Question: The NRC mandates that licensees ensure the SSCs are returned to an operable state within a specified time frame following an incident.

Administrative Variation: Salem Functional Unit 3.a.1) correlates to ISTS Function 3.a.(1). The RAs both invoke returning the SSCs to operable status in 48 hours.

Administrative Variation: Salem Functional Unit 4.b correlates to ISTS Function 4.b. The RAs both invoke restoring the channel to operable status within 24 hours.


Question: What is a Loss of Reactor or Secondary Coolant (LOCA)?

A Loss of Reactor or Secondary Coolant (LOCA) refers to a significant failure in the reactor coolant system where coolant escapes, causing a drop in pressure and temperature, which could lead to the release of radioactive materials and impact plant safety.


Question: Are the procedures for the measurement and test equipment sufficient for compliance with the standards?

Verify that measures are established and implemented for the control of measuring and test equipment (M&TE).

Assess if measures established are commensurate with the complexity of the items/services provided.


Question: Item Relied On For Safety

The inspectors evaluated 11 Item Relied On For Safety (IROFS) to verify that they were being implemented as described in the ISA and that the licensee was in compliance with the License Application.

During the inspection period, normal production activities were ongoing.


Question: What are the decommissioning costs for a PWR reactor?

- LLW Burial Location: South Carolina and Utah.


Question: The NRC mandates that certain radioactive devices must be registered to ensure proper tracking and safety compliance.

Devices subject to registration include those that emit radiation and are used in various applications, requiring regulatory oversight by the NRC to ensure public safety.

The distributor license number for the device is L03524, issued to Thermo Measurement, reflecting its authorized status in the distribution of radioactive devices.


Question: What is the requirement for reporting unavailability data when a train is restored to MSPI monitoring?

If a previously excluded train/segment is restored to monitoring as a result of a PRA model revision, unavailability data only for the quarter in which the train/segment is restored (and future quarters) needs to be included for reporting purposes.

The NRC evaluates all aspects of safety and technical specifications to ensure public health protection.


Question: COBRA-FLX critical heat flux correlation

The COBRA-FLX code incorporates a new critical heat flux correlation derived from extensive thermal-hydraulic analyses, which significantly improves the prediction accuracy under steady-state and transient conditions.

The report details the methodology used to assess the performance of various pump designs under different operating conditions.


Question: What is the role of engineering safety feature activation system relays in nuclear safety?

The NRC identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, procedures, and drawings," when the licensee failed to prescribe and accomplish procedures to meet the requirements of ASME NQA-1, Subpart 2.18, and RG 1.33, App. A, Section 9, for the preventative maintenance of engineered safety feature activation system (ESFAS) relays.

On June 19, 2024, after inspector questioning, the licensee performed the evaluation and determined that there were several relays in the Unit 1 and 2 ESFAS system that would result in a plant trip at 100% power, similar to the event at Waterford.


Question: weighted mean activity in radiological measurements

Weighted Mean Activity (pCi/grams) represents the average activity of a nuclide in terms of decay rate per unit mass, which is crucial for understanding the radiological impact of that nuclide in safety assessments.

Errors quoted at 2.00 sigma are standard deviations used to indicate the statistical uncertainty of the measurements, providing an estimate of how much the results might vary.


Question: RWCU system isolation

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


Question: What are non-power reactors in nuclear licensing?

In § 55.40, this final rule revises paragraph (d) to replace the phrase "test and research reactors" with the phrase "non-power reactors."

This final rule adds new § 51.56 to clarify existing requirements for the submittal and content of environmental reports by applicants seeking a permit to construct, a license to operate, or a renewal of a license to operate, a non-power production or utilization facility.


Question: What is a 15x15 fuel rod design in nuclear reactors?

Analyses to evaluate normal operation criteria are performed on a sample design representative of a 15x15 fuel rod design with M5 cladding irradiated for up to three 24-month cycles in a Babcock & Wilcox reactor.

The report includes various calculations and methodologies used to assess the efficiency of the cooling system for the reactor, ensuring adequate heat removal during operation.


Question: What are the aging effects on components in the cooling systems due to water exposure?

Reduction of heat transfer due to fouling

There are no stainless steel piping, piping components exposed to closed-cycle cooling water >140°F in the screened-in portions of the Steam and Power Conversion Systems.


Question: Please update the records to reflect any changes concerning the devices registered under your name.

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

Please complete Section 4 for devices not in your possession.


Question: Tennessee Macroinvertebrate Index

Tennessee Macroinvertebrate Index scores were determined for each of the six monitoring stations. Tennessee Macroinvertebrate Index scores for East Fork Poplar Creek ranged from 14 (C rating, moderately impaired) to 42 (A rating, non-impaired).

Aquatic habitats present within Lot 6b of the study area include one ephemeral stream and one intermittent stream.


Question: safety light specifications

The Safety Light was manufactured on January 1, 2006, under model 2000 and has a serial number of 354169, with a decay rate of 9.500 and a value of 4.52.

The reactor core's thermal neutron flux must be monitored continuously to ensure optimal operating conditions.


Question: What is the purpose of the Reactor Coolant Drain Tank in nuclear reactors?

One tank is provided for each unit. The purpose of the RCDT is to collect leakoff-type drains inside the containment at a central collection point for further disposition through a single penetration via the RCDT pumps.

The tank size is adequate to accommodate 11 days of expected influent during normal operation.


Question: The NRC mandates comprehensive evaluations for risks to ensure safety in operations.

The NRC requires all licensees to conduct thorough safety assessments to determine the effects of isotopes and related products on the public and environment.

Eckert & Ziegler Isotope Products, a provider of isotopic products, operates under various codes and dates as part of their product descriptions.


Question: What is LCO 3.8.4 in the context of nuclear safety?

LCO 3.8.4 DC Sources - Operating is not met for Unit 1 and Unit 2. All four channels of vital DC subsystems are required to be operable. 1-AOI-21.03, Loss of 125V DC Vital Battery Bd III, directs the performance of Appendix A, Transfer of 125V DC Buses, to restore control power.

Plausible because the first part is correct. LCO 3.8.4 DC Sources - Operating is not met for Unit 1 and Unit 2. All four channels of vital DC subsystems are required to be operable. The second part is plausible because the applicant may incorrectly determine that 2-AOI-21.03, Loss of 125V DC Vital Battery Bd III, Appendix A, Major Equipment Availability, will restore control power.


Question: The studied pressure values reflect significant hoop stress on the nozzle of the component.

**Figure 7-17**
Stress contours due to unit internal pressure for PWR pressurizer vessel top head (CE design)


Question: What is the process for requesting document disclosure under FOIA?

I formally request such a waiver. I request a waiver of all costs pursuant to 5 U.S.C. §552(a)(4)(A)(iii) (“Documents shall be furnished without any charge … if disclosure of the information is in the public interest because it is likely to contribute significantly to public understanding of the operations or activities of the government and is not primarily in the commercial interest of the requester.”)

I am willing to pay any reasonable expenses associated with this request; however, as the purpose of the requested disclosure is in full conformity with the statutory requirements for a waiver of fees.


Question: What is heat flux in nuclear reactor systems?

Heat flux is defined as the amount of heat energy transferred per unit area per unit time, and it is typically measured in terms of fractions of nominal values for reactor safety analysis.

The table illustrates the startup sequence of an inactive reactor coolant pump, detailing the time in seconds and corresponding heat flux values.


Question: what is raw water in nuclear safety

Raw water is water that has not been demineralized or treated to any significant extent. This can include water for use in open-cycle cooling water and fire protection systems.

The Groundwater/Soil environment is the external environment for structural components buried in the soil where there is groundwater present.


Question: The NRC mandates procedures to maintain the reliability of critical safety systems.

The Residual Heat Removal System is a safety-related system designed to remove residual heat from the reactor core and maintain safe operating conditions.

The approved project plan includes a detailed timeline and budget allocated for each phase of construction.


Question: occupational dose monitoring

We will monitor individuals in accordance with the criteria in the section titled, ‘Radiation Safety Program–Occupational Dose’ in NUREG–1556, Vol. 9, Rev. 3, ‘Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Medical Use Licensees.’

We have developed and will implement and maintain written procedures for a program for training required under 10 CFR 19.12 for each group of workers, including - (i) topics covered, - (ii) qualifications of the instructors, - (iii) method of training, - (iv) method for assessing the success of the training, - (v) initial training, - (vi) annual refresher training.


Question: What are the safety inspections for Class II research and test reactors?

The audit evaluates the U.S. Nuclear Regulatory Commission's safety inspections of Class II research and test reactors, focusing on compliance with regulatory standards.

The distribution of this report includes various NRC offices and personnel who are involved in nuclear regulatory processes and oversight.


Question: The NRC outlines the importance of analyzing transient core power levels in determining safety margins.

Figure 4-12 illustrates the relationship between Transient Core Power, F₀, and Fuel Rim Temperature, particularly for GAIA Fuel during an accident scenario.

The investigation revealed the design features of the reactor system, including the advanced control systems installed to enhance operational efficiency.


Question: Reactor building non-essential return isolation valve

Reactor Bldg. Non-Ess. Return Auto. Iso. to Tran A and B Cooling Water HX's


Question: The NRC evaluates drainage management as a crucial component of environmental safety in uranium project licensing.

The crownpoint uranium project includes detailed provisions for managing drainage areas to protect the environment as part of the SUA-1580 license renewal application.

The scale of the project site is outlined with measurements in feet, which illustrates the geographical dimensions important for planning and safety assessments.


Question: What is the role of the main and startup feedwater system in a nuclear reactor?

Knowledge of the physical or control/protection logic relationship between the main and startup feedwater system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Knowledge of bus or division power supplies to the following: (CFR: 41.7)


Question: What are the anticipated cycle counts for the various types of scram transients during a 60-year operational period?

The staff noted that USAR Section 3.9.1.1.3 indicates that the following design cycles related to scrams were evaluated in the existing fatigue analysis for the HCUs: (1) 140 cycles of the "scram test" transient; (2) 160 cycles of "startup scram" transient; and (3) 300 cycles of "operational scram" transient.

LRA Section 4.7.5 indicates that USAR Section 3.9.2.2.1.6.4 "Hydraulic Control Unit (HCU)," documents that the HCUs were analyzed for faulted conditions including the effects of seismic and hydrodynamic loads.


Question: unclassified records in nuclear regulatory filings

NSIR staff informed us that, although no classified records were located, they did locate unclassified records.

These records, with some redactions, are enclosed.


Question: What are the requirements of a general license in nuclear regulation?

C. 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: What is the approach taken towards harvesting materials from decommissioned reactors?

The purpose of this effort is to develop a more systematic proactive "strategic" approach to ex-plant material harvesting, rather than the more reactive opportunistic approach to date.

I am working with Amy Hull in my branch on an effort associated with strategic harvesting of ex-plant materials.


Question: What is the significance of NUREG-0654 in regulatory filings?

The Common Emergency Plan must be consistent with NUREG-0654, Revision 2, which outlines the guidelines for emergency planning for nuclear power plants.

Florida Power & Light Company operates the St. Lucie Plant, which consists of Units Nos. 1 and 2.


Question: impingement area of influence

The impingement AOI is the region extending outward from the intake screens in which impingeable-sized aquatic organisms (i.e., juvenile and adult fish and shellfish) would not be capable of overcoming the velocities created by water withdrawals at the cooling water intake structure and, thus, would have a higher probability of being impinged upon an intake screen.

Duke Energy proposes no changes to the cooling system and no changes to the amount of cooling water withdrawals as part of SLR.


Question: What is the safety injection system (SIS) in nuclear reactors?

Control pressurizer water level with safety injection system (SIS).

Monitor system pressure and temperature.


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

| Nuclide Name | Energy (keV) | Yield (%) | Activity (pCi/grams) | Nuclide MDA (pCi/grams) | Line MDA (pCi/grams) |


Question: What are the acceptance criteria for pump vibration readings in nuclear plants?

Compare pump vibration readings to the following Acceptance Criteria, AND CHECK the appropriate box.

The performance step is critical to determine if Motor Vibration meets Acceptance Criteria.


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

Wt mean Activity (pCi/grams) provides a measure of the radioactive decay of a nuclide, which is crucial in assessing the safety and regulatory compliance of nuclear materials.

Errors quoted at 2.00 sigma represent the uncertainty in the measurements but do not pertain directly to the concept of nuclide activity.


Question: Reactor water level control during emergency scenarios

Once Reactor pressure lowers below 129 psig, the crew will re-inject and restore Reactor water level above -19°.

Main Turbine high vibrations will develop. The crew will enter AOP-66 to address the vibrations.


Question: proprietary information withholding

Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses.

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.


Question: What is the function of containment isolation valves in nuclear facilities?

Each containment isolation valve shall be OPERABLE.

Note: Enter applicable ACTION(s) for system(s) made inoperable by containment isolation valves.


Question: What are the requirements for a reactor license application?

The application must include a description of the detailed design, safety analysis, and operational plan that demonstrates compliance with regulatory requirements.

The system architecture should account for environmental assessments, including the impact on local wildlife and ecosystems.


Question: Nominal Trip Setpoint Calculation

The Nominal Trip Setpoint (NTSP) should be calculated using the equations below depending on the direction of process variable change when approaching the Allowable Value.

This determination of setpoints and allowable values has been moved to IP-E-0032 (Input 4.5.2), so these equations are no longer utilized within this calculation and have been deleted from this section.


Question: What best practices should be followed in preparing estimates for complicated technical projects?

Choose the best estimating method for each WBS element.

Estimate data sources and associated guidance can be found in Sections B.4, B.7.3, and B.7.4 and in Table B-1.


Question: What are the key components of a draft Safety Evaluation (SE) for nuclear projects?

The Safety Evaluation (SE) is a document that assesses the safety aspects of the proposed project and is an essential part of the licensing process for nuclear reactors.

Ngola Otto is involved with the contract management for the Office of Nuclear Reactor Regulation.


Question: What is the COBRA-FLX correlation in nuclear safety assessments?

The COBRA-FLX correlation is a method used to evaluate the critical heat flux in nuclear reactor safety analyses, particularly under various test conditions to ensure adequate cooling and prevent overheating of reactor components.

The report includes detailed information about the test assembly conditions used for the non-mixing grid tests, highlighting parameters such as pressure, temperature, and flow rates.


Question: isolation system response time

The instrument response times must be added to the PCIV closure times to obtain the ISOLATION SYSTEM RESPONSE TIME. ISOLATION SYSTEM RESPONSE TIME acceptance criteria are included in Reference 6.

The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.


Question: The impact of temperature on aquatic ecosystems is a crucial aspect of cooling systems at nuclear plants.

Temperature effects on sediment transport capacity are a significant concern for the environmental analysis of cooling systems at the Monticello Nuclear Generating Plant.

The assessment of groundwater contamination is vital for understanding the broader ecological impacts of the Monticello facility.


Question: What is the significance of the Nuclear Service Cooling Water Tower in reactor safety?

The Nuclear Service Cooling Water Tower el 128 ft 0 in., SSE Horizontal is critical for maintaining temperature control and ensuring that service cooling water is adequately supplied to various systems within the reactor facility.

The Control Building el 173 ft 0 in., OBE Vertical is designed to house the control operations of the reactor, ensuring a secure environment for personnel and equipment.


Question: The requirement for licensees to include specific contact information is crucial for establishing standing in petitions.

The Alliance and Collective also argued that applying the address and telephone number requirement, and other regulatory elements of standing, is *ultra vires* and extra-statutory.

In the Reply, the Alliance and Collective argued that the requirement to include an address and telephone number in the petition simply was a clerical omission and included supplemental affidavits with addresses and telephone numbers for themselves, but not for the members on whose behalf they seek representational standing.


Question: The analysis involves identifying different energy levels corresponding to various nuclides present in the sample being evaluated.

Identifying the ambient background is crucial for accurate measurements, as it ensures that the signals from the nuclides are clearly distinguished from external noise.


Question: What is the unit of measurement for radioactivity commonly used in licensing documents?


Question: The specified detection limits for radionuclides can significantly impact safety assessments.

Nuclide MDA (pCi/grams) defines the minimum detectable activity for each radionuclide, which plays a crucial role in regulatory compliance and safety evaluations.

The energies of different peaks are displayed with their corresponding counts per second in the specified table.


Question: The review process for environmental impacts during license renewal is influenced by specific categorization of issues such as Category 1 and Category 2.

The Staff's license renewal environmental review is guided by the 2013 GEIS (NUREG-1437, Rev. 1) and the “Standard Review Plan for Environmental Review of Nuclear Power Plants—Operating License Renewal” (“ESRP-LR”) (NUREG-1555, Supp. 1, Rev. 1) (June 2013). Like the applicant, the NRC staff is not required to address Category 1 impacts in its plant-specific EIS (SEIS), which it publishes as a supplement to the GEIS.

Contentions raising environmental issues in a license renewal proceeding are limited to those issues that are affected by license renewal and have not been addressed by rulemaking or on a generic basis.


Question: What is the significance of water content measurements in geological borings for nuclear facilities?

The water content, expressed as a percentage, is crucial in evaluating the properties of subsurface materials during site assessments for nuclear facility expansions.

The boring record indicates the depths at which samples were collected and their respective split spoon sampling recovery rates in inches.


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

Activity concentration is reported in picocuries per gram (pCi/g) for various radioisotopes, indicating the amount of radioactive material present in a given mass of sample.

The analytical results show various counts for radionuclides, with appropriate uncertainty values associated with each measurement.


Question: Site Area Emergency definition

Upon receipt of a classification of Site Area Emergency by RBS, the Parish dispatcher/communicator will notify the Director of the Office of Homeland Security and Emergency Preparedness, other key individuals, and supporting response organizations consistent with this classification and in accordance with established procedures.

All supporting response organizations will maintain standby status until verbal closeout or escalation of the emergency.


Question: The aging management program for structural bolting related to spent fuel pool gates will monitor material loss.

The DCPP Structures Monitoring AMP (B.2.3.33) will be used to manage loss of material for structural bolting associated with the spent fuel pool gates.

The spent fuel pool is critical for safely storing nuclear materials after they have been used in the reactor.


Question: What are the environmental conditions considered for the Seismic Isolation System?

The SIS design shall accommodate environmental degradation due to aging effects, creep, fatigue, operating temperature, radiation and exposure to moisture or damaging substances.

Structures directly interfacing with the SIS shall be designed with adequate rigidity to ensure all seismic isolators are engaged.


Question: What prevents LB-LOCA occurrences in nuclear reactors?

Layers of Defense that support prevention of LB-LOCA: NSSS piping system design (e.g. material selection, geometry…) NSSS piping system fabrication (Q/A, welding procedures, welder qualification, weld inspection…) NSSS normal and abnormal operating procedures that limit piping loads In-service Inspection Leak Rate Detection Many months for detection of small leak Multiple independent methods of detection Leak rate progression makes leakage more evident Undetected leakage to the point of piping rupture is not plausible

Confirmed by xLPR analysis probabilistic fracture mechanics analysis


Question: The analysis of Nickel-63 concentration is essential for understanding radiological safety.

Nickel-63 is reported with a result of -0.478 pCi/g and an uncertainty of +/-1.80 in the rad liquid scintillation analysis.

The report includes various analytical methods performed, which range from DOE EML HALS-300 to EPA 906.0 Modified.


Question: The NRC mandates that valves in the reactor coolant system be designed for maximum safety and reliability during operation.

All manual and motor-operated valves of the reactor coolant system that are 3 inches and larger are provided with double-packed stuffing boxes and stem intermediate lantern gland leakoff connections.

Materials of construction are specified to minimize corrosion/erosion and to ensure compatibility with the environment.


Question: JIS function activation process

The JIS function is established at this point. Activation of the JIC is coordinated with the offsite agencies and carries no time requirement.

IT personnel monitor critical digital assets remotely and respond any time an issue is identified.


Question: What are the average measurements from the Site Meteorological Tower data over the years?

The NRC maintains rigorous standards for the safety of nuclear facilities.


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

The Nuclear Regulatory Commission (NRC) reviews applications for nuclear power plant operating licenses to ensure compliance with safety regulations.

Weather conditions and geological assessments are crucial for determining the site suitability for construction projects.


Question: What are the aging effects on core barrel assembly welds?

Loss of fracture toughness due to neutron irradiation embrittlement

The environment surrounding the reactor coolant and neutron flux plays a critical role in assessing material performance.


Question: The analysis of the radionuclide concentration is expressed in pCi/g Dry.

The count date and time for the sample analysis was recorded as 09/23/2020 12:26.


Question: The qualifications of personnel in charge of fire protection should meet specific engineering and training standards.

THE POSITION RESPONSIBLE FOR FORMULATION AND IMPLEMENTATION OF THE FIRE PROTECTION PROGRAM SHOULD HAVE, WITHIN HIS ORGANIZATION, OR AS A CONSULTANT, A FIRE PROTECTION ENGINEER (WHO) IS A GRADUATE OF AN ENGINEERING CURRICULUM OF ACCEPTED STANDING AND (WHO) SHALL HAVE COMPLETED NOT LESS THAN SIX YEARS OF ENGINEERING ATTAINMENT INDICATIVE OF GROWTH IN ENGINEERING COMPETENCY AND ACHIEVEMENT, THREE OF WHICH SHALL HAVE BEEN IN RESPONSIBLE CHARGE OF FIRE PROTECTION ENGINEERING WORK.

THE RECOMMENDATIONS FOR ORGANIZATION, TRAINING, AND EQUIPMENT OF 'PRIVATE FIRE BRIGADES' AS SPECIFIED IN NFPA NO. 27, ARE CONSIDERED APPROPRIATE CRITERIA FOR ORGANIZING, TRAINING, AND OPERATING A PLANT FIRE BRIGADE.


Question: What is the role of SGDB sample isolation valves in managing steam generator safety?

SGDB sample isolation valve HV-955B isolates SGDB sample from steam generator No. 1 upon remote-manual operation.


Question: What does it mean to transfer a nuclear device to another general licensee?

Transferred to another general licensee (Complete Parts 2 and 3)

The NRC Device Key is a unique identifier used in regulatory filings.


Question: The evaluation score reflects the overall scenic quality and environmental impact of the area around the HCS.

The BLM Scenic Quality Inventory protocol allows for assigning a score at any integer value within the maximum and minimum score range listed for an individual key factor.

While there are no buildings or structures, modifications have previously cleared surrounding areas of forest and natural vegetation.


Question: The NRC mandates that when losing Automatic Depressurization System functionality, the associated valves must be declared inoperable.

F.1 Declare Automatic Depressurization System (ADS) valves inoperable.


Question: nuclear materials safety regulations

Room in accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390, “Public Inspections, Exemptions, Requests for Withholding.”

Sincerely, David E. Hills, Chief Materials Control, ISFSI, and Decommissioning Branch Division of Nuclear Materials Safety


Question: threshold limit concentration for Nickel in monitoring wells

Nickel has an ACL (Acceptable Concentration Level) of 98 mg/L in Monitoring Well 32-59.

The table lists various dates along with their respective concentration levels for different metals.


Question: The NRC stipulates that licensees must take specific actions when testing components of the coolant system.

**NOTE** While 1-LPF-62-93C, CHARGING HEADER FLOW CONT, is out of service for removal of M&TE, LCO 3.3.4 may be applicable.

Personnel injury and contamination could result if safety and radiological control precautions are NOT observed.


Question: What are the requirements for inservice testing of pumps and valves in nuclear power plants?

Specification 4.0.5 establishes the requirement that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with a periodically updated component of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a.

The purpose of this specification is to ensure that system and component OPERATIONAL requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility.


Question: What is a Performance-Based Regulatory Approach in nuclear licensing?

A Performance-Based Regulatory Approach (PBRA) emphasizes the use of performance indicators and risk assessments to evaluate the safety and adequacy of nuclear facilities, focusing on actual performance outcomes rather than prescriptive regulations.

The proposal outlines the need for advanced technology integration in the next generation of nuclear reactors to enhance efficiency and reduce operational costs.


Question: The removal of a power generation facility from service and addressing residual radioactivity are critical for process safety during decommissioning.

Activities that are unique to the termination of operations and decommissioning of a nuclear power generating facility include the safe removal of the facility from service and the reduction of residual radioactivity to a level that permits release of the property under restricted conditions or unrestricted use and termination of the license.

The environmental consequences of decommissioning can generally include the following: short-term impacts on air quality and noise from the deconstruction of facility structures.


Question: water table elevation

Water Table Elevation1 (ft msl) refers to the level at which the ground is saturated with water and is measured in feet above mean sea level.

The moisture content is measured by weight percent indicating the amount of water present in a sample relative to its total weight.


Question: radiolytic gas generation in radioactive waste systems

Radiolytic gas generation in highly loaded submerged demineralizer system (SDS) vessels was calculated to be significant.

A pressure relief system, consisting of a burst diaphragm and micropore graphite filter, was also added to each SDS vessel to prevent the uncontrolled, long-term buildup of noncombustible gas mixtures.


Question: The AACDG serves as a backup power source to ensure continuity in case of primary power loss.

The AACDG (commonly referred to as the station blackout diesel generator) can also be started and connected, from the Control Room, to power either 4.16 kV switchgear.

With respect to uniform loading, each train is relatively balanced with the other, with the significant exception being the motor-driven EFW pump, which is powered from red train 4.16 kV switchgear 2A3.


Question: What is the significance of rod average burnup in nuclear reactors?

The burnup values applied in the analysis do not exceed the rod average burnup of [ ] gigawatt-days per metric ton of uranium (GWd/MTU) for Zircaloy-4 and up to [ ] GWd/MTU for M5 cladding.

This analysis was performed using fuel with a rod outside diameter of 9.5 mm.


Question: What constitutes emergency classification levels in nuclear safety protocols?

In the context of a nuclear facility, emergencies are classified into three distinct levels: Alert, Site Area Emergency, and General Emergency, each denoting the severity and nature of the incident and the required response actions.

The checklist outlines the process for monitoring and managing public information during an emergency, ensuring effective communication and coordination among the participants.


Question: materials used in nuclear reactor construction

Concrete from Zorita is a specialized material harvested for use in nuclear reactor construction, ensuring structural integrity and safety compliance.

The timeline for project completion spans multiple phases, including initial discussions, harvesting, and testing, which involve extensive coordination efforts.


Question: control rod movement during alarm conditions

The NSO selects a control rod and observes the white backlight is LIT on the Select Matrix and Four Rod Display, but NOT LIT on the Full Core Display.

Unit 1 is operating at 50% power with a startup in progress when the following occurs:


Question: ITAAC deletion process


Question: The NRC mandates that licensees monitor the Heat Flux Hot Channel Factor to ensure compliance with safety regulations.

LCO 3.2.1 Fₓ(Z), as approximated by Fₓⁱ(Z) and Fₓʳ(Z), shall be within the limit specified in the COLR.

In the event that the neutron flux exceeds specified limits, the operators must perform a series of safety checks and adjustments to maintain operational integrity.


Question: What is the NRC's approach to determining the emergency planning zone size for small modular reactors and other advanced technologies?

The NRC is establishing a consequence-oriented, technology-inclusive approach to EPZ size determinations for SMRs and ONTs.

As mentioned in the “Technical Basis” section of this document, NUREG-0396 established the planning basis for EP and established EPZs for large LWRs based on the conclusion that the objective of emergency response plans should be to provide dose savings for a spectrum of accidents that could produce offsite doses in excess of the EPA PAGs.


Question: LOCA frequency assessment

So, in terms of plant changes and LOCA frequencies, again they need to demonstrate that they don’t -- There’s, sorry, a couple different options here.

Option one is to directly evaluate the effects of the change on those variables that are most important that affect both the frequencies that are articulated in 1829.


Question: seismic isolation system qualification

The overall RIPB SIS seismic design and seismic classification process is illustrated in Figure 7-1.

Feedback from SPRA results used to evaluate SSC seismic risk significance and update SSC fragilities to meet NEI 18-04 risk targets.


Question: What is the function of the Service Air System in a nuclear plant?

The Service Air System is a plant-shared system. The Service Air System consists of two air compressors located in the turbine building and associated piping that extends throughout the plant including into the reactor buildings.

The aging management review results for these components are provided in Table 3.3.2-11, Control Air System - Summary of Aging Management Evaluation.


Question: The NRC stipulates that licensees must evaluate the safety and environmental impacts of their operations.

The General Design Criteria for nuclear power plants serves as the basis for licensing and regulation by the NRC, ensuring public health and safety are protected during operation.

The report also includes detailed thermal performance data measured during the test, indicating various temperature gradients and their impacts on efficiency.


Question: What is a Limiting Fault in nuclear safety analysis?

This event is considered to be a Limiting Fault.

The case of recirculation pump shaft breaking represents the extremely unlikely event of instantaneous stoppage of the one recirculation pump motor and is considered as an accident.


Question: What are the average aircraft flights at local airports near GGNS?

The average number of aircraft flights is reported as 134 per week at Tensas Parish Airport, which services single-engine aircraft.

The Vicksburg Municipal Airport has 18 based aircraft and accommodates 57 flights a day, serving primarily local and transient traffic.


Question: What are the candidates for NRC examination scenarios?

The validation for the examination scenario was performed by E Crew.


Question: The NRC mandates that licensees must demonstrate compliance with technical standards.

The Nuclear Regulatory Commission (NRC) requires that all licensees maintain compliance with established technical standards governing reactor operations and safety protocols.

This software is designed to enhance user experience by providing various customization options.


Question: What is the importance of Replacement Steam Generators in nuclear power plants?

The Replacement Steam Generator (RSG) Project at Watts Bar Nuclear Plant (WBN) Unit 2 involves the installation of new steam generators to replace the existing units in order to enhance operational efficiency and safety.

The teleconference provided an overview of the WBN Unit 2 operational history, emphasizing the milestones achieved in the last inspection cycle.


Question: What is the Risk Informed Completion Time Program?

Specifically, the Risk Informed Completion Time Program is contained in TS 6.5.20 rather than 5.5.20 as shown in the TSTF-591 SER.

The differences for the proposed change are administrative and do not affect the applicability of TSTF-591 to the WSES TS.


Question: The assessment of FLEX and HCVS components requires careful consideration of their associated failure rates.

“FALSE” indicates FLEX and HCVS are credited, with component failure probabilities based on similar types of equipment; “TRUE” indicates no credit for FLEX or HCVS (which is similar to assigning failure probabilities of 1.0 to FLEX and HCVS components).

The table summarizes data on various events and their associated probabilities, which is essential for risk assessment in nuclear regulatory activities.


Question: transition break size definition

Transition break size (TBS) for reactors licensed under this part before December 31, 2015, is a break area equal to the largest cross-sectional flow area of the reactor coolant pressure boundary piping excluding the hot leg, cold leg, or crossover leg piping for a pressurized water reactor, or the largest cross-sectional flow area of either the feedwater line or residual heat removal line inside containment for a boiling water reactor, or a plant-specific alternative break area.

The staff plans to continue engaging with industry on other approaches (e.g. modified Alternatives 4 and 5) via licensing interactions and workshops.


Question: Review your conditions outlined in the attached license renewal following the recent update.

Please find enclosed Amendment No. 33 to NRC License No. 50-27667-01 renewing your license in accordance with application dated June 26, 2020, letter with attachments dated November 20, 2020, and E-mail with attachments dated December 1, 2020.

The NRC’s Safety Culture Policy Statement became effective in June 2011.


Question: The permit stipulates that samples must be indicative of the discharge and collected under normal operating conditions.

Samples shall be collected as the following location just prior to discharge and shall be representative of the discharge.

Total Suspended Solids is limited to 30 mg/L and must be monitored monthly.


Question: What is the purpose of the Structures Monitoring program in managing concrete aging effects?

The Structures Monitoring (B.2.1.33) program is expected to adequately manage the loss of material (spalling, scaling) and cracking due to freeze-thaw that could occur in below grade inaccessible concrete areas of Group 6 structures.

Further evaluation is recommended to determine the need for a plant-specific AMP or plant-specific acceptance criteria to Structures Monitoring AMP, to manage this aging effect.


Question: The NRC sets specific thresholds for radiation levels to ensure safety within different zones.

The Action Level is a threshold that, if exceeded, indicates that additional actions are required to manage radiological safety.

The logged readings are recorded in counts per minute (cpm), which are essential for monitoring various operational parameters.


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory reviews and approvals conducted by the Nuclear Regulatory Commission to ensure that a facility can operate safely and in compliance with federal standards.

In addition to safety concerns, the economic impact of nuclear energy on local communities must be considered during planning and development.


Question: What are the requirements for job performance measures in reactor operations?

Determine the number of control room system and in-plant system job performance measures (JPMs) to develop using the following table:

Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function.


Question: What is the Minimum Detectable Concentration in radiological analysis?

The Report Date for the Certificate of Analysis is August 15, 2019.


Question: What are the components of decommissioning costs according to NRC regulations?

The elements of decommissioning costs under 50.75(c)(2) are assigned to three categories: 1. those that are proportional to labor costs, \(L_x\); 2. those that are proportional to energy costs, \(E_x\); 3. those that are proportional to burial costs, \(B_x\).

Effective July 1, 2000, the South Carolina LLW burial site applied different price schedules for waste from states within and outside the then newly created Atlantic Compact (comprised of South Carolina, Connecticut, and New Jersey).


Question: pressurizer design criteria

The tank design is based on the requirement to accept a steam discharge from the pressurizer equal to 110% of the pressurizer steam volume at full load.

The pressurizer is designed and constructed in accordance with the ASME Code, Section III, 1968.


Question: What is the significance of the energy measured in keV for nuclear isotopes?

The table lists various nuclear isotopes along with their energy levels measured in kilo-electronvolts (keV), which indicates the energy of the gamma radiation emitted by each nuclide.

The yield percentage indicates the efficiency of production for each nuclide, showing how much of the isotope can be successfully produced or detected.


Question: transfer to cold leg recirculation procedures

4. Determine the indications that are monitored to ensure proper system/component operation for each step in 2-EOP-LOCA-3.

The proposed references to be provided: None.


Question: What is the significance of a certificate of service in regulatory filings?

Pursuant to 10 CFR 2.305, I certify that, on this date, a copy of the foregoing filing was served through the Electronic Information Exchange (the NRC’s E-Filing System) in the above-captioned docket.

The NRC requires that all operational reactors comply with the safety standards outlined in 10 CFR Part 50, ensuring that they operate within prescribed limits to protect public health and safety.


Question: The NRC requires that every environmental report includes a detailed assessment of fuel transport and radioactive waste movement.

The impacts associated with transporting fresh fuel to one 1,000 MWe model light-water reactor and with transporting spent fuel and radioactive waste (LLRW and mixed waste) from that light water reactor are provided in Table S-4 in 10 CFR 51.52.

Based on the February 24, 2022, NRC Memoranda and Orders, the GEIS did not address SLR.


Question: steam jet air ejectors

Steam Jet Air Ejectors are an essential component of the Condenser Air Ejection System, designed to maintain a specified pressure within the condenser by removing air and non-condensable gases.

The auxiliary steam has a pressure of 170 psig and provides 1238 Btu/lb, which is important for various operational processes in the system.


Question: airborne radioactive material evaluation

This section deals with the models, parameters, and sources required to evaluate airborne concentration of radionuclides during plant operations in various plant radiation areas where personnel occupancy is expected.

The exposure rates at the surface of the reactor cavity and spent fuel pool water are dependent on the purification capabilities of the reactor cavity and spent fuel pool cleanup systems.


Question: proposed transmission corridor

The Proposed Transmission Corridor is identified by the symbol ■■■ in the site legend, indicating the planned route for the transmission line.

The data sources include several reports from the IDNR dating back to 1984, as well as census data from the U.S. Census Bureau, used to provide demographic and geographic context.


Question: boric acid corrosion management

The DCPP Boric Acid Corrosion AMP is an existing condition monitoring AMP that manages loss of material due to boric acid leakage. The DCPP Boric Acid Corrosion AMP monitors mechanical, electrical, and structural components within the scope of LR that are susceptible to boric acid corrosion from systems that contain reactor coolant or treated borated water.

The reactor fuel assemblies were designed to withstand a range of operational conditions, ensuring safety and efficiency throughout their operational life.


Question: radionuclide activity measurement

The table includes various radionuclides along with their respective activity values measured in g dry.

The NRC report also provides an overview of safety protocols implemented during reactor operation.


Question: The evaluation of MEK degradation indicates a half-life span between several days to months, depending on site data.

Using a range of half-lives of 20 to 200 days (which is consistent with literature values discussed earlier)—and assuming no further increases of MEK concentrations (consistent with concentration trends at YO-119 and YO-117D)—it would take between 3 months and 3 years for the June 2010 MEK concentrations to drop below cleanup criteria (Exhibit 13).

The historic acetone data at downgradient wells showed sporadic hits above the 5 µg/L performance standards, with a maximum of 50 µg/L detected in April 2001 at YO-12A.


Question: The NRC mandates that all operational equipment is regularly inspected and assessed to ensure safety compliance.

The equipment listed includes operational air handling units such as CRA-P-1 and SGR-AHU-1A, which are critical for maintaining the environmental conditions within the facility.

The design of the terminal includes a sophisticated communication system that enhances data transmission and feedback among the various components.


Question: To evaluate the potential impact of tephra falls on stream channels, a bulking factor is applied to account for the additional volume that can be integrated into existing debris flows.

The bulking factor typically evaluates the additional amount of material incorporated into an already established debris flow, generated by instantaneously released material upstream and rapid generation of peak-flow conditions.

This simplified approach indicates that a potential debris flow would not overtop the 6–8-m-deep inactive channels during flooding events, following deposition of 54–113 cm of tephra from a Yellowstone caldera eruption.


Question: What are the regulatory considerations for fusion reactors according to the NRC?

The 2009 NRC Memo suggested that an “additional consideration involves the potential benefits of the NRC establishing a national regulatory framework for fusion devices instead of requiring various State and local agencies to develop programs to address this new technology.”

States already handle radioactive sources under Parts 20 and 30 through the Agreement State Program (with 39 states participating) and the NRC exerts oversight through regular audits, so national consistency is already maintained.


Question: The NRC mandates that licensees monitor the Reactor Vessel Water Level to ensure safe operating conditions.

4. Reactor Vessel Water Level (RVWL) – Low, Level 3 (Four channels)

One RVWL – High channel in each RPS trip system


Question: bounding power history envelope

Figure 6-1: Bounding Power History Envelope for CE 14x14 Creep Collapse Initialization Sample Case

The reactor pressure vessel is a crucial component in ensuring that the reactor operates within safe pressure levels.


Question: The NRC mandates that a minimum detectable amount for radiological samples is calculated based on specific formulas that take into account background counts.

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

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


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

A safety analysis report (SAR) provides detailed information on the safety measures and analyses that have been performed to evaluate and ensure the safety of a nuclear facility in compliance with regulatory requirements.

The economic impact of new nuclear reactors can significantly influence energy prices and market competition.


Question: What is the activity level of isotopes used in nuclear devices?

The manufacturer name for the device listed is THERMO MEASURETECH.


Question: defense in depth and diversity analysis

The D3 coping analysis is part of a planned license amendment request to support a digital modernization project at Limerick.

The NRC staff initiated a Regulatory Audit in accordance with NRR Office Instruction LIC-111 to support review of the D3 evaluation.


Question: The reactor recirculation system is essential for maintaining proper flow and temperature control during operation.

The Reactor Recirculation Loop A is part of the systems designed to ensure efficient operation and safety of the reactor.

The table lists various piping configurations and their respective isolation valve operators.


Question: what is the boron concentration during refueling

The boron concentration of the refueling water and the reactor coolant is maintained at a concentration that will, together with the negative reactivity of control rods, keep the core 5 percent or more Δk/k subcritical during the refueling operations.

Fuel is moved between core positions or between the reactor vessel and the refueling canal by the reactor manipulator crane.


Question: What procedures are in place to ensure the safety of workers in radiological controlled areas?

Rad Haz Areas Exposure Controls procedures are implemented annually to manage and mitigate exposure risks for workers in areas with radiological hazards.

The status of inspections can vary greatly depending on safety protocols and the type of equipment being assessed.


Question: What is the purpose of a safety inspection report in the nuclear regulatory process?

The U.S. Nuclear Regulatory Commission (NRC) conducts safety inspections to ensure that licensees comply with regulations and standards for safe nuclear operations.

The NRC holds various workshops every year to discuss advancements in nuclear technology and safety measures.


Question: What is the role of the Division of Safety Systems in nuclear regulatory processes?

The Division of Safety Systems is responsible for overseeing the safety and regulatory compliance of nuclear facilities.

Theresa Clark manages various operational aspects within the division.


Question: seismic probabilistic risk assessment

SNC is approved to utilize the seismic probabilistic risk assessment (SPRA) model for use in the categorization process rather than the previously approved seismic margin approach.

Lead test assemblies (LTAs) 7S1, 7S2, 7ST3, and 7ST4 contain fuel rods that include advanced coated cladding features and doped or standard fuel material.


Question: What are NHPA Section 106 conclusions in nuclear regulatory filings?

The NHPA Section 106 conclusion letters are formal documents issued as part of the compliance process with the National Historic Preservation Act, which require federal agencies to consider the effects of their undertakings on historic properties.

The reactor containment structure is designed to withstand significant external events, providing protection for both the reactor and the surrounding environment.


Question: Surveillance Frequency Control Program

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

The Control Room Envelope Habitability Program includes procedures for assessing air quality and maintaining the integrity of the control room environment.


Question: failure response of safety systems

The PPS protective provision in case the safety system fails to accomplish its safety function is to fail to a known state. In the case of RPS functions, the failed state is to initiate a reactor trip.

The LGS design basis events as defined in Chapter 15 for LGS are not changed as a result of this PPS project.


Question: nuclear regulatory commission audit schedule

The schedule details various activities, including meetings and site visits between the U.S. Nuclear Regulatory Commission (NRC), Energy Harbor, and contractor subject matter experts (SMEs).

The entrance meeting was held to discuss regulatory compliance and safety protocols related to the plant operations.


Question: The temperature of the ultimate heat sink (UHS) plays a crucial role in determining the heat removal capability of emergency systems following a loss of coolant accident.

For the limiting UHS post-LOCA peak temperature case the calculated UHS temperature profile shows the temperature initially decreases and remains below the starting temperature until about 36 hours after the event.

Routed through the SX cooling coils, the Chilled Water (WO) cooling coils, and the fan and discharge duct (one for each RCFC unit).


Question: What is the significance of thermal discharge in aquatic ecosystems near nuclear power plants?

The DNPS thermal plume has not caused a significant increase in abundance. Striped bass, silver carp, and grass carp have been collected only during one or two years since 1991.

The two entrainment and two impingement studies conducted at DNPS over a 13-year period (2005–2018) demonstrate a stable aquatic community at the site, with a relatively unchanged fish species composition.


Question: The report presents findings on the concentration of Americium-241 in soil samples collected from the site during the restoration project.

The project involves multiple phases, including site restoration and environmental assessments to ensure compliance with regulatory requirements.


Question: emergency action level

Given the following sections of EPIP-EPP-01-EAL, Emergency Action Level Matrix / Unit 1 (see next page).

The plant is operating at 100% power with the following sequence of events:


Question: It is required that containment isolation valves must remain functional during Mod-1 to Mod-4 operations.

LCO 3.6.3 Each containment isolation valve shall be OPERABLE.

Penetration flow path(s) except for containment shutdown purge valve flow paths may be unisolated intermittently under administrative controls.


Question: The evaluation of flood protection features was conducted to ensure compliance with the licensing basis.

Deficiencies and non-conforming or unanalized conditions were not observed during the walkdown of flood protection features that are the subject of the FWR.

Channel diversions are not expected to produce a flood event that could affect the site. No flooding elevation is given in the VEGP 1 & 2 FSAR (Section 2.4.9 of Reference 1).


Question: What is the Control Room Envelope volume for nuclear facilities?

Volume of Control Room Envelope, ft³ (Ref. 5, 4, Attachment L): 240,000

Assumed Unfiltered In-leakage, cfm (Ref. 28-31): 436


Question: What is the significance of the design control process in nuclear licensing?

The design control process ensures that any design changes affecting safety will be properly analyzed, reviewed, and implemented with appropriate regulatory oversight.

Nuclear power plants are equipped with numerous safety systems designed to prevent reactor core damage during operation.


Question: The NRC has specified parameters for the reset and drop-out values of degraded voltage relays in their documentation.

New Allowable Values in terms of secondary voltage at the relays:


Question: To comply with health regulations, water samples must be analyzed for contaminants such as lead and nitrate levels.

Nitrate (NO3) and nitrite (NO2) levels are monitored and must adhere to the Maximum Contaminant Level (MCL) or action level standards set by regulatory authorities.

The data collected from wells shows varying levels of carbon disulfide and trichloroethene, but does not provide specific action levels for those substances.


Question: What is a Nuclear Condition Report (NCR) in regulatory assessments?

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function: Nuclear Condition Report (NCR) 2354739, Investigate/repair overload on 1WL-39A, on October 25, 2020.

The inspectors observed and evaluated licensed operator performance in the control room during Unit 1 refueling outage, on October 15, 2020.


Question: What are the consequences of making false statements in NRC applications?

It is a criminal offense to knowingly and willfully make a false statement or representation on applications or correspondence (18 U.S.C. 1001).

When the application references commitments, those items will be incorporated into the license and therefore will become binding regulatory requirements.


Question: basement faults

Faults mapped within the site vicinity fall into two categories: Piedmont faults exposed in northernmost DE, and basement faults beneath the Coastal Plain strata.

The 5-mi. radius of the PSEG Site contains no tectonic faults, folds, or structures.


Question: time-limited aging analysis

A fatigue analysis is considered to be a time-limited aging analysis (TLAA) as defined in 10 CFR 54.3 when it is within the current licensing basis and is based upon transient cycle assumptions associated with 60 years of plant operation.

Table 2 contains all of the Aging Management Review information for the plant, including whether or not it aligns with NUREG-2191 recommendations.


Question: What is the role of the UFSAR in nuclear plant operations?

The station shall be operated in accordance with the license granted by the United States Nuclear Regulatory Commission (NRC), the Technical Specifications, and the Updated Final Safety Analysis Report (UFSAR), and the Operational Quality Assurance Program.

In addition to these, Virginia Power retained the following consultants: Site geology, hydrology, and seismology: Dames & Moore, Inc.


Question: What are the functions of structural fire barriers in nuclear facilities?

Structural Fire Barriers: Walls, Ceilings, and Floors - Fire Barrier - Flood Barrier - HELB Shielding - Shelter, Protection - Structural Support

Concrete (Inaccessible Areas): All - Fire Barrier - Flood Barrier - HELB Shielding - Missile Barrier - Shelter, Protection - Structural Support


Question: What is the significance of the MDA value in nuclear measurements?

MDA value not calculated; Half-life too short to be able to perform the decay correction; CAUTION: MDA value is inconsistent with Currie MDA at 95% confidence level.

The Energy (keV) of Y-88 is 898.04, and it has a yield of 93.70%.


Question: risk-informed categorization in nuclear power

10 CFR 50.69 "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," becomes effective

Regulatory Guide (RG) 1.201, Rev 1 is published and endorses NEI 00-04 as an acceptable method for complying with the requirements in 50.69.


Question: The NRC has outlined provisions regarding the compliance with ASME Code requirements for examination coverage in the Inservice Inspection program.

In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Entergy Operations, Inc. (Entergy), requests NRC approval of a proposed relief request associated with the Inservice Inspection (ISI) Programs for the cited Entergy nuclear power plants.

The proposed alternative will allow Entergy to ensure safety while adhering to regulatory requirements.


Question: What are the hazards that affect low-pressure ECCS suction strainers during recirculation?

Hazards of particular interest are those that may introduce phenomena leading to significant challenges to the effectiveness of low-pressure ECCS suction strainers during recirculation (i.e., ECCS water suction from suppression pool, and return to suppression pool via LOCA breaks).

Design changes and procedure changes are reviewed for their impact on the PRA model.


Question: The calculation of margins between the ATSP and ALs is essential for determining system performance.

Margin to Lower AL (MARRL) ATSP = AL + (Z + * MAR) Rearranged and with dropout terms: MARDO = ATSPDO - AL - Z* = ATSPDO - NTSPDO

Performing the same for each relay, new ATSPR_4K in V at the 4160 V bus:


Question: The calculation of counting uncertainty is performed to ensure confidence in the results reported.

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

The project includes various client sample IDs to maintain organization and tracking of the samples.


Question: nuclear licensing terminology

The Browns Ferry Nuclear Plant Subsequent License Renewal Application is a document submitted to the Nuclear Regulatory Commission to request the continued operation of the facility under a renewed operating license, demonstrating compliance with safety and regulatory requirements.

The Tennessee Valley Authority conducts various environmental protection activities to manage construction and maintenance operations effectively.


Question: What are the hydrologic characteristics required for an ESP application?

Describe the hydrologic characteristics of the proposed site.

Provide the existing and projected population profiles for the area around the site.


Question: The operator must ensure that the levels in the RVLIS are maintained within specified limits during operations.

Check RVLIS Level - GREATER THAN RCP-DEPENDENT VALUE

Operator monitors parameters while venting and terminates venting when criteria reached.


Question: What does severity level imply in nuclear regulation?

Severity Level means violations are categorized in terms of five levels of severity with Severity Level I assigned to the most significant violations and Severity Level IV and Level V violations are minor violations that if left uncorrected would lead to more serious violations.

The technical staff members are responsible for conducting inspections and reactive inspections (investigations) in accordance with applicable procedures, rules, regulations and instructions and for categorizing the apparent violations, if any, observed during these inspections.


Question: The NRC is updating their research related to dissimilar metal welds, specifically focusing on CASS.

NRC: Update on CASS (dissimilar metal welds)

IRSN will re-connect NRC (Carol Nove) with Gerard Cattiaux.


Question: The NRC requires a detailed application for the renewal of operating licenses for nuclear power plants to ensure compliance with safety and regulatory standards.

CEG has prepared this environmental report (ER) in conjunction with its application to the NRC for a subsequent license renewal (SLR) of the DNPS OLs, as provided by the following NRC regulations and guidance (NRC 2013a):

DNPS is located along the Illinois, Des Plaines, and Kankakee Rivers in Grundy and Will Counties, Illinois.


Question: occupancy factor in nuclear safety assessments

Location Occupancy Factor Data is critical for assessing how occupancy influences potential radiation exposure in various compartments over time.


Question: spent resin disposal process

The resin in the spent resin storage tank is loosened before disposal by sending pressurized nitrogen or sluice water through the six sparging nozzles in the tank.

When operating without steam generator leakage or when the system effluent is being recycled back to the main condenser, there will be essentially zero release of radioactivity from this system.


Question: The regulation stipulates that licensees must provide data on reactor operations to ensure compliance with safety protocols.

In accordance with 10 CFR 50.34, applicants for a construction permit or an operating license shall submit a safety analysis report that evaluates the potential hazards associated with the proposed nuclear facility.

The amount of cooling water required for a reactor can significantly impact its thermal efficiency and operational costs.


Question: What is the purpose of a Probabilistic Risk Assessment (PRA) in nuclear safety?

Probabilistic Risk Assessment (PRA) is a systematic process for evaluating the risks associated with nuclear power operations, focusing on identifying potential accident scenarios and their likelihoods, assessing plant responses, and estimating the consequences of such events.

The NRC requires all nuclear facilities to develop comprehensive environmental impact statements, which include detailed evaluations of potential ecological consequences.


Question: acceptance criteria for reactor measurements

The acceptance criteria for replicate static measurements is that the same conclusion is reached for each measurement. This is defined as the replicate measurement being within 20% of the standard measurement.

The standard measurement values show variations in assessed activities across different quality control cases.


Question: The regulations stipulate that licensees must demonstrate compliance with technical specifications.

Under 10 CFR Part 50, a licensee must submit an application containing a safety analysis and a clear description of the proposed facility's systems and structures.

The cooling system of the reactor uses both forced and natural circulation to maintain optimal operating conditions.


Question: What criteria are used to assess examination outlines for nuclear reactor operators?

Assess whether the outline was systematically and randomly prepared in accordance with Section D.1 of ES-401 or ES-401N and whether all K/A categories are appropriately sampled.

Verify that the systems walkthrough outline meets the criteria specified on Form ES-301-3:


Question: nuclear regulatory licensing process

The licensing process for nuclear facilities involves a thorough review of the applicant's qualifications, safety assessments, and compliance with established regulatory standards.

The analysis includes operational data that reflects the efficiency of the current reactor systems under varying temperature and pressure conditions.


Question: The collected rainfall data at various measurement stations is crucial for hydrological assessments.

Hourly Rainfall Distribution - Thissen Polygon - Lake Anna Drainage Basin.

The readings at each station are critical for understanding the area's topography.


Question: What are the licensing requests for regulatory relief during the COVID-19 pandemic?

The NRC staff has reviewed and approved several licensing requests seeking temporary flexibilities to maintain the safe and secure operation of the U.S. nuclear fleet during the pandemic.

A summary of approved licensing actions provided during this reporting period are summarized below.


Question: Understanding the operational responses related to abnormalities during reactor operations is crucial.

Knowledge of system setpoints, interlocks, and automatic actions associated with emergency and abnormal operating procedure entry conditions.

Drop control rods may lead to changes in plant configuration but require careful monitoring of overall reactor stability.


Question: The aging management program will focus on identifying and mitigating issues in inaccessible concrete areas.

The Structures Monitoring (B.2.1.33) program is expected to adequately manage the loss of material (spalling, scaling) and cracking due to freeze-thaw that could occur in below grade inaccessible concrete areas of Group 6 structures.

As a result, the Structures Monitoring (B.2.1.33) program is expected to adequately manage the loss of material (spalling, scaling) and cracking due to freeze-thaw that could occur in below grade inaccessible concrete areas of Group 6 structures.


Question: The integrity of piping components, particularly those subject to loss of material due to environmental factors, is crucial for maintaining safety standards.

Piping, Piping Components | Leakage Boundary (spatial) | Stainless Steel | Demineralized Water (Int) | Loss of Material | Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19) | VIII.E.SP-87 | 3.4-1, 016 | A

The aging effect on buried piping, including cracking and changes in material properties, necessitates specific inspection protocols.


Question: Total pounds per acre for nuclear site planting

Total pounds per acre for various plant species recommended by NRC staff in CUP FEIS is 20.0 for Northern Desert Plant Species and 21.5 for Pinyon/Juniper and Ponderosa Pine species.

Western wheatgrass (Arriba) is recommended for sandy sites, while Pinyon/Juniper species thrive in clay and loamy sites.


Question: weighted mean activity of nuclides

The weighted mean activity is calculated using the contribution of each nuclide's activity based on its presence in a sample, as shown by the values for K-40, PB-212, BI-214, PB-214, and RA-226 in the table.

The comments section indicates potential processing notes and additional information related to the accuracy of the nuclide analysis, such as uncertainties and exclusions from the results.


Question: nuclear security processes

Nuclear security is an essential part of safeguarding nuclear facilities and materials from unauthorized access and potential threats, ensuring both public safety and regulatory compliance.

The versatility of the reactors allows for various configurations that can accommodate different types of fuels and operational requirements.


Question: What is critical heat flux in nuclear engineering?

The Critical Heat Flux (CHF) is the minimum heat flux at which a transition from nucleate boiling to film boiling occurs, which can lead to overheating and damage in nuclear reactors.

The 2006 CHF Look-up Table provides an extensive analysis of various thermal hydraulic parameters for different reactor configurations.


Question: nuclide activity measurement

The weighted mean activity of CS-137 is reported as 3.95E-02 pCi/grams, indicating its concentration in a specific gram of sample material.

Confidence levels for different nuclides can vary; for instance, PB-212 has a confidence of 0.552, which implies it is less reliable than CS-137 with a confidence of 0.999.


Question: What is k-eff in nuclear engineering?

The k-eff value, or effective neutron multiplication factor, is a measure of the efficiency of a nuclear reactor in sustaining a chain reaction. It is a critical parameter for reactor safety and performance evaluation.

Nuclear reactor coolants are used to transfer heat away from the reactor core to prevent overheating and maintain safe operation.


Question: What is the minimum cooling time required for PWR fuel assemblies?

The NRC monitors various compliance programs to ensure that nuclear facilities operate safely.


Question: Groundwater monitoring is essential following the installation of new monitoring wells to assess environmental contamination.

The next groundwater sampling periods are Fall 2020 and Spring 2021.

There are no known processes in this area that currently or historically used PCE.


Question: The NRC staff will conduct an audit to understand how the steam generator reactor coolant system is maintained despite weld cracking issues.

In accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” and NUREG-2172, appendix A, “License Renewal Commitments,” the NRC staff will perform an audit to gain a better understanding of: (1) how the Callaway steam generator reactor coolant system (RCS) pressure boundary is adequately maintained in the presence of divider plate weld cracking, and (2) how the susceptibility of the Callaway steam generator tube-to-tubesheet welds to primary water stress corrosion cracking (PWSCC).

In March 2015, the NRC published NUREG-2172 “Safety Evaluation Report Related to the License Renewal of Callaway Plant, Unit 1,” Volume 1, which documents the technical review of the Callaway license renewal application (ML15068A342).


Question: What is greater-than-Class C (GTCC) material in radioactive waste regulations?

Those materials whose activity concentrations exceed the limits for Class C LLW are identified by footnote as greater-than-Class C (GTCC) material.

The analyses in this NUREG postulate placing this material in a LLW disposal facility, the disposal costs for this material may be significantly overestimated compared with high-density packaging and geologic repository disposal.


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

A safety analysis report (SAR) provides a comprehensive assessment of the potential hazards of a proposed facility and outlines the measures that will be taken to mitigate those hazards under normal and accident conditions.

The reactor's cooling system was designed to operate under extreme conditions, ensuring that temperature regulation is maintained even during unexpected events.


Question: What is the effective porosity in groundwater studies?

The effective porosity is 0.2 (dimensionless).

The flow in river is measured at 533,333 cfs (46,076,971,200 ft³/day).


Question: The Nuclear Regulatory Commission mandates that a thorough assessment be conducted regarding a Power Range Monitor failure before proceeding with further actions.

Power Range Monitor (PRM) Failure (continued)

The Work Control team should be notified to troubleshoot the issue and ensure all requirements are addressed promptly.


Question: The NRC's regulations require a robust framework for assessing emergency categories during nuclear plant incidents.

The Emergency Classification program establishes a structured approach to determine and categorize the emergency level based on the scenario and its potential impact on safety and occupational health.

The Emergency Medical Response Plan outlines the protocols necessary to ensure proper medical treatment and evacuation procedures during significant incidents.


Question: Source Range Neutron Flux trip Function

The Source Range Neutron Flux trip Function provides protection for uncontrolled control rod withdrawal from subcritical and control rod ejection events.

The LCO requires two channels of Source Range Neutron Flux to be OPERABLE.


Question: The NRC needs to keep track of radioactive devices, including their activity levels and isotope types.

Please update the information as necessary.


Question: Loss of Offsite Power

Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: CFR 41.8 / 41.10 / 45.3

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.


Question: The review of alternatives considered by Xcel Energy focuses on ensuring reliable electricity generation to meet the region's needs.

Xcel Energy developed the following set of evaluation criteria to review MNGP replacement alternatives: - The purpose of the proposed action (SLR) is the continued generation of approximately 640 MWe net baseload power beyond MNGP’s current license term to meet future system generating needs.

The following generation sources were selected as reasonable replacement alternatives based on capability to provide reliable baseload power: 1. Natural Gas and Renewables Alternative - Natural gas combustion turbine (CT) units with MDCTs located offsite - Wind turbines located offsite - Solar panels located onsite and at offsite locations.


Question: thermal neutron fluence threshold

(f) This Case is only for welding in applications predicted not to have exceeded thermal neutron (E < 0.5 eV) fluence of 1 × 10^17 neutrons per cm² prior to welding.

The size of the flaw detected in the base metal will be used to define the life of each overlay.


Question: nuclear safety analysis process

The nuclear safety analysis process involves a systematic examination of safety-related systems to ensure compliance with regulatory standards and to identify potential hazards.

The quarterly maintenance schedule outlines the timelines and necessary procedures for routine checks on non-safety systems and equipment.


Question: The NRC stipulates that operators must establish and maintain clearance boundaries for critical safety systems.

NRC2018-A2  Determine Clearance Boundaries for Safety Injection System Leak

None of the training aids were applicable to this task.


Question: The NRC has invited feedback on public comments related to regulatory petitions PRM-50-93 and PRM-50-95.

Petitioner has responded to NEI’s comments on PRM-50-93 and PRM-50-95 promptly.

In these comments on PRM-50-93 and PRM-50-95, Petitioner responds to the Nuclear Energy Institute’s ("NEI") comments on PRM-50-93 and PRM-50-95, dated November 24, 2010.


Question: The NRC's audit of the Final Safety Analysis Report identified numerous issues that needed to be addressed by the licensee.

From March 2023 to August 2024 the NRC conducted a regulatory audit of FSAR Chapter 16 and Part 4 which generated 68 audit issues.

This slide lists the sections in the FSAR Chapter 16 and Part 4.


Question: tectonic stress in the mid-continent region

Expert teams that participated in the 1986 EPRI evaluation of intra-plate stress found that tectonic stress in the CEUS primarily is characterized by NE-SW-directed horizontal compression.

The primary tectonic elements of the region are off east rift systems such as the Reelfoot Rift, or formal tectonic zones such as the Paleozoic Ouachita Orogenic Belt and Appalachian Mountains.


Question: nuclear licensing requirements

The licensing process for nuclear facilities is governed by a set of regulations designed to ensure safety and compliance with federal and state laws.

The equipment for cooling systems must be regularly inspected for efficiency and compatibility with existing infrastructure.


Question: The NRC establishes criteria for training personnel involved in regulatory oversight of nuclear facilities.

App C Training Requirements for Fuel Facility Inspectors in the Nuclear Material Safety and Safeguards Program Area 10/28/14 (14-026)

App D1 Fuel Facility Information Security Inspector Specialized Qualification Program Training and Qualification Journal 10/28/14 (14-026)


Question: nuclide identification

Nuclide identified during the nuclide identification

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


Question: 10 CFR 55.43(b)(5) for SRO-only processes

Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only

Can the question be answered solely by knowing immediate operator actions?


Question: What is the significance of Source Range Monitors in nuclear power plants?

The requested changes would modify TS Section 3.3.1.2, “Source Range Monitors (SRM) Instrumentation.” Specifically, the proposed change will allow for a lower required count rate for Surveillance Requirement (SR) 3.3.1.2.4 of ≥ 0.7 counts per second (cps) with a signal to noise ratio ≥ 20:1.

A regulatory audit is a planned license- or regulation-related activity that includes the examination and evaluation of information that provides the technical basis for the LAR.


Question: maximum initial enrichment in nuclear fuel assemblies

Maximum Initial Enrichment – 37-Assembly Undamaged Fuel Configuration – 0.036 g/cm² 10B Absorber

Design Characteristics of the Fuel Basket Assemblies


Question: loss of ADS initiation capability

96 hours from discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND 8 days


Question: What is a Probabilistic Seismic Hazard Analysis in the context of nuclear plants?

Probabilistic Seismic Hazard Analysis: Background Information.

Environmental Impact Statement for an Early Site Permit (ESP) at the Clinch River Nuclear Site.


Question: What is the aging management evaluation for steel piping in nuclear facilities?

The Lubricating Oil Analysis program (B.2.1.25) and One-Time Inspection program (B.1.20) will be used to manage loss of material of the carbon steel and gray cast iron piping, piping components, and tanks exposed to lubricating oil in the Emergency High Pressure Makeup, Residual Heat Removal Service Water, Raw Cooling Water, High Pressure Fire Protection (Diesel Driven Pump), Standby Liquid Control, Standby Diesel Generators, Supplemental Diesel Generator, Control Rod Drive and Diesel Generator Starting Air Systems.

Not Applicable. There are no steel heat exchanger components exposed to condensation in screened in portions of Auxiliary Systems.


Question: The inspections mandated by ASME Section XI ensure the integrity of pressure boundaries in nuclear facilities.

The RV Nozzle Safe End Welds are designed as a pressure boundary component, made from nickel alloys and subjected to reactor coolant conditions, particularly focusing on aging effects like cracking.

Diablo Canyon Power Plant Units 1 and 2 have undergone extensive operational assessments to maintain regulatory compliance.


Question: What are the units of measurement used in nuclear device activity reporting?

The following table lists isotopes along with their activity measured in millicuries (mCi): | Isotope (e.g AM241) | Activity (e.g. 1005) | Unit (e.g. mCi) | |---------------------|----------------------|------------------| | CS137               | 100                  | mCi              |

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


Question: technical information needed for informed harvesting decisions

Technical information needed for informed harvesting decisions

Why our organization is interested in harvesting


Question: The evaluation of liquid radwaste management is crucial for both regulatory compliance and plant safety.

Refer to BCP 400-TWX01, Liquid Radwaste Release Form.

The candidate should determine the release cannot occur until the missing signature is resolved.


Question: What are the proposed areas for an ESP facility?

Proposed Areas for EGC ESP Facility Structures - Vicinity: 6 mi radius around site.

The Bureau of Transportation Statistics reported data in 2000 regarding various transportation networks.


Question: What is the purpose of AOP-43 actions in reactor operations?

AOP-43 actions result in closing of the RCIC trip throttle valve, which prevents RCIC from automatically injecting to the Reactor given low water level.


Question: What is thermal power in the context of nuclear reactors?

THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components: a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.


Question: What are the implications of declining block rates for electricity consumption?

Historically, utility rate structures were designed to reflect the cost of service of providing electricity to a customer by using the declining block rate system.

For example, it is recognized that TVA's involvement in the promotion of conservation measures was directly responsible for pending legislation in the State of Tennessee that sets new improved insulation standards for new homes.


Question: net peak area in nuclear regulatory filings

Net Peak Area is a significant value used in spectrometry to quantify the amount of a specific substance present in a sample, as indicated in the data tables documenting peak analysis at various energy levels in keV.

The spectrometry data includes a range of energy values measured during the experimental process, providing insights into the background noise reported from ambient conditions without detailing the net peak area.


Question: What are the types of nuclear waste and their measurements?

Spent Resins: 2400 Chemical resins and fission and corrosion products - includes powdex

The NRC requires licensees to analyze emergency planning zones to ensure public safety during potential reactor emergencies.


Question: In regulatory proceedings, applicants must provide a rationale for using exemptions in their financial projections.

In our license transfer adjudications, we have long held that financial assurance will be acceptable "if it is based on plausible assumptions and forecasts."

New York argues that an applicant cannot rely on a prospective exemption to support its cost estimates, but it does not provide support for this argument.


Question: The process for validating job performance measures is critical for ensuring operational effectiveness.

**SRRS:** 3D.105 (when utilized for operator initial or continuing training)

**Reviewed By:** Brian Payne /s/  *Operations Representative*  **Date:** 12/21/2018


Question: What is the significance of cracking due to stress corrosion cracking (SCC) in PWRs?

Cracking due to SCC could occur in PWR SS bottom-mounted instrument guide tubes exposed to reactor coolant. Further evaluation is recommended to ensure that these aging effects are adequately managed.

Cracking due to cyclic loading could occur in steel and SS BWR isolation condenser components exposed to reactor coolant.


Question: analytical methods for isotopic analysis

- **Analytical Method:** DOE EML HASL-300, Am-05-RC Modified

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


Question: analytical results for water quality parameters

The Semiannual Report includes analytical results and field measurements for various parameters measured in water, including Draft B. Water, pH, Dissolved Oxygen, and Electrical Conductivity.

Values marked with '<' indicate that the result is less than the limit of detection (LOD).


Question: calibration frequency for radiation measurement instruments

A licensee committed to a routine or emergency radiation survey program should perform an acceptable calibration of all radiation measurement instruments and equipment at the frequency specified in NRC regulations, annually, or at the frequency recommended by the manufacturer, whichever period is shorter.

Routine maintenance of radiation measurement instruments should be performed as recommended by the manufacturer.


Question: What are the requirements for facility licensee examination reviews?

Yes. The independent managerial or supervisory reviewer is confirming and signing that the written examinations and operating tests meet the requirements of NUREG-1021.

NUREG-1021 eliminated the limits on written question repetition from quizzes given during the training program, thereby eliminating the need to track question histories.


Question: The design of the confinement system for spent nuclear fuel storage must adhere to specified safety standards and regulations set by the NRC.

The TSC is the confinement system for the stored fuel. The TSC assembly consists of a right circular cylindrical shell with a welded bottom plate, a fuel basket, a closure lid, a closure ring, and two sets of redundant penetration port covers.

NAC International Inc., Final Safety Analysis Report for the MAGNASTOR® System, Docket No. 72-1031


Question: Licensees must conduct rigorous statistical evaluations to ensure measurements meet established release criteria.

The evaluation of the Sign test results clearly demonstrates that the survey unit passes the unrestricted release criteria; thus, the null hypothesis is rejected.

No anomalies were observed during the performance or analyses of the survey.


Question: What is a Neutron Source Assembly (NSA) in nuclear storage?

One Neutron Source, or Neutron Source Assembly (NSA) is permitted to be loaded in a TSC in fuel storage locations No. 11, 12, 13, 18, 19, 20, 25, 26 or 27. Neutron source assemblies may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets.

Fuel assemblies may contain any number of unirradiated (i.e., not inserted in-core) nonfuel solid filler fuel replacement rods. Activated stainless steel rods are limited to 5 per assembly at a maximum burnup/exposure of 32.5 GWd/MTU.


Question: The regulatory requirements for evaluating transient behavior in reactor cores are crucial for ensuring safety compliance.

SVEA-96 Optima3 transient validation for   \( J^c \), measured temperature vs minimum CPR simulated.

Measured and POLCA-T calculated time of dryout for FRIGG SVEA-96 Optima3 transient dryout tests.


Question: The regulatory considerations for microreactors have become increasingly important due to their unique implications for safety and siting.

Siting is usually one of those issues that's discussed a lot in terms of microreactors.

Another activity that's going on right now is microreactors, which are receiving a lot of attention.


Question: What is the significance of time frame in analyzing precipitation data for nuclear safety?

Time frame is critical in precipitation data analysis as it determines the resolution and reliability of the hydrological models used in safety assessments.

Base Map Added Precipitation Grid corresponds to an updated mapping method used for visualizing rainfall distribution over a defined area.


Question: The calibration constant of this detector was found to be 3.894109E+09 during tests.

The meter readings indicate several values, including NA mR/hr and 14.7 mR/hr for various reference points.


Question: Nuclear Fire PRA Methodology

Methods may have been used in the FPRA that deviate from guidance in NUREG/CR-6850, “EPRI/NRC-RES [Office of Nuclear Regulatory Research] Fire PRA Methodology for Nuclear Power Facilities” (References 22, 23, and 24), or other acceptable guidance (e.g., frequently asked questions (FAQs), NUREGs, or interim guidance documents).

The NRC staff evaluates the acceptability of the PRA for each new risk-informed application, and as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision-making may vary with the relative weight given to the risk assessment element of the decision-making process.


Question: What are DBA forces in nuclear structural design?

The DBA forces are 2 to 5 times larger than other forces that are applied to the lower support structure.

The stress in the various structural members for all of the design load cases was found to be below the design criteria as specified in Section 6.2.2.16.


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

The safety analysis must evaluate the potential hazards associated with the reactor, including internal and external events that could affect safety and operational reliability.

The regulatory filing process includes submissions of license applications, amendments, and reports to ensure compliance with safety regulations.


Question: uncertainty approach for licensing parameters

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

The results of the study indicated significant improvements in the efficiency of the reactor's thermal performance through optimized fuel management strategies.


Question: What is the definition of a manual action in reactor operations?

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

The NRC requires a comprehensive analysis of potential accident scenarios at nuclear facilities to ensure operational safety and regulatory compliance.


Question: The NRC's evaluation includes measuring airborne radioactive concentrations to ensure compliance with safety standards.

Parameters and Assumptions for Calculating Airborne Radioactive Concentrations

Average Individual Exposure Based on PWR Plant Age


Question: nuclide identification in nuclear safety

The identified nuclides included K-40, CS-137, PB-212, and BI-214, with their respective confidence levels, energy outputs, yields, and activity measured in pCi/gram.


Question: The design of the AFWS incorporates features to ensure redundancy and dependability during operation.

As recommended in NUREG-0800, the reliability evaluation of the AFWS was performed utilizing the methodology in Appendix III and Annex I of Appendix X in NUREG-0611.

The motor-operated valves in the miniflow lines of trains A and B (5154 and 5155) are subject to maintenance for calibration of the flow element actuation device in these valves.


Question: What is the role of radiological environmental monitoring in nuclear plant safety?

Duke Energy collects and analyzes samples of air, water, sediment, fish, vegetation, and milk, if available, for radioactivity as part of its ongoing, comprehensive Radiological Environmental Monitoring Program.

The special pathway receptors analysis is an important part of the environmental justice analysis because consumption patterns may reflect the traditional or cultural practices of minority and low-income populations in the area.


Question: What is covered in the Initial License Examination Outline for nuclear plants?

Indiana Michigan Power Company (I&M), the Licensee for Donald C. Cook Nuclear Plant Units 1 and 2, hereby submits a copy of the Initial License Examination Outline for the planned July/August 2018 Examination at Donald C. Cook Nuclear Plant (CNP).

The examination is being prepared in accordance with NUREG 1021, Revision 11.


Question: The NRC guidance allows for a risk-informed methodology to handle issues related to GSI-191 and GL 2004-02.

The risk-informed approach is consistent with the guidance of Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.174 and SECY-12-0093.

The primary purpose of the ECCS is to ensure the long-term removal of decay heat from the reactor fuel following a loss of coolant accident (LOCA).


Question: What are the requirements for transportation risk management in nuclear projects?

It is recommended that all delivery trucks used to transport project materials [loaded resin, yellowcake, process chemicals and 11e.(2) by-product material] should carry the appropriate certifications of safety inspections.

If birds begin to use the retention ponds, it is recommended that any bank vegetation that would provide cover for birds should be removed, and visual or sound devices should be installed to ward off birds.


Question: What is the purpose of the HEC-HMS model in nuclear safety analysis?

HEC-HMS models are used to calculate runoff hydrographs from each subarea and maximum water surface levels of storage areas in the LIP analysis.

The software is dedicated in accordance with QA procedures.


Question: What are the alternative requirements for nozzle inner radius inspections in nuclear reactors?

In lieu of an ultrasonic examination, licensees may perform a VT-1 examination in accordance with the Code of Record for the ISI program using the allowable flaw length criteria of Section XI, Table IWB-3512-1, with limiting assumptions on the flaw aspect ratio.

Prior to welding, an examination or verification must be performed to ensure proper preparation of the base metal, and that the surface is properly contoured so that an acceptable weld can be produced. This verification is to be required in the welding procedure.


Question: What is the significance of the 500ml Marinelli geometry in gamma spectrum analysis?


Question: What is meant by the term 'transfer' in signal processing?

THE OUTPUT EQUALS THE SIGNAL WHICH HAS BEEN SELECTED BY TRANSFER. THE EXIT OF THE TRANSFER IS IDENTIFIED BY AN ASSIGNMENT.

THE OUTPUT IS DIRECTLY PROPORTIONAL TO THE INPUT'S VALUE.