ML13204A035

Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specification 4/4.7.11, Ultimate Heat Sink

ML13204A035
Person / Time
Site:	Millstone
Issue date:	07/19/2013
From:	Scace S Dominion Nuclear Connecticut, Dominion
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
13-419
Download: ML13204A035 (186)

Text

{{#Wiki_filter:Dominion Nuclear Connecticut, Inc.h.mii Millstone Power Station JUL 19 2013 Rope Ferry Road, Waterford, CT 06385 U. S. Nuclear Regulatory Commission Serial No. 13-419 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATION 3/4.7.11, "ULTIMATE HEAT SINK" By letter dated May 3, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 2 (MPS2). The proposed amendment would modify Technical Specification (TS) 3/4.7.11, "Ultimate Heat Sink," to increase the current ultimate heat sink (UHS) water temperature limit from 75 0F to 80OF and change the TS Action to state, "With the ultimate heat sink water temperature greater than 80'F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours." In a letter dated June 26, 2013, the Nuclear Regulatory Commission (NRC) provided DNC an opportunity to supplement the LAR identified above. Supplemental information was provided to the NRC in a letter dated June 27, 2013. In a letter dated July 18, 2013, the NRC transmitted a request for additional information (RAI) related to the LAR. provides DNC's response to the NRC's RAIs. If you have any questions or require additional information, please contact Wanda Craft at (804) 273-4687. Sincerely, S. . Scace" Site Vice President - Millstone Power Station STATE OF CONNECTICUT COUNTY OF NEW LONDON The foregoing document was acknowledged before me, in and for the County and State aforesaid, today by S. E. Scace, who is Site Vice President of Millstone Power Station. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief. Acknowledged before me this j1 day of " 013. My Commission Expires: ( al 21 NotaryP c WM. E. BROWN NOTARYPUBLIC My COMMISSION EXPIRES MAR. 31.,2018

Serial No. 13-419 Docket No. 50-336 Page 2 of 3 Commitments made in this letter: None Attachments:

1.

Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink"

2.

Excerpts from Calculation S-M2CNT-04326S2

3.

Excerpts from Calculation S-M2CNT-04325S2

4.

Excerpts from Calculation 12-347

5.

Excerpts from Calculation 97-ENG-01962-M2, Rev. 1

6.

Excerpts from Calculation 97-169

7.

Excerpts from Calculation 97-ENG-01862-M2

8.

Excerpts from Calculation 12-001

9.

Excerpts from Calculation 12-328

10.

Updated Final Safety Analysis Report, Section 1.4, "Principal Architectural and Engineering Criteria for Design" (Revision 31.1)

11.

Updated Final Safety Analysis Report, Section 9.4, "Reactor Building Closed Cooling Water System" (Revision 31.1)

12.

Updated Final Safety Analysis Report, Section 14.8, "Millstone Unit 2 FSAR Events Not Contained in the Standard Review Plan" (Revision 31.1) cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 James S. Kim Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No. 13-419 Docket No. 50-336 Page 3 of 3 Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No 13-419 Docket No. 50-336 Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink" Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 1 of 8 Response to Request for Additional Information Regarding License Amendment Request for Changes to Technical Specifications 3/4.7.11, "Ultimate Heat Sink" By letter dated May 3, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 2 (MPS2). The proposed amendment would modify Technical Specification (TS) 3/4.7.11, "Ultimate Heat Sink," to increase the current ultimate heat sink (UHS) water temperature limit from 750F to 80°F and change the TS Action to state, "With the ultimate heat sink water temperature greater than 800F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours." In a letter dated June 26, 2013, the Nuclear Regulatory Commission (NRC) provided DNC an opportunity to supplement the LAR identified above. Supplemental information was provided to the NRC in a letter dated June 27, 2013. In a letter dated July 18, 2013, the NRC transmitted a request for additional information (RAI) related to the LAR. This attachment provides DNC's response to the NRC's RAI.

RAI-1

Background The vendor heat exchanger data sheet for the reactor building closed cooling water (RBCCW) heat exchanger provided in the licensee's letter dated June 27, 2013 states that the design service water (SW) inlet temperature is 75°F. The Updated Final Safety Analysis Report (UFSAR) Section 9.7.2 states that the maximum service water temperature is 75°F. The UFSAR Section 9.4.2.1 states that the maximum RBCCW heat load during the injection mode following Loss of Coolant Accident (LOCA) is 213.4 x 106 Btu (British Thermal Unit)/hr and 146.0 x 106 Btu/hr during recirculation mode. The table provided as Enclosure 2, page 1 of 17 of MPS2 June 27, 2013 letter lists the design heat load of the RBCCW heat exchangers as 2.044 x 108 BTU/hr [204.4 x 1061. Issue The licensee has proposed raising the SW temperature limit from 75°F to 80°F. The increase in SW temperature will cause a corresponding increase in the RBCCW heat exchanger outlet temperature which supplies the RBCCW cooling loads. The maximum heat load stated in the UFSAR is different than the heat load presented in, page 1 of 17 of MPS2 June 27, 2013 letter. Request Discuss the quantitative effects and acceptability of the increase in RBCCW cooling water to all the safety related loads cooled by RBCCW during normal and accident conditions assuming the maximum RBCCW heat loads described above. In the discussion of the acceptability of the increased RBCCW cooling water, please provide the following:

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 2 of 8 a) all the design inputs, assumptions and acceptance criteria of the design calculations that proved that the increase in RBCCW cooling water is acceptable for each safety related load that is cooled by RBCCW cooling water. b) Explain how the acceptance criteria are met to provide satisfactory results. c) Described [sic] how you verified the RBCCW cooling water flow and SW flow to the RBCCW heat exchangers in the design calculations referred in a) above. d) explain why there is a difference in the design heat loads listed in Enclosure 2, page 1 of 17 of MPS2 June 27, 2013 letter and the UFSAR. DNC Response to RAI-la The design inputs, assumptions, and acceptance criteria for the design calculations that establish the basis that the increase in RBCCW cooling water is acceptable for each safety related load cooled by RBCCW are included in the attached calculation excerpts. The following provides a brief description of the calculations included: " Calculation S-M2CNT-04326S2 (Attachment 2): This calculation determined the LOCA summer minimum required SW flowrate to RBCCW using 7570 gpm (3,874,840 Ibm/hr) and 80°F for SW flow to the RBCCW heat exchangers. The 7570 gpm flow rate is listed as an initial input in Calculation S-M2CNT-04325S2 (Attachment 3) that was used to develop the GOTHIC containment analysis model for MPS2. Calculation 12-347 (Attachment 4): This calculation evaluated the ability of the RBCCW heat exchangers to remove the required heat load and established the minimum required SW flow at 80'F. Section 5.1 benchmarked the required flowrate calculated by the LOCA analysis in Calculation S-M2CNT-04326S2. Section 5.2 includes a calculation for a new required winter LOCA flow rate based on the new LOCA heat loads. Section 5.3 includes a calculation for seismic and shutdown cooling (SDC) (one SW pump to two RBCCW heat exchangers) required flow rates. " Calculation 97-ENG-01962-M2, Revision 1 (Attachment 5): This calculation performed the reactor coolant system (RCS) cooldown analysis and assumed an RBCCW flowrate of 6400 gpm. Calculation 97-169 (Attachment 6): This calculation determined the flow distribution in the RBCCW system and compares it to the required flow. It also determined the outlet temperatures for each of the RBCCW components. This calculation assumed a normal operating SW flowrate to RBCCW of 7500/7650 gpm for Facility (train) 1 and 2, respectively. As discussed in the response to RAI-4, this did not result in RBCCW temperature remaining at or below 85 0F for normal operation. Calculation 97-169 uses heat load inputs from Calculation 97-ENG-01862-M2 (Attachment 7). It should be noted that some heat loads were taken directly from the LOCA analysis contained in Calculation S-M2CNT-04326S2. Calculation 12-001 (Attachment 8): This calculation determined the predicted SW flows to the SW-cooled heat exchangers, including RBCCW. Section 11.1 provides a description of

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 3 of 8 the benchmarking performed during MPS2 Refueling Outage 21 (2R21) in October 2012. Section 12 summarizes the required flows. DNC Response to RAI-1 b The calculations discussed in response to RAI-1 above show how the acceptance criteria are met. The table below summarizes the results of the attached calculation excerpts. Minimum Required Flow Predicted Flow RBCCW Heat Exchanger Flow rates Flowrate with Ref 3 Condition Description [gpm] Ref Case uncertainty Page No. [gpm] X18A LOCA Summer 7570 1 2a2 7655.18 63 X18B LOCA Summer 7570 1 2a3 7720.67 63 X18C LOCA Summer 7570 1 2c2 7734.89 64 X18A LOCA Winter < 60°F 5423 2 2e2 6397.75 65 X18C LOCA Winter < 60°F 5423 2 2g3 6595.94 66 X18A Normal Operations 7500 4 4b 7529.16 67 X18C Normal Operations 7650 4 4b 7688.99 67 X18A Seismic Summer w/o LNP 5150 2 6b 6661.87 71 X18C Seismic Summer w/o LNP 5150 2 6b 6786.31 71 X18A Cold Shutdown w/ Lo-Tide 6400 5 9 6955.70 74 X18C Cold Shutdown w/ Lo-Tide 6400 5 9 7090.19 74 X18A Single SW Pump to Cool Two 4450 2 10b 5349.20 75 SDC trains X18C Single SW pump to Cool Two 4450 2 10a 5362.99 75 SDC trains Table References

1. Calculation S-M2CNT-04326S2, Rev 0, LOCA GOTHIC Containment Analysis for Millstone Unit 2

2. Calculation 12-347, Rev 0, Thermal Performance of the Unit 2 RBCCW Heat Exchangers for UHS Temperature Increase

3. Calculation 12-001, Rev 0, MP2 SW Model and Design Basis Analysis

4. Calculation 97-169, Rev 4, MP2 RBCCW-Design Basis Flow Distribution

5. Calculation 97-ENG-01 962-M2, Rev 1, RCS Cool Down Time with Reduced RBCCW Flow during Normal Shutdown DNC Response to RAI-lc One train of RBCCW is analyzed and benchmarked against the predicted RBCCW flows every refueling outage using station procedures. The SW system was analyzed and benchmarked during 2R20 (B train only) and 2R21 (A and B trains). The results of the benchmarking efforts are provided in Calculation 12-001, Section 11 (Attachment 8).

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 4 of 8 DNC Response to RAI-ld The design heat load of 204.4 x 106 Btu/hr reported in Enclosure 2 of DNC letter dated June 27, 2013, is based on the LOCA containment analysis reflected in FSAR Revision 31.1, dated July 10, 2013. The values stated in NRC's RAI-ld are consistent with FSAR Revision 30.2, dated March 31, 2013. Subsequent to FSAR Revision 30.2, MPS2 implemented revised LOCA and Main Steam Line Break (MSLB) containment response analyses under the provisions of 10 CFR 50.59 to correct nuclear steam supply system vendor errors in the mass and energy release analyses and to bring the analyses in-house using the NRC-approved analysis methodology described in Dominion topical report DOM-NAF-3-0.0-P-A. The containment analyses assumed 80'F UHS temperature as part of the update. The updated analyses results were incorporated into Revision 31.1 of the MPS2 FSAR and therefore represent the current plant licensing basis. FSAR Sections 1.4, 9.4 and 14.8 were modified as a result of the updated analyses. The affected sections from Revision 31.1 are provided in Attachments 10, 11 and 12 for MPS2 FSAR Sections 1.4, 9.4 and 14.8, respectively. The design heat load of 204.4x1 06 Btu/hr during the injection mode following a LOCA is shown in Attachment 11 (MPS2 FSAR Page 9.4-6). Thus, the information in our letter dated June 27, 2013 is consistent with the current plant licensing basis as reflected in MPS2 FSAR Revision 31.1.

RAI-2

Background Technical Specification 3.9.3.1, REFUELING OPERATIONS DECAY TIME, states that the reactor shall be subcritical for a minimum of 100 hours prior to movement of irradiated fuel in the reactor pressure vessel. Issue Minimum time for sub criticality is dependent upon the capability to remove heat from the spent fuel pool which is dependent upon RBCCW cooling capability which is dependent on ultimate heat sink (UHS) temperature. Request Discuss the impact of raising the maximum UHS temperature to 80°F upon TS 3.9.3.1 and whether the proposed UHS temperature change is acceptable. DNC Response to RAI-2 The vendor proprietary analysis performed to support the 100 hour decay time minimum limit in TS 3.9.3.1 did not use UHS temperature as an input but rather assumed an initial RBCCW temperature. In this analysis, core offload at 100 hours is predicated on maintaining 75 0F RBCCW temperature prior to the start of core offload, while core offload after 150 hours of decay time is permitted with 85°F RBCCW temperature. This limit ensures that the 150°F

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 5 of 8 maximum normal operating temperature for the spent fuel pool (SFP) is not exceeded. The MPS2 refueling procedure provides specific steps to ensure that the RBCCW temperature is less than or equal to 75 0F when moving fuel between 100 and 150 hours after the reactor is subcritical and RBCCW temperature is less than or equal to 850F when moving fuel after 150 hours. This refueling procedure does not allow fuel movement if the RBCCW temperature exceeds these limits. In general, this restriction does not present a challenge for spring refueling outages because the UHS temperature is lower at this time of year. However, for fall refueling outages, a longer period of time may be needed prior to core offload since the UHS temperature is typically higher than in the spring.

RAI-3

Background The emergency diesel generator (EDG) vendor data sheets state that SW flow to the diesel generator coolers is 700 gallons per minute (gpm). The SW inlet temperatures to the in-series air cooler heat exchanger, lube oil cooler and jacket water cooler are 750F, 830F and 91.60F, respectively. Heat removed is 2,769,000 BTU/hr, 2,890,000 BTU/hr and 3,522,000 BTU/hr respectively. Issue The licensee has stated in the application that the SW supports EDG operation with 5% tube plugging, 80°F SW inlet temperature and 672 GPM SW flow. Request Please provide all assumptions and design inputs and acceptance criteria of the calculation that proves the Issue discussed above is accurate. Explain how the acceptance criteria are met. Explain how you verify that the diesels are each receiving at the minimum SW required by the design calculation. DNC Response to RAI-3 Calculation 12-328 (Attachment 9) determined the required SW flow to the EDGs is 637 gpm. The assumptions, acceptance criteria and design inputs are contained in Attachment 9. The predicted SW flow of 672 gpm to the EDGs was determined by Calculation 12-001 (Attachment 8). See the table below for a summary of this information. SW flow tests performed during 2R20 and 2R21 confirmed that SW flow to the EDGs were meeting the minimum predicted flowrates. Calculation 12-001, Section 11.1, contains a discussion of the flow test data which confirms the measured EDG flowrate is within the 10% model uncertainty'. 1 Predicted flowrate is reduced by 10% before comparing it to the acceptance criteria.

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 6 of 8 Minimum Required Flow Predicted Flow Component rates Flowrate with Reference 3 Condition Description [gpm] Reference Case uncertainty Page [gpm] EDG A Hx LOCA Summer 637 1 2a2a 672.4 64 EDG B Hx LOCA Summer 637 1 2c2a 677.7 65 EDG A Hx LOCA Winter < 601F 337 2 2e2 '481.1 65 EDG B Hx LOCA Winter < 601F 337 2 2g3 495.1 66 Loss of Normal Power 5 Seismic 679.8 68 EDG A Hx Ls of 276 1 9 Lo-Tide 492.2 74 (@1800KW) 10a SDC 340.8 75 Loss of Normal Power 5 Seismic 679.0 68 EDG B Hx Ls of 276 1 9 Lo-Tide 511.1 74 (@ 1800KW) 10aSDC 357.7 75 Table References

1. Calculation 12-328, Rev 0, MP2 Equivalent Thermal Performance of the U2 EDG Hx for UHS Temp Inc

2. Calculation 92-125, Rev 00-01, Calculation of Min Required SWS cooling for EDG HX based on SWS inlet temperature of 60OF-2750kW

3. Calculation 12-001, Rev 0, MP2 MP2 SW Model and Design Basis Analysis

RAI-4

Background The application states that an RBCCW supply temperature of 850F will be included in the FSAR as a design requirement of the RBCCW system when in Modes 1, 2 and 3. Operating procedures will be modified to maximize SW flow to RBCCW and RBCCW loads will be minimized as appropriate. Request Please provide the design inputs, assumptions, and acceptance criteria of the design calculation that shows that an RBCCW temperature of 850F can be achieved with SW temperature at 800F. Describe what modified operating procedures will do to maximize SW flow to RBCCW and minimize RBCCW loads as appropriate. DNC Response to RAI-4 The limit of 850F is invoked since it is used as an input to the water hammer analyses performed in response to Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions." This is an initial condition only as RBCCW supply temperature increases significantly post accident. Containment analyses for LOCA and MSLB do not require this initial condition. Note that a limit of 85 0F RBCCW supply temperature is not new. This has existed for some time and is only being emphasized due to margin reduction.

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 7 of 8 During previous periods of operation with high UHS temperature conditions, it was observed that the difference between UHS temperature and RBCCW temperature was less than 50F. Since RBCCW supply temperature is observed under normal operating conditions and can be continuously monitored, maintaining the RBCCW supply temperature within limits is readily achievable. An analysis that demonstrates that 85°F RBCCW supply temperature can be maintained with 80'F UHS temperature, does not currently exist. Calculation 97-169 (Attachment 6) shows that the UHS temperature cannot exceed 79 0F to meet this criterion assuming design basis heat loads 1 during normal operations. Further, empirical evidence demonstrates that this requirement can be met during normal operation. Operating procedures will be revised to include actions that can be taken to maximize cooling of the RBCCW and to mitigate this limited concern. These procedures will be made effective upon implementation of this LAR. Actions to be performed are as follows: " There are four containment air recirculation (CAR) fans, two on each train of RBCCW. Normal operation has three fans running and one in standby. There are two SFP coolers, one on each train of RBCCW. Loads can be shifted as necessary to have the in-service SFP cooler on the RBCCW train that has only one CAR cooler in service. This alignment balances the heat load and reduces RBCCW supply temperature. During a LOCA or MSLB, the fourth CAR fan will start and RBCCW flow to its cooler will be initiated. There are two valves in parallel that control the flow of SW through the RBCCW heat exchangers; the winter valve (which is smaller and normally in service when the UHS is colder and flow requirements are less) and the summer valve (which is in service when the UHS temperature is elevated and opens automatically on a safety injection signal). During high UHS temperatures, the winter valve can be opened 2 to provide several hundred gpm of additional flow. Modifications to the RBCCW temperature control valves will be made to allow increased SW flow to the RBCCW heat exchangers to provide additional margin. Currently, SFP decay heat loads are lower due to the elapsed time since the fall 2012 refueling outage. 1 This analysis assumes each SFP heat exchanger has the full SFP heat load. If both pumps are in operation, the SFP heat load will be split between them (reducing heat load by over 3 MBTU/hr) or if one is in operation, it carries the full heat load while the opposite train has none. Due to the difference between the analytical results and the empirical evidence, this overly conservative assumption will be removed in analyses currently underway. Preliminary results indicate that 850F RBCCW supply temperature can be maintained with 80°F service water temperature. The SFP heat load for this calculation assumes fuel has been discharged from a recently completed spring outage. This occurs once every three years. Additionally, there is conservatism in the SFP heat loads because design heat loads are used instead of best estimate heat loads (reducing heat load by at least 1 MBTU/hr). 2 The SW flow analysis included this assumption to ensure other safety related components would still obtain their minimum required flows for normal and accident cases.

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink, Page 8 of 8

RAI-5

Background The licensee has stated that allowable SW side differential pressure (DP) limit for the 'A' RBCCW heat exchanger is being lowered from lO-psid to 8-psid. Issue Generic Letter 89-13 Supplement I states "With regard to the testing of containment spray heat exchangers, as of all safety-related heat exchangers, a pressure drop test alone is not sufficient to satisfy the indicated heat transfer capability concerns. [Request Please explain your use of pressure drop to indicate satisfactory heat exchange performance of the RBCCW heat exchangers. What other parameters in conjunction with pressure drop testing do you use to verify adequate heat transfer capability? (included in a July 16, 2013 NRC email).] DNC Response to RAI-5 Heat exchanger performance can be affected by macrofouling (e.g., seaweed, shells, etc. that can impede flow) and microfouling (e.g., silt, biological growth, etc. that can build up on the tube walls and impede heat transfer). Pressure drop surveillances are used to monitor for macrofouling. Due to changes in the environment, macrofouling does not always proceed at a predictable rate; therefore this surveillance verifies that flow is not impeded by this mechanism. The surveillance frequency varies from heat exchanger to heat exchanger depending on previously observed rates of fouling, inspections done during cleaning activities, and on time of year. For the most sensitive heat exchangers (i.e., RBCCW), this surveillance will be conducted weekly on the in-service heat exchangers during the warmer months (July through September). Microfouling proceeds at a more predictable rate, although significant interruptions in hypochlorite treatment may affect this mechanism. Periodic cleaning (every three months in the case of the RBCCW heat exchangers) addresses microfouling. Microfouling is modeled in the heat exchanger design performance analysis. For RBCCW, a fouling resistance of 0.0005 hr-ft2-°F/BTU has been validated assuming a three month cleaning frequency. If hypochlorite treatment is affected, procedural controls require the system engineer to be notified so that he/she can assess the impact and recommend an increase in the cleaning frequency, if necessary, to ensure acceptable heat exchanger performance.

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation S-M2CNT-04325S2 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 1 ENGINEERING WORK SHEET I ilc Nummber: SM2CNT-U4'325S2 R.,. 0 jAdd. IN/A S ct; i Jhoi21 Titke; GOTHIC Containment Analysis Models for Millstone Unit 2 4 1 - 1637( -= 7.03 fit As noted below, this value Is ulUmately increased to 10.8 tt2 to obtain good agreement with the heat exchanger performance data. The hydraulic diameter is taken as the tube outer diameter of 0.75 in. The shell side heat transfer area is calculated similarly to the tube side heat transfer area.. A,,,= A,,,,, :x 2 7,470) ft x 2 0.75 )I,99 ID +OD) .65+ 0,75)191ft The tube (primary) and shell (secondary) side fouling factors are both 0.0005 [8]. The operating pressure is set to 11S psla to match the RBCCW pressure from the SOC HX.

model, The Builtin HTC is applied to both the tube and shell sides of the RBCCW HX.

Section 6.4.1.2 of Reference [51 provides Information about the GOTHIC Buliltin HTC, A multiplier of 0.946 is applied via a forcing function On the tube side HTC to account.for the fact that the tube side fluid Is. salt water as described below. There are no fins entered for either side of the HX. The primary (tube) side flow is given as 3,874,840.Ibm/hr,. While the secondary (shell) flow Is 2,553,887 Ibm/hr [8], The tube flow is generated by a boundary condition In the GOTHIC. model. The. correspondIng flow rate of 1,076.34 Ibm/sec Is. altered to.1,030.01 Ibm/sec as shown below to account.for the fact that the tube side fluid Is salt water. The shell flow value is converted to 709.413 Ibm/sec. Effect of Salt Water: The tube side fluid In the RBCCW HX Is the service water (SW), which is salt water. The salt water affects several of the water properties. The fluid properties In GOTHIC are based on fresh water, which are Internal to the code and therefore unable to be changed directly. Therefore, compensating measures will be taken to account for the effect of the salt water. Appendix D of CCN1 of Reference (201 provides properties for normal and salt water as a function of temperature. Item I.D of Reference [8] shows that the service water experiences a temperature range of 75 OF to about 120 OF; Properties for the salty service water and fresh water from Appendix D of CCN1 Reference (20] are presented In the table below, Note that the Prandtl number is a calculated value, and is calculated from the: other properties listed in the table by the following equation:

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation S-M2CNT-04326S2 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 13 ENOINEERINO WORK SHEIIET CaicNumber: S-M2CNIT-0412&92 Rev. 0 Add. N/A Sheet: 9 of223 Tide: LOCA GOTHIC containment Analysis for Millstone Unit 2 Notes: 1.2.1 RBCCW Flows and Heat Loads - established consistent with Table I of[5] and do not need to be changed from the base models, section 6.6 of [2]. The tables below are obtained from section 6,6 of [2], showing the flows and heat loads relative to SRAS and SFPC actuation, and repeated here for convenience.. Changes to the flows below specific to the RBCCW temperature calculation are provided In Section 5.2. Flows: RBCCW Flow to CARsO 2 ) Time Interval Flow Flow (sec) pn) (ft 3/sec) <SRAS 4000 8.912 >SRAS <SFPC 3200 7,1296 >SFPC 280 6.2384 (2) - Flows are to two coolers, hence twice the values In Table 1 of (5) RBCCW Fixed Loads Flow-Header A Time Interval Flow Flow 1 (sec) (qpm) (ft 3/sec) I <SRAS 1200 2.6736 l >SRAS <SFPC 200 0.4456 >SFPC 200 0.4456 RBCCW Fixed Loads Flow-Header B Time Interval Flow Flow (sec) (*pr.) ft. sac <SRAS 400 0.8912 >SRAS <SFPC 100 0.2228 >SFPC 100 0.2228 RBCCW SDC HX Flow (1) Time Interval Flow Flow (sec) L CL (Q - ) (ft?/sec) <SRAS .L 0 0 >SRAS <SFPC 2000 4.456 >SFPC 1800 4,0104 (1) - Header A only for single CS pump RBCCW SFP HX Flow Time Interval Flow Flow (sec) l2 pm) yft3/sech <SFPC 0 0

Serial No 13-419 Docket No. 50-336, Page 2 of 13 ENGINEERING WORK SHPRET CIu Number; S-2CNoT-04320S2 IRv. 0 Add. N/A Sheet; Ioo73 Tictle LOCA GOTHIC Containment Analysis for Millstone Unit 2 Time Interval Flow Flow I (gcrn) Cft/sec >SFPC 1100

• 2.4508 Heat Loads:

RBCCW A Fixed Heat Load Time Interval 1 Heat Rate Heat Rate _.(sc) (Btu/hr) (Btu/sec <SRAS 34.8e6"* 9666.7 >SRAS <SFPC 2,Oe6 555.6 >SFPC 2,0e6 555.6

• Assumes 3 Control Element Drive Mechanism (CEDM) coolers, Value Is 24.3E6 Btu/hr with 2 CEDM coolers In operation (see p.12 of [121)

RBCCW B Fixed Heat Load Time Interval Heat Rate I-Heat Rate (secR (AtuSt11) (Btu3sec) <SRASe) 3 t0er.6 833.3 >SRAS <SFPC I 2.0e6 555.6. >SFPC [ 2.0e6 555,6 SFP Heat Load Time Heat Rate Heat Rate .sec) (Btu/hr) (.tuJse_ <SFPC 0 0 >SFPC 11.0e6 3055.6E5 11.1.b Single Failure - The failure of 1 EDG Is referred to as 'mln 51' configuration and credits operation of I LHSI, 1 HHSI and I CS pump. With 2 EDGs available ('max SI' configuration), 2 LHSI, 2 HHSI and 2 CS pumps are available except that single failures as assumed to result In itbr 2 CARs and 2 CS pumps available pr 4 CARS and 1 CS pump available. II,2.a Containment Spray - Only one CS pump Is conservatively credited post-SPAS to accommodate the possibility of failure of one RBCCW pump which would result In the loss of two CAR cooling units but would not affect CS pump operation until after SPAS when RBCCW is required to cool CS pump flow, at which time one CS pump would be

lost, I.13.d Pump Heat - Pump heat Is not explicitly Included In the GOTHIC models used for this analysis because significant margin exists to offset this effect. For example, from

[18] the brake horsepower (BHP) for a HPSI or LPSI pump Is 400 to 450 HP (318 Btu/sec for 450 HP based on 1 HP= 0.707 Btu/sec) while the BHP for a CS pump Is about 200 HP

Serial No 13-419 Docket No. 50-336, Page 3 of 13 ENGINEERING WORK SHEET oCal Number: S-MCNT-04 'R' U Add. N/ he° oi I.*'ill: LOCA GOTHIC Containment Analysis for Millstone Unit 2 (141 Btu/sec). For single train operation, this yields a heat input of 1100 HP (778 Btulsec) or 3, 1E6 Btu over 4000 secs, when SRAS will have occurred, which Is the time of limiting sump temperature, After that time, the LHSI pump is not running and the CS pump Is eventually turned off, Two train operation yields higher heat Input but SRAS occurs proportionately earlier. As discussed In notes h and j of Table 5.1-2 and shown In Attachment A, there is significant conservatism In raising the Initial vessel level In the RCS model to the bottom of the hot leg and assuming that this fluid (which Is removed from the available R.WST fluid) Is at saturated conditions, This conservatism Is about 20E6 to 30E6 Btus. Next, for all cases except the RBCCW cases, an RBCCW train A fixed heat load of 34,BE6 Btu/hr Is assumed Instead of an allowable value of 24.3E6 Btu/hr (see note 1,2.1 of Table 3-1). This yields an additional 10.5E6 Btu/hr of conservatism or 11,7E6 Btu over 4000 seconds for comparison to the example above. Also, an RWST temperature of 100 OF is assumed In the analysis; however, the plant surveillance procedure Includes a maximum administrative limit of 95 OF for this parameter, [19] Other margln exists In the assumption of 80OF SW temperature although this margin may not be available In the future if It becomes necessary to Justify this higher SW temperature. Similar arguments apply for other components such as the RBCCW pumps and CAR fan motors. For additional pump heat sensitivity discussion, see Section 6.2, Table 6.2-1c,

Serial No 13-419 Docket No. 50-336, Page 4 of 13 ENGINEERING WORtSHEIrT Caic Nimb,,; S-M.2NT-0432(hS2 I Rev. f) 1Add. N/A [Ihect!,_7of7 Lril-:~ LOCA GOTHIC Contaii'ment Analysls for Millstone Unit.2 5.2 RBCCW Temperature The RBCCW fluid temperature response Is required as Input tO separate calculations to ensure that RBCCW piping and component design limits are not exceeded, The RBCCW temperature response cases are based on the models used for the P/T analysis from Section 5.1, but with certain modifications as discussed below, Design Inputs and Assumptions As described In (12), It Is generally most limiting to assume the faliure-of 1 RBCCW train, which maximizes the heat load Into a single RBCCW train Instead of distributing the load between both trains. For similar reasons, the fouling associated with the CAR heat exchangers and SDC heat exchangers is minimized, The RBCCW train failure limits RBCCW cooling to two CAR fans and one CS pump. Asecond Cs pump Is available during the Injection phase (prior to SRAS) as long as A/C power Is available. Cases are run with both minimum and maximum CS flow rates since this can affect the distribution and timing of heat.idads. Table 5.2-1 summarizes the design inputs for the RBCCW temperature analysis. These Inputs represent potential changes for the RBCCW cases and supplement information provided InTable 3-1. Method of Analysis Table 5.2-2 provides details on modifying the GOTHIC models to reflect the design Input changes, As noted above, not all changes are applicable for every case. Table 5.2-3 provides an RBCCW case list with configuration assumptions. This list was determined by first reviewing the comparable case list from [121.. The list was modified by considering the RBCCW heat exchanger outlet temperature response for the pressure/temperature cases summarized In Table 6,1-1. The GOTHIC models are created from analogous named models used in the P/T analysis sections (5.1, 6.1), with the addition of RBCCW specific changes. The model naming remains the same except that'RB' Is appended, In addition, the letter'a' is used before 'RB' to designate cases that assume maximum CS flow. One case was added, hlGmx22aRB, created from P/T minilmum S. flow case hiGmn with changes to

Serial No 13-419 Docket No. 50-336, Page 5 of 13 ENGINEERING WORK SlIEI3T Csid Number: S-M2CNT-04326S2 IRev. 0 1Add. N/A [sheet: _2_8 or Ti:- LOCA GOTHIC Containment Analysis tor Mlilstone Unit 2 accommodate maximum SI flow In addition to the usual RBCCW changes. During testing It was noted that the CAR HX heat transfer appeared low relative to values listed in [14). This is discussed in more detail In Attachment C. Additional sensitivity cases were run based on limiting RBCCW cases (e.g. slSmx22aRB, h]Gmx22aRB, diGmx22RB) that Increase the heat transfer by reducing Heat Exchanger Type 2 Film Thickness Multiplier from 1.45 to 1.1 (per Attachment C). These cases are Identified by the additional letter 'Q' to the case name. It was also determined during testing that It Is slightly more conservative to Initialize the containment to maximum relative humidity (100%) and minimum initial pressure (14.27 psla [7]) as discussed In Section 6i1. The limiting cases Included these changes. Finally, these same three (modified)cases were run again, but first adding increased SW temperature (8u0 F), and then also adding Increased SFP heat load (16E6 Btu/hr). A final sensitivity case was run to represent added pump head loads. The RBCCW temperature results are summarized In Section 6.2, Table 5.2-1 RBCCW Design Inputs I. DESIGN INPUTS

1. Initial Conditions.

a. Containment Total Pressure (psla) 14.27 Minimum (7].

Determined by sensitivity cases, Both mIn/max cases

included,

c. Relative Humidity (%)

100 T Maximum. (7]. Determined by sensitivity, cases, Both min/max cases Included

2. System/Component

a. Service Water (SW) temperature (F) 75 J.D,12 of [10] and Table V-LA, pg,44 I

I of L12] 80 Sensitivity to evaluate Increased SW temperature

b. CS Flow (gpm per pump)

-pre-SRAS / post-SRAS 1900/ 1900 1,B.1,3,b [10], Values represent maximum CS flows assuming low containment back-pressure. Cases based on minimum flows will continue to use the Table 3-1 values (eg., 1300 / 1350 gprn per pump). Cases that assume two spray trains will use

Serial No 13-419 Docket No. 50-336, Page 6 of 13 ENGINEERING WORK SHI-ET Caic Number: S-lvl2CNT-0432682 I.Rcv. 0 1Add.N/ ec:2o2

Title:

LOCA GOTHIC Conlainment Analysis for Millstone Unit 2 twice the stated flow value pre-SRAS

c. RBCCW Flows / Heat Load See note. below for values to use when assuming mimum flow, Cases based on minimum flows will continue to use the Table 3-1, Item 1.2.1 values.

d. CAR delay foilowing SIAS (sec) is.

1.8.2.3 [5], A/C power available case only (see Table 3-1 for LOOP value) e, CS Pump maximum delay time to IB.1.4 [5], A/C power available case reach full speed (sec)

16.

only, add to pipe fill time. (see Table 3-1 for LOOP value) II. ASSUMPTIONS

1. System/Component

a. CAR Cooler heat exchangers

0.

Heat transfer Into RBCCW fluid Is - fouling factor maximized. (also consistent with assumption of [12]), b, SDC heat exchanger

0.

1.C.7,8 of [10]. -fouling factor Note, per I.DiB,9 of [10.] fouling remains 0,0005 foe RBCCW heat exchangers.

c. Single Failure i train Results In avaliability of; RBCCW 2 CAR fan coolers 1 CS spray pump (after SRAS)

Maximum heat load to one RBCCW train. d, CEDM (L.e, control rod) heat loads 24.3E6 With AC power available-prior to (Btu/hir) SRAS, IV.B of (10] 4E6 With loss of AC power - prior to: SRAS. !__VB of [10] 2E6 All cases, after SRAS. IV.B of [10]

e. Offsite Power Availabillty See note Cases are considered both with and without A/C power available.

See note 1II,.e below,

Serial No 13-419 Docket No. 50-336, Page 7 of 13 ENGINEERING WORK SIUtET Caic Number: S-M2C-N\\T-04326S2 Rev, 0 Add, NIA t SIec: ritiq: LOCA GOTHIC Containment Analysis for Millstone Unit 2 Notes: 1.2.c RBCCW Flows: Maximum flow assumptions are from Table 1B of [10]. Changes relative to the values presented In Section 3, Item 1.2.1 are highlighted In bold. All flows are assumed to be applicable to Header A only since one train (RBCCW B) Is assumed to be failed. Cases based on minimum flow assumptions will continue to. use the values provided: In Secton 3. RBCCW Flow to CARs - Maximum Flow Time (gpm) (se-c) (1J Flow (re/sec) <SRAS 4400 9.803 >SRAS <SFPC 4000 8.912 >SFPC 3800 8.466 (1) - Flows are to two coolers, hence twice the values In Table 1B of [10] RBCCW Fixed Loads Flow-MaximumM') Time Flow Flow (sec) (gpm)_ (ft*/sec) <SRAS 1400 3.119 I>SRAS <SFPC 200 0.4456 >SFPC 200 0,456 RBCCW SDCHX Flow - Maximum Time Flow Flow §Lsec) I }gpm (ft 1/sec) <SRAS 0 0 >SRAS <SFPC 2900 6.461 >SFPC 2600 5.793 RBCCW SFP HX Flow - Maximum Time Flow Flow (sec) (fpro) ft'/sec) <SFPC. 0 0 >SFPC 1200 2.6736 Heat Loads: RBCCW A Fixed Heat Load Time Heat Rate Heat Rate (sec) (Btu/hr -(Btu/sec) <SRAS 24.,3E61) - 6750 1

Serial No 13-419 Docket No. 50-336, Page 8 of 13 GNOINr1ERING WORK SH1EET Cuic Number: SN2CNIT-04320S2 IRev. 0 1 Add. NIA sbeeL; '1102 1'itle: LOCA GOTHIC Containment Analyis for Millstone Unit 2 Time Heat Rate Heat Rate (sec) (Btu hr) 1Btu/sec) 4.e6l") 11111 >SRAS <SFPC 2.0e6 555.6 >SFPC j 2.0e6 555.6 1 -With power cases assumes 2 CEDM coolers in operation. For 3 CEDM coolers, the value Is 34.8E6 Btu/hr (see Section 3). Cases without A/C power assume a heat load of 4.E6 BtU/hr. See section IVB of [l0], 11.2.e OffsIte Power assumption -In general, cases with or without A/C power will assume the following configuration: A/C power available: 2 SI trains ('Max' St), 1.2.b of Table 3-1 2 CS pumps prior to SRAS Fixed heat load = 24.3E6 (prior to SRAS), based on 2 CEDMs In service CAR starting delay= 15 sec CS pump start delay= 16 sec (plus pipe fill time of 33 secs per Table 3.1) Loss of A/C power and single E6G failure I SI trains (Min SI), 1,2,b of Table 3-1 I CS pump prior to SRAS Fixed heat load = 4.E6 (prior to SRAS), based on 0 CEDMs in service CARP starting delay= 26 sec CS pump start delay= 26 sec (plus pipe fill time of 33 secs per Table 3-1)

Serial No 13-419 Docket No. 50-336, Page 9 of 13 ENGINEERING WORK SH1ET Calc Numher: S-M2CNT-04326S2 Rev. 0 1 Add. N/A lSheet: 32af. 3 Tide: LOCA GOTHIC Containment Analysis for MilIslone Unit 2 Table 5.2-2 RBCCW Model Changes Parameter GOTHIC... Elem Parameter Notes ..Element No(s) Value SW BC 5, 75 F SW Inlet temperature to Temperature 9 RBHX. Heat Exchanger HX Type 2, CAR 0 h-ft2-F/Btu Value applies to both Fouling 3, SDC primary and secondary for Resistance both CARs and SDC HX. RBCCW Flow to CVAR 21 9,8032 ft3/s Initial Value. Corresponds CARs to 4400 gpm 2=4400 gpm Function Component 3= 4000 gpm values. 4= 3800 gpm RBCCW Flow to CVAR 22 3.1192 ft/s Initial Value., Corresponds Fixed Loads to 1400 gpm. 2= 1400 gpm Function Component value. .RBCCW Flow to CVAR 24 3= 2900 gpm Function Component SDC HX 4= 2600 gpm values. RBCCW A Fixed CVAR 25 6750 Btu/s Initial vaiue,= 24.3E6 btu/h Heat Load-2 (1111 Btu/s) Value In () Is for loss of A/C. CEDMs 24.3E6 Btu/hr FUnction Component 2. (4.E6 Btu/hr) Value In () Is for loss of A/C RBCCW Flow to Pump Type 5 2.6736 ft3/s Flow value, Corresponds to SFP HX 1200 gpm. CS spray flow BC 3 4.2332 ft3/s Corresponds to 1900 gpm. rate Two spray trains yield 8.4665. t3/s Film thickness Component-2 iii Sensitivity cases reduce multiplier Heat Ex Value from 1.45 to Increase (CARHX) Type f heat transfer

Serial No 13-419 Docket No. 50-336, Page 10 of 13 ENGINEEWINO WOR K.SHEIr CauuNumber: S-M2CNT-o4326S2 Rev. 0 1Add. N/A Shoo,.t:.aof

Title:

LOCA GOTHIC Conlainment Analysis for Millstone Unit 2 .No I Case Bre Sis* Table 5.2-3 RBCCW GOTHIC Case Summary ak A/C SW CEDM SI Flow No, No, JCSI RECCW a Pwr Temp Ht Load (1) CAR Spray Plow Plow ? (F) <SRAS Fens pumps (2) (3) (etu/hi) <SRAS diGainR3B DEG I4.... N 75 4E6 MIN 2 1 MIN MIN 2a dlGmx22RB DEG Y 75 24.3E6 MAX 2 2 MIN MAX 2b dlGmx22QRB 4 ) DEG Y 75 24.3E6 MAX 2 2 MIN MAX 2c dIGrnx22Q DEG Y.7S 24,3E6 MAX 2 2 MIN MAX SWOORB5)_ 3 diGmx22aR8 DEG Y 75 24,3E6 _MAX 2 2 MAX MAX 4a hlGmxm22aRB DEG Y 75 24.3E6 MAX 2 2 MAX MAX 4h h1Gnmx22aQRB(4j E Y 75 24,3E6 MAX 2 2 MAX MAX 4c hlGmx22Ao DEG Y 75 24.3E6 MAX 2 2 MAX. MAX SW8ORB( 5) 5 hlGmnRB DEG N 75 4.E6 MIN 2 1 MIN MIN 6 h[GrrnaR DEG N 75 .4.EG HIMN 2 1 MAX HIN 7 hl2mxaRB 2ft2 jY 75 24.3E6 MAX 2 2 MAX MAX

• • 8a

Smx22aRB DES Y

75 24.3E6 MAX 2 2 MAX MAX eb .s~nmx22 aRBQa4F DES Y 75 243E6 MAX 2 2 MAX MAX ac slSmx22aRBO DES Y 75-2-4'3E6 MAX 2 2 MAX MAX sWo( 5) (1) Max S1= two SI trains (2 HPSI/2 LPSI) In operation; Min SI= 1 SI train (2) Max CS = 1900/1900 gpm per pump pre/post-SRAS; Min CS= 1300/1350 gpm (3) Max/Mln RBCCW - flows to CARs and other components per Table 5.2-1 (Max) and Table 3-1 (Min) (4) Sensitivity cases 2bi2c, 4b,4c and Bb,8c assume conservatively high CAR HX heat transfer capability. They are also configured for 100% RH and minimum Initi.al pressure. (5) Sensitivity cases with 800F SW temperature. Not shown In this table are three additional cases, 2d, 4d, Bd that Increased SFP heat load from I E6 Btu/hr to 16E6 Btu/hr and one additional case Be that evaluates the effect of CS and SI pump heat.

Serial No 13-419 Docket No. 50-336, Page 11 of 13 IONG7NEERING WORK SHEET Calc Number: S-M2CNT-04326S2 IRe. 0 Add, N/AR eShet: 52 of 73

Title:

LOCA C(OTHIC Cuntainment Analysi for Mlltstnc Unit 2 6.2 RBCCW Temperature Results Maximum RBCCW system temperatures during specified Intervals (e.gý, pre-SRAS, post-SRAS, post-4 hours) are summarized for each case In Table 6.2-1a. A composite of maximum RBCCW heat exchanger outlet temperatures Is provided In Table 6,2-2 for SW temperature of 75 0 F and In Table 6.2-3 for SW temperature of 800 F. A plot is provided In Figure 6.2-1. A plot of RBCCW system temperature response is shown for a representative case sISmx22aQRB In Figure 6.2-2. As discussed In Section 5.2, the three limiting cases (2b, 4b, and 8b) assume a more conservative CAR HX heat removal capability, In addition, they are Initialized to maximum Initial relative humidity and minimum pressure in containment, See also note 4 to Table 6.2-la. These same cases are run at 800 F SW Inlet temperature (2c, 4c, and 8c) to establish limiting values for the higher SW temperature. Additional sensitivity cases (2d, 4d, 8d) with an increased SFP heat load of 16E6 Btu/hr (and 80°F SW) are summarized In Table 6.2-lb and the effect of pump heat Is provided by case 8e, Table 6,2-1c. Table 6.2-1a GOTHIC RBCCW Case Summary(l) Case HD(CW remperatuies (ii No Case BRAS (sTi) Trip 7 RBCCW Tormperstuteos (F) RB HX In (Vol 2) Re HX Oul (Vall3) eSRAS .SRA5 ,4 hta CAR IIX Out (Val 4) <SRAS >.SlAS >4hrs SFP HX > 4hrs (Vol 27) la dIGmnR8 3a57.2 199,3 135.9 119.3 117.8 218.1 157.3 157.3 136.4 2a dIGmx22RB 2291.8 208.7 140.1 119.8 118.0 219.4 158.0 1,52,6 137.5 .2b dIQmx22QRB 2281,5 220.2 145.1 122.2 120.0 233.0 164.5 156.4 138.7 2c dlGmx220 SWc 0RB 2282,0 221.7 148.1 125.5 123,8 234.0 686.8 159.7 142.4 3 dlGmx228R8 2038.5 207,9 139.8 121.9 119.9 218.5 158.8 151.7 138., Ih-G-rx22alR8 2008.3 203,9 130.0 l15.8 118.8 215.3 161.6 149.3 137.4 4b hIGmx22aQRB2 (Note 4) 2004.6 215.4 143.0 120.3 119.7 229.3 156.6 152.3 138.3 -4c NllGrmx22idO SWaOlB2 2004.9 216.9 146.0 123.7 123,5 230,3 159.2 155.7 142.1 6 hlGmnRB 3790.2 196:9 130.7 114,3 114.3 216.8 158.5 158.5 134.6 6 hiGmnaRS 3426.1 195.0 131.2 115.4 115.4 218,0 158.2 158.2 135.7 7 hl2mxaRB 2023.9 200.7 139.2 121.8 120.0 217.4 158.1 151.7 138.6 a elSmx22aR8 2047.4 206.1. 139.0 123.0 120.5 217.1 161.1 153.0 139.1 Bb slSmx22aRBQ 2042.1 { 216.5 143.5 124.9 121.3 229.5 166.6 2042.3 216.5 143.5 124.9 23051.3. swao 2042,3 2t BO 146.5 1 128.0 125.0 230.5 16B.7 156.0 139.9 159.3 143.7

Serial No 13-419 Docket No. 50-336, Page 12 of 13 POGINEE,,JT WORK.UEF" CeIc Number: S-M2CNT-O4326S2 Rev. 0 A I^dd. NIA [Shee,: L3 3 I I Tidte: ] OCA GOTHIC Containment Analysis for Millstonc I'nit2 I No Case RBCCW Tamp (F) 1 SW Outlet Temp (F) HX Heal Rate (MBlulh) SDC HX Out (RB HX) RO(SW) CAR SDC (Vol 26) (Vol 6 or ti1h) (ovar76) Noin ¶ (oearl7) Nole 2 (car78) SRAS > 4 hra J SRAS <SRAS >SRAS <SRA8 >BRAS c 4 hra la dlGmnRffl 149.5 147.8 124.1 107.7 177,3 118.1 163.2 36.9 52,4 2a dlGmx22RO 149.8 149.2 128.0 108.1 191.4 119.4 193.1 37.4 49.2 2b dIGmX22QRB 151.6 149.3 132.8 110.0 208.6 126.4 196.8 40.4 48.8 2c dl13mx22Q

5.

123 .SWO0RS 154.0 15213 J38,7 114,1 204.4 123.0 193.7 40.3 48,2 3 di3mx22aRB 158.1 155.4 127.7 109.8 190.3 135.9 193.1 34.0 70.3 4a hlKmx22aRB 156.7 153.9 128.0 107.1 184.2 115.9 186.0 32.4 62.9 4b hlGmx22aORB2 157.5 153.6 130.8 108.5 201.4 121,0 188.1 36.6 62.4 4c hiGnix22aQ SWh0RB 159.6 156.5 134.7 112,7 197.2 117.8 186.1 34.7 10.5 5 hiGmnRl 156.0 153.5 121.0 104.9 168.1 108.1 146.5 35.3 47.7 6 hlamnaRB 1654,(5) 160,6(5) 121.4 106.2 167.0 1127 151.5 33.4 57,8 7 hl2mxsaRB 164.2 156.1 127.2 109,7 188.4 1254 153.7 36.4 68.7 8a slSmxu22an 160A1 156.3 128.9 110.7 190.3 125.5 193.1 36,1 61.2 m8b lSrnX22aRDO 161.0 165.7 131.2 112.2 203.1 134.3 182.8 40,8 60,8 ac slSmx221kHSO 163.0 158,6 135.1 116.2 198.9 130,7 180.0 39,7 1 59.1 (Ii) Results obtained from Data files. The GOT files may have been used for same values. (2) CAR heat rates are shown for one heat exchanger and represent the maximum value at CAR Initiation, Values generally decrease significantly over-several seconds as system temperatures approach new values. (3) CVAR numbers used to determine heat rates are provided for convenience. In some cases the CVAR numbering Is slightly different although the sequence (RBHX, CAR, SDCHX) remains the same. (4) cases 4b and 4c are InItialized with maximum containment pressure to avoid spray termination at SRAS, The effect or assuming maximum containment pressure on temperatures prior to SRAS Is very small (-, -0.1-O.2 0F) and the cases are not limiting. (5) sensitivity case hlGmnaQSW80sfpRB was run based on hIGmnaRB but Including the configuration changes of cases 2d, 4d, and 8d from Table 652-1b. This Includes a) Increased heat transfer b) maximum relative humidity c) 80 0F SW temperature and d) Increased SFP heat load. To prevent Gothic from terminating sprays at SRAS, the Initial pressure was retained at 15.7 psla. The resulting SDC HX outlet temperature (Vol 26) Increased from 1165.4 0F to 167.0"F (>SRAS) and 160.5 0F to 162.6 0F (>4 hrs). Table 6.2-1b SFP Heat Load Case Summarym') SCase RIB HXOuI fCARHX RB Out I BFPH X No. (-RB Out I SDC HX-RB Out T2d rCd1MX220 SWBOR1P116 >4 hr (vol s1 125.0 I.4rs (Vol 4 1 > 4hrs(al o27) 1 4 lirs (Vol 26) ='"- "" 152.2 153.0 4d hIGmX22aQ 124.6 1564 1.51.8 1 157.1 d sISa22aflBQ 126.1 159,9 153.4 159.2 SSW83slp I (1) SFP heat load at 4hrs = 16E6 Btu/hr (increase from 1IE5 Btu/hr), 4c, and 8c respectively from Table 6.2-1a based on cases 2c,

Serial No 13-419 Docket No. 50-336, Page 13 of 13 ENGINEERING WORK SHEET Caic Number: S-M2CNT-04326S2 Rev. 0 1Add. N/A [Shcer: 5.*6of f3 Tide: LOCA (GOTHIC Contninment Analysis for Millstone Unit 2 Table 6.2-3 Maximum RBCCW Heat Exchanger Outlet Temperature (composite). (Service Water Te perature=80*f' Trbccw Trbcaw Time (F) Time Trbccw SPP (M) (seea (hr} Note l csec) (hrs) e d)jes (F) Note 2 I.E-05 130.0 .. 6000 1.7 128.3 5 113.2 10000 2.8 126.1 !.0 107.0 15000 4.2 125.9 127,1 20 105.4 20000 5.6 124.0 125.3 30 1 127.0 30000 8.3 120.6 122.1 50 143.9 35000 9.7 114.8 120.8 75 147.0 40000 11.1 113.5 114.2 100 147.8 50000 13.9 111;9 113.5 150 149.0 60000 16.7 111.1 112.8 200 149.1 70000 19,4 110.3 112.1 250 148,7 80000 22,2 1097 111.5. 300 148,1 90000 25.0 109.5 111.3 350 147.2 100000 27.8 1.2 109.6 111.3 400 146,1) 11000,0 30.6 1.3 109.5 111.3 500 143,9 130000 36.1 1.5 109.1 110.9 750 138.9, 150000 41.7 1.7 108.6 110*3 1000 134,3 175000 48,6 2.0 107,8 109.6 1250 130,0 200000 55.6 2.3 107,0 108.8 1500 126.1 225000 62.5 2.6 105.3 108.1 1750 122.6 250000 69,4 2.9 105.8 107.7 2000 119.5 375000 104.2 4.3 104.0 105.9 2250 126.0 500000 138.9 5.8 102.2 104.1 2500 127.5 750000 208.3 8.7 99.2 101.1 2750 128.2 1.00E+06 277.8 11.6 97.0 98.9 3000 128.6 1.40E+06 388.9 16.2 95.7 97.6 3500 129,0 1.806+06 500.0 20.8 94.6 96.6 4000 1.1 120.2.206+06 611.1 25.5 93.8 95.7 5000 1.4 128,7 2,59E+06 720.0 30.0 93.2 95.1 (1) - A margin of 1.0F Is added to the temperatures. (2) - Column labeled 'SFP' based on SFP heat load of 16E6 Btu/hr (increase from 11E6 Btu.hr)

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 12-347 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 6

0.

12-347 0 oP 9 ZACHRY 7.ACHRY

NUCLEAR, INC.

TITLE Thermat Performance of the Unit 2 RBCCW Heat Exchangers for UHS Temperature increase ZN Doc.ument Tyýp: QAPD 1.0 PURPOSE The purpose of (his calculation is to determine operating conditions for the Millstone Unit 2 Reactor-Building Closed Cooling Water (RBCCW) Heat Exchangers following a Loss of Coolant Accident (LOCA), during a seismic event with offsite power available, and when a single service water pump Is supplying two RBCCW trains during Modes 5-6. 2.0 APPROACH The analysis. of the RBCCW heat exchangers Will be performed using Zachry Nuclear Englneering's thermal performance heat exchanger modeling software, Proto-HX Shell & Tube Module Version 4.10. The Proto-HX software was developed and validated in accordance with Zachry's Nuclear Software Quality Assurance Program persreference 7.1. This program meets the requirements of 1fCFR50 Appendix B, 10CFR21, and ANSI NQAwl. It was developed In accordancewith the guidelines and standards contained in ANSI/IEEE Standard 730119134 and ANSI NQA-2b-1991. The use of the application will be in accordance with reference 7.2. Reference 7.3 developed and benchmarked a Proto-HX model for the RBCCW Heat Exchangers. This model, Rb-strw.phx, will be used for the analysis In:thls calculation. Several cases will be run to determine the required Service Water (SW) flow rates or temperatures for the RBCCW heat exchangers for the situations given in Section 1. For all cases, a heat load and RBCCW flow are specified. For the LOCA case, the SW flow rate at the appropriate temperature is iterated until the specified RBCCW temperature is met. The SW flow is then reiterated until the specified SW outlet temperature Is met. For the seismic case and the case of a single service water pump supplying two RBCCW trains, the available SW flow rate Is entered and outlet temperatures are calculated. The analyses will use a Service Water inlet temperature of 800F, With an addltlnal LOCA case performed with a Service Water inlet temperature of 60"F. 3.0 DESIGN INPUTS 3.1 The RBCCW heat exchanger Design Requirements used for all cases are listed below, along with their sources; PerTable 5 of Reference 7.4, the maximum RBCCW HX heat load occurs during peak RBCCW temperature analysis. Reference 7.5 updates several of the design requirements for an increased service water temperature of B801. Tube-side fouling resistance: 0.0005 hrf*t"F/BTU (Rof 7.3) Shell-side fouling resistance: 0.0005 hr'ftt¶IBTU (Ref 7.3) Maximum allowable # of plugged tubes: 164 (10%) (Rel 7.4) Total Number of Tubes: 1637 (Ref 7.4) Reference 7.5 does not list an RBCCW shell-side flow updated for 80"F service water. Examining section 5.2 of reference 7.6 confirms that the Reference 7.4 value Is unchanged. 3.2 The following design inputs are used lor the LOCA cases: Form: NO3O1FCS Revision: 00.00 Date; 10-28-2011 Page 1 of *1 F~orm: N0301F05 Revision: 00-00 Date: 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 2 of 6 ',LCM. 12-347 O 0 .F 9 ZACVRY NUCLeA. fNC. TT':. Thermal Performance of the Unit 2 RBCCW Hea: Exchangers for UHS Temperature In roeaso ZN? Vocurment Typo: QAPD Shell-side inlet temperature: 222"F (Ref 7.5) (goal when Iterating to~shell inlet tamp) Tube-side outlet temperature: 137"F (Ref 7.5) (goal when iterating to tube outlet Lamp) Heat transfer rate: 204,400,000 BTU/hr (Ref 7.6) Shell-side flow: 5800 gprn (Ref 7.4)4 This is converted a mass flow rate using the density of water at the shell Inlet temperature, interpolated from data in Reference 7.7. Additional Iteration is required when iterating to the tube-side outlet temperature since shell side flow rate affects the calculated shell-side Inlet temperature. 3.3 The following design inputs are used for the case of a seismic event with oftsite power avallable: Shell-side flow' 5800 gpm (Ref 7.4), This is converted to a mass flow rate as described in Sec 5.3. Heat transfer rate; 23,444,000 BTU/hr. The Train B heat load from Reference 7.8 was used since it Is higher than the Train A heat load in the same reference. 3.4 The following design Inputs are used for the case of a single service water pump supplying two R8CCW trains during Modes 5-6: Shell-side flow: 5800 gprn (Ref 7.4). This is converted to a mass flow rate as described in Sec 5.3. Heat transfer rate: 40,750,000 BTU/hr from Train A, Case IlIof Ref 7.8 CON 2 since. Train A is slightly higher than Train B 3.5 The RBCCW Proto-IAX model was obtained from Reference 7.3. To be sure the model being used matched the Reference 7.3 model, the model was run using the same parameters as Attachment L of Reference 7.3 to be sure the results were the same. The reports for this benchmarking are included In Attachment C. The Reference 7.3 model deals with plugged tubes by reducing the total number of tubes, where this model deals with plugged tubes by reducing the number of active tubes. Thus, the "Total Number of Tubes" and 'Total Effective Area Per Unit" fields In the Input Parameters Report do not match, but the results between the two calculations are the same. 4.0 ASSUMPTIONS 4.1 The Service Water inlet temperature (tube temp) is B0*F per Reference 7.10. For an additional LOCA case, the Service Water Inlet temperature is assumed to be 601F as explained in Section 5.2. 4.2 Using the maximum normal operations heat load ftr the seismic case is bounding because a trip following a seismic event without a safety injection actuation signal will have the same or less heat inputs than the normal operations case. This heat load Is Immediately following the trip and does not include shutdown cooling. The TBCCW breaks assumed in the seismic case would be isolated prior to shutdown cooling being placed in service, Form: N0301 FOS Revision: 00-DO Date: 10-28-2011 Page I.of I

Serial No 13-419 Docket No. 50-336, Page 3 of 6 CXJ:ýNV-. 12-347 0 1 PA8. G or 9 ZACEIRY ORcINAMA C ZACHRY NUCLEAR, INC. TMT E Thermal Pedormance of the Unit 2 RBCCW Heat Exchangers for UHS Temperature Increase ZM Document Typo. QAPD 4.3 LOCA cases will be analyzed at 0%, 5% (1555 active tubes), 7% (1522 active tubes), and 10% (1473 active tubes) tube plugging. The seismic and single service water pump cases will only be analyzed at 10% tube plugging since the heat loads (or these cases are Considerably below the LOCA case. 4.4 For the seismic case, the available SW flow is assumed to be 5150 gpm. This Is based on Table 12 of Reference 7.11. The table stales that the lowest flow rate to any of the RBCCW HXs for a seismic event is 6066 gpm. 85% or this value is used to leave margin. 4.5 For a single SW pump cooling two RBCCW HX trains, the available SW flow is assumed to be 4450 gpm. This is based on Table 15 of Reference 7.11. The table states that the lowest flow rate to any Of the RBCCW HXs with one SW pump supplying 2. RBCCW HXs is 5230 gpm. 85% of this value is used to leave margin; 5.0 ANALYSIS 5.1 REQUIRED SERVICE WATER FLOW AT 8,0 F lOR LOCA The overall fouling factor was calculated from the Design Input 3.1 fouling factors using the following formula from Reference 7.2: R, = Ro(d./di)- Rio Where: R, is the overall fouling resistance 8 N is the inside fouling resistance (0-0005 hr*fe*'/BTU) Rio is the outside fouling resistance (0.0005 hr'ft"F/BTU) do is the outside diameter (0.75") dfis the inside diameter (0.652') Using these values, the overall fouling resistance is 0,001075 hr*ft°"F/BTU. The use of a fouling, limit of 0.0005 continues to support the 3 month heat exchanger tube-side cleaning frequency as identified in Ret. 7.3. The user specified extrapolations shell flow, constant heat load and tube inlet temperature were entered for the LOCA case as specified In Sections 3 and 4. The user specified tube flow (service water), in tb/hr, was iterated until the Shell Temp In matched 2220 F. Reference 7.4 states that the design basis for the RBCCW heat exchanger is that 10% of the tubes are plugged. Reference 7.6 also uses 10% tube plugging in its analysis. This Iteration was performed for tube plugging levels of 0, 5%, 7% and 10% for the LOCA case with the service water at its maximum allowable temperature of 801F. These results are Included In Attachment B. Reference 7.5 states that the maximum service water outlet temperature Will be 137"F. The results of this analysis show that the maximum service water outlet temperatute is exceeded. Therefore, the cases are re-analyzed with the tube flow (SW) iterated until the service water outlet Is 137'F, These results are shown in Table 1. Note that the RBGCW Inlet temperature is below the maximum analyzed value of 222TF for these cases. Form: N0301F05 Revision: 00-00 Dale: l0-28-2b1 I Page 1 of I Form. N030IF05 Revision: 00-00 Mile: 10-28-2011 Page 1 ofI

Serial No 13-419 Docket No. 50-336, Page 4 of 6 CA! 1CN. -347 RV I 7o ZACHRY

NUCLEAR, UNC, ri E Thermal Performance of the Unit 2 HBCCW Heat Exchangers for UHS Temperature Increase ument Type: OAPD Table 1 - Service Water Flow Rates for LOCA with SW at 809I: and SW Outlet at 1379F i# Active Tubes Service Water Flow IR BCCW Inlet Temp 1,637 (0 plugged) 3,750,000 Ib/hr (7333 gpmL_

2.15.91F 1555 (5% plugged)._3P*750O000 Ib/hr (7333gpm) 218.6F 1!.522 (7%/ plugd) 3,750,000 Ib/hr (7333 gpm) - 21g.81F... 1473 (110% plugged) 3,750,000 Ib/hr i73332mL_ 221.6-F Proto-HX reports for these runs are included in Attachment B. 5.2 REOUIREO SERVICE WATER FLOW AT 60"F FOR LOCA An additional LOCA case with 10% of the tubes plugged is run with a Service Water temperature of 60"F to pýrovide SW flow rates when the UHS is below its maximum allowed value of 80IF. This analysis is analogous to the Section 5.1 analysis with the exception of the SW inlet temperature. The results of this analysis are included in Attachment S. They showthat the maximum service water outlet temperature of 1379F Is exceeded. Therefore, the cases are re-analyzed with the tube flow (SW) Iterated until the service water outlet is 1371F. These results are shown In Table 3. Table 2 - Service Water Flow Rates for LOCA with SW at 601F and SW Outlet at 1376: SW Inlet Temp Service Water Flow RBCCW Inlet Temp 60"F 2,780,0001b/hr 5423 m 218.9"F Proto-HX reports for these runs are Included in Attachment B. 5.3 RBCCW OUTLET TEMPERATURES WITH SW AT 801: FOR SEISMIC EVENT AND SINGLE SW Pump For the seismic event and a single SW pump supplying two. RBCCW trains, it was desired to have the outlet temperature of the RBCCW match the outlet temperature reported for the analogous case in Reference 7.12. However, using the available service water flows from Assumption 4A and 4.5, those temperatures could not be met. The outlet temperatures that occur with the available service water flow using the applicable information from section 4 with 10% tube plugging are given in Tables 3 and 4. VolumetrIc flow rates were converted to mass flow rates using the water density (from Ref. 7,7 for fresh water andRef 7,13 for.salt water) at, the corresponding heat exchanger inlet temperature as calculated by Proto-HX. Since the shell side mass flow rate affects the shell side inlet temperature, a guessed mass. flow rate was entered, and Proto-HX calculated the shell Inlet temperature; The shell mass flow was then iterated until the mass flow was equal to. 5800 gpm at the calculated inlet temperature. Form N0301F05 Revision: Oct-Oct l)ata: 10-28-2011 Page 1 of 1 Form ý N0301 F05 Revision: 00-00 Date: 10-28-2011 Pago I of 1

Serial No 13-419 Docket No. 50-336, Page 5 of 6 DAM. WX 12-347 0V 9 ZACHRY AMOR .IVI ZACHRY NUCLEAR, INC. T1T7E Thermal Performance of the Unit 2 RBCCW Heat Exchangers for UHS Temperature Increase ZNI Vocumunl Type. OAPD Table 3 -Results for Seismic Event with Off site Power Available Service WaterFlow RBCCW Water Service Water Outlet Outlet Temp 1 Temp 2r633,500 lb/hr (5150 gpm) 91.4-F 89,3F Table 4 -Results for Sin.le SW Pump Supplyin_ Two RBCCW Trains [Service Water Flow RBCCW ' Water Service WaterOu.. uteTapem lb/hr 1 u tl0 W ater T em p u 1 2,276,000 lb/hr (4450 gp2ý) j .... _1 01.50F I98,80F Proto-HX Reports for these runs are Included In Attachment B.

6.0 CONCLUSION

The Millstone Unit 2 RBCCW heat exchanger thermal performance model, Rb-strw.phx developed in Reference 7.3, has been prepared, documented,. and independently verified in accordance with Zachry.Nuclear Engineering's Quality Assurance (QA) Program. This model was run in Proto-HX Shell and Tube Module V4.10, The model is complete and its output is suitable for use in QA calculations. Any conclusions-from those calculations must take into account validation test data measurement uncertainty as appropriate. The Roactor Building Closed Cooling Water (RBCCW) Heat Exchanger operating conditions following a LOCA, a seismic event with offsite power available, and a single service water pump supplying two RBCCW trains during Modes 5-6 have been analyzed, with results reported in Section 5.0. For the LOCA cases, required service water flow rates at the reported temperatures to maintain the SW outlet at 137 'F were obtained. For the seismic event and a single service water pump supplying two RBCCW trains, RBCCW and SW outlet temperatures were calculated using expected SW flow rates for these cases. 6.1 PRECAUTIONS AND LIMITATIONS The current minimum required flowrates for RBCCW listed in Reference 7.4 are based in part on the former LOCA analysis Ihat Is being superseded by Reference.7.6. This will occur when ETE-NAF-2012-0152 (Ref 7,9) has been approved. At that time, this calculation will become the calculation of record for the minimum SW flowrates for RBCCW.

7.0 REFERENCES

7.1 Proto-Power Corporation, Software Validation and Verification Report (SVVR) for Heat Exchanger Thermal Performance Modeling Software - Proto-HX - Shell and Tube Module, SVVR-93948&02-SheIl and Tube, Revision H Version 4.10 7.2 Proto-HX Version 4.10. Shell and Tube Heat Exchangers User Documentation, Copyright Zachry Nuclear Engineering 2010 Form: NO3O1FOS Re~lslon: 00-00 Date: 10-28-2011 Page.1 of 1 Form: N0301 FOS Revision: 00-00 Date: 10-28-2011 Page.I of I

Serial No 13-419 Docket No. 50-336, Page 6 of 6 CAUNO, 12-347 REY o G - C. ACHRY CATOR VERkTA Z*CHRY NUCLEAR, INC. ,TI E Thermal Performance of the Unit 2 RBCCW Heat Exchangers for UHS 'inet )ý: 5A-D -Temperature Increase ',neBU Type: QAPO 7.3 Calculation 00-067, Rev 0 Including CCN 1, 'Analysis of Xk18A and X-18B Thermal Performance Test Results" 7.4 Millstone Calculation 03-ENG-04035M2, Rev 0, 'MP2 Service Water System Design Basis Summary Calculation" 7.5 Millstone Engineering Technical Evaluation ETE-NAF-2012-0117, Rev. 1, '-Transmittal of Millstone 2 LOCA Containment Reanalysis Results using the Dominion GOTHIC Methodology With an Ultimate Heat Sink Temperature of 80'F" 7.6 Calculation S-M2CNTw0436S2, Rev 0, "LOCA GOTHIC Containment Analysis for Millstone Unit 2" 7.7 Flow of Fluids Through Valves, Fittings and Pipe, Technical Paper No. 410, Crane Co, 2010. 7.8 Calculation 97-ENG-01862, Rev 000 Including CCNs t 4, 'RBCCW System Heat Load Flow Rate" 7.9 ETE-NAF-2012-0152, "Implementation of Revised M2 Containment Analysis Results Required to Close 00 00281 and Support Increase in UHS to 80'F" Approval date pending 7.10 Millstone DC MP2-12-01205, "Unit 2! Increase In Ultimate Heat Sink Temperature Limit from. 751F to 80*F" 7.11 Calculation 92-120, Rev 4, 'MP2 SWS Design Basis Alignments - Summer & Winter" 7.12 Bechtel letter dated December 11, 1992, "Millstone Nuclear Plant Unit 2 Bechtel Job 11867-036 NUSCo P.O, 883424 Performance Evaluation of Service Water System Heat Exchangers" 7.13 Proto-Power Calculation 93-049, Rev A, "Fluid Properties - Salt Water-Range 32F tO 320 'IF - Salinity 35g/kg" Form: N0301 F05 Revision ý 00-00 Dame: 10-28-2011 Page I of 1

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 97-ENG-01962, Revision 1 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 8 12,359 (Zachry) AEV I PAGE 5 OF 50 ZACHRY ENG 01962-M2 Rev I forilnon) ZACHRY

NUCLEAR, INCC.VFI9WEn RCS Cool Down Time with Reduced RBCCW Flow During Normal Shutdown ZN1 Document Type: QAPD 1.0 PURPOSE The Shutdown Cooling System (SDC) is designed to cool the Reactor Coolant System (RCS) beginning when SDC Is inltiated (as early as 3.5 hours after reactor shutdown) to cold shutdown temperature. SDC heat exchangers are cooled by the Reactor Building Closed Cooling Water System (RBCCW) which discharges SOC, Spent Fuel Pool Cooling (SFPC) and other miscellaneous heal loads to the service water system (SWS) via the RBCCW Hx.

SDC Hx design RBCCW flow Is 4,820 gpm; however the flow may vary depending on Spent Fuel Pool (SFP) cooling water requirements. During RIS cool down, in order to supply the design flow t I IN0 gpm to the SFP Hx It Is necessary to reduce flow to the SDC Hx which may lengthen the time to reach cold shutdown, During normal operation with two RBCCW trains there is no Technical Specification (TS) time limit to reach Reactor Coolant System (RFS) cold shutdown temperature (I.e.,, RFS temperature below 200:F).. However, it the Refueling Water Storage Tank (RWST) Is not operable, or If one RBCCW train is not operable, cold shutdown must be attained within 30 hours (TS 3/4.5.4) or within 36 hours (TS 3/4.7.3), respectively; The purpose ot this calculation Is to evaluate the Impact of reduced RBCCW flow on the ability of the SDCS to cool down the Reactor Coolant System (RCS) within Technical Specilicatlon time requirements from Its temperalute at the initiation of the Shutdown Cooling System (SDC) until cold shutdown temperature is reached. For the purpose of this calculation the available RBGCW flow to the SDC Hx Is limited to 3,500 gpm. The calculation also determines, the impact of National Pollutant Discharge Elimination System (NPDES) service water temperature discharge criteria on limiting the RBCCW Hx heat load. Two RCS cool down cases at 3,500 gpm RBCCW (low to the SDC Hx are evaluated: Case I - Cool down operallon with two Reactor Building Closed Cooling Water (RBCCW) trains in service, each train supplying 3500 gpm cooling water to each SDC Hx. Both RCS loops, LPSI pumps and SDC heat exchangers are also operating. Case II - Cool down operation with one RBCCW train supplying 3500 gpm to one SOC Hx. For the purpose of this calculation RBCCW Hx available flows are 6,300 gpm RBCCW and 9,000 GPM Service Water at 75 IF as well as 6,400 GPM Service Water at 80'F This calculation also provides an evaluation to demonstrate that the plant can achieve cold shutdown within 72 hours following the most limiling Appendix R fOre scenario. It is noted that the 751 analyses are: maintained for historical purposes only and are used to demonstrate calculation methodology. The 80'F analyses will serve to support plant design basis documentation. Form: N0301F05 Revision: 00-00 Date: 10-28-201 1 Page lol 1 Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page 1.el 1

Serial No 13-419 Docket No. 50-336, Page 2 of 8 LALC O 12-359 (Zachry) REY 0 PAl 6 OP s0 C IC H R Y OR GNA ENG-0 1962-M2 Rev I (Domninon) 0 'ACHRY OUC;LEAR, INC. RCS Cool Down Time with Reduced RBCCW Flow During Normal Shutdown ittet Type: QAPO SUMMAR'*

SUMMARY

OF RESULTS Max SW

Max,

[W Max. RCS R8CCW No. of SRC TiTpme from r Tieo Ro m Case Available INlet Outlet Flow to Flow to B Ta"C, at RA Trip to Mx Trip to A Ch SW Flow ramp. ramp. SOC HX SC HX TrainsC nl (r*om (F) t'F Ipm {pro (IF) {11,ours I hoursll 2.0 l-7 8. 91)00 1751 95 3o00 3000 2 F 300 T ___, __5 il-75 90S00 75 105 3 3000 300 1I 290 4 15.5 2 -0 8288 S.o. 105 5 95 3000 2 300 5.5 a, e 11480 6288 go 8 105 3000 3000 1 280 jIS 25.5 7 AooR j B28 j 80so 105 3000 3000 1 300 1 S 52.4 8 2.1 Graph of RGS temperature versus time after shutdown with two RBOCW trains in service for 751F Service Water Is given in Attachment 1 Figure I. Based on the maximum allowable cool down rate of 80 'F/h above 230 F (Ref. 3.1.19), shutdown cooling Is initiated 3.5. hours after reactor shutdown at 300'F RCS temperature. Estimated time after shutdown to reach cold shutdown temperature of 200IF is 9 hours. This time Is based on both ROPs In operation above 190IF RCS temperature. The plant operator controls the cool down rate by adjusting SOC Hx outlet flow control valve SI-657 within two limitations,. whichever is more restrictive: (1) maximum RCS cool down rate of 80 *F/h above 2301F and 30 "F/h at or below 230'F, or (2) NPDES permit maximum allowable service water discharge temporature of 95?F. Based on the RBCCW Hx available service water of 9,0 00 gpm at 75'F, maximum RBCCW Hx duty Is 87 MMBtu/h, With other RB3CCW heat loads to total 17 MMBtu/h, the allowable SDC Hx duty is 70 MMBtu/h. At the beginning of SOC operation the estimated ROS cool down rate is approximately 6.2 'F/h and reaches 30 'F/h in 9,5 hr. (Refuoling temperature of 140OF Is reached in about 18.5 hours) 2,2 With one RBCCW train In service with 75OF Service Water and shutdown cooling beginning 4 hours alter reactor shutdown at 260F FICS temperature and no RCPs operating, the estimated time to reach cold shutdown temperature is 15.5 hours after shutdown. Maximum SOC Hx duty is 104.1 MMBtu/h. Throughout SDC operation neither the RCS cool down rate nor service water discharge temperature are limiting and SDC (RCS) flow Is maximum at 3000 gpm (Attachment 2 Figure. 2). 2.3 Graph of FIGS temperature versus time after shutdown with two RBCCW trains In service with 80 F Service is given in Attachment 6 Figure 1. The service Water system capacity is sufficient to remove shutdown cooling hoal loads while maintaining discharge tomperature below 105'F, postulating an 80*F intake temperature with two RBCCW trains In service. Estimated time after shutdown to reach cold shutdown temperature of 2001F Is 14 hours after shutdown. Based on the RBCCW Hx available service water of 6288 gpm at 80°F, maximum RBCCW Hx duty is 77 MMBtu/h. This flow rate is within the system capacity given operator action to reduce service water flow through the non-vital TBCCW heat exchangers using the Temperature Control Valve adjustable setting. Service water flow to the RBCCW heat exchanger may be reduced due to pump degradation, heat exchanger touling, or diversion of flow through the TSCCW heat exchanger. With other RBCCW heat loads to total 17 MMBtu/h, the allowable. SOC Hx duty is.60 MMBtulh. At the beginning of SDC operation the estimated RCS cool down rate Is approximately 0.3 'F/h and reaches 30'F/h in 14.8 his after shutdown when the reactor coolant pumps are secured. (Refueling temperature of 1401f Is reached in about 25 hours) 2.4 With one RBCCW train In service and shutdown cooling beginning 19 hours after reactor shutdown at 260 IF R0CS temperature and no RCPs operating, the estimated lime to reach cold Form: N0301F05 Revision: 00-00 Date: 10-28.2011 Page I 011 Form: N030IF05 Revision: 00-00 Date: 10-28-2011 page I of 1

Serial No 13-419 Docket No. 50-336, Page 3 of 8 r,413; N,. 12-359 (Zachry) qFv 0 7 50 Z A CE IR Y ORGNATO 97.ENG.01962.M2 Ray 1 (Dominaon) 1 ZACHRY NUCLEAR, INC. Zn,.e RCS Cool Down Time with Reduced RBCCW Flow During Normal Shutdown ZNI Dneumrnfn Type: GAPe shutdown temperature is 25.6 hours alter shutdown. Maximum SOC Hx duty is 60 MMBtu/h to ensure a service water: discharge temperature below 1051F. These times are based on an available service water flow of 6288 gpm at 801F to the RBCCW Hx (Attachment 7). The RCS relueling temperature Is not reached In a reasonable time In this constrained alignment. Either a second train of RBCCW would be necessary or the service water flow rate to the RBCCW heat exchanger would need to Increase. 2.5 With one RBCGW train in service, the time it takes to reach an RCS temperature of 206,F from SDC entry conditions after an Appendix R Control Room Fire Is 2.4 hours (Attachment 8). This duratIon should be added to the maximum amount of time Reference 3,1.32 Indicates as the time to reach SoC entry conditlons tor an R-1 Fire Scenario (52.52 hra). Therefore, the total time It takes to reach cold shutdown following an R-i Fire Scenario is approximately 55 hrs.

3.0 REFERENCES

AND DESIGN INPUTS

3. 1 References 3.1.1 Engineers & Fabricators Exchanger Specification Sheet, Shutdown Cooling Heat Exchanger, Rev. B (6-3-75) 3.1.2 Struthers Wells Exchanger Specification Sheet,. Reactor BUilding Closed Cooling Water Heat Exchanger, Revised 6/30/72 3.1.3 Strurhers Wells Calculation for Bechtel Corporation, "Thermal Performance Analysis of Reactor Building Closed Cooling Water Heat Exchanger Supplement 92, Rev; 0, 10/29193.

[RBCCW Hx analysis with 75SF SW and 10% tube pluggodI 3.1,4 Branch Technical Position ASB 8-2 Residual Decay Energy for Light-Water Reactors ior Long-Term Cooling, Rev. 2 July 1981. [Curve.of Residual Decay Heat Release vs Time After Shutdownj 3.1.5 NPDES Permit, CT Department of Environmental Protection, expiring date 12J14197. 3.1.6 NU Memorandum, Jim Foertch/Milan Keser (NUEL, Millstone) to Marcel. Ranieri (BOP Engineering, Berlin) 'Potential to exceed NPDES permit temperature limit In MPS SWS during accident scenario", September 2,1993 3.1.7 Technical Specification 3/4.4,9 Pressure/Temperature Limits Reactor Coolant System, PTSCR 2-17-97. 3.1.8 Technical Specification 3/4.7,3 Reactor Building Closed Cooling Water System, August 1, 1975 3.1.9 Technical Specification 3/4.5A4 Refueling Water Storage Tank, 8/1/75 3.1.10 Deleted 3.1,11 Standard ot the Tuibular: Exchanger Manufacture's Association, Seventh Edition, 1998, Pages 103-112. 178 Form' N0301 F05 Revision: 00-00 Date: 10-28-201b Pafe I of I

Serial No 13-419 Docket No. 50-336, Page 4 of 8 ALG

• NO 12"359 (Zachry)

REV 0 R VAGE 8 OF 50 IH R Y RINNO 97-ENG-0,i 62-M2 Rev 1 (Dominiz~on) /ei~ )CLEAR, INC. Til RCS Cool Down Time with Reduced RBCCW Flow During Normal Shutdown P. QAPD 3.1.12 Tsou, J.L., Power Plant -Heat Exchanger Performance Prediction A Simplified Method, ASME Paper 84-JPGC-NE-14. 3.1.13 Rohsenow, Warren M., Choi, Harry Yj Heat, Mass and Momentum Transfer, Prentice-Hall, 1961 3.1.14 Saline Water Convorsion Engineoring Data Book, M.W. Kellogg Co., 1975 3.1.15 Bowmanj RA., Mueller, A.C. and Nagle, W.M., Mean Temperature Dilference In Design, Transactions of ASME, May 1940, pp. 283-293 3.1.16 Perry's Chemical Engineer's Handbook, Sixth Edition, McGraw Hill, 1984 3.1.17 "Shutdown Cooling/RBCCW Head Loads", Combustion Engineering Letter MP2-CE-186 dated 8/8/69 3.1.18 RE&C/NUSCO RBCCW System Meoeing Minutes, 10/31/97 3.1.19 Calculation 97-SDS-1760 M2 Roy. 1, Millstone 2 Prossure/Tomperature Umif Curves for 20 EFPY 3.1.20 Core Heat Balance EN21002 Royv 10, 3/19/97 3.1.21 Calc 97-ENG-01862-M2 RBCCW Rev. 0. System Heat Loads and Flow Rates. 3.1.22 OP 2310 R20, Operating Procedure Shutdown Cooling and Flows 3.1,23 Technical Specilication Table 1.1 Operational Modes, Amendment 72, 2/22/82 3.1.24 Steam Surface Condenser Standards, Heat Exchange Institute, Eight Edillon, 1989 3;1.25 Design Basis Document Package Reactor Coolant System, DBDP MP2-RCS Rev. 1, 811/95 3.1.26 Heatup/Cooldown Operator Calculations, Calculation HUCD-O 1800M2. Rev. 0 3.1.27 Design Basis Document Package DBOP MP2-ECCS Rev. 0, 1/1/93 (For informalion only) 3.1.28 Deleted. 3.1.29 Calculation 92-120 Revision 5 MP2 SWS Design Basis Alignments - Summer & Winter 3.1.30 Calculation 99-005 Revision 00 MP2 - SWS Evaluation of LOCA without LNP Flows During Summer Operation and RBCCW TCV Required Positions 3,1.31 NPDES Permit ID CT0003283, CT Department of Environmental Protection, expiring data 8/31/2015. 3.1.32 Calculation W2-517-00744RE Revision 3 - MP2 Appendix R Cooldown Form: N0301 FOS Revision: 00-00 Date: 10-28&2011 Page I of 1

Serial No 13-419 Docket No. 50-336, Page 5 of 8 3.1.33 3.1.34 3.1.35 3.2 Design 3.2.1 Calculation S-02824S2 Revision 2 - Millstone Unit 2 R2 Fire Appendix R Analysis Engineering Record Correspondence 25203-ER-98-0089 Revision 0 - Millstone Unit 2 RCS P-T Limits for Figures 3.2 & 3.5 of the EOP(s) Calculation 92-120 Revision 5 MP2 SWS Design Basis Alignments - Summer + Winter Inputs [REFERENCE] Reactor Coolant System

1.

Reactor design thermal power 2700 MW

2.

Operating conditions at the start ot shutdown cooling operation Two RBCCW Trains Operaltin Start of SDC hr 3.5 RCS Temperature IF 300 Number of RCPs Operating 2 RCP Heat Input Power KW 8500 One RBCGW Train Operatlng Start of SDC hr 4 RCS Temperature

• F 280 Number of RCPs Operating 0

RCP Heat Input Power kW 0

3.

RCS Maximum Cool Down Rate RCS temperature above 2301F 80 CFih RCS temperature 230 0F and less 30 0F/h

4.

RCS Heat Capacity With 2 RCPs operating 1,2000,000.Blu/'F With 0 RCPs operating 317,000 BtuI0F

5.

Residual Decay Heat Release 3,2.2 Service Water System Temperatures

1.

Service Water Temperalure 75"F 801F [Rot. 11.251 [Ret 3.1,17] [Rat 3.1.23] [Rot 3,1.26] [Ref 3.1.18] (Rot 3.1,18] [Ref 3,1.181 [Ref 3.1,18] [Ref 3.1.19] [Ret 3.1.19] [Ret 3.1.261 [Ref 3.1.26] tRet 3.1.41 (Rot 3.1.31 Assumption 4.8 Form: N0301F05 RevIsion; 00.00 Date: 10-28-201 I Page I of I Farm: N0301 F05 Revision: 00-00 Date.- 10-28-2011 Page i.of i

Serial No 13-419 Docket No. 50-336, Page 6 of 8

2.

Expired NPDES Permit DischargeTemp. Limit: 95IF max Current NPDES Permit Discharge Temp. Limit: 105IF max

3.

Discharge Temperature wilh 1 RBCCW Hx: 105'F max 3.2.3 RBCCW Heat Exchanger Operating Parameters During Shutdown Heat Duty Other Than SDC Hx (Train B) 14,449,000 Btu/h (Conservatively use) 17 MMBtufh Shell side fouling resistance 0.0005 h-ft'- F/Btu Tube side fouling resistance 0.0005 h-ft-FtiBtu Tubes Plugged 10% RBCCW Flow Train A 8 6,769 gpm Train B = 6,709 gpm Service Water Flow Minimum 3,150,000lb/h Maximum 9,000 gpm Limiting Condilions 6400 gpm 3.2.4 Shutdown Heat Exchanger Operating Parameters During Shutdown Shell side fouling resistance 0.0005 h-ft'-eFIBtu Tube side fouling resistance 0.0005 h-ftt-ýFlBtu Tubes Plugged 10% RBCCW Flow 3,500 gpm SDCHx Flow 3,000 gpm maximum LPSI Pump Heat Input 400 hp (each) [Ref 3.1.5]T (Ret 3.1.31] [Ref 3.1.31] [Ref 3.1,21] (Ret 3.1.3] [Ref 3.1.3] [Ref 3.1.3] [Ref 3.1,21 ] [Ret 3.1.2] Assumption 4.7 Assumption 4.7 [Ret 3.1.1] [Ret 3.1.1] (Ret 3.1.18] tRot 3.1.22] [Rot 3.1.20] 3.2.5 RBCCW Heat Exchanger Speciilcation Parameters Manufacturer Heat Transferred Btulh Heat Transfer Rate-Clean Btu/h-Iltt-F Heat Transfer Rate - Service. Btuthwftf-F No. Tubes Tubes Plugged Effeclive Surface Area itt Tube Material Tube Outside Diameter In Tube Wall Thickness BWG uin) 'Tube Inside Diametor in [Ret 3.1.21 Struthers Wells 36,900,000 425.8 292.1 1473 10% 7468.5 AI-Brass (8-11 1-8) 0.75 18 (0.04g) 0.652 Expired Permit Discharge Limit cases kept for h*sllorcal purposes because.the case which uses 95"F Outlet temperature was used to validate spread sheet calculations. Fonn: N030IF05 Revision: 00-00 Date: 10-28-2011 Page 1 of 1

Serial No 13-419 Docket No. 50-336, Page 7 of 8 Service Conditions Shellid Tubeio. Fluid Water Seawater No. Passes I I Flow b/lh 3,539,309 2,370,240 Temperature In I 105.4 75 Temperature Out IF 95 91.3 Fouling Resistance h-ltt-F/Btu 0,0005 0.0005 3,2.6 RBCCW Heat Exchanger Specification Parameters [Rot 3. 1. t11 .Man ufacturer Heat Transferred Heal Transfer Rate - Clean Heat Transfer Rate Service No. Tubes Tubes Plugged Effective Surface Area Tube Material Tube Outside Diameter Tube Wall Thickness Tube Inside Diameter Service Conditions Fluid No. Passes Flow Temperature In Temperaturo Out Fouling Resistance Btulh Btu/h-l'- "F In BWG (in) in Ib/h IF IF ,-e-t/= t Engineers & Fabricators, Inc. 27,200,000 400 256 659 0% 5790 304 Stainless Sleet U 0.75 18 (0.049) 0.652 Sholl Side Tube Side RBCCW Wir Borated Water 1 2 2,410,000 1,500,000 95 130 106.3 111.9 0.0005 0.0005 3.2.7 Appendix R Fire Shutdown Coo!lng Entry Conditions Start of SDC hr 52.52 RCS Temperature 'F 277.8 [Ref 3.1.321 [Ret 3,1.321 4.0 ASSUMPTronS 4.1 No RCS heat loss to ambient. This conservatively maximizes the required SDC heal load. 4.2 No heal loss to ambiont tram SOC System.*This conservatively maximizes SDC heat load. 4.3 Conservatively assume higher RCP total brake power of 8,550 kW versus design input 8,500 kW [Ref. 3.1.41. 4.4 Assume that RBCCW flow to the RBCCW Hx Is 6,300 gpm based on enveloping the design input elows [3.2.31. Feorm: N0301 FOS Revision: 00-00 Date: 10-28-2011 Page i of I

Serial No 13-419 Docket No. 50-336, Page 8 of 8 A 0 12-359 (Zachry) RE 0 1 A~ 12 50 Z A C H R Y AMAO 97-ENG-0I 562-M2 Rev I(DmIin ZACHAV NUCLEAR, MC. Tr RFS Coal Down Time with Reduced RB5CW Flow During Normal Shutdown ZN1 Document Type: OAPO 4.5 Similar to RBCCW Hx performance calculations [Ref. 3.1.3] assume that 10% of SOC Hxtubes are plugged to provide realistic performance margin. 4.6 Conservatively assume RBCCW heat loads other than SDC Hx total 17 MMBIu/h per train. [Para. 3.2.3]. 4,7 Conservatively assume that RBCCW Hx service water Ilow ranges from the Hx rated flow of 3.150,000 lb/h [Raf. 3.1.2]. When analyzing conditions with 75$F service water, a maximum flow rate to 9.000 gpm Will be used, These cases are used for validation of methodology only and rolect earlier revisions of this calculation. To demonstrate limiting conditions of operations, service water flow will not exceed 6,400 gpm when analyzing conditions with 80°F service water supply temperatures. 4.8 In order to demonstrate coal down capability with elevated ultimate heat sink temperatures, analyses will be performed assuming both 75"F and 80"F service witer supply temperatures. 4.9 The current NPOES permit (Ref. 3.1.31] service water.discharge temperature limit is 105*F. This is an Increase over the limit defined in the expired permit [Ref. 3.1.6j; 95"F. The service water discharge temperature limit in the expired permit applied to normal operation or expected off-normal operations (e.g., refueling outages and cleanings) and was not intended to limit activities during an accident or abnormal operation. Two-train analysts with 751F service water will be based on the expired limit (limit service water discharge temperature to 951F). Single train analysis and all analyses with 804 service water will be based on the current temperature limit of 105$F 4.10 in order to conservatively bound RFS conditions from the standpoint of the amount of sensible and decay heatthat needs to be removed by SDC following an R-1 Fire Scenario (Design Input 3.2,7), a SOC entry time of 50 hrs after reactor trip and an ROS temperature of 300F will be Considered. Using these values means that the time found to cool from SDC entry to cold shutdown will need to be added to the time it takes to reach SDOC entry conditions under an Appendix R tire. 5.0 METHOD OF CALCULATION 5.1 Nomenclature The following abbreviations are used In cool down and heat exchanger performancecalculations, Symbol Unit A Heat exchanger effective surface area ftV At Tube inside area ftt Ct Shell fluid specific heat Btullb-'F Ct Tuba fluid specific heat Btullb-IF Os Shell fluid density tbft0 ot Tube fluid density ib/et 5 do Tube outside dlameter in di Tube inside dlarniator in dt RCS cool down time Increment H dT/dt ACS cool down rate

• F/h F

LMTD Correction Factor Hi Tube inside film coefficient Biulh-ft1 -'F Ho Tube outside film coelticient Btu/h-1l1-'F Form: NO30 IF05 Revision: 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 97-169 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

1.0 Serial No 13-419 Docket No. 50-336, Page 1 of 42 L C HR Y "ALC" Basi 9 4[1PACE'6 OF 48 ZACHRY NUC.EAR, INC. TITLE MP2 RBCCW - Design Basis Flow Distribution nernt Type: QAPD PURPOSE The purpose of thisocalculation is to determine the flow balance and maximum component outlet temperatures.for the Millstone Point Unit 2 (MP2) Reactor Building Closed Cooling Water (RBCCW) System using the PROTO-FLO-' model of the system documented-in Reference 4.1 for the following design basis operating conditions:: I Normal Operation II 3.5 Hours after Shutdown Operation Ill 27.5 Hours after Shutdown Operation IV Loss of Coolant Accident (LOCA) Injection Operation V LOCA Recirculation Operation It is noted that due to the need to periorm operator action, Case V actually consists ot three discrete sequential operating conditions: V w/NE Operation with Non-Essential Components on-line. V w/o NE Operation with Non-Essential Components secured V w/SFP Operation with spent fuel pool cooling restored Flow balance results for Case II with a Loss of Instrument Air (LIA):are also provided. This additional condition is designated: It w/LIA 3.5 Hours after Shutdown Operation with Loss of Instrument Air In the event that operators elect to draw a sample via the Post-Accident Sampling System (PASS) following an accident, RBCCW cooling water will be aligned to the sample coolers:X-64 and X-65 by opening Isolation valve 2-RB-210. This will result in a redistribution of flow through the system. The following cases have been analyzed to document the system.flow distribution with the PASS sample coolers in the flow path: IV (PASS) LOCA Injection Operation (with flow to PASS sample coolers) V w/NE (PASS) LOCA Recirculation Operation with Nan-Essenlial Components on-line (with flow to PASS sample coolers) V w/oNE (PASS) LOCA Recirculation Operation with Non-Essential. Components secured (with flow to PASS sample coolers) V w/SFP (PASS) LOCA Recirculation Operation with spent fuel pool cooling testored (with flow to PASS sample coolers) Form: N0301F05 Revision: 00-00 Date: 10-28-2011 Page-i at 1 Form: NO0301F05 Revision: 00-00 Date: 10-28-2011 Page'1I of 1

Serial No 13-419 Docket No. 50-336, Page 2 of 42 ,CALCNO. 97-169 1" .4 r^GA 7 o[W 48 ZACHRY

NUCLEAR, INC Tmi MP2 RBCCW -Design Basis Flow Distribution ZNI ocument* Typa QAPV All system alignments are analyzed for 75I Service Water (SW) conditions as well as 80IF SW conditions except for the Normal Operation case.

The maximum allowable RBCCW outlet temperature is 851F for Normal Operation. This temperature is not achievable for Normal Operation with a SW temperature of 801F with 10% tube plugging, design fouling, and -the assumed current SW flow rates of 7500 gpm for the A Train and 7650 gpm for the B train, in accordance with Assumption 5.3. As a result, an analysis of the RBCCW HXs using the Proto-HX model documented in Reference 4.19 Was Included to determine the fouling, tube plugging, and SW conditions at which the 85OF RBCCW temperature can be maintained. All.Proto-HX output reports are included in Attachment L. In addition to performing the system flow balance and maximum temperattre analysis, this I calculation will: determine required throttle positions for the Shutdown Cooling Heat Exchanger (HX) (X-23A/B) throttle valves (2-RB-13A/B, 2wRB-i3.1A/B, and 2-RB-14A/8) for both shutdown cooling (Cases It and (l1) and LOCA recirculation (Cases V w/NE, V w/o NE and V w/SFP) operations, = assess the potential cavitation in the Shutdown Cooling HX (X-23A/B) throttle valves,, evaluate available net positive suction head (NPSHA) versus the RBCCW pump's required net positive suction head (NPSHR).for worst case temperature conditions, and evaluate the required brake horsepower for the RBCCW pumps for each operating

case, This calculation supersedes Proto-Power Calculation 97-038 as the design basis flow distribution for the MP2 RBCCW system. This calculation can also be used to supersede calculation 97-064 as the maximum outlet temperature analysis for the Unit 2 RBCCW.

2.0

SUMMARY

OF REsuLTS The RBCCW system was successfully balanced to the maximum flows consistent with Reference 4.47, and minimum required flows provided in Reference 4.2 (provided as Attachment A) with margin as prescribed in Reference 4.3 (provided as Attachment B) and as modified by Reference 4.8 (provided as Attachment J), with the following exceptions: For Case II, 3.5 hours following shutdown, the required design flow of 165 gpm is based upon maximum operating conditions where the containment air temperature is at 1201F. Currently, the flow at an 804F UHS temperature forea degraded pump falls between 1 and 2 gpm below the design limit, to 163 gpm. As the design flow is based upon a containment air temperature of 120"F, which is significantly higher than containment temperatures observed following shutdown, the difference ot flow of 2 gpm from the design flow is considered insignificant. Additionally, during this period two Reactor Coolant Pumps instead of four will be operating 3.5 hours following a shutdown. This condition coupled with a significantly lower RCS temperature (as a result. of the plant shutdown) also demonstrates that the 2 gpm discrepancy Is insignificant. X-192 Sample Cooler in Case II and Case V w SFP alignments could not be throttled to required flows. With the throttle valve for this flow path, 2-RB-359, 100% open, this Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 3 of 42 CAISN 97-169 4 8 48 ZACINRY OWMT4-&f,, m ZACHRY NUCLEAR. INC. T".E MP2 RBCCW - Design Basis Flow Distribution ZN1 Voaument Type: QAPD component receives 4.96 gpm of the 5 gpm required In Case II. Although notable, the H1ow of 0.04 gpm less than the required is not considered significant. Flow to the PASS sample coolers (X-64 and X-65) are indicated as having less than the design flow for the Normal Operations Case (29 gpm versus 31 gpm). Samples are drawn from the RCS and cooled through these heat exchangers. In addition to the minimal usage time of these coolers (<3 hrs per sample), the use of these sample coolers is procedurally limited during elevated UHS temperatures (>750F) in order to maintain the RBCCW outlet temperature at or below 859F. This impact is considered minimal as the sample may be cooled longer to reach the optimal sample temperature. With the CAR cooler throttle valves positioned to maintain maximum flow rates for nominal pump performance, flow margin above the design flow rates cannot be maintained for all degraded pump performance cases. The A Train components (X-35A and X-35C) for Case II receive the design flow with a 97 gpm of the 100 gpm margin for a total of 597 gpm instead of 600 gpm. For Case IV these components receive design flow with 68 gpm of the 100 gpin margin for a total of 2068 gpm instead of 2100 gpm. Although the available margin for tho CAR coolers has decreased as a. result of the changes to the modal, it Is still above the design minimum and therefore is acceptable. The system flow balance takes into account the Valve position changes documented in References 4.5 and 4.6. A table of the predicted flow rates using both a nominal pump curve and a degraded pump curve in accordance with Reference 4.4 is provided in Attachment C. Except as noted later, this proves the existing RBCCW pumps to be adequate for all design basis operating conditions. During Normal Operation and LOCA Injeclion, the Shutdown Cooling (SDC) HXs 2-RBr13.TA/B are closed while 2-RB-13A/B and 2-RB-14AIB are fully opened, Isolating flow to the SDC HXs. To control flow through the SDC Hx, valves 2-RB-13AJB and 2-RBOf4ANB are manually throttled, coincident with the preset open position of valves 2-RB-13.11AB. During normal shutdown conditions, valves 2-RB-1 3A/B and 2-RB-14A/B are fully opened per Reference 4.6, where 2-RB-13.1A/B will open to their present position. Once normal shutdown is completed and prior to restoring the plant to normal power, valves 2-RB-13A/B and 2-RB-1 4A/B are restoaed to their throttle accident flow condition. The required positions for Shutdown Cooling HX throttle valves 2-RB-13A/B, 2-RB-13.1A/B, and 2-RB-14A/B are provided in Table 1: Form: N0301F05 Revision: 00-00 Date: 10.28-2011 Page I of I Form: N0301 FOS Revision: 00-OD Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 4 of 42 Table 1 - Shutdown HX Required Valve P sitions Valve Tag Approximate Position Applicable Cases(') 2-RB-13A 100_ 9% Open _____ . I1,.ll I.. 2-RB-13A 13.8% Open V 2-RB-138 100% Open .1, I1. III,. IV 2-RB-13B 12.7% Open V 2-RB-13.1A 16.75% Open III IIIV 2-RB-13.1A Shut I, IV 2-RB-13.1B 16.68%"Open II, Ill, V 2-RB-13.1 B Shut I, IV 2-RB-14A 100% Open 1, 11, Ill, IV 2-RB-14A 13.8% Open V 2-RB-14B 100% Ope

__IIII, 2-RB-14B 12.7% Open V

h~ole. (I) Case II encompasses bolth d t nndII W1,LIA, Case IV anlompasses both IV and IV (PA5SS. Case V encompasses VwINl_, V wto NE. and V WISFP tall wIlh and withaul PAWSS. Cao I and IV aiignmoenls do el lno.lude Iow to the Shutdown Heat Exchangers. Following: the method of Tullis per Reference 4.10, valves 2-RB-13.1A/B will operate in the "incipient cavitation" regime (as defined in Reference 4.10 and section 6.4 of this calculation) for Cases II and Ill. Valve 2-RB-14A operates In the incipient regime for LOCA Recirculation prior to isolating non-essential components (Case V wiNE), for operation with 80"F SW. For all other cases valves 2-RB-13A/B, 2-RB-13.1A/B, and 2-RB-14NB operate outside the cavitation regime. Cavitation damage is not expected, however, due to: the valves being nearly shut, regions of high flow velocity may exist within the valve which may cause damage. Based on Section 2.1.2 of Reference 4.6, incipient cavitation, vibralioni and noise are not of any concern. The RBCCW System provides sufficient NPSHA to operate the RBCCW pumps for all examined operating cases with a minimum margin of 77.56ft for the 'A' pump in Case IV with SW at 80 0F and nominal pumps. When recalculated for a conservative inlet temperature of 2501F in accordance with Assumption 5.1, this margin is 28.21 ft. NPSHA and NPSHR data for all operating cases and both RBCCW trains is provided in Section 7.4. The RBCCW pumps require a maximum brake horsepower of 315 hp in a variety of cases. Brake horsepower requirements for all operating cases and both RMOMW trains are provided in Section 7.5. Alignment of the RBCCW system to provide cooling water to the PASS sample coolers X-64 and X-65 during Cases IV and V results in a net decrease in the calculated flow to the other components in the "B" header. In most cases, the required flow to safety related components is maintained In spite of the flow diversion to the branch line supplying the PASS sample coolers. However, the delivered flow has dropped below the required flow (design flow with margin) for the CAR HXs for the cases shown In Table 2 and the B-Train Shutdown Cooling HXs for the. cases shown In Table 3. Form: N0301 F05 Revision. 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 5 of 42 T i 3 AU ik1 A. 0 9 1 rIg At Il aJ LIf IVPMI Delivered Flow (gpm) 4---.-.. 4 1 Component CAR Unit A (X-35A).56 CAR Unit C IX-35Cl Design Flow (gpm) 2000 2000 Design Flow wl Margin (gpm) IV (PASS) 75'F SW Degraded Pump IV (PASS) 80'F SW Degraded Pump 2100 2068 2100 2072

4.

4 CAR Unit B (X-35B) 2067 2071 2071 2069 2000 2100 CAR Unit D (X.35D) 2000 2000 2072 2070 2100 Table 3 - PASS Cases with B-TLrain SDC (X-236) Flow eslow Marqin I -Design Flow Design Flow Delivered Case (gpm) w/ Margin Flow (gpm) V w/NE (PASS) 75"F SW 2000 2100 2087 _DPegraded Pump___ V w/NE (PASS) 80IF SW 2000 2100 2086 V W/ SFP (PASS) 75'F SW 1800 1900 1850 ..Degra~ded pu.mp........ V w/ SFP (PASS) 80'F SW 1800 1900 1850 Degraded Pump I __ All component flow margins can be found in Table 7 and Table 8. The above conditions are acceptable because all components still receive design flow. All component outlet lemperatures are listed in Attachment L. The 85 1F maximum RBCCW supply temperature is not achievable for Normal Operation for 80"F SW with the expected flow rates listed in Assumption 5.3j the minimum tube plugging limit for each heat exchanger (3% for X-18A and 5% for X-18C) and fouling equivalent to the clean condition analyzed in Reference 4.19. The 859F RBCCW supply temperature can be maintained for a SW temperature of 79OF Form: N030 IFOS Revision: 00-00 Date: 10-28-2DI I Page I of I

Serial No 13-419 Docket No. 50-336, Page 6 of 42 with design fouling and 7% tube plugging for both heat exchangers at nominal and degraded RFBCCW tlow rates and for a SW temperature of 78 0F with design fouling, and 10% tube plugging for both heat exchangers at nominal and degraded RBCCW flow rates. :For a SW temperature of 791F, the RBCCW supply temperature can be achieved with 10% tube plugging with lower fouling limits for each heat exchanger and RBCCW flow condition, as shown in Table 4 below. The overall fouling for a clean heat exchanger is 0.00082.(hr-e-fF/13TU) (Reference 4.19). For each new overall fouling limit listed, a time to reach that level of touling has been estimated based on the linear touling rate calculated from the results of reference 4.19 as described in Design Input 3.4. The conditions and iterated variable are described in Section 6.8. All component outlet temperatures and Proto-HX reports are included in Attachment L. Table 4: Proto-HX RBCCW.Suool Temperature-FEesuits Overall Tube RCCW Cleaning NX Heat Load SW Flow SW Temp RBCCW Fouling Plugging supp Frequench (mBTU/hr) (gpm) (F) Pump Curve (hr-ft=-F/BTU) Temp(MF X-18A 15.369 7500 80 Degraded Design 10 86 X-18A 15,369 7500 80 Nominal Design 10 86.1 X-18A 15.369 7500 80 Degraded 0.00082 3 85.2 0 (clean) X-18A 15.369 7500 so Nominal 0.00082 3 85.2 0 (clean) X-18A 15.369 7500 79 Degraded Design 10 85.1 X-18A 15.369 7500 79 Nominal Design 10 85.1 X-18A 15.369 7500 79 Degraded 0.00101 10 84.9 3.6 X-18A 15.369 7500 79 Nominal 0.00101 10 84.9 3.6 X-18A 15,369 7500 79 Degraded Design 7 84.9 X-18A 15.369 7500 79 Nominal Design 7 B4.9 X-18A 15.369 7500 78 Degraded Design 10 84.1 X-18A 15.369 7500 78 Nominal Design 10 84.1 X-18C 15.084 7650 80 Degraded Design 10 85.9 X-18C 15.084 7650 80 Nominal Design 10 86.0 X-18C 15.084 7650 80 Degraded 0.00082 5 85.2 0 (clean) X-18C 15.084 7650 80 Nominal 0.00082 5 85.2 0 (clean) X-18C 15.084 7650 79 Degraded Design 10 85.0 X-18C 15.084 7650 79 Nominal Design 10 85.0 X-1BC 15.084 7650 79 Degraded 0.00105 10 84.9 4.4 I Form: N0301F05 Revision: 00-00 Date; 10-28-2011 Page 1 at 1 Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 page 1 el I

Serial No 13-419 Docket No. 50-336, Page 7 of 42 NX Heat Load (mBTU/hr) SW Flow SW Temp RBCCW (gpm) (7F 1 Pump Curve overall Fouling (hr-ftz-*F/BTU) Tube Plugging X R8CCW Supply Temp (IF) cleaning Frequency Months)" X-18C 15.084 7650 79 Nominal 0.00105 10 84.9 4.4 X-18C 15.084 7650 79 Degraded Design 7 84.8 X-18C 15.084 7650 79 Nominal Design 7 84.8 X-]8C 15.084 7650 78 Degraded Design 10 84.0. X-18C 15.084 7650 78 Nominal Design 10 84.0 Note 1: Approximato clearing froeclioncy provided only for overall fouling oiler than design fouling: The SW flow rate required to achieve less than 85OF RBCCW supply temperature was determined for each heat exchanger and RBCCW pump performance for design fouling and 10% tube plugging, and then for a condition equivalent to approximately 3 month cleaning frequency and 5% tube plugging. The resulting required SW flow rates are shown in Table 5 below. Table 5: Proto-HY RBCCW Sut iy Temoerature Results Overall Tube RBCCW cleaning lHeat Load SW Flow SW Temp RBCCW Fouling Plugging Supply Frequency (mBTU/hr) (gpm) (*F) Pump Curve (rft.F/BTU) Te F reunc Tep (Manths)Ul X-19A 15.369 10750 80 Degraded Design 10 84.9 X-LBA 15,369. 10850 80 Nominal Design 10 84.9. X-18A 15.369 9300 80 Degraded 0.00098 5 84.9 3.1 X-18A 15.369 .9300 80 Nominal 0.00098 5 84.9 3.1 X-18C 15.084 10700 80 Degraded Design 10 84.9 X-18C 15.084 10800 80 Nominal Design 10 84.9 X-18C 15.084 9200 80 Degraded 0.00098 5 84.9 3.1 X-18C 15.084 9250 80 Nominal 0.00098 5 84.9 3.1 Nola 1: Approxrrnate cleaning frequency provdes only for overall fouling otheirr Inan design fouling: The fouling resistance rate of.change for the RBCCW heat exchanger is 2.3572 x 10" (hr-1t1-0F/BTUJ) over 4.5 months based on the test data documented in Table 21 of Reference 4.19., as stated in Design Input 3.4. For a linear fouling rate of change the time to reach the overall fouling limit from a clean fouling of 0.00082 (hr-ft7-'F/BTU) is estimated for each condition above. The estimated cleaning frequency should be determined by rounding down from these Values. Form: N0301 FOS Revision: 00-00 Dale: 10-28-2011 Page I of 1

3.0 Serial No 13-419 Docket No. 50-336, Page 8 of 42 '0' 97-169 mV 4 PAGE 13 48 ICEIRY R11AR EIA ZAC4RY NUCLEAR. INC. TIRE MP2 RBCCW - Design Basis Flow Distribution ment Type: QAPO DESIGN INPUTS 3.1 The Proto-Flo ihermal :hydraulic model used in this analysis was originally documented in Reference 4.1. Changes were made: to the model in Revision 4 of this calculation for consistency with other system analyses and modifications to the RBC0W system and to aid in model convergence. 3.2 The RBCCW shell and tube heat exchanger model used in this analysis was imported from Reference 4.19 using Proto-HX Shell and Tube Module Version 4.10, developed in accordance with Reference 4.45. 3.3 Revision 4 of this calculation supersedes the component outlet temperature analysis originally performed in Revision 3 of 97-064. Temperature analysis was added by assigning fixed heat loads to the ProtowFlo heat exchangers consistent with Reference 4.2 using the methodology from Reference 4.1, except where updated in accordance with References 4,48, 4.49, 4.50, and 4.51. LOCA heat loads for the SFP, SDC, and CAR heat exchangers provided in Reference 4.48 are consistent with a UHS temperature of 80'F and are bounding for 75°F SW cases. The CEOM heat load was varied based on the table of maximum CEOM heat loads versus RBCCW supply temperature presented in Reference 4.49. Heat load assignment methodology Is further described in Section 6.8. 3.4 The fouling rate of change for the RBCCW heat exchanger is 2.3572 x 10'4 (hr-ft, "F/BTU) over 4.5 months based on the test data documented In Table 21 Reference 4.19. 3.5 The Proto-Flo case alignment included in Attachment D for Normal Operation with elevated SW temperature uses a SW flow rate of 7500 gpm for the A Train HX, X-18A, and 7650 gpm for the B Train RBCCW HX X-18C, as described In Assumption 5,3. The SW flow rate for LOCA conditions is 7570 gpm based on the LOCA containment analyses documented in Reference 4.44. All SW flows were assigned to Ihe tube-side of the RBCCW heat exchanger as shown In Table 6 below: Form,' NO301F05. Revision. 00-00 Date: 10-28-201 g Page I of 1

Serial No 13-419 Docket No. 50-336, Page 9 of 42 Table 6: Service Water. Flow Rates Case Case Description SW Flow Reference (GPM) 7500 I Normal Operation 7650 Assumption 5.3 II 3.5 Hours After Shutdown 9000 4,43 1II 27.5 Hours Alter Shutdown 6000 4.43 IV LOCA Injectlon 7570 4.44 V w'NE LOGA Reciro w/ Non-7570 4.44 Essential

• V wIot1E LOCA Retire w/o Non; EwA on/Iar 7570 4.44 v w/SFP LOCA Roc1rc w!SFP 'IlXs 7=570 4.44 3.6 The system was balanced to minimum required flow rates documented in Reference 4.2 with margins consistent with the guidance in References 4.3 and 4.8. This was done while also maintaining maximum required flow rates in accordance with Reference 4.47.

3.7 Changes made to the model to reflect system modifications completed since the last revision of the Reference 4.1 model documentation are based on References 4.20, 4.21, 4.33, 4.34, 4.37, 4.39, and 4.40.

4.0 REFERENCES

4.1 Proto-Power Corporallon Calculation 97-170, MP2 R1CCW - Documentation of PROTO-FLOWv Thermal-Hydraulic Model, Revision A, dated December 23, 1997 4:2 NU Calculation 97-ENG-01862-M2, RBCCW System - Heal Loads and Flow Rates, Revision 0 including CCNs 01-04 (Attachment A) 4.3 NNECo ERC 252D3-ER-97-0370, Design Input for RBCCW Flow Balancing Calculation, dated November 18, 1997 4.4 MP2 Surveillance Procedure SP2611A Revision 6 4.5 M2-97038, Revision 0, RBCCW System Letdown Heat Exchanger Valve Failure Position Change 4.6 M2,97039, Revision 0, RBCCW System SDC Heat Exchanger Inlet and Outlet Valve Position Change 4.7 NNECo ERG 25203-ER-97-002, Millstone Unit 2 Reactor Building Closed Cooling Water Design Inputs for Development of RBCCW Thermal Hydraulic Computer Model., dated January 21, 1997 Form: NOOO1FO5 Revision: 00-00 Date: 10-28-2011 Page I at I Form: N0301 F05 Revision: OMO Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 10 of 42 CALC"C 97R1E9 RV 4 1 PA1E5 'F 48 OCHRY RVERFIER ZACHRY NUCLEAR. INC. rai MP2 RBCCW - Design Basis Flow Distribution ment Type: QAPV 4.8 NNECo ERC 25203-ER-98-0034, Design Input for MP2 RBCCW Flow Distribution Calculation, dated February 10, 1998 (Attachment J) 4.9 NUSCo Purchase Order 554145, dated October 31, 1998 4.10 Tullis, J. Paul. 1989. Hydraulics, of Pipelines: Pumps, Valves, Cavitation, Transients. John Wiley & Sons, Inc. 4.11 Prolo-Power Corporation Software Validation and Verification Report (SVVR) for Thermal-Hydraulic Modeling Software - Proto-Flo, SVVR-93948-01 Steady State, Revision N 4;12 Crane Technical Paper No. 410. 1988. Flow of Fluids through Valves, Fittings, and Pipe. Joliet, IL: Crane Co. 4.13 PROTO-FLO TM Software Corrective Action Request 93948-01-07, dated January 10, 1997 4.14 Karassik, Igor J., et al. 1986. Pump Handbook. Second Edition. McGraw-Hill Book Company. 4.15 Van Wylen, Gordon J. and Sonniag, Richard E. 1978; Fundamentals of Classical. Thermodynamics. SI Version 2e, revised prinling. John Wiley & Sons, Inc. 4.16 NUSCo Purchase Order 02025384, Release 153 4.17 EN 21242 Rev 001 -03, "Reactor Building Closed Cooling Water System Facility 2.Flow Balance Verification 4.18 25203-MP2-SFR, Safety Functional Requirements Manual, Rev 09 4.19 Proto-Power Calculation 00-067, Analysis of X-18A and X-18B Thermal Performance Test Results, Revision 00 CCN 0t 4.20 DM2-00-0222-04, Modify RBCCW. Piping Servicing CEDM Cooler X-34A, Including .supplemental documents DM2-01-0222-04 and DM2-02-0222-04 4.21 DM2-00-0223-04, Modify RBCCW Piping Servicing CEDM Coolers X-34B & X-34C, including supplemental document DM2ý01.0223-04 4.22 25203-20150 SH. 881, RBCCW Supply to CEA Drive Coolers, Rev 9 4.23 25203-20150 SH. 361, RBCCW Supply to CEA Drive Coolers, Rev 9 4,24 25203-20150 SH. 882, RBCCW Return From CEA Drive Coolers; Rev 10 4.25 25203-20150 SH: 362, RBCCW Return to CEA Drive Coolers, Rev 13 Form; N03011F05 Revision: 00-00 Date: 10-28-2011 Page I of 1

Serial No 13-419 Docket No. 50-336, Page 11 of 42 C H Y CALCN0 9716 4 16' i 48j NCHRY o0,o.G+,-S ZACHRYNUCLEAR, INC. MP2 RBCCW -Design Basis Flow Distribution ernt Type: QAPD 4.26 25203-20150 SH. 879, RBCCW Supply to CEA Drive Coolers, Rev 9 4.27 25203-20150 SH. 359, RBCCW Supply to CEA Drive Coolers, Rev 10 4.28 25203-20150 SH. 883, RBCCW Return to CEA Drive Coolers, Rev 11 4.29 25203-20150 SH. 363, RBCCW Return From CEA Drive Coolers, Rev 10 4.30 25203-20150 SH. 880, RBCOW Supply to CEA Drive Coolers, Rev 10: 4.31 25203-20150 SH. 884, RBCCW Return From CEA Drive Coolers, Rev 10 4.32 25203-20150 SH. 364. RBCCW Return From CEA Drive Coolers, Rev 17 4.33 DM2-00-0154-01, Install Flow Orifice in "A" Train RBCCW Surge Tank Outlet line 8"- HBD(B)-1 15, including supplemental document DM2-02-0154-01 Supplement 4.34 DM2-00-0155-01, Install Flow Orilice In "B" Train RBCCW Surge Tank Outlet line 8 HBD(B)-1 15, including supplement document DM2-01-0155-01. 4.35 25203-20194, SH. 276, Pump Suction From RBCCW Surge Tank 4.36 25203-20194, SH. 272, Pump Suction From RBCCW Surge Tank 4,37 DM2-00-0802-97, Replace RBCCW Flow Element FE&6732, including supplement document DM2-01-0802-97 4.38 25203-29119, SH. 158, Orllice Plate 21W Schedule 40 Pipe Fig. 21," -300#, Rev 1 4.39 DM2-00-0380-01, Unit 2 Containment Air Recirculation (CAR) Fan Cooling Water Outlet Valve. 2 'RB-29B and 2-RB-290 Replacement, Including supplemental document DM2-0.1-0380-01 4.40 DM2-00-0381-01, Unit 2 Containment Air Recirculation CAR Fan Cooling Water Outlet Valve 2-RB-29A and 2-RB-29C Replacement, including supplemental document DM2-01-0381-01 4.41 25203-29053 SH. 125, PERMASEAT VALVE ASSY 10 IN.-150 LB. WAK, WAFER WITH 6 IN. INTERNALS, Rev A 4.42 25203-29053 SH. 126, PERMASEAT VALVE ASSY 10 IN-150 LB. WAK, WAFER WITH 6 IN. INTERNALS, Rev A 4A43 97-ENG-01962-M2, RCS Cooldown Time With Reduced RBCCW Flow During Normal Shutdown, Revision 0 including CCN 1 and 2 4.44 S-M2CNT-04326S2, LOCAGOTHIC Containment Analysis for Millstone Unit 2, Rev 0 Form: N0301F05 Revision: 00-00 Deto: 10-28-2011 Page I 0$ 1 Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 12 of 42 0' 97-169 REV 4 PAG 17 48 zACHRV NUCLEAR, INC. TIr NA MP2 RBCCW -Design Basis Flow Distribution ?NI Document Type: QAPO 4.45 Proto-Power Corporation Software Validation and Verification Reports (SVVR) for Heat Exchanger Thermal Performance Modeling Software - Proto-HX, SVVR-93948-02, Shell and Tube Rev H, Version 4.10 4.46 ETE-NAF-2012-0117, Transmittal of Millstone 2 LOCA Containment Reanalysis Results using the Dominion GOTHIC Methodology with an Ultimate Heat Sink Temrnperature of 80'F, Rev 2 4.47 ERC 25203-ER-97..0391, Transmittal of Design Inputs for Millstone Unit 2 Peak RBCCW Temperature Analysis, Rev 1 4.48 ETE-NAF-2013-0027, Transmittal of the RBCCW Heat Load Results of the Millstone 2 LOCA and MSLB Containment Reanalysis using the Dominion GOTHIC Methodology with an Ultimate Heat Sinik Temperature of 80*F, Rev 0 4.49 97-057, MP2 CEDM Cooler Performance using Proto-HX, Rev 01 4.50 M2-EV-99-0105, Technical Evaluation for Post-LOCA Heatup of the. Spent, Fuel. Pool, Rev 0 4.51 Zachry Calculation 97-083, Determination of Maximum Room Temperature for ESF Rooms "A", "B", and "C" During a LOCA. Rev 2 4.52 25203-26022 SH. 3. Piping & Instrumentafion Diagram. R.B.C.C.W, System Cntmt Spray Prop and S.I. Prop Seal Coolers, Rev 14 4.53 EOP 2532, Loss Of Coolant Accident, Rev 029-01 5.0 ASSUMPTIONS 5.1 The NPSHA for the most limiting case is recalculated for an assumed RBCCW pump suction temperature of 250*F. As shown In the Attachment D output. reports, the RBCCW pump suction temperature will not exceed 2501F in any operational mode. This is a reasonable assumption based upon the results of Reference 4.46 which determined a maximum R8CCW pump suction temperature of 2220F which exceeds the maximum temperatures in this analysis. 5.2 The value of specific heat for fresh water is assumed to be constant fram 70"F to 250*F having a value of 1.0 BTU/Ibm-*F. 5.3 It is assumed that during Normal Operation a SW flow rate of 7500 gpm is currently achieved or exceeded for the A Train and a SW flow rate of 7650 gpm is currently achieved or exceeded for the B Train. The results of this calculation are dependent on these conditions. Form: N0301 F05 Revision- 00-00 Date: 10-28-2011 Page I of 1

Serial No 13-419 Docket No. 50-336, Page 13 of 42 W.CH R Y , 97-169 gV 4 JPE t8 o' 48 ZACH'RY HANt NUCEAR INC TM-ElIF ZACHF# NUCLEAR ,NC, TIrE MP2 RBCCW - Design Basis Flow Distribution ZN1 Document Typee: QAPD 5.4 All cases were analyzed for 75TF and 80!F SW temperatures with the exception of the Normal Operation alignment. Normal Operation cases were analyzed at 751F and a range of SW temperatures between 78"F and 80*F. For Normal Operation cases at elevated SW temperature In both Proto-Flo and Proto.HX, a tube plugging limit of 3%, 5% or 7% was assumed for the A Train and 5% or 7% for the B Train. Allowable tube plugging was reduced from the design tube plugging limit of 10% (as indicated in Reference 4.19) to demonstrate conditions that would result in an RBCCW temperature of 85'F for the Normal Operations Case. All cases other than Normal Operation with 80OF SW temperature are analyzed with 10% tube plugging. 5,5 Design fouling is assumed for the RBCCW heat exchanger consistent with.Reference 4.19. An overall fouling value of 0.00082 (hr-ft'-'F/BTU) for a clean ticat exchanger is assumed to be the minimum possible resistance based on the Reference 4.19 test results. For several Proto-HX analyses included In this calculation, the fouling Is varied between clean, and design values to determine the conditions at which 85:F RBCCW temperature can be achieved. 6.0 APPROACH 6.1 Model Chanqes The RBCCW model was updated to Proto-Flo Version 4.60 in Revision 4 of this calculation. Additionally, several changes were made to the model to, reflect modifications to the system and to aid in model convergence. These changes are not expected to have a significant impact on the flow balancing of the system, A diflerence report reflecting the changes between the Revision D and Revision 4 models is included in Attachment D. The "default" database in this difference report is not the original Revision D database but a default database (DBD) created Immediately after conversion to 4.60, therefore this report is included for descriptive purposes only. 6.1.1 Node Diameter Agreement Fluid velocities at the nodes in the model are calculated based on nodal diameter. Subsequent to updating the RBCCW system model to version 4.60 of Proto-Flo. all node diameter Inconsistencies as assigned by the connecting pipes were corrected by reordering or adding reducers, diffusers, and abrupt diameter changes. The resulting changes in hydraulic resistance represent a negligible Impact on flow balancing. All changes are shown in the difference report included in Attachment D. 6.1.2 CEDM Supply and Return The elevations of the CEDM coolers were changed and as a result the supply and return piping was modified as documented in Design Change Notice (DCN) References 4.20 and 4.21 and Drawing References 4.22-4.32. The pressure indicators associated with these heat exchangers were also removed. Miscellaneous fixed resistance values assigned to the inlet and outlet pipes were recalculated based on the Reference 4.1 methodology. Form: NO3O1FoS RGvision: 00-00 Date: 10-28-201 1 Page 1 of I Form: N0301 FOS Revision: 00-00 Date: 10-28-2UI I Page I of I

Serial No 13-419 Docket No. 50-336, Page 14 of 42 AhMIl 97-169 1 EV 4 1-19 C' 48 ZACHRY NUCLEAR. INC. TrrLE MP2 RBCCW - Design Basis Flow Distribution zNi Docume,, Tylpe CJ.p. 6.1.3 Surge Tank Outlet Line Flow Orifice Flow Orifices have been installed in the surge lines per the Reference 4.33 and 4.34 DCNs. Pipes 156-159 were updated based on References 4.33 and 4.34 and the Reference 4.35 and 4.36 drawings. 6.1.4 Flow Element FE-6732 The flow element FE-6732 upstream of the Engineered Safety Features (ESF) Room Air Cooling Coil was replaced per the Reference 4.37 DCN and Reference 4.38 drawing; Pipe 234.5 in the Proto-Flo model was updated to reflect this change. 6.1.5 Heat Exchangers The temperature analysis previously performed in a separate calculation was Incorporated into this calculation. The RBCCW shell and tube model was imported from the Reference 4.19 heat exchanger test data analysis. All other heat exchangers are modeled as fixed heat loads according to Reference 4.2 applied using the methodology documented in Reference 4.1. 6.1.6 CAR Outlet Valves 2.RB-29A/B/C/D The Containment Air Recirculation (CAR) Valves 2-RB-29NB/C/D were replaced according to DCN References 4.39 and 4.40, and drawing References 4.41 and 4.42. The replacement valves have a lower associated Cv and theretore do not need to be throttled as severely, preventing excessive degradatlon; The valve information and Cv for each of these valves was updated accordingly. The assumed 6'-l V0' butterfly valve curve was still appropriate in the absence of any other information and was not changed. Due to a lack of information, there is no longer a cavitation Index or recovery coefficient assigned to these valves as they were not checked for choking and cavitation In Proto-Flo. 6.1.7 CAR Outlet Valves 2-RB-29A/B/CiD An error in the RBCCW system model was identified In which the HPSI B. cooler was assigned to the B Train. The HPSI B cooler is a swing component that can be assigned to either train. It is never in operation at the same time as another HPSI component on the same train, however as shown in the Reference 4.52 P&ID, it is normally aligned to the A Train pump. For all cases, valves 2-RB-15C and 2-RB-17D were closed and valves 2-RB-1 5D and 2-RB-1 7C were opened to align the component to the A Train. 6.2 FlowBalancingo The RBCCW system model was balanced to maintain both minimum and maximum .component flows with all valves remaining in the same throttle positions across all alignments. This was achieved byfirst balancing the system to minimum required flows with margin for the limiting most limiting flow condition, degraded pump.performance and Form: N0301 F05 Revision',00-00. Date; 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 15 of 42 rA o'97-169 REV 4 1A 20 OF 48 CALCNO.019RVO ZACHRY 0__31_A_09 UCA.INC Once all MIP2 RBCCW - Design Basiseow Distribution ZN1 Document Type: QAPV 80OF SW. Once all minimum required flow rates were met, the limiting maximum flow condtltonsi nominal pump curves with 75¶F SW, were run to determine with maximum flow rates were exceeded in any case. The components exceeding maximum flow limits were then throttled down and resulting adjustments to other components were performed based on the resulting flow balance. The PROTO-FLOt" model of the RBCCW System documented in Reference 4.1 was aligned to Case I1, and the system throttle valves were adjusted Io achieve the Case Ii required flows with margin. For some components, the required flow margin was based on a +/-2% of full range error on the component flow meters for those components with installed instrumentation; for other components, It was based on +/-2% of the design flow per Reference 4.3; and for others still, it was based on "1:5% of the accident flow rates per Reference 4.8. The required margin for those components with installed flow instrumentation is provided in Table 7 while the required flows with and without margin are provided in Table 8. The system valve alignments for each case have taken into account the valve position changes documented in References 4.5 and 4.6. Table 7 - RBCCW Svstem Reouir Flow Margi Flow Maier Full Range Flow Mietr % component (GPm) Error Marein RBCCW Pump A (P-1 1A) N/A N/A NIA Sldn Ouaench Tank Cir (X-82) Wo 2% 11. Spaml Fuel Pool CIr B iX-2013) 2000 2% 40 Sthutdown Cooing HlX R (X.23B) Note 1 N/A 1,00-ESF Room Coole B (X-36B) Note 1 N/A 3 Sluldown CGolirg HX A (X-22A) Note 1 NIA 1i(1( ESF Roam Cuoley A (X-38A) Note I N/A 3 Spent Fuel Pool CIr A IX.20A) 2000 P% 40 WGC B Aftowooler NWA 2% N/A WGC A Aftea'ooler MIA 2% NMA HPSI Pump I,,Sat CIr (P-41B8 Note I N/A I HPSI Pump C Sei Cit iP.4t4C) Note I NIA I LPSI Pump [ Seal CI, iP-42B) Note 1 N/A I CS Pump B Seal CIr (P-438) Note I NiA I CS Pump A Seal CO, (P-.43A) Nola I N/A 1 HPSl Pump A Seat*Ck OP-41A) Notae N/A I LPSI Pump A Seal Clr iP-42A Note 1 N/A I Prl Ofn TWOuench Tk CIr 1X-24) 400 2% 8 RCP A (P-4(*tAX-7Ae) 200 2% 4A RCP B (P-40C/X-73C} 200 2% 4 RCP C (P-400/X-73C) 200 2% 4 RCEEýJDP-4ooAX-734A) 200 1% 4 C EOM Cooler A tt(-34A) 250 1 20/1 5 Form: N0301F05 Revision: 00-00 Date: 10-28-2011 .page I of 1

Serial No 13-419 Docket No. 50-336, Page 16 of 42 CEDM Cooler B (X-34B) 250 2% 5 CEOM Cooler C (X-34C) 250 2% 5 So pla Cole, (X-19921 NA 2% N/A CAR Unit A (X.35A) No t N/A 100 CAR..Uý.C (x Note I N/A 100 SDgas Yen Cond (H-24) N/A 2% N/A Lotdowj.t.x (X.22) N/A 2% N/A Degas Eli CIr (X-51) NIA 2% NIA Sample coolem rX-i4/05) NWA 2% NIA C A R U nit B.JX U*_5 Note I N/A 100 CAR Unit DOX-35D) Note 1 NIA 100 RV Suppor Cooler ;1A 33.3 2% ___,. 1 RV Supporl Cooler 2A 33.3 2% 1 RV Support Cooler 3A 33.3 2% 1 RV Supporl Cooler 1B 33.3 2% 1 AV Support Cooler 2B 33.3 2% I RV Support Cooler 3B 33.3 2%. 1 R5CCW Pump C (P-0 'C) NiA N/A NWA Nolt 1: Margin baetd on Refrlenn*e 4.8 Table 8 - RBCCW System Required Flows (GPM) and Maroins, Case I Case II Case III CaSe IV 14 Component Design wiMargin Design w/Margln ODesgn wIMargtn Design wlMorgln FIBCCW Pump C (P-t IC) no na nt ne tie no no no. RBCCW Pump A jP-I IA) ne na nt ni. ne no ne nt Bldn Quernch Tank CIr (X-t2) 480 491 FlOW t3) Flow t 3i Fiow(3) Flow t 3) 0 0 Spenl Fuel Pool CBr (.X.20B) 1100 1140 1100 1140 1100 1140 0 0 Shutdown Coeli*tg HX B (X-23B)' 0 0 3500 3000 3500 360r) 0 0 ESF Room Cooler B (X-35B)' 0 .0 59 62 59 62 59 62 Shuldown Cooling H X A (X-23A) 0 0 3500 3600 3500 3600 0 0 ESF Room Cooler A (X-36A)" 0 0 59 62 59 62 59 62 Spent Fuel Pool CrA tX-2OA) 1100 1140 1100 1140 1100 1140 0 0 WGC B Alercooler 5 5 5 5 5 5 5 6 WGC A Aflorcoolor 5 5 5 5 5 b 5 5 IPSI Pump B Seal Cr (P-41B)' Flow' F l Flow") Flwi 3 Flow( 3 Flow(

3)

Flow 3 ' .5 16 ,-PSI Pump C Seal COr (P-41C)4 Flow 3 ) Flow t3) FlowM 3t Flow) F*ow"' Flow() 18 -PSI Pump S Seal Cr (P-42B) 4 Flow"

0) lr-ow"3) 3 4

3 4 3 4 CS Pump B Seal Or (P-438)' Flow (3 Flow(3) Flow(3) Flow t 3) Flow() IFlow t 3) 11 12 CS Pump A Seal CIr (Pý43A) 4 Flow() Flow" rlow(") Fjowl') Flown Flow) 12 Form: NO301F05 Revision: OD-00 Date: 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 17 of 42 Case I Case II Case III Case IV.no Component Oeslgn wlMargln Design w/Margin Design wlMargin Design w/Margin HPSI Pump A Seal Uir P41 A)' Floow" w"I Flow" J Flow"' Flow(y) Flowi 1 ) is 16 LPSI Pump A Seal Cir (P-42A) Flow( 1 ) Flow) 3

4.

3 4 3 4 Pit Cm Th/Ouenci Tk CIr (X-24) 2DO 208 200 218 0

0.

200 208 RCP A (P-40AJX-73A) 110 114 110 114 0 0 110 114 RCP B (P-4001X-738) 110 114 110 114: 0 0 110 114 RCP C (P.40CXA73C) lIt 114 10 114 0 0 110 114 RCP D (P-400/X-73D) ItO 114 ItO 114 0 0 110 .114 CEUM Cooler A (X-34A) 165 170 165 170 0 0 165 M70 CEOM Cooler 8 (X-34B) 165 170 185 170 0 0 165 170 CEOM Cooler C WX-340) 165 170 165 170 0 0 185 170 Sample Cooler (X-192) 5 6 5 5 5 5 0 0 CAR Unit A (X-35A)' 500 800 500 600 500 600 2000 2100 CAR Unit C (X-35C)" 500 500 500 600 500 G00 20DO 2100 Degas Veto Cand (H-24) 5 5 Flow(") Flow") Flow"' Flow") 1 Flaw") Flowo:) Letdown HX (X-22) 1200 1224 110 184 0 0 0 0 Dogas Eft Crr (X-51) 250 255 0 0 0 0 .0. 0 Simple Coolers (X-646b) 311 32 0 0 0 0 31 32 CAR Unll 8 (X-35B) '500 .B00 500 D00 50o 600 2000 2100 CAR Unit 0 (X-35D)0 500 600 500 000 500 600 202.0 2100 RV Support Cooler IA 12 13 1 2 1.3 0 0" 12 13 RV Supporl Cooler ?A 19 13 12 1i 0 0 12 13 RV Suppter Coolor 3A 1;-. 13 12 13 0 0 12 13 RV Spporrl Cooler 1B 12 13 12 13 0 0 12 13 RV Supporl Cooler 2l 12 13 12 13 0 0 12 13 RV Support Cooler 3B 12 13 12 13 0 0 12 13 Noles: (i) The roqulmed flow* for Case IV and all Casa V s.b-c'oes Are oha same with or without Row to the PASS simple coolers. (2) For cases IV and V, lho flows shown in bold Italics are lhe flows which are essential for safe shutdown. (3) I'Flow"e indicates that there is Slow delivered to the componentl lul no flow requlremenl. (4) Margin based on Reference 4.8 Form :N0301F05 Revision: 00.00 D~to: 1O~28-201 1 PagelOl 1 Form: N0301 FOS Revision: 00-00 Date: 10-28-2011 Page.1 of I

Serial No 13-419 Docket No. 50-336, Page 18 of 42 Table 8 - RBCCW System Required Flows (GPM) and Margins (continuedl (l, Caea V WINE m" Case V wlaNE 0 2 Case V w/SFP " Component Design wiM-rgln ODegn wIMargin Design w/Margln A8CCW Pump C (P-IIC) na no na na no no R8CCW Pump A (P. tI A) na "a vM a na no Bldn Quench TanR Clr (X-32) 0 0 0 0 Flowl') Flow"I Spent Fuol Pool Cir 6 (X-20B8 0 0 a 0 1100 1140 ShuldOWn 4oolnlri HX S (X-23B) 4 2000 2100 2W00 2100 1800 1909 ESF Room Cooler B (X-36B)' 59 82 59 62 59 62 Shuldown Coohlvg HX A IX-23A)' 2000 2100 2000 2100 1800 19DO ESF Room Cooler A (X-36A) 4 59 62 59 62 59 60 Spent Fuel Pool Cir A (X-20A) 0 0 0 D 1100 1140 WGC 0 Aflerwoolo 5 5 5 5 9 5 WGC A Afircvolet 5 5 5 5 5 -IPSI Pump B Seal Clr (P-41B)' 15 16 15 16 is .16 HPSI Pump C Seal CIr (P.41 C)' is 16 is 16 is 16 .PSi Pdmp 1 Seal Clr (P42B) 4 3 4 3 4 3 4 CS Pump 8 Seal CIr (P.43B,) 11 12 11 12 i1 12 CS Pump A Seal Cir (P-43A)

1.

12 11 12 1I 12 HPSI Pump A Seal CIr (P-41A)' 15 16 is 16 15 t6 LPSI Pump A SeEl Clr (P-42A)l 3 4 3 4 3 4 Pr! Om Th/Ouench 1 k CIr (X-24) Flow (a) Fiowi 0 ) 0

• 0 u

0 RCP A (P-AA.743A) Flow(3) Flowi3' 0 0 1] 0 RCP 0 (P.408/X. 73) Flow(.) Flowo31 0 0 0 0 RCP C IP-40U.X-73C) FlowN 1 Flawi'l 0 0 o a RCP 0 (P-40D.X773D) Flow( 13 Flow121 0 a 0 0 CEOM Cooler A (X-34A) Flow(3) Fkbw(3 0 0 0: .0 CEDM Cooler B (X-348) Flow(') Flow( 3" 0 0 0 0 CEDM Cooler C (X-34C) Flow 13-Flow") 0 0 0 0 Sample Cooler (X-192) 0 .0 a 0 5* 5 CAR Unill A (X-35A)' 1600 170,0 f6ew 100 1400 1500. CAR Unil C (X-35yC 1500 16 0f 1600 17f0 140 15w Degas Vent Cond (H-24) Flow 13 ) Flow(') 1w 131 Flow.() Flow( 3 1 Flow9) Lotdown HX (X-22) 0 0 0 0 0 0 Form'* N0301 FOS RevisRon: 00-00 Date: 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 19 of 42 Case V WINE '* Case V wloNE "I Casa V w/SFP I" Component Design w/Margin Design wlMergin Design wtMargln Dagas Eli Cir (X-51) 0 0 D 0 0 0 Saroplo Coolers (Xl.64/65) 31 32 31 32 31 32 CAR Uitl B (X.35B)1 16t1 1700 1600 1700 1400 1500 CAR Unit 0 (X-50) 4 1600 1700 Iw0o 1700 1400 1500 RV Support Cooler IA Fow(3) rlowi'l 0 a 0 0 RIV Suppvrl Cooler 2A Flow" Flow3* 0 0 0 0 RV Support Cooler 3A Flow(3) Flow"' t 0 a 0 0 RV Support Cooler IS Flow() Flow" 3 ' a U 0 .0

AV support Croolor 21B FloWt31 FlowI) 0 0

PlV Support Cooler 3B Flow'3 1 Flow"1 0 0 0 0 Notes: (9) 1he required flows for Case IV and all Case V asu-cases are the same Iwith or withoUt how 10 the PASS simpirr cootlrs, (2) For cases IV and V. the flows shown In bold Italics are bhe flows which efr essntilal for safe shutdown. (3) -FloW indicates that horno Is flow donllvord to the component bul no flow requtnierenh. (4) Margin basod on Rrtlrre,*ce 4.8 After balancing the valves to the Case II required flows, the model was realigned to Case IV using the balanced valve positions from Case. I1. The throttle valve for any component which did not receive adequate Ilow was adjusted to provide the required flow. The model was then realigned to Case II using the throttle valve positions determined in Case IV. This process continued until all components received adequate flow for Cases II and IV. The model was then aligned to Case I using the throttle valve positions from'Casesll and IV. Once again, the throttle valve for any component which did not receive adequate flow was adjusted to provide the required flow. After each valve adjustment, Cases II and IV were verified to still achieve the required flow balance. Once Cases II, IV, and I were balanced, the model was aligned to Case V w/SFP using the balanced valve positions. The throttle valve for any component which did not receive adequate flow was adjusted to provide the required flow.. After each valve adjustment, Cases II, IV and I were verified to still achieve the required flow balance. The final flow balancing was conducted in Case V wINE. The throttle valve for any component which did not receive adequate flow was adjusted to provide the required flow. After each valve adjustment, Cases II, IV, I, and V w/SFP were verified to still achieve the required flow balance. Once this final balance was complete, Cases III and V w/o NE were verified to provide adequate flows, then Case 11 w/LIA was run. I Form: N0301F05 Revision: 00-00 Date: 10-28-2011 Page 1 of ~l Form: N030 iF05 Revision: 00-00 Date: W-28-201111 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 20 of 42 CALO NO. 97-1 69 REV 4 1 F'A2G 5 F48 lI Y OnlaINATOR -t~rF T U MP2 RBCCW - Design Basis Flow Distribution e* QAPD When the system was balanced to achieve all minimum required flow rates, all cases were aligned to nominal pump performance and lower SW temperatures to create bounding maximum flow conditions and compared to the maximum flow limits documented in Reference.4.47 shown in Table 9 below. Case V w/o NE was found to be the limiting maximum flow case for the CAR coolers for the A Train. These components were throttle to just below their maximum flow limits and all cases were rerun to determine the effect on the overall flow balance. Where it was not possible to maintain both the maximum and minimum flow limits with the same throttle position, the margin for the minimum flow rate was reduced and the maximum flow rate was maintained. Table 9: Maximum Flow Limits Component Pre-SRAS-RBCCW Posi-SRASIRMCW Post-SRASROCCW Flow (Uim) Flow (gpm) Flow with SFP (gpm) CAR fX 2200. .2000 lw00 SDcHX 0 209( 2600 SFP HX 0 0 1200 ROCCW HX Sso0 7100 7800 PROTO-FLOT M Version 4.60 is verified and validated in accordance with: Proto-Power Corporation's Software Quality Assurance Program per Reference 4.11. 6.3 Shutdown Cooling HX Reguired Throttle Valve Positions The Shutdown Cooling HX will be throttled using a combination of three valves. For Cases It, and III, flow Is throttled using valves 2-RB-13.1A and 2-RB-13.18 and valves 2-RB-13A, 2-RB-14A, 2-RB-13B and 2-RB-14B are 100% open. For all LOCA Recirc cases (Case V) valves 2-RB-13.1A and 2-RB-13.1B are held in the same position and flow is throttled using valves 2-RB-13A, 2-RB-14A, 2-RB-13B and 2-RB-14B. Valves 2-RB-13A and 14A are throttled equally, and valves 2-RB-13B and 14B are throttled equally. In order to determine the positions of these valves, two approaches Were taken, For valves 2-RB-13.1NAB, the required positions were taken directly from the balancing runs. For 2-RB-13A/B and 2-RB-14NB, only valves 2-RB-13A/B were used during the initial balancing runs with 2-RB-14A/B left fully open. Since these valves are identical, the resulting position for 2-RB-13AIB was used to determine the position of both valves In the train assuming they are throttled equally. Thus, knowing the position of 2-RB-1 3A/B, and using the % Open vs. % Cv curve (closure curve) in the PROTO-FLO II model, the. required C, tar the valve was determined by linear interpolation from the curve. For small changes to the position of these valves they can be throttled manually in unison for each balancing run. The methodology for initially calculating the valve positions is shown below. Form: NO3OIFOS Revision: 00-00 Dale: 10-28-2011 Page 1 of 1 Form: N0301 F05 Revision: 00-00 Dale: 10-28-2011 Page l of I

Serial No 13-419 Docket No. 50-336, Page 21 of 42 C.AL' "o 97169 j '4tv 4 2rAct6 Or 48 XACHRY oRINOPI o-RIFIES zACHY NUCLEAINC. PILE MP2 RBCCW - Design Basis Flow Distribution ZN) Documnen Type; QAPD The pressure drop was divided equally between the two valves. Per Reference 4.12, knowing the total Cv, the pressure drop can be expressed as: 62.4 CJ Where: AP = Pressure drop (psi) Q = Flow rate (gprn) Cv = Valve flow coefficient Now, knowing that the final pressure drop for each valve will be one-half that found from the flow balance, the required C, for each valve may be lound: P QWWV.iwot!lv

• ,**-*-=0.5 = 62.4 C Recognizing that the density and flow rate will be unchanged yields:

0.5 Knowing the C4 for each valve, the required valve position may be found by linear Interpolation from the closure curve. 6.4 Assessment o. Potential for Cavitation In Shutdown Coolino HX Throttle Valves The cavitation assessment performed in this calculation used valve Inlet temperatures from the Proto-Flo model output. The RBCcW flow rate was obtained from the Flow Summary Reports In Attachment D for the nominal pump cases for 751F SW and elevated UHS temperature conditions. Using the valve inlet temperature, the vapor pressure and density of water at this temperature are found. Now, the valve inlet and outlet (or upstream and downstream) pressures are obtained from the Combined Output or Node Summary Reports in Attachment D. With this information.the potential for cavitation In the Shutdown Cooling HX throttle valves can be assessed. Form: N030tF05 Revision: 00-00 Date: ID-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 22 of 42 VAICN1 97-169 REV P4O1 27 or "48 ZACHRiYOC ACWR ZACHRY NUCLEAR, INC. IRE. MP2 RBCCW - Design Basis Row Distribution ZN) Document Type: aAPD Cavitation analysis for each of the valves was performed using the method outlined in Reference 4.10. This method evaluates the level of cavitation by comparing the asytm with the t calculated for the various levels of cavitation. If the is less than ithe calculated for a certain level of cavitation, that level of cavitation is said to =exist at the valve. Tihe following equations were used: -P V =(2gAH 4. Vj" LT =0.57+1lC~ -11CI +13Cý a 0.40 +5,8C, - 6.2C,' +7.t4C, =0.07 + 2.3C, - 1.46C; +.7, =~~(P P PSE2 . 82 q,= ISE, c". SSE Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 Reference 4.10 c,1 SSEa,. PSEj Form; N0301 F05 Revislon:.00-00 Date: 10-28-2011 Page loll Form. N0301 F05 Revislow.00-00 Date: 10-28-2011 Page :t of I

Serial No 13-419 Docket No. 50-336, Page 23 of 42 97189 REV 4 1 AE 28 o 48 1NHRY oRoN.Oo VERIERE8 PUCLEARINC. MP2 RBCCW - Design Basis Flow Distribution i:.AP0 Where. Pý. = downstream pressure corrected for velocity head (psia) P. = vapor pressure (psia) V = average velocity (ft/sec) Pug = upstream pressure corrected for velocity head (psia) AP = pressure drop across valve (psld) d = pipe diameter (inches) H - Head Loss (ft) SSE = Size Scale Factor PSEý = Pressure Scale Factor for o&, and oa PSE2 = Pressure Scale Factor for am amo = Uncorrected Cavitation Level In addition to the equations from Reference 4.10 above, the following corrections to tihe PROTO-FLO'T pressure results must :be taken into account in accordance with Reference 4.13 to correct the results for Velocity head: H" = Pda-Pmdadow*ml - Hv Where: H, - Velocity Head (psi) The levels of cavitation used in this method are defined in Reference 4.1(0 as follows: Incipient cavitation a,: Usually occurs intermittently.over.a restricted area There is no objectionable noise and there Is no damage except when only a small part of the flow Is involved, such as a small step or roughness element where serious erosion may occur. Incipient is a conservative design limit and Is suggested for use only when no noise or other disturbances can be tolerated. Critical cavitation ,,: Associated with continuous light cavitation. Noise and vibration are at acceptable levels and only minor damage is expected after long periods of operation (months to years). This levelis often adopted as the design criterion for a component& Per Reference 4.10, the cavitation at critical wouldnot be objectionable and would hot decrease valve life, Incipient damage Old: The onset of surface pitting after short periods of operation; the noise levels may be objectionable. Farni: NO3O1FOS Revision: 00.00 Oste: 10-28.2011 Page 1 of 1 Form: N0301FG5 Revisiow 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 24 of 42 'c No' 97-169 I 4 1 PAGC 29 OF 48 ZACHYU"EARINC. MP2 RBCCW - Design Basis Flow Distribution ZNI DaCumenif Type* QAPO Choking cavitation ayh: Associated wilh the lowering of the outlet pressure on the device to vapor pressure, after which the flow through the device is unaffected by downstream pressure. Steady state pressure, flow and pressure loss relationships no longer apply. An Excel spreadsheet was developed to evaluate these equations for the Shutdown Cooling HX throttle valves and Is Included as Attachment E. All cases were also analyzed for elevated UHS conditions with degraded pump curves. A detailed description of the spreadsheet is included in Attachment E. Note that the methodology contained In Attachment E includes an alternate calculation of Inlet temperatures from heat loads and flow rates although the temperatures were obtained from the output reports. 6.5 Evaluation of Available Net Positive Suction Head to the RBCCW Pumps The NPSHA results reported In the tables in Section 7.4. are taken directly from Proto-Flo reports. The most limiting of these cases was then recalculated based on a suction temperature of 250*1 in accordance with Assumption 5.1 using the methodology below. NPSHA Is defined in Reference 4.14 as: NPSHA = + Y 2, Where: P, = Atmospheric pressure (psia) Pv = Vapor pressure of liquid being pumped at pump suction (psia) P= = Pressure at pump suction (psig) Zpý Elevation ot pump suction gauge relative to NPSH datum (f1) V = Fluid velocity at pump suction (ft/e) g Gravitational acceleration (fVs') V = Specific weight of fluid at pumping temperature Since velocity head is included in the definition of NPSHA, and the pressures reported by PROTO-FLOTM include velocity head per Reference 4.13 and are at the pump Inlet datum, the NPSHA equation reduces to: P -P P NPSHA= "L +. r Y Form: N0301 F05 Revisionm 00-00 Date:. 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 25 of 42 CLACM] 97-169 RE 4 ,Ar3 OF 4 IC H K Y ORIGNACI V-I*IAR ZACHnRYNUCLEAR, INC. , r. MP2 RBCCW - Design Basis Flow Distribution 7tent Type. QAPD 6,6 Evaluation of Brake Horseoower Reguired by the RBCCW Pumps Brake horsepower of the RBCCW pumps for each case will be evaluated using! the RBCCW pump total dynamic head, efficiency, and brake horsepower curves per Reference 4.7. The curves are provided as Attachment F. 6.7 Evaluation of Flow Distribution with PASS In Service Operation of the PASS sample coolers requires that RBCCW cooling flow be supplied via Isolation valve 2-RB-210. Valve 2-RB-210 is a branch line Isolation valve that, when opened, will supply RBCCW to several components.. The result Is a flow redistribution that decreases flow to safety related components In the "B" header. The flow re-distribution is quantified by re-running the accident cases with Isolation valve 2-RB-210 100% open. No other valve. position changes were made to restore flow to safety-related components. A required flow tate for the PASS sample coolers for the accident cases has not been specified in Reference 4.2 since the PASS sample coolers are normally isolated from the "B" header in an accident scenario. For purposes of evaluating the flows delivered to the PASS sample coolers when 2-RB-21 0 Is opened, the normal operation (Case I) required flow rate is used. Accordingly, the Case IV and.Case V required flow rates for the PASS sample coolers are taken as 31 gpm (design) and 32 gpm (design with margin). 6.8 Component Outlet Temperature and RBccW Suoolv Temperature Analysis In the current version of the model, fixed heat load Information was incorporated, consistent with Reference 4.2 with the exceptions described below. This was done using the same heat load asslgnment.methodology documented in Reference 4.1 model calculation, with the exceptions described below. The heat load for the Control Element Drive Assemblies for Case IV is considered to be approximately 10.3 MBtu/hr per assembly based upon Reference 4.49, Calculation 97-057. Specific heat loads for each case were determined based on the RBCCW supply temperature ior each case and the maximum heat rejection to RBCCW versus tube inlet temperature documented In Attachment D of Reference 4.49. Additionally, the Reference 4.53 EOP 2532 Isolates the non-essential equipment from the RBCCW system during the recirculation phase of a LOCA (Case V)..Although indicated differently than Rel. 4.2, the CEDM heat load of approximately 10.3 mBfu/hr Is assumed to conservatively contribute to the heat load of the RBCCW during the LOCA Injection phase (Case IV) and the initial portion of the LOCA recirculation phase When non-essential equipment is aligned (Case V wINE.) This methodology also applies to the RPV Support Cooling Coils and the RCP Thermal Barrier and Lube Oil Coolers. Fmrm; N030 iFOS Revision: 00-00 Date: 10-28-2011 Page 1 01.1

Serial No 13-419 Docket No. 50-336, Page 26 of 42 ZACH Y A " 97-169 R=v 4 -r_ 31 OF 40 ZACHRY NUCLEAIR M. flity. MP2 RBCCW - Design Basis Flow Distribution ZN1 Documeoi Type: QAP0 The Spent Fuel Pool heat load for LOCA Recirc was updated to 10.16 mBTU/hr In accordance with Reference 4.50. The ESF (X-36AB) heat load was updated to 0.393 mBTU/hr based on the maximum heat rejection to RBCCW documented in Reference 4,51. For all LOCA cases, "Clean" heat exchanger heat loads were assigned for maximum flow cases, or nominal pump performance, and "fouled' heat loads were assigned for the degraded pump cases for the SDC and CAR heat exchangers; In accordance with Reference 4.48. These heat loads correspond to an 80"F SW temperature and therefore bound the 75'F SW condition, and so were assigned for both sets of cases. The Proto-HX shell and tube model of the RBCCW heat exchanger documented in Reference 4.19 was Imported directly into the Proto-Flo RBCCW system model. All cases were run at a SW temperature of 751F and 80¶1, except for the Normal Operation case which was run at 75"F SW and then at elevated SW temperature conditions which would maintain the RBCCW supply temperature limit of 850F, as determined by Proto-HX. A series of evaluation cases using the total heat load and flow rates from Case 1 were performed using the Reference 4.19 Proto-HX model to determine the conditions at 'which the 851F RBCCW temperature was achievable. For each heat exchanger, first nominal and degraded pump flows were determined from the Proto-Flo output reports Included In Attachment D. For each RBCCW heat exchanger, for both nominal and degraded R8CCW flow rates, first Ihe model was run with 80OF SW, 10% tube plugging, and design fouling. If the 85IF RBCCW could not be maintained, tube plugging was then reduced. If a reduction to the minimum tube plugging could not achieve 851F RBCCW supply temperature, next the fouling was reduced. If the supply temperature still could not be achieved, the SW temperature was reduced to 79F and the process was repeated. The results of these iterations are shown in Table 4. The required flow rate to achieve 851F RB3CCW supply temperature with a 801F SW temperature was determined for each heat exchanger and RBCCW pump curve, first for design fouling and 10% tube plugging, and then for 5% tube plugging and an overall fouling corresponding to approximately 3 months cleaning frequency, The results are shown in Table 5. All Proto-HX output reports and component outlet temperatures are included in Attachment L. 7.0 ANALYSIS 7:1 FLOW BALANCING The flow balancing was performed entirely using the PROTO-FLOTM model of the RBCCW System documented in Reference 4.1, The various model reports are included In Attachment D In both Crystal Reports and Excel file formats; The files are named per the following convention: ABBCDEEEE.ZZZ Where: Foim: NO3OIFOS l~evlsion: 00-00 Date:.10-28-201 .l'age1 of I Farm,. N030FG5 Revision: 00-00 Date: 10-28-2011 .page-t ot I

Serial No 13-419 Docket No. 50-336, Page 27 of 42 'A' 'B6' 'C' 'D' 'EEEE' 'zzz' - Pump Curve ('n'=Nominal, 'd'=Degraded) = Case (As defined in Table 10) = Subcase (As defined In Table 10) = SW Temp ('1'=750 F, '2'=Elevated UHS: except for Normal Case at 79"F) = Report.Type (As defined in Table 11) = File Format ('rpt = Crystal Reports,. 'xis' = Excel 5.0) 801F for all cases For example, a. file named 'nO521flow.rpt' would contain the PROTO-FLOTM Flow Summary Report for Case V w/o NE run with the nominal RBCCW pump curve and 75*F SW in Crystal Reports formal. A Crystal Report viewing application is provided in Attachment D. The model used in Revision B to this analysis had been compared to the model documented In Reference 4.1. Additionally a difference report has been included comparing the Rev 4 and Rev B models for information only. This report is not a direct difference report to the model used In previous revisions but to the Default Database (DBD) created directly after conversion to Proto-Flo Version 4.60. Both sets of differences reports are provided on Attachment D. As noted In Section 2.0, a table of the balanced flows for each case is provided in Attachment C. Table 10.- Case and Subcase Number Descriptions Case Subcase Case ID Description II I I Normal Operation 2 1_1 1I 3.5 hours alter Shutdown Operation____________ .........2..................

• 5 o..

u...l o..... ..h_._ o-............... 12 /. .LIA 3.5 hours alter Shutdown Operation with Loss of Instrument Alt 3 1 Ill 27.5 hours after Shutdown Operation 4 1 IV LOCA Iniection Operation 4 2 IV LOCA Injection Operation (with flow to PASS sample coolers) (PASS) 5 1 V w/NE LOCA Recirculation Operation with Non-Essentlal Components on-line 5 2 V w/oNE LOCA Recirculation Operation with Non-Essential Components secured 5 3 V wISFP LOCA Rocirculation Operation with spent luel pool cooling restored 5 4 V wINE LOCA Recirculation Operation with Non-Essential Components _____{ASS) on-line (with flow to PASS sample coolers) 5 5 V wtoNE LOCA Recirculation Operation with Non-Essential Components (PASS) - secured (wih flow to PASS sample coolers). 5 6 V w/SFP LOCA Recirculatlon Operation with spent fuel pool cooling (PASS) restored (with flow to PASS sample coolers) Form: N0301 FOS Revision: 00-00 Date: 10-28-2011 Page I of 1

Serial No 13-419 Docket No. 50-336, Page 28 of 42 Table II - PROTO-FLOT Report Title Abbreviations Abbreviation [ Report Title comb Combined Output RepRort oval Control Valve Line-Up Report diTf Default Dilffaronces Report (low Flow Summary R9epodrt.......... heat Heat Exchanger Data Report q *........... .......e t E xa R e or Hydraulic ResistanceRep *t. mval Manual Valve Line-UpReport node Node SummarRport pump.__ _ Pump Status Report summ Calculation Summary Reporl 7.2 SHUTDOWN COOLING HX REoUIRED THROTTLE VALVE POSmONS The positions for 2-RB-13.1A and B may be determined directly from the PROTOýFLOTM output in Attachment D. The required positions for these valves is 16.75% open for 2-RB-13.1A and 16.68% open for 2-RB-13.1B. The methodology described In Section 6.3 was originally used to determine the positions of 2-RB-1 3A/B and 2-RB-14A/B. Subsequent small adjustments tothe valve positions were done manually on both valves in each pair at once. The methodology used to Initially throttle these valves is shown below and should be used for any major adjustments to the valves moving forward. During the flow balancing, the positions for 2-RB-13A/B were found to be 9.42% open and 8.64%/6 open, respectively. Thus for the valves., linearly interpolating from the Valve closure curve in PROTO-FL0Th (provided In Attachment G) and knowing the valve's fully open C, Is 17617 per the Valve Data Report (also provided in Attachmnenl G): 2-RB-13A: %C, = 2.40-0 (9.4236 -0)+0 =.2.0357%

• 11.11l-0 C, = 020357(17617) = 358.63 2-RB-1 3B:

I Form: NO3O1FOS Revision: 00-00 Dale: 10-28-2011 Page 1 of 1 Form.'N0301F0S Revision: 00-00 Dale: 10-28-2011

• .Page t of 1

Serial No 13-419 Docket No. 50-336, Page 29 of 42 I"_~o 97ý119 REV 4 1PAGE 34 OF 48 WCHRY O-Nf1o: V+NHEHE8 ZACHRY NUCLEAR.

INC, IITLE MP2 RBCCW - Design Basis Flow Distribution 7ment Type: OAPD 0 (8.6427 -o)+o = 1.8670%

11.11.-0 C,, =.018670(17617) =328.91 Now, using the equation developed in Section 6.4 to determine the C. for throttling both I 2-RB-13A/B and 2-RB-14ANB equally and linear interpolating the valve closure curve to lind the valve positions: 2-RB-13A and 2-RB-I4A: ,F2 =. = 35.63)4 =: 507.1X %C= 'O-"009=2.8799%

• (.17617)

%Open = 22.22-11.11 (2.9719-2.40)+ 11.1 I 13.454% 13.45% 4.67-2.40 2-RB-13B and 2-RB-1 4B: = (328.911 42 465.15 ( 465.15 100% = 2.6403% i, 17617 ) %Open-22.22-li.1(2.6403 - 2.40)+ 11.11 = 12.286% 12.29% 4.67 - 2.40

7.3 ASSESSMENT

OF POTENTIAL FOR CAVITATION IN SHUTDOWN COOLING HX THROTTLE VALVES In order to demonstrate the cavitation assessment technique. the Case II assessment for 2-RB-13.1A for the 75¶F SW conditions Will be performed in detail here. Attachment E I contains the complete Excel spreadsheet used to perform these evaluations and details of the application of the various equations in the spreadsheet. The methodology documented in Attachment E Includes the calculation of valve inlet temperatures for use when necessary, however the current Attachment E assessment uses temperatures determined from Proto-Flo heat load analysis, Form: N0301F05 Revision: 00-00 Dale~ 10-28-2011 Page 1 oIl Form: N0301 F05 Revision' 00-00 Date' 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 30 of 42 CALCf NO. IFV 36 Or Z AIL.HI'IY ORIINA ) VER FR ZACHRY

NUCLEAR, INC.

• 'T.E MP2 RBCCW - Design Basis Flow Distribution ZNI Documfent "ryes, QAPO From Attachment D results;

= 3708.53 gprn PWM~MM = 101.49 psia P.,-I,. = 66&80 psia V = 5.091 IVs d = 17.25 in Per Attachment D, the RBCCW supply temperature is;. = 103.720F Per Reference 4.15, density at 850F Is: p = 62.17 Ib/ftW Per Attachment D, the valve inlet temperature Is: = 142.13"" Now, per Reference 4.15, vapor pressure and densily at 142.131F are: P,. = 3.1322 psia P142.F = 61.34 Ib/ft3 Correcting the valve upstream and downstream pressures for velocity head; 17.25' ( H )61.34(370S.53)2 6663psa JP*, =66.80-(i.8[0 ).t "_""=66pi AP 101.32-66.63 = 34.69psid All = 80.34ft 62.17 Now, knowing these.inputs, the cavitation indices may be calculated: 66.63-3.1338 S, 3Ir*.6 83 34.6Q Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I of f'

Serial No 13-419 Docket No. 50-336, Page 31 of 42 CAt'N"' 97-169 PAGE 0 r. 48 ZACIHIRY oRIoNAMR I

36o, ZACIRY NUCLEAR (NC.

TME MP2 RBCCW - Design Basis Flow Distribution ZNI Docurnent.Type: CAMO 5.091 =.0706 (2 x32.2 X 80.34 + 5.0912 ' S101.31-3.1322)"'^

PSE,

.I! ~~ I =1.05 k 82' ) S ('101.31-3.1338y"' 8*2 ) ý; =1.09 a,., = 1.2 + 16x0.0706 - 23(0.0706)' +31(0.0706)' 2.23 q, =1.09x2.23x1.05 =2.55 or" = 0.57 +1 I.xO.070- 1](0.0706)2 + 13(0.0706) = 1.30 qc, = !.09xL.30x1.05.=L148 r,0 = 0.40+ 5.8x 0.070- 6.2((1.0706)' +7.8(0.0706)Y = 0.78 ',* = 1.03x0.78 =0.81 a,,, =0.07 +2.3x0.0706-1.46(0.0706)' +4.7(0.0706)'--0.23 Comparing the system cavitation Index to the threshold cavitation indices, it Is clear that this valve will operate within the incipient cavitation regime. The cavitation information for all of the Shutdown Cooling HX throttle valves Is contained in Attachment E. 7.4 EVALUATION OF AVAILABLE NET POSITIVE SUCTION HEAD) TO THE RBCCW PuMPS Using the equations from Section 6.5, pump suction pressures from the Node Summary I Reports in Attachment D, and the following constants; P. 14.7 psla P = 29.82 psia @ 250*F per Reference 4.15 y = 58.8201 Ibtft @ 2500F per Reference 4.15 Form: N0301F05 RevisIon: 00-00 Date: 10.28-2011 Page I at 1 Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 32 of 42 S.%ICND. 97-169 PEV 4 P1A-37 C' 48 ZACHR NUCLEAR, INC. MP2 RBCCW - Design Basis Flow Distribution ZNt Document Typo: QAPO NPSHA was recalculated for the most limiting case for a conservative temperature of 250"F in accordance with Assumption 5.1, The most limiting condition for NPSH margin was Pump A in Case IV (PASS) with nominal pump curve and 80IF service water. (14.7-29.M2 57.35 -14.7 144 = 67.39JI NPlIA I 58.8201 58.8201) The NPSHR was determined by PROTO-FLO'u and was obtained from the Pump Status Reports in Attachment D. Where noted, the calculated pump flow rates were greater than the limit of the NPSHR curve. In these cases, the estimated NPSHR was taken as the linear extrapolation of this curve based on the last two points. The NPSHA, NPSHR and resulting margin for both pumps in all cases is provided in Table 12 and Table 14 for the degraded pump runs and in Table 13 and Table 15 for the nominal pump runs. Form: N0301F05 Revision: 00-00 Date: 10-28-2011 Page I of !

Serial No 13-419 Docket No. 50-336, Page 33 of 42 Table 12 - Net Positive Suction Head Results - Degraded Pump - 750F SW Pump j Case T.(VF) Ps (psi) NPSHA(tt) NPSHR (ft) Margin (ft) A r 86.52 57.26 131.14 20.77 110.37 1 1 90.51 57.25 131.01 25.22 105.79 A 11 127.04 56.93 128.13. 26.06 102.07 C 1 11 127.32 57.19 128.71 25.86 102.85 A 11 w/LIA 121.95 56.83 128.35 27.9 100.45 C' w/LIA 120.74 57.17 129.24 26.48 102.76 A

1.

104.89 57.12 130.17 22.89 107.28 C IT[ 104.59 57.26 130.49 24.78 105.71 IV 202.83 57.47 108.24 17.48 90.76 C IV 193.89 57.86 113.74 15.27 98.47 A TV (PASS) 202.84 57.47 108.23 17.48 90.75 C 1V (PASS) 192.19 57.78 114.34 15.98 98.36 A V w/NE 167.1 56.96 121.28 25.02 96.26 C V w/NE I154.63 57.41 125.26 20.76 1t04.5 A V wo NE 153.77 57.21 124.96 20,77 104.19 C V wlo NE 154.47 57.47 125.44 19.89 105.55 A V w/SFP 150.41 56.95 124.98 25.48 99.5 C V wiSFP 150,24 57.17 125.54 25.85 99.69 A V w/NE (PASS) 167.11 j56.96 121.28 25.02 96.26 C V w/NE (PASS) 154.77 57.35 125.09 21.99 103.1 A V w/o NE (PASS) 153.76 57.21 124.96 20.77 104.19 C V w/o NE (PASS) 154.55 57.41 125.27 20.81 104.46 A V w/SFP (PASS) 150.4 56.95 124.98 25.48 99.5 C V w/SFP (PASS) 150.83 57.15 125.37 26.38 98.99 Form: N0301 F05 Revision: 00-00 Date' 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 34 of 42 Table 13 - Net Positive Suction Head Results - Nominal Pump - 75:F SW Pump Case T.(") P,(psia) NPSHA(ft) NPSHR (It) Margin (ft) A I 86A43 57.16 130.92 22.02 108.9 C 1 90.43 57A19 130.89 25.95* 104,94 A 11 126.71 56.85 127.99 27.01' 100.98 C 11 127 57.13 128.6 2.76* 101.84 A II w/LIA 121.5 56.7 128.09 30.451 97.64 C Ii w/LIA 120,42 57.09 129.09 27.557 101.54 A It 104.71 57.05 130 24.02 105.98 C 111 104.48 57.21 130.38 25.52 104.86 A IV 217.26 57.36 98.79 18.04 80.75 C IV 210.06 57.76 104.63 15.73 88.9 A IV (PASS) 217.25 57.36 98.79 18.04 80.75 C IV (PASS) 208.1 57.68 105.66 16.58 89,08 A V wiNE 184.67 56.89 115;37 25.72 89.65 C. V wINE 172.88 57.3 120.46 22.02 98.44 A V wi/ NE 172.04 57.1 120.22 22.05 98.17 C V win NE 172.78 57.36 120.62 20.86 99.76 A V w/SFP 160.55 56.87 122.72 26.28 96A44 C V w/SFP 160.33 57.1 123.3 26.76T 96.54 A V wINE (PASS) 18467 56.89 115.37 25.72 89.65 .C V w/NE (PASS) 173.01 .57.26 120.29 23.23 97.06 A V wi/ NE 172.02 57.1 120.22 22.05 98.17 (PASS) C V w/o NE 172.8 57.31 120.48 22.1 98.38 (PASS) A V w/SFP 160.53 56.87 122,72 26.28 96.44 ,(PASS) C V w/SFP 160,77 57.06 123,11 27,44 95.67 (PASS) Note 1: Linear extrapolation beyond end of NPSHR curve. Foini: N0301F05 Rl3vision: 00-00 Date: 10.28.2011 Page 1 all Forom: N0301 F05 Revision: 00-00 DaW 10-28-2011 Page I of 1

Serial No 13-419 Docket No. 50-336, Page 35 of 42 Table 14 - Net Positive Suction Head Results - Degraded Pump - 80 F SW Pump [ Case T.(T) P. (psla) NPSHA (ft) NPSHR (ft) Margin (ft) A m 89,4 57.26 131.06 20.77 110.29 C

10) 88.86 57.25 131.05 25.22 105.83 A

11 131.56 56.93 127.65 26.06 101.59 C 1I 131.84 57.19 128.23 25.86 102.37 A I. w/L]A 126.5 56.82 127.93 27.9 100.03 C II w/LTA 125.29 57.16 128.84 26.48 102.36 A 111 109.5 57.12 129.92 22.9 107.02 C U1] 109.21 57.26 130.24 24.78 105.46 A IV 207.1 57.46 105.75 17.48 88.27 C IV 198.17 57.85 111.63 15.27 96.36 A IV (PASS) 207.1 57.46 105.75 17.48 88.27 C IV (PASS) 196.47 57.78 112.3 15M98 96.32 A V wINE 171.46 56.96 120.04 25.02 95.02 C V wINE 158.99 57.41 124.33 20.76 103.57 A V w/o NF 158.15 57.21 124.04 20.77 103.27 C V w/o NE 158.85 57.47 124.51 19.9 104.61 A V w/SFP 154.8 56.94 124.13 25.48 98.65 C V w/SFP 154.64 57.17 124.69 25.85 98.84 A V wINE (PASS) 171.46 56.96 120.04 25.02 95.02 C V wINE (PASS) 159.15 57.34 124.15 21.99 102.16 A V w/o.NE (PASS) 158.15 57.21 124.04 20.77 103.27 C V w/o NE (PASS) 158.94 57.4 124.33 20.81 103.52 A V w/SFP (PASS) 154.81 56.94 124.13 25.48 98.65. C V w/SFP (PASS) 155.23 57.14 124.5 26.39 .98.11 Note 1: Service Water Is 78IF or 79IF for the Normal Operation case. Form: N0301F05 Revision: 00-00 Data: 10.28~20l 1 Page I oil Forrm-N0301F05 Revision: 00-00 Date: 10-28-2011 Page of I

Serial No 13-419 Docket No. 50-336, Page 36 of 42 GALNO. 97-169 FIýV 4 1PAP 41 48 ZACHRY OIaNATOR vmie ZACHRV NUCLEAR, MC. 117R[ MP2 RBCCW - Design Basis Flow Distribution LNI Document Iype.W GA'0 Table 15 - Net Positive Suction Head Results - Nominal Pump -80'F SW Pu I Case T&F ) j P.(psla) NPSHA(ft) NPSHR (f) Margin (ift) A P.)J 89,33 57.16 130.84 22.02 108.82 C tIM 88.82 57.19 130.93 25.95 104.98 A It 131,24 56.85 127.51 27.017 100,5 C [1 131.53 57.12 128.11 26.76& 101.35* A II w/LIA 126.07 56.7 127.68 30.46 97.22 C 11w/LIA 124.98 57.08 128.69 27.56' 101.13 A 1.11 109.33 57.04 129.75 24.02 105.73 C I1 109.11 57.21 130.13 25;53 104.6 A IV 221.49 57.35 95.6 18.04 77.56 C IV 214.3 57.75 101.81 15,73 86.08 A IV WASS) 221.48 57.35 95.6 18.04 77.56 C IV (PASS) 212.34 57.67 102.94 16.58

86.36 A

V w/NE 1 188.95 58.88 113.62 25.72 87.9 C V wINE 177.18 57.3 119.07 22.03 97.04 A V w/o NE 176.35 57.1 118.85 22,05 96.8 C V w/o NE 177.08 57.36 119.24

20.87 98.37 A

V w/SFP 164.9 56.87 121.65 26.29 95.36 C V w/SFP 164.66 57.09 122.24 26.76T 95.48 A V w/NE (PASS) 188.97 56.88 113.61 25.72 87.89 C V w/NE (PASS) 177.32 57.25 118.9 23.23 95.67 A V w/o:NE (PASS) 176.34 57.1 118.85 22.05 96.8 C V w/o NE (PASS) 177.13 57.3 119.08 22.1 96.98 A V w/SF*P (PASS) 164K9 56.87 121.65 26.29 95.36 C V w/SFP (PASS) 165.14 57,08 122.03 27.44 94.59 Note 1: Linear extrapolation beyond end of NPSHR curve. Note 2: Service Waler is 78F or 791F for the Normal Operation case. Formi:N030F05 Revision: 00-00 Date: 10-28-2011 Page 1 of I

Serial No 13-419 Docket No. 50-336, Page 37 of 42 7.5 EVALUATION OF BRAKE HORSEPOWER.REQUIRED BY THE RBCCW PUMPS Using the flow results from the degraded and nominal pump runs, the brake: horsepower for each pump in each case was determined using the certifiedshop curves provided In Attachment F. The pump flows obtained from the Pump Status Reports in Attachment D and the resulting brake horsepower requirements are shown in Table 16, Table 17, and Table. 18. Where noted, the calculated pump flow rates were greater than the limit of the brake horsepower curve provided in Attachment F. In these cases, the estimated brake horsepower was taken as the linear extrapolation of this curve. I T;,*L*.I^ -t* OD/"*pIA/ I*l=.m* O**t.* '7r~ CC Ot IdbleIt~ - [lU'%AaLVV rUIIIIJ BrakeU HrIUI~UUwe, ReutiImeits - LJvoyedU ruiisip I JI V Pump Case Approximate Flow (gpm) Required Brake.Horsepower A 1 6453.08 315 C I 7031,77 315 A 11 7150,87 315 C II 7122.7 315 A 11 w/LIA 7414.18. 310 C 11 w/LIA 7211.1 315 A Iil 6736.84 315 C III 6972.54 315 A IV 5658,53 305 C IV 4884.9 290 A IV (PASS) 5658.23 305 C IV (PASS) 5160.47 295 A V wINE 7002.91 315 c V wINE 6451 315 A V wIo NE 6454.29. 315 C V w/o NE 6278.99 315 A V w/SFP 7068.1 315 C V w/SFP 7121.29 315 _ A V w/NE (PASS) 7002,92 315 C V wINE (PASS) 6623.89 315 A V wlo NE (PASS) 6454.3 315 c V w/o NE (PASS) 6462.33 315 A V w/SFP (PASS) 7068.1.1 315 C V w/SFP (PASS) 7197.67 315 Note 1: Linear extrapolation beyond end of BHP curve. Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I.of I

Serial No 13-419 Docket No. 50-336, Page 38 of 42 Table 17 - RBCCW Pump Brake Horsepower Requirements - Nominal Pumn - 75OF SW Pump Case ApproxImate Flow (gpm) Required Brake Horsepower A I 6627.02 315 C I 7135.96 315' A II 7287.42 3101 C I1 7251.35 3101 A II wILIA 7630.1 310' C II wILIA 7364.85

310, A

III 6876.91 315 C III 7074.9 315 A IV 5847.83 305 C IV 5075.67 290 A IV (PASS) 5847.83 305 C IV (PASS) 5358.34 295 A V wINE 7102.24 315 C V wINE 6627.89 315 A V w/o NE 6630.84 315 C V wo NE 6472.9 315 A V w/SFP 7183.35 315 C V w/SFP 7251 310' A V w/NE (PASS) 7102.25 315 C V w/NE (PASS) 6778.37 315 A V wo NE (PASS) 6630.85 315 C V w/o NE (PASS) 6637.67 315 A V w/SFP (PASS) 7183.35 315 c V w/SFP (PASS) 7348.94 1'310' Note 1:. Linear extrapolation beyond end of BHP curve, Form: NO3O1FOS Revision: 00-00 Dale: 10-28-201 I Page 1 of I Form: ND30 I F05 Revision: OMO Date: 10-23-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 39 of 42 Table 18 - RBCCW Pump Brake Horsepower Requirements - Degraded Pump - 80F SW Pump Case Approximate Flow (gpm) Required Brake Horsepower A I1 6453.5 315 O 111 7031.63 315 A II 7151.11 315 C II 7122.89 315 A 1 w/LIA 7414.49 310 c 11 w/LIA 7211.35 315 A I11 6737.09 315 C III 6972.8 315 A IV 5658.97 305 C IV 4885.31 290 A IV (PASS) 5658.97 305 C IV (PASS) 5160.86 295 A V w/NE 7003.16 315 c V w/NE 6451.4 315 A V w/o NE 6454.72 315 C V w/o NE 6279.38 315 A V w/SFP 7068.33 315 C V wISFP 7121.5 315 A V wINE (PASS) 7003.16 315 C V w/NE (PASS) 6624.29 315 A V w/o NE (PASS) 6454.73 315 C V w/o NE (PASS) 6462.72 315 A V w/SFP (PASS) 7068.34 315 C V w/SFP (PASS) 7197.88 315 Note 1: Service Water is 78"F or 79TF for the Normal Operation case. Form: N0301F05 Revision: 0D-00 Date: 10-28-2011 Page I 01 I

Serial No 13-419 Docket No. 50-336, Page 40 of 42 S CALC 97-169 R&V 4 1'E. 45 48 ZAGMRY

NUCLEAR, INC.

r, MP2 RBCCW - Design Basis Flow Distribution ZNt Document Type: QAPO Table 19 - RBCCW Pump rake Horsepower Requirements - Nominal Pup - 80*F SW. Pump Case Approximate Flow (gpm) Required Brake Horsepower A I 6627.39 315 C 1i' 7135.79 315 A If 7287.73 310' C II 7251.6 310' A 1i w/LIA 7630.49 3101 C II wtLIA 7365,17 310' A III 6877.13 315 C III 7075.1 315 A IV 5848.26 305 C IV 5076.08 290 A IV (PASS) .5848.26 305 C IV (PASS) 5358.73 295 A V wINE 7102.55 315 C V w/NE 6628.25 315 A V w/o NE 6631.21 315 C. V wo NE 6473.33 315 A V wISFP 7183.63 315 C V w/SFP 7251.27

310, A

V wINE (PASS) 71.02.56 315 C V w/NE (PASS) 6778.74 315 A V w/o NE (PASS) 6631.22 315 C V w/o NE (PASS) 6638.02 315 A V w/SFP (PASS) 7163.64 315 C V wISFP (PASS) 7349.22 3101 Note 1: Linear extrapolation beyond end of BHP curve. Note 2: ServiceWater is 791F for the Normal Operation case. 7.6 EVALUATION OF FLOW DISTRIBUTION WITH PASS IN SERVICE The flow re-distribution caused by operation of the PASS sample coolers Is quantified by re-running the accident cases with Isolation valve 2-RB-210 100% open. No other valve position changes were made to restore flow to safety-related components. The results of the PASS runs for Cases IV, V wINE, V w/oNE, and V w/SFP are included in Attachments C, D and E. 7.7 COMPONENT OUTLET TEMPERATURE AND RBCCW SUPPLY TEMPERATURE ANALYSIS In some instances, component outlet temperatures and RBCCW supply temperatures tor the LOCA cases exceed the system temperature limits determined by the RBCCW Peak Temperature Analysis as documented In Reference 4.48. The heat loads assigned to the CAR and SDC heat exchangers in accordance with Reference 4.48 are based. on the maximum RBCCW flow rate through the components and therefore result in a higher Form: N0301FO5 rRevision: 00-00 Data: 10-28-2011 Pagel1 of 1

Serial No 13-419 Docket No. 50-336, Page 41 of 42 outlet temperature when used in conjunction with the lower flow rates reported by Proto-Flo. This is conservative for maximum temperature analysis. All temperature results are shown in the output reports included in Attachment D, A table of component outlet temperatures Is Included in Attachment L. The RBCCW Proto-HX evaluation output reports are included in Attachment L and a summary Of the results is shown in Table 4 and Table 5. The RBCCW supply temperatures for each case are shown in Table 20 and Table 21 below. Tabile 20 - RBCCW Sup ly Temperaturc - 75'F SW Case Pump Curve J X-18A Supply X-18C Supply (IF) (OF) I Degraded 81.51 84.03 I Nominal 81-55 84.05 II Degraded 103.62 103.7 II Nominal. 103.72 103.80 II w/LIA Degraded 101.42 100.39 II w/LIA Nominal 101.57 100.50 III Degraded 93.03 93.14 III Nominal 93,10 93.20 IV Degraded 135.43 125.93 IV Nominal 143.00 133.69 IV (PASS) Degraded 135.44 127*37 IV (PASS) Nominal 143.00 135.29 V wINE Degraded 125.89 117.38 V wiNE Nominal 135.37 127.22 V w/o NE Degraded 116.96 116.64 V w/o NE Nominal 126.81 126.45 V w/SFP Degraded 117.35 117.43 V w/SFP Nominal 123.13 123.25 V wiNE (PASS) Degraded 125.89 118.10 V w/NE (PASS) Nominal 135.37 127.96 V w/o NE (PASS) Degraded 116.96 117.39 V weo NE (PASS) Nominal 126.81 127.22 V w/SFP (PASS) Degraded 117.34 117.99 V w/SFP (PASS) I Nominal 123.11 123.83 Form: N0301 F05 Revision: 00-00 Date: 10-2B-2011P Page I of I

Serial No 13-419 Docket No. 50-336, Page 42 of 42 'Fable 21 - RBCCW Sur jly Temperature - 80'F SW Case Pump Curve X-18A Supply X-18C Supply I (M (OF 11 Degraded 84,83 _8476 II Nominal 84.88 84.78 II Degraded 108.15 108.24 11 Nominal 108.27 108.35 II wILIA Degraded 106.00 104,97 II wILIA Nominal 106.14 105,08 III Degraded 97.68 97.80 III Nominal 97.76 97.86 IV Degraded 139.65 130.18 IV Nominal 147.1W 137.88 IV (PASS) Degraded 139.64 131.62 IV (PASS) Nominal 147.16 139.48 V wINE Degraded 130.23 121.75 V wINE Nominal 139.63 131.51 V w/o NE Degraded 121.35 121.02 V'w/o NE Nominal 131.11 130.74 V w/SFP Degraded 121.74 121.83 V w/SFP Nominal 127.47 127.58 V W/NE (PASS) Degraded 130.23 122.48 V WINE (PASS) Nominal 139.64 132,25 V wo NE (PASS) Degraded 121.35 121.77 V w/o NE (PASS) Nominal .131.11 131.53. V w/SFP (PASS) Degraded 121.75 122.38 V w/SFP (PASS) Nominal .127.47 128.18 Note 1: Conditions necessary to maintain RBCCW supply at 850F shown in Table 4. Proto-Flo cases were assigned a SW temperature of 791F, design fouling, and 7% tube plugging for an RBCCW supply temperature just under 851F. 8.0 PRECAUTIONS ANtI LIMrrATIONS This calculation does not represent the approved design basis for MPS2 until such time as ETE-CME-2013-1003is issued and the follow-on licensing activities associated with Ultimale Heat Sink reanalysis are completed. It Is assumed that during Normal Operation a SW flow rate of 7500 gpm is currently achieved or exceeded for the X-18A RBC0W heat exchanger and 7650 gpm for the X-18C RBCCW heat exchanger. It is important to note that the model documented in Reference 4.1 has not been calibrated. In order to verify the results of this calculation, calibration of the model to RBCCW system flow test Form; N0301F05 Revlstori: 00.00 Date: 10-28-2011 Page lot I Form: N030tIF05 Revision, 00-00 Date: 10-28-2011 Page 1. 0i I

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 97-ENG-01862-M2 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 6

• Dohu.hlo Dominion Nuclear Connecticut AI ROGOW System Heat Load Flow Role MIltn Uni 2 Pg io2 Cmloulalonll`ra97ENG01832IV2L Revssion 0 I

CCN 4 1 01122/2013 MWd.ofOperatlion Case I Case I-A-acC"W > 8SaF dP Flow H Flow Teiep N liv tcaud Flow Tamp N Heat Lead Equipment on Header A Tag 0 a 0 lie a (pi) lepm) t term) (.r) I a (gpo)* (1F) % (10' SthhI) Wulsiet as compreesnn* F-1A 2.00 5 0' 55 0012.5 5; 9 .1 Afterwoolera Waste Gas Compressors Aftercooler F-MA 2.00 5 5 95 .012 5 S SS 0.0125 BoricAdd Evaprat Pectikae H-26 Reactor Ve ssel su pport N,1 conceteCooll Con i NOTs '8 4.50 1 52 6 12 0.001333 12 0.001333 Reactor Vessel Support Noco~rnlepr Tag 4.50 0' 12 6 11 6M .001333 12 0' 0.001333 Concrete CDoolnr Coils 46-" 3 Rct Vessel uppot No Tag 9 4.50 32 6 12 O.G01330 12 0.001333 Contrite cooling CoIls Reactor Coolant Pumps and l.40A 20.00 4 Thermal Barner ,4 0.00 4 0' 2 65 I0.26 RCP Lube OilCoo1W$ X-73A 18.00

1. D 6S5 GS OA0 6i-04 Reactor Coolant Pumps and 3.40!

20.00 45 45 0 0.25 45 0.2 Thermal Barrier 4 4 45 RCP Lube OIl Coolers X-73C 1]o,0 65 65 0' 0*40 65 0.40 lll'SI Pump Se-al Cooler~s 91G 3,0 Q 0.4 (P-41A) -120 3,0 0.' 74 .'O HPSI PumpSeal Coolers X-217B 3.00

1

'14 7 0.00 1 7 0.00 (P418) ___!_7 1 _.0 LPSI Pump Seal Cotlers .X-ZISA 3.00 L 14 71 0,00 7 0.00 (P41A) ,,0 Contianment Slay Pum'p Seat Cooelersn PrayP X.214A 7.60 I2 7 0.00 7 0.00 Coolant (P-4.04)J Spent Fuel Pool lent X2OA 4.50 ) 1100 1100 Het 7,0 1100 0 7.3 Exchanger X-2 DO OD 120 10

16 1

Erheor. Shutdown Heat Exchanger X-23A 25.00 4020 0 0 0 0 PrmerDr alnank &Quelt* 1X-Z4 7.00 200 200 1 200 0 Tan, Coolern C*oM Coolers X-34A I.54 I 160 16 0' 0.S1 165 0-0.81 CEOM Coolers X-34C 6.84 0 1*5 15 ' 1

0.

76 a J 71 0 Containment Air Redlc. (CAR) 0-54 10 o s 22 500 0' and Coolant Unit -A 10 "0 So S 2.2 Containment Air Ileci. (CAR) X-35C 1.30 0 500 500 0 2.2 500 end Coolant Unit So Engr. Safety Features Room Air X-36A 3,42 $9 O 0 0 0 01 0: CoolinR Coil Sample Cooner X-155 -5 _0 0O RICCW Pump Heat P.0.A 0' 0.52 0.7 RUCCW Circulation Flow P-or Note 0' Total Before 16.n79 After 15.369 The Table above provides the heat load on the "A" Train header for Case I and Case I-A, befbre and after the heat.load reduction when IR3CCW temperatures may exceed 85'F. Note: All notes listed in the above table arc piovided in Revision 0 ofthls.calculation.

Serial No 13-419 Docket No. 50-336, Page 2 of 6 5004tInlo, Dominion Nuclear Connecticut A0 RBCCW*alSysem Heat Load H2w Rat. Page 12. U203 Caku!haicin 97-ENG-01862M2 fevlelun 0 CcN 4 011/22120Q3 Modes of Operatlon Case I Case I-A-RBCCW > eSF dP Flow N Flow Temp N HeatLoad Flow Temp N Heal Load Equipment on Header B Tag (1N (0 a (PSI) (Rpm) t (pie) (I s, (op).m) _' 2 _o'./,, 0egasifier Vent Condenser 11-24 0,07 S S 0.05 S 0 0..5 Aerated Waste Evaporated 4-25 ReactorVesselSupport No Tg 0It 4.5 12 6 12 @2 .0.1..1... 0.001321 Connrk.i f.vonl,,I Calls T_.-," Reactor Vessel Support NorTagA 4.5 12 6 12 a, 0,001333 1 0.001333 Concrete Cooling Coils Reactor Vessel Support NoTagO 4.5 12 6 12 V, 0.003-12 0.0113 Concrete Cooling Colls 0.00131 ___@__00 _3 Reactor Coolant Pump and Ilherlal Bantler P-4O 20 45 .45 =6 45 0,26 ACP Luba CIl Coolers X-73B 8 is 65 65 0,40 65 0,40 Reactor Coolant Pump told recr.real Bantrlru P.40D 70 ow 45 45 0.26 45 3, 0.26 RCP Lube Oil Ceolers X-730 18 05 1 65 0.40 65 0.40 IlIPSI PumopSeal Coolers x-7168 a 4 14.2 7 ole0 o 0.00 6P31 Pump Seal Coolers (P-42131 x.2158 7 0.00 7 0.00 Contalnmenltlpray.PumpSeal

14.

7.6 L 1 p, i t 7 0,00 0" 7 0.011 Spent Fuel Pool Heat X-2711 4 lion

loo, 7.6

,.3 Eschanger Fl23 0 Letdown Heat Exchanger X-52 10.2 IV 120 02.5 570 7.115 Shutdown Heat Exchanger X-230 is 4820 .0 1 0 a 0 Contalltment rc. 1,- I SOA 0 530 2,[ and Coolant Unit 2_"___._* Cottalnment Air Rccirc. (CARI 0-50 1.3 ip 0 Sao22 1 and Coolant Unit _-_50____@ .2 W3 _0 0 Engr. Safely Features Room Air .4" 0 IV Q~wle CllX-36B 3.42 S9 l Ill 0 0 0 I : 0 Cooling Call ODag !Uafer ýEflunt Cooler -3 .3 75a 250 8 9 0 SamplaeColers X.Fa 1 I 3 31 6p 0.65 31 F@ Sample Coolers rX-65 10 31 1 l 0.05 31 US "l l Quench Tank Heat tEchanger X-82 3.8 480 480 C' ____._'_1 a WtlCfXJ Pump Heet P-B215 0.92 Ig 0.73 T58CCW Pump terirculatlilr .P-IlAor

.,i Fle BorC

'or 4 C: I Total Before 23A44 After 35,034 The' able above provides the heat load on (he "B" Train header fbr Case I and Case 1-A, before and after the heat load reduction when RBCCW temperatures may exceed 856'F. Note: All notes listed -in the above table are provided in Revision 0 of this calculation.

Caic, N, 97-ENG4 M-k2-CaloNo 9-~NOO1M2U~? age f 38ATTACHMENT -A TABLE A -RBCCW SYSTEM COMPOUENT CONDITIONS44EADER X' 123 Page of 3..8 modes a, Opmrao Case.) as'iCe bp Cmse;l I/2 Hours Aff-rSh.itcown Case il 27-1I/2 Houn;A~fter Shutdownr Case IV -I*dz phas ter LOCA .Ca"V -Rachr, P~e After LOCA

• ote :

~ ~~For Car. IV aro V. Mimtfrough tm*equip-rf Idlerilfildby a.las ý*J.), fe requU kf, r saute '*w, n .Re fearoc* = R!N, ZM252 3-**C22 (Sh t s1 through bm) - P a& D R B.C X..W -Syun Equipmet on Header A Tag dP (V31 Neat.cad LOW H FtL.d ow Heat Load Row Kat Load Heat Load 9Pr" 5-Mw xI l'ov, PM XIO'BYIthr GPM XIOeBTUAM m 14 X1G STUbr PM XX8B'lilbr a

• Ga]s opms.

m "'F, A 2.00 '" .00 1 0.01.25 S,.00 0.0125 5.0D 0.01M5 5.C0. 0.0125 0 0.0125 1' E EGas C E 7s F-4 B 2.W 5.00 a0. 5.00 0::.012 5 %Go0 0125 S.00 0.0_25 ftRermoler -5. 1 (.Q)1 Cm

• Acid Evapomo H-.25i DEM SEE PDCR.

,Paokaqe 4___ Reactor Vessel Support NO TAGW 4.5 (Note6) 12.00 0.001333 12 0.001333 0 a 12 0.D01333 a 0 26 o.ncrele cQ.Inq..Cos (104.12 (Ian(1 ) cactmVo,ssei.Supporn NOTAG# 4.5 (Nate6) 12.00 3.0C01333 12 0.00333 0 a 12 0o001333 0 0 25 01C*_ 12 1_ (04.% _107 MforVevs.0 Suppor 140TAG# 4.5 (NoteG) 12.00 0.031333 12 *0.00333 0 0 12 0.00J333 0 0 26 crete Cooling Coll 12 (104.5) (107) -evaor cooan Pungs P-4QA 20.,0 45.CC 02e 45.00 0M 2 0 0 45.00 0.8 0 0 12,14 R F Aow Dra ss. emop ThemV. Bmer 0 45 from ref. I2.. heat I ""' i I ýad from re.14 RCPLubOlCsoolars X-73A latO --

5.

0.040 65.00 0,40 0 0 67.C0 0.4O 0 0 12.14 Fwlo&.press. drop 065 frem tat. 1 ; I Ia Tram ref.1 k o-Coolant Pumps P-40C 20.00 45.0D 0U26 45.00 026 a 45.00 02 0 0 12 14 Flows & p-ess. drop aru Thwmal Barler . 45 f8m W. 12 had I_ load frarn W.14 iCP Lub Oil Coders X-73C 18.00 65.00 0.40 mm.00 0.40 0 0 55.00 0.40 0 0 12,14 FpI*& prso. drop .6 from ref. 12, hea I____ I____ WdFrom_.14 ,PSI Pum*r s eu Coi lea X-2168 3.0.0 Row C 0 S CC M W 15.00"* f 13 'P41A) I Q 142 No"~t I_(We-71 1__ t.+/-Z) 1~'L~ IPSI Purnp Seal Cooles X-217B 3.00 1 .00 ,low 0.00 0.00 is5" 15.00 'P-419) 0142 Note 7) (Note 7) Woe 7) (1I)'m O (147) =0.192 P$1PlmpSelCocolrs PX-21SA 3.00 oW .0.00 03.00 0.012 3.0 012 3.0 0.012 D3. P-42A) -I_, A-I CC &3 O-L e 00 =) 300 CD CD m)0) C,) CD, -I 00 CID

Qatc. No.9-w-0a2M ATTACHMENT -A pane ar.fo TABLE-' A "RBCCW SYSTEM COMPONENT CONDMONS-HEADER N' modes ofoperassa Caw I Normal Opeavdam ICawl 1/28crimAfterShrutdowne Caso ii: 1I2 Hawrs After Shtfown Case IV - Injectione Pluse Afte LOCA CaeWV Rwirm. Phase After L0CA

4ote, FOT Came TVWnd V, byw

¶~~i h qiaettello an majakftowe/a. r reqrui~ed ter ~shilte oaw Floms shmavfr for to he a/lwer pen"Ot SM for reference atily. Gw.2S2G3.260= (Stresicizat/h 6)- P & 1 RBC.C.W Sydem Modies of OperAtion: Cass I CaseD____ Casse (1 case re____ case V Referen~ce Rearnks a 3mPM x g, BTUIbr PM e XIOtLhr 11 XIOBSTU.'h PM XDS~TU/h, ;PM FXi/101WLWh OtF) F) O F) 9COF) a I a ateS I a__ oti IdeR a_ x~re Spa Rump1k 12 T. Fm 0,0ow ro 0.0 1,0 0 oo twme tioa* plegib 13 tjl son AppU.eMX UWWolor (P-43A) i jB1jflf ____fl61n,1__ Ilae ) 5i2 olIz X-20A 14.00 1100.001 7.W 11100 1 7.5D 1!. 25 ~ rCckAna W 10 1 l.....I I _tj...j-ioo nNo ac I&NC~ thasw/ '~e1Ejr~a~e X-M~A 11.0 (LOOiLO &500 70.0 0 k-MG %I 40 For CaseV. refc Do Pivr rlTak X-24 7.00 200M0 1.00 200 1.00 0 0 200 1.00 Hl 1W2No auenmti Tank Coolers k%__ 2M

EiwM Coelers x.34 6.84 185.00 08*~

166 "1 0 0 165 0.81 0 0 7 Note 5 EOM Coes XA48 8.4 165.00 0.8.1 165 ".1 0 0 16s 13.51 0 a 7 MNARS EA)M Coolers X-4C 56-4 Flow 0 Flow 0 0 0o Flow a a 0 7.39 SPARE a 155 (Note 1 Mai 7___1 "Note_ -) SPARE ýoretaimenl Air Reaka X-5A 30 500.00 2.20 WO0 0.7 500O 00 60 10 /- 815.& 26 Refer to Asswrpfim I CAR) and C4411ant Una I___ jo S:t/nNti

QrftinfrnenAlr Reatro.

X-35C 11.30 50000 220 sm 0.7 60 .0 ~ 0 1300 (.00 82 Sectona ossmte 1 CART~~~1a (1306 Cel6 Sectio & .I2JBL S/wNotalO 10 ý.erv. Safely Feahmes X..38A 3.A2 0.00 0.00 59 0203 fa0~ 5 .34 .4 9.2.4 Note 1.oCmAitCOr ICooC06 Ss 0,4 0-14 11~ 9_2_ 43Nt ~~arel~~~~~~a ifor30 O5/bo I./ e Ramge cocier /0-lIg 5A .00 1Nt9/%/bC 5.00 N"l4W Z.00 I Nespove 11 u.uu Nte7ow 0.0 NoteT .ea Serzlon 5.3,2.C 1SCCW"PLup Heat P-1IA NIA N/A 0.82 WA o0.2 N/A 0.82 N/A 0.82 MrNA 0,82 3 Note i 'SCCW PUMP P-t"A or (Note 14) Flow NIA Flow N/A Flow N/A Flow N/A Flow WA 41 rEC/FCULALT1O FILOW S,21 4075 5C54 't,515 NateiS rOTAL 210 16,789 6,7W89 O9 s~l! 40.75 _____4 Note___ 15 O7-r7 C~t 6~ ' 0LPý 0) 3C) CD uZ Z cc 9 CD C O)Oh(9 A-2

Pape of 39 ATTACHMENT - B TABLE B "RBCCWSYSTEM COMPONENT CONDITIONS-HEADER B" WId., ofOaer'a Cas! I No 'al O'..a... Case I; 12 HOW* After Slohiown Casme Ill 112 No= Alter Shutdcum C-s IV -Io*-o Phae AlTr LOCA Case V - Ra'c=. PhaseMAter LOCA Nots:: Far Cae.V and V, llaom thupoh tne eoqtpment irfrMed by a,* sleeacn(, am ;equked for safe sýuwmsw Fl hw far (or the fte aine

Reference:

2m - (Shg hh 1 618-P & ID R.S.C system Eimoden of OpHeader. Ce I Tagase Case I Case CV CaseV Refirence Remarks EquipmenI o HeaderS TaRP dP (Psi) Fiow Heat Load Heat Lead Flow Hea Load Hut Load HeatdLoad

• gps GPM Xl1BTUt/r X1STUlhs PM Xlotnulhrs 3PM X10VU/hr M

XtOBThhr gOF) CF) OF) (OF) OF) 8 ______ HolI`[e 8 Lt8 8 aeS NO* DsslarVes taidienrt N-24 0.07 5,0 0.00 IWO 0.00D 0.00 2.26 @ 5.00 ________Nte i.eoted Wa*oe Eva ParaBd H-25 NIA NA W NIA WA WA W/A N/A. WNIA NIA NIA 'Reire fn place,

• '=kag*e

____PDCR 2-.03-79 ResýFtw Vesst Sppma NO "AGR 4.5 (Note 6) 12.00 0.001333 12 0.l0013 0 0 12 0.00333 0 0 26 .ConcreteCoo(tng cob @12 _104."1 (_ _i__ 0 2 RectarVes*s$*Srppot NOTAGI 4,5 (Noe 61 12.00 0.001333 12 0.C01333. 0 0 12 0.001333 0 0 25 Conwre,, coooSg CaDs I 12 U I I IMKS) 1 (107) 1o_ R""AVF

• o"NPumpwid P-408 2000 4,.00 0.25 45.00 0.25 0

.00 45.00 0.26 0 a 12,14 Flaws & press. drop

• t 45 ArS ref. 12. hea.

RCP. Lload fro, ret 14 RCP Lube OCooers X-73B 1l.00 06.00 0.40 65.00 0,40 0 0.00 85.00 0.40 0 0.00 a2,14 Flssa&Fress dro @65 ftom res. 12, hreat _ __ad rawr ref.. 14 Ratr Cooantr*Pump UAn P-400 20.00 46.00 0.26 46.00 0-28 0 0.M 4500 0.28 D 0ox0 12.14 FWsaress.Orup arnil Barier @ 45 rom ref. 12. *,'w I____ od froms ref(4.. RCP Lube OI Co,*em X-73D 18,00 S5.00 ID40 66.00 0A 10 0000 65.00 0.40 0 0.00 '1214 F'mi s & n. drop 66 frommret. 12. we ________load fnromre?.14 HPSPu Se*l Coole" X-21R 3.&00 lw 00.00 lw 0.00 15,00" (-'CB 15.00. .M.6. 13 1CC.1 (P-41C) @42 Noten O (NOLt7 Oi 4: LFSI.:mvSeatCoolers X-2158 9 3.I 0 a aw 0.00 3.M0 Q.012 3.00 0.0122 2M 0.012 0 13 CosmntnrlSp'ayl'rnp X-2145 T.IM 00.00 0.00 Flw 0.00 11.o I¶-.00(#4u; l-1 SeeAppendad I N S es6b-43B) @ 11 Note 7 s7) I "te ).1 75%,u SperSi Fuel Pool Meat X-203 4.00 1 I0.00 7.82 ¶100.00 "70 1 1100.00 7.60 0 0,0 1100 9

3. 11.& 25 E av-harkr 0

Notel .1.,, n-etdown leat arEager 2 020 570,.V 2.85 110.00 3.Ca 0 D w 0 0 0 4.e6 Note 4 =_ 1200 (Note 7) 00 30 CD0 f ~0 0 ~-ICD 0D w -D6 CYIO)C0 CCd r-PI

CIALD No.0-~19iM Fpos2z o.36 ATTACHMENT. B-TABLE B "RBCCW sYs_.M COMPONENT CONDITIONS-HEADER Bw case; I, -.3-1rZHowa After hUWl~rh rcase l 112 Ho=~ After Shutdown Case IV - InjectO Phase A~ftr LOCA Case V - ftedra Prase After LOCA Note: For Casa IV and V, flows throuigh thfe equipwmefE ldRdtld by an asteffsll4). Wa reqdrod~bW fo Shee~~dowf fioA'e siheen for tar the WAr nequprnae it aarermoe: Ownm. 25203,2q= (ShoeMl 1 Ithrough, 8)- P A ID R.8.CC.W Sverern WMoe-s of flperatioor, case I aseii 5case It -- due IT Case V Raforcilce Flfernrkx

• quf unnt an Header a Tai dp IW ow Megt Lead low ItaLoa

low Heak Load

'ow Heat Load Ba~a @gpm PM Xl'ST)Urhr ;M Xi~eS~flt, 3t XQ-Wle M Xle Extu hr 3N eTh,1 XllO io h OF) RO01 WOF) M10E) IF) a_ 1__ &4 a iotaf a Me8 _B Shtdw i~~ ~Exat a X-238 5.00 0.00 0C 3500 MO.O 3500 31 a 00 06 5.40 For Case V. rcber to ~o4M3 Note3 3Z ~Assrnt'on 03r 1C nM.ArReoc.- (CAR) X W35 1.30 5 00.0 2.20 SOO080 0.7 5 00.00 OM8 f2 oo 80 3800 ý 815,4928 R efer to As sumption

and Coolant unit-_ -___

I 1& 10 130) So~ Ng~til Er Safrfie Feateem X-340 3.42 0.00 0.00 59 0.293 59 0293 '-5,9-0.349 5g * -W4 9,4Z &843 Ouegaier lqhwnt C WMT X-52

9.

0 0.00 3.1o 0.00. G aw 0.0 a 0.0 ao a 620 Seeoo&.. .Ejelfate 7W Hol0 9___ Hol s~pBCoolerspH 10.0D 31.00/ 0.32 4100 0.002 0./A 0.82 0/ 0.2 Wa 0.8 38Noe R8CCW P.MP tP-1IA or (NoteI14) raw NIA Plow WEA Flow W/A Flow MA Plow N/A, 41 roTAt. 3,852 23.440 SAM9 84.M 84M JTOr-Koat 15 02.. /U&ý ~ ~ ~ 400-Z ac"cJ a? 00 0, ~0 CD ZZj )9 Po (1W -Pý h L0 O) 0)

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 12-001 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 43

• D~nonl."

Dominion Nuclear Connecticut AM MP2 SW Modal and D0,h Basis Ama~s.s XMISrwtn Untri 2 P;a2 30 0l.S8 Caleuuatiln 12-001 Rsvhion 1) Addondurn NIA 011/'2013 Mow, gpni 10858.5 9326 900 121 73.5 58 'B' Trnin SWV Flow 'rest Data for 'C' RBCCW (from Ref 3.9.8) B SW Ildr C RHCCW B EDG X183 Flow, 2pm 10878 9529 920 42. 'BTrain SW Flow Test Data for 'B' RBCCW (from Ref 3.9.8) It SW itir Bt RBCCW B EDO X183 Flow. gpm 10934 9544 960 43 This Addendum determined the appropriate position settings for the RBCCW summer TCVs to be as follows Valve It) Thnie Pmition [% Open] 2-SW-8. I A 67.87 2-SW-8. 1 68.97 2-sw-8.tc 65.7. Additionally, this addendum provides additional pump curves for 3%. 5% and 7% which are discussed in further detail in this calculation. B. This addendum incorporated the changes to the MP2 SWS system from DC MP2-12-01 177 (Ref 3.5.37). DC MP2-12-01177 removed WEKO Seals A2 and A3 in the Millstone Unit 2 Service Water System and replaced these seals with a single seal and sleeve arrangement, approximately 7 ft in length. WEKO Seal A4 was replaced by two WEKO seals, This addendum ensured that with the change in the WEKO seal configuration, the Millstone Unit 2 Service Water Systerm still meets the required flows to each of the necessary components in the most limiting case for the A train of Vhe system. The A train was evaluated due to the modified seal being located in the A train flow path. 8.0 Dusiga Inputs 8.1 The SWS model fluid is 'Salt Water - Salinity 35G/KG". 8.2 Nodes - The node names are depicted on the SWS schematic in Attachment A. A list of nodes as well as their boundary type and elevations is provided in Attachment C, 8.2.1 Temperature - The SWS model assumes a maximum service water inlet temperature of 830F for both summer find. winter conditions (Ref 3.1.1 and .3.5.39). This is applied to the mondel by setting the Temperature (oI-) in the Boundary Conditions tab for nodes P5Asuc, P5Bsuc and PSCsuc. The maxinmum winter operating temperature is considered to be 60 "F. Reference 3.6.6 allows slipstcam flow through the swing RBCCW and

Serial No 13-419 Docket No. 50-336, Page 2 of 43 MIN. Dominion Nuclear Connecicut ,ll M?2 SW Modei anid Deoipn Basis Ans ysls MdIlSto10' l u' Page 31 Q1 95. Calculation 12-Dur Raeslon 0 Addendum NWA 01/31t2013 TBCCW heat exchangers when the temperature is less than 58"F. These heat exchanges are isolated before temperatures exceed 58"F. 8.2.2 Node Elevations - Elevations are entered for every node in the system. This field is used in the pre-ssure loop to account for the hydraulic head djfferences within the piping system. The elevations were first entered into the DOS program in Ref 3.2.7. Additional nudes and corresponding elevations were added as part of the hypochlorite and gland scal modeling changes in Ref.3.4.4. The change to the diesel discharge piping (Ref 3.5.36) also added additional nodes and elevations. In older versions of the model (versions prior to Rev 2 of Ref 3.4,11), the. elevation for node 'TromLIC" was found to be inconsistent with nodes "FromLIA" and "FronmL1W' and upon review of the as*sociaited drawing (Ref. 3.7.1), was found to be in error, The nodal elevation is corrected from 22 ft to.16.59 ft. The elevation* of node ARBCCW has been corrected to 2.1 feet in accordance with Ref 3.7.40. 8.2.3 Boundary Conditions - Boundary conditions are specified in the Nodes tab of the PROTO-FLO model. The Boundary Conditions Reports arc included in Attachmnent P. Because the SWS Is an open system,,discharge boundary conditions roust be specified in the model. A list of both the suction uan discharge MP2 SWS boundary node names, boundary condition types, boundary condition elevation, boundary condition static pressure and boundary condition temperatures is provided below,

Serial No 13-419 Docket No. 50-336, Page 3 of 43 87DOMINO Dominion Nuclear ConnecticutAl MP2 SW Modal and Oes*tgn Basis Analysis MIllstone Uanit Page 32 of 96 Calculailon 12-001 aovisian 0 Addendum N/A 01/3142013 Table 8.2.3: Suction and Discharge Boundary Conditions Nutie Num

• im lxlnaMy Corndition Elovution rni Static PI' iial Tehmpr tll'4r G.At.Quoiry.

piswactirg(fl, Nving) 0N 14.7 so DOcliQuarry Discbharge nuwtnsi 0,9 14.7 HO Proir~l05i Ilinu FnIc1llh)t sL~choar (t Oittji) 9.72 1."7 LI Anul Oiwlbaage jflnwlnlý 113213 14.7 911 LIUout Iisdcalgo itnvwig) 16.33 14.7 N1O I-lCet I3lscharge(ftlowin 8 ) ,913 14.7 NO 1'5Asuc Suction (natic) ,%en intake 14.69 it3 ,5I'S ¢ Soction(slatic) elevation i1IM .0i p.csuc Sut;ion (tItatC) bw 14.69 t0 PAScai DISC~tar f.lowing) t0h79 14.7 s0 PSTISc.a Discharge (lowinjg) tL-.79 1,4.7 80 PSh15CRl OMdritc (3oiwing) 14.79 14.7 A0 P*I) Disnl M.,nn (fluwing) 14.79 14.7 80 RBLCWoulA I inrle.rge (1o1"Rni) I 14,7 86 R-UCWots Misd+a.Te (fluwing) I 1W7 116 "T1ICC*Wo Dthrge.(nowin) I 1417 8 8.2.4 Line breaks are modeled in the node conditions by setting the node from 'Flow or Free' to 'Pressure Discharge (Flowing)' and changing the static pressure to 14.7 psia. The isolation valves downstream or the break are closed. No exit losses are assumed at the break points. In order for the break flow rates to be shown in the surnmary reports, flow summary titles are provided in the infu tabs of the pipe sections upstream Of the breakos. Since thesie titles are recorded globally (i.e., for all ease.s) in the databases. they are deleted, once the seismic runs are completed and filed. 8.2.5 Piping is considered to remain intact for a design basis event (LOCA) and if the flow is not automatically isolated, flow into the branch is controlled by the system resistance in the branch and predicted by the tlow model (Ref 3.3.5). 8,2.6 Line breaks are assumed 10 occur in all non-safety related branch connections for a seismic event. For the TBCCW system, the break is assumed at the inlet of the X-17AIB/C heat exchangers (Nodes: Xl7Ain, X17Bin, and X17Cin, respectively) and the corresponding valves 2-SW-OS8AIBIC are closed. The specific pipe sections arc 34, 35 and 36 for TBCCW breaks. Since this break is not modeled at the intersection of the QA/non-QA piping, the justification, is included in Ref 3.3.7. This Note tilla while ilhe l:hundaiy conlditions lib ill tIle auitdel Witt) rL't1llil'es a tut11l presure, t'ROTO-FLO only uses this 1t0111 pressure t1s all initial gtm1. (Rof 3.9.4). The 1ot1ll pressure needs to be calulculted from the elevaition head and the vel(lity head, wahich PRO'rt)FIXt) determnines during each culculation.(Rtef 3.8.4). Ilier*orae th1 total preostsr it ntut c1'1sidered tIere to be a criticul design input and is not provided. The suction temperatures are tho only critical input. The other inpuls are used as an initinl guless.

Serial No 13-419 Docket No. 50-336, Page 4 of 43

• Oamlnlan" Dominion Nuclear Connecticut All MP2 SW MorJel and Design Basis Anasis Millstone Unil 2 Page 33 of 96 Ca!culallon 1-M!R-oviinn n

Addemaum N/A 0t53112013 evaluation has demonstrated that the MP2 TBCCW Heat Exchanger (X-17A, X-17-B and X-17C) anchorage is seismically adequate to prevent the heat exchangers from excessive movement daring a SSE (DBE) seismic event. "'herefore, the piping nuzzles attached to the heat exchanger will not experience any significant displacement that could induce significant additional stress in the attached piping. This supports the analysis and conclusions contained within pipe stress analysis calculation Ref 3.4.41. This ETE in combination with Ref 3.4.41 demonstrates that the TBCCW supply piping pressure boundary will not fail considering the limiting seismic design basis event. Thus, while seismic failure of the TUCCW supply piping has been precluded, failure of the piping, at the TBCCW nozzles can be conservatively postulated for the purposes of the hydraulic. analysis (i.e., failure could be postulated on the discharge nozzles of the TBCCW heat exchangers based this assessment). 8.2.7 Sodium Hypochlorile Injection LineBreaks-For Sodium Hypochlorite injection, breaks are assumed at the nodes R066670ut and R066680ut immediately downstream of the orifices RO-6607 and RO-6668, respectively, and the corresponding downstream valves 2-SW-84AIB are closed. The specific pipe sections 128 and 129 for Sodium Hypoehlorite line breaks. 8.2;8 CW Pump Lubrication Line Breaks - For the supply line to CVW pump lubrication, the break is assumed at nodes ToPDI-A and ToPDI-B and downstream valves 2-SW-134A/B are closed. The specific pipe sections are 228 and 239 for cite water lubrication line breaks. 8.2.9 Switchgear Room Cooler and Vital Chiller Line Breaks - For the limiting case for the switehgear room coolers (Case 6b Seismic Summer), breaks are assumed in the soisuiiclnon-selsmic interface downstream of the SW return piping from the switchgear room coolers. and the vital chillers (downstream of valves I-SW-192 and.2-SW-1.97, respectively). Assuming a break in these high point locations reduces the flow to the branchcs supplying SW to the switchgcar room coolers and the vital chillers. Two dummy valves (dummyvlvl and dummyvlv2) are created in pipe sections 190 and 172, respectively, to isolate the downstream piping to simulate the break. The nodes SW192Out and SW197Out at the hreak locations arc set to atmospheric pressure by changing the node from Flow or Free to Pressure Discharge (Flowing). 8.2.10 Siphon Breaks ý A siphnn break is assumed for large hare piping at high points of discharge piping when the flow rate is less than the critical flow rate required to establish siphon. Tesn data has shown that a siphon exists in the lIDG discharge piping for flow rates greater than 500 gpm (Ref 3.4.101). Since the SW. flow rates in the EDM discharge outlet are above 700 gpm, a biphon break is not considered on the-high point of the EJDG discharge piping for summer operation runs. T'his 'is conservative since it biases towards further minimizing the flow

Serial No 13-419 Docket No. 50-336, Page 5 of 43 WD. nll Dominion Nucear Connecicut All MP2 SW Modat aid e008n B90 ia B rasySs M INslons trUni! 2 Page,34 ot 96 Caailatinrn 12-001 Ravlslon 0 Addendum NI/A 01/31/2013 rates to the RJ3CCW heat exchanger. A siphon break on the high point of the IDG discharge piping is considered Ifr all winter cases. This is conservative in minimi.ing the flow rates to the FDG heat exchangers, The boundary conditions in the model for a siphon break are similar to those for a pipe break. Siphon break is assumed at nodes DGAout and DGBout by forcing the boundary condition to atmospheric pressure by changing the noda from Flow or Free to Pressure Discharge (flowing). Dummy valves dummyvlv3 and dummyvlv4 are created downstream of the siphon break in pipe sections 113 and 114, respectively. Clusing (hese valves prevents unrealistic flow rates into the system at the siphon breuk poin due to low pressures. 8.2.11 Intake Elevations - The intake elevations are provided into the :SWS model at the pumip suction nodes P5Asuc, P5BSuc and P5Csuc. The static pressure for the suction nodes is set to atmospheric pressure (14.69 psia). Originally, the tide levels were determined based on Ref 3.2.3. A review of this guidance determined that the tide level of -1.4 ft MSL for the design basis accidents was non-conservative. According to the Official tide data for New London, CT (provided in Attactment L) the lowest tide since January 1983 (the beginning of the current NOAA tidal epoch) wut (-)2.77 ft MLW. A review of the lowest IVMLW tides in 73 years shows that there are tides that have been below this value. However, it was decided this value was sut'ficiently conservative considering that the tide must occur concurrently with a LOCA or earthquake and while the SW pumps are degraded 10% and the SW heat exchangers are at their maximum debris loaded condition aid water temperature is at 80' F. All of the 10 lowest tides recorded at New London occurred between December 1" and April I" when water is cold. A review uf the winter seismic case determined that the. +Aft MSL should be replaced with the mean high water level identified as +1.3 ft MSL per Ref 3.1.5. A review of the Lo-Lo tide case shows that -5.85 ft does not bound the -7 ft identified in MP2 FSAR Section 2.5.2,1 (Ref 3.1.3). Since the -7 ft is for the maximum sctdown condition during a hurricane, it is assumed the circulating pumps will be off in accordance with Ref 3.6.10. The SWS intake levels in the SW model are provided as follows;

Serial No 13-419 Docket No. 50-336, Page 6 of 43 Dominion Nuclear Connecticut All MP2 SW Modnl end 0asgn Besis AnalysIs Millstone Uit02 Page 35 of N COleultlion '.2Gl Revis*lmo 0 Adoendinm N/A. Ot1"l/2013 Table 8.2.7: Intake Ieveisror Case Alignments Mean Sea Model Itnake Sreern Calculated Case Case Nsne Levelt3 toreclion4 D~ifferential' IntakeLe-ve.t 2ai-k J.OCA w/o ULP -412 5.0 -2.5 11.7 Nornmal Ops. 4a (Diesul B yprav -1.4 '5.1) -2.5 -8.9 Open) Normal Ops. 41b (Diesel Bypass 0 -5.0. -2.5 -7.5 Closed) 61, Seismic Summer .4.2 -5.0 -2.5 -11.7 6c-1l Serinkt Winter 1.3 -5.0 -2,5 -6,2 60 Seismic Summer -4,2 -5.0 0 -9.2 w/ LNP L-LoU Tide wto -7.0 -5.1. (1' -12 LNP ll1v-Single Puita b Operation -1.4 -5.11 0 ,6,4 (SI) Cooling) 8.3 Pipe Sections - The pipe section numbers are depicted on the SWS schematic in Auachment A. Attachment C provides the details of each pipe section in the SWS model. 8.3.1 To account for the piping liner material and to provide additional documenation of the basis for.the pipe section inside diameters, all standard schedule piping was input as nominal size and schedule. 'The PROTO-FLO program automatically adjusts the inner pipe diameter when the pipe schedule information and liner thickness is selected.

However, this f*eature only works when the pipe diameters are selected after the pipe thickness. The following piping materials are used for this analysis: CS Acc6rding to Attachment L, the New London NOAA slasion's Meti Sea Level is 53.0 ft and Mean Low Water Level is 3.71 (1 Thus t.) convert MI.W ito MSL, (5.06..371=)t.35 ft mnust be subttacted frn the LLW level. These values have been rounded up. 'there:

has been a small increase in MSL since* MPS was built but the difference between MSJ. and MLW is essentially constant. 4 Based n the riduction and analysis or the A and B Header Test data collected during the 1992 rlfueling,oulag., empirical adjustments have been made to the flow model. Model correction rector ol'-5.0 1t per Hers 3.4.9 andt 34.10 'rThe operation of the CW pumps in conjunction with the debris ladeo trash roeks said traveling screens will cauec the level in the inlte sntewre to droip. Front Ref.3.7,32 & 3.7.3N it is concluded that the CW pumps will trip offl tine automatically when the level drops 30 inches (this is the trip set point) across the screens. Therefore, ir the CW pumps are in operation, a drop in level of 2.5 ft (i.e. in cases without LNP) is incorporated into the intake level value. 6.Cutculnited Intake Level = MSL - 5 ft (model correction) + CW Pmnp Condiiton (Intake Screei l)iflhretaihl) 'Since the -7 ft MSL is fur the msximum secdown condition during a hurricane (FSAR Section 2.5 and 9.7), it is assumcd the circulating pamps will b* off in accordsance with Ret-36.,11. 'This is a sbutdown risk eame t i 6 acceptable to tim Ilte hnurmttal operalion tide level.

Serial No 13-419 Docket No. 50-336, Page 7 of 43 9011tfl. Dominion Nuclear Connecticut I MP2SW Model and Design Basis Analysis Millstone Unit 2 Page 36 do 96 Caljulalian 12-001 Reisiorv 0 Addendum N/A 01131/2013 Epoxy Lined, lnsitufonrn. Copper Nickel and AL6XN (carbon steel). Each pipe section data sheet has the applicable pipe class, size, and liner material and thickness in the "Notes" field. The epoxy lined material is used for all JGD class piping, and all PVC and Arcor coating piping, unless the piping has been replaced with AL6XN. The material is assigned a liner thickness of 1/16" (0.0625"). Thus, for example, a 24" class.IGD pipe (standard schedule is schedule 20) has an inside diameter nf:24.00 - 2*(.375) - 2*(0.0625) = 23.125" The Insitu'nrm material is used for the appropriate sections of class KE piping and was assigned an ubsolute roughness of 0.00026 ft based on Reference 3.4.2. The material is assigned a liner thickness of 0.5" also based on Reference 3.4.2. All class HUD piping is copper nickel and is assigned an absolute roughness of 0.0002 ft and thickness zero based un Reference 3.4.2. For AL6XN, the material roughness was changed to unlined carbon steel and thickness zero (e=.000150 ft t=0) in the piping section. Specific details of which pipes have been changed from lined to AL6XN are provided below. 8.3.2 Piping Changes - Table 8.3.2 summarizes the changes made to the pipe data for the MP2 SWS PROTO-FLO model as documented in 3.4.11 and 3.4.3. The spools that have been replaced with AL6XN are identified. (AL6XN is modeled as Carbon Steel with liner thickness of 0.0.) Table 8.3.2 : Summary of Pipe Section Changes in the MP2 SWS PROTO-FLO Model Mrwf NW, i S.d. I 10d. I ID LI,.,Ib 4 6 03 Ilt"

d) S3 1

WCA d OM 0r2' 21.25 Inc. M OI.WN RI-A II544ýM,* d i .J1P I,.*Ul ) I .2' C.b.,w, 23.2* 73.235 W: N* 14 1N~41W I R 1.4 MO(WIINTI 111 Rq,MII,, q 1 1.0 S-.147W V.. 01 2.2 1 2% 2 lie N 14 a() lRl.0 ri f 1.*.0IJ'4.,l't RNhcc ',oI,l l.4.p.* I.I7..jIi,.ll l lI~...U,,I 2nt 1 .10..~ 25.175 NrC N Y, KI-") Ml. I.IN Iplb... l V,6.n I AI I .IN

• 1II Ini,

Isr.v o! 5., I ., I C. N, II. P I Q 08 .It I MW (KII.I*10 klI.,111. II.U.I2*,HI l oI 172. I).= NC NC, II IN 111 kl*I iIý%I S r*l It-I ln I l.I.Il,192O I. o IleI I2I 2.1,t 2 NV NU IN 2I' R5. i2 rl. 50.* I 7SIII 4**f¢ l r t pl2,,lt el 2.mt

li-'J0 1.2 1

.,45I "*l**,i ~ t'* 1.0. lC NO C II ,N.:AII.S 11112 D1121 M A 11I111-II Is II*-I*I

• . 4 A.3 11471

,24 l2't 1.21 Vons'.l..M NIl W. 5 2 'N WI.-% K 12

*52 711115?

ILh*rp -b"1131S-I-1*tA.*". (2,.".Ml0. 22.,25 n1,u7 NC NI:. 112 1.1.. 2A 112 ',*-.'l I L)* "*4'OLr..,L*2 51-*t 100I 1-4.~d pM 91m IG 1 -SA t-I.'1 f JIMU I.*t NC~ 012..11III2 Sr.-1, 1111)1 M-.. po 1XN1SSI2"IIl .17 0 fI 1t-I, 2 -I

l.

,105111) I rt I I .IMi. 14 5ý*,d I *Al ON S I I"1I1 l.121-*I', ,l L A1...)11, I-4O.IrjI-7l IAA27l~ 4o,I4HIi .flI1..I I

• C

,N 'NC = Nn Chtinge 10 Epoxy Lined has thickness ofO.0625 in., Node Diameter is calculated as ID-2*0.0625in,

Serial No 13-419 Docket No. 50-336, Page 8 of 43 1111omlnn Dominion Nuclear Connecticut MP2 SW Mo-*i and Oesign Basin Analysi s Mhilstone Unit. Calculatlio Ia-t01 -Rsion 0 Addendum WA All Page 37 of 06 01/3l/2013 mi.sI CA. t, S In[//1 Dnep,tp/ 4501./ll 511.1 M ,tiNI O 1/ i 2' 4 10 114 0 l raa/ t ite W i6 t/h - Wl .., /5.n/ l f lr ./ 91 noW IA R.62- '4/b/.60/,,a* 2: W.: ra/tI, 4t//n S46.1/$8.114 1 DM) a-,.nlil5 .p-t1 l CD*-. I 1Ml -a'i Al M7285. !7 9109H4.0 A-IYA2AMk/tiRJ-I 119,1-0/ 1,-4 '/1)76. U11.4 AM-Wl5tfl//i si ,IH4 do 156 rt2N./i111 2-1362/la)-] nrkt, an, 26/k/l1 6-1 MN ./ a ati) 6 /It Rin, Olla~isiC.-,o Me.il tc/ul m t i,nntiaa. nWa5 /1.ha wa 61./8///16/lI-1. /6)41,'/Nifl6/li4 X-/tfiiaA" // la/inn/," /aZ//t/J// 8f ltlltf (nd.2 t i A/ 2/II11.0) 44S2Ai)146-00M 2 81 as n/aIA"i tnn tnr uu22 ai M.I oll I*)hO l dtýll'w q*' ll 1, t field I d6 - I WARMI"n1; c/naffA 6 rl /4c P, .1 2 '. 54/I It II ("IAt W04140 /,51*, t"A. urib r '4. 16i / '. £4t.1I-1 44-1/ft lIt 11612.11 .M 7 e .i 6/i-i /v K Int* 4316/ I4/. A.Fil 10inW. An.,T ke0c/.t 1156 is4)! 18 4: IA95 IUN 2258/ .*t.l J - /46-/Itf* , i.lpa Wil dc. I sinl I, 144 1J -,)/i/a 11 I 51682-t / i/K/ / 6t 1 t 1A.6 n/A0w I, t t //"ia/t,ý- a/nt 7 3' h./ Pa,. n1.t/vt 56 164318! 162 1MIlA-) Wa.$ills I /tA +te q/I 26/l)4.1 0 1411" uI tins:U a-/ 1 /O,., A6 149-22 t1tDWItMODt RlnI'.I.p W IU- &/ISxti. ./a flM

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Serial No 13-419 Docket No. 50-336, Page 9 of 43 Dominion Nuclear Connecticut MP2 SW Modal and Desagn f3asis Analysis Wilsitone Unit All Pago 38 o196 01/31/2013 CaitIintiton 1 2411 .212114(3 92-.:-It i 5 Revision 0 Addendum NIA iNm Itptrlp'i 12333 '1 4#,31.111A.tInk.nil. 1151-1-uS Nm i Di 1 2 4 2. nfl W I4 I r p o C a l t. I I 4 ' nt"in 4-P1t (u/Ilb 'if Ismit W "wt c iln. Si j1S124:l1-42124 jalt-It-n7 Wip It3.11-Li pit..231-m ivcn .J i h1nirnhI IYoN I to .rm.J I11111&NP 0P.it4 ii10 e. 41.J 40ii ran,, SIn tlgl ti roNat WAl 1.081 .a ill: Na: lii: NU 9q-L4.i PA.. A17' 111 4.211 PIP. I-I.t I... !I. 041in Ill kZ 155 t III) 12A U-' I ia.l pi-1-1 Mmtw~j.pnl i.,I]nWebii IWskýIji ll P1 1 1-ivm Un.t P3iiti4 I, tihWt Al1U 1(1. 32.1min( L2110I.LL1-I dC Svp~iaw. ld'iirulkfl uiri 9kna, nnlan"ila l II21 7247FNia.1iii Itan. Al*1 Wfi A-W-1.14. ai PVll ýIi, ,iaa Si-4 k,-an

d. 71-1 4,!,

II .nbb ta 'lnnI nlO? ltf ~ lIr a liII ~ i,,.tW-MnJi -1, Al naXI. Its. q1~~) 2 i"Iqfl W2I-15il.(. lipt.[It iLt.i.1 XM(4 N S ill' 02.Stlij Wili4iilt1ilY Wlpiul-t AUliS? IN4 q2-39 K2 imkl4-1*W't RcPhonla,.h ATiAil .1173 nip 1:1 R.- s~ UN2-lnS.liai4 P.-na -tAit.11*1b1At(MN Liiuin U-n 0-,mi No v*A ?.Val 1:0111 CC! NV: 61163 He1* MC .-I*oal NQC NV tint 2,06 1W10 2307 itini 1.07 2.0107 10*7 2.04;7 an:. NP: MC NC NC Ni NI Ps13.L NC NC N11 VC NC NC. Nit. Vi. 8.3.3 WEKO Seals - A resistance of each WEKO seal installed has been calculated and input as a misc Kfixed in the pipe data input screen, fittings

tab, A skeich of a praotyypic WEKO Seal croes-settiei is shown below:

D, 0.75' Whcrc: Do Orifice Inside Diameter (in.) Di = Pipe Inside Diameter (in.) = Orifice Length (in.) ' No*s I n'un Pipe 17 which is nuulinold.

Serial No 13-419 Docket No. 50-336, Page 10 of 43 ?Daominion' Domlnitn Nuclear Connecticut All MP2 SW Model and Design Basis Analysis Milislone Ulms Paga 38 aof l Caloulavol 12-00a Revision 0 Addneium N/A 01131,2013 The WEKO Seal is modeled as a thick-walled, square-edged orifice, conservatively ignoring the beveled edges oin the seal and assuming a constant inside diameter that is based on the smallest diameter of the seal. The resistance is calculated using formula (4-15) From Reference 3.8.2 (see Attachment I): K = [o.s(I-_ Ao)075 + I A.) t.375 + (I A). +I Q:Y A El]A At, All A) + Dj+ (A. ( 1 Where: K = Seal Plow Resistlance Coefficient Ali = Orifice Flow Area Ai = Pipe Flow Area fT = Pipe Friction Factor (Ref 3.8.1 "K" Factor Fable pg. A-26) D,, = Orifice. Diameter = Orifice Lengtd T = Coefficient Representing the Orifice Length (Ref 3.8.2) ,c is determined by the table in Ref 3.8.2 Diagram 4-15. To determine E, one mast calculate. /Di where Di = and nio is the cross-sectional wetted no perimeter. Fnr a pipe. the llydraulic Diameter, Da. is equal to the Orifice Diameter. D. Table 8.3.3-1 provides the data for each WEKO seal and flhe calculated K. The direction of the tiow determines, which diameter should be used since this diameter is not the same before and after the WEKO seal for all

seals, Table 8.3.3-1: WEKO Seal Resistance 1.1117r.0 1

tIk t M-1, I Plý Aý 4 P;x K. M. r. Is, ft l*s t In* Itlow [o lid N1 , I 1.it I1 Ii 050 t-I 2.5 Vide. Al I 1W; -1. 1,0721 10100 0 Z .01.M X 070 33* -I M72 21 0411 .'24.A 34641 ft nill -.1 tlI k*- AIC 101II2 2.o t.W*.. I Oitn i i

1. 32 4 20.5 144.:

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• 1t.4n101,.0.l 20164 0

2i i 101"i 111 1 2.1,i5 1*fm. 10.15 M4. 5 1 013 W. i £ 4112 D .2 I-.) A' 1l.12.10.017.7-00 ?V46 ' 0l 1 I 2024 3 2 123I 4 l 003ISA 447.7 illl tL4 01t12 22 41112 IH .U

• l-1I2 "-411171412 2 '

1V HIM1 211224 .1 A 7.A. .2*1 14 71 .4111 510 WY, (It'll 22 1Rig I.' AS M lI'!21241 17-1 ??A 2 2 M 1201

a 2

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2.

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1.

20223 0 1 220" 02.110 .4,7.4 k17.1 71"12 01 10 ln 1 021a I 411l1` U11020011 a 21 2400i 1 201" I 1G224 2 04.07 .2222 5.71 461.0 IVAl. l6t Will 23

1121, 1

WCit A W1., s

12. 11%.t2 M"210 21 01 1252241 1

23.10 22 '4.1* 441)0 01.1~ Q %I I 5 2

16)

Lin I IsI. ý triclevointO 41i:+/-oJ lItocluie.,r AM We n iiolictt rogoreltoo of O1,:0 eim itt il~ 19 IC x flow direcIoet w u onoble to his iJrenooned so i1 oI tnect "aeevtti diw-t'iin.o s wrvt oll Ifrd its citeitlot*I slt

Serial No 13-419 Docket No. 50-336, Page 11 of 43 Dom~nlon.= Dominion Nuclear Connecticut MP2 SW ModdI and Desisn'Ba*kl Aralyria MIllstone (;i.1 2 All Page 40.01 as 01113W2013

• lculatlon 12-001 Ravisloll 0 Addendum N/A A4.A 6G A&

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1.

I'.l I n M-2 6? 1161,76 1 Table 8.3.3-2 provides the total Kmisc for each pipe by sunmming up the WEKO seals installed on each pipe. Attachment C provides, the Knisc of each pipe with a WEKO seal. Table 8.3.3-2: WEKO Seal Re.sistance Summary Pipe Description (inchfded ir Pipe Section Dtui Mise K Fixed InformanI 01 IabJ 20 3 WRK() seals (jioint A 1, DCN DM2-00-0609-0.37.10.21 70.22= 0.80 98;,joii*is A 11) and All..117CN. DMI-11-0-096i1J)8-21 1 WEKO Nal uljiritM VC 6.- CN DM2-Ill-01934-98) 0122 22 5i WI3KO seals (jo~ints AZ, A3, A4. A5 (DCN 0.21+0,19+0.34+0.10+0,21+0.31= DMV12-00-0609-98/ DC MP2-12-0I 1177) and A9 145 (DCN DM2-Ol -0609-98)) 23 5 WP.K()Owals (110171 MI 40 3,4. 5 (DCN DJM2-0.21 +0J21+021+13.26+0l,26 = 1.15 01).0132-08), 7 and 9 (DUN )M2-01 -01 32.0$)) 47 2 WVEKO Seal joints.Ab arid A7 (DCN DM2.00-0.264-0.21+0.281+0,28=1=.0)3 06014-98)). KC is doubled 1t) a2ccount1 for 2 seals iii series i11 eaehjloirif; 48 5 WiiKU.166IOints MFC 1. 2(DCN 0M2 0.2l1:,W.26 t0.357W.22*sW.20 1.24 0370.98).,9. 10. andl It (IJON lM2-01-0132-08)) 40 1 WhKO Sca2.2 (J&int As orig5inall6y installe11d per1 0.36 DJCN DM2-00-0609.98 and ioplaccd by DCN WA2-04-0394-05) SI 2 WEKO seals0 (Spool 912)installed by DC. WP-0.38+1D.38=0176 11-01214 8.3.4 rFiw Orifices

Serial No 13-419 Docket No. 50-336, Page 12 of 43 "DOOM1,,10 Dominion Nuclear Connecticut All MP2 SW Model ard Dt_~n Basil l MAIe tone Utni Page. 4 1 of 6e OAtIalion 12.00!1 R ltsion 0 AC*,l.im NIA 013112013 FE 6397 WPipe 92) and FE 6389 (Pipe 91) have been modeled as a flow orifice with a bore of 4.658 which was original plant installed equipment. The flow orifice plates were changed to an Annubar and replttced per Ref 3.5.13 and 3.5,17. According to Ref 3.7.41, the annubars have a K=.6038. The flow orificein the model has been replaced with u miscellaneous K=.6038. RO-6667 (Pipe 129) and RO-6668 (Pipe 128) have a bore of "1.98 inchcs in accordance with Ref 3.7.42. The bore size in the flow model has been corrected with this value. 8.3.5 Different Pipe Diameters - Pipe Sections 64, 65, 66, 145, and 155 contained significant lengths of pipe with an inside diameter different from that of die main pipe section. In order to account for the resistance of these sections, an LID ratio has been calculated and input as a Misr ,iD in the pipe section data screen, fittings tab. LAD is essentially a K factor without including the effects of the friction. Factor. thusa it acts like a length of pipe in the model. Like a K factor, however, L/D must be corrected from the actual pipe diameter of the length of pipe to the diameter of the main pipe section. From Reference (3.4.3) the correction is as follows: K, K

• .)

(F~q2) Where; I = smaller pipe 2 = larger pipe

• d = diameter (in)

'Table 11,1.4-1 contains the detils, of the L/D calculations for these five pipe sections, Attachment C provides the lID tor each of these pipes. Table 8.3.4: LID Calculations Pipe L ID Pipe ID LID (d~ld,)4 CorC eceted

Seci11,

[fin [in] [in) I - I I -1 (lit)) 04 8.75 13.25 23.25 0.66 0,101548 6.26 65 9.9375 13.25 23;25 0:75 0.1MW48 7.11 66 9.9375 13.15 23.25 0.75 0.10548. 7.11 145 37.25 6.0)6!5 5.911 6Ji4 0.92007 5.65 155 79.25 (1.065 5.94 13.07 0,92(M7 12.02 8.3.6 Exit Losses, for Pipe Breaks: Reference.3.4.8 added an ekil loss. resistance to nodes "FromLI A",. "FromnAiB", and "FromLIC", to account

Serial No 13-419 Docket No. 50-336, Page 13 of 43 DOOmInlgo" Dominion Nuclear Connectlicut All MP2 SW Modal and DeOsign Basis Analysis Aillsions Unit 2 Page 42 of 96 Gabulalion 12-001 Rtevlslon 0 Addrnidum N/A 01S1/2D 13 for the exit losses when modeling pipe breaks in the SWS strainer backwash piping. Revision 2 or Ref 3.4.11 included this exit K=1 for pipes 120.5,. 121.5 arid 122.5. 8.3.7 Miter Bends - PROTO-FLO program does not have a standard option for a mitered bend. therefore the resistance for the 9 degree miters in pipe sections 119 tind 249 JUSL be modeled using a miscellaneous K. From Ref.3.8.1 (Crane p.A-29) a miter bend K=2PT for 0"; K=4FT for 15' where FT = 014 for 10" pipe (Ref 3.8.1, pA-26). Interpolating for 9' gives K=3.2F-r= 3.2*.014=.0448 Since there are two miters in each train, a miscellaneous K of 0.0896 will be modeled. 8.3.8 The tee which existed in pipe section 117 has been. capped off using a blind flange per Ref 3.5,36, Pipe section 117 had conlained the run portion of the tee, resulting in a K = 20fr according to Ref 3.8.1. and Ref 3.8.9. (Nute: previously, pipe section 118 contained the branch portion of the tee) In prior versions of the SW model (prior to Rev 0 of 12-001), the flow resistance of the capped tee in pipe section 117 was modeled as a welded 90' elbow instead of as a tee. This decreased the "K" factor of pipe section 117 to K = A4fT according to Ref 3.8.9. However, further review indicates only a stub of the tee remains in the branch portion of the tee, and this portion is capped while the main flow path is still through the run portion Of the tee in pipe section 117. The process fluid is not making an abrupt directional change as would be experienced in a. 90' elbow, as the run portion of the lee in pipe section 11.7 is essenlially unmodified, Therefore, to be both conservative and to more jiccunately reflect the physical piping configuration in the field, flow is rnodeled ibm a tee in pipe section 117. See Attachment C. pipe 117 to see how. this was modeled. 8.4 Eduetor L-429 - Pressure drop across eductor L-429 was assumed as I psid at 9 gpm. based on data from the manufacturer.(Penberthy). In the nortnal mode, valve 2-CL-80 is throttled to limit flow to L-429 to 18 gptn (Re' 3.2.7); the resistance attributed to L-429 is irrelevatit if L-429 receives the required flow in the normal alignment. The pressure drop across this component does not independently affect the results of this calculution. only the total pressure drop across vulve 2-CL-80 and L-429 is significant. For.the PROTO-FLO model in nortial mode. flow was balanced to 18 gpnt through valve 2-CL-80 and the valve was set at this throttled position for all cases, This method accurately simulates the actual resistance of this Set of pipe sections, including the eductor. and has no impact un the results of this calculation; 8.5 SWS Pumps - The conditions of the operating SW systcm pump curves for each case are provided below,

Serial No 13-419 Docket No. 50-336, Page 14 of 43 50OIn)am"hat Dominion Nuclear Connectlcut Al MP2 SW Mocdal anid Daolgn tlosis Analysis Caoulallon 12-001 Ravislan 0 Mihatorie W~it 7l Ad~adrl.mw NI/A Page 4131 of 01113/2013 Table 8.5: Pump Configuration Case .a k 2U 1 2a2 & 2ai3 2c & 2c I 2c2 & 26 U2 & 201 2g & 2g I 2FP2 & 2g3 6h & '6b 64: & 6dr 9 10h PSA ON OFF OFF OFF ON OFP OFF OFF ON ON ON ON ON 01.1; PUmp P51A OFI (.N ON OFF ON OFF ON OFF 01 R OFF OFF. OFL PSC OFF ON OFF OFF OFF ON OFF ON ON ON ON (FF ON Pump Charuicterisfic Curve I Of Dlrira~dd I (11A6 Dgraded 10%6 Degraded 10% Degraded 10% Degraded 10%6 Degraded I 0rk Degrarded 10%6 Degraded Nominal 10% Degidrred Nominal 101A Degraded 10.1f D~egranded 10% Degraded 8.5.1 Pump Curves - The vendor pump curve-0 (Ref 3.7,39 and 3.4.7) is used as the nominal (or design) curve as shown below. Ref. 3.4.7 created a 10% degraded curve to use as the acceptlance criteria for the In-service Testing (IMS) program tor the SW pumps. Using the same methodology (degrading on totil. pump head) additional pump curves for 3%. 5% and 7% degraded have since been created and added to the model. These pump curves ure provided below, Attaclunent D provides the PROTO-FLO pump curves reports for each curve provided in tie model. Table 8.5.1: SW Pump Curvet Nominaiel 21)10 1/0A i96q Dsjrsrdtri 194 18M.18 111(1.2 172.66 164.9 MAI1 5%6 Deg**ded 190 184.3 176.7 16I9. 1 161.5 154.85 Herd 796 Degraded 186 180.42 172A91 1511.51 101% Degrudtd 1741.6 167.4 1.53 14637 1296. Degraded 116 170.72 163.68 1561rA J49.11 143.44 Ipil ?00 40C0D 510N) "Ii Vendor curvesr arr provided Mi" all three pumps, However, the curve is from the weakest pump in ten,,& or pump hydraulic pclf.rl*nnur:e at the accident runge hais been nsed for till the pumps.

Serial No 13-419 Docket No. 50-336, Page 15 of 43 "WDomlnloi Dominion Nuclear Connecticul MP2 SW Model and Dea-Iqn Basis Analysis Millstone U-it4 2 All CGaculallon 12-tOCI Revision 0 Addemdum NWA Nurnimil 155 147 1 38 3% LUeradcd 1501.35 142.59 1333.86 5% 147.25 139,65 131.1 796 144.15 136.71 1281.34 .I10%A Dq..radoi 139.5 1.42.J 124.2 12%6. W)grsdetl 136.4 129.16 121.41 Psep 44 of 96 li020193 Flow tgprnlj 1011 70013 aiml 130 126.1 123.5 120.9 117 114.4 goo() 121 117.37 114.95

112.53

.13 109.61 10735 105.109 1111.9 106.48 1 OU10 101.7 99.44 1 iO0 I(13 99.91 97.85 95.79 92.7 90.64 121100 94 91.18 89.3 87.42 ,4,6 82.72 t19011 83 80.51 78,85 77.19 74.7 73.04 140013 72 69.F4 68.4 66.96 64.8 63,36 15000 815.2 NPSH Required Curve - The NPSH Required curve is taken from References 3,4.12 and 3.8.13 and is shown below. Attachment D provides the PROTO-FLO pump curve report for this curve. Table 8.5.2: NPSHR Curve Flow (gpni) 0 to 8000 10000 12000 14000 14400 15300 19800 NPSHR (f0 13 16 20 24.5 25.75 28 40 8.6 Valves - Attachment F provides a list of the valves in the system. For the control valves in the system, the control parameters are provided. The position of the valves for each case is depicted :in Attachment B for each Case. If the valve is a throtlle valve, the throttle position of the valve is provided for caih case the valve is relevant to in Attachment B. The Cv chat is modeled in PROTO-FLO is provided on Attachment b along with its reference and valve vendor documentation as appropriate. 8.6.1 Control valves allow the user to modify the control valve parameters (such as temperature and flow). Foe the most part, the MP2 model overrides ihis feature and instead speciries a valve position. The current exception is the valves to the cire water pump Iubrication (2-SW-130A-D). These are controlled to balance the flow in each line at 12 gpm each for all of the non-lOCA cases. For the ILOCA cases, the valve position is set based on the normal operations case and placed in override.

Serial No 13-419 Docket No. 50-336, Page 16 of 43 LAD t Dominion Nuclear Connecticut j MP2 SW Model and Ocsign 8*aui Analysis MilinlonE U1:1l 2 Page 45 of 98 Cftelu1allon 12-001" ReviS/ln 0 Addendum NIA bt,3ull13i 1.6.2 Some valves were identified as being diaphragm type valves in the DOS model (prior to Rev 0 ot" Ref. 3.8.41. PROTO-FLO does not support this valve type and all of these valves are identified as either globe valves or plug valves. The nonlenclature used for the valves does not impact the results. 8,6.3 A library of valve closure curves has been added to the model as part of the original model development in Ref 3.8.4. If a valve is to be throttled in a case, then a curve must be selected. The valve curve is identified on the Valve Data Report provided in Attaelhment E, 8,6,4 A summary of valve changes in the model as documented in Ref-3.4.l I is provided in 'able 8.6.4. ,rable 8.6.4: Summary of Previous Valve Changes Cale Rev ICN 0 tescrlplton or Change chimt Mmlde iL 2.-SW-177A, 179A, I RA, and CV chtanged lIruz nNI2-fla*on22.9.-2 i I :A replaced with Noein.-Inch 27.40 1.,30 Butterfly Vutlcs DMM0O-0318-02 r2M2.0D.D3 /9.02 InsLmals have been removed fronm cheuk vnlveu 2-SW-I I A/l and 2-Valves detiiued Dh1DM O-.t320.(r SW-t3A/B DM2-00-0321-02 Rev 2 or Rer 3.4.3 DM2-00-0253-04 EVG bypass valves 2-SW-CV changed from DM2 -00025 4.04 23 1 AfB replaced with 8,-eich 2256 to 2064 DM2-0)0-0254-t,4 Crime Nuclear butterfly valvc* OM2-00-0361-06 SW pomp dilcharge chuck valves CV changed from 0M2-01-0301-06 2.SW-IA/B3Crcpluccd with Zfff0 in 23000 0MN1.lOM.0301-06 Dorvi Water-Check vwlves t)M2.*6.0027-fl 2.SW-I 2A/RJKC/D replaced with CV changed from, Enettrhei hutteilly valvc, 32280 to 1200 SW pump tsol1aion valves 2-SW-V changed frm Rev t DM7 )34 ".08 2A and 2-SW -3A ioplaeod whith 2 11701) o 200).c AddfRendf WEIR huterfly valves A of Ref 3.4.11 DM3-00d-n43-08 2-SWrTA replaced with WVI'tR CV changed from molor aperalted butietfly valve 27700 to 21500 DM240-0098-01 2-SW-SB. 2-SW-t1C, 2-SWT7Ai Rev I DM24JO,0064-10 2-SW-9t, 2-SW-9C. 2-SW. IA, CV changed from Addenduir M2-00-rt065. I6 2-SW-3rl, 2-SW-2C replaced 26700 to 20500) ) or Ref DM2-00-0170-t:ii Aiih Weir Tricnitrit: Valves DM2-O0-f069z 1 2-SW-97B replaced with WEIR CV changed From Mo'or operalted ttilierfly valva 27 700 t) 20.M10 Re I 131M2-{M2-0198-09 2-SW-MA rupltwc with Weir CV changed from Addcndamn T"iccnlric Valve 26701) to2050X) E of Ref 3.4.) "In addition Io these chianges, [tic mnu~fact~urer. rvinodt numbher wist frawinig nunihers were revised in itievalvE input iscrecn etid the tKCN noninha was included in the intlrilniation tub of tlit pipe section.

Serial No 13-419 Docket No. 50-336, Page 17 of 43 Do minion Nuclear Connectliut MP2 SW Model and Design B3s;s Analysis Mlllsione hUn 2 Page 46 of aS Caiculation 12.001 Rowsion 0 Addendum NIA 01/3112:3 Cole Rrv DCN VIscrpiion ol'pClo uge Clujsge Made to Model" Rev 5 of DM2-It1t.0O3O-2.SW.-7ti, 249W-A, 2-SW-i01l CV chnnged from Rtf 3.4,1 I replaced with Weir Valvet. 2710)6 to 20500 8.6.5 The summer RBCCW TCVs (2-SW-8.lA/B/C) valve position is controlled during normal operations to maintain RBCCW temperatures S 85'F. Upon receipt of an LNP/SIAS or failure of the TCV controller, these valves open to a pre-determined position (set in the field during 2R21 by Ref 3.9.7 and 3.9.8). Prior to Rev 2 of Ref 3.4.11, the Sumuner RBCCW TCVs throttle position was get to ensure adequate SW flow was provided to the EDGs. Plant changes (primarily, removal of the. DG strainers and valve charnges) have made it possible to open these valves fully and still provide adequate flow to all the safety related components. During 2R21 when these valves were opened fully, it was discovered that the flow data did not match the expected flows predicted by the model (off S901) gpm). 'hliese valves have a diverter plate as showil on Ref 3.7.44 (does not include CV on drawing). Ref 3.7.45 shows the CV for these. valves as 4000. Observed data seemed to indicate the valve. CV may not he 4)00). However, discussions with the valve manufacturer indicate that the CV is 4000 with the diverter plate.installed, It hasbeen postulated that turbulence from the diverter plate may be the cause for this reduction in fiow, The flow test left the valves in the throttle position that achieved the maximum tlowrates. Revision 5 Addendum A of Ref 3.4.3 evaltated the flow test data and determined the model positions of 2-SW-8.IA/B/C (i.e. the valve position is used to empirically adjust the flow model to match the flow data). Due to changes made to the. model, these positions have been recalculated in Sectinn 11.1. 8.7 Hvat Exchangers - The SW system heat exchangers are modeled either as shell and tube units or fixed. heat loads. The RBCCW (X-18A/B/C). 'BCCW(X-17lAI1/C) and EDG (X-83AIB Air Coolers. X-53A1B Lube Oil Coolers. X-45A/B Jacket Water) heat exchanges are m.deled using the "Shell & Tube" model within PROTO-FLO. Initial mudeling was done in Ref 3.8.3 using the data contained in the heat exchanger data sheets (included in Attachment 0). The RBC"W.Heat Exchangers data was updated by Ref 3.4.5 and 3.4.23. The EDG Heat Exchangers data was updated by Ref 3A4.6 and 34.24. The TBCCW heat exchangers were updated in Ref 3.4.1.6. 8.7.1 For heat exchangers modeled as shell and tube heat exchangers in PROTO-FLO. the prtocess Jlttw 2 and process inlet temperature must he entered. This data is taken rrom either the Heat Exchanger Data Sheets (provided in Attuchment.G of this calculation) or the current accident e This term is called service flow and service Llrnlparture in IPRlI7YO-t'LO" it.is the non-Service Water Iluid that is serviced "cooled" by the system being modeled. In other words it:is RBCCW,TBCCW or coolant tie DGs:it is not Service Waler,

Serial No 13-419 Docket No. 50-336, Page 18 of 43 "*DcnthuIos-Dominion Nuclear. Connecticut 411 MF?,SW MWel end Design Banta A.*l/*l, Millstone UtOln 2 Page -17 V' 9Q3 Calsalalion 12-aOl Revislon 0 .AOendum NIA 011/5/2012 analysis documented in the calculation provided with each heat exchanger depending on which is higher. These values arc provided below. Table 8.7.1: SW Floivs and Temperatures in Shell & Tube SWS UX Process Side Pioperties lixchanger IDescription Flow InletTenmp RefarencP Lpml [IF} X-17AID/C 3280" 108.5 3.5.40 (included in AMG) Normal Ops 70263 92.5 3.8.14 (included in Ait G) X-18A/B/C [,OCA 58M(1 222 3.3.613,1.6 X.t3A 4010 134 X-53A/1 500 215 318I 5 (:included in Att G) X-45AiII fia1)) 185 8.7.2 The AC Switchgeor Room coolers and Vilal Room chillers are modeled as fixed load heat exchangers, For'fixed load heat exchangers, the only input in PROTO-FLO is the heat load listed in the heat exchanger data screen, status tab. The heat loads are shown below. Table 8.7.2: SW Flows.and Temperatures in Fixed Load SWS Heat Exchangers Heat Load (BTU/hr) Heat LOCA wlo L.NP - Normal Opentiowm Rkeerenra Faecbnmyer -Seismic COWSR -Shutdown Cases X-18i 271568'" 28X501"I X-181A 18 1046 1923342' 3.4.28 X-15M 90522' 96167"' X.-182 131153 123006 3.4.31. X-183 121536 93949 3.4.32 X-169Afll 138140 13H140

3.4.29 & 3.4.30 8.7.3 Heat Exchanger Difrerenlial Pressure - The hydraulic resistance of heat exchangers is expressed in PROTO-FLO as a known pressure drop for a 1*

i.03 MBTUilbin eotvetts to 3284.08 gpm @ 108.5'F Value is 3.500,00(1 lbm/hr converted to Pgnm @ 92.5"F. The Operating proeeute. (3.6.1 I) 1% the optimtmu RBCCW header flow between 6000 and 7300 gpm. Since the 3,500,0001bm/hr from the FiX datasheet Is witind this. range, i is appropriate: tU Aise Ibis value. Ties number iC from Rev 3 of WR13.4,28 which uonservatively.hounds the Rev 4 value of 266,534 BTU/hr " This number is frorn Rev 3 of Ref 3.4,28 which conservatively bounds:the Rev 4 value or 207,284l BTIJ/hu " X-181A has.2 coils therefore LOCA heat Ioadfor X-181A is 2/3'271,568 BITU/hr X-1" AIA has 2 cnils therefore normal ups heet load for X-181A Is 213"288,50l BTU/hr X-1811B has I coil thleefore LOCA heat toad for X-I8LB is 1/3"271,568 BTU/hr 7X-18 IB has I coil there-fore normal ops heat loatd for X-181 Bl is 1113288,501 BTU/hr

Serial No 13-419 Docket No. 50-336, Page 19 of 43 PIPDomtntafl, Dominion Nuclear Connecticut All MP2 SW Model ar*d Oesign Basis Analysis Mlllarosa Inlt) 2 Page 48 of 96 Calculaiion 12.001 Oevsldn 0 Addendum NIA lui/ai/2013 known flow rate at a specified temperature. This fixed resistance is entered under the component lab in the pipe section data screen. Switchgear Room Cooler XISIA is not modeled in this manner, see Section 8.7.5. 8.7.4 The heat exchangers are modeled cither clean or debris loaded as. assumed in the DP surveillance Imacro fouling (debris in the heat exchanger), not micro fouling (tube fouling)lisee Assumption 9.131. When the case calls for the 'clean Ap' across a heat exchanger, the design differential pressure for the heat exchanger in question is used. This informiation is entered in the Fixed Resistance box on the Components lab of a pipe section in the PROTO-FLO model, The 'clean Ap' across the SWS heat exchangers are as follows: Table 8.7.4 Heat Exchanger Clean DP Clean DP P Volurnetmnc Pipe teat Oxchanger ipsidl Flow Rell: smikin tgpmJ RECCW X-I.A 58 HICCW X-lXl 4 11800 59 J.A.10 R *tcrw X.isc 60 E11G X-83A 105 2 IUD) inO X-t3H 106 EDO X-53A ItW I 700

1. 9. 12 13DG X-53B 10.

ED1G X-45A 109 2 7041 3410 X-45H 7f 110 SWOR Ciorles X-18 1A 0I F) IR4 3.8.11 SWOW. ConlerlX.l.NI B 5 F0 115 SW0R 1ocra X-1 82 5

31) 197

.3.8.9 SWGR C.olers - 183 5 25 206

1. I I Vital Chiller X.I 69A 148 V*

.41 1'i.4.27.+ Vital Chiller.X-1619R 158

• rtCCW X. 17A 37 TBrCW X-17B

4.

4591 38 3.5A0 T1HC:W X. 17V? 39 8.7.5 X-18IA DII - The hydraulic resistance for heat exchanger X-ISIA is modeled as part of the mtisc K value in pipe section 184 instead of a 31 CleannuP is based on original dalashesi provided on p. A8 ufl RHr 3.4,27: 4.7 psi itL 41 gpm rotindedt ip it) 5 psi. This calculation revised the dnlasheet - which changed ihe clean DP to 6 psi allowed/2.597 caluhluled (1).C9). The clean value, was; kept at 5 instead uf 6 psid 6ince this maximizes nilw through the heat tixchanger l*ir the pttrp run-out cuse. Since the accident cases use addebria loaded I)P 6paid is horunded hy Ilime cases. 32 Note thl'tIBCCW tIX presmsuw drop is calculated based on 60IIF sal-waler density per Section 17.4 of Ref. 3.8.8,

Serial No 13-419 Docket No. 50-336, Page 20 of 43 10FDam,.,on Dominion Nuclear Connecticut All MP2 SW Model and Design Basis Analysis MNlstn*le l*fh Page 49 o 26. Carsilatlon 12-001 Revision 0 Addendum N/A 01131,2013 separate component resistance (see calculation below). The misc K for the debris loaded HX is 6.23 and for a clean HX is 4.37. This value is input as a misc Kfixed in the pipe section data screen, fittings tab for. Pipe 184. Pipe Section 184 contains two parallel branches to the two parts of heat exchanger X-181A. This piping has received extensive modification and therefore the equivalent resistance. coefficient of the parallel branches have been recalculated since the initial model development. For two parallel branches. the rcsistance.coetlicient of an equivalent single branch can be shown to be: K,.= K; K, ((q 3) KI~+K,+2 KK2 Where the I and 2 subscripts refer to branches I and 2, respectively, and the 'eq' subscript refers to the equivalent resistance coefficient. Piping is shown on drawing Ref 3.7,30. Branch I (upper heat exchanger) contains (K and fT values are from Ref 3.8.1): 109,75" or 2.067" ID pipe: K=ft-r(LID)=f.r(109.75/2.067)=53,IfT where ir = 0.019 K=l.0 .3 SW 90' Elbows: K=3f1{30) =1.71 3 Tec Branches: .K=3fr(60) =3.42 I HX: Debris Loaded DP 8 psid [see Assumption 9.191 & Clean DP 5 psid [Ref 3.8.111 @ 80 gpm and 70'F K=d4Ap/(0 00001799pQ2 )=(2.067)4 (8)/(0.00001799(63.8)(80)') K= 19.88 for Debris Lotded DP K= 12.43 for Clean DP Branch 2 (lower heat exchanger) contains (K and fr values are from Ref 3.8.1): 89.375" ot'2.067" ID pipe: K=0.82 3 SW 90' elbows: K=1371 I Run Tee: K=fT(20) =0,38 1 Tee Branches: K=fr(60) = 1. 14 1 HIX: K=19.88 'or Debris Loaded DP K=12.43 for Clean TP '3See Attaphinent. X Ijord lrival irsni ofIhis t/qutifon

Serial No 13-419 Docket No. 50-336, Page 21 of 43 =..,nit, Dominion Nuclear ConnecticUt All MP2 SW Mael -and Deelg Basis Anayvstr Milla*iron UDi I Page 50 1 B96 Cabilallon 12-001 Revision 0 Addnduui N/A 013132013 Summing the resistances for each branch and inserting them in Eq, 1, provides the following resistances. See Attachment C. pipe 184 for how this Kmisc. is modelcd. Table 8.7.5, Pipe 184 Resistance Coefficients KI K2 K,, IK.,,, for pilm. 184.1 Debris Loaded DP 26.01 23.93 6.23 Clean DP 18.56 16.48 4.37 8.8 Strainers - Strainers can be modeled either as a fixed (shown on the Pipe Summary Report - Attachment F) or variable resistance (shown on the Variable Resistance Data Report - Attachment F) in the pipe. 8.8.1 Service Water Strainers (L1AI/BC) are automatic self-cleaning strainers, located on the discharge header of the SW pumps, The, strainers are automatically backwashcd when the differential pressure (Ap) across the strainer reachrs 3 psid (Ref 3.6.3 and 3.7.32). The strainers arc modeled as a variable resistance in pipe sections 11-13 heading towards the service water strainer headers and as a fixed resistance on pipes 120.5, 121.5 and 122.5 heading to backwash blowdown. The: differential pressure (Ap) gauge that actuates the backwash at 3 paid is a Bartor model 288A differential pressure-indicating, switch. The accuracy at the point of switch actuation is +/-1.5% and the accuracy of repeatability is +/-0.2% (Ref 3.8.16). The full-scale reading of the instrument is 10 psid. Therefore the backwash could actuate at a differential pressure as high as 3+ 0.015 x 10 psid,+/- 0.002 x 10 psid = 3.17 psid. Conservatively the model assumes 3.5 psid as the starting point for the maximum differential pressure. Reference 3.3.2 identified that test results show that beyond 12,000 gpm, the strainer flow DP relationship is such that Ap exceeds the ;3 psid backwash initiation nominal setpoint. This technical evaluation recommended a curve based on the following data: Table 8.8.1 SW Strainer Maximum DP Flow (gpm) DP (psid) 0 3.5 8000 3.5 10000 3.55 12000 5.i 1 14000 6.96 1600M 9.09 18000 l L5 The SWS flow uses a clean SW strainer for the nonmal operations and seismic winter (pump runout) cases. A lower differential pressure conservatively maximizes pump flow for the seismic case. The normal operatiOns cast! Uses nornitnal inputs. The SWS models a constant clean 2

Serial No 13-419 Docket No. 50-336, Page 22 of 43 " oinlo. Dominion Nuclear Connecticut All MP2 SW Model and Design Basls Ana'ysts MiIsiene Unit,. Page 51 of 9i Cat;*hlatlon 1-Oti1 9evislin U Addendum NIA 00112f01:3 psid differential pressure. For volunctric flow rates of 0 to 25000 gpm across the strainer, According to Ref 3.7.43 and 3.6.12, the strainer motor will start between 2.85 to 3.15 psid. Based on its switch deadband of 5% of full scale DP (0.0 to 10 psid), the switch will shutoff.5psid lower than where it started. Thus if it starts at 2,85 psid it will stop at 2.35 psid> However, according to Ref 3.7,43, the motor will continue: to run for 5 rntinutes once the switch reaches this setpoint, Therefore, it is reasonable to assume the clean DP may be at 2 psid for nominal conditions and it is conservative for pump niout. 8.8.2 During backwash, a Ap of 50 psid at 775 gpm (pipes 120.5, 1215 and 122.5) is modeled to account for backwash losses in the strainer. No source could he found but value is consistent and similar to the strainer at MP3 (Ref 3.4.14). In addition, backwash has been placed in manual (itn-service) during the bcnchmarking done in References 3.4.9 and 3.4.10, as well as during the flow testing done ror 2R21 (Ref 3.4.3 Rev 5 Addendum A). Althuugh not directly measured, the other large components flows were within the 10% model accuracy. 8.3 Service Water Strainer to Cire Witer Pump Lubrication (L-64ABi) These strainers are located on the service water piping to tChe circulating water putmp.seals. The strainers are basket type manufactured by Mueller Steam Company. The flow model inputs a clean strainer differential pressure of t psid @ 85 gpm (Ref 3.4.4) across these strainers. 8.9 PROTO-FLO allows flow to be model into or nut of a node. Ref.34.4 modeled the NaOCI solution as I gpm flow into node L-429 (the Sodium Hypochlorite Jet Pump). The basis foa I gpm is provided in Ref 3.237 which is attached to Ref 3.4.4. 8.10 The volumetric flow rate to the Sodium Hypochlorite system is assumed to be 18 gpm (Ruf 3.2.7) in the flow model. 8.11 The volumetric flow rate to the CW pumps is assumed to be 48 gpm in the flow model (Ref 3.4.18). Reference 3.6.7 ensures that SW flow is maintained between 1.0 and 12 gpm per circulating water pump, 8.12 SW flow to the spare TBCCW heat exchanger in slip flow is < 2000 gpm (Ref 3.6.6). Instrument measurement uncertainty is not added to this value because these heat exchangers will be isolated on a LOCA or LM', The.10% model uncertainty is added to this value when determining the valve setting. With uncertainty the value used for balancing was 2222.22 gpm. 8.13 SW flow to the spare RBCCW heat exchanger in slip flow is set to < 1500 gpm (Ref 3.6.6). 1lowever, the instruments (FI-6433-6435) used to align the HX are accurate for high flow (IST pump test), According to Ref 3.9.12 these instruments have a large measurement uncertainty at low flow. Taking into

Serial No 13-419 Docket No. 50-336, Page 23 of 43 " Do1minien Dominion Nuclear Connecticut All MP2 SW Modot and Deslgn Basis Analysis Mnflslooe UaIt Page 52 of 96 Caicutfiani 12-001 PRovisioni 0 Addridum NIA 0111I2013 account the uncertainty from Assumption 9.14, a flow of 2555.56 gpm was used to determine the valve settings for the spare RBCCW heat exchanger. 9.0 A.sttsnptions 9.1 When benehmarking plant data (not. performed in Rev 0 of this ciaaulatioo), the flow instrument accuracies (Ref3.3.3) are as follows; RBCCW within +/- 150 gpm EDG within +/- 65 gpm Vital Switchgear Coolers within +/- 3 gpm (used same accuracy as FI-6923) 9.2 Flow Model Accuracies: The flow accuracy is based on benehmarking performed in Ref 3.4.9 and 3.4.10. This benchmarking resulted in the -5 ft model: correction to the pump suction level. Also from this henchmarking an additional resistance was added to the FDG lines. This resistance is no longer necessary since the duplex strainers have been. removed. Ref 3,4,9 and 3.4.10 had insufficient data. to determine the model accuracy for the switchgcar coolers. Section 1t. 1 provides a comparison of the Calculated Flows vs I-low obtained in 2R21 (References 3.9.7 and 3.9.8). This data demonstrates that the model predictions remain within the 10% of the actual test results for RBCCW, the EDGs and X 181 AB, Although no data exists for TBCCW, 10% uncertainty will also be assumed for these heat exchangers due to them being similar in size to RBCCW. However, the switchgear coolers X182 and X183 uncertainty must be increased to 15%. Since no bcnchmarking was done on X 169A/B the uncertainty will be increased from 10% to 15% to be consistent with XI82.and X183. The benchmarking performed in Ref 3.4.9 and 3.4, 10 and in 2R21 (Ref 3.4.3 R5 Addendum A) both had the strainer backwash Inservice. It has been identified that the SW strainer is not modeled properly but since the model has been benehmarked with plant data and a correction factor and 10% uncertainty have been added to account for the discrepancies, it is not necessary to cOnrect this error at this time. .9.3 A WYvKO Seal may be modeled as u thick-walled, square-edged orifice ignoring the beveled edges and assuming a constant inside diameter. This is a conservative assumption due to the sudden cross-sectional area change experienced in a thick-wailed, square edged orifice is large in comparison to the mnore gradual cross-sectional area change of a beveled edge. Additionally, by assuming the inside diameter of the WEKO seal is constant and the inside diameter is the smallest inside diameter of the seal, the WEKO seal is modeled in a configuration which is more rcstrictive to the available flow than exists in an actual WEKO seal. 9.4 For back-to-back WFKO Seals, the K factor ftr a single seal may be doubled if the WEKO dimensions are the same. This is a conservative assumption when compared to doubling the length of a single seal.

Serial No 13-419 Docket No. 50-336, Page 24 of 43 lDomnianlhm" lDominion Nuclear Connectlcut .ll MP2. SW.Miodei and Oesigt Basla Anayl9s MIIltone Ufit I Page 53 el 96 Calculation 12,,001 vlsion a Adrtendurn NIA OI/'I/2013 9.5 The inside diameter decrease catsed by a rubber sleeve located between two WEKO seals can be ignored due to conservatism in modeling a square-edged oritice instead of a beveled-edge orifice and because tile smlaller inside diameter reduces the flow resistance. 9.6 Where a pipe section contains lincd and unlited pipe or a variety of liners of varying thickness, the pipe section are assumed tolhave the inside diameter of the thickest linet (i.e. the smallest inside diameter. This ie s a coservative assumption from a deliverable flow perspective, 9.7 For piping runs that have only partially been replaced with AL6X.N (less than 75%), the entire length of pipe is still modeled as lined pipe. In previous versions of the MP2 Service Water model, the pipe was left lined if any portion of the pipe. was still lined. This was considered conservative because the material roughness of both epoxy lined and AL6XN (same as carbon steel) pipe is equal. In addition. the inside pipe diameter of AL6XN is slightly larger since the AL6XN pipe wall is thinner and unlined. 9.8 AU component pressure drops provided by Rcf 3.2.2 arc assumed to be calculated hased on 70*F salt-water density. The impact of using a single temperature will he negligible due to the small change in density of normal concentration sail water over the temnperature range of the system. 9%9 All check valves are assumed to he rully open. This is a reasonable assumption due to the high flow rates throughout.the SWS. 9.1.0 The settings f'or valves 2-SW-..40A/B/C/D (to the CW pump lubrication) are determined by modeling the valves as flow control valves (FCVs) in a. normal alignment and limiting flow to 12 gpmn through each valve. Per Ret 3.6.7. to prevent shaft scorching, a continuous flow of lube water to ilhe 'B', 'C' and VD' on-line Circulating Water Pump, at a flow normally between 10 and 1.2 gpm. must he maintained. The valves are then set. to "Override" for all LOCA alignments, this configuration accurately simulates a LOCA scenario. 9.11 All bends for the NaCL and CW Pump Seat Injection pipe sections are axsunted as RID = 5 (standard bend geometry) per Ref 3,4.4. 9.12 Negligible resistance is assumed for the rotameters present in Pipe Sections 242. 244, 247, and 248. 9.13 When the case calls for the 'debris loaded! or 'maxinium Ap' across a heat exchanger, t(le following DPs were assumed because they provide acceptable results in lhis analysis. They were then used in Ref 3,4.20 to create the Surveillance Action limit. The 'maximum Ap' across the SWS heat exchangers are listed in Table 9.13. Based on the analysis in this calculation the current debris loading limit of 10 psid (ACTION limit) fbr the A RHCCW HX will no longer be. acceptable. A review of three years of data (Rcf 3.6.13 completed surveillances) shows that the

Serial No 13-419 Docket No. 50-336, Page 25 of 43 S llorinionl" Dominion Nuclear Connecticut All MPP SW Model end Design Basis Anayvsis tblne uIit Page 54 of 95 Ca.,alian 12-001 Revision 0 Addenotwi N/A 01,wi013 ACTION limit has not been exceeded wid seldom is the ALERT limit reached before its quarterly cleaning frequency. See Section 14 for further details. Table 9.13 Heat Exchanger Debris Loaded DP fllct Exdlongerd 4 Debris Loaded @ Vnluinletric Pipe DPF [peidJ Flow Rate fgpm] Section kBCCW X-19A 8,1$ 5s RBCCW X-ISB IO INN) 59 RBCCWX-INC 10 60 EDG X-83A 4 700 105 EDG X-831 t06 EFDG X-53A 7111 107 LDG X-531J 108 1D0 X-45A 4 700 109 EDO X45B HOI Vital Chillim X(- 169A 2,4,148 Vital Chillers X-1691i 158 SWR Coolers X-J8IA 0 810 184 SWGR Coolers X-181F1 8 80 I85 SWGR Coulem X-182 8 30 197 SWGR Cnoaers X-183 8 25 .2116 9.14 According to Rcf 3912, the RBCCW flow instrumtents have an accuracy of ÷/- 1700 gpm ait 0 gpm and +1-200 gpm at 60U0 gpm. Preliminary evaluations indicate these instrunients are accurate within +/- lUg0 gpin. It is assumed the accuracy is +1055.56 gpm, which makes design input 8.13 2555.56 gpm, When the formal calculation is complete, the procedure (Ref 3.6.6) will be adjusted as neces*ary so that the analysis bounds the flow %etting plus uncertainty. 10.0 Methodology 110.1 Case Analysis Methodology 10.1.1 Maxinmm Ptunp Flow (Seisinic Winier Cases) The purpose of these cases is to determine whether the pumps are susceptible to pump run out conditions. When analyzing for maximum pump flow the following assumptions are used: I. Nominal Design Pump curves

2.

The Bay is at maximum tide level of 1.3 ft MSL..

3.

SW strainer(s) is (are) operating in the clean condition; 4, The heat exchangers are operating at their clean differential presure. SNobte lt.i in the cawsO in which ith milnisiitnn.Ap is utilized, the TBCCW heal exchangers are isolated (riot hi Itw Iflow path) and thereort no Ap is required. "S This value htas nt yel bmei clhaoge in R6t 3.4120. it will chnrige its a resullt W this revision.

Serial No 13-419 Docket No. 50-336, Page 26 of 43 0mlnft,* Dominion Nuclear Connecticut All MP2 SW Model *rnd Design Basis Ana!yise Miklons Unit 2 Pag 5e of 95 Calculallon 12-001 RevIsion: 0 Addendum NIA Ota3t'2013 11.0 Calculations 11.1 Empirical Detcrmination of RBCCW Summer TCVs Positions Based off 2R21 Flow 'rest Data As a result of model corrections made in design inputs 8.3.4 (flow orifices)i 8.5.1 (pump curves) and 8.8.1 (clean strainer differential pressure), it was decided that RBCCW Sununer TCV (2-SW-8. IA/B/C) valve positions should be recalculated. As discussed in Section 7.2.6, Rev 5 Addendum A of Ref 3.4.3 the positions of these valves were empirically deterrnined based on flows determined during a SWS flow test performed in

2R21,

,this calculation reviewed the IST pump data during the outage and determined that PSA Was operating on its 7% degraded pump curve while P5B and P5C were operating on their 3% degraded pump curve. These pump curvcs Ir provided in Design Input 8.5,1. The following inputs were applied to all three eases in order to match plant conditions during the test: ' Temperature 56F (Based on PPC data for 1l/I 112 see Attaclument M)

• Hypochlorite balanced to I kgpm (Design Input 8.10)

Circ Water luhe balanced to 48gpm (Design Input 8.11)

• All heat exchangers set to clean DP (Dewign Input 8.7.4 & 8,7.5)
• SW strainers set to clean DP (Design InputS8.8.1)

SProcess Flow to RBCCW HX set to zero (Ref 3.9.7 and 319.8 had RBCCW flow not in service on the facility being tested) The following case specific inputs were applied (See Flow Diagrams in Attachment N): Case. I i 'A' Train through 2-SW-8.lA (Ref 3,9;7)

• 7% degraded pump curve (discussed abnve)

Tide +-1.9 with correction suction modeled at -3 1ft

• XI8A flow balanced to 9326 gpm Case I a 'B' Train through 2-SW-8.lB (Ref,3.9.8) 3% degraded pump curve (discussed above)

Tide -1.825 with correction suction modeled at -6.825fl

• XI B flow balanced to 9529 gpm Case lIb B' Train through 2-SW-8..IC (Re[.3.9.8)
• 3% degraded pump curve (discussed above)
• Tide -I.825 with correction stuction modeled at -6,825fi X18C flow balanced to 9544 gpm Table 11, 1 provides the results of these runs and the model uncertainties based the flow instrument uncertairties. Table 11.2 provides the valve positions determined by these.

runs. Flow Summary Reports and Difference Reports for these cases are provided in Attaclunent 0.

Serial No 13-419 Docket No. 50-336, Page 27 of 43 "WWV n Dominion Nuclear Connecticut MP2 SW Model and Design Basis Analysis Millsltone nlit2 All Page 60 0.1 96 01/1/2013 COnculatlon 12,001 RovIsiorn 0 Addendum NIA Table 11.1 Results of Model Runs of Flow Test Configuration & Model Uncertainties A SW Ildr 'A' Train SW Flow Tu. Data L08585 +1Shypof-24 Flow Igpml cire cornlinm= 10890,5 Case lIi a 04/0D./2013 19.39 flow [pn'l 0922,27 Rlnw i'gpni] DOflferclcc Uncertiinty ift + Urc (Diff+Uu)/CIdh: Flow I SW Hdr 'W' Train SW Flow TtL Data 1093'4l8hypo+?A Flow Igpl'I circ colingl-10976 Case I I" (14/0m2013 2111: 3 0952.32 Flow a[,n1 Dilference Unccrtanity D.rf + Unc (DifI*Un*c)/tC, Hlow A IB.CCW A blK XISIA X.IlR X182 9326 900 121 73.5 5.8 9325.27 9)17.48 124.43 75.06 53.67 -(1.73 7,48 3.43 1.56

• 4.13

+/- 156 +!-65 +1-3 1/-3 4-3

• 1511.73 72.48 6.43 4.56

-7.33 - 1,62% 79f0 5, 17% 6.1J8% -13.66% 0 RBCCW B.EDO X183. 9529 920 472 9528.33 907..01 45.82 -0,67 -12.99 3.82 +/-.- 150 +-165 41.3 .150.67 -77,99 6,82 -1.58% -8.60% 14.689. B SW Ildr B RBCCW 8 EDG X583 'V' Trin~a SW Flow Tamt D0mi 1(134+18 hypo+24 9544 960 43 Flow [Ipull tire cooling= 10976 CasI( lib 04/03h12013 20:56 Flow tgpml I 71,27 i'143.13 940.1 44.64 .0iflferma', -0.87 -19,9 1.14 Unce'rlainl1 +I-151) 14-65 +/. 3 Diff i. Unse -150.87 -84,9 4.84 (Liff+lJU.)/Cale How .1.58% -9.03% 10.79% 'Table 11.2 RBCCW Summer TCV Model Posldton4,2 Valve PInilion in SWS Model 2-SW-8. I A 2-SW-R. I B 2-SW-8.1C 6K.53% 69.43% 71.52% 11.2. Design Basis WSW Case-s 4'ThIis pocliotn ill (It SWS model macebts ihe flow rate obserVCle during titt flow tests in Ref 3.0.7 and 3.9.8.

Serial No 13-419 Docket No. 50-336, Page 28 of 43 "O roDominion Nuclear Connecticut All MP2 SW Model and D"alan Basis Anatysls Millstone UnI 2 Pago st nt 96 Calculation 12-001 Revision a Addendum N/A 0113112013 The ltlowing descriptions de&rihe the rationale for each of the cases anaiwyed as well as the boundary condition settings, valve lineups and other unique feulures relevant to each casc r1on. The SW system provides cooling water to both safety-related and non-saFety related components during normal plant operation, plant shutdown and following design basis accidents and events. The heat removal requirement for each heat exchanger was determined by calculation or specified in the heat exchanger design specifications (see Section 12 for Acceptance Criteria). The minimum SW flow required to provide these heat removal requirements has been calculated for each beat exchanger and is summarized in tables provided with each case description, in the following-A total of ten (10) design basis SW system alignments are identified for analysis& I. LOCA coincident with Loss of Normal Power (LNP)

2.

LOCA without LNP

3.

LNP

4.

Normal system operation

5.

Seismic event with LNP

6.

Seismic event without LNP

7.

Loss of Instrument Air (LIA)

8.

Lo Tide - Cold Shutdown with LNP

9.

Lo Tide.- Cold Shutdown with LNP

10.

Single pump operation fbr two SDC trains in Modes 5 & 6. 11.2.1 LOCA coincident with Loss of Normal Power (LNP) The purpose of this case is to ensure that minimum required flow rates are. available to all essential components during a Design Basis Accident. T17his case is bounded by case 2 since case 2 tuses a lower intake level 43. 11.2.2 LOCA without LNP (See Ilow Diagrams in Attachment B; pgs. 1-19) The purpose of this case is to ensurethat minimum required flow rates are available to all essential components during a Design Ba.is Accident. bi these case alignment,-, both service water headers are in operatinn. The flow model analyzes each header independently. Each facility delivers flow to the essential components such its RBCCW heat exchangers, EDG heat exchangers, vital chillers and the vital AC switchgcar Mrom coolers. The SW system A train provitdes flow to the switchgear room coils XIS and X 182, while the SW system B train provides tlow to,X183. As discussed in Section 10,2.2, flow to CW pump lubrication and sodium hypochlorite system is determined by 13As described in DcuIgii Intput H.2. 1) the i niuke level is reducred by 2.5.tt whecn tte cirvutatiuaa punps tire in Ioperaling.

Serial No 13-419 Docket No. 50-336, Page 29 of 43 "#" J.dnIo Dominion Nuclear Connecticut All MP2 SW Model ann Design Fala; Anyaysle M6iqlore tloiti 1 Page 62 eis 9 Calculation 12-001 Rev&sIon 0 AddlendumNIA 01131 M013 PROTO-FLO using the prewdetermined throttlc valve setpoinrs. No breaks are assumed in thenon-scismic piping. This euse is analyzed for both surner and winter alignments. The following settings are used for the summer.alignments (all valve settings are shown in Attachment B): I, Summer RBCCW TCVs are open to their LOCA settings (see Section 11. 1).

2. Winter RBCCW TCVs (2-SW-2471246/245) are closed for the liniting RBCCW cases and open for the limiting EDG cases.

3, TBCCW heat exchangers are isolated (automatic closure of 2-SW-3.2A/B).

4. The EDG bypass valves 2-SW-231A/B are closed (upon EDG start) and the EDO HXs are aligned.

5.

One SW pmnp is aligned / in operation using the 10% degraded pump curves (Design Input 8.5.1).

6. A tide level of-I 1L7 ft is used. (Design Input 8.2.11).

7. SW strLiner backwash -is in operation and the strainers are modeled at their maximum differential pressure (Design Inputs 8.8. 1).

8.

Heat exchangers are modeled at.. their maximunm hydraulic resistance, provided in Assumption 9.13. The following settings are used for the winter alignment: I. Summer RBCCW TCVs are open to their LOCA settings.

2. Winter RBCCW TCVs (2-SW-24712461245) are closed.

3. TBCCW heat exchangers are isolated (automatic elosure.of2-SW-3.2AIB).

4. The EDG bypass valves 2-SW-231A1B are closed (upon EDGOstart) and EDG Hx are

.aligned. Since oafsitc power is available, the EDGs start but are not loaded, 5ý One SW pump is aligned I in operation using the 10% degraded pump curves (Design Input 8.5.1).

6. A tide level of-I 1.7 ft. (Design Input 8.2.11)

7. A siphon break is assumed on the high point or the EDG discharge piping it the flow rate is below 600 gpm (Design Input 8.2.10).

8.

Valves 2-SW-7B(A) and 2-SW-IOB(A) are open to align Facility 1(2) it the spare RBCCW heat exchanger (XIR1B1) for slip flow and set in accordance. with Section 10.2.2. The manual outlet valve 2-SW-9A, 2-SW-9B or 2-SW-9C is throttled in accordance with Section 10.12. 9, SW strainer back wash is in operation and L the strainers are modeled at their maximum differential pressure (Desiý,n Inputs 8.8.1).

10. Heat exchangers are modeled at their debris loaded differential pressure. provided in Assumption 9.13.

Serial No 13-419 Docket No. 50-336, Page 30 of 43 "*aoi,,0 Dominion Nuclear Connecticut MP2 SW MOM2 and Doeilgn Basis Analolys MIlIoslne Unitd 2 Ai Page M. al 026 0011012013 CaIoubilon iz.021 Revision 0 Addendum. NIA A comparison of the, predicted flow rates to the required flow rates for the safety related components are provided in the tables below. Flow Summary Reports and Differential Reports for the LOCA scemarios arc documented in Attachment F. Table 112.2-1; LOCA wNI LNP Sunmmer Operation Limiting IIBCCW Cases-Fadiity 1 Caw 2b (4),Jt511 17:0,J) CaN zI 1/(01 )/1il 17.013ý I'M w X3)18A(AP= 8p~NdT.Vra LOCA (3uminerF'aaiil/ I 115A to X-IRA i'5A tXIID (At-, ir~opi'r~w LUC3A 511nrnr FPutlity I 11A a, N-IBI Predicted ,~ Recqlmrd ftedicted Reaciriwd V,4umd / Volumetria Ntwgin Vatuametrt, w Vuuitric Nwia o22w Rate Fle~,n ow Rafe Irp'ul M'IW Rule i'lo Rule. [a0m] fcpml Igpo fgtllj Apm] (gpml oj MH/A 33541.49 M67.34 XI/92 KOA H% T77,10 (,99.329 X-lolA 117.43 1056410 X-1815 67.41 602.67 X181 45.49 1302.6 V-169A 47.62 432.46 Nuntp P51A 1=20/.917 6/02127iA 75705 117.3-1 NIA NtA NVA N/A 16615M 77 53,57 7.70 1 M3.57 637 62.39 773.67 WA96A4) 637 51).48 60 45,69 I 21.00) 1015.24

60) 45.29 M0 (1.67 67./.1 ft4o 30 /

10.44 15 23.67 45112 .38.5? 15 23.52 26.9 13.581 47,45 40.33 .26.9 13..43 NIA NIA 10090.90 9091A.6 N/A WtA Caý 222 2/4A1)4/)3 17:76) C= WnS (01104113 37:35i P50 16t X H1 A (At'. prd T 0A'M P5 lo X I18B (AP = I0 poK T. 90l'H Irj32CA *siirnna. '.ility I P511 21 X. I KA LOCA Summer I'inllity I P58. XX-I 9il Comnponent p/i7ld Rcqkdred Predicted Reau/ited 'Vohulocir1t "V/ VJtlrtlcffc Alialgiri V"a umir A*n¢illeltll Vodlumctkiý Marg '*in] ow Rat unmly Fluwv Raic lgprl Fl'ow NIew tIa-lw Raf [jgiil t-p]A 9/pi1 /It. l lar3pm1 %piJ 19p90 XI8A 6505.75 7655.12 7570 1,15. 1 N/A N/A X 6181 N/A NIA 8B.".5. 77207.07 7570 150.67 1200 A t(A 77.1.4 6R6.46 637 .,4/Af' 770.57 (.93.5/1 &17 56.51 X./Il/A 1.6.94 3 U5.25 6IM 45.25 2/6.49 1/04.4 60 44.64 .Xh I IS W. 12 60,,1 M2 301 6.41 67 60.6IR 10 30.18 X-IK2 45.30 73.51 1* 3 3 3/%

45.

.16,31KI I 23,,36 .'-22/36A 37.42 40.31 26.9 0,41 47.25 40.106 26.9 13.26 Pump M 0 IS

3) tn 32067M.Y I2I NIA NWA lat'367,16 9.060.44 NI2A NWA

Serial No 13-419 Docket No. 50-336, Page 31 of 43 !P' .m0i" Dominion Nuclear Connecticut All MP2 SW Model ard Dosign Basis Analysis Millstone Uail i Pago 64 01 96 Galculallon 12-001 Revision 0 Addwiidum N/A 01/31/2013 Table 11.2.2-2: LOCA w/o LNP Summer Operation Limiting EDO Cases-Facility I Case 2W2i (04//111-17*43) P5U to X[5A (AP1' 8 pdid T= 80*17 LOCA Suntanr Fi P51 io WIAA Min Eon* Cemponesi Poetlicted Rteuired Vsilaheiric fl(ow w/ uncrilaimy Vlumicic Flow Margin Rate. (gpi] Wgpall 1wa.e (4prl XIIA 9063.49 8iS7.lJ 7570 587.14 XlIl. NIA N/A I.[X; A Hx 747.08 6723.37 637 35.37 '-l11A. 11330 101.97 60 11.97 X-114111 651,4 58.54 30 26.34 X-I-E2 43.99 37.31 15 22.31 X-16QA 45.95 39.116 26.9 12.16 Pump PlS1 10 11.43 9460.29 N/A NIA Tuble 1112.2-3: LOCA wio LNP Summer Operation Limiting RBCCW Cases - Facility 2 Cawlic (IM1 I/I.Ol0) Case 2le (148/4 IM: 161 POCloXIBC PSC Iol xlI LOCA Suntnler.Facility 2 PSC-X18C I.OCA Satlenr Frcilly2 P5C-X Ilill Predicted R R.aitmed POrediced RL9;ltircil Volumait,. Vonrli.'. Marli. V.!wmld W/ VWltmneeric Margin Flow Iate aecetty lt ale [m a Ilwatae untaitly Iflaw Rate Ipml Lapml [gpm] [wIolnJ lpel isi'll P ,n.pro1 XISB NIA NIA 9746J 35 7671.72 757/0,o NIA X18C 8619.53 7757,5B 7570 89738-A NIA N/A LUG U IIa 7813.67 702.60 631" 55,6) 774.96 69746 6W7 60,4f, x 1W3 33,90 33.07 17 16.07 18,64 32.8,4 17 15.74 .'-169B 48.22 40.99 26.9 1.+09 47.90 40.72 26,9 13.82 Pump p17C 0948,63 9953,23 N/A N/4sA (61W5.R6 9059.27 N/A NWA Case 2c2 (04/l/I3 18:27) Coaw 2s9 (04 I 11:37) P51Ito XIO W, F51amoXIMf LOCA SItamler Facility 2 P5C-Xi 9iC I.LICA Sumrstr Fci/lity 2 P51-XI 181 Cornotelnl" Ptl;dlr*dJ lIoulrd P'redictle wl Relqtzi %'olanen/c Volumetric Margin VolsassetliY VYohtticileu "tigli Flow kle iTl ncetmoly w lgnce fm] lnlyq loe Kale 11181 P low Rate jpit] lFlosW RFle Kae I l-,,,n] [-plil IEPrI W alplt [Prof lpi ,I\\l8 N/A N/A 1718.146 7341.61 7370 N/A X1C 6S9.32 7734.lYO 7,5711 164,89 NIA N/A UDG h 1113 778.39 70034 637 63.34 772.47 695,22 637 58.22 X-i1)3 30.79 3.97 17 15.97 38.52 32.74 17 15.74 x.- I/1 'lOIRK .01)7 26 1t.97 47.7.% 403M9 26.1 13.61) PsIt., 8P5 931)(19.02 11911,.62 NIA NIA 1(9)19.()1 9017.10 N/A NIA

Serial No 13-419 Docket No. 50-336, Page 32 of 43

0Fbaminien, Dominion Nuclear Connecticut All MP2 SW Model and Design Basls Analysis Mitlslone LVhdhI Page 65 of 96 Calculalion 12-001 Revision 0 Addendum W/A 0V3112013 Table 11.2.2-4

LOCA win LNP Summer Operation Limiting EDC CUses-, Fucility 2 Cse 2¢vL [/.104/iNiJ I,:171 P52131 RXlI IJ3CA Su nswrlki'ciliiy 2 P3U.-C ICClr Min EI00 I-'ejw Cnmpnaeni; 1'r tfedlua R* equired Volunelric Flow w/.onCrtauty Vluimlo How Nlargi Ra o

Rl~pmlm]01 gae ln X liIi) N/A NIA x I C 91,10.70 8Z66.63 7570 (156,63 UDG H llx 753.02 677,71 637 10.71 X-1923 37.03 31.99 17 14.99 X-169D 41,4/6 39,66 21.76 Ptruap P5I2 IN14 i 1.4n, 9370.16 N/A NIA Tnble 11.2.2-5: LOCA wNo LNP Winier Operation - Facilty.1 Case.e 121.91413 19:1)3 Ci;houel (04/0.1/13 19:3.) PIA to X 18A P5 A to X*IC LDCA Wturer Faolilt, I PSA-XIlNA w'X-t1L2 Silp LLCA Winteri 'silily I PSA-I 9F1R ./ X-18ASlip C :1J rt nIt iPll PredIkted w/ Required Pirdisnd W/ Requited VllaI',*nli .i Volutmtiriu Mwgin Volumeic U. Volumeme Msrgin nlow R=s CCI1002y Howm Rate pvpel FR,,. Rnl, "o"Y wu Rule. jpnj XI2A 7145.15 6430Ar 5,1230111 10027.64 2699.16 2.429.24. 2305.362 12'l.24 X.I28 3346.19 .1011.57 7.1011,00 71t..7 7154.53 6799.09 5423.011 137.012 R110 A Hs 538.901 465.35 337 )48.05 559.I1 501.20 337 166.20 K-131A I213.25 92.93 60 32.93 105,41 94,87 60 34.87 X-181"4 3q.27 53.34 70 2134 l0.51 5.4A& it) 23.46 X. 182 40.11 347.410 25 I YA,1KI 4.192 74.71 Is 19.71 X-169A 42.26 3.5.92 26.9 9,102 4.142 316.0 26.6 9,0 Puomp iA 1 0667,26 10500.5.1 N/A NIA 11459,02 103.13.12 NIA X/A Cus:, 2c2 1)0(/N13 1941) Case 2703 (0/00413 19:51) P51=3w XIl A P5R1 t XISB LOCA Wi.wr F.witity 1. PV52-XISA.wl X-11S5 Slip .OCA Wlitser laviilY I P.5&-XISR2wl X-i9A51li, CNrPnenp Pr edicted ws Required Prediuted i ve Reqtuied VotUIuRlrlc. owetrlni Volu:merie Mrin Volumetric Vnluielri, Mmagrie Plow Raum y RFow Rune [spi) Flow Rate Flow aeo fpin) [(ePn.J 2pm1 lutrnl Igpn,] )gpno1 111A 710W.61 639775 5.123.X) 974.75 2605.92 2417-33 230000 127.33 X118B 7329.21 2996.20 2300.910 696.20 7527,41 6769.07 5423.00 1342.67 EDG A I, 534.351 9.0.4! 337 2".11) 554."5 499.39 337 162.35 X+J1iA 102.32 92.25 1it) 31.-5 12)4.q !94.411 60 3.1.40 x-281B 58.96 533.06 30 23.10k 6.71 54 I3 30 24.19. X-I22 391.901 .33.81 15 I31.93 40,60 14.M 15 19.54 X. 169A 42.2U 3.1.73 26.9 8.83 42.61 36.22 20.9 9:12 ringP*g i n 216.,76 90264.04 N/A N/A 1141,9970 101277,73 N/A N/A

Serial No 13-419 Docket No. 50-336, Page 33 of 43 Dominion Nuclear Connecticut MP2 SW Model and Designi Basis Analysis Millstone Unie Calculation 12-001 Hevfsasn 0 Addendum NIA All Page 85 of 96 0113112013 Table 1,12.2-6: LOCA w/o LNP Winter Operation - Facility 2 Cas 29 (g1104IL13 21). PSC Ito X IAc 1.2CA Winter F.10ily 2 CC tI XIC whitl X-188 Slip Cavs.gj (1/041l3 2111201 C2C ao X1 uB LUCEA Winter Facility 2 P512 to X18D with X,18C Slip. creipascill YINt ~l X-16911 Pumnp PiC 1INllcled Veilameoric now Raite 31,59932 7402.96 4"52 118128.23 30753,4 1,662.66 29.17 36511 RLquired Volumeioc I-low Rate 21:101) 2,1111 337 17.K) 26.1" .NIA Martin 723.04 1611.90 1 7.0D 11.61 NIA Predicted Volunctric Flow Rate I1pje! 7712.5. 2074.62 575.4.5 314.7h 43.16 1 1466.99 itneerta ly ([pm] 0'14127 2407., f. 317.905 29.56 .36,69 10320.28 1{equire-d Volumiraic I*iw Riate tsp'"] 5423 1*011 337 26.4 N/A Martn [*pml 13!R-121 1072.16 180.91 h.756 9.7U3 X/A Caw; 292 1({4614 3 21135) Case 2.' (MlAlK13 2DA'I2) FSl io XiSC 1511Io X IalII LOCA Winlerl'acility2 P.O w X 18C wilh X-18C Slip I WTA WIniI" 'acilIly 2 PS5Rto X51911 wii X-I. Ia Slip Cmponpen' Predlcted Rcquired Volaul ise or wte Valeunailir Marlin Hlow Rale Flow kale Ippo] [Rpm!* I gi"rl~ 5161 7712,i68 6941,A1 5421 1511.41 5111C 2613.613 2370,27 2*81 711.27 6D20a 1x 512.25 515,1135 337 178,03 X-183 34.64 29.A4 l.0 12.44 X-16911 4.00 36X55 2W9 9.65 P op MIR 11408.14 1026733 NKtA NMA I.'rdiciOe Rcqlirtd VlSmn.ic W'. Volumer h Flow Ra ancrilily Fow K1.1 [Sn] [~poml rep".] ?1290.013 293.203 21f1) 73214.2 &.195.94 54.92 5f0.15 495.135 .337 33.83 28.76 17.00 42.34 35.99 28O I 041.6s 10177.51 NIA Margiu' fd2.11231 1172.94 158.14 11.7K .9.)14 NiA 11.2.3 Loss of Normal Power Case 1 bound. this case (which is hounded by case 2). For the loss of normal power case; the SW flow rates to the EDG heat exchangers and RBCCW heat exchangers require much lower flow than for a LOCA with LNP. The required flow rates are desriTbed in Section 12. Therefore no additional analyses are required for this case. 11.2.4 Normal Operation (See Flow Diagram in Attachtment B, pg. 19) The intent of Ihis case ik. to show that. all components receive at. least the minimum required flow during normal operation. Two trains are normally in operation for both the suomner and winter alignments. One SW pu1mfp for eac-h train is in operation. During normal operation SW flow in each train is delivered to the TBCCW heat exchanger, RBCCW hIcat exchanger, EDG heat exchanger assembly, switchgear room cooling coils and t[l: vital chillers. SerVice water valves to the CW pumps for seal lubrication and the Sodium Hypochlorite injection are set.as determined in Section 10;2.2. RBCCW flow is maximized for 8F by closing EDGY bypass valves and adjusting the SW-5s toobtain the rain TBCCW flow.

Serial No 13-419 Docket No. 50-336, Page 34 of 43 .Daminton Dominion Nuclear Connecticut All MP2 SW MC~del and Oefql Basis Analysis Mitxoere Upic 2 Page G7 ot 96 Caklul~aon 12t.tD Ravitton 0 Mdapndum WA 01131,2013 Reference 3.6.4 requires the EDG bypass valves to be closed when the SW temperature im greater than 650F. Normal Operation for the winter alignment is not included as design basis. case sihce this case is bounded by the sumnner nligtnsient cases with respect to minimum flow requirements and bounded by the seismic event-winter slipstream cases with -respect to maximum pump flow. However, winter cases were run to balance flows. as discussed in Section 10.2.2. .Nonrvital chillers are in operation during, normal operation. Therefore the flow requirements are not specified for the X-169A1B he-al exchangers (the vital chillers). The following settings are used: I. Summer RBCCW TCVs (2-SW-8.IA/B/C). are throttled to their ILOCA settings (see Section 11.1).

2.

Winter RBCCW TCVs (2--,W-247/24(-245) are closed,

3.

TBCCW valves 2-.SW-5A/C are open to 40.98% 1 41;07%* in order to achieve tloe minimum 4316 gpm (with uncertainty 4795.gpm) through the TBCCW heat exchanger required by Section 12,

4.

The EDG bypass valves 2-SW-23 IA/B are closed. 5, Two SW pumps are aligned /in operation using the nominlal.pump.curves (Design Input 8.5.1). 65 A tide level of 0 ft MSL is used (modeled as -7.5 ft MSL ýintake level). (Design Input 8.2.11)

7.

SW strainer backwash is in operation and the strainers are modeled clean (Design Input 8.8.1). K. I-leat exchangers are modeled dean.. and the input DP values are provided in Design lnputs 8.7.4 and 8.7.5. 9, Vital Chiller valves 2-SW-I I I and -113 are closed. Flow Summary Report% for the normal nperation scenarios are documented in case 4b presented in.Attachment F. Table 11.2.4-1, Normal Summer Operation w/o LNP Caw,c 4h (0'tM1N2013 415) M05 Mn XIS, NMxaud Op*i vsuntrwr Fa'hly I & 2 -%4ix RtCCW Flow Compone.ru ReouMd Predhtnd Re Iiitw vetlkt iel ] Flow

• 0*2 V,. ulis.rnaicIIv Vlu~ncact* Ikiw Ma~in Rt twa!

"XIMA U0.1 75T9.16 I(OI' 29.16 xl.;' 854,32. 76M.99 9str' 38,99 KItA .l795.27 4315.7-i 4316 hSAtnted ".4 Rel'r it Scclion P12.I Acueplace Criteria is Retfr to Section 12.0 Aeccpitnr. Crilt:rha

Serial No 13-419 Docket No. 50-336, Page 35 of 43 lvP2 SW Mfldal 1nn6 Design Beale Annly3si: CaVculathw! 12-001 X17C W10K A liD F.DG 33 Hx. X-PC 1 A X-182H 5182 X, 16911 I-Amp MS

Pump, 1 PSI Dominion Nuclear, Connecticut All Revlsioq U 47)53,6 13 0o 1.1.131

'01855 591.45 42.51 35.92 4*174 '1.94 1.7241.311 13174.25 4315.73 11.78 1,82 89.70 511A3 4.2n 12367.63 12372.53 M115wene 11,1112 Addendum N/A 43M6 NIA N./A 76 13 15 17 NIA NI/A W/A NIA Page 88 of 98 0113112M11 N/A N4/A NIA 11.2.5 Seismic Event with LNP (See Flow Diagram in Attachment B. pg. 24) TBCCW isolation valves 2-SW-3.2A/B close on a LNP signal. Therefore a break in TBCCW line is automatically isolated. Case 6 'Seismic Event without LNP' bounds this ease. III Case 6, the Vital Chillers do not automatically start because they have not received either an LNP or SIAS signal. Cast 5 is run to validate that the Vital Chillers receive adequate flow, Table 11.2.5-1: Seismic Event Summer Operation w/ LNP Cawc 5 (04A4/13 14:44) PSA & MC Sicie Smi nrner will, i.NP XI RA Xl SC E100 A 1 1211 BIIx I SI1A X-18 in x-10a K-IS.! X-I 69A K-1I MU 1'.,ml Y5A Nmp Me Predlav' aane thpn] 8M132 15513],23 75j;29 754.41 M1.91 61.31 3,11.7t SlI 87 40.74 11.103 I II)4 Reqtbircdi Yohimelrie l1' ~I~ix84Ii~y15,w Raw.C Margin S8137.51 9 1 M.91. 679.76 1;?8A1 ob, Is 35.1 K M5.17 29.14 'U.03 34.88 r,035.7 92727.7,1 04)0 276 276 26 13 is 17 26,9 -22.9 NWA N/A 2887.-1 311.91 403,76 10 297, 710.13 42.18 20,17 12,54 7.73. 3.98:. N/I. 11.2.6 Seismic Event without LNP (See now Diagrams in Attachment B, pgs.25-27) The purpose of this case is Lu ptedict the flow rates to the safety related components as well as maximum pump output for NPSH concerns, assuming breaks in the seismic/non-seismie boundary interfaces. See Design Inputs 8.2.4 - 8.2,10 for the description of how these breaks are modeled. Titles iuust be added in order to print out these break flows on1 the flow suimimary reports generated by PROTO-FLO. The analysis is performed for both

Serial No 13-419 Docket No. 50-336, Page 36 of 43 Dominion Nuclear Connecticul A11 MP2 SW Modael an Design Basis Analysis Millsione Utl3 "I. Page 69 6l g9 Calculation 12-001 RvIslson 0 Addendum N/A 01131/20i3 summer and winter operation. For all cases.analyzcd it is assumed that the thrnltlIU settings are maintained the set in accordance with Section 10.2, Summer Operation The purpose of the seismic case for summer operation is to predict the mininsiw, flow rettes to the swirthgwar rnoou coolers46. The following settings are used. for the summer alignment: I Summer RBCCW TCVs (2ýSWý8. I A/B/C) are open to their LOCA settitigs (see Section 1.1l). 2.. Winter RBCCW TCVs (2ýSW-245-247) are assumed to fail open which mnximizes flow diverted away from the switclhgear coolers.

3.

TBCCW isolation valves 2-SW-3.2A/B fail/remain as is (full open) since neither SIAS nor LNP ccur.

4.

TBCCW valves 2-SW-5A/C are set in accordance with Section 10.2.2, 5, The EDG bypass valves 2-SW-231A1B are modeled closed. Reference 3.6.4 and 3.6.5 requires these valves to he closed when service water temperature is above 65°F. 6, Two SW pumps arc aligned / in operation using the 110% degraded pump curves. (Design Input 8.5)

7.

A tide level of -11.7 ft is used. (Design Input 8.2.11 ).

8.

The following pipe breaks are postulated (Design Inputs 8,2.4 to 8.2. 10): Upstream Pipe Node nldctitrying. Steatio .Component CW t(u1p 228 ToPDI-B 2"SW4l30A Lubricsling IiIJe.9 239 ToPDI-A 2-S'W-131ý11 34 XI7Ain X-17A }UtCCW tIIXs 36 XI7Cin X. 17C: Sodium 129 R.66(7tut RU-6667 HypoChlorite tlnjL,"cion 128 R0i6668Ouit NO-668. SWGR Coolers 189,5 SW192Oul Dummy VivI Vital Chillers 171.5 SW 197Out Dummy Vlv2

9.

SW strainer back wash is in operation and the strainers are modeled at their maximrnum differential pressure (Design Input 8.8. I). "The limiting case. for the West,480V Swilchgettr Room Cooltrs (X.18tA1/B) is now the winter LOCA case since there is now a higher flow oaquitemnent during: I.OCA.oha.,n during normal operations (heat toad rot seismic is. m'mluivatlot to normal ops),

Serial No 13-419 Docket No. 50-336, Page 37 of 43 Dominion Nuclear Connecticut All MP2 SW Modlo atnd Desigrn Basis Analyss wistone Ut1h a Page ?0 of 96 Caloulation 12-001 Revision 0 Addendum NIA DIA1/a2o'11

10.

Heat exchungera are mudeled at their dtbris loaded differential pressure. provided in Assumption 9.13. Winter Operation A winter minimum flow case (degraded pumps, lower tide, debris loaded heat exchangers) provides flows that are lower than the summer minirmum flow case.. However, because the required flows"7 are less due to the lower process fluid temperatures, there is additional margin available before the mrintimum flow limit is reached. Therefore, the stumner minimum flow case is used as the limiting case for the Switchgear Room Coolers antd bounds this case. The winter maximum flow case is analyzed to determine whether the pumps are susceptible to pump run out conditions. The results show winter: alignment is more bounding for pump run out conditions than the summer alignment due to slipstream flow through the idle RBCCW and 'T'CCW heat exchangers. Cases are run tao Facility I and 2 with each using the swing heat exchangers for slipstream flow. The following settings are used for the winter alignment:

1.

Summer RBCCW TCVs (2-SW-S.A/B/C) are open to their LOCA settings (see Section 10.2.1).

2.

Winter RBCCW TCVs (2-SW-247/246t245) are assumed to fail open.

3.

TBCCW isolation valves 2-SW-3.2A/B fail/remain as is (full open) wince neither SIAS nor LNP occur.

4.

TBCCW valves 2-SW-5A/B/C are set in accordance with Section 10.2.2.

5.

Valves 2-SW-7B(A) and 2-SW-IOB(A) are open to align Facility 1(2) to the idle RBCCW heat exchanger (XI 8B). The manual outlet valves are set in accordance with Section 10.2.2.

6.

The EDG bypass valves 2-SW-231A/B are assumed full open since neither a: SIAS nor LNP has occurred.

7.

Two SW pumps are aligned I in operation using the design pump curves. (Design Input 8.5.1)

8.

A tide level of +1.3 ft is assumed (modeled as -6.2). (Design Input 8.2.11)

9.

No siphon break is modeled in the EDG piping (this would reduce flow),

10.

The following pipe breaks are postulated (Design Inputs 8.2.4 to 8.2.10): Upstream Pipe Node Identifying Section Component CW Pump 228 ToPDI-It 2-SW-130A Lubricating 239 ToPDI-A 2-SW-130B "Current unuayso does not have it roquiret flow lbr winter caseq (60TF), however Ref 1.2.4 shows that req~uired tbows tliulpped hya ii.wtm 50% wheiii fenqwrlutreI1t

k. tLtwete rwuitt 751F to WE~

Serial No 13-419 Docket No. 50-336, Page 38 of 43 "DoMnIl0 Dominion Nuclear Connecticut All MP2 SW Moda and Cesign Basis Analysis Mll!siona Inau 2 a.e 71 El96 Cak-ulalion 12-001 Revision 0 Addondum N/A 01,31/2013 Lines. 34 X]7Ain X-17A I'Bcx.'W [I (X, 35 X I7B in X-17B 36

X170i, X-17C Sodium 129 RO6667Oue RO-6667 HypoChlrite Iniection 128 RO6668Out RO-6668 II.

SW strainer Nack wash is in operoion and the strainers are modeled clean (Design Input 8.8,1). .12. Heat exchangers are modeled clean to minimize the pressure drop, provided in Design Input 8.7.4 and 87.5. Flow Summary Reports for the seismic scenarios ure documented in.Attachmen( F.. Table 11.2.6-11: Seismic Event Summer Operation w/o LNP' Caw 6b 104A/1l1 14:'-29V P5 A& PIC sam1..~e imbmlerl Sliln**IIdo'iCIt w unvenssry11l sNn Ftwdki:t.'l RcqU11kd V'dusneric Haw o/1snCEAlf' Volumetric Flo Mg.-,in Rael IgpaI [spmj 5WW. (tpnml [ gpn~l1 x 18A 7412.118$ 60W.68/1 5150 3580,19 XISC 75401.34 6746b.3 5150 1703.Jl XI7A Hre*k 4241.50 NIA N/A NA XI7C Hmreek 4257.09 NIA N/A N/A FlG A RIi 1235 3 1j 15 N/A M/A ECCI B h 1231I. 11.142 NIA NIA A I*ypulre*"k 124)77 N/A N/A N/A B I lbVU w':k 32...92 N/A NIA N/A A CW Seal (n'ri 83.11 N/A N/A NVA I9 CW scuI RpmaA 78.32 NIA NIA N/A X-.IRA 84.32 7.8.99 26 49.89 X-I91l3 49.,30 43.56 i 3 30.5h X,112 72.07 V.77 15 12.77 X-18. 27.55 23.42 17 6.42 X-19A 1.9t) 3.32 NIA N/A X-.16311 3,8 3.301 N/A N/A Purnpl'SA 12555,6"1 I1M.10WI) N/A A Pinmp Pic 12572.67 11315.411 NWA NIA

Serial No 13-419 Docket No. 50-336, Page 39 of 43 J D ominion Dominion Nuclear Connecticut MP2 SW Mdel and Dosagn Bes9SAnelysis Millstone Unit 2 Page 72 af as Calculatlon 12-001 Revlson C Addendum N/A 01/31/2013 Table 11.2.6-2: Seismic Event Winter Operation (Slip Stream Flows) CAc6& i,*l&ib.Oi3 13;05) Cite 6J (QI/01013 1131) OA & PSC PMA&IP.¶C Sei anic Winier F.i/lly I 'ith X 10 Slip Seismic Wirlet Fa/il y I v/Sh XlJS Slip LAIIIIC Pl 4 taeLd wicq Requited Pr'dicmd w/ Required

• Volumetric W:.in Vuilvlucr:e M'gie VeilieClric Une,ly Vhlummerin Margin
• Haow Kale ls, Kae Ipiml Flow Rate How ItRmv. j*le

[gpm]

apr, laval larval jm]

iwml XIXA 6160.84 6359.7M XI1H. 2516919 2518.34 xISc 7123.56 6361.3B XI17A (544ir 43954 B l~reak X1 7 413.66 536!.16 Brleak SOIA H* 1.28 A H.3K (5J)G 1H.* 1 3327 II A. 1710.84 0r26.24 E10 0 11839,94) 1666.76 Iiypash Alipo 3101.113 2WOA5 Break B Hyp 2'83.62 Z21.*1 Break ACWSeAI 76.72 71.11 Break B CW S 66.71-6. X,-I8IA "/i. 74.43 X-18iH. 46.39 44.90 X.182 33.17 32,11) X-103 30, I 27,XV X-lI/IA 4.95 4.98 X-1013 5.41 4.96 Puirp PSA )5S92.64 17151.90 17527 373.10 15655.09 1722(1,59 17527 "316.41 Putmp PSC 15163,72 I6lrfl.4iA 1752.7 846.91 15614M9 17170i.1 I P7.M7 M0.69 11.2.7 Loss of Instrument Air Case 1 bounds this case. The SW supply inlet valves to the TBCCW heat exchaugers are-designed to "Fail as is" on loss of air and have a back-up air accumulator. to close valves for SIAS or LNP, ilfinslrument air is lost, The flow alignment is the same as the LOCA with LNP alignment. Therefore no additional analysis is required for this ease. 11.2.8 Lo-Tide Cold Shutdown Operation w/o LNP This case is similar to case 9 and bound this one because ihe TBCCW heat exchangers are still inservice and the EDG heat exchangers are required for LNP. 11.2,9 La-Tide Cold Shutdown Operation wi LNP (See flow Diagram in Attacnhment B, pg, 28)

Serial No 13-419 Docket No. 50-336, Page 40 of 43 '*DnmtuIn Dominion Nuclear Connecticut All MP2 SW Modal and DeOaLn Basis Anal,.*i Millstone titit 2 PFafa 73 nf.96 ,atculation 12,001 Revision 0 Audonduni NIA 010*311013 The purpose of this case is to ensure minimum SW flow to the RBCCW heat exchtngers is provided for safe shutdown of the plant assuming the maximum setdown condition during a hurricane; tide = -7 ft MSL (FSAR Section 2.5 and 9.7). In order to perform a controlled cool down to cold shutdown within the "rech Spec action statement limit of 30 hours, the minimum SW flow rate must be at least 6400 gpm to the RBCCW heat exchanger (Reference 3.4.33). This case is different from the normal operation case becaue this case assumeg that the RCS is cooled with reduced RBCCW flow of 3500 gpm (Reference 3.4.33). Although TBCCW cooling can be isolated in the shutdown mode of operation, conservatively, reduced SW flow to the TBCCW heat exchangers. is assumed. The winter cases are not analyzed biince theservice water inlet temperature is below 6(0F which will improve the cooldown rate. Since the EDO bypass valves arc not allowed to be open when the SW temperature is greater than 65°F, these valves are modeled closed. A siphon break is modeled in the EDG discharge piping since the flow to the EDG heat exchangers is less than the critical flow for maintaining the siphon. The following settings are used for the summer alignment: I Sunmer RBCCW TCVM (2-SW-8.IA/B/C) are to their LOCA settings (see Section 11.1).

2.

Winter RBCCW TCVs (2-SW-247/246/245) are closed,

3.

A tide level of -7 ft MSL is used (modeled as-12 ft. MSL). (Design Input 8.2.11) ,4, Two pumps are aligned / in ope'ation using 10% Degraded pump curves. (Design: Input 8.5, 1)

5.

T"BCCW isolation valves 2-SW-5AIHIC are throttled for low load operation aW 25%. 6.. SW strainei back wash is in operation and the strainers are modeled at their maximum differential pressure (Design Input 8,8 1). 7.. Heat exchangers are modeled debris loaded. (Assumption 9.1.3)

8.

Vital Chillers tre notin-service due to the lack of SIAS or LNP (Ref 3.2.6). Case 9 show.q that the predicted SW flow rates corrected for uncertainties arm greater than the required flow rates. The Flow Suwm,'ary Report for the Lo-Tide cold shutdown case is documented in Attachment F.

Serial No 13-419 Docket No. 50-336, Page 41 of 43 V ventl"am6, Dominion Nuclear Connecticut MP2 SW Modal and Design Bas% Analysis Millstone Unli I Calculallon 12-001 Revlsiorp 0 Addanaum N/A Table 111.2.9.1: Lo-Tide Cold Shutdown Summer Operation w/o LNP P3A & PWC. All Page 74 of196 01131/2013 1.n-tsi' COW ShtiuOtW1 OPeNIIDA PAUNEYIy & 2 Componcent Flow 1010. Igpinl nowml XI SA X lt. .X1?A X1 7C* EDt' A Ho 1100 0 fly X-IRIA X-18111 X-12 X..143 X-169A. X-16914 Pump1 PM IPomp I'Sc 3726.55 71177.OJ X130.99 346.911 567.93 106.56 61.17 472.73 ,13.11 117314.74 11245.68 b.955.701 2097.91) 4192.15 5 11.14 1.410a 6.1161 NIA NIA .176. 276 N/A NIA 235.14 42.00 12.61 9.42 91.3 INIA NI/A. 29.61 ,136.,I 17 !6i9 I1.5 NIA NIA iflI n.*1R7 101121,11 11.2.10 Single Pump Summer Operation to cool two SDC trains (See Flow Diagram in Attachment B pgs.29-30) The.purpose of this case is to assure that the two operable RBCCW heat exchangers will receive adequate flow rates from a single operable SW train to cool two shutdown cooling trains in Modes 5 and 6. This enables increased inspection lime for the inoperable SW hceader. One train is in operation for both summer and winter allgnmenib. The summer case is bounding and therefore the winter case is not analyzed. TBccw is isolated and the EDG heat exchangers and Vital Chillcrs48 are aligned on an LNP. A siphon break is assumed for all runs at the high.point of the EDO discharge line since the flow to the EDG HlXs is less than the critical flow for maintaining the siphon. The following settings arc used for this alignment:

1. Summer RBCC.W TCVs (2-SW-1.IA/R/C) are to their LOCA setting.i (see Section 10.2. 1).

"2. Winter RI3CCW TCVs (2-SW-247/246/245) are closed, ' ýAcv2 Ihru 4 l' Ref3,4,3 only hod the hyplss valves (2-SW-125/127) op'.n, this is. oI consisllnt wilti II.NP liniupf herafomr Rev 5 ofl Raf 3.4.3 has 2-SW-Ill and -113 open.

Serial No 13-419 Docket No. 50-336, Page 42 of 43 9WDe% nlrr Dominion Nuclear. Connecticut 41I MP2 SW ModMl ants OQergn 6sls Analysis MIIalslns 1.hi14 2 Page 75 cl.9B Ca*culal'on 12-001 Ruvvsion 0 Addendum NIA 0113112013

3.

A tide level of -1.4 ft is used (modeled as -6.4 It). (Design Input 8.2,11)

4. One pump is in operation / aligned using the 10% Degraded pump curve (Design Input 8.5. 1 ).

5. TBCCW heat exchangers are isolated (2-SW-3.2A/B are modeled closed).

6. SW strainer back wash is In operation and the strainers are modeled at their maximum differential pressure. (Design Input 8.8. 1)

7.

Heat exchangers are modeled debris loaded (Design Inpnts 8.7.5 and Assumption 9.13).

8. Siphon break is assumed in EDG discharge piping.

9; The RBCCW isolation valves 2-SW-.AIB/C are 100% open,

10. The operable. SW headeris aligned to all room coolers and chillers. To achieve this alignment. 2-SW-194 and 2-SW-175 are opun and 2-SW-195 is closed when these heat exchangers are supplied by Facility 1. 2-SW-195 and 2-SW-175 are open and 2-SW-194 is clused when these heat exchangers are supplied by Facility 2.

Flow Summary Reports for the single pump operation scenarios are documented in Attachment F. Table 11.2.10-1: Single Pump Summer Operation to Cool Two SDC Trains Can I1SD (0,/0411 15:.7) Cae lob (-04/av!3 16:39) PsA 15C Facilily; it 2 Shhutloii Coolinp I.olm Facllty 2 to2 S2hbuifn CiiOltng Loops I.'o*np(Mw*l Irndicied ktquired Prdicted Required V*m4irln V.,umetvti Maroin Vluinmui W Volutmeaic Matrgie Rari.lntl lury Riw tl Fn'.i R1.n6 ycefily w R (win) tsv*.[p,-I [3911 h 601f91m lgpl .[pl [prn

• XlA 6024.25 5421.83 44.50 971,83 5943-56 5349.20 4450 8992.

XlSC 595.LI 5362.99 .145D 91t2.99 (73.7 54168.13 4,0, 1016.13 tIll; A Hx 379.61 340.75 276 64.15 382.33 344,10 276 69.J0 ED(t BI Hi 397.41) 357.66 27A 82.0* 4003.318 A1.16 276 X4.16 X +ltA 97.,1A 7814AK 21" 52.01 W,149 77A14 26 51..4. (181*.1t 50.3L" 45.28 13 32.29 49.65 44.0) 13 31 69) X-2192 33.96 2U.87 15 13.H7 33,52 25.49 13 13.49 X-183 28,36 24.11 17 7.1t 21.1h 23.95 17 ,.95 X. '69A ,25.39 30,0 26.9 3.18 34.42 29.26 28.9 2.16 X.Itf3R 35.107 29.91 269 2.91 35.35 30.091 NO,9 3.35 liti:np MSA .[3334.71 17046.'.57 NIA KIA W N/A NIA Pump 1'5C NflW NIA 21N492. 3 1214.35 N/A N/A 12.0 Acceptance Criteria (Optional) Tube Phigging

Serial No 13-419 Docket No. 50-336, Page 43 of 43 Dominion Nuclear Connecticut All MP2 SW Model and eosign Basis Analysis Milltoine una il P*ps 76 of 98 Calculalton 12.001 Revision 0 Addendum NWA 01/312Ot3 The tube plugging limits are listed in each individual heat exchanger calculation. These limits are used to determine the required the flow rates through each :heal. exchanger which are presented in Table 12-1. Required Flows The required flow rates are shown in Table 12-1 as follows: Table 12-1: Acceptance Criteria - Required Flow Rates Cnomirirenn Crouliks Driptio Required %un flo Reference XI OAJ/C LOCA Summer< BO"I 7370 3.4.3413. 1.r XIlIAIC LOCA Winter< 6frF: 5423 1.14.27-A S8,VI"/C N1n,11.1l Oipr ,,,citly I iAcilily 2 - 7.65t)1 1.4,38 XIRAM/11C Samnit 5150 3A,23 XISAiW/C Slhutdown Coolig (C.,

9)

(Ann 1A.11 Xl RAiDZC Single Pump 2 Shulttown CIxlhrg in'ops WCse. I{11 44.50n 3,A.23 XITA/I/C Normal Olt

4311, lA.16 EDII A/ Hx LOCA Summar / SL'mric w/LNPI, 60P 637 34.24 EDG,AM Hx I.JCA Winter< OFF 337 3,1.39 LDG ANO IIx LI4P (W IR00KW)ICaw. 9 & 10) 276 3,4.24 X-IBIA 26

.3,d35 Normal tOtV 05Li.mlc (Cases 4.l,6. 9 & 10) AXl-ISIn I'l 11,1.35 X.IRIA 63 3.4.35 sci Lt)lCAo3,j X-I1I1H

31) 3,4,35 X-i12 I1 3.4,36 X-LIOCA

( uw 3, "Il cwos) 17 3,4.35 1-1O6AIIa IXf 3.4.37 SW I'Pump ,ontwut Flow 17,527

l.4.T!

13.0 Results and/or Conclusions Design basis Flows Section 11.2 providex a discussion as well as tables for each csisc which shows a comparison of the predicted versus required flow rates for each component. The predicted flow rates for each case are obtained from the Flow Summary Reports in Attachment F, The margin reported in Section 11.2 is margin relative to the conservatIve design inputs and assumptions used in calculating the predicted and required flow rates. , According tit Rcteficu'x 3.4.38, at hifs service wuler flow rate whern the inlet tempernaure Ix 7.9'F, the RBCCW temperature will be maintained at 84.9"F with the RBCCW Pump in a dcituded cnditiiin arid tile twvealil roaling in the. X-81A heat exchanger set tit.00101 hr-.'-"F/'ru with 10% tubr. plugging. "n According to Reference 3,4*38. at this service water flow rate when tile inlet temperature is 71)IF, the I*tC'W temperiauroe will be nmintained at 85.0'F widh the RBCCW Pump in I design Or degraded condition and thc overall fouling in the X-81C heat exchanger get at the design fouling with 10% tulle plugging.

Serial No 13-419 Docket No. 50-336 Excerpts from Calculation 12-328 Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336, Page 1 of 7 CALcNO. 12-328 f'Fv 0 5 OF. t ZACHIRYo ZACHRV NUcI6AR, INC. M Equivalent Thermal Performance of the Unit 2 EDG Heat Exchangers for UHS Temperature Increase ZN) Document Type: OAPO 1.0 PURPOSE The purpose of this calculation is 1o determine the required service water flow rates for the Millstone Unit 2 Emergency Diesel Generator Heat Exchangers at Ultimate Heat Sink temperatures ranging from 78"F to 80"F. 2.0 APPROACH The analysis of the Emergency Diesel Generator (EDG) heat exchangers will be performed using Zachry Nuclear Engineering's thermal performance heat exchanger modeling software, PROTO-HX Shell & Tube Module Version 4.10. The PROTO-HX software was developed and validated In accordance with Zachry's Nuclear Software Qualily Assurance Program per References 8.1 and 8.2. This program meets the requirements of 10CFR50 Appendix B. 1 OCFR21, and ANSI NQA-1. It was developed in accordance with the guidelines and standards contained In ANSIIIEEE Standard 730/1984 and ANSI NQA-2b-1991. The use of the application will be in accordance with Reference 83. The EDG is cooled by the Intercooler (X-83 A/B), Lube Oil Cooler (X-53 NB), and the. Jacket Water Cooler (X-45 NB) per Reference 8.5. The EDG heat exchangers are cooled in series on the Service Water (SW) System side. Service Water enters the Intercooler waler cooler, exits and then enters the lube oil cooler, exits and enters the jacket water cooler. Since the coolers are installed in series on the service water side, the flow rate will be the same for each: component. The SW system takes suction from the Ultimate Heat Sink (UHS), the Niantic Bay. The components that are cooled by the service-water system have previously been analyzed for a maximum UHS temperature of 770F. This analysis will review the EDG heat exchangers for UHS temperatures up to 80"F.. The following steps will be taken to analyze the performance of the EDG heat exchangers: Perform a comparison benchmark case run to the original model to ensure that the correct model has been selected. Update the model configuration for the analysis. Determine the required service water flow rate that needs to be supplied.to maintain the heat removal rate. 3.0 DESIGN INPUTS 3.1 The Intercooler Proto-HX Model is documented in Reference 8.4. 3.2 The Inlercooler has a heat duty of 2.067x1 0' Btu/hr for the 2750 KW generator analysis and 1.6874x1 06 Btu/hr for the 1800 KW generator analysis with a process-side flow rate of 400 gpm at a process-side inlet temperature of 1340F per Reference 8.6. 3.2.1 The correct shell-side mass flow rate for 400 gpm at 134*F is correctly modeled by inputting a shell-side flow rate of 399.7 gpm as described in Reference 8.4. Form' N0301FP05 Revision: 00-00 Date, 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336, Page 2 of 7 VArNO. 12-328 'v^ '0 o .ZA C H RY.ORNA°"I°EIF ?AcHm, MUCLEAR, INC. TTLE Equivalent Thermal Performance of the Unit 2 EDG Heat Exc for UHS Temperature Increase Z 3N RMeum8nr Tyt n e 04PD hangers 3.3 Reference 8.5 reports the minimum required SW flow rate is 507 gpm for 5 tubes plugged (5%) at 77IF; these values were calculated In Reference 8,6. 4.0 ASSUMPTIONS 4.1 The analysis will assume 5 tubes plugged for the intercooler.. Reference 8.5 summarizes the design basis for the components and provides flow rate requirements for 5% tube plugging. The vendor data sheet for the intercooler lists 110 tubes; Iherefore 5 (110

• 0.05) tubes are plugged, consistent With Reference 8.5.

4.2 It is assumed that Intercooler X-83 A and X-83-B are Identical.. This assumption Is reasonable as bounding conditions will be used in this analysis. 5.0 ANALYSIS The EDG Is cooled by both the intercooler water cooler, lube oil cooler, and the jacket water cooler. The intercooler water cooler Is the controlling component as discussed In Reference 8.4. This means that if the required heat load Is achieved for the intercooler water cooler, then it will also be achieved for the lube oil cooler and the jacket water cooler. To provide a bounding analysis for the coolers, this analysis will focus on the performance of the Interceoler water cooler for 5% tubes plugged. As outlined In Section 2.0, a comparison to the previous documented benchmark results is performed to ensure that the correct model Is being used. The model is benchmarked against the case 'Design Case (511 gpm @ 8OF, 10 plugs) from Attachment A of Relerence 8.4. The output reports for Ihe cooler are included in Attachment B. Once [he model Is verified, it needs to be configured for the analysis. The Calculation Settings are configured as follows: Fouling Factor: Use Design Fouling Factors Extrapolation Method: Constant Inlet Temperatures Extrapolation Conditions : User Specified Conditions These settings properly configure the model for the analysis. Reference 8.4 documents the evaluation of test data for the intercoolers prior to and following tube-side cleaning for t997. The test data evaluation of Reference 8.4 was based on the heat exchanger data sheet values for fouling, i.e. a process-side design fouling factors of 0 001 and service water-side fouling factors of 0.0005. Those same fouling factors were used in the. Reference 8.6 analysis reported in Reference 8.5. Those fouling factors were applied to the limiting conditions to determine the predicted heat duty of the cooler. The limiting conditions based on an UHS temperature of 77F are summarized as follows (Design Inputs 3.2 and 3.3): Form: N0301 F05 Revision: OD-00 Date: 10-28-2011 .Pago I of I

Serial No 13-419 Docket No. 50-336, Page 3 of 7 CNIt, 12-328 REV a a 7 o" 11 ZACIHRY OFPKIATOR I ZACHRY NUCLEAR. INC. N.CF Equivalent Thermal Performance of the Unit 2 EDG Heat Exchangers for UHS Temperature Increase ZNI Daoument Type: GAPD SW Flow = 507 gpm (5 tubes plugged) SW Inlet Temperatures= 770F Intercooler Process-Side Flow rate = 399.7 gpm intercooler Water Inlet Temperature = 1340F The analysis documented herein will re-evaluate the heat exchanger performance and determine a required flow rate undor limiting conditions with an ultimate heat sink temperature from 78"F to 809F. The required flow rate shall satisfy the previous predicted heat duty for the cooler which was for an ultimate heat sink temperature of 770F. The results of this analysis will provide an Indication as to how much impact a three degree change in temperature has on the flow rate. The service water flow rate entered into PROTO-HX Is a mass flow rate. This ensures the mass flow rate used by PROTO:HX is based on density as a function, of the service water inlet temperature. The volumetric flow rate used to calculate the mass flow rate is reported in Table 1 through Table 31 The mass flow rate is calculated as follows: th = qp.6I017.481 Where q = volumetric flow rate (gpm) p = density at inlet temperature (Ibm/ft") 60 --,conversion from minutes to hours 7,481 = conversion from gallons to cubic feet The density of salt water is calculated using the curve fit equation provided in the PROTO-HX Module library: p = 63.614879 + 0.02339221.5(T)- 0.0025181423(T)'-' + 9.235363,A (")* - 155,78811 Where 'T'is the inlet temperature. The density for each temperature Is as follows: p @78F 63.78 Ib/ft3 p @79F 63.77 1b/ft3 Forin:NO3O1FIJS RevisIon: 00-00 Date; l0-2a-20 II Par~e I of 1 Form: N0301 F05 Revision: 00-00 Date. 10-28-2011 Page t of 1

Serial No 13-419 Docket No. 50-336, Page 4 of 7 CALcIo' 12.328 oL 0 1PAGI 8 O8 ZACHI Y., 1I ZACt.$Y OUCLEAn, INC. lnti Equivalent Thermal Performance of the Unit 2 EDG Heat Exchangers for UHS Temperature Increase Dn1 Document Type: QAPO p @8OF = 63.76 Ib/ft3 The analysis is accomplished in the following steps for 5% tubes plugged.

1. Enter the number of active tubes as 105 (5 tubes plugged).

2. Guess an initial volumetric flow rate greater than 507 gpm and convert the flow rale to a mass flow rate.

3. Select "Calculate." and enter the 'guessed' value for the tube side mass flow rate, enter a tube side temperature for the corresponding UHS temperature, and enter the limiting conditions for the intercooler.

4. Iterate the flow rate until the calculated heatl transfer rate meets the heat duty listed in Reference 8.6 (Inpul 3.2).

The results from the analysis are shown In the following tables. Table 1 through Table 3 provide the results for the 5% tube plugging analysis. Table 1: X-83 ANB Results for 5% Tubes Plugged at an UHS Temp of 78°F 275D KW Generator 1800 KW 2 Load I Generator Load Required Volumetric Flow (gpm) 547 252 Required Mass Flow Rate (tb/hr) 279,810 128,907 Extrapolaeed Heat Transfer (Btu/hr) 2,067,136 1,688,419 Required Heat Transfer (Btu/hr) 2,067,000 1,687,400 Pefrent Difference (%) 0.007 0.060 Table 2: X-83 NB Resulls for 5% Tubes Plugged at an UHS Temp of 79"F 2750 KW Generator 1800 KW Generator Load Load Required Volumetric Flow (gpm) .589 264 Required Mass Flow Rate (lbihr) ,101,247 135,024 Extrapolated Heat Transfer (Btu/hr) 2,067,398 1,689,212 Required Heat Transfer (Btuhr) 2,067,000 1,687,400 Percont Difference (%) 0.019 0.107 Form: N0301F05 Hevlsior,~ 00-00 Onto: 10-28-2011 Page 1 of I Form. N030IF05 Revision. 60-00 Date: 10-28-2011 Page 11 of I

Serial No 13-419 Docket No. 50-336, Page 5 of 7 CAL lNO 12-328 n Dv a PACE 0 l ZACH'RY IDAG°WO 'TVFE ZACHRY NUCLEAR. INC. TfE Equivalent Thermal Performance of the Unit 2 EDG Heat Exchangers for UHS Temperature Increase MZN Document Type: OAPD Table 3: X-83 A/B Results for 5% Tubes Plugged at an UHS Temp of 80F 2750 KW Generator 1800KW Generator Load Load Required Volumetric Flow,(gprnj 637 276 Required Mass Flow Rate (lb/hr) 325,746 141,139 Extiapolated Heat Transfer (Btu/hr) 2,067,166 1,687,489 iequrod Heat Transfer (ButTr). 2,067,000 1,687,400 _ý__ 6 -00 Pecnt Difference (%) 0=080.5 The output reports for the results are provided In Attachments C through H. 640 DISCUSSION OF RESULTS The required flow rate was determined for the design limiting conditions for ultimate heat sink temperatures ranging from 78°F to 801F. The chart below shows the increase in flow rate as the ultimate heat sink temperature is increased for the intercDolers with 5% tubes plugged with a 2750 KW -load on the diesel generators. The curve Is mostly linear with the largest increase in flow of 48 gpm from 79"F to 80*F. The trend shows an approximate 8% Increase in flow for every degree the temperature-is increased. Flowrate vs. UHS Temperature 2750 KW Generator Load, 5% Tubes Plugged 640 620 E 580 540 77.5 79 78.5 79 79.5 s0 80.5 Tempera wrei F) The chart below shows the increase in flow rate as the ultimate heat sink temperature is increased for the intercoolers with 5% tubes plugged with a 1800 KW load on the diesel generators. The curve Is linear. The trend shows an approximate 5% increase in flow for every Form: N0301 FOS Revision: 00-00 Date; 10-28-2011 Page Vl o I

Serial No 13-419 Docket No. 50-336, Page 6 of 7 CALC NO. 12-328 REV a PAGE 10 Or II IZACHRY.ORSNATO'R I________ %WROPIER ZACHRY NUCLEAR, INC. TME Equivalent Thermal Performance of the Unit 2 EDG Heal Exchangers for UHS Temperature Increase ZN1 DMoumnnt Type:.QAPP degree the temperature is increased. Flowrate vs. UHS Temperature 1800 KW Generator Load, 5% Tubes Plugged 280 270 2rO M.o 775 78 78.5 19 79.5 80 80.5 Temperature (F)

7.0 CONCLUSION

This calculation performs an equivalent thermal performance evaluation by taking design data and predicting the performance of the coolers at design limiting conditions for a change in ultimate heat sink temperature. The analysis focuses on the intercooler water cooler which is, the limiting component compared to the Jacket water cooler or the lube oil cooler. The maximum increase in :flow rate that is expected is approximately 8% for every increase In degree. Table 4: Required Service Water Flow Rates for 5% Tubes Plugged Minimum SW Flow Rate Minimum SW Flow Rate UHS Temperature 2750 KW Generator Load 180D KW Generator Load (gpm) (gpm) 252 78 547 264 79 589 276 60 637 Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page :1 of 1

Serial No 13-419 Docket No. 50-336, Page 7 of 7 CALCE 112-328 ReV 0

iiAC, 11 CF 11 Z A HO IIGI.ATOR IVERFIVII ZACHRY NUCLEAR, INC.

WLE Equivalent Thermal Performance of the Unit 2 EDG Heat Exchangers for UHS Temperature Increase ZN! DocumeMn Type: QAPD

8.0 REFERENCES

8.1 Proto-Power Corporation, Software Validation and Verification Report (SVVR) tar Heat Exchanger Modeling Software - Proto-HX. SWR-93948-02, Revision G, May 28, 1999. 82 Proto-Power Corporation, Software Validation and Verilication Report (SVVR) for Heat Exchanger Thermal Performance Modeling Software - Proto-HX, SVVR-93948-02-ShelI and Tube, Revision H. March 23. 2004. 8.3 ProtafPower Corporation, PROTO-HX. Shell and Tube Heat Exchangers Module Version 4.10 User Documentation. 8,4 Proto-Power Corporation Calculation 98-119, Revision B, Analysis of MP2 EDG Heal Exchancqer Thermal Perlormance Test Results. 8.5 Dominion Calculation 03-ENG-04035M2, Revision 00 Including CCN 1 through CCN 3, MP2 :Service Water System Design Basis Summary Calcula3tion. 8.6 Dominion Calculation 94-DES.11111-M2, Revision 00 Including CCN 1, MP2 SWS Maximum Allowable SWS Temperature to the EDG Heat Exchangers 0 1800KW and 2750 KW Electrical Load Levels with 5% Tubes Plugged in each Unit Assuminq 507 GPM 0 2750 KW and 265 aom (01 ROD KW Form: N0301F05* Revision: 00-00 Data: iO~28*2O1 I Page I oft Form: N0301 F05 Revision: 00-00 Date: 10-28-2011 Page I of I

Serial No 13-419 Docket No. 50-336 0 Updated Final Safety Analysis Report Section 1.4 "Principal Architectural and Engineering Criteria for Design" (Revision 31.1) Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336 0, Page 1 of 8 NIPS-2 ISARI 1.4 PRINCIPA. ARCHITECTURAL AND ENGINEERING CRITERIA FOR DESIGNS 11we principal architectural and engineering.feutures used in the design of Unit 2.ofthe Millstone Nuclear Power Station are sununarized in the following material. 1.4.1 PLANT DESIGN Principal structures mid equipment which may serve either to prevent accidents or to mitigate their consequences have been designed.. fabricated and erected in accordance with applicable codes so as to withstand the most severe earthquakes, flooding conditions,. windstonns. ice conditions, temperature and other deleterious natural phenomnena which could be reasonably assumed to occur at the site during the lifetlime ofthis plant. Systems and components designed .for Seismic Category I requiremenLs are listed in l'eabl I.4-I. It should be noted that the terms 'Category* and 'Class' are used interchangeably throughout the MP2 FSAR in defining seismic design classifications of Structures, Systemns and Components. Unit 2 was designed so that the safiety of one unit will not he impaired in the unlikely event of an accident in the otherunit. Principal structures and equipment were sized for die maximum expected nuclear stearn supply system (NSSS) and turbine outputs. Redundancy is provided in the reactor and safety systems so that the single failure ofany active component ofeither system cannot prevent the action necessary to avoid an unsali condition. "llie unit is designed tofaceilitate inspection and testing of systemls and components whose reliabilities are important Ito he protection of the public and plant personnel. Provisions have been niade to protect against the hazards ofsuch events as fires or.explosions Sysiems and components which are significant froni.the standpoint.of nuclear safety are designed,

• fabricated and erected to quality standards.conimensurate with the safety function to he peribnned..Appendix IA of this FSAR addresses the implementation of Atomic Energy Comm ission (AEC) General Design Criteria for Nuclear Power Plants. 10 CFR Part 50, Appendix A. Chapter 12.q describes the Qtuility Assurance Program.

1.4.2 REACTOR The tollowing.crileria (see Chapter 3) apply to die reactor:

a.

The reactor is of the pressurized water-type, designed to provide heal to stean generator,- which, in turn, provide steam to drive a turbine generator. The initial full powercore thennal output was 2560 inegawatts (the NSSS rating was 2570 megawatts) prior to its uprating to the current 2700 megawatts thermal power level (NSSS rating of 2715 niegawatts).

b.

The reactor is refueled with slightly enriched uranium dioxide containedin zircalloy luhes_ 1.4-1 Rev. 3t.1

Serial No 13-419 Docket No. 50-336 0, Page 2 of 8 NIPS-2 FSAR Minimun departure from nucleate boiling ratio during nornal operation mid anticipated transients will riot be below that value which could lead to fiuel rod failure or damage. "The niasimurn fuel centerline tmpenrature evaluated at tlhe design overpower condition will be below that value which could lead to fuel rod failure. The melling point of the UO2 will not be reached during routine operation and anticipated transients. d.. Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within Are rods and other factors affecting design lilfe will be considered for the maximum expected exposures. '[lie reactor and control systems are designed so that any xenon transients can be adequately damped. .r. The reactor is designed to accommudate the anticipated transients safely and wilhott fuel damage. g T nie reactor coolmat system (RCS) is designed and constructed to maintain its inLegrity throughout the expected plant life. Appropriate means of test and inspection are pro\\idcd. h.. PoNwer excursions which could result fromiany credible reactivily addition.accident will not cause damage. cither by deformation or rupture, to the pressure vessel or impair operation of the engineered safely features (ESF).

i.

Control element assemblies (CEA) are capable of holding the core suberiliial at .hot zero power conditions following a trip. and providing a safety margin even with the most reactive CEA stuck in the fill. withdrawn position.

j.

'ie chemical and volume control system (CVCS) can add boric acid to the reactor coolant at a sulficient rate to maintain an adequate shutdown margin when the RCS is cooling down hillowing a reactor trip. "ibis is accomplisbed at a maximuni design rate. This system is independent of the CEA system. k.. The combined response of the fuel temperature. coefficient. the moderator temperature cocfficient. the moderator void coefficient and the moderator pressure ~edllicient to an increase in reactor thermal power is.a decrease in reactivity. In addition. tlhe reactor power transient remains bounded zud damped in response to any expected changes in any operating variable, 1..4-2 Rev. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 3 of 8 MI'S-2 FSAAR 1.4.3 REACTOR COOLANTA AND AUXILIARY SYSTEMS 1-4.3.1 Reactor Coolant System The design bases in this section are those used lbr the integrated design of the RCS or those which apply to all of the system components. The design bass" unique to each component arc discussed in Section 4.3. The RCS is designed fbr the nornial operation of transferring 2715 MWt (9.26 x 10 BtnIhr) firom the reactor core (2700 iWI) and reactor cioolant pumnps (15 MWt) to the steam generators. In the steam generator, this heat is transinrred to the secondary system Ibrning 5.9 x 106 Ib;hI" of 880 psia saturated steam per generator with a 0.2 percent maxinum moisture content. The RCS is designed to acconmnodate tile noninal design transients listed. These transients include conservative estimates ol'the operational requirements of the systems and are used to make the required component fatigue anaiyses.

a.

500 heatup and cooldown cycles at winaximum heating and cooling rate of 100Fihr. The pressurizer is designed for a maximum cooldown rate of 200'Fiir,

b.

Pressurizer spray piping is limited to 160 plant heatup and cooldown cycles. Primary manmay studs oI'the replaced steamn generators are limited to 200 licatup and cooldom~n cycles.. C. 15,000 power change cycles in the range between 15 and 100 percent of fidl load with a ramp load change of 5 percent of full load per minute increasing or decreasing. This will occur without reactor trip.

d.

Primary manway studs for the replaced steamn generators arelimihed to 1,000 cycles with a ramp load change ofl5% per minute decreasing and 30% per hour increasing (plant 1oading~unloading).

e.

2,001) step power changes or I0 percent both increasing and decreasing between 15 and 100 percent offull load. Primary manway studs for the replaced steam generator are limited to 1.500 step power changes. .10 cycles of hydrostatic testing at 3.110 psig and a temperature at least 60'F above the nil ductility transition temperature (NDT') of the. component having the highest NDYT g1 2(X) cycles of leak testing at 2.485 psig and a temperature. at least 60TP greater than the NDTT of the component with the highest NDDT. 1h, Primary aianway studs for tie replaced steam generators are limited to 80 cycles of leak testing at 2.485 psig. 1..4-3 Rev. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 4 of 8 MPS-2 FSAR i, l06 cycles of operating pressure variations of +/-100 psi from the normal 2,235 psig operating pressure and +/-61F at operating temperature and pressure.

j.

400 reactor trips when at 100 percent power. Primary manway studs tbr the replaced steam generator are limited to 200 reactor trips when at 100% power., In addition to these tnormal design rnusients, the fbllowing ahnormal transientLs are also considered to arrive at a satisfactory usage factor as defined in Section IlL Nuclear Vessels. of the ASME Boiler and Pressure Vessel Code:

a.

40 cycles of loss of tu'bine load from 100 percent power. b.: 40 cycles of loss ofreactor cooltmnt flow when itt 100 percent.

c.

5 cycles of loss of main steam system pressure. Components of the RCS are designed and will be operated so that no deleterious pressure or thermal stress will be imposed on the structural materials. The necessary consideration has been given to tie ductile characteristics or' the materials at low temperature. 1.4.3.2 Chemical and Volume Control System 'lbe major Iftnctions of the CVCS (see Section 9.2) arc toi

a.

Maintain tile required volume of water in the RCS.

b.

Maintain dle* chemistry and purity of the reactor coolant.

c.

Maintain tile desired boric acid concentration in the reactor coolant.

d.

Provide a controlled path to thie wvaste processing system. The system is designed to accept die discharge when die reactor coolant is heated at tile design rate of I 00Fuir and to provide the required makeup when the reactor coolant is cooled at the design rate of 100*1,aiFhr. Discharge is automatically diverted to the waste processing system when the volume control tank is at its highest permissible level. The system will also supply makeup or accept discharge due to power decreases or increases. Tlhe design transients are +/-10 percent of lull power step changes and ramp changes. of+/-5 percent of filtl power per minute between 15 to I 00 percent power. Omi power increases, the letdown flow is automatically diverted to thie waste processing system whe the volume control Iank. reaches die highest permissible level. On power decreases, sufficient coolant is hi tite volume control tank to allow a full to zero power decrease without additional makeup. in the event of a makeup systetim failure or override. For anm assumed 1 percent failed fuel condition. the activity in the reactor coolant does not exceed 411 itCi/ce at 771'. Trhe systcm is also designed to maintain the reactor coolant chemistry within the limits speciied in Subsection 14,3. 1.4-4 Rev. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 5 of 8 INIPS-2 FSARI "he rawt of boron addition is sufficient to counteract the. maximum reactivity increase due to cooldown and Xenon decay. Any one of the three charging pumps is capable of: injeefing tie required boron (as boric acid). lThe maximum rate at which the reactor coolant boron concentration cal be reduced must be substantially less than the equivaklet maximium rate of reactivity insertion by the CEA. Prior to refueling, the system is capable Of incrcasing flhe reactor coolant boron concentration firom zero to 1720 ppm by feed and bleed when the reactor coolant is at hot stuadby operating temperat ure. Provisions to fIailitate the plant hydro.sttiic testing and to leak test the RCS are included:. 1A4.3 Shutdown Cooling System The shutdown cooling system (see Section 9.3) is designed to cool the RCS from approxinmately 300' to 130"tFin 24 hours, assuming that the component cooling water inlet.temperature is at its ,nax'iiniin design value of 95"1. The design RCS cooldown rate is I00°F/hr. A temperature of 1 30'*F or less call be achieved 27.5 hours aller reactor shutdown. assuming an infinitely exposed core, The maxinumr allowable pressure lor lhe RCS during shutdown cooling is approximately 285 psig. 1,4.4 CONTAINMENT SYSTEM 'The ontainment (see Sctkins 5.2 and 14.8). including the associated access openings and penetrations, is designed to contain pressures and temperatures resulting from a postulated main steamline break (MSIL) in which:

a.

k range of power level break sizes, and single failures are considered.

b.

Cases with the loss of oti'ife power and with AC power available are analyzed to 13-2 determine which scenario maximizes the energy removal into containment. C. Saf'ety injection is nolt assuned since it would tend to reduce.the energy released into containment.

d.

Thec containment air recirculation cooling system and the containment spray 13-2 system are credited to mitigale. the containment pressure and temperature coileiuenices. Contairnent response to a loss-of coolant (LOCA) accident was also analyzed. It was flound that the peak containment pressure and temperature of the MSIA1 accident houndthe IL.OA. I13-2 hlie containment is designed to assure integrity against postulated missiles laon equipment Failures and against postulated missiles froin external sources. 1.4-5 Rev'. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 6 of 8 MPS-2 FSAR Means are provided fbr pressure and leak rate testing of the containmnent system. This includes provisions for leak rate testhig of individual piping and electrical penetrations that rely onl gestated,eals, sealing compounds, expansion bellows, and the interior of the containment. The enclosure building (see S*eliou 5.3) is designed to withstand a Wind loading of 115 mph, with gusts of 140 mph, snow load of 60 psf and seismic loads. 11hC EnClosurM Building is designed so that is structural framing will withstand tornado loads, but the sidingwill be blown away (see Section 5.3,3). 1.4.5 EN(INEER.ED SAFETY FEATURES SYSTEMS" The design incorporates redundant independent full capacity engineered safety lfiatures systems (ESFS). l'hese. in conjtunction with the containment ensure that the release of fission products. following any postulated occurrcncec at least the minimum ESF required to temninatw that occurTence, are operable. The lillowing are required as minimum safety features: One high pressure safety injection (IIPSI) train One low pressure safety injection (LPSI) train Foursactly injection tanks (water quantity of three is required to reach the core) One containment spray and two containment air recirculation and cooling subsystems, or equivalent (Section 6.4) One hydrogen control subsystem One cnclosure building filtration traiin One auxiliary teedwater trains Each of these subsystems is independent of its redundant counterpart with the exception of the safety inlijction subsystems. 'Ilie HPSI and IPSI subsystems.(S ctl1in 6.3) are independent up to the common pipe connections to the Iburreactor coolant cold legs. Remote imanualy operated valves provide appropriate cross-eonneciions between redundant subsystems for backup and to allow maintenance. Redundant components are physically separated. Th1e ESFS are designed to perform their functions for all break sizes in the RCS piping up to and including the double-ended rupture of the largest reactor coolant pipe. The safety injection system limits fule, and cladding damage to an amount which will not interfere with adequate emergency core cooling and holds metal-water reactions to minimal amounts. Two full capacity systems, based on different principles remove heat from the containment to maintain containment integrity. the containme~ut spraysystern (Sections 6.4) and the containment air recirculation and cooling systenm (Section 6.5). he enclosure building filtration.systern (EIFS) (Section 6.7) maintains the enclosure building filtration region (EB13FR) at aslightly negative pressure and filters the exhaust forom this space. The containuncnt postaccidetu hydrogen control system (Seclion 6.6) mixes and 1.4-6 Rev. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 7 of 8 NIPS-2 ESA I monitors the accumulation oi'hydrogcn gases within the cortainiment. Purge and recombiners are not credited for any mitigating fumction. 1.4.6 PROTECTION. CONTROL.A'D INSTRUMENTATION SYSTEM A reactor protective.system (RPS) (sce Sect ion 7.2) is provided which initiates reactor trip ifthe reactor approaches an unsafe condition, Interlocks and automatic protective systems are provided along with administrative controls to ensure safe operation of the plant. Sufficient redundancy is installed to permit periodic tesling of the RPS so that failure or removal froin service of any one proltective.system component or portion of the system will. nut preclude reactor trip or other safety action when required. tlle protective system is isolated from the control instrumentation systems so that failure or removal front service of an-control instrumentation system component or channel does not inhibit the funetion ofrthe protective system: 1.4.7 ELECTRICAL. SYSTEMS Normal, reserve and emergency sources of auxiliary electrical power are provided.to assure safe and orderly shutdown of the plant and to maintain a safe shutdown condition wider all credible. circumstances, Onsitc electrical power sources and systems arc designed to provide dependability, independence, redundancy and testability in accordance with the requirements of 10 CFR Part 50. Appendix A. 'lhe load-carrying capability and other electrical and mechanical characteristics of emergency power systems are in accordance with the requirements of Safety Guide Number 9. 'Two redundant, independent, full capacity emergency power sources and distribution subsystems are provided. Each of these subsystems powers all equipment in the associated safety related subsystems as described in Stilxreetion 1.4.5. 1.4.8 RADIOACTIVE WASTE PROCESSING SYSTEM 'llie radioactive waste processing system (see Section 11. 1) is designed so that discharges of radioactivity to the environment are minimized. and are in accordance with the requirements of Sections 1301 and 1302 and Appendix B of 10 CFR Part 20 and Appendix I of 10 CFR Palr 50. L.4.9 RAD)IATION PROTECTION Millstone Unit 2.is. provided wvith a centralized control room which has adequate shielding (see. Section.1.1.2.2.3).and ventilation system features (see Section 9.9; 10) to permit occupancy during all poslulated accidents involving radiation releases. lI'lie radiation shielding in Millstone Unit 2 and the radiation control procedures ensure that. operating personnel do not receive exposures during normal operation and maintenance in excess of the applicable limits of 10 CFR Part 20. 1.4-7 Rev. 31.1

Serial No 13-419 Docket No. 50-336 0, Page 8 of 8 NIPS-2 FSAR 1.4.10 ULIEL IIANDLING A.,ND STORAGE Fune handling mid storage Facilities (see *cioi n9,X) are provided for the sare handling and storage of fuel. The design precludes accideutal criticality. Rev. 3 1.1

Serial No 13-419 Docket No. 50-336 1 Updated Final Safety Analysis Report Section 9.4 "Reactor Building Closed Cooling Water System" (Revision 31.1) Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Docket No. 50-336 1, Page 1 of 12 MPtS-2 FSAR 9.4 REACTOR 13UILDING CLOSED COOLING WATER SYSTEM 9,4.1 hSl(iN B-ASES 94.1.1 Functional Requirement.s The function of the reactor building closed cooling water (RBCCW) system is to transfer heat firomn satty related structures. systems, and components to an ultimate heat sink. The RBCCW systern.saletv fuiction is to transtfer the combined heat load of'these structures, system and components under inonnal operating and loss-of-coolant accidenit (LOCA) conditions. 9.4.1.2 Design Criteria The following criteria have beei used in the design of the R13CCW system:

a.

The svstemn shall have two independent redundant subsystems having 100 percent heat removal capacity following a LOCA.

b.

• fhe system shall have suitable subsystem and component alignaments to assure operation f'1the complete subilsystem with associated components.

C. Capabilities shall be provided to assure the system operation with either onsite prover (assUningoflkite power is not available) or with off site power.

d.

A single failure in either subty*tem shall not affect the limational capability ofthe other stibsysteni.

0.

llic RBCCW system shall be designed to permit periodic inspection ol important components, such as RBCCW pumps, heat exchangers. valves and piping to assure the integrity and capability ofthe system.

f.

The RBCCW system shall be designed to permit appropriate periodic pressure and lunCtional testing to assNur: (1) the structural and lkaktight integrity of its componenLts- (2) the operability and perlbrmance ol'the active components of the systent, and. (3) the operability of!the system as a whole. Underconditions as close to ftie design as practical. perfornance of the full operational sequenec that brings the system into operation, including operation of applicable portions of the protection system, and the tran-lrr between normal and emergency p6wer sources. shall be demonstrated.

g.

The system shall be designed to the general criteria as described inSectiotn 61. Ii. The RBCCW system shall be an intermediate barrier between the service.water (seawater) system and the radioactive or potentially radioactive fluids contained in the systems and components being cooled by the RBC-CW.systemn. 9.4-1 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 2 of 12 MNPS-2 FSAR

i.

'hec RIICCW system componcnts shall be designed to operate within the environment to which they are exposed. .j. ýrTh RB3CCW system shall be evaluated to assure that it maintains structural integrity and pressure boutdary for the postulated water hammer loading transient resulting from concurrent occurrences ofa Design HBasis I.,OCA event (i.e., IOCA or MSLB) with Loss of Nomial Power cycnt (i.e., LNP) as described in NRC Generic Letter GL 96-06. The structural design or piping and supporting components shall accommodate the aforementioned water haimmer loading transient. with other applicable concurrent loads, so that continued function of the system is assuired to mitigate lhe consequences orthe allorenlentioned accident.

k.

[he CAR and.CEHM Cooler units shall be evaluated to assure the integrity ofithe pressure boundary and structural adequacy for the GI. 96-06 based water hammer loading condition. 9.4.2 SYSTEM DESCRIPTION 9.4.2.1. System "1bc RBCCW systoe is showvn schematically in Figuires 9.4-1 through 9.4-6. The logic diagram is shown in Figure 7,3-,. 'Ibe R.BCCW system consists of two independent headers. each including one motor-driven RBCCW pump, one RBCCW heat exchanger and associated piping. valves, instrumentation, te ank. th ird RI3C( pupadbt controls and a downeomner from the RHCCW surgt A r CW pump and heat exchanger is provided atc a spare for the system. Redundant safety feature components, cooled by the RBCCW system, are split between the two independent.RBCCW headers. The other systems and components cooled by the RBCCW system are divided between the R13CCW headers to cquialize header heat loads. 'l1e items cooled by the RHCCW system are listed in the following. Further inlruonation about each maybe round in the section referenced in the parentheses. Containment air recirculation and cooling writs (Section 6.5) Refactor vessel support cooling coils (Section 6:5.4. 1) Containment spray punp seal coolers (Svction 6.4) lHigh and low pressnre sat1y injection pttnp seal coolers (Sectioli 6.3) Shutdown cooling beat exchangers (Section 6.4. 6.3) Engineered safey' features room air recirculation and cooling units (Section 9,9. 7) Reactor co0lani pump thermal harrier and oil coolers (Scetii.n 4.3.3) 9.4-2 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 3 of 12 MPS-2 FSAR Primary drain and quench tankk heat exchanger (Section.11. 1) CEI)M coolers (SMtionI 9.9:1 ) Letdown heat exchanger (Scetiun 9.2) Degasilier effluent cooler (Sectiow Ir.I 1) Degasificr vent condenser Sample coolers (Section 9.6.2. 1) Spent fuel pool heat exchangers (Section 9.5) Waste gas compressor aftercoolers Quench tank heat exchanger (Section I 0A4:6) Each RBCCW pump is designed to circulate 7000 gpm of'water through the RBCCW system :for removing heat Ironi systems and components that handle the reactor coolant-. containment atmosphere and engineered safety features. The RBCCW is coolcd in the R3CCW heat exchangers by the servi cewater (seawater) (Subseciion9 7 2. 2. 1). The RBCCW System is designed to provide cooling lbr the following system operating conditions:

1.

Nonmal Operation I. 3.5 Hours after Shutdown Il. 27.5 Hlours alter Shutdown IV. Loss of Coolant Accident (LOCA) Injection Operation V-LOCA. Recirculation Operation System hydraulic analysis and flow balancing were performed to ensure diat the minimum required Flows can he provided to essential safety related loads :for the above operating conditions. The flow balance was. field tested to demonstrate that. when the safety injection actuation signal (SIAS).and the sump recirculation actuation signal (SRAS) automatically realigns the system, the minimum required flows will be. provided to the essential safety related loads during LOCA injection operation (condition IV) and IOCA recirCulation operation (condition V). Thble 9.4-3 identities the minimum required flows for these two post-accident operating conditions. "blie R.ICCW pump and heat exchanger design is hased on requirement.% during normal operation, nomal shutdown cooling and a LOCA. This results in a design which provides an adequate cooling capacity for normal and emergency conditions, including reactor shutdown. A R13CCW surge tank provides sufficient NPSH for the RI3CCW pumps and absorbs the vohltnetric change due to temperature changes imposed on the water within the RBCCW system. 9.4-3 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 4 of 12 MPS-2 FSAR 'Ihe lower half oIuthc RI3CCW surge lank is divided into two equal sections by a vertical weir. If one of the two independent RBCCW lhezaders nrptures. the vertical weir in the RBCCW surge tank assures that 1000 gallons of makeup water will he availahle to the lunctional header. Demincralized water is used for makeup and comes front the prinary water storage tankk (Section 9.121). A flow orilice is located in each surge tank. discharge header.downsreamn Of the outlet isolation valves to dampen pressure surges that result from cold starts of the RBCCW Pumps that could occur following an LNP scenario. Makeup water supplied to the RBCCW surge tank enters firon the lop of the surge lank at two locations on either side of the weir, therefore, makeup water can he.supplied to both sections ofh the tank. Tlo stop the flow of water to the rifiled header, manual valves can be closed. 'Ihe rollnoNiing instrtnicntation and alarms louated on the main control board give.sufflicient infornation to assure the operator that makeup water is being supplied to the R13CCW surge tank section supplying the functional header: I. Position lights ZS-6000 on makeup valve LV-6000 indicate whether the valve is opon or closed. A timed alarm ZAII-6000 indicates that the valve has been open longer taitn required for normal tmakeup.

2.

l.evel controller IC-6000) controls 6 inches above the weir. so regardless of which side tfils., the controller will keep the makeup water valve LV-6000 open until the I'ailed.header is repaired or until the valve is closed by the hand switch HS-6)0(0 or by using manual valves.

3.

I.evel indicators 11-6001 and [U-6730 indicate the water level in each of the two sections of the surge tank. Alarm LAHL-6892 indicates high and low water level in the surge tank. The valves, piping. instrumentation and atarnis described above are shown on Figure 9.4. t. Corrosion inhibitor can be injected into the RBCCW system to increase the optimal concentration during honnaT. operation..., retquired. Exleinal leaks, relief valve disdharges and drains are collected by the drains system and processed through the radioactive waste processing system. Leakage into the RBCCW system can be determined by a radiation monitor installed on the discharge of the RI3CCW pumps. The RBCCW system components located in tile containment building are subject to a maximum temperature of 289'F and a maximunm relative !tumidity of 100 percent during post-incident ope ration. "lic RIBCCW system components located in the auxiliary building are subject to a temperature of I 1WT and a maximum relative hutnidity of 100 percent. The RBCCW pumps and motors are subjected to a maximum temperature of 1357 after a ILOCA. l1e RBCCW sysecm is designed to permit venting o1 components, obtainiing system water samples. and draining of components to the radioactive waste system. 9.4-4 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 5 of 12 NIPS-2 FSAHI Components and heat exchangvrs served by the RBCCW system, which can be isolated arc equipped with self-actuated, spring-loaded relief valves for overpressure protection. 9A.2.2 Components A description ofthe nmjor RIBCCW system components is given in 'lIC 94.-I. 9.4.3 SYSTEM OPERATION 9.4.3.1 Normal Operation During nomnal operation, RBCCW is supplied totthe components described in Subsection 9.4,2.1 with the exception of the shutdown heat exchangers and the cngineered safely features room air recirculation and cooling.units. Two RBCCW pumps and two RBCCW heat exchangers arC required for cDOling service during normal operation. The heatload per RBCCW hcat exchanger during nortnal operation is 16.8 x 106 BtWl-r (header.A) and 23.4 x 106 B.tiuhr (header "B') and the RBCCW discharge temperature is 85"F whien the RBCCW heat exchangers are cooled by service water up to maximtan temperature of( 75°F. The operation of the RBCCW system is. monitored with the tbltowinginstrulnntation:

a.

A temperature detector in the inlet and outlet line of the R.BCCWS heat exchangers and temperature alarms in the outlet lines. h.. Pressure indicators in the lines hctween the pumps and the RHCCWS heat exchangers.

c.

I,evel alanm, and a controlleron the surge tanV.

d.

Temperature detectors located on the outlet lines of the components being cooled.

e.

Flow indicators wnd high flow alarmns in both headers on the discharge side of the !UICCW heat exchangers.

f.

1 ow pressure alani in hoth headers on the discharge side of the RBlCCW heat exchanger.

g.

Handswitches and indicating lights for the pumps and remotely operated control valves. 9-4.3.2 Emergency Conditions Following a I..OCA, the R13CCW system is automatically aligned for post-incident cooling.by the safety injection actatmion signal (Section 7.3), the containment isolation actuation signi.I 9.4-5

• Rev. 3 1.. 1

Serial No 13-419 Docket No. 50-336 1, Page 6 of 12 NIPS-2 1'SAI (Secion 7.3) and the sunmp recirculation actuation signal (Scction 7.3). "1'e SIAS and CIAS actuates valves to stop the RBCC*W :flowto components not required during a LOCA and to initiate R13CCW hlow to the engineered safety Icualures room air recireulation and cooling unit,. Ilic SRAS signal will open the R13CCW shutdown heat exchanger outle't valves. Prior to the. SRAS signal, the RBCCW shutdown heat exchanger outlet valves have a safety ibuction to remain closed (during the injection phase of a LOCA) to ensure adequate RBCCW flow is dirccted to the CAR coolers. The RBCCW is also supplied to the HIPSI LPSI and containment spray pump seal coolers. 't1he R HCCW system operation during a IL.CCA Uses tWo RIHCCW pumps, two RHCC.W heat exchangers andtwo header% tbr LOCA cooling. Ifrone RBCCW header is losL one R.BCCW ptunp. heat exchanger and one header can provide adequate cooling. Calculations provide the basis for the FSAR Section 14.8.2 Containment Pressurization.Analysis. To maximize the containment.pressurization consequences Ftollowinga MSLB or LOCA accident, these. analvses.minimize the energy removed from containment hy the CAR cooling units and the SDC heat exchangers. This is accomplished by assuming design fouling of the CAR cooling units, the SI)C heat exchanger., and the RI1CCW to Service Water heat exchangers, and minimum RBCCW flow distributions assuming degraded RBCCW pumps. Prior to the stunp. recirculation (SRAS), two CAR cooling uniks traisifr a maximum of 160 million RTUliour ol' energy to the RhLCCI system. Following SRAS, two CAR cooling units transfer a maximum of70 million BTU/hir and the S.DC heal. exchanger transfers a maximumn of 47 million. BTUthour of1 energy to the RBCCW System. The RBCCW peak temperature analysis-which maximizes the RBCCW temperatures by assuming clean CAR cooling unit, and SDC heat eXchangers, with maximum R8CCW flow distributions was performed to evaluate the system design capability. Based on the RBC.CW peak tenmperature analysis, the maxinmum RBCCW heat load from the two containment air recirculation and cooling units plus the balance of plant components is 204.4 x j 13-2 106 Dtuhr during the injcetion mode following a LOCA. During thc recirculation mode following a 1,OCA the maximum calculated heat load of 130,.7 x 106 htulhr is translelned to the RBCCW 13-2 system through the containment air recirculauion and cooling units. the shutdown heat exchanger plus the balance of plant components. "Ilic design heat removal capacity of the R13CCW heat exchanger is 160 x 106 Blut1hr duringthe injection mode, but calculations demonstrate that the heat exchangers can adequately handle the 204.4 x 1.06 ltu/hr heat load. 13-2 The RBCCW peak temperature analysis calculated the maximum system supply temperatures based on maximum RBCCW llow distribution values, clean CAR coolers and Si)0 heat exchangers, and fooled (design) RBCCW heat exchangers. The component maximum outlet temperatures wvere then determined based on design heat loads. 't1c system components and piping, as well as the affect on the equipmenl and rooms being cooled post-accident were, evaluated at the maxinian RHCCW temperatures. llie calculated peak.RJCCW temperature supplied to the components during a LOC.AMSLI3 with LNP (GL 96-06 scenario), and considering no fouling in the CAR. Cooler units is less than 1491'F. t32 9.4-6 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 7 of 12 IMPS-2 FSAR Tlhe following is a list of essential components cooled by the IWCCW System during a LOCA:. Containment Air Recirculation And Cooling I inits Shutdown Cooling Ilcat Exchangers Engineered Safety Features Room Air CoolingCoils High Pressure Safety Injection Pump Seal Coolers Low Pressure Safety Injection Pump Seal Cooler'. Coniainment Spray Pump Seal Coolers 'l1ic nonessential components which are on line olter a ILOCA tre the waste gas compressors, Primary drain and quench tank cooler. CEIDM coolers, Reactor verse! support concrete cooling coils. reactor coolant punmp thenliharriers and lube oil coolcrs, and degasvificr vent condenser. After SRAS all the nonessential components except for. the waste gas compressors and degasilier vent condenser s-hall he off'the line. l-owever, the spent fuel pool heat exchanger will be retun*.d to service %sithin 4 to. 5 hours after the start of the LOCA. The Primary Sample Coolers may also be returned to se.vice as necessary to support PASS cooling. A failure mode aialysis is given in FSAR Table 9.4-2 which describes how leaks and malfunctions are detected and what corrective actions must be taken to insure that the safety functions performed by the RBCCW System are not impaired. Once the operator has detected a leak or malfiunction, the operator will isolate the equipment.for repair maintenance if environmental conditions pennit operation of manual valves. If environmental conditions do not permit isolation of the equipment, the entire header may be isolated remotely from. the main control room. Tlie safety functions ofthe.RCCW sysleln can be perl'onbmd with only one header in operation. If the leak or realiunection involves the RI3CCW pump and heat exchanger, the pump and heal exchanger can be isolated remotely froom the main control room. Also. equipment in the containment can be isolated remotely i.n the main control room by closing the. containment isolation valves. Relief valves. on the RBCCW supply and return hieaders serving the reactor Coolant Pumps integral heat exchangers, ensure that the RBCCW containment isolation valves, on these headers, can be closed during an intersystem LOCA, by ensuring that the valves will not be overpresstuized to the point lhat the valve motors camnot close the valves. The relief valves, which have a set pressure of 165 psig may open upon a rupture of an integral heat exchanger tube (a beyond design basis event)- This is required in response to NRC IN 89-54 to preclude an uttisolable release of radioactive fluid outside of the containment tlhat may exceed 10CFRI00 limnits. Tlhe operation oflthe RIBCCW system during a LOCA is monitored as described in suheclion 9,4,3.1. Tlhe RBCCW system, consisting of two independent headers, is at a higher prussure than the service water system during normal operations to prevent service water leaking into the RBCCW 9.4-7 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 8 of 12 MPS-2 ISAR system. 'llie only time that the R13CC W.systern pressure will be lower than the service water system is (1.) during shutdown flr routine maintenance. (2) itf the RBCCW pumps malfunction or (3) ifthere is a maior RI3CCW leak greater than 400 gpm. Loss of'RICCW syslem pressure is indicated by low pressure alarms (PAL-6036 & PAL-6037) located in the RBCCW heat exchanger discharge headers. Once the operator has observed the low pressure alarm, it will be decided whether this branch can he easily isolated; if this is not possible. the RHCCW header will be taken out of service and at the same, time the shell sidc of the RBCC7W hieat exchanger will be isolated. Therefore, the possibility of service water leaking into the RBCCW heat exchanger is remote. If service water does leak into the RBCCW system during periods of1low pressure, itwill he detected by Imanual samplinig on a routine basis. Since the RBCCW system is potentially capable of contaminating the servicewater systemn with radioatlivity due to leakage, a.detection syslem is located within the RICCW system. The RBCCW Sys*tem is continuously monitored for radioactivity by using a self-contained closed-loop radiation monitoring system consisting oF(l) a gamnma Nal(Tfl) scintillation deteelor assembly, (2) a 4 inch schedule 40 stainless steel sample chamnber shielded with lead to reduce the ambient background radiation level, (3) solid-state, control room readout module (4) local meter indication with visual mid audible alarns, and (5) auxiliary support equipment such As piping, val ves and recorder. The RBCCW sample is continuously taken from. the RBCCW pump discharge and is circulated through the sample chamber and detector assembly.,The sample returns to the R HCCW pump suction. "llie radioactivity present in the system is mionitord by the gamma detector assembly. displayed and recorded in the control room. Alarm set points based on the limits established by 10 CFR Part 2(0, will result in alann annunciations in the control room and at the monitoring site. 9.4.3.3 Shutdown Tlhe RIBCCW system is required for continuous operation during normal unit shutdown. cooling .(Secti on.6.3) beginn ing 3.5 hours after the start of unit shutdown. The RBCCWsysten supplies cooling water to the shutdown heat exthangcrs to cool'the reactor coolant from 300 to 130'F in 24 hours. During normal shutdown, such as for refueling, two RBCCW ptumps and two RBCCW heat exchangers are used I'r cooling. The heat load per RHCCW heat exchanger is 84.1 x 1006 ltu/hr (header:'A') and 84.5 x 106 l3tihr((header '"3) at3.5 hours after start oft't.nit shutdown, The RBCCW heat exchanger discharge is less than [30TF when the RBCCW heat exchangers are cooled by service water at a maximutm temperature of 75"F. At 27.5 hours after the start of Unit shutdown, the heat load per RBCCW heat exchanger is reduced to 40.8 x IOr; Bto/hr. 9.4-8 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 9 of 12 MIPS-2 I"SARI lfonc.RBCCW heat exchanger orpump should mailunction. the spare can be immediately placed in service. orlthe unit can be cooled down at a slower rate. will one heat,exchanger and one pump. During sluldo%.n the operation of the RBCCW systcm is monitored us described in Suhscclion 944A1. 94.4 AVA'llAHIliTY AND REl.IABlLIT, 9.4.4.1 Special Features Ilie components of the 13BCC.W system are designed to general design requirements including seismic response as described in Section 6.1. All components are protected I1roml missile dam*age and pipe whip fiom high pressure piping ms described in Section.6.1. To ensure pump availability, each RBCCW pump motor is supplied.from a specific 4160.volt bus. The buses may be powered from alternate sources (i.e.. NSST,. RSST. )/QG or via Unit. 3 bus 34A or 34H). In the event that reserve station service power-is lost the RI3CCW pumps, required for engineered safety features systenis cooling or normal shutdown cooling, arc autoinalically supplied by the emergency buses. The RBCCW pump motor is designed to withstand a maximumu temperature of" 557F during a IOCA and maxinum relative humidity of 100 percent. The RFHCCW system is provided with a radiation monitor to alarm if radioactive fluids leak into the RBCCW. A failure mode analysis is given in Table 9.4-2. A rupture in the system is considered an initiating event only; it is not postulated concurrent with a LOCA (or any other Chapter 14 event). System redundancy and header separation have been provided to maintain continuous cooling in thie evnt olf'a single passive filure during post-accident long tenn cooling. 9.4.4.2 TeILst and Inspection One RBCCW pump was shop-tested for hydraulic perfonnance. Seven lest points were used to generate the perfomiance curve which is shown in ligure 9.4-7. Each pump was hydrostatically tested at 375 psi.. 1.5 times the design pressure of the pump. Nondestructive testing for the RB3CCW heat exchangers were perl'ormed in accordance with the ASME Code Section VIII. The RBCCW heat exchangers were leak tested as follows:

a.

Shell side gas. test (inert gas) at 188 psig.which is 1.25 times the design pressure.

b.

Shell side hydrostatic test at 225 psig which is 1.5 times the design pressure. 9.4-9 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 10 of 12 NIIPS-2 F*SA R C. Channels hydrostaltic test at 250 psig which is twice the tube side design pressure. d-Tubes, hydrostatic test at 150 psig, which is 1.2 tims., tile tuhe side design presstre. Each component is inspected and cleaned prior to installing it in the system. Following installation., hc RI3CCW s,'stean undergoes a preoperational test before startup. The detail lest procedure is described in Section 1.3. lnstrumneLs are calibrated during testing and the automatic controls are tested for actuation at the proper set points-Alarm fimctions are checked For operability mad Ilimits duringpreoperational testing. inilially die system was opcrated and tested with regard to flow path.:, flow capacity and mechanical operability. At least one pump was tested online to demonstrate head and capacity (Section* 13). Data are taken periodically during nornial operation to confimi the heat transfer capabilities: Major components olthe system such as pumps, tanks, and heat exchangers are accessible For periodic inspection during operation. Thle RHCCW system and components are periodically tested to demonstrate the operability, performance, structural and leaktight integrity of the system mid components. "Ie testing is accomplished as Follows: During nornial operation the RBCCW system components and piping are cons tanlly prCsstlri/.ed by the RIiCCW punip, therefore, they are constantly being tested For structural and leaktighl integrity. 'lhe operability and perforrmance.or all components except the enginecied safety teatlu'es room air recirculation and cooling units, the shutdown cooling heat exchanger, the spare RHCCW pump and the spare RICCW heat exchanger is determined by observing the temperature. pressure and flow instrtmentation associated willi each component. Online testing of the spare RBCCW component(s) (pump andlor heat exchanger) is pertbmled by bringing the spare component(s) to be tested into operation w\\'hile removing the operating component(s) fi'om service. The operation and performance of the spare RBCCW pttmip and heat exchanger can be determined by observing the pressure and flow instruimentation. Online testing of the reiotely operated valves actuated by SIAS signals and located on the discharge side ofthe containment air recirculation and cooling units, the spent lVlel pool heat exclangersm the safeguards room air cooler units. and the degasifier effluent cooler is perfonned by operating cac valves manual control 9.4-10 Rev. 31.1

Serial No 13-419 Docket No. 50-336 1, Page 11 of 12 .MPS-2 FSAR swijch. Tlheposition of the valves is verified by position indication lights in the control room. Testing these valves during power operation by initiating SIAS signals is not perlormed as it results in unnecessary starting of the emergency diesel generators and causes RBCCW system flow transients which operators must respond to. Online testing of the other remolely operated valves actuated by S.RAS is pertbnned by closing the appropriatermanual valves and by mnanuallyviniiating the SIRAS signals for the valves on the discharge side of the shutdown heat exchangers. The position ol'the valves is verified by position indicated lights in the conlrol room. b.. During shutdown the operability and performance ofthe RHCCW components and the RICC\\V system as a %whole is determined by observing the flow. pressure and temperature instrutnentation. C. After normal unit shutdown for refueling, the applicable portion of the RBCCW system used for engineered saflety ficature systems cooling is tested by actuating SIAS and S'RAS signals to demonstrate operability.and position ofvalves on a per facility basis. Structural and lealtighl integrityoflthe. s.m.tem is tested separately as part of the tinit 2 Inservice Inspection Program. The RI3CCW containment isolation valves for the supply and return lines lbr the Containment Air Recirculation and Cooling heat exchangers have "-lRingC seats. In lieu of test resuls under IOCFR50. Appendix J. Type C lesting, the "T-Ring" seats will be replaced based on observed degradation. Iliese valves are not included in the "l'pe C testing program because the valves are open during accident conditions. This eliminates a commitment made under letter A06107. dated t/16/V7 under D)ocket Number 50-336. 9.4.5 CODE STRUCThTRATL QUALIFICATION. The code of record for the R3CCW piping system is USAS B31.1-1967, The portion of piping in the containment penetration area is designated as ANSI B.31.7. Class 2. The piping and support components are designed fbr the design basis loads for the appropriate load combinations and criteria. The original code qualification was performed fbrthe following load conditions and eftltcts: (a) internal pressure. (b) dead weight. (c) thermal, expansion at maximumn tempcratures, (d) 013P inertia loads, (c) DBE inertia loads, and (f) seismicanchor movements for ODE and DBE. as applicable. The system and components are evaluated for the newly identified LOCA/water hammer loads, including the original design basis loads. considering temperature changes based on zero fouling in the CAR Cooler units. The postulated r OCA induced dynamic.water hammer loads based on the scenario destribed in. G. 96-06 is classified as "Faulted Plant Occasional Mechanical load" because (OCA is a "Faulted Plant Condition." Althotigh not part ofthe original design basis. this load is rceognized as a design basis load. 9.4-11 Rev. 3 IA

Serial No 13-419 Docket No. 50-336 1, Page 12 of 12 MIPS-2 FSAR Ile acceptance criteria for piping is based on maintaining the combined primary stress due to concurrent dead weighl. internal pressure and dynamic LOCA/water hamnnertransient within the material yield stress of piping at operating temperature. 'Ilie acceptance criteria Fbr design verification of support comnponecnlts corresponds to conservative Emergency strcws limits (i.eC.. ASME Level C Limits) as follows, with a minor exception as noted in item b; (a) structural steel eleim ents-within 90% of yield per AiSC Code (h) vendor supplied components I A)ad Capacity per Load Capacity Data Sheets (LCD) Level C, except those supports whfich are not protecting the safety related function of the RBCCW system in the post-accident scenario may be evaluated using LCD Level D (c) anchor bolls safety factors shall be 4 for Wedge type (e.g., IILTI KWIK BO1 I') and 5 for shell type (e.g.. Phillips Red Head) consistent willi NRC I.E. Hulletin 79-02. The LOCAvwater hammer is analyzed using dynamnic time history analysis melhod. The tmne history forking functiorns applicable to various segments ef RHCCW piping are determined by detailed thermal hydraulic analysis of the system using RELAP 5/Mod 3.2 computer code. 9.4-12 Rev. 31.1

Serial No 13-419 Docket No. 50-336 2 Updated Final Safety Analysis Report Section 14.8 "Millstone Unit 2 FSAR Events Not Contained in the Standard Review Plan" (Revision 31.1) Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 1 of 17 MPS-2 FSAR 14,8 MILLSTONE UNIT 2 TSAR EVENTS NOT CON'IAINED IN'T'IIE S'I'ANI).AD REVIEW PLAN 14.8.1 FAILURES OF EQUIPMENT WIIICIH PROVIDES JOLNT CONTROL/SAFETY FUNCTIONS 5Millstone Unit 2 has no instrumentation Which serves a combined function of process control and o' initiation o0'emrgency satytv systems. 14.S.2 CONTAINMENT ANAIYSIS 14,8.2.1 Main Steam Line Break Analysis 14.8.2.1..1 Event Initiator In the event or a Main Stearn Line Break (NMS1.J3), the release or stearn into containment will result in a rise in both temperature and pressure. The break is assumed to occur in the piping between ihe steAni generator and the containment wall penetration. Mams and energy ruleases are limited by the flow reslrictor in the steam.generator outlet nozzle. 14.8.2.1.2 Protecivc Systems Eingineered Sal'ely Fealures (ESF) systems which will operate to tenninale the mass andenergy release to containmient and suppress containment atmosphere temperature and pressure art the Main Steam Isolation Signal (MSIS). Safety injection Actuation Signal (SIAS), and Containment Spray Actuation Signal (CSAS). A MSIS will actuate on receipt ola containment high pressure signal to shut the floiowing valves and trip the main feedwater pumps: I. Steaim Generator I & 2 Isolation Valves. (HIV-4217 & [V-4221 or MS-64A&B)

2.

Steam Generator 1&2 Isolation Valves Bypass (HV-4218 & HV-4222or MS-65A&B)

3.

Steam Generator I & 2 Feedwater Isolation Valves (1.'-5419 & HV-5420.or FW-SA&H)

4.

Steam Generator I & 2 Feedwater Regulating Valves (FV-5268 & FV-5269 or FW-5lA&B3)

5.

Main Steam Leg.Low PT. Drains (HIV-4193 & IIV.4209 or MS-26513 & MS-26613) 6, Steam Generator 1. & 2 Feedwater Regulating Bypass Valves 14,&.1 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 2 of 17 NIPS-2 FSAR (FV-52.15 & FV-5216 or I"W-41 A&13) 7, Feedwator Block Valve to Steam Generators I & 2 (IIV-5263 & IIV.5264 or FW-42A&B) S. Siearn Generator 1 & 2 Feed Punip Discharge Valves (I'IV-5245 & [IV-5247) I113-2 A SIAS will activate the Containment Air Recirculation (CAR) fans and give a start signal to the emergency diesel generators (l DGs)., 14.8.2.1.3 Method (if Analysis A complete MISLB spectrum study has been performed to determine the limiting eases tfr peak containment pressure- 'the NRC approved methodology (References 14.8-2 and 14.8-3) 132 associated with the Westinghouse SGN-Ill computer program was used to determine the mass and energy releases to containment. Using Ihese nmass.and energy releases, the NRC approved Dominion GOTIIIC methodology (Relerence 14.8.4) was used to determine the containmnent. pressurctemperature consequences ol'the MSI.B.This methodology includes consideration fIr the. fbllowing. (a) inclusion ol'tho steam line and. reed line volumes into the overall determination of blowdown voltune available. (b) determination of temperature/pressure expansion factor ror the S(is and R CS to maximize the volume available for blowdown,: (c) increase in i'cedwater Ihow to th, affected SG due to the increasing pressure imbalance between the affected and intact SG; (d) inclusion of SG shell metal heat transfer as part ofthe energy release: and lastly. (e) a complete determination of the effects of diflerent component single failures during the accidenL 14.8.2:1.4. Major Assumiptions The maJor assumptions are as follows: I Otfsite power is assumed to be available for most of the cases. This increases the 13=2 primary to secondary hial transfer since lhe reactor coolant pUMps (RCPs) are operating. 'tb verify this assumption, loss orollmsite power eases were included as pan ol'the single failure analysis. "2. For determination of peak containmniet pressure., the initial contaimnent pressure" temperature is conservatively assumed to be at. the Technical Specification maxi.mwin of 15.7 psia and 120"F.

3.

Consistent with the NRC Standard Review Plan (SRP) Section 6.2.1.4, break. spectrtun studies were used to address moisture carryover 4 Preferential addition of feedwater to [he affected SG is accounted ror by Lonservalively doubling the initial feedwater flowrate as long ax the. MFW systean is operating. 14.8-2. Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 3 of 17 INPS-2 FSAR 5.. Credit is taken fbr the main steam non-return valves to prevent blowdown ofrthe unaffected SG into the containment.

6.

nhe maxinuan RCS flow rate was conservatively assumed to ia.\\inize theheat tmnstfr from the primary to secondar" side.

7.

Cases initiated from.0% power'assunied auxiliary" feedwater (AFW) is tile.solc source of steam generator inventory control and AFW flow tothe affected steam gelerator is maximized lkon the beginning (time

0) of the analysis. All other cases initiate AFW ala conservative miniuniu time of l120.seconds.

.9, Reactor coolant pump heat was included.

9.

All actuation signals are redundant and safety grade. In sonic cases credit is taken for actuation of nonsatfety grade components initiated by the safety grade signals.

10.

Relative humidity is assumed at 25 percent.

11.

A 0.75" auxiliary steam line located between the two steam generators remains unisolated during the events. Thiis causes tie intact steam generator Lo continue to blowdown even after the MSIS.

12.

A cavilating venturi installed in each AFW discharge line will limit AFW Ilow to a steam generator to 530 gpm. 13-1

13.

Operator action to isolate the AF W is assumed to occur no greater than 30.niinutes rollowing MSIS. L3-

14.

The fixedwiter flow rate tor the cases that initiate witlhthe main feedwater syslem operating conservatively account the following:: Feedwater flow increases as the affected steam generator pressure decreases. ULpon receipt nfa MSIS. the feedwater pimps are assumed to coast down at a rate based on plant operating. experience. If tho ruptured gteam generator pressure dccreosme to the discharge pressure of the still running condensate and heater drain pumnps, flow will again begin to increase. Flow is eventually stopped aftler the conlainment high pressure signal actuates the MSIS and the main feedwater is isolated. 14.8-3 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 4 of 17 N PS"-2 FSA IR 14.8.2.1.5 Initial Conditions and hiput D)ata Initial. conditions-and input data are given in Tahles 14,X.2-i, i 4.8.2-2 and 14.9.2-5S Table I.4.8.2-3 gives al accounting of the amounts of steel assumed to be hiside of contaiinmint. Since this material acts as a heat sink to reduce containment temperature and pressure., minimumn 13-2 amounts are used. 14.8.2.1.6 Results In order to detemrine the limiling conditions, ibur different spectrum studies were perfonned. 'Ilqese are as tIllows: I. Power level Mid break siMe:

2.

Feed system single failures,

3.

Containment heat removal systems single failures.

4.

Spectnum study tfr peak containment temperature. 14.8.2.1.6.1 Power Level and lrcak Size A comprehensive sensitivity study was perlbnied to detemnine ihe limiting break.size for each power level. A sensitivity study was needed because of tie interaction of power level with SG hiventory and moisture carryover. The limiting break size at a given power level is the largest break size that would result in a pure steam blowdown, since a pure steam blowdown results in the greatest amount of energy beingtransftrred to the containment almosphere in a short period of time. 'Th1e limiting results for each power level show that a maximum br'ak size of 3.51 112 is liniting for 25 percent, 50 percent. 75 percent. and 1 O0 percent power. At 0 percent power the 13-2 limiting size break is 1.89 ft.. 1 14.8.2.1.6.2 Feed System Single Failtires A comprehensive feedwater system isolatiun single failure study was perlronned.. Foreach single failure, a range of steady state initial power levels was analyzed. using the insights from the Power Level/Break Size sensitivity study. I. Feed Pump Failure to Trip - he failure of a feed pump to trip on MSIS results in additional feed water being pumped preferentially into the affected SG until the Feedwater Regulating Valves (FRV) and isolation valves shut. For 25 percent and 50 percent power levels, only one feedwater pump was assumed toberunning when the accident commences. This event was not applicable to 0% cases (see assumption 7 of Section 14.8.2.1.4). 14.8-4 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 5 of 17 NMPS-2 FSAR

2.

Inadvcrtent Initiation ofAFW Feedwaterto Affected Steam GOnerator-Maximum AFW Ilow was assumed to be inadvertently initiated at thestart ofthe event for 13-2 cases initiated from greater than 0% powcr.

3.

Feedwater Bypass Valve Fails Open -This failure is only credible when the FW bypass valve is initially open (cases initiated from 25% power and below). TIhe failed open feedwaxcr bypass results in additional feedwater being pumped preferentially into the affected SG until the FW pump discharge valves shut. In addition. even with the feed pump discharge xalves shut, flashing in the feedwater lines continues to add energy into the affected steam generator. This effect has been taken into account. 4.. Failure of Vital Bus Cabinet VA-JO or VA-20t This Ihilure could prevent closure of the FRVs and results in flie loss of one train of the Contaimnent Hleat Removal Systems. Feedwateraddition to the affected SG will continue until closure of the main reed pump discharge valves. 13-2 14.8.2.1.6.3 Containment lfeat Removal Systems Single Failures A comprehensilve containment heat removal systems single fisilure.swudy was perlfmnned. For each single failure, arange, o steady state initial power levels was analyzed, using the insights from the Power Level/Break Size sensilivit' study. 1.. Failure of'lWo CAR falns to start - TIhis failure is bounded by Section 1:4.8.2,1.6.2, item 4 described above.

2.

Failure of one spray train to start - This failure is bounded by Section 14.8.2.1.6.2. item 4 described above.

3.

Failure of the Vital: Bus Transtlr lMechanism - This failure results in a loss of the normal off-site power supply Ior the vital buses. llhus initiation ofthe containment sprays and C.AR falns is delayed until the EDGs are poweringthe vital buses and auto sequencing has occurred. Since tie.FRVs have a backup l)C power source, they are unaffected by this failure and will isolate the alflctcd SG 'Phe RCPs and certain other nonvital loads are also unaffected by this failure. which contributes to tile severity of this accident by providing more rapidimeca transfer from the primary to the atfected SG I03-2 .4. Loss of Of0iite Power with a Loss of One EDGO-A loss of ofl'site power will result in loss of power to the RCPs, the condensate pumps amid feedwater heater drain pwnps. While onlyone train otcontainment heat removal Systems is available, the loss of power to these pumps results in a greatly degraded heat transfer in the allected SO and less limiting results. Feedwater isolation will.be unalfected since. the FRVs are powered by DC backup power supplies. 14.8-5 Rev. 3t.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 6 of 17 NIPS-2 FSAR 5.. Loss of OfTsitc Power with a Loss of VA-10/20 -This case is similar to item 4 (the preceding paragraph), with the exception of the ef. et on feedwater isolation to the affected S(i With this failure. there is the potential for I hilure of'the.FRV and the Other isolation valves to close. llowever. with the loss of the e*ndcnsate mnd teedwater heater drain pumps, feedwater addition to the affected SG is termitnated. '11he effect of continued energy addition to the. aftherted S(O from flashing in the feedwater lines has been taken into account. 14.8.2.1.6.4 Maximum Containment'Temperature Spectrum Study lbe limiting peak pressure cases were re-run with the tbilowing modilied ttssumption. to 1)3"2 maxilllize resultant containment temperature. I. The initial containment pressure was reduced to 14.27 psia. This results in the maximum delay in containment spray actuation.

2.

The relative humidity was increased to 100 percent.

3.

lThe.MSLB mass and energy releases model the Steam Generator steam super 13-2 heating as it passes tile uncovered portion o' the Steam Generator tubes belbre exiting the break to addrss 11' Information Notice 84"90. "11e containment wall re-evaporation is modeled using the GOTHIC built-in models for calculating the vaporization of the liquid in containment as described in Reference 14.8.4. 14.8.2.1.7 Conclusions The results ofa MSLB initiated from 102 percent reactor power with coitncident loss of.ffsite 13-2 power and the failure oflthe Vital Bis Cabinet VA-10 or VA-20 produces the limiting containment peak pressure of 53.8 psig. The peak containment atmospheric temperature for this case is 325. l.1. With the loss of oftfite power and this single tailure, the condensate and feedwater heater 13-2 drain pumps are lost, and pumped 1eedwater addition to the ait'cted steam generatortis quickly terminated. However, since the FRV and the other feedwater iotalion valves fail to close on the NiSIS, a significant volume of feedwater system remains connected to the affected steam generator. As the affected steam generator depressurizes. feedwater in this system flashes and adds significant mass and energy to the alrected steamn generaior. The portion of tile feedwater that flashes to steam is conservatively assumed to be directly added to the containment atmosphere separate from the mass and energy releases Irom the steam generator, The portion of the feedwater liqtuid that reaches the affected steam generator as it depressurizes increases the affected steam generator liquid mass, which increases the steamn generator mass and energy releases Ito containment. The plant response for the limiting peak pressure ease is shown in Figures 14.8.2-1 thirough 14.8_2-9 mid the sequence otf eents is given inTable 14.8.2-4. The results of MSLB initiated front. 102 percent reactor power with offsite power available amd 13,2 the failure of Vital Bus Cabinet VA-10 or VA-20produces the limiting contaiunment peak atmospheric temperature of 360.9'F. 14.8-6 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 7 of 17 NIPS-2 FSAR the conmainment pr,'ssure remains below the design pressure of 54 psig. Although peak containment atmospheric temperature exceeds the 289ý'F designtemperature, it is only ftr a short period of time and does not raise the containment structure above 289'F. 14.8.2.2 Loss of Coolant Accident Analysis 14.8.2.2.1 Events.. nalyzcd Twenty four separate cases of Loss of Coolant Accidents (LOCAs) were analyzed with variations 13-2 for break locations, break. sizes, single failures and availability ololl'site power. Break locations anal)zed.are the reactor coolant pump suction leg. pump discharge leg ard hot leg. Break sizes include double-ended guillotine and slot breaks (9.82 sq. it. area fbr the reactor coolant pump suctioni and discharge leg breaks, 19.24 sq. 1t. area for a hot leg break) and smraller break. sizes for hot leg breaks (10 sq. ft. and 2 sq. ft.) and reactor coolant pump suction and discharge leg breaks (5 sq. ft. and 2 sq. t.), Single failures considered are tailure oftan emergency diesel generator (EhDO) (for a loss of power (L OP) case) which fails I train ofcontainment heat removal systems or (for no LOP). failure of either 1 spray system or 2 CAR fans. With an LOP I wain ofECCS will operate, with no l.OP both trains will operate. S13-2 14.8.2.2.2 Method of Analysis The NRC approved. Westinghouse containment analysis methodology was used f6r the I 132 development of the shour t rn mass and energy releases following a L.OCA (Rdelfrenee 14.8-3). Mass and energy input are provided through the End-of-&Blowdown (EOB) from CEFLASH-4A, and from the EOB to End-of-Post Reflood (I'.OPR) from Fi,.(.X)I)3. Ilie long term boil-olyphase mass and energy input was calculated using the Doininion GOri TIC code (Rel~rence 14.8-4). The 113-2 containment pressure and temperature response for the entire LOCA transient was calculated using the Dominion GOTHIC compLtttr code. 13-2 14.N.2.2.3 Input and Assumptions

a.

Containment input data, such as:.hentsink area., spray fow rate. CAR fal cooler heat removal rate, spray water temperature-containment volume and initial containment temperalureare tlhe same s, oroimore coniservative than, tlht i.sed for 113"2 the MSLB in section 14.8.2. 1.

b.

Initial containment pressure is 15.7 psia for the peak pressure case and 14.27 psia 13-2 for the peak temperature case. C. Initial containment humidity is assumed to be 25 percent as in the MSI B,1 Ijr the. peak pressure case and 100 percent for the peak temperature case. 14.8-7 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 8 of 17 M PS-2 IFSAR

d.

'1110 minimui usablu Refueling Water Storage Tank (RWST) volume assumed for calculation of the time of Sump RecirculationActuation Signal (SALkS) is 370:060 I 13"2 gallons. eý Both the HPSI and the LPSI pumps operate prior to SRAS. Following SRAS. the I PSI are automaticaIly stopped.

f.

The Reactor Building Closed Cooling Water (RBCCW) is modeled with assumed flows before and after SRAS; Tihe RI3CCW is cooled by Service Water at 801'. 13-2

g.

"llie heat removal from the CAR ftin cooler is modeled in the Dominion (iOTlIC Code using a fan cooler mnodel benelnm-uked to the post-LOCA specification data. Th1C specification identifies that one CAR fan is capable of removing a rinimwum of 80 million BTU/hr based on a containment air'inIct 1cmperature of 289"17 aind a lfn llowrate ot 34,800 cthm. along with a cooling water inlet temperature or 130('F and a Ilowrale of 2000 gpn. .11 A minimum spray flow or 1300 gpm is credited prior to SRAS, and 1350 gpm following SRAS. 14.8.2.2.4 Results 'Tlhe limiting L.OCA was dtennined to be the 5 square foot discharge leg break with the lOE the 113-2 lailure ol'two CAR fans and one spray train, and mhiimum. ECCS. "The maximum containment presslue and temperature of this limiting LOCA are bounded by the M SLB results provided in section 14.8.2.1. 14.8.2.2.5 Conclusion The maiimum containment pressure and temperature of1'he. LOCA are less than the containment design pressure mnd temperature of 54 psig and 2899' 14.8.3 DEILETED 14.8.4 RADIOLOGICAL CONSEQUENCES OF THE DESIGN BASIS ACCIDENT 14.8.4.1. General A LOCA would increase the pressure in the containment resulting in a coontainment isolation and initiation of the ECCS and containment spray systems. A SIAS signal automatically starts the Enclosure Building Filtration Syslem (FIFS) which maintains a negative pressure within the enclosure building during accident conditioms. The nuclide inventory assumed to be initially available (or release is consistent with the requirements of Regulatory Guide 1. 183 (Reference 14.8-5). A SIAS also isolates the control romon by closing the firesh air dampers wilhin 20 seconds. Within 1 hour alter control room isolation, the control room emergency ventilation (CREV) is 14.8-8 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 9 of 17 NI PS-2 ESA R properly aligned. CREV recirculates air within the control room through a*charcoal ilter at 21500 clmn (+/-10%) to remove iodines from the control room envelope. The radiological consequences of a Design Hasis LOCA at Millstone 2 were previously analyzed lor a low mid high wind spced condition based on guidance front Regulatory Guide 14 (Reference 14.8-6) and SRP 6.5.3 (Reference 14.8-7). The low wind speed case was found to bound the high wind speed case. Therefore, the low wind speed case is the design basis for a LOCA and the high wind speed case is no longer analyzed. 14.8.4.2 Release Pathways The release pathways to lhe environmeni subsequent to a LOCA are leakages fromn containment and the enclosure building, which are collected and processed by EIWFS. and leakages.firont containment and the RWST which bypass FliES. Containnent Leakage Thelc containment is assuned to leak-at the design leak rate for 24 hours after the accident. Aller 24 hours. since the pressure has been decreased significantly. Regulatory.Guide 1.183. allows for the leak rate to be reduced to one-half the design leakage rate. All containment leakage for the first 110 seconds is assumed to bypass I.BFS and is released directly out the MP-2 containment. This is due to the fact thatit takes 110 seconds for EBFS to achieve the required negative pressure in the enclosure buildiig, thereby, ensuring that leakage will be into the enclosure building rather tham out. TU3FS collects most ofthe conlainent leakage and processes it through IIEPA and charcoal ..filters and releases it up the Millstone stack. All containnment leakage is collected and filtered by I-1IFS except for the small amount that is assumed to bypass iCI'S and is released directly out the MP-2 containment. Credit is taken for iodine removal due to containment sprays. The sprays are effective after 75 seconds post-I OCA. "lic effeetiveness of' the sprays in renioving elemental iodine ends at 3.03 hours and in removing particulate iodine at 3.23 hours. Credit is taken lbr iodine retention in the: containment sump based on pUst-L.OCA sump pH Ž 7.0 as discussed in Scctioin.62,4. 1. ESF Svs-lem Leakage Palhwav Post-accident radioactive releases ti-em the ESF system are derived friom fluid leakages assumed during recirculation of the containment sutmp water through systems located oLutside containment. The nuclideinventory assutmed to be available for release friom this pathway consists of 40% of the core iodines. The quantity of leakage is based on the assumption that the ESF.equipment leaks at twice the niaxinumn expected operational leak rate. and that 10 perceen of the iodine nuclides contained in the leakage fluid become airborne in the enclosure building. The nuclides which become airborne are collected and released to the environment through EBFS to -the Millstone stack. 14.8-9 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 10 of 17 ,M PS"-2 FSA R RWVST 3ackleakage Pathwav Post-accident radioactive releases firorn the ECCS system area result of ECCS subs-ystenIs containing recirculated sump fluid backlkaking to the RWST. hlie backflow rate to the RWST; as a result of isolation valve leakage, is predefined and time dependent. Due to this time dependency, the contaminated StMlP fluid from backleakage does not enter into the RWST until 6.45 hours post-LOCA. Since the RWST is vented to atmosphere. the release is a result of the breathing rate ol the RWST due to solar heating. 14.8.4.3 Control Room Habitability The radiation design objective of the control room is to limit the dose to persontel inside the control room to 5 rem T-EIDE. during a DIBA. lte potential radiation dose to a %ontrol room operator is evaluated for the LOCA. The analysis is based on the assumptions and meteorological parameters (XiQ values) given in a :ltbles 14,8.4-3 and.14.9.4-4. 'lte control room is designed to be conttiouoISly..occlpied for the duration of the accident. 30 days..Two basic sources of radiation have been evaluated: leakage of airbonic activity into the control room from sources described in 14.8.4.2and direct dose Frotn sources outside tihe control room. tlie control room shielding serves to protect the operators From direct radiation due to the passing cloud of radioactivelcffluent assumed to have leaked front containent, enclosure building and the RWST. The control room walls also provide shielding proleetion for radiation emanating froom the CREV filters and contaiunent shine. A SIAS front Millstone 2 initiates control room isolation Within 20 secondsby securingthe fresh air intake dampers. Within 1 hour, CREV is ht operation recirculating air in the control room envelope through charcoal filters to remove radioactive iodines froni.theatinosphere.. "1he calculated TEDE dose from a Millsione 2 LOCA is presented in Table 14.8,4-5 and Li below the General Designi Criteria 19 and 10 CFR 50.67 limits. 'l1ie calculated TEIE dose from a Millstone 3 LOCA to the Millstone 2 control room is below the General Design Criteria 19 limits and bounded by the dose conseqtluenees front the Millstone 2 I,OCA. No credit has been taken for control room isolation or CREV operation. 14.8.4.4 0fTlite Dose Computation The radiological ofllite dose consequences resulting fitoni a postulated Millstone 2 LOCA are reported in Table : 4.8.4-2. The ofi'site dose analysis shows that the consequences to the EAB (highest 2 hour) and LPZ (0-30 day) are less thatle limit ol"25 ren TFDI" as specilied in 10 CFR 50.67. lte assuntptionti used to perform the radiological analysis are stunmarized in.Table 14.8.4-1. 14.8-4.5 Concltsion Analysis shows that the oltsite and control room radiological consequences are within 10 CFR 50.67 criteria. 14.8-10 Rev. 3 1.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 11 of 17 NIPS-2 FAR 14.

8.5 REFERENCES

14.9-I Deleted. 14.8-2 Preliminar, Safety Analysis Report (PSAR) to CESSAR. Appendix 6B. "'Description of I 3-2 the SCON-I Il Digital Computer Code Used In Developing Main:Stam Line Break Mass,'EnerE: Release Data For Containment.Analysis," 14.8-3 NRC Safely Evaluation Report. - Stwndard Reference System CESSAR System 80. Combustion Engineering. ln., December 1975. 14.8-4 Dominion Topical Report DOM-NAF-3, Revision 0.0-P-A. '6OTHIC Methodologi 1,2 For Analyting the Response to PTstulated Pipe Ruptures h.iside Containmenk't Nptember 2006. 14.8-5 Regulatory (vuide

1. 193-Alternati've Radiological Source T'emis for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000, 14,8-6 Regulatory Guide 1.4, Assumptions ttmsed for Evaluating the Potential Radiological Consequences ofa Loss or Coolant Accident for Pressurized Water Reactors Rev. 2.

June 1974. 14.8-7 SRI' 6.5.3, Fission Product Control Systems and Structures., 14.8-11 Rev. 3.L.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 12 of 17 M PS-2 FiA R TABLE 14.8,2-1 CONTAINMENT DESIGN PAIR\\METERS [iternai dimensions (feet) Cyvliinder wall diameter 130.0 Cvflildcr wall height 132.4 Curved dome helighl 43.3 1372 Net fiee internal voluime 1:899.000 cubic Ceel Page I of). Rev. 3 1 I

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 13 of 17 NIPS-2 FSAR IABLE 14.8.2-2 INITIALCONDITIONS FOR PRESSURE ANALYSES Reactor Coolanut Systern Corethermal power level (MWtI,/% of rated power). 2754+102 Reactor coolant Pump heat (M.WI) 17.1 Coolant pressure (psig) 2300 Inlet coolant temperature (OF) .551.25 Internal coolant volume (culbic rect).(exludes the pressurizer) 10.104.4 Contaimnent System Pressure (psia) 15.7 Relative humidity (%) 25 inside temperature (F) 120 Outside temperature ('1F) 'N.A Service Water Inlet temperature IF) 80 Refiueling Water Stoiagc Tank RWST) water temperature ('F) IQ)) Salety Injection Tank (SIT) water temperature (T) 120 1 13-1 B 1-2 I13-21 I No heat transfer crediled From contain*ent, structure to outside environment. Page I of RRev'. 3 1.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 14 of 17 TABLE-2 I3MM C AIMR TABLE 14.8.2-3 MINIMIUM-CONTAINMENT HlEAT SINK DATA Heat Sink Exposed Surface Area (sq tt) I. Containment Cylinder and Dome Cylinder 52,X00 Dome 19,07 71.870

2. Unlined ConUTCel Sicamn Geerator Compartinent Walls 26,114 Miscellaneous Slabs 4.488 ElevaLor :Foundation 643 Pressurizer Wall and Roof 2.159 Refueling Canal (Outside) 10,043 Steam Generator Pedestals 21860 Steam Generator lilutresses 3,840 Fuel Canal Buttresses 3X3 53,420

3. Reactor Support Concrete 3,486 (3 inches, exposed on one side to the containment atmosphere and on the other to a 150"F source to acucount Fbr the higher reactor cavity operating lemperaturej

4. Galvanized Steel 1A16,497

5. Painted Steel Less than 0-12 in, "lbl 5,605

6. Painted Steel 0.12 to 0.16 in Thick 16,863

7. Painted Steel 0,16 to 0.24 in. Thick 36,713

8. Painted Steel 0.24 to 0.3 in. 'Thick 10,289

9. Painted Steel 0.3 to 0.4 in. Thick 19.366

10. Painted Steel 0.4.to 0.5 in. Thick 4,525 I13.21 113-1 I113-1 I13-2 Page I of 2 Rev'. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 15 of 17 NJ PS-2 FSA R TABLE 14.8.2-3 MINIMUM CONTAINMENT HEAT SINK DATA Heat Siik Exposed Surface Area (sq ft)

11. Painted Steel 0.5 to 0.625 in. T11ick 5,338

12. Paited Steel 0.625 to 0.75 in..ik 2.243

13. Painted Steel 0.75 to 1.0 in. Thick 2862

14. Painted Steel 1.0 to 1.5 in. Thick 4,322

.15. Painted Steel Greater than 1.5 in, Thick 1.031

16. I.tipainted Stainless Steel 18,464

17. COntainment Flnor 8,102

18. Safety Injecticm Tanks 2.541 13-2 13-2 13-2 13,2 I13-2 I13-2 Page 2 0o2 Rev. 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 16 of 17 MPS-2 FSA'R TABLE 14.8.2-4 SEOUENCE OF EVENTS. MP2-MSLB: LOSS OF OFFSITE POWER AND THE FAILURE OF VITAL BUS VA-10 OR 01VA-20 FROM 102% POWVER TIME (,*eonds) EVENT SETPOINTIVAIUE ,0.0 MSLB occurs from 102%power, break size is 3.51 lt2. 0.0 Loss of oftfiile power. 1.29 Low RCS flow reactor trip condition reached. 89.7% oFinitial RCS flow 1.94 Low RCS flow reactor trip signal generated. 0.65 seond delay 2.50 Containment High Pressure Signal (ClIPS) 5,83 psig with condition is reached. (Due to the assumed loss of tmcertainty offsite power and failure of VA-10 or VA-20. the aiffeted steam generator FRV and main ieedwater isolation valves do not losc.) 3.40 HIigh Contaiiunent Pressure MSIS generated. 0.9 second delay 6.15 Containment High-fHigh Pressure Signal 11.08 psig with (CIIlI PS) condition reached. Uncertainty 28.5 Containment cooling fans energize. Time based on CHPS, 26 second delay 37:1 Peak Containment Pressure reached. 326.50F 74.76 Containment spray flow commences. See Note I Time based on CHHPS ý 68.6 seconds tbr pump start, valve stroke time, and header rill lime. 1 80 Maximum AFW Flow to 1heAfteeled Steam 5.50 gpn. 10("F (enerator. 552.1 Peak Contaimnent Pressure reached. 53.8 psig 1000 Simulation ended. 13-2 I13-: 13.2 13.2 13.2 13-2 13-2 13-2 113-2 I13--" 113-2 113-!. (1) The. minimum conlainment. spray flows used in this ulalysis are provided in Table 14.8.2-5. 113-2 Page 1 ofI3 Rev., 31.1

Serial No 13-419 Response to RAI for MPS2 Ultimate Heat Sink 2, Page 17 of 17 MPS-2 FSAR TABLE 14.8.2-5 ENGINEERED SAFETY FEATURES PERFORMANCE FOR MSLB C()NTAINMENr ANALYSIS 13.' ,afeAy ,'enturts Value Notes

1. Containment spray

- water temperature 1 QooE - CSAS setpoint 11.08 psig TS1 value plus Lincertainty - minimum flow rate 1361 gpm at 54 psig Values based on a minimum 1375 gpm at 51 psig RWST level of 30. feet above 1394 gpim at 47 psig t1nk. bottom. 1428 gpm at 40 psig - delay time with nornal AC 49 seconds Includes 33 seconds for header power availablk rill time and 16 sccunds fur signal generation, pump slarl and valve stroke. - delay lime with the loss of 68.6 secondis Includes 33 second-.,ror header. normal AC power fill time and 35.6:seconds for signal generation. pitnip start and valve stroke.

2. Containment Air Recirculation (CAR) Cooling Falls

- number of fans 4 2. or certain singlO faihure clues -.activation selpoint 5.83 psig I'S value plus umcertainty - delay time with normal AC 15 seconds power available - delaytime with toss of normal 26 seconds AC power -beat removal capability of one.80 million BTUlir based on A (C)TH IC Fan Cooler Model CAR Fau air inlet temperature of is tsed. T11his model is 289TF and a fan flow rate of bLnchniarkcd to the citc'd 34,800 cfi. along with a specificaliou data. cooling water inlet lemperature of' 130'F and flow rate of 2000 gpm.

3. Safety injection None Ags*um'ed.

Page 1 of I Rev. 3 11.}}