ML17220A214
| ML17220A214 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 08/08/2017 |
| From: | David Gudger Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MF9289, TAC MF9290 | |
| Download: ML17220A214 (13) | |
Text
Exelon Generation@
200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 10 CFR 50, Appendix K August 8, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Measurement Uncertainty Recapture License Amendment Request -
Supplement 3 Response to Request for Additional Information
References:
1.
Exelon letter to the NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated February 17, 2017 (ADAMS Accession No. ML17048A444)
- 2.
- 3.
Email from R. Ennis (USNRC) to D. Neff (Exelon), "Peach Bottom Atomic Power Station, Units 2 and 3 - Request for Additional Information Regarding License Amendment Request for Measurement Uncertainty Recapture Power Uprate (TAC Nos. MF9289 and MF9290)," dated July 10, 2017 (ADAMS Accession No. ML17191A349)
Exelon letter to NRC, "Measurement Uncertainty Recapture License Amendment Request - Supplement 2 Response to Request for Additional Information," dated July 13, 2017 (ADAMS Accession No. ML17195A285)
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 the Renewed Facility Operating Licenses to implement an increase in rated thermal power from 3951 Megawatts-Thermal (MWt) to 4016 MWt. During their technical review of the application, the NRC Staff identified the need for additional information. Reference 2 provided the Requests for Additional Information (RAls) from the NRC Reactor Systems Branch (SRXB). Attachment 1 to this letter provides the responses to the RAls.
During the preparation of responses to the RAls, a typographical error was identified in a figure that was provided in Attachments 5, 7 and 14 of Reference 1. Specifically, in each of these attachments to the LAR, Figure 1-2, "Reactor Heat Balance - TPO Power, 100% Core
MUR LAA Supplement 3 Response to Request for Additional Information August 8, 2017 Page2
- Flow, 11 has an incorrect core inlet enthalpy of 516.8 Btu/lbm. The correct value is 521.3 Btu/lbm.
This error is limited to the core inlet enthalpy value contained in Figure 1-2 and all related LAA analysis and information are unaffected.
Additionally, an error was identified in Figure 2 on page 8 in the Attachment of Reference 3.
The operating point described in the box is correct, however, the operating point shown on the graph was inadvertently shifted. A replacement page 8 of the Attachment of Reference 3 with the corrected Figure 2 is provided in Attachment 2 to this response.
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 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 10 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 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 9th day of August 2017.
Respectfully, David T. Gudger Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1 - Response to Requests for Information from NRC Review Branch SRXB 2 - Replacement Page 8 for PBAPS MUR LAA Supplement 2 Attachment cc:
USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Pennsylvania Bureau of Radiation Protection S. T. Gray, State of Maryland Peach Bottom Atomic Power Station, Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 Response to Request for Additional Information from NRC Review Branch SRXB
MUR LAR Supplement 3 Responses to Requests for Information Responses to NRC Staff's Request for Additional Information Page 1of8 By application dated February 17, 2017, as supplemented by letter dated March 20, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos.
ML17048A444 and ML17080A067, respectively), Exelon Generation Company, LLC (Exelon, the licensee) submitted a License Amendment Request (LAR) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The amendments would revise the Renewed Facility Operating Licenses and Technical Specifications (TSs) to implement a Measurement Uncertainty Recapture (MUR) power uprate. Specifically, the amendments would authorize an increase in the maximum licensed thermal power level from 3,951 Megawatts-Thermal (MWt) to 4,016 MWt which is an increase of approximately 1.66%.
In an email dated July 10, 2017, from the NRC (Rick Ennis) to Exelon (David Neff) (ADAMS Accession No. ML17191A349), the NRC provided Requests for Additional Information (RAls) seeking clarification of certain issues related to those RAls. A conference call between the NRC staff and the Exelon staff was held on July 10, 2017, to discuss the draft RAI questions. Based on this call, RAl-SRXB-1 was withdrawn since the information needed has already been docketed. Exelon agreed to provide a response to the RAls by August 9, 2017.
SRXB-RAl-1 Question withdrawn.
SRXB-RAl-2 On page 3-16 of the TSAR 1, it states that:
The conclusion in the LOFW [loss-of-feedwater] analysis-of-record based on SAFER/GSTRM will remain valid with SAFER/PRIME as the water level response between the SAFER/GSTRM and the SAFER/PRIME methodologies are expected to be essentially the same.
The NRC staff understands that the currently approved thermal conductivity degradation (TCD) model is incorporated in the PRIME code, not in the GSTRM code. The staff also believes that degraded fuel thermal conductivity may result in higher fuel stored energy, and that this additional stored energy as an initial condition in the fuel may lead to a higher boil-off rate resulting in a reduced water level in the core during a LOFW event. Explain why TCD is expected to have no impact on the calculation of water level during a LOFW event.
1 Attachment 5 to the licensee's application, GE - Hitachi Nuclear Energy (GEH), "Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2, and 3, Thermal Power Optimization," NEDC-33873P, Revision 0, dated February 2017, summarizes the evaluations performed for PBAPS for the proposed MUR. This proprietary report is referred to as the TSAR (i.e., Thermal Power Optimization Safety Analysis Report). A public version of the TSAR, GEH report NED0-33873, is contained in Attachment 7 to Exelon's application.
MUR LAR Supplement 3 Responses to Requests for Information
RESPONSE
Page 2of8 During the LOFW event, which includes a reactor scram at the beginning of the event, the predicted fuel temperatures in the core would be higher with a fuel thermal model that accounts for TCD. This will affect the stored energy in the core at the start of the event. However, the stored energy dissipates over the period of time that the reactor water level falls to the low-level setting which initiates the Reactor Core Isolation Cooling (RCIC) system. At the time of the RCIC system initiation (approximately 68 seconds after the SCRAM), the predominant source of energy to the coolant is decay heat, which is unaffected by the initial stored energy and TCD.
Therefore, for the LOFW event analysis, the effect of stored energy (due to TCD) is small and insignificant considering the available margin to top of active fuel during the LOFW event analysis.
SRXB-RAl-3 In Section 4.2, "Emergency Core Cooling Systems (ECCS)," of Attachment 14 to the LAR (i.e.,
redline/strikeout of the TSAR), the statement below was deleted for the High Pressure Coolant Injection (HPCI) system, the Core Spray (CS) system and the Low Pressure Coolant Injection (LPCI) system:
The ability of the....... system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the......... system are within previous evaluations and the requirements are unchanged for TPO uprate conditions.
For the Automatic Depressurization System (ADS), the statement was replaced by a similar statement. Explain why the above statement is not valid for the HPCI, CS, and LPCI systems of the ECCS.
RESPONSE
The plant-specific evaluations for the HPCI, CS and LPCI systems conclude the systems continue to perform required safety functions because analyses performed at 102% of CL TP bound the TPO uprate conditions including thermal power uncertainty. Although the statement to this effect was inadvertently removed from TSAR Sections 4.2.1, 4.2.2 and 4.2.3 for HPCI, CS and LPCI systems, respectively, in MUR LAR Attachments 5, 7, and 14, it remains valid and applies to these systems. This conclusion is supported by the discussion in TSAR Section 4.3, "Emergency Core Cooling System Performance," and supplements the information provided in TSAR Sections 4.2.1, 4.2.2 and 4.2.3.
MUR LAR Supplement 3 Responses to Requests for Information SRXB-RAl-4 Page 3of8 In Table 9-2 of the TSAR, results for the anticipated transient without scram (ATWS) analysis at TPO were provided. Provide the following additional information:
- a. Specify at which state point in the power/flow map (TSAR Figure 1-1a) the following limiting results occur at TPO power:
i)
Peak vessel bottom pressure ii) Peak cladding temperatures for A TWS and A TWS with instability (A TWSI)
If the state points are different from that of the current analysis-of-record for CL TP/MELLL~+, explain the reason.
RESPONSE
Limiting ATWS events for peak vessel bottom pressure, ATWS Peak Cladding Temperature (PCT), and ATWSI PCT, are all performed at the same statepoint corresponding to point J' for MELLLA+ and point J forTPO on TSAR Figure 1-1a. The statepoints are the same (rated power, minimum flow) other than the slightly higher TPO power and higher core flow on the same MELLLA+ boundary.
SRXB-RAl-5 The table in Section 4.1 of the TSAR does not provide any information regarding the containment temperature response for Equipment Environmental Qualification (EEQ). For the CL TP level (i.e., EPU thermal power level), the most limiting containment temperature response for the EEQ of equipment was obtained in the long term analysis for a Small Steam Line Break (SSLB) accident. Provide the analysis results, including the limiting break size under the TPO uprate condition.
RESPONSE
The existing plant-specific containment analyses for Extended Power Uprate (EPU), including the SSLB analysis, were performed at 102% of CL TP (EPU) as documented in Attachment 6 to the EPU LAR (Reference 5-1), Section 2.6.3 and Table 2.6-2, as well as in the response to EPU SCVB-RAl-2 in EPU LAR Supplement 7 (Reference 5-2). Because 102% of CL TP bounds the TPO uprate power, the containment analyses, including the most limiting containment temperature response for EEQ and the limiting break size, bound those at TPO power; therefore, no additional containment analyses at TPO power are needed.
REFERENCES 5-1 Letter from Kevin F. Borton (Exelon) to NRC Document Control Desk, "License Amendment Request-Extended Power Uprate," September 28, 2012(ML12286A012).
5-2 Letter from Kevin F. Borton (Exelon) to NRC Document Control Desk, "Extended Power Uprate License Amendment Request - Supplement 7 Response to Request for Additional Information - Extended Power Uprate," July 31, 2013 (ML13213A285).
MUR LAR Supplement 3 Responses to Requests for Information SRXB-RAl-6 Page 4 of 8 The table in Section 4.1 of the TSAR does not provide any information regarding the peak containment wall temperature for structural analysis. For the CL TP level (i.e., EPU thermal power level), the most limiting containment wall temperature was obtained in the long term analysis for a SSLB accident. Provide the analysis results, including the limiting break size under the TPO uprate condition.
RESPONSE
The existing plant-specific containment analyses for EPU, including the SSLB) analysis, were performed at 102% of CLTP as documented in Attachment 6 to the EPU LAR (Reference 5-1),
Section 2.6.3 and Table 2.6-2, as well as in the responses to EPU SCVB-RAl-2 and SCVB-RAl-6 in EPU LAR Supplement 7 (Reference 5-2). Because 102% of CL TP (EPU) power bounds the TPO uprate power, the containment analyses, including the peak containment wall temperature for structural analysis and limiting break size, bound those at TPO power; therefore, no additional containment analyses at TPO power are needed.
SRXB-RAl-7 With respect to TSAR Section 4.1.2, provide justification why the TPO uprate is determined to have no effect on the current evaluation of Generic Letter (GL) 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves."
RESPONSE
Safety related Motor Operated Valves (MOVs) in the GL 96-05 Program are contained in the following systems: Main Steam (MS), Reactor Recirculation (RR), Feedwater (FW), Reactor Water Cleanup (RWCU), Residual Heat Removal (RHR), Reactor Core Isolation Cooling (RCIC), Core Spray (CS), High Pressure Coolant Injection (HPCI), High Pressure Service Water (HPSW), Emergency Service Water (ESW), Reactor Building Closed Cooling Water (RBCCW),
Drywell Cooling, and Emergency Cooling Tower (ECT).
Reactor pressure is not changed by TPO, therefore analyzed MOV conditions are not impacted for MS, RR and RWCU systems.
MOVs in the RHR, RCIC, CS, HPCI and RBCCW systems are evaluated for post-accident conditions which are analyzed at 102% of CL TP which bounds TPO. Additionally, RCIC and HPCI system valves are analyzed and found to be acceptable for increased peak reactor pressure under A TWS conditions at TPO.
TPO has no effect on the HPSW, ESW and ECT systems' maximum motor ambient temperatures, maximum line pressures or differential pressures. Therefore, analyzed MOV conditions are not impacted for HPSW, ESW, and ECT systems.
The only GL 96-05 Program MOVs in the FW system are the startup recirculation isolation valves, which have a safety function to close only during startup and shutdown and are normally closed at full power operation. Operating conditions during startup and shutdown are not impacted by TPO.
MUR LAR Supplement 3 Responses to Requests for Information Page 5of8 The GL 96-05 Program MOVs in the Drywall Cooling system are evaluated for post-LOCA drywall pressure, which is analyzed at 102% of CL TP and therefore bounds TPO. The maximum analyzed ambient temperature is based on a ruptured main steam line. As the main steam conditions are not changed by TPO, the analyzed MOV ambient temperature is not changed by TPO.
SRXB-RAl-8 NRC GL 96-06, "Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions, 11 identifies the following potential problems with equipment operability and containment integrity during design-basis accident (OBA) conditions: (1) cooling water systems serving the containment air coolers may be exposed to water hammer during postulated accident conditions; (2) cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated accident conditions; and (3) thermally induced over-pressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. GL 96-06 questioned whether the higher heat loads at accident conditions could potentially cause steam bubbles, water hammer, and two-phase flow due to the higher outlet temperatures from cooled components, particularly the containment fan coolers.
With respect to TSAR Section 4.1.4, provide justification why the TPO uprate does not affect the current evaluation of the problems identified in GL 96-06 described above.
RESPONSE
Post-accident drywall conditions are analyzed at 102% of CL TP, which bounds TPO uprate.
The PBAPS EPU analysis performed to support the response to SCVB RAl-21 (Reference 8-1) for GL 96-06 is not impacted by TPO.
REFERENCE 8-1 Supplement 10 to Extended Power Uprate License Amendment Request, NRC Docket Nos. 50-277 and 50-278, Response to Request for Additional Information - SCVB RAl-21 and RAl-25(ML13241A418).
SRXB-RAl-9 With respect to TSAR Section 4.1.5, provide justification why the TPO uprate is determined to have no effect on the GL 89-16, "Installation of a Hardened Wetwell Vent.
11
RESPONSE
The required relieving capacity of the hardened wetwell vent is analyzed at 102% of CL TP which bounds TPO conditions.
MUR LAR Supplement 3 Responses to Requests for Information SRXB-RAl-10 Section 4.2.5 of the TSAR states:
Page 6of8 A conservative error was identified in the NPSH [net positive suction head]
evaluations for Appendix R Cases A 1, C 1 A and C 1 B at EPU conditions. These evaluations should have used a service water (SW) temperature of 86°F as indicated in Table 9.2f of the EPU LAR, but instead used 92°F, and therefore have been re-performed with the corrected temperature. The SW temperature of 92°F is a TS limit, while 86°F is a nominal value based on a statistical analysis of a five-year sampling of data for the months of June, July, August and September.
- a. Indicate the Attachment number to the EPU LAR letter dated September 28, 2012, to which Table 9.2f belongs. Note that the Table 2.5-1 of the Attachment 6 to the EPU LAR letter dated September 28, 2012 (i.e., PUSAR), states the correct SW temperature of 86°F.
- b. Table 2.5-1 of Attachment 6 of EPU LAR dated September 28, 2012, provides Appendix R fire event key input parameters including the SW temperature of 86°F. As stated above, the EPU analysis was erroneously performed using a SW temperature of 92°F instead of 86°F.
Confirm that in the revised analysis at the TPO uprate, the remaining parameters, besides core thermal power and SW temperature, listed in Table 2.5-1 of the PUSAR, were not changed. If parameters were changed, provide justification in case the change in any of the parameters reduces the conservatism in the suppression pool temperature response.
RESPONSE
1 O.a. Table 9-2f, "Key Changes in Inputs to NPSH Calculations for EPU - Appendix R Fire Safe Shutdown Events" appears in Attachment 9 to the EPU LAR, "Planned Modifications." Although this table was revised in Attachment 8 to Supplement 1 to the EPU LAR, the Service Water (SW) temperature was not changed and correctly indicated 86°F.
1 O.b. Except for the reactor thermal power, the parameters listed in Table 2.5-1 of Attachment 6 (PUSAR) of the EPU LAR (Reference 5-1) were not changed for the TPO uprate analysis. The error was in the use of the incorrect service water temperature (92°F) in the EPU Net Positive Suction Head (NPSH) evaluations for the cited Appendix R cases. The correct service water temperature (86°F), as indicated in PUSAR Table 2.5-1, has now been used in these same evaluations for the MUR LAR, as indicated in TSAR Section 4.2.5. The TPO uprate evaluations therefore are based on the Appendix R inputs provided in PUSAR Table 2.5-1 other than the EPU core thermal power value.
SRXB-RAl-11 With respect to TSAR Section 9.3.1.2, please provide justification for the statement in the first sentence of the third paragraph (last paragraph on page 9-3) regarding the loss-of-offsite power (LOOP) event.
MUR LAR Supplement 3 Responses to Requests for Information
RESPONSE
Page 7of8 The Loss of Offsite Power (LOOP) event does not result in a reduction in the RHR suppression pool cooling capability relative to the Main Steam Isolation Valve Closure (MSIVC) and Pressure Regulator Failure Open (PRFO) events because the same RHR suppression pool cooling capability that is credited in the analysis of the MSIVC and PRFO events is provided during the LOOP event by equipment powered by the Standby AC Power Supply. Thus the analyses for the MSIVC, PRFO, and LOOP events all credit the same RHR suppression pool cooling capability.
SRXB-RAl-12 With respect to TSAR Section 9.3.1, explain the basis for the acceptance criteria of the limiting temperature 180°F of the containment structure and justify that it is conservative. In addition, provide the Heat Capacity Temperature Limit (HCTL) at the normal suppression pool level and the Safety-Relief Valve (SRV) opening pressure.
RESPONSE
Maintaining suppression pool temperature below 180°F is applied as an acceptance criterion during an A TWS event in order to avoid damage to the HPCI pump and ensure availability of the HPCI system. As indicated in TSAR Table 9-2, the peak suppression pool temperature during an ATWS event at TPO/MELLLA+ conditions is analyzed to be less than 180°F. The limiting temperature of 180°F is conservatively lower than the 281°F design temperature of the containment structure (drywell and suppression chamber) given in Table 5.2.1 of the PBAPS UFSAR.
For CL TP, the HCTL is 178.5°F at the minimum suppression pool water level of 14.5 ft that is allowed by Technical Specifications, and the lowest SRV opening pressure of 1135 psig. For TPO, under the same suppression pool water level and SRV opening pressure conditions, the HCTL will be 178.4°F for PBAPS Unit 2. The TPO HCTL for PBAPS Unit 3 will be determined as part of the core reload process and is expected to be similar to the Unit 2 limit given that both cores are comprised of all GNF2 fuel.
SRXB-RAl-13 With respect to TSAR Section 9.3.1.6; with the boron injection rate remaining unchanged from CLTP to TPO, it is not clear from the CLTR (ADAMS Accession No. ML032170332) that the suppression pool temperature following an A TWS event meets all CL TR dispositions. Please explain and provide the section number of the CL TR which mentions this disposition.
RESPONSE
For the proposed PBAPS TPO, the CL TR (NEDC-33004P-A, Revision 4) requires a plant-specific analysis to ensure sufficient standby liquid control system capability. The intent of TSAR Section 9.3.1.6 is to state that this was accomplished. A plant-specific analysis was
MUR LAR Supplement 3 Responses to Requests for Information Page 8of8 performed, and Section 9.3.1.2 and Table 9-2 show that all acceptance criteria are met with the same boron injection rate as was used for CLTP/ MELLLA+. Since a plant-specific analysis was performed with acceptable results, further dispositioning of the CL TR is not required.
SRXB-RAl-14 TSAR Section 10.1 states:
Vessel dome pressure and other portions of the RCPB [Reactor Coolant Pressure Boundary] remain at current operating pressure or lower.
Describe the portions of the RCPB under the TPO conditions that would be operating at a lower pressure than the operating pressure at the CL TP, and provide reason(s).
RESPONSE
As the main steam flow rate increases due to the TPO uprate, the frictional pressure drop increases while the reactor vessel dome pressure remains constant from CL TP to TPO.
Therefore, at TPO conditions, main steam downstream of the reactor will operate at slightly lower pressures than at CL TP.
SRXB-RAl-15 TSAR Section 10.1, second paragraph states:
At the TPO RTP, HELBs [High Energy Line Breaks] outside the drywell would result in an insignificant change in the sub-compartment pressure and temperature profiles. The affected building and cubicles that support safety-related functions are designed to withstand the resulting pressure and thermal loading following a HELB at the TPO RTP.
Please explain why there would be a change in the sub-compartment pressure and temperature profiles at the TPO uprate conditions. Also explain why the change is insignificant compared to the current profiles and the existing design margins.
RESPONSE
TSAR Section 10.1.2.1 discusses the change in feedwater mass and energy release due to TPO uprate in the main steam tunnel and states the increase is insignificant. However, the analyzed pressure and temperature profiles in the main steam tunnel due to a Main Steam Line Break (MSLB) bound those from a feedwater line break, and are unchanged for TPO. Sub-compartment pressure and temperature profiles for all analyzed HELBs remain unchanged with TPO.
Peach Bottom Atomic Power Station, Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 Replacement Page 8 for PBAPS MUR LAR Supplement 2 Attachment
MUR LAR Supplement 3 Responses to Requests for Information MUR LAR Supplement 2 Responses to Requests for Information 1500 1400 1300 1200 1100 Figure 2-Unit 3 Main Generator Capability Curve
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Max TPO Operating 1000
..._-+-_....___....._..__.....__,...__... Point at Lowest CJ ct00 z a
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300 2
200 100 0
0 500 1000 MEGAWATTS 1500 Historical Power Factor (1388 MWe I 0.945 PF I 481 MVAR / 1469 MVA)
- b. The projected gross generator output for both Units 2 and 3 at licensed TPO power level is approximately 1323 MWe (summer operation) and 1388 MWe (winter operation). At the lowest historical power factor (0.945), the generators would operate between 1400 MVA (summer operation) and 1469 MVA (winter operation). Typically, the generators' MVA loadings would be much less than this, with operation at power factors above 0.98. These MVA loadings are below the 1530 MVA generator rating.
EEOB-R.Al-2 In Table f>-1, "Plant Electrical Equipment Ratings," of the TSAR, the licensee provided the ampere ratings for the isolated phase bus duct (generator bus, main section, delta section, and auxiliary section). Section 6.1.1 of the TSAR states that the isolated phase bus duct is adequate for both rated voltage and low voltage current output.
This is a replacement page 8 for the Attachment to Exelon letter to NRC, "Measurement Uncertainty Recapture License Amendment Request - Supplement 2 Response to Request for Additional Information," dated July 13, 2017 (ADAMS Accession No. ML17195A285). The operating point that the text box pointed to was incorrect in Supplement 2. The above page corrects the operating point on the graph. The data in the text box was correct and is unchanged.