ML20055F489

Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod. Response to Public Comments - 2nd Comment Period

ML20055F489
Person / Time
Issue date:	06/10/2020
From:	NRC/RES/DE
To:
Edward Odonnell
Shared Package
ML20055F488	List: ML20055F489 ML20055F490 ML20055F496 ML20077M781 ML20084N202 ML20119A614 ML20275A227 ML20275A228 ML20275A229
References
DG-1327, NRC Docket-2016-0233 RG-1.236
Download: ML20055F489 (21)

Text

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 Based on comments received, the radiological source Krypton to better reflect experimental data. This modification would term information in Appendix B was removed. The NRC 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 staff is considering adding this information to a future Section B-1 .2 on pp. B-4 and B-5, and the example calculation on p. B-9. update to RG 1.183. This comment will be considered in any future update to RG 1.183.

Based on comments received, the radiological source term information in Appendix B was removed. The NRC staff is considering adding this information to a future Framatome 7 Source Terms App. B Provide definition of terms from Reference 1 for clarity and completeness. N/A update to RG 1.183. This comment will be considered in any future 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 Based on comments received, the radiological source calculated with the maximum core radial peaking factor."

term information in Appendix B was removed. The NRC Westinghouse 5a Source Terms App B N/A staff is considering adding this information to a future This is a simple but conservative assumption. The guidance requiring the update to RG 1.183. This comment will be considered in use of the maximum radial power factor should be revised to allow an any future update to RG 1.183.

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." Based on comments received, the radiological source term information in Appendix B was removed. The NRC Westinghouse 5b Source Terms App B Both cited regulatory guides have Appendix H for rod ejection defining N/A staff is considering adding this information to a future 100% noble gas and 25% for iodine for containment leakage and 50% for update to RG 1.183. This comment will be considered in iodine for secondary releases. But there is no guidance there relating to any future update to RG 1.183.

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 GWd/MTU target burnup. The staffs assessment (ADAMS ML20090A308) concluded that the guidance in The gas release calculation only supports fuel rod average burnup of 65 RG 1.236 was applicable up to a fuel burnup of 68 GWd/MTU (third paragraph on page B-1, and Figure B-1 on page B-2).

GWd/MTU (rod average), provided the cladding did not exhibit localized imperfections (e.g., spallation, hydride This should be extended to support the industry efforts to go to higher blisters) due to excessive oxidation. The staff also allowable burnups. It is recommended that the figure support burnup to at identified that the guidance was not applicable to fuel Westinghouse 5c Source Terms App B least 68 GWd/MTU which industry seeks to achieve in the near-term. Accept rods with pre-existing cladding failure (i.e., leaking, However, it would be ideal for the figure to support burnup to 75 waterlogged). In addition, the staff identified data gaps to GWd/MTU which the industry seeks to achieve in the next 5 to 7 years.

support 75 GWd/MTU (rod average).

The new ANS5.4 standard is based on database above rod average burnup of 70 GWd/MTU [9-page 4.10].

Based on comments received, the radiological source term information in Appendix B was removed. The NRC staff is considering adding this information to a future update to RG 1.183. This comment will be considered in any future update to RG 1.183.

Based on comments received, the radiological source Item B-5 on page B-7 states that, "Rod power histories used in the fuel rod term information in Appendix B was removed. The NRC design analysis based on core operating limits report thermal-mechanical Westinghouse 5d Source Terms App B N/A staff is considering adding this information to a future operating limits or radial falloff curves should be used."

update to RG 1.183. This comment will be considered in any future update to RG 1.183.

9 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject This text should be deleted as it is just one way to bound anticipated operation. The rest of the texts provides sufficient guidance for using conservative rod power histories.

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 Based on comments received, the radiological source based on. However, the text does not pick up on the next page. term information in Appendix B was removed. The NRC Westinghouse 5e Source Terms App B N/A staff is considering adding this information to a future There are issues with the break between page B-7 and page B-8. It update to RG 1.183. This comment will be considered in seems that some text may be missing from page B-8. Please correct the any future update to RG 1.183.

page break.

The table on Page B-10 has all the gap fractions set to the maximum Based on comments received, the radiological source value.

term information in Appendix B was removed. The NRC Westinghouse 5f Source Terms App B N/A staff is considering adding this information to a future This is too conservative for short life isotopes. Burnup dependent gap update to RG 1.183. This comment will be considered in fractions (similar to power fall-off) are more appropriate.

any future update to RG 1.183.

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.

10 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject Section 2.2.1.4 of the draft RG discusses that the maximum uncontrolled worth of an ejected rod should be calculated based on fully or partially inserted misaligned or inoperable rod or rods if allowed. When referring to the phrase " .. .fully or partially inserted misaligned or inoperable rod or rods if allowed" The NRC staff agreed with the comment and modified Analytical the text. Applicants do not need to consider dropped or Exelon 1 C.2.2.1.4 Exelon is requesting further clarification regarding whether this is limited to Accept Methods misaligned rods which are being recovered within TS rods in which the safety analysis has been performed to justify its non-LCO completion times.

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 Based on comments received, the radiological source extensive overlap between the two documents. If DG-1327 steady state term information in Appendix B was removed. The NRC Exelon 3 Source Terms App. B and transient gap releases supersede the NRC's previous positions, N/A staff is considering adding this information to a future Exelon recommends that DG-1327, Appendix B, should acknowledge this update to RG 1.183. This comment will be considered in fact in an effort to prevent any misunderstanding of the NRC's any future update to RG 1.183.

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.

11 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject Characterization of public comments in the Background section implies the public comments were made on the NRC memorandum supporting the technical and regulatory basis which is not appropriate. The public comments were provided on the initial DG- 1327(Reference 4). Please General/ remove text indicating it was amended by public comments as shown The NRC staff agreed with this comment and has NEI 1 B Accept Editorial below. clarified the text.

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 NRC staff agreed with this comment. The term Non-LOCA was removed.

In Appendix B, replace all instances of the term Non-LOCA with RIA, General/ Based on comments received, the radiological source NEI 2b C.1 as events other than RIA are not germane to the Regulatory Guide. Accept Editorial term information in Appendix B was removed. The NRC staff is considering adding this information to a future update to RG 1.183. This comment will be considered in any future update to 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 General/ C.2.2.1.2 the references to zero power in Items C.2.2.1.2 for PWRs and C.2.2.2.2 The NRC staff agreed with this comment and the text NEI 4 Accept Editorial C.2.2.2.2 for BWRs be updated to include hot zero power for PWRs and cold zero has been modified.

power for BWRs.

12 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject 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.

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 staff added Item C.2.3.7 in response to comment AREVA-17 from the first public comment period. The comment requested clarification on the The NRC staff agreed with this comment. Section treatment of the potential pressure reduction caused by the assumed C.2.3.7 was deleted. A slightly modified version of failure of the control rod pressure housing for criterion other than RCS AREVAs proposed text was adopted.

peak pressure.

The NRC agreed with the comment and indicated the NRC staff believes the original CRE design basis should be preserved, and plants existing license basis should be maintained (i.e., consideration of high worth rod ejections).

Additionally, comment GE-11 on the same section as comment AREVA-17 to which the NRC agreed, identified this item as only being applicable Analytical to PWRs.

NEI 9 C.2.3.7 Accept Methods Item C.2.3.7 as currently written implies the need to perform additional analyses of the control rod housing which are beyond the scope of the DG. Specifically, the NRC cited NUREG-0800, Section 3.9.4 and the requirements of GDC 14 as the basis for the additional requirements in the response to comment AREVA-17.

It is requested the NRC replace Item C.2.3.7 with the suggested text below and relocate it to Section C.2.2.1, such that there is no confusion with BWRs.

Fuel failure predictions do not need to consider any reactor coolant system depressurization resulting from the assumed failure of the control rod pressure housing.

13 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject 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 Based on comments received, the radiological source the rods in the bundle are at 62 GWd/MTU. If an approved CRE/CRDA term information in Appendix B was removed. The NRC Analytical NEI 10 C.2.4 method is applied on a cycle-specific basis, is it acceptable to use cycle N/A staff is considering adding this information to a future Methods specific rod source terms as cycle specific rod worths are already used? update to RG 1.183. This comment will be considered in any future update to 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 The NRC staff partially agreed with this comment. To be Analytical additional changes. The revised text from the response to comments from consistent with the staffs response to NEI Comment NEI 11 C.3 Partially Accept Methods the first public comment period is shown below: #23, the term bounding was deleted. The term conservative remains appropriate.

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.

The staff did not agree with the comment. Both temperature and pulse width effects were considered in the technical bases of the PCMI cladding failure curves.

Key test data that defines the proposed limits were generated under Additional RIA tests are planned to further investigate conditions far from prototypical of a commercial reactor rod ejection/rod Failure these effects. This future data may help to improve NEI 12 C.3.2 drop design basis accident. The atypical test conditions, from which the Reject Thresholds these curves.

NRC proposed limits are based, produces results not representative of commercial LWR.

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

100°C. The brittle-to-ductile transition temperature is well demonstrated in

14 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject the LS-series of tests conduct at the JAEA NSRR. Fuel from the same parent rod was tested at room temperature and 280°C. The test conducted at room temperature, LS-1, failed at an energy deposition of 53 cal/g while LS-1, conducted at 280°C, survived a maximum energy deposition of 89 cal/g.

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 Some licensees use 10 CFR 100 radiological consequences acceptance guidance related to radiological consequences is NEI 15 Source Terms C.4 criteria Revise Section 4 to include reference to 10 CFR 100 along with Reject provided in the cited RGs which provide an acceptable RG-1.183 or RG-1.195. method to satisfy applicable regulations (e.g., 10 CFR 50.67).

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 NRC staff partially agreed with this comment and NEI 17 Implementation D The industry is concerned that the extensive RIA guidance in the DG will Accept the text has been modified to reflect the intention on use be used in the future by the NRC staff for license amendment requests of the RG.

that do not specifically involve RIA-related plant changes. The types of LARs that do involve RIA and DG-1327 evaluations have been identified

15 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject 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 Based on comments received, the radiological source fission gas gap fraction calculations, a technical basis document would be term information in Appendix B was removed. The NRC NEI 18 Source Terms App. B beneficial. Please revise PNNL- 18212 to use the FAST code per N/A staff is considering adding this information to a future ML19154A226. update to RG 1.183. This comment will be considered in any future update to 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 Based on comments received, the radiological source calculations for steady state non- LOCA gap fractions. Based on a review term information in Appendix B was removed. The NRC NEI 19 Source Terms App. B of the reported results in ML19154A226, and using conventional rounding N/A staff is considering adding this information to a future techniques, appropriate gap fractions for I-131 and other Halogens would update to RG 1.183. This comment will be considered in be 0.05 and 0.03, respectively. Update the gap fractions to reflect the any future update to RG 1.183.

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.

Page B-2 Paragraph 1 last sentence: The sentence uses the phrase described in the attachment. Please replace in the attachment with within this appendix.

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

Based on comments received, the radiological source While calibrated and validated against a large empirical database, FAST term information in Appendix B was removed. The NRC and its predecessors are not NRC-approved codes and may not be NEI 20 Source Terms App. B N/A staff is considering adding this information to a future utilized to calculate that plant-specific, fuel-specific, or cycle-specific gap update to RG 1.183. This comment will be considered in inventories that are in accordance with the acceptable analytical any future update to RG 1.183.

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.

Based on comments received, the radiological source If that were to occur then an Appendix B-based methodology submitted by term information in Appendix B was removed. The NRC a vendor/licensee and approved by the NRC may not be consistent with NEI 21 Source Terms App. B N/A staff is considering adding this information to a future the updated RG 1.183.

update to RG 1.183. This comment will be considered in any future update to RG 1.183.

The industry requests the update to RG 1.183 and the deletion of Appendix B be an administrative change only, and that no technical changes are included.

The industry is also concerned that there is no indication a DG-1327 Appendix B Dose assessment is sufficient to demonstrate compliance to

16 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject 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.

Based on comments received, the radiological source term information in Appendix B was removed. The NRC Please clarify the exposures discussed in the figure are pellet exposure, staff is considering adding this information to a future NEI 22 Source Terms App. B not rod exposure. Clearly identify exposure basis and application. N/A update to RG 1.183. This comment will be considered in any future update to RG 1.183.

The conservative or bounding terminology are relative terms. So, what are they relative too? Specifically, Section C.2.3 is entitled Predicting the total number of fuel rod failures. Is the conservative or bounding The NRC staff did not agree with this comment. The terminology supposed to be with respect to the number of rods failed, or is terms conservative or bounding do not appear in it really supposed to be with respect to dose consequence?

Section C.2.3.

When the failure criteria for a fuel rod was a constant with respect to With respect to calculating the number of failed rods, the exposure, a failed number of rods could be thought of as a surrogate for word conservative only appears once (Section C.3):

dose, and dose could be a surrogate for failed rods. The new non-linear failure criteria breaks that line of reasoning. It is possible to envision To ensure a conservative assessment of onsite scenarios with higher dose consequence with fewer rod failures and not and offsite radiological consequences, each of just from the rod eject / rod drop perspective, but from all non-LOCA these failure modes should be quantified, and the events.

sum total number of failed fuel rods should not Analytical NEI 23 All Reject be underestimated.

Methods Please clarify the basis for DG-1327, and explicitly express what the appropriate metric is for assessing terminology such as conservative or Its use in this instance is judged appropriate.

bounding.

With respect to calculating radiological consequences, This issue is important with respect to how RG 1.183 comes into play. If I the word bounding appears in many places within am doing an AST analysis defending fuel bundles at the exposure limits Appendix B. Based on comments received, the for source term purposes, then maybe I do want conservative/bounding radiological source term information in Appendix B was choices with respect to failed rods because the source term is essentially removed. The NRC staff is considering adding this fixed.

information to a future update to RG 1.183. This On the other hand, if analyses described in DG-1327 are automatically comment will be considered in any future update to RG acceptable for satisfying RG 1.183, then I probably want 1.183.

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 We should not confuse a statistical curve fit of data, with the nature of the testing must encompass a broad range of initial test itself. A best estimate curve fit does not mean the proposed limit is conditions, materials, and transient conditions to provide best estimate, unless the experimentally derived data represent the Failure criteria with broad applicability.

NEI 24 nominal application condition. Data used for the purposes of input to the Reject Thresholds curve fit are conservative because the nature of the testing doesnt Furthermore, while the staff elected to employ more of a necessarily represent actual operating conditions. While the curve fit may lower bound of the failure data, as opposed to a best-fit be best estimate, the proposed limit is conservative.

of the failure data, there was no attempt to quantify and

17 Accept/

Commenter # Category Section Summation of Comment NRC Response Reject apply uncertainties in the reported initial conditions (e.g.,

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

18 : 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.

19 Revised PCMI Cladding Failure Curves

20 Comparison of PCMI Failure Curves

21 : 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.