Question: design basis floods for nuclear power plants

Regulatory Guide 1.59, Revision 2, August 1977, provides guidelines for determining design basis floods for nuclear power plants, ensuring that structures can withstand flooding conditions.

Regulatory Guide 1.58, Revision 1, September 1980, outlines the qualification requirements for inspection, examination, and testing personnel at nuclear power plants.


Question: The calibration of the trip bistable components involves monitoring output for proper actuation as inputs are simulated.

It is assumed that the “calibration adjustments” for trip bistable (1E31-N611A&B) are performed using the same method that is currently used for trip bistable (1E31-N609A&B) which is described in Section 8.19 of Input 4.9.1. This method is defined as slowly raising a simulated input test voltage at the input of the trip bistable, while monitoring the trip bistable output for trip actuation within an acceptable as-left tolerance.

This method is defined as applying three (3) simulated 1 to 5 vdc input test voltages at input of the summer, while monitoring the summer output voltages for proper readings within an acceptable as-left tolerance.


Question: The observed issues with thermocouple measurements at elevated temperatures led to significant findings regarding temperature deviations.

Further examination of such high temperatures measured by thermocouples gave rise to the detection of a cable shunting effect which is defined in "Experiment Analysis and Summary Report for OECD LOFT Project Fission Product Experiment LP-FP-2," as the formation of a new thermocouple junction on the thermocouple cable due to exposure of the cable to high temperature.

From the analyses of core temperature measurements in [the LOFT] LP-FP-2 [experiment], the rapid increase in temperature shown in fig 14 was a result of the oxidation of zircaloy which became rapid at temperatures in excess of 1400 K.


Question: The survey unit was evaluated for conditions that could impact the reliability of measurements.

For survey unit B2-010-103, compliance with the unrestricted release criteria was demonstrated through a combination of surface scanning and surface static measurements with a Ludlum Model 44-116 beta/gamma detector.

The number of systematic measurements conducted was 112, ensuring a robust statistical foundation for the survey analysis.


Question: The NRC emphasizes the importance of a comprehensive overview of rules related to regulatory filings.

The staff does a high level overview of the rule.

There's only two areas where I think I could really help.


Question: The NRC defines several operational scenarios including anticipated operational occurrences that must be analyzed by licensees.

AOO | Anticipated Operational Occurrence: An expected event that may occur during plant operation that is analyzed to ensure it does not cause unsafe conditions.

The EOL (End of Life) refers to the point in time when a component or system can no longer perform its intended function.


Question: What impact do climatological changes have on health risks in regulatory filings?

Will expected climatological changes affect any health impacts from the presence of etiological agents?


Question: The NRC specifies a flow rate for the Reactor Coolant System during refueling that must be equal to or greater than 1000 gallons per minute.

Reactor Coolant System flow rate during refueling mode, gpm | ≥ 1000, and as required to maintain RCS temperature ≤ 140°F

Cooldown time, hours after initiation of RHRS operation | ~16


Question: What is the significance of the Exclusion Area Boundary in nuclear licensing?

A reactor licensee is required by 10 CFR 100.21(a) to designate an exclusion area and to have authority to determine all activities within that area, including removal of personnel and property.

Meteorological data are used to determine various postulated accident conditions as specified in RG 1.145.


Question: PWR vessel internals

Enhancement 1 in LRA AMP B2.1.7, “PWR Vessel Internals,” which states: “Procedure(s) will be revised to include a list of the components that require inspections for the Primary, Expansion, and Existing Programs categories specified in MRP-227, Revision 1-A.”

During staff’s review of the OARs for V.C. Summer, the staff noted that in Refueling Outage (RF) 17, there were 24 studs, washers, and hydrants replaced for RCP C.


Question: Safety injection (SI) is triggered immediately to manage flow to the RCS, utilizing both high-head centrifugal pumps and backup normal charging pumps, affecting system pressure dynamics.

Safety injection (SI) is actuated at time zero, with flow injected to the RCS from two high-head centrifugal charging pumps plus the normal charging pump.

The transient responses for the DNB and limiting pressurizer filling cases are shown in figure 15.5.1-1. Table 15.5.1-1 shows the calculated sequence of events.


Question: nuclear material testing methods

Analytical Methods Performed include a variety of standardized testing protocols such as DOE EML HALS-300 and ASTM C 1475-00 which ensure the accurate assessment of nuclear materials.

The Recovery% values in the table indicate the efficiency of tracer recovery tests conducted during the analysis, with standard acceptable limits defined for each isotope.


Question: The NRC assesses whether the ITAAC, as described in the ICN, truly reflects the requirements laid out in the combined license.

c. NRC personnel with the requisite technical and engineering knowledge have determined that the ITAAC Determination Basis (IDB) of the ICN contains sufficient information including a summary of the methodology for performing the ITAAC, to demonstrate that the licensee has successfully performed the inspection, test, and/or analysis stated in the ITAAC. The methodology described in the IDB of the ICN either was reviewed and approved by the NRC, or is acceptable based on sound scientific, mathematical, and/or engineering principles, and is repeatable, if necessary, without any significant change to the reported result.

Enter "Yes" in the blank at the beginning of a statement below if the whole statement is true, "No" if the whole or part of the statement is not true, and "N/A" if the statement is not applicable.


Question: The NRC specifies that heat exchangers are crucial components that must be monitored for signs of loss of material.

Heat Exchanger – (Regen Heat Exchanger) Shell Side Components | Leakage Boundary   | Stainless Steel | Air - Indoor, Uncontrolled (External) | Cracking | One-Time Inspection (B.2.1.20) | VII.E4.AP-209a   | 3.3.1-004                | C

The reactor water cleanup system is designed to ensure the purity of water used in the reactor systems.


Question: What is the significance of the peak energy in radiological assessments?

Peak energy is an important characteristic in radiological assessments, indicating the specific energy level at which radioactive decay emissions occur.

The units of measurement in radiological assessments are critical for ensuring safety and compliance with regulatory standards.


Question: What is required to comply with the Backfit Rule in nuclear regulation?

The Backfit Rule establishes a more difficult standard for imposing new regulations and positions than the cost benefit standard used in regulatory analysis.

Mainly limited cases, it may not be possible or practical to quantify some of the factors to be addressed in the analysis, and the Backfit Rule permits the judicious and disciplined consideration of other relevant and material factors, including qualitative factors (see Appendix A).


Question: What is the significance of ASME OM Code Case OMN-26 in the context of nuclear safety?

Code Case OMN-26 better aligns OM Code Mandatory Appendix III to the Risk and Margin Based Licensee Motor Operated Valve (MOV) Programs developed in response to NRC Generic Letter 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves," that have been in effect since 1998.

Over the past twenty years, Exelon MOV Programs have demonstrated many margin stable MOVs that can be readily justified to extend from their current MOV Program maximum inservice test intervals of six years (for High Risk) and ten years (for Low Risk).


Question: The Engineered Safety Features at the Virgil C. Summer Nuclear Station were assessed for functionality after a reported incident.

On September 17, 2024, ESF in the Control Building were conservatively declared inoperable upon discovery of degraded steam propagation barrier door latch.

Dominion Energy South Carolina, Inc. is submitting this letter to formally withdraw its Licensee Event Report, for Virgil C. Summer Nuclear Station.


Question: Certificate of Compliance TN-RAM

TN Americas LLC requested renewal of Certificate of Compliance No. 9233 for the Model No. TN-RAM package without changes to the package design, operating procedures, acceptance tests, and maintenance program of the package.

The certificate holder has an NRC-approved quality assurance program.


Question: What are the aging management programs for steel components in nuclear reactors?

The Structures Monitoring (B.2.1.33) program will be used to manage loss of material of carbon steel and galvanized steel in concrete elements (embeddments), equipment supports and foundations, hatches/plugs, manhole/valve enclosures, manholes, handholes & duct banks, pipe whip restraints and jet impingement shields, steel components (including trash racks, grills, gates, ring girder assemblies, stabilizers, truss arrangements, Reactor Building vent stack, transmission towers), steel elements (including liners, anchors, and integral attachments for reactor pedestal and reactor shield wall and the refueling bulkhead) and other miscellaneous structural applications (including closure plates, siding, and blowout panels) exposed to air - indoor uncontrolled and air - outdoor in the Crib Houses, Diesel Generator & HPCI Building, Isolation Condenser Pump House, Main Control Room and Auxiliary Electric Equipment Room, Primary Containment, Radwaste Structures, Reactor Building, Stacks, Switchyard Structures, Turbine Building, and Yard Structures.


Question: corrective action program guidelines

The Corrective Action Program Pre-Screening Meeting on 11/18/2020 is an important process for evaluating potential issues and ensuring issues are addressed appropriately.

The Performance Assessment Review Board (PARB) Meeting held on 11/24/2020 reviewed the overall performance metrics and safety compliance of the plant.


Question: What does Code Case OMN-26 specify for MOV test intervals?

Exelon will implement the relief request (RR) in compliance with Code Case OMN-26 in its entirety, including all tables and associated notes.

The stations included in this relief request are on either an 18- or a 24-month refueling cycle.


Question: What are the methods for seismic analysis in nuclear safety?

The Seismic Analysis Methods described in section 3.7.B.2.1 outline procedures used to evaluate the structural integrity of seismic Category 1 structures under seismic loading conditions, ensuring safety and compliance with regulatory standards.

Nuclear power plants must maintain a detailed maintenance program that outlines routines for equipment checks, procedure updates, and operational parameters to ensure optimal function and compliance with regulations.


Question: Licensees must submit certain documents for approval before initiating regulated activities.

In order to ensure safe operations, nuclear facility licensees are required to submit various applications and reports to the Nuclear Regulatory Commission for review and approval.

The reactor cooling system plays a crucial role in maintaining optimal temperature levels throughout the reactor core.


Question: What is the application of boron substitution in nuclear safety analysis?

Using the boron substitution methodology, the following statistical comparison can be made between the equivalent boron (0.008 gB10/cm²) and the 10% reduced values.

From these cases, the 1/v-absorber cross-section treatment using boron-10 at an areal density of 0.008 gB10/cm² was demonstrated to yield statistically equivalent results as shown below.


Question: reactivity control systems

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the safety-significant elements of the primary coolant boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor vessel internals to impair significantly the capability to cool the core.

The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.


Question: What is the EPA 200.7 testing method for metals in water?


Question: acceptance criteria for nuclear regulatory filings

The acceptance criteria for technical specifications must be clearly defined and adhered to throughout the operational processes to ensure compliance with regulatory requirements, particularly during safety evaluations and inspections.

The records generated during these activities must be organized and maintained effectively to support audits and inspections as required by regulatory authorities.


Question: Transportation impacts for nuclear power plants are assessed based on workforce changes and local infrastructure capabilities.

The NRC found that transportation impacts depend on the anticipated change in workforce, the capacity of the local road network, traffic patterns, and the availability of alternate commuting routes to and from the plant.

In its reviews, NRC staff identified that it is unlikely that additional workers will be needed for continued operations of a nuclear power plant.


Question: What does gamma spectrum analysis entail in nuclear safety?

Gamma spectrum analysis is performed using a 500 ml Marinelli detector to measure the radioactive isotopes present in a sample taken from Dairyland_NPP.

The sample size was measured at 7.410E+02 grams and the acquisition started on 5/28/2019 at 12:05:06 PM.


Question: The toughness characteristics of the material are crucial for understanding how they respond under stress conditions, such as during earthquakes or pipe ruptures.

The toughness properties of this material have been obtained by Dr. Stout.

There may be five operating basis earthquakes, however, the critical one is that where we combine the one design basis earthquake with pipe rupture with the deadload.


Question: What are the qualifications required for personnel performing independent quality verification in nuclear facilities?

In lieu of being certified as Level I, II, or III in accordance with NQA-1-1994, personnel performing independent quality verification inspections, examinations, measurements, or tests of material, products, or activities will be required to possess qualifications equal to or better than those required for performing the task being verified; and the verification is within the skills of these personnel and/or is addressed by procedures.

High school diploma for those independent review personnel who are required to review problems in administrative control and quality assurance practices, training, and emergency plans and related procedures and equipment.


Question: fuel assembly reconstitution process

Fuel assembly reconstitution is performed as needed. The methodology for evaluating fuel assembly reconstitution is provided in Section 16 of Reference 1.

The proposed changes to the fuel cycle will enhance the efficiency of energy production and reduce waste generation.


Question: What is the importance of the TMD code in analyzing fluid dynamics in nuclear reactor systems?

As part of the review of the TMD code, additional effects are considered. Changes to the analytical model required for these studies are described in Reference 1.

The performance of the TMD code was verified against the 1/24 scale air tests and the 1968 Waltz Mill tests.


Question: average enrichment of uranium-235

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

A robust safety culture is essential for ensuring operational safety and reliability in nuclear facilities.


Question: Strontium-90 analysis

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


Question: What is the role of the Chemical and Volume Control System in aging management?

The Chemical and Volume Control System is designed to manage the chemical and volume conditions necessary for reactor operation, which includes the integrity of components exposed to various environments, such as sodium hydroxide and treated borated water.

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


Question: The Virginia Electric and Power Company has provided the Updated Final Safety Analysis Report (UFSAR) to comply with regulatory requirements.

Pursuant to 10 CFR 50.71(e), the Virginia Electric and Power Company hereby submits the Updated Final Safety Analysis Report (UFSAR) for the North Anna Power Station Units No. 1 and 2.

The enclosed UFSAR is a completely new document without the pages from the original FSAR.


Question: RCS subcooling criteria

Checks if SI Termination Criteria is MET:

RCS Subcooling >40°F

Secondary Heat Sink (SG >14% or AFW Flow >240x10³)

RCS Pressure rising or stable - NO

Pressurizer Level >21%

US/RO | Announces transition to 1-OHP-4023-E-1, Loss of Reactor or Secondary Coolant and directs operator actions.


Question: What are operability determinations in nuclear regulatory context?

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:


Question: reactor coolant pump shutdown seals

Each RCP in both units have the Westinghouse Low Leakage Shutdown RCP seals (GEN III SDS seals) installed. All hazard models (internal events including internal flooding, fire, and seismic) credit these seals.

The PRA model used for the shutdown seals performance is PWROG-14001-P-A, 'PRA Model for the Generation III Westinghouse Shutdown Seal,' Revision 1 (Reference 7).


Question: What activities are involved in preventing fouling in containment spray heat exchangers?

The purpose of the Performance Testing Activities – Containment Spray Heat Exchangers is to manage fouling of stainless steel and titanium heat exchanger tubes that are exposed to raw water.

The purpose of the Heat Exchanger Preventive Maintenance Activities – Control Area Chilled Water is to manage fouling and loss of material of parts of the control room area chillers exposed to raw water.


Question: The NRC oversees the transfer of licenses and the implications for corporate ownership structures in facilities like Metropolis.

The first step includes both a direct and indirect transfer of control. In the first step: 1. The direct transfer of control will occur when Honeywell transfers its materials license for the Metropolis Facility (and ownership of the relevant assets and operations related to Metropolis) to NewCo Corp., and NewCo Corp. will become the new licensee.

As a result of the indirect transfer described in clause 2 of the first step, NewCo Corp. will become an indirect intermediate corporate parent of Honeywell’s indirect 50 percent ownership interest in the ConverDyn joint venture.


Question: gamma measurement in nuclear regulatory filings

**Sampling Location:** Pickard Hall
**Boring Number:** 16

**Date:** 8-10-20
**Time:** 10:30


Question: What is the activity of radioactive isotopes measured in?

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

Manufacturer name: TN TECHNOLOGIES, INC.


Question: nuclear safety review process

The Nuclear Safety Review is a critical step in the regulatory process by which the safety of a proposed nuclear facility is assessed before authorization can be granted by the NRC.

The cooling system in the reactor must operate within specified parameters to ensure optimal thermal performance and prevent overheating.


Question: DNPS site boundary definition

The DNPS Site Boundary is delineated by a red box, which indicates the extent of the designated area for the facility operations.

The report also discusses the operational status of the reactor and its compliance with regulatory standards.


Question: What is the meaning of characteristic recurrence time in seismic analysis?

Characteristic Recurrence Time (years) refers to the estimated time interval in which an earthquake of a certain magnitude is expected to occur at a specific location, which is crucial for seismic risk assessments in nuclear facility licensing.

The table lists various characteristic magnitudes along with their corresponding displacement, slip rate, and recurrence time, essential for understanding seismic activity.


Question: What are the safety systems required for nuclear reactors?

Nuclear reactor safety systems include multiple redundant systems that can detect and control abnormal conditions to ensure safe operation and prevent accidents.

Windows is either a registered trademark or a trademark of Microsoft Corporation in the United States and/or other countries.


Question: The NRC mandates that all devices containing radioactive isotopes must be registered and accounted for in compliance with safety regulations.

Isotope (e.g. AM241) | Activity (e.g. 1005) | Unit (e.g. mCi) | AM241 | 150 | mCi

Manufacturer Name: NDC INFRARED ENGINEERING, INC.


Question: The analysis of the gamma spectrum using a 500ml Marinelli geometry allows for precise identification of spectral peaks.

The gamma spectrum analysis was conducted using a 500ml Marinelli geometry to ensure accurate measurement and identification of the radionuclides present in the sample.

The sample size was determined to be 408.4 grams, highlighting the importance of precise measurement in nuclear evaluations.


Question: failure probability during reactor operation

The Failure Probability (demand) for the system components is crucial for evaluating the reliability of the reactor's safety systems and understanding operational risks.

The nominal pipe diameter for all listed weld tags is consistently recorded at 12 inches, providing a standard reference for construction and maintenance activities.


Question: What is gamma measurement in environmental sampling?

Gamma measurement refers to the detection of gamma radiation levels in a given interval, as exemplified by the sample collection logs for various depths below the slab in Pickard Hall.

The boring number for the sample collected in Pickard Hall Basement Room 1 is P6, which indicates the specific drilling location for the sampling.


Question: The impact assessments for continued operations and building activities at Palisades highlight that current water quality standards will be upheld throughout the projected changes.

Accounting for midcentury projections, climatological changes would not result in significant impacts from building activities and continued operations of Palisades.

Midcentury projections may result in noticeable changes in runoff patterns but the impacts from plant activities would continue to be permitted, managed, and monitored under applicable regulations that are protective of the State's water quality standards.


Question: What regulatory framework governs the storage of spent nuclear fuel in the U.S.?

The NRC regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste," govern the storage of spent nuclear fuel (spent fuel) generated at commercial nuclear power reactors licensed by the NRC.

The NRC staff is currently reviewing these updated DFPs submitted by Ameren Missouri.


Question: The Permittee is required to report results for total residual oxidants when using certain treatments for the condenser cooling water.

If samples are collected for total residual oxidants when bromine or bromine and chlorine are used for intermittent treatment of condenser cooling water, results for those samples shall be reported on the DMR.

This permit authorizes the Permittee to treat and dispose of industrial process wastewater in accordance with the provisions of this chapter.


Question: What is the RICT estimate in nuclear regulatory filings?

Table E9-17 summarizes the results of these select RICT calculations.

As expected, given the small sensitivity of the model to the change, the technical specifications are unaffected by this sensitivity.


Question: byproduct material management in nuclear operations

That introduces obviously shielding questions. It raises the complexity of transport and installation at the final place of operation. So, how much could we limit operations associated with this testing in order to minimize the amount of byproduct material that would be introduced?

If you set out these kind of controls, limited power levels, loading with fresh fuel, then what current limitations on operations in Part 53, the proposed Part 53, might be relaxed?


Question: loss of control room air conditioning

Knowledge of the relationship between a loss of control room air conditioning and the following systems or components: (CFR: 41.8 / 41.10 / 45.3)

Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to a loss of control room air conditioning: (CFR: 41.5 / 41.7 / 45.7 / 45.8)


Question: The NRC mandates visual inspections to assess structural integrity at defined intervals.

Inspected intersection and bottom edge of beams at 5 core locations per BWRVIP-183 utilizing a new visual inspection tool and rim area 0° - 90°. No indication of cracking.

No indication of cracking was reported during the inspections of the bottom edge of beams.


Question: What is the archaeological area of potential effect in a nuclear plant project?

TVA considers the archaeological area of potential effect (APE) to be the footprint where ground disturbance would take place (1LI535 is outside of the APE).

In addition, the cold water discharge canal is proposed between the north end of the spoil pile and the existing western perimeter ditch, and approximately a five-acre construction staging area is necessary.


Question: Containment Spray System functionality

The Containment Spray System (CSS) is designed to mitigate the pressure buildup in the containment structure during a postulated loss-of-coolant accident by spraying water to cool the containment atmosphere and condense steam.

The importance of manual operations in emergency situations is critical for ensuring that all personnel are familiar with their roles and responsibilities to maintain operational safety.


Question: What are Emergency Operating Procedures (EOPs) in nuclear regulation?

Supplement 1 to Revision 9 of NUREG-1021 erroneously listed "emergency operating procedures" as an example of the type of information that could be evaluated under the "Emergency Procedures/Plan" Administrative Topic of the operating test.

The NRC regrets any confusion that this error may have caused.


Question: Do we need to ensure that a clean copy of the examination package is submitted according to the regulations?

Moreover, Section D.2.a of ES-403 instructs the grader to make a copy before marking the original, and Section C.1.a of ES-501 instructs the facility licensee to submit the clean copy with the examination package.

Please add note to Form ES-403-1 for the grader to copy the answer sheets. I would also suggest making two copies, NRC and facility to have.


Question: What is the weighted mean activity of a nuclide?

The weighted mean activity for a nuclide is calculated using the activity measurements adjusted for their uncertainties, expressed in picocuries per gram (pCi/gram).

The table lists various nuclides along with their confidence levels and associated comments, but does not delve into specifics regarding regulatory requirements for nuclear materials.


Question: What are the equations governing thermal energy in nuclear reactor systems?

The one-dimensional mass, momentum, and thermal energy equations are summarized as follows:

The regulators ensure that all safety protocols are followed during routine inspections of the facility.


Question: nuclide minimum detectable activity


Question: Licensees must submit safety analysis reports to demonstrate compliance with safety regulations.

The NRC encourages applicants to provide sufficient information in their license applications to ensure that safety and environmental impacts are adequately assessed.

The pressure vessel must be inspected regularly to ensure structural integrity and prevent any potential failures during normal operations.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors in the United States involves multiple steps including the submission of a license application, safety reviews, and public hearings conducted by the Nuclear Regulatory Commission (NRC).

The cooling system is designed to remove excess heat from the reactor core during normal operations and in accident scenarios.


Question: wound assessment procedures

At the time of processing, the specimen was in very poor condition, exhibiting evidence of the early stages of decomposition. Some minor bruising of surface tissue was observed on the torso (refer to illustration and specimen photographs), however no major injuries indicating a cause of death were identified.

Submit completed forms (within 24 hours of observation of fish): by e-mail to IncidentalTake@noaa.gov or by fax (978-281-9394). Questions can be directed to NMFS Protected Resources Division at 978-281-9328.


Question: The organization of personnel involved in radiation protection is crucial for maintaining safety standards.

The Radiation Protection Supervisor oversees the In-Plant Monitoring Team and the Offsite Monitoring Teams to ensure compliance with safety regulations.

The Chemistry Team plays a critical role in analyzing materials and ensuring quality control in operations.


Question: nuclide identification

The identified nuclides include K-40 with a confidence level of 0.96 and an energy of 1460.75 keV, along with PB-212 which has a high confidence of 0.99 and an energy of 17.50 keV.

The flow rate of the coolant system must be maintained at optimal levels to ensure efficient heat removal during operation.


Question: The structures at the switchyard are considered crucial for the operation of safety-related systems due to their roles in compliance with various regulations.

The Switchyard Structures are in scope under 10 CFR 54.4(a)(2) because failure of nonsafety-related portions of the station blackout facility could prevent satisfactory accomplishment of function(s) identified for 10 CFR 54.4(a)(1).

Environmental qualification is a critical aspect to ensure all components can withstand the conditions they will face during their operational lifespan.


Question: What is a Technical Specification in nuclear licensing?

Technical Specifications are a set of requirements included in the operating license that delineate the specific operational parameters and limitations of a nuclear power reactor.

The reactor pressure vessel is designed to withstand high pressure and temperature, ensuring the integrity of the reactor system during operation.


Question: The licensee must ensure compliance with the GSI-191 guidance to maintain adequate safety margins.

Maintain GSI-191 compliance is essential to ensure that the facility's design adequately addresses the potential impacts of debris on safety-related systems.

The table of contents provides a structured overview of the different sections included in the licensing submittal.


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

The application for a nuclear reactor license must include a detailed description of the proposed facility, an assessment of its potential environmental impacts, and compliance with the requirements set forth in the relevant safety regulations.

The safety culture within the organization is essential for promoting effective communication and employee engagement across all levels.


Question: What is the purpose of a regulatory order in nuclear licensing?

The NRC issues regulatory orders to impose specific requirements on licensees to ensure compliance with the applicable regulations and to protect public health and safety.

The letter outlines the basic administrative procedures for submitting documents to the NRC and includes contact information for further inquiries.


Question: What is the process of background subtraction in nuclear measurements?

The background subtraction report provides the original area, uncertainty, ambient background, and the subtracted area along with their uncertainties for each identified peak.

The identified nuclides are recorded in the nuclide library used, which is essential for ensuring accurate measurements.


Question: The NRC requires a facility diagram to be included in applications for new licenses.

This location is not listed in the NRC license. Therefore, provide the information described in page C-23 (Item 9: Facility Diagram) of NUREG-1556, Volume 9, revision 3.

Your letter dated January 25, 2021 (Revised) states that Y-90 microspheres will be used and stored at Kootenai Health, 2003 Lincoln Way, Coeur d’Alene, Idaho 83814.


Question: What are the requirements for contention admissibility in nuclear regulatory proceedings?

The rules require “a clear statement as to the basis for the contentions and the submission of … supporting information and references to specific documents and sources that establish the validity of the contention.”

“Failure to satisfy any of the six pleading requirements renders a contention inadmissible.”


Question: enrichment limits for SNF assembly

Preferential Load/Underload Combination Enrichment Limits: Maximum planar average enrichment limits for the BWR 81-Assembly Basket SNF assembly are specified out in the table.

Locations C, D, F, G, H, I in Figure B2-5 are pertinent to identifying certain assembly configurations.


Question: The Office of Nuclear Material Safety and Safeguards oversees the funding for decommissioning to ensure sites are safely managed after contamination.

He's in the Office of Nuclear Material Safety and Safeguard, and he's going to talk to us about decommissioning funding assurance considering existing radiological contamination at some of these brownfield sites.

And we're big fans of preemptive mitigation, especially when dealing with volatile chemicals, say, at a former dry cleaner, attacking source or installing vapor intrusion mitigation tools.


Question: What does it mean for a proposed change to present no significant hazards according to 10 CFR 50.92?

The Dominion Energy Virginia evaluation against each of the criteria in 10 CFR 50.92 follows, determining that the proposed amendment presents no significant hazards.

The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a change in the method of plant operation, or new operator actions.


Question: What are Neutron Source Assemblies in nuclear fuel storage?

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

Fuel assemblies may contain any number of unirradiated (i.e., not inserted in-core) nonfuel solid filler fuel replacement rods.


Question: cathodic protection in nuclear facilities

Discuss cathodic protection, if any, and the power needed to support cathodic protection.

Since the alternative decommissioning schedule exemption request is tied to the pending license renewal applications for Peach Bottom Units 2 and 3 describe the limitations on the time frame that should be considered, if granted.


Question: Containment Purge and Ventilation System

The Containment Purge and Ventilation Systems provide adequate capacity to assure that purge temperatures are maintained in the various portions of the Containment under operating and shutdown conditions in all types of weather.

In an accident, the containment purge and exhaust isolation valves are assumed to be closed (Modes 1-4).


Question: An evaluation of thermocouple measurements during the Zr2K test raises questions about their accuracy regarding cladding temperatures.

One of the more difficult aspects of evaluation of Zr2K test results is associated with the fundamental data for the tests, the recorded thermocouple…responses. GE has been very *liberal* with their accreditation of observed [thermocouple] responses as erratic. However, several proffered examples of erratic response seem to show well-defined inter-rod correlations.

It should be noted, however, that the recorded temperatures of rod 16, which developed the first electrical anomaly after the official start of the test, were almost identical to those of rod 24, which was given credit for the maximum temperature measurement.


Question: What is involved in measuring dose rates for storage casks?

Measure dose rates at approximate 70-inch diameter at four points approximately on 90-degree axes.

Measure dose rates at four target points (approximately 0, 90, 180 & 270 degrees) on the mid-plane.


Question: The estimated social costs associated with radiation risks from the Bellefonte facility are considered minimal.

The radiation risk due to transportation of fuel and wastes within and beyond the 50-mile radius is about 14 man-rem per year.

The location of 170 new families in the Scottsboro area for maintenance and operation of the Bellefonte Nuclear Plant will necessitate some expansion in the provision of public goods and services by the Scottsboro municipal government and the city and Jackson County school districts.


Question: The ADVANCE Act involves assessing how various regulatory components interact within the nuclear licensing framework.

The ADVANCE Act aims to create a streamlined process for regulatory filings related to nuclear safety and operational efficiency.

The meeting included gratitude for participants and plans for future collaboration on unrelated topics such as general communication.


Question: What is the role of postulated initiating events in nuclear reactor safety?

The white paper "IMSR® Postulated Initiating Events" discusses various scenarios that could potentially lead to challenges in reactor safety and operation, providing a framework for understanding these events within the broader safety analysis context.

The Nuclear Regulatory Commission is responsible for regulating commercial nuclear power plants and other uses of nuclear materials.


Question: What is a DFC designated location in nuclear fuel storage?

DFC designated locations may contain a loaded DFC or a PWR UNDAMAGED SNF ASSEMBLY.

The schematic shows various storage locations for safety analysis.


Question: What is the importance of the Internal Events/Internal Flooding PRA model in nuclear safety assessments?

The Internal Events/Internal Flooding (IE/IF) PRA model was subject to a self-assessment and a full peer review conducted in August 2009 against ASME/ANS RA-S-2009 and RG 1.200, Revision 2.

In September/October of 2019, an F&O Closure Review was conducted for ANO-1.


Question: What are the considerations for construction activities near transmission lines in nuclear projects?

Any results available from the transmission line studies (see Section 6), and any revisions in the TVA plans for construction and maintenance on the lines covered under Steps 2 and 3, shall be reviewed with the AEC staff prior to initiation of any construction activities on these lines.

Construction personnel should be encouraged not to enter the non-construction portions of the site, e.g., the overbank area and adjacent forested hills.


Question: flood magnitude and frequency analysis

Table 2-4 presents the calculated peak flood magnitudes and frequencies at Rowell Gauging Station and at the Clinton Lake Dam, showing values for different recurrence intervals, expressed in cubic feet per second (cfs).

The results of the environmental assessments indicated that the overall impact on surrounding wildlife habitats would be minimal, with no significant barriers to migration identified.


Question: isotope activity measurement

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

NRC Device Key: 595424 (Internal Control Number)


Question: Circulating Water Intake Structure

Descriptive location (be specific): Removed from Circulating Water Intake Structure (CWIS) intake trash racks, intake bay 12B, during routine trash rack cleaning at the Salem Generating Station, PSEG Nuclear LLC.

A deceased Atlantic sturgeon was collected from outside the Salem Circulating Water Intake Structure during routine trash rack cleaning on 01/30/25, at approximately 08:45 AM.


Question: nuclide yield

The yield of PB-214 is 19.20%, which indicates the percentage of the nuclide detected in relation to other components in the sample.

The half-life too short to be able to perform the decay correction makes certain measurements impractical, as it limits the time available for observation.


Question: groundwater monitoring program

The Offsite Dose Calculation Manual (ODCM) for the existing units establishes the requirements for the Radiological Environmental Monitoring Program and specifies the conduct of the program.

Seven open-tube standpipe piezometers (P-10, P-14, and P-18 through 22) are currently monitored in accordance with the test procedure, along with two additional piezometers (P-23 and P-24) that are monitored for information purposes only Figure 2.4-16.


Question: The capability of assessing various factors related to ADS Stage 1–3 functions is crucial for operational safety.

Ability to evaluate the following parameters and/or conditions as they apply to ADS Stage 1–3 Actuation: (CFR: 41.7 / 43.5 / 45.6)

Fuel pool cask loading pit level is critical for ensuring the safe handling and storage of spent nuclear fuel.


Question: The design and layout of systems must ensure that safe shutdown capability is maintained even in the event of a fire, as per the regulatory standards.

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

Normal unit shutdown would be considered available as it is unlikely a fire would damage both trains.


Question: What are the requirements for declaring equipment inoperable under nuclear regulatory guidelines?

Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.


Question: nrc device transfer documentation

The company is expected to submit periodic reports to comply with the regulatory requirements.


Question: What are ITAAC notifications in nuclear licensing?

ITAAC notifications, specifically the UIN and ICN types, are essential for documenting the compliance of construction with the design and regulatory requirements set forth by the NRC.


Question: Changes to regulations can have significant implications for both current license holders and new applicants in the nuclear industry.

In my view, based on regulatory ambiguity, it is appropriate to read the language of Part 51, the regulatory history, regulatory analysis, and agency guidance holistically.

I have serious concerns about the message this action sends to the public, applicants, licensees, and other stakeholders.


Question: What is required before re-enabling an annunciator in the safety system?

Ensure there is a Condition Report (CR) to correct the problem if Annunciator Disablement is not being tracked by an existing WO, CR, or site procedure, for the equipment deficiency and/or a CR for the Annunciator Disablement.

When all required approvals for disabling alarm are obtained, perform the following and complete Form OPDP-4-1, Item 9: Review disabling steps.


Question: What is a calibration unit in nuclear safety?

Atomergie Chemicals Corporation markets the Rad-Lab 1023 Calibration Unit, which is used for ensuring the accuracy of radiation measurement instruments.

The vendor list includes various item numbers from different companies, each detailing various product specifications.


Question: What is the significance of aging effects for structures and components in nuclear licensing?

Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

During its audit, the staff noted that Devtar 247 is a high build epoxy mastic coating.


Question: technical specifications requiring justification

Table E1-3 lists the TSTF-505 Revision 2 Table 1 Technical Specifications (TS) that require additional justification along with a description of how the additional justification is provided in the LAR.

The licensee must also ensure that the components of the reactor coolant system are properly classified and maintained to avoid degradation over time.


Question: tritium distillation process

Product: LSC, Tritium Distillation, Solid

The following samples were analyzed using the above methods and analytical procedure(s).


Question: Regulatory submissions for nuclear reactors involve a careful analysis of the safety and environmental impacts associated with operation.

A safety analysis report is required by regulations to demonstrate that the proposed design meets the necessary safety margins for reactor operations.

The reactor core consists of fuel assemblies that must be monitored for any signs of degradation or overheating.


Question: What does minimum detectable concentration (MDC) mean in nuclear analysis?

MDC refers to Minimum Detectable Concentration, which is the smallest amount of a substance that can be reliably measured under stated conditions, usually defined in analytical methods for nuclear materials.

The LaCrosseSolutions company is located in Genoa, Wisconsin, and specializes in various analytical services.


Question: The calibration and validation database for low stress creep behavior specifically includes data for PWR rod designs.

The M5 calibration database used to determine the model uncertainties is made of PWR rods (see Table 30-1, M5 cladding). This database is a sub-part of the data used in Figure 5-11 in Revision 0 of the methodology document (Reference 1) which uses creep data.

This database is used in Figure 10-47 in Revision 2 of the V&V document (Reference 4).


Question: The National Wetlands Inventory provides essential information regarding wetland status and characteristics within the U.S.

Wetlands are defined as areas inundated or saturated by surface or groundwater at a frequency and duration sufficient to support (and that under normal circumstances do support) a prevalence of plants typically adapted for life in saturated soil conditions.

Physical features (e.g., length, width, route) of in-scope transmission lines are described in Section 2.2.5.1. All in-scope transmission lines are located completely within the DNPS site, as shown in Figure 2.2-2.


Question: What is the significance of the Action Level in radiation monitoring?

The Action Level is defined as the predetermined threshold of 33,706 counts per minute (cpm) that, if exceeded, triggers specific regulatory actions or investigations.

The logged readings for each scan area were recorded, indicating the level of radioactivity detected during monitoring.


Question: effluent release limits

For the purpose of evaluating compliance with 10 CFR 20, the boundary of the restricted area for the EGC ESP Site is defined as the EGC ESP Site EAB. This boundary distance is established as a site characteristic (see Table 1.4-1, Section 9.4.4).

The CPS cooling water intake is about 3 mi northeast of the confluence of Salt Creek and the North Fork of Salt Creek.


Question: The NRC evaluates potential dose contributions in various locations to ensure they remain within regulatory limits.

The evaluation includes assessing dose calculations at designated locations such as the EAB and other specified areas.

The calculation pathways listed provide varying indices and values that represent different assessments.


Question: loss of component cooling water

RCPs are required to be stopped due to loss of cooling flow to the RCP Oil Coolers.

Which ONE of the following completes the statements in accordance with 1-AOI-15, Loss of Component Cooling Water (CCS)?


Question: HEC-RAS model application in nuclear hazard reevaluation

TVA is using the HEC-RAS model for flood analyses performed in response to the post-Fukushima request for information letters under 10CFR50.54(f) because the HEC-RAS model is a well-documented and supported industry standard program; it allows the entire watershed to be modeled in a single continuous simulation.

In steady flow mode, HEC-RAS explicitly solves the energy equation with an iterative procedure called the standard step method.


Question: What are Generic Safety Issues (GSIs) in the context of nuclear licensing?

In accordance with the guidance in NEI 95-10 and Appendix A.3 of NUREG-1800 (Reference 1.7.25), Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants, review of NRC Generic Safety Issues (GSIs) as part of the LR process is required to satisfy a finding per 10 CFR 54.29.

DCPP IPA as reflected in Section 3.1, Aging Management of Reactor Vessel, Internals, and Reactor Coolant System, and in Appendix B, Aging Management Programs.


Question: What is the role of the Radiation Protection Coordinator in emergency response?

The TSC Radiation Protection Coordinator relieves the Shift Manager of the Supervision of Radiation Protection Staff and Site Radiation Protection function.

NextEra has added a drill/exercise to the 8-year cycle to validate that an operating crew can perform an EP drill scenario without any assistance from the augmenting ERO for 90 minutes as a verification that the shift can perform their function with no assistance.


Question: initial ground support system definition

The BWRX-300 design does not rely on the resistance provided by the initial ground support system or the material used to fill the annular gap between the below-grade RB shaft exterior wall and the excavated soil and rock.

Erosion control measures will be required during excavation through the soil overburden and weathered rock.


Question: tensioner pressure and stud stress relationship

K = XS*Z2/XP ! Relate tensioner pressure to stud stress

/COM, Check for flange contact links which came out of compression


Question: What is the purpose of the Containment Spray System in nuclear reactors?

The primary purpose of the spray system is to spray cool water into the Containment atmosphere, when appropriate, in the event of a loss-of-coolant accident (LOCA) and thereby assure the Containment pressure cannot exceed containment design pressure of 15 psig.

Detection of leakage from the Containment Spray System involves the use of sump level instrumentation and floor drain tank level instrumentation, and Waste Evaporator Feed Tank level instrumentation.


Question: human health impacts of NGCC components

Impacts on human health from construction and operation of the NGCC component are presented in Section 7.2.3.1.10. They would be SMALL for construction and SMALL to MODERATE for operations.

The construction of the solar and wind installations would create sanitary and industrial waste in smaller than or similar quantities to the NGCC plant.


Question: Part 70 License Application for enrichment facility

Orano will be submitting a Part 70 License Application for an Enrichment Facility.

Orano requested a pre-application meeting related to the Environmental Report submittal for Project IKE in Oak Ridge Tennessee.


Question: What are the critical heat flux correlations in nuclear reactor safety?

The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations are essential for evaluating the thermal hydraulic performance of nuclear reactors under various operational conditions, ensuring safe and efficient reactor operation.

The Dominion Energy Fleet Report includes numerous appendices that outline various methodologies and evaluations related to reactor operations, but they do not specifically address heat flux correlations.


Question: What is the purpose of the CIM in nuclear safety systems?

The CIM receives commands and transmits status on three ports to/from interfacing systems.

The equipment was tested under various environmental conditions to ensure reliability and performance.


Question: nuclear safety analysis methodologies

The methodology used for safety analysis is based on the latest regulatory guidelines and integrates probabilistic safety assessment techniques to evaluate potential risks associated with reactor operations.

The oxygen level in the containment building can have significant impacts on the efficiency of combustion processes.


Question: The system is designed to provide means for obtaining samples to facilitate analysis of chemistry and radioactivity levels in the reactor.

The Nuclear Sampling System is designed to provide means to obtain samples from various systems in each of the two units for chemistry and radioactivity analysis.

A sample room is provided for each unit.


Question: measured ambient noise levels

The measured daytime(a) ambient noise level; event 1 (Leq, dBA) was 45.3, while the measured nighttime(b) ambient noise level; event 3 (Leq, dBA) was 45.3.

Daytime events occur between 7:00 a.m. and 10:00 p.m.


Question: What is a nuclide identification report?

The identification report lists the nuclides detected during analysis, including details on energy levels and associated areas, which are crucial for regulatory compliance and safety assessments in nuclear operations.

The report also provides detailed statistics on peak energy levels and uncertainties related to measurements, which are essential for understanding overall detector performance.


Question: What is the ALARA principle in nuclear safety?

ALARA, which stands for As Low As Reasonably Achievable, is a safety principle designed to minimize radiation exposure and associated risk to workers and the public by implementing strategies that are economically and socially acceptable.

The central portion of the figure comes directly from Regulatory Guide 1.174, serving to visually represent how the different portions of the Part 53 proposed rule relate to the principles of integrated risk-informed decision making.


Question: What is fuel rod ballooning in reactor safety analysis?

The possibility of ballooning during an AOO transient or accident needs to be examined.

Load reinsertability is a third criterion that must be satisfied.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves several key steps as mandated by regulatory bodies such as the NRC, including the submission of safety analysis reports and environmental impact assessments.

The cooling system is designed to remove heat generated from the reactor core and ensure efficient operation under various conditions.


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

Regulatory filings in nuclear licensing serve to ensure that all safety and compliance standards are met before a reactor can operate, allowing for the review of technical specifications and assessments critical to public safety.

The construction of a new reactor typically involves multiple phases, including design, procurement, and assembly, which are overseen by professional engineers and project managers.


Question: What is the process for Licensees to respond to an Errata?

Where good cause is shown, consideration will be given to extending the time to request a hearing. A request for extension of time in which to submit an answer or request a hearing must be made in writing to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and include a statement of good cause for the extension.

This Errata does not in any way alter any substantive provision of the Order or requirements imposed thereby on any Licensee.


Question: nuclear device registration

Enter the company name and the street address for the physical location of use for your device(s). For portable devices, specify the primary storage location. Do not use P.O. Boxes.

For NRC Use Only
(Do not write here)
**Category:**
**Packet Receipt Date (MMDDYYYY):**
**Accession Number:**


Question: The NRC allows for a hearing request to be submitted within 60 days following the publication of a notice related to regulatory actions.

Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action.

If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the *Federal Register* a notice of issuance.


Question: What is the Practical Quantitation Limit in analytical chemistry?

PQL: Practical Quantitation Limit. Synonymous with the EPA term 'minimum level'.

The analysis exceeded method hold time. pH is a field test with an immediate hold time.


Question: systematic sampling soil analysis

The systematic locations of the soil samples were selected using Visual Sample Plan (VSP), in accordance with LC-FS-PR-002.

In total, twenty-four (24) soil samples were collected for the FSS of survey unit L1-SUB-TDS.


Question: What are the minimum allowable core reflood rates during a loss-of-coolant accident?

Specifically, the petitioner requests that the NRC amend its regulations based on data from multi-rod (assembly) severe fuel damage experiments and promulgate a regulation that will stipulate minimum allowable core reflood rates in the event of a loss-of-coolant accident.

The attachment to this letter provides comments from the Nuclear Energy Institute (NEI) on behalf of the nuclear energy industry on the Petition for Rulemaking (PRM-50-93), in response to the Federal Register notice of January 25, 2010.


Question: What is the minimum cooling time for assemblies with a burnup between 50 and 51 GWd/MTU?

50 < Assembly Average Burnup ≤ 51 GWd/MTU Minimum Cooling Time (years)

Certificate of Compliance No. 1031 Amendment No. 6, Revision No. 2


Question: high activity in reactor coolant

The INDIVIDUAL alarming Area Radiation Monitors are able to be read in the MCR. RCS Activity level is able to be DIRECTLY monitored in the MCR.


Question: What is the average enrichment of Uranium-235 in PWR fuel assemblies?

The following is a loading table which lists various parameters for different reactor fuel assemblies and conditions.


Question: What is the function of the Intermediate Range High Flux Rod Stop in a nuclear reactor?

When reactor power goes above P-10, Nuclear At Power, C-1, the Intermediate Range High Flux Rod Stop, is BLOCKED ...


Question: The NRC mandates a thorough examination of groundwater withdrawal limits for licensees.

Groundwater withdrawal limits are regulated to ensure that environmental impacts are mitigated and that the health and safety of the public and environment are protected.

The facility's operations involve significant use of mechanical systems that require continuous monitoring and maintenance.


Question: transient and safety analysis methodologies

The Transient and Safety Analysis Methodologies Framework outlines the systematic approach used for evaluating the safety impacts of various operational transients and accidents within the Xe-100 reactor design.

The calculations involved in thermal hydraulic analyses require intricate knowledge of fluid dynamics and heat transfer principles that govern reactor operation.


Question: The NRC mandates that licensees assess the impact of environmental factors on structural integrity.

Loss of material due to general, pitting, crevice corrosion

Not Applicable. This Item Number is not applicable to the BFN Mark I steel containment. See Subsection 3.5.2.2.1.1.


Question: What is a loss of coolant accident in nuclear reactors?

To date, there has only been one loss of coolant accident of significant consequence at a nuclear power reactor licensed by the NRC – the March 1979 accident at Three Mile Island Unit 2.

UCS recognizes and appreciates the unselfish commitment to public health and safety this petition represents.


Question: The NRC evaluates aging effects on safety feature components to ensure integrity over time.

Not Applicable. There are no titanium (ASTM Grades 1, 2, 7, 9, 11, or 12) heat exchanger components other than tubes, piping, piping components exposed to treated water in the Engineered Safety Features systems.

The NRC requires submittals to include comprehensive engineering analyses to support regulatory decisions.


Question: risk-informed approach in nuclear licensing

In its LAR, Dominion Energy Virginia proposed to apply a risk-informed approach to demonstrate that the Fuel Handling Trolley Support Structure (FHTSS), as designed, meets the intent of a tornado resistant structure under the current Surry licensing basis for a 360 miles per hour maximum tornado wind speed.

The NRC issued construction permits for Surry, Units 1 and 2, before May 21, 1971.


Question: Strontium-90 reporting levels

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


Question: minimum required prestress force for dome hoop tendons

The minimum required prestress force (MRV) for the dome hoop tendons is 1119 kips.

The Dome Hoop Tendon Liftoff Data table includes various dates and force measurements pertaining to the Wolf Creek Generating Station.


Question: nuclear regulatory filings process

Pursuant to 10 C.F.R. § 2.305, I certify that, on this date, a copy of the foregoing 'Applicants’ Answer Opposing Request for Hearing, Petition to Intervene, and Petition for Waiver Submitted by Beyond Nuclear, Sierra Club, and Alliance for Progressive Virginia' was served upon the Electronic Information Exchange (the NRC’s E-Filing System), in the above-captioned docket.

Signed (electronically) by Ryan K. Lighty
Ryan K. Lighty, Esq.
MORGAN, LEWIS & BOCKIUS LLP
1111 Pennsylvania Avenue, N.W.
Washington, D.C. 20004
(202) 739-5274
ryan.lighty@morganlewis.com


Question: nuclide identification

Nuclide identified during the nuclide identification

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


Question: The Nuclear Regulatory Commission oversees the licensing process for power plants across the nation.

The NRC requires licensees to analyze the potential impact of their operations on the surrounding environment and public health.

The total population of Ludlow in 2000 was 324, which reflects a small community size with limited resources.


Question: The NRC has mandated that all licensees must enhance their security measures to protect against radiological threats, particularly in relation to the storage of spent fuel.

Specific physical security requirements are contained in 10 CFR 73.51 or 73.55, as applicable.

The Commission has also communicated with other Federal, State, and local government agencies and industry representatives to discuss and evaluate the current threat environment in order to assess the adequacy of security measures at licensed facilities.


Question: What is the average hourly speed of vehicles recorded?

The average hourly speed of vehicles recorded over the years shows fluctuations, with an average of 7.0 mph noted for the year.

The prevailing direction of winds recorded shows a dominant direction of SW throughout most months, with variations noted in October and September.


Question: digital modernization in nuclear plants

Digital modernization at Florida Power & Light Co. Turkey Point Unit 3&4 involves implementing advanced technology solutions to enhance operational efficiency, improve safety systems, and support regulatory compliance.

The financial implications of the operational overhaul must be carefully assessed to ensure long-term profitability and sustainability of the energy production.


Question: What is the role of the borated water in the refueling process?

The refueling water storage tank provides a source of borated water for use during refueling and a loss-of-coolant accident.

Two small submersible sump pumps are located in the pipe trench to remove rainwater which collects in the trench and prevent damage to heat tracing due to submersion in water.


Question: What is chain of custody in analytical testing?

Chain of custody refers to the sequential documentation of the handling and possession of samples from the point of collection until testing, ensuring that the samples are properly tracked and their integrity is maintained.

The client name listed in this submittal is La Crosse Solutions, indicating the organization requesting the analytical testing services.


Question: What is a CHANNEL CHECK in nuclear instrumentation?

A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious.

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


Question: The NRC mandates regular gamma radiation measurements at various depth intervals during waste management assessments.

Gamma measurement is a critical process in assessing the levels of radiation present in collected samples, as illustrated by the recorded measurements taken at various depths in the sample collection log.

The locations listed for sampling include specific basement rooms, each assigned a unique boring number, which is essential for cataloging and tracking contamination assessments.


Question: The operator must enter 0-AOI-29, which pertains to handling issues related to dropped or damaged fuel assemblies in situations where refueling operations are ongoing.

The NUSO will enter 0-AOI-29, Dropped or Damaged Fuel Or Refueling Cavity Seal Failure, which will direct the crew to close 2-ISV-78-600, FUEL TRANSFER TUBE ISOL.

Plausible because 1-ARI-124-130, MISC, for annunciator SFP LEVEL HI/LO [128-A], directs the NUSO to enter either 0-AOI-29, Dropped or Damaged Fuel Or Refueling Cavity Seal Failure, or 0-AOI-45, Loss of Spent Fuel Pool Level Or Cooling, but the Purpose of 0-AOI-45 directs the NUSO to use 0-AOI-29 due to refueling operations being in progress.


Question: reactor pressure vessel embrittlement

PSR argues that the “heart of Contention 2 is that [Point Beach] has inadequate physical samples (metal capsules/coupons from inside the reactor vessels of the two units) to enable metallurgical testing to understand the extent of embrittlement of the vessels and related components in order to assure the reactors will last” through the period of extended operation.

The Board also found Contention 2 inadmissible because it challenged NRC regulations and the current licensing basis for Point Beach and thus raised issues outside the scope of the proceeding.


Question: The company has specified that certain radiochemical parameters are not available for certification under NELAC guidelines.

The following parameters are not offered for certification or are not covered by NELAC certificate #ACZ.

The Rio Algom Mining Company is collaborating with ACZ Laboratories to advance their project.


Question: Nuclear facilities must adhere to strict licensing requirements as part of the regulatory process to ensure safety and compliance.

The licensing process for a nuclear facility involves multiple steps, including the submission of detailed safety analyses and compliance reports to the NRC for review.

The plant operators regularly perform maintenance checks to ensure all equipment is operating correctly and efficiently.


Question: What is the significance of reactor coolant letdown flow in nuclear reactor safety?

Reactor coolant letdown flow (purification) is measured at 2.23E+04 lb/hr, which is essential for maintaining chemical composition and radiological integrity within the reactor system.

The secondary coolant concentrations are based on a primary-to-secondary leakage of 100 lb/day.


Question: What is the concentration of Nitrogen-16 as it exits the core?

Leaving core, the Nitrogen-16 Activity is 189 µCi/g.

The intensity of the energy emitted at 6.13 MeV/gamma is 60.0%.


Question: The process of analyzing the spectrum involves identifying peaks and assessing their significance.

Peak Area Range (in channels): 100 - 4096


Question: What are the maximum water surface elevations during tropical storm conditions?

PMH surge maximum water surface elevation occurs at 41.1 ft NAVD, considering the metrics provided in the calculations.

Calculations are performed using more significant figures than shown in the Table, and the result is rounded to tenths of a ft. If intermediate results were rounded prior to addition the result would not be correct.


Question: What are equipment control procedures in a nuclear plant?

These procedures provide instructions for releasing plant equipment or systems for maintenance, testing, or inspection; they establish the shift supervisor as the responsible authority for issuing and releasing clearances for equipment to be taken out of service within his assigned unit.

These procedures ensure that a comprehensive exchange of information takes place between the oncoming and offgoing shift personnel so that the oncoming shift is aware of critical plant status information and system availability prior to assuming duty.


Question: metal matrix composites in nuclear storage

Metal matrix composites (MMCs), as described in Chapter 10 of the SAR, are limited to 0.5% open porosity, as determined by qualification testing.

Each lift of a MAGNASTOR® TSC, transfer cask, or overpack must be made in accordance with the heavy loads requirements and procedures of the licensed facility at which the lift is made.


Question: The NRC mandates that any piping component designated as Class 1 must comply with rigorous safety standards to ensure its integrity under operating conditions.

Piping (Class 1)                              | Pressure Boundary                              |

The effectiveness of the reactor coolant pumps is critical for maintaining operational efficiency and safety standards.


Question: pellet to clad gap

A radial gap is provided between the fuel pellet and cladding to allow for pellet growth due to thermal expansion and irradiation swelling. This will prevent excessive cladding strain that can be caused by a pellet-clad interaction.

The primary purpose of the pellet to clad gap between the UO₂ fuel pellet and Zircaloy cladding is to accommodate differential axial expansion of the fuel pellet and the cladding.


Question: What is the capacity of steam generator safety valves?

The steam generator safety valve capacity is sized to remove the steam flow at the engineered safety features rating (105 percent of steam flow at rated power) from the steam generator, without exceeding 110 percent of the steam system design pressure.

A loss-of-external-load event results in a nuclear steam supply system transient that is bounded by the turbine trip event analyzed in subsection 15.2.3.


Question: What is the significance of power histories in the safety analyses of fuel rods?

Within thermal-mechanical analysis of LWR fuel, the power history of the rod being modeled has the largest impact on the results of concern from a safety perspective.

The specific thermal-mechanical analyses that should be performed are listed below.


Question: What are the total observations for wind direction E-NE?


Question: The evaluation of safety systems is an essential part of the licensing process.

A comprehensive assessment of the digital modernization at Turkey Point Units 3 and 4 was conducted to ensure compliance with current safety standards and regulations.

The project involved extensive community outreach efforts to inform local residents about the planned upgrades and changes.


Question: What is a Nuclide Minimum Detectable Activity (MDA)?

The Nuclide MDA (Minimum Detectable Activity) is a measure used to determine the lowest activity level of a nuclide that can be reliably distinguished from background noise during radiological measurements.

The peak size for energy measurements recorded during the analysis can vary based on the properties of the detected radiation.


Question: What is the demographic profile of the population in the Oconee region?

Total Population: 78,607 (Oconee County), 131,404 (Pickens County), Region of Influence: 210,011.

The table shows percentages of population changes in both counties over multiple decades, reflecting growth trends.


Question: nuclear criticality safety

Nuclear criticality safety is the prevention of an uncontrolled nuclear chain reaction in materials capable of sustaining such reactions, which includes ensuring that sufficient spacing is maintained between fissile materials.

Effluent Control and Environmental Protection involves managing the release of radioactive materials into the environment to comply with regulatory requirements.


Question: The NRC evaluates potential environmental impacts of license renewals for nuclear facilities.

The Generic Environmental Impact Statement for License Renewal of Nuclear Plants provides a comprehensive assessment of the potential environmental impacts associated with license renewal for the differing types of nuclear plants.

Each of these supplements is intended to inform the public and assist the NRC in its decision-making process.


Question: The distribution of gross activity levels is presented in the histogram format, illustrating the frequency of occurrences across various upper end values.

The histogram shows the frequency of gross activity measurements in disintegrations per minute per 100 cubic centimeters (DPM/100CM³).

The report includes various figures and tables that summarize the operational history of the reactor and its compliance with safety protocols.


Question: Containment Pressure failure response

Containment Pressure Channel II has failed high and it has been removed from service in accordance with the applicable procedure.


Question: The NRC ensured that the program for Unit 2 incorporated findings from Unit 1 to prevent issues during operation.

In addition, the chosen tests were selected to encompass the physics test goals listed in the NRC Branch Technical Position DOR-1, *Guidance for Reload Submittals, Draft - Spring, 1978*.

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


Question: What are the fuel performance criteria defined under 50.46a?

Fuel performance criteria. Must have NRC-approved limits that: i. Address cladding degradation phenomena ii. Maintain fuel coolability iii. Avoid explosive concentration of combustible gas iv. Ensure long-term cooling

Evaluate applicability of the TBS to the licensee’s facility.


Question: In areas where work with nuclear material occurs, significant contamination levels require immediate assessment through smear tests to evaluate removable alpha contamination.

If the total alpha survey indicates total contamination greater than 1,000 dpm/100 cm², a smear survey will be performed to determine the removable contamination.

Routine surveys in the CCP and satellite facilities will consist of both a visual inspection for obvious signs of contamination and instrument surveys to determine total alpha contamination.


Question: Part 53 rulemaking process

The Nuclear Energy Institute (NEI) and its members appreciate the Nuclear Regulatory Commission’s (NRC) efforts to establish a technology-inclusive, risk-informed, and performance-based regulatory framework for advanced reactors, commonly referred to as the Part 53 rulemaking.

While there is still much work ahead to achieve a technology-inclusive, risk-informed and performance-based Part 53 rule that is transformational and meets the needs of future new reactor designers, owners and operators, we believe that the NRC’s approach is headed in the right direction.


Question: audit activities in nuclear licensing

The audit period will begin approximately 2 weeks from the date of this audit plan and will last through November 10, 2022.

The NRC’s licensing project manager will inform the licensee of the entrance and exit meeting dates when they are established.


Question: The NRC evaluates the integrity of materials used in safety-related systems to ensure compliance with regulations.


Question: What are the regulations for radioactive waste management in nuclear facilities?

- **IMC 0350**: Oversight of Reactor Facilities in a Shutdown Condition Due to Significant Performance and/or Operational Concerns


Question: The inspectors assessed the quality of the biennial requalification written examinations for licensed operators.

An inspection was performed to assess the effectiveness of the facility licensee in implementing requalification requirements identified in 10 CFR Part 55, "Operators’ Licenses." Each of the following inspection activities was conducted in accordance with IP 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance."

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:


Question: What are the requirements for withholding information under 10 CFR 2.390?

The information sought to be withheld from public disclosure is owned and has been held in confidence by the University.

The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR 2.390; it is to be received in confidence by the Commission.


Question: What does MDC mean in the context of nuclear analytical methods?

The client sample ID is vital for tracking the analytical results through the testing process.


Question: What is the purpose of condenser inspections in nuclear plants?

The secondary side water chemistry program inital will include a comprehensive inspection program of the condenser and will be developed to ensure condenser integrity.

Sampling points are identified in table 9.3.2-3 (grab sample points and process instrumentation) and table 9.3.2-4 (grab sample points). These sample points include hotwells, condensate storage tank, condensate, feedwater, auxiliary feedwater system, steam generator blowdown, reheat steam, and heater drains.


Question: What materials are used in the Sampling and Water Quality System?

The materials of construction for the Sampling and Water Quality System components are:
- Cast Austenitic Stainless Steel
- Copper Alloy
- Copper Alloy (>15% Zn or >8% Al)
- Glass
- Gray Cast Iron, Ductile Iron
- High-Strength Steel
- Nickel Alloy
- Polymeric
- Stainless Steel
- Steel

The following aging management programs manage the aging effects for the Station Drainage System components:
- Bolting Integrity (B.2.1.10)
- One-Time Inspection (B.2.1.20)
- Selective Leaching (B.2.1.21)
- External Surfaces Monitoring of Mechanical Components (B.2.1.23)
- Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24)


Question: lead blanket shielding storage boxes

New section to describe lead blanket shielding storage boxes, explain why they are not designated Seismic Category I, and provide references to drawings that depict storage box locations.

Eliminated seismic qualification requirements for the Containment Gaseous and Particulate Radiation Monitoring System.


Question: geologic and seismologic environment in nuclear licensing

Geological and seismological considerations in licensing actions are principally matters concerning safety. These are summarized in the applicant's Preliminary Safety Analysis Report, and will be fully evaluated in the course of the staff's safety evaluation.

This type of investigation would be more appropriate for a regional or state governmental body.


Question: What are the important input parameters in a bounding analysis for nuclear safety?

Bounding (Worst Case) Direction for Important Input Parameters includes key parameters such as ZOI Debris Quantities, Unqualified and ADQ Epoxy Coatings Debris Quantity, and Containment Pressure.

The Table 6-1 provides an overview of Bounding (Worst Case) Direction for various inputs, which don't involve qualitative assessments.


Question: PZR PORV Block valves

The Unit 1 Unit Supervisor directs you to locally close the U-1 PZR PORV Block valves per 1BOA PRI-5, step 15.

The timeclock starts when the candidate acknowledges the initiating cue.


Question: The NRC mandates specific preparations for the differential pressure test conducted on components like charging pumps.

INSTALL differential pressure (dp) test gauge (0 to 500 in. WC minimum range) across 1-FE-62-257 as follows:

LOCATE and IDENTIFY the following components (EL 692 Pipe Chase, approx 7 ft above floor):


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) is the smallest quantity of a nuclide that can be detected with a specific level of confidence in a given sample.

The energy levels of different nuclides are measured in keV, which helps in identifying their respective characteristics and applications.


Question: NE Inspection Letter definition

A Letter issued to the licensee, to document a clear inspection or an inspection resulting in Severity Level IV and/or Level V violations that are neither willful nor repetitive and that were corrected while the inspector was present or the licensee committed to correct the violations within 30 days.

Attachment 3.05-1 - M:\rad_hlt\RadMat\RAM Procedure Manual\Current


Question: The NRC mandates that specific valves must be closed to ensure safety during operations.

Ensure closed the following valves: MSIVs 29AOV-80(A-D), MSIVs 29AOV-86(A-D), MAIN STM DRN 29MOV-74, 77, 78 and 79, MAIN STM DRN VLV 29MOV-101(A-D), RFPT A and B HP STOP VLV HP SVA-1 and SVB-1, TSV-(1-4), TO PCV 96MOV-S1, PCV BYP 96MOV-S2, and 29MST-105 (SJAE MST supply 29PCV-107 outlet isol valve) (remote operated from East Electric Bay).

Candidate closes and/or observes green light on and red light off at the following locations: MSIVs 29AOV-80(A-D) at 09-3 and 09-4, MSIVs 29AOV-86(A-D) at 09-3 and 09-4, MAIN STM DRN 29MOV-74, 77, 78 and 79 at 09-3 and 09-4 (77 requires closing) and MAIN STM DRN VLV 29MOV-101(A-D) at 09-7.


Question: The quantification of human errors in post-initiator scenarios involves using different methodologies based on action types.

The execution part of the post-initiator HEP is quantified using THERP.

In general, for procedure-directed actions, the cognitive part of the post-initiator HEP is quantified using the Cause Based Decision Tree Methodology (CBDTM).


Question: The evaluation of upper shelf energy is critical to assess the material properties of reactor vessels.

Evaluation of Upper Shelf Energy for McGuire Unit 2 Reactor Vessel Materials at 54 EFYP discusses the unirradiated USE and projected USE decrease for different materials used in the reactor vessel.

The table provides detailed information about the weight percentage of copper in various materials.


Question: The operator may place rod control in automatic mode at the request of the Control Room Supervisor.

Rod control can be placed in Auto at this time if the CRS directs.

After the above steps are complete Maintenance support is needed to complete the rest of the procedure for tripping bistables.


Question: FIVs and FRVs in reactor safety

The operability of the FIVs, FRVs, FRBVs, and SGFP turbine steam stop valves ensures that the valves will be capable of performing their intended safety function.

Technical Variation: Salem LCO 3.7.13, "Main FIVs, Main FRVs, FRBVs, and SGFP Turbine Steam Stop Valves," is commensurate to ISTS 3.7.3, with the addition of the SGFP Turbine Steam Stop Valves being within scope of the LCO.


Question: Process Sampling and Radiation Monitoring System

The Process Sampling and Radiation Monitoring (PRM) System is a normally operating system that is in scope for license renewal. The PRM System includes the high radiation sampling system, turbine building sampling system, reactor building sampling system, off gas filter building air particulate sample system, primary containment particulate sample system, turbine building and radwaste air particulate sample system, feedwater particulate sampling system, and the process and effluent radiation monitoring system. Those portions of the PRM System that do not perform an intended function are not in the scope of license renewal.

The reactor building ventilation system Geiger-Mueller monitoring system supports this function.


Question: The NRC mandates that licensees assess the critical heat flux for safety during reactor operations.

The COBRA-FLX code provides a critical heat flux correlation that is used to ensure safe thermal hydraulic performance in nuclear reactors.

The table includes experimental conditions for various grid tests conducted using the COBRA-FLX simulation method, detailing factors like coolant temperature and pressure.


Question: What are technical specifications for nuclear reactors?

In addition, T.S. information "below the line" is NOT required to be memorized (question concepts must have "a direct, important relationship to the ability to perform the job"). (ES-4.2, C.1)

Is passing through 0 inches (and, therefore, getting the Group III isolation) a guarantee on any/all reactor scrams from 100% power?


Question: What is involved in the fire protection program for nuclear plants?

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

The inspectors observed and evaluated licensed operator performance in the Unit 2 control room during down power for a rod pattern adjustment on December 19, 2024.


Question: How can the control rod drive system be restored after an alarm indicates low suction pressure?

Swap suction filters and restart the 1A CRD pump per QCOA 0300-01, CONTROL ROD DRIVE PUMP FAILURE.

Unit 2 is in MODE 4 during an outage with Buses 23 and 24 de-energized.


Question: What does dry weight basis mean in sample analysis?

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

There are no exceptions, anomalies or deviations from the specified methods.


Question: What are the technical considerations for the application of commercial standards in nuclear reactors?

The NRC perspectives on the 2023 ASME Section III Division 5 Endorsement and Technical Considerations for application of Commercial Standards.

Closing remarks will conclude the meeting and no further discussions will take place.


Question: government operating costs

The analyst should perform three steps for estimating the other government operating costs: 1. Determine the activities that other governments should perform after the proposed action is implemented. 2. Estimate government staff labor, contractor support, and any special equipment and material required. 3. Estimate the costs of the required resources, discount (usually over the remaining lifetimes of the affected facilities) to yield present value, and then sum.

A major effort beyond the standard analysis would proceed along the lines described above; however, a more detailed and complete accounting would be expected. The analyst could ask the responsible government agencies to provide available information.


Question: The NRC mandates that an evaluation of the auxiliary feedwater system must address its aging management implications.

The Auxiliary Feedwater System is critical for ensuring the plant's safety by providing backup cooling and must undergo rigorous aging management evaluations.

The insulation around piping components is significant for maintaining overall system efficiency and preventing energy loss.


Question: The ability to forecast the effects of unexpected activations on the Engineered Safety Features Actuation System is crucial.

The applicant is required to predict the impacts on the ESFAS of an inadvertent Safety Injection, and use procedures to mitigate the consequences.

I am not familiar with the specifications for reactor coolant systems during normal operations.


Question: reactivity management plans

Review the current controlling Reactivity Management Plans (N/A for AUOs)

Unit 1 EEOS Risk Matrix is Green. Grid is Qualified.


Question: The NRC performs audits to better grasp submitted material, enhancing the efficiency of their review process.

The amendments request the use of new criticality safety analysis (CSA) methodology for performing criticality safety evaluations for fuel in the spent fuel pool.

While they are an optional tool, they have been found to greatly reduce the number of requests for additional information and total number of hours necessary in performing the review and are generally seen by both the NRC staff and the licensee as a large efficiency gain.


Question: The electrical signals monitored in the plant are crucial for assessing component health and system performance.

The values recorded for electrical readings are essential indicators of system performance and stability in the reactor.

The NRC must ensure that licensees adhere to strict environmental regulations to protect public health and safety.


Question: Nuclear power plants must adhere to strict regulations regarding the operation of their safety systems.

The regulatory framework governing nuclear facilities is designed to ensure that safety-related systems function as intended under all operational conditions.

A reactor's power output is a crucial factor in its operational efficiency, influencing both economic viability and environmental impact.


Question: The NRC mandates that all licensees conduct safety assessments to demonstrate compliance with regulatory standards.

Nuclear reactors must undergo rigorous licensing processes to ensure they meet safety and operational standards as set forth by the regulatory authority.

The computer system requires regular updates to maintain security and functionality.


Question: We need to ensure that all evaluations are properly documented and submitted for review.

It looks like we are missing an evaluation on [missing text].

Bruce will take care of these early next week.


Question: decommissioning funding plan

Subpart 72.30, "Financial assurance and recordkeeping for decommissioning," requires that each holder of, or applicant for, a license under this part must submit for NRC review and approval a decommissioning funding plan that contains information on how reasonable assurance will be provided that funds will be available to decommission the Independent Spent Fuel Storage Installation (ISFSI).

The operating licenses for the two SSES units are currently set to expire on July 17, 2042 and March 23, 2044, respectively.


Question: wind speed distribution

Monthly and Annual Joint Frequency Distributions of Wind Speed and Direction by Atmospheric Stability Class

The stability class is defined based on the vertical temperature gradient in the atmosphere, which affects the dispersal of pollutants.


Question: The MNGP's service water system is essential for maintaining cooling across various components in the plant.

The plant service water system supplies cooling water for the plant main generator, reactor and turbine building air conditioning units, turbine lube oil coolers, reactor building dosed cooling water system heat exchangers, various plant motor-generator sets, reactor feedwater system pumps, and the condensate pump motor bearings.

Low river flow rates are more critical than high flow rates when considering temperature impacts.


Question: IEEE Std 603 safety system requirements

IEEE Std 603-1991 Clause 6.3.1 states: Where a single credible event, including all direct and consequential results of that event, can cause a non-safety system action that results in a condition requiring protective action and can concurrently prevent the protective action in those sense and command feature channels designated to provide principal protection against the condition, one of the following requirements shall be met:

The operator or maintenance technician can call up the CRC/System Load Display (Section 3.2.12.2) to view the software CRC identification for each PM646A.


Question: What are the acceptance criteria for cathodic protection in nuclear facilities?

Revise DCPP cathodic protection procedure to specify that for steel components, where the acceptance criteria for the effectiveness of the cathodic protection is other than -850 mV instant off, loss of material rates will be measured per the recommendations in LR-ISG-2015-01, Appendix B, Section 6.m.

PG&E evaluates industry OE items for applicability per the OE Programs and takes appropriate corrective actions.


Question: I-131 dose measurements

The table displays dose measurements of I-131 in Curies over various time intervals, illustrating both control room levels and unsprayed dump levels.

The Cumulative Dose Summary shows the hourly readings for EAB, LP2, and the Control Room for specified times, tracking radiation exposure without specific reference to I-131.


Question: radioactive material license transfer requirements

At the time of transfer of a radioactive material license authorizing possession of radioactive material with a half-life of greater than 120 days, in an unsealed form, transferred or assigned in accordance with He-P 4030.15 to a new licensee, transfer the following records to the new licensee and the new licensee will be responsible for maintaining these records until the license is terminated, the following:

Ensure that all pertinent information, including stamps, initials, and signatures, are included on all required records, including letters, drawings, specifications;


Question: What are the isotopes related to nuclear devices?

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


Question: The NRC evaluates various systems that contribute to plant operations, such as the Fire Protection System and its importance for safety.

The following systems are addressed in this section: Cranes and Fuel Handling System (Section 2.3.3.1), Spent Fuel Pool Cooling System (Section 2.3.3.2), Saltwater and Chlorination System (Section 2.3.3.3), Component Cooling Water System (Section 2.3.3.4), Makeup Water System (Section 2.3.3.5), Nuclear Steam Supply Sampling System (Section 2.3.3.6), Compressed Air System (Section 2.3.3.7), Chemical and Volume Control System (Section 2.3.3.8), Miscellaneous HVAC System (Section 2.3.3.9), Control Room HVAC System (Section 2.3.3.10), Auxiliary Building HVAC System (Section 2.3.3.11), Fire Protection System (Section 2.3.3.12), Diesel Generator Fuel Oil System (Section 2.3.3.13), Diesel Generator System (Section 2.3.3.14), Lubricating Oil System (Section 2.3.3.15), Gaseous Radwaste System (Section 2.3.3.16), Liquid Radwaste System (Section 2.3.3.17), Extraction Steam and Heater Drip System (Section 2.3.3.18), Radiation Monitoring System (Section 2.3.3.19), Sanitary Sewage System (Section 2.3.3.20), Secondary Sampling System (Section 2.3.3.21), Service Cooling Water System (Section 2.3.3.22), Solid Radwaste System (Section 2.3.3.23), Turbine Generator Associated Systems (Section 2.3.3.24), Oily Water and Turbine Sump (Section 2.3.3.25).

The component types include Valve Body, which serves multiple functions such as Leakage Boundary (Spatial), Pressure Boundary, Structural Integrity (Attached), and Structural Support.


Question: What is a nuclide in the context of nuclear safety?

A nuclide is a distinct nuclear species defined by its number of protons and neutrons, with specific energy levels and decay properties that are crucial for evaluating safety and regulatory compliance in nuclear operations.


Question: requirements for reporting nuclear plant events

1. **§50.73(a)(2)(i)(A)** - The completion of any nuclear plant shutdown required by the plant's Technical Specifications.

Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.


Question: equipment qualification in nuclear licensing

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

Addressed in PSAI 7.4 and satisfied by: Licensing Technical Report