ML20119A614

Response to Comments for RG 1.236 for ACRS

ML20119A614
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Issue date:	04/28/2020
From:	Office of Nuclear Regulatory Research
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ML20055F488	List: ML20055F489 ML20055F490 ML20055F496 ML20077M781 ML20084N202 ML20119A614 ML20275A227 ML20275A228 ML20275A229
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2016-0233 DG-1327
Download: ML20119A614 (22)

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1 Response to Second Round of Public Comments Draft Regulatory Guide (DG)-1327 (NRC Docket-2016-0233; ADAMS Accession No. ML20055F489)

Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents Proposed New Regulatory Guide On July 30, 2019, the NRC published a notice in the Federal Register (84 FR 36961) that Draft Regulatory Guide, DG-1327, a proposed new Regulatory Guide, was available for public comment. The public comment period ended on November 19, 2019 and comments were received from seven stakeholders representing a total of 54 individual comments. The comment submitters are listed below. To facilitate the identification and disposition of each comment received, the NRC staff compiled the six submissions. For example, the Global Nuclear Fuel (GNF) submission contained three comments, which were annotated by the staff as GNF-1 through GNF-3. Each submission is available in Agencywide Documents Access and Management System (ADAMS). The NRC staff responses are contained in the following table.

Comment Submissions

1. Kent Halac, Global Nuclear Fuel - Americas, LLC, Comments on Draft Regulatory Guide DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents, dated October 14, 2019 (ADAMS Accession No. ML19297G296). [GNF-1 through GNF-3]

2. David P. Helker, Exelon, Comments on Draft Regulatory Guide (RG) DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents (Federal Register 84FR36961, dated July 30, 2019, Docket ID NRC-2016-0233), dated October 28, 2019 (ADAMS Accession No. ML19304A376). [Exelon-1 through Exelon-3]

3. Korey L. Hosack, Westinghouse Electric Company, Transmittal of Westinghouse Electric Company Comments on draft regulatory guide DG-1327 [Docket ID NRC-2016-0233], dated October 31, 2019 (ADAMS Accession No. ML19318E698). [WEC-1 through WEC-5]

4. B.E. Standley, Dominion Energy Services, Inc., Comments on DG-1327, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents (Docket ID NRC-2016-0233)

(Federal Register Notice 84 FR 36961), dated November 12, 2019 (ADAMS Accession No. ML19344C063). [Dominion-1 through Dominion-2]

5. Gary Peters, Framatome Inc., Framatome Inc. Response to Request for Public Comment on the Draft Regulatory Guide DG-1327, 'Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accident (Federal Register Vol. 84, No. 146, 36961, dated July 30, 2019; Docket ID NRC-2016-0233), dated November 13, 2019 (ADAMS Accession No. ML19324E586). [Framatome-1 through Framatome-7]

6. Frances Pimentel, Nuclear Energy Institute, Industry Comments on Draft Regulatory Guide DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents (Federal Register 84FR49125, dated September 18, 2019 and 7590-01-P, dated September 18, 2019, Docket ID NRC-2016-0233), dated November 18, 2019 (ADAMS Accession No. ML20054B702). [NEI-1 through NEI-26]

7. Breaha, David, Chininau, CO. 12345, davidl@example.com. (Adams Accession No. ML19304A040).

The pie chart below illustrates the distribution of comments received during this second round (2019) of public comments on DG-1327.

2 3

Accept/

Commenter # Category Section Summation of Comment NRC Response Reject This section declares that analyses should consider the potential for wider operating conditions as the result of xenon oscillations or plant Analytical maneuvering. The phrase xenon oscillations is applicable only to PWRs The NRC staff agreed with this comment and deleted GNF 1 C.2.2.2.3 Accept Methods as BWRs do not experience spatial xenon oscillations. This phrase should the phrase xenon oscillations from the text.

be removed, or the DG should state that this is applicable only to PWRs.

This section states Credit for additional control blade banking within the bank position withdrawal sequence (BPWS) may be used to reduce the control blade reactivity worth during the event. The licensees reload analysis should fully reflect any additional control blade banking beyond the minimum required in the BPWS.

The cladding failure criteria in the DG are more limiting than those behind BPWS and some sequence variations are allowed beyond those specifically analyzed in the BPWS LTR.

Analytical The NRC staff agreed with this comment and GNF 2 C.2.2.2.4 Accept Methods incorporated the proposed language into 2.2.2.4.

Therefore, the DG should be more generic on this topic.

Credit for additional control blade banking, such as from within the banked position withdrawal sequence (BPWS) or another similar banking scheme may be used to reduce the control blade reactivity worth during the event. The licensees reload analysis should fully reflect the required bank positions that were assumed in the CRDA analysis any additional control blade banking beyond the minimum required in the BPWS.

Regarding Figure 4, the staff elected to replace the previous piecewise linear (PWL) relationship with a curve fit through the data. To facilitate the curve fitting process, it was necessary to treat the highest non-failure enthalpy/hydrogen content point (72 wppm, 150 cal/g) as a presumed failure point. This presumed failure point should serve as an anchor point for the curve fit. The current curve instead omits three other non-failure The NRC staff agreed with the comment and modified points.

the failure threshold lines on Figures 2 through 5 to Failure GNF 3 C.3.2 Accept encompass more of the non-failure data points and also Thresholds The primary response from a CRDA is often from the fresh fuel (i.e. lower adopted the proposed exponential function form. See exposure) with highly exposed fuel reacting less energetically. Thus, the Attachment 1 for further details.

purposed failure threshold is less accurate in the area of interest particularly between 55-to-100 wppm.

Please redraw the curve to encompass more of the actual non-failure data points. While this comment directly pertains to Figure 4, the same concept applies to the other fitted curves.

The last phrase can be misinterpreted to suggest that hydrogen enhances General/ PCMI. The impact of hydrogen is on the potential for cladding failure.

The NRC staff agreed with this comment and the Framatome 1 Editorial - Accept proposed language was adopted.

Background Proposal: "The prompt thermal expansion of the fuel pellet, which can be exacerbated at high burnups by gaseous fission product swelling, may

4 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject cause the fuel cladding to fail by PCMI. The potential for failure is enhanced by the presence of hydrogen in the cladding."

The NRC did not agree with this comment. While the The guidance limits the applicability of the RXA PCMI failure curves to 70 commenter provided technically reasonable arguments, wppm excess hydrogen for non-lined cladding. A limit of 70 wppm of the staff still considers the liner (and the absorbed Failure excess hydrogen creates undue burden for M5 cladding material. Based hydrogen within the liner) to have an impact on the initial Framatome 2a C.1.2.3 Reject Thresholds on the information presented in this comment, Framatome requests that condition of the cladding. Where the liner not present, this applicability limit be deleted. more of the hydrogen would likely reside toward the exterior of the cladding which would likely influence the failure enthalpy.

Alternative Proposal: If the proposal above is not considered an acceptable proposal, then revise Section 1.2.3 to provide adjusted PCMI Failure limits for non-liner RXA cladding. PCMI limit curves for non-liner RXA The NRC agreed with this comment. See Attachment 2 Framatome 2b C.1.2.3 Accept Thresholds cladding materials can be developed by adjusting the hydrogen content of for further details.

the liner RXA cladding tests to account for the influence of the liner on the hydrogen distribution in the bulk cladding material.

The statement in the guidance is vague and does not agreed with the earlier public comment resolution. Without specific guidance this item is subject to misinterpretation.

"Each applicant should address the possibility of hydride reorientation because of power maneuvering or reactor shutdown."

The NRC staff agreed with this comment. Reference to Analytical Framatome 3 C.2.3.5 Proposed revision: Accept NUREG-0800, SRP Chapter 4 was added. The revision Methods "Each applicant should address the possibility of hydride reorientation was omitted to avoid confusion with future revisions.

because of power maneuvering or reactor shutdown consistent with the requirements in NUREG 0800 Section 4.2 II. ACCEPTANCE CRITERIA, SRP Acceptance Criteria, 1.A. Fuel System Damage: vi (2), page 4.2-7, Revision 3, March 2007."

The RIA criteria for High-Temperature Cladding Failure Threshold are The NRC agreed with this comment and has modified derived from RIA testing of fuel pins undergoing prompt critical power the text. Framatomes technical bases for the proposed excursions. These power pulses have pulse widths < 0.05 seconds. The change seems logical. To confirm the proposed change initial power level does not define whether there is a power pulse that in the applicability of the high-temperature cladding needs to be evaluated against these criteria. The RAI criterion for High failure curve, the staff completed a series of calculations Temperature Cladding Failure Threshold in the DG 1327 Figure 1 is not with a simple point-kinetics model with simulated fuel applicable to conditions where there is no power pulse.

Failure temperature reactivity feedback (i.e., Doppler). The Framatome 4 C.3.1 Accept Thresholds figure below illustrates the impact of changes in Starting at extremely low powers with a non-prompt reactivity insertion reactivity insertion on the resulting power profile. As does not result in a power pulse. A non-prompt reactivity insertion results shown in the figure, a prompt power excursion resulting in a prompt jump proportional to the original flux with a subsequent period from a large positive reactivity insertion exhibits a large, of seconds rather than milliseconds and may not have a distinct pulse narrow pulse. Whereas, non-prompt power excursions shape. For example, a reactivity insertion of 0.95$ would cause the flux to exhibit a lesser, broader power profile.

jump by a factor of 20 with a Period of -1 sec. An initial power of 0.00001

% power would jump to 0.00020% power and increase with a period of -1

5 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject second. The power would slowly approach sensible heat and reach an asymptotic power as the feedback balances the reactivity insertion. Some overshoot in the neutron power may occur resulting in a very broad power response. The DNBR and fuel centerline melt (FCM) criteria are adequate to assess the integrity of the fuel for this type of event. This type of power excursion can be very similar to other HZP events that use the DNBR and FCM fuel failure criteria.

The NRC approved Framatome topical report ANP-10338P-A (December 2017) reflects this by applying this criteria only to prompt critical pulses.

DNBR is evaluated for non-prompt power pulses.

The following wording is proposed.

3.1 High-Temperature Cladding Failure Threshold Figure 1 shows the empirically based high-temperature cladding failure threshold. This composite failure threshold encompasses both brittle and ductile failure modes and should be applied for events with prompt critical excursions These confirmatory calculations support the proposed (i.e. the ejected rod worth or drop rod worth >1.0$). Because ductile changes.

failure depends on cladding temperature and differential pressure (i .e.,

rod internal pressure minus reactor pressure), the composite failure threshold is expressed in peak radial average fuel enthalpy (calories per gram (cal/g)) versus fuel cladding differential pressure (megapascals (MPa)). If the event reaches a significant power after the pulse where the heat flux and neutron power remain relatively constant, fuel cladding failure is presumed if local heat flux exceeds thermal design limits (e.g.,

departure from nucleate boiling and critical power ratios) . For non-prompt critical excursions, fuel cladding failure is presumed if local heat flux exceeds thermal design limits (e.g., departure from nucleate boiling and critical power ratios) .

Use of the term "centerline" creates potential confusion for application to annular fuel pellets. This word could be deleted with no change to the intended meaning.

Coolable The NRC staff agreed with this comment and the Framatome 5 C.6 Accept Geometry Proposal: "If fuel melting occurs, the peak fuel temperature in the outer 90 proposed language has been adopted.

percent of the fuel volume should remain below incipient fuel melting conditions."

Modify the Cesium FGR fraction to be based on Xenon rather than The radiological source term information in Appendix B Krypton to better reflect experimental data. This modification would was moved to RG 1.183. The NRC staff plans to Framatome 6 Source Terms App. B change Table B-1 and footnote 81 on p. B-1, Section B.2 on p. B-3, N/A consider this comment while performing an update to Section B-1 .2 on pp. B-4 and B-5, and the example calculation on p. B-9. RG 1.183.

The radiological source term information in Appendix B was moved to RG 1.183. The NRC staff plans to Framatome 7 Source Terms App. B Provide definition of terms from Reference 1 for clarity and completeness. N/A consider this comment while performing an update to RG 1.183.

6 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject It is suggested that Section 1.1.1 be modified as The applicability of this The NRC staff did not agree with this comment. The guidance to LWR fuel rod designs, different from the UO2 fuel rod designs empirical database used to establish the guidance and described in Section 1.1 (e.g., doped pellets, changes in fuel pellet limits included in this regulatory guide is limited to the microstructure or density, changes in zirconium alloy cladding applicability described in Section 1 as written.

microstructure or composition, coated zirconium alloy cladding), will be addressed on a case-by-case basis.

Westinghouse 1 Applicability C.1.1.1 Reject As stated, the guidance is applicable to currently approved designs (e.g., Optimized ZIRLOTM). Applicants The doped pellets and changes in zirconium alloy have been used already requesting approval for future fuel designs (e.g., above in light water reactor (LWR) fuel designs. Some of those fuel designs have 5.0% 235U enrichment) would need to demonstrate the been previously addressed and approved as similar to the UO2 fuel rod continued applicability of the guidance or provide an design, such as the Optimized ZIRLO' fuel cladding.

alternative.

Westinghouse Optimized ZIRLO high performance cladding, which has a partial recrystallized anneal (pRXA) final heat treatment, currently represents the majority of reload pressurized water reactor (PWR) fuel in the United States (US). Therefore, this significant population of US PWR fuel does not have defined cladding failure criteria under the draft guide.

Westinghouse has supporting documentation, which can be readily provided or audited upon request, that demonstrates that the trend in The NRC staff agreed with this comment. On February cladding ductility as a function of hydrogen, hydrides distribution after 13, 2020, the staff conducted an audit of the irradiation, and hydride reorientation behavior under stress are similar for Westinghouse documentation supporting the application ZIRLO cladding. This data supports a position that Optimized ZIRLO of the SRA PCMI cladding failure threshold to Optimized 2 Applicability C.1.2.2 Accept cladding should have similar reactivity initiated accident (RIA) cladding ZIRLO. Based on the results of the audit (ADAMS failure limits as ZIRLO cladding. A significant part of this data has been Accession No. ML20049F944), the staff accepts the use previously submitted as part of the PAD5 licensing and in public domain of the SRA PCMI cladding failure thresholds for papers. Optimized ZIRLO cladding.

Westinghouse requests that the proposed pellet-clad mechanical interaction (PCMI) cladding failure criteria for SRA ZIRLO cladding be applicable to pRXA Optimized ZIRLO cladding. If the NRC needs time to review the Optimized ZIRLO cladding data, then the Reg Guide should be delayed until that review is complete to ensure that the final guidance is applicable to the fuel products currently in many US PWRs.

The NRC staff partially agreed with the comment. The guidance in Section 3.1 defines a breakpoint between the use of the high-temperature failure curve (Figure 1) and design thermal limits (e.g., DNB). The breakpoint was set to 5% power. There is no breakpoint related to Please clarify the technical basis of the 5% power level for the switch PCMI.

Failure Westinghouse 3a C.3.1 between the PCMI failure criterion and the high temperature cladding Partially Accept Thresholds failure criterion. Framatome comment #4 requested a change to application of the high-temperature failure curve. The requested change is similar to the Westinghouse comment in that it involved the 5% power breakpoint.

The Framatome comment was accepted.

7 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject The NRC staff did not agree with this comment. PCMI is a separate failure mechanism from boiling crisis induced, high cladding temperature failure. Thus, PCMI should have its own analytical failure threshold.

The PCMI failure threshold curves in Figures 2 through 5 are based on prompt critical power excursions. Hence, these curves are directly applicable to RIA scenarios which experience a prompt critical power excursion. The application of these failure curves has been judged to be conservative to non-prompt power excursions.

PCMI should be used as the only failure criteria when an ejected rod Failure Westinghouse 3b C.3.1 worth is equal to or greater than $1. DNB should be used as the only Reject There are no approved fuel rod thermal-mechanical Thresholds failure criteria when ejected rod worth is less than $1. models to predict the fuels behavior under RIA conditions. Also, there are no approved analytical limits (e.g., cladding strain) to define acceptable performance under RIA conditions. Lacking an alternative PCMI failure limit, the NRC accepts the use of Figures 2 through 5 for all RIA scenarios.

The guidance states alternative fuel rod cladding failure criteria may be used if they are adequately justified by analytical methods and supported by sufficient experimental data.

The NRC staff did not agree with this comment. PCMI (i.e., strain driven) and cladding temperature are separate failure mechanisms and each deserves a Based on earlier studies documented in NUREG/CR-0269, it can be separate failure threshold.

concluded that in the range from 120 cal/g to 240 cal/g energy deposition, DNB occurs but the cladding damage is not sufficient to result in cladding Framatome comment #4 was accepted by the staff and failure in a short transient period. includes changes related to the application of DNB failure to prompt and non-prompt scenarios. This change For prompt critical events, the proposed DG-1327 limits for PCMI addresses a portion of the Westinghouse comment.

conservatively envelope time at temperature failure from DNB. For prompt critical cases that do not fail due to PCMI and remain at a significant During the transient, the time in boiling crisis depends on Failure Westinghouse 4 C.3.1 power level (such as cases that do not trip), the DNB criterion will be Reject several variables including the rate and magnitude of Thresholds applied following the prompt critical power excursion. deposited energy and local thermal-hydraulic conditions.

A firm technical basis for a time-at-temperature criteria Westinghouse proposes for non-prompt critical RIA at part power does not exist.

operation, specific time at temperature cladding limits can conservatively be applied to calculated cladding time and temperature in DNB and The guidance states alternative fuel rod cladding failure separate failure from non-failure. criteria may be used if they are adequately justified by analytical methods and supported by sufficient experimental data. The NRC would consider alternate criteria, including a time-at-temperature failure threshold, on a case-by-case basis.

8 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject In the first paragraph of page B-1 in the "Steady-State Fission Product Gap Inventory" section it states that, "The gap fractions from Table B-1 are used in conjunction with the calculated fission product inventory The radiological source term information in Appendix B calculated with the maximum core radial peaking factor."

was moved to RG 1.183. The NRC staff plans to Westinghouse 5a Source Terms App B N/A consider this comment while performing an update to This is a simple but conservative assumption. The guidance requiring the RG 1.183.

use of the maximum radial power factor should be revised to allow an alternative (more realistic but still conservative) calculation using actual power history.

Page B-1, Paragraph 2 states that, "For fuel that melts, the combined fission product inventory (steady-state gap plus transient release) is added to the release resulting from fuel melting. RG 1.183 (Ref. B-1) and 1.195 (Ref. B-2) provide additional guidance on fuel melt source term." The radiological source term information in Appendix B was moved to RG 1.183. The NRC staff plans to Westinghouse 5b Source Terms App B Both cited regulatory guides have Appendix H for rod ejection defining N/A consider this comment while performing an update to 100% noble gas and 25% for iodine for containment leakage and 50% for RG 1.183.

iodine for secondary releases. But there is no guidance there relating to melt for Alkali metals. The NRC should provide guidance for alkali metals fission product inventories.

The NRC staff agreed to this comment. The transient fission gas release correlations, as well as the entirety of the guidance, was reviewed to determine whether its range of applicability extended to the industrys 68 The gas release calculation only supports fuel rod average burnup of 65 GWd/MTU target burnup. The staffs assessment GWd/MTU (third paragraph on page B-1, and Figure B-1 on page B-2). (ADAMS ML20090A308) concluded that the guidance in RG 1.236 was applicable up to a fuel burnup of 68 This should be extended to support the industry efforts to go to higher GWd/MTU (rod average), provided the cladding did not allowable burnups. It is recommended that the figure support burnup to at exhibit localized imperfections (e.g., spallation, hydride Westinghouse 5c Source Terms App B least 68 GWd/MTU which industry seeks to achieve in the near-term. accept blisters) due to excessive oxidation. The staff also However, it would be ideal for the figure to support burnup to 75 identified that the guidance was not applicable to fuel GWd/MTU which the industry seeks to achieve in the next 5 to 7 years. rods with pre-existing cladding failure (i.e., leaking, The new ANS5.4 standard is based on database above rod average waterlogged). In addition, the staff identified data gaps to burnup of 70 GWd/MTU [9-page 4.10]. support 75 GWd/MTU (rod average). Radiological source term information was moved to RG 1.183. The NRC staff plans to consider this comment while performing an update of RG 1.183.

Item B-5 on page B-7 states that, "Rod power histories used in the fuel rod design analysis based on core operating limits report thermal-mechanical The radiological source term information in Appendix B operating limits or radial falloff curves should be used."

was moved to RG 1.183. The NRC staff plans to Westinghouse 5d Source Terms App B N/A consider this comment while performing an update to This text should be deleted as it is just one way to bound anticipated RG 1.183.

operation. The rest of the texts provides sufficient guidance for using conservative rod power histories.

9 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject On Page B-7 it states, This example illustrates the potential improvement in the radiological source term from calculating less bounding gap fractions. For this example, the licensee elects to calculate gap inventories The radiological source term information in Appendix B based on. However, the text does not pick up on the next page. was moved to RG 1.183. The NRC staff plans to Westinghouse 5e Source Terms App B N/A consider this comment while performing an update to There are issues with the break between page B-7 and page B-8. It RG 1.183.

seems that some text may be missing from page B-8. Please correct the page break.

The table on Page B-10 has all the gap fractions set to the maximum The radiological source term information in Appendix B value.

was moved to RG 1.183. The NRC staff plans to Westinghouse 5f Source Terms App B N/A consider this comment while performing an update to This is too conservative for short life isotopes. Burnup dependent gap RG 1.183.

fractions (similar to power fall-off) are more appropriate.

Dominion Energy also endorses the Framatome and Westinghouse comment on High-Temperature Cladding Failure Threshold as to when cladding failure due to the local heat flux exceeding thermal design limits Failure Dominion 1 C.3.1 (e.g., departure from nucleate boiling and critical power ratios) should be Accept See response to Westinghouse 3a and Framatome 4.

Thresholds evaluated and would appreciate the NRCs consideration of those comments.

The NRC staff agreed with several commenters that the fission product release fraction guidance contained in DG-1327 Appendix B should reside in RG 1.183 and RG 1.195. The NRC staff indicated their reason for not updating RG 1.183 and RG 1.195 to incorporate the fission product release fraction guidance was that it would take additional calendar time which would further delay fully implementing revised guidance for CRD and CRE. This justification for maintaining the fission product release guidance in DG-1327 is not sufficient justification for a change that will result in unclear regulatory requirements due to different fission product The NRC staff agreed with this comment and plans to Dominion 2 Source Terms App B release fraction guidance existing in the different regulatory guides. In Accept move Appendix B to RG 1.183.

2004, the NRC issued Research Information Letter (RIL) 0401 that compiled available reactivity-initiated accidents (RIA) test results and completed a safety assessment of currently operating reactors. RIL 0401 concluded there was no concern related to protecting the health and safety of the public for the operating reactors due to RIA. Therefore, it seems reasonable and warranted to take the additional time now to properly update RG 1.183 and RG 1.195 to ensure clear, transparent and consistent regulatory guidance is presented to the public.

Section 2.2.1.4 of the draft RG discusses that the maximum uncontrolled The NRC staff agreed with the comment and modified worth of an ejected rod should be calculated based on fully or partially Analytical the text. Applicants do not need to consider dropped or Exelon 1 C.2.2.1.4 inserted misaligned or inoperable rod or rods if allowed. When referring to Accept Methods misaligned rods which are being recovered within TS the phrase " .. .fully or partially inserted misaligned or inoperable rod or LCO completion times.

rods if allowed"

10 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject Exelon is requesting further clarification regarding whether this is limited to rods in which the safety analysis has been performed to justify its non-normal position, or whether it also includes rods that are dropped or misaligned but are being recovered within the associated Technical Specifications (TS) Limiting Conditions for Operation (LCO) Completion Times (CTs).

Section 2.2.2.4 of the draft RG discusses that the maximum uncontrolled worth for a dropped blade should be calculated based on the following conditions: (1) the range of control blade positions allowed at a given The NRC staff agreed with the comment and modified power level, (2) additional fully or partially inserted misaligned or the text. Applicants do not need to consider uncontrolled inoperable blades if allowed, and (3) any out-of-sequence control blades withdrawal (as the initiating event) of an inoperable Analytical that may be inserted for fuel leaker power suppression. When referring to Exelon 2 C.2.2.2.4 Accept blade that has been locked in place and cannot Methods the phrase " .. .fully or partially inserted misaligned or inoperable rod or physically move. However, the impact of that inoperable rods if allowed .. . ," Exelon is requesting further clarification whether this blade on the worth of other blades needs to be is also intended to include inoperable blades that have been locked in considered.

place and cannot physically move or be dropped in accordance with the associated TS.

The NRC recently published a Sandia National Laboratories (SNL) technical document entitled, "Release Fractions in Non-LOCA Accidents in Draft Regulatory Guide 1. 183 DG-1199, "dated April 10, 2019 (ML19094A336). This SNL technical document is dated after the previous revision of DG-1327 was issued (i.e., issued in November 2016) that updated Appendix B, "Fission Product Release Fractions." The proposed Appendix B in DG-1327 states: " .. . The fission product release guidance contained in Appendix B for CRE and CRD accidents should be used instead of the gap fractions provided in RG 1. 183, Revision 0, for a CRE and CRD accident until RG 1. 183 is updated. " There appears to be an The NRC staff agreed with this comment and has extensive overlap between the two documents. If DG-1327 steady state removed Appendix B from this guidance document. The Exelon 3 Source Terms App. B and transient gap releases supersede the NRC's previous positions, Accept staff plans to incorporate the information into Regulatory Exelon recommends that DG-1327, Appendix B, should acknowledge this Guide 1.183 to prevent confusion.

fact in an effort to prevent any misunderstanding of the NRC's expectations regarding the gap release assumptions that are acceptable under use of Alternative Source Term (AST). Exelon further recommends that DG-1327 should consider listing the many previous gap release technical basis documents, including ML19094A336, and state that they are superseded by Appendix B of DG-1327, if applicable. This would allow the radiological safety analysis practitioner to have a clear understanding with respect to acceptable NRC guidance related to steady-state and transient gap release fractions.

Characterization of public comments in the Background section implies General/ the public comments were made on the NRC memorandum supporting The NRC staff agreed with this comment and has NEI 1 B Accept Editorial the technical and regulatory basis which is not appropriate. The public clarified the text.

comments were provided on the initial DG- 1327(Reference 4). Please

11 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject remove text indicating it was amended by public comments as shown below.

This memorandum documents the empirical database, as well as the technical and regulatory bases for this guide.

The NRC had provided Revised RG Text in the response to comments from the first public comment period that was not incorporated in to the DG posted for the second public comment period. The Revised RG Text provided in the NRC response read as:

The analytical limits and guidance described are not applicable to General/ anticipated operational occurrences (AOOs) and other postulated The NRC staff agreed with this comment and the NEI 2a C.1 Accept Editorial accidents involving positive reactivity insertion (e.g., PWR excess load, proposed language has been adopted.

PWR inadvertent bank withdrawal, PWR steam line rupture, BWR turbine trip without bypass, BWR rod withdrawal error).

Please incorporate the revised NRC RG Text as indicated above into Section C.1, page 7, paragraph 1.

The radiological source term information in Appendix B In Appendix B, replace all instances of the term Non-LOCA with RIA, was moved to RG 1.183. The NRC staff plans to General/

NEI 2b C.1 as events other than RIA are not germane to the Regulatory Guide. Accept consider this comment while performing an update to Editorial RG 1.183.

It is requested the NRC indicate RG 1.203 does not need to be applied when the guidance of DG-1327 is employed for the evaluation of postulated CRE and CRD accidents, regardless of existing Vendor models/methods. The NRC staff partially agreed with this comment and Related the text has been modified. Note there are portions of NEI 3 C.2.1.1 Partially Accept Guidance Add the following to the end of section C.2.1.1 RG 1.203 (e.g., QA, documentation) which are not addressed within this guidance.

Note, if the guidance provided in this section is employed for the evaluation of postulated CRE and CRD accidents, the staff recognizes that RG 1.203 does not need to be applied.

For consistency with NRC memorandum supporting the technical and regulatory basis for RIA acceptance criteria and guidance, it is requested the references to zero power in Items C.2.2.1.2 for PWRs and C.2.2.2.2 for BWRs be updated to include hot zero power for PWRs and cold zero General/ C.2.2.1.2 The NRC staff agreed with this comment and the text NEI 4 power for BWRs. Accept Editorial C.2.2.2.2 has been modified.

For example: Accident analyses at zero power should encompass both (1)

BOC following core reload hot zero power for PWRs and cold zero power for BWRs and (2) restart following recent power operation.

General/ The NRC staff agreed with this comment and the NEI 5 C.2.2.1.5 Section C.2.2.4 should be Section C.2.2.1.4 Accept Editorial proposed language has been adopted.

12 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject C.2.2.1.10 Removing terms coefficients and coefficient of with a more generic General/ C.2.2.1.11 The NRC staff agreed with this comment and the NEI 6 term such as reactivity feedback, as there are multiple ways to simulate Accept Editorial C.2.2.2.10 proposed language has been adopted.

the reactivity mechanisms within an analysis.

C.2.2.2.11 The segmenting of the axial length uses the word several. It is expected General/ C.2.3.3 that the number of axial nodes would be much larger than several. The NRC staff agreed with this comment and the NEI 7 Accept Editorial C.2.4 Replace several with selected. proposed language has been adopted.

Section C.2.3.4 states than an NRC-approved hydrogen uptake model should be used. The hydrogen uptake model in Appendix C is designated The NRC staff agreed with the sentiment of this General/ as acceptable. The concern is that a vendor/licensee submittal of the comment. The language in section C.2.3.4 has been NEI 8 C.2.3.4 Accept Editorial Appendix C hydrogen uptake model would be subject to additional NRC modified to add (or the appropriate model from review. In Appendix C replace acceptable with NRC-approved. Appendix C of this guidance).

The NRC staff agreed with this comment. The pertinent sections of the staffs response to AREVA-17 (1st round of public comments) are summarized below.

The staff added Item C.2.3.7 in response to comment AREVA-17 from the In accordance with GDC-14, the CRDM shall be first public comment period. The comment requested clarification on the designed, fabricated, erected, and tested so as to have treatment of the potential pressure reduction caused by the assumed an extremely low probability of abnormal leakage, of failure of the control rod pressure housing for criterion other than RCS rapidly propagating failure, and of gross rupture.

peak pressure.

Regardless of the extremely low probability of gross The NRC agreed with the comment and indicated the NRC staff believes failure, the staff has maintained that a non-mechanistic, the original CRE design basis should be preserved, and plants existing postulated CRE accident be included in the plants license basis should be maintained (i.e., consideration of high worth rod design and license basis. In general, the existing PWR ejections). fleet analyzes a non-mechanistic, postulated CRE over a Additionally, comment GE-11 on the same section as comment AREVA- short time duration to demonstrate compliance to GDC-17 to which the NRC agreed, identified this item as only being applicable 28 and limits on radiological consequences.

to PWRs.

Analytical NEI 9 C.2.3.7 Accept Specific modeling of a small break in the reactor vessel Methods Item C.2.3.7 as currently written implies the need to perform additional upper head (i.e. CRDM failure) will promote a gradual analyses of the control rod housing which are beyond the scope of the loss of inventory and depressurization of the RCS. For DG. Specifically, the NRC cited NUREG-0800, Section 3.9.4 and the the simulation of a high worth CRE scenario with its requirements of GDC 14 as the basis for the additional requirements in rapid (prompt critical, or close to prompt critical) power the response to comment AREVA-17. excursion, the accident progression and consequences would not be significantly impacted by this small break in It is requested the NRC replace Item C.2.3.7 with the suggested text the RCS upper head. The power excursion, inherent below and relocate it to Section C.2.2.1, such that there is no confusion doppler feedback, and response of safety-related SSCs with BWRs. (e.g., high power reactor trip, scram, pressure relief values) occur in a short time frame.

Fuel failure predictions do not need to consider any reactor coolant system depressurization resulting from the assumed failure of the control For the simulation of a low worth CRE scenario, initial rod pressure housing. conditions could be orchestrated to produce a slowly evolving transient which delays or even avoids a timely reactor trip. For this scenario, the gradual loss of inventory and depressurization become more important.

RCS depressurization would promote DNB degradation,

13 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject potentially leading to cladding failures and cladding burst. This may potentially increase the predicted number of failed rods which is input to the dose calculation.

The CRE event originates from concerns associated with rapid, or prompt, power excursions which could significantly disturb the fuel bundle array, pulverize or melt fuel pellets, disperse fuel particles, promote a rapid generation of steam, and challenge the integrity of the reactor vessel and its internals and the ability to cool the core. The CRE serves as a design basis accident for the reactor pressure boundary, safety-related pressure relief functions, RPS trip functions (i.e. excore detector high-flux, high pressurizer pressure), control rod design and insertion limits, and fuel design and core loading pattern.

Evolving the CRE design basis to explicitly analyze a long-term scenario involving a benign power excursion with RCS depressurization would de-emphasize the original basis. Safety-related SCCs never before associated with CRE now become important. RPS trip functions (e.g., low pressurizer pressure, Thot saturation) and emergency core cooling system (ECCS) actuations would be relied upon to mitigate the severity of the accident. The overall accident progression and potential response of control room operators would mimic a SBLOCA.

10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, defines ECCS performance requirements under loss-of-coolant accident conditions. This regulation does not require ECCS performance demonstration for breaks in the reactor vessel. Hence, ECCS is not designed nor its performance judged on the ability to mitigate a CRDM housing failure induced loss-of-coolant scenario.

Portions of the ECCS have be credited to mitigate non-loss-of-coolant scenarios (e.g., main steam line break).

However, expanding the loss-of-coolant spectrum of breaks to include those within the reactor vessel is beyond the requirements in § 50.46.

From a public safety perspective, the long-term CRE scenario involving a benign power excursion with RCS depressurization is not limiting. Reactor vessel integrity and coolable geometry would not be challenged by a benign power excursion. Furthermore, a breach in the reactor vessel would provide a radiological pathway similar to LOCA (i.e., RCS activity released into large

14 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject containment structure with small leakage to atmosphere). Any additional fuel rod cladding failures associated with the RCS depressurization would be insignificant with respect to the LOCA source term (a.k.a. Maximum Hypothetical Accident) which assumes significant core wide damage and fuel melt. Hence, this type of long-term scenario is bounded by SBLOCA.

Based on the extremely low probability and the potential consequences described above, the NRC staff believes that the original CRE design basis should be preserved and plants existing license basis should be maintained.

Plants with CRDM housings and reactor head penetrations designed to GDC 14 requirements should evaluate a non-mechanistic CRE scenario to demonstrate compliance with GDC 28 and applicable on-site and off-site dose limits. Initial conditions and assumptions should be selected to maximize the challenge to these requirements. A mechanistic long-term scenario involving a relatively benign power excursion and RCS depressurization is not required.

Guidance associated with radiological release paths in RG 1.183 and RG 1.195 continue to apply.

In the context of the proposed Section C.2.4 wording, to what extent will realistic rod power histories be allowed in the context of AST? It makes no physical sense to say all bundles are at 54 MWd/MTU exposure, and all The radiological source term information in Appendix B the rods in the bundle are at 62 GWd/MTU. If an approved CRE/CRDA was moved to RG 1.183. The NRC staff plans to Analytical NEI 10 C.2.4 method is applied on a cycle-specific basis, is it acceptable to use cycle N/A consider this comment while performing an update to Methods specific rod source terms as cycle specific rod worths are already used? RG 1.183.

Please clarify the expectations between DG-1327 and RG1.183.

The addition of the words Conservative and bounding to the allowance to propose alternate fuel failure criterion creates confusion and is not consistent with the move towards more performance-based requirements.

It is recommended that the NRC use the wording from the response to comments from the first public comment period (AREVA-18) without any Analytical additional changes. The revised text from the response to comments from The NRC staff agreed with this comment and the NEI 11 C.3 Accept Methods the first public comment period is shown below: proposed language has been adopted.

Alternative fuel rod cladding failure criteria may be used if they are adequately justified by analytical methods and supported by sufficient experimental data. Alternative cladding failure criteria will be addressed on a case-by-case basis.

Failure Key test data that defines the proposed limits were generated under The staff did not agree with the comment. Both NEI 12 C.3.2 Reject Thresholds conditions far from prototypical of a commercial reactor rod ejection/rod temperature and pulse width effects were considered in

15 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject drop design basis accident. The atypical test conditions, from which the the technical bases of the PCMI cladding failure curves.

NRC proposed limits are based, produces results not representative of Additional RIA tests are planned to further investigate commercial LWR. these effects. This future data may help to improve these curves.

The guidance allows for alternate failure criteria and will be assessed on a case-by-case basis.

Numerous publications, including NRC sponsored research, show a brittle-to-ductile recovery temperature of less than 150°C [1-6] for cladding with radial hydride components. In a 2012 NRC sponsored research report, the brittle-to-ductile transition was determined to be influenced by the applied stressed used to re-orient hydride. In this report, a brittle-to-ductile transition temperature of 125°C was reported for an applied hydride re-orientation stress of 110 MPa for ZIRLO and Zircaloy-4 with The NRC disagreed with this comment. This guidance high hydrogen concentration. A ductile-to-brittle transition temperature of does not define a minimal measure of cladding ductility less 100°C was later presented by the same author in 2013 for M5 at such as a DBTT. But instead, addresses the changing lower hydrogen concentration. The reported transition temperature is degrees of ductility necessary to avoid cladding failure consistent with 100°C determined under RIA conditions in reference, for as a function of increasing fuel enthalpy (and associated pulse width greater than 10 ms. In the past a NSRR RIA test was pellet thermal expansion). Since zirconium hydrides conducted at 85C but did not show noticeable improvement in energy have a dominant effect on cladding ductility, the cladding absorption capacity. Test data from reference would indicate at the 4-5 ms Failure failure threshold is provided as a function of excess NEI 13 C.3.2 pulse width the brittle-to-ductile transition temperature is higher than Reject Thresholds hydrogen. The NRCs investigation found that the impact 100°C. The brittle-to-ductile transition temperature is well demonstrated in of initial cladding temperature on PCMI failure threshold the LS-series of tests conduct at the JAEA NSRR. Fuel from the same was only 18 cal/g between cold (room temperature) parent rod was tested at room temperature and 280°C. The test testing and hot (above 500°F) testing. The NRC would conducted at room temperature, LS-1, failed at an energy deposition of 53 consider, on a case-by-case basis, further scaling cal/g while LS-1, conducted at 280°C, survived a maximum energy between 500°F and a lower temperature (corresponding deposition of 89 cal/g.

to plant-specific BWR startup conditions).

The brittle-to-ductile transition temperature of ~100C is too low for significant hydride dissolution and ductility recovery is through other mechanism. The brittle-to-ductile transition behavior is a well-documented phenomenon. The RIA simulation tests merely provide a method to load the cladding. Test results under RIA loading conditions have been produced and verifies test data at other conditions.

The NRC staff agreed with the comment and has modified the failure threshold lines on Figures 2 through Failure 5 to encompass more of the non-failure data points and NEI 14 C.3.2 Same comment as GNF3 (above). Accept Thresholds adopted the proposed exponential function form. See Attachment 1 for further details.

The NRC staff did not agree with this comment. Current guidance related to radiological consequences is Some licensees use 10 CFR 100 radiological consequences acceptance provided in the cited RGs which provide an acceptable NEI 15 Source Terms C.4 criteria Revise Section 4 to include reference to 10 CFR 100 along with Reject method to satisfy applicable regulations (e.g., 10 CFR RG-1.183 or RG-1.195. 50.67). The fact that some licensees may have 10 CFR 100 acceptance criteria within their licensing basis is not relevant to future applicants using this guidance.

16 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject Does the content of this DG present a safety concern related to protecting the health and safety of the public for the operating reactors? The NRC staff did not agree with this comment. RIL-0401 was based on a limited, realistic assessment of The NRC staff initially performed an assessment of postulated reactivity- PWR control rod worths. The assessment concluded initiated accidents for operating reactors in the US in Research that legacy methods (e.g. point kinetics, 1D) were Information Letter 0401, dated March 31, 2004. The 2004 assessment sufficiently conservative to compensate for the new concluded that there was no concern related to protecting the health and research findings. However, like any safety assessment, safety of the public for the operating reactors. The NRC has issued two its a snap-shot in time. Fuel designs, materials, NEI 16 Implementation D memorandums (dated January 17, 2007 and March 16, 2015) on the Reject analytical methods, and fuel utilization are not stagnant.

proposed technical and regulatory basis for reactivity-initiated accident For the past 16 years, new fuel assembly designs (e.g.,

acceptance criteria since the 2004 assessment. The two memorandums GNF2, GNF3, Atrium 10XM, Atrium 11, SVEA Optima2, continued to reference the 2004 safety assessment. Given the conclusion GAIA), cladding materials (e.g., Optimized ZIRLO), and of the 2004 assessment and the continued reliance upon it, it is believed analytical methods (e.g., 3D realistic) have been that NRC staff does not have a safety concern related to protecting the implemented. Conclusions from RIL-0401 may no longer Health and Safety of the public for the operating reactors based on the be valid. This new guidance provides an acceptable path issuance of the guidance contained in DG-1327. to support all of these new technology improvements.

Include the staff requirements regarding forward fitting as defined in Management Directive 8.4 in the Use by NRC Staff section.

The industry is concerned that the extensive RIA guidance in the DG will The NRC staff partially agreed with this comment and be used in the future by the NRC staff for license amendment requests NEI 17 Implementation D Accept the text has been modified to reflect the intention on use that do not specifically involve RIA-related plant changes. The types of of the RG.

LARs that do involve RIA and DG-1327 evaluations have been identified by the staff in the NRC response to the first round of DG- 1327 comments, (p. 45 Item e).

Since RG 1.183 is not consistent with current codes and the consensus of The radiological source term information in Appendix B fission gas gap fraction calculations, a technical basis document would be was moved to RG 1.183. The NRC staff plans to NEI 18 Source Terms App. B beneficial. Please revise PNNL- 18212 to use the FAST code per N/A consider this comment while performing an update to ML19154A226. RG 1.183.

Table B-1 presents recommended steady state gap fractions documented in ML19154A226 for I-131 and other Halogens of 0.08 and 0.05, respectively. ML19154A226 reports the results of bounding FAST The radiological source term information in Appendix B calculations for steady state non- LOCA gap fractions. Based on a review was moved to RG 1.183. The NRC staff plans to NEI 19 Source Terms App. B of the reported results in ML19154A226, and using conventional rounding N/A consider this comment while performing an update to techniques, appropriate gap fractions for I-131 and other Halogens would RG 1.183.

be 0.05 and 0.03, respectively. Update the gap fractions to reflect the results of ML19154A226 using conventional rounding techniques.

Page B-1 Paragraph 1 last sentence: It is confusing to refer to Appendix B within Appendix B. Please replace Appendix B with this appendix. The radiological source term information in Appendix B Page B-2 Paragraph 1 last sentence: The sentence uses the phrase was moved to RG 1.183. The NRC staff plans to NEI 20 Source Terms App. B described in the attachment. Please replace in the attachment with N/A consider this comment while performing an update to within this appendix. RG 1.183.

Page B-4 Paragraph 3 last sentence: Please make the following changes:

17 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject While calibrated and validated against a large empirical database, FAST and its predecessors are not NRC-approved codes and may not be utilized to calculate that plant-specific, fuel-specific, or cycle-specific gap inventories that are in accordance with the acceptable analytical procedure below without further justification.

Page B-7: Start the sample calculation on a new page.

Page B-8: Is this page intentionally blank?

Page B-11: Earlier in Appendix B a footnote was designated B1 on page

1. Yet, the footnotes on page B-11 are designated 1 and 2. Please adopt a consistent standard.

The industry is concerned the guidance in the final RG-1327 Appendix B may be subsequently changed by the NRC staff with the ongoing update to RG 1.183 and a subsequent deletion of Appendix B at a future point from DG-1327.

If that were to occur then an Appendix B-based methodology submitted by a vendor/licensee and approved by the NRC may not be consistent with the updated RG 1.183.

The radiological source term information in Appendix B The industry requests the update to RG 1.183 and the deletion of was moved to RG 1.183. The NRC staff plans to NEI 21 Source Terms App. B Appendix B be an administrative change only, and that no technical N/A consider this comment while performing an update to changes are included. RG 1.183.

The industry is also concerned that there is no indication a DG-1327 Appendix B Dose assessment is sufficient to demonstrate compliance to RG 1.183 which effectively requires use of source term values at the highest exposure limits while pin failure is being effectively tied to much lower exposures via the non-linear hydrogen uptake phenomenon.

The industry needs assurance that only ONE dose assessment is required to meet both RG 1.183, and future DG-1327 requirements.

The radiological source term information in Appendix B Please clarify the exposures discussed in the figure are pellet exposure, was moved to RG 1.183. The NRC staff plans to NEI 22 Source Terms App. B not rod exposure. Clearly identify exposure basis and application. N/A consider this comment while performing an update to RG 1.183.

The NRC staff did not agree with this comment. The The conservative or bounding terminology are relative terms. So, what terms conservative or bounding do not appear in are they relative too? Specifically, Section C.2.3 is entitled Predicting the Section C.2.3.

total number of fuel rod failures. Is the conservative or bounding terminology supposed to be with respect to the number of rods failed, or is With respect to calculating the number of failed rods, the it really supposed to be with respect to dose consequence?

Analytical word conservative only appears once (Section C.3):

NEI 23 All Reject Methods When the failure criteria for a fuel rod was a constant with respect to To ensure a conservative assessment of onsite exposure, a failed number of rods could be thought of as a surrogate for and offsite radiological consequences, each of dose, and dose could be a surrogate for failed rods. The new non-linear these failure modes should be quantified, and the failure criteria breaks that line of reasoning. It is possible to envision sum total number of failed fuel rods should not scenarios with higher dose consequence with fewer rod failures and not be underestimated.

18 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject just from the rod eject / rod drop perspective, but from all non-LOCA events. Its use in this instance is judged appropriate.

Please clarify the basis for DG-1327, and explicitly express what the With respect to calculating radiological consequences, appropriate metric is for assessing terminology such as conservative or the word bounding appears in many places within bounding. Appendix B. The radiological source term information in Appendix B was moved to RG 1.183. The NRC staff This issue is important with respect to how RG 1.183 comes into play. If I plans to consider this comment while performing an am doing an AST analysis defending fuel bundles at the exposure limits update to RG 1.183.

for source term purposes, then maybe I do want conservative/bounding choices with respect to failed rods because the source term is essentially fixed.

On the other hand, if analyses described in DG-1327 are automatically acceptable for satisfying RG 1.183, then I probably want conservative/bounding to be with respect to Dose, as not every contributing bundle/rod will be at the exposure limit of operation during the event.

The NRC staff did not agree with this comment. RIA testing must encompass a broad range of initial conditions, materials, and transient conditions to provide We should not confuse a statistical curve fit of data, with the nature of the criteria with broad applicability.

test itself. A best estimate curve fit does not mean the proposed limit is best estimate, unless the experimentally derived data represent the Failure Furthermore, while the staff elected to employ more of a NEI 24 nominal application condition. Data used for the purposes of input to the Reject Thresholds lower bound of the failure data, as opposed to a best-fit curve fit are conservative because the nature of the testing doesnt of the failure data, there was no attempt to quantify and necessarily represent actual operating conditions. While the curve fit may apply uncertainties in the reported initial conditions (e.g.,

be best estimate, the proposed limit is conservative.

hydrogen content) nor transient conditions (e.g., failure enthalpy). Application of such uncertainties would certainly result in more restrictive failure thresholds.

19 : Revised PCMI Cladding Failure Threshold Curves Comments GNF3 and NEI14 request a change to Figure 4, PCMI Cladding Failure ThresholdRXA Cladding below 500 Degrees F. Specifically, the commenters noted that the proposed curve omits important non-failed data points and hence is less accurate in the area of interest (i.e., lower cladding excess hydrogen). The commenters proposed an alternate, exponential function (i.e., a

• Hb + c) for the failure threshold curve, along with coefficients for a best-fit and lower-bound. The commenters requested that the all of the PCMI failure curves (i.e., Figures 2 through 5) be redrawn.

The NRC staff agreed with the comment. Figure A-1 below illustrates the proposed best-fit and lower bound exponential function of the failure threshold curves along with the empirical database. The proposed exponential function appears to better represent the data. As suggested by the commenters, this equation incorporates the non-failed data and hence provides a more accurate failure threshold in the area of most interest. The exponential function also improves the curve by better representing (1) the rapid loss in RXA cladding ductility as zirconium hydrides form and (2) the saturation-effect at higher concentrations of zirconium hydrides. The commenters provided both best-fit and lower-bound coefficients. As shown in Figure A-1, both sets of coefficients improve the curve at lower concentrations of hydrides. However, at higher concentrations of excess hydrogen, the best-fit curve remains above the sole data point beyond 300 ppm (i.e., NSRR VA-6). The lower-bound coefficients intersect the VA-6 failure enthalpy.

FIGURE A-1 Regulatory stability is a concern when defining cladding failure thresholds (or any safety-related criteria) based upon a best-fit approach with a limited empirical database. Future test results will likely prompt continuous re-assessment and may invalidate best-fit failure thresholds. For example, the reported failure enthalpy for NSRR VA-5, which was published after DG-1327 was initially issued for public comment, slightly shifts the best-fit failure threshold for SRA cladding materials (in the non-conservative direction). Given that NSRR continues to conduct tests and that both CABRI and TREAT have restarted their test programs, the staff has elected to employ engineering judgement to define failure thresholds which are more representative of a lower-bound than a best-fit, but do not necessary bound all failure data. These failure curves should provide improved regulatory stability.

As a result of the GNF and NEI comment, the NRC revised all four PCMI cladding failure threshold curves using the proposed form of the equation. Coefficients were selected to better represent the non-failed data at low hydrogen levels and bound much of the failed data at higher hydrogen levels. The revised failure thresholds are shown below along with the DG-1327 curves and supporting empirical database.

20 Revised PCMI Cladding Failure Curves

21 Comparison of PCMI Failure Curves

22 : RXA Cladding PCMI Failure Curves - Adjusted for Presence of Liner Framatome Comment 2b provides an alternative proposal. Revise Section 1.2.3 to provide adjusted PCMI limits for non-liner RXA cladding. PCMI limit curves for non-liner RXA cladding materials can be developed by adjusting the hydrogen content of the liner RXA cladding tests to account for the influence of the liner on the hydrogen distribution in the bulk cladding material. In their proposal, Framatome provides a strong technical bases for an upper bound adjustment of 30% (i.e., 30% of total hydrogen content resides in liner). Based on this information, the staff adjusted the NSRR tests with lined cladding (i.e., FK series). The original and adjusted data is shown in the figure below. The excess hydrogen content in the blue symbols was reduced by 30% to account for the liner.

Using the same exponential form, the RXA cladding failure threshold curves were adjusted. The revised curves are shown below.