Question: inservice testing program alternative requests

This submittal includes requests for alternatives to the inservice testing program for pumps and valves, specifically for the RHR and Core Spray Pump suction pressure at Edwin I. Hatch Nuclear Plant - Units 1 and 2, as outlined in 10 CFR 50.55a(z)(2).

The inspection has been completed, and the performance metrics have shown a satisfactory operational efficiency across all systems.


Question: steam supply valve safety functions

Steam supply valve HV-3019 from SG 2 to TDP P4-001 turbine, normally open dc-powered MOV, train A

Check valve 1301-U4-404 in steam supply line to TDP turbine from SG 1


Question: The qualification of the Reactor Seismic Isolation System is crucial for ensuring safety during seismic events.

The Reactor Seismic Isolation System must be qualified in accordance with industry-related codes and standards to ensure its effectiveness during seismic conditions.

The efficiency of the turbine system can significantly affect the overall performance of the power plant, impacting electricity generation.


Question: What are the power supply requirements for the isolation valves in the Ice Condenser System?

The applicant is required to have knowledge of the power supply to the Train A Glycol Return Containment Isolation Valve, 1-FCV-61-193, GLYCOL FROM AHU OUTSIDE CIV ØA.

The Electrical Power Supply systems must be maintained to ensure all critical safety functions remain operative during all conditions.


Question: Prior to granting a license, the NRC assesses compliance with safety regulations.

Nuclear licensing involves a comprehensive review process where safety, environmental, and technical aspects of the proposed reactor design are evaluated to ensure compliance with regulatory standards.

The reactor's coolant system is designed to maintain optimal thermal efficiency, ensuring that operational limits are not exceeded under normal conditions.


Question: What is the process for reviewing HMC's groundwater model in regulatory audits?

The scope of the proposed audit includes the review of files and documents related to HMC’s groundwater model.

The audit will focus on reviewing the relevant modeling files and documents, as appropriate, for conformance with appropriate supporting documentation guidance and modeling standards.


Question: What is the minimum ECCS flow rate after a reactor trip?

Minimum ECCS flow rates are critical for ensuring that emergency core cooling systems operate effectively during a reactor trip, providing essential cooling to prevent core damage.

The table represents the projected temperature fluctuations over a 24-hour period, indicating how ambient conditions can affect reactor performance.


Question: The NRC assesses the energy levels and corresponding uncertainties of gamma-ray peaks in regulatory filings.

Energy (keV) is measured to determine the peak centroid, providing essential data for safety analysis in nuclear regulatory filings.

Errors quoted at 2.00 sigma relate to statistical methods used in error analysis, not specific to gamma-ray measurements.


Question: dry-loading equipment requirements

The dry-loading equipment included some general requirements that were applicable to all components. These requirements included the following: All lifting and handling equipment was designed to meet the requirements in NUREG-0612 and American National Standards Institute N14.6, "Radioactive Materials—Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More."

Left: Fuel canister lowered into the Model 125-B shipping cask from a fuel transfer cask. Right: Shipping cask loading tower lowered the shipping cask to a horizontal position onto a skid that would be attached to the rail car.


Question: What are power uprate-related licensing submittals in nuclear regulation?

The U.S. Nuclear Regulatory Commission (NRC) is issuing this regulatory issue summary (RIS) to request information on planned power uprate-related licensing submittals for all power reactor licensees.

This RIS is intended to promote early communication with NRC licensees regarding planned power uprate-related licensing activities.


Question: What is the purpose of observation wells in nuclear regulatory context?

Observation wells are key components used to monitor groundwater levels and quality near nuclear facilities, ensuring environmental safety and compliance with regulatory requirements.

The Vogtle Site Area includes various planned structures and existing ones, which are critical for operations and management, but do not specifically pertain to groundwater monitoring.


Question: nuclide activity measurement

The activity of a nuclide is expressed in pCi/grams, which indicates the rate of decay of the radioactive material present per gram of sample.

The reactor design that employs a pressurized water system requires careful consideration of thermal efficiency and reactivity control.


Question: What is the safety function of the AFW pump suction valve HV-5118?

AFW pump suction valve HV-5118, normally closed MOV, train B is crucial for ensuring that the motor-driven AFW pump P4-002 operates effectively.

The methods used for failure detection, including status indications and testing protocols, are fundamental to maintaining the overall system integrity.


Question: What are the quantities listed in the regulatory filings?

The total quantity of items in the document is 433 for Eternal Sun 67 - HP 800 and 425 for Eternal Sun 105 - 800W.

The Unit Price varies between 0.02 and 0.15 across different submissions.


Question: What is meant by spectral acceleration and how is it measured in seismic evaluations?

Spectral acceleration is a measure of the maximum response of a structure to earthquake ground motion as a function of the frequency of the structure. It is typically expressed in units of g (gravitational acceleration).

In a seismic hazard assessment, it is essential to consider the different types of ground motions and how they may vary geographically.


Question: The evaluation of criticality safety examines parameters and conditions that affect the effective multiplication factor in the storage racks.

Calculated values of K_{eff} for the storage arrays, including the effects of calculational and geometrical uncertainties, are less than those required by ANSI N18.2-1973, Section 5.7.4.1 when a full loading of the assemblies listed in assumption #5 below is considered.

The methods, codes, and specifications used in the design of the New Fuel Storage Vault are also given in Section 3.8.4.2.


Question: What is the significance of liquid velocity in nuclear reactor safety analysis?

Liquid velocity, measured in feet per second (ft/s), is crucial in ensuring proper heat transfer within the reactor core and affects the overall thermal hydraulic behavior during operation.

The numerical values indicate the edit frequencies for different steps in the data management process, which are essential to maintaining an efficient control setup.


Question: What is the Westinghouse Cobalt-60 Digital Serialization Approach?

This white paper presents for NRC staff review Westinghouse’s proposed method of compliance with the applicable NRC regulations regarding NSTS tracking of Cobalt-60.

The subject document was prepared and classified as Westinghouse Proprietary Class 2.


Question: What is the purpose of a reactor pressure vessel in nuclear reactors?

The reactor pressure vessel is a critical component that houses the nuclear reactor core and is designed to contain the nuclear reaction while maintaining the necessary pressure and temperature for the reactor's operation.

Radiation shielding is a crucial aspect of nuclear safety, ensuring that harmful radiation is contained and does not leak into surrounding areas.


Question: maximum debris loads in nuclear safety

As shown in the table, the plant debris quantities shown in Table 3.f.7-1 represent the maximum quantity of each debris type for the breaks of 10 inches and smaller.

The strainer flow rate used for the head loss analysis was 9100 gpm, which bounds the maximum strainer flow rate of 8,710 gpm from combining the maximum RHR and CS pump flow rates of 4,760 and 3,950 gpm.


Question: The NRC stipulates that licensees must conduct thorough inspections and testing of safety systems such as diesel generators in accordance with the Technical Specifications.

The various modes of operation of the diesel generator fuel oil system are as follows: 1. manual start 2. automatic start 3. normal shutdown 4. emergency shutdown 5. recirculation.

Safety evaluations for underground storage tanks indicate that they do not need extra protective measures against environmental conditions.


Question: plant-specific aging management programs

The DCPP Water Chemistry AMP (B.2.3.2) and DCPP ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD AMP (B.2.3.1) will be used to manage this aging effect for CASS components.

There is a gas blanket of pressurized N2, thus the vapor content is negligible.


Question: What is the input signal range for the NUSI 2000 Series Four-Input Summer?

Input Signal: 1 V to 5 Vdc normal; ±10 Vdc maximum


Question: How do climatic changes potentially impact the transportation of radioactive substances?

Expected climatological changes can influence the impact level assigned to the transportation of radioactive materials in Chapters 3 or 4.

The transportation of radioactive materials involves complex logistics and regulations that are essential for ensuring public safety.


Question: The sample receipt process includes a list of bottle types that do not undergo preservation checks.

The preservation of the following bottle types is not checked at sample receipt: Orange (oil and grease), Purple (total cyanide), Pink (dissolved cyanide), Brown (arsenic speciation), Sterile (fecal coliform), EDTA (sulfite), HCl preserved vial (organics), Na2S2O3 preserved vial (organics), and HG-1631 (total/dissolved mercury by method 1631).

Client must contact an ACZ Project Manager if analysis should not proceed for samples received outside of their thermal preservation acceptance criteria.


Question: control rod withdrawal mechanisms

Blocks automatic and manual control rod withdrawal


Question: Probable Maximum Flood definition

The Probable Maximum Flood (PMF) is defined as the flood that may be expected from the most severe combination of meteorological and hydrologic conditions that are reasonably possible in the geographical area.

The underlying tectonic structures of the site play a significant role in determining how the area will respond during an earthquake.


Question: What are the specified cladding diameters for nuclear reactor components?

The cladding diameters for nuclear reactor components must meet specific regulatory standards to ensure safety and performance in the reactor environment.

The safety assessments conducted by the NRC include a comprehensive evaluation of operational procedures and maintenance protocols.


Question: What are the required actions if concrete cask average surface dose rates are not met?

A. CONCRETE CASK average surface dose rate limits not met | A.1 Administratively verify correct fuel loading | 24 hours | A.2 Perform analysis to verify compliance with the ISFSI offsite radiation protection requirements of 10 CFR Part 20 and 10 CFR Part 72. | 7 days

B. Required Action and associated Completion Time not met | B.1 Remove all fuel assemblies from the NAC-MPC SYSTEM | 30 days


Question: What are the procedures for simulator initialization in nuclear plant operations?

This JPM has been pre-shot in IC #287. Should IC # 287 be erased or fail to perform as expected then use the following set-up instructions: a. Initialize to IC# 50 b. Insert Remote rmr13 erfds met data source selection to simulator c. Place simulator in run and acknowledge all alarms d. Place simulator in freeze e. After performer indicates understanding of task, place simulator in run.


Question: What is Fuel Bearing Material (FBM) in nuclear licensing?

Fuel Bearing Material (FBM) is any component or pieces of components associated with Three Mile Island Unit 2 (TMI-2) reactor operations that have been contaminated by used (spent) nuclear fuel and or the associated isotopes in used (spent) nuclear fuel.

FBM Vessels fabricated from Stainless Steel, containing no liner, and sealed with mechanical closures containing used filter-media used to extract FBM are approved contents for loading into the MAGNASTOR FBM TSC provided that they have demonstrated to be free of moisture and hydrogen generating materials as described below prior to being sealed within the FBM TSC.


Question: The assessment of aquatic species, including mussels and snails, is a critical aspect of environmental evaluations conducted during the licensing process.

**Mussels** are an important group of organisms that must be assessed to evaluate the ecological impacts of a facility's operation.

The total mean density of benthic macroinvertebrates downstream was reported as 961 per square meter.


Question: significant hazards consideration

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

As a result, plant safety is either improved or unaffected.


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

The NRC requires licensees to submit detailed analyses of the safety features of the reactor system as part of the regulatory review process.

The Arkansas Nuclear One facilities are crucial for the local energy supply and support numerous jobs within the community.


Question: ITAAC notification process

The ITAAC notification process is outlined under the NRC regulations, specifying the conditions under which notifications must be submitted for the construction of approved reactor designs.

The federal inspection reports detail various safety assessments and outcomes but do not address the ITAAC notification process.


Question: irradiation embrittlement in nuclear components

C7 CSBFW was not identified as having potential irradiation embrittlement (IE) in MRP-227, Table 4-2.

Item (1) under "Linked Expansion Components and Items" states that the C5.5 CSBFW in Table 4-5 is "linked to the Primary Item for CE-design CSB UGs in MRP-227, Revision 2, Table 4-2, Item C5." This is not correct.


Question: The extended allowed outage time (AOT) is critical for ensuring the operational reliability of the diesel generators.

The second Completion Time (14 days) applies to an inoperable Division 1 or Division 2 DG and is a risk-informed allowed outage time (AOT) based on a plant-specific risk analysis.

Operating crews will be briefed on the DG work plan whenever the extended AOT period is used, with consideration given to key procedural actions that would be required in the event of a loss of offsite power or station blackout.


Question: What are Nickel-based alloy filler metals and their significance in nuclear safety?

These dowels welds are made from Nickel-based alloy filler metals that are potentially susceptible to stress corrosion cracking (SCC) mechanism.

40-60: Only the SSHT bolts and bolt locking devices at Davis Besse were Expansion category components for the UCB, LCB, and FD bolts and bolt locking devices.


Question: The NRC mandates that licensees must compute and confirm pump flow rates to assure compliance with acceptance criteria.

Applicant performs calculation and verifies that it is 144.53 gpm. (below the specified band)


Question: What are the requirements for transfer cask radiation shielding?

For the MTC and PMTC Transfer Casks, 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 ASME Code, 2001 Edition with Addenda through 2003, Section III, Subsection NG, is the governing Code for the design, material procurement, fabrication and testing of the spent fuel baskets.


Question: The regulatory limit for an individual's annual dose is set at 100 mrem, and Joe notes the potential dose is significantly higher than this limit.

If the sum in Step 8 exceeds 100 mrem per year, proceed to Part 2 of the calculational method.

Joe reviews his assumptions and recognizes that the QC supervisor is not at the bottling control panel 24 hours per day.


Question: tritium groundwater monitoring

MNGP monitors groundwater for tritium as part of the GWPP. The groundwater monitoring wells are sampled at different frequencies ranging from monthly to quarterly to annually.

Shallow wells in outwash and alluvial deposits supply water for many farms and residences in the surrounding countryside. The water usually is generally very mineral rich containing calcium, magnesium, and bicarbonates, with small amounts of sodium, potassium, sulfates, and chlorides.


Question: hydrologic description in nuclear regulation

The hydrologic description section provides an overview of the site’s hydrology, including the water bodies, groundwater flow, and the implications for facility operations and environmental protection.

The safety analysis report outlines the necessary measures to ensure the reactor's structural integrity under seismic events, detailing design criteria and load considerations.


Question: nuclear reactor safety systems

The safety systems of a nuclear reactor are designed to prevent the release of radioactive materials into the environment under both normal and accident conditions.

Compliance with the proposed changes will be assessed during regular inspections to ensure that all activities conform to regulatory standards.


Question: thermal aging embrittlement

Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems


Question: nuclide activity measurement

Wt mean Activity (pCi/grams) shows the weighted mean activity of a nuclide, measured in picocuries per gram.

The table lists various nuclides alongside their associated Confidence values, which reflect the reliability of the measurements.


Question: The NRC mandates that entities conduct thorough evaluations of seismic risks associated with their facilities.

CEUS SSC refers to the Central and Eastern U.S. Seismic Source Characterization for Nuclear Facilities, which evaluates the seismic risks that nuclear plants may face.

CNS stands for Consolidated Nuclear Security, LLC, which is a company that provides security and safety services to nuclear power plants.


Question: nuclear regulatory inspection process

The U.S. Nuclear Regulatory Commission (NRC) conducts inspections to assess compliance with regulatory requirements and standards related to the operation of nuclear facilities.

The analysis of thermal hydraulics is essential for ensuring that the reactor core remains within safe operational limits under all conditions.


Question: What is a 50.69 LAR in nuclear licensing?

The 50.69 License Amendment Request (LAR) process allows nuclear plant operators to risk-inform their regulatory submittals in compliance with NRC guidelines, aiming for more efficient reviews.

Nuclear reactors operate under strict safety protocols that ensure the integrity of the reactor core during normal and emergency conditions.


Question: What is the Maintenance Rule in nuclear regulation?

10 CFR 50.65 (a)(1), the Maintenance Rule, requires each licensee to monitor the performance or condition of SSCs against licensee-established goals to ensure that the SSCs are capable of fulfilling their intended functions. If the performance or condition of an SSC does not meet established goals, appropriate corrective action is required to be taken.

However, these second Completion Times became a problem when the Industry proposed risk-informed Completion Times for some of the Specifications that contained the second Completion Times, specifically TSTF-409, "Containment Spray System Completion Time Extension (CE NPSD-1045-A)," and TSTF-430, "AOT Extension to 7 Days for LPI and Containment Spray (BAW-2295-A, Rev. 1)."


Question: What is TWINKLE used for in reactor analysis?

The spatial kinetics computer code, TWINKLE, is used for the average core transient analysis.

The pressure surge may therefore be calculated on the basis of conventional heat transfer from the fuel and prompt heat absorption by the coolant.


Question: What is the purpose of the Exclusion Area Boundary in nuclear regulation?

The Exclusion Area Boundary (EAB) is the line beyond which the public is generally not permitted to enter during a nuclear incident to ensure their safety.

The cumulative frequency data presented indicates the measured values at varying percentages, which helps in understanding statistical distributions.


Question: MagnaSTOR system specifications

The MAGNASTOR® System is designed to ensure the safe storage of spent nuclear fuel, incorporating advanced features like multiple layer containment and robust monitoring capabilities.

The amendments made to the Certificate of Compliance No. 1031 include updates to the administrative practices for License Renewal applications in the Nuclear Regulatory Commission submissions.


Question: Presentations on the current research plans and results regarding PWSCC are critical for understanding industry challenges.

The PWSCC Crack Growth – Current research plans and results are essential topics for discussion in the materials exchange meeting.

The harvesting and IAD topics could be presented in quick 10-minute segments at the upcoming meeting.


Question: The correlations between annual wind speeds and the established 15-year trends demonstrate a strong consistency for validity evaluations.

that yearly wind speed distributions correlated strongly with the 15-year distribution, with R² values ranging from 85% to 98%.

Figure 13 shows even stronger correlations between wind direction distributions, with R² values ranging from 95% to 99%.


Question: What requirements are there for environmental testing laboratories in relation to certifications?

The NRC requires all laboratories to maintain proper certification as part of the compliance verification process.

Each state maintains its own certification number for environmental laboratories, which is distinct and separate from federal requirements.


Question: What is the purpose of the Reactor Trip System Instrumentation?

3.3.1 Reactor Trip System Instrumentation | As shown in Table 3.3-1. ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT.

As shown in Table 3.3-1. ACTION 10 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours or be in at least HOT STANDBY in the next 6 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.


Question: decommissioning plants information gathering

Contacts for people that should be invited to the workshop as well as people who could be useful for getting information from decommissioning plants.

The workshop - 1 page "plan" attached to meeting notice. Any thoughts on topics, participants, location, dates, etc. are appreciated.


Question: The impact of losing offsite power can significantly affect the core flow rate in a reactor.

Figure 15-213 illustrates the correlation between time in seconds and core flow rate in gallons per minute during a loss of offsite power situation.

The overall efficiency of a reactor can be influenced by various design features and operational protocols that ensure optimal performance.


Question: What is the purpose of a Peer Review in a PRA standard?

The standard requires a peer review process that identifies and assesses where the technical requirements of the standard are not met.

The standard is a living document. Consequently, it should not impede research. It is structured so that, when improvements in the state of knowledge occur, the standard can easily be updated.


Question: The NRC evaluates the potential impact of biological events on systems that utilize raw cooling water to ensure operational reliability.

Biological Events  | Accumulation or deposition of vegetation or organisms (e.g., zebra mussels, clams, fish, algae) on an intake structure or internal to a system that uses raw cooling water from a source of surface water, causing its functional failure.

The reservoir and cooling tower are periodically inspected for macro and micro biological fouling and are treated, as required.


Question: The NRC has set regulations that require timely analysis of the operational metrics of steam generators during the licensing process.

NRC letter to TVA, "Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment No. 147 Regarding Change to Steam Generator Tube Inspection Frequency and Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-510 (EPID L-2020-LLA-0161)," dated July 26, 2021 (ML21153A049).

The TVA has submitted various applications outlining proposed changes to the technical specifications for water usage within the facility.


Question: civil penalty process in nuclear regulation

The notice of violation shall specify the date or dates, facts, and the nature of the alleged act or omission with which the person is charged, and shall identify specifically the particular provision or provisions of the law, rule, regulation, license, permit, part 76 certificate of compliance or compliance plan, or cease and desist order involved in the alleged violation and must state the amount of each proposed penalty.

If the person charged with violation requests a hearing, the Commission will issue an order designating the time and place of hearing.


Question: What is the purpose of the CIM in nuclear regulatory filings?

The CIM technical report disposition directly applies for the PPS application.

The IRIG data link does not need to be a real time data link.


Question: What are the implementing procedures for quality assurance in nuclear plants?

Curtiss-Wright / RPC Radics LLC Implementing Procedures provide a comprehensive framework for the quality assurance program, ensuring that all processes meet regulatory standards and operational safety requirements.

Quality monitoring involves systematic examinations, tests, and measurements to assess the performance and reliability of equipment and systems across various industries.


Question: door dynamic analysis

The top deck doors were dynamically analyzed to determine the loads and structural integrity of the door for the design basis load conditions.

Hinge bars are installed. Blankets are clamped. Static insulation is attached.


Question: The shipping paper must include the total activity of each package, represented in SI units.

- Total activity contained in each package in SI units (e.g. Bq, TBq, etc.), or in bold SI and customary units (e.g. Ci, mCi, etc.) with customary units in parentheses following the SI units

These are basic reference charts; refer to current U.S. DOT & NRC regulations for complete requirements.


Question: nuclide identification report

Nuclide confidence index threshold = 0.30


Question: What is the purpose of a FSS release record in nuclear operations?

The FSS Release Record documents the site's radiological conditions and confirms compliance with the release criteria established for the site closure process.

The emergency response plan outlines the actions to be taken in case of an incident, detailing evacuation routes and communication protocols.


Question: What is NEPA analysis in the context of nuclear projects?

the Staff’s NEPA analysis of the environmental effects of the Dewey-Burdock project on Native American cultural, religious, and historic resources was adequate

the Tribe concedes that many of the cases have distinguishing characteristics from the instant case.


Question: What are the aging effects requiring management for structural bolting materials in nuclear facilities?

Loss of material due to general, pitting, crevice corrosion is an aging effect requiring management for structural bolting.

The NRC reviews applications to ensure that licensees implement effective oversight programs.


Question: What temperature limits are set for cladding after a Loss-of-Coolant Accident?

The calculated maximum fuel element cladding temperature shall not EXCEED __(1)__ °F, and the maximum cladding oxidation shall not EXCEED __(2)__ percent of the total cladding thickness.

The NRC's regulations require that licensees adhere to stringent guidelines to ensure the safety and integrity of nuclear facilities.


Question: What are emergency operating procedures in nuclear plants?

Emergency Operating Procedures are vital protocols established to guide operators in responding to emergencies that threaten the safety of the nuclear power plant.

Nuclear power plants require a variety of support systems to maintain safe operation and extend system life.


Question: pressure regulator in nuclear reactors

The Pressure Regulator (PPR) is an essential component of the reactor coolant system that maintains the desired pressure levels within the system.

Various flow and pressure indicators are critical for monitoring the system's performance and ensuring operational safety.


Question: renewal of radioactive materials license

Enclosed is Amendment No. 9 renewing your U.S. Nuclear Regulatory Commission (NRC) Materials License No. 48-26564-01, in accordance with your January 14, 2025 renewal application.

An environmental assessment for this action is not required because this action is categorically excluded under Title 10 of the Code of Federal Regulations (CFR) Section 51.22(c).


Question: plant-specific aging management program

For units with thermally treated Alloy 690 SG tubes and with tubesheet cladding using Alloy 600 type material, a plant-specific AMP is necessary unless the applicant confirms that the industry’s analyses for tube-to-tubesheet weld cracking (e.g., chromium content for the tube-to-tubesheet welds is approximately 22 percent and the tubesheet primary face is in compression as discussed in EPRI 3000202850) are applicable and bounding for the unit, and the applicant will perform general visual inspections of the tubesheet region looking for evidence of cracking (e.g., rust stains on the tubesheet cladding) as part of the steam generator program.

The existing programs rely on control of reactor water chemistry to mitigate cracking due to PWSSC and visual inspections of the steam generator head interior surfaces.


Question: What is the role of k_eff in nuclear criticality safety calculations?

The normal distribution was used in this analysis for the calculated k_eff values.

Table 2 lists various parameters, including Fuel Type and Clad Material, essential for evaluating LWR type critical experiments.


Question: Licensees must perform a thorough assessment of the potential impacts of their operations to ensure compliance with safety regulations.

The regulatory framework mandates that all licensees must conduct comprehensive safety evaluations and submit detailed reports for NRC approval.

The new reactor design includes advanced cooling systems and innovative materials that enhance operational efficiency.


Question: The NRC assesses the performance of the diesel generating systems to ensure reliable power supply during emergency situations.

The Diesel Generator System is designed to provide backup power during unforeseen disruptions, thereby maintaining essential safety functions.

The emergency lighting systems are essential for ensuring visibility during power outages and must comply with the relevant safety standards.


Question: What is the significance of unsprayed containment volumes in nuclear safety assessments?

The unsprayed containment volume is quantified as 5.0000e+05, indicating the portion of the containment that is not directly sprayed for cooling in the event of a loss-of-coolant accident (LOCA).

The Control Room has a volume of 8.6300e+04, which is important for ensuring the safety and operational effectiveness of emergency response efforts.


Question: environmental qualification requirements for cable insulation

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

For inaccessible and underground instrumentation and control cables exposed to significant moisture where testing is required, a one-time test will be performed.


Question: I-131 containment measurements


Question: emergency shutdown procedures

Emergency Shutdown (Core Spray Injection)  |  0  |  1  |  ---  |  (Note 5)

Turbine Roll and Increase to Rated Power  |  174  |  235  |  ---  |  (Note 5)


Question: flaw growth rate in reactor materials

The NRC staff finds the uncertainty parameter of 0.467 acceptable for the Vogtle plant-specific relief request because it is based on over 1,000 FCG rate data for low alloy steels.

Even though EPRI assumed 0.001 flaw per nozzle at the NIR for the base case PFM analyses, EPRI performed a SS on the impact of 0.1 flaw per nozzle.


Question: What are the voltage-based alternate repair criteria for nuclear reactors?

The Voltage-Based Alternate Repair Criteria provides guidelines for addressing specific conditions related to voltage performance in electrical and instrumentation systems at Beaver Valley Unit 2 during Refueling Outage 24.

The refueling outage at Beaver Valley Unit 2 is scheduled to enhance safety measures and perform standard maintenance procedures.


Question: What is the impact of debris blockage on emergency recirculation during design basis accidents?

Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.

Nuclear Industry Procedure IP-ENG-001. Standard Design Process. Revision C, September 8, 2016.


Question: The proposed change suggests extending the time between examinations of the containment post-tensioning system from five years to ten years.

The foregoing analyses and evaluations, and the conclusions derived therefrom, support the proposed extension of the interval between containment post-tensioning system examinations to 10 years from the current 5 years.

The trend line, which is based on the previously stated postulate, remains well above the minimum line at T = 100, which is 54 years after the latest date for completion of the next surveillance if the interval is extended to 10 years.


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

Energy Calibration Used Done On: 7/8/2014


Question: What information is required when a nuclear device is no longer in possession?

Provide Information about devices listed in Section 2 or 6, but no longer in your possession.

- NRC Device Key:
- Transfer Date:
  - MM:
  - DD:
  - YYYY:

- Location of the Device:
  - [ ] Whereabouts Unknown (Complete Part 1 only)


Question: best estimate methodology in nuclear licensing

...I do think you have to account for uncertainties. Maybe not 95/95, but you have to have uncertainties in the definition of a best estimate.

Vendors, they prefer to do something maybe a little more conservative than we would accept to make it easier for the review and quick in the review because they don't need all that margin.


Question: The NRC oversees licenses related to industrial radiography and safety requirements involved in this field.

Licenses for Industrial Radiography and Radiation Safety Requirements for Industrial Radiography

These regulation changes are not required to be adopted for purposes of Compatibility.


Question: The evaluation of the effects of debris on ECCS and CSS system components focuses on the wear and blockage risks in the context of LOCA scenarios.

An analysis bounding all break sizes was performed to evaluate the impact of debris on the wear or blockage of the ECCS and CSS piping and components downstream of the strainer following a LOCA.

The containment sump pool water volume following a LOCA was determined by considering all water sources (i.e., the refueling water storage tank, reactor coolant system (RCS), the safety injection accumulators, and the spray additive tank) and subtracting various holdup volumes.


Question: What is dose estimation in nuclear safety?

Dose estimation would have to be done using some Monte Carlo-based software.

There is no additional risk from extravasations of radiopharmaceuticals as from gadolinium or iodinated contrast media.


Question: What information is required in a chain of custody report for NRC licensed materials?

Check box if samples include NRC licensed material?

Please refer to ACZ’s terms & conditions located on the reverse side of this COC.


Question: environmental monitoring programs

Chapter 6 discusses the Environmental Measurements and Monitoring Programs designed to assess the impact of operations at the TRISO-X Fuel Fabrication Facility on surrounding environments.

The estimated viewshed boundaries for the TRISO-X Fuel Fabrication Facility are presented in Figure 4.9.2-3, indicating potential visual impacts from various locations.


Question: digital modernization in nuclear reactors

Digital modernization refers to the incorporation of digital technologies to enhance the operational performance and safety of nuclear reactors, including monitoring and control systems.

The evaluation of defenses in safety systems should account for both expected and unexpected operational conditions, ensuring robust reactor performance under various scenarios.


Question: What is the significance of the acceptance criteria in regulatory filings?

The Acceptance Criteria is a standard that the survey unit must meet to demonstrate that the unit complies with the regulatory requirements established by the NRC.

The survey unit consistently shows various measurement values over multiple tests, which highlights the reliability of the methodologies used.


Question: The NRC staff highlighted potential impacts on the timing of manual actions during scenario events due to limitations in the testbed.

The NRC staff questioned how the limitations of the testbed may affect the timing associated with manual action in the scenario events.

The staff noted that the simulator model has not yet been fully developed by the vendor, which necessitated the use of simulator staff manipulation to facilitate the validation activities.


Question: The performance of the EOC-RPT trip function must remain within operational parameters to ensure safety.

B. One or more Functions with EOC-RPT trip capability not maintained.

Remove the associated recirculation pump fast speed breaker from service.


Question: What is LCO 3.0.4 in nuclear regulatory terms?

DENC has established TS Bases for LCO 3.0.4 and SR 4.0.4 which state that use of the TS mode change limitation flexibility established by LCO 3.0.4 and SR 4.0.4 is not to be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to operable status before entering an associated Mode or other specified condition in the TS Applicability.

DENC has concluded that the proposed NSHCD presented in the Federal Register notice is applicable to Millstone Power Station Unit 2 (MPS2) and Millstone Power Station Unit 3 (MPS3) and is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).


Question: What does MMD/MTU mean in nuclear licensing?

MWD/MTU refers to megawatt days per metric ton uranium, a measure used in the regulatory assessment of uranium utilization in nuclear reactors.

The Minnesota Pollution Control Agency (MPCA) is responsible for ensuring methods are in place for managing environmental quality in Minnesota.


Question: The classification system for emergencies is crucial for effective response management.

The Emergency Classification System is designed to categorize emergencies based on their actual or potential impact on public health and safety.

Medical and public health support is essential for ensuring the well-being of individuals affected by emergencies.


Question: What geological processes have shaped the Piedmont Physiographic Province?

Since Early Paleozoic time, rocks of the Piedmont Physiographic Province have undergone extensive tectonic activity, primarily from three compressional orogenies (Taconic, Acadian, and Allegheny) during the Paleozoic Era and one extensional episode during the Mesozoic Era (as presented in Section 2.5.1.1.2).

During the Penobscot orogeny westward and northward thrusting of the island-arc terrain was initiated.


Question: assembly average burnup

49 < Assembly Average Burnup ≤ 50 GWd/MTU

The certificate ensures compliance with NRC regulations.


Question: The NRC requires applicants to justify specific re-inspection strategies for critical components as part of their licensing submission.

The review concluded that the BWR Vessel Internals aging management program will provide reasonable assurance that the applicable aging effects will be adequately managed so that the intended functions of components within the scope of license renewal are maintained consistent with the current licensing basis during the subsequent period of extended operation.

Inspections required by BWRVIP-47-A relative to CRD housings are further discussed in the BWR Penetrations program (B.2.1.6).


Question: What are the specifications for the Model 8040?

RadioMed Corporation's Model 8040 is a part of their comprehensive line of medical equipment designed for nuclear diagnostics and therapy.

The P192 SERIES of Radionics, Inc. includes various devices ranging from high-output imaging systems to specialized diagnostic equipment.


Question: The NRC requires projections of flow rates for groundwater treatment systems based on past performance and present capacity.

Expected flowrate projections for the groundwater treatment systems are provided in the “Groundwater Flow and Transport Model Report” in Appendix F to the revised GCAP.

Cost estimates in the GCAP were prepared for the purpose of comparing the proposed corrective action alternatives.


Question: fire protection program implementation

THIS POSITION SHOULD BE RESPONSIBLE FOR ASSURING THE EFFECTIVE IMPLEMENTATION OF THE FIRE PROTECTION PROGRAM BY PLANNED INSPECTIONS AND SCHEDULED AUDITS.

THE RESPONSIBILITIES OF EACH FIRE BRIGADE POSITION SHOULD CORRESPOND WITH THE ACTIONS REQUIRED BY THE FIRE FIGHTING PROCEDURES.


Question: What is the purpose of secondary containment in nuclear facilities?

The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment into secondary containment following a Design Basis Accident (DBA).

The secondary containment boundary consists of the reactor building structure and associated removable walls and panels, hatches, doors, dampers, sealed penetrations and valves.


Question: The system used to remove residual heat from the reactor coolant is crucial during the cooldown process after shutdown.

The residual heat removal system is designed to reduce the temperature of the reactor coolant from 350°F to 140°F over a period of 16 hours.

Plant shutdown encompasses the operations that bring the reactor from hot no-load temperature and pressure to cold shutdown conditions.


Question: Strontium-90 analysis in environmental samples

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


Question: What is the yield percentage of EU-152?

EU-152 has a yield percentage of 28.40%.

The activity of PA-228 at the energy of 911.23 keV is 3.89E+00 pCi/grams.


Question: What limitations arise when the RPIS system is inoperative?

Per QCAN 901-5 G-5, RPIS INOP, the automatic action is a Reactor Manual Control Select Block. Relay -28/-138 drops out on an RPIS INOP signal. This prevents the Rod Select Relays from energizing when a control rod is selected on the Pushbutton Matrix.

The control rod movement can only be restored by resetting the Select Block, which is crucial for proper reactor operation.


Question: What is a degraded condition in nuclear safety systems?

A degraded condition is one in which the qualification of an SSC or its functional capability is reduced. Examples of degraded conditions are failures, malfunctions, deficiencies, deviations, and defective material and equipment.

During changing plant conditions (plant refueling or maintenance outages), the frequency and scope of plant status tours may be increased to tour areas not normally accessible and to observe material condition and equipment in an abnormal lineup.


Question: What is CASS in nuclear materials research?

NRC: Update on CASS (dissimilar metal welds)


Question: What is the Risk-Informed Completion Time (RICT) in nuclear operations?

The Risk Managed Technical Specifications (RMTS) implementation, and specifically discuss the decision process for risk management action (RMA) implementation during a Risk-Informed Completion Time (RICT).

More detailed procedures will provide specific responsibilities, limitations, and instructions for implementing the RICT program.


Question: The Online Monitoring Program must adhere to specific guidelines involving the evaluation and analysis of transmitter performance.

This program provides controls to determine the need for calibration for pressure, level, and flow transmitters using condition monitoring based on drift analysis.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.


Question: The fire water system is designed to ensure the operability of fire pumps, especially through the effective management of material loss and flow blockage.

SLRA Section 2.3.3.7 states that the circulating water system 'contains traveling screens which are credited for filtration of fire pump supply water to support fire pump operability.'

The NRC staff is unclear on what a fire window is and what procedures address fire windows. Please discuss what a fire window is, including material and where located, and what procedures address their inspection.


Question: nuclear fuel cycle

The nuclear fuel cycle includes the processes of mining, milling, enrichment, and fabrication of fuel, followed by its use in nuclear reactors and the management of spent fuel and radioactive waste.

The safety protocols for chemical handling ensure that all materials are stored in compliance with environmental regulations and that personnel are trained to manage hazardous substances effectively.


Question: The NRC mandates specific analytical procedures for assessing thermal margins in licensing applications.

The NRC requires licensees to analyze thermal hydraulic performance under various design basis conditions to ensure safety and compliance in their regulatory filings.

The data shows varying flow rates and energy measurements at different time intervals, displaying the complex interactions within a reactor's steam cycle.


Question: What is the role of the Emergency Technical Director during emergencies?

The Emergency Technical Director (ETD) reports to the SEM in the TSC as part of the augmentation of the on-shift emergency organization. He directs the activities of the Technical Support Team.

The Shift Technical Advisor (Control Room) will remain in the Control Room to advise the Shift Manager or Unit Supervisor on operations activities.


Question: In the context of safety systems, redundancy is crucial for mitigating failures in critical operations.

The High Pressure Core Spray (HPCS) System ensures reactor vessel water is maintained at safe levels during emergency conditions through multiple channels and redundant instrumentation.

The Turbine Trip can occur due to various operational anomalies, leading to immediate reactor safety protocols being enacted.


Question: What measurement units are used for reactor coolant temperatures?

The coolant temperatures in the reactors are measured in degrees Celsius (°C).

The reports frequently summarize operational data over time spans such as weeks and months.


Question: What is the activity of Cs-137 in a regulatory context?

The activity (pCi/g) of Cs-137 is -2.42E-01 in the sample L101-10A-JACS-003-SS.

The activity (pCi/g) of Fe-55 is 3.979E-01 in sample L101-10A-JACS-003-SS.


Question: Diablo Canyon nuclear reactor licensing

The Diablo Canyon Power Plant is a nuclear power facility located in California, and it operates with two boiling water reactors (BWRs) that were commissioned in the late 1980s.

NRC Document Control Desk is responsible for managing the documentation and administrative processes associated with regulatory filings.


Question: The measurements recorded show the activity levels of various radionuclides in different samples.

Gamma results were measured in disintegrations per minute per 100 cm² (dpm/100 cm²) to assess the radioactive contamination in samples.

The reactor design includes advanced safety features that mitigate the risks associated with operational abnormalities.


Question: The NRC's examination process requires thorough interpretation and execution of procedures relevant to operating reactors.

**Comply with and Use Procedures (1)**      | 1-8         | 1.3        | 5-8         | 7-10       |

The competencies checklist outlines various scenarios for different competencies that are necessary for operation but does not detail licensing procedures.


Question: License applications must include detailed information on safety protocols and operational procedures.

The NRC reviews regulatory filings to ensure compliance with safety standards and licensing requirements for nuclear facilities.

The maintenance of electrical systems is crucial for the overall functioning of a nuclear reactor.


Question: What is the significance of oxidation reduction potential (ORP) in water quality assessments?

ORP (Oxidation Reduction Potential) measures the cleanliness of water and its ability to decompose organic materials; it is typically measured in millivolts (mV).

The field pH values recorded ranged from 4.59 to 7.33 across various sampling dates, indicating a range of acid to neutral conditions.


Question: The NRC mandates that licensees conduct safety evaluations to ensure compliance with technical specifications.

The safety analysis reports must demonstrate that the facility can adequately respond to both design basis and beyond-design-basis events.

The history of nuclear power development spans several decades and includes numerous technological innovations.


Question: how to determine the active or inactive status of a senior reactor operator license

To maintain an active status, the licensee shall actively perform the functions of an operator or senior operator for a minimum of seven 8 hour shifts a calendar quarter or five 12 hour shifts a calendar quarter in a position credited for watch-standing proficiency.

The performance evaluation details the actions and responsibilities required for license maintenance and efficacy of the reactor operations.


Question: minimum cooling time after assembly enrichment

Minimum Cooling Time (years) is specified for each range of Minimum Initial Assembly Avg. Enrichment in the loading table for PWR fuel.

The coefficients observed in the cooling time data depend directly on the reactor type and operational parameters.


Question: What are uranium screening levels for external exposure?

The U.S. Environmental Protection Agency (EPA) has developed uranium screening levels for external exposure for different frequencies of exposure, including decay products.

The highest concentration of uranium observed in residual soils at the site was 65 mg/kg in the vicinity of the former raffinate pond where highly concentrated milling fluids were disposed.


Question: The emergency air conditioning system for the control room must have two functional air filtration trains to ensure safe operation.

3.7.6 The common control room emergency air conditioning system (CREACS) shall be OPERABLE with: a. Two independent air conditioning filtration trains (one from each unit) consisting of: 1. Two fans and associated outlet dampers, 2. One cooling coil, 3. One charcoal adsorber and HEPA filter array, 4. Return air isolation damper.

The control room envelope (CRE) boundary may be opened intermittently under administrative control.


Question: If a system fails to meet its Limiting Condition for Operation, action must be taken to return it to compliance.

When a system is determined not to meet the LCO, Condition A is entered.

The completion times for required actions are crucial for maintaining safety protocols.


Question: roof ponding in nuclear plant design

There are two effects which must be considered to assure that roof ponding has no detrimental effects on safe operation of the plant. They are: The ability of the roof structure to carry the additional weight, locally and globally, resulting from the depth of ponded water; and Potential in-leakage through roof joints such as air intakes/exhausts which would not be submerged under normal rainfall but which may become submerged under intense rains such as the PMP.

In the flood studies performed by the firm of Tippets-Abbett-McCarthy-Stratton (TAMS) Hydrologic Engineers and Architects, it was assumed that the PMF hydrograph at Conowingo Dam is the same as that at Harrisburg, or a peak discharge of 1,750,000 cubic feet per second (cfs).


Question: dynamic load factors

To account for the dynamic nature of the blowdown forces, dynamic load factors are applied to the DBA forces applied statically to the finite element representation of the lower support structure.

The tangential force acting on the ice baskets due to cross flow in the ice condenser chamber has been taken to be 23.8 lbs per linear foot.


Question: The NRC stipulates that licensees must evaluate the presence and levels of certain nuclides in their environmental monitoring programs...

The table lists various nuclides along with their energy levels measured in keV, percent yield, and minimum detectable activity (MDA) measured in pCi/m².

The Energy (keV) values shown in the table are essential for understanding the radiological profile of the materials in question.


Question: The NRC evaluates the levels of radioactive isotopes such as Co-60 and Cs-137 during regulatory assessments to ensure compliance with safety standards.

Measurement results indicate the levels of radioactivity, specifically showing the gamma results for various isotopes including Co-60 and Cs-137 measured in disintegrations per minute per 100 square centimeters (dpm/100 cm²).

Gamma measurements can vary widely based on environmental factors and the specific calibration of detection equipment used in the field.


Question: The Auxiliary Feedwater system consists of multiple trains to ensure adequate coolant supply during various operational modes.

LCO 3.0.4.b is not applicable when entering MODE 1.


Question: What is the purpose of a detector setup in nuclear regulatory filings?

The detector setup is configured to ensure accurate measurement of radiation levels to support regulatory compliance and safety evaluations.

The logging mode selected during the tests was continuous, enabling regular updates of the logged units recorded during the measurement.


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

The identification report contains the nuclide confidence index threshold, which is set at 0.30, indicating the minimum level of assurance required for the identification of nuclides in regulatory contexts.

The report lists the Energy Tolerance as 1.000 keV which is critical for ensuring accurate energy readings during spectrum analysis.


Question: The logic design ensures that Emergency Feedwater is isolated based on conditions from signal feedback, preventing overcooling risks.

The function of the EFIC vector logic is to determine whether EFW should not be fed to one or the other SG. This is to preclude the continued addition of EFW to a depressurized SG and, thus, minimize the overcooling effects.

The LCOs also support meeting the requirements of GDC-17, "Electric Power Systems."


Question: cyber security controls in nuclear infrastructure

It may be appropriate for licensees to consider protecting to the level of a CDA when upgrading AA infrastructure and applications as it is less impactful to address and implement cyber security controls early in the design phase of the life cycle.

The reactor coolant system is designed to remove heat from the reactor core and transfer it to the steam generators.


Question: Aging effects must be proactively managed in reactor components to ensure structural integrity throughout their operational life.

Aging Management Review assures that the components maintain their intended function, such as the structural integrity of piping and associated parts under various conditions.

Piping components are made from stainless steel and are exposed to different internal environments.


Question: The NRC expects licensees to cultivate and sustain a positive safety culture in their operations.

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

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


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves multiple phases, including pre-application consultations, formal application submission, safety reviews, and public comment periods, ensuring transparency and thorough evaluation of safety protocols.

Many thermal power plants operate on fossil fuels, generating electricity through combustion processes that release gases into the atmosphere, impacting air quality and climate change.


Question: What are the design characteristics preventing steam line breaks from affecting lower ice condenser inlet doors?

Postulated steam line breaks inside Containment cannot result in direct steam impingement of the lower ice condenser inlet doors due to the following design characteristics: 1. Piping arrangement with respect to the inlet doors was a major pipe routing consideration. 2. Guard piping of portions of the Main Steam System inside Containment is utilized to preclude: a. steam burn-through of the ice condenser due to jet impingement on the lower inlet doors. b. a LOCA caused by a postulated steam line break. c. a subsequent pipe break of either the Main Steam or Feedwater System of unaffected steam generators. d. local overpressurization caused by excessive flow rates. 3. Restraints and energy absorbers are utilized to restrict excessive movements from a postulated rupture and prevent direct steam impingement on the lower inlet doors.

The developed ice condenser cold head will then compress the gasket seals, and will also serve to re-close the doors should the panels briefly and inadvertently break away from the seal during normal operation.


Question: Reactor Pressure Control systems

**SYSTEM:** Reactor Pressure Control (SF 3) – Pressurizer Pressure Control System

The operator continues the depressurization.


Question: AlOOH solution volumes in TB tests

The total amount of AlOOH added to the test tank was 15.53 kg at test scale or 147.1 kg at the plant scale.

Pressure differential across the test strainer, flow rate, and test water temperature were measured and recorded continuously throughout the entire duration of both tests.


Question: requirements for applicants to submit environmental reports

Licensees are required by 10 CFR 51.53(c)(1) to submit with its application a separate document entitled 'Applicant’s Environmental Report—Operating License Renewal Stage.'

On August 6, 2024, the NRC published a final rule (89 FR 64166) revising its environmental protection regulations in 10 CFR 51, 'Environmental protection regulations for domestic licensing and related regulatory functions.'


Question: The program must establish protocols to test the containment leakage rate as required by regulatory guidelines.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig for Unit 1 and 38.4 psig for Unit 2.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves a series of steps including pre-application consultation, application submission, acceptance review, and safety review to ensure compliance with regulatory standards.

The reactor core consists of fuel assemblies, control rods, and a coolant system, which play critical roles in maintaining safe operational conditions.


Question: How do main turbine stop valves function during reactor testing?

The MicroNet System will perform the following functions during testing: Maintain the 21, 23, and 24MS29 (GV) to maintain load during testing.

The evaluator noted that local observation of valve movement is required for proper evaluation.


Question: Fermi-2 PRA model history

Table B-1 shows a brief history of the Fermi-2 PRA model.

Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10CFR50.65(a)(4)).


Question: The Zone IIA saprolitic soil has been examined, leading to the finding that it does not show signs of being susceptible to liquefaction.

Soil liquefaction is a process by which loose, saturated, granular deposits lose a significant portion of their shear strength due to pore pressure buildup resulting from cyclic loading, such as that caused by an earthquake.

For Profile 1, which is used in the liquefaction and slope stability analysis, the ZPA at the ground surface increased from 0.46g in Table 2.5-46 for the low frequency case to 0.57g using 150 percent Gmax.


Question: inadvertent opening of a pressurizer safety valve

An inadvertent opening of a pressurizer safety or relief valve incident is classified as an ANS Condition II event, a fault of moderate frequency. See Section 3.10 for a discussion of Condition II events. As described in Section 4.0, the inadvertent opening of a pressurizer safety or relief valve scenario is analyzed to demonstrate that the DNBR SAFDL, the FCM SAFDL, and the TCS SAFDL are not exceeded as a result of the fuel transition.

The pressure continues to decrease throughout the transient. The effect of the pressure decrease is to decrease power via the moderator density feedback, but the reactor control system (if in the automatic mode) functions to attempt to maintain the power and average coolant temperature by withdrawing control rods until reactor trip occurs.


Question: The NRC welcomes input on the environmental impact findings and the processes involved in cultural resources assessment.

The NRC staff invites you to submit any comments that you may have on the enclosed cultural resources report, the NRC staff’s finding of Adverse Effect, and the development of an MOA.

If you have any questions, comments, or concerns, please contact Sarah Lopas, Senior Environmental Project Manager, by phone at 301-415-6360 or via email at Sarah.Lopas@nrc.gov; or Patricia Vokoun, Environmental Project Manager, by phone at 301-415-3470 or via email at Patricia.Vokoun@nrc.gov.


Question: The NRC evaluates the potential risks associated with temporary modifications to safety systems at nuclear facilities.

Virginia Electric and Power Company (Dominion Energy Virginia) has reviewed the proposed change for Surry Power Station (SPS) Units 1 and 2 against the criteria of 10 CFR 50.92 and has concluded the change does not pose a significant safety hazards consideration as defined therein.

Existing station abnormal procedures already address a loss of component cooling and the use of alternate cooling for a loss of decay heat removal in the unlikely event they are required.


Question: advanced licensing system

ALS helps move forward with the immediate problem in the schedule that the industry is pursuing that was described to us yesterday to allow them to move forward with increased enrichment.

It removes a large piece -- uncertainty is not the right word probably because -- a large piece of the regulatory uncertainty.


Question: safety-significant systems cooling

In addition to the heat rejection capability of the passive residual heat removal system, systems to transfer heat from structures, systems, and components important to safety to an ultimate heat sink shall be provided, as necessary, to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

The single failure criterion language is deleted consistent with the guidance in NEI 18-04 as endorsed by RG 1.233.


Question: The report includes the measured activities of various radionuclides, such as their mean values and associated uncertainties.

The report details various isotopes without discussing their effects on structural integrity.


Question: What is the process for submitting an Agreement Application to the NRC?

The State is revising their draft Agreement Application to the NRC and expects to have it submitted by the end of 2024.

Indiana staff continue to attend NRC training courses and accompany NRC and Agreement State inspectors on inspections in the State.


Question: What is the purpose of a Gaussian Peak Locate Threshold in gamma spectrum analysis?

Peak Locate Threshold is defined as the minimum number of counts for identifying a peak, set to 3.00 for the analysis.

The sample was acquired on 5/8/2019 at 4:15:48 PM and the live time of the detection was exactly 1800.0 seconds.


Question: Clarifications requested by NRC staff pertain to structural responses related to CoC 1042 Amendment No. 4.

TN provided supplemental RAI responses for EOS Amendment No. 4 on November 6, 2024 (ML24311A209, Package ML24311A208).

Don Shaw is the contact person from TN Americas LLC, which is a registered organization related to the licensing process.


Question: The documentation stipulates that all licensees must adhere to specific guidelines for radiation safety assessments.

The NRC mandates that licensees conduct thorough safety evaluations to minimize risks associated with nuclear facilities.

The update to Chart A will now reflect changes in the utilization metrics used for health services.


Question: What is NRC incensable material?

4) Are any samples NRC incensable material?

The Chain of Custody form or other directive shipping papers must be present for all samples.


Question: What are the environmental impacts of postulated accidents in nuclear licensing?

Environmental Impacts of Postulated Accidents are addressed in Appendix F which discusses potential risks and considerations.

Environmental Issues and Impact Findings Contained in the Final Rule, 10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions” cover broader regulatory functions.


Question: aerosol removal data


Question: What is the significance of nuclide yield in nuclear regulatory assessments?

Nuclear reactors must adhere to strict regulations to ensure the safety and security of the facility.


Question: What is ARITA in nuclear regulatory context?

The ARITA program is designed to enhance the safety and reliability of nuclear reactor operations by improving the standards for regulatory filings and technical safety evaluations.

The meeting options include various links to join the conference remotely or to call in using a telephone number.


Question: The accurate evaluation of the centerline temperature is essential for understanding the thermal behavior of nuclear fuel during operation.

The best-estimate character of the results for UO₂ and gadolinium fuel is confirmed by the centerline temperature predictions shown in Section 4.3.2.

The NRC emphasizes the need for continuous monitoring to ensure compliance with safety regulations.


Question: HMC's commitment to address groundwater impacts was outlined following NRC guidelines.

The Groundwater Corrective Action Program aims to mitigate and manage potential contamination issues as mandated by regulatory oversight.

The attendees of the meeting included representatives from multiple agencies to ensure comprehensive discussion and decision-making.


Question: What is the significance of the Baker-Just equation in nuclear safety assessments?

In its posted denial of PRM-50-76, the NRC states, "The remaining data from Bostrum (`The High Temperature Oxidation of Zircaloy in Water," W. A. Bostrup, WAPD-104 March 1954) and Lemmon (`Studies Relating to the Reaction Between Zirconium and Water at High Temperatures," A. W. Lemmon, Jr., BMI-1154, January 1957), at more relevant zirconium cladding conditions, were used by Baker and Just in the derivation of their equation."

It is more likely that those authors merely lifted the description of those references from the Baker-Just report, ML050550198.


Question: NRC’s Regulatory Issue Summary addresses the importance of protecting security-related sensitive information during regulatory processes.

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

A copy of this letter will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s ADAMS, accessible from the NRC Web site at https://www.nrc.gov/reading-rm/adams.html.


Question: The requirements for separating non-class 1E circuits from class 1E circuits are outlined in IEEE 384 section 4.6.1.

IEEE 384 section 4.6.1 requires non-class 1E circuits to be separated from class 1E circuits by the minimum separation requirements of sections 5.1.3, 5.1.4, or 5.6.

The previous design also periodically tested the trip circuit during quarterly testing.


Question: What are the consequences of a Loss of Coolant Accident (LOCA) in a nuclear reactor?

The crew will insert a manual scram on high Drywell pressure due to a break in the 'A' Recirc Loop Suction Line. After the reactor scram and main turbine trip, the RAT fails to transfer to Bus 11. The 1B RFP trips on overcurrent resulting in a loss of all RFPs. Actions per QGA 100 and QGA 100 Alternate Level Control will be taken to try and maintain RPV water level above -142 inches (MSCVRL). Containment parameters will be controlled per QGA 200 actions. When conditions permit, the crew will blowdown the RPV, and restore level with low pressure ECCS systems.

The plant is currently at 50% power with the 1C Reactor Feed Pump out of service. Electrical Maintenance has requested the 1A EHC pump secured and removed from service for a motor inspection. The DFWL Operator Work Station (OWS) is out of service.


Question: The NRC mandates a comprehensive safety evaluation of nuclear reactor designs before granting operational licenses.

The Nuclear Regulatory Commission (NRC) requires that license applications for nuclear reactors include detailed information about the design, safety analysis, and environmental assessments to ensure public safety and compliance with regulatory standards.

Nuclear power plants are designed to operate under high temperatures, utilizing advanced cooling systems to maintain safety during operations.


Question: gamma spectrum analysis

The gamma spectrum analysis was performed using a 500 ml Marinelli geometry, with the sample identified as L1-SUB-TDS-FSGS-B09B-SB and weighed 7.858E+02 grams.

The sample was taken on 5/23/2019 at 10:21:00 AM, and the acquisition started later on 6/8/2019 at 1:37:52 PM.


Question: Licensees must conduct thorough risk assessments to ensure that safety systems are reliable and effective.

The licensing process for nuclear facilities requires a comprehensive review of safety measures and regulatory compliance to protect public health and safety.

Regular maintenance of the reactor core is crucial to the overall efficiency of energy production and minimizing operational downtime.


Question: What is the authorized use of Californium-252 in nuclear applications?

Authorized use: For use in Tracer Electronic portable gauging device for measuring physical properties of materials.

Individuals who have received the training required under 10 CFR 32.210 or by an Agreement State, in the absence of a record of training should be used for not exceeding six months.


Question: nuclear reactor emergency core cooling systems

The Emergency Core Cooling System (ECCS) is designed to provide sufficient cooling to the nuclear reactor core in the event of a loss of coolant accident, ensuring that fuel damage is prevented.

The cooling systems in many industrial processes are critical for efficiency, but they operate under different principles and do not apply to nuclear reactor safety.


Question: The ability to suspend an individual's license comes with specific conditions under the general licensing authority for GLROs.

Individuals operating under license subject to conditions.

Facilities must continue to meet the criteria of 53.800 (failure would be a reportable unanalyzed condition).


Question: Plutonium isotopes

Client Sample ID: L1-010-104-AJGS-013-SS


Question: What role do the Baker-Just and Cathcart-Pawel equations play in ECCS evaluation models?

Data from cited experiments indicate that the Baker-Just and Cathcart-Pawel equations used to calculate the metal-water reaction rate in ECCS evaluation models that the NRC has determined to be acceptable for use in evaluating ECCS cooling performance are not conservative.

The NRC disagrees that the Baker-Just equation is not conservative.


Question: The spent fuel pool cooling system is equipped with features to prevent siphoning and ensure a continuous water level.

Loss of forced cooling calculations have been performed to verify the adequacy of the spent fuel pool as an assured emergency source of primary coolant pump seal water.

The amount of time before the pool begins to boil depends on the heat load and the initial pool temperature.


Question: The NRC mandates inspections on critical safety systems during operational outages to ensure compliance with safety regulations.

The 71124.04 procedure focuses on Occupational Dose Assessment to maintain the safety of workers and ensure regulatory compliance during the operation of nuclear facilities.

The inspection procedure 57050 addresses the visual testing examination aspect of nondestructive testing ensuring structural integrity.


Question: The evaluation of potential environmental impacts must adhere to regulatory guidelines.

The NRC conducts environmental reviews to ensure that the licensing of nuclear facilities complies with applicable laws and regulations regarding environmental protection.

Reduced rainfall can influence local agriculture and water supply, significantly impacting crop yields and food security.


Question: wind speed analysis in regulatory filings

The average wind speed recorded from 2008 to 2024 in Gallup, New Mexico, is 1.65 m/s, as indicated in the wind rose plot.

The total count of hours recorded during this period was 110993 hrs, reflecting the extensive data collection for wind direction and speed.


Question: Who are the key personnel involved in the project team?

- **Michael Kosman** – Project Manager
- **Robert Ammon** – Technical Director / Project Engineer
- **Dan Hunt** – QA Manager

The NRC requires comprehensive documentation for all safety analyses conducted at the facility.


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

Regulatory filings are essential for the nuclear licensing process, as they provide the necessary documentation that demonstrates compliance with safety standards and regulatory requirements set forth by oversight authorities like the NRC.

Thermo Process Instruments, LP, a subsidiary of Thermo Fisher Scientific, specializes in high-performance measurement solutions for industrial applications.


Question: What is a reactor seismic isolation system?

The purpose of this report is to provide a description of the methodology and requirements to establish the design criteria and qualification of the Natrium™ reactor seismic isolation system (SIS) for review and approval by the U.S. Nuclear Regulatory Commission (NRC).

There are no assumptions used in the development of this report requiring verification. There are no open items that require future actions to verify and close.


Question: What is the threshold for the nuclide confidence index used in the identification process?

**Nuclide confidence index threshold:** 0.30

The table lists various identified nuclides along with their associated data such as confidence, energy, yield, and activity.


Question: The NRC mandates licensees to regularly submit documentation that ensures compliance with safety protocols.

Licensees must provide detailed technical information regarding the safety performance of their systems and components as part of the regulatory filings.

The financial report detailed the annual expenditures and budget allocations for various departments.


Question: PWR pressurizer vessel top head

3-D finite element model and mesh for PWR pressurizer vessel top head (CE design)


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

The Nuclear Regulatory Commission (NRC) reviews applications for nuclear licenses to ensure compliance with safety regulations and environmental standards.

The temperature of the coolant is crucial for maintaining reactor stability and ensuring safe operations.


Question: The regulatory process requires thorough reviews before issuing permits to operators.

The NRC's licensing process involves careful evaluation of safety analyses submitted by nuclear power plant operators to ensure compliance with regulatory requirements.

The office supports reactor modifications that enhance operational efficiency without compromising safety measures.


Question: What is the significance of the Heat Flux Hot Channel Factor (FQ(Z)) in reactor operation?

The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core.

These LCOs, along with LCO 3.1.4, 'Rod Group Alignment Limits,' LCO 3.1.5, 'Shutdown Bank Insertion Limits,' and LCO 3.1.6, 'Control Bank Insertion Limits,' maintain the core limits on power distributions on a continuous basis.


Question: What is COBRA-FLX and its significance in critical heat flux correlation?

The COBRA-FLX code is a thermal-hydraulic computer code used to calculate the behavior of coolant flow and heat transfer in nuclear reactor core assemblies, particularly during critical heat flux conditions.

Table B-1 provides a detailed overview of the test assembly conditions used in various non-mixing grid tests conducted to assess reactor performance.


Question: What are Protective Action Recommendations (PARs) in emergency response?

Evaluate recommended protective actions with the Parish President, RADEF Officer and EOC staff. Assist the Parish President in reaching a decision on protective action(s) to implement.

At a General Emergency, be prepared to discuss protective action recommendations (PARs) with RBS and LDEQ over the RBS Hotline (Refer to Public Alert procedure).


Question: general licensed nuclear devices

Find enclosed the required annual update for the general licensed nuclear devices at Mount Storm Power Station (GL-701314-26).

If you have any questions, please contact me at 304-813-0430.


Question: What is the ECCS Supplemental Boron System?

We also added an LCO, or what we call ECCS Supplemental Boron System, which is important for extended cooling periods to maintain shutdown margin -- maintain a reactor shutdown.

Next slide, some more changes. A couple things to note here, one of them -- and most of these will be addressed in other discussions, obviously, the design itself or the analyses.


Question: What is the importance of the safety analysis in nuclear power plants?

The safety analysis in nuclear power plants is conducted to evaluate the potential risks associated with the design and operation of the facility, ensuring that adequate safety measures are implemented to protect public health and safety.

The turbine generator operates by converting mechanical energy into electrical energy, and this process is essential for the overall efficiency of the power plant.


Question: transfer of source or byproduct material

(a) No licensee shall transfer source or byproduct material except as authorized pursuant to Part He-P 4030.

In considering an application by a licensee to renew or amend its license, DHHS/RHS shall apply the criteria set forth in this chapter for granting of an initial license.


Question: In the event of a de-energized condition, it is essential to consult the appropriate operational procedures, such as the Turbine Runback Response.

IF 6.9kV Unit Bd 1C is de-energized, OR motor breakers have tripped, THEN: REFER TO 1-AOI-37, Turbine Runback Response, as necessary.

Performer determines that 6.9kV Unit Bd 1C is de-energized, and refers to 1-AOI-37, Turbine Runback Response.


Question: What is atmospheric stability and how is it estimated?

For estimates of average dispersion over extended periods, the joint probability of occurrence of wind speed, wind direction, and atmospheric stability must be known.

The construction and operation of the EGC ESP Facility will have the potential to influence the local micrometeorology of the area in the immediate vicinity of the CPS Facility.


Question: What is LBB in the context of BWRs?

And we've never granted an LBB for BWR. And while that's not part of the rule, there may be other benefits that the BWR community could glean from LBB.

So two things. There is an alternative path for BWRs that does not require, that's an LBB evaluation.


Question: The NRC assessed how the uprate would affect the discharge temperatures and their environmental impact.

Xcel Energy projected that the power uprate would increase the temperature of the heated water released to the discharge canal by a maximum of 4.5°F over the pre-uprate temperature range of circulated cooling water released to the discharge canal of 66 to 95°F.

Only nine outbreaks occurred from exposure in lakes and rivers. More than half (52 percent) of the outbreaks were caused by Cryptosporidium. Other agents associated with outbreaks were E. coli, Legionella, norovirus, Pseudomonas, Giardia, and chemicals.


Question: radioactivity measurements for radionuclides

The table provides various measurements of radionuclides, including K-40 with an activity of 1.22E+01 pCi/g Dry and a measurement uncertainty of 1.61E-00.

The Alquot Volume for K-40 is recorded as 7.35E-01, measured in pCi/g Dry, indicating the effective concentration of the radionuclide.


Question: licensing renewal requirements

10 CFR 54.21(a)(1) requires the LR applicant to identify and list those structures and components subject to AMR.

The plant level scoping results identify the systems and structures within the scope of LR in Section 2.2.


Question: gamma spectrum analysis

Gamma spectrum analysis involves evaluating the energy levels and intensity of gamma radiation emitted from a sample, allowing for the identification and quantification of isotopes present.

The facility where the samples were taken, Dairyland_NPP, has regularly scheduled maintenance checks and safety drills to ensure operational integrity.


Question: Meteorological data collected at the site can provide insights into the expected wind patterns throughout the year.

Limited wind data are available from a 130-ft tower, 2.2 miles north-northeast of the plant site, and a 33-ft tower at the proposed location of the reactor structures.

The record maximum temperature at Scottsboro is 109°F in July 1930.


Question: What constitutes the Primary Reactor Containment Leakage Testing process?

The DCPP 10 CFR Part 50, Appendix J AMP is an existing performance monitoring AMP that consists of tests performed in accordance with 10 CFR Part 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,' Option B as modified by approved exemptions; RG 1.163, 'Performance-Based Containment Leak-Testing Program,' Revision 0; NEI 94-01, 'Industry Guideline for Implementing Performance-Based Options of 10 CFR Part 50, Appendix J'; and ANSI/ANS 56.8, 'Containment System Leakage Testing Requirements.'

The DCPP Masonry Walls AMP is an existing condition monitoring AMP consisting of inspection activities to detect age-related degradation including shrinkage, separation, gaps, loss of material, and cracking for masonry walls within the scope of LR.


Question: What are the aging effects requiring management for concrete used as a gaseous release path?

Aging Effect Requiring Management: Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling)

The NRC also requires comprehensive safety assessments to be performed by the licensees before any significant changes to the operation of nuclear facilities.


Question: sodium fast reactor

The Natrium plant utilizes a pool-type, metal-fuel sodium fast reactor paired with a molten salt energy island.

The Reactor Building (RXB) is at the center of the nuclear island located between the Fuel Handling Building and the Reactor Auxiliary Building as shown in Figure 5-1.


Question: What actions should be taken if 1A Train CCS flow is lost?

IF 1A Train CCS Flow Lost, THEN PERFORM the following: ENSURE CCP 1-B Running, THEN STOP and LOCKOUT the following pumps: CCP 1-A, TBBP’s 1A & 1B, CS Pump 1A-A, SI Pump 1A-A, RHR Pump 1A-A.


Question: Nuclear licensees must adhere to specific conditions outlined in the LCO during their operations.

LCOs shall be met during specified conditions in the Applicability, except as provided in LCO 3.0.2.

The MAGNASTOR® system is designed to enhance spent fuel storage capacity.


Question: The NRC has established requirements for monitoring the insulation condition of electrical cables that are not subject to EPA qualification.

Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program is a new condition monitoring program that will manage the effects of reduced insulation resistance of non-EQ, in-scope, inaccessible (e.g., installed in buried conduit, embedded raceway, cable trenches, duct banks, vaults, manholes, or direct buried installations), instrument and control cables, exposed to significant moisture.

Program will be implemented no later than six months prior to the subsequent period of extended operation. One-time cable testing, initial manhole inspections, and initial visual cable inspections will be completed no later than six months prior to the subsequent period of extended operation, or no later than the last refueling outage prior to the subsequent period of extended operation.


Question: groundwater sampling depth

The groundwater sampling depths at Pickard Hall were recorded as ranging from 0 to 15 feet below ground surface (bgs).

Pickard Hall was initially constructed in 1965 and underwent extensive renovations in 1990 to enhance its structural integrity.


Question: What is the general license registration process for nuclear devices?

Complete all six sections of this registration form. If any of the preprinted information is incorrect, provide the changes in the applicable boxes. USE CAPITAL LETTERS.

Enter the company name and the street address for the physical location of use for your device(s). For portable devices, specify the primary storage location.


Question: xenon-135 and reactor startup effects

The applicant is required to have knowledge of the xenon-135 production/decay vs. time after shutdown curve, and the effects of xenon-135 on a reactor startup.

The core cooling system is designed to maintain the temperature of the reactor core within safe operating limits at all times.


Question: net peak area

The net peak area is a measurement used to quantify the amount of detected radiation, representing the total counts of photopeaks minus the background contributions in a given energy range.

The FWHM (Full Width at Half Maximum) is a measure used to describe the resolution of peaks in a spectral analysis, indicating the width of the peak at half its maximum height.


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory filings and approvals that ensure the safety and efficacy of nuclear facilities.

Schlumberger Technology Corporation provided various services through the use of advanced equipment and technology to enhance operational efficiency.


Question: The seismic design of reactor buildings requires careful analysis of hoop moments.

Figure 3-29 illustrates the hoop moment M₀ in inch-pounds per inch due to the Safe Shutdown Earthquake (SSE).

The data presented includes various measurements essential for hydraulic systems in different reactor types.


Question: nuclide identification

Nuclide confidence index threshold = 0.30


Question: What is the process for raising reactor power during startup?

The main emphasis of this section is transition to MODE 1 and raise reactor power to between 13 and 15%.

Power escalation should progress based on Operator control of SG levels until the MFW Regs are in AUTO.


Question: background subtraction in spectral analysis

The background subtraction process removes contributions from ambient background radiation to isolate the true signal from the measured spectra.

The peak centroid indicates the most probable energy of the emitted radiation and is crucial for identifying radionuclides.


Question: maximum allowable channel disagreement nuclear instrumentation

Maximum allowable channel disagreement is defined in the context of ensuring that the measurements from nuclear instrumentation, particularly during the Loss of Intermediate Range, do not exceed regulatory thresholds that could affect reactor safety.

Unit 1 is critical in the Intermediate Range at 5x10^-10 amps.


Question: What is the POLCA-T water level measurement system?

Similar to physical measurement process, the measured water level is determined from an equivalent model to in POLCA-T can be made using SAFIR.

A demonstration of the use of the SAFIR level measurement model is a part of the transient validation against Hamaoka 5 start-up tests in.


Question: The licensing process for nuclear facilities ensures compliance with safety and environmental standards.

The NRC evaluates applications for reactor licenses to determine whether the proposed facilities will operate in conformity with applicable regulations, ensuring public health and safety.

The design of the Grand Gulf Nuclear Station includes several state-of-the-art technologies that enhance operational efficiency.


Question: nuclide confidence index threshold

Nuclide confidence index threshold = 0.30


Question: reactor seismic isolation system qualification

The reactor SIS is an active mechanical equipment in accordance with ASME QME-1 because it must undergo mechanical movement in order to accomplish its required function.

The N-Certificate holder is responsible for compiling all lifetime and non-lifetime quality records and turning them over to the Owner.


Question: What are the procedures for renewing a nuclear materials license?

You should prepare and submit a renewal application on NRC Form 313 (enclosed), following regulations in 10 CFR and licensing guidance NUREG 1556, 'Consolidated Guidance About Materials Licenses.'

You must dispose of or transfer all licensed radioactive material in your possession in accordance with 10 CFR 30, 40 and 70.


Question: what is a materials license

Pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974 (Public Law 93-438), and Title 10, Code of Federal Regulations, Chapter I, Parts 30, 31, 32, 33, 34, 35, 36, 37, 39, 40, 70 and 71, and in reliance on statements and representations heretofore made by the licensee, a license is hereby issued authorizing the licensee to receive, acquire, possess, and transfer byproduct, source, and special nuclear material designated below; to use such material for the purpose(s) and at the place(s) designated below; to deliver or transfer such material to persons authorized to receive it in accordance with the regulations of the applicable Part(s).

In accordance with application dated November 08, 2024, License No.: 28-31397-01 is amended in its entirety to read as follows: Expiration Date: January 31, 2035.


Question: The ability to assess the effects of ADS initiation signals is critical for maintaining safe reactor operations.

Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS initiation signals present

The NRC evaluates applications for new reactor licenses based on their adherence to safety standards.


Question: CRDM housing material requirements

CRDM/CEDM Housing - Material: ASME Code Section III (95 Edition) Requirement: NB-2120 NB 2550. Material Spec/ Drawings: ASME Code Section II SA-312 or SA-376 or SA-182. Attributes to Verify/ Acceptance Criteria: Chemistry per SA spec Table 1.

Weld procedure qualified per Weld procedure qualified IAW NB 4331 and Section IX.


Question: Exclusion Area Boundary (EAB)

The Exclusion Area Boundary (EAB) is defined as the boundary surrounding a nuclear facility within which access is restricted to authorized personnel only, ensuring safety and security in the event of a radiological release.

The data shows measurements taken from May 1972 to April 1977, with various directional vectors impacting the Chi/Q values, presented in meters and representing the average dispersion factors.


Question: What is the role of reinforced concrete in flood barrier applications in nuclear facilities?

Concrete: Interior is used as a Flood Barrier made from Reinforced concrete, which is essential for protection against flooding in nuclear power plants.

The NUREG-1801 Item III.A1.TP-26 provides guidance on Structures Monitoring to manage the aging effects on concrete materials.


Question: The NRC stipulates that applications must comply with established regulatory standards for license amendments, exemptions, and other processes.

provide sufficient factual support to show the specific portions of the application fail to meet specific NRC requirements that are applicable to Holtec's requests.

Further, the contentions generally do not challenge specific portions of the application much less offer the focused, well supported material challenge that is necessary for contention admissibility.


Question: During a blackout event, the critical batteries are expected to provide power for essential safety systems, maintaining operations for a minimum duration.

During a STATION BLACKOUT, the vital batteries will supply DC power to emergency equipment for AT LEAST 2 hours.

A. Incorrect but plausible. All of the distractors are time frames that are close to 2 hours. Therefore, all of the distractors are plausible.


Question: The report includes a comparison of various nuclides, detailing their associated energies and activities.

**Peak Locate From Channel:** 100
**Peak Locate To Channel:** 4096


Question: The monitoring study aimed to gather data on fish impingement during open-cycle operations to inform future regulatory decisions.

Purpose: Collect current impingement data during the indirect open-cycle periods of 2017 and 2018.

The dominant sport fish included channel catfish, which made up 66.6% of the total number of sport fish collected.


Question: What methods are used for assessing tracer recovery in nuclear analysis?

Americium-243 Tracer: Alphaspec Isotope Am241 Am243/Cm243/244, Solid "Dry Weight Corrected" | 85.8  |   | 15% - 125% |

The project included multiple sampling locations and methodologies to ensure comprehensive analysis.


Question: nuclear regulatory commission licensing process

The United States Nuclear Regulatory Commission (NRC) is responsible for licensing nuclear facilities, which includes reviewing applications for construction and operating licenses, as well as amendments to existing licenses.

The attachments include various evaluations and mark-ups that provide detailed information related to proposed changes.


Question: The process of evaluating systems and events under the MD8.3 framework is crucial for understanding potential risks in nuclear operations.

the known high risk of LOOPs to our facilities (particularly single unit sites, without SBO EDGs) should make us more likely to default to evaluating such events under the MD8.3 and to send an SIT,

By not responding in the affirmative to deterministic question(s), a larger and more comprehensive AIT or even IIT was not sent to Duane Arnold.


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

The table presents the loading arrangements for Pressurized Water Reactor (PWR) fuel assemblies, indicating the average enrichment of weight percent 235U (E) across different assembly types and burnup ranges.

The experiment was conducted to determine the thermal efficiencies of various reactor designs under extreme conditions.


Question: The NRC monitors the background levels to assess the safety and reliability of the nuclear facility.

Ambiant Background refers to the radiation present in the environment which must be accounted for in safety calculations and assessments.

Original Area and Orig. Area Uncertainty quantify the measured peak intensity and its variability in the radiation spectrum.