ML20330A328

Public Comment Resolution Table for DG-1363

ML20330A328
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Issue date:	01/31/2021
From:	Office of Nuclear Regulatory Research
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20330A304	List: ML20330A328
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DG-1363, RG-1.105, Rev 4
Download: ML20330A328 (26)

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Response to Public Comments on Draft Regulatory Guide DG-1363 Setpoints for Safety-Related Instrumentation Proposed Revision 4 of Regulatory Guide (RG) 1.105 On August 14, 2020, the NRC published a notice in the Federal Register (85 FR 49685) that Draft Regulatory Guide, DG-1363 (Proposed Revision 4 of RG 1.105), was available for public comment. The public comment period ended on September 14, 2020. The NRC received comments from the organizations listed below. The NRC has combined the comments and NRC staff responses in the following table.

Comments were received from the following:

1. Anonymous 2. Anonymous Agencywide Document and Management System ADAMS Accession No. ML20234A198 (ADAMS) Accession No. ML20234A197

3. Anonymous 4. Anonymous ADAMS Accession No. ML20234A200 ADAMS Accession No. ML20234A202

5. Anonymous 6. Anonymous ADAMS Accession No. ML20234A204 ADAMS Accession No. ML20234A697

7. Anonymous 8. Anonymous ADAMS Accession No. ML20240A256 ADAMS Accession No. ML20240A257

9. Anonymous 10. Anonymous ADAMS Accession No. ML20240A259 ADAMS Accession No. ML20240A260

11. Anonymous 12. Anonymous ADAMS Accession No. ML20241A164 ADAMS Accession No. ML20245E263

13. Anonymous 14. Mendy Maxey ADAMS Accession No. ML20245E265 Meenterprise 4 Evergreen Dr.

Pine Bluff, AR 71602 ADAMS Accession No. ML20246E509 January 2021

15. Mendy Maxey 16. Anonymous Meenterprise ADAMS Accession No. ML20246G653 4 Evergreen Dr.

Pine Bluff, AR 71602 ADAMS Accession No. ML20246E530

17. Anonymous 18. Anonymous ADAMS Accession No. ML20247J618 ADAMS Accession No. ML20253A004

19. Anonymous 20. Anonymous ADAMS Accession No. ML20253A229 ADAMS Accession No. ML20253A231

21. Anonymous 22. Anonymous ADAMS Accession No. ML20254A107 ADAMS Accession No. ML20255A303

23. Anonymous 24. Stephen J. Vaughn ADAMS Accession No. ML20261H514 Senior Project Manager Engineering and Risk Nuclear Energy Institue 1201 F Street, NW, Suite 1100 Washington, DC 20004 ADAMS Accession No. ML20261H516 Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

1. Anonymous General The public review and comment period for this matter was shortened The NRC staff disagrees with this from the normal 60 days to 30 days. The rationale for this change comment. The NRC does not normally given was the assertion that the NRC has previously interacted with hold public meetings for the proposed stakeholders on related industry and NRC guidance and the proposed revisions to Regulatory Guide updates.

revision endorses ANSI/ISA 67.04.01-2018 without any exceptions An explanation of the transition from or clarifications. There have not been a public meeting on this matter DG-1141 to DG-1363 is provided in since August 2014 on DG-1141 so many of those who have detail within the Federal Register notice commented on the previous draft regulatory guide are at a for this proposed revision issued on disadvantage first to become aware of this notice and then to review August 8, 2020 (85 FR 49685). The the new version and prepare appropriate comments. The staff has had International Society of Automation sufficient time since December 2018 when ISA Standard 67.04.01- (ISA) Standards Development January 2021

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) 2018 was issued to conduct a public meeting to gather feedback from Organization (SDO) follows the other stakeholders who have similar systems and components which American National Standards Institute perform safety-critical functions. (ANSI) endorsed process for developing standards, which includes opportunities for the public to submit comments to the standards organization(s) during the development of the draft standard.

Further, because the NRC is endorsing this consensus ANSI/ISA 67.04.01-2018, Setpoints for Nuclear Safety Related Instrumentation, without any exceptions, the 30 day comment period was deemed appropriate.

The NRC staff made no changes to DG-1363 as a result of this comment.

2. Anonymous General Obsolescence of systems and components, and market conditions, is The NRC staff partially agrees with this incentivizing nuclear power plant (NPP) owners to upgrade outdated comment with respect to the need for analog Instrumentation and Control (I&C) systems with digital digital technology to be used along with technology. Most U.S. operating plants have extended their license its associated uncertainties and that DG-to 60 years (potential extension to 80 years), so replacing outdated 1363 does not address quantization error.

1960s-70s technology is unavoidable. Digital components (e.g., I/O However, the ANSI/ISA 67.04.01-2018 modules, software) may introduce new errors into the measurement standard is technology inclusive and due to digital technology. The regulatory analysis of the draft provides the bounding requirements and regulatory guide states that the revision would incorporate the latest attributes one needs to analyze while information in setpoint determination. However, the draft regulatory evaluating uncertainties. DG-1363 guide does not address Quantization Error which may be introduced endorses this standard and Section 4.4 of due to digital upgrade changes in NPPs. the standard states that all sources of uncertainties need to be addressed.

The process by which an analog signal (sampled and held at a Section B.2.2.8 of DG-1363 also constant value), is approximated to a set of values meant to represent identifies that additional information for the signal and dependent on number of bits used to represent the determining total loop uncertainties may signal. The analog signal and its digital representation after be found in the ISA Recommended 3

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) quantization is the Quantization Error (Rounding/Round off and Practice (RP) document ISA-Truncation/Truncating errors). RP67.04.02-2010, Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation, which implements the set of requirements specified in the standard.

The identification and treatment of digital processing-related channel uncertainties are covered in depth within Section 6.2.9 and Annex H of ISA-RP67.04.02-2010. In addition, ISA periodically reviews their recommended practices to look for needed revisions.

However, as stated in DG-1363, the NRC staff does not endorse any version of ISA RP67.04.02, but the staff believes those versions contain useful information.

The NRC staff made no changes to DG-1363 as a result of this comment.

3. Anonymous Section C The endorsed standard (i.e., ISA 67.04.01-2018) contains a figure that The NRC staff disagrees with this is misleading (i.e., Figure 1, "Relation between setpoint parameters") comment. Both the safety limit and which should be clarified in the RG. Typically, the accident analysis analytical limit are established by a presumes a particular protective action is initiated at a particular safety analysis outside the scope of ISA process parameter value (i.e., the analytical limit) and takes an 67.04.01-2018. The safety limit and the assumed amount of time to achieve the protective action. The analytical limit in Figure 1 are only accident analysis then determines the most extreme values that all the provided for illustrative purposes as a process parameters reach; these extreme values are then compared to reference point.

the associated safety limits. That is, the process parameter that initiates the protective action (e.g., primary coolant temperature) The NRC staff made no changes to DG-generally is not the same process parameter that has an associated 1363 as a result of this comment.

safety limit; that is, the safety limits (typically listed in the technical 4

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) specifications) are very few, and are often not directly measurable by sensors (i.e., they are calculated). Therefore, the analytical limit and the associated safety limit are generally on two different process parameters. An explanation superior to this one should be included in the RG.

4. Anonymous General For analog to digital system upgrades the analog signals received by The NRC staff disagrees with these three digital processor; filtered, digitized, manipulated, converted back into comments because ISA 67.04.01-2018 analog form, filtered again, sent out for safety-related purposes. The and DG-1363 are not intended to address associated filter component reduces aliasing noise introduced by any particular type of error. The signal frequencies high relative to fixed sampling rate and the identification and treatment of digital amplitude of signal is held long enough to permit conversion to a processing-related channel uncertainties digital word. For a sampling rate higher than twice analog signal are covered in depth within Section 6.2.9 bandwidth, then the sampled signal is a good representation of analog and Annex H of the ISA-RP 67.04.02-input signal. Analog signals containing frequencies too high versus 2010, and Section B.2.2.8 of DG-1363 the sampling rate, aliasing uncertainty will be introduced. Either anti- identifies that additional information aliasing band limiting filters should be used to minimize aliasing may be found in the ISA RP. In general, uncertainty or this error should be accounted for in setpoint all uncertainties, including digital calculations. The draft regulatory guide should address this type of uncertainties, within a instrument error. channel need to be addressed when establishing safety-related instrument

5. Anonymous For analogy to digital upgrades in aging nuclear power plants the setpoints. In addition, ISA periodically analog to digital converter (A/D Converter) is a source of errors reviews their recommended practices to associated with digital technology. For example, (1) Digitizing look for needed revisions. However, as Uncertainty - associated with A/D Converters such that sampled stated in DG-1363, the NRC staff does signal amplitude at that time divided into a finite number of levels, not endorse any version of ISA digital word n bits long. The lower the numbers of bits, the greater the RP67.04.02, but the staff believes those digitizing uncertainty; (2) Linearity Error -maximum deviation of the versions contain useful information.

A/D converter from ideal to the actual; and (3) Gain Error - deviation between full scale actual change in input signal and output of the A/D The NRC staff made no changes to DG-converter. The draft regulatory guide should address these types of 1363 as a result of this comment.

errors.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

6. Anonymous Analog signal approximated to a quantization level value higher than the original analog signal Truncation Error: Analog signal above the nearest quantization level is dropped. Usually the main source of error in numerical integration or solution of differential equations.

Errors can be amplified as they propagate through a computation: (1)

Loss of precision in displayed or monitored parameter, and (2) Cause oscillation in closed loop control systems; control error (difference between measured value and control setpoint) inaccurately represented and output signal either set too high or too low, depending on the error. Overflow Error: Result of a computation that cannot be held in the accumulator. It may result in wraparound error.

Such type errors have been implicated in two high-visibility rocket accidents: (1) Failure of U.S. Patriot missile to intercept Iraqi-launched Scud missile during Gulf War, and (2) Failure of Ariane 5 launch vehicle during maiden flight. Indicating Reading Error: Error applied to accuracy when reading analog and digital indications in an instrument loop or on M&TE. The draft regulatory guide should address these types of errors.

7. Anonymous DG-1141, the predecessor to the subject draft regulatory guide DG- The NRC staff disagrees with this 1363, defined a random variable as follows: comment. The exclusion of total loop uncertainty (TLU) in DG-1141 was Trippoint = {Measured Setpoint} + {unknown errors] Trippoint is the associated with a particular set of value of the process variable at which a channel actually does trip circumstances that are explained within under operating conditions DG-1141 and not always applicable.

(including design basis condition). Section 4.4 of ANSI/ISA 67.04.01-2018 states, The TLU shall account for the Figure 2, Trippoint Probability Distribution, of DG-1141 depicts an effects of all applicable design-basis Actual Trippoint (ATP) distribution intended to show the importance events and the following process of separating the limiting setpoint from the analytical limit (AL) by instrument uncertainties unless they were an amount not less than the total loop uncertainty (TLU). included in the determination of the analytical limit, considering as a The industry standard, American National Standard Institute (ANSI)/ minimum. Therefore, the concerns International Society of Automation raised in the comment with respect to 6

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

(ISA) 67.04.01-2018, Setpoints for Nuclear Safety-Related DG-1141 are adequately addressed in the Instrumentation, being endorsed by DG-1363 does not define a revised standard.

random variable equivalent to ATP. The subject standard in Section 4.4 stipulates that the Limiting Trip Setpoint (LTSP) for a trip or In addition, the NRC (as a voting actuation on an increasing process would be LTSP = AL - TLU member of the ISA Nuclear Standards The subject standard states that the data used to calculate the TLU Committee) was aware that the ISA SDO should be obtained from appropriate sources, which may include any considered the proposed guidance that of the following: operating experience, equipment qualification tests, was contained in DG-1141 and equipment specifications, engineering analysis, laboratory tests, and addressed those criteria considered to be engineering drawings. relevant in the revision to ANSI/ISA 67.04.01-2018.

TLU cited in the subject standard should account for the effects of all applicable design-basis events and process instrument uncertainties The NRC staff made no changes to DG-unless they are included in the determination of the analytical limit. In 1363 as a result of this comment.

contrast, per Figure 2 of DG-1141, TLU is characterized as Bias and Random factors.

Further, Section C of DG-1141 utilizes the ATP variable to determine the staff regulatory positions. For example, one regulatory position in Section C.8.d states:

As used to determine the limiting setpoint, the TLU does not need to include setting tolerance. If the setting tolerance is included in the TLU but is not to be included in the determination of the limiting setpoint, then, for the purpose of determining the limiting setpoint, the setting tolerance should be removed The subject standard includes the setting tolerance in the determination of the TLU. Given that the regulatory analysis for DG-1363 commits to address the technical issues related to the issuance of DG-1141, the draft regulatory guide should disposition the staff's position on ATP.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

8. Anonymous Advanced sensors and portable devices which interconnect with Although the NRC staff agrees that internet of things (IoT) and industrial IoT (IIoT) devices can improve digital technology and interconnectivity nuclear power plant operation, maintenance, decommissioning and are increasing in use, the NRC staff data storage. Big data technology offers a better understanding of disagrees with the recommendation of system and equipment performance by remote monitoring and data this comment as it relates to collection. The use of artificial intelligence technology to analyze Cybersecurity guidance. Cybersecurity real-time data could inform the decision-making process related to considerations and RG 5.71 are outside preventive maintenance and plant operations (e.g. manual and the scope of DG-1363. The scope of automatic response). regulations and requirements applicable to DG-1363 are contained within Section These innovations which may reduce plant operating costs for an A of DG-1363. In contrast, cybersecurity industry beset by economic drivers to become more competitive in regulations and requirements are order to continue to stay in business may also introduce new cyber contained in 10 CFR 73.54, Protection security attack surfaces associated with the setpoint control of nuclear of digital computer and communication safety-related instrumentation. systems and networks, and RG 5.71.would not be an appropriate related Since one of the purposes of Regulatory Guide 1.105 is to ensure that guidance to list in DG-1363.

the nuclear safety-related instrumentation protect nuclear power plant safety and remain within the appropriate analytical limits, the draft The NRC staff made no changes to DG-regulatory guide revision should not be silent on potential safety 1363 as a result of this comment.

issues arising from a cybersecurity threat.

The draft regulatory guide revision should reference Regulatory Guide 5.71, Cyber Security Programs for Nuclear Facilities, and the staff should seek future opportunities to address safety issues from cyber security threats in future regulatory guidance documents.

9. Anonymous DG-1141 defines a new term Deviation (or Setpoint Deviation) as the During the development of DG-1363, the amount of change in a setpoint during the interval between scheduled staff considered and addressed technical setpoint assessments (i.e., the difference between the as-found value issues and public comments related to and the previous as-left value). The Deviation is used as described in the issuance of DG-1141. In addition, as B.5.1 to evaluate the acceptability of the asfound setpoint. discussed in the August 14, 2020 Federal Register notice, the NRC staff elected not to finalize DG-1141 as a 8

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

The industry standard, American National Standard Institute (ANSI)/ revision to RG 1.105 and chose instead International Society of Automation (ISA) 67.04.01-2018, Setpoints to evaluate the ANSI/ISA 67.04.01-2018 for Nuclear Safety-Related Instrumentation, being endorsed by DG- standard for endorsement. When the 1363 does not define any term like Deviation but uses the previously NRC issued DG-1363 as a replacement determined as-found tolerance to evaluate the asfound setpoint. for DG-1141, the draft staff positions in DG-1141 were replaced by the draft staff According to the Staff Positions C.7.a and C.7.d(2) restated below the positions in DG-1363.

as-found tolerance may need to be recalculated depending on any change to uncertainty components: As a result of public comments and concerns raised regarding DG-1141, the C.7.a The limiting value for acceptable setpoint deviation the as- ISA Standards Committee opted to found tolerance, should be computed in the setpoint uncertainty revise the standard considering all of the analysis. issues raised during the development of the 2018 revision to the standard.

C.7.d(2) The as-found tolerance should include only those uncertainty Therefore, the NRC initiated a review of components which are applicable to the as-found value measurement the 2018 version of the standard while at the time the measurement is taken. considering the previous concerns raised with DG 1141, which resulted in the In addition, Figure 1 Note 3 cited below also adds uncertainty in the issuance of DG-1363 and the evaluation of as-fund setpoints: endorsement of ANSI/ISA 67.04.01-2018.

Note 3 Section C.7c of this RG addresses the acceptability of occasional deviation in excess of the as-found tolerance (+AFT), The NRC staff made no changes to DG-provided that the deviations are neither too large nor too frequent. 1363 as a result of this comment.

Section C.7e (3) of thisRG recommends that the deviation should be deemed excessive if the as-found value (AsF) of the setpoint is less conservative than the allowable value (AV) regardless of whether or not the as-found tolerance is exceeded and whether or not the occurrence of this condition is chronic.

The industry standard methodology to evaluate as-found setpoint for acceptable performance provides more clarity and predictability in terms of implementation but DG-1363 does not withdraw the staff position on the use of the deviation methodology issued in DG-1141.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

Given that the approach in DG-1141 is significantly different from that utilized by the subject industry standard is the deviation methodology replaced with the endorsement stated in Section C.1 of DG-1363?

10. Anonymous Given the introduction of software in the digital systems and The NRC staff disagrees with the components now being used in the replacement of analog recommendation of this comment. DG-instrumentation and control (I&C) equipment in operating nuclear 1363 does not need to reference other power plants and being designed for the advanced and new reactors regulatory guidance regarding the use of neither DG-1363 or the industry standard, American National software or the effects of common cause Standard Institute (ANSI)/ International Society of Automation (ISA) failure in digital I&C since those 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation, concerns are considered to be outside the being endorsed by DG-1363 adequately addresses the safety issues scope of DG-1363. The scope of DG-associated with todays technology. 1363 is to establish and maintain limiting safety system settings (LSSS). Hazards SECY-18-0090, Plan for Addressing Potential Common Cause that may result in a common cause Failure in Digital Instrumentation and Controls, identified that SRM- failure would be considered only if it SECY-93-087 provided flexibility for the treatment of digital affects total loop uncertainty common cause failures in digital I&C systems through the regulatory determination.

tools available to the NRC staff. One of those tools is the issuance of regulatory guides to provide one acceptable method for licensees and The NRC made no changes to DG-1363 applicants to meet the agencys regulations. as a result of this comment.

In addition to the potential to introduce new common cause failure modes, software allows highly complex systems to be created and the potential for unintended ad unexpected interactions among components. The more interactions between system components and the more complex the functional design, the more the opportunities for unintended effects and consequently, the more opportunities for unsafe control actions that can lead to hazards.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

If there are regulatory guides which can provide adequate guidance to licensees and applicants on the use of software in digital I&C then DG-1363 should reference these documents.

11. Anonymous The adoption of Figure 1 of ANSI/ISA 67,04,01-2018 to illustrate the The NRC staff agrees with the comment.

setpoint relationship for nuclear safetyrelated setpoints is a welcomed The NRC staff made no changes to DG-clarification and disposition over Figure 1 in DG-1141. Some of the 1363 as a result of this comment.

key differences are

1. Figure 1 of DG-1141 does not identify the discretionary margin typically included to maintain conservativism between the Limiting Trip Setpoint (LTSP or LSP) and the Nominal Trip Setpoint (NTSP or NSP). Note also that the Setting Tolerance (ST) is included in Total Loop Uncertainty (TLU) per the industry standard whereas DG-1141 associates ST with the NSP.

2. Figure 1 of DG-1141 depicts an anticipated excursions band which is not defined in the document nor is its applicability to the NRC regulations discussed for the reader.

3. Figure 1 of the subject industry standard includes the Safety Limit is defined in 10 CFR 50.36(c)(1)(i)(A).

4. Figure 1 of the subject industry standard indicates that the as-found tolerance (AFT) is a probabilistic band taken as a maximum value above and below the desired output such that the as-left value can quickly be assessed as acceptable or not during the calibration of an instrument or instrumentation channel.

The staff is commended for providing a clear depiction of the relative position and relationship of the various setpoints.

12. Anonymous NRC Safety Evaluation dated 10/14/16 (ADAMS Accession No. Westinghouse representatives ML16256A788) which was issued after Westinghouse comments on participated in the ISA committee for the DG-1141 [Letter from Mr. James A. Gresham, Manager Regulatory 2018 revision to the standard and their 11

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

Compliance Westinghouse to Ms. Cindy Bladey, NRC dated concerns related to LSSS were discussed 10/1/2014] stated that The NRC staff has found that WCAP-17503 and addressed while preparing the final WCAP-17504-P/NP, Westinghouse Generic Setpoint Methodology version of ANSI/ISA 67.04.01-2018 and are acceptable for referring in licensing applicationswith the proper no reconciliation document was created documentation. The subject Westinghouse Letter cited WCAP-17504 for the Westinghouse comments on DG-in their comments on DG-1141. 1141.

The regulatory analysis for DG-1363 stated that the draft regulatory During the development of DG-1363, the guide would consider and address technical issues and public staff considered and addressed technical comments related to the issuance of DG-1141. Given the significant issues and public comments related to staff involvement with WCAP-17504 was there a reconciliation the issuance of DG-1141. In addition, as document issued on the Westinghouse comments on DG-1141? discussed in the August 14, 2020 Federal Register notice, the NRC staff elected not to finalize DG-1141 as a revision to RG 1.105 and chose instead to evaluate the ANSI/ISA 67.04.01-2018 standard for endorsement and issued DG-1363 as a replacement for DG-1141.

As a result of public comments and concerns raised regarding DG-1141, the ISA Standards Committee opted to revise the standard considering all of the issues raised during the development of the 2018 revision to the standard.

Therefore, the NRC initiated a review of the 2018 version of the standard while considering the previous concerns raised with DG 1141, which resulted in the issuance of DG-1363 and the endorsement of ANSI/ISA 67.04.01-2018.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

The NRC staff made no changes to DG-1363 as a result of this comment.

13. Anonymous NuScale Power LLC Letter dated September 30, 2014 provided 26 The NRC staff disagrees with the comments on DG-1141. Although most comments seem applicable to recommendation in this comment. As a DG-1141 Comments # 1, 19, 23 and 25 seems to be applicable to the result of public comments and concerns current scope of DG-1363. The points raised by NuScale regarding a) raised regarding DG-1141, the ISA the expansive treatment of small instrument errors with little safety Standards Committee opted to revise the significance; b) use of two-sided statistical approach effectively standard considering all of the issues establishing a 97.5% probability which may increase plant raised during the development of the trip/transient probability; c) NRC expectations regarding bounding 2018 revision to the standard. Therefore, values for environmental testing required for digital I&C equipment the NRC initiated a review of the 2018 per RG 1.209; and d) consistency with Standard Review Plan version of the standard while considering (NUREG-0800) and draft Branch Technical Position (BTP) 7-12. The the previous concerns raised with DG draft regulatory guide should address these technical issues 1141, which resulted in the issuance of previously raised by NuScale on DG-1141. DG-1363 and the endorsement of ANSI/ISA 67.04.01-2018.

The NRC staff made no changes to DG-1363 as a result of this comment.

14. Mendy General NRC -2020-0171 .Thanks BG MD Maxey- West The NRC staff did not consider this Maxey comment within the scope of DG-1363.

The NRC staff made no changes to DG-1363 as a result of this comment.

15. Mendy General NRC -2020-0171 .Thanks BG MD Maxey- West The NRC staff did not consider this Maxey comment within the scope of DG-1363.

The NRC staff made no changes to DG-1363 as a result of this comment.

16. Anonymous General 6 out of 8 documents in Section A authored by Mr. Paul J. Rebstock. The NRC staff will make these Non-Concurring Employee for NCP-2020-004, Non-concurrence on documents publicly available.

DG-1363 (RG 1.105) Setpoints for Safety-Related Instrumentation (NRC Form 757), listed as references are not available to members of The NRC staff made no changes to DG-the public in the NRC ADAMS database. 1363 as a result of this comment.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

Package containing all cited ADAMS documents: ML20006f031 Response Document (RD) (from 9Mar2016 concurrence pkg)

ML15335a085 (ML19317D837 includes notes)

Draft Regulatory Guide 1.105r4 (from 9Mar2016 concurrence pkg)

ML15135a255 (ML19317d838 includes tracking to RD commitments) 95/95 and SSS --- ML19239a261 Can the above documents be made publically available to clearly understand the safety and/or regulatory concerns of the Non-Concurring Employee?

17. Anonymous DG Section During the Advisory Committee for Reactor Safeguards (ACRS) The NRC staff disagrees with this C, Meeting held on November 21, 2019, one ACRS Member asked why comment. The staff believes that the subsection the Electric Power Research Institute (EPRI) which has studied standard is not deficient in the manner 2.2.3 instrument drift for years could not share that data with the NRC staff stated by the commenter. The purpose of to help address the decades-long concerns on this issue. The EPRI the standard is not to specify what data to representative stated that non-disclosure agreements prevented the use but rather to identify appropriate sharing of the raw data with the NRC staff. attributes of the instrument channel performance when determining the TLU.

Revision 1 of Regulatory Guide (RG) 1.105 noted in the Discussion The instrument performance data used to Section that Abnormal Occurrence Reports submitted by operating determine the TLU is not within the utilities between January 1972 and June 1973 record the most scope of this RG. In addition, the staff frequent abnormal occurrence as the drift of the protective instrument finds that the footnote in Section 4.6 of setpoint outside the limits specified in the technical specifications. ANSI/ISA 67.04.01-2018 is acceptable for clarifying the guidance in the In Section 2.2.3 (Evaluation of the Allowance for Drift) of DG-1363, standard and the staffs endorsement of the NRC staff provided additional direction beyond what is stated in the standard in Section C of DG-1363 ANSI/ISA 67.04.01-2018 with the following statements: does not need any clarifications.

As described in the footnote in Section 4.6 of the standard, when However, to avoid confusion, the staff determining the magnitude of drift to be included when establishing made the following edits to DG-1363 the AFT, licensees should estimate on the low side so as not to and replaced some of the current 14

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) potentially mask the ability to detect a degrading instrument during a language in DG Section B.2.2.3. with the required surveillance. In addition, when establishing an appropriate exact language from the Section 4.6 allowance for total loop uncertainty between the analytical limit and footnote of the standard, stated as the limiting trip setpoint, licensees should estimate drift on the high follows:

side so as to ensure adequate margin for instrument channel performance in achieving the safety objectives. Here, estimates of more conservative performance test acceptance criteria are These statements imply that the subject standard is deficient in some those that result in acceptance tolerances manner. Why not identify this additional direction in Section C as a that tend to be on the smaller side, so as regulatory position to clarify the staffs intent? Given that instrument not to mask any adverse performance.

drift is just another random uncertainty why not provide acceptable For estimates of TLU, more conservative ways of handling this uncertainty as provided using probabilistic risk estimates of total uncertainty are those assessment techniques (i.e., RG 1.174)? that tend to be on the larger side, so as not to underestimate the required minimal allowance for instrument channel performance uncertainty between the analytical limit and the limiting trip setpoint.

18. Anonymous DG-1363 While it is not clear why the NRC staff chose in Section 2.2.2, The NRC staff agrees with this Section B, Adequacy of the Allowance for Channel Uncertainties between the comment. The language in DG-1363 subsection Limiting Trip Setpoint and the Analytical Limit, of DG-1363 to Section B.2.2.2 is consistent with the 2.2.2 paraphrase portions of ANSI/ISA 67.04.01-2018 there is the potential language in ANSI/ISA 67.04.01-2018.

for this section to cause confusion. The ISA Standard 4.4 states If However, to avoid any confusion, the there is not sufficient data to justify a statistical estimate but Section staff has replaced the second sentence of 2.2.2 states for cases in which the sample population is not large the second paragraph in Section B.2.2.2, enough to support a usable statistical estimate and replaced it with the following text from Section 4.4 of the ISA standard:

Section 2.2.2 omits noting that drift and reference accuracy should not be among those uncertainty terms which may have insufficient If there is not sufficient data to justify a data to estimate the confidence level (see ISA Standard Section 4.4, statistical estimate of the uncertainty page 19). In describing the ISA Standard, Section 2.2.2 states that the tolerance interval at the 95/95 level, then result of the combination represents a value of the random uncertainty a bounding uncertainty term shall be 15

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) performance of the instrument channel at a 95-percent probability at a determined, and the basis for 95-percent confidence level versus the clearer statement in the ISA determining the bounds of the Standard that The result of the combination shall be a value that uncertainty shall be documented. The represents the performance of the instrumentation with a 95% bounding estimates shall be treated as a probability at a 95% confidence level, 95/95 term in the uncertainty analysis.

19. Anonymous General NRC should seize this transformational opportunity to use DG-1363 The NRC staff agrees with the to define the content of the information that new/advanced reactor conclusion of this comment. However, applicants and operating reactor licensees need to submit in order to licensees are only required to comply adequately address instrumentation uncertainties. In the past there with the requirements in their licensing have been a range of outcomes centered on the proper communication basis and any new reactor applicants are and understanding of the instrument uncertainties for a change to the encouraged to follow the latest NRC licensing basis of a nuclear power plant. guidance. In addition, the ANSI/ISA 67.04.01-2018 standard endorsed in DG-For example, one outcome involving good communication of the 1363 is applicable to the establishment uncertainties pertained to the measurement uncertainty recapture and maintenance of LSSS as addressed (MUR) power uprates program as described in Regulatory in 10 CFR 50.36. The discussion Information Summary 2002-03, Guidance on the Conduct of presented in the comment deals with Measurement Uncertainty Recapture Power Update Applications. operational issues of plants within their Several licensee applications utilized the NRC approved the Caldon licensed condition and does not relate to Ultrasonic Inc. (Caldon)), the leading edge flow meter (LEFM) LSSS.

CheckPlus System, [ER-80P with ER-157P design] to increase their maximum power up to 1.7% depending on being able to adequately The NRC staff made no changes to the accounted for all instrumentation uncertainties in the reactor thermal DG-1363 as a result of this comment.

power measurement uncertainty calculations. Specifically, from the earliest to later application the following nuclear power plants (%

power increase approved) utilized the guidance provided by the NRC staff for the above ultrasonic flowmeter design: Waterford (1.5);

Sequoyah (1.3); Grand Gulf (1.7); H. B. Robinson (1.7); Peach Bottom (1.62=>1.66); Point Beach (1.4); D. C. Cook (1.66/1.7);

River Bend (1.7); Seabrook (1.7) Crystal River (1.7); Vogtle (1.7);

Cooper (1.62); David-Besse (1.63); Calvert Cliffs (1.38); North Anna (1.6); Prairie Island (1.64); LaSalle County (1.65); Surry (1.6);

Limerick (1.65); Shearon Harris (1.66); McGuire (1.7); Braidwood 16

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

(1.63); Fermi (1.64); Catawba (1.7); Columbia (1.66);and Hope Creek (1.62). The guidance provided in RIS 2002-03 and Regulatory Guide 1.105, Revision 3 was used by the NRC staff and licensees for these completed licensing actions.

In another example. in 2006, LaSalle County Nuclear Power Plant was denied its licensee amendment request to revise Technical Specification (TS) 3.7.3, Ultimate Heat Sink [UHS], to increase the temperature limit of the cooling water supplied to the plant from the core standby cooling system pond (i.e., UHS) from 100 degrees Fahrenheit to 101.5 degrees Fahrenheit. The licensee proposed to reduce the temperature measurement uncertainty by replacing the existing thermocouples with higher precision temperature measuring equipment. The NRC Safety Evaluation (Agencywide Document and Management System (ADAMS) Accession No. ML062760617),

concluded that the licensee had not shown that the instrumentation is sufficiently accurate to justify the requested reduction in margin and provided adequate assurance that anticipated measurement errors would not exceed the margin between the proposed TS limit and the temperature assumed in the plant safety analyses. It should be noted that current TS 3.7.3 for LaSalle County Nuclear Power Plant does reflect an increase temperature limit from 100 degrees Fahrenheit to 101.25 degrees Fahrenheit.

Therefore, it would benefit both new and advanced reactor applicants and operating reactor licensees to obtain informed guidance from the NRC staff on how to address instrumentation uncertainties in their future submittals.

20. Anonymous As noted below there is a more stringent regulatory requirement in The NRC staff agrees in part with this the reactor area (i.e., guidance document) than the nuclear materials comment in that the criteria specified in area (i.e., NRC regulation) proposed for licensees and applicants by 10 CFR 50.68 forms the basis for the 95-DG-1141, although Regulatory Guide 1.105, Revision 3 stipulates the 95 criteria for setpoints. This criteria is applied to each instrument loop 17

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) same 95 percent probability, 95 percent confidence level criteria as independently. Therefore, the use of specified in 10 CFR 50.68. redundant channels in combination with the 95% probability and 95% confidence Background Section of NRC Information Notice 2011-03, level provides a reasonable assurance of Nonconservative Criticality Safety Analyses for Fuel Storage, states adequate protection relating to bounding the following: uncertainties.

Paragraph 50.68(b)(4) of 10 CFR 50.68, Criticality Accident However, the NRC staff disagrees with Requirements, requires the following: If no credit for soluble boron is the comment with respect to the taken, the k-effective of the spent fuel storage racks loaded with fuel associated non-concurrence NCP-2020-of the maximum fuel assembly reactivity must not exceed 0.95, at a 004 (ADAMS Accession No.

95 percent probability, 95 percent confidence level, if flooded with ML20181A524). The non-concurrence unborated water. If credit is taken for soluble boron, the k-effective of was dispositioned using the NRCs the spent fuel storage racks loaded with fuel of the maximum fuel established policies and procedures assembly reactivity must not exceed 0.95, at a 95 percent probability, located in Management Directive 10.158, 95 percent confidence level, if flooded with borated water, and the k- NRC Non-Concurrence Process, effective must remain below 1.0 subcritical), at a 95 percent (ADAMS Accession No.

probability, 95 percent confidence level, if flooded with unborated ML18073A296). NCP-2020-004 water. concluded that no changes to DG-1363 were required due to the non-NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety concurrence. Further, the NRC cannot Calculational Methodology, January 2001 (Agencywide Document use a guidance document to change a and Management System (ADAMS) Accession No. ML050250061), regulation, as the commenter suggests by provides guidance on determining the bias uncertainty for Monte saying the NRC staff should make the Carlo codes. safety case for a more stringent requirement than what is in place for 10 The primary NRC staff guidance regarding the depletion uncertainty CFR 50.68.

is an internal NRC memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of The NRC staff made no changes to DG-Fuel Storage at Light-Water Reactor Power Plants, dated August 19, 1363 as a result of this comment.

1998 (ADAMS Accession No. ML003728001) (Kopp Letter). The Kopp Letter is referenced by virtually all spent fuel pool criticality license amendment requests submitted since its issuance.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

Regarding the depletion uncertainty, the Kopp Letter states the following:

A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption.

Although DG-1363 through the endorsement of ANSI/ISA 67.04.01-2018 returned to the Revision 3 criteria for probability and confidence level sought for instrumentation performance, the associated Non-Concurrence statement by a NRC Senior Instrumentation & Control Engineer indicates the desire for a more stringent requirement.

Section 2 of DG-1363 should either make the safety case for a more stringent requirement than what is in place for 10 CFR 50.68, Criticality Accident Requirements, or discuss the acceptability of the 5 percent uncertainty.

21. Anonymous DG-1363 In DG-1363, Section 2.2.6, Graded Approach Based on Safety The NRC staff disagrees with this Section C, Significance, there is the following statement: comment. ANSI/ISA 67.04.01-2018 is a subsection deterministic standard for establishing 2.2.6 The NRC staff has not identified a specific position on the and maintaining LSSS for compliance appropriate technical methodology to be used when establishing with 10 CFR 50.36 requirements. 10 setpoints for nonlimiting safety system setting-related instrument CFR 50.36 does not have any provisions channels. for risk informing the LSSS determination.

However, the Staff Position in Interim Staff Guidance (ISG) DI&C-ISG-03, Task Working Group #3: Review of New Reactor Digital Although the risk-informed approaches Instrumentation and Control Probabilistic Risk Assessments, in ISG DI&C-ISG-03 and draft BTP 7-Revision 0 (Initial Issue for Use) which utilize a risk-informed 19, Revision 8, are available, these approach to review digital instrumentation and controls systems and approaches are not within the scope of components identifies any attributes that warrant additional DG-1363.

19

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) regulatory attention and whether there are highlevel, risk-significant problems, including the existence of risk outliers. Therefore, DI&C- The NRC staff made no changes to DG-ISG-03 should be referenced as an acceptable guidance document to 1363 as a result of this comment.

assist applicants and licensees in the development their own design-specific safety significant graded approach. In addition, the Draft Branch Technical Position (BTP) 7-19, Guidance for Evaluation of Potential Common Cause Failure Due to Latent Software Defects in Digital Instrumentation and Control Systems, Revision 8, June 2020, introduced a graded risk-informed approach which could also assist applicants and licensees in developing a graded approach for setpoints. The existence of ISG DI&C-ISG-03 and draft BTP 7-19, Revision 8, seem to contradict the above mentioned statement that there is no applicable staff position.

22. Anonymous Excerpts from the comment submitted by Mr. Jerald Head on October The NRC staff disagrees with this 8, 2014 for General Electric Hitachi (GEH) [Agencywide Document comment. GEH representatives and Management System (ADAMS) Accession No. ML14283A501] participated in the ISA committee for the are cited below. The draft regulatory guide should address the 2018 revision to the standard and their technical issues previously raised by GEH on DG-1141. concerns related to LSSS were discussed and addressed while preparing the final GEH COMMENT version of ANSI/ISA 67.04.01-2018. In The Draft DG-1141 appears to impose a requirement of 97.5% addition, as discussed in the August 14, probability of single channel trip before the AL is reached. This is 2020 Federal Register notice, the NRC inconsistent with the current and previous revisions of RG 1.105 staff elected not to finalize DG-1141 as a (Revision 3 and earlier) which clearly define the requirement of trip revision to RG 1.105 and chose instead before AL is reached to be 95% probability. The previous 95% to evaluate the ANSI/ISA 67.04.01-2018 probability requirement is the basis of the licensed GEH safety standard for endorsement and issued analyses, and the basis of the NRCapproved GEH setpoint DG-1363 as a replacement for DG-1141.

methodology (Reference 2). As a result of public comments and concerns raised regarding DG-1141, the The GEH safety analysis application methodologies use the same ISA Standards Committee opted to 95/95 definition. This is evidenced by a letter from the NRC to GE revise the standard considering all of the (Reference 3) which states, in part, "This procedure provides for a issues raised during the development of statistical determination of the pressurization transient CPR/ICPR the 2018 revision to the standard.

20

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) such that there is a 95% probability with 95% confidence (95/95) that Therefore, the NRC initiated a review of the event will not cause the critical power ratio to fall below the the 2018 version of the standard while MCPR Fuel Cladding Integrity Safety Limit." Thus, 95% is the non- considering the previous concerns raised exceedance %/probability. GEH has consistently used this with DG 1141, which resulted in the 95/95=95% non-exceedance definition in analysis of Anticipated issuance of DG-1363 and the Operational Occurrences. The 97.5% probability is a different endorsement of ANSI/ISA 67.04.01-definition of 95% probability/95% confidence level from that already 2018.

being applied by the NRC.

The NRC staff made no changes to the Note also that basing the setpoint on the 97.5% probability criterion DG-1363 as a result of this comment.

instead of the 95% probability criterion could also decrease the margin between the setpoint and the normal operating limit (OL), and that would result in an undesirable increase in the spurious trip probability.

These calculations show that basing the LSP on the 97.5% probability criterion rather than the historical 95% probability criterion results in an insignificant increase in probability of tripping before the AL is reached, but could lead to a significant detrimental increase in spurious trip probability. Moreover, the licensed GEH safety analyses are based on LSPs that meet the 95% probability criterion, so no increase in trip probability is required from the safety point of view.

The 97.5% probability criterion is the consequence of using two-sided" statistics, whereas using "singlesided" statistics would correctly locate the setpoint such that it meets the historical 95%

probability requirement for not exceeding the AL. Note that the NRC's statistical handbook (Reference 4,NUREG-1475 Rev 1,"Applying Statistics") indicates that use of single-sided statistics is appropriate for the usual case where the variable approaches a safety related setpoint, or limit, in one direction from the safe side (see description of Critical Power Ratio in example 9.4 of Reference 4.

and see Section 9.13 of Reference 4 for a description of how to determine with high confidence the upper limit of the population 21

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) standard deviation from the standard deviation obtained from a limited size sample).

23. Anonymous The NRC staff is commended on clarifying the mathematically The NRC staff disagrees with the incorrect assumptions in DG-1141. However, Section B should comment. ANSI/ISA 67.04.01-2018 clarify whether the desired statistical objective for licenses and requires both (1) a statistical limit applicants is a tolerance interval rather than a confidence interval as between the analytical limits and the requirements for safety-related instrument setpoints. (See LSSS and (2) a tolerance interval with https://www.graphpad.com/support/faq/the-distinction-between- endpoints determined at a 95%

confidence-intervals-prediction-intervalsand-tolerance- confidence level to encompass 95%

intervals/#:~:text=If%20you%20set%20the%20first,wider%20than% probability distribution of interest.

20a%20prediction%20interval) During the development of DG-1363, the

). staff considered and addressed technical The incorrect assumptions center on Figure 2 in DG-1141. For issues and public comments related to example, characterization of the as-found trip setpoint in terms of an the issuance of DG-1141. In addition, as error band reflects a deterministic not a probabilistic engineering discussed in the August 14, 2020 expectation. Trip setpoint which is subject to the various random Federal Register notice, the NRC staff based uncertainties would be expected to be found within the total elected not to finalize DG-1141 as a loop uncertainty as defined in the ISA Standard Section 4.4. revision to RG 1.105 and chose instead Similarly, given the nonrandom terms in Equation 2 of the ISA to evaluate the ANSI/ISA 67.04.01-2018 Standard the assumption of a normal distribution is an incorrect one. standard for endorsement and issued Lastly, the notion that 21/2% probability can be assigned or allocated DG-1363 as a replacement for DG-1141.

to one side or another of a distribution runs contrary to the The relevant concerns associated with laws of probability. DG-1141 were adequately addressed by the ISA committee while preparing the On confidence intervals final version of 2018 standard and which (https://en.wikipedia.org/wiki/Confidence_interval) has been fully endorsed by the NRC in DG-1363.

A 95% confidence level does not mean that for a given realized interval there is a 95% probability that the population parameter lies The NRC staff made no changes to DG-within the interval (i.e., a 95% probability that the interval covers the 1363 as a result of this comment.

population parameter). According to the strict frequentist interpretation, once an interval is calculated, this interval either covers the parameter value or it does not; it is no longer a matter of 22

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) probability. The 95% probability relates to the reliability of the estimation procedure, not to a specific calculated interval. Neyman himself (the original proponent of confidence intervals) made this point in his original paper:

"It will be noticed that in the above description, the probability statements refer to the problems of estimation with which the statistician will be concerned in the future. In fact, I have repeatedly stated that the frequency of correct results will tend to . Consider now the case when a sample is already drawn, and the calculations have given [particular limits]. Can we say that in this particular case the probability of the true value [falling between these limits] is equal to

? The answer is obviously in the negative. The parameter is an unknown constant, and no probability statement concerning its value may be made..." Deborah Mayo expands on this further as follows:

"It must be stressed, however, that having seen the value [of the data],

NeymanPearson theory never permits one to conclude that the specific confidence interval formed covers the true value of 0 with either (1 )100% probability or (1 )100% degree of confidence.

Seidenfeld's remark seems rooted in a (not uncommon) desire for NeymanPearson confidence intervals to provide something which they cannot legitimately provide; namely, a measure of the degree of probability, belief, or support that an unknown parameter value lies in a specific interval. Following Savage (1962), the probability that a parameter lies in a specific interval may be referred to as a measure of final precision. While a measure of final precision may seem desirable, and while confidence levels are often (wrongly) interpreted as providing such a measure, no such interpretation is warranted.

Admittedly, such a misinterpretation is encouraged by the word

'confidence'."

A 95% confidence level does not mean that 95% of the sample data lie within the confidence interval.

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Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

A confidence interval is not a definitive range of plausible values for the sample parameter, though it may be understood as an estimate of plausible values for the population parameter.

A particular confidence level of 95% calculated from an experiment does not mean that there is a 95% probability of a sample parameter from a repeat of the experiment falling within this interval.

Bottom Line: If the NRC expects licensees or applicants to meet a statistical limit with respect to the Analytical Limits in nuclear power plant technical specifications then a tolerance interval should be required rather than a confidence interval.

24. Stephen J. B.1 ANSI/ISA S67.04-2018 should be ANSI/ISA S67.04.01-2018 The NRC staff agrees with this comment Vaughn and made the editorial change to Section B.1 of DG-1363.

B.2.1 First sentence. The term anticipated conditions is not commonly The NRC staff agrees in part with the used and causes confusion. The term accident conditions is a more comment. The term, anticipated appropriate term because any of the non-normal conditions are not conditions, is not the correct term to anticipated. use. However, the staff does not agree that the correct term is accident conditions. The staff changed anticipated conditions in Section B.2.1 of DG-1363 to design basis event conditions, which encompasses both anticipated operational occurrences and postulated accidents, as opposed to just accident conditions. Therefore, Section B.2.1 of DG-1363 was updated to change anticipated conditions to design basis event conditions.

B.2.2.1 Paragraph 3 reaffirms the NRC staffs approval of Option B of TSTF- Option A of Technical Specifications 493, Revision 4 but there is no mention throughout the DG-1363 of Task Force (TSTF) Traveler 493 is Option A of TSTF-493, Revision 4. discussed in 75 FR 26294, Notice of 24

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission)

Availability of Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-493, Revision 4, Clarify Application of Setpoint Methodology for LSSS Functions, dated May 11, 2010. The staff believes that Option A is not relevant to setpoint determination and is outside the scope of DG-1363. However, TSTF-493 Option B is intended to establish and maintain safety-related setpoints.

The NRC staff made no changes to DG-1363 as a result of this comment.

B.2.2.3 This section refers to drift as though is mentioned in the footnote The NRC staff disagrees with this under 4.6 of the standard; however the footnote in the standard only comment. The discussion in Section 4.6 mentions test acceptance criteria and TLU. of the standard adequately addresses consideration for drift. However, the footnote under Section 4.6 of the ANSI/ISA 67.04.01-2018 standard provides caution on appropriately calculating the drift value. The discussion in Section B.2.2.3 of DG-1363 and the footnote in Section 4.6 of the standard are both addressing the potential for masking any adverse performance.

Refer to the response to comment 17 on the footnote in Section 4.6 of the standard and the change made to DG-1363 indicated there. No additional 25

Commenter Section of Specific Comments NRC Resolution DG-1363 (These are the full comments as provided in each submission) changes were made to the DG-1363 as a result of this comment.

B 2.2.5 The term (Performance Acceptance Criteria) should be The NRC staff agrees with this comment (Performance Test Acceptance Criteria) and made the editorial change to Section B.2.2.5 of DG-1363.

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