ML20071G486

ABWR DC Renewal Supplemental FSER Section 6.2.1.3 Short-Term Pressure Response OGC Final

ML20071G486
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Site:	05200045
Issue date:	03/31/2020
From:	James Shea NRC/NRR/DNRL/NRLB
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Shea JJ / 415-1388
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ML20071F088, ML20105A349	List: ML20071G486
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6-i 6 ENGINEERED SAFETY FEATURES Appendix A, Design Certification Rule for the U.S. Advanced Boiling Water Reactor, to Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, constitutes the standard design certification (DC) for the U.S. Advanced Boiling Water Reactor (ABWR) design. To document the U.S. Nuclear Regulatory Commission (NRC) staffs review supporting initial certification of the ABWR, the staff issued a final safety evaluation report (FSER) in NUREG-1503, Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, in July 1994 and NUREG-1503, Supplement 1, in May 1997.

The staff is documenting its review of the GE-Hitachi Nuclear Energy (GEH or the applicant) application for renewal of the ABWR DC in Supplement 2 to NUREG-1503. Chapter 1 of this supplemental FSER describes the staffs review process for the ABWR DC renewal. This supplemental FSER section documents the NRC staffs review specifically related to Chapter 6, Engineered Safety Features, Section 6.2.1.3, Short-Term Pressure Response, of the GEH Design Control Document (DCD), Revision 7. Except as modified by this supplement to the FSER, the findings made in NUREG-1503 and its Supplement 1 remain in full effect.

6.2.1.3 Short-Term Pressure Response 6.2.1.3.1 Regulatory Criteria The applicant for the ABWR DC renewal, completed design changes to the certified ABWR DCD in Revision 7, after identifying an error in the containment peak pressure analysis as discussed in a letter from GEH dated June 8, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100640164). In Enclosure 1 of the letter dated December 7, 2010, transmitting its application to renew the ABWR DC (ADAMS Accession No. ML110040176), the applicant stated, in part, the following:

the containment peak pressure re-analysis complies with NRC regulations that were in place at the time of certification, as required by 10 CFR 52.59(a), the amendment also complies with current applicable NRC regulations. GEH expects that the applicable regulations will remain the same during the NRC review of the application. However, if the NRC amends those regulations during the time period of its review, GEH will review such amendments to determine if any further changes are necessary.

The staff assessed the design changes associated with the containment peak pressure reanalysis and determined that some of the changes would meet the criteria for modifications while others would be identified as amendments, as these terms are defined in Chapter 1 of this FSER supplement. However, due to the interrelationship of the design changes, the staff decided to treat all the changes as amendments to the certified design and will correspondingly evaluate the changes using the regulations applicable and in effect at renewal.

GEHs statement above regarding compliance with current regulations supports this decision.

In addition, the staff determined that the pertinent requirements in current regulations and associated staff guidance for the review of the changes are not substantially different than the regulations and associated guidance in effect at the time of the original ABWR DC. Therefore, 6-1

by conducting the review against current regulations, the staffs evaluation also supports a finding of compliance with the applicable regulations in effect at initial certification.

The NRC staffs requirements for its review are specified in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, (GDC) 16, Containment Design, and GDC 50, Containment design basis, as they relate to the containment and its associated systems being able to accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. In NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Section 6.2.1.1.C, "Pressure-Suppression Type BWR Containments," Revision 7, issued March 2007, provides guidance on acceptable analytical models for calculating the containment peak pressure and temperature.

6.2.1.3.2 Summary of Technical Information In the ABWR DCD, Revision 6 submittal, GEH included the following changes from the original ABWR certification, incorporating changes contained in GEHs response to request for additional information (RAI) 06.02.01.01.C-1, Revision 1 dated August 11, 2015 (ADAMS Accession No. ML14239A137):

• a change in decay heat curves assumed for the long-term containment analysis from nominal values in the 1979 version of the American National Standards Institute (ANSI)/American Nuclear Society (ANS)-5.1, Decay Heat Power in Light Water Reactors to the 1994 version with a two standard deviation uncertainty on decay heat

• containment vent system modeling changes to include the drywell connecting vent (DCV) loss coefficients to correct the modeling of horizontal vents

• the feedwater line break (FWLB) flow changes to remove the initial 3.75-second inventory depletion period in the original DCD Tier 2, Figure 6.2-3

• a change in the suppression pool water level assumed for the long-term containment response analysis from 7 meters (equivalent to a volume of 3,580 cubic meters) to 6.9 meters (equivalent to a volume of 3,455 cubic meters)

• a change in the residual heat removal system (RHR) heat exchanger overall heat transfer coefficient assumed for the long-term containment response from 3.7x105 watts per degree Celsius (W/°C) to 4.27x105 W/°C (an approximately 15 percent (%) increase)

• Wetwell design temperature change from 104°C to 124°C

• negative pressure design evaluation changes including (1) eliminating analyses for events with inadvertent initiation of containment (drywell/wetwell) spray during normal operation, (2) taking credit for heating of emergency core cooling system (ECCS) flow in the reactor pressure vessel before being discharged into the drywell, and (3) using the GEH SUPERHEX computer code instead of the previous analyses, which used a series of end-point calculations to generate a set of conditions that produces a bounding prediction of the peak negative [wetwell to reactor building] differential pressure 6-2

In Supplements 1 and 2 of the applicants response to RAI 06.02.01.01.C-1, Revision 1 dated May 6, 2016 (ADAMS Accession No. ML16127A032) and Revision 2 dated June 22, 2016 (ADAMS Accession No. ML16174A179), respectively, the applicant made changes to DCD Tier 2, including the following as ABWR DCD, Revision 6, markups:

• adding text in DCD Tier 2, Section 5.4.7.3.2, Worst Case Transient, to state that [t]he normal shutdown condition is used to establish the limiting heat exchanger capacity and is evaluated in Appendix 5B.3

• replacing text in DCD Tier 2, Section 5.4.7.3.2, where rather than stating that RHR heat exchanger size was established to limit the suppression pool peak temperature to 97°C, the text will instead state that the heat exchanger size will also support the safety function of limiting suppression pool peak temperature to 97°C

• In DCD Tier 2, Table 6.2-2, Containment Design Parameters: the vent loss coefficient (VLC) is changed from 2.5 - 5.0 to 4.2 - 6.7 and a footnote is added to the table to state that the overall vent system loss coefficient includes a contribution from flow loss coefficient of 1.7 for DCV 6.2.1.3.3 Technical Evaluation The staff reviewed the final ABWR DCD, Revision 7, changes in DCD Tier 2, Sections 6.2.1 and 6.2.2 to determine compliance with GDC 16 and 50, using the guidance in SRP Section 6.2.1.1.C, Revision 7, issued March 2007. The staff determined that additional information was needed to complete its review and issued RAI 06.02.01.01.C-1, dated April 24 (ADAMS Accession No. ML14114A566). GEH responded in a letter dated August 27, 2014 (ADAMS Accession No. ML14239A137), which it revised and replaced by the letter dated August 11, 2015. GEH supplemented its response further in the letter dated June 22, 2016. of the DC renewal application dated December 7, 2010, the applicant made DCD changes to correct the containment peak pressure analysis to reflect a more limiting line break that GEH identified and discussed in the letter dated June 8, 2009. The limiting line breaks for the short-term accident response did not change from the certified design to the revised design.

However, for the long-term accident response, revisions to FWLB analysis resulted in a change to the drywell peak pressure and revisions to the main steamline break (MSLB) analysis resulted in changes to the drywell peak temperature. The June 8, 2009, letter, refers to NEDO-33372, Advanced Boiling Water Reactor (ABWR) Containment Analysis," which was later withdrawn from NRC topical report review by letter dated March 30, 2010 (ADAMS Accession No. ML100890313). As such, the staff was not clear about the documentation supporting the ABWR design certification renewal application changes to DCD Tier 2, Sections 6.2.1 and 6.2.2.

Therefore, in Part (1) of RAI 06.02.01.01.C-1, the staff requested GEH to provide documentation supporting the containment reanalysis changes of the ABWR DCD.

In its response dated August 11, 2015, GEH stated the following:

There are no new documents that have been issued or new references cited that were required to support the changes for the DCD revision. Although NEDO-33372 is no longer directly applicable to the ABWR for the reasons discussed above, there is certain information that remains applicable to the 6-3

ABWR renewal application. Therefore, rather than revise NEDO-33372, the information is proposed to be included in the ABWR DCD.

GEHs response identified two major and four minor changes associated with containment analysis. Major changes were associated with the decay heat used for the long-term containment analyses and modeling of the containment vent system. Minor changes were associated with FWLB flow, suppression pool volume margin, the overall heat transfer coefficient for the RHR heat exchanger, and wetwell design temperature.

The original certified ABWR DCD long-term containment analysis was based on nominal ANSI/ANS-5.1 (1979) decay heat curves. GEH determined that additional actinides and activation products not accounted for in the ANSI/ANS-5.1 (1979) standard, affect the decay heat curves. Therefore, in the revised DCD Tier 2, Section 6.2, Containment Systems, for long-term containment analysis, GEH used ANSI/ANS-5.1 (1994), which includes contributions from additional actinides and activation products. In addition, GEH conservatively used a two standard deviation uncertainty on decay heat when performing the revised long-term containment analysis. The staff finds that using the ANSI/ANS-5.1 (1994) decay heat model with a two standard deviation uncertainty for the long-term containment analysis is acceptable since the addition of decay heat from actinide decay and activation products is conservative for containment pressure and temperature analysis.

In the applicants RAI response dated August 11, 2015, GEH stated the following about the changes to the containment vents model:

In the containment analysis for the certified ABWR DCD, the main vent system model did not capture some of the key features that impact the short-term containment response and thus the pool swell loads. The model for DCD Revision 4 did not properly simulate the horizontal vent portion of the vent system and consequently incorrectly modeled the vent clearing time. These deficiencies are the major contributor to the difference between the previous certified ABWR and the ABWR revised containment analysis results.

The revised ABWR containment analysis correctly models the horizontal vents and was performed with DCV loss coefficients included. The total DCV loss coefficient is based on a summation of losses. The entrance loss coefficient accounts for the presence of the biological shield wall that is next to the upper drywell entrance to the DCV. The flow loss coefficient accounts for trash racks at the entrance to the vents to block insulation from entering the vents and flowing into the suppression pool. The friction loss through the DCV is the maximum pressure loss coefficient due to piping, cabling and supports routed in the DCV. The exit loss coefficient can be neglected since each DCV is directly above a Drywell-Wetwell (DW-WW) vertical vent. These flow losses are then summed and included in the containment analysis model for the DCV.

The dimensions of the horizontal vents were included in the revised analysis and confirmation of the vent clearing was performed to ensure the revised model was correct. These modifications were the major contributors to the revised analysis results for the wetwell pressure and drywell-to-wetwell differential pressures.

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GEH needed to change the containment vents model to correct self-identified errors in the containment analysis. The staff finds that the above features, which were missing in the containment analysis for the certified ABWR DCD by error, were needed to correctly model the GEH ABWR design, and therefore, determines that these modeling changes are acceptable.

DCD Tier 2, Table 6.2-2 lists the VLC range between 2.5 to 3.5. The applicant changed the VLC range to between 2.5 to 5.0. GEH cited NEDO-33372 and indicated that the applicable information was extracted from this document and put into the DCD. The staff noted that NEDO-33372 lists the VLC as between 2.5 to 3.5, which is different from the range of values provided in the ABWR DCD, Revision 5, specifically the upper limit. Therefore, in a public teleconference on April 6, 2016, the NRC staff requested GEH to clarify this difference.

In the applicants response to RAI 06.02.01.01.C-1, Revision 1, Supplement 2, dated June 22, 2016, GEH reiterated that it does not intend for NEDO-33372 to be part of the licensing basis for ABWR DC renewal and the ABWR DCD will contain all pertinent content and that the range of VLC values shown in the markup for DCD Tier 2, Table 6.2-2 that was included in NEDO-33372 does show a VLC range of 2.5 - 5.0. The original upper end value of 3.5 is shown crossed out in the markup.

The original range of 2.5 - 3.5 was first developed for use with the GEH Mark III Containment Pressure and Temperature (M3CPT) code for analyses of the Mark III short-term containment response. It was then applied in the ABWR M3CPT analyses due to the similarity in the Mark III and ABWR horizontal vent system geometry. A subsequent evaluation updated the range of VLCs for Mark III M3CPT analysis to 2.5 - 5.0. The revised values were then also applied to the ABWR M3CPT containment analysis. The values shown in DCD Tier 2, Table 6.2.2 (2.5 - 5.0) only included the losses associated with the ABWR vent system. It did not include or identify a 1.7 loss coefficient adder to the values shown in DCD Tier 2, Table 6.2-2 that was applied to account for flow losses associated with the DCV that connects the upper drywell to the vent system. The applicant provided a markup for the ABWR DCD, Revision 6, identifying the range of overall VLCs used for the analyses for the ABWR DCD that includes the 1.7 loss coefficient adder (4.2 - 6.7).

The applicant provided the ABWR DCD, Revision 7, value for VLC as (4.2 - 6.7), that the staff found conservative and therefore acceptable, as incorporated from the applicants response to RAI 06.02.01.01.C-1. Therefore, Confirmatory Item 6.2.3.1-1 from the staff advanced safety evaluation with no open items for the ABWR DC renewal is resolved and closed.

The FWLB flow change was to increase the 116% nuclear boiler rated (NBR) flow from the balance of plant side during the initial 3.75-second feedwater inventory depletion period to 164% NBR flow, as assumed for the inventory depletion period after the 3.75-second period and shown in the certified DCD Tier 2, Figure 6.2-3. The specific enthalpy of feed water flow as shown in DCD Tier 2, Figure 6.2-4 was unchanged. This increase in mass flow is conservative because it produces a higher energy flow into the containment than that used in the certified ABWR design during a FWLB, resulting in higher short-term containment peak pressures.

Therefore, the staff finds that the FWLB flow change is acceptable.

In the applicants RAI response dated August 11, 2015, GEH stated the following about the change in suppression pool volume margin:

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The water volume in the suppression pool including the vents is required to be equal to or greater than 3,580 cubic meters, as stated in the Tier 1 Section 2.14.1. The ABWR revised [long-term] containment analyses of scenarios with low initial suppression pool water level were performed with a smaller water volume (3,455 cubic meters) to ensure analysis/operational margin. This smaller volume is based on a suppression pool water level of 6.9 meters. The volume of 3,580 cubic meters is equivalent to a 7-meter water level. The technical specification for suppression pool water level (LCO 3.6.2.2) is greater than or equal to 7 meters and less than or equal to 7.1 meters. This is a very tight band to control the suppression pool water level; so additional margin (0.1 meters) has been built-in to the safety analysis. It is conservative to base the safety analysis for scenarios with a lower initial suppression pool water level based on a smaller water volume as this results in higher pool temperatures.

The staff determined that the change in the suppression pool volume margin in the safety analysis is conservative for as described by the applicant, and therefore, is acceptable.

As part of the applicants response to RAI 06.02.01.01.C-1 dated August 11, 2015, and supplemented with the response dated May 6, 2016, GEH increased the RHR heat exchangers heat transfer coefficient. However, cooldown rates for the ABWR are administratively controlled and are governed by technical specifications; therefore, the staff concluded that the increase in the heat transfer coefficient RHR does not affect the safety of the reactor or the containment analysis.

The staff reviewed the ABWR DCD, Revision 6, markups and confirmed that the applicant has made the appropriate changes in the ABWR DCD, Revision 7, from the response to RAI 06.02.01.01.C-1 in the May 6, 2016 supplemental letter. Therefore, Confirmatory Item 6.2.3.1-1 from the staff advanced safety evaluation report with no open items for the ABWR DC renewal is resolved and closed.

On the wetwell design temperature change, GEHs response dated August 11, 2015, stated the following:

The certified ABWR wetwell gas space design temperature was 104°C. The containment structural analysis design value is 124°C; therefore the Tier 2 DCD wetwell chamber design temperature was revised to 124°C.

The staff finds this change acceptable because it is more protective from a safety standpoint and makes the containment structural analysis and wetwell chamber design temperatures consistent.

As described above, the staff finds that GEHs response to Part (1) of RAI 06.02.01.01.C-1 acceptable.

In DCD Tier 2, Chapter 6, Change List Item 18 (which is related to DCD Tier 2, Section 6.2.1.1.3.3.1.2) it is stated that lower drywell flooding is not modeled. The staff was not clear why lower drywell flooding was not modeled. Therefore, in Part (2) of RAI 06.02.01.01.C-1, the staff requested GEH to justify not modeling lower drywell flooding.

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In the applicants response letter dated August 11, 2015, GEH described two mechanisms causing lower drywell flooding. The first was spilling of break flow water from the upper drywell to the lower drywell through the DCV connection. GEH assumed that water, that can spill, into the lower drywell would flow into the suppression pool instead. This assumption is conservative because water that flows out from the break during suppression pool drawdown during ECCS injection will be hotter than the water in the suppression pool and adding it back to the suppression pool would heat the suppression pool water.

A second mechanism is the potential for reverse vent flow from the suppression pool to the lower drywell through the lower drywell overflow orifice connection to the vertical vent. GEH showed that extended periods of large negative DW-WW pressure gradients would not exist because of opening of wetwell-to-drywell vacuum breakers. Further DCD Tier 2, Section 6.2.1.1.10.3 states the following:

The interconnection between the lower drywell and the wetwell is at elevation -4.55 m,

[which is] 8.6 m above the floor of the suppression pool. Thus, approximately 7.2E5 kg of water must be added from outside the containment for the suppression pool to overflow into the lower drywell.

As such, reverse vent flow from the suppression pool to the lower drywell would be unlikely to occur. Therefore, the staff finds that GEHs justification for not modeling the lower drywell flooding as provided in the response to Part (2) of RAI 06.02.01.01.C-1 is acceptable.

In the ABWR DC renewal application, ABWR DCD, Revision 5, Tier 2, Change List Item 19 (which is related to DCD Tier 2, Section 6.2.1.1.3.3.1.2, Assumptions for Long-Term Cooling Analysis), the applicant deleted previous assumption No. 7 from the certified design and inserted an assumption stating that the structural heat sinks are credited. The previous assumption, which was deleted, stated that at 70 seconds, the feedwater specific enthalpy becomes 418.7 Joules per gram (J/g) (i.e., saturation fluid enthalpy at 100 degrees Celsius (°C).

The staff finds that removing previous assumption No. 7 is acceptable because the applicant used DCD Tier 2, Figure 6.2-22, from the ABWR DCD, Revision 4, certified ABWR DCD instead, which provides a more limiting value for feedwater specific enthalpy. DCD Tier 2, Figure 6.2-22 shows that the feedwater specific enthalpy drops below 418.7 J/g only after 86 seconds.

However, the application did not provide the details for its modeling of the heat sinks.

Therefore, in Part (3) of RAI 06.02.01.01.C-1, the staff asked GEH to provide this information.

In the applicants response dated August 11, 2015, GEH provided details for its modeling of the structural heat sinks in the drywell airspace, wetwell airspace, and suppression pool. The applicant has modeled heat transfer in the drywell and wetwell air space by natural convection and condensation. The applicant modeled heat transfer from the suppression pool water to the suppression pool heat sinks. The applicants response included tables of heat sink parameters for the modeled heat sinks in the drywell airspace, wetwell airspace, and suppression pool. The applicant stated that the crediting of the heat sinks remains valid for as-built plants unless there is a change in plant dimensions. However, a COL applicant will include inputs for heat sinks in the standard form that the applicant uses to confirm inputs to the containment analysis and confirm the validity of the ABWR DCD analysis for the as-built plant. Design Commitment 4 in DCD Tier 1, and DCD Tier 1, Table 2.14.1 states that [t]he maximum calculated pressures and temperatures for the design basis accident are less than design conditions. The discussion of inspections, tests, and analyses for this commitment states that [a]nalyses of the design basis 6-7

accident will be performed using as-built [primary containment system] data. The applicant provided tables with properties of heat sinks in response to Part (3) of RAI 06.02.01.01.C-1.

The staff reviewed these properties to confirm that the applicant used correct thermal properties and correctly calculated the mass and internal thermal resistance for the heat sinks. Based on its review, the staff finds that the applicants response to Part (3) of RAI 06.02.01.01.C-1 is acceptable.

DCD Tier 2, Chapter 6, Revision 5, Change List Item 23 (which is related to the main steamline break discussion in DCD Section 6.2.1.1.3.3.2) changed assumption (5). Assumption (5) in the certified ABWR DCD, Revision 4, stated that MSIVs are completely closed at a conservative closing time of 5.5 seconds (0.5 seconds greater than the maximum closing time plus instrument delay), in order to maximize the break flow. ABWR DCD, Revision 5, changed the closing time to 5 seconds and eliminated the reference to 0.5 seconds delay. The staff was not clear whether the 0.5-seconds delay was included in the MSIV closing time. Therefore, in Part (4) to RAI 06.02.01.01.C-1, the staff requested GEH to clarify the MSIV closing time.

In the applicants response dated August 11, 2015, GEH stated that the instrument delay of 0.5 seconds to begin closing the MSIVs is included in the total 5.0 second duration for MSIV closure from the start of the event. This clarifies how the instrument delay of 0.5 seconds is accounted for. The staff finds that closing the MSIVs sooner (i.e., in 5 seconds versus 5.5 seconds used in the certified ABWR DCD) is conservative because it reduces radioactive releases through MSIVs during a design basis accident. Based on its review, the staff finds that GEHs response to Part (4) of RAI 06.02.01.01.C-1 acceptable.

DCD Tier 2, Revision 5, Chapter 6, Change List Item 24 relates to changing assumptions used in short-term containment analysis in DCD Tier 2, Section 6.2.1.1.3.3.2.1. GEH deleted the following assumptions:

• Assumption 1. The vessel depressurization flow rates are calculated using the Moodys

[homogeneous equilibrium model (HEM)] for the critical break flow.

• Assumption 2. The turbine stop valve closes at 0.2 second. This determines how much steam flows out of the RPV, but does not affect the inventory depletion time on the piping side.

• Assumption 4. The feedwater mass flow rate for a [main steam line] break was assumed to be 130 percent of NBR for 120 seconds. This is a standard [MSLB]

containment analysis assumption based on a conservative estimate of the total available feedwater inventory and the maximum flow available from the feedwater pumps with discharge pressure equal to the [reactor pressure vessel] pressure. The feedwater enthalpy was calculated as described for the [FWLB] (Subsection 6.2.1.1.3.3.1.1) for 130 percent of NBR flow, and is shown in Figure 6.2-11.

The reason for these deletions was not clear to the staff. Therefore, in Part (5) of RAI 06.02.01.01.C-1, the staff asked GEH to explain these changes. In the applicants response dated August 11, 2015, GEH noted that Assumption 1 was listed as an exception to the assumptions identified for the FWLB analysis in DCD Tier 2, Section 6.2.1.1.3.3.1.1. GEH deleted this assumption in DCD Tier 2, Section 6.2.1.1.3.3.2.1 because the same assumption is listed under DCD Tier 2, Section 6.2.1.1.3.3.1.1. The staff finds that deletion of Assumption 1 acceptable because the deletion was to remove a repetitive assumption in the ABWR DCD.

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Concerning the deletion of Assumption 2, GEH noted that the use of the turbine stop valve closure time is not applied for the revised MSLB analysis to establish the vessel isolation time, and Assumption 5 in Section 6.2.1.1.3.3.2 states that MSIVs are completely closed at a conservative closing time of 5 seconds in order to maximize the break flow. The staff finds GEHs deletion of Assumption 2 acceptable because it is not used for the revised analysis.

Concerning the deletion of Assumption 4, GEH stated that Assumption 4 describes feedwater injection to the vessel for the MSLB, which is not modeled in the current short-term MSLB analysis. Injecting relatively colder feedwater into the reactor pressure vessel will tend to reduce the short-term vessel pressure due to reduced steaming that in turn reduces the break flow into the containment, thereby lowering the predicted short-term MSLB containment pressure and temperature. Therefore, to produce a more conservative short-term MSLB pressure and temperature response, the applicant has not included feedwater injection in the MSLB short-term analysis. The staff finds the applicants deletion of Assumption 4 conservative and acceptable.

As described above, the staff finds GEHs response to Part (5) of RAI 06.02.01.01.C-1 acceptable.

DCD Tier 2, Revision 5, Chapter 6, Change List Item 26 (which is related to the discussion of short-term accident response in DCD Tier 2, Section 6.2.1.1.3.3.2.3) indicates that the short-term MSLB has a more severe drywell temperature response than before as it increased from 169.7 °C in ABWR DCD, Revision 4, to 177.2 °C in ABWR DCD, Revision 5. The reason for this change was not clear to the staff. Therefore, in Part (6) to RAI 06.02.01.01.C-1, the staff requested GEH to explain.

In the applicants response dated August 11, 2015, GEH stated the following:

The revised analysis included corrections to the vent system modeling that had a significant impact on both the peak drywell pressure and peak drywell temperature. The length of the horizontal vent was not correctly accounted for in the original calculation. In addition, the overall flow loss coefficient for the ABWR vent system did not account for the flow losses associated with the drywell connecting vents (DCV). The corrections that were implemented in the revised calculations produced a delay in clearing of the horizontal vents and an increase in the vent flow resistance after vent clearing. These changes produced the higher values for predicted peak MSLB drywell pressure and temperature.

The peak calculated MSLB drywell temperature of 177.2 °C is higher than the design limit of 171.1 °C. However, this value represents the peak predicted MSLB drywell atmosphere temperature. A review of the analysis shows that predicted drywell atmosphere temperatures are above 171.1 °C for approximately only 1 second during the early, steam break flow only phase of the MSLB. The MSLB analysis assumes level swell of the vessel liquid due to voiding, which produces a two-phase break flow mixture after two seconds into the event. Thereafter, drywell temperatures fall rapidly (see DCD Tier 2, Figure 6.2-13). The very short predicted duration of atmosphere temperature 6-9

above 171.1 °C will not result in drywell structural temperatures that are above the drywell structure design limit.

The applicant corrected a self-identified error in modeling the overall flow loss coefficient for the ABWR vent system. The staff reviewed these modeling changes under Part (1) to RAI 06.02.01.01.C-1 and found them acceptable. The peak calculated MSLB drywell atmosphere temperature of 177.2 °C exceeds the drywell design limit of 171.1 °C for a 1 second duration. However, due to thermal inertia, components in the drywell structures (in particular, the upper head seals) will not have sufficient time to reach the design limit temperature during this 1 second period. Therefore, the staff finds that containment atmosphere temperature exceeding the structural design temperature in this case is acceptable.

Based on its review, the staff finds GEHs response to Part (6) of RAI 06.02.01.01.C-1 is acceptable.

DCD Tier 2, Chapter 6, Revision 5, Change List Items 30 and 31 (which are related to DCD Tier 2, Section 6.2.1.1.4.1 and Section 6.2.1.1.4.2 on the negative pressure design evaluation) states that the applicant replaced each section except the first two paragraphs. The applicant did not state the reasons for the changes. Therefore, in Part (7) to RAI 06.02.01.01.C-1, the staff requested the applicant provide details justifying the changes. In the applicants response dated August 11, 2015, GEH stated that it performed the revised calculations to provide a more accurate and realistic simulation of negative pressurization events consistent with the ABWR plant system design, plant system operation and plant operating conditions. The main changes made in the revised analysis are as follows:

(1) The applicant eliminated analyses for events with inadvertent initiation of containment (drywell/wetwell) spray during normal operation. As described in DCD Tier 2, Section 6.2.1.1.4, the ABWR design has features that prevent the initiation of the RHR mode of the drywell spray(s) during normal plant operation.

(2) The revised analyses start at time zero of the postulated loss-of-coolant accident event with normal operating conditions as the initial conditions. The analysis itself is used to predict the initial conditions prior to ECCS reflood or drywell (DW) spray initiation as opposed to using user-defined conditions at the time of ECCS reflood or spray.

(3) Drywell break flow rate and break flow enthalpy during periods of ECCS injection are mechanistically calculated considering the effects of ECCS injection rates, ECCS source temperature, and heatup in the vessel before discharge to the drywell.

(4) The revised analyses include modeling of DW spray with suction from the suppression pool. The DW spray temperature is established by the calculated exit temperature of the modeled RHR heat exchanger and accounts for the heat exchanger heat removal characteristics (heat exchanger coefficient), calculated suppression pool temperature, RHR service water temperature and containment spray flow rate.

(5) The new analyses include a small steamline break with DW spray operation to provide the containment negative pressure response due to operation of drywell spray in a superheated steam drywell environment, which would occur during a small steam break, and which is potentially limiting for containment negative pressure.

The following provides the staffs evaluation of the above changes:

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(1) As stated in DCD Tier 2, Section 6.2.1.1.4, of the ABWR DCD, Revision 6, an interlock on the drywell spray injection valves that requires high drywell pressure to be present before the valves are allowed to be opened and a time delay in the logic will allow initiation of drywell spray 60 seconds after the drywell high pressure signal (2 psig) is received. In addition, the RHR system can only be manually initiated in the drywell spray mode from the main control room by two methods, both requiring two independent actions. Therefore, the staff finds that a likelihood of a spurious initiation of drywell spray during normal plant operation to be remote and the elimination of such activation from analysis to be acceptable.

(2) The applicant used the analysis itself, rather than user defined conditions, to establish initiation of the negative design pressure evaluation. The staff finds that this approach is less subjective, and therefore, acceptable.

(3) The applicant stated in its RAI response dated August 11, 2015, that [i]n the original DCD analysis it was assumed that 100 percent of ECCS flow (including [high pressure core flooder, low pressure core flooder and reactor core isolation cooling]) is taken from the [condensate storage tank] (at 60°F) and discharged directly into the drywell without heating of the ECCS injection fluid in the vessel. Using mechanistically calculated drywell break flow rate and break-flow enthalpy during periods of ECCS injection produces a less conservative result than that provided in the certified ABWR DCD.

However, the staff finds the applicants mechanistic calculation consistent with SRP Section 6.2.1.1.C, and therefore, acceptable.

(4) The analysis presented in the certified ABWR DCD did not assume the operation of drywell sprays. The staff finds that the operation of drywell sprays would lower the drywell temperature and pressure by condensing steam in the drywell, which conservatively increases the DW-WW negative pressure, and therefore, is acceptable.

(5) As stated under Item 4 above, operation of drywell sprays in a steam environment would lower the drywell pressure, and thus, increases the DW-WW negative pressure. The small steamline break with DW spray operation is a new analysis which is potentially limiting for the negative containment pressure. The staff finds this change is acceptable because it was done to seek more conservative analysis for the containment negative pressure.

The results of the revised calculation show a significantly smaller calculated peak DW-WW negative differential pressure relative to the value reported previously, -3.86 versus -9.8 kilopascal (kPaD). GEH attributes this change to a less conservative analysis approach as described above. Based on its review the staff finds that the applicants negative pressure design evaluation is consistent with SRP Section 6.2.1.1.C guidance, and therefore, is acceptable.

The results of the revised calculation show a smaller calculated peak wetwell-to-reactor building (WW-RB) negative differential pressure relative to the value reported previously, -8.76 versus -9.8 kPaD. The applicant attributes this to the SHEX code used to generate transient responses; the previous analyses used a series of end-point calculations to generate a set of conditions that produces a bounding prediction of the peak negative WW-RB differential 6-11

pressure. The GEH states the following on using the SHEX code for calculating the WW-RB negative differential pressure:

The GEH SHEX computer code was used for the revised calculations of the ABWR negative containment pressure for ABWR DCD Revision 5. The SHEX code has models for all containment, safety and auxiliary systems needed for the ABWR DCD negative pressure analysis. This is the code that corresponds to the Long-Term Cooling model identified in DCD Tier 2, Section 6.2.1.1.3.4.2. The GEH SHEX code has been verified and validated for general use in compliance with the GEH Nuclear Energy Quality Assurance Program.

The GEH calculations of the ABWR containment negative pressure response with the SHEX code and evidence of verification for the calculations are contained within the GEH electronic archives of the design records.

Although, the original ABWR DCD did not name the computer code used for analyzing the containment long-term cooling, as stated above, GEH identified it as SHEX. Considering that the SHEX code has been verified and validated for general use, and it was used for analyzing the long-term containment response in the original ABWR DCD, Revision 4, the staff finds it acceptable to use the SHEX code for calculating the peak negative WW-RB differential pressure, which is another application of containment long-term response. As such, the staff finds GEHs response to Part (7) of RAI 06.02.01.01.C-1 acceptable.

The applicant provided the necessary information in the ABWR DCD, Revision 7, which incorporated the appropriate changes described in the applicants response to Part (7) of RAI 06.02.01.01.C-1, that was found acceptable to the staff. Therefore, Confirmatory Item 6.2.3.1-1 from the staff advanced safety evaluation with no open items for the ABWR DC renewal is resolved and closed.

6.2.1.3.4 Conclusion Based on the evaluation provided in this supplemental FSER section, the staff concludes that the changes in DCD Tier 2, Sections 6.2.1 and 6.2.2, related to short-term containment pressure response do not alter the safety findings made in NUREG-1503 and are consistent with SRP Section 6.2.1.1.C, Revision 7, issued March 2007. Therefore, the staff finds that the changes reviewed in ABWR DCD, Revision 7, resulting from containment re-analysis are acceptable and meet the requirements in GDC 16 and 50 and therefore are acceptable.

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References

1. 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants.

2. 10 CFR Part 50, Appendix A, GDC 16, Containment Design,

3. 10 CFR Part 50, Appendix A, GDC 50, Containment design basis,

4. 10 CFR Part 52, Appendix A, Design Certification Rule for the U.S. Advanced Boiling Water Reactor.

5. 10 CFR 52.59, Criteria for Renewal.

6. NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 6.2.1.1.C, "Pressure-Suppression Type BWR Containments," Revision 7, March 2007(ADAMS Accession No. ML063600403).

7. NRC, NUREG-1503, Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, July 1994 (ADAMS Accession No. ML080670592).

8. NRC, NUREG-1503, Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, Supplement 1, May 1997 (ADAMS Accession No. ML080710134).

9. GEH, ABWR Standard Plant Design Certification Renewal Application Design Control Document, Revision 5, Tier 1 and Tier 2, December 2010 (ADAMS Accession No. ML110040323).

10. GEH, ABWR Standard Plant Design Certification Renewal Application Design Control Document, Revision 6, Tier 1 and Tier 2, February 2016 (ADAMS Accession No. ML16214A015).

11. GEH, ABWR Standard Plant Design Certification Renewal Application Design Control Document, Revision 7, Tier 1 and Tier 2, December 2019 (ADAMS Accession No. ML20007E371).

12. GEH, NEDO-33372, Advanced Boiling Water Reactor (ABWR) Containment Analysis."

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