ML16083A394
| ML16083A394 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/23/2016 |
| From: | David Helker Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CAC MF6774, CAC MF6775, TAC ME9631, TAC ME9632 | |
| Download: ML16083A394 (11) | |
Text
Exelon Generation 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 March 23, 2016 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License No. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Changes to High Pressure Coolant Injection and Reactor Core Isolation Cooling Surveillance Requirements - Supplement 1 Response to Request for Additional Information
References:
- 1.
Exelon letter to the NRC, "License Amendment Request - Proposed Changes to the Technical Specifications for High Pressure Coolant Injection and Reactor Core Isolation Cooling Surveillance Test Pressure and Clarification of Surveillance Requirements," dated October 2, 2015 (ADAMS Accession No. ML15275A265)
- 2.
NRC Letter to Exelon, "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Extended Power Uprate (Amendments 293 and 296) (TAC Nos. ME9631 and ME9632)", dated August 25, 2014 (ADAMS Accession No. ML14133A046)
- 3.
E-mail from R. Ennis (NRC) to D. Neff (Exelon), "Draft RAI - HPCI &
RCIC SRs (CAC MF6774 & MF6775)," dated February 8, 2016 (ADAMS Accession No. ML16056A596)
In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) requested amendments to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3, respectively (Reference 1 ).
Specifically, the proposed changes would revise PBAPS Unit 2 and Unit 3 TS Surveillance Requirement (SR) 3.5.1.8 and SR 3.5.3.3 for the High Pressure Coolant Injection and Reactor Core Isolation Cooling Surveillance Test Pressure bands, respectively.
Additionally, the proposed changes would revise PBAPS Unit 2 and Unit 3 TS SRs 3.1. 7.10, 3.6.2.4.3 and 3.6.2.5.3 to remove changes provided in the PBAPS EPU Amendments (Reference 2) that have been determined to be unnecessary or overly conservative.
U.S. Nuclear Regulatory Commission Changes to HPCI and RCIC Surveillance Requirements - Supplement 1 Response to Request for Additional Information March 23, 2016 Page2 Reference 3 provided NRG Requests for Additional Information (RAls). Exelon agreed to provide a response to the questions by March 23, 2016. The attachment to this letter provides responses to the Reference 3 RAls.
Exelon has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the U.S. Nuclear Regulatory Commission in the referenced License Amendment Request (Reference 1 ).
The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Further, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania and the State of Maryland of this response by transmitting a copy of this letter along with the attachment to the designated State Officials.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact Mr. David Neff at (610) 765-5631.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rd day of March 2016.
Respectfully, David P. Helker Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC
Attachment:
Response to NRG Staff's Request for Additional Information cc:
USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Commonwealth of Pennsylvania S. T. Gray, State of Maryland w/attachment w/attachment w/attachment w/attachment w/attachment
Attachment Changes to HPCI and RCIC Surveillance Requirements - Supplement 1 Peach Bottom Atomic Power Station Units 2 and 3 NRC Docket Nos. so-2n and 50-788 Responses to NRC Staff's Request for Additional Information
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI Responses to NRC Staff's Request for Additional Information Attachment Page 1of8 By letter dated October 2, 2015, Exelon Generation Company, LLC (Exelon) submitted a License Amendment Request (LAR) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 (Reference 1 ). The proposed changes would revise PBAPS Unit 2 and Unit 3 TS Surveillance Requirement (SR) 3.5.1.8 and SR 3.5.3.3 for the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Surveillance Test Pressure bands, respectively. Additionally, the proposed changes would revise PBAPS Unit 2 and Unit 3 TS SRs 3.1.7.10, 3.6.2.4.3 and 3.6.2.5.3 to remove changes provided in the PBAPS Extended Power Uprate (EPU) Amendments (Reference 6) that have been determined to be unnecessary or overly conservative.
In an email dated February 8, 2016, from the NRC (Rick Ennis) to Exelon (David Neff)
(Reference 4), the NRC provided Requests for Additional Information (RAls) seeking clarification of certain issues related to the LAR. This attachment provides responses to those RA ls.
SCVB-RAl-1 The SPS and the DWS are safety-related subsystems or operating modes of the RHR system which perform containment cooling function during a design basis accident. These modes have associated Limiting Conditions of Operation (LCOs) with SRs. In the licensee's letter dated October, 2, 2015, Attachment 1, Section 3.0, last paragraph under heading "SR 3.6.2.4.3 and SR 3.6.2.5.3 Deletions" provides the following justification for deleting SR 3.6.2.4.3 and SR 3.6.2.5.3 for the LCOs for these RHR modes:
While the use of the alternate power supply could be used for operation of the SPS and DWS sub-systems, the need for this capability would involve more than a single failure. The alternate power supply capability for the specific RHR valves is not relied upon for the design function of the SPS and DWS subsystems. The accident analyses do not rely on the use of the alternate power supply for the SPS and DWS sub-systems assuming any single failure. The License Amendment Request for the PBAPS EPU [Extended Power Uprate]
misinterpreted the FMEA [Failure Modes and Effects Analysis] to rely on the use of the alternate power supply for all modes of containment cooling instead of only the SPC function.
a) The above statement does not clearly justify deletion of SR 3.6.2.4.3 and SR 3.6.2.5.3 for the SPS and DWS modes while similar SR 3.6.2.3.3 for the RHR Suppression Pool Cooling (SPC) mode is required, even though all three (SPC, SPS, and DWS) are safety-related modes of RHR system credited for containment cooling function during a design basis accident. Describe the safety design basis functions of the SPS and DWS modes and further explanation and reasons for deleting the SRs.
b) Please explain why the need for the alternate power supply capability for the SPS and DWS modes would involve more than a single failure. Describe the possible failures that would require the need for SPS and DWS to perform their safety design basis functions.
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI
RESPONSE
Attachment Page 2 of 8 a) The safety design basis functions for the RHR system, including the Suppression Pool Spray (SPS) and Drywell Spray (DWS) modes of RHR, are relied upon in the accident analysis to maintain primary containment conditions within design limits. The design basis functions of the SPS and DWS modes of RHR are described primarily in Updated Final Safety Analysis Report (US FAR) Sections 5.2, 14.1 O and Appendix M.
The following background information is provided to describe the relationships between the SPS and DWS modes of RHR, the emergency electrical power supplies and the need for the alternate power supplies to compensate for a postulated emergency electrical power supply failure. The RHR system is designed with two subsystems, with the A subsystem comprised of the A and C RHR trains powered by the E1 and E3 Diesel Generators (DGs),
and the B subsystem compromised of the Band D RHR trains powered by the E2 and E4 DGs. The A subsystem SPS and DWS injection valves are powered by the E3 DG. The corresponding B subsystem valves are powered by the E4 DG. In the event of a Loss of Offsite Power (LOOP) and a single active failure of the E3 or E4 DG, the SPS and DWS valves in the subsystem with the failed DG will not have emergency electrical power.
The need for the alternate power supply for the RHR Heat Exchanger (Hx) cross-tie, RHR flow control and RHR Hx High Pressure Service Water (HPSW) outlet valves was discussed in Supplement 5 of the PBAPS EPU LAR (Reference 8). The RHR alternate power supply was installed to ensure that these valves are provided with the capability of receiving power from either their normal source or from an alternate safety related power supply. For the Recirculation System Line Break (RSLB) Design Basis Accident (OBA) Loss of Coolant Accident (LOCA) analysis, including EPU power level and no Containment Atmosphere Pressure (CAP) credit, primary containment heat removal is fulfilled utilizing one RHR pump, the RHR Hx cross-tie, two RHR Hx's and two HPSW pumps for primary containment heat removal. Before the EPU amendment, this analysis required only one RHR pump and one HPSW pump at one hour after the OBA LOCA.
The bounding accident analysis for the Large Break LOCA is the RSLB OBA LOCA in one unit with a LOOP and safe shutdown of the second unit. The bounding accident analysis for the peak primary containment suppression pool, Drywell airspace and OW shell temperature is the Small Steam Line Break (SSLB) LOCA in one unit with a LOOP and safe shutdown of the second unit. Both events must be accomplished with the limiting single failure of the loss of one emergency electrical AC power source (i.e., loss of one DG). The results of the accident analyses for primary containment temperature and pressure and use of the RHR system are discussed in UFSAR Chapter 5 Section 5.2.4.3.1 and Chapter 14, Sections 14.10.3 and 14.10.4.
In the case of the RSLB OBA LOCA with safe shutdown of the second unit and the loss of either the E3 or E4 DG, the analysis assumes that one hour into the event, only one RHR pump will be left running on the accident unit and it is powered by the division with the failed diesel. The loading of the emergency electrical AC power supply is done to levelize DG loading. At this time, the DWS and SPS function will no longer be available since the failed diesel would power the necessary valves in that subsystem. The RHR cross-tie is placed into service at one hour, using the valves powered by the alternate power supply, and the containment heat removal function is performed by the Coolant Injection Cooling (CIC) mode of RHR. Therefore, in this scenario, even if the DWS and SPS mode was used in the
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI Attachment Page 3 of 8 first hour, the DWS and SPS function is not utilized after one hour, when the alternate power supply is needed.
In the case of the RSLB OBA LOCA with safe shutdown of the second unit and loss of either the E1 or E2 DG, the DWS and SPS would remain available for the containment cooling function; however, this function could also be performed by the CIC mode as described for the loss of E3 or E4 DG. Therefore, it is not necessary to utilize the DWS or SPS mode with the RHR Hx cross-tie in service after one hour after the event. This analysis does not rely on the SPS and DWS modes of AHR coincident with use of the AHR Hx cross-tie operation. Consequently, the alternate power supply is not relied upon for the SPS and DWS modes of operation for the RSLB OBA LOCA analysis.
For the SSLB LOCA analysis, the SPS and DWS modes are credited, including use of the AHR Hx cross-tie, AHR flow control and AHR Hx HPSW outlet valves. For this event, Core Spray pumps are not required, and therefore, more emergency electrical power is available to equalize electrical loading between both units. This allows use of the AHR subsystem, including the AHR Hx cross-tie, AHR flow control and AHR Hx HPSW outlet valves, in the emergency electrical AC power division with two DGs to mitigate the accident unit using CIC, SPC and/or SPS with DWS while maintaining adequate emergency power to allow safe shutdown of the non-accident unit. In this analysis, the alternate power supply to the AHR Hx cross-tie, AHR flow control and AHR Hx HPSW outlet valves is not needed for any of the modes of AHR including SPS and DWS.
Since there is no safety design analysis that relies on the SPS and DWS mode of AHR concurrently with the alternate power supply for the AHR Hx cross-tie, AHR flow control and AHR Hx HPSW outlet valves, SR 3.6.2.4.3 and SR 3.6.2.5.3 are unnecessary to assure the quality of the SPS and DWS system components are maintained to meet the system design basis.
b) Since the safety design analyses for the SPS and DWS functions of AHR already include the limiting single failure, an event that could result in the use of the alternate power supply capability to support operation of the DWS and SPS would need to involve an additional failure. This additional failure consideration is beyond the SPS and DWS design basis.
There are no possible single failures that would require use of the alternate power supply for the AHR flow control and RHR Hx HPSW outlet valves, to support the SPS or DWS modes of AHR in the performance of their safety design basis functions.
SCVB-RAl-2 Is SPS and DWS credited for containment atmosphere cleanup as required by 1 O CFR Part 50 Appendix A, General Design Criteria (GDC)-41 1, which requires systems to control fission products released into the reactor containment be provided to reduce the concentration and quality of fission products released to the environment following postulated accidents?
a) In case SPS and DWS are not credited, please explain which systems are provided to meet the AEC draft GDCs equivalent to current GDC-41.
1 The PBAPS evaluation of the comparable 1967 Atomic Energy Commission (AEC) proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H, Draft GDC-11, Draft GDC-17, Draft GDC-69, and Draft GDC-70.
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI Attachment Page 4 of 8 b) In case the SPS and DWS are credited for fission product cleanup, the RHR control valve and cross-tie valves must be used for achieving a positive net positive suction head (NPSH) margin for the RHR pumps without using containment accident pressure (CAP). Please explain why an alternate safety-related power source for these valves would not be needed in case of loss of the normal safety-related power source.
RESPONSE
Regarding questions a) and b), the SPS and DWS modes of RHR are not credited for fission product cleanup. To meet the draft GDC-70 (comparable to final GDC 41) the licensing basis and safety analysis for the containment atmosphere cleanup relies on the fission product control systems and structures (UFSAR Section H.2.9). The systems that are credited are described in UFSAR Chapters 3, 5 and 14 and include Primary Containment and Primary Containment Isolation Valves (UFSAR 5.2 and 14.9), Standby Liquid Control (SLC) system (UFSAR 3.8 and 14.9) and Secondary Containment I Standby Gas Treatment System (UFSAR Sections 5.3 and 14.9). Section 3.2.2.4 of NRC SER for PBAPS application of Alternative Source Terms (AST)
License Amendments 269 and 273 (Reference 5) details the NRC approval of the radioactive material deposition modelling in the primary containment. The approved model is based on gravitational deposition of aerosols and elemental iodine deposition on wetted surfaces inside primary containment due to natural convection and condensation. The atmosphere scrubbing and additional wetted surface area due to SPS and DWS operation are not credited in the AST analysis. The SLC system is used to maintain the Suppression Pool water pH above 7.0 to keep ionic iodine dissolved in water and lower the release of elemental iodine inside of primary containment. This AST modelling and SLC system operation was applied in the EPU analysis as approved in PBAPS Unit 2 and Unit 3 EPU License Amendments 293 and 296 (Reference 6).
SCVB-RAl-3 Section 4.3 of Attachment 1 to the licensee's letter dated October 2, 2015, states, in part, that:
These specific changes delete the SR 3.6.2.4.3 from the Suppression Pool Spray (SPS) TS section and delete SR 3.6.2.5.3 from the Drywell Spray (DWS) TS section since the alternate power supply capability is not necessary for the SPS or DWS sub-systems of the AHR system to perform their design functions.
Please explain why alternate power supply capability is not necessary for the SPS or DWS sub-systems of the RHR system to perform their design functions.
RESPONSE
As described in the response to SCVB-RAl-1 above, the alternate power supply for the RHR Hx cross-tie, AHR flow control and RHR Hx HPSW outlet valves is not relied upon in the design basis analysis for the SPS and DWS modes of the RHR system to perform their design functions.
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI SCVB-RAl-4 Attachment Page 5 of 8 Section 4.3 of Attachment 1 to the licensee's letter dated October 2, 2015, states, in part, that:
... deletion of the SPS and OWS sub-system SRs 3.6.2.4.3 and 3.6.2.5.3 do not affect the ability of these systems to perform their design functions. These proposed changes are administrative in nature and have no effect on plant operation.
Please explain why the deletion of SRs 3.6.2.4.3 and 3.6.2.5.3 is considered as administrative in nature.
RESPONSE
The original LAR (Reference 1) and the response to SCVB-RAl-1 above show that the alternate power supply capability is not relied upon for the SPS and OWS modes of the RHR system to meet their design functions. The routine test of this power supply transfer capability does not meet the 1 O CFR 50.36(c)(3) criteria to be included as a SR for the SPS and OWS subsystems.
The TS requirement for testing the alternate power supply capability is retained under TS SR 3.6.2.3.3, Suppression Pool Cooling. Removal of the two unnecessary SRs is considered non-technical, and hence administrative, based on the lack of a requirement to have the alternate power supply to support the design functions of the SPS and OWS modes of the RHR system.
SCVB-RAl-5 As discussed in Reference 3, Section 2.6.1.2.1, under the heading "Loss-of-Coolant Accident Loads", the Emergency Operating Procedures (EOPs) require OWS prior to Wetwell (WW) pressure exceeding 9 psig as stated below:
However, emergency operating procedures (EOPs) for PBAPS include direction to initiate OW sprays prior to WW pressure exceeding 9.0 psig. Containment analyses performed for PBAPS EPU have shown that WW [Wetwell] pressure will exceed this OW [Drywell] spray initiation pressure of 9.0 psig before 900 seconds following initiation of the event. Initiation of OW sprays will rapidly reduce OW pressure and stop chugging. [emphasis added]
As discussed in Reference 3, Section 2.6.5.1, under the heading "Pool Temperature Response
- RSLB [recirculation Suction Line Break] OBLOCA," the shutdown cooling of the non-accident unit under a dual unit interaction uses the OWS and SWS spray modes to maintain the containment shell temperature less than the shell design temperature as stated below:
The non-accident unit containment cooling interruption is assumed for a period of 10-minutes due to the dual unit interaction. After the 10-minute interruption, containment cooling on the non-accident unit is restored with the same containment cooling configuration as existed prior to the dual unit interaction interruption. When RPV pressure reaches 150 psig, the analysis assumes that the operators will maintain the RPV at this pressure. Containment cooling is maintained using RHR SPC mode. The containment spray mode of RHR may be initiated to maintain the containment shell temperature less than the containment design temperature of 281°F. [emphasis added]
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI Attachment Page 6 of 8 As discussed in Reference 3, Section 2.6.5.1, under the heading "Pool Temperature Response
- Small Steam Break LOCA," for single unit analysis, it states that:
At 10 minutes, operators turn off two of the RHR pumps, and align the remaining RHR pump to provide containment cooling with a flow of 8600 gpm through one RHR heat exchanger, and one HPSW pump providing cooling flow to the RHR heat exchanger. When DW pressure exceeds 2.0 psig, operators initiate WW spray, with the remainder of the RHR flow remaining in SPC mode. Prior to WW pressure exceeding 9.0 psig. operators initiate OW spray and stop all SPC flow.
When bulk suppression pool temperature exceeds 110°F (but not earlier than 1 O minutes following initiation of the event), operators initiate a controlled reactor vessel cooldown at 100°F per hour. At one hour from initiation of the event, operators establish the RHR heat exchanger cross-tie to the other RHR heat exchanger in the same loop, such that a total RHR flow rate of 8600 gpm is maintained to the OW and WW spray headers. When reactor pressure is decreased below the pressure permissive for NSOC (70 psig), operators maintain OW and WW spray cooling with the RHR heat exchanger cross-tie in service. and maintain reactor vessel pressure as low as possible to limit steam flow from the break. When bulk suppression pool temperature is below a pre-determined value (170°F for EPU), operators open one or more ADS valves and increase vessel water level in the reactor vessel to the MSL nozzles using the one loop of CS until water flows from the open ADS valves back to the suppression pool, establishing ASOC. OW and WW sprays continue to be used to provide containment and SPC, bulk reactor water temperature decreases below 200°F and cold shutdown is achieved prior to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> from initiation of the event, which is conservative for plants with cold shutdown defined at a higher temperature. The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-11 and the peak bulk suppression pool temperature at 187°F. [emphasis added]
As discussed in Reference 3, Section 2.6.5.1, under the heading "Pool Temperature Response
- Small Steam Break LOCA," for dual unit interaction analysis, it states that:
At 10 minutes, operators turn off two of the RHR pumps and align the remaining RHR pump to provide containment cooling using with a flow of 8600 gpm through one RHR heat exchanger and one HPSW pump providing cooling flow to the RHR heat exchanger. When OW pressure exceeds 2.0 psig. operators initiate WW spray, with the remainder of the RHR flow remaining in SPC mode. Prior to WW pressure exceeding 9.0 psig. operators initiate OW spray and stop all SPC flow. When bulk suppression pool temperature exceeds 110°F (but not earlier than 1 O minutes following initiation of the event), operators initiate a controlled reactor vessel cooldown at 100°F per hour. At one hour from initiation of the event. operators establish the RHR heat exchanger cross-tie to the other RHR heat exchanger in the same loop, such that a total RHR flow rate of 8600 gpm is maintained to the OW and WW spray headers. At one-hour following the start of reactor depressurization it is assumed that a LOCA signal on the SSLB LOCA unit may occur on HOWP commensurate with low reactor pressure. This timing for the HOWP/low reactor pressure LOCA signal on the small break
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI Attachment Page 7 of 8 LOCA/accident unit is based on the depressurization rate of 100°F/hr mentioned above. [emphasis added]
As stated in the underlined portions of the above Sections of Reference 3, the large break LOCA and the Small Steam Break LOCA containment cooling analysis credit the DWS and SPS modes of the AHR system. In case these modes are not available, due to loss of normal power source to the AHR cross-tie, AHR flow control and AHR heat exchanger HPSW outlet valves, describe the alternate safety system for reducing chugging, mitigating large and small steam break LOCAs, and safe shutdown cooling of the non-accident unit that should be credited in the
- analysis,
RESPONSE
Regarding the first cited section of Reference 3 in SCVB-RAl-5, the LOCA primary containment dynamic loads evaluation, including chugging loads, is discussed in PBAPS UFSAR Appendix M, Section M.3.7. This analysis is primarily based on the short-term RSLB OBA LOCA analyses and compliance with generic criteria developed through testing programs. The Mark I Containment Program Load Definition Report (LOR) (Reference 7) defines the onset and duration times for chugging based on break size. Although the chugging duration times used in the original PBAPS LOR remained applicable and bounding for the design analysis at EPU power levels, the basis of the chugging duration times for the small break LOCA (SBA) and intermediate break LOCA (IBA) changed from manual operation of Automatic Depressurization System (ADS) to operation of the DWS mode of AHR. For the RSLB OBA LOCA, the LOR is unaffected for the design analysis at EPU power levels. Since use of the DWS mode of AHR for the SBA and IBA will be completed within the first hour of the event, use of the alternate power supply to the AHR cross-tie, AHR flow control and AHR Hx HPSW outlet valves is not needed for this function.
Regarding the second cited section of Reference 3 in SCVB-RAl-5, the analysis for shutdown of the non-accident unit does not rely on the AHR Hx cross-tie capability and does not need the alternate power supply to the AHR cross-tie, AHR flow control and AHR Hx HPSW outlet valves to support operation of the SPS and DWS modes of AHR.
Regarding the third and fourth cited sections of Reference 3 in SCVB-RAl-5, the SSLB LOCA analysis does not need to rely on the alternate power supply to the AHR cross-tie, AHR flow control and AHR Hx HPSW outlet valves to support operation of the SPS and DWS modes of AHR. This is discussed in the response to SCVB-RAl-1 above.
In conclusion, there is no case within the design basis where the SPS or DWS modes of AHR rely on the use of the alternate power supply to the AHR Hx cross-tie, AHR flow control and AHR Hx HPSW outlet valves.
Changes to HPCI and RCIC Surveillance Requirements Supplement 1 Responses to NRC Staff's RAI REFERENCES Attachment Page 8 of 8
- 1. Exelon letter to NRC dated October 2, 2015, "License Amendment Request Proposed Changes to the Technical Specifications for High Pressure Coolant Injection and Reactor Core Isolation Cooling Surveillance Test Pressure and Clarification of Surveillance Requirements," (ADAMS Accession No. ML15275A265).
- 2. Exelon letter to NRC dated September 28, 2012, "License Amendment Request - Extended Power Uprate," (ADAMS Accession No. ML12286A012).
- 3. General Electric-Hitachi Report NEDC-33566, Revision 0, Attachment 6 to Exelon letter to NRC dated September 28, 2012 (Reference 2), "NEDC-33566P, Safety Analysis Report for Exelon Peach Bottom Atomic Power Stations Units 2 and 3 Constant Pressure Power Uprate," (Non-Public, Proprietary).
- 4. NRC Memorandum from R. Ennis (NRC) to D. Broaddus (NRC), "Peach Bottom Atomic Power Station, Units 2 and 3, Draft Requests for Additional Information (CAC MF6774 &
MF6775)," dated February 22, 2016 (ADAMS Accession No. ML16056A596).
- 5. NRC Letter to Exelon, "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Application of Alternative Source Term Methodology," (Amendments No.
269 and 273), September 5, 2008 (ADAMS Accession No. ML082320406).
- 6. NRC Letter to Exelon, Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Extended Power Uprate (Amendments 293 and 296) (TAC Nos. ME9631 and ME9632), dated August 25, 2014 (ADAMS Accession No.ML14133A046).
- 7. General Electric Company, "Mark I Containment Program Load Definition Report," NED0-21888, Revision 2, November 1981.
- 8. Exelon letter to NRC dated July 27, 2013, "Supplemental Information Supporting Request for License Amendment Request -Extended Power Uprate - Supplement 5," (ADAMS Accession No. ML13182A025).