ML22174A072

Comment (11) of Frances A. Pimentel on Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants

ML22174A072
Person / Time
Site:	Nuclear Energy Institute
Issue date:	06/21/2022
From:	Pimentel F Nuclear Energy Institute
To:	Office of Administration
References
87FR23891 00011, DG-1389, NRC-2021-0179
Download: ML22174A072 (22)

Text

6/23/22, 6:59 AM blob:https://www.fdms.gov/f7b0535d-bc59-408c-bf70-821a6738efa6 SUNI Review Complete Template=ADM-013 As of: 6/23/22 6:59 AM E-RIDS=ADM-03 Received: June 21, 2022 PUBLIC SUBMISSION ADD: Michael Eudy, Kevin Hsueh, Sean Meighan, Status: Pending_Post Tracking No. l4o-dzh7-8x7o Mark Blumberg, Bridget Comments Due: June 21, 2022 Curran, Mary Neely Comment (11) Submission Type: Web Publication Date:

4/21/2022 Docket: NRC-2021-0179 Citation: 87 FR 23891 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants Comment On: NRC-2021-0179-0001 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Document: NRC-2021-0179-DRAFT-0013 Comment on FR Doc # 2022-08519 Submitter Information Email: atb@nei.org Organization: Nuclear Energy Institute General Comment See attached file(s)

Attachments 06-21-22_NRC_Industry Comments on DG-1389 blob:https://www.fdms.gov/f7b0535d-bc59-408c-bf70-821a6738efa6 1/1

FRANCES PIMENTEL Senior Project Manager, Risk and Technical Support 1201 F Street, NW, Suite 1100 Washington, DC 20004 NUCLEAR ENERGY INSTITUTE P: 202.739.8132 fap@nei.org June 21, 2022 Office of Administration Mail Stop: TWFN-7-A60M U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Program Management, Announcements and Editing Staff Project Number: 689

Subject:

NEI Comments on draft Regulatory Guide (DG), DG-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Docket ID NRC-2021-0179.

Subm itted via regulations.gov

Dear Program Management,

Announcements and Editing Staff, The Nuclear Energy Institute (NEI) 1, on behalf of our members, appreciates the opportunity to provide 0F comments on the subject draft regulatory guide (DG), DG-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. This DG is proposed Revision 1 to Regulatory Guide (RG) 1.183 which describes a method that the NRC staff considers acceptable for complying with regulations for design basis accident dose consequence analysis using an Alternative Source Term. The purpose of this letter is to provide the attached comments which recommend several changes to improve clarity and consistency on the recommended approaches, methods, and analysis this guidance provides.

These comments were developed by a nuclear energy industry task force and reflect a substantial body of industry technical expertise, experience, and lessons learned gained from successful licensing actions utilizing alternative radiological source terms as part of their supporting design basis accident analyses.

Of particular concern is that the DG does not include the results of the staffs PWR and BWR analyses to enable an assessment of the impact to analyzed doses resulting from the updated guidance, such as impact from the new release fractions and timing. Industry views the proposed changes in this DG just as significant as when the guidance transitioned from the TID Source Term to the Alternative Source Term. As proposed, the conservatisms and changes incorporated in this revision precludes its use by many plants that 1

The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.

NUCLEAR. CLEAN AIR ENERGY

June 21, 2021 Page 2 are interested in implementing near-term ATF design concepts, fuel burnup extension to 68 GWd/MTU (peak rod average), and 235U enrichments up to 8.0 wt%.

A comparison of the changes in MHA release fractions between BWRs (Table 1) and PWRs (Table 2) identified a significant increase in the BWR halogen release fractions with no indication in either SAND2011-0128 or DG-1389 as to the cause. This increase may adversely affect the ability of BWRs to comply with RG 1.183 Rev. 1. A comparison to the PWR analyses suggests that the accident sequences may be responsible for this impact. SAND2011-0128 updates NUREG-1465 with higher core exposures utilizing the latest NRCs MELCOR methodology. The accident sequences that were analyzed to develop the PWR release fractions are listed in Table 5 of the Sandia report and include a variety of accident types. However, for the BWR release fractions, Table 3 of the Sandia report indicates that nearly all the evaluations were based on station blackout (SBO) sequences. However, the risk from SBO events has been substantially reduced by the industrys implementation of FLEX. Consistent with the PWR analysis in SAND2011-0128, the BWR release fractions should be re-evaluated to ensure that they are based on an appropriate set of accident sequences that more accurately reflect BWR risk profiles.

Additionally, it is unclear from the SAND2011-0128 report whether suppression pool scrubbing was credited in determining the release fractions from a BWR SBO. Credit for suppression pool scrubbing can significantly decrease the airborne activity since the SBO-related releases would be released via spargers submerged in the suppression pool. Therefore, any sequences that involve releases through the pool spargers should take credit for suppression pool scrubbing.

Also, although new guidance was added for crediting holdup and retention of MSIV leakage within the main steamlines and condenser for BWRs, it contains more conservative assumptions than previous models that limits its effectiveness. For example, Section A-5 of the DG presents three acceptable methods for calculating aerosol deposition within the main steam lines, but also states, these methods are not valid if credit has been taken for aerosol removal from drywell sprays. No technical justification is provided for why aerosol removal from drywell sprays is not valid if used with credit for main steamline deposition. As a result, the effectiveness of the application of these models is significantly reduced. Considering the number of BWRs currently modeling both removal mechanisms, the DG should provide guidance for crediting both of these important mitigative features.

Further, the DG states, Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations. A combination of the methods contained in revision 0 or revision 1 of RG 1.183 would need additional justification. However, it is not clear in the proposed document if licensees can transition one (or a few) analyses to revision 1, or if full implementation for all design basis accidents is required. Clarification should be added to discuss that selective implementation is acceptable, provided that each accident analysis uses either revision 0 or revision 1 and to specify that a combination of the methods contained in revision 0 or revision 1 of RG 1.183, in a single analysis, would need additional justification.

June 21, 2021 Page 3 Finally, we appreciate the staffs efforts to include guidance for accident tolerant fuel (ATF), high-burnup fuel, and increased enrichment source term analyses and in revising the transport and decontamination models for the fuel handling DBA when developing this draft guidance. We encourage your consideration of all stakeholder comments prior to finalizing this draft RG. Given the long-lasting impact of the final document, we recommend a public meeting be scheduled to discuss in more detail the comments included in the attachment and how the staff plans to address them. We trust that you will find these comments useful and informative as you finalize the draft, and we look forward to future engagement on this important matter.

Please contact me at fap@nei.org or (202) 739-8132 with any questions or comments about the content of this letter or the attached comments.

Sincerely, Frances A. Pimentel Attachment c: Sean Meighan, NRR/DRA/ARCB, NRC Mike Franovich, NRR/DRA, NRC

Consolidated Comments on Draft Regulatory Guide DG-1389 (RG 1.183, Rev.1)

Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors No. Section Comment/Basis Recommendation Section B The NRC has identified that licensees may elect to use risk insights in support of proposed NRC should provide some additional (Background) changes to the design basis that are not addressed in currently approved NRC staff explanation on the acceptable uses of

[p. 5] and positions. The NRC should provide some more explanation on the parameters/limitations risk insight in conjunction with the

1. Section 1.4 of using risk insights. radiological analyses performed with Risk this RG.

Implications

[p. 14-15]

The last paragraph of the section states: Revision 0 of RG 1.183 will continue to be NRC should include a statement like the available for use by licensees and applicants as a method acceptable to the NRC staff for following: For example, across a demonstrating compliance with the regulations. A combination of the methods contained licensing basis, different revision in Revision 0 or Revision 1 of RG 1.183 would need additional justification. methods may be adopted provided the individual analyses fully adopt a single Section B.

The NRC should provide additional clarifications in this paragraph (or Section 1.2) to revision (i.e., individual analysis inputs,

2. (Background) inform the licensee as to acceptable instances for combination of Revision 0 and 1 assumptions, or modeling aspects may

[p. 6]

methods. not adopt different revision methods within the same analysis).

Additional details should also be included in Section 1.2 - Scope of Implementation.

The last sentence states: Once the staff has approved the initial AST implementation and The NRC should clarify the sentence as:

Section 1.1.1 it has become part of the facility design basis Once the staff has approved the initial

3. (Safety This sentence should be clarified to identify the initial AST implementation associated AST implementation in accordance with Margins) [p.8] with Revision 1 of RG 1.183. Revision 1 and it has become part of the facility design basis SECY-98-154, Reference 19, performed a re-baselining of sample radiological Supplement Section 1.3.2 by consequences analyses when transitioning from TID to AST source terms. Has the NRC incorporating results from supporting Section 1.3.2 performed similar re-baselining studies for the major method changes being made in DG- analyses performed by staff when (Reanalysis 1389 compared to RG 1.183, Revision 0? Notably, the NRC should investigate sample PWR developing the proposed revisions 4.

Guidance) and BWR analyses for the LOCA release fraction and timings, non-LOCA release fractions contained in DG-1389.

[p.13] (including TFGR components), updated FHA modeling, etc. If this has already been performed, the wording for this section should be supplemented to include reference to these studies.

Page 1

No. Section Comment/Basis Recommendation Section 3.1 RG 1.183, Revision 0 identified specific computer codes acceptable for use in core The NRC should continue to include (Fission inventory calculations (ORIGEN2 and ORIGEN-ARP). These code examples were deleted examples of core inventory codes which

5. Product for DG-1389. The NRC should continue to include examples of core inventory codes which are acceptable for use in core inventory Inventory) are acceptable for use. calculations.

[p.17]

Section 3.1 Based on the directive to "maximize fission product inventory" it is implied that bounding The RG should be consistent in Fission core parameters should be used, however downstream discussion (e.g., Section 3.2) prescribing if a bounding or an

6. Product states that the inventory should be for an equilibrium core. The RG should be consistent equilibrium core should be applied to Inventory - in prescribing if a bounding or an equilibrium core should be applied to dose calculations. dose calculations.

[p.18]

Section 3.1 Its not clear that this directive can be applied. Many BWR units do not have radial Remove discussion of COLR/TS from this Fission peaking factors in the COLR nor Technical Specifications. section.

7. Product Inventory -

[p.18]

Footnote 10 identifies that the data in this section does not apply to cores containing Footnote 10 should also include these Section 3.2 mixed oxide fuel (MOX). The section also identifies that Accident Tolerant Fuel (ATF) ATF concepts as not being applicable for (Release

8. concepts, excluding near-term ATF concepts, are also not applicable to the data in this the data in Section 3.2.

Fractions) section. Footnote 10 should also include these ATF concepts as not being applicable for

[p.18]

the data in Section 3.2.

It states the steady-state fission product release fractions in Table 3 can only be used if Please remove the requirement that BWR part-length rods are treated as full-length rods with respect to overall quantity of BWR part-length rods are treated as full-Section 3.2, fission products. BWR fuel bundles can have up to 20 part-length rods. Assuming all the length rods with respect to overall

9. 4th paragraph part length rods have the same fission products of a full-length rod penalizes the source quantity of fission products.

[p.18]

term by a large amount. The consequence of this requirement would be the source term has many pins more worth of inventory than it actually has.

Regulatory Position 3.2 includes the following statement: The NRC should clarify the level of Section 3.2 If it can be demonstrated that local power level, rate of fission gas release, and justification needed to confirm fission (Release

10. cumulative fission gas release remain less than the limiting co-resident UO2 fuel rod, then gas release rates and cumulative fission Fractions)

Table 3 and 4 steady-state fission product release fractions apply to fuel rod designs gas release for fuel designs containing

[p.18]

containing integral burnable absorbers (e.g., Gadolinia). integral burnable absorbers.

Section 3.2 A factor is presented to adjust the transient fission product release correlations for short- The NRC should present the factor for (Release lived isotopes. This factor is not prominently presented with the correlations and may be short-lived isotopes more prominently

11. missed by applicants or licensees. with the correlations presented in the Fractions)

[p.19] section.

Page 2

No. Section Comment/Basis Recommendation Section 3.2 contains the following guidance: Recommend that the current draft regulatory guide be updated to For the remaining non-LOCA DBAs which predict fuel rod cladding failure, such as PWR provide complete guidance for these reactor coolant pump locked rotor and fuel handling accident, additional fission product remaining accidents.

Section 3.2 releases may occur as a result of fuel pellet fragmentation (e.g., fracturing of high burnup (Release rim region) due to loss of pellet-to-cladding mechanical constraint or impact loads. TFGR 12.

Fractions) has been experimentally observed under a variety of accident conditions and should be

[p.19] addressed in future applications.

However, the regulatory guide does not provide guidance regarding an acceptable treatment for any additional fission product releases for these accidents. Therefore, the current guidance within the draft regulatory guide is incomplete.

The paragraph starting with For the remaining non-LOCA DBs which predict fuel rod Include background documents for cladding failure is not clear on the background of the issue or the success pathways further description of the issue and/or Section 3.2 afforded to the licensees to address the NRCs concerns on the topic in question. Ideally, documents identifying acceptable (Release background documents for further description of the issue and/or documents identifying methods of evaluation so that licensees 13.

Fractions) acceptable methods of evaluation are cited. The NRC should include these documents in can fully understand and properly

[p.19] order for the licensees to fully understand and properly address the issue in question. If address the issue in question. If no no documents exist, the NRC should develop these positions. documents exist, the NRC should develop these positions.

Page 3

No. Section Comment/Basis Recommendation Tables 1 and 2 present core inventory fraction releases into containment for BWRs and The NRC should clarify that the Te, Ba, PWRs. The release groups are updated from Revision 0 of RG 1.183. One significant and Sr groups do contribute releases difference is that the tellurium group (both PWRs and BWRs) and barium, strontium during the gap release phase for LOCA group (PWRs only) identify a release fraction during the gap release phase. releases in Tables 1 and 2; however, these releases are more accurately From page 34 of the SAND-2011-0128 report: At the same time, the Gap Release Phase classified as early in-vessel releases and was calculated to be long enough that some core degradation characteristic of the In- these groups do not need to be vessel Phase of release as prescribed in the NUREG-1465 Source Term did take place. This considered in gap releases for other is indicated by small amounts of tellurium release during the Gap Release Phase and in events.

Section 3.2 the case of PWR accidents small amounts of alkaline earth release. Ordinarily tellurium (Release and alkaline earths are not thought to be contributors to the gap release. They contribute 14.

Fractions) here because some portions of the core had entered into what would be categorized

[p.20] phenomenologically as in-vessel release before the criterion to terminate gap release had been reached."

This confirms that, while the timing of the Te, Ba, and Sr groups does release during the gap fraction phase, these releases are more accurately associated with the early in-vessel phase. The continued identification of these groups release as gap phase release creates confusion with subsequent gap fraction tables (Tables 3 and 4) which do not identify these nuclide groups as being released.

[Note that it is not requested to add Te, Ba, and Sr to the gap releases.]

Tables 1 and 2 report release fractions from a new Molybdenum group; however, the It is expected that the final grouping is Section 3.2 elements in this new group are not listed in Table 6. consistent with SAND-2011-0128 since Tables 1 & 2 that is the basis for the release fractions 15.

(p20), and Which release group is Zirconium (Zr) considered to reside? DG-1389 is consistent with in Tables 1 & 2. Update the Table 6 for Table 6 (p22) Revision 0 and indicates it is part of the Lanthanides while Table 14 in the underlying the new proposed Molybdenum group.

Sandia report (SAND 2011-0128) reports Zirconium as part of the Cerium group. Change Zr grouping to Cerium group.

Page 4

No. Section Comment/Basis Recommendation Figure 1 presents the maximum allowable power operating envelope for steady-state Include the following statement in a release fractions. The envelope is identified by rod average power (kW/ft) and rod footnote for Figure 1 or incorporated average burnup (GWD/MTU). The figure further identifies a Peak LHGR (Linear Heat into the discussion of Section 3:

Generation Rate) of 15.0 kW/ft for BWRs and 14.0 kW/ft for PWRs.

From Ref. 24, the Peak LHGR is defined From the Ref. 24 Technical Basis for Non-LOCA Fission Product Release Fractions, the Peak as the product of the peak fuel rod LHGR is a combination of rod average power and the axial power profile maximum average power and the peak fuel rod factors. For example, using a rod average power of 12.2 kW/ft from Figure 1 and an axial axial power distribution. This Peak LHGR Section 3.2 power profile maximum factor of approximately 1.15 for PWRs from Figure 2 of Ref. 24, may differ from the definition used in (Release

16. the resulting Peak LHGR would be approximately 14.0 kW/ft. licensees Technical Specification or Fractions)

Core Operating Limits Report. A Peak

[p.21]

The maximum linear heat rate is a typical term for licensees (generally defined in safety LHGR derived consistent with the analyses and/or Technical Specifications) and is the product of the rod average power (or definition of Ref. 24 should be used for linear heat generation rate determined by the core rated thermal power and the linear comparison of applicability to Figure 1.

component of all power producing rods in the core) and the hot channel factor (FQ).

These maximum linear heat rates can exceed the definition of Peak LHGR associated with Figure 1 of DG-1389. Without further explanation of the definition of Peak LHGR associated with the figure, misinterpretation and confusion of the Figure 1 envelope may result between the licensee and regulator.

The Figure 1 rod average power envelopes were derived in Ref. 24. Ref. 24 does not The NRC should confirm that indicate if uncertainties were applied to the rod average powers or normalized axial uncertainties do not need to be applied power distributions in Figure 2 of Ref. 24. to the power inputs used in comparing to the bounding power profile and It is assumed that uncertainties are not applied; however, the NRC should confirm that include this clarifying information in the Section 3.2 uncertainties do not need to be considered when comparing against the bounding power discussion in Section 3.2.

(Release

17. profile.

Fractions)

[p.21]

Note that for typical applications for determining release inventories, uncertainties may be accounted for in peaking factors such as the radial peaking factor (applied uncertainty), which is applied separately to the core inventory from the gap fractions in the radiological analyses. As such, applying uncertainties in determining the gap fractions may result in double-accounting and should not be advised.

Section 3.3 Table 5 in Section 3.3 lists the duration of the gap release phase is 0.22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> for PWRs The NRC should delete or correct the (Timing of and 0.16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> for BWRs. However, the last sentence of the following paragraph says: sentence in question as it contradicts

18. Regardless of delays in the onset, the duration of the gap release phase is 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. gap phase durations in Table 5.

Release Phases) [p. 22] The values in Table 5 are not consistent with this statement.

Page 5

No. Section Comment/Basis Recommendation DG-1389 states: The activity released from the core during each release phase should be Clarify that the statement in the modeled as increasing in a linear fashion over the duration of the phase. guidance excludes the effects of decay.

Section 3.3 19.

[p. 22] RADTRAD models core release as a constant fraction of the core inventory over the release duration. Including decay of the core inventory will make this release slightly non-linear, decreasing over time as the core inventory decreases due to decay.

Section 3.4 Table 6 presents the radionuclide groups which should be considered in design basis The NRC should present molybdenum as (Radionuclide analyses. The table presents molybdenum as included in the noble metals group. This is a separate group in Table 6 in order to

20. inconsistent with Tables 1 and 2 which present molybdenum as a separate group with maintain consistency with Tables 1 and Composition)

[p. 22] different release fractions than the noble metals group. 2.

DG-1389 includes a statement that the transit dose to personnel traveling to and from the The NRC should remove these control room should be considered for licensees whose licensing basis includes transit statements from the RG on the basis dose. This statement should be clarified or removed based on the following points: that transit doses would be addressed

• For licensees who do not currently include transit dose in their licensing basis, will by ALARA principles through Emergency the NRC force the adoption of transit dose if this regulatory guide is adopted? Planning. These measures may include

• For licensees who do currently include transit dose in their licensing basis, will the alternate travel pathways to and from NRC allow for removal of the transit dose based on the positions summarized the control room such that operators below: are not traversing the radioactive plume

• GDC 19 identifies that Adequate radiation protection shall be provided and personal protective equipment to permit access and occupancy of the control room. GDC 19 does not (e.g., respirators).

Section 4.2 identify that access includes the need for the control room to provide (Control Room protections outside of the control room envelope. Additionally,

21. Dose radiological consequences analyses are intended to provide an evaluation Consequences)

[p. 24]

of the design and performance of structures, systems, and components of the facility. This intent is to confirm the design, construction, and siting of the facility and applicable safety features are adequate to limit dose exposure; thus, operator actions outside of the CR envelope are not historically considered in the design basis analyses.

• During emergency situations, personnel dose is governed by ALARA principles such that Emergency Planning will limit the dose received by operators during transit to and from the control room. 10 CFR 50.47(a)(9), (10), and (11) provide sufficient requirements for the Emergency Response organizations to measure and limit doses to emergency response personnel including control room operators.

Page 6

No. Section Comment/Basis Recommendation The occupancy factors in RG 1.183 R0 are based on RG 1.183 R0 Reference 22. Reference Remove the following discussion as it is 22 is the paper by K. G. Murphy and K. W. Campe, "Nuclear Power Plant Control Room unnecessary:

Ventilation System Design for Meeting General Design Criterion 19," published August 1974. In that paper, the occupancy factors are described as follows: The following guidance should be used in determining the TEDE dose for

"...an allowance may be considered for the time the operator leaves the plant vicinity. demonstrating compliance with 10 CFR This is described as the occupancy factor." 50.67(b)(2)(iii). For the purpose of this RG, a transit dose is considered to be Table 1 in the paper by K. G. Murphy and K. W. Campe provides the occupancy factors the dose that is accumulated as included in the determination of the X/Q values using the K. G. Murphy and K. W. Campe personnel travel to and from the control methodology. room for the duration of an accident once onsite (e.g., dose from site Therefore, it is reasonable to conclude the occupancy factors (1, 0.6 and 0.4) assumed for boundary to the control room).

Section 4.2 the control room analysis already implicitly account for the time for the operator travel Licensees whose licensing basis includes (Control Room from the parking lot to the control room. The occupancy factor discussion is in the X/Q transit dose should include the transit

22. Dose section (page 13) which is consistent with this argument. No additional discussion is dose to demonstrate compliance with Consequences required with those occupancy factors. 10 CFR 50.67(b)(2)(iii). The licensees

) [p. 24]

results for the evaluation of transit dose Further the Murphy and Campe paper interprets the GDC in this way: Whole body will be evaluated on a case-by-case gamma radiation from direct shine radiation sources external to the control room and basis. New reactor licensees that are from the airborne activity within the control room should not exceed a total of 5 rem. required to show compliance with GDC Direct shine here is from outside radioactivity directly to operators within the control 19 or similar control room radiological room. This would further exclude a transit dose for the operators. habitability principal design criteria may use this guidance.

As well, in NUREG-0800, Section 6.4, referring to individuals within the control room says:

In accordance with GDC 19, these doses to an individual should not be excluded for any postulated accident. The whole body gamma doses consists of the contributions from airborne radioactivity inside and outside the control room, as well as direct shine from all radiation sources. Industry has general evaluated the other sources in terms of containment shine and filter shine.

Page 7

No. Section Comment/Basis Recommendation Footnote 15 states that nuclides used for modeling dose from airborne radioactivity The NRC should provide clarification for inside the control room may not be conservative for determining the dose from Footnote 15 in terms of which nuclides radioactivity outside of the control room. are of concern for the different aspects Section 4.2.3 of the control room dose. Lack of (Transport

23. This statement is vague and does not provide the licensee with details as to which explanation infers that the NRC accepts Models) [p.

nuclides the NRC is concerned about for the different control room dose aspects the licensees currently analyzed source 25]

(inhalation/immersion versus plume shine dose). Without further explanation, it is terms and no updates are needed.

assumed that the NRC is in agreement that the sets of nuclides used by licensees acceptably addresses the footnote.

This section provides the expression to correct the semi-infinite cloud dose to a finite The NRC should note that the expression Section 4.2.7 cloud dose for external exposure. This expression is generally incorporated into may be incorporated into radiological (Dose radiological consequences codes (e.g., RADTRAD) and does not specifically need to be consequences codes (e.g., RADTRAD) 24.

Conversion considered as an additional factor applied to the control room dose results. and does not need to be specifically Factor) [p. 26] applied to the control room dose results.

Table 7 provides the analysis release duration for the various accidents presented in the For consistency with the event-specific regulatory guide. The following items should be addressed by the NRC: guidance presented in Appendices E, F,

• The PWR Steam Generator Tube Rupture, PWR Main Steamline Break, and PWR and G, change wording from the analysis Locked Rotor identify that the analysis release duration is Until cold shutdown is release duration is Until cold shutdown Section 4.4 established. For consistency with the event-specific guidance presented in is established to Until shutdown (Acceptance Appendices E, F, and G, this wording should be changed to Until shutdown cooling is in operation and releases from 25.

Criteria) [p. cooling is in operation and releases from the SG(s) have been terminated. the SG(s) have been terminated.

28]

• For Fuel Handling Accident, the analysis release duration should be updated for consistency with the updated FHA model presented in Appendix B. The updated In the Fuel Handling Accident, update model presents releases in two phases which potentially span the entire standard the analysis release duration with the 30 days of event duration. updated FHA model presented in Appendix B.

Page 8

Guidance for the modeling of the limiting X/Qs for the LPZ and control room is presented The NRC should re-assess the need to and directs the licensee to model the period of most unfavorable atmospheric dispersion include this guidance for atmospheric factors coincident with the time period of most adverse environmental release. This dispersion factor modeling. The guidance has been included previously for the control room location in RG 1.194 (and guidance, as presented, may result in a identified in the 2015 periodic review of RG 1.194 as an item for alignment). However, the misappropriation of the limiting two-guidance has not been previously incorporated in or identified as needed for RG 1.183. hour X/Q values for the time period of

[Also, no identification of this guidance has been made for RG 1.145 for the LPZ location.] greatest release as this may not align with the time period of greatest dose The current widely used practice for modeling X/Q values for the LPZ and control room is contribution for the control room.

to align the values in correct time period order, e.g., 0-2 hrs, 2-8 hrs, etc. (similar to the first portion of Figure 2 in DG-1389). However, it is recognized that this may not align with It is recommended that the NRC remove the period of most adverse environmental release in instances of delayed releases from this position from DG-1389 and maintain damaged/melted fuel (in cases of an MHA) or prolonged buildup and transit from the the current practice of applying LPZ and primary to secondary systems to the environment (in cases of a locked RCP rotor). CR values from highest X/Q at event initiation (T=0) for all events, regardless The interpretation of the guidance presented in DG-1389 is that the period of most of release magnitude. It is also Section 5.3 adverse release aligns with the period of highest dose. However, the NRC has not recommended that this position be (Atmospheric provided sufficient guidance for consideration of situations in which the period of most removed from RG 1.194.

Dispersion adverse environmental release does not align with the period of highest dose. This may

26. Modeling and be the case for instances in the control room depending on specific modeling Meteorology assumptions.

Assumptions)

[p. 30-32] For instance, for a PWR MHA, a staged release is modeled in accordance with Table 2 of RG 1.183. This results in the limiting environmental release typically occurring after the initial 0-2 hour time period. The control room is typically isolated early in the event, accounting for some delay, such that only gap activity enters the control room envelope during unfiltered, normal ventilation mode. Therefore, during the period of greatest environmental release, the activity entering the control room would be subject to emergency mode flows (possibly lower than normal mode), filtration (~90-95% iodine and particulates), and lower X/Q values associated with emergency mode intakes (depending on the receptor positioning). As a result, applying the proposed guidance would decrease the control room dose consequences relative to the current practices.

Additionally, no guidance has been provided for cases in which multiple release pathways are modeled with varying periods of limiting environmental releases and/or dose.

For simplicity, and to avoid potential non-conservative applications of the guidance, it is recommended that the NRC remove this position from DG-1389 and maintain the current practice of applying LPZ and CR values from highest X/Q at event initiation (T=0) for all Page 9

No. Section Comment/Basis Recommendation events, regardless of release magnitude. It is also recommended that this position be removed from RG 1.194.

This section indicates that the aerosol deposition models in NUREG/CR-6189 are still Confirm the continued applicability of Appendix A, applicable; however, based on Section 1 of NUREG/CR-6189, the models are based on the the NUREG/CR-6189 aerosol removal Section A-2.2,

27. release fractions and timing in NUREG-1465. Considering the significant changes to the rates.

Page A-2 release fractions and timings in DG-1389, there may be significant impacts to the

[p.39]

deposition rates in this NUREG.

Revise A-2.5 to allow credit for suppression pool scrubbing based on: Change A-2.5 to reference NUREG/CR-NUREG/CR-6153 - provides models for accident dose calculations using the AST 6153 and allow credit for reduction in for the purposes of crediting iodine decontamination provided by suppression airborne radioactivity in the Appendix A, pools containment by suppression pool Section A-2.5,

• Accounts for the changing aerosol distribution following passage through scrubbing in BWRs.

28.

Page A-3 the pool

[p.40] State-of-the-Art Reactor Consequence Analyses (SOARCA) project results (ML20304A339- NRC Brochure) include the suppression pool in their models and indicate that all modeled accident scenarios, progress more slowly and release smaller amounts of radioactive material than calculated in earlier studies.

For gas-filled secondary containment bypass leakage paths, this section states deposition Add deposition of aerosol radioactivity Appendix A, of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis. and plate-out of elemental halogens Section A-3.5,

29. Plate-out of elemental iodine may also be a significant removal mechanism in gas-filled Page A-4 secondary bypass leakage pathways. In addition to aerosol deposition, the plate-out of

[p.41]

elemental iodine should not be excluded.

Multiple BWRs currently have credit for aerosol removal from drywell sprays as well as Provide a model where it is acceptable Appendix A, aerosol deposition within in the main steam lines (some also have Condenser removal) in to credit aerosol removal from drywell A-5. Main their current licensing basis. Section A-5 presents three acceptable methods for sprays as well as aerosol deposition Steam calculating aerosol deposition within the main steam lines, but this section states, within in the main steam lines.

Isolation Valve however, these methods are not valid if credit has been taken for aerosol removal

30. Leakage in from drywell sprays.

Boiling-Water Reactors, Given the prevalence of credit for both sprays and steam line deposition, why is there not Page A-6 a model presented that the Staff finds acceptable for crediting both, or modifications to

[p.43] the presented models if the licensee wants to credit spray removal (e.g., different aerosol size distribution)?

Page 10

No. Section Comment/Basis Recommendation Appendix A, What is the technical justification for drywell sprays aerosol removal not being valid if Revise words to state that the impact of A-5. Main used alongside main steam isolation valve leakage? The removal mechanisms associated sprays on aerosol removal should be Steam with spray described in NUREG/CR-5966 are largely different than the removal determined in the submittal, and do not Isolation Valve mechanisms associated with deposition described in SAND2008-6601 and AEB 98-03. state that removal by both sprays and

31. Leakage in Section 6.2 of SAND2008-6601 states that sprays should not be used in conjunction with deposition should not be used.

Boiling-Water main steam line deposition, but this discussion is based on the steam dome being the Reactors, source of radioactivity rather than the accepted position of the drywell being the source Page A-6 of radioactivity.

[p.43]

Appendix A, DG-1389 states that the reported 3 aerosol methods are not valid if credit has been taken Confirm acceptability of applying A-5. Main for aerosol removal by drywell sprays. Are other aerosol removal mechanisms similarly NUREG/CR-6189 aerosol deposition in Steam affected? Can a licensee credit the natural removal mechanisms in NUREG/CR-6189 addition to reported MSL deposition Isolation Valve consistent with Section A-2.2 and apply the reported MSL models? Although the models.

32. Leakage in deposition models in NUREG/CR-6189 may affect the aerosol size distribution, this impact Boiling-Water would be expected to be similar to the impact of the MSL models themselves such that Reactors, the aerosol distribution entering the MSLs are not significantly different from that Page A-6 assumed in the models.

[p.43]

DG-1389 reports the first MSL aerosol deposition model as: Direct adoption of the SAND Revise first aerosol deposition model to 2008-6601 (Ref. A-11) recommendations without scaling R*-factors; without scaling RM and R* factors.

Appendix A, This approach needs additional explanation. As defined in Section 1.1 of SAND2008-6601, A-5. Main the R* factor is defined as the ratio of NUREG-1465 containment airborne concentrations Steam to MELCOR containment airborne concentrations. A separate factor, RM, models the ratio Isolation Valve of the steam dome concentration to the drywell concentration determined by the

33. Leakage in MELCOR full plant analyses. Backing the R* factor out of the process does not appear to Boiling-Water be possible since, per Section 5.2 of SAND2008-6601, the RM and R* factors are combined Reactors, and their product is applied to develop the results in Table 5-3.

Page A-6

[p.43]

This section should adopt the MHA approach where the drywell and steam dome are well-mixed. As such, there should be no scaling factor applied to model the increased concentration in the steam dome.

Page 11

No. Section Comment/Basis Recommendation It is understood that there are several deposition and removal mechanisms (see Table Change footnote 4 to state that 4.1-3 of the State-of-the-Art Report for examples) present inside of steam lines and it is previously approved methods are not challenging to provide a comprehensive model due to lack of experimental data and superseded by the RG 1.183 Revision 1 disassociation between the reality of core cooling being maintained during a design basis methods if unaltered during submittals accident and fuel damage being assumed per DG-1389. The State-of-the-Art Report associated with licensing actions Appendix A, attempts to capture all of these uncertainties but cannot and does not form a conclusion unrelated to steam line removal.

34. A-5.6, Page A-based on all of the available literature.

9 [p.46]

Based on previous approval of AST applications and the lack of concrete knowledge surrounding removal in steam lines, it is recommended to allow licensees to continue to use their existing main steam line removal models if pursuing licensing actions unrelated to the steam lines themselves.

Aerosol deposition in vertical volumes is non-zero. SAND2008-6601 Section 3.3 states that Remove in horizontal volumes or some deposition occurs in vertical surfaces. This is confirmed by Figure 2-16 and Figure provide different guidance for horizontal Appendix A, 2-17 of the State-of-the-Art Report which show strong correlations between temperature vs. vertical volumes when applying

35. A-5.6, Page A-driven deposition by thermophoresis and condensing vapor deposition by Method 1: Direct adoption of the SAND 9 [p.46]

diffusiophoresis. These removal mechanisms are largely independent of gravitational 2008-6601 recommendations without settling in horizontal segments. scaling RM and R* factors.

Section A-5.6 implies that only plants other than Mark I, II, or III will be considered on a Revise statement to allow case by case Appendix A, case-by-case basis. All models should be considered on a case-by-case basis considering review for all technologies.

36. A-5.6, Page A- the models described by positions A-5.6.1, A-5.6.2, and A-5.6.3 do not consider all 9 [p.46] removal mechanisms present in the steam lines.

Section A-5.6 indicates that the multi-group method will be evaluated on an individual Confirm that the approaches in Sections case-by-case basis. A-5.6.2 and A-5.6.3 are acceptable Appendix A, models for aerosol deposition in the

37. A-5.6, Page A- This statement implies that the application of this method may not be fully approved, MSLs.

9 [p.46] thereby increasing the potential for regulatory uncertainty in future submittals. Why does the approach described in Section A-5.6.2, including an AMMD of 2.0 µm, 2,000 groups, and 10,000 trials, require additional regulatory review?

Page 12

No. Section Comment/Basis Recommendation The basis for not being able to credit the piping upstream of the inboard MSIV seems to Revise table to be consistent with come from Section 6.3 of SAND2008-6601 which states that at times in the simulation SAND2008-6601 and add in calculated the temperature of portions of the in-board MSL piping are predicted to be high enough removal for in-board lines.

to vaporize fission products that had been previously deposited. The MELCOR simulation Appendix A, in SAND2008-6601 is not representative of plant specific thermo-hydraulic conditions.

38. A-5.6.1, Page (e.g., Figure 2-19 of SAND2008-6601 shows a long term temperature of approximately A-9 [p.46] 800F in the steam dome while most BWR analyses would show long term temperatures of less than 300F in the core with ECCS operational). Credit should be able to be taken for deposition in the inboard lines if plant specific analysis shows low temperatures considering RIS 2006-04 states that deposition of particles in the inboard volume occurs.

In addition, the time periods in DG-1389 are inconsistent with SAND2008-6601.

Calculations typically consider a single failure which is either assumed to be a break in a Clarify that there is no in-board pipe Appendix A, steam line leading to not modeling holdup/deposition in a steam line upstream of the break in the MHA scenario.

A-5.6.2, Page inboard MSIV or assumed to be a stuck open MSIV which leads to combining the inboard 39.

A-9 to A-11 volume with the between MSIV volume in a single steam line. The discussion states that a

[p.46-48] pipe break is not assumed, so should all applicants consider the limiting single failure to be a stuck open MSIV (with respect to deposition in the steam lines)?

The State-of-the-Art report recommends a AMMD of 3 µm for containment and Section Revise the text to apply a 3.0 µm AMMD A-5.6.2 describes the approved approach as applying an aerodynamic mass mean or explain the basis for the suggested Appendix A, diameter (AMMD) of 2.0 µm. This appears to be an average value of the RCS and value of 2.0 µm.

A-5.6.2, Page

40. containment AMMD values of 1.0 and 3.0 µm respectively. Considering the MHA A-9 to A-11 assumption of a well-mixed drywell, the containment AMMD of 3.0 µm would be the

[p.46-48]

applicable parameter on which to base the aerosol size distribution entering the main steam lines.

These methods are based on the AEB 98-03 methodology which does not account for Clarify that other associated removal Appendix A, thermophoresis, impaction, diffusiophoresis, flow irregularities, and hygroscopicity. The mechanisms may be calculated A-5.6.2 and A-removal calculated is largely correlated to particle size but because it does not include separately and included in the TEARE.

41. 5.6.3 Page A-9 these other mechanisms it is overly conservative. For example, Section 2.6.3 of the State-to A-12 [p.46-of-the-Art Report states that thermophoretic deposition velocity is not an especially 49]

strong function of particle size.

Page 13

No. Section Comment/Basis Recommendation Section A-5.6.4 indicates that aerosol deposition in the condenser using a multi-group or Clarify use of Table A-1 condenser numerical integration approach needs to be evaluated on an individual case basis. deposition coefficients or clarify why the multi-group method is not applicable to Appendix A, Why would the multi-group or numerical integration approaches not be applicable in the the condenser.

42. A-5.6.4, Page condenser? These models consider the impact of easier-to-remove particles being A-12 [p.49]

removed in the upstream compartments, leading to less deposition in the condenser. Can the condenser deposition coefficients in Table A-1 be applied instead of a value developed with the multi-group or numerical integration approaches?

Footnote 1 in Appendix B presents that if a postulated event (heavy load drop) occurs for Replace the footnote wording with the an FHA, the activity release may be based on core average gap fractions presented for the following to clarify the gap fractions and LOCA in Tables 1 and 2. The following issues should be addressed by the NRC: timings to apply for the heavy load drop accident:

• Tables 1 and 2 identify gap fraction releases for the Te, Ba, and Sr groups. As identified in a previous comment, the SAND-2011-0128 report, in which Tables 1 These assumptions may also be used in and 2 are based, clarified that these releases listed in the gap phase were more assessing the radiological consequences accurately classified as in-vessel releases that occurred during the gap release of a heavy load drop over fuel accident.

phase. As a heavy load drop event is a shutdown event and fuel damage should If the event is postulated to damage all only result from a load impact, there would not be a driver for further fuel failure of the rods in the core, the release Appendix B associated with in-vessel releases (which results from fuel melt). Therefore, these activity may be based on the core-(Fuel Handling

43. nuclide groups should be excluded from consideration in the fuel handling average gap fractions of Tables 1 and 2 Accident) [p.

accidents involving the entire core. in the main text of the guide, and the 52]

• The Tables 1 and 2 release fractions are associated with the timings presented in radial peaking factor may be omitted.

Table 5. Table 5 presents gap phase durations of 0.22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> (PWRs) and 0.16 Gap fractions may be limited to the hours (BWRs). The timings in Table 5 are based on simulations of various plant noble gas, halogen, and alkali metal transients which do not include fuel handling accidents / heavy load drop events. groups, consistent with Tables 3 and 4.

Therefore, it should be specified that the use of Tables 1 and 2 do not need to Additionally, the release timings and explicitly follow the gap phase durations from Table 5 and can instead use the durations associated with this appendix durations presented in this appendix (Appendix B). may be used in lieu of release timings and durations associated with Regulatory Position 3.2 for Tables 1 and 2.

Page 14

No. Section Comment/Basis Recommendation In the current licensing basis fuel handling accidents, the impacts of stable and long-lived Confirm the applicability of Tables 3 and iodine isotopes (e.g., I-127 and I-129) are not typically considered due to their negligible 4 for I-127 and I-129 as other dose consequences. For the revised fuel handling accident approach, these isotopes are halogens.

Appendix B, B- important to develop the pool iodine concentration as they represent most of the iodine

44. 1.3, Page B-1 inventory in the fuel rods.

[p. 52]

The NRCs assessment in the staffs example in ML21190A040 appears to apply a much larger value of 23% for Iodine-129 which is significantly higher than the 4% from Table 4 for other halogens.

DG-1389 only reports the iodine species distribution released into the pool water. Reg Provide additional guidance on Guide 1.183 Rev. 0 reported that the iodine chemical species above the water is 57% acceptable airborne iodine species elemental and 43% organic. assumptions.

For Rev. 1 applications, what iodine species distribution should be applied for the early 2-

45. Section B.1-3 hour airborne release considering the new fuel handling accident model? Based on the calculated overall pool DF from Equations 3-1, 3-2, and 3-3, is a species-dependent DF approach acceptable with organic iodine having a DF of 1?

Is a purely elemental iodine species appropriate for the long-term release since this is comprised of re-evolved elemental iodine?

There can be fuel handling accident scenarios where the water depth may be greater than Revise statement as follows:

23 feet such as drops over the core. However, Section B-2 states that the DF can be calculated from Equations in B-1, B-2, and B-3, if the water depth is between 19 and 23 For water depths greater than or equal Appendix B, B-feet. There is no guidance in the event the water depth is greater than 23 feet. to 19 feet, an overall iodine DF based on

46. 2, Page B-2 [p.

pin pressure is computed as follows:

52]

Since the DG-1389 equations are not dependent on depth, it would be expected that they would yield a conservatively low DF than an actual value for cases that credit the additional scrubbing depth.

Appendix B-3 The mass transfer coefficient (KL) is presented as a constant value (3.66x10-6 m/s). From The NRC should clarify that the mass (Phase 2 Enclosure 4 of Reference B-1, this is a combined liquid-gas phase coefficient that does not transfer coefficient remains applicable Release - Re- consider recirculation in the pool and is based on the assumption of a high flow rate to for all conditions (i.e., recirculation in 47.

evolution clear the building air. the pool and low flow rate to clear the Release) [p. building air) and no adjustments need to 54, B-3] be made for changing conditions.

Page 15

No. Section Comment/Basis Recommendation Step 2 of the FHA model instructs the analyst to calculate the amount of iodine in the fuel The NRC should specify that additional pin gap. The calculations consider I-131 for the amount of radioactive iodine (with non-radioactive iodine isotopes subsequent clarification to include other radioactive iodine isotopes to the total) and (especially I-127) should be considered consider I-129 for the amount of non-radioactive iodine (with no subsequent clarification to maximize the amount of iodine Appendix B-3 to include other non-radioactive iodine isotopes). The radioactive and non-radioactive considered in Step 2 of the FHA model (Phase 2 iodines are then combined for a total iodine which is used in the downstream calculation.

Release - Re-

48. calculations. [Note that the individual components are not used in any downstream evolution calculations within the model and calculations involving only the radioactivity iodine can Release) [p.

be removed from the model development.]

54, B-3]

It was observed in the calculations that a higher total iodine amount is conservative for calculating the release rate to be used in the analysis. As such, additional non-radioactive iodine isotopes (I-127) should conservatively be considered.

Equation B-10 in Step 2 calculates the [I2]/[I-]2 concentration ratio using the following The NRC should provide clarification as Appendix B-3 equation: to the discrepancy in constants between (Phase 2 R.i = [hJ/[I-]2 = ch2 1 6.0603E-14 + 1.4 osE-09 ch 1 NUREG/CR-5950 and DG-1389 or update Release - Re- Equation B-10 to present the constants

49. Constants 6.0603E-14 and 1.4708E-09 originate from Ref. B-2 (NUREG/CR-5950);

evolution to the degree of precision from Release) [p. however, the constants are not presented in the reference to that degree of accuracy.

NUREG/CR-5950.

55, B-4] The constants presented in NUREG/CR-5950 are 6.05E-14 and 1.47E-09 [NUREG/CR-5950 page 13 and Appendix C, page C.3]. It is unclear where the discrepancy is occurring.

Position B-5.1 provides a release duration time for the first phase releases (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). The The NRC should clarify the starting time Appendix B-5 position does not provide any specifics for onset and duration timing for the second and duration of the re-evolution phase.

(Fuel Handling phase releases (re-evolution releases). From the Ref. B-1 model background, this release Accidents phase should begin immediately after the first phase (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and be modeled until the 50.

within the typical end of the radiological considerations (30 days). It is assumed that there is no Fuel Building) overlap of release phases.

[p. 56, B-5]

Note that this comment is also applicable to Position B-6.3.

Appendix B-5 Footnote 4 breaks across pages B-5 and B-6. This footnote should be contained on a single See comment for the items to address.

(Fuel Handling page.

Accidents

51. within Note that this footnote is referenced in Positions B-6.1 and B-6.2 which are currently on Containment) separate pages, so Footnote 4 may have to be repeated on separate pages.

[p. 56-57, B 6] Likewise, Footnote 3 for Position B-6.4 is not located on the same page as the position.

Page 16

No. Section Comment/Basis Recommendation Appendix E Revision 0 of RG 1.183 presented the PWR Main Steamline Break event in Appendix E. The NRC should consider maintaining (Steam There is no indication as to the motive behind the switch in appendices for DG-1389. This consistency in the event order

52. Generator may cause confusion to licensees which need to compare/contrast the RG revisions or presented in the regulatory guide Tube Rupture) those that elect to implement and maintain different RG revision methods for these appendices.

[p. 63] events.

Appendix E The page numbering for Appendix E is not correct as the first page of the appendix is The NRC should correct the page (Steam identified as page E-3. number in Appendix E and also review

53. Generator the entirety of the regulatory guide to Tube Rupture) correct any further editorial errors.

[p. 63]

Position E-6.1 states that the primary-to-secondary leak rate in the steam generators The NRC should add a statement that should be assumed to be the leak rate for the limiting condition for operation specified in licensees may credit a reduction in the the TSs. This leak rate is associated with significant pressure differentials across the steam TS leakage rates for later periods of the Appendix E-6 generator tubes which typically occur early in the transient. In later periods of the transient when pressure differentials (Transport -

transient, and especially during longer-term steaming for cooldown, the pressure across the steam generator tubes are Steam

54. differential is significantly decreased. Licensees should be able to credit this significant significantly reduced based on adequate Generator reduction in leakage if it is shown to be credible. technical justifications.

Tube Rupture)

[p. 64]

This comment is also applicable for other steam release events which model primary-to-secondary leakage. [MSLB in Appendix F, Locked Rotor in Appendix G, and Rod Ejection in Appendix H]

Page 17

No. Section Comment/Basis Recommendation Position E-6.6 identifies that the potential impact of tube uncovery on the transport The NRC should update the wording of model parameters needs to be considered. this position to acknowledge for U-tube steam generators that the issue of The issue of tube uncovery was addressed by the Westinghouse Owners Group (WOG) in short-term tube uncovery has previously WCAP-13247, "Report on the Methodology for the Resolution of the Steam Generator been resolved and does not need to be Tube Uncovery Issue," March 1992. The WOG program concluded that the effect of tube considered in future applications.

Appendix E-6 uncovery would be essentially negligible and the issue could be closed without any (Transport - further investigation or generic restrictions. This position was accepted by the NRC in a Steam letter dated March 10, 1993, from Robert C. Jones, Chief of the Reactor System Branch, to 55.

Generator Lawrence A. Walsh, Chairman of the WOG. The letter states " ... the Westinghouse Tube Rupture) analyses demonstrate that the effects of partial steam generator tube uncovery on the

[p. 66] iodine release for SGTR and non-SGTR events is negligible. Therefore, we agree with your position on this matter and consider this issue to be resolved." Consistent with this position, tube uncovery should not need to be addressed further for U-tube style steam generators.

This comment also applies to Positions F-6.6.3 for MSLB, G-5.5 for Locked Rotor, and H-7.4 for Rod Ejection.

Appendix I is not referenced within DG-1389 and the purpose and importance of this NRC should delete this appendix or Appendix I [p. appendix is not understood. Given the unspecified use and importance of this appendix, it provide reference in document as to

56. is recommended the NRC remove the appendix and flowchart. why this is needed.

76]

The first paragraph of this appendix cross-references Section C of the main body of this The NRC should correct this cross-Appendix J [p. guide for the release fractions in Tables 3 and 4. This is an editorial error as Tables 3 and 4 reference and review DG-1389 for any

57. are located in Section 3.2. other potential cross-referencing errors.

77]

Section C position 3.2.

Page 18