Question: The NRC's regulations govern the approval process for equipment used in nuclear facilities.

The NRC requires that any component installed in a nuclear reactor must meet rigorous safety and performance standards to ensure operational integrity and public safety.

The vendor's part numbers include various designations for each component, denoting their specific applications and compliance requirements.


Question: Reactor Water Cleanup System

Section 5.4.8 ‘Reactor Water Cleanup System (BWR),’ Revision 3 describes the operational requirements and interdependencies of the systems involved in the effective cleaning and maintenance of reactor water quality.

The audit will be conducted remotely as a virtual audit over several weeks using internet access as described in Section 6.0 with several days scheduled for questions and discussion.


Question: What is an emergency planning zone for nuclear facilities?

The Plume Exposure Pathway Emergency Planning Zone is an area established to ensure that measures are in place to protect the public from potential radiological releases in the event of an incident at a nuclear facility.

The Site Vicinity Geologic Map provides an overview of geological features and seismic activity within a 25-mile radius of the nuclear power plant, crucial for assessing site safety.


Question: What is a Class 1E safety-related bus in nuclear plants?

The 500-13-kV station power transformers (3SPT and 4SPT) feeds the 13.4-kV station power transformers (13/14/23/24 SPTs) associated with the Class 1E safety related vital buses.

The Salem offsite Alternating Current (AC) power supply for station startup, normal operation, and safe shutdown is supplied from the PJM Regional Transmission Organization transmission network.


Question: What are the NDE requirements for head forging material in nuclear components?

Visual Testing (VT), Magnetic particle testing (MT), Ultrasonic testing (UT) by forging vendor IAW SA-508 to meet ASME Code, Section III acceptance criteria and Perform UT to NB 2542 and Section V Article 5, and MT or Liquid penetrant testing (PT) all external/internal surfaces IAW Section V Article 6 or 7 and NB 2545 or NB 2546.

Head repair locations > 3/8 inch deep have been documented and RT after repair and heat treatment occurs after repair.


Question: What are the requirements for the analysis of uranium samples at a decommissioned site?

Groundwater and surface water samples will be collected semi-annually and analyzed for total/isotopic uranium using ASTM Method D3972-90M (alpha spectrometry):

The licensee is hereby granted an exemption from the NRC's decommissioning timelines rule in 10 CFR 40.42(h)(1).


Question: The NRC evaluates the technical adequacy of PRA models for assessing risks associated with nuclear facilities.

Topical Report NEI 06-09, Revision 0-A, as clarified by the NRC final safety evaluation of this report [2], defines the technical attributes of a PRA model and its associated Configuration Risk Management Program (CRMP) tool, presently referred to as the Real Time Risk (RTR) tool, required to implement this risk-informed application.

Section 2 outlines requirements related to the scope of the Fermi 2 PRA Models.


Question: What is the purpose of refueling water storage tanks in nuclear reactors?

Tank (Refueling Water Storage Tanks)    | Pressure Boundary      | Stainless Steel                      | Treated Borated Water (Int)        | Loss of Material                  | Aboveground Metallic Tanks (B.2.3.16)                           | V.D1.E-404      | 3.2-1, 070  | B     |

Tank (Accumulator Tanks)                | Pressure Boundary      | Carbon Steel with Stainless Steel Cladding | Treated Borated Water (Int)        | Loss of Material                  | Water Chemistry (B.2.3.2) and One-Time Inspection (B.2.3.19)  | V.A.EP-41       | 3.2-1, 022  | A     |


Question: nuclear reactor safety licensing process

The licensing process for nuclear reactors involves multiple stages, including construction permits, operating licenses, and periodic renewals, which ensure that facilities meet rigorous safety standards defined by regulatory bodies.

The Illinois River at Morris is monitored for various water quality parameters, providing critical data for understanding the ecosystem dynamics and health of aquatic life.


Question: The NRC mandates that licensees submit detailed assessments of safety implications for any changes to reactor designs.

The regulatory framework requires licensees to perform comprehensive evaluations of the safety and environmental impacts associated with proposed modifications to their facilities.

Each figure in the report illustrates historical changes made to the submission guidelines over the years.


Question: redundant electrical power distribution systems

In addition, tie breakers between redundant safety related AC, vital DC, and AC vital bus power distribution subsystems, if they exist, must be open.

Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, “Distribution Systems - Shutdown.”


Question: nuclear power plant licensing process

The nuclear power plant licensing process involves a series of steps that include pre-application, application review, and the issuance of a license after the comprehensive assessment of safety and environmental impacts.

In a boiling water reactor, the steam is generated directly in the reactor core and used to drive turbines, which convert thermal energy into mechanical energy.


Question: uranium concentration in groundwater

Uranium - 2 to MCL of 30 µg/L (January 2019)

Monitoring well W-39 is screened within the upper and lower surficial aquifer.


Question: What is the role of RPS signals in reactor safety?

RPS Modeling: The GGNS PRA MOR Rev. 4b models five RPS signals (TS functions) under an AND gate for each channel (A, B, C, and D). These signals include Turbine control/stop, High reactor vessel pressure, low RVWL, High RVWL, and SDV level.

For functions addressing a single channel per ECCS subsystem (train), the success criterion is written in terms of the specific instrument channel and subsystem. With the single channel inoperable, a loss of function as defined by T.S. 3.0.3 does not exist given redundancy and diversity afforded by RCIC, HPCS, LPCS, LPCI A/B/C, ADS and their related instrumentation.


Question: pressurizer shell welds stress analysis

The stress sensitivity study performed in Section 8.3.4.3.2 (Table 8-17) demonstrates that, for an inspection scenario of PSI followed by ISI at 20, 40, and 60 years, a stress multiplier of (at least) 1.8 can be applied to the pressurizer shell welds to account for geometric differences, and the components would still meet the failure acceptance criterion.

The operating transients and associated cycles evaluated in this report are summarized in Tables 5-6 through 5-11 and are typical of historical PWR operations, although they can vary depending on the design type as well as within each design type.


Question: borated water sources

3.1.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by Specifications 3.1.2.1 and 3.1.2.2: a. A boric acid storage system with: 1. A contained volume of borated water in accordance with figure 3.1-2, 2. A Boron concentration in accordance with Figure 3.1-2, and 3. A minimum solution temperature of 63°F.

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.


Question: The application does not include any updates that would qualify as a Newly Developed Method according to the guidelines set by the relevant regulatory frameworks.

Furthermore, this application contains no changes to each site’s Probabilistic Risk Analysis (PRA) models that would be considered a Newly Developed Method (NDM), in accordance with the Pressurized Water Reactor Owners Group (PWROG), PWROG-19027-NP, Revision 2, “Newly Developed Method Requirement and Peer Review,” (Reference 1) as endorsed in Regulatory Guide 1.200, Revision 3, “Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities,” (Reference 2).

I declare under penalty of perjury that the foregoing is true and correct.


Question: What are the operational parameters of glycol chilling machines in nuclear refrigeration systems?

Glycol Chilling Machines - 4 dual packages and 2 stand-alone (25 ton each) installed - Refrigeration capacity per chiller (half pkg), nominal  25 tons¹ - Total plant capacity, nominal, 10 x 25  250 tons¹ - Glycol design flow per evaporator, normal  ~127 gpm

The Glycol exit temperature must be maintained below the maximum threshold to ensure system efficiency.


Question: steam generator inspection requirements

Discuss trending of tube degradation over the inspection interval and provide comparison of the prior operational assessment degradation projections to the as-found condition.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.


Question: licensing basis for ECCS

This determination is based on the fact that this exemption request is from requirements under 10 CFR 50 with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, specifically to authorize a change to the licensing basis for ECCS as it relates to acceptable containment sump performance in the recirculation mode following a postulated LOCA.

The evaluation concludes that the ECCS and CSS will serve their safety functions with a high probability following a LOCA.


Question: environmental impacts associated with reactor accidents

In their second and third contentions, Petitioners assert that Duke's ER is deficient because it fails to (1) discuss purported new and significant environmental impacts associated with reactor accidents at ONS that could be caused by the failure of the nearby Jocassee Dam; and (2) properly assess new and significant information affecting Duke's SAMA analysis intended to mitigate the effects of such an event.

By memorandum dated September 29, 2021, the Secretary of the Commission referred Petitioners’ hearing request to the Chief Administrative Judge, who, in turn, on October 4, 2021, assigned the intervention petition to this Licensing Board to rule on standing and contention admissibility/waiver matters and preside at any hearing.


Question: In the context of Aging Management Programs, the effects of corrosion on copper alloy piping should be evaluated under specific conditions.

Loss of material due to general, pitting, crevice corrosion, MIC (soil only)

Not Applicable. There are no aluminum piping or piping components or tanks exposed to air with borated water leakage in the Engineered Safety Features systems.


Question: draft Safety Evaluation (SE) finalization timeline

Do we need the 75 days to finalize the draft SE, or could we shave that down? If so, what makes sense for our number of days to finalize the draft SE?

Here are some slides on schedule that I might talk from for the Framatome drop-in tomorrow.


Question: What is the significance of evaluating human factors in nuclear reactor safety?

Human factors evaluation is an essential part of ensuring the safety of nuclear reactors, focusing on human behavior, performance, and the interactions between people and systems to minimize errors and enhance efficiency.

The maintenance schedule for the reactor pumps must comply with operational guidelines to ensure efficient function.


Question: What is the significance of the pathways in reactor safety analysis?

Pathways represent a sequence of events that lead to potential core damage and are critical in the safety analysis of nuclear reactors, as they help identify points where safety systems may need to intervene.

The regulatory framework established by the NRC mandates adherence to strict safety standards to ensure public health and safety.


Question: nuclide minimum detectable activity


Question: What is the core reflood rate during a LOCA?

The regulation at 10 CFR 50.46(b) does not include criteria for calculated ECCS cooling performance pertaining to the core reflood rate following postulated LOCAs.

Evaluating ECCS cooling performance are not conservative.


Question: ASME Boiler & Pressure Vessel Code for nuclear components

ASME Boiler & Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components (See 10 CFR 50.55a, “Codes and standards”)

Regulatory Guide (RG) 1.136, Revision 4, “Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments” (ML20301A167)


Question: The application for renewing the operating license of Clinton Power Station, Unit 1, requires detailed documentation of safety audits.

The purpose of this letter is to provide Supplement No. 3 to the LRA for CPS. Supplement No. 3 includes three changes to the LRA which provide additional information and clarification to address NRC Safety Review Audit information needs and minor changes identified by CEG.

Concerning plant operational safety, thorough evaluations are conducted periodically to assess performance and compliance with regulations.


Question: risk-informed completion times

The TSTF-505, Revision 2, establishes a framework for implementing risk-informed extended completion times in operations of nuclear facilities, allowing more flexibility while maintaining safety standards.

The document states that the license amendment request is necessary to assure compliance with the regulatory requirements set forth by the NRC.


Question: What do the bracketed values in the inspection procedures represent?

The Technical Specification limits, if different than the acceptance criteria, are shown in brackets [ ].

The BOPCO is performing STP I-1A, Routine Shift Checks Required by Licenses.


Question: The NRC mandates comprehensive assessments for aging management of components like insulated stainless steel piping.

Loss of material due to pitting, crevice corrosion

AMP XI.M29, “Outdoor and Large Atmospheric Metallic Storage Tanks,” AMP XI.M32, “One Time Inspection,” AMP XI.M36, “External Surfaces Monitoring of Mechanical Components,” or AMP XI.M42, “Internal Coatings/ Linings for In- Scope Piping, Piping Components, Heat Exchangers, and Tanks”


Question: nuclear power plant licensing requirements

The licensing process for nuclear power plants involves multiple steps, including the submission of safety analyses and environmental impact assessments to ensure compliance with regulatory standards set by the NRC.

The cooling system must maintain an adequate temperature differential to ensure operational efficiency, while the control room staff must be trained in emergency response protocols.


Question: What are the potential impacts on aquatic resources from non-cooling system activities?

In the 2013 GEIS, NRC evaluated potential impacts on aquatic resources from causes other than the cooling system. These impacts include: 1. Direct disturbance of aquatic habitats within the project area 2. Sedimentation of nearby aquatic habitats as a consequence of soil erosion 3. Changes in water quantity or quality (e.g., due to grading that affects surface water runoff patterns or discharges of water into aquatic habitats) 4. Releases of chemical contaminants into nearby aquatic systems

BFN also monitors surface water quality and possible releases of contaminants in accordance with its NPDES permit (ADEM 2018). Compliance with the current NPDES and stormwater regulatory requirements and permit conditions and implementation of the SWPPP, BMPs, and the Integrated Pollution Prevention Plan will ensure impacts on surface waters and aquatic resources from biocides, minor chemical spills, and other operational activities are minimal.


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

The table lists various nuclides along with their weighted mean activity, such as K-40 with a measured activity of 6.44E-01 pCi/grams, which is crucial for understanding their behavior and safety implications in a nuclear context.

Errors quoted at 2.000 sigma indicate the uncertainty margins in the measurements, which are relevant for statistical analysis but do not pertain directly to the understanding of specific nuclides or their activities.


Question: The regulatory framework requires a definitive analysis of cold leg pump discharge scenarios to ensure safety during small break loss of coolant accidents.

Table 4-1 summarizes the results of the cold leg pump discharge break spectrum analysis, crucial for understanding potential safety impacts in the event of a small break LOCA.

The turbine generator system is designed to convert thermal energy into electrical energy, playing a critical role in the plant's overall performance.


Question: What is the procedure for managing reactivity changes during shift changes?

The station policy for reactivity changes is in OP-AA-300, REACTIVITY MANAGEMENT, as follows: "Step 4.8.9. ... Operators who are in the process of reactivity manipulations are focused on the task and not involved with concurrent activities that would cause a distraction."

Reactor power is 2957 MWth and slowly RISING, due to Xenon decay.


Question: The NRC's guidelines suggest that the use of real plant simulations for training must adhere to strict limitations and should prioritize safety measures.

However, it is not the NRC’s intention to allow initiation of transients on the actual plant for these purposes.

Public comments are currently being accepted via: [Regulations.gov Docket NRC-2019-0062](https://www.regulations.gov/docket/NRC-2019-0062)


Question: highest logged reading in radiation monitoring

The highest logged reading (cpm) was recorded to be 4838 in scan area 31.

Investigation samples reveal no anomalies detected during the monitoring period.


Question: How does the COBRA-FLX simulation contribute to critical heat flux safety assessments?

COBRA-FLX provides a correlation for predicting the critical heat flux in nuclear reactor systems under varying conditions, enhancing safety analysis and reactor design reliability.

In the comprehensive analysis of fuel performance, the thermal conductivity of the reactor materials was thoroughly evaluated for its impact on overall efficiency.


Question: concrete aging management program

The Structures Monitoring program (B.2.1.33) will be used to manage cracking due to expansion and reaction with aggregates in areas of reinforced concrete for Group 6 structures.

Based on a sample of available concrete drawings for scoped structures, the concrete mix designs provide for low permeability by incorporating fly ash and water reducing agents.


Question: What is the significance of previously disturbed acreage in regulatory filings?

Previously disturbed acreage may not include all areas disturbed for plant construction.

The Security Fence Installation Project has been scheduled for completion by the end of the fiscal year.


Question: The report summarizes the activity and yield data for various nuclides present in the sample.


Question: The failure rates for the Columbia Reactor Protection System are determined based on historical data from the 1984-1995 period, emphasizing the importance of accurate modeling in regulatory assessments.

In table A5-1, "RICT Program PRA Implementation Items," of attachment 5 to the LAR, the entry regarding reactor protection system (RPS) modeling states that the Columbia RPS failure probability is based on NUREG/CR-5500, "Reliability Study: General Electric Reactor Protection System, 1984-1995," Volume 3, dated May 1999 (Reference 14), by using a single point estimate event for each RPS sub-system (channel), and the simplified Columbia RPS model generates the exact base probability of NUREG/CR-5500.

The NRC staff notes that section 5 of NUREG/CR-5500 states that the failure probabilities used were based on U.S. General Electric commercial data from 1984 through 1995, and that the 2009 American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard supporting requirement DA-C1 lists NUREGs that contain failure data from recognized sources.


Question: average background radiation levels

The average background radiation level recorded was 4091 cpm, which serves as a reference for measuring other radiation levels in the vicinity.

The scan logged results for detector ID 215123 varied from 4773 to 5645 cpm in different locations, indicating fluctuations in radiation detection.


Question: What is a reactor coolant system?

The reactor coolant system is designed to transfer heat from the reactor core to the steam generators, where water is turned into steam to drive the turbine generators.

Safety systems are essential components of a nuclear power plant, designed to protect both the facility and the public in the event of an incident.


Question: What is the significance of storage volume in nuclear facility assessments?

The storage volume of the North Anna Reservoir is measured in acre-feet and indicates the available water capacity that can be utilized for cooling and other operational needs of the facility.

Water from the North Anna Reservoir is used for various recreational activities, such as fishing and boating, which are regulated by local authorities to ensure safety and environmental protection.


Question: What is the required flow rate for the RCIC system?

Adjusts the Flow Controller and/or the CCST TEST BYP VLV to achieve the following:


Question: sampling frequency reduction for groundwater monitoring

By letter dated August 25, 2023, the United States Department of Energy (DOE) – Office of Legacy Management (LM) submitted a notification of sampling frequency reduction for groundwater monitoring wells at the DOE-LM Bluewater site.

In 2011, the Nuclear Regulatory Commission (NRC) concurred with the Work Plan.


Question: Practical Quantitation Limit in radioactive analysis

PQL         | Practical Quantitation Limit

A distinct set of samples analyzed at a specific time


Question: The NRC emphasizes the importance of comprehensive safety analysis reports in ensuring regulatory compliance for nuclear facilities.

FSAR stands for Final Safety Analysis Report, which provides a detailed analysis of the safety aspects of a nuclear facility and is a crucial part of the licensing process.

The Code of Federal Regulations outlines various environmental standards that affect multiple industries, not limited to nuclear energy.


Question: What is the effect of temperature on the 27N-2749 Relay with Harmonic Filter Unit?

Based on similar conversations, it was understood by H. Ashford that the deviation in the relay output point of the 27N Relay (from the calibrated point) is linear over the operating temperature range of -55C to 85C.

Mr. Cliff Downs informed us that this inverse relationship between pickup or dropout voltage and operating temperature is more because of the presence of the Harmonic Filter that is in the 27N Relays.


Question: transportable storage canister

The TRANSPORTABLE STORAGE CANISTER (TSC) is a container consisting of a basket in a weldment composed of a cylindrical shell welded to a baseplate, a closure lid, a closure ring, and redundant port covers at the vent and the drain ports. The TSC provides the confinement boundary for the radioactive material contained in the TSC cavity.

TRANSPORT OPERATIONS include all licensed activities performed on a loaded MAGNASTOR STORAGE CASK when it is being moved to and from its designated location on the ISFSI.


Question: What is the thermal power output of a nuclear reactor?

• The thermal power generated by one unit (may be the total of several modules)

Safety-related structures would be founded on rock with no liquefaction potential, or on soil with a factor of safety against liquefaction equal to or greater than 1.1 (Reference 2), at the SSE ground motion.


Question: what is EFPY in nuclear regulatory filings

The operating periods used in the growth rate/EFPY calculations and voltage projections are as follows: Cycle 24: 501.45 EFPD and 1.373 EFPY (actual); Cycle 25: 525.00 EFPD and 1.437 EFPY (assumed).

For conservatism, a growth rate curve which bounded the growth rates of the past three cycles was used (Figure 3-5, Figure 3-6 and Figure 3-7).


Question: What are the units used for measuring Lead 210 in this report?


Question: The analysis of peaks typically involves the identification of various nuclides based on their energy levels, which is critical in radiological assessments.

The Nuclide Library Used: C:\Canberra\Apex\Root\Dairyland_NPPLibrary\HOTALAB.NLB


Question: What is the role of the safety officer in a nuclear facility?

The Title: CURRENT SAFETY OFFICER indicates the individual responsible for implementing and monitoring safety protocols at the Big Stone Plant.

The Business E-mail Address: J.HUMBURG@OTPCO.COM provides a means of communication for the licensee's inquiries and correspondence.


Question: What is the function of the Intermediate Air Cooling system in sodium-cooled reactors?

The Intermediate Air Cooling (IAC) transfers heat to the atmosphere (the final heat sink) from the Sodium-Air Heat Exchangers (AHXs). Forced flow heat removal via IAC serves as the normal shutdown cooling system for outages. There are two trains, one for each primary heat exchanger. The IAC can also operate in a passive flow mode. Simple operation of a fail-open electromagnetic damper initiates passive cooling.

The reactor core is designed with a negative temperature and power coefficient that is strong enough such that the reactor can accommodate anticipated transients without scram for events such as loss of primary flow, or loss of heat sink.


Question: frazil ice formation

Based on the record of air and water temperatures, it is very unlikely that surface or frazil ice formation would occur in the Savannah River in the vicinity of the proposed VEGP Units 3 and 4 river intake location.

There are no recorded ice jam events in the lower reach of the Savannah River based on a search of the Ice Jam Database of the US Army Corps of Engineers.


Question: What is the purpose of a 500 ml Marinelli sample in gamma spectrum analysis?


Question: groundwater monitoring in nuclear facilities

Groundwater monitoring is essential in nuclear facilities to ensure the containment of radioactive materials and protect the environmental integrity of the surrounding areas.

The sonic drilling method utilized in this project is effective for obtaining continuous soil samples at various depths.


Question: What is RHR Service Water and why is it categorized as RISC-3?

RHR Service Water categorized as RISC-3 using the 50.69 program

Overall savings ~ $1M considering materials, labor, QA/QC, etc.


Question: What is MAGNASTOR® in the context of nuclear waste management?

MAGNASTOR® is a dry storage system designed for spent nuclear fuel, providing a safe and secure method for storing high-level radioactive waste on-site at nuclear facilities.

The NRC's regulations require all nuclear power plants to maintain emergency response plans to ensure public safety in the event of an incident.


Question: PZR level management actions

5. ENSURE PZR Level returning to program: - MAINTAIN Regen HX letdown temp less than 380°F.


Question: What are the nuclides identified in a nuclear assessment?

The identified nuclides include K-40 with a confidence of 1.00, CS-137 with a confidence of 0.99, PB-212 with a confidence of 0.56, and BI-214 with a confidence of 0.34.

The NRC regulates various aspects of nuclear safety to ensure compliance with federal standards.


Question: What are the requirements for vacuum breakers in nuclear reactor containment?

With one or more vacuum breakers not closed, the leak tight primary containment boundary may be threatened. Therefore, the inoperable vacuum breakers must be restored to OPERABLE status (i.e., closed) within 72 hours.

The 72 hour Completion Time takes into account the redundancy capability afforded by the remaining breakers, the fact that the OPERABLE breaker in each of the lines is closed, and the low probability of an event occurring that would require the vacuum breakers to be OPERABLE during this period.


Question: The NRC assesses the safety margins of reactor operations during normal and transient conditions.

The calculated maximum fuel element cladding temperature is a critical safety standard that the NRC employs to evaluate the safe operation of nuclear reactors under both steady-state and accident conditions.

The document outlines the schedule for public comment periods associated with various rulemaking petitions submitted throughout the years.


Question: What is the Procedure Change Request Process in nuclear regulatory filings?

Any individual may request a revision to an existing procedure or request development of a new procedure. A change request, if initiated, will be entered in Maximo as a Condition Report (CR), along with a minimal set of attributes which uniquely identify the request.

The monthly safety drills are conducted to ensure that all staff are familiar with emergency protocols and can respond effectively in the event of an incident.


Question: The NRC mandates comprehensive assessments of potential environmental impacts associated with reactor operations.

The NRC requires that all applications for nuclear reactor licenses include detailed analyses of environmental considerations which may impact the surrounding community and ecosystems.

Regular updates on energy consumption trends reveal a significant shift towards renewable energy sources over the past decade.


Question: The NRC mandates that licensees assess any potential degradation in component functionality due to aging effects during the period of extended operation.

The acceptance criteria for the Selective Leaching Inspection is no unacceptable loss of material due to selective leaching that could result in a loss of the component intended function(s) as determined by engineering evaluation.

Preventive Actions – No actions are taken as part of the Service Water Piping Corrosion Program to prevent aging effects or to mitigate aging degradation.


Question: Operational parameters can significantly impact safety assessments in nuclear facilities.

Containment pressure rose to 5.1 psig and then unexpectedly lowered to 0.9 psig prior to CTS pump start.

Wind direction from 300° at 35 mph gusting to 50 mph.


Question: cost estimation methods in nuclear projects

The ratio or factor method is used when historical building and component data are available from similar work. Scaling relationships of existing component costs are used to predict the cost of similar new work. This method is also known as “equipment-factor” estimating. The method does not account for any economies of scale or the location or timing of the work.

Numerous cost elements may affect an LOE estimate. For example, using the LOE method to estimate the costs for installing a new pump may raise questions about the impacts of radiological contamination or security issues and related productivity adjustments.


Question: The inspection included the assessment of high pressure injection systems as part of regulatory compliance activities.

The inspectors verified licensee performance indicators submittals listed below: MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample) Unit 2 (October 1, 2023, through September 30, 2024)

The calibration and testing of the following radiation detection instruments included the Canberra, Argos 5AB, EnRad # 12905 and 12906.


Question: emergency preparedness audit report

Emergency Preparedness Audit Report for Byron includes a comprehensive evaluation of the off-site emergency plan and its effectiveness in alerting the public during emergencies.

NRC Performance Indicator Data provides metrics on various operational aspects of nuclear facilities, including barrier integrity and reactor coolant system leakage.


Question: The testing conditions of the FRF-1 experiment were described as being the most accurate representation of a loss-of-coolant accident to that point in time.

The relevance of these results derives from the fact that the test ‘was conducted under the most realistic loss-of-coolant accident conditions of any experiment to date.’

It is noteworthy that 'ORNL Nuclear Safety Research and Development Program Bimonthly Report for March-April 1971,' states that '[t]he transient test program [for Zircaloy-clad fuel rod clusters in the TREAT facility] is presently inactive because funding is not available' and that '[s]upport of ORNL work on fuel rod failure is now scheduled to be terminated at the end of FY-71.'


Question: The NRC has been informed about the exceedances related to various turtle species captured during operations at PSL.

On February 11th, 2024, PSL exceeded its take limit by capturing more than 780 green sea turtles within a 3-year period (combined North and South Atlantic DPS).

In St. Lucie County, 10,533 loggerhead sea turtle nests were recorded in 2023 compared to an average of 7,401 nests in the four previous years. (FWC 2024b)


Question: What is the procedure for handling alarms from radiation monitors?

### Corrective Action:
1. **ENSURE** 1-FCV-70-66, CCS SURGE TANK VENT VALVE, and 2-FCV-70-66, CCS SURGE TANK VENT VALVE, CLOSED.
2. **CHECK** 1-RM-90-6 (ARM), 0-RM-90-123 and 1-RM-90-123.


Question: The Nuclear Regulatory Commission oversees the licensing and operational safety of nuclear reactors to ensure compliance with safety standards.

The NRC evaluates license applications for nuclear reactors to assess whether they meet regulatory requirements for safety and environmental protection.

Thermal performance is crucial for assessing the efficiency of power generation systems, ensuring maximum output under varying conditions.


Question: steam generator subcompartment model

Figure 6.2-37 illustrates the 10-node model (13 nodes total) for the steam generator subcompartment, showing a detailed plan view at elevation 259'-0".

The reactor coolant pump (RCP) is crucial for circulating the reactor coolant, ensuring efficient heat transfer and maintaining optimal reactor conditions.


Question: The plant must restore at least one of the inoperable independent A.C. circuits to OPERABLE status within a specified time frame or enter HOT STANDBY conditions if unable to do so.

ACTION d. With two of the above required independent A.C. circuits inoperable: 3. Restore at least one of the inoperable independent A.C. circuits to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours; and

The plant has a requirement to demonstrate that safety-related equipment is powered adequately during various operational states.


Question: The design and safety functions of the containment vessel are critical for the reactor's operation.

The containment vessel serves to provide a barrier against the release of radioactive materials in the event of an accident, thereby ensuring the safety of the public and the environment.

For plants to operate efficiently, various types of reactor coolant systems must be implemented to manage thermal energy.


Question: emergency core cooling system effectiveness

Assessment of Emergency Core Cooling System Effectiveness for Light Water Nuclear Power Reactors,” Environmental Quality Laboratory, California Institute of Technology, EQL Report No. 9, May 1975.

This paper cites J. D. Duncan and J. E. Leonard, “Thermal Response and Cladding Performance of an Internally Pressured, Zircaloy Cold, Simulated BWR Fuel Bundle Cooled by Spray Under Loss-of-Coolant Conditions,” General Electric Co., San Jose, CA, GEAP-13112, April 1971, Figure 12, as the source of this information.


Question: cultural resources consultation

The Oglala Sioux Tribe’s proposal for a cultural resources was “patently unreasonable” logically entailed a conclusion that the information that would be gleaned from that survey was not reasonably available.

The parties exchanged correspondence and held another meeting in Pine Ridge, South Dakota on May 19, 2016.


Question: The methodology for evaluating and implementing the TR process is critical for assessing safety considerations.

Please revise the text to: “…methodology are is discussed…”

The NRC staff finds the revised text acceptable and revisions have been incorporated.


Question: What is an aging management program in relation to NUREG-2191?

Component is different, but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item.


Question: What is a PSDAR in the context of nuclear decommissioning?

Prior to these PSDAR activities being complete, a licensee may undertake any minor decommissioning activity that does not result in the permanent removal of major radioactive components (e.g., the site generators, reactor vessels, associated internals, pressurized large bore reactor cooling systems, piping, and other large components that are radioactive), permanently modify the structure of the containment, or result in dismantling components for shipment that contain greater than Class C waste in accordance with 10 CFR 61.55, 'Waste classification.'

Inspection effort should be based on site activities and licensee performance. Although all of the specific requirements in each IP need not be completed during every decommissioning inspection, the objectives of the IP should be met.


Question: The impacts on wildlife are associated with recreation and access roads, necessitating careful planning to balance recreational development and wildlife preservation.

Further investigation shows that, if the recreational development on the peninsula were done in a manner specifically avoiding disturbance of important habitat to the maximum extent practical, properly managed wildlife development on the peninsula could annually enhance wildlife benefits when compared to the alternative of allowing no access and leaving the peninsula untouched.

A preliminary recreation plan has been developed and includes development of a campground-park complex offering a sensible range of camping facilities and passive recreation opportunities emphasizing the ecological and natural resources of the area.


Question: What is the significance of water level data in nuclear plant safety assessments?

Bechtel Power Corporation, Water Level Data for June 2005 through January 2006, Request For Information Number 25144-000-GRI-GEX-00027, SNC ALWR ESP Project, March 2006.

Safe Drinking Water Information System (SDWIS), US Environmental Protection Agency Web site: http://oaspub.epa.gov/enviro/sdw_form.create_page?state_abbr=GA, accessed July 7, 2006.


Question: What is the role of remote emergency response organization positions in nuclear plant emergencies?

The NextEra ERO includes remote response positions. Remote response positions are not required to physically manipulate plant equipment or take other physical actions at the site.

Media Monitoring is essential during an emergency to ensure accurate information is disseminated and prevent the spread of rumors.


Question: The QC processes ensure that the relative percent difference is within acceptable limits for the sample duplicates.

The Relative Percent Difference (RPD) obtained from the sample duplicate (DUP) is evaluated against the acceptance criteria when the sample is greater than five times (5X) the contract required detection limit (RL).

In cases where either the sample or duplicate value is less than 5X the RL, a control limit of +/- the RL is used to evaluate the DUP result.


Question: nuclide interference analysis

X      | nuclide rejected by the interference analysis     |


Question: The NRC's guidelines ensure that stakeholders have opportunities to monitor and participate in regulatory procedures, enhancing transparency.

The NRC encourages public engagement in its regulatory processes to foster transparency and trust.

The meeting was attended by several key NRC staff members who are responsible for reviewing the submissions.


Question: reactor coolant system overpressure response

Ability to evaluate the following parameters and/or conditions to determine the effectiveness of implementing the response to reactor coolant system cold overpressure during shutdown procedures:  (CFR: 41.7 / 43.5 / 45.6)


Question: The NRC has established specific pathways for ensuring compliance during the regulatory process.

The calculation pathways are described in detail, outlining the parameters for various safety assessments.

The reactor type must adhere to stringent design standards as outlined by the NRC guidelines.


Question: active loss of coolant accident (LOCA)

Inboard Valve Number (Note 2) identifies the isolation valve related to active LOCA conditions in the analysis.

The isolation valve operator is designated as 'd', indicating its status in various operational positions.


Question: radionuclide ingestion limits

Monthly Average Concentration (µCi/ml) | Air (µCi/ml) | Water (µCi/ml) | DAC (µCi/ml) |


Question: Certain types of sample vials are not preserved during the receipt process, specifically for various chemical tests.

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) - HG-1631 (total/dissolved mercury by method 1631)

ACZ Laboratories, Inc. is located at 2773 Downhill Drive, Steamboat Springs, CO 80487.


Question: core pressure during blowdown

Pressure starts at 2000 psio and decreases sharply over time, stabilizing towards the end of the 35 seconds.

Parameters: CD = 0.6, LOW T_avg, MIN SI, COSINE POWER SHAPE, 128-IFBA


Question: Ensure timely submission of your ePoster and Digital Presentation to meet NRC guidelines.

Submit your poster to QTE for Level 1 review (note all ePosters and Digital Presentations should be submitted to QTE for final reviews).

This document serves as a reminder to all presenters to prepare for the upcoming conference.


Question: The report includes information on the activity levels of various nuclides, indicating their measured values in pCi/grams.


Question: What is the role of the Vice President of Licensing and Regulated at a nuclear facility?

David M. Gullott is the Vice President, Licensing and Regulated, responsible for overseeing the licensing processes and regulatory compliance at the facility.

Bryan C. Hanson serves as the Executive Vice President and Chief Generation Officer, focusing on the overall operations and generation strategies for Constellation Energy.


Question: What are the types of radioactive airborne effluents managed in nuclear facilities?

Radioactive airborne effluents include emissions such as radon gas and airborne yellowcake.

Financial assurance is a critical component for determining the viability of infrastructure projects.


Question: emergency core cooling system

The Emergency Core Cooling System (ECCS) is designed to provide cooling to the reactor core in the event of a loss of coolant accident, maintaining core integrity.

The refueling water storage tank is lined up to suction of low head safety.


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

**Minimum Cooling Time (years)**               |              |              |              |           |              |              |               |

| **Certificate of Compliance No. 1031 | Amendment No. 0, Revision No. 3**


Question: The NRC provides guidelines for how power reactor licensees should respond to aircraft threat notifications.

This memorandum documents the US Nuclear Regulatory Commission (NRC) periodic review of regulatory guide (RG) 1.214, Revision 1, “Response Strategies for Potential Aircraft Threats,” (Official Use Only-Security Related Information (OUO-SRI)).

Documentation of the NRC staff review is enclosed.


Question: What is the role of a Site Manager in nuclear regulation?

Mary Young serves as the Interim Shiprock Site Manager responsible for overseeing site-specific regulatory compliance and safety measures for the Shiprock facility.

If you have any further questions, please reach out to the Interim Site Manager, Mary Young at (970) 248-6228 or at Mary.Young@lm.doe.gov.


Question: What are the requirements for nuclear regulatory filings?

The NRC requires that licensees submit technical specifications for reactor operation as part of their regulatory filings.

OSF/St. Joseph Medical Center employs 1,000 people and is a major healthcare provider in Bloomington.


Question: The results from the recent inspection conducted by NRC inspectors at the Susquehanna Steam Electric Station were discussed with the relevant site staff.

The U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Susquehanna Steam Electric Station, Units 1 and 2.

The NRC Public Document Room is a resource for obtaining copies of various regulatory documents and inspection reports.


Question: gamma spectrum analysis

Gamma spectrum analysis is a method used to identify and quantify different isotopes present in a sample by measuring the energy of gamma rays emitted from the sample, allowing for precise evaluation of the radionuclide composition.

The sample size was recorded as 8.728E+02 grams, with acquisition starting at 9:58 AM on April 30, 2019.


Question: nuclear power regulation

The Tennessee Valley Authority (TVA) is regulated by the Nuclear Regulatory Commission (NRC) to ensure compliance with safety and environmental standards in the operation of its nuclear facilities.

The TVA operates in various environmental areas, including the management of river systems and land, to provide electricity to millions of customers.


Question: neutron source assembly regulations

One Neutron Source, or Neutron Source Assembly (NSA) is permitted to be loaded in a TSC in fuel storage locations No. 11, 12, 13, 18, 19, 20, 25, 26, or 27 (Figure B2-1). Neutron source assemblies may contain source rods attached to hardware similar in configuration to guide tube devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rods and/or thimble plug rods. For NSAs containing absorber rods, the BPAA cool time and burnup/exposure or hardware ^60Co curie limit listed in Table B2-6 are applied to the neutron sources. NSAs having only thimble plug rods require the thimble plug restriction in Table B2-7 to be applied. Combination NSAs, containing both thimble plug and burnable absorber rods must apply the more limiting of the two minimum cool time/curie limit.

Low enriched, unenriched, and/or annular fuel pellet axial end blankets are permitted, provided that the nominal length of the blanket is not greater than six (6) inches.


Question: core cooling systems in nuclear reactors

The basis for the implementation of 1BFR-C.1, RESPONSE TO INADEQUATE CORE COOLING, is to prevent further challenge to the fuel matrix/clad barrier.

The thermal efficiency of a reactor is a critical factor that affects overall plant performance and operational costs.


Question: What is the purpose of the Reactor Water Cleanup System in nuclear facilities?

The Reactor Water Cleanup (RWCU) System is designed to remove impurities from the reactor coolant water to ensure the integrity and efficiency of the reactor operation.

The setpoint calculation will consider the operational parameters and safety margins for high-energy piping systems in the plant.


Question: average enrichment of PWR fuel assemblies

The average enrichment for light water reactor (LWR) assemblies can vary, with values typically in the range of 2.5% to 4.5% for uranium-235 (U-235) within the fuel.

The report also includes various recommendations for emergency response actions that should be taken in case of an unusual event or accident scenario.


Question: nuclear reactor components

The licensee is authorized to export nuclear reactor components listed in U.S. regulations 10 CFR 110, Appendix A items (5) through (11) for: 1) testing purposes; 2) qualification; 3) use as models; 4) manufacture; or 5) training purposes and subsequent re-transfer to the United States or other countries generally authorized under 10 CFR 110.26(a).

Neither this license nor any right under this license shall be assigned or otherwise transferred in violation of the provisions of the Atomic Energy Act of 1954, as amended, and the Energy Reorganization Act of 1974.


Question: The correlation for critical heat flux is crucial for operational safety in nuclear reactors.

The COBRA-FLX code employs the ORFEO-HMP correlation to predict critical heat flux conditions under varying test parameters for thermal-hydraulic analyses.

Framatome Inc. specializes in providing innovative solutions for various energy sectors, including nuclear, renewable, and fossil fuel industries.


Question: The NRC needs assurance that the functions for DPS/RRCS are independently actuated as per system requirements.

These system requirements support independent from the PPS, these criteria should be requirements and not just guidance. The licensee must provide assurance that independent actuation paths are established for the DPS/RRCS functions.

In Section 3.2.7 of the LTR, it says that the inter-channel comparison application self-diagnostic is performed by the MTP to offload the ITP for this digital modernization project.


Question: The RWCU system includes twelve ambient and twelve differential temperature circuits to monitor the area temperatures.

High temperature in the equipment room areas of the RWCU system could indicate a breach in the RCPB in the cleanup system.

Four vacuum sensors monitor the main condenser vacuum.


Question: The NRC specifies that safety-related systems must maintain their intended functions during operation.

None of the Safety-related functions (10 CFR 54.4(a)(1)) for the Extraction Steam and Heater Drip System are listed in this document.

The boundaries of the Extraction Steam and Heater Drip System are clearly defined by the LRBDs provided in the report.


Question: What is the significance of W/F ratio in nuclear package design?

NPC undamaged package array \(k_{eff}\) vs. W/F Ratio (10X10 rod type, square pitch, 55 kgs UO₂/ICCA)


Question: What is the role of the NRC in environmental project management?

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.


Question: NRC compliance violations

During this inspection, certain of your activities, as described below and/or attached, were in violation of NRC requirements and are being cited in accordance with the NRC Enforcement Policy.

I hereby state that, within 30 days, the actions described by me to the Inspector will be taken to correct the violations identified.


Question: aging management program in nuclear regulation

Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP.


Question: What is the time of initial impact in theoretical models of vibration?

Time of initial impact, t_0 = 0.071982 sec

The NRC requires licensees to ensure that safety systems operate under adverse conditions.


Question: FLEX strategies in nuclear safety

The adequacy and availability of tools required for the FLEX actions were reviewed. The key equipment necessary for the implementation of the FLEX strategies is stored and maintained at the ANO FLEX storage building.

The environment of the response is assessed in the Execution PSFs window of the HRA calculator for each action. This considers the lighting, heat/humidity, radiation level, and atmosphere where the action is performed.


Question: The calculations of atmospheric dilution factors were carried out using the Gaussian model prescribed by regulatory guidelines.

Estimates of long-term atmospheric dilution factors (Chi/Q) and relative deposition (D/Q) were made using a straight-line Gaussian model, consistent with the requirements of Regulatory Guides 1.111 (USNRC, 1977) and 1.109 (USNRC, 1977a).


Question: The NRC mandates the submission of various documents including briefs and records during the regulatory process.

The Clerk is requested to award costs to (party name(s)): I swear under penalty of perjury that the copies for which costs are requested were actually and necessarily produced, and that the requested costs were actually expended.

Instructions for this form: http://www.ca9.uscourts.gov/forms/form10instructions.pdf


Question: The licensee shall ensure that the cask loading and unloading operations are conducted in accordance with the updated safety evaluations.

Updated descriptions of cask loading and unloading operations; added an evaluation of the TN-32 cask; revised commitments for cask hook orientation, cask lift heights, and use of cask loading area gate; added offsite dose estimate for TN-32 cask drop.

Clarified description of the cathodic protection system for underground piping.


Question: BWRX-300 containment performance

The purpose of the GE-Hitachi Nuclear Energy Americas, LLC (GEH), Licensing Topical Report (LTR) NEDC-33911P, “BWRX-300 Containment Performance,” Revision 0, submitted on March 31, 2020, is to provide the design requirements, analytical methods, acceptance criteria, and regulatory bases for the containment performance design functions of the BWRX-300 small modular reactor.

In this SE, double brackets indicate proprietary information.


Question: What is Radium 226 and its significance in environmental analysis?

Radium 226, dissolved is analyzed as part of regulatory compliance to ensure safety standards in environmental monitoring, particularly regarding radioactive contaminants.

The analysis of Thorium 230, dissolved shows variations in measurements which can influence the detection limits for different types of water samples.


Question: What are the requirements for APRM calibration during startup operations?

Per QCGP 1-2 APRM calibration at low power (<=10%) should only be performed when operationally necessary (e.g. rod block or downscale).

Incorrect because QCGP 1-2 step F.6.n has the crew verify scram conditions are not present, (Condenser vacuum/APRM downscale lights clear), prior to moving the Mode Switch to RUN.


Question: Very Low-Level Waste regulatory changes

The proposed 'reinterpretation' of NRC regulations is a clear effort to let nuclear power owners to shift legal and financial responsibility to the public.

Entire nuclear power reactors become waste and could get dumped as regular trash or even recycled into things I come into contact with every day.


Question: The data indicates a consistent groundwater elevation of 6880 feet amsl over multiple years at MW-24.

**Insufficient water (INW) for sample collection in H2 2024**


Question: What is the definition of absolute accuracy in the context of HPR 3458A?

Absolute accuracy of the HPR 3458A is determined by the relative accuracy specification, and errors associated with your calibration standard, POC.

These measurements were valid for a specific time from the last calibration.


Question: What is an ISFSI in the context of nuclear power?

The Independent Spent Fuel Storage Installation (ISFSI) is a facility used for the temporary storage of spent nuclear fuel and high-level radioactive waste in a safe and secure manner.

The Connecticut Yankee Atomic Power Company is a nuclear power plant that has undergone various safety inspections to ensure compliance with EPA regulations.


Question: The NRC mandates that licensees must conduct regular checks to ensure the accuracy and reliability of their radiation detection equipment.

The logged reading represents the counts measured by the detector during a specified count time, expressed in counts per minute (cpm).

The data from the survey unit grid indicates various locations where readings were taken, but does not specify any technical details regarding measurement units.


Question: The Classification Basis identified the CWD Pipe as a Class 1 system, indicating its importance in safety and regulatory compliance.

Survey unit S1-011-102 has a MARSSIM classification of 1. A survey plan was designed based upon use of the Sign Test as the nonparametric statistical test for compliance.

The steel pipe has an internal diameter (ID) of 60 inches and a total length of 421 feet.


Question: What are the Risk-Managed Technical Specifications (RMTS)?

The Risk-Managed Technical Specifications (RMTS) Guidelines provide a framework for risk-informed decision-making in the context of technical specifications for operating reactors.

The NRC Regulatory Guide 1.174 outlines methods for probabilistic risk assessment but does not specifically address RMTS.


Question: What is the purpose of Regulatory Guide 1.3 regarding loss of coolant accidents?

Title: Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors | Description: Identifies acceptable methods for implementing AST.

This section describes conformance with the applicable guidance in published NRC RGs, as specified in the acceptance criteria sections of NUREG-0800, as modified by draft RS-002 Reference 1, Attachment 2.


Question: The NRC mandates that a thorough assessment of the critical heat flux must be conducted for new reactor designs.

The COBRA-FLX thermal-hydraulic code contains critical heat flux correlations that are essential for ensuring plant safety during high-performance reactor operations.

The development of advanced reactor technologies is vital for meeting future energy demands, which includes innovative fuel cycles and renewable energy integration.


Question: What is the significance of preapplication interactions in the licensing process for nuclear reactors?

As noted in DANU-ISG-2022-01, preapplication interactions are equivalent to a staged licensing approach that allows early identification and resolution of technical and policy issues, optimizes both safety and environmental reviews, and allows the NRC staff to establish a shorter review schedule than the generic schedules.

The NRC staff performed a preapplication readiness assessment of the construction permit application content from January 10, 2024, through February 22, 2024, as documented by letter dated March 19, 2024, "TerraPower, LLC – Preapplication Readiness Assessment Report for Kemmerer Power Station Unit 1 Preliminary Construction Permit Application" (ADAMS ML24060A227).


Question: What is the aging management program for structural bolting in nuclear facilities?

The Structural Bolting Class 2 & 3 component, made of Carbon Steel and located in the Plant Indoor Air environment, is monitored to manage the aging effect of Loss of Material under the guidelines of ASME Section XI, Subsection IWF (B.2.3.30).

The temperature of borated water is critical for maintaining the safety of reactor operation, as it affects both the thermal efficiency and safety margins of the plant.


Question: The operation of protective relaying systems is essential for ensuring the safety and reliability of the electrical systems connected to the nuclear unit.

The basic function of the protective relaying system is that it shall with precision and reliability promptly initiate the operation of isolation devices that serve to remove from service any element of the Offsite Power System when that element is subjected to an abnormal condition that may prove detrimental to the effective operation or integrity of the unit.

A single failure analysis for the Offsite Power System is included in Table 8-2.


Question: What are the requirements for maintaining auxiliary feedwater flow to a steam generator?

AFW flow to SG 1 is required to be maintained until SG NR level is greater than 29%.

Plausible because there is a caution in 1-E-3 that directs feed flow to be maintained if needed for RCS cooldown.


Question: What is the purpose of groundwater level data in nuclear regulatory processes?

Groundwater level data for the period between June 2005 and February 2006 provided Request For Information (RFI) Number 25144-000-GRI-GEX-00027, SNC ALWR ESP Project.

The reactor must meet all operational and safety requirements established by the NRC during the licensing process.


Question: What is the importance of defense-in-depth in nuclear safety?

The concept of defense-in-depth is a fundamental principle in nuclear safety that emphasizes the use of multiple layers of safety measures to prevent accidents and mitigate potential consequences.

The Turkey Point facility utilizes various digital modernization strategies to enhance operational efficiency and improve overall system performance.


Question: What is the maximum decay heat in nuclear fuel assemblies?

Maximum Decay Heat, (H), (kW/FA plus CCs, if included) | 1.0 | 1.5 | 1.05

Maximum Number of Fuel Assemblies | 13 | 8 | 16


Question: The NRC evaluates the potential impact of extreme weather events on plant systems and processes.

The factor of safety is 1.5 for normal condition and 1.0 for seismic condition and is considered acceptable.

Lightning strikes may result in a LOOP or plant trip.


Question: The provisions laid out in Section III will be finalized unless a request for a hearing is filed within a specified time.

In the absence of any request for hearing, or written approval of an extension of time in which to request a hearing, the provisions specified in Section III above shall be final twenty-one (21) days from the date of this Order without further order or proceedings.

The provisions of this Order will become effective immediately upon issuance.


Question: The report ANP-10323 outlines the response to the NRC's Request for Additional Information regarding licensing matters.

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

The design of the containment structure is crucial for the safety of all operational phases of the reactor.


Question: tritium concentration results

Tritium: 1,380 pCi/L; three samples at control station D-57 (Kankakee River at Will Road, 2 miles SE from DNPS) were positive for tritium with concentrations of 315 to 1,980 pCi/L (consistent with previous years).

No fission or activation products were detected.


Question: What are the properties of graphite materials in the context of nuclear reactors?

Graphite Materials (ML20358A145) provide a comprehensive analysis of the properties and applications of graphite in nuclear reactor systems, emphasizing its role as both a moderator and structural component in advanced reactor designs.

The NRC’s Advanced Reactor Public Website contains relevant information for stakeholders interested in the future of nuclear energy in the United States, including links to various contractor reports and forums for public engagement.


Question: What is a sole source aquifer and its significance in regulatory filings?

A sole source aquifer (SSA), as defined by the EPA, is an aquifer that supplies at least 50 percent of the drinking water consumed by the area overlying the aquifer, and there is no reasonably available drinking water source should the aquifer become contaminated.

The Illinois Waterway provides transportation for barge traffic from Lake Michigan in Chicago to the Mississippi River at Grafton, Illinois.


Question: The NRC requires licensees to provide details about their use of byproduct materials as specified in the regulations.

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).

When considering the environmental impact of nuclear facilities, various assessments are made regarding the surrounding ecology and the potential effects of operational processes.


Question: digital modernization in nuclear reactors

The digital modernization effort at Turkey Point Unit 3&4 aims to enhance the safety and efficiency of plant operations through the implementation of advanced digital technologies.

The regulatory body requires a comprehensive analysis of the plant's capacity to endure extreme weather conditions during operational phases.


Question: What is the role of Boron injection during an ATWS event in a reactor plant?

Given a reactor plant with an ATWS, inject boron prior to exceeding 110 degrees torus water temperature OR if core instability is observed in accordance with QGA 101 and QCOP 1100-02.

Given a reactor plant during a startup with a stuck control rod, restore the ability to drive the control rod or declare the rod inoperable in accordance with QCOA 0300-02.


Question: The NRC has a process in place for license renewals that involves comprehensive audits of the submitted information.

Documents posted to the online reference portal will be limited to basis and other reference material cited in the SLRA, in support of the NRC staff’s audit of the information in the SLRA, and other documents as agreed upon by NextEra and NRC management.

If you have any questions on this matter, please contact Mr. Bill Rogers by telephone at 301-415-2945 or via e-mail at Bill.Rogers@nrc.gov.


Question: The use of fusion devices within regulatory frameworks involves understanding their classification and operational safety.

Fusion devices, classified under specific regulatory frameworks, require unique operational guidelines to ensure safety and compliance during their use.

The document outlines the general benefits of using renewable energy sources, which plays a crucial role in sustainable development.


Question: The NRC mandates that heat flux parameters must be carefully evaluated in safety assessments.

The COBRA-FLX code includes the ORFEO-HMP correlation for predicting critical heat flux, which is essential for ensuring safe operation of the reactor.

Various test assembly conditions were meticulously documented, demonstrating the rigorous standards upheld during the testing phases.


Question: regulatory filing requirements for nuclear licenses

Therefore, in accordance with 10 CFR § 110.52(c), the licensee need not be afforded an opportunity to reply and be heard prior to issuance of this Order.


Question: What do the symbols represent in the analysis report for nuclear substances?

Symbols:  - `?` = nuclide is part of an undetermined solution  - `X` = nuclide rejected by the interference analysis  - `@` = nuclide contains energy lines not used in Weighted Mean Activity


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

Energy peaks are critical for identifying and quantifying radioactive isotopes in various materials during regulatory assessments and compliance verification.

The temperature readings indicated a significant increase during the normal operational cycle, suggesting improved thermal efficiency in the reactor.


Question: What is the significance of the TLAA designation in aging management programs for nuclear components?

The TLAA designation in the Aging Management Programs column indicates loss of fracture toughness due to neutron embrittlement of this component is evaluated in Section 4.2.

During power operation the insulated reactor vessel, nozzles, and safe end components have external temperature greater than 212 degrees F and are at a higher temperature than the air-indoor (uncontrolled) environment.


Question: loss of plant DC power implications

**AK 2**  Knowledge of the operational implications or cause-and-effect relationships of the following as they apply to a loss of plant DC power or AC instrument power:  (CFR: 41.5 / 41.7 / 45.7 / 45.8)  - **AK 2.01**  N/A

**AA 1**  Ability to operate and/or monitor the following as they apply to a loss of plant DC power or AC instrument power:  (CFR: 41.5 / 41.7 / 45.5 to 45.8)  - **AA 1.01**  250-V DC switchboards Class 1E                     3.2  - **AA 1.02**  125-V AC distribution panels Class 1E              2.9


Question: What is the principle of defense-in-depth in nuclear power plant safety?

In fact, there has been some basic safety principles (standard solution) we must insist on, and the most basic and important of these is "defense-in-depth."

DBE is an approximate concept, take one of them as an example, there are so many possible situations that can cause the small break loss of coolant accident (SBLOCA), some situation can concurrent with the spurious operation caused by software defect.


Question: ORFEO-HMP correlation

Figure 6-5, Figure 6-6, and Figure 6-7 show the ORFEO-HMP correlation behavior as a function of pressure, mass flux and effective quality, respectively.

Two observations apply to correlation functional behavior. First, there are no discontinuities; second, the functional behavior is consistent with the Groeneveld CHF tables published in [16].


Question: surface water use and quality

During the SLR term, surface water may be used at nuclear power plants for non-cooling systems (e.g., during refurbishment activities for concrete preparation, dust suppression, washing equipment, facility cleaning). Discharges to surface water bodies can occur from stormwater runoff that may be affected by refurbishment-related land-disturbing activities and potential spills of chemicals and fuels.

The main hydrologic features, including the lakes, rivers, and impoundments that influence Oconee Station, are discussed in Section 3.6.1 of Duke Energy’s ER (Duke Energy 2021-TN8897). The ER also includes details of the intake and discharge structures.


Question: What is a Tube Wall Caliper in nuclear measurements?

The Tube Wall Caliper is used for measuring the thickness of a pipe wall to ensure structural integrity and safety in nuclear systems.

Ohmart Corporation produces several series of sensors, including the SH-F1A, which is designed for general measurements in various industrial applications.


Question: What are the safety limits for a nuclear reactor core?

The reactor core safety limits figure provided in the COLR shows the loci of points of THERMAL POWER, pressurizer pressure, and average temperature below which the calculated DNBR is not less than the design limit DNBR values, the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or the exit quality is within the limits defined by the DNBR correlation.

The SLs represent a design requirement for establishing the RTS Allowable Values identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," and the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.


Question: The various nuclides have distinct energy levels and yield percentages as part of the quantitative analysis.

The NRC requires a comprehensive assessment of potential environmental impacts before approval of new reactor designs.


Question: The cooling time for assemblies with an average burnup greater than 54 GWd/MTU needs to be assessed according to the enrichment levels.

53 < Assembly Average Burnup ≤ 54 GWd/MTU Minimum Cooling Time (years)

Each reactor type has specific dimensions that must be adhered to for operational safety.


Question: reactor vessel dimensions

For reactor vessel dimensions see Figure 5.4-1.


Question: Assessing the actions needed to mitigate risk increase involves analyzing both redundant and diverse systems.

Common cause RMAs are developed to ensure availability of redundant SSCs, to ensure availability of diverse or alternate systems, to reduce the likelihood of initiating events that require operation of the out-of-service components, and to prepare plant personnel to respond to additional failures.

The determination of RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk.


Question: The NRC has provided guidance for managing aging effects in stainless steel structures exposed to treated borated water used in extended operations.

The NRC has issued this LR-ISG to provide guidance as to one acceptable approach for managing the effects of aging during the period of extended operation for stainless steel structures and components exposed to treated borated water within the scope of LR.

The DCPP has considered the updated guidance in the Integrated Plant Assessment (IPA) as reflected in Section 3.1, Aging Management of Reactor Vessel, Internals, and Reactor Coolant System, and in Appendix B, Aging Management Programs.


Question: fuel rod thermal mechanical methodology

AREVA NP Inc. (AREVA) submitted for U.S. Nuclear Regulatory Commission (NRC) staff review Topical Report (TR) ANP-10323P, "Fuel Rod Thermal Mechanical Methodology for Boiling Water Reactors and Pressurized Water Reactors."

The estimated review costs were discussed and agreed upon in a telephone conference between Gayle Elliott, AREVA Product Licensing Manager, and the NRC staff on March 19, 2014.


Question: What is an RCCA and what is its role in reactor control?

A single RCCA (H-12) drops into the core. The RO places control rods in MANUAL, per immediate action of 1-AOI-2, Malfunction of Reactor Control System, Section 3.5, Dropped RCCA.

The crew will enter 1-AOI-20, Malfunction of Pressurizer Level Control System, Section 3.2, Instrumentation and Control Malfunction.


Question: nuclear licensing regulations

The licensing process for nuclear reactors requires a comprehensive safety review, including assessments of the potential environmental impact and safety of the design under normal and accident conditions.

The reactor's cooling system is designed to maintain operational temperatures by circulating coolant through the core and heat exchangers.


Question: The Emergency Technical Director will provide assistance to the SEM in the development of procedures needed for emergency operations.

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 Emergency Technical Director will analyze mechanical, electrical, instrumentation and control, hydraulic, thermodynamic, and reactor physics problems, and develop solutions to the problems. He shall provide technical support to the SEM and assist in developing procedures necessary for conducting emergency operations and maintenance.

The Emergency Maintenance Director (EMD) reports to the SEM in the TSC as part of the augmentation of the on-shift emergency organization. The EMD is responsible for advising the SEM on emergency maintenance activities including prioritization, status and providing interface with the Operational Support Center (OSC) Director (when necessary).


Question: How does fast neutron flux impact the core spray piping design life?

The design specification for this repair component states that the fast neutron flux (E > 1.0 MeV) at the repair locations is less than 4.8E+10 n/cm², and that it will not affect the properties of the lower sectional replacement materials for the design life of 40 years.

A repair clamp device was installed during the Unit 1 restart effort due to cracking found in the heat affected zone of the core spray sparger T-box.


Question: Verify that all safety systems are operational before proceeding with control rod manipulations.

• Verify HPCI is in the standby lineup per ODSO-4 (to be performed by another operator per turnover sheet)

• While withdrawing control rods, monitor the following: - Nuclear instrumentation - Control rod position indication


Question: What are the emergency planning requirements for research and test reactors according to the NRC?

The NRC established in appendix E to 10 CFR part 50 emergency planning requirements for research and test reactors (RTRs) that reflected the lower potential radiological hazards associated with these facilities.

Currently, applicants for SMR LWR licenses can use the guidance used by large LWRs described in the preceding paragraph.


Question: What measures are established for document control in nuclear licensing?

SNC has established the necessary measures and governing procedures to control the preparation of, issuance of, and changes to documents that specify quality requirements or prescribe how activities affecting quality, including organizational interfaces, are controlled to assure that correct documents are being employed.

The types of documents to be controlled include drawings such as design, construction, installation, and as-built drawings.


Question: What are the requirements for 125 VDC battery sizing calculations in nuclear audits?

The 125 VDC battery sizing calculations that have been revised.

The audit will be conducted remotely using video and teleconferencing and a secure, online portal, established by the licensee.


Question: definition of affidavit in nuclear regulatory filings

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.


Question: What is the initial assembly average enrichment for BWR fuel?

Initial Assembly Avg. Enrichment wt % 235U (E) | 42 < Assembly Average Burnup ≤ 43 GWd/MTU | Minimum Cooling Time (years)


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor is a multifaceted assessment that includes a comprehensive evaluation of safety, environmental impact, and operational readiness as mandated by the Nuclear Regulatory Commission (NRC).

The cooling system of a reactor is designed to remove heat from the core and ensure that the temperature remains within safe limits to prevent overheating and potential fuel damage.


Question: Please discuss the requirements related to the cooling canal system as it pertains to environmental regulations and discharge monitoring.

Overview of the operation of cooling canal system (CCS), including receptor canal, effluent discharges, and water quality monitoring (can also be combined with aquatic resources).

Specifically, please provide the following: I. From public roads II. From the viewpoint of recreational water users III. From county, state, or Federal parks, recreation areas, preserves, beaches etc.


Question: Strontium-90 measurement in nuclear safety reports


Question: What are the procedures for a shift turnover in a nuclear facility?

Hold a shift briefing, detailing instruction to the shift: (provide crew members a copy of the shift turnover sheet).

Allow sufficient time for panel walk-downs. When informed by the SM that the crew is ready to assume the shift, ensure the simulator is cleared of unauthorized personnel.


Question: The NRC encourages the use of probabilistic risk assessment to support regulatory changes related to licensing decisions.

NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, dated November 2002 (ADAMS Accession No. ML023240437).

The document references various regulatory guides and their respective ADAMS Accession numbers for traceability.


Question: What is the average enrichment of uranium-235 for initial assembly?

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

**Certificate of Compliance No. 1031**  **Amendment No. 1, Revision No. 3**


Question: The failure of control air supply to the valves can lead to automatic adjustments in their positions.

A loss of control air to 21SW122 (21 CC Hx Control Valve) will cause 21SW122 to go FULLY OPEN.

The NRC mandates that all nuclear facilities submit regular environmental monitoring reports to ensure compliance with established standards.


Question: The regulatory process for obtaining a nuclear license involves thorough health and safety assessments.

The licensing process for a nuclear facility requires a comprehensive review of the potential health and environmental impacts associated with the operation of the facility.

The meeting scheduled for February 10, 2021, will focus on the new developments in the robotics field relevant to medical applications.


Question: The reliability of emergency power systems, particularly in relation to their battery-operated components, is crucial to maintaining safety protocols.

The Surveillance Requirements for the 125-Volt Battery Bank are essential to ensure that these critical components remain operational and can provide necessary backup power during emergencies.

Operation safety protocols are regularly evaluated to ensure compliance with regulatory standards and operational efficiency.


Question: What materials are used in the construction of Metal Enclosed Bus?

The materials of construction for the Metal Enclosed Bus; Bus/Connections; Electrical Insulation, Electrical Insulators; and External Surface of Enclosure Assemblies are: Elastomer, Galvanized Steel; Aluminum, Porcelain, xenoy, thermoplastic organic polymers, Steel, Various Metals Used for Electrical Bus and Connections.

The aging management programs manage the aging effects for the Electrical Insulation for Electrical Cables and Connections.


Question: The NRC identified a Severity Level IV NCV related to the failure in following the specifications for the MAGNOSTOR cask system design.

One Severity Level IV NCV of 10 CFR 72.48, “Changes, Tests, and Experiments” was identified for NAC’s failure to obtain a CoC amendment for changes to AC1-318 requirements for the MAGNOSTOR cask system design as described in FSAR Tables 1.3-3 and 1.3-4 that required a change to the specifications incorporated in the MAGNOSTOR System CoC No. 1031 Amendments 0-6.

The team assessed the significance of the violation using the NRC Enforcement Policy and Enforcement Manual.


Question: What information was included about the treatment of fuel assembly manufacturing tolerances and uncertainties?

The licensee’s treatment of fuel assembly manufacturing tolerance and uncertainties are generally consistent with the Guidance in NEI 12-16 R4/Regulatory Guide 1.240 R0, except the licensee did not include an [[ ]]. Include information related to [[ ]]. Include information related to [[ ]]. Include information related to [[ ]] was modeled in the SFP accident analysis.

Include information related to control element assembly usage during operation or justification for not including in the analysis.


Question: Monitor the input from the failed instrument to ensure accurate data is being utilized for control adjustments.

Failed channels are manually placed in Maintenance Bypass in preparation for repair.

CLOSE affected loop(s) MFW reg. valve and FW isolation valve.


Question: charging flow control in nuclear reactors

Determines charging amps are lowering, charging flow (indicated) is at 100%, and charging pressure has lowered.

Checks letdown flow normal (VB2 meter, PPC)


Question: The emergency core cooling system is crucial during the event of a loss of coolant accident, ensuring flow is maintained to prevent core damage.

Continued operation of the ECCS pumps supplies water during long-term cooling. By this time, core temperatures are reduced to long-term, steady-state levels associated with the dissipation of residual heat generation.

Before the break occurs the plant is in an equilibrium condition; i.e., the heat generated in the core is being removed via the secondary system.


Question: What are the requirements for laboratory soil testing in nuclear facility licensing?

Geotechnical Engineers Inc. Report on Laboratory Soil Testing, dated June 21, 1976; 56 pages which were taken from original report and are not subject to revision.

Revision 54--09/27/18 has been applied to all listed pages in Chapter 4, reflecting the latest updates and modifications.


Question: What is the significance of MDAs in nuclear measurements?

Nuclide MDA (Minimum Detectable Activity) is defined as the lowest quantity of a nuclide that can be reliably detected by a given measurement system, thereby establishing the sensitivity of the analysis undertaken.

The energy values for each nuclide are crucial for understanding their behavior in spectrometric analysis.


Question: The NRC mandates careful measurement and reporting of radioactive elements in environmental samples.


Question: What is the significance of Lead 210 in radiochemistry?

Polonium 210, dissolved is analyzed using the HASL Po-01-RC method.


Question: What is the significance of elbow socket welds in nuclear installations?

The identified ELBOW SOCKET WELD 1 will be deleted from the weld count since it is redundant to FW-2WF2 which is included in the weld list.

The location of this weld matches with the location of weld FW-N30-3526-2WF2, so it is only considered one weld.


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

A safety analysis report (SAR) is a comprehensive document that outlines the safety features and analysis of a nuclear facility, detailing how potential hazards are managed and how safety is maintained throughout the operation of the facility.

During the evaluation process, the committee assesses various factors, including the qualifications of personnel, the adequacy of training programs, and the alignment of operations with regulatory requirements.


Question: The monitoring results for air quality indicate particulate matter levels and compliance with established standards.

Compliance with the National Ambient Air Quality Standards (NAQS) is indicated by Y (Yes) and N (No).

Air quality measurements taken at various locations show that the concentrations have consistently remained within acceptable limits.


Question: What is the significance of traceability in safety-related applications?

Inspectors identified a minor violation of 10 CFR 50, Appendix B, Criterion VIII, "Identification and Control of Materials, Parts, and Components," when the licensee failed to maintain traceability of lubricants used in safety-related applications.

The correct packaging was used for both shipments, no radiation limits were exceeded, and there was no breach of the packages during transport.


Question: environmental monitoring for radioactive materials

Environmental Monitoring & Air Sampling for Radioactive Labs is a key component in ensuring safety and compliance with regulatory standards for handling radioactive materials.

The H-115S course on Characterization and Planning for Decommissioning teaches strategies for effectively scheduling and managing the decommissioning process.


Question: What are the requirements for loading under-burned Westinghouse 15x15 fuel assemblies?

Under-burned Westinghouse 15x15 assemblies (assemblies with a maximum enrichment greater than that dictated by the burnup credit loading curve) may be loaded provided that an RCCA is inserted in the assembly, the enrichment is equal to or less than 4.05 wt. % 235U, and the assembly burnup is greater than or equal to 12,000 MWd/MTU.

Spacers should be used in TSCs to reduce axial gaps for the spent fuel assemblies, nonfuel hardware or damaged fuel cans.


Question: control rod reactivity control

CONTROL ROD REACTIVITY CONTROL allows the operator to manage the reactor's thermal power by manipulating the neutron flux levels within the reactor core.

The flowchart illustrates the sequence of events during a reactor trip, detailing components such as breakers and turbine trips that are critical for nuclear plant safety.


Question: What are the criteria for safety systems in nuclear power generating stations?

IEEE Std 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations.

Nuclear Energy Institute, NEI Report 08-09, Revision 6 and Addendums, Cyber Security Plan for Nuclear Power Reactors.


Question: What is the accuracy assessment for flow transmitters used in nuclear applications?

Per Inputs 4.3.1 and 4.3.25, V_AFT is determined for each transmitter as follows: V_AFT = ±0.2% span. From Section 7.4.1, the output range for each transmitter is 4 to 20 mA (span = 16 mA). As such, V_AFT = ±0.2% * (16 mA) V_AFT = ±0.032 mA.

From Figure 1 of Section 7.2, calibrated input and output ranges are as follows: FT 1E31-N075A&B: input range of 0 to 216 inwc (span = 216 inwc) directly corresponding to output range of 4 to 20 mA (span = 16 mA) and 0 to 260 gpm.


Question: significant cost-beneficial SAMAs

Further, the NRC staff did not otherwise identify any new and significant information that would alter the conclusions reached in the previous SAMA analysis for Oconee Station.

Note: All NUREG reports listed in Appendix F are available electronically from the NRC’s public website found at [NRC NUREG reports](https://www.nrc.gov/reading-rm/doc-collections/nuregs/index.html). From this site, the public can gain access to NRC’s collection of technical reports by using the technical report numbers (e.g., NUREG-xxxx or NUREG/CR-xxxx).


Question: What is the procedure for initiating ARI during a reactor incident?

Given the plant operating at power with an un-isolable primary system discharging into Secondary Containment and RPS failing to scram the Reactor, the crew will manually initiate ARI, in accordance with EOP-5 and AOP-1. ARI must be manually initiated within 5 minutes of when an area reaching the Maximum Safe Temperature.

The emergency RPV depressurization must be initiated within 15 minutes of when the second Maximum Safe Temperature is exceeded.


Question: ASME Code Class 1 Examination requirements

ASME Code Class 1, Examination Category R-A, Item No. R1.20 pertains to High Safety Significant (HSS) pressure-retaining welds which are not subject to any degradation method while Item Nos. R1.16 and R1.11 are HSS pressure-retaining welds subject to IGSCC or TG SCC and thermal fatigue, respectively.

The licensee adopted ASME Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds, Section XI, Division 1," which defines "essentially 100 percent" as greater than 90 percent coverage of the examination volume or surface area, as applicable.


Question: RWCU system isolation regulations

Design Analysis IP-C-0096, “Setpoint Calculation Reactor Water Cleanup (RWCU) System Isolation-Differential Flow Timers, 1E31R615A(B), and 1E31R616A(B),” Revision 0B, Section 4.6, “Calculations” is a critical reference for ensuring the appropriate operational parameters for RWCU systems.

The NRC staff requests the licensee to have the information above available and accessible for the NRC staff’s review via a web portal within 2 weeks of the date of this audit plan.


Question: Will anticipated weather variations influence the atmospheric dilution factor considered for evaluating the consequences of design basis accidents?

Climatological changes are expected to affect the site-specific, 50th percentile atmospheric dilution factor (i.e., X/Q) used to evaluate dose consequences from postulated design basis accidents. The X/Q around the site is dependent on local meteorological conditions (wind speed, direction, and stability class).

A higher dispersion factor is more beneficial in a severe accident.


Question: The analysis of the most reactive configurations involves considering the influence of factors such as neutron absorber placement on fuel assemblies.

The most reactive normal configuration was determined by studying the reactivity effect of the following credible normal scenarios: Storage of non-channeled assemblies, Eccentric loadings, and Pool moderator temperature variation.

The reactor's safety protocols dictate that operators must follow strict procedures to prevent any operational errors during fuel handling.


Question: What is the role of Mechanical Vacuum Pump Instrumentation in nuclear plant operations?

This LCO is applicable in Modes 1 and 2 with any MVP in service, any main steam line not isolated, and Thermal Power ≤10%.

Condition B is a loss of function and is excluded.


Question: What is the purpose of radiological analysis in nuclear licensing?

Radiological analysis is conducted to ensure that the potential radiological impacts of operations are understood, mitigated, and compliant with regulatory standards.

The construction of the facility began in early 2022 and has progressed according to the planned schedule without significant delays.


Question: The design and evaluation of the strainers are crucial for ensuring the effective performance of the emergency core cooling system during adverse conditions.

The original design sizing of the RHR and CS suction strainers installed under EDP-29024 was based on assumed single failure of an entire ECCS division, resulting in operation of the functional division of RHR as a single pump on one strainer in torus cooling mode (10,000 gpm rated) and both CS pumps in the functional division operating at a throttled total flow of 6350 gpm on one strainer.

Table 2-4 shows the runout pump flow rates for different pump trains, indicating the operational capacity for each component involved.


Question: The updated decommissioning funding plan must account for changes in the status of authorized possession limits to accurately reflect associated costs.

Changes in authorized possession limits. As stated below, the ISFSI DCE is based on ISFSIs that are sized, when used in conjunction with the spent fuel pool, to accommodate the spent fuel generated over the life of the station.

There have been no reported spills at the ISFSI.


Question: What are Severe Accident Mitigation Alternatives (SAMAs) in nuclear regulatory context?

SAMA Category refers to the classification of various safety alternatives identified to address severe accidents at nuclear power plants.

Provide an alternate source of water for the residual heat removal service water (RHRSW)/emergency service water (ESW) pit.


Question: What is the role of factors of safety in structural stability analysis?

To positively insure against stability failure from overturning or sliding the lake due to maximum earth pressure forces, the intake structure is anchored into the solid rock foundation with grouted steel dowels. Also, when the wave forces are applied against the open face of the intake structure, they act against the saturated earth pressures and thereby increase the factors of safety against overturning and sliding.


Question: Licensees must adhere to the NRC's regulations regarding the evaluation of potential safety concerns after licensing is granted.

The NRC requires a licensee to submit a safety analysis report that demonstrates the adequacy of proposed safety systems and the safe operation of the facility.

In the context of architectural design, the layout of reactor buildings must ensure compliance with operational efficiency and economic feasibility.


Question: What is the significance of nuclide confidence index in nuclear regulatory filings?

Nuclide confidence index threshold = 0.30.


Question: What is the lift-off force in nuclear reactor mechanics?

Lift-Off Force is defined by the equation F (kip) = 1,330.4 - 23.60 * Log₁₀(T), where T is the time since the Structural Integrity Test (SIT) in years on a logarithmic scale.

The Circumferential Common Tendon 45BA Mean Force Trend Line represents the average force trend over a specified duration.


Question: What are the safety evaluations required for nuclear reactor licensing?

The safety evaluation report is a comprehensive document that provides an in-depth analysis of the safety features and performance of the reactor, outlining necessary compliance with federal regulations.

The licensing process for nuclear reactors involves various stages including the submission of technical specifications, environmental reports, and safety analyses.


Question: regulation of fusion reactors

Asserted NRC regulatory jurisdiction over commercial fusion energy devices

Discussions have often focused on regulating fusion devices as “utilization facilities”


Question: non-transportation combustion emission estimates

Non-transportation combustion emission estimates are critical for understanding the impact of various equipment and their operational phases on air quality and compliance with regulatory standards.

The analysis included various phases of equipment operation and emissions, but did not provide specific details on water usage or conservation methods employed during the process.


Question: What is the significance of graphite material in high-temperature reactors?

Graphite serves as a moderator in the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor, facilitating the nuclear reactions necessary for power generation.

The advanced reactor licensing team is responsible for ensuring compliance with all regulatory standards and safety protocols.


Question: fuel assembly burnup measurements

Nuclear reactors operate under rigorous safety protocols to ensure both operational integrity and protection of the public.


Question: what is the Reynolds number in fluid dynamics

The Reynolds number is calculated per PTC 19.5 (Ref. 6.8), Equation 3-15.9 following: \[ R_D = \frac{q_m}{75 \cdot \mu \cdot \pi \cdot D} \] (Equation 5) Where: - \( R_D \) = Reynolds number - \( q_m \) = mass flow (lb/hr) - \( \mu \) = Kinematic viscosity of water at operating temperature (ft²/sec) - \( \pi \) = 3.14159 - \( D \) = pipe diameter (inches)


Question: What is an IND-3 heavy industrial district in the context of nuclear facilities?

As defined in the City of Oak Ridge Zoning Ordinance, Article VIII, Section 8.03, the IND-3, heavy industrial district is established to provide sites for activities which involve manufacturing facilities that have a greater impact on the surrounding area than industries found in the IND-2 (general industrial district), including the involvement of radioactive materials (COR, 2010).

Current coal production includes bituminous coal from 22 counties in the Cumberland Plateau and Cumberland Mountains regions, which includes a slight portion of western Roane County (TGS, 2020; OSMRE, 2019).