ML13032A550

NEDO-33808, Revision 0, Peach Bottom Atomic Power Station Units 2 and 3 SIL 636 Evaluation of Small Break Loca.

ML13032A550
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Site:	Peach Bottom
Issue date:	01/31/2013
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NEDO-33808, Rev 0
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ATTACHMENT 3 GE Hitachi Nuclear Energy Report GE-NE-0000-0011-4483, Revision 0, "Project Task Report- Peach Bottom Atomic Power Station Units 2 and 3 SIL 636 Evaluation," dated March 2003 Excerpted Information (NEDC-33808P - Peach Bottom Atomic Power Station Units 2 and 3 SIL 636 Evaluation of Small Break LOCA)

(Non-Proprietary Information and Affidavit)

• HITACHI GE Hitachi Nuclear Energy NED0-33808 Revision 0 DRF Section 0000-0157-2718 RO January 2013 Non-proprietary Information - Class I (Public)

Peach Bottom Atomic Power Station Units 2 and 3 SIL 636 Evaluation of Small Break LOCA Copyright 2013, GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

NED0-33808 Revision 0 Non-proprietary Information Class I (Public)

NON-PROPRIETARY NOTICE This is a non-proprietary version of the document NEDC-33808P Revision 0, which has the prop1ietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of supporting the Peach Bottom Residual Heat Removal Drywell Spray License Amendment Request before the U. S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contract between GEH and Exelon, and nothing contained in this document shall be construed as changing that contract.

The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

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TABLE OF CONTENTS Section EXECUTIVE SUM:MAR Y ...................................................................................................... 1 1.0 OVERVIEW ..................................................................................................... 2 1.1 Introduction ............................................................................................................... 2 I. .2 Evaluation ................................................................................................................. 3 2.0 DECAY EIEAT VALUES ................................................................................ 6 2.1 Introduction ............................................................................................................... 6 2.2 Evaluation Results .................................................................................................... 6 3.0 SMALL STEAM LINE BREAK ANALYSIS ................................................. 7 3.1 Introduction ............................................................................................................... 7 3.2 Inputs and Assumptions ............................................................................................ 8 3.2.1 Method of Analysis ........................................................................................... 8 3.2.2 Assumptions ........................... ,.......................................................................... 8 3 .2.3 Decay Heat Models ......................................................................................... 11 3.2.4 Break Sizes Analyzed ..................................................................................... 11 3.3 Evaluation Results .................................................................................................. 13 3.3.1 SSLB Analyses ............................................................................................... 13 3.3.2 SSLB vs DBA-LOCA Results ........................................................................ 13 3.4 Conclusion .............................................................................................................. 39

4.0 REFERENCES

............................................................................................... 40 111

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LIST OF TABLES Table 1 Key Input Parameters and Initial Conditions for SSLB Analysis ........................ 12 Table 2 Summary Key Results: With Containment Sprays .............................................. 14 lV

NED0-33808 Revision 0 Non-proprietary Infonnation -Class l (Public)

LIST OF FIGURES Figures Figure 1 Drywell Airspace Temperature Response For 0.01 ft2 Steam Line Break ......... 15 Figure 2 Suppression Pool Temperature Response For 0.01 ft? Steam Line Break .......... 16 Figure 3 Wetwell Airspace Temperature Response For 0.01 ft 2 Steam Line Break ........ 17 Figure 4 Drywell and Wetwell Airspace Pressure Response For 0.01 ft 2 Steam Line Break................................................................................................ 18 Figure 5 Drywell Airspace Temperature Response For 0.05 fe Steam Line Break ......... 19 Figure 6 Suppression Pool Temperature Response For 0.05 ft 2 Steam Line Break .......... 20 Figure 7 Wetwell Airspace Temperature Response For 0.05 t1? Steam Line Break ........ 21 Figure 8 Drywell and Wetwell Airspace Pressure Response For 0.05 ft 2 Steam Line Break................................................................................................ 22 Figure 9 Drywell Airspace Temperature Response For 0.10 ft2 Steam Line Break ......... 23 Figure 10 Suppression Pool Temperature Response For 0.10 ft2 Steam Line Break .......... 24 Figure 11 Wetwell Airspace Temperature Response For 0.10 ft 2 Steam Line Break ........ 25 Figure 12 Drywell and Wetwell Airspace Pressure Response For 0.10 ft2 Steam Line Break................................................................................................ 26 Figure 13 Drywell Airspace Temperature Response For 0.25 ft 2 Steam Line Break ......... 27 Figure 14 Suppression Pool Temperature Response For 0.25 ft2 Steam Line Break .......... 28 Figure 15 Wetwell Airspace Temperature Response For 0.25 ft2 Steam Line Break ........ 29 Figure 16 Drywell and Wetwell Airspace Pressure Response For 0.25 ft2 Steam Line Break................................................................................................ 30 Figure 17 Drywell Airspace Temperature Response For 0.50 ft2 Steam Line Break ......... 31 Figure 18 Suppression Pool Temperature Response For 0.50 ft2 Steam Line Break .......... 32 Figure 19 Wetwell Airspace Temperature Response For 0.50 ft2 Steam Line Break ........ 33 Figure 20 Drywell and Wetwell Airspace Pressure Response For 0.50 ft2 Steam Line Break................................................................................................ 34 Figure 21 Drywell Airspace Temperature Response For 1.0 ft 2 Steam Line Break ........... 35 Figure 22 Suppression Pool Temperature Response For 1.0 ft2 Steam Line Break ............ 36 Figure 23 Wetwell Airspace Temperature Response For 1.0 ft2 Steam Line Break .......... 37 Figure 24 Drywell and Wetwell Airspace Pressure Response For 1.0 ft2 Steam Line Break................................................................................................ 38 v

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ACRONYMS AND ABBREVIATIONS,,,,

Term Dctlnition ',,' ...

ANS American Nuclear Society ANSI American National Standards Institute ATWS Anticipated Transients Without Scram BWR Boiling Water Reactor cs Core Spray CST Condensate Storage Tank DBA Design Basis Accident ECCS Emergency Core Cooling System EOPs Emergency Operating Procedures GE GE Nuclear Energy gpm Gallons per Minute HPCI High Pressure Coolant Injection HX Heat Exchanger LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray MSIV Main Steam Isolation Valve MWt Mega-Watts Thermal NFI New Fuel Introduction NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission OPL Operating Parameters for Licensing PBAPS Peach Bottom Atomic Power Station PCT Peak Cladding Temperature RCIC Reactor Core Isolation Cooling RG Regulatory Guide RHR Residual Heat Removal RHRSW RHR Service Water RPV Reactor Pressure Vessel SIL Service Information Letter SSLB Small Steam Line Break SRV Safety/Relief Valve Vl

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Term Definition TPO Thermal Power Optimization UHS Ultimate Heat Sink UFSAR Updated Final Safety Analysis Report

---~------ *- - - - - "

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EXECUTIVE

SUMMARY

The work in this report was performed by UE Nuclear Energy (UE) in 2002 and is being documented in this GE-Hitachi Nuclear Energy Americas LLC report to support Peach Bottom Residual Heat Removal (RHR) Drywell Spray License Amendment Request (LAR).

During the Peach Bottom power rerate project, engineering analyses were performed to support rerating the licensed thermal power at Peach Bottom Atomic Power Station (PBAPS) Units 2

& 3 to 3458 MWt (1 os*y.) of the original licensed power of3293 MWt). In addition, selected analyses were performed conservatively assuming 3622 MWt (llO(Yo of the original licensed power of3293 MWt). These analyses arc described and documented in the GE report, NEDC-32230P, March 1994 (Reference 1). In 2001, GE issued Service Information Letter (SIL)

No. 636 to inform utilities with GE boiling water reactors (BWRs) of a change in the GE method for calculating the decay heat values using the American National Standards Institute/American Nuclear Society (ANSVANS) 5.1-1979 standard (Reference 2). The revised method now includes decay heat from additional actinides and activation products.

GE provided support for an assessment ofthe SIL 636 affect, relative to the design and licensing analyses potentially affected by the revised decay heat calculation method described in SIL 636.

In addition, supporting evaluations tor a service water temperature increase from 90 °F to 92 °F.

were performed by GE to assess the effect of SIL 636 plus 2cr decay heat uncertainty and a 92 °F service water temperature on the current analyses; this effort included generation of new decay heat values based on the SIL 636 method, and an update of the containment design input data to rebaseline the containment analysis calculations. Also, the residual heat removal (RHR) heat exchanger K-value in this SIL 636 evaluation was assumed to range from its current value of 244.5 Btu/sec-°F to the maximum attainable value of270 Btu/sec-°F planned by Exelon.

The engineering tasks performed tor this evaluation are summarized in Section 1.0 of this report, and the results are provided in Section 3.3. The containment analysis calculations were performed using the GEH computer code SHEX with the updated design input data and assumptions. Overall, the analysis results from this SIL 636 evaluation show no adverse effect on the current licensing/design basis.

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NED0-33808 Revision 0 Non-proprietary lnfi:>rmation -Class I (Public) 1.0 OVERVIEW 1.1 Introduction During the power rerate project (Reference l ), engineering analyses were performed to justify rerating the licensed thermal power at PBAPS Units 2 & 3 to 3458 MWt (105% of the original licensed power of 3293 MWt). Selected analyses were also performed assuming 3622 MWt (110% of the original licensed power of3293 MWt) as the initial reactor thermal power. These analyses used the decay heat values based on the ANSI/ANS 5.1-1979 standard and did not include 2cr uncertainty, but only a 2% uncertainty in core thermal power (i.e., assuming a power level of 3694 MWt for power level of 3622 MWt). Subsequently, GE issued SIL 636 (Reference

3) to inform utilities with GE BWRs of a change in the GE method for calculating the decay heat values using the ANSVANS 5.1-1979 standard. The revised method now requires the inclusion of 2cr uncertainty adder and decay heat from additional actinides and activation products.

GE provided support for an assessment of the SIL 636 effect. This assessment is relative to the design and licensing analyses potentially affected by the changes in the decay heat calculation method described in SIL 636. In addition to the SIL-related assessment, GE performed supporting evaluations for service water temperature increase from 90 °F to 92 °F. Also, in anticipation of any adverse effect of SIL 636 and 92 °F service water temperature on design basis accident (DBA) loss-of-coolant accident (LOCA) peak pool temperature response, Exelon has planned to upgrade the RHR heat exchanger K-value from its current value of244.5 Btu/sec-°F to a maximum attainable value of270.0 Btu/sec-°F. This planned maximum attainable K-value of 270 Btu/sec-°F is expected to offer more than sufficient effect to offset the effects of SIL 636 plus 2cr uncertainty and a 92 °F service water temperature on DBA-LOCA peak pool temperature response, and produce the peak pool temperature response comparable to that calculated for the 110% of the original rated power level in Reference 1.

Accordingly, GE has performed engineering analyses to assess the affect of both SIL 636 with 2cr uncertainty adder and the 92 °F service water temperature on pertinent design and licensing analyses. The containment analysis input data and basis for use in this SIL 636 evaluation were accepted by Exelon, and are documented as Operating Parameters for Licensing-4A (OPL-4A) document (Reference 4). The long-term containment analyses in this evaluation were performed using the base deck of the GE computer code SHEX, which was generated based on the updated design input data in OPL-4A. Specific cases were analyzed with SHEX by adding case-specific input overlays to the base deck. The plant rated thermal power for this evaluation is 3458 MWt (i.e., 105% of the original rated thermal power), and the calculations include sensitivity analyses to determine the K-value that will be adequate to obtain a DBA-LOCA peak pool temperature response comparable to the target peak pool temperature value obtained with 3622 MWt (i.e.,

110% of the original rated power) thermal power level reported in Reference 1.

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NED0-33808 Revision 0 Non-proprietary lnl'otmation Class I (Public) 1.2 Evaluation The long-term DBA-LOCA analysis results for a post-LDCA period of 100 days, conclude that a K-value of263.0 Btu/sec-°F for the RIIR heat exchanger would produce peak suppression pool temperature response comparable to the target value of (( )) calculated at 102% of I l 0% of the original rated power of 3293 MWt under the power rerate project. Therefore, given that all other pertinent design inputs and assumptions remain the same as used in this SIL 636 evaluation, a K-value of270.0 Btu/sec-°F would result in a peak suppression pool temperature below that target value. The containment pressure and temperature responses exhibit a monotonic decreasing trend (( ]J.

Engineering analysis calculations were performed to generate DBA-LOCA containment response data, which provide input to net positive suction head (NPSH) evaluations to be performed by Exelon. Two time periods were analyzed: short-term (before initiation of containment sprays), and long-term (after initiation of containment sprays). The initial conditions and analysis assumptions for this event scenario were selected such that the containment pressure response is minimized, while the suppression pool temperature response is maximized, and a constant leakage rate of 0.5% by weight per day, irrespective of containment pressure, was factored into these analysis calculations.

The calculated results from the analysis calculations performed for the NUREG-0783 limiting event, with SIL 636 decay heat plus 2cr decay heat uncertainty adder and 92 °F service water temperature, show compliance with the NUREG-0783 requirements for the local suppression pool temperature limits; both the peak local suppression pool temperature requirement of 200 op and the requirement that the local pool temperature must be at least 20 op below the local saturation temperature at the Safety/Relief Valves (SRV) discharge quencher location are satisfied. The calculated suppression pool temperature responses were found to be minimally sensitive to the RHR heat exchanger K-value.

Engineering analysis calculations were performed to calculate the containment pressure and temperature response for a spectrum of small steam line break (SSLB) sizes, with SIL 636 decay heat plus 2cr uncertainty adder and 92 op service water temperature. The steam line breaks are the most limiting events for drywell temperature response, since steam has higher energy content than liquid. These analyses, with primary focus on the drywell temperature response, took credit for containment sprays and structural heat sinks in the drywell and the wetwell airspace. The calculated results show a maximum drywell airspace temperature of ((

))

The engineering evaluation determined that the SIL 636 decay heat with 2cr uncertainty adder has no adverse effect on the emergency core cooling system (ECCS)-LOCA performance, for both peak cladding temperatures (PCTs) and long-term core cooling performance. The maximum allowable core uncovered times (fuel can remain uncovered before the cladding temperature exceeds 1500 °F) as a function of time after shutdown, applicable for GE14 fuel, were calculated.

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The results from engineering evaluations indicated no effect on the short-term DBA-LOCA containment response and plant transients, and no adverse effect on the anticipated transient without scram (ATWS) analyses. The short-term DBA-LOCA containment response calculations use a more conservative decay heat characteristic than the ANSIIANS 5.1-1979 standard discussed in STL 636, and the service water temperature increase to 92 °F is of no significance since the analyses are applicable for time period before initiation of the RHR systems. The analyses of the plant transients also use a more conservative decay heat characteristic, and the service water temperature increase to 92 °F has no effect since this parameter docs not enter into the transient analysis calculations. The current ATWS analysis conservatively assumes only two RHR heat exchangers (out of four available heat exchangers),

and this inherent conservatism is found to be sufficient to offset the effect of SIL 636 decay heat plus 2cr decay heat uncertainty adder and the service water temperature increase from 90 °F to 92 °F.

Normal shutdown cooling analyses were performed to assess the effect of SIL 636 with 2cr uncertainty adder and the 92 °F service water temperature on the RHR normal shutdown cooling capability within these two prescribed bounds: 1) the time when the reactor pressure vessel (RPV) coolant temperature reaches 125 °F with both RHR loops operating, and 2) the time when the RPV temperature reaches 200 °F with one RHR loop operating. These calculations used an RHR heat exchanger K-value of 270.0 Btu/sec-°F, and assumed an initial reactor power level of 3458 MWt. The calculated results showed that with the operation of both RHR loops, the RPV coolant temperature could be reduced to 125 °F within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, which is two hours less than the cool down time reported in PBAPS Updated Final Safety Analysis Report (UFSAR). With one RHR loop operating, the calculated results showed that the RPV coolant temperature could be reduced to (( )) less than the Technical Specification criterion of achieving cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These results demonstrate and confirm that with the K-value of270.0 Btu/sec-°F, SIL 636 with 2cr decay heat uncertainty adder and the 92 °F service water temperature would have no adverse effect on the RHR system capability to achieve cold shutdown conditions.

Alternate shutdown cooling analysis calculations were performed to determine peak bulk suppression pool temperature and the time to cool the RPV coolant temperature to cold shutdown condition (<200 °F), assuming that the normal shutdown cooling mode is not available. The calculated time to reach the cold shutdown condition was compared against the requirement of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> specified in Regulatory Guide (RG) 1.139. These calculations were performed with the GE computer code SHEX and the updated containment design input data. These calculations used the RHR heat exchanger K-value of 270.0 Btu/sec-°F, and assumed an initial reactor power of 3458 MWt. The results show that the cold shutdown condition can be achieved in

(( )) (less than the RG 1.139 requirement of36 hours), and the maximum bulk pool temperature was (( )) (less than the design limit of281 °F). These results confirm that with the K-value of270.0 Btu/sec-°F, SIL 636 with a 2cr decay heat uncertainty adder, and the 92 °F service water temperature would have no adverse effect on the plant alternate shutdown cooling capability.

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The Appendix R analyses, perfonned previously during the power rerate project and the recent GE14 New Fuel Introduction (NFI) project, were re-performed to evaluate the effect of SIL 636 and the service water temperature increase from 90 °F to 92 °F. The primary focus was to detennine ifthe plant meets the key performance goals of maintaining reactor coolant inventory and removing decay heat, in order to demonstrate compliance with the requirements of 10CFR 50.48 and 10CFR 50 Appendix R. These calculations used an RHR heat exchanger K-value of 270.0 Btu!sec-°F, service water temperature of 92 °F, and assumed an initial reactor power of 3528 MWt. The results show compliance with the requirements of 10 CFR 50.48 and 10 CFR 50 Appendix R. The calculated peak vessel pressure is bounded by the Technical Specification limit of 1375 psig, and the PCT value remains below the requirement of 1500 °F, thus precluding fuel failure. Also, the calculated peak suppression pool temperature and the wetwcll airspace pressure remain below their respective design limits. The results confirm that by using an RHR heat exchanger K-value of270.0 Btu/sec-°F, SIL 636 with 2cr decay heat uncertainty adder and the 92 °F service water temperature would have no adverse effect on the Peach Bottom compliance with the performance goals as required in 10 CFR 50.48 and 10 CFR 50 Appendix R.

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NED0-33808 Revision 0 Non-proprietary Information -Class I (Public) 2.0 DECAY HEAT VALUES 2.1 Introduction The containment analysis calculations, documented in Reference 1, used the decay heat values based on the ANSI/ANS-5.1-1979 Decay Heat Standard and those values did not include 2a uncertainty adder. Subsequently, GE issued SlL 636 (Reference 3) to inform utilities with GE BWRs of a change in the GE method for calculating the decay heat values using the ANSI/ANS-5.1-1979 standard. The revised method now includes decay heat from additional actinides and activation products.

GE generated new decay heat tables for PBAPS Units 2 & 3, based on the ANSI/ANS 5.1-1979 standard, including the decay heat from actinide and activation products. Best-estimate decay heat values, with and without a 2a uncertainty adder are to be calculated, for a representative core conliguration.

2.2 Evaluation Results GE performed engineering analysis calculations to generate a new decay heat table based upon the ANSI/ ANS-5 .1-1979 standard, and an added conservatism corresponding to two sigma uncertainty was applied. The new table was prepared with an allowance for miscellaneous actinides and activation products, consistent with the recommendations of STL 636.

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NED0-33808 Revision 0 Non-proprietary Information -Class I (Public) 3.0 SMALL STEAM LINE BREAK ANALYSIS 3.1 Introduction GE has performed SSLB analyses utilizing the revised OPL-4A input parameters documented in Reference 4. The primary purpose of these analyses is to generate the drywcll temperature response following SSLBs, which produce the limiting drywell temperature response. A total of six steam line break sizes (0.01 re, 0.05 tt2 , 0.1 ft2 , 0.25 ft2 , 0.5 ft 2 , and 1.0 ft2 ) are analyzed, taking credit for containment sprays (both in the drywell and the wetwell) after ten ( 10) minutes.

This spectrum of break sizes is chosen to develop the limiting drywell temperature conditions with respect to both temperature and duration. The largest break size (typically 1.0 ft 2) is chosen to be the largest expected break size where reactor vessel level swell does not occur. This break size will produce the maximum initial steam only break tlow and superheated conditions in the drywell. Larger break sizes will induce vessel level swell which results in two-phase break flow into the drywell and, consequently, producing saturated (instead of superheated) steam conditions in the drywellleading to lower drywell temperatures associated with saturated conditions. The smallest break size chosen is typically 0.01 ft 2, which is regarded as small enough to ensure that the vessel is not depressurized due to the break flow. Breaks of this size result in a much slower drywell temperature rise; however, if action is not taken to depressurize the reactor vessel or tum on drywell sprays, these breaks can result in elevated temperature condition for very long duration. Therefore, various break sizes between the smallest and largest break sizes are chosen in these analyses to envelop the limiting temperahue and duration conditions. The only heat sinks credited in these analyses are the ((

))

These analyses used PBAPS-unique decay heat curve generated using the ANSl/ANS 5.1-1979 decay heat model with two standard deviations (2a) of uncertainty, as described in Section 2.

The normalized decay heat values are combined with additional heat sources (fuel relaxation energy and metal-water reaction energy), and the combined heat sources are used as a single decay power input in the analysis calculations.

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NEDO-J380~ Revision 0 Non-proprietary Information -Class I (Public) 3.2 Inputs and Assumptions 3.2.1 Method of Analysis To generate drywell temperah1re response as a function of time for a post-LOCA duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the SSLB analyses are performed using the GE computer code, Sl I EX.

The GE computer code SHEX (i.e., SHEX-05A) is used to perform analyses to calculate the long-term containment pressure and temperature responses to SSLBs. The SHEX code uses a coupled pressure vessel and containment model. The code performs fluid mass and energy balances on the reactor primary system, the suppression pool, and the drywell and wetwell airspaces. The BWR primary system, feedwater system, ECCS, and SRVs are also modeled to the extent that their response affects that of the containment system, including operator actions.

The code calculates the suppression pool bulk temperature, and the pressures and temperatures in the drywell and wetwell airspace. The use of the SHEX code has been accepted by the Nuclear Regulatory Commission (NRC) for calculating the response of the containment during an accident or a transient event and the same code has been applied to the evaluation of containment response for other BWR plants. ((

))

3.2.2 Assumptions The assumptions, event sequence, and key inputs noted here in this section are based on those specified in Reference 4. The assumptions and initial conditions used for the DBA-LOCA apply equally to the steam line breaks, except that these are steam breaks instead of the DBA-LOCA recirculation suction line break, ((

))

((

))

The RHR heat exchanger K-value given in OPL-4A (Reference 4) is 270 Btu/sec-°F, which represents the maximum attainable value. The DBA-LOCA is the limiting event for setting the RHR heat exchanger K-value. As part of the containment response for DBA-LOCA analysis, the limiting RHR heat exchanger K-value, which produced the peak suppression pool temperature comparable to that obtained with 110% of the original rated power reported in Reference 1, was determined to be 263 Btu/sec-°F. For conservatism, the SSLB analysis uses a K-value of263 Btu/sec-°F, (( ))

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NED0-33808 Revision 0 Non-proprietary In formation -Class I (Public) 3.2.2.1 Event Sequence for Small Steam Line Breaks:

The following is the event sequence for steam line breaks, as stated in OPL-4A (Reference 4),

Section 2.3.2, Appendix A with credit for containment sprays:

1. The plant is operating at 102% of the rated thermal power (i.e., 1.02 x 3458 MWt) when a steam I ine break of specific size occurs. There is also a concurrent loss of offsite power and only minimum diesel power is available. Reactor scram occurs.

2. Main steam isolation valves (MSIVs) start closing at 0.5 second and close completely at 3.5 seconds.

3. For the first 10 minutes following the accident, two low pressure coolant injection (LPCI) pumps (one in each RHR loop, inject with a flow rate of 10,000 gpm/pump) and one core spray (CS) loop (with a flow rate of 6,250 gpm) are available for vessel makeup. The LPCI and CS pumps are assumed to inject water into the vessel during the first 10 minutes, if the vessel pressure is below the permissive pressure for vessel injection.

4. At I 0 minutes, operator activates the RHR heat exchanger in one RHR loop. One LPCI pump is turned off. The second LPCI pump is re-aligned from LPCI mode to RHR mode so that RHR pump flow goes through the heat exchanger before discharging to the drywell and wetwell in the form of drywell and wetwell sprays. Of the total9500 gpm RHR flow, 8690 gpm goes to the drywell spray and the remaining 810 gpm goes to the wetwell spray.

For the small steam break (0.01 t1:2) it is expected that the drywell and wetwell sprays will be initiated later than 10 minutes depending on drywell and wetwell conditions relative to the initiating conditions given in the emergency operating procedures (EOPs). The conditions at l 0 minutes for the larger SSLB breaks will also be reviewed to confirm that conditions at 10 minutes are consistent with the initiating conditions given EOPs. Once containment cooling with drywell and wetwell sprays is initiated, this configuration is maintained throughout the accident.

5. After 10 minutes, the CS pump flow is maintained at 6,250 gpm.

6. When the suppression pool temperature reaches 120 °F, operator initiates controlled vessel depressurization at 100 °F/hr using the SRVs. For those breaks that depressurize the vessel faster than 100 °Fihr, no such operator action is required.

7. When the reactor is depressurized below the low pressure permissive for the CS, the CS will be initiated.

8. When the reactor is depressurized to 50 psig, the operator will maintain the reactor at this pressure for the remainder of the event.

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NED0-3380g Revision 0 Non-proprietary lnt<.nmation *-Class I (Public) 3.2.2.2 Assumptions for Small Steam Line Breaks The following are the key assumptions for the long-term containment response analysis for the SSLB. The assumptions arc defined in Appcndix A Sections 2.1.1 and 2.3 .1 of the OPL-4A (Reference 4). The key assumptions are summarized here.

1. The reactor is operating at 102% of rated thermal power (i.e., l.02 x 3458 MWt) with an initial reactor pressure of 1053 psi a.

2. The reactor core power includes fission energy, fuel relaxation energy, metal-water reaction energy, and decay heat. The decay heat is based upon the ANSIIANS 5.1-1979 standard consistent with the recommendations ofSIL 636 plus the 2a uncertainty adder applicable for fuel up to and including GE 14; (( ))

3. Reactor blowdown flow rates are based on ((

))

4. The reactor vessel control volume includes ((

))

5. Concurrent with the postulated event, a loss of offsite power occurs. Also, only minimum diesel power is available. This results in operation of one RHR cooling loop with one RHR pump, one RHR SW pump, and one RHR heat exchanger available for containment cooling after I 0 minutes. The RHR pump and RHR Service Water pump are aligned to the RHR heat exchanger at 10 minutes to initiate containment cooling.

6. The portion of the feed water inventory ((

))

7. For the SSLB analysis, heat and mass transfer from the suppression pool to the wetwell (suppression chamber) airspace is ((

))

8. A spectrum of steam line breaks occurs (0.01 ft 2 , 0.05 ft 2 , 0.1 ft2 , 0.25 ft2 , 0.5 Wand 1.0 ft 2).

9. The initial suppression pool water volume corresponds to the Technical Specification Low Water Level to maximize the suppression pool temperature response.

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10. The initial suppression pool temperature is 95 °F.

II. With one RHR pump at 9,500 gpm and one RHR Service Water pump (4500 gpm), the corresponding RHR heat exchanger K-value is 270.0 Btu/sec-°F. However, as noted earlier in Section 9.2.2, a K-value of263.0 Btu/sec-°F is used in the SSLB analysis. Containment cooling is achieved by operating the RHR loop, with heat exchanger, in the containment spray mode (drywell and wetwell sprays) only.

12. Containment spray (for both drywell and wetwell sprays) ((

))

13. The RHR service water temperature is at the proposed maximum Technical Specification value of92 op (to maximize the suppression pool temperature)

14. ((

))

15. Drywell fan coolers are inactive.

16. (( ))

17. All Core Spray (CS) and LPCl!RHR pumps have 100% of their motor horsepower rating converted to pump heat, ((

))

18. MSN closure starts at 0.5 seconds after the initiation of the event and full closure is achieved at 3.0 seconds after closure is initiated.

19. Condensate storage tank (CST) water inventory is not available for vessel makeup.

20. (( ))

The key input parameters and initial conditions associated with these assumptions are summarized in Table 1.

3.2.3 Decay Heat Models The decay heat curve was generated specifically for PBAPS, using the ANSI/ANS 5.1-1979 standard with a 2cr uncertainty adder (See Section 2). The normalized decay heat values (Section

2) are combined with additional heat sources (fuel relaxation energy and metal-water reaction energy), and the total combined decay power is used as decay heat power input in these analyses.

3.2.4 Break Sizes Analyzed A spectrum ofSSLBs, comprising a total of six break sizes (0.01 fi?, 0.05 ft2 , 0.1 ft 2 , 0.25 ft2 , 0.5 ft2 , and 1.0 ft2) were analyzed with credit for containment sprays at 10 minutes.

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Table 1 Key Input Parameters and Initial Conditions for SSLB Analysis Parameter Unit ..

Value Initial Core Thermal Power MWt 3,528 (I 02% of Rated Power 3458 MWt)

Initial Vesse! Dome Pressure psta 1,053 Total Drywell Free Volume re 175,800 Initial Wetwell Free (Airspace) Volume (Low Water Level) nJ 132,000 Initial Suppression Pool Volume ft3 122,900 (Low Water Level)

Initial Drywell Pressure psm 17.2 Initial Drywell Temperature OF 145 Initial Drywell Relative Humidity  % 20 Initial W ctwell Pressure psta 17.2 Initial Wetwell Temperature OF 95 Initial Wetwell Relative Humidity  % 100 Initial Suppression Pool Temperature "F 95 Break Type - Steam Break Break size(!) ft2 1.0 LPCI!Containment Cooling Pump Heat (per pump) hp 2000 Core Spray Pump Heat (per pump) hp 600 RHR Heat Exchanger Initiation Time sec 600 UHS (service water) Temperahtre "F 92 RHR Heat Exchanger K-Value per Loop (Containment Spray, 1 RHR loop, 1 RHR pump Btu/sec-°F 263 (9500 gpm), 1 RHR SW pump (4500 gpm), 1 HX)

LPCI Flow Rate per Pump 10,000 gpm RHR Flow to Drywell Spray in Spray Cooling gpm 8,690 Mode per Pump (After 600 sec.)

Drywell Spray Initiation Time sec 600 RHR Flow to Wetwell Spray in Spray Cooling gpm 810 Mode per Pump (After 600 sec.)

Wetwell Spray Initiaion Time sec 600 Core Spray Flow Rate per Pump gpm 6,250 NOTE (1): A total of six break sizes: 0.01, 0.05, 0.1, 0.25, 0.5, and 1.0 ft 2 .

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NED0-3380~ Revision 0 Non-proprietary Information -Class l (Public) 3.3 Evaluation Results The results ofthe SSLB analysis performed are presented and discussed below.

3.3.1 SSl,B Analyses As shown in Table 2, a spectrum of breaks with break sizes ofO.Ol ft 2, 0.05 ft2, 0.1 ft2, 0.25 ft 2, 0.50 fe, and l.O ttl at a Main Stearn Line have been analyzed, and containment pressure and temperature profiles have been obtained. The steam line breaks are the most limiting events for drywell temperature response since steam has higher energy content than liquid, on per unit mass basis.

The key results summarized in Table 2 demonstrate that the maximum drywell airspace temperature is (( l], maximum wetwell airspace temperature is (( )), and the maximum suppression pool temperature is (( )).

Figures 1 through 24 show containment pressure and temperature responses up to 1 day (86,400 seconds) for the six steam line break cases.

((

))

3.3.2 SSLB vs DBA-LOCA Results A review of the SSLB analysis results, summarized in Table 2, revealed that the peak pool temperature of (( )) for the 0.01 ft2steam line break is less than the peak pool temperature value of (( )) calculated for the DBA-LOCA case reported in Section 1.2.

It is to be noted that both SSLB cases and the DBA-LOCA case were analyzed using the RHR heat exchanger K-value of263.0 Btu/sec-°F.

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Table 2 Summary Key Results: With Containment Sprays

> ..J*>'*>*. . .; *.::

0.25 i*'"o.so**

.* .. * . c.:icl!: .. .:****'*:*

0.05 !***. 0.1 Break Size 0.01 ft 2 n7 rr 1.0 ft~ * *.*

';. ....*.*,...  ; *. >~ .**:.<*.*. .*.  ;

i

ftz *..*.

• . 'r< ..

Parameter

• Peak drywell temperature, °F ((

At time, seconds

• Peak drywell pressure, psia At time, seconds

• Peak wetwell temperature, op At time, seconds

~-Peak wctwellQ~:~ssurc_,_psia At time, seconds

• Peak pool temperature, op At time, seconds ))

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((

))

Figure 1 Drywell Airspace Temperature Response For 0.01 ce Steam Line Break 15

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((

))

Figure 2 Suppression Pool Temperature Response For 0.01 ft2 Steam Line Break 16

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((

))

Figure 3 Wetwell Airspace Temperature Response For 0.01 ft 2 Steam Line Break 17

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((

))

Figure 4 Drywell and Wetwell Airspace Pressure Response For 0.01 ft 2 Steam Line Break 18

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((

))

Figure 5 Drywell Airspace Temperature Response For 0.05 ft2 Steam Line Break 19

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((

))

Figure 6 Suppression Pool Temperature Response For 0.05 ft 2 Steam Line Break 20

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((

))

Figure 7 Wetwell Airspace Temperature Response For 0.05 ff Steam Line Break 21

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((

))

Figure 8 Drywell and \Vetwell Airspace P1*essure Response For 0.05 ft2 Steam Line Break 22

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[(

))

Figure 9 Drywell Airspace Temperature Response For 0.10 fr Steam Line Break 23

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((

))

Figure 10 Suppression Pool Temperature Response For 0.10 ft 2 Steam Line Break 24

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((

))

Figure 11 Wetwell Airspace Temperature Response For 0.10 ft 2 Steam Line Break 25

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((

))

Figure 12 Drywell and Wetwell Airspace Pressure Response For 0.10 ff Steam Line Break 26

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((

))

Figure 13 Drywell Airspace Temperature Response For 0.25 te Steam Line Break 27

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((

))

Figure 14 Suppression Pool Temperature Response For 0.25 ft 2 Steam Line Break 28

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((

))

Figure 15 Wetwell Airspace Temperature Response For 0.25 re Steam Line Break 29

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((

))

Figure 16 Drywell and W etwell Airspace Pressure Response For 0.25 ft 2 Steam Line Break 30

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((

))

Figure 17 Drywell Airspace Temperature Response For 0.50 ft2 Steam Line Break 31

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((

))

Figure 18 Suppression Pool Temperature Response For 0.50 ff Steam Line Break 32

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((

))

Figure 19 Wetwell Airspace Temperature Response For 0.50 ft 2 Steam Line Break 33

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((

))

Figure 20 Drywell and Wetwell Airspace Pressure Response For 0.50 ft2 Steam Line Break 34

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((

))

Figure 21 Drywell Airspace Temperature Response For 1.0 te Steam Line Break 35

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((

))

Figure 22 Suppression Pool Temperature Response For 1.0 re Steam Line Break 36

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((

))

Figure 23 Wetwell Airspace Temperature Response For 1.0 ft 2 Steam Line Break 37

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((

))

Figure 24 Drywell and Wetwell Airspace Pressure Response For 1.0 ft 2 Steam Line Break 38

NED0-33808 Revision 0 Non-proprietary Infmmation -('lass I (Public) 3.4 Conclusion Based on the results described above, it is concluded that the maximum drywell temperature with credit lor containment sprays (drywell and wetwell) is (( )), which remains below the 340 °F limit.

As noted above, the SSLB analyses are performed at 3528 MWt (102% of3458 MWt), which is equivalent to 100.4% of the Thermal Power Optimization (TPO) power level of35l4 MWt and it meets the power uncertainty requirement defined in Reference 6. Therefore, these SSLB analysis results support the TPO power level of 3514 MWt.

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4.0 REFERENCES

I. GE Nuclear Energy, Peach Boltom Power R.crate Project Engineering Report",

NEDC-32230P, March 1994.

2. ANSVANS-5.1-1979, "American National Standard f(H Decay Heat Power in Light Water Reactors",.

3. SIL-636, "Additional Terms included in Reactor Decay Heat Calculations," Revision I, June 6, 2001.

4. Peach Bottom 2&3 OPL-4A, Revision 0, dated September 5, 2002.

5. Bechtel Power Corporation, BN TOP-3, "Performance and Sizing of Dry Pressure Containments," Revision 4.

6. GE Nuclear Energy, "Safety Analysis Report for Peach Bottom Power Station Units 2 & 3 Thermal Power Optimization," NEDC-33064P, Revision 2,0ctober 2002 40

AFFIDAVIT GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, James F. Harrison, state as follows:

(l) I am the Vice President Fuel Licensing of GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) that is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH proprietary report NEDC-33808P, "Peach Bottom Atomic Power Station Units 2 and 3 SIL 636 Evaluation of Small Break LOCA," Revision 0, dated January 2013. GEH proprietary information in NEDC-33808P is identified by a dark red dotted underline inside double square brackets,

(([l}l~...l~~!lt\.:l}~~_j~_.mL(,':)(_~tmP.t~,:>_:_)). Figure and large equation objects containing GEH proprietary information are identified with double square brackets before and after the object In each case, the superscript notation m refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relics upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C.

Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 as decided in Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over GEH or other companies.

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, that may include potential products of GEH.

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

NEDC-33808P, Revision 0 Affidavit Page 1 of 3

(5) To address I 0 CFR 2.390(b)(4), the infonuation sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by GEII, and is in l~1ct so held. The information sought to be withheld has, to the best of my knowledge and belicC consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of an analysis performed by GEH to support the Peach Bottom Residual Heat Removal Drywell Spray License Amendment Request. This analysis is part of the GEH LOCA methodology. Development of the LOCA methodology and supporting analysis, techniques, and information and their application for the design, modification, and processes were achieved at a significant cost GEH.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

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The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 25 1h day of January 2013.

~0--..h~~

James F. Harrison Vice President Fuel Licensing GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd Wilmington, NC 28401 james.harrison@ge.com NEDC-33808P, Revision 0 Affidavit Page 3 of 3