ML24306A053

DG-1354 (RG 1.238 Rev. 0) and DG-1420 (RG 1.32 Rev. 4) Comment Resolution Table

ML24306A053
Person / Time
Issue date:	01/14/2024
From:	Sheila Ray NRC/NRR/DEX/EEEB
To:
References
DG-1354, DG-1420 RG-1.238, Rev. 0, RG-1.032, Rev. 4
Download: ML24306A053 (12)

Text

1 Responses to Public Comments on Draft Regulatory Guide (DG)-1420 Criteria for Power Systems for Nuclear Power Plants, and DG-1354, Criteria for the Protection of Class 1E Power Systems and Equipment for Nuclear Power Plants On August 28, 2024, the U.S Regulatory Commission (NRC) published a notice in the Federal Register (89 FR 68787) that two Draft Regulatory Guides were available for public comment. DG-1420, Criteria for Power Systems for Nuclear Power Plants, (proposed Revision 4 to Regulatory Guide (RG) 1.32), and DG-1354, Criteria for Protection of Class 1E Power Systems and Equipment for Nuclear Power Plants, (newly proposed RG-1.238). The public comment period ended September 27th, 2024. The NRC received comments from the organizations and individuals listed below. The NRC has combined the comments and NRC staff responses in this document.

Comment Submission 1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML24262A008 Name: Gregg Reimers Email: greimers@gmail.com Comment 1-1 Gregg Reimers (DG-1354)

The design requirement stated in the third paragraph of IEEE 741 Section 5.1.1 begins with If the power distribution design allows for automatic bus transfers While automatically initiated transfer capability may not be necessary for all system designs, the ability to transfer between offsite power circuits is required. Standard Review Plan, NUREG 0800, Section 8.1, Table 8-2, states Each of the two required offsite power circuits shall be designed to be available in sufficient time to effect safe shutdown in the event of a loss of all onsite power and the loss of the other offsite circuit. Therefore, a transfer scheme to switch back and forth is not an optional feature. Regardless of how a transfer is initiated (i.e. automatic vs manual), protection concerns (e.g. IEEE 741, Annex C) would apply.

Why doesnt the NRC identify an exception and / or offer an interpretation regarding IEEE 741 Section 5.1.1 in the DG 1354 endorsement? Specifically, the protection of the Class 1E electrical distribution systems during the transfer between offsite power circuits is a valid concern for all designs.

NRC Comment 1-1 Response The staff disagrees with the comment. Some designs use alternate approaches where bus transfer is not required (see reference below).

This Standard Review Plan (SRP), NUREG-0800, establishes criteria that the staff responsible for the review of applications to construct and operate nuclear power plants intends to use in evaluating whether an applicant/licensee meets the NRC's regulations. The SRP is not a substitute for the NRC's regulations, and compliance with it is not required.

The regulation in 10 CFR Part 50, Appendix A, GDC 17, in part, states that:

Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other

2 offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

NUREG-0800, Section 8.1, Table 8-2, item d provides, in part, the NRC staffs interpretation states:

Each of the two required offsite power circuits shall be designed to be available in sufficient time to effect safe shutdown in the event of a loss of all onsite power and the loss of the other offsite circuit. (The staff has designated the second circuit as the delayed access circuit. The offsite power system (i.e., the two circuits considered together) must meet the single failure criterion, but only with respect to the delayed access circuit function.)

Neither the regulation nor SRP require a transfer between the two offsite circuits. A transfer scheme could be used to meet the regulations but is not the only method of meeting the regulations.

IEEE Std. 741-2022, Section 5.1.1, Switchgear and bus protection states that If the power distribution design allows for automatic bus transfers, consideration shall be given to the impact of the bus transfer on the coordination of protection devices (see Annex C). Further, Annex C, Auxiliary system automatic bus transfer - protection concerns, discusses bus transfer and protection concerns. The staff finds that the standard sufficiently addresses bus transfer.

While the NRC staff does not endorse Annex C, it may still be used as useful information and good engineering practice.

No changes were made to DG-1354 as a result of this comment.

Reference (for information purposes only)1:

N. K. Trehan, "Fast bus transfer scheme and its performance in nuclear power generating stations," 2002 IEEE Nuclear Science Symposium Conference Record, Norfolk, VA, USA, 2002, pp. 1909-1913 vol.3, doi: 10.1109/NSSMIC.2002.1239698.

Comment Submission 2 ADAMS Accession No ML24262A009 Name: Gregg Reimers Email: greim416@gmail.com Comment 2-1 Gregg Reimers (DG-1354)

The analysis described in IEEE 741 Annex A is incomplete regarding electrical distribution designs based on one immediately available offsite power circuit. While system level coordination of the LVR voltage and time setpoints is important to prevent loss of safety equipment functionality, they must also be coordinated to ensure sufficient time is available to complete design basis bus transfers from one offsite power circuit to another.

Why doesnt the NRC identify an exception and / or offer an interpretation in the DG 1354 1 This document can be found at the IEEE Xplore website: https://ieeexplore.ieee.org/document/1239698

3 endorsement regarding the limited applicability of the IEEE 741, Annex A, LVR analysis?

Specifically, the LVR setpoint determination has competing constraints that IEEE does not address.

NRC Comment 2-1 Response The staff disagrees with the comment. IEEE Std. 741-2022, Annex A.1, states that However, it is recognized, because of the diversity of nuclear power plant auxiliary system designs, there are other protection schemes that provide the desired level of protection, Also, this annex does not address the capability of various relay types, but rather discusses the philosophy behind the desired actuation times and voltage levels. IEEE Std. 741-2022, Section 5.1.1, Switchgear and bus protection states that If the power distribution design allows for automatic bus transfers, consideration shall be given to the impact of the bus transfer on the coordination of protection devices (see Annex C). Further, Annex C, Auxiliary system automatic bus transfer - protection concerns, discusses bus transfer and protection concerns. The staff finds that the standard sufficiently addresses coordination of the LVRs to ensure sufficient time is available to complete bus transfers from one offsite power circuit to another, as indicated in Section 5.1.1 and Annex C of IEEE Std 741-2022.

While the NRC staff does not endorse Annex C, it may still be used as useful information and good engineering practice.

No changes were made to DG-1354 as a result of this comment.

Comment Submission 3 ADAMS Accession No ML24270A013 Name: Gregg Reimers Email: greim416@gmail.com Comment 3-1 Gregg Reimers (DG-1420)

Regarding the AC power distribution system of a nuclear unit, the scope of IEEE 308-2020 is limited to the Class 1E portion. This includes the establishment what is required from the preferred power supply (PPS) at the interface. Therefore, the requirements defined in Section 5.2.3 are limited to the Class 1E system interface with the PPS, not the PPS itself. How those interface requirements are satisfied by the PPS is the scope of IEEE 765, which is not part of the DG 1420 endorsement.

DG 1420 states that the purpose of regulatory guides is to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses.

The endorsement of IEEE 308-2020, Section 5.2.3, without any exception or clarification fails to satisfy the regulatory guide purpose. Specifically, since IEEE 308-2020, Section 5.2.3, neither establishes any functional nor performance requirements pertaining to capacity, capability, or availability at the PPS interface; what methods, techniques, and information is DG 1420 endorsing? This absence of clarification complicates the proposed Section 6.3 addition of preoperational testing requirements for the various possible combinations of power sources It appears that either substantial DG 1420 clarification is warranted, or IEEE 308 should not be endorsed at this time.

4 NRC Comment 3-1 Response The staff disagrees with the comment. General Design Criterion (GDC) 17 of 10 CFR Part 50, Appendix A addresses offsite power systems. In addition, as indicated during the August 5th, 2020 NRC Public Meeting, the NRC is considering endorsing IEEE Std 765(see ML20212L534 and ML20189A599),

which relates to this comment. The contents of DG-1420, which endorses IEEE Std. 308-2020, should not be overly prescriptive of offsite power capacity, capabilities and abilities which are addressed or will be endorsed elsewhere.

No changes were made to DG-1420 as a result of this comment.

Comment Submission 4 ADAMS Accession No ML24271A033 Name: Gurcharan Matharu Email: Lawnak90@gmail.com Comment 4-1 Gurcharan Matharu (DG-1420)

The NRC Staff need to review the applicability of new regulations that are currently applicable to the existing fleet of operating plants.

The NRC Staff need to define any exemptions or unique requirements that have been granted or imposed on the design of new passive design nuclear power plants and small modulator reactors.

NRC Comment 4-1 Response The staff disagrees with the comment. DG-1420 includes current regulations applicable to operating and new reactors, including small modular reactors and passive designs. New and small modular reactor designs are being considered by the NRC on a case-by-case basis in regards to the particular design features and respective safety analyses.

No changes were made to DG-1420 as a result of this comment.

Comment 4-2 Gurcharan Matharu (DG-1420)

For the Draft Guide, the NRC Staff need to review the applicability of following 10CFR 50 Regulations:

1. Appendix R - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 which has specific requirements for:

a. Alternative or dedicated shutdown capability.

b. One train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) to be free of fire damage

c. Safety, non-safety and associated cables physical and electrical separation to preclude damage to redundant train of equipment and spurious actuations.

NOTE: NRC RG 1.189 provides guidance for Appendix R related conformance.

2. 50.155 Mitigation of beyond design basis events which has specific requirements for DC systems and temporary or portable AC systems.

5

3. 50.63 Loss of all alternating current power. (IEEE Std 308-2020 Note g in Table 1) which requires in part that each light-water-cooled nuclear power plant licensed under subpart C of 10 CFR part 52 after the Commission makes the finding under § 52.103(g) of this chapter, and each design for a light-water-cooled nuclear power plant approved under a standard design approval, standard design certification, and manufacturing license under part 52 of this chapter must be able to withstand for a specified duration and recover from a station blackout as defined in § 50.2. The specified station blackout duration shall be based on the following factors:

(i) The redundancy of the onsite emergency ac power sources; (ii) The reliability of the onsite emergency ac power sources; iii) The expected frequency of loss of offsite power; and iv) The probable time needed to restore offsite power.

NOTE: NRC RG 1.155 provides further guidance for conformance with Station Blackout Rule.

4. GDC 23, Protection System Failure Modes requires that the protection system be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, radiation) are experienced

5. 50.71 Maintenance of records, making of reports. (Reference Section 8 of IEEE 308-2020 refers to documentation and Test Records)

NRC Comment 4-2 Response The staff partially agrees with the comment and is adding 10 CFR Part 50, Appendix R, 10 CFR 50.63, 10 CFR 50.71, and GDC 23 to the applicable regulations section, as well as RGs 1.155 & 1.189 to the related guidance section.

IEEE Std. 308-2020, Section 4.1, General states that Class 1E power systems shall be designed to provide that no design basis event causes: (1) A loss of electric power to a number of engineered safety features, surveillance devices, or protection system devices so that a required safety function cannot be performed, and (2) A loss of electric power to equipment that could result in a reactor transient capable of causing significant damage to the fuel cladding or to the reactor coolant pressure boundary. As such, the scope of IEEE Std. 308-2020 does not include beyond design basis events. Therefore, the staff disagrees with the inclusion of 10 CFR 50.155 because it is outside the scope of this RG.

Comment 4-3 Gurcharan Matharu (DG-1420)

In the Draft Guide, the staff has elaborated and provided further clarification on harmonics in power systems. There are other phenomena that can degrade power systems and adversely impact safe shutdown capability. Of particular interest are consequences of Geomagnetic Storms and Electromagnetic Pulse.

NRC staff should elaborate on actions required for new plants.

NRC Comment 4-3 Response The staff disagrees with the comment. Background section of DG-1420 states: NRC published Research Information Letter 2023-09, An Assessment of the Harmonics Effects on Undervoltage Relays in Nuclear Power Plants that contains information on the effects of harmonic distortion on electrical

6 protection relay operation. Research Information Letter 2023-09 includes considerations regarding Geomagnetic storms, lightening, etc. NUREG 0933, Resolution of Generic Safety Issues, Issue 20:

Effects of Electromagnetic Pulse (EMP) on Nuclear Power Plants (Rev. 1) addresses electromagnetic pulse events.

No changes were made to DG-1420 as a result of this comment.

Comment 4-4 Gurcharan Matharu (DG-1420)

Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

4.10 Single-failure criterion This section refers to IEEE Std 352' and IEEE Std 577' which provide guidance for probabilistic assessment. Is there a RG approving the use of these Standards? Is this acceptable in a deterministic design basis?

4.8 Independence Independence of redundant equipment and circuits shall be in accordance with IEEE Std 384'. Please clarify if this is acceptable in view of requirements noted in RG 1.75 and Appendix R.

NRC Comment 4-4 Response The staff disagrees with the comment. There are currently no RGs endorsing standards IEEE Std. 352 and IEEE Std. 577. However, the NRC staff is considering endorsement of these two standards in a separate RG (see ML20212L534 and ML20189A599).

Regarding independence of redundant equipment, RG 1.75, endorsing IEEE Std. 384, is referenced in DG-1420 No changes were made to DG-1420 as a result of this comment.

Comment 4-5 Gurcharan Matharu (DG-1420)

Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

5.3.4.7 Transient protection Consider adding the flowing clarification: Provisions shall be made for the battery charger to prevent transients from the ac system from unacceptably affecting the dc system, and vice versa. Elaborate on the events that occurred at Forsmark nuclear plant whereby a transient event in the offsite power disabled redundant battery chargers and uninterruptible power supplies.

7 5.3.4.3 Capability NRC Comment 4-5 Response The staff disagrees with comment. Operating experience is considered as part of licensees 10 CFR Part 50, Appendix B programs. There are existing RGs that address battery chargers and uninterruptible power supplies, some of which are mentioned in DG-1420.

No changes were made to DG-1420 as a result of this comment.

Comment 4-6 Gurcharan Matharu (DG-1420)

Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

5.3.3.6 Test provisions Means shall be provided to perform battery capacity tests in accordance with IEEE Std 450'. Discuss applicability of RG 1.129 for battery capacity testing.

NRC Comment 4-6 Response The staff disagrees with the comment. RG 1.129, endorsing IEEE Std. 450, is referenced in DG-1420.

No changes were made to DG-1420 as a result of this comment.

Comment 4-7 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

5.4.2.3 Independence Distribution circuits to redundant equipment shall be physically and electrically independent of each other in accordance with IEEE Std 384. Please clarify if this is acceptable in view of requirements noted in RG 1.75 and Appendix R.

NRC Comment 4-7 Response The staff disagrees with the comment. RG 1.75, endorsing IEEE Std. 384, is referenced in DG-1420.

No changes were made to DG-1420 as a result of this comment.

Comment 4-8 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

8 6.4 Periodic tests The periodic tests shall be performed at scheduled intervals in accordance with IEEE Std 338'. Please clarify that the periodic test shall be performed in accordance with licensing basis of the plant.

NRC Comment 4-8 Response The staff disagrees with the comment. RG 1.118, endorsing IEEE Std. 338, is referenced in DG-1420.

No changes were made to DG-1420 as a result of this comment.

Comment 4-9 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

5.2.4.5 Energy storage Stored energy (fuel) at the site shall be of sufficient quantity to operate the standby power supply while supplying post-accident power requirements to a unit for the longer of the following:

Seven days The time required to replenish the energy from sources away from the generating units site following the limiting design basis event.

Elaborate on guidance provided in RG 1.137 for Fuel oil Systems for Emergency Power Supplies NRC Comment 4-9 Response The staff agrees with comment. RG 1.137 is added to DG-1420 in the related guidance section.

Comment 4-10 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

5.3.3.5 Stored energy Stored energy shall be sufficient to provide an adequate source of power for starting and operating all required connected loads and for operating all necessary circuit breakers during an interval of time when either of the following occur:

AC to the battery charger is lost for the time stated in the design basis. Clarify the statement that stored energy requirements should satisfy SBO and BDB requirements and any additional requirements for passive designs.

NRC Comment 4-10 Response The staff disagrees with the comment. 10 CFR 50.63 and 50.155 address SBO and mitigation of beyond design basis scenarios. Specifically, 10 CFR 50.63(a)(2) states, in part, that the reactor core and

9 associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled, and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. 10 CFR 50.155(c) requires, in part, that equipment relied on for mitigation strategies must have sufficient capacity and capability, which could include DC systems. Therefore, stored energy requirements are included in the scope of these rules.

No changes were made to DG-1420 as a result of this comment.

Comment 4-11 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard.

6.3 Preoperational system test Consider elaborating more as proposed in italics below:

The preoperational system tests shall be performed with all components installed. These tests shall demonstrate that the equipment operates within design limits and that the system is operational and can meet its performance specification. These tests shall be performed after the preoperational equipment tests and shall demonstrate that:

a) All required coincident Class 1E and non-Class 1E loads can operate acceptably within the in the allowable range of voltage and frequency and voltage of preferred power supply. The Class 1E and non-Class1E loads should not be adversely impacted by power systems transients expected in the offsite power systems b) Degraded and loss of the preferred power supply can be detected and transfer to the Class 1E power source accomplished in time consistent with assumptions in safety analyses (or design basis) without degrading the safety class equipment that may be operating.

e) Transfer between preferred and standby power supplies can be accomplished in a time commensurate with assumptions in design basis or safety case.

f) The batteries of the dc power supply can meet the design requirements of their connected load without the charger(s) in operation for the duration that complete loss of onsite and offsite power sources is postulated in safety analyses.

NRC Comment 4-11 Response The staff agrees with the comments. Changes were made to DG-1420 according to the italicized text above.

Comment 4-12 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard. Consider elaborating more as proposed in italics below:

6.4 Periodic tests Tests shall be performed at scheduled intervals to:

10 Detect within practical limits the deterioration of the equipment toward an unacceptable condition.

Demonstrate that standby power equipment and other components that are not exercised during normal operation of the station are operable for the duration assumed in safety analyses.

The testing of Class 1E equipment shall be scheduled to confirm that sufficient equipment is available at all times to fulfill the safety function.

The periodic tests shall be performed at scheduled intervals in accordance with licensing basis of the plant. IEEE Std 338'.

NRC Comment 4-12 Response The staff agrees with the comment. DG-1420 is changed to include the suggested changes.

Comment 4-13 Gurcharan Matharu Please consider elaborating in a similar level of detail on the following observations on the various sections of the IEEE Standard. Consider elaborating more as proposed in italics below:

8.1 Design documentation records a) Steady-state and voltage profile studies that show the voltages throughout the power system various modes of plant operation, including design basis events, at the time of normal and degraded voltage conditions. The studies should also demonstrate the ability of the plant power system to withstand transients in the offsite power system (switching, lightning, etc.) that will be experienced before automatic protective actions and let through currents adversely impact operating equipment.

c) An I&C power system study that examines loading and voltages in the ac and dc systems for postulated design basis conditions. The power system analyses should also document the ability of power sources (including the uninterruptible power source) to withstand power transients in the AC power supply that can result in common mode failure of redundant equipment (Forsmark event)

NRC Comment 4-13 Response The staff disagrees with the comment. IEEE Std. 308-2020, Section 4.4, Design Basis, item f) states that malfunctions, accidents, environmental events, and operating modes that could physically damage Class 1E power systems shall be addressed, and item g) states that the acceptable ranges for transient and steady-state conditions shall be addressed. Furthermore, IEEE Std. 308-2020, Table 1, Illustrative malfunctions, accidents, etc. includes lighting. Therefore, the NRC staff notes that transient conditions are addressed in the standard and no further clarification is needed.

No changes were made to DG-1420 as a result of this comment.

11 Comment Submission 5 ADAMS Accession No. ML24271A192 Name: IEEE PES NPEC Working Group 4.7, Protection of Class 1E Power Systems Email: mdbowman@ieee.org Comment 5-1 IEEE PES NPEC Working Group 4.7 The following comment is being submitted as the consensus position of the working group and does not necessarily reflect the views of IEEE, IEEE-SA, PES, or the Nuclear Power Engineering Committee (NPEC).

We consider the Section 5.1.2C supplemental guidance on RG1.105 only applies to Degraded Grid Voltage and Loss of Voltage settings because they are in plant technical specifications, where applicable.

We are concerned that this supplemental guidance could set an unwanted precedent that other electrical protection devices are commensurate with GDC Section III Protection and Reactivity Control Systems.

We recommend this supplemental guidance be reworded or removed.

NRC Comment 5-1 Response The staff disagrees with the comment. RG 1.105, Setpoints for Safety-Related Instrumentation describes an approach that is acceptable to the NRC to meet regulatory requirements to ensure that: a) setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and b) the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant technical specifications. RG 1.105, in Section A, Applicable Regulations, lists GDC 13, Instrumentation and Control, and GDC 20, Protection System Functions. Therefore, RG 1.105 is applicable to safety-related instrumentation to protect nuclear power safety and analytical limits, which could include electrical protection devices or instrumentation pertaining to electrical protection devices that comply with GDC 20.

Regulatory guides are not NRC regulations and compliance with them is not required, as stated in DG-1354, Section A, Purpose of Regulatory Guides. Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission. Therefore, RG 1.105 could be applied but is not required for other electrical protection devices commensurate with GDCs 20-29, as listed in 10 CFR Part 50, Appendix A, Section III, Protection and Reactivity Control Systems.

No changes were made to DG-1354 as a result of this comment; however, since the issuance of DG-1354, the NRC staff has updated generic language in Section A, Introduction, subsection, Purpose for Regulatory Guides, for all DGs and RGs as follows:

The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if the applicant provides sufficient basis and information for the NRC staff to verify that the alternative methods comply with the applicable NRC regulations.

In addition, the NRC staff has updated the generic language in Section D, Implementation, of all DGs and RGs as follows:

12 Licensees generally are not required to comply with the guidance in this regulatory guide.

If the NRC proposes to use this regulatory guide in an action that would constitute backfitting, as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests; affect the issue finality of an approval issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants; or constitute forward fitting, as that term is defined in Management Directive 8.4, then the NRC staff will apply the applicable policy in Management Directive 8.4 to justify the action. If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may inform the NRC staff in accordance with Management Directive 8.4.

Both DG-1354 and DG-1420 will be updated to reflect the new generic language identified above.