ML20055F489
| ML20055F489 | |
| Person / Time | |
|---|---|
| Issue date: | 06/10/2020 |
| From: | NRC/RES/DE |
| To: | |
| Edward Odonnell | |
| Shared Package | |
| ML20055F488 | List: |
| 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 Commenter Category Section Summation of Comment Accept/
Reject NRC Response GNF 1
Analytical Methods C.2.2.2.3 This section declares that analyses should consider the potential for wider operating conditions as the result of xenon oscillations or plant maneuvering. The phrase xenon oscillations is applicable only to PWRs as BWRs do not experience spatial xenon oscillations. This phrase should be removed, or the DG should state that this is applicable only to PWRs.
Accept The NRC staff agreed with this comment and deleted the phrase xenon oscillations from the text.
GNF 2
Analytical Methods C.2.2.2.4 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.
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.
Accept The NRC staff agreed with this comment and incorporated the proposed language into 2.2.2.4.
GNF 3
Failure Thresholds C.3.2 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 points.
The primary response from a CRDA is often from the fresh fuel (i.e. lower exposure) with highly exposed fuel reacting less energetically. Thus, the 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.
Accept The NRC staff agreed with the comment and modified the failure threshold lines on Figures 2 through 5 to encompass more of the non-failure data points and also adopted the proposed exponential function form. See for further details.
General/
Editorial -
Background
The last phrase can be misinterpreted to suggest that hydrogen enhances PCMI. The impact of hydrogen is on the potential for cladding failure.
Proposal: "The prompt thermal expansion of the fuel pellet, which can be exacerbated at high burnups by gaseous fission product swelling, may Accept The NRC staff agreed with this comment and the proposed language was adopted.
4 Commenter Category Section Summation of Comment Accept/
Reject NRC Response cause the fuel cladding to fail by PCMI. The potential for failure is enhanced by the presence of hydrogen in the cladding."
Framatome 2a Failure Thresholds C.1.2.3 The guidance limits the applicability of the RXA PCMI failure curves to 70 wppm excess hydrogen for non-lined cladding. A limit of 70 wppm of excess hydrogen creates undue burden for M5 cladding material. Based on the information presented in this comment, Framatome requests that this applicability limit be deleted.
Reject The NRC did not agree with this comment. While the commenter provided technically reasonable arguments, the staff still considers the liner (and the absorbed hydrogen within the liner) to have an impact on the initial condition of the cladding. Where the liner not present, more of the hydrogen would likely reside toward the exterior of the cladding which would likely influence the failure enthalpy.
Framatome 2b Failure Thresholds C.1.2.3 Alternative Proposal: If the proposal above is not considered an acceptable proposal, then 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.
Accept The NRC agreed with this comment. See Attachment 2 for further details.
Analytical Methods C.2.3.5 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."
Proposed revision:
"Each applicant should address the possibility of hydride reorientation 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."
Accept The NRC staff agreed with this comment. Reference to NUREG-0800, SRP Chapter 4 was added. The revision was omitted to avoid confusion with future revisions.
Failure Thresholds C.3.1 The RIA criteria for High-Temperature Cladding Failure Threshold are derived from RIA testing of fuel pins undergoing prompt critical power excursions. These power pulses have pulse widths < 0.05 seconds. The initial power level does not define whether there is a power pulse that needs to be evaluated against these criteria. The RAI criterion for High Temperature Cladding Failure Threshold in the DG 1327 Figure 1 is not applicable to conditions where there is no power pulse.
Starting at extremely low powers with a non-prompt reactivity insertion does not result in a power pulse. A non-prompt reactivity insertion results in a prompt jump proportional to the original flux with a subsequent period of seconds rather than milliseconds and may not have a distinct pulse shape. For example, a reactivity insertion of 0.95$ would cause the flux to 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 Accept The NRC agreed with this comment and has modified the text. Framatomes technical bases for the proposed change seems logical. To confirm the proposed change in the applicability of the high-temperature cladding failure curve, the staff completed a series of calculations with a simple point-kinetics model with simulated fuel temperature reactivity feedback (i.e., Doppler). The figure below illustrates the impact of changes in reactivity insertion on the resulting power profile. As shown in the figure, a prompt power excursion resulting from a large positive reactivity insertion exhibits a large, narrow pulse. Whereas, non-prompt power excursions exhibit a lesser, broader power profile.
5 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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 (i.e. the ejected rod worth or drop rod worth >1.0$). Because ductile 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).
These confirmatory calculations support the proposed changes.
Coolable Geometry C.6 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.
Proposal: "If fuel melting occurs, the peak fuel temperature in the outer 90 percent of the fuel volume should remain below incipient fuel melting conditions."
Accept The NRC staff agreed with this comment and the proposed language has been adopted.
Source Terms App. B Modify the Cesium FGR fraction to be based on Xenon rather than Krypton to better reflect experimental data. This modification would change Table B-1 and footnote 81 on p. B-1, Section B.2 on p. B-3, Section B-1.2 on pp. B-4 and B-5, and the example calculation on p. B-9.
N/A 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.
Source Terms App. B Provide definition of terms from Reference 1 for clarity and completeness.
N/A 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.
6 Commenter Category Section Summation of Comment Accept/
Reject NRC Response Westinghouse 1
Applicability C.1.1.1 It is suggested that Section 1.1.1 be modified as The applicability of this guidance to LWR fuel rod designs, different from the UO2 fuel rod designs described in Section 1.1 (e.g., doped pellets, changes in fuel pellet microstructure or density, changes in zirconium alloy cladding microstructure or composition, coated zirconium alloy cladding), will be addressed on a case-by-case basis.
The doped pellets and changes in zirconium alloy have been used already in light water reactor (LWR) fuel designs. Some of those fuel designs have been previously addressed and approved as similar to the UO2 fuel rod design, such as the Optimized ZIRLO' fuel cladding.
Reject The NRC staff did not agree with this comment. The empirical database used to establish the guidance and limits included in this regulatory guide is limited to the applicability described in Section 1 as written.
As stated, the guidance is applicable to currently approved designs (e.g., Optimized ZIRLOTM). Applicants requesting approval for future fuel designs (e.g., above 5.0% 235U enrichment) would need to demonstrate the continued applicability of the guidance or provide an alternative.
2 Applicability C.1.2.2 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 cladding ductility as a function of hydrogen, hydrides distribution after irradiation, and hydride reorientation behavior under stress are similar for ZIRLO cladding. This data supports a position that Optimized ZIRLO cladding should have similar reactivity initiated accident (RIA) cladding failure limits as ZIRLO cladding. A significant part of this data has been previously submitted as part of the PAD5 licensing and in public domain papers.
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.
Accept The NRC staff agreed with this comment. On February 13, 2020, the staff conducted an audit of the Westinghouse documentation supporting the application of the SRA PCMI cladding failure threshold to Optimized ZIRLO. Based on the results of the audit (ADAMS Accession No. ML20049F944), the staff accepts the use of the SRA PCMI cladding failure thresholds for Optimized ZIRLO cladding.
Westinghouse 3a Failure Thresholds C.3.1 Please clarify the technical basis of the 5% power level for the switch between the PCMI failure criterion and the high temperature cladding failure criterion.
Partially Accept 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 PCMI.
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 Commenter Category Section Summation of Comment Accept/
Reject NRC Response Westinghouse 3b Failure Thresholds C.3.1 PCMI should be used as the only failure criteria when an ejected rod worth is equal to or greater than $1. DNB should be used as the only failure criteria when ejected rod worth is less than $1.
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.
There are no approved fuel rod thermal-mechanical 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.
Failure Thresholds C.3.1 Based on earlier studies documented in NUREG/CR-0269, it can be 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 failure in a short transient period.
For prompt critical events, the proposed DG-1327 limits for PCMI conservatively envelope time at temperature failure from DNB. For prompt critical cases that do not fail due to PCMI and remain at a significant power level (such as cases that do not trip), the DNB criterion will be applied following the prompt critical power excursion.
Westinghouse proposes for non-prompt critical RIA at part power operation, specific time at temperature cladding limits can conservatively be applied to calculated cladding time and temperature in DNB and separate failure from non-failure.
Reject 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 separate failure threshold.
Framatome comment #4 was accepted by the staff and includes changes related to the application of DNB failure to prompt and non-prompt scenarios. This change addresses a portion of the Westinghouse comment.
During the transient, the time in boiling crisis depends on several variables including the rate and magnitude of deposited energy and local thermal-hydraulic conditions.
A firm technical basis for a time-at-temperature criteria does not exist.
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 would consider alternate criteria, including a time-at-temperature failure threshold, on a case-by-case basis.
8 Commenter Category Section Summation of Comment Accept/
Reject NRC Response Westinghouse 5a Source Terms App B 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 calculated with the maximum core radial peaking factor."
This is a simple but conservative assumption. The guidance requiring the use of the maximum radial power factor should be revised to allow an alternative (more realistic but still conservative) calculation using actual power history.
N/A 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.
Westinghouse 5b Source Terms App B 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."
Both cited regulatory guides have Appendix H for rod ejection defining 100% noble gas and 25% for iodine for containment leakage and 50% for 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.
N/A 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.
Westinghouse 5c Source Terms App B The gas release calculation only supports fuel rod average burnup of 65 GWd/MTU (third paragraph on page B-1, and Figure B-1 on page B-2).
This should be extended to support the industry efforts to go to higher allowable burnups. It is recommended that the figure support burnup to at least 68 GWd/MTU which industry seeks to achieve in the near-term.
However, it would be ideal for the figure to support burnup to 75 GWd/MTU which the industry seeks to achieve in the next 5 to 7 years.
The new ANS5.4 standard is based on database above rod average burnup of 70 GWd/MTU [9-page 4.10].
Accept 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 RG 1.236 was applicable up to a fuel burnup of 68 GWd/MTU (rod average), provided the cladding did not exhibit localized imperfections (e.g., spallation, hydride blisters) due to excessive oxidation. The staff also identified that the guidance was not applicable to fuel rods with pre-existing cladding failure (i.e., leaking, waterlogged). In addition, the staff identified data gaps to support 75 GWd/MTU (rod average).
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.
Westinghouse 5d Source Terms App B 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 operating limits or radial falloff curves should be used."
N/A 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.
9 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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.
Westinghouse 5e Source Terms App B 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. However, the text does not pick up on the next page.
There are issues with the break between page B-7 and page B-8. It seems that some text may be missing from page B-8. Please correct the page break.
N/A 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.
Westinghouse 5f Source Terms App B The table on Page B-10 has all the gap fractions set to the maximum value.
This is too conservative for short life isotopes. Burnup dependent gap fractions (similar to power fall-off) are more appropriate.
N/A 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.
Dominion 1
Failure Thresholds C.3.1 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 (e.g., departure from nucleate boiling and critical power ratios) should be evaluated and would appreciate the NRCs consideration of those comments.
Accept See response to Westinghouse 3a and Framatome 4.
Dominion 2
Source Terms App B 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 release fraction guidance existing in the different regulatory guides. In 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.
Accept The NRC staff agreed with this comment and plans to move Appendix B to RG 1.183.
10 Commenter Category Section Summation of Comment Accept/
Reject NRC Response Exelon 1
Analytical Methods C.2.2.1.4 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" 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).
Accept The NRC staff agreed with the comment and modified the text. Applicants do not need to consider dropped or misaligned rods which are being recovered within TS LCO completion times.
Exelon 2
Analytical Methods C.2.2.2.4 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 power level, (2) additional fully or partially inserted misaligned or inoperable blades if allowed, and (3) any out-of-sequence control blades that may be inserted for fuel leaker power suppression. When referring to the phrase "...fully or partially inserted misaligned or inoperable rod or rods if allowed...," Exelon is requesting further clarification whether this is also intended to include inoperable blades that have been locked in place and cannot physically move or be dropped in accordance with the associated TS.
Accept The NRC staff agreed with the comment and modified the text. Applicants do not need to consider uncontrolled withdrawal (as the initiating event) of an inoperable blade that has been locked in place and cannot physically move. However, the impact of that inoperable blade on the worth of other blades needs to be considered.
Exelon 3
Source Terms App. B 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 extensive overlap between the two documents. If DG-1327 steady state and transient gap releases supersede the NRC's previous positions, Exelon recommends that DG-1327, Appendix B, should acknowledge this 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.
N/A 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.
11 Commenter Category Section Summation of Comment Accept/
Reject NRC Response NEI 1
General/
Editorial B
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 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.
Accept The NRC staff agreed with this comment and has clarified the text.
NEI 2a General/
Editorial C.1 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 anticipated operational occurrences (AOOs) and other postulated accidents involving positive reactivity insertion (e.g., PWR excess load, 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.
Accept The NRC staff agreed with this comment and the proposed language has been adopted.
NEI 2b General/
Editorial C.1 In Appendix B, replace all instances of the term Non-LOCA with RIA, as events other than RIA are not germane to the Regulatory Guide.
Accept The NRC staff agreed with this comment. The term Non-LOCA was removed.
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.
NEI 3
Related Guidance C.2.1.1 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.
Add the following to the end of section C.2.1.1 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.
Partially Accept The NRC staff partially agreed with this comment and the text has been modified. Note there are portions of RG 1.203 (e.g., QA, documentation) which are not addressed within this guidance.
NEI 4
General/
Editorial C.2.2.1.2 C.2.2.2.2 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 power for BWRs.
Accept The NRC staff agreed with this comment and the text has been modified.
12 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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.
NEI 5
General/
Editorial C.2.2.1.5 Section C.2.2.4 should be Section C.2.2.1.4 Accept The NRC staff agreed with this comment and the proposed language has been adopted.
NEI 6
General/
Editorial C.2.2.1.10 C.2.2.1.11 C.2.2.2.10 C.2.2.2.11 Removing terms coefficients and coefficient of with a more generic term such as reactivity feedback, as there are multiple ways to simulate the reactivity mechanisms within an analysis.
Accept The NRC staff agreed with this comment and the proposed language has been adopted.
NEI 7
General/
Editorial C.2.3.3 C.2.4 The segmenting of the axial length uses the word several. It is expected that the number of axial nodes would be much larger than several.
Replace several with selected.
Accept The NRC staff agreed with this comment and the proposed language has been adopted.
NEI 8
General/
Editorial C.2.3.4 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 as acceptable. The concern is that a vendor/licensee submittal of the Appendix C hydrogen uptake model would be subject to additional NRC review. In Appendix C replace acceptable with NRC-approved.
Accept The NRC staff agreed with the sentiment of this comment. The language in section C.2.3.4 has been modified to add (or the appropriate model from Appendix C of this guidance).
NEI 9
Analytical Methods C.2.3.7 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 treatment of the potential pressure reduction caused by the assumed failure of the control rod pressure housing for criterion other than RCS 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 to PWRs.
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.
Accept The NRC staff agreed with this comment. Section C.2.3.7 was deleted. A slightly modified version of AREVAs proposed text was adopted.
13 Commenter Category Section Summation of Comment Accept/
Reject NRC Response NEI 10 Analytical Methods C.2.4 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 rods in the bundle are at 62 GWd/MTU. If an approved CRE/CRDA method is applied on a cycle-specific basis, is it acceptable to use cycle specific rod source terms as cycle specific rod worths are already used?
Please clarify the expectations between DG-1327 and RG1.183.
N/A 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.
NEI 11 Analytical Methods C.3 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 additional changes. The revised text from the response to comments from the first public comment period is shown below:
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.
Partially Accept The NRC staff partially agreed with this comment. To be consistent with the staffs response to NEI Comment
- 23, the term bounding was deleted. The term conservative remains appropriate.
NEI 12 Failure Thresholds C.3.2 Key test data that defines the proposed limits were generated under conditions far from prototypical of a commercial reactor rod ejection/rod drop design basis accident. The atypical test conditions, from which the NRC proposed limits are based, produces results not representative of commercial LWR.
Reject 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.
Additional RIA tests are planned to further investigate 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.
NEI 13 Failure Thresholds C.3.2 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 high hydrogen concentration. A ductile-to-brittle transition temperature of less 100°C was later presented by the same author in 2013 for M5 at lower hydrogen concentration. The reported transition temperature is consistent with 100°C determined under RIA conditions in reference, for pulse width greater than 10 ms. In the past a NSRR RIA test was conducted at 85C but did not show noticeable improvement in energy absorption capacity. Test data from reference would indicate at the 4-5 ms pulse width the brittle-to-ductile transition temperature is higher than 100°C. The brittle-to-ductile transition temperature is well demonstrated in Reject The NRC disagreed with this comment. This guidance does not define a minimal measure of cladding ductility such as a DBTT. But instead, addresses the changing degrees of ductility necessary to avoid cladding failure as a function of increasing fuel enthalpy (and associated pellet thermal expansion). Since zirconium hydrides have a dominant effect on cladding ductility, the cladding failure threshold is provided as a function of excess hydrogen. The NRCs investigation found that the impact of initial cladding temperature on PCMI failure threshold was only 18 cal/g between cold (room temperature) testing and hot (above 500°F) testing. The NRC would consider, on a case-by-case basis, further scaling between 500°F and a lower temperature (corresponding to plant-specific BWR startup conditions).
14 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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.
NEI 14 Failure Thresholds C.3.2 Same comment as GNF3 (above).
Accept The NRC staff agreed with the comment and has modified the failure threshold lines on Figures 2 through 5 to encompass more of the non-failure data points and adopted the proposed exponential function form. See for further details.
NEI 15 Source Terms C.4 Some licensees use 10 CFR 100 radiological consequences acceptance criteria Revise Section 4 to include reference to 10 CFR 100 along with RG-1.183 or RG-1.195.
Reject The NRC staff did not agree with this comment. Current guidance related to radiological consequences is provided in the cited RGs which provide an acceptable method to satisfy applicable regulations (e.g., 10 CFR 50.67).
NEI 16 Implementation D
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 initially performed an assessment of postulated reactivity-initiated accidents for operating reactors in the US in Research Information Letter 0401, dated March 31, 2004. The 2004 assessment concluded that there was no concern related to protecting the health and safety of the public for the operating reactors. The NRC has issued two memorandums (dated January 17, 2007 and March 16, 2015) on the proposed technical and regulatory basis for reactivity-initiated accident acceptance criteria since the 2004 assessment. The two memorandums continued to reference the 2004 safety assessment. Given the conclusion of the 2004 assessment and the continued reliance upon it, it is believed that NRC staff does not have a safety concern related to protecting the Health and Safety of the public for the operating reactors based on the issuance of the guidance contained in DG-1327.
Reject The NRC staff did not agree with this comment. RIL-0401 was based on a limited, realistic assessment of PWR control rod worths. The assessment concluded that legacy methods (e.g. point kinetics, 1D) were sufficiently conservative to compensate for the new research findings. However, like any safety assessment, its a snap-shot in time. Fuel designs, materials, analytical methods, and fuel utilization are not stagnant.
For the past 16 years, new fuel assembly designs (e.g.,
GNF2, GNF3, Atrium 10XM, Atrium 11, SVEA Optima2, GAIA), cladding materials (e.g., Optimized ZIRLO), and analytical methods (e.g., 3D realistic) have been implemented. Conclusions from RIL-0401 may no longer be valid. This new guidance provides an acceptable path to support all of these new technology improvements.
NEI 17 Implementation D
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 be used in the future by the NRC staff for license amendment requests that do not specifically involve RIA-related plant changes. The types of LARs that do involve RIA and DG-1327 evaluations have been identified Accept The NRC staff partially agreed with this comment and the text has been modified to reflect the intention on use of the RG.
15 Commenter Category Section Summation of Comment Accept/
Reject NRC Response by the staff in the NRC response to the first round of DG-1327 comments, (p. 45 Item e).
NEI 18 Source Terms App. B Since RG 1.183 is not consistent with current codes and the consensus of fission gas gap fraction calculations, a technical basis document would be beneficial. Please revise PNNL-18212 to use the FAST code per ML19154A226.
N/A 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.
NEI 19 Source Terms App. B 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 calculations for steady state non-LOCA gap fractions. Based on a review of the reported results in ML19154A226, and using conventional rounding techniques, appropriate gap fractions for I-131 and other Halogens would be 0.05 and 0.03, respectively. Update the gap fractions to reflect the results of ML19154A226 using conventional rounding techniques.
N/A 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.
NEI 20 Source Terms App. B 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:
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.
N/A 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.
NEI 21 Source Terms App. B 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 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 N/A 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.
16 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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.
NEI 22 Source Terms App. B Please clarify the exposures discussed in the figure are pellet exposure, not rod exposure. Clearly identify exposure basis and application.
N/A 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.
NEI 23 Analytical Methods All 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 terminology supposed to be with respect to the number of rods failed, or is it really supposed to be with respect to dose consequence?
When the failure criteria for a fuel rod was a constant with respect to exposure, a failed number of rods could be thought of as a surrogate for 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 scenarios with higher dose consequence with fewer rod failures and not just from the rod eject / rod drop perspective, but from all non-LOCA events.
Please clarify the basis for DG-1327, and explicitly express what the appropriate metric is for assessing terminology such as conservative or bounding.
This issue is important with respect to how RG 1.183 comes into play. If I am doing an AST analysis defending fuel bundles at the exposure limits 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.
Reject The NRC staff did not agree with this comment. The terms conservative or bounding do not appear in Section C.2.3.
With respect to calculating the number of failed rods, the word conservative only appears once (Section C.3):
To ensure a conservative assessment of onsite and offsite radiological consequences, each of these failure modes should be quantified, and the sum total number of failed fuel rods should not be underestimated.
Its use in this instance is judged appropriate.
With respect to calculating radiological consequences, the word bounding appears in many places within Appendix B. 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.
NEI 24 Failure Thresholds We should not confuse a statistical curve fit of data, with the nature of the test itself. A best estimate curve fit does not mean the proposed limit is best estimate, unless the experimentally derived data represent the nominal application condition. Data used for the purposes of input to the curve fit are conservative because the nature of the testing doesnt necessarily represent actual operating conditions. While the curve fit may be best estimate, the proposed limit is conservative.
Reject 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 criteria with broad applicability.
Furthermore, while the staff elected to employ more of a lower bound of the failure data, as opposed to a best-fit of the failure data, there was no attempt to quantify and
17 Commenter Category Section Summation of Comment Accept/
Reject NRC Response 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.