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December 19, 2018 Presentation Slides - ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting
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Issue date:	12/19/2018
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GE Hitachi Nuclear Energy ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting December 19, 2018 GE Hitachi Nuclear Energy Copyright © 2018, GE Hitachi Nuclear Energy Americas LLC December 19, 2018 Slide 1

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting Purpose

- Address open items resulting from the regulatory audit of the GE Hitachi Nuclear Energy Advanced Boiling Water Reactor design certification renewal application relating to Title 10 CFR § 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors peak cladding temperature reports.

- Clarify and re-evaluate the supplemental information provided by GEH in letter M170071, dated March 20, 2017, using the approved methods applied in the original ABWR design certification.

- Discuss any open items needed for the staff to complete its Safety Evaluation Report for the Emergency Core Cooling System related to peak cladding temperature.

ABWR Cert Renewal PCT Error Estimate NRC Staff Observations The following observations were raised during the audit with a request that clarifications be provided.

1. Letter 2002-02: (a) Determine whether the SAFER model was used for the vessel water level dryer design pressure drop. (b) Provide a clarification of the level impact of this analysis for the ABWR and, determine if smaller breaks also are shown to uncover the core during a LOCA event. (c) Provide additional information in the DCD to address potential bleed-over to other break sizes and potential change in limiting break sizes. (d) Eliminate applicability of this error to ABWR, if possible.

2. Letter 2006-01: (a) Determine whether the value used for the Top Peaked Power Shape was the same as that specified for the GE-7 fuel design for ABWR. (b) Because Large Break LOCA for ABWR looks more like Small Break LOCA in other plants determine whether the assignment of the error magnitude may need to reflect this difference. (c) Address the possibility of some other break size is limiting for ABWR.

3. Letter 2012-01: (a) Determine whether PRIME was used to resolve Fuel Thermal Conductivity Degradation (TCD). (b) Remove the reference to PRIME from the DCD if it was not used. (c)

State that new methodology is not being used to make the justification. Add that the original software, which was contemporary with the certification, can be relied upon.

4. Letter 2014-03: (a) Determine if SAFER04E4 was used for the LOCA analysis. (b) Determine the minimum core DP model, in particular, whether the 10o F benefit calculated is applicable for ABWR .

5. Station Blackout - Identify the time-step convergence basis. (compared with the ECCS-LOCA issue of 2001-02).

6. Ensure that the errors identified do not invalidate the ABWR Evaluation Model (EM).

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

1. Audit Consensus: Letter 2002-02: (a) Determine whether the SAFER model was used for the vessel water level dryer design pressure drop. (b) Provide a clarification of the level impact of this analysis for the ABWR and, determine if smaller breaks also are shown to uncover the core during a LOCA event. (c)

Provide additional information in the DCD to address potential bleed-over to other break sizes and potential change in limiting break sizes. (d) Eliminate applicability of this error to ABWR, if possible.

1. Response:

Subsequent review of the historic design file finds that a calculated value was introduced in the ABWR calculation to account for this difference. It is concluded that the Notification Letter 2002-02 would not be applicable to the ABWR analysis.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

2. Letter 2006-01: (a) Determine whether the value used for the Top Peaked Power Shape was the same as that specified for the GE-7 fuel design for ABWR.

(b) Because Large Break LOCA for ABWR looks more like Small Break LOCA in other plants, determine whether the assignment of the error magnitude may need to reflect this difference. (c) Address the possibility of some other break size is limiting for ABWR.

2. Response:

-The Steam Line Outside Containment is the limiting ABWR break location. The combination of the rapid depressurization and the trip of all reactor internal pumps with subsequent fast reduction in core flow results in a short duration departure from nucleate boiling at the initiation of the event. The heat up is a very short duration with a relatively low temperature excursion to a peak of 1149F within 10 seconds into the event with no core uncovery.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

2. Response (continued):

- The depressurization results in liquid entrainment that rapidly cools the fuel back to saturation conditions. This same phenomena would be expected with this method irrespective of the initial power shape.

- This is much different than the phenomena for 2006-01 where there is a long term core uncovery for a small break LOCA scenario, where a more top peaked shape had a large impact.

- Therefore, the effect of top-peaked power shape is concluded to show no impact on the ECCS-LOCA analysis; Notification Letter 2006-01 would not be applicable to the ABWR analysis.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

3. Audit Consensus: Letter 2012-01: (a) Whether PRIME was used to resolve Fuel Thermal Conductivity Degradation (TCD). (b) Remove the reference to PRIME from the DCD if it was not used. (c) State that new methodology is not being used to make the justification. Add that the original software, which was contemporary with the certification can be relied upon.

3. Response:

- TCD influence becomes notable at later exposure times of fuel residence in the core. The ECCS-LOCA analysis predicts bounding PCT at early exposure times (as a function of heat generation, gap conductance, stored energy). The Thermal Mechanical Operating Limit (TMOL) curve applied to the fuel reduces the allowed linear heat generation rate (LHGR) as exposure increases; the lower LHGR results in reduced PCT for analysis at later exposure conditions.

TCD was considered and evaluated to conclude an increase in PCT postulated by TCD - at later exposures - would not be greater than the reduction in PCT from the TMOL curve. The ABWR ECCS-LOCA analysis, performed at a bounding early time in life, would remain bounding, regardless of affect by TCD.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

3. Response (continued):

- PRIME was not used to arrive at the foregoing conclusion. However, subsequent generic model comparisons of GESTR (as used for ABWR) with PRIME, which explicitly addresses TCD, affirms this conclusion for limiting PCT.

- In the previous letter report, M170071, PRIME was used to demonstrate that the ABWR would still fall within 10 CFR § 50.46 acceptance criteria even if modern calculation methods were used. PRIME is not used to make justification, no new methodology.

- The results are not changed from the original DCD. The change reported in the 2012-01 letter does not apply to the cumulative PCT reported for ABWR.

All reference to PRIME has been removed from the next DCD revision.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

4. Audit Consensus: Letter 2014-03: (a) Determine if SAFER04E4 was used for the LOCA analysis. (b) Determine the minimum core DP model, in particular, whether the 10o F benefit calculated is applicable for ABWR .

4. Response:

- The ABWR changes per 10 CFR § 50.46, identified in PCT Notification letter 2014-03, are removed from the cumulative listing.

- During the period of the original DCD PCT analysis (1986-87), the later version code (04) was not yet approved, although it was under development. Because of this, SAFER04 would not have been available for use in the original ABWR PCT analysis. The original analysis was performed with the prior SAFER03 version. The change reported in the 2014-03 letter would not be applicable to the ABWR analysis.

- Moreover, the sensitivity reported for BWR/5-6 plants was noted generally for small break limiting results, which is not a factor for the ABWR.

- The 10o F benefit was not applicable to ABWR, because the conditions evaluated were not applicable for the ABWR.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

5. Audit Consensus: For Station Blackout, identify the time-step convergence basis (compared with the ECCS-LOCA issue of 2001-02).

5. Response:

- Because SAFER was not used for SBO, the issue associated with this error is not relevant for the SBO analysis.

- The Station Blackout (SBO) analysis did not use the same models as the ECCS-LOCA analysis. The ABWR SBO is designed to have alternate AC power source available to be connected to the equipment providing core inventory and decay heat removal within 10 minutes. Therefore, the core is always covered, and the peak cladding temperature is the initial fuel steady state temperature.

No coping analysis is required according to 10 CFR § 50.63 rule. Thus, the original DCD analysis for station blackout (SBO) was based on an engineering evaluation, and was not calculated using any code.

- The time-step convergence error noted in 10 CFR § 50.46 report 2001-02 for SAFER does not affect the SBO evaluation.

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

6. Audit Consensus: Ensure that the errors identified do not invalidate the ABWR Evaluation Model (EM).

6. Response:

- The total of the cumulative errors and changes in peak cladding temperature reported in the revised 10 CFR § 50.46 annual reporting are seen to potentially affect the calculated results based on the evaluation model as applied in the analysis of the DCD.

- It is concluded that the existing evaluation model for the ABWR DCD fuel type remains valid.

- The ABWR DCD will be modified to reflect these impacts in the next revision.

The expected changes follow (slide 16):

ABWR Certification Renewal Peak Cladding Temperature Increase NRC Public Meeting

6. Response (cont):

- Letter 2003-03 regarding separator loss coefficient error was evaluated for impact on Transients and ATWS.

>> The ABWR DCD Transient and ATWS analyses basedecks are developed separately from the LOCA analysis. The ABWR DCD Transient and ATWS analyses basedecks used the correct separator pressure drop. No impact.

- Fuel Thermal Conductivity Degradation (TCD), LOCA issue 2012-01

>> The ABWR DCD Transient and ATWS analyses are not significantly affected by fuel TCD. This is based on GEH experience and was previously evaluated by the USNRC GE Hitachi Nuclear Energy Copyright © 2018, GE Hitachi Nuclear Energy Americas LLC December 19, 2018 Slide 12

Re-evaluation of ABWR Cert Renewal PCT Error Estimates Description Year Calculated PCT Value Analysis of Record Licensing 1,149 °F Basis PCT 1999-02 CCFL in upper tie plate, 1999 + 25 °F pre-GE11 2001-02 Time step change for 2001 + 25 °F convergence 2001-04 Steam condensed by 2004 + 10 °F ECCS injection 2003-01 SAFER Level Volume 2003 + 10 °F Table 2003-03 Steam Separator 2003 + 5 °F Pressure Drop Cumulative and Absolute Sum of 2018 + 75 °F 10 CFR § 50.46 Changes Projected Licensing Basis PCT 2018 1,224°F Based on These Changes GE Hitachi Nuclear Energy Copyright © 2018, GE Hitachi Nuclear Energy Americas LLC December 19, 2018 Slide 13

ABWR LOCA Max Oxidation and Hydrogen Generation The cumulative changes and errors reported to the evaluation model have resulted in changes to Peak Cladding Temperature (PCT). No reportable effect has been noted in either the LOCA maximum oxidation or the hydrogen generation reported in the DCD.

The peak oxidation was determined for the ABWR using the original, approved DCD analysis. The maximum cladding oxidation reported for the ABWR remains unchanged due to these changes and errors. The ABWR maintains significant margin below to the allowable limit of 17%.

The calculated total hydrogen generated from the chemical reaction of the cladding with the water or steam remains unchanged from that shown in the original DCD. Again, a substantial margin is maintained to the 1% core wide limit.

For ABWR Design Basis Accident, there is no core uncovery. Hence, a coolable core geometry is maintained. Long-term core cooling is ensured by the use of redundant reactor heat removal systems with adequate water sources to keep the core covered and the transfer of decay heat in the core to the ultimate heat sink.

Result of the Re-evaluation of 10 CFR § 50.46 Changes and their effect on the DCD Safety Analysis In summary, the effects of the changes and errors discovered in the approved ECCS-LOCA evaluation model, as applied for the original ABWR DCD analysis -

with identified PCT applied - are the following:

1. The only accident analysis results affected by changes or errors found in the approved ECCS-LOCA evaluation model are confined to the ECCS-LOCA analysis.

2. The bounding cumulative change in PCT was 75°F (42°C), resulting in a projected PCT total of 1,224°F. This result remains well below the criterion of 10 CFR § 50.46 of 2,200°F.

3. The other 10 CFR § 50.46 criteria remain unaffected - thus, oxidation and hydrogen generation remain compliant.

4. The ABWR DCD will be modified to reflect these impacts in the next revision.

The expected changes follow:

Expected DCD changes (cont.)

Questions?

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