Question: liquid scintillation analysis

The Rad Liquid Scintillation Analysis reports the results of Nickel-63 with a result of 0.171 pCi/g, having an uncertainty of +/-1.27.

The Company LaCrosseSolutions is located at S4601 State Hwy 35, Genoa, Wisconsin 54632.


Question: reactor core isolation cooling system

The inspectors evaluated the following licensee’s event reporting determinations to ensure it complied with reporting requirements. (1) Licensee Event Reports (LERs) 2023-002-00 and 2023-002-01 for Clinton Power Station, Unit 1, "Reactor Core Isolation Cooling System Inoperable Prior to Mode Change" (ADAMS [Agencywide Document Access Management System] Accession Nos. ML24051A083 [LER 2023-002-00] and ML24270A035 [LER 2023-002-01]).

The inspectors verified licensee performance indicators submittals listed below: **BI02: RCS Leak Rate Sample (IP Section 02.11)** (1 Sample) (1) Unit 1 (October 1, 2023, through September 30, 2024)


Question: What is the significance of probabilistic analysis in assessing hazards at nuclear power plants?

A probabilistic analysis is used to determine the frequency of hazards due to chemicals that are transported on the Delaware River.

The total number of allowable trips is calculated using Equation 2.2-3:


Question: The extensive review of Holtec Licensing Documents is crucial to ensure all scientific and legal considerations are well understood by involved parties.

The licensing of the proposed Holtec CISF is a massive project. The recent release of over a thousand pages of Holtec Licensing Documents, including Holtec RAI Responses to highly complex NRC-issued questions, implicate important issues that warrant time-consuming review by scientific and technical experts as well as counsel.

A comprehensive presentation of opposing viewpoints during this shortened window prejudices the interests of those who may be affected by the proposed Holtec CISF.


Question: Request for Additional Information process

RAI-27 was addressed in the following report: ANP-10323, Rev. 1, Q1P, Revision 0, Response to Request for Additional Information – ANP-10323P, December 2018.

The reactor design incorporates advanced safety features to enhance operational reliability and reduce the likelihood of incidents.


Question: The evaluation of safety margins is crucial in transient analyses for a BWR/6 reactor, particularly in relation to the AOO acceptance criteria.

A Pressure Regulator Failure - Closed (PRFC) event is chosen for the demonstration. The turbine bypass valves are by the event definition not opening during the transient. This is a transient that may challenge the AOO acceptance criteria.

This section describes an AOO demonstration analysis using POLCA-T and the analysis methodology presented in Reference 2. The purpose of the analyses is to demonstrate the application of the methodology of evaluating the operating limits using POLCA-T.


Question: The NRC stipulates specific conditions for the assessment of flow rates and temperatures for safety testing in reactor operations.

The temperature and flow data provides insights into the operational range, yet the details of head loss calculations are critical for understanding system performance.


Question: radiation exposures to the public

Radiation doses to the public from continued operations and refurbishment associated with license renewal are expected to continue at current levels and would be well below regulatory limits.

These organisms are not expected to be a problem at most operating plants, possibly at plants using cooling ponds, lakes, or canals that discharge into rivers.


Question: The NRC stipulates that licensees must adhere to specific voltage setpoints for relays to ensure operational safety.

Primary Voltage = (Relay voltage + Wire Voltage Drop dV) * RCF * Nominal VT Ratio of 35

During routine inspections, the NRC identified several instances of non-compliance that must be addressed in future reports.


Question: What are the responsibilities of the Rad Protection Manager in nuclear facilities?

The Rad Protection Manager is responsible for ensuring compliance with radiation safety regulations and implementing radiation protection measures at the facility.

The mailing address for correspondence regarding the device includes the North Anna Power Station located at 1022 Haley Drive in Mineral, VA.


Question: What is the maximum GDT release rate for radioactive waste gas?

The actual maximum GDT release rate is 32 SCFM based on system design. The calculated maximum allowable release rate shall be >32 SCFM. Tanks with a calculated maximum allowable release rate of ≤32 SCFM shall not be released.

Inoperability of the Noble Gas Activity Monitor (2R12A, 2R41A & D), Iodine Sampler, Particulate Sampler, Flow Rate Monitor, or Sampler Flow Rate Monitor for >30 days is reportable IAW NC.WM-AP.ZZ-0002(Q), Performance Improvement Process (IAW the ODCM).


Question: minimum required cooling time for nuclear fuel

The minimum cooling time is 1 year. For fuel transferred in the EOS-TC108, the minimum cooling time is 2.0 years.

Linear interpolation is allowed to obtain a cooling time within the specified range of burnup and enrichment values.


Question: What is defined as radioactive liquid effluents in nuclear facilities?

Radioactive liquid effluents are regulated by the NRC and include any liquid waste that contains radioactive material and is released into the environment.

The NRC mandates that licensees must maintain accurate and detailed records of their training programs to ensure compliance with operational standards.


Question: The NRC evaluates loss-of-coolant accident frequencies using various time frames to assess safety in reactor operations.

NUREG-1829 contains “25-year” or “current-day” LOCA frequencies and “40-year” or “end-of-plant-license” LOCA frequencies.

Breaks at the reactor vessel nozzles do not need to be considered as discussed in Section 3.a.1 of Attachment VIII.


Question: What is the significance of gamma measurement in nuclear safety?

Gamma measurements are crucial in assessing radiation levels and ensuring compliance with safety standards in nuclear facilities.

The photograph collected provides a visual representation of the boring location and sampling activities conducted at Pickard Hall.


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


Question: What is the significance of ADS initiation capability in ECCS?

In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability in both trip systems.

The completion time is intended to allow the operator time to evaluate and repair any discovered inoperabilities.


Question: The design of the flow restrictor assembly ensures that it properly manages steam flow during system operations.

The flow restrictor is an assembly of seven smaller nozzles installed within the steam outlet nozzle of the steam generator.

The Residual Heat Removal System transfers heat from the Reactor Coolant System to the Component Cooling System to reduce the temperature of the reactor coolant to the cold shutdown temperature.


Question: What is the NRC Device Key for registered devices?

**NRC Device Key:** 719660 (Internal Control Number)

Please update the information as necessary.


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


Question: The NRC provides acceptance guidelines that outline acceptable levels of risk for one-time-only TS CT changes.

The licensee has implemented a risk-informed plant configuration control program, including procedures to use, maintain, and control such a program (Tier 3).

Available alternatives are sometimes compared to justify a TS change. For changes in TS CTs, such cases primarily involve comparing the risk of shutting down with the risk of continuing power.


Question: reactivity insertion rate

The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the two sequential control banks having the greatest combined worth at maximum speed (45 in./min).

In the event of an RCCA withdrawal accident from the subcritical condition, the core and the RCS are not adversely affected, since the combination of thermal power and the coolant temperature results in a DNBR greater than the limiting value.


Question: What is the average enrichment weight percentage of 235U in PWR fuel assemblies?

The table provides crucial information regarding the assembly average burnup measured in GWd/MTU.


Question: What are the requirements for emergency stop devices in diesel generator systems?

An Emergency Stop device is provided to allow the diesel generator to be shut down in the event of a fire or other occurrence which would necessitate manually stopping the diesel.

The criteria established for the motor starting torque provide for minimum starting time while at the same time limiting the inrush kVA to a value compatible with the engine capability.


Question: on-shift staffing analysis

Regulatory Issue Summary (RIS) 2016-10, License Amendment Requests for Changes to Emergency Response Organization Staffing and Augmentation, (Reference 6.13), states that an on-shift staffing analysis under 10 CFR 50, Appendix E, Section IV.A.9 should not be used to provide the primary basis to support the Technical Evaluation of a License Amendment Request (LAR).

In conjunction with this License Amendment Request, NextEra has performed and documented an OSA per 10 CFR 50, Appendix E, Section IV.A.9 following the guidance of NEI 10-05, Assessment of On-Shift Emergency Response Organization Staffing and Capabilities, (Reference 6.14).


Question: What are the requirements for NRR approval in nuclear reactor operations?

The criteria for NRR approval include the following: 1. The reactor in which the materials will be irradiated and the reactor for which the materials are being irradiated must have sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

A proposed withdrawal schedule must be submitted with a technical justification as specified in [10 CFR] § 50.4. The proposed schedule must be approved prior to implementation.


Question: What is the function of the Auxiliary Decay Heat Removal System in nuclear reactors?

The Auxiliary Decay Heat Removal System is designed to remove decay heat from the reactor core after shutdown to prevent overheating and ensure safety.

The results also indicate that the Security System must undergo regular reviews to ensure it meets the standards set forth in regulatory guidelines.


Question: What are reactivity manipulations in nuclear operations?

Reactivity manipulations may be conducted under normal or controlled/abnormal conditions (refer to Section 5.d) but must be significant per Section C.2.a of Appendix D.

Both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.


Question: What are the ROD INSERTION LIMITS in nuclear operations?

At 80% power, the ROD INSERTION LIMIT for Control Bank D is LOWER than it was at 100% power.

A Shutdown Margin Determination is FIRST required to be performed at time 00:04.


Question: Containment vessel second horizontal mode

Figure 3-92. Containment Vessel, Second Horizontal Mode, f = 25.660 cps (Finite Elements)


Question: What is the Certificate of Compliance in the context of nuclear licensing?

A Certificate of Compliance is an official document issued by the regulatory body that authorizes the use of specific materials or designs in licensed nuclear activities.

The amendment outlines the regulatory requirements for emergency preparedness audits, which must be conducted biannually.


Question: The NRC has determined that potential collisions of bats with structures at nuclear facilities are unlikely given the specific habitat conditions.

Listed bats can be vulnerable to mortality or injury from collisions with nuclear power plant structures and vehicles.

This section addresses each of these stressors below.


Question: Coordination with various agencies is essential for determining reentry conditions post-emergency.

Contact LDEQ to determine whether or not reentry is recommended and the extent of reentry, i.e., limited or unlimited reentry.

Emergency medical assistance in case of accidents during reentry.


Question: Plant Computer System in nuclear power plants

The Plant Computer System (PCS) was installed in order to support the data acquisition needs of the emergency response facilities. The PCS will provide plant monitoring, data acquisition, and critical plant data in the form of real-time status displays for the purpose of making a rapid evaluation of the reactor plant's safety status.

The Station's Fire Protection System is designed to furnish water and other extinguishing agents with the capability of extinguishing any single or probable combination of simultaneous fires that might occur.


Question: What are risk-informed completion times in nuclear licensing?

The NRC has identified a need for licensees to adopt Risk-Informed Completion Times to enhance safety while maintaining operational flexibility.

The Columbia Generating Station has undergone several technical specification revisions over the years to improve compliance and performance.


Question: The NRC's review of the FMEDA showed that it includes reliability information crucial for confirming plant-specific reliability standards.

The NRC staff determined the RadICS platform reliability analysis results of the FMEDA contain platform reliability information that can be used to demonstrate conformance to plant-specific reliability goals.

Addressed in PSAI 7.6 and satisfied by: Licensing Technical Report Project Implementation Deliverables.


Question: nuclear safety analysis report

The Updated Final Safety Analysis Report (UFSAR) provides a comprehensive safety analysis for the North Anna Power Station, detailing the design, operational, and regulatory aspects essential for maintaining safety standards.

The reactor employs a closed-loop cooling system designed to efficiently transfer heat away from the core during normal operations.


Question: The contour lines represent variations in groundwater levels within the Tertiary Aquifer.

Figure 2.4.12-16 shows the Piezometric Contour Map for the Tertiary Aquifer, indicating groundwater levels for December 2005.

The observation well data helps in monitoring surface water interactions with the wetland areas near the site.


Question: What is the purpose of the Supplemental Diesel Generator System in nuclear facilities?

The Supplemental Diesel Generator (SDG) System provides a non-safety related highly reliable backup source of power for the Emergency High Pressure Makeup (EHPM) System and auxiliaries.

The aging management review results for these components are provided in Table 3.3.2-29, Standby Diesel Generators - Summary of Aging Management Evaluation.


Question: The NRC specifies a seismic design criterion that ensures a uniform risk against failure across all SSCs by using a derived uniform risk spectrum.

The target mean annual performance goal for nuclear plants of mean 10^{-5} per yr is achieved by coupling site-specific DRS with the deterministic seismic design criteria and procedures specified by NUREG-0800.

The procedure for derivation of the DRS is as follows.


Question: What does cfs stand for in nuclear safety assessments?

Discharge (cfs) is measured in cubic feet per second, which quantifies the flow rate of water out of a body such as a lake during storm events.

The table provides a detailed hourly breakdown of the storm outflow hydrograph, comparing the actual and computed outflow rates for accurate analysis.


Question: The NRC assesses the emergency planning zone for various reactor types to ensure adequate safety measures are in place.

Both guidance documents state that the EPZs for facilities operating above 50 MWt are to be considered on a case-by-case basis.

These EP standards provide additional information for developing the consequence-oriented approach to establishing EPZs and the planning commensurate with the radiological risk for SMRs and ONTs.


Question: The NRC provides specifics regarding the Minimum Detectable Activity (MDA) for different nuclides involved in nuclear safety assessments.

CAUTION: MDA value is inconsistent with Currie MDA at 95% confidence level.


Question: groundwater monitoring program

Vogtle Electric Generating Plant, *Groundwater Monitoring Program, Procedure Number 30140-C, Revision 22*, January 2006.

National Climatic Data Center Web site (http://www.ncdc.noaa.gov/oaa/climate/research/monitoring.html) (accessed March 25, 2007).


Question: The failure of Vital Instrument Power Board 2-IV leads to the opening of the RWST valve, which directly affects reactor coolant system operations.

Loss of the 120V AC Vital Board 2-IV results in a loss of power to all Channel IV SSSP bitables.

Upon restoration of power to 120V AC Vital Board 2-IV, the following can occur: - MD AFW pump 2B-B master level controllers for SGs 3 and 4 will have to be reset to AUTO.


Question: The NRC emphasizes the importance of accurately identifying nuclides in safety data submissions for nuclear facilities.

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

The half-life of certain isotopes can complicate the assessment of activity before and after decay corrections are applied.


Question: nuclear safety systems

Nuclear safety systems are designed to prevent accidents and mitigate potential consequences in the event of a failure.

The document describes the budgetary allocations for the next fiscal year, outlining expected expenditures and revenue sources.


Question: What is the unit of measurement for isotopes used in NRC filings?


Question: What are the shielding requirements for transfer casks in nuclear applications?

the nominal configuration transfer cask radial bulk shielding (i.e., shielding integral to the transfer cask; excludes supplemental shielding) must provide a minimum radiation shield equivalent to 2 inches of carbon steel or stainless steel and 3.2 inches of lead gamma shielding and 2.25 inches of NS-4-FR (with 0.6 wt % B4C and 6.0 wt % H) neutron shielding.

The American Concrete Institute Specifications ACI-349 and ACI-318 govern the CONCRETE CASK design and construction, respectively.


Question: What are the requirements for petition admissibility in NRC hearings?

These contention admissibility requirements are strict. They are intended to ensure that adjudicatory hearings are triggered only by substantive safety or environmental issues that raise a supported dispute with the application on a matter material to the NRC’s decision on the challenged action.

New York proposes three contentions in this proceeding. In Contention 1, New York argues that HDI’s cost estimates are inadequate because HDI impermissibly assumes that the decommissioning trusts will grow at a 2% annual real rate of return.


Question: NRC evaluation criteria for regulatory filings

When evaluating the question from both tier 1 and tier 3 perspectives, the NRC determined that answer choice ‘D’ remains the only correct answer.

The safety analysis report must include detailed descriptions of all systems, components, and their interactions to ensure proper regulatory compliance.


Question: What are the inspection procedures used for safety reviews at permanently shut-down reactors?

IP 37801: Safety Reviews, Design Changes, and Modifications at Permanently Shut-down Reactors

Inspection of Remedial and Final Surveys at Permanently Shut-down Reactors are critical for maintaining safety standards.


Question: Please confirm if the RPV materials listed in the SLRA are excluded from the beltline as defined in 10 CFR Part 50.

Confirm that these RPV materials are NOT part of the RPV beltline (i.e., exceed neutron fluence 1E17n/cm2).

For the RPV materials in which the initial USE increased - that are NOT highlighted in yellow – Explain what data/information was used to determine the 'original' Initial USE value.


Question: What is the role of the Work Control Shift Foreman in assessing radiological work?

The Work Control Shift Foreman (WCSFM) is assessing support work at hot spot #202 at 73’ elevation for the 2-2 RHR Heat Exchanger under RWP-0003, O & M in the RCA (Minor RP Impact).

Transit time dose may be considered negligible.


Question: What are EPRI Guidelines in the context of nuclear safety?

EPRI Guidelines revisions are crucial for ensuring compliance with safety standards in nuclear power plant operations.

RAI #1: 100% inspection "past or future" (Alloy 600TT) is a procedural step taken to ensure material integrity.


Question: The NRC requires an understanding of the processes affecting ice within the ice condenser, including vapor deposition and sublimation.

An analytical model of the sublimation process has been developed to provide an estimate of the expected sublimation rate as well as identify the significant parameters affecting the sublimation rate.

Sodium tetraborate is used as a chemical additive to the ice in the plant.


Question: What are the internal sources of hydrogen controlled during manufacturing in nuclear safety?

Internal sources of hydrogen are controlled during manufacturing by product specifications that limit the moisture content.

The fuel rod temperature is not directly evaluated in this methodology.


Question: What is the purpose of the Thermal Aging Embrittlement of Cast Austenitic Stainless Steel AMP?

The DCPP Thermal Aging Embrittlement of Cast Austenitic Stainless Steel AMP is a new AMP that manages embrittlement of ASME Code Class 1 CASS components due to thermal aging.

For components that are not susceptible to thermal aging embrittlement, no additional inspections or evaluations will be required.


Question: What is the average enrichment level for fuel assemblies?

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


Question: The NRC stipulates that specific actions must be taken based on RPV level trends in order to ensure core cooling is maintained effectively.

If the RPV level trend is not reversible with an L3 system capable of providing adequate core cooling, INITIATE an emergency depressurization before Minimum Steam Cooling RPV Water Level (MSCRWL).

Crew resource availability and priorities should also be considered in this evaluation.


Question: Nuclear power plants must assess the radiological impacts of their operations on surrounding ecosystems.

Radiological impacts are a critical consideration in evaluating the environmental effects of plant operation, particularly regarding the potential effects on man and biota.

The construction of transmission lines can lead to significant changes in local terrestrial environments, affecting flora and fauna.


Question: The steps outlined in the operating procedure to restore the normal power supply to L-34 are critical for ensuring reactor safety.

Direct BOP to restore L-34 to the normal supply per OP-46A section G.12.3.

Ensure closed L14 L24 L34 L44 4KV FDR BKR 10440.


Question: What actions should be taken when a steam generator tube leak is detected at a nuclear plant?

When contacted as Chemistry acknowledge request to sample SGs. Wait 10 minutes and report indication of tube leakage on SG#1.

At the discretion of the Lead Examiner, proceed to Event 3.


Question: nuclide confidence index

Nuclide confidence index threshold = 0.30


Question: What is the Minimum Detectable Activity in nuclear measurements?

The report includes various parameters related to sample analysis and the detection limits.


Question: What are OTTP values in nuclear licensing?

The OTTp values represent the operational transient temperature parameters used in the safety analysis of nuclear reactors, which are critical in evaluating the thermal performance during operational conditions.

The NRC evaluates numerous factors when assessing new technology in various industrial applications, ensuring that safety standards are uniformly applied.


Question: BWRVIP examination requirements

BWRVIP recommended examinations of integrally welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. As a minimum, the same or superior visual examination technique is required for examination at the same frequency as the code examination requirements.

The BWRVIP Guidelines require a volumetric examination of the susceptible welds at a frequency identical to the Code requirement.


Question: radioactive isotopes in nuclear devices


Question: The crew must initiate an Emergency RPV Depressurization if Primary Containment conditions cannot be upheld.

Given the inability to maintain Primary Containment conditions inside the Pressure Suppression Pressure, the crew will perform an Emergency RPV Depressurization, in accordance with EOP-4. Emergency RPV Depressurization must be initiated within 15 minutes of Pressure Suppression Pressure being exceeded.

Ensure Trigger 5 automatically goes active when the Reactor scrams to cause the SRV tailpipe break.


Question: license amendment request process

A request for an amendment to a specific license may not be on a Nebraska State Form. The request may be a letter plus attachments or a formal application; however, it shall be signed by the individual in the position, or higher, that signed the application for license or the request shall be returned for proper signature.

The initial review of the request for amendment shall determine if the request is so broad that it should be processed as a rewrite of the current license or as a new license.


Question: nuclear inventory power


Question: How was the design life for the extended operation determined in relation to fatigue crack growth?

Please clarify how the design life of the plant was determined for the stress and crack growth analyses with respect to the renewed license, which is scheduled to expire in 2044.

During our audit of the validation of the AREVACSC code, the direct comparison with the [[ ]] was unclear.


Question: The NRC mandates that licensees assess the performance of emergency core cooling systems during potential accidents.

NRC Generic Letter (GL) 2004-02 (Reference 1) required licensees to perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions, and the flow paths necessary to support those functions, based on the potential susceptibility of sump strainers to debris blockage during design basis accidents that require recirculation operation of ECCS or CSS.

The evaluation meets the acceptance guidelines for a very small risk impact as defined in Regulatory Guide (RG) 1.174 (Reference 2).


Question: The NRC continuously oversees the management of nuclear waste to ensure public safety and environmental protection.

The NRC requires licensees to maintain strict regulations on the disposal and management of nuclear waste to safeguard public health.

The public has repeatedly rejected proposals that would allow the nuclear industry to evade responsibility for waste management and safety.


Question: The NRC staff identified the necessity for a regulatory audit to review the non-docketed information related to the license amendment request.

A regulatory audit is a planned license- or regulation-related activity that includes the examination and evaluation of primarily non-docketed information associated with the LAR.

The hope creek generating station operates under specific guidelines defined by the NRC to ensure safety and compliance with established regulations.


Question: Coordination with state regulators is crucial for the effective management of licenses and safety inspections in the nuclear sector.

Provide the opportunity for license reviewers and inspectors from the State seeking an agreement to review regional licensing casework in progress and observe NRC inspections at the facilities of licensees in that State.

The policy and guidance in this directive and handbook apply to all NRC employees who process a request for a Section 274b. agreement.


Question: gamma spectrum analysis process

Gamma spectrum analysis is a technique used to identify and quantify the radioactive isotopes present in a sample by analyzing the energies and intensities of gamma rays emitted.

The detection system was calibrated to ensure accurate readings, with a focus on maintaining optimal performance standards during the test.


Question: The NRC Staff released their findings on environmental impacts related to the project.

The Staff issued a Safety Evaluation Report (SER) in March 2013.

Both the Tribe and individual members of Consolidated Intervenors (Susan Henderson and Dayton Hody) commented on the DSEIS and later filed proposed contentions relating to the DSEIS.


Question: What are the surveillance requirements for nuclear safety systems?

Verify \( F_g(Z) \) is within limit specified in the COLR.

The NRC requires a comprehensive evaluation of the proposed changes before any action can be taken.


Question: What type of fuel is used in Oconee Station?

Oconee Station operates with low-enriched uranium dioxide fuel. With the NRC approval of Zircaloy-4 and M5 cladding fuel usage, Duke Energy operates the reactor cores to yield an equilibrium cycle (normal cycle) burnup of approximately 20,854 megawatt-days per metric ton uranium (MWd/MTU) for 24 months and maximum burnup rate of 62,000 MWd/MTU.

Nuclear power plants are served by controlled access roads that are connected to U.S. highways and Interstate highways.


Question: The NRC establishes a framework for inspections that includes determining the necessity of special or reactive inspections.

Reports inspection statistics to Upper Management or their designee as requested.


Question: What are the PCT and Burst Elevations in nuclear reactor systems?

- **PCT Elevation** = 7.25 ft
- **Burst Elevation** = 5.75 ft


Question: The values of VFO-KEFF and VFO-Sigma are crucial in determining the effective neutron multiplication factor in nuclear systems.

| VFO-Name | VFO-KEFF | VFO-Sigma | K+2S  | Bias    | K+2S - B |

The NRC regulations stipulate comprehensive assessments for operational readiness and safety measures.


Question: request for hearing under NRC regulations

If the answer includes a request for a hearing, it shall, under oath or affirmation, specifically set forth the matters of fact and law on which AmerenUE relies and the reasons as to why the Order should not have been issued.

To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request (1) a digital


Question: The system employs two independent trip mechanisms that include four channels for monitoring Reactor Vessel Water Level and Reactor Steam Dome Pressure.

The ATWS-RPT consists of two independent trip systems, with two channels of the Reactor Steam Dome Pressure-High function and two channels of Reactor Vessel Water Level-Low Low, Level 2 function, in each trip system.

The design ensures that the reactor operates efficiently under various load conditions.


Question: What is the purpose of Risk-Informed Completion Time in nuclear operations?

Constellation Energy Generation, LLC is proposing to apply the Risk-Informed Completion Time Program only in Modes 1 and 2 and not in the shutdown modes.

This enclosure is not applicable to the Dresden Nuclear Power Station, Units 2 and 3 submittal.


Question: criteria for identifying security-related sensitive information

NRC's Regulatory Issue Summary (RIS) 2005-31 provides criteria to identify security-related sensitive information and guidance for handling and marking of such documents.

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


Question: The NRC mandates that all nuclear facility operators demonstrate compliance with safety regulations and operational protocols.

Licensees must ensure adherence to regulatory requirements by submitting detailed safety analyses for review.

The weather patterns in the area have shown significant variability over the past decade, impacting local agriculture.


Question: What are the procedures for verifying containment air lock door status?

Verify an OPERABLE door is locked closed in the affected air lock.


Question: nuclear channel dimensions

| Channel Section | Zone 1 | Zone 2 |
| Dimension      | mm    | in    | mm    | in    |

The reactor's core is designed to withstand high pressures and temperatures, ensuring optimal performance and safety throughout its operational life.


Question: The analysis of nuclides is essential for understanding the radioactivity in solutions.

The interference analysis rejected certain nuclides that could compromise the accuracy of the results.

The system reported data errors that were corrected at 1.00 sigma for precision in measurements.


Question: The interlocks among the 1/2CV459 and 1/2CV460 valves are significant, as they prevent unnecessary operations during normal power conditions.

Due to the interlocks between the 1/2CV459, 1/2CV460, & the 1/2CV8149A-C valves, exercising these valves during normal operation results in (multiple) total letdown flow isolation events.

The Component Function(s) section specifies that the valves are normally OPEN with the Unit at power, allowing letdown flow to occur.


Question: What is the significance of water content in soil samples for nuclear safety assessments?

Water Content (%) is a crucial parameter that indicates the amount of water present in a soil sample, which can affect its mechanical properties and stability during nuclear facility operations.

The maximum dry unit weight is important for determining the load-bearing capacity of the soil without considering its moisture content.


Question: What are the requirements for containment isolation valves in nuclear facilities?

All containment isolation valves purchased by Stone & Webster are factory tested and inspected. A written specification defines the specific requirements for valve procurement, which include the following: 1. Welding and NDT procedure qualification. 2. Welder and NDT operator qualifications. 3. Mill test reports. 4. Dye penetrant test (as required). 5. Magnetic particle test (as required). 6. Radiography (as required). 7. Body hydrostatic test. 8. Seat and valve steam leakage test. 9. Performance test. 10. Dimensional check. 11. Cleaning. 12. Preparation for shipment. 13. Seismic qualification.

The status of the valves during normal, shutdown, and accident conditions is given in Table 6.2-37.


Question: nuclide minimum detectable activity


Question: license renewal amendments

In accordance with 10 CFR 54.21(b), each year following submittal of the LRA and at least three months before scheduled completion of the staff’s review, PG&E will submit an LRA amendment identifying any current licensing basis changes that materially affect the contents of the LRA, including the Final Safety Analysis Report supplement.

If you have any questions or require additional information, please contact Mr. Philippe Soenen at (805) 459-3701.


Question: What is the purpose of the Generic Fundamentals Examination for nuclear operator licensing?

The Generic Fundamentals Examination is designed to assess the knowledge of candidates in the fundamental principles underlying the operation of nuclear reactors and their associated systems.

The email distribution list for the document includes names from the NRR and DRS offices, indicating its administrative handling and regulatory oversight.


Question: What are the types of licenses regulated in nuclear material licensing?

Please provide a list of licensing actions completed during the review period, including license number, license reviewer, and license type (e.g., medical, academic, commercial, R&D, industrial radiography, gauges, etc.).

The licensees we have that may fall into this category:


Question: What is the fail-safe testing protocol for valves during refueling outages?

These valves will be fail-safe tested closed as an augmented IST test during refueling outages in accordance with Braidwood Station Technical Specifications.

The closure of the Main Feedwater Regulating Valves (1/2FW510, 1/2FW520, 1/2FW530, 1/2FW540) during Unit operation would cause a loss of feedwater to the steam generators, resulting in a plant transient with a possible reactor trip as a result.


Question: frequency of tornadoes in Mississippi

Tornado frequency did not change significantly over the time range, so data from all years were used to calculate the annual frequencies.

Grand Gulf is in Claiborne County. Warren County, directly to the north, is included because Vicksburg is in that county, while Hinds, to the East, is included because Jackson is in that county.


Question: What are the maintenance guidelines for respirators used in nuclear safety?

NuFuels will develop a Standard Operating Procedure (SOP) which addresses updated procedures for the respiratory program.

Washing the respirator in a dish washer using liquid soap, such as LIQUI-NOX. Following the wash, all parts will be allowed to air dry at room temperature.


Question: What is the purpose of the Average Power Range Monitor (APRM) in a nuclear reactor?

A rod block is issued when any APRM is downscale with the Reactor Mode switch in RUN.

The NRC requires that all reactor systems must be regularly inspected and maintained to ensure safety and compliance.


Question: What actions should be taken in response to a Steam Generator Tube Rupture?

The SG 2 MSIV has failed to close after a Steam Generator Tube Rupture on SG 2, necessitating immediate action for steamline isolation.

Main Steam line warming is NOT in progress and does not affect the isolation procedure.


Question: emergency exposure authorization

IF it is necessary to exceed normal occupational radiation exposure limits (5 Rem TEDE) to save a life, protect the public, or save vital equipment, THEN AUTHORIZE emergency exposure limits using EP EEA, Emergency Exposure Authorization.

NOTIFY the NRC as soon as practical, AND in all cases within 60 minutes of the occurrence of any deviation from the station's Physical Security Plan authorized pursuant 10 CFR 50.54(x) using NRC Form 361, Reactor Plant Event Notification Worksheet.


Question: The NRC expects that licensees maintain compliance with Technical Specification requirements related to water system operations.

Proposed License Amendment Request to Revise Technical Specifications 3.7.8 'Station Service Water System,' and 3.8.1 'AC Sources – Operating'

The list of attendees includes representatives from both the U.S. Nuclear Regulatory Commission and Vistra Operations Company LLC.


Question: The NRC evaluates whether new information requirements imposed by regulations can enhance the safety standards of nuclear facilities.

The analysis and administrative burden is given on P87009: "The NRC is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues"

Withdraw, reduce and revise all such guidelines and sub-references unless: (a) justified by data and PRA/RIDM analyses (b) verified as generically useful, applicable and effective by the prototype demonstration process suggested (c) not prescriptive regulation, guidance, expectation and de facto requirements and ‘preferred’ submission content for NARS licenses.


Question: What procedures are in place for full recovery and unlimited reentry after an emergency?

When informed that full recovery and unlimited access are recommended, the Director of the Office of Emergency Preparedness will perform the following: A. Meet with the EOC staff and Parish agencies to ensure all personnel are prepared to accomplish an orderly reentry.

The areas that may be reentered and the scheduled times for limited reentry will be posted in the EOC.


Question: What are the administrative requirements for personnel entering a High Radiation Area?

Determine Personnel Exposure and Special Requirements for Entry into A High Radiation Area (HRA)

CUE: WHEN operator informs you the task is complete, OR the JPM has been terminated for other reasons, THEN RECORD the STOP TIME.


Question: What is the significance of K-40 in nuclear assessments?

The identified nuclide K-40 has an energy of 1460.75 keV, a yield of approximately 10.67%, and an activity of 5.63E+00 pCi/grams, with an uncertainty of 6.47E-02.

The table outlines a variety of identified nuclides, including their corresponding energy levels, yields, and activity measurements.


Question: What is the significance of Operating Flow Rates in nuclear safety analysis?

Operating flow rates are critical parameters in determining the hydraulic performance and safety of the reactor coolant system, ensuring that adequate cooling is maintained during operation.

The appendix includes a detailed overview of the vendor qualifications necessary for supplying equipment to the nuclear power plant.


Question: The NRC requires analyses to determine the levels of specific radionuclides in waste material.

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


Question: The NRC evaluates the net area of peaks to determine radiation levels...

The net peak area is a critical factor in the evaluation of radiation measurements, specifically calculated to reflect the actual signal amidst background noise.

The original area of the peak is calculated before subtracting background noise, and the uncertainty related to this original area is also documented.


Question: How does the reactor coolant system manage steady-state heat generation during full-power operation?

The reactor coolant system is designed on the basis of steady-state operation at full-power heat load.

Figure 5.3-1 shows a temperature-power operating map depicting T_{avg} as a function of percent of full power.


Question: What is the significance of energy release rate in nuclear reactor operations?

Energy Release Rate (Btu/s) represents the amount of energy released per unit of time during reactor operation, which is critical for understanding thermal dynamics and safety margins in the facility.

CD: 0.6 represents a characteristic data point used in various regulatory assessments that do not directly pertain to energy metrics in reactor operations.


Question: nuclear power plant warning

Nuclear Power Plant Warning is issued by Local Municipalities, Emergency Management to ensure public safety during potential incidents at nuclear facilities.

Civil Unrest/Civil Danger alerts are coordinated by Law Enforcement and Emergency Management to inform the public about disturbances or threats to safety.


Question: The report lists various nuclides along with their respective energies and activities, essential for assessing radiological safety.

The peak locate was performed early in the morning to ensure optimal conditions for measurements.


Question: Torus ECCS Suction Header

Unit 2 has a small break Torus ECCS Suction Header break on the Torus and an ATWS.


Question: In determining the number of soil samples, the optimal range for the relative shift is outlined as being between one and three.

The relative shift (Δσ) for the survey unit data set is defined as shift (Δ), which is the Upper Boundary of the Gray Region (UBGR), or the DCGL (SOF of 1), minus the Lower Bound of the Gray Region (LBGR) (SOF of 0.5), divided by sigma (σ), which is the standard deviation of the data set used for survey design.

The sample size from Table 5.5 of MARSSIM that equates to the Type I and Type II error of 0.05 for use with the Sign Test is an N value of fourteen (14).


Question: What types of analysis are conducted for radioactivity in soil samples?

Rad Alpha Spec Analysis includes measurements for isotopes such as Americium-241, Americium-243, Curium-242/244, Neptunium-237, Plutonium-238, and Plutonium-239/240.

The Liquid Scintillation Analysis provided results for Tritium Distillation and counted various samples without specifying the types of isotopes measured.


Question: The process of responding to an anticipated transient requires understanding the critical elements needed to ensure safety.

The job performance measure (JPM) titled 'RESPOND to an Anticipated Transient Without Trip' outlines the necessary actions for operators when facing potential reactor challenges.

Operator performances are evaluated regularly to determine compliance with established safety protocols and operational standards.


Question: What is nuclear reactor licensing and its regulatory importance?

The licensing process for nuclear reactors involves comprehensive safety evaluations to ensure compliance with established engineering standards and regulatory requirements set forth by organizations such as the NRC.

The comment from Robert H. Leyse on PRM-50-93 emphasizes the need for transparency in the regulatory process, highlighting concerns about public involvement.


Question: power reactivity decrement

R&D on Power Reactivity Decrement informed possible deformations (i.e. within core restraint system)

Assembly rotation can rectify potentially unacceptable trends (done at EBR II)


Question: Understanding how the steam generator system failure could impact various critical systems is necessary for reactor safety.

Knowledge of the effect that a loss or malfunction of the steam generator system will have on the following systems or system parameters: (CFR: 41.7 / 45.6) - K 3.01  Steam generator blowdown system 3.0

The Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) governs the cleanup of contaminated sites.


Question: What is the significance of Power Range Neutron Flux - High trip setpoints in nuclear reactor safety?

A reduction of the Power Range Neutron Flux - High trip setpoints by ≥ 1% for each 1% by which F_{Q}(Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions.

The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.


Question: What does AFT represent in nuclear licensing?

A_{FT} = 	ext{the calculated uncertainty associated with the flow measurement from the flow transmitter.}

The flow rate to the feedwater is closely monitored to ensure safety and efficiency during operation.


Question: Figures-of-Merit for Anticipated Operational Occurrences (AOO)

Table C-2-1 provides a comprehensive overview of the Figures-of-Merit for Anticipated Operational Occurrences (AOO), which are critical in assessing safety and operational performance metrics.

Table C-6-1 lists the model parameters and uncertainties for all high ranked AOO phenomena, focusing on statistical distributions and their implications for safety analyses.


Question: The NRC establishes certain criteria for operators to follow during significant operational events.

The Entry Conditions of the Emergency Operating Procedures (EOP) provide critical guidelines for operators during reactor emergencies, ensuring that safety systems are effectively engaged.

The Westinghouse Non-Proprietary Class 3 report outlines various operational procedures and configurations relevant to nuclear power generation.


Question: What is the significance of lift-off forces in nuclear reactor safety assessments?

Lift-off Force (Typ) is a critical parameter in ensuring structural integrity during dynamic conditions experienced by the reactor components.

Time since sit: Years is a measurement used in various assessments across engineering disciplines, but it does not directly pertain to nuclear reactor safety or lift-off forces.


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


Question: What is the function of remotely operated valves in nuclear systems?

Remotely operated valves are critical components in the safety systems of nuclear reactors, serving functions such as managing coolant flow and maintaining pressure within the containment structure.

The component cooling water from the RHR system plays a vital role in transferring heat away from the reactor, ensuring that operational temperatures remain within safe limits.


Question: The NRC staff has determined that the requests related to the testing intervals for pressure relief valves must comply with established safety standards.

The NRC staff finds that the licensee’s SRV Best Practices Maintenance Program, in sharing applicable SRV test data between the licensee’s nuclear power plants, has resulted in successful SRV as-left and as-found set pressure testing to provide confidence in the operational readiness of the SRVs at QCNPS, Units 1 and 2.

The licensee will document the failure in its Corrective Action Program and the requirements of ASME OM Code, appendix I, paragraph I-1320(c), will be followed.


Question: What are the responsibilities of the Technical Support Manager in emergency scenarios?

The Technical Support Manager (TSM) reports to the CRM and is responsible for the command functions related to prompt and accurate dose assessments; notifying state and local governments of the emergency status and any changes in a timely manner; and assessing and providing protective action recommendations to offsite authorities.

Emergency positions will be designated by corporate management based on the technical requirements of the position. Guidance for selection of emergency responders is provided in administrative procedures.


Question: The report includes a table detailing the energy levels and activity for various nuclides, emphasizing the importance of accurate measurements for safety assessments.

The NRC requires licensees to submit documentation of compliance with all safety regulations and operational standards.


Question: The average enrichment of uranium fuel assemblies must meet specific criteria outlined in regulatory filings.

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

The table lists the corresponding values for specific burnup categories related to the performance of the fuel assembly.


Question: The assessment of fuel fragment sizes and their importance for fragmentation thresholds is crucial in analyzing reactor behavior during operations.

The selection of the fuel fragment size at which fragmentation is deemed to have occurred and its impacts on the burnup threshold.

The basis for assuming a reactor coolant pump trip time of 5 minutes for intermediate breaks.


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

48 < Assembly Average Burnup ≤ 49 GWd/MTU


Question: What is TSTF-439 in relation to nuclear licensing?

TSTF-439, "Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO," Revision 2, allows licensees to eliminate the second completion time requirements back to the discovery of a failure to meet a Limiting Condition for Operation (LCO).

The Columbia Generating Station is an important asset in the Pacific Northwest that generates clean energy through nuclear fission.


Question: The system utilizes austenitic stainless steel piping due to its compatibility with reactor coolant under recirculation operation.

The Containment Spray System piping is austenitic stainless steel. This is required because the piping contains reactor coolant during recirculation operation.

After initiation of the ECCS recirculation phase, in conjunction with an enable permissive from the Containment Pressure Control System, Containment Spray pump suction is manually aligned to the Containment sump and one train of Containment Spray is manually initiated.


Question: ARITA methodology in nuclear safety assessments

For each analyzed event, the applicability range of the ARITA methodology shall be limited to the range over which its constituent computational codes and models have been assessed and validated to provide acceptable predictions of relevant phenomena and processes. (Reference to Section 5.1 of Reference 1)

The investigation concluded that additional efforts are needed to enhance the efficiency of the cooling system.


Question: nuclear regulatory compliance

The Commission has concluded, based on the considerations discussed above, that: 1. there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 2. there is reasonable assurance that such activities will be conducted in compliance with the Commission’s regulations, and 3. the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: - J. Medoff, NRR - J. Budynski, NRR


Question: The NRC emphasizes the importance of effective groundwater treatment technologies such as reverse osmosis in the restoration of production zones.

With the reverse osmosis techniques, injection and extraction operations continue at the facility except produced water will be processed through a RO unit which produces a deionized fluid for reinjection.

Key elements of the restoration demonstration will be as follows:


Question: environmental audit schedule

The Turkey Point Nuclear Station Environmental Audit Schedule includes a series of planned virtual meetings between the NRC and Florida Power & Light Company to discuss environmental compliance and safety assessments.

The virtual tours and meetings are designed to facilitate collaboration and information sharing among the subject matter experts.


Question: What is the purpose of detector calibration in nuclear instrumentation?

Calibration of the Ludlum 2221 was performed to ensure accurate measurements of radiation levels and to establish efficiency with Sr-90.

The physical condition of the calibration equipment was recorded as 'Good' after a series of checks including battery and response verification.


Question: nuclide activity

The table lists nuclide activity with specific weights and uncertainties, providing a comprehensive analysis of various nuclides and their respective measurements.

The errors quoted at 2.000 sigma indicate the level of uncertainty associated with the measurements.


Question: fuel assembly burnup requirements

For fuel assembly average burnup greater than 45 GWd/MTU, cool time tables have been revised to account for a 5% margin in heat load.

Certificate of Compliance No. 1031  B2-80 Amendment No. 13, Revision No. 1


Question: Will climate change influence the radiological dose exposure to nearby populations due to the transport of nuclear materials?

Will expected climatological changes affect the radiological dose to the population in the region of interest due to transportation of radioactive materials?

Will expected climatological changes affect average environmental risks of severe accidents due to either changes in severe accident probabilities or associated consequences?


Question: What is the significance of monitoring extravasation in radiopharmaceutical treatments?

There have been several reports by clinics demonstrating this improvement from monitoring.

A 40-year pass for infiltration exposure that could be higher than a dose now requiring reporting for a spill of the radioisotope on the body seems wrong.


Question: HPCI recovery after maintenance

Restoration of Steam Supply to HPCI Turbine after system maintenance

Radiation Monitoring operates continuously during plant operation


Question: Measurements of gamma radiation levels were recorded using cpm as the unit of measurement.


Question: What is a Clathrate Gas Detector and its purpose?

A Clathrate Gas Detector is a specialized device used for detecting the presence of clathrate compounds, which are solid structures formed from gas and water under specific conditions.

LFE Industrial Systems Corp. has developed various industrial systems for diverse applications in engineering, contributing to advancements in safety and efficiency.


Question: flow comparator function

The flow comparator >10% difference function only causes a rod block.

The outputs of Flow Units B and D are compared and if they differ by more than 10%, they result in a rod block.


Question: Main Feedwater Pump failure response

Main Feedwater Pump can fail, in which case it is necessary to bring plant to safe shutdown conditions if shutdown trip limits are exceeded using Auxiliary Feedwater System.

Any failure of pump or valves requires isolation to repair and use a redundant pump.


Question: The Nuclear Regulatory Commission oversees the evaluation of safety systems in nuclear plants.

The safety systems in a nuclear power plant are designed to prevent accidents and mitigate their consequences, ensuring compliance with regulatory standards.

The water supply in Pickens County is managed by local authorities to maintain a balanced system for the community's needs.


Question: The regulation of medical imaging products involves various stipulations regarding their safety and efficacy.

Bristol-Myers Squibb Medical Imaging offers a range of products designed for safe and effective medical imaging applications.

The product codes listed above are part of a broader inventory for pharmaceutical products and do not pertain to specific medical imaging technologies.


Question: What is the role of the Division of Operating Reactor Licensing in nuclear regulation?

The Division of Operating Reactor Licensing is responsible for the review and approval of applications for licenses to operate commercial nuclear power plants, as well as modifications and renewals of those licenses.

The team may also interview the licensee’s subject matter experts using video and teleconferencing if necessary.


Question: The mass flow rate of the system is critical for understanding the performance during operational events.

The Intact Loop Accumulator Mass Flow Rate during Blowdown is a crucial parameter that affects the reactor's safety and operational efficiency.

Temperature measurements should be accurately recorded to ensure the system operates within its design limits.


Question: What are the requirements for implementing ASME Section XI Code Cases?

The regulation at 10 CFR 50.55a(b)(5) provides requirements related to the implementation of ASME Section XI Code Cases.

A broad spectrum of stakeholders participates in the ASME process to help ensure consideration of the various interests.


Question: flood protection measures for dam failure

The interim mitigation and monitoring plan was based on the fact that the U.S. Army Corps of Engineers (USACE) Upper Missouri Reservoir System has a very large storage capacity and that it is both unlikely and predictable in advance that the reservoirs will reach the hydrologic or sunny-day reservoir levels over the course of a water year.

Several yearly forecast conditions are published by USACE for their mainstem dams.


Question: What is the significance of safety licensing processes in nuclear facilities?

The NRC reviews and approves applications for licenses, permits, and certifications to ensure that they meet safety and regulatory requirements essential for the operation of nuclear power plants.

Operating a nuclear plant involves various technical challenges and considerations around energy efficiency.


Question: What is the weighted mean activity of a nuclide in pCi/grams?

The weighted mean activity (pCi/grams) is calculated based on the individual activities of isotopes considering their respective uncertainties.

The errors quoted at 2.000 sigma indicate the confidence interval for the measurements of the nuclides listed in the document.


Question: once-through cooling system

No statute authorizes NRC not to comply with NEPA on equitable grounds; CLI-22-6, 95 NRC 111 (2022)

Prior to having an admitted contention, there is no adjudicatory discovery available by which petitioner could obtain related information; LBP-22-1, 95 NRC 49 (2022)


Question: What are the required details for devices that fall under regulatory registration?

A device subject to registration must include the Manufacturer Name, Initial Transferor Name, Device Model Number, Device Serial Number, acquisition details, and specified isotopes with their respective activity and unit of measurement.

The completion of the application should be handled promptly to ensure compliance with regulatory standards.


Question: The analysis of radioactive materials in soil samples is essential for ensuring compliance with environmental safety standards.

SW846 3050B/6020A Uranium Soil "Dry Weight Corrected"


Question: reactor coolant pressure boundary

contention that continued operation violates 10 C.F.R. Part 50, Appendix A, Criterion 14 because the reactor coolant pressure boundary has not been tested is inadmissible; CLI-22-5, 95 NRC 97 (2022)

NRC cannot require members of the public to parse language in a regulatory analysis or glossary to discern that the agency is making a major shift in the scope of an EIS and rulemaking; CLI-22-2, 95 NRC 26 (2022)


Question: What are groundwater protection programs in nuclear plants?

GWPPs have been established at all operating nuclear power plants to minimize the potential impact from any inadvertent releases.

Surface water consumption for non-cooling water-related operational activities is limited to such uses as facility and equipment cleaning.


Question: What is the purpose of the Evaluation Model Development in EMDAP?

The Evaluation Model Development is a crucial aspect of EMDAP Element 3, which focuses on creating a structured evaluation model that adheres to regulatory requirements and safety analysis standards.

The sample plant description provides insights into the overall design and operational factors considered during the initial licensing phase.


Question: What procedures are involved in nuclear reactor licensing?

The Office of Nuclear Reactor Regulation oversees the licensing process of nuclear reactors to ensure compliance with safety standards and regulations.

The Chief Information Officer is responsible for managing the agency's information technology systems and ensuring their security.


Question: The NRC oversees the regulatory framework for reactor licensing.

The NRC's licensing processes are critical for ensuring that nuclear facilities operate in compliance with safety standards and regulatory requirements.

Hi Viz, LLC offers a diverse range of shooting systems and products tailored for various applications in the market.


Question: Main Steam Line Tunnel temperature monitoring

The Main Steam Line Tunnel Temperature—High Function receives input from 16 channels, four for each of the four tunnel areas. The logic is arranged similar to the Main Steam Line Flow—High Function. One channel from each steam tunnel area inputs to one of four trip strings. Two trip strings make up a trip system and both trip systems must trip to cause an isolation.

The Drywell Radiation—High Function receives input from two radiation detector assemblies each connected to a switch. Each switch actuates two contacts. Each contact inputs to one of four trip strings.


Question: The regulation requires that meteorological data be accurately recorded and analyzed for safety assessments.

The reactor’s cooling system is designed to efficiently transfer heat from the core to the environment.


Question: How will climate change affect air quality around nuclear plants?

Will expected climatological changes affect the sources, types, and estimates of annual air emissions from the operating plant and transmission lines?

Will expected climatological changes affect cooling system impacts from the operating plant on local weather, including plume lengths, additional hours of fogging and icing, and salt deposition?


Question: What does a Marinelli beaker indicate in nuclear sample analysis?

**Energy Calibration Used Done On:** 7/8/2014


Question: What is the weight mean activity of nuclides?


Question: What is the importance value in nuclear regulatory documents?

EK2.02 Importance value needs to be moved to right

4.2-49 Missing KA AK 1.01 Booster/main feedwater pumps 2.4 value


Question: The NRC evaluates laboratory tests to determine the characteristics of undisturbed soil samples for various depths.

Summary of Laboratory Testing of Undisturbed Samples from Boring P-15

Sample Number | Elevation (ft) | Depth (ft) | USCS Group Symbol | Fines (%) | Water Content (%) | Dry Unit Weight (pcf) | Void Ratio | Saturation (%) | Specimen Diameter (in.) | (q_u) (kef) | (u_f) (%) | (t_f) (kef) | Mode of Failure | Condition of Tube Cutting Edge


Question: What is the activity measurement in nuclear licensing?

Distributor/Distributed By: Thermo Process Instruments, L.P


Question: The NRC mandates that all design changes undergo rigorous evaluation to confirm adherence to safety standards and design requirements.

Specifically, the licensee failed to perform an engineering change evaluation in accordance with MFP-QA-1220, “Engineering Change Process Handbook,” to ensure that the design changes were acceptable for Susquehanna design requirements.

In 2013, AREVA recommended and performed changes to the Cutler Hammer Type DHP-VR 4.16kV breakers that they provided to the licensee.


Question: inspection duration and frequency in nuclear regulation

Consider the industry-proposed inspection duration, frequency, and team complement.

Incorporate principles within the ADVANCE Act (Section 507) focused on an innovative, efficient, and effective cycle.


Question: The operability of accumulators in the ECCS is critical for maintaining safety in various operational modes.

**LCO 3.5.1** Four ECCS accumulators shall be OPERABLE.

Several actions must be taken if an accumulator is inoperable due to conditions outside of boron concentration limits.


Question: What are the requirements for alternative examinations of the torus metallic liner?

Duke Energy Progress, LLC (Duke Energy, the licensee) requested an alternative (request RA-22-0308) to certain requirements of the 2007 Edition through 2008 Addenda of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Subsection IWE to conduct alternative visual examinations of the torus metallic liner at Brunswick Steam Electric Plant, Unit Nos. 1 and 2 (Brunswick).

The audit team will review the licensee’s records, analyses, and calculations related to proposed alternative examinations required by ASME Code, subsection IWE, for visual examination of the metallic liner surrounding each torus suppression chamber (torus) of Brunswick, Unit Nos. 1 and 2.


Question: What is the purpose of the barge facility in nuclear plant operations?

An upgraded barge facility could eventually be used to facilitate transport of spent fuel canisters offsite for disposal in a national repository.

Section 3.2.1 of the ER states that the meteorological facility consists of a 91-meter (300-foot) instrumented tower for wind and temperature measurements, a separate 10-meter (33-foot) tower for dew point measurements, a ground-based instrument for rainfall measurements, and a data collection system in an instrument building.


Question: nuclear reactor safety licensing process

The licensing process for nuclear reactors is governed by regulations that ensure safety and compliance with standards set by the NRC.

The financial implications of nuclear energy development can significantly impact investment decisions and project timelines.


Question: What are the methodologies for evaluating loss-of-coolant accidents in nuclear reactors?

The proposed changes would update the list of approved methodologies in MPS3 TS 6.9.1.6.b to establish the core operating limits included in the Core Operating Limits Report for the GAIA fuel with M5 cladding.

The licensee responded to this request on April 1, 2024 (ML24093A216).


Question: laser power measurement techniques

All laser beam spot size measurement techniques (CCD, aperture, and knife-edge) are discrete measurements, and can only be obtained prior to and after a build, and can only be measured perpendicular to the laser beam (i.e., in one location on the build plate directly below the laser).

When discussing mirror control, the distinction between dimensional control and location control is important to note.


Question: core inlet and outlet mass flow rate

At time 0s: ~40000 (Core Inlet), At time 5s: ~20000 (Core Inlet), At time 10s: ~0 (Core Inlet), At time 15s: ~-10000 (Core Outlet), At time 20s: ~-15000 (Core Outlet), At time 25s: ~-20000 (Core Outlet)

Solid Line: Core Inlet, Dashed Line: Core Outlet.


Question: Each shipment of thoriated tungsten welding rods will include important safety notices required by industry standards.

Thoriated tungsten welding rods are manufactured in China and imported in sealed retailed type package, typically three or ten welding rods of a given length, usually 3 inches or 7 inches, per packet.

Radiological training is not applicable under the distribution license since the use and possession of thoriated tungsten welding rods are exempt pursuant to 10 CFR 40.13(c)(1) and equivalent agreement state regulations.


Question: What are the risks involved in policy changes for nuclear licensing?

We really like that term, because what we get is the same thing that we saw with the three flow charts that were discussed this morning.


Question: The NRC sets limits for radioactive contamination based on disintegrations per minute recorded at various measurements.

Disintegrations Per Minute Per 100 Square Centimeters (dpm/100 cm²) is a measure used to quantify the specific activity of radioactive materials in wipe samples.

The table outlines sample identifiers along with corresponding disintegration rates, which are crucial for assessing contamination levels.


Question: The NRC emphasizes the importance of fostering a safety culture among organizations involved in regulated activities.

The NRC's Safety Culture Policy Statement became effective in June 2011. While a policy statement and not a regulation, it sets forth the agency’s expectations for individuals and organizations to establish and maintain a positive safety culture.

In accordance with 10 CFR 2.390 of the NRC’s 'Agency Rules of Practice and Procedure,' a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s document system (ADAMS). ADAMS is accessible from the NRC Web Site at: [NRC Reading Room](http://www.nrc.gov/reading-rm/adams.html).


Question: background reference area

Background Reference Area - Boring 2 - 5-10 ft bgs


Question: What is the purpose of LCO in ECCS flow management?

The intent of the LCO limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient ECCS centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2).

In addition, 2 psid is added to accommodate instrument uncertainty in the pressure drop measurement.


Question: What is counting uncertainty in nuclear testing?

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

The parameter for Detection Limit (DL) indicates the lowest concentration of a substance that can be reliably measured by an analytical procedure.


Question: What analysis techniques are used for Framatome fuel properties in nuclear safety assessments?

The pressure drop can impact the core flow due to the presence of different types of fuel assemblies. Consistent with the RLBLOCA EM hydrodynamic modeling features (Section A.1.3.6.2.4 of EMF-2103(P)(A) Revision 3),

Similarly, the embedded fuel rod code models and cladding models in the S-RELAP5 calculation in ANP-4032P were specific to Framatome fuel properties and design.


Question: suction pressure test gauge

INSTALL suction pressure test gauge at test tee between 1-PI-62-109 and 1-ISIV-62-352C as follows:

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


Question: The fuel handling system's safety features must ensure proper operation during mishandling events.

See Section 9.1.4.2.2, under Fuel Handling Bridge Crane, and Section 9.1.4.3.1, for a discussion of Regulatory Position 3. Note that, “the spent fuel crane has identical safety features to those on the manipulation crane with the exception that the spent fuel crane does not have an RCC hoist system.”

Each unit of the McGuire Station has an independent fuel storage system. The Fuel Handling System associated with the pool is discussed in Section 9.1.4, and Spent Fuel Cooling System is presented in Section 9.1.3. Radiation shielding and monitoring are presented in Sections 12.1 and 11.4, respectively.


Question: What are the requirements for Quality Assurance Program Descriptions (QAPDs) in nuclear applications?

Quality Assurance Program Descriptions (QAPDs) must meet specific requirements outlined in regulatory documentation to ensure consistency and reliability in nuclear reactor operations.

The flexibility of design and operation for various reactor types allows for innovation in energy production, influencing the overall efficiency of the energy market.


Question: core average temperature in nuclear reactors

The graph shows the core average temperature, stabilizing around 570 °F over time.


Question: sensitivity studies in nuclear regulatory submissions

Clarification on applicability and conclusions to be drawn from sensitivity studies. For example, on page 5-17, WCAP-18850-P appears to make conclusions concerning “2-Loop PWRs.” While the actual context and analysis appears to be indicative of Westinghouse 2-Loop pressurized-water reactors, in a literal sense, other pressurized-water reactors within the proposed scope of WCAP-18850-P could arguably be categorized as having two reactor coolant system loops (i.e., Combustion Engineering plants).

Particular focus is intended upon the validation of the modeling approaches and the completion of appropriate validation analyses on both a separate effects and integral basis for the full suite of validation comparisons.


Question: rod position sequences

The analyzed rod position sequence will be established pursuant to the approved methods controlling normal plant operations.

The proposed change ensures that analyzed rod position sequences are developed to minimize incremental control rod reactivity worth in accordance with the Reference 1 NRC approved methodology implemented in Susquehanna TS 5.6.


Question: What are the limitations and conditions of the GALILEO methodology?

There are four L&Cs imposed on the GALILEO methodology (Section 5.0 of the SER of Reference 3). The method of adherence for each L&C is described below.

The NRC staff does not approve the ability to make changes to the model parameters without NRC approval.


Question: licensing actions during IMPEP review

List of all licensing actions completed during the review period (sorted by license reviewer, if possible).

Please have the following information available for use by the IMPEP review team when they arrive at your office:


Question: What is the significance of the Materials Licensing Section in nuclear regulatory processes?

The Materials Licensing Section is responsible for the oversight and regulatory licensing of materials used in the nuclear industry, ensuring that materials are handled and controlled in compliance with safety standards.

The medical physics consultants in this context offer services related to radiation therapy and diagnostic imaging, focusing on optimizing treatment plans and ensuring patient safety.


Question: automatic turbine trip response

The closure of the MSIVs leads to a pressurization transient with the corresponding increase in reactor power.

The event is initiated at time one second, when the TCVs and TBVs start to open.


Question: remainder coolant system operation

Two RCS loops shall be OPERABLE, and either: - Two RCS loops shall be in operation when the Rod Control System is capable of rod withdrawal; or - One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

All reactor coolant pumps may be removed from operation for ≤ 1 hour per 8 hour period provided: - No operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SDM of LCO 3.1.1; and - Core outlet temperature is maintained at least 10°F below saturation temperature.


Question: definition of Non-light-water reactor

- **Non-light-water reactor** means a nuclear power reactor using a coolant other than light water.

- **Small modular reactor** means a power reactor, which may be of modular design as defined in § 52.1 of this chapter, licensed under § 50.21 or § 50.22 to produce heat energy up to 1,000 megawatts thermal per module.


Question: revisions to the RBS Attachment

This section of the Parish Administrative Procedure addresses the revisions to the Enclosures of the RBS Attachment to the LPRRP. All information contained within the RBS Attachment and Enclosures will be reviewed annually by the Parish OHSEP.

When a revision is made to the RBS Attachment and Enclosures, any and all EIPs that may be affected by the revision are revised.


Question: The criteria for initial enrichment levels for spent nuclear fuel assemblies in the BWR reactor design are detailed in the documentation.

| 89-Assy | Max. Initial Enrichment (wt % 235U) |


Question: gamma spectrum analysis

The gamma spectrum analysis was performed using a 500 ml Marinelli geometry, which is standard for assessing the radioactive content of samples.

The sample was taken on 8/23/2019 at 8:08:00 AM, and the acquisition started on 8/26/2019 at 9:12:31 AM.


Question: What is the significance of surveyed building locations in nuclear regulatory filings?

Figure 1 shows the satellite view of surveyed building locations which are crucial for the assessment of spatial relationships in compliance with regulatory standards.

The table presents numerical data associated with surveyed points, which might be used for various analyses in the regulatory process.


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

Activity (pCi/grams) indicates the strength of a radioactive source, which is critical for ensuring compliance with safety regulations in nuclear operations.

The nuclide name table displays various isotopes with distinct energy levels and yields, serving as a reference for identifying radioactive materials.


Question: What are the requirements for evaluating changes to the emergency plan in nuclear facilities?

The inspectors identified multiple examples of a Severity Level IV non-cited violation of 10 CFR 50.54(q)(3), for changes to the licensee’s emergency plan associated with protective action recommendation (PAR) and emergency classification procedures, for failures to demonstrate that the changes would not reduce the effectiveness of the plan.

The inspectors reviewed the licensee’s implementation of its corrective action program related to the following issues: Unit 1, leakage past emergency core cooling system check valves, Unit 2, leakage past emergency core cooling system check valves, Unit 1, train B of high head safety injection pump circuit breaker, Qualified display parameter system failures.


Question: What is the purpose of the Aging Management Audit in nuclear licensing?

The Aging Management Audit will be conducted primarily using telecommunications from June 17, 2024, to January 31, 2025, with potentially a portion of the On-Site Audit (Aging Management Audit Part #2), as determined appropriate, conducted at the applicant’s facility.

Entrance and exit briefings will be held at the beginning and end of the audit via telecommunications.


Question: What is the maximum decay heat for the 61BTH Type 2 DSC?

| Max Decay Heat per DSC (kW)   |        |        |        |        |        | 19.4   |

The NRC requires thorough evaluation of all safety-related systems to ensure compliance with regulatory standards.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors is governed by the rules established by the Nuclear Regulatory Commission (NRC), which evaluates applications for construction and operation to ensure public safety.

The complexity of neutron scattering experiments requires precise calibration of instruments and careful control of experimental conditions to obtain accurate results.


Question: What materials are used for the construction of the Off-Gas Treatment Building?

The materials of construction for the Off-Gas Treatment Building components are: Concrete, Concrete Block.

The Equipment Access Lock components are exposed to the following environments: Air - indoor uncontrolled, Air - outdoor.


Question: Transportable Storage Canister conditions

A. TSC cavity vacuum drying pressure limit not met. A.1 Perform an engineering evaluation to determine the quantity of moisture remaining in the TSC.

B. TSC helium backfill density limit not met. B.1 Perform an engineering evaluation to determine the effect of helium density differential.


Question: harvesting materials from nuclear power plants

The US NRC would like to propose one more presentation for this autumn's PLiM conference, on harvesting of materials from operating and decommissioning power plants.

I understand that the document must adhere to strict regulatory standards to ensure safety and compliance with nuclear energy policies.


Question: The impact of constructed structures on bird populations near nuclear facilities was explored in the NRC's evaluations.

In the 2013 license renewal GEIS, the NRC reviewed the impact on avian mortality from birds colliding with cooling towers and transmission lines. It evaluated the primary literature for avian collision mortality associated with all types of constructed objects, as well as the results of monitoring studies conducted at six nuclear plants.

However, during the summer months, supplemental cooling is provided by three mechanical draft helper towers along the hot and cold canals.


Question: The report details various measurement intervals for gamma and GFP radiation during the sampling process.

### Gamma Measurement (in sleeve):
- 10-11 ft interval: 6,800 cpm
- 11-12 ft interval: 6,800 cpm
- 12-13 ft interval: 6,900 cpm
- 13-14 ft interval: 6,900 cpm
- 14-15 ft interval: N/A

**Sampling Location:** Pickard Hall
**Boring Number:** 7
**Date:** 8-7-20
**Time:** 13:27


Question: graphite material properties

| Property                        | Values*                    |
|---------------------------------|---------------------------|
| Grain size, µm                 | 40µm max, 15µm avg       |
| Bulk density, g/cm³            | 1.75                      |
| Electrical resistivity, µΩm     | 14                        |
| Thermal conductivity, W/mK      | 104.4                     |
| Coefficient of Thermal Expansion, x10⁶/°C | 3.8            |
| Tensile strength, MPa           | 34.3                      |
| Flexural strength, MPa (3 points) | 58.8                     |
| Compressive strength, MPa       | 98.0                      |
| Elastic modulus, GPa            | 10.8                      |
| Ash content, ppm                | [[    ]]                 |
| Boron Content, ppm              | [[    ]]                 |

The reactor core requires proper shielding to ensure that radiation levels remain within safety limits during operation.


Question: What are underground injection control (UIC) wells?

This condition requires a listing of the Homestake underground injection control (UIC) wells, their status and location.

The document is addressed to various regulatory agencies and highlights compliance with environmental standards.


Question: What is required for nuclear plant licensing?

The NRC requires that licensees submit a comprehensive application that includes a description of the facility, proposed operations, and compliance with relevant safety regulations.

The Vice President of R. E. Ginna Nuclear Power Plant oversees operations and ensures the plant meets environmental and safety standards, but does not directly engage in the licensing process.


Question: import license requirements for nuclear materials


Question: The NRC indicated that various documents were submitted over the years regarding the licensing of the Crownpoint Uranium Project.

The purpose of the CONSOLIDATED OPERATIONS PLAN (COP) was to extract and combine the information in previously submitted documents into one consolidated specification report.

The processing and drying facility will be located at the Crownpoint Central Plant (CCP).


Question: nuclear reactor safety system

In a nuclear reactor, the safety system is designed to prevent the release of radioactive materials in the event of an incident.

The diversity of organisms in marine ecosystems is crucial for maintaining ecological balance.


Question: What are the requirements for the LMTC Neutron Shield in nuclear plant operations?

If draining the LMTC Neutron Shield is required to meet the plant architectural limits, the LMTC Neutron Shield shall be verified to be filled after completion of the critical lift.

A training program for the MAGNASTOR® system shall be developed under the general licensee’s systematic approach to training (SAT).


Question: The facility includes multiple storage areas for gauges and construction materials.

Gauge Storage is crucial for ensuring the accurate measurement and monitoring of material quantities during construction projects.

Asphalt Lab is equipped to analyze the composition and quality of asphalt samples for road construction.


Question: The storage design of new fuel must ensure that criticality is managed properly through careful arrangement of components.

In accordance with BFN Technical Specification 4.3.1.2, “The new fuel storage vault shall not be used for fuel storage. New fuel shall be stored in the spent fuel storage racks.”

Servicing equipment is provided to facilitate refueling, fuel inspection and fuel maintenance.


Question: The NRC evaluates information from licensees to ensure compliance with established safety standards during reactor construction.

The NRC requires licensees to analyze emergency zones to verify compliance with emergency preparedness regulations.

The status of submissions and notifications varies, with some classified as 'Sufficient Information' while others remain 'Not Received'.


Question: reactor coolant pressure boundary

5.2 Integrity of Reactor Coolant Pressure Boundary includes critical evaluations and compliance checks to ensure the safety and integrity of the reactor coolant system.

4.6.2 Evaluations of the CRDS focuses on the performance and testing of control rod drive systems, crucial for managing reactivity within the reactor core.


Question: What is the Probable Maximum Flood (PMF) in nuclear regulation?

The PMF is the hypothetical flood (peak discharge, volume, and hydrograph shape) that is considered to be the most severe reasonably possible, based on comprehensive hydrometeorological application of PMP and other hydrologic factors favorable for maximum flood runoff such as sequential storms and snowmelt.

The precipitation for Alternatives 1, 2, and 3 for Rock Run Creek is determined in: - Calculation PEAS-FLOOD-02, "BDBEE – Flood Re-Evaluation – Rock Run Creek Probable Maximum Precipitation (PMP) and Local Intense Precipitation (LIP)" (Reference 21).


Question: SAMA cost-benefit analysis in nuclear regulatory filings

Although an increase in risk could change the cost benefit values in a SAMA analysis. NRC regulations only require the SAMA analysis to be performed once. The 2024 LR GEIS Rule states, in part, that severe accident mitigation alternatives do not warrant further plant specific analysis because the demonstrated reductions in population dose risk and continued severe accident regulatory improvements substantially reduce the likelihood of finding cost-effective significant plant improvements. Palisades demonstrated the reduction in population dose risk in Table E.3-1 of the 2024 LR GEIS.

Climatalogical changes might affect the average environmental risks of severe accidents because of changes in either severe accident probabilities due to an increase in the rate of severe natural phenomena and/or associated consequences due to altered patterns of atmospheric dispersion. While the potential severity of storms and other natural phenomena might increase, nuclear power plants must be designed to withstand all credible natural events at the site of concern.


Question: What is the significance of harvested materials from decommissioned nuclear plants?

Ex-plant materials are valuable because they have been exposed to actual in-service plant operating conditions.

With plants shutting down both in the U.S. and internationally, there are increasing opportunities to harvest components from decommissioning plants.


Question: What is the significance of 10 CFR Part 21 in regard to nuclear plant safety?

The following table reflects the status of US nuclear plants to which SF1154 fluid from Batch 17BLVS293 was shipped by Lake Engineering Company (LEC) in accordance with purchase orders which specified that the requirements of 10 CFR Part 21 are applicable.

This letter is a follow up to our previous notification letter dated October 6, 2020 pertaining to the above subject.


Question: The classification process must occur quickly, typically within a set time frame after an emergency is identified.

Classifies the event less than 15 min. of the Event Initiation Time.


Question: What are the safety features related to fuel handling accidents in nuclear facilities?

The probability of a fuel handling accident is very low because of the safety features, administrative controls, and design characteristics of the facility, as previously mentioned.

In addition to the area and effluent radiation monitors, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations.


Question: What are the requirements for the Reactor Vessel Material Surveillance Program?

10 CFR Part 50, Appendix H, “Reactor Vessel Material Surveillance Program Requirements,” requires that peak neutron fluence at the end of the design life of the vessel will not exceed 1.0 x 10¹⁷ n/cm² (E > 1.0 MeV), or that reactor vessel beltline materials be monitored by a surveillance program.

The tanks within the scope of this AMP are: EDG fuel oil day tanks, emergency diesel fuel oil pump head priming tanks, diesel driven fire pump fuel oil storage tanks, portable diesel electric generator fuel oil tanks, and the portable caddy fuel oil tanks.


Question: What is the target mean annual performance goal for nuclear plants regarding core damage risk?

The target mean annual performance goal for nuclear plants of mean \( 10^{-5} \) per yr is achieved by coupling site-specific DRS with the deterministic seismic design criteria and procedures specified by NUREG-0800.

The procedure for derivation of the DRS is as follows.


Question: What is a SAMA category in nuclear regulatory context?

A SAMA Category identifies specific areas of concern in nuclear safety assessments, such as A: AC/DC, B: Core Cooling, C: FW/Condensate, and others, where various risk management strategies may be applied based on the plant's operational characteristics.

Dresden has integrated FLEX and B.5.b equipment into the DEOPs such that it may be used as an alternate injection source when the operators deem it to be useful.


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

The quench spray system is designed to decrease the temperature and pressure within the containment structure by spraying water onto steam to ensure that any potential release of radioactive materials is contained and effectively managed.

Figure 6.2-71 presents the Containment Temperature Outside RS Pump NPSH Available Analysis as part of a series of evaluations related to the pump's operational capacity.


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

Energy Flow is measured in units of million BTU per second (10⁶ BTU/Sec) and is crucial for assessing the performance and safety of nuclear reactor operations.

The table shows a sequence of measurements relating to mass flow and energy flow values at various times with specific numerical data.


Question: The MDA value indicates the minimum detectable activity for a nuclide in a given sample.

Nuclide MDA (pCi/grams) refers to the minimum detectable activity of a nuclide in terms of picoCuries per gram.

The energy levels of the listed nuclides provide insights into their potential applications in radiation therapy.


Question: The process of evaluating the transfer mechanism of the refueling machine is essential for operational integrity during fuel handling.

To verify the ability of the refueling machine, new fuel elevator, fuel transfer car, spent fuel bridge crane, spent fuel cask handling crane, and associated fuel handling tools to transfer a dummy fuel assembly.

Verify the operability of the control circuits and associated interlocks of the cranes and elevator.


Question: What is the maximum allowable activity concentration for tritium in the Waste Gas Decay Tank?

The 500 curie limit for Xe-133 equivalent activity and 5 curie limit for tritium have been converted to µCi/cc based on the volume of a full WGDT.

**IF** discharge path is 1-RE-90-400 and is operable, THEN **OBTAIN** the averaged µCi/s background for 1-RE-90-400A from ICS by entering 'CHEM4' on the yellow bar AND **RECORD** on Attachment 1.


Question: The NRC suggests that small aircraft impacts may pose a heightened accident risk due to their commonality and weaker security measures compared to large aircraft.

In our opinion, small aircraft impacts may rise to a higher level of accident risk because such aircraft are relatively common, with security measures somewhat less robust than those employed with large commercial aircraft.

The likelihood of a large commercial aircraft impact then becomes some form of accident, as opposed to a malicious act.


Question: What is the Risk Informed Completion Time (RICT) program in nuclear regulation?

A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.


Question: structural support in nuclear power plants

Equipment supports and foundations provide structural support and are made of reinforced concrete, particularly in air - indoor, uncontrolled environments, where they may face aging effects such as cracking, loss of bond, and loss of material.

Masonry walls, often built with concrete blocks, are important for shelter and protection in indoor environments and may experience issues such as cracking over time.


Question: For pressurized water reactors, it is essential to maintain a minimum level of dissolved boron in the system to ensure safety protocols during operations involving fuel loading or unloading.

LCO 3.2.1 The dissolved boron concentration in the water in the PWR TSC cavity shall be greater than, or equal to, the concentration specified in Appendix B, Table B2-4. A minimum concentration of 1,500 ppm is required for all PWR fuel types. Higher concentrations are required, depending on the fuel type and enrichment.

The nuclear facility's security measures include multiple layers of physical protection and personnel screening to prevent unauthorized access.


Question: To maintain safety, the plant layout includes control measures for access to different radiation zones, which may vary during abnormal occurrences to limit personnel exposure.

In accordance with Southern Nuclear Operating Company (SNC) policy, and consistent with the recommendations of Regulatory Guides 8.8 and 8.10, the radiation exposure of plant personnel will be kept as low as reasonably achievable (ALARA) by means of the health physics program discussed in section 12.5. The radiation protection policies and practices contained therein are initiated through the training program discussed in section 13.2 and through the plant procedures manual discussed in subsection 12.1.1 and section 12.5.

Providing labyrinth entrances to radioactive pump, equipment, and valve rooms. Adequate space is provided in labyrinth entrances for easy access. Highly radioactive passive components with minimal maintenance requirements are located in completely enclosed compartments and are provided with access via a shielded hatch or removable blocks.


Question: The regulatory frameworks guiding nuclear energy facilities are outlined by the Atomic Energy Act.

The Atomic Energy Act (AEA) provides the framework for the regulation of atomic energy and establishes the Nuclear Regulatory Commission (NRC) to oversee the licensing and regulation of nuclear facilities.

The American Society of Mechanical Engineers (ASME) is a professional organization that develops codes and standards for mechanical engineering.


Question: soil classification for nuclear applications

SAND (SW), tan, fine to medium grained, trace silt, trace mica, well graded, saturated.


Question: environmental review process

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

This meeting will not be recorded or transcribed.


Question: The requirements for determining the Levels of Detail (LOD) must be accurately defined in the context of safety assessments.

Low LOD based on picture. Replace Q to address KA.