Question: Reactor Coolant Pressure Boundary System

Reactor Vessel Flange Leak Detection Line   | Pressure Boundary      | Stainless Steel | Air - Indoor, Uncontrolled (External)  | Cracking                          | One-Time Inspection (B.2.1.20)    | IV.A1.R-61a     | 3.1.1-016                | A     |

Valve Body                                 | Leakage Boundary       | Carbon Steel   | Air - Indoor, Uncontrolled (External)  | Loss of Material                  | External Surfaces Monitoring of Mechanical Components (B.2.1.23) | IV.C1.R-431     | 3.1.1-124                | A     |


Question: The NRC requires that licensee submissions include detailed information on the instruments used to measure safety parameters.

The regulatory framework mandates that all measurement devices used in nuclear safety applications meet specific performance criteria established by the NRC.

The sale of industrial nuclear products is often preceded by a thorough market analysis to assess demand and competition.


Question: Are participants allowed to observe the webinars without restrictions?

We’re not making it public in terms of meeting announcement and everything, but we haven’t been turning away any participants via webinar.

Please let me know if you have any questions or input.


Question: What is the role of operating experience in nuclear safety programs?

The review and assessment of relevant operating experience (OE) for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227, Revision 1-A.

The NRC requires licensees to comply with the regulations outlined in 10 CFR Part 50, Appendix B, which sets forth quality assurance criteria for nuclear power plants.


Question: What does the double contingency principle entail in nuclear criticality safety?

The double contingency principle of the NRC guidance, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants" contained in an NRC letter to all power reactor licensees from Laurence Kopp dated August 19, 1998 states that it is not required to assume two unlikely, independent concurrent events to ensure protection against a criticality accident.

Although the dropped assembly is more reactive outside rather than inside the poisoned storage cell, the assembly is no more reactive dropped on top of the storage rack than located anywhere else in the pool outside the storage rack.


Question: What is boron concentration in the context of spent fuel pools?

Spent Fuel Pool Boron Concentration is a critical aspect of maintaining the safety and criticality control of the spent fuel pool, ensuring that the concentration of boron in the water does not fall below the required levels to prevent a nuclear chain reaction.

The distribution systems for electrical power are vital for the operation of various plant systems, ensuring that power is supplied reliably during both normal and emergency conditions.


Question: The NRC evaluates safety aspects associated with the use of various radionuclides in nuclear facilities.

Radionuclides such as Dysprosium-155, Dysprosium-157, and Dysprosium-159 are classified under 'W, all compounds' indicating their usage in nuclear safety applications.

Erbium-165 is noted for its various physical properties, which make it suitable for distinct technological applications outside the nuclear sector.


Question: What is the acceptable method for setting valve control in safety functions?

Valves with safety function to close are being set by limit switch control instead of torque switch. This is acceptable where certain considerations are addressed: If a valve has a leakage requirement and is required to close under dynamic conditions (flow and pressure), setting up the limit switch control under static conditions (no pressure or flow) needs to have reasonable assurance that the valve will meet its leakage requirement when it performs its safety function. This can be accomplished via setting the valve close limit under expected dynamic conditions in-situ or in a test facility.

The report includes various configurations of reactor systems and their performance under different scenarios.


Question: What is a Maximum Demand scenario in nuclear plants?

15.1.2 (Section 3.2) - FWC Failure- Max Demand without Turbine Bypass

| 15.1.3 (Section 3.4) - Pressure Regulator Failure- Open |


Question: What is Tritium Distillation in nuclear safety analysis?

The product: LSC, Tritium Distillation, Solid refers to the separation process that is utilized to extract tritium from other substances for analysis in nuclear safety assessments.

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


Question: The Emergency Core Cooling System (ECCS) is essential for maintaining reactor safety during emergencies.

3.5.2 ECCS – Operating | 2 Centrifugal Charging (CH) pumps (high pressure) | (1) Injection from the refueling water storage tank into cold legs | (2) Cold leg recirculation from the containment sumps | (3) Hot leg recirculation from the containment sumps

Design Success Criteria | (1) (a) 1 of 2 CH pumps for small/medium loss of coolable accident (LOCA) until the RCS is depressurized to allow injection from 1 of 2 SI pumps | (b) 1 of 2 RHR pumps | (c) 1 of 2 CH pumps for a steam generator tube rupture (SGTR) or main steam line break (MSLB)


Question: What is Steam Line Isolation in nuclear safety protocols?

Steam Line Isolation also mitigates the effects of a feed line break and ensures a source of steam for the turbine-driven AFW pump during a feed line break.

Closure of the MSIVs limits the accident to the blowdown from only the affected SG.


Question: What is the PRIME model used for in nuclear fuel analysis?

The PRIME Model for Analysis of Fuel Rod Thermal – Mechanical Performance is a suite of methodologies developed to evaluate the thermal and mechanical performance of fuel rods under various operational conditions.

Global Nuclear Fuel NEDC-33256-P-A, Revision 2, outlines the detailed processes involved in the licensing of new reactor designs and the applicable regulations that must be adhered to.


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

The COBRA-FLX is a thermal-hydraulic analysis code that includes the ORFEO-HMP critical heat flux correlation, which is used to predict critical heat flux in boiling water reactors during normal and accidental conditions.

The test assembly conditions for HTP grid tests include parameters such as temperature, pressure, and flow rate, which are critical for evaluating the performance of nuclear fuel under various operational scenarios.


Question: Nuclear power plants must adhere to strict guidelines when submitting technical documents for licensing approval.

The NRC evaluates applications for licenses and amendments to ensure compliance with safety and environmental regulations.

The quick brown fox jumps over the lazy dog, demonstrating the playful nature of language and its structure.


Question: weighted mean activity of nuclides

The weighted mean activity is calculated by taking into account the contribution of each nuclide's activity and its uncertainty, as shown in the corresponding table for individual nuclides like PB-212 and BI-214.

The table lists nuclide names and their corresponding activities along with uncertainties, which provide essential data for different radiation monitoring applications.


Question: maximum allowable impurity concentrations in reactor condensate

The maximum condenser leak for allowable water quality maintenance that can be accommodated continuously at full-load operation with continuous recoating (no 2-h-lag allowance) and with a 0.1-gal/min steam generator leak of primary coolant is 61 gal/min.

b. This might be as much as 4000 ppb for several hours at initial plant startup. This value is used for equipment specification purposes, since without condenser tube leaks, sodium and chloride concentrations are much lower as a result of cleanup by the steam generator blowdown demineralizer.


Question: risk-informed approach in nuclear licensing

The proposed ECCS rule change will allow use of a risk-informed approach, addressed in 10 CFR 50.46(e), in lieu of a deterministic evaluation.

Prior to the risk-informed approach, deterministic methods were used to evaluate the effects of accident-generated and transported debris in order to meet the current licensing basis assumptions for analyzing the effects of post-accident debris blockage in the sump and in-vessel.


Question: What is an NRC license transfer process?

We are requesting a Transfer of Control amendment.

If you have any questions or require additional information, please contact our medical physics consultant, Laura Luna at 734-662-3197, or lluna@mpcphysics.com.


Question: The NRC mandates aging management for components exposed to operational environments to ensure safety and functionality.

The aging effect requiring management for Carbon and Low Alloy Steel Bolting is loss of material as indicated in the assessment for the Main Generator and Auxiliaries System.

The summary table outlines various components of the system, each of which plays a critical role in the functionality and safety of nuclear plants.


Question: What are the aging effects on the steel steam generator shell assembly?

Loss of material due to general, pitting, and crevice corrosion.


Question: safety-related air conditioning compressors

The inspectors reviewed a self-revealed Green finding and associated non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI, "Corrective Action," for the licensee’s failure to promptly identify and correct a significant condition adverse to quality.

The inspectors considered this to constitute a negative trend in software implementation issues. The licensee concurrently identified a similar trend which they documented in condition report 204407239.


Question: chemical analysis during power ascension

The total core received at the specified control settings...individual response...


Question: connecting nodes in nuclear systems

Connecting nodes in a reactor system are crucial for determining the flow characteristics between various components, with specific measurements indicating their interrelationship, such as L/A (ft⁻¹).

The reactor's cooling system operates through a series of pumps that maintain the appropriate temperature during normal and emergency conditions.


Question: What is the significance of defense in depth in nuclear plant safety?

This section, 53.250, would establish requirements for defense in depth and those are based on longstanding philosophy of providing defense in depth to address uncertainties about the design, operation, and performance of the commercial nuclear plants.

The analysis of LBEs confirm the adequacy of design features and programmatic controls in meeting the safety criteria under 53.210 and 220 and that they would be used to establish related functional requirements for SSCs, personnel, and programs.


Question: What are state values in reactor safety evaluations?

Table 5-2 lists the state values for the increase in steam flow event, which are essential for evaluating the potential impact of the event on reactor safety.

Framatome Inc. conducted extensive analyses to ensure the reliability and performance of critical systems under various operational conditions.


Question: What are the components involved in the main steam lines for a nuclear reactor?

Main Steam Loop A consists of several components detailed in the drawings, including 731E756, Sheet 1, which outlines the configuration of the steam loop.

The list of drawings includes revisions for various components and is essential for tracking changes in design specifications.


Question: What is the total annual natural radiation whole-body dose measured in mrem?

J. M. Richardson, "Alabama Encyclopedia," Vol. I, copyright 1965.


Question: loss of normal AC power acceptance criteria

This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient.

The biases and conservatisms presented in Table 7-44 are considered in identifying the bounding transient simulation for primary and secondary pressure.


Question: What is the purpose of the RICT Program in nuclear plant operations?

The scope of the training for the RICT Program will include rules for the new TS program, Configuration Risk Management Program (CRMP) software, TS Actions included in the program, and procedures.

Documentation requirements may vary according to the specific type of regulatory filing submitted to the NRC.


Question: What are the actions required for aligning control room pressurization dampers?

Align the following intake dampers one partial open and one closed: - 1-HV-ACR-DA-2, Control Room Pressurization Intake Damper - 1-HV-ACR-DA-2A, Control Room Pressurization Intake Damper - Open 1-HV-ACR-DA-3, Control Room Pressurization Clean-up Recirc Damper

START Control Room Pressurization/Cleanup Filter this Ventilation Fan 1-HV-ACR-I or 1-HV-ACR-2


Question: The evaluation of head loss criteria is critical to ensure structural integrity in a nuclear facility.

Starter Head Loss (SHL) = Conventional HL + Chemical HL

Degradation void fraction is measured to assess the performance of the reactor under various operational conditions.


Question: The regulatory requirements for obtaining a nuclear license include comprehensive safety evaluations and environmental considerations...

The NRC requires that all license applications contain sufficient information and analyses to support the safety of the proposed activities.

Oglethorpe Power Corporation holds multiple units within its portfolio, which influence its operational strategy significantly.


Question: What is ECCS leakage and how is it quantified in nuclear reactor safety assessments?

ECCS leakage is administratively controlled to a value of 138,000 cc/hour [Ref. 23]. The AST accident analysis basis is 276,000 cc/hour [Ref. 4, Item 4.12]. This leak rate is considered an upper bound that would still allow ECCS operability after 30 days, without makeup (a 12% inventory loss).

The revised Dose Conversion Factors (DCFs) from the U.S. Federal Guidance Report 11 & 12 (Ref. 11, 12) are used for this analysis.


Question: seismic probabilistic risk assessment

Section 8.6: Seismic probabilistic risk assessment (SPRA) focuses on evaluating the seismic hazards and their impact on the safety of nuclear reactors, ensuring that potential risks are systematically assessed and addressed in regulatory filings.

Clearance to the stop is a critical part of the design process, ensuring that all components are appropriately positioned to prevent any interference during operation.


Question: The licensee must adhere to submitted documents and regulations as stipulated by the U.S. Nuclear Regulatory Commission.

Except as specifically provided otherwise in this license, the licensee shall conduct its program in accordance with the statements, representations, and procedures contained in the documents, including any enclosures, listed below.

This license condition applies only to those statements, representations, and procedures that are required to be submitted in accordance with the regulations.


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

Reactor Coolant System - Monitor (Chemical and Volume Control System) - Before heatup and at temperature.

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


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

The energy peaks measured during the radiological analysis are essential for accurately identifying the nuclides present in the sample, where each peak corresponds to a specific energy level associated with a distinct nuclide.

The subtracted area values indicate the net counts after accounting for ambient background, which is critical for understanding the overall radiological profile.


Question: What is the purpose of tracer yield in nuclear sample analysis?

The Cm-244 traced portion of sample, (see below), did not meet the client’s yield requirement due to the matrix of the sample; however, the combined Am-243 & Cm-244 traced yields do meet the requirement.

All sample data provided in this report met the acceptance criteria specified in the analytical methods and procedures for initial calibration, continuing calibration, instrument controls and process controls where applicable, with the following exceptions.


Question: What is the role of the Auxiliary Feedwater System (AFWS) in nuclear safety?

The AFWS supports transfer of core decay heat, RCP heat, and sensible heat from the RCS to the UHS during plant cooldown.

Maintain integrity of nonsafety-related components such that no interaction with safety-related components could prevent satisfactory accomplishment of a safety function through structural integrity or leakage boundary.


Question: What is the One-Time Inspection program for aging effects in nuclear systems?

BFN will implement the One-Time Inspection program (B.2.1.20) to demonstrate that the aging effect of loss of material is not occurring on the external surfaces in the Air Conditioning system.

Deficiencies will be documented in accordance with 10 CFR Part 50, Appendix B Corrective Action Program.


Question: containment hydrogen sample

MISV6910  | CONTAINMENT HYDROGEN SAMPLE     | M239C          | SV             | OPEN                    | CLOSED | OPEN  |          |                              |

MIMV7010  | ILRT TEST CONNECTION            | MANUAL         | CLOSED         | AS IS                   | CLOSED |       | CLOSED   |                              |


Question: What is the significance of feedwater system pipe breaks in nuclear reactor safety?

Figure 15-43 illustrates the pressure level in the pressurizer over time following a feedwater system pipe break, which is critical for long-term core cooling analysis.

The table lists the values for time and corresponding pressurizer level, which are necessary for preparing system response evaluations.


Question: ASME Section XI Inservice Inspection process

The DCPP ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD AMP is an existing AMP that is performed to manage cracking, loss of fracture toughness, and loss of material in Class 1, 2, and 3 piping and components within the scope of LR.

Indications, relevant conditions, and resolution of rejectable indications and relevant conditions are evaluated as required by ASME Section XI IWB-3000, IWC-3000, and IWD-3000.


Question: Licensees must evaluate the potential impact of design changes on safety calculations and compliance.

The safety analysis is essential for understanding the effects of proposed modifications on nuclear facility operation.

The report outlines the staffing and organizational structure required for the successful operation of the facility.


Question: What are the atmospheric effects of cooling towers at nuclear power plants?

Cooling towers would have air emissions and atmospheric effects from drift and plumes. Cooling tower drift consists of liquid droplets entrained in the exhaust air stream.

The size of the DNPS site would allow considerable sound-level attenuation to offsite receptors.


Question: The current regulatory framework permits the usage of low-enriched uranium fuel, with a cap set at 5.0 weight percent uranium-235.

Current licensing framework allows for the use of ≤ 5.0 weight percent uranium-235; however, technology developments may require numerous exemptions to utilize fuel enriched above 5.0 weight percent.

Staff request to pursue rulemaking and develop a regulatory basis was approved by the Commission via SRM-SECY-21-0109.


Question: What is the energy yield of CE-144 in MeV?

CE-144 has an energy of 80.12 MeV and a yield of 1.36%.

The energy of 1238.11 MeV is associated with a different nuclide, which has a yield of 5.94%.


Question: empty locations in DSC configurations

Location identified as "A" is for empty placement in 88-FA Loading.

The regulatory body requires that all submissions be made electronically using the NRC's web-based system.


Question: The NRC mandates that licensees must perform comprehensive evaluations to assess potential flooding scenarios.

The probable maximum flood (PMF) hydrograph at Three Mile Island was developed to evaluate the potential impact of flooding on plant safety.

The figures provided illustrate a variety of hydrographs and dam analysis but do not specifically pertain to the regulatory processes for nuclear licensing.


Question: The NRC oversees the submission of emergency plans as part of license amendment requests for nuclear plants.

The NRC requires licensees to submit a common emergency plan that is consistent with NUREG-0654, which is essential for ensuring safety protocols in nuclear facilities.

The documents provided to the NRC include detailed financial assessments and revenue projections related to plant operations, which are crucial for evaluating regulatory compliance.


Question: departure from nucleate boiling

The effect of the 17 x 17 fuel assembly geometry on the departure from nucleate boiling (DNB) heat flux has been determined experimentally and has been incorporated into a modified spacer factor for use with the W-3 correlation.

The above testing was conducted in the Westinghouse Test Engineering Laboratory Facility (see Section 1.5.3).


Question: What is a nuclide and how is it measured in a nuclear context?

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


Question: The facility will utilize liquid sodium as a coolant instead of traditional water in its reactor design.

The facility is planned to be a pool-type reactor that uses liquid sodium as the coolant instead of water.

The map of the area of potential effects (APE) is crucial for visualizing potential impacts on surrounding land.


Question: microbiological hazards in nuclear power plants

In the 2013 license renewal GEIS, the NRC reviewed microbiological hazards to plant workers. The GEIS discussion of microbiological hazards focuses on the thermophilic microorganisms *Legionella spp.* (which can be a hazard during such activities as cleaning condenser tubes and cooling towers) and the pathogenic amoeba, *Naegleria fowleri* (which can be a hazard in cooling water discharges and can pose a public health hazard, as addressed under a separate Category 2 issue). No change in existing microbiological hazards is expected over the license renewal term.

CEG has a comprehensive health and safety program with procedures that implement industrial hygiene practices to minimize the potential for station worker exposure.


Question: The NRC has established acceptable practices for Section XI inservice inspection programs through various code cases.

The NRC determined that the Code Cases listed in Table 1 are acceptable for application in licensee Section XI ISI programs.

Quality assurance programs are essential for ensuring that all components meet the required standards for nuclear reactors.


Question: The regulatory filings require a comprehensive assessment of the surrounding land use near the EGC ESP site.

The land use and land cover in the region surrounding the site includes various classifications such as open water, residential areas, and commercial zones.

The U.S. Census Bureau collects data on demographic characteristics and economic indicators that provide insights into the local population.


Question: CCW pump malfunction assessment

Specifically, the candidate must analyze the CCW Bezel and determine that the 22 and 23 CCW Pumps have tripped trip which is causing CC Header Low Pressure alarms.

The question meets the SRO requirements as described in NUREG 1021 ES-401 Attachment 2.


Question: nuclide identification report

The Nuclide Library Used is C:\Canberra\Apex\Root\Dairyland_NPP\Library\HOTLAB.NLB.

Peak No. 2 has an Original Area of 2.81E+01 with an Orig. Area Uncertainty of 16.78.


Question: The licensing process for nuclear reactors requires thorough documentation and analysis of technical specifications; this ensures compliance with safety standards.

The technical specifications for the nuclear test reactor include detailed descriptions of reactor operation, safety systems, and regulatory compliance requirements to ensure safe operation and adherence to licensing criteria.

During the audit, the regulatory team reviewed the operating procedures and safety protocols, ensuring they align with general industry standards for operation and maintenance.


Question: What steps should be taken to isolate the RBCLC supply to the Drywell?

Task Standard: The RBCLC supply to the Drywell is isolated.

Ensure closed the following valves at panel 09-4:


Question: service water system regulations

At least two independent service water loops shall be OPERABLE.

With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.


Question: What is the function of the instrumentation in the Boron Recycle System?

Instrumentation aligns the valves to bypass the demineralizers. Local temperature indication and a high temperature alarm on the Boron Recycle System panel are provided by this instrumentation.

During on-line operation, periodic radioisotopic sampling is performed using the Nuclear Sampling System as described in Section 9.3.2 to identify failed fuel.


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

The Auxiliary Feedwater System is designed to provide cooling water to the steam generators during a loss of normal feedwater supply, thereby maintaining safe reactor cooling in post-accident conditions.

Plant Indoor Air (Ext) refers to environments that may impact the performance of plant systems, but it does not directly relate to water chemistry considerations for auxiliary systems.


Question: The determination of FGR model parameter uncertainties relies on a comprehensive database of fuel rod data from various reactor types.

The large number of cases in the FGR database offers well-behaved statistics and allows for proper uncertainty estimation of the model parameters.

Section 5.4.6 discusses each parameter as referenced to the PIRT and its disposition on the manner uncertainties are addressed in the application of GALILEO.


Question: maximum primary pressure turbine trip

In Figure 15-32, the data illustrate the maximum primary pressure case during a turbine trip event, showing the pressure values decreasing over time.

The reactor cools down over a period of several hours after the shutdown, which is necessary to maintain safe conditions and allow safe maintenance operations.


Question: The NRC's assessment emphasizes the need for emergency preparedness and nearby community support in the event of a nuclear incident.

However, there is precedent for direct aid to the local school districts by TVA.

According to the Economic Atlas (October 1972) prepared by the Top of Alabama Regional Council of Governments (TARCOG), the 1970-1971 enrollment in Jackson County schools was 6959 and in Scottsboro 3183 for a total of 10,142.


Question: ground motion analyses

Ground Motion Analyses are essential for evaluating the seismic performance of structures and systems, ensuring they can withstand seismic events and maintain their safety functions.

The design and implementation of effective emergency response plans are vital for ensuring public safety in the event of unforeseen incidents.


Question: What is the appropriate abnormal procedure for control room evacuation?

Given a set of initial plant conditions: Determine the appropriate abnormal procedure.

The learning objective for this examination focuses on determining the cognitive level for control room evacuation.


Question: The NRC specifies coefficients for estimating decommissioning costs that account for labor, energy, and burial expenses for reactor types.

Per Table 3-1, the C (LLW burial associated) coefficient, or that percentage representing the portion of decommissioning cost attributable to LLW burial charges, are the same (.22) for both PWRs and BWRs.

Values for L_x and E_x for years subsequent to 1986 are based on the national product price indexes, national consumer price indexes, and local conditions for a given site, as described below in Sections 3.2 and 3.3, respectively.


Question: The reactor's power output is measured as a fraction of its nominal power.


Question: CHF test correlation

COBRA-FLX is a critical heat flux correlation developed by Framatome, which is used for assessing the thermal-hydraulic performance of nuclear reactors.

Figure A-29 depicts the axial geometry utilized in the CHF test SI070, providing a visual representation of the experimental setup.


Question: weighted mean activity of nuclides

The weighted mean activity is defined as the average activity calculated by weighting the individual activities of the nuclides based on their respective confidence levels and uncertainties.

The table lists various nuclides along with their respective confidence levels and other measurements, but does not provide a direct link to how these values affect reactor safety protocols.


Question: What is the HTP CHF test matrix in the context of nuclear licensing?

The HTP CHF test matrix supporting the ORFEO-HMP CHF correlation is shown in Table 4-4. All tests feature HTP grids and some tests include IFMs. The data set covers the majority of these CHF tests were previously used in other NRC-approved CHF correlations as following: tests K2200 - K4001, except K2202: ACH-2, tests K9000 - K9300: NuScale, tests with prefix “SP” except SP010 and SP660: BHTP.

It is essential to monitor environmental conditions to ensure the safety of nuclear facilities, including the assessment of natural disaster impacts and infrastructure resilience.


Question: What is the RICT Program and its purpose in nuclear regulatory processes?

The RICT Program is designed to provide guidance on recalculating RICT and risk management action time (RMAT) within 12 hours or within the most limiting front-stop CT after a plant configuration change.

Documentation requirements related to individual RICT evaluations are essential for maintaining compliance with safety and operational standards.


Question: compressive strength of polyurethane foam

The primary polyurethane foam physical property of interest for the HAC drop and puncture tests is its compressive strength.

The actual gross weights of the CTUs were: 2,788 pounds (CTU-1), 2,758 pounds (CTU-2), 2,752 pounds (CTU-3), and 2,754 pounds (CTU-4).


Question: What are the regulatory considerations related to inspections and aging management programs in nuclear plants?

Regulatory considerations related to inspections and Aging Management Programs (AMPs) are crucial for ensuring the safety and reliability of nuclear facilities throughout their operational life.

The availability of materials for nurturing is an important factor in the development of any project, influencing both the cost and complexity of the undertaking.


Question: The NRC regulations stipulate that surface scans must cover the entire area of the survey unit for Class 1 units.

The LTP, Section 5.6.4.4 and Table 5-15 specifies that for Class 1 open land survey units, surface scans will be performed on 100% of the surface area in the survey unit.

Sixty-four (64) scan lanes were established.


Question: What aging effects require management for valve bodies in nuclear systems?

Aging effects requiring management for valve bodies include loss of material and cracking, particularly in environments such as secondary water or sodium hydroxide.

The valve body functions as a leakage boundary in a spatial context.


Question: RCS unidentified leakage

RCS Leakage is considered RCS unidentified leakage if it is NOT any of the following: Leakage, such as from pump seals or valve packing that is captured or directed to a collection tank or sump.

The following steps check indications to locate the leak and identify affected areas.


Question: The NRC requires emissions to be reported in terms of pounds per hour for various pollutants.

The fuel consumption rate for the gas turbine is documented as 121,200 lb/hr.


Question: Changes in climate can affect various ecosystems and the services they provide to humans, impacting aquaculture and fisheries.

Increased vulnerability of biodiversity, food security, human health, and water quality due to altered biogeochemical cycles and climate change

The projections suggest severe impacts on human health due to extremes in climate and weather patterns that affect availability of clean water.


Question: What is 10CFR50 Appendix R Safe Shutdown?

This Fire Area complies with 10CFR50 Appendix R, Section III.G.3.

The capability to achieve and maintain hot standby conditions utilizing the SSS is assured in this area by virtue of location of SSS required equipment and cabling outside this fire area and breaker coordination on non-required SSS loads.


Question: What is the purpose of the certification process for specialty boards in nuclear licensing?

The certification process of specialty boards ensures that qualified individuals are recognized for their expertise in specific technical areas related to nuclear safety and operations.

The document outlines various administrative protocols for managing records and documentation relevant to the certification procedures.


Question: The NRC has issued an Early Site Permit for the construction of an advanced nuclear technology park.

The park would contain one or more advanced nuclear reactors with a cumulative electrical output not to exceed 800 megawatts electric for the site.

Planning and installation of the Calicner Project began in 2014 and is ongoing.


Question: The NRC has determined that Westinghouse's methodology for analyzing fuel licensing meets the necessary requirements based on the performance of PARAGON2.

The NRC staff finds that Westinghouse has adequately demonstrated acceptable performance for the PARAGON2 method to analyze lattices with various BP loadings.

Overall, the agreement between the PARAGON2 and MCNP results is excellent.


Question: The HVAC system is designed to maintain the necessary conditions to ensure the habitability of the Control Room during operational challenges.

The Control, Equipment, and Cable Rooms Heating, Ventilation and Air Conditioning (CRA HVAC) system design is described in Section 6.4 and shown in Figure 6-191.

High smoke level in the discharge duct of the Control Room AHU or the Control Room Area AHU stops the AHU with the high smoke level.


Question: Licensees must ensure compliance with safety standards throughout the operational lifecycle of a nuclear plant.

The NRC mandates that all plant operational practices adhere to stringent safety regulations, ensuring the protection of public health and the environment.

To maintain efficient operations, the facility utilizes advanced software tools for data management and analysis.


Question: shutdown safety color code for Reactor decay heat removal

A portion of OU-NM-103-101, Shutdown Safety Management Program, Attachment 1, Unit 1 Key Safety Functions, is provided on the next page.

Preparations are underway for Reactor head de-tensioning.


Question: What actions should be taken to handle a seismically induced loss-of-coolant accident?

When requested Execute Commands in Field Action branch to isolate and clear CFCU 1-2.


Question: requirements for maintaining an active license for operators

In accordance with **10 CFR 55.53(f)(2)**, an operator who fails to maintain an active license must, before resuming licensed duties, complete a minimum of 40 hours of shift functions under the direction of an operator or senior operator, as appropriate, and in the position to which the individual will be assigned; for senior operators limited to fuel handling under **10 CFR 55.53(c)**, one shift must have been completed.

The possibility that both no-solo operators standing watch together would become incapacitated at the same time is pretty remote. Therefore, yes, it would be acceptable for two no-solo operators to back each other up.


Question: What is the aging management program for iron components in nuclear systems?

Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.27)

Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)


Question: What is the initial reactivity worth in nuclear fuel assemblies?

Initial Reactivity Worth, %Δρ (hot): 0.0 – 3.0 (typical)

The weight of Boron – 10 per foot of rod, lb/ft: Variable


Question: The NRC analyzes the adequacy of safety evaluations in relation to external hazards such as flooding.

According to Petitioners’ expert report, the April 2009 Giitter Letter is significant because it indicates that the Staff planned to evaluate the results of Duke’s inundation model benchmarking 'to determine whether further regulatory actions are necessary to ensure there is adequate protection against external flooding at Oconee' and states the Staff’s expectation that it would receive sensitivity analyses 'which would establish an adequate licensing basis for external flooding.' Mitman Report at 12 (quoting April 2009 Giitter Letter at 3).

Nonetheless, in the context of their section 2.335 petition seeking to show a “significant” environmental matter that requires a regulatory waiver, in our view this does not relieve Petitioners of the responsibility to address, to the extent they can, the adequacy of the post-Fukushima evaluations in the context of whatever relevant public information is available.


Question: Main Condenser backpressure

**Key Parameter Response:** Main Condenser backpressure ↑, Main Chimney flowrate ↑, Off-Gas radiation level ↑, Main Generator MWe ↓

**Automatic Actions:** Main Turbine trip at 10 in. Hg and Rx scram at 9 in. Hg backpressure


Question: What is meant by the term 'End State' in nuclear licensing?

The “End State” is defined as the configuration of the remaining below-ground basements, above-ground structures, buried piping, and open land areas at the time of license termination.

Figure 2.2 provides the location of the planned remaining below-ground basements, above-grade buildings, and buried piping that will remain at the time of license termination.


Question: This application does not include any materials that are referenced from external sources.

No material has been incorporated by reference in this application.

None of the sections in this document provide external citations or references to other works.


Question: What are the implications of changing reactor coolant system temperature?

Changing reactor coolant system temperature and the effect on pressurizer pressure and level  3.6

The NRC requires licensees to perform regular inspections and maintain records to ensure compliance with established regulations.


Question: energy calibration in nuclear measurement

Peak analysis performed on: 9/24/2019 6:47:48AM includes measurements of energy in keV, detailing original area, area uncertainty, ambient background, and the subtracted area and uncertainty for specific peaks.

The table contains a series of numerical values corresponding to different properties measured during the analysis.


Question: Each APRM must function with a specified number of operable LPRMs to ensure safety.

Technical Specification 3.3.1.1 bases require each APRM to have 11 total operable LPRMs and 2 operable LPRMs per detector level.

The safety analysis report must include an evaluation of the potential consequences of a design basis accident.


Question: What are the independence features of safety systems in nuclear power plants?

The following types of interfaces between the PPS and external systems are summarized below along with independence features that protect the safety system from failures of external systems:

Refer to Section 3.3.2.5 for compliance to IEEE Std 603-1991, Clause 5.9.


Question: The updates to LRA Table 3.5.1 include items addressing increased porosity and permeability as well as material loss due to chemical factors.

LRA Table 3.5.1 Items 3.5.1-24 and 3.5.1-67 are associated with numerous NUREG-1801 items for management of increase in porosity and permeability; cracking; and loss of material (spalling, scaling) due to aggressive chemical attack.

The list of affected LRA sections includes multiple tables that require revisions for clarity and accuracy.


Question: The NRC mandates that plants must maintain structural integrity and leakage boundary to uphold safety functions.

Nonsafety-related components that could affect safety-related functions (10 CFR 54.4(a)(2)): 1. Maintain integrity of nonsafety-related components such that no interaction with safety-related components could prevent satisfactory accomplishment of a safety function through structural integrity or leakage boundary.

Table 3.2.2-2 provides the results of the AMR.


Question: What is the purpose of the NRC's physical security program?

The physical security program (PSP) outlines the measures that licensees must implement to ensure the physical protection of their facilities and materials against unauthorized access and potential security threats.

The NRC's regulations and guidelines are documented in various publications, including the Federal Register, where stakeholders can review proposed rules and submit comments.


Question: What are the additional security measures for independent spent fuel storage installations?

The Order requires compliance with two sets of requirements: (a) 'Additional Security Measures for Physical Protection of Dry Independent Spent Fuel Storage Installations' (Enclosure 1, Attachment 1), and (b) 'Additional Security Measures for Access Authorization and Fingerprinting at Independent Spent Fuel Storage Installations' (Enclosure 1, Attachment 2).

The Commission, following the terrorist events of September 11, 2001, determined that the additional security measures (ASMs) included in this Order are required to ensure national safety and security.


Question: reactor protection system

The ESFS and reactor protection systems operate as designed to initiate the appropriate reactor trip and safety actuation signals. (See NSSS Startup Manual, SU-2.7.2 and SU-2.7.3.)

Required slave output relay field wiring has been electrically isolated to prevent operation of components controlled by the relays except where specifically removed in the preoperational test procedure.


Question: The activity concentration of cesium-137 in terms of pCi per gram is a critical measurement in nuclear safety assessments.

CS-137 has an Energy of 661.65 keV, a Yield of 85.12%, and an Activity of 1.77E-01 pCi/grams.

The yield percentages for europium isotopes indicate their relative production in nuclear reactions, with EU-152 at 28.40%.


Question: What are tone alert monitors in emergency response procedures?

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

The reactor's coolant system is designed to maintain safe operating temperatures and pressures during normal and emergency situations.


Question: What is the significance of Ultrasonic Examination in nuclear weld testing?

Until studies are complete that demonstrate the ability of UT to replace RT for construction, the NRC will not endorse UT in lieu of RT Code Cases or generically allow the substitution of UT in lieu of RT for construction examinations.

LWR experience has shown that numerous in-service defects and failures initiate at the location of fabrication weld repairs.


Question: What species should be considered in an effects analysis for nuclear projects?

There is a total of 16 threatened, endangered, or candidate species on this species list.

Please contact the designated FWS office if you have questions.


Question: What is the significance of 4.16 kV emergency bus undervoltage in nuclear safety systems?

Two events for Division 3 are used to model a loss of Emergency Bus Undervoltage Signal in the PRA. This event is used as a surrogate for inoperability of any channel in these functions. This is conservative because diverse signals and redundant channels are not explicitly credited to start the standby DGs and transfer the Emergency Bus source.

Time Delay for power recovery is a crucial aspect of ensuring that safety systems can restore necessary power levels after a disturbance.


Question: What is the role of primary containment in nuclear reactors?

Primary containment is intended to function as a flood barrier, missile barrier, and shelter/protection in addition to providing shielding and structural pressure barrier functionalities.

The ASME Section XI, Subsection IWL provides guidelines for aging management programs to ensure structural integrity and safety compliance.


Question: The procedures for managing radiological emergencies are integral to ensuring public safety.

List of EBRP Radiological Emergency Implementing Procedures

List of EBRP Radiological Standard Operating Procedures Cross Reference


Question: post-LOCA boron precipitation analysis

The post-LOCA boron precipitation analysis determines the acceptable time to provide an active dilution mechanism to prevent precipitation of boric acid in core.

The change to GAIA fuel necessitates an assessment of Generic Letter 2004-02 (also referred to as Generic Safety Issue (GSI)-191, ML042360586). Attachments 6 (non-proprietary) and 12 (proprietary) contain the detailed evaluation.


Question: The design of the reactor helium pressure boundary must ensure that operational limits are maintained under all expected scenarios.

All systems that are part of the reactor helium pressure boundary shall be designed with sufficient margin to ensure that the design conditions of the reactor helium pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

The modified PDC accommodates Safety-Significant PSF 1 Control Radionuclide Transport from the Primary Boundary.


Question: The DEIS's overall assessment of transportation impacts does not address all potential risks associated with shipments near the proposed site.

The NRC requires licensees to analyze emergency zones, including the impacts of transportation on public health and safety.

All are premised on information that was available to Fasken long before publication of the DEIS.


Question: matrix solid swelling model

In regards to this RAI, the matrix solid swelling model is classified as an independent model, typically based on the swelling rate coefficient, and is described in Section 7.4.1 of Reference 5.

Fuel diameter and length changes are fuel rod behaviors (results) which the code has to predict as these phenomena are not explicitly modeled.


Question: What are the aging effects on aluminum piping in nuclear auxiliary systems?

Loss of material due to pitting and crevice corrosion associated with aluminum piping, piping components, and piping elements exposed to Treated water is noted in the aging management evaluations.


Question: GAIA fuel temperature analysis

Transient Fuel, Fuel Rim, and Clad Temperature – GAIA Fuel.

The Byron and Braidwood Rod Ejection Accident Analysis provides a detailed assessment of the potential consequences of a control rod ejection accident.


Question: The functionality of the containment sumps is crucial for maintaining reactor safety during operation.

**LCO 3.6.8** Two containment sumps shall be OPERABLE.

Initiating action to mitigate incidents is vital for the ongoing safety and compliance of the reactor.


Question: The SLC System must function properly to keep offsite doses within the limits set by 10 CFR 50.67 after a loss of coolant accident, ensuring iodine retention in the suppression pool.

The SLC System satisfies Criteria 3 and 4 of Reference 3.

In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref. 5) limits following a LOCA involving significant fission product releases.


Question: What is cadmium worth in nuclear calculations?

The GEMER ENDF/B-IV cross-section library treatment of cadmium is based on the 190-multigroup structure, as shown in Figure 6.22. This figure is generated directly from the GEMER cross-section library.

Based on the data provided in Table 6.12, calculations can be performed with the following equations to derive the population mean (μ), population variance (σ²) and population standard deviation (σ) for the sixteen calculated eigenvalues (N=16). This is done for the flux weighting distribution.


Question: The evaluation of conditional failure probabilities (CFPs) is a crucial part of the risk-informed approach to safety assessments.

The risk-informed approach identifies scenarios that fail any one of the following GSI-191 acceptance criteria, as determined by break-specific analysis:

The methodology includes conservatisms in the plant-specific testing and in the assumption that all unbounded breaks result in loss of core cooling.


Question: seiche hazard

Seiche refers to flooding from water displaced by an oscillation of the surface of a landlocked body of water, such as a lake, that can vary in period from minutes to several hours.

Sudden ground motion or vibration of the Earth as produced by a rapid release of stored-up energy along an active fault.


Question: What is the purpose of HEC-RAS in hydrologic modeling?

The HEC-RAS model is used for simulating the hydraulics of water flow through natural rivers and streams, improving the design of hydraulic structures and floodplain management.

The table presents elevation data collected at various stations along the VEGP site, showing significant variations in elevation that are crucial for geological surveys.


Question: peak energy measurements in nuclear safety reports

The energy peak measurements, such as F 1 at 294.99 keV and F 4 at 1460.82 keV, indicate the energies at which specific interactions occur in the detection process.

The subtracted area is calculated by taking into account the ambient background and applying respective uncertainties to ensure accuracy in the results.


Question: The regulatory requirements for nuclear licenses include detailed safety analyses and risk assessments.

Nuclear license applicants must submit comprehensive documentation that includes safety analysis reports, which are evaluated by the NRC to determine compliance with regulations.

The reactor needs to be shut down for maintenance based on the standard operational procedures that ensure safety during the process.


Question: The Commission encourages public engagement and transparency in regulatory processes.

The NRC values the input of stakeholders and aims to incorporate public participation into its decision-making processes regarding nuclear regulatory matters.

The Federal Register provides a diverse range of notices and proposed rules that may impact various industries, indicating the breadth of government oversight.


Question: What are general arrangement drawings in nuclear regulatory filings?

General Arrangement Drawings are detailed illustrations that depict the layout and organization of major components within a nuclear facility, including critical safety systems and structures.

The reactor pressure vessel plays a crucial role in maintaining safety by ensuring that the nuclear reaction occurs under controlled conditions.


Question: safety equipment requirements for nuclear facility operations

Tools/Equipment/Procedures Needed: Hard hat, safety glasses, gloves and approved shoes.

The Watts Bar Nuclear Plant is located in Spring City, Tennessee, and is part of the Tennessee Valley Authority.


Question: What is the COBRA-FLX Critical Heat Flux Correlation?

COBRA-FLX is a thermal-hydraulic code used to predict critical heat flux in nuclear reactor cores, allowing for the assessment of safety margins in nuclear designs.

Table B-1 lists the test assembly conditions for non-mixing grid tests, detailing parameters such as inlet temperature and pressure for various experimental setups.


Question: Revisions to documents related to the River Bend facility must be reviewed periodically to ensure accuracy and compliance.

All information contained within the RBS Attachment and Enclosures will be reviewed annually by the Parish OHSEP.

Changes such as misspelled words, names, etc., must be corrected, but are not a revision by themselves.


Question: The difference in Relative Percent Difference (RPD) is considered for samples with a concentration above five times the required detection limit.

The Relative Percent Difference (RPD) obtained from the sample duplicate (DUP) is evaluated against the acceptance criteria when the sample is greater than five times (5X) the contract required detection limit (RL). In cases where either the sample or duplicate value is less than 5X the RL, a control limit of +/- the RL is used to evaluate the DUP result.

Consult Case Narrative, Data Summary package, or Project Manager concerning this qualifier.


Question: RCCA bank withdrawal effects

Figure 15-71 illustrates the uncontrolled withdrawal of a control rod cluster assembly (RCCA) bank from the reactor core, showing the average temperature in the reactor coolant system (RCS) as a function of time.

The reactor coolant system (RCS) plays a critical role in maintaining safe operating conditions by removing heat from the reactor core during normal operations.


Question: The NRC mandates that licensees ensure the Reactor Protection System operates correctly during critical functions.

Preset: M:RP01AB, Reactor Protection System Automatic Scram Failure

The plant is operating at approximately 80% power.


Question: The NRC establishes guidelines for the management and disposal of radioactive waste under various regulations.

The Resource Conservation and Recovery Act mandates proper waste management practices to ensure the safe disposal of hazardous materials.

The Hobbs Act provides regulations related to the jurisdiction over cases regarding licensing and legal appeals.


Question: What is the significance of internal flooding in nuclear reactor safety assessments?

3 - Combined. A bounding assessment of internal flooding improvements is addressed in SLR SAMA 52.

3 - Combined. SLR SAMA 21 addresses SAMAs related to extending the time DC power is available.


Question: nuclear safety analysis

The safety analysis report shall evaluate the potential risks and identify appropriate safety measures to ensure the protection of public health and safety in the event of a nuclear incident.

The cooling system is designed to maintain optimal temperatures during regular reactor operations and is monitored continuously.


Question: Reporting criteria for nonconformances in nuclear licensing

Nonconforming conditions that make the containment sump inoperable will meet 10 CFR 50.73 reporting criteria for a condition prohibited by Technical Specifications (TS).

The specific requirements for the license application described in RG 1.174 (Reference 4) are addressed in Enclosures 1 and 2.


Question: The Commission acknowledged gaps in knowledge regarding reactor safety issues, exacerbated by the pace of reactor design advancements.

Witnesses from that laboratory affirmed that reactor designs have gotten ahead of the Commission’s understanding of reactor safety problems, because reactor safety research has not kept abreast with the requirements for it in the expanding U.S. nuclear power program.

At the conclusion of the hearing, the A.E.C. staff, led by Stephen Hanauer, recommended that the A.E.C. reaffirm its June 1971 approval of the existing E.C.C.S. designs—and the Commission did so.


Question: What is a 500 ml Marinelli sample used for in gamma spectrum analysis?


Question: Strontium-90 measurement in soil samples

GFPC, Sr90, Solid "Dry Weight Corrected" | Strontium=90 | U | +/0.0546 | 0.0954 | 0.400 | pCi/g

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


Question: What is the function of valve bodies in nuclear safety systems?

Valve Body      | Pressure Boundary        | Copper Alloy (≤15% Zinc and ≤8% Aluminum)  | Plant Indoor Air (Int)  | Loss of Material                  | Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) | VII.G.AP-143   | 3.3-1, 089  | B, 1

The document outlines various components that play essential roles in ensuring the structural integrity of the plant.


Question: What is the purpose of using isotopes in nuclear applications?

Isotopes, such as I-123 and I-131, are used in various nuclear applications because they have specific decay properties that make them suitable for medical diagnostics and treatment.

The total heat output from the system is measured at 22.0 x 10^0, which indicates the efficiency of the thermal conversion process.


Question: The NRC stipulates that investigations must be made into safety measures for electrical systems within the context of fire-related assessments.

Meeting to discuss options for harvesting electrical distribution components to support fire research activities.

When: Tuesday, February 28, 2017 2:00 PM-3:00 PM (UTC-05:00) Eastern Time (US & Canada).


Question: Licensees must submit detailed plans to ensure compliance with safety regulations before commencing operations.

The licensing process for nuclear facilities involves rigorous evaluation of safety, operational plans, and compliance with established regulations.

Cooling systems are essential for maintaining reactor temperatures and preventing overheating during operation.


Question: Regulatory requirements for RadICS platform-based safety systems

The RadICS platform TR does not address traceability activities associated with application specific logic. Therefore, plant application requirements traceability activities for RadICS platform-based safety systems must be performed during plant application development and thus could not be evaluated in this SE.

Addressed in PSAI 7.2 and satisfied by: [[ - - - - ]].


Question: aging management programs in nuclear facilities

The DCPP Protective Coating Monitoring and Maintenance AMP, with enhancement, will provide reasonable assurance that the effects of aging will be managed so that the intended function(s) of components within the scope of the AMP will be maintained consistent with the CLB during the PEO.

when necessary through the systematic and ongoing review of both plant-specific and industry OE, including research and development, such that the effectiveness of the AMP is periodically evaluated.


Question: bolting integrity management in nuclear power plants

The bolting integrity management program encompasses measures to monitor the integrity of bolting (closure) components, which can experience loss of material or preload.

The heat exchanger (air side seal oil cooler) shell side components serve as a leakage boundary and are subjected to loss of material due to lubricating oil exposure.


Question: What is the purpose of the signed form in nuclear regulatory compliance?

The signed form is essential for the Nuclear Regulatory Commission's compliance processes, ensuring appropriate documentation and communication with the relevant agencies.

Once the signed form is received, please send it to IPAWS@FEMA.DHS.GOV.


Question: What are the requirements for testing centrifugal charging pumps in nuclear power units?

A. This Instruction is required to be performed quarterly while the Unit is in Mode 1, 2, or 3.

B. Only ONE CCP can be operable while the Unit is in Mode 4, 5 or 6. This Instruction or 1-SI-62-901-B is required to be performed quarterly when the Unit is in these modes.


Question: How can the potential impacts of TENORM waste be mitigated during construction?

TENORM will be mitigated by minimizing the quantity of drilling fluids. Other mitigation measures that will minimize potential impacts from TENORM waste disposal include disposal in mud pits in accordance with EPA TENORM regulations, backfilling, restoring and reseeding mud pits, typically within a single construction season, using sediment control BMPs, avoiding construction in areas with previously identified, potentially NRHP-eligible cultural sites, and stopping work if any previously undiscovered cultural resources are encountered during construction or reclamation of mud pits.

Potential impacts from domestic sewage will be mitigated by complying with NMED standards for designing septic tank/leach field systems and performing routine maintenance such as septic tank pumping when needed and leach field inspection for ponding.


Question: The NRC oversees the licensing of nuclear facilities to ensure compliance with safety and security regulations.

The NRC requires licensees to analyze emergency zones and submit detailed reports as part of their regulatory filings.

Kevin Williams serves as the Director of the Division of Materials Safety, Security, State, and Tribal Programs.


Question: What is the significance of failure probability in nuclear systems?

The failure probability (demand) for the Main Steam system components is calculated to be 3.63E-04, indicating a low likelihood of failure during operational demands.

The isolation valve operator is critical to maintaining safety during a loss of coolant accident (LOCA), but it operates independently of the failure probability calculations.


Question: The NRC mandates that all licensees must adhere to strict guidelines when filing their reports.

The regulatory framework for nuclear licensing requires that every applicant submit detailed technical specifications to ensure safety and compliance.

The reactor core is designed to operate at a maximum temperature of 300 degrees Celsius, providing efficient thermal management during operations.


Question: The TRM stipulates that the Refueling Machine must have a capacity of at least 3150 pounds and defines specific electrical overload cutoff limits.

2-FHI-4, Refueling Machine, requires that interlocks are to be bypassed ONLY under direct supervision of the Fuel Handling Supervisor (FHS). The TRM requires that the Refueling Machine shall have a capacity of ≥ 3150 pounds, and two electrical overload cutoff limits of ≤ 2650 pounds and ≤ 2800 pounds, respectively. With the overload cutoff limit interlock bypassed, the refueling machine is inoperable, and refueling operations must not be resumed.

Plausible because Shift Manager permission is required to perform core alterations. The second part is plausible because it is possible to defeat the interlock and resume refueling operations; it is prohibited by procedures and the TRM. This would be the correct answer if the fuel assembly is outside of the reactor pressure vessel and the FHS wanted to place the Fuel Assembly in a safe location (RCCA Change Fixture).


Question: The system designed for maintaining the cleanliness of the containment atmosphere is crucial for operational integrity.

The Containment Atmosphere Cleanup System is responsible for ensuring that the atmosphere within the containment structure is maintained at acceptable levels.

The containment structure itself is designed to withstand external pressures and prevent radioactive leaks to the environment.


Question: What is the pellet-to-clad gap in nuclear reactor fuel rods?

The pellet-to-clad gap in the PWR rod design for both Zr-4 and M5 cladding typically starts to close around a rod burnup of [ ] GWd/tM.

In the development of the response to RAI 30, a. The gap is considered open if there is no pellet-cladding contact and [ ] These criteria ensure free creep of the cladding due to external pressure effects.


Question: The NRC anticipates seeing additional applications concerning advanced reactor designs.

This rulemaking would accommodate the licensing of any nuclear power reactor design, including any potential micro-reactor and/or small modular reactor applications.

The preliminary priority for Alternative 3, Rulemaking, is medium, which would be a significant contributor to supporting the NRC’s mission for conformance with external regulations and NRC licensing initiatives.


Question: What is an audit summary report in nuclear licensing?

The NRC staff will issue an audit summary report within 90 days of the audit exit.

The licensee must submit their updated technical specifications to demonstrate compliance with regulatory standards.


Question: What is the willingness to pay principle in regulatory analyses?

The WTP principle captures the notion of opportunity cost by measuring what individuals are willing to forgo or pay to enjoy a particular benefit.

A formal data collection plan may be useful. Such a plan would include tasks to identify the types of data available; to acquire the data with supporting documentation; to determine which estimating methods and models will be used with which dataset; and to verify, validate, and normalize the data.


Question: wind speed observations

Total No. Observations: 95 for the wind direction E-NE were recorded, showing varying wind speeds and their corresponding frequencies.

In the report, various conditions of the cooling water system were examined to evaluate its operational efficiency and compliance with safety regulations.


Question: The NRC assessed various methods to manage radioactive materials effectively, including sprayer systems and remote technologies.

The use of the containment emergency spray system to spray down large volumes of water, detergents, chemicals, or steam on the structures, systems, and components inside the containment building was not pursued.

Experience showed that water volume per unit area required for run off was considerably less than that for horizontal surfaces.


Question: nozzle-to-vessel welds in pressurizers

Item No. B3.110 (Pressurizer, Nozzle-to-Vessel Welds) is applicable to all pressurizer nozzle-to-vessel welds. Every PWR plant performs these examinations unless the nozzles were integrally forged with the vessel.

The surge nozzle is located in the bottom head of the pressurizer and is subjected to Insurge/Outsurge events due to inventory changes within the pressurizer.


Question: What is the role of a Radiation Safety Officer in managing radioactive materials?

You, Michelle Kritzman, have been appointed Radiation Safety Officer and are responsible for ensuring the safe and secure use of radiation and radioactive material.

In addition, you are free to raise issues with the U.S. Nuclear Regulatory Commission at any time.


Question: A request for a license renewal must be submitted at least 30 days prior to the expiration date to ensure the license remains valid during the review.

Means receipt of an application for renewal of a license that has been filed 30 days or more before the license's expiration date. The license remains in effect until processing of the application for renewal has been completed.

The technical staff members are responsible for processing the assigned licensing actions in accordance with the priorities.


Question: What does the NRC require for alternate methods of disposal according to 10 CFR Part 50 or 52?

In the case of 10 CFR Part 50 or 52 licensees, th[e] requirement [to seek NRC approval for alternate methods of disposal] is based on the NRC’s jurisdiction over the operation of nuclear power plants, which cannot be delegated to an Agreement State.

But this assertion flies in the face of the plain text of the 2019 Letter, which, by its terms, is a 'respon[se]' to NEI’s 'request[]' that the NRC 'rescind' the Issue Summary, and a decision not to do so.


Question: The proposed ALWR units and SMR unit will utilize advanced cooling technologies to minimize water consumption.

ALWR: Cooling water withdrawal—50 mgd <br> Consumptive water use—35 mgd (NRC 2013-TN6435)

Six units for a total of approximately 2,600 MWe


Question: The NRC evaluates potential environmental impacts and alternatives during the license renewal process for nuclear plants.

As documented therein, the NRC staff's recommendation is that the adverse environmental impacts of the SLR for Oconee Station (i.e., the continued operation of Oconee Station for a period of 20 years beyond current expiration dates) are not so great that preserving the option for energy-planning decision-makers would be unreasonable.

In accordance with Title 36 of the Code of Federal Regulations Section 800.8(c), the NRC staff elected to use the National Environmental Policy Act of 1969, as amended, process to comply with its obligations under Section 106 of the National Historic Preservation Act of 1966, as amended.


Question: The NRC mandates that operators conduct detailed assessments of safety-critical systems, such as reactor containment.

The updated procedure utilized, BOP RH-6, to revision 49, reflects the current standards in reactor operation and maintenance.

Change JPM designator, this JPM was previously utilized as CR-d on 2011 Cert Exam.


Question: Connecticut is proposing to adopt certain definitions into regulation, including terms related to site closure.

**Definitions Decommissioning**
Connecticut states that the definitions are not included in the proposed legislation but will be adopted in regulation. However, this definition was omitted from the list.
Connecticut needs to provide an explanation or justification as to the reason for the omissions; or supply the missing information.


Question: The sump strainer must meet the design codes and withstand various loads including pressure differentials and missile impacts.

The objective of the sump structural analysis section is to verify the structural adequacy of the sump strainer including seismic loads and loads due to differential pressure, missiles, and jet forces.

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


Question: radiation exposure limits in emergencies

The required dose to perform the job is 17,000 mRem.

An emergency is in progress with the following: - An Alert has been declared.


Question: The NRC mandates that reactor trips occur at specific temperature limits to prevent safety hazards.

The Low Pressure Reactor Trip is designed to activate under certain conditions to maintain thermal limits and protect reactor integrity.

Nominal Opening Pressure refers to the pressure level at which a valve begins to open, irrespective of reactor safety mechanisms.


Question: What are the qualities that define the recovery limits in lab analyses?

Upper Recovery Limit, in % (except for LCSS, mg/Kg) and Lower Recovery Limit, in % (except for LCSS, mg/Kg) are critical parameters for evaluating the accuracy of quantitative analysis in the lab context.

Continuing Calibration Verification standard is utilized to ensure that the calibration remains valid throughout the testing process.


Question: The La Crosse Boiling Water Reactor has undergone rigorous safety evaluations...

The La Crosse Boiling Water Reactor is designed with multiple layers of safety features to prevent and mitigate accidents.

This document outlines the administrative procedures for submitting filings to the NRC.


Question: What are the sources of Probable Maximum Precipitation Estimates?

NOAA, 1978. Hydrometeorological Report No. 51, Probable Maximum Precipitation Estimates, United States East of the 105th Meridian, U.S. Department of Commerce, NOAA, USACE, Washington, D.C.

The University of Missouri offers an extensive range of resources through the Missouri Spatial Data Information Service, accessible for various research purposes.


Question: What is the significance of environmental reports in nuclear license applications?

The Environmental Report provides an analysis of the potential environmental impacts of the proposed license renewal, including assessments of water, air quality, and ecological effects.

The U.S. Environmental Protection Agency regulates emissions from various industries to protect air quality and human health.


Question: What is the impact of void fraction on NPSH margin in nuclear systems?

Void fraction at the pump suction could affect pump NPSHr and NPSH margin. The methodology in NUREG 0897 (Reference 55), along with the void fractions shown in Table 3.1.14-3, was used to correct the NPSHr obtained from the vendor curves.

The total strainer head loss was calculated by combining the clean strainer and debris head losses (see the Response to 3.f.10). The head loss of the ECCS circuit piping between the strainer exit and the pump suction was accounted for in a hydraulic model using the Fathom software.


Question: byproduct material licensing

Any byproduct material permitted by 10 CFR 35.100

In accordance with letter dated October 29, 2020.


Question: What is the heat exchange coefficient in nuclear safety assessments?

The heat exchange coefficient is the net rate at which heat is lost or gained by a body of water, is defined by the following equation (Ref 1): K = 15.7 + (0.26 + B) f(w), BTU/FT²/Day/F.

The meteorological data from 1951 to 1972 was scanned in order to find the highest monthly average dewpoint (71°F in July 1961), and the lowest average wind speed (4.7 mph in August 1953).


Question: The materials used in the reactor coolant pressure boundary must meet specific standards for safety and reliability.

The closure bolting for the primary side consists of studs made from ASME SA-193 Gr. B7 and nuts from ASME SA-194 Gr. 7, which are specially designed for use with hydraulic stud tensioners.

The steam generator components include parts such as the lower assembly shell, which is made from ASME SA-508, Class 3 Forging, and the transition cone, also constructed from ASME SA-508, Class 3 Forging.


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

Wind speed measurements are critical for assessing the dispersion of radioactive materials and ensuring proper emergency response planning in nuclear facility operations.

Nuclear reactors operate under several regulatory frameworks that ensure their design and operation meet strict safety standards.


Question: The primary coolant loop model for Hamaoka 5 includes steam line connections and feedwater boundary locations.

A schematic of the POLCA-T Hamaoka 5 primary coolant loop model, steam line connections and feedwater boundary location are shown in Figure C.5-19.

A schematic representation is necessary for understanding the thermal performance of different reactor designs.


Question: Flood Hazard Reevaluation Report

This completes the flooding hazard reevaluation for FNP as required by Reference 1.

Mr. C. R. Pierce states he is the Regulatory Affairs Director for Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.


Question: 10 CFR 50.91 state consultation

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), CEG is notifying the Commonwealth of Pennsylvania of this supplemental letter by transmitting a copy of this letter to the designated State Official.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 5th day of February 2025.


Question: What is General License Registration Number in the context of nuclear regulation?

General License Registration Number: GL-705978-26

The address should include both the city and state for accuracy.


Question: What is the purpose of the Heat Exchanger in the Component Cooling Water System?

Heat Exchanger (CCW Heat Exchanger) Tubes are essential for Heat Transfer, using Copper Alloy materials in a Closed-Cycle Cooling Water environment, which helps prevent reduction of heat transfer efficiency as part of the aging management evaluation.

The NRC conducts regular inspections of various safety systems within the plant, ensuring compliance with operational standards.


Question: Revised Technical Specifications for nuclear power plants

The Revised Technical Specifications for the Edwin I. Hatch Nuclear Power Plant Unit 1 address the operational requirements necessary for ensuring the safety and security of the facility.

The cooling system utilizes a series of pumps and valves to circulate coolant through the reactor core and maintain stable temperatures.


Question: Residual Heat Removal Service Water System

The RHRSW System functions to remove heat from the RHR System and Emergency Equipment Cooling Water (EECW) System components by pumping water from Wheeler Reservoir through the Residual Heat Removal (RHR) heat exchangers and Emergency Equipment Cooling Water (EECW) System components and discharges back to Wheeler Reservoir.

The tanks serve as a pressure boundary for various substances in the fuel oil system and are integral to the mechanical closure functions.


Question: What are the components involved in the Reactor Protection System?

The Reactor Protection System (RPS) is designed to monitor and control reactor operations to ensure safety and prevent accidents.

The NRC reviews various regulatory guides to ensure compliance with safety standards.


Question: nuclear device registration

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

The thermodynamic properties of metals are crucial in understanding their behavior under different temperatures.


Question: What is the purpose of the Reactor Building Exhaust Radiation Monitor?

When reactor building ventilation is not isolated; While OPERATION, or Movement of recently irradiated fuel within the secondary containment is in progress & during Inservice Leak and Hydrostatic Testing Operation/ Every 12 hours

Every 12 hours, various checks are performed on the reactor building to ensure compliance with technical specifications.


Question: What is the peak flow measurement for dam C-6 during the breach analysis?

The peak outflow from the breach of composite dam C-6 is 113,000 cfs, which is directed into time with no reach model in between.

In HEC-HMS, a conservative approach was taken to model composite dam C-6 and Dam in series.


Question: reactor vessel nozzle aging management

(N-3A-D Steam Outlet) Reactor Vessel Nozzle, Safe Ends, and Welds | Pressure Boundary | Steel | Reactor Coolant (Internal) | Cumulative fatigue damage: cracking due to fatigue, cyclic loading | Section 4.3 "Metal Fatigue" | IV.A1.R-04 | 3.1-1, 007 | A

(N-4A-F Feedwater) Reactor Vessel Nozzle | Pressure Boundary | Steel | Reactor Coolant (Internal) | Cumulative fatigue damage: cracking due to fatigue, cyclic loading | Section 4.3 "Metal Fatigue" | IV.A1.R-04 | 3.1-1, 007 | A


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

A Safety Analysis Report (SAR) provides a comprehensive description of the facility design, operational details, and safety evaluations necessary to support the licensing of a nuclear facility.

The communications regarding the file storage were primarily about the location and access of a large PDF document in an internal directory.


Question: assembly average burnup

30 ≤ Assembly Average Burnup ≤ 32.5 GWd/MTU


Question: What does PMP Storm Extent refer to in nuclear licensing?

PMP Storm Extent refers to the parameters that define the extent of probable maximum precipitation as it relates to risk assessment and safety analysis in nuclear facility licensing.

The Missouri River Basin from Sioux City to Omaha is a vital waterway that supports various ecological and recreational activities.


Question: peak load in energy planning

Peak load refers to the highest level of demand for electricity in a specified time period, which is critical for energy planning and resource allocation.

The diversity served metric indicates the variety of customer categories that receive electricity, helping utilities understand and manage their customer base.


Question: The NRC defines measurement points for various system flows, including the flow rates for the condenser and feedwater.

The values of “I” (transmitter output current in mA) that corresponds with the above operating flow rates of interest are determined from above Equation 2.7 as follows:

The NRC requires strict adherence to safety protocols during the operation of nuclear reactors to ensure the protection of public health and the environment.


Question: What is gas bubble disease in aquatic organisms?

The resulting condition is known as gas bubble disease.

Eutrophication (nutrient loading) and the resulting effects on chemical and biological oxygen demands have not been found to be a problem at operating nuclear power plants.


Question: What is the role of the Reactor Oversight Process in nuclear safety regulation?

The Reactor Oversight Process (ROP) is a systematic approach established by the NRC to assess the safety performance of commercial nuclear power plants and ensure their compliance with regulatory requirements.

The administration of nuclear substances requires strict adherence to guidelines that govern their storage and disposal.


Question: The analysis focused on the concentrations of gamma emitters such as Sr-90 in the selected samples.

The off-site laboratory, GEL Laboratories, processed the three (3) samples selected for HTD ROC analysis. Samples L2-011-104-FSGS-001-SS, L2-011-104-FSGS-010-SS, and L2-011-104-FSGS-011-SS were selected. Only the HTD ROC Sr-90 was included in the analysis. All analyses met the required MDC.

The statistical properties of the systematic sample population included calculations such as the mean, median, and max values for each radionuclide.


Question: The operation of the automated level control systems ensures that water levels within the feedwater system remain stable during normal conditions.

Each heater drain tank is installed below the number 4 heater and receives the total drain flow by gravity flow.

Isolation valves are provided which allow each steam generator feedwater pump to be individually removed from service, while continuing operations at reduced capacity with the parallel pump.


Question: aging management program for nuclear components

Loss of material due to pitting and crevice corrosion, cracking due to SCC is an aging effect requiring management for structural support members and connections made of stainless steel in conducive environments.

Support members; welds; bolted connections; support anchorage to building structure serve as structural support in various material types and environments.


Question: RadICS integration plans

The NRC staff determined the RadICS integration plans provide an acceptably documented method for performing product integration activities needed for safety related digital I&C system development.

Addressed in PSAI 7.2 and satisfied by: [[ - - - - - ]]a,c,e


Question: The NRC emphasizes the importance of monitoring radioactive leakage in the system and provides various methods for detection.

Detection of radioactive leakage into and out of the system is facilitated by area radiation monitoring (discussed in subsection 12.3.4), process radiation monitoring (discussed in section 11.5), and steam generator blowdown sampling (discussed in subsection 10.4.8).

However, these lines are designed to withstand a safe shutdown earthquake, and 100 percent volumetric inspection at installation is provided (i.e., 100 percent volumetric examination of shop and field longitudinal and circumferential welds).


Question: reactor power control

Raise Reactor Power with Recirculation Flow

Secure Unit Substation Transformer 71T-9 load monitoring


Question: The NRC mandates that academic institutions and facilities maintain precise identification methods for radionuclide samples.

The Peak Analysis was performed on June 24, 2019, at 11:32:59 AM and ranged from channel 100 to channel 4096.


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

The COBRA-FLX critical heat flux correlation is designed to predict the onset of critical heat flux in various thermal-hydraulic conditions and is essential for ensuring safety in reactor operations.

The test assembly conditions for HTP grid tests include various parameters such as inlet temperatures, pressures, and flow rates, which are crucial for evaluating thermal hydraulic performance.


Question: asw pump operation

- (6.1.1) Determines current configuration: ASW Pump 1-2 running and CCW HX 1-2 in service.

- (6.1.12) Closes FCV-603, CCW HX 1-2 saltwater inlet valve.


Question: The average activity level for the nuclide K-40 is measured at 4.70E+00 pCi/grams.

K-40 has a yield of 10.67% and an energy level of 1460.75 keV, with an activity of 4.70E+00 pCi/grams.

Comments on the nuclides include notes on their respective confidence and activity levels, but do not discuss any specific measurement techniques.


Question: What is the activity of PB-214 nuclide measured in pCi/grams?

The weighted mean activity of PB-214 is measured at 5.83E-02 pCi/grams.


Question: The NRC's licensing review encompasses stored equipment that must be included if its availability is critical for operation or function, especially as per 10 CFR 54.4(a)(2).

In some cases, stored equipment may be included in the scope of LR. If an SSC is considered stored and availability of that SSC and/or follow on operation or function of that SSC is credited for design bases or regulated events, those stored SSCs are considered in scope for LR.

The evaluation process for consumables is consistent with the guidance provided in NUREG-1800.


Question: What is the significance of the CWA Section 316(a) variance for nuclear facilities?

With each NPDES permit renewal application, Duke Energy has requested, and the SCDHEC has granted, continuation of the CWA Section 316(a) variance.

Because the characteristics of the thermal effluent would remain the same under the proposed action, the NRC staff anticipates similar effects.


Question: What actions should the operator take when the electrical system misaligns following a turbine trip?

Which one of the following describes the required operator action, in accordance with the Abnormal Operating Procedures?

The plant is operating at 100% power with the following: The Main Turbine trips.


Question: safety system design requirements

Since the RadICS TR does not address a specific application or establish a definitive safety system design, the evaluation against this requirement is limited to consideration of the means provided within the platform to enable testing and calibration of redundant portions of a safety system.

Addressed in PSAI 7.11 and satisfied by: - Licensing Technical Report


Question: What are the characteristics of water wells located near nuclear facilities?

Offsite Registered Water Wells within 2 Miles of DNPS Center Point (Sheet 5 of 8) include various well identifiers, distances from the center point, their depths in feet, status descriptions, and formations from which the water is drawn.

The unit of measurement used for well depth is feet, which reveals how far the water source extends below the surface.


Question: The NRC's assessment indicates that the modifications do not lead to a significant increase in risk.

Based on the above, DTE concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DTE has reviewed the environmental evaluation included in the model safety evaluation published on November 21, 2018 (ADAMS Accession No. ML 18267A259) as part of the Notice of Availability.


Question: The analysis of water quality includes evaluating parameters such as conductivity which is measured in umho/cm.

Conductivity is measured in umho/cm, and it is essential for assessing the quality of the water used in various processes.

The analysis of the sample chloride levels shows variability across different test conditions.


Question: NextEra Common Emergency Plan

The proposed NextEra Common Emergency Plan establishes an updated licensing basis for the NextEra sites that complies with current NRC regulations in 10 CFR 50.47(b) and 10 CFR 50, Appendix E.

This update also incorporates the four supplemental documents and addenda that have been issued in the intervening years and is intended to modernize and consolidate the guidance making it easier for users to understand.


Question: The offsite emergency operations facility is designed to ensure effective communication and assessment during an incident.

Each EOF provides communications to the Control Room, TSC, field monitoring teams, NRC, and OROs. The EOF has the capability for the acquisition, display, and evaluation of unit, radiological and meteorological conditions necessary to perform accident assessment and determine protective measures.

Coordination of radiological and environmental assessment activities with offsite agencies.


Question: The evaluation of internal events considers modeling uncertainties as part of the Probabilistic Risk Assessment (PRA).

Modeling uncertainties are considered in both the base PRA and in specific risk-informed applications.

HCVS failure rates are based on rates for like-components from NUREG/CR-6928 (updated with plant-specific experience when available), such as for vent valves and batteries.


Question: What is organizational standing in regulatory filings?

The Black Hills Clean Water Alliance has organizational standing.

In her declaration, Dr. Jarding asserts that she and other Alliance members "view and photograph scenery and wild plant life, appreciate and value the cultural and historical resources at the site, and generally enjoy using the area of the Project for recreational, cultural, historical, conservation, and aesthetic purposes."


Question: electric shock safety in nuclear plants

Design criteria for nuclear power plants that limit hazards from steady-state currents are based on the NESC, adherence to which requires that utility companies design transmission lines so that the short-circuit current to ground produced from the largest anticipated vehicle or object is limited to less than 5 milliamperes.

Xcel Energy finds that impacts to electric shock hazards for the proposed SPEO are SMALL.


Question: What is the purpose of sodium hydroxide in the Spray Add Tank?

To prevent re-evolution of iodine collected in the sump following a DBA.

The control board operator reports the Spray Add Tank Level is at 75%.


Question: BWR fuel assembly design

☐ = Fuel Assembly Locations
- ◻ = Vent/Drain Port Locations
- ✖ = Designated Nonfuel Locations

The reactor's cooling system utilizes multiple redundant pumps to ensure constant flow, which enhances efficiency and minimizes downtime during operations.


Question: What does pCi/g represent in nuclear analysis?

The measurement unit used for the concentration of radioactive isotopes in a sample is picocuries per gram (pCi/g).

The LaCBWR Site Restoration Project involves extensive environmental clean-up and restoration efforts at the site.


Question: What is the role of the Pressurizer Pressure Control System in nuclear reactors?

The Pressurizer Pressure Control System (PZR PCS) is designed to maintain the pressure of the reactor coolant system within a prescribed range during various operational conditions, including startup, operation, and shutdown phases.

The NRC requires all licensees to conduct regular staff training to ensure compliance with regulatory requirements and safety protocols.


Question: What is a Technical Evaluation Report (TER) in the nuclear licensing process?

The Technical Evaluation Report (TER) provides a comprehensive assessment of the technical aspects of a proposed action or project by the NRC staff, which is critical for making informed regulatory decisions.

The NRC staff needs additional time to provide feedback to PNNL on the current version of the Technical Evaluation Report (TER).


Question: What is the role of nuclides in nuclear regulatory assessments?

PB-214 has a confidence level of 0.992 and a weighted mean activity of 6.59E-02 pCi/grams.


Question: nuclear licensing process

The nuclear licensing process establishes the necessary framework for a facility to obtain legal authorization from the NRC to operate.

The North Anna Power Station has two reactors that use Westinghouse pressurized water reactor technology.


Question: nuclear operator licensing

The Operator Licensing Exam Schedule outlines the examination process for licensed nuclear reactor operators, detailing the site, exam week, docket number, and number of candidates participating in the exam.

The table above displays the scheduled exam weeks along with the names of the exam authors and chief examiners involved in the operator licensing process, providing a summary of the personnel assignments.


Question: chromium coated fuel rods

The corrosion rate is determined by the outer surface material of the fuel rods, and the normal operation corrosion of chromium coated AXIOM cladding.

No impact is anticipated on the lost parts analysis. There will not be any change or impact to the storage of the LTAs as the LTA weight is minimally changed.


Question: What is the significance of the ALARA principle in nuclear safety?

Pursuant to UNC-NRC License SUA-1475, submitted herewith are the results of our Annual ALARA Audit conducted on December 18, 2024.

If you have any questions or recommendations, please advise.


Question: requirements for standing under NRC regulations

the requirements for standing under 2.309(d) are different from the requirements for contention admissibility under 2.309(f).

the language that the Commission has used when describing the inextricably intertwined doctrine mirrors the language in the contention in the stability regulation.


Question: What is nuclear licensing and its importance in safety regulations?

Nuclear licensing is a regulatory process through which nuclear power plants must submit comprehensive documentation to ensure their compliance with safety, environmental and operational standards set forth by the Nuclear Regulatory Commission (NRC).

The software update for the monitoring system was successfully implemented earlier this month, allowing for enhanced data collection and analysis.


Question: What are ARITA parameters in nuclear regulatory filings?

Here is the summary of our technical review of the ARITA parameters.

Also for color coding, the yellow highlights are the areas we have concerns while the blue highlights are areas we think we understand what Framatone has proposed, but will probably check with them that we are not misunderstanding.


Question: The required Shutdown Margin must be at least 1.3% in MODES 3 and 4, indicating safety standards for reactor operation.

The COLR provides the required SDM limits for MODES 3, 4, and 5, and MODE 2 with Keff < 1.0. Per the Cycle 2 COLR, SDM must be greater than or equal to 1.3% in MODES 3 and 4, and it must be greater than or equal to 1.0% in MODE 5.

The Minimum Required Shutdown Boron Concentration values are crucial for maintaining reactor safety, particularly considering the temperature variations in BOC, MOC, and EOC scenarios.


Question: The NRC's regulatory framework includes requirements for analyzing the safety of the systems in place.

The ESB Dissolver CNV wall is designed to accommodate specific flow rates and ensure the proper control of condensate.

This figure illustrates the layout of equipment in the nuclear power plant, showcasing various components in their operational configurations.


Question: What are EPRI MRP reports in relation to RVI gap analysis?

Thus, the licensee should be prepared to identify and discuss all EPRI MRP reports that were being relied upon for the development of RVI gap analysis in the AMP basis document, and if more recent EPRI MRP reports are being relied on for the gap analysis (e.g., the proprietary MRP-191, Rev. 2 or MRP-232, Rev. 2 reports), to discuss how differences between MRP 2018-022 and the more recent EPRI MRP report bases are being reconciled for development of the component-specific aging mechanism screening bases, component-specific inspection categories, and I&E criteria for managed components identified in the RVI gap analysis of the AMP (i.e., as cited in Appendix A of the AMP basis document).

The licensee shall also provide an assessment of the potential impact on non-impacted structures, systems, and components that may result from the implementation of changes proposed in the SLRA.


Question: The system is designed to promptly trip the recirculation pumps in response to significant signals like high reactor pressure.

Low water level 2 or high reactor pressure RRCS signals cause a trip of the recirculation pump drive motor breakers.

The entire trip circuits may be tested during plant operation, except for opening of the pump breakers.


Question: What special considerations are there regarding the prescribed analysis for hazardous materials?

The analysis method requested involves the use of 10% Hydrochloric Acid, Nitric Acid, and Sulfuric Acid as preservatives for the samples.

Sample IDs are collected regularly and analyzed to ensure compliance with safety protocols.


Question: The NRC requires operators to monitor the performance of safety systems, including proper functioning of trip lights.

Annunciator window 115-E, POWER RANGE FLUX RATE HI, will clear if the positive rate trip light is LIT.

REPLACE N-42 in Maintenance Bypass using DCS Operator Display: CHANGE to either NSSS Operator or BOP Operator environment.


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

The Westinghouse Emergency Core Cooling System Evaluation Model provides an analysis of the emergency core cooling systems designed to be used in the event of an accident to ensure the reactor core remains cool and does not overheat.

The Two-Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices and Short Tubes addresses the flow characteristics of fluids in various components, but does not involve specific safety systems related to reactor operation.