NRC, detailing the status of specific construction inspections or tests that demonstrate compliance with the design and licensing requirements.

The NRC conducts regular inspections to ensure that nuclear facilities operate safely and in accordance with regulatory standards, focusing on areas such as radiation protection and emergency preparedness.


Question: The facility authorized for the safekeeping of spent fuel within NAC-MPC SYSTEMs is known as an independent spent fuel storage installation (ISFSI).

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) The facility within the perimeter fence licensed for storage of spent fuel within NAC-MPC SYSTEMs (see also 10 CFR 72.3).

A handling container for moving up to 60 individual INTACT FUEL RODS or DAMAGED FUEL RODS in stainless steel tubes into a CY-MPC DAMAGED FUEL CAN.


Question: The NRC regulations stipulate procedures for changes in control of licensed materials.

Per 10 CFR 30.34 (b), I am writing to provide notice of a possible change of control of our NRC Materials License.

The Hospital will remain the licensee. There will be no change to the Hospital’s name, mailing address or contact information, including phone numbers.


Question: How does the NRC regulate information requests for nuclear facility licensing?

Procedures for NRC information requests directed to production and utilization facility licensees appear in 10 CFR 50.54(f).

The Regulatory Flexibility Act requires Federal agencies to prepare a regulatory flexibility analysis, to be made available for public comment, if a proposed rule will have a significant economic impact on a substantial number of small entities.


Question: The NRC plans to conduct a meeting dedicated to discussing the environmental assessment and finding of no significant impact with Tribal representatives.

If it is determined that a final EA and final FONSI should be issued, it is expected that they would be published in June 2025.

This meeting will not be recorded or transcribed.


Question: What are the surveillance requirements for auxiliary spray and purification line isolation valves?

This SR demonstrates that the auxiliary spray and purification line isolation valves actuate to the isolation position in response to an actual or simulated actuation signal.

The Frequency of 24 months is based on the need to perform this surveillance during periods in which the plant is shutdown for refueling to prevent any upsets of plant operation.


Question: Licensees must ensure that all safety systems are properly designed and tested before operation.

Before granting a nuclear license, the NRC requires a comprehensive evaluation of the proposed facility's compliance with safety regulations.

The report detailed the progress of construction on the new fuel storage facility, emphasizing the importance of adhering to local building codes.


Question: The measurement of core heat flux is essential for evaluating reactor performance during transient conditions.

Core heat flux indicates the fraction of the nominal heat value generated in the reactor core, measured over time.

The steam flow measurement in a nuclear plant reflects the fraction of nominal values recorded for the low-pressure systems during operation.


Question: competency evaluation in nuclear regulation

The applicants' license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.


Question: The licensing process involves thorough evaluations to ensure compliance with safety standards.

The NRC reviews applications for obtaining a nuclear reactor license to ensure that all safety and regulatory requirements are met before the facility becomes operational.

To monitor the environmental impact of the reactor, continuous assessments are conducted to evaluate the effects on local wildlife.


Question: What are the specific testing conditions required for conducting a toxicity test?

Test conditions must include the method utilized, end points of the test, and any deviations from the referenced method along with the reasons for such deviations.

Effluent samples include details about the sampling point, collection dates, collection method, and the physical and chemical data of undiluted effluent samples.


Question: The sump pump system includes components that must be regularly tested to ensure they function correctly during an emergency.

The Diesel Generator Room Sump Pump System is designed to remove leakage and equipment drains in the Diesel Building and to protect the diesel generators from flooding due to a nuclear service water pipe rupture in the adjacent diesel room acting simultaneously with a turbine building flood.

Visual inspections by station personnel are performed on a scheduled basis.


Question: The inspectors assessed the readiness of crucial systems to withstand severe weather conditions during hurricane preparations.

The inspectors evaluated the adequacy of the overall preparations to protect risk significant systems from impending severe weather, Hurricane Milton, on October 8 and 9, 2024.

On November 15, the unit was shut down due to a manual reactor trip as a result of three dropped control rods during rod control maintenance.


Question: What is the process for local leak rate testing of a primary containment isolation valve?

Local leak rate testing of primary containment isolation valve 2VQ005B, as documented in WO 5053661 during the week ending October 12, 2024.

The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes.


Question: nuclear licensing processes

The nuclear licensing process involves several phases, including pre-application consultation, application review, and ultimately, evaluation by the regulatory authority to ensure safety and compliance with established standards.

The cooling system in nuclear reactors is designed to efficiently transfer heat away from the reactor core, but the specifics of its operation can vary significantly between different reactor designs.


Question: The NRC evaluates the limits of enthalpy rise in advanced nuclear fuel systems.

Enthalpy Rise Limits for M5 Fuel Based on Excess Hydrogen are specified to ensure that safety margins are maintained during reactor operations.

Transient core power is measured to assess the performance of the reactor under different operational conditions.


Question: What is the significance of the source term in nuclear reactor safety analysis?

More recent source term information indicates that the timing from dominant severe accident sequences, as quantified in the state-of-the-art reactor consequence analysis (NRC 2012-TN3092), is much later than the analysis forming the basis of the 1996 LR GEIS.

Operating at a higher reactor power level results in a larger fission product radionuclide inventory in the core than if the reactor were operating at a lower power level.


Question: aging management program

The Open Cycle Cooling Water System (B.2.1.11) program will be used to manage the aging effects applicable to this component type, material and environment combination.

The review of 10-years of plant specific operating experience has not identified cracking in copper alloy with 15% zinc or more in air environments due to potential exposure to ammonia or ammonium compounds.


Question: What are the concentrations of various radionuclides in characterization samples?

**Note:** Bold values indicate concentration greater than MDC. Unbolded values indicate the MDC value. All values in pCi/kg. Blank cells indicate that a particular radionuclide was not included in the analysis.


Question: Has your agency received comments on this regulatory rule?

With respect to this rule, did your agency solicit public comments? Yes ☐ No ☐ N/A ☐

The regulation identifier number or other unique identifier is essential for tracking and compliance.


Question: What is the role of the Chief Nuclear Officer in a company?

The Chief Nuclear Officer oversees all activities related to nuclear operations and ensures compliance with regulatory requirements.

The Executive Vice President of Finance manages the financial strategy and planning of the company, focusing on budgeting and investment.


Question: CUP operational environmental monitoring

Table 6.1-1 summarizes the CUP Operational Environmental Monitoring activities, which include systematic data collection and analysis to ensure compliance with environmental regulations and standards.

Table 3.10.4 compares the income and poverty status in 2023 between McKinley County and the State of New Mexico, illustrating socioeconomic variations across regions.


Question: emergency communication systems at River Bend Station

These systems will be used in the event of an emergency at River Bend Station (RBS) as the communications links between RBS, State agencies, the Parish dispatch point and the Parish EOC. To ensure these systems are operational, should an emergency occur at RBS, portions of these systems will be tested on a regular basis.

Revisions to the Emergency Resources Databook may be covered at the next Parish training for EOC staff or individually as the case warrants.


Question: What is an Audit Plan in nuclear licensing?

An Audit Plan outlines the objectives and scope of an audit related to regulatory compliance and operational safety assessments for nuclear facilities.

The project manager oversees the licensing activity for the reactor and ensures all documentation is accurate and submitted in a timely manner.


Question: The RBCLC Pump A requires proper tagging of its suction isolation valve position to ensure safety during operations.

Identifies RBCLC Pump A suction isolation valve (15RBC-40A) should be tagged in the closed position.

The team must verify the operational parameters of the backup generator before proceeding with the maintenance check.


Question: Replacement steam generators are built to comply with specific ASME Code sections governing their performance under faulted conditions.

Babcock and Wilcox replacement steam generators are designed to the requirements of Section III of the ASME Code for faulted conditions.

Testing of components is crucial to ensure safety and operational efficiency in nuclear facilities.


Question: ASME Section XI inspections

A detailed schedule (including preliminary dates) of: 1.1 Nondestructive examinations planned for ASME Code Class Components performed as part of your ASME Section XI, risk informed (if applicable), and augmented inservice inspection programs during the upcoming outage.

It is important that all of these documents are up to date and complete in order to minimize the number of additional documents requested during the preparation and/or the inspection.


Question: What are the requirements for a licensing application in nuclear regulation?

The NRC requires that all license applications include detailed safety analysis reports that assess the potential risks associated with the operation of the reactor.

The latest updates on technology trends have shown significant advancements in artificial intelligence and machine learning capabilities, impacting various industries.


Question: Endangered Species Act requirements for nuclear power plants

Consultation with the U.S. Fish and Wildlife Service under Endangered Species Act Section 7(a)(2) would be required if license renewal may affect listed species or critical habitats under this agency’s jurisdiction.

The NRC considered the potential effects of particular concern for listed terrestrial species, including habitat loss, degradation, disturbance, or fragmentation resulting from construction, refurbishment, or other site activities (including site maintenance and infrastructure repairs); noise and vibration and general human disturbance; and mortality or injury from collisions with plant structures and vehicles.


Question: MRP-227 technical approval process

The revised versions of Primary Items B16 and B17 allow for alternate aging management disposition of the Primary category component type by supplemental analysis (for NRC approval) in lieu of inspection.

Based on the staff's acceptance of MRP-227, Revision 2, criteria changes to Items B20 and B20.1 – B20.6 (as identified and made in EPRI letter MRP 2024-013), the issues in RAI MRP-227-B&W-3 and RAI 3 are resolved.


Question: What is the significance of monitoring SO₂ levels in regulatory filings?

The provided data table summarizes the daily high 1-hour SO₂ monitoring results at the USBR Shiprock Substation, indicating varying levels of sulfur dioxide emissions over the years against the National Ambient Air Quality Standards (NAAQS).

The reactor system's efficiency is assessed through several performance indicators, including thermal efficiency and fuel utilization metrics.


Question: What is the State Preparedness Network Radio System?

The State Preparedness Network Radio System serves as the backup communications system to the RBS Internet-based Electronic Notifications and State and Local Hotline.

Fixed, mobile and hand-held radios operating on law enforcement and fire/rescue frequencies provide for communication means between the Parish EOC and representatives from the following organizations:


Question: The NRC mandates that operators maintain comprehensive records related to changes in reactor operation.

The regulatory process for nuclear licensing involves submitting detailed documentation to the NRC, demonstrating compliance with safety standards and operational procedures.

The revision history included in the first page was necessary for clarity on the simulator setup and other administrative details.


Question: The NUREG-1903 analysis results can be applied directly if certain conditions are fulfilled.

If the limiting location is part of the existing inservice inspection program, the approach is greatly simplified. As shown on the leftside branch of the slide, the only remaining step is to check whether the limiting location meets the requirements of the ASME Section XI IWB-3500 or 3600. If these requirements are met, then NUREG-1903 results are directly applicable to the limiting location of interest, and no new analyses are required.

So, again, for each of the limiting locations there are two possible options or pathways available to demonstrate the applicability of the NUREG-1903 analysis results.


Question: RPV blowdown procedure

RPV blowdown refers to the controlled release of coolant from the reactor pressure vessel to reduce pressure and allow for safe operation under emergency conditions.

The total number of malfunctions recorded during the assessment period was 7, with 2 occurring after the emergency operating procedures were initiated.


Question: The design and testing of air handling units are critical for ensuring that they meet specified cooling capacity requirements under both normal and accident conditions.

The gross cooling capacity of each AHU package is 30,000 Btu/hr with the plenum air entering at 19°F maximum and cooled by the AHU to 10°F nominal.

In order to limit seismically induced loads the AHU and supports are designed to have a natural frequency in excess of 20 Hz.


Question: The NRC evaluates nuclide confidence for regulatory compliance in safety assessments.

The nuclide confidence index threshold is set at 0.30, meaning that any nuclide with a confidence value below this threshold is not considered reliable for regulatory purposes.

Errors quoted at 2.00 sigma provide an indication of the reliability of the reported measurements.


Question: What are the materials used in the Open Cycle Cooling Water System?

The materials of construction for the Open Cycle Cooling Water System components are: Aluminum Alloy, Carbon and Low Alloy Steel Bolting, Carbon Fiber Reinforced Polymer, Carbon Steel, Cast Iron, Copper Alloy with 15% Zinc or More, Copper Alloy with less than 15% Zinc, Elastomer, Galvanized Steel, Glass, Polymer, Stainless Steel.

The following aging management programs manage the aging effects for the Off Gas System components: Bolting Integrity (B.2.1.10), External Surfaces Monitoring of Mechanical Components (B.2.1.23), One-Time Inspection (B.2.1.20), Water Chemistry (B.2.1.2).


Question: What is the importance of simulator setup in nuclear training?

This is a NEW JPM. Added revision history and simulator setup pages.

Editorial comments from IP 711111.11 FASA.


Question: What is the significance of the reactor containment structure in nuclear safety?

The containment structure is designed to prevent the release of radioactive material to the environment in the event of a loss of coolant accident or other safety-related incident.

The operators initiated a routine inspection of the cooling systems to ensure optimal performance during normal operations.


Question: What is the ARITA methodology used for in nuclear reactor analysis?

ARITA is Framatome's next-generation multi-physics, multi-code methodology for the analysis of PWR non-LOCA events in Chapter 15 of the SRP (except for the Control Rod Ejection event) using a statistical approach.

The NRC prepares draft safety evaluations (SE) as part of the regulatory process to ensure compliance with safety standards.


Question: What is the role of manual trip switches in Engineered Safety Feature Actuation Systems?

The ESFAS manual initiation switches provide the operator with the capability to actuate ESFAS functions from the control room in the absence of any other initiation condition.


Question: The licensing process mandates that all essential safety measures are duly documented by the nuclear facility.

The regulatory framework ensures that nuclear licensees maintain documentation of their compliance with safety standards.

Nuclear power plants are designed to withstand extreme weather events and seismic activity, ensuring the structural integrity of the facility.


Question: Nusselt number correlation

The Gnielinski Nusselt number is 6% accurate for the lower Prandtl range and 10% for the higher Prandtl range. The friction factor used in this correlation is the Fanning friction factor and it is recommended that Eq. (173) be used to calculate it.

The Blasius Friction Factor Correlation for Circular Pipes simplifies the calculation of the friction factor as it provides a direct relationship with the Reynolds number for varying flow conditions.


Question: The NRC is evaluating siting and safety requirements concerning the environmental impacts of nuclear facility locations.

NRC already looking closely at Environmental Requirements for Siting and Safety Siting Requirements

Brownfield sites are already impacted and site reuse may be environmentally preferable to additional land disturbance.


Question: PCT elevation measurement

The PCT elevation is set at 7.25 FT, which is crucial for understanding the system's hydraulic performance under various conditions.

The mass flux is measured in lbm/s·ft², and this parameter is essential for evaluating flow characteristics in thermal-hydraulic analyses.


Question: The flow of coolant during a reflood operation is critical for maintaining safety standards.

Mass Flow Rate (lbm/s) values during the reflood process are essential for ensuring proper cooling of the reactor.

The temperatures recorded in various reactor components are crucial for maintaining operational efficiency.


Question: What measures are in place to ensure that the fire barrier functions correctly during extended operation?

SLRA Tables 3.5.2-1, 3.5.2-2, 3.5.2-6, 3.5.2-7, 3.5.2-12, 3.5.2-13, and 3.5.2-20 include steel elements with a fire barrier intended function, however, the Fire Protection program is not credited to manage the aging effects for these steel elements.

SLRA Tables 2.4.1-14 and 3.5.2-14 include fireproofing and fire barriers with intended functions other than fire barrier, including enclosure protection, flood barrier, and pressure boundary.


Question: Fire PRA model uncertainty

The purpose of the following discussion is to address the epistemic uncertainty in the DNPS Fire PRA (FPRA). The FPRA model includes various sources of uncertainty that exist because there is both inherent randomness in elements that comprise the FPRA, and because the state of knowledge in these elements continues to evolve.

A detailed evaluation of assumptions and sources of uncertainty for the IE PRA Model is included in Reference [6]. The summary evaluation of those IE PRA assumptions and sources of uncertainty determined to be "key" for this application are shown in Section 5.


Question: What is the significance of kinetic impact energy absorption in nuclear safety?

Table 2.10.2-3 presents the Percentage of Total Kinetic Impact Energy Absorbed for various materials used in the OCA structure, including polyurethane foam and ceramic fiber board, critical for assessing the safety performance during impact events.

The various compressive foam strengths listed provide a comparison of different material types under standard test conditions, although they do not specifically address the applications within nuclear regulatory contexts.


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) is reported in pCi/grams for various isotopes and represents the minimum levels of activity that can be measured within a sample.

Peak Locate Performed on: 12/7/2017 8:40:53 PM, indicates the specific time when the analysis was conducted for various energy peaks detected from the samples.


Question: nuclear offgas system

The Turbine Building is an essential structure where the turbine generator and associated systems are located.


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

A containment system is designed to prevent the release of radioactive materials to the environment in the event of a reactor accident.

The reactor's core is composed of fuel elements that provide the necessary nuclear reactions for energy production.


Question: What is an audit plan in the context of nuclear licensing?

Audit plans are essential documents that outline the objectives, scope, methodology, and criteria for audits conducted on nuclear facilities to ensure compliance with regulatory standards.

The Office of Nuclear Reactor Regulation oversees the licensing and inspection of reactors to ensure public safety and environmental protection.


Question: final license applications

Final license applications, including any waiver/excusal requests, due and Form 2.2-1 prepared (ES-2.2 C and E).

Examination preparatory week conducted by the NRC and the facility licensee (I).


Question: What does the NRC require for reevaluating flooding hazards?

The NRC requested that licensees ‘reevaluate the flooding hazards at their sites against present-day regulatory guidance and methodologies being used for early site permits and combined license reviews.’

The NTTF reported a set of recommendations that were intended to clarify and strengthen the regulatory framework for protection against natural phenomena.


Question: aging management programs for electrical components

A plant-specific aging management program is to be evaluated

The plant-specific AMP used to manage loss of material and increased resistance of connection is the DCPP Transmission Conductor and Connections, Switchyard Bus and Connections, and High-Voltage Insulators AMP (B.2.3.42). See further evaluation in Section 3.6.2.2.3.


Question: What are the average discharge temperatures at the Monticello Nuclear Generating Plant?

The average discharge temperatures at MNGP are reported to be between 50°F and 56°F, which is critical for assessing thermal impacts on local aquatic ecosystems.

Historic seismic events between 1970–2021 indicate the Monticello Nuclear Generating Plant is located in an area that has experienced various levels of seismic activity, impacting emergency planning.


Question: registration of nuclear devices

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

Distributor/Distributed By: VEGA Americas, Inc.


Question: Plume Exposure Pathway

The Plume Exposure Pathway (10 mile) EPZ is also partitioned into 16 sectors A through R with each sector with 22.5 degrees.

Protective action recommendations are issued by River Bend Station (RBS) directly to the Parishes and State agencies concurrently, but the State may separately agree or disagree with the recommendation issued by RBS.


Question: What are the required safety systems for nuclear reactor operation?

The safety systems for nuclear reactors include the Emergency Core Cooling System (ECCS), which is designed to provide cooling to the reactor core in the event of a loss-of-coolant accident (LOCA).

The analysis of operational data is critical for ensuring efficient reactor performance and conducting maintenance.


Question: nuclear power plant safety analysis

The purpose of a safety analysis is to evaluate the potential risks and impacts of design and operational changes at the nuclear power plant to ensure compliance with safety regulations.

The reactor core consists of fuel assemblies that produce heat through nuclear fission, which is then used to generate steam for electricity production.


Question: nuclear licensing requirements

The licensing process for a nuclear facility is governed by a series of regulations aimed at ensuring public safety and environmental protection as stipulated by the Nuclear Regulatory Commission.

Capital expenditures for the project are being evaluated based on current market conditions as well as long-term growth projections.


Question: What is included in a nuclear reactor's licensing documentation?

Docket No. 05000483, License No. NPF-30, signifies the official regulatory documentation associated with the operational license of the reactor, which includes various safety assessments and regulatory compliance information.

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


Question: What is the importance of understanding the worth of control rods and their relation to reactor operation?

Define control rod worth, differential control rod worth, and integral control rod worth  3.1

State the purpose of rod sequencing and overlap 3.0


Question: The NRC has evaluated the findings from the radiological surveys conducted at the Menlo Park site, which are crucial for determining safety compliance.

The NRC staff reviewed the four final status survey reports and compared the information provided in the reports to several NRC guidance documents.

The memorandum discusses formal correspondence between the NRC and licensees regarding various compliance matters.


Question: reactor vessel integrity

Integrity/Safety Function develops (magenta path on F-0.4 RCS Integrity) by completing the following: Events which lead to a relatively rapid and severe reactor vessel downcomer cooldown can result in a thermal shock to the vessel that may lead to a small flaw, which may already exist in the vessel wall, growing into a larger crack.

Isolate feedwater by ensuring closed: LCV-107 (TDAFW Level Control Valve)


Question: jurisdictional status in nuclear licensing

Applicants are responsible for determining, in advance, the jurisdictional status of the specific areas where they plan to conduct licensed operations.

Once on the Web site, use the link for “NMSS Procedures” in the left-hand column under “Resources & Tools.”


Question: The NRC sets specific requirements for the content and characteristics of PWR fuel assemblies, including enrichment and decay heat.

1. **Uranium PWR UNDAMAGED FUEL ASSEMBLIES** listed in Tables B2-2 and B2-3 and meeting the following specifications:

Up to 37 PWR UNDAMAGED SNF ASSEMBLIES as shown in Figure B2-1. For a TSC that is less than fully loaded, empty fuel storage locations shall begin with location No. 19 (refer to Figure B2-1), followed by location Nos. 18, 20, 12, 26, 11, 27, 13, 25 and continuing outward, as required, in an approximately symmetric pattern.


Question: The NRC reviews demographic data to assess the impacts on minority groups residing within a specific radius from nuclear facilities.

According to the CEQ definition, a minority population exists if the percentage of the minority population of an area (e.g., census block group) exceeds 50 percent or is meaningfully greater than the minority population percentage in the general population.

The analysis shows that the majority of block groups within the radius are not classified as minority populations.


Question: What are the enrichment limits for BWR assembly loading?

Outer Assembly Enrichment Limit (wt % 235U) is set at 4.6%, 4.7%, and 4.8% for different assembly types as outlined in the criteria.

Certificate of Compliance No. 1031 is crucial for documenting compliance with regulatory requirements and ensuring safety standards are met.


Question: The NRC's review process for radiation safety assessments is crucial for maintaining operational standards.

The Electric Power Research Institute (EPRI) has developed a performance prediction methodology for motor-operated valves (MOVs) which is essential for ensuring their operational reliability in nuclear plants.

The official record copy is maintained for regulatory compliance and can be requested through the NRC's document control center.


Question: What is the upper growth rate for steam generators in nuclear reactors?

The upper 5% of Growth/EFPY for SG-A, SG-B, and SG-C are compared in Figures 3-9, 3-10, and 3-11 to analyze their performance over time.

The bounding growth curve provides a statistical representation of the expected variations in growth across different operating conditions.


Question: The NRC has specific guidelines regarding environmental evaluations during the licensing process.

The Nuclear Regulatory Commission (NRC) requires that environmental reviews are completed as part of the licensing process for new nuclear power plants.

The soil sampling for Boring 11 indicates a depth of 10-15 feet below ground surface.


Question: What does ICCDP mean in the context of nuclear risk assessment?

**Incremental core damage probability (ICCDP)** for the additional 28 hours is below 5E-07 at 3.0E-07. This includes increase attributable to full power internal events (FPIE) and internal fire (FPRA).

8. Confirm that the facility organization that normally reviews safety issues has reviewed and approved this request and that a written NOED request will be submitted within two working days, excluding licensee holidays, of the NRC staff’s decision regarding the NOED.


Question: What is the significance of Probabilistic Risk Assessment in nuclear safety?

Probabilistic Risk Assessment (PRA) is a systematic methodology to evaluate risks associated with nuclear reactor operations and design, providing insights into potential failures and their consequences.

The report describes various fiscal year operational experiences of different reactor systems, including maintenance schedules and compliance inspections, which do not pertain directly to risk assessment methodologies.


Question: soil classification decision rules

The decision rule addressing PSQ2 is stated as: If soil concentrations indicate that a Class 2 or Class 3 should be reclassified to a higher classification, summarize confirmatory data for NRC’s evaluation.

The sixth step in the DQO process examines the consequences of making an incorrect decision and establishes bounds on decision errors.


Question: What are the key aspects of nuclear safety regulatory processes?

The U.S. Nuclear Regulatory Commission (NRC) oversees the licensing and regulatory processes for nuclear facilities to ensure compliance with safety standards and environmental protection measures.

The primary function of the Environmental Technical Review Branch 1 is to evaluate environmental impact assessments submitted by various industries.


Question: What is the significance of maintaining a specific spent fuel pool level during reactor operations?

RS2.1 Basis Level 3 corresponds to an indicated SFP level of 260.7 ft. which is the top of the SFP racks (ref. 2). (DRR 2345116)

Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or Alert and Notification System Design Report?


Question: ECCS operability requirements

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to a minimum of two cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low.


Question: Regulatory compliance for nuclear facilities involves stringent safety protocols to ensure public protection.

The Nuclear Regulatory Commission (NRC) requires timely and accurate submissions from licensees to demonstrate compliance with applicable regulations and technical requirements.

Installed in the reactor building, the containment structure is a critical barrier to prevent radioactive release and ensure operational integrity.


Question: What is the function of the low-pressure safety injection system?

The low-pressure safety injection system is designed to provide water to the reactor core in the event of a loss of coolant, ensuring that the core remains adequately cooled during emergency conditions.

The pump system characteristics include specifications such as the flow rate, pressure ranges, and motor specifications that ensure efficient operation but do not specifically mention safety functions.


Question: Changes made to reactor safety systems and components related to operational modifications are tracked under specific NRC provisions.

Update includes removal of reactor trip on steam generator water-low coincident with steam flow/feedwater flow mismatch and addition of median signal selection circuitry to the feedwater control system.

Update reflects the reinstallation of the valve internals on 1-RC-MOV-1595.


Question: What is the role of the Accident Assessment Team in nuclear emergency response?

The Accident Assessment Team includes various roles such as Accident Assessment Team Leader, Dose Assessor, and Reactor Core Analysis to ensure effective evaluation and response during nuclear incidents.

The Corporate Emergency Response Center Organization outlines the structure and roles of various support managers and teams involved in emergency management and operational coordination.


Question: What is the GALILEO methodology in nuclear safety?

The NRC staff has reviewed the Framatome's best-estimate fuel performance code GALILEO, which models the thermal-mechanical behavior of fuel rods.

The LTR has stated that the "GALILEO methodology provides fuel rod temperatures that can be used to initialize the ECCS analysis.


Question: The total precipitation accumulation at the measurement stations is crucial for understanding weather impacts.

**ACCUMULATION** refers to the total amount of precipitation that falls within a specified period, which is essential for hydrological assessments and environmental monitoring.

The table shows various hourly data points for each station, which helps in analyzing weather patterns over time.


Question: Westinghouse AP300 reactor licensing process

The Westinghouse AP300 is a small modular reactor design that is currently under development, aiming to provide safer and more efficient nuclear power generation.

Public comments on the application must be submitted within 30 days of the license amendment notice for agencies to consider community concerns.


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

Energy peaks were identified at various levels, such as 77.31 keV for Peak F 1, which helps in determining the response and behavior of materials under radiation exposure.

The distribution of subtracted area is crucial in analyzing background variations during the assessment process.


Question: What are the surveillance requirements for channel calibration in nuclear facilities?

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program.

Neutron detectors are excluded from CHANNEL CALIBRATION.


Question: What is an Evacuation Time Estimate (ETE) in the context of nuclear safety?

The Seabrook Station site specific ETE report is documented in EP-SBK-143, SBK Evacuation Time Estimate Study.

Incorporated either by reference or in their entirety in the emergency plan.


Question: What do the Pasquill Stability Classes indicate in the context of nuclear safety?

Percent Frequency of Occurrence by Stability Pasquill Stability Class indicates the distribution of atmospheric stability conditions at the site.

Classes are as follows: Class A: Extremely unstable, Class B: Moderately unstable, Class C: Slightly unstable, Class D: Neutral, Class E: Slightly stable, Class F: Moderately stable, Class G: Extremely stable.


Question: What is the procedure for evaluating the residual filterable solids in nuclear safety filings?

Residue, Filterable (TDS) @ 180C is evaluated according to SM 2540 C-2011 standards, which provide guidelines for measuring total dissolved solids in water samples.

The analysis for potassium dissolved in water follows EPA 200.7, where various samples are tested for concentration and recovery percentages.


Question: What is a Seismic Probabilistic Risk Assessment (SPRA)?

The staff’s review concluded that the SPRA is of sufficient technical adequacy to support Phase 2 regulatory decision-making in accordance with the intent of the 50.54(f) letter.

Probabilistic health consequences, such as population dose risk, are not available because this information was not requested in the response to NTTF Recommendation 2.1.


Question: What is Polonium 210 and how is it measured in groundwater?

ACZ Laboratories, Inc. **2773 Downhill Drive** **Steamboat Springs, CO 80487** **(800) 334-5493**


Question: liquefaction evaluations

The liquefaction assessment was made using a peak ground acceleration of 0.3g consistent with the value set forth in Regulatory Guide 1.60 (USAEC, 1973), which represents the peak acceptable value for the plant that form the basis of the Plant Parameter Envelope (see Section 1.4), and using a range of earthquake magnitudes (M = 5.5, 6.5, and 8) consistent with the range of source mechanisms potentially causing ground shaking at the EGC ESP Site.

Normal design requirements preclude location of a critical structure on material that could liquefy.


Question: The NRC is requesting feedback on the environmental impact report and the associated findings regarding cultural resources.

The NRC staff’s finding of Adverse Effect is integral to understanding the impact of regulatory actions on cultural resources.

The enclosed documents include an APE map and various contacts from the environmental team regarding data submissions.


Question: loss of material due to corrosion in nuclear systems

The One-Time Inspection (B.2.1.20) and Water Chemistry (B.2.1.2) program will be used to manage loss of material of the carbon steel tanks exposed to treated water in the Condensate System, Feedwater System, and Main Turbine and Auxiliaries System.

The One-Time Inspection (B.2.1.20) and Water Chemistry (B.2.1.2) program will be used to manage various types of material loss, which are critical for maintaining the integrity of nuclear systems.


Question: nuclear dispersal modeling

You can offer dispersal. But you've got to show us that as well.

To see if there's a pathway to no burst as a criterion or whether the flexibility will need to be there.


Question: The guidelines for assessing natural phenomena impacts on nuclear plant sites include considerations for flooding.

U.S. Nuclear Regulatory Commission (NRC). 2011. NUREG/CR-7046, PNNL-20091, Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America. ML113211495, November 2011.

Southern Nuclear Operating Company (2015), "Joseph M. Farley Nuclear Plant, Units 1 & 2 Final Safety Analysis Report (FSAR)," Rev 26.


Question: What are the police services provided in the CUP project area?

Police protection around the CUP is provided by the Navajo Nation Police Department and the McKinley County sheriff department.

The treatment facility consists of one lagoon and is operating at full capacity.


Question: Control Rod Drop Accident

The primary consideration for the Control Rod Drop Accident (CRDA) is radiological release to the main control room and to the environment.

This accident involves the postulation of a small steam or liquid line pipe break inside or outside primary containment but within a controlled release structure.


Question: The NRC sets guidelines about the insertion of unirradiated fuel assemblies in fuel storage locations.

Unenriched fuel assemblies and unirradiated (i.e., not inserted in-core) fuel assemblies are not authorized for loading.

RCCs are restricted to fuel storage locations No. 11, 12, 13, 18, 19, 20, 25, 26 and 27 (Figure B2-1). Minimum RCC cool times are:


Question: What is the significance of nuclide confidence in nuclear measurements?

Nuclide confidence index threshold = 0.30.


Question: The accuracy of the random flow transmitter is adjusted based on the flow rate, as described in the technical specifications.

Random flow transmitter accuracy (from Section 7.4.1.11) is propagated through the square root extractor at the specified operating flow rate points of interest by application of Equations 2.4 and 2.6 (as applicable) from Appendix 2 as follows:

An operating flow rate value of 327.24 gpm provides corresponding 'I' value of 17.986745 mA, which does not result in a negative value for the denominator of Equation 2.4 of Appendix 2.


Question: aging management program

The component environment is nonradioactive waste drains associated with the Sanitary Sewage System that have been evaluated as a raw water environment. Loss of material as a result of internal component surface exposed to floor and equipment drain environment will be managed by the DCPP Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components AMP instead of the Open Cycle Cooling Water AMP.

Consistent with component, material, environment, aging effect, and AMP listed for NUREG-1801 line item. AMP is consistent with NUREG-1801 AMP description.


Question: nuclear power plant licensing process

The licensing process for a nuclear power plant involves multiple steps, including pre-application consultation, submission of a license application, safety reviews by the NRC, and public involvement throughout the process.

The operational phases of a nuclear reactor include startup, ramp-up, steady state, and shutdown, each requiring specific control measures and monitoring protocols.


Question: initial assembly average enrichment weight percentage 235U

Initial Assembly Avg. Enrichment wt % 235U (E) | 50 < Assembly Average Burnup ≤ 51 GWd/MTU Minimum Cooling Time (years) | CE 14x14    | WE 14x14    | WE 15x15    | B&W          | CE 16x16    | WE 17x17    | B&W           |

The reactor type, known as a Pressurized Water Reactor (PWR), is designed to use water under high pressure as a coolant.


Question: Please verify the details of the registration for devices such as the one identified with the NRC Device Key.

**NRC Device Key:** 664929 (Internal Control Number)

The manufacturer of the device is TN Technologies, Inc. and holds a license number associated with distribution.


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

Americium-241 is a radioactive isotope that is often analyzed in nuclear safety assessments due to its potential presence in nuclear waste and its radiological hazards.

Tritium is a radioactive isotope of hydrogen that is used in nuclear fusion and as a tracer in environmental studies.


Question: The calculations for flow characteristics must account for factors like contraction ratios and flow areas between components.

The flow area of the path from 1 to 2 is 217.7 ft² and has a contraction ratio of 0.765.

The volumes of Elements 1 and 2 are 5298 ft³ and 6370 ft³ respectively.


Question: ARITA final TER

The ARITA final TER represents the culmination of discussions and analyses related to the Assessment of Regulatory Impacts on Technical Activities.

The period of performance ends September 2023.


Question: What is the definition of meteorite/satellite strikes in relation to nuclear safety?

A release of energy due to the impact of a space object such as a meteoroid, comet, or human-caused satellite falling through the Earth’s atmosphere, to result in a direct impact with the Earth’s surface, or a combination of these effects. This hazard is analyzed with respect to direct effects and is necessarily limited to frost, ice cover, and snow.

The frequency of a meteor or satellite strike is judged to be so low as to make the risk impact from such events insignificant.


Question: What is the significance of pH in nuclear licensing?

The pH, measured in standard units (S.U.), is a critical parameter in assessing the chemical properties of water, especially in the context of nuclear reactor operations where water quality is essential for safety and regulatory compliance.

Temperature measurements are essential for evaluating the thermal dynamics of a reactor system, impacting various safety assessments and operational limits.


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

The safety analysis report is a key document required by the Nuclear Regulatory Commission that evaluates the potential risks associated with the operation of a nuclear facility and provides a framework for ensuring the safety of both the public and the environment.

The nuclear facility's operation requires a comprehensive assessment of its environmental impact to ensure compliance with all regulations and safety standards.


Question: harvesting materials from decommissioning plants

Available materials from decommissioning plants and past harvesting programs

Perspective on future harvesting planning


Question: radioactive materials transportation

Transportation of Radioactive Materials is a critical aspect of ensuring safety and compliance in nuclear operations, requiring strict adherence to legal regulations and safety protocols.

The Inspection Procedures course provides a comprehensive overview of the protocols and methodologies used by inspectors to evaluate facilities and operations.


Question: What is the role of the Corrosion and Metallurgy Branch in nuclear safety?

The Corrosion and Metallurgy Branch is responsible for evaluating the effects of corrosion and other materials-related issues on nuclear reactor safety and regulatory compliance.

When: Tuesday, September 27, 2016 10:00 AM-11:00 AM (UTC-05:00) Eastern Time (US & Canada).


Question: scoping process for license renewal

The scoping process is the systematic approach used to identify the DNPS systems, structures, and components within the scope of subsequent license renewal.

The reactor coolant system plays a crucial role in regulating the temperature of the reactor during operations.


Question: What is the purpose of feedwater heaters in a nuclear power plant?

Feedwater heaters are essential components in the condensate and feedwater system, designed to improve the overall thermal efficiency of the power plant by preheating the feedwater before it enters the steam generator.

The condensate pump has a capacity of 10,640 gal/min at a 1250 ft head, ensuring adequate supply to the system.


Question: What are the surveillance requirements for calculating maximum surface dose rates of storage casks?

SR 3.3.1.1 Verify maximum surface dose rates of STORAGE CASK loaded with a TSC containing fuel assemblies are within limits.

The storage cask is an essential component in the secure containment of nuclear materials.


Question: What is a specific exemption request in nuclear licensing?

On the specific exemption, we don't contest that in accordance with the full Berka interpretation or full Berka denial that it would be possible to use the specific exemption request to return Palisades to service.

Everybody wants our petitions to be categorically dismissed. We're here at the invitation of the Palisades NRC restart panel.


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

The unit of measurement used for the activity of a radioactive isotope is the millicurie (mCi).

The manufacturer of the device in question is RONAN ENGINEERING COMPANY.


Question: What are the regulatory limits for radioactive materials releases?

The licensee controlled, monitored, and quantified releases of radioactive materials released to the environment to ensure offsite doses were within regulatory limits and ALARA.

The inspectors verified that when issues were identified, licensee personnel appropriately documented the issue in the CAP.


Question: What is the role of the Dispatcher/Communicator in nuclear regulatory procedures?

The Dispatcher/Communicator plays a crucial role in coordinating communication during emergency situations as outlined in the emergency response procedures.

The procedure outlines the responsibilities of various administrative roles generally, without specific mention to the Dispatcher/Communicator.


Question: radiological assessment of soil samples

A Radiological Assessment (RA) in survey unit L1-SUB-TDS A was performed in June of 2019. Nine (9) soil samples were collected and analyzed by the on-site gamma spectroscopy system.

The results of the off-site analysis are provided in Table 3-2 below.


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

Net Peak Area is calculated from the energy spectra to quantify the amount of radiation detected at specific energy levels.

The Env. Background File is crucial for accurate radiation analysis and is located at C:\Canberra\Apex\Root\Dairyland_NPP\Data\0000001364.CNF.


Question: The use of a limited number of LTAs is carefully considered to ensure safety within the reactor's design parameters.

The proposed change involves the use of a small number of LTAs which are very similar in all aspects to the co-resident fuel.

The analyses of record remain bounding, and there will not be a significant impact on the operation of any plant SSC or on the progression of any operational transient or design basis accident.


Question: The organizations established under GEH are structured to ensure compliance with regulatory standards in their operations.

The GEH business includes Operational Organizations that are comprised of operational segments responsible for project management, product and services management and delivery, continued plant operations support, and development of new plants, technologies and related offerings to the industry. These organizations are established to comply with regulatory requirements.

The Quality Organization is responsible for managing and supporting implementation of the OMS and the QAPD with responsibility for Quality Programs and Quality Oversight.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves a series of steps including the submission of applications for construction and operating licenses, which must demonstrate compliance with safety and environmental regulations set by the NRC.

The reactor core consists of various components like fuel rods, control rods, and coolant systems that work together to maintain safe and efficient operation.


Question: What are Partial Length Shield Assemblies in nuclear reactors?

PARTIAL LENGTH SHIELD ASSEMBLIES (PLSA) are PWR fuel assemblies that contain stainless steel inserts in the bottom of each fuel rod, reducing the active fuel length, and a natural uranium blanket at the top of the active core. PLSAs are sometimes used in reactors to reduce fast neutron fluence reaching the pressure vessel wall.

A STORAGE CASK is either a CONCRETE CASK with a CONCRETE CASK LID or an MSO.


Question: What are the proposed changes to Millstone Power Station Unit 3?

Dominion Energy Nuclear Connecticut, Inc. has proposed changes regarding operational protocols at Millstone Power Station Unit 3 to improve safety and efficiency.

Operational protocols at nuclear power plants are critical for maintaining safety, but they must also be balanced with effective cost management strategies.


Question: What are the enrichment limits for low spent nuclear fuel assemblies?

Low SNF Assembly Average Burnup Enrichment Limits can include requirements such as a minimum assembly average initial enrichment of 1.3 wt% 235U for a maximum assembly average burnup of 10,000 MWd/MTU.

The chemical properties of uranium can vary significantly depending on its isotopic composition and the method of enrichment.


Question: Vacuum Pipe Building

The Vacuum Pipe Building is an underground Class 1 structure that provides structural support and shelter/protection for the condenser circulating water system vacuum breaker components that prevent backflow from the warm water channel to the intake channel.

The aging management review results for these components are provided in Table 3.5.2-8, Equipment Access Lock - Summary of Aging Management Evaluation.


Question: What are the requirements for nuclear reactor licensing?

In accordance with NRC regulations, an application for a nuclear reactor license must include detailed information on the design, construction, and operation of the proposed facility to ensure compliance with safety standards.

The report highlights various temperature dependencies of the materials within the reactor core, which are crucial for understanding the overall thermal performance.


Question: What does DPM/100CM² measure in nuclear safety context?

The table lists the upper end value for gross gamma activity measured in disintegrations per minute per 100 square centimeters (DPM/100CM²), which is a unit used to quantify radiological contamination levels.

The data collected indicates that the frequency of occurrences at various upper end values ranged from 0 to 31, providing a distribution for analysis.


Question: The regulatory body mandates that all nuclear facilities comply with specific safety requirements during their operations.

The NRC established licensing procedures to ensure that operators maintain a high standard of safety and compliance with regulatory requirements.

The report outlines various administrative processes and required documentation for licensing approval.


Question: The NRC mandates stringent guidelines for the analysis of control samples to ensure accuracy in results.

Data Summary: There are no exceptions, anomalies or deviations from the specified methods. All sample data provided in this report meet the acceptance criteria specified in the analytical methods and procedures for initial calibration, continuing calibration, instrument controls and process controls where applicable.

Product: LSC, Tritium Distillation, Solid


Question: What is the GALL-SLR Report?

Dominion’s SLRA relies upon the GALL-SLR Report and the SLR-SRP—both of which incorporate revisions 'to reflect aging differences for increased operating time from 60-80 years' as well as revisions 'to consider new operating experience and provide information identified as missing since the release of GALL Report Revision 2.'

In addition, the Staff considered and reviewed results from aging management program (AMP) audits, domestic and international operating experience, industry viewpoints and recommendations, and public comments in drafting the GALL-SLR Report.


Question: Regulatory filings require careful consideration of neutron radiation effects on material properties such as toughness.

ΔRTNDT means the transition temperature shift, or change in RTNDT, due to neutron radiation effects, which is evaluated as the difference in the 30 ft-lb (41 J) index temperatures from the average Charpy curves measured before and after irradiation.

The PWROG stated that the TR methodology is more conservative than that required by Appendix G to Section XI of the ASME Code, because the margin term is greater than what is required by the ASME Code.


Question: What is TBS in the context of nuclear regulatory evaluations?

Evaluation required to demonstrate that breaks greater than TBS remain unlikely.

DG-1428 provides several options for demonstrating TBS applicability.


Question: Refueling Water Storage Tank

On receipt of a safety actuation signal, the system is realigned to isolate the Refueling Water Storage Tank (RWST) and assure that the water is available for the Emergency Core Cooling System (ECCS).

There are no interlocks capable of blocking the SSPS signal from performing its intended functions.


Question: detector current readings in nuclear regulatory documents

Applicant obtains the detector current readings from handout and enters the appropriate data into Attachment 4.


Question: The analysis of the Cm-241 in the sample involved a recount due to unexpected results which fell outside the safety parameters.

The Cm-241 portion of sample 1204388647 (MB) was recounted due to results more negative than the three sigma TPU. The recount is reported.

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


Question: The NRC mandates that applicants for operating licenses provide an evaluation of potential environmental impacts throughout the life of the facility.

The applicant must assess the environmental impacts of their proposed actions as part of the licensing process, ensuring compliance with applicable regulations.

Federal regulations require strict adherence to safety protocols during the operation of wastewater treatment facilities to prevent any potential environmental pollution.


Question: What is the importance of safety analysis in nuclear regulatory filings?

Safety analysis is an essential process in demonstrating compliance with regulatory requirements and ensuring the safe operation of nuclear facilities.

The design of the reactor core must accommodate various thermal and hydrodynamic conditions during operation.