RBG-47823, Request for Review and Approval of Relocation of the Reactor Core Isolation Cooling (RCIC) Injection Point from the Reactor Vessel Head Spray Nozzle to the a Feedwater Line Via the a Residual Heat Removal (RHR) Shutdown Cooling

Request for Review and Approval of Relocation of the Reactor Core Isolation Cooling (RCIC) Injection Point from the Reactor Vessel Head Spray Nozzle to the a Feedwater Line Via the a Residual Heat Removal (RHR) Shutdown Cooling

ML18029A187
Person / Time
Site:	River Bend
Issue date:	01/29/2018
From:	Maguire W Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
IR 2017009, RBG-47823
Download: ML18029A187 (91)

Text

{{#Wiki_filter:,. .~. Enterw. January 29, 2018 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 RBG-47823 Entergy Operations, Inc. R'IIeI Bend Station 5485 U S Highway 61 N St FrancIsVille LA 70775 Tel 225*381*4374 William F. Maguire Site Vice President R'IIer Bend Station

Subject:

Request For Review And Approval Of Relocation Of The Reactor Core Isolation Cooling (RCIC) Injection Point From The Reactor Vessel Head Spray Nozzle To The 'A' Feedwater Line Via The 'A' Residual Heat Removal (RHR) Shutdown Cooling Return Line River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47

References:

1.

2.

NRC Letter to Entergy, "NRC Problem Identification & Resolution Inspection Report 05000458/2017009" dated June 9, 2017 Entergy Letter to NRC, "Reply to a Notice of Violation 05000458/2017-009-01 " dated July 6, 2017 Pursuant to 10 CFR 50.90 and Entergy Operations, Inc. (Entergy) hereby requests approval of a revision to the Facility Operating License NPF-47 for River Bend Station (River Bend), Unit 1. The change revised the River Bend Updated Safety Analysis Report (USAR), Technical Specification Bases (TS Bases) and Technical Requirements Manual (TRM) to relocation of the Reactor Core Isolation Cooling (RCIC) injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' Residual Heat Removal (RHR) shutdown cooling return line. This LAR is not a voluntary request from a licensee to change its licensing basis and it is not subject to 'forward fit' considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 (ADAMS Accession Number ML01960180)." The LAR is in response to a Notice of Violation related to relocation of the Reactor Core Isolation Cooling (RCIC) injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' Residual Heat Removal (RHR) shutdown cooling return line. Based on the previous, inadequate 50.59, the NRC could not conclude that a spurious RCIC actuation into the feedwater line would not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the USAR.

RBG-47823 Page 2 of 3 The purpose of this LAR is to evaluate and document that the change does not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident. includes the description of the change, no significant hazards consideration determination, and evaluation for environmental impact. Attachment 3 to the Enclosure provides a copy of the marked up USAR, attachment 2 provides a copy of the marked up Technical Specification Bases (TS Bases), and attachment 1 provides a copy of the marked up Technical Requirements Manual (TRM). The proposed change has been evaluated in accordance with 1 OCFR50.91 (a)( 1) using criteria in 1 OCFR50.92(c) and it has been determined that the proposed change involves no significant hazards consideration. Although this request is neither exigent nor emergency, your prompt review is requested. Entergy will implement the amendment within 30 days of the NRC approval date. In accordance with 10 CFR 50.91 (b)( 1), Entergy is notifying the State of Louisiana and the State of Texas of this LAR by transmitting a copy of this letter and enclosure to the designated State Official. This letter does not contain any new commitments. If you have any questions or require additional information, please contact Mr. Tim Schenk at (225) 381-4177 or tschenk@entergy.com. I declare under penalty of perjury that the foregoing is true and correct. Executed on January 29, 2018. Sincerely, WFM/alc : Evaluation Of The Proposed Change - Relocation Of The Reactor Core Isolation Cooling (RCIC) Injection Point From The Reactor Vessel Head Spray Nozzle To The 'A' Feedwater Line Via The 'A' Residual Heat Removal (RHR) Shutdown Cooling Return Line : Technical Requirements Manual (TRM) Page Markups : Technical Specification Bases (TS Bases) Page Markups : Updated Safety Analysis Report (USAR) Page Markups cc: U.S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd. Arlington, TX 76011-4511 Ms. Lisa M. Regner, Project Manager U.S. Nuclear Regulatory Commission

RBG-47823 Page 3 of 3 MS-8-H4 One White Flint North 11555 Rockville Pike Rockville, MD 20852 NRC Senior Resident Inspector PO Box 1050 St. Francisville, LA 70775 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312 Public Utility Commission of Texas Attn: PUC Filing Clerk 1701 N. Congress Avenue P. O. Box 13326 Austin, TX 78711-3326 RBF1 0011 LAR 2017-09

RBG-47823 Page 1 of 10 ENCLOSURE 1 Evaluation Of The Proposed Change - Relocation Of The Reactor Core Isolation Cooling (RCIC) Injection Point From The Reactor Vessel Head Spray Nozzle To The 'A' Feedwater Line Via The 'A' Residual Heat Removal (RHR) Shutdown Cooling Return Line

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 System Design and Operation / Background 2.2 Reason for the Proposed Change 2.3 Description of the Proposed Change

3.

TECHNICAL EVALUATION 3.1 Previous and CurrenUProposed Configuration 3.2 Previous and CurrenUProposed Changes to the Licensing Basis

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration Analysis 4.3 Conclusions

5.

ENVIRONMENTAL CONSIDERATION ATTACHMENTS: : Technical Requirements Manual (TRM) Page Markups : Technical Specification Bases (TS Bases) Page Markups : Updated Safety Analysis Report (USAR) Page Markups

RBG-47823 Page 2 of 10

1.

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests a license amendment to the Facility Operating License NPF-47 for River Bend Station (River Bend), Unit

1. Entergy requests review and approval for a change to the River Bend Updated Safety Analysis Report (USAR), Technical Specification Bases (TS Bases), and the Technical Requirements Manual (TRM). This LAR is not a voluntary request from a licensee to change its licensing basis and it is not subject to 'forward fit' considerations as described in the letter from S. Burns (NRC) to E. Ginsberg (NEI), dated July 14, 2010 (ADAMS Accession Number ML01960180)." The LAR is in response to a Notice of Violation related to relocation of the Reactor Core Isolation Cooling (RCIC) injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' Residual Heat Removal (RHR) shutdown cooling return line.

It was determined that a spurious RCIC actuation into the feedwater line could result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the USAR.

2. DETAILED DESCRIPTION 2.1 System Design and Operation I Background The function of the Reactor Core Isolation Cooling (RCIC) system is to respond to transient events by providing makeup coolant to the reactor. The RCIC system is not an Engineered Safety Feature system. Based on its contribution to the reduction of overall plant risk, however, the system is included in the Technical Specifications as required by the NRC Policy Statement.

The RCIC system is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of Reactor Pressure Vessel (RPV) water level. Under these conditions, the High Pressure Core Spray (HPCS) and RCIC systems perform similar functions. The RCIC System consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core. Suction piping is provided from the Condensate Storage Tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from main steam line A, upstream of the inboard main steam line isolation valve. On July 3, 1999, River Bend implemented a design change to RCIC (MR 96-0069). The portion of the modification relevant to this amendment is the relocation of the RCIC injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' RHR shutdown cooling return line. The change was originally evaluated under 10 CFR 50.59 (SEN 98-0021).

RBG-47823 Page 3 of 10 The purpose of the reroute was twofold: (1) a result of a GE Potentially Reportable Condition (PRC) (GE PRC 87-04, RBS PRC 87-

11) on May 5, 1987 related to RCIC injection through the reactor head spray (Reference EEAR 87-RO 195). As reported in the GE PRC, the flow injected through the reactor head spray nozzle by the RCIC system can induce water level measurement errors. The cause of the level errors was determined to be water from the RCIC injection being drawn into the line from the vessel to the condensing chamber resulting in an indicated vessel level higher than the actual vessel level. RBS experienced this phenomenon on one of the four narrow range level instruments during the startup test program when a RCIC injection test was performed at a reactor pressure of 150 psig. In the evaluation of the GE PRC, Engineering Analysis evaluated several corrective actions for this phenomenon and recommended that the RCIC injection pOint be rerouted to the feedwater line (EEAR 87-R0195, EA-ES-0009).

(2) eliminate problems associated with check valves E51 -AOVF065 "Outboard Containment Isolation Check Valve" and E51-AOVF066 "Inboard Containment Isolation Check Valve". These check valves were containment isolation valves and were required to receive a Local Leak Rate Test (LLRT) every refueling outage. These check valves had never passed an as found LLRT and had been reworked during every refueling outage. An additional benefit of rerouting the RCIC injection point is that the Main Turbine trip that is initiated when RCIC injects to the vessel would no longer be required. The Main Turbine was tripped on a RCIC initiation to protect the turbine from damage. Damage could be caused by carryover of water that is injected by the RCIC system, becoming entrained in the steam leaving the reactor. With the RCIC system injecting into feedwater, carryover from this source was no longer a concern and the Main Turbine trip was deleted. This eliminated the possibility of a Turbine Trip and Reactor Scram, should a spurious initiation of RCIC occur. 2.2 Reason for the Proposed Change In October of 2015, during the NRC 50.59 and Permanent Plant Modification Inspection, the inspector identified a deficiency in the 10 CFR 50.59 Evaluation (SEN 98-0021) performed in 1998 for relocation of the RCIC injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' RHR shutdown cooling return line. River Bend received a non-cited violation of 10 CFR 50.59 for failure to obtain a license amendment. Specifically, River Bend failed to correctly evaluate a spurious reactor core isolation cooling actuation injecting into the feedwater line. Based on the previous, inadequate 50.59, the NRC could not conclude that a spurious RCIC actuation into the feedwater line would not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the USAR. River Bend failed to address the issue identified in October 2015 in a timely manner and in April of 2017, during the PI&R Inspection, received a Notice of Violation for failure to implement timely corrective actions to address a Severity Level IV violation issued during the 2015 50.59 and Permanent Plant Modification Inspection. In response to the Notice of Violation, River Bend committed to submit a License Amendment Request for relocation of the RCIC injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' RHR shutdown cooling return line.

RBG-47823 Page 4 of 10 The purpose of this LAR is to evaluate and document that the change does not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident. 2.3 Description of the Proposed Change 10 CFR 50.59, Changes, Tests, and Experiments, states that a licensee may make changes in the facility and procedures as described in the USAR and conduct tests or experiments not described in the USAR without obtaining a license amendment pursuant to 10 CFR 50.90 if a change to the technical specifications is not required, and the change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of 10 CFR 50.59. Otherwise, a license amendment pursuant to 10 CFR 50.90 will be obtained prior to implementing a proposed change, test, or experiment. River Bend is requesting approval of the USAR, TS Bases, and TRM changes made in 1999 (USAR Rev 12, TRM Rev 49, Bases Rev 4-9), with supplemental changes made to the USAR in 2001 (USAR Rev 14), 2002 (USAR Rev 15), and 2015 (USAR Rev 24), and any future document changes determined by the licensee to be necessary, as a result of the relocation of the RCIC injection pOint from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' RHR shutdown cooling return line. Document Description RBS Revision Implementing Document TRM Table 3.3.3.2-2 page 2 of 3. Delete the controls for valve E12-LAR 98-01 TRM Revision 49 1 Remote Shutdown Controls MOVF023 RHR/RCIC Head Spray TRM Table 3.4.6-1, Reactor Delete E51 -AOV065 RHR/RCIC Head Coolant System Pressure Isolation Spray, E51-MOVF013 RHR/RCIC Head LAR 98-01 TRM Revision 49 Spray, and E12-MOVF023 RHR/RCIC 2 Valves Head Spray TRM Table 3.4.6-1-1. Reactor Delete E51 -PTN052 RCIC Pump Coolant System Interface Valves LAR 98-01 TRM Revision 49 3 Leakino Pressure Monitor Suction Pressure High Delete E12-MOVF023 RHR/RCIC Head Spray, E51-AOVF065 RHR/RCIC Head TRM Table 3.6.1.3-1 pages 1, 3, & Spray and E51 -AOVF066 RHR/RCIC 5 of 6, Primary Containment Head Spray, rename E51 -MOVF013 LAR 98-01 TRM Revision 49 Isolation Valves from RHR/RCIC Head Spray to RHR/RCIC Injection, and add E12-4 MOVF023 RHR/RCIC Injection TS Bases B 3.5.3 page B 3.5-20 Revised to show the injection of RCIC LAR 98-01 Bases Rev 4-9 5 through Feedwater 6 USAR Table 3.5-28 Remove E51 -AOVF066 from table LBDCR 05.04-142 USAR Rev 12 7 USAR Table 3.6A-9a Revise Max Stress Range values LBDCR 05.04-142 USAR Rev 12 8 USAR Table 3.6A-10a Revise Stress values LBDCR 05.04-142 USAR Rev 12 9 USAR Table 3.6A-11a Delete Table LBDCR 05.04-142 USAR Rev 12 10 USAR Table 3.6A-11 b Delete Table LBDCR 05.04-142 USAR Rev 12 USAR Table 3.6A-21 Remove RCIC from High Energy Piping LBDCR 05.04-142 USAR Rev 12 11 in Containment list 12 USAR Table 3.6A-30b sh 1 Add Break Point 100 to table LBDCR 05.04-142 USAR Rev 12 13 USAR Table 3.6A-39b sh 1 Remove E51 -AOVF066 from table LBDCR 05.04-142 USAR Rev 12 14 USAR Table 3.6A-39b sh 3 Remove E51 -AOVF066 from table LBDCR 05.04-142 USAR Rev 12 15 USAR Table 3.6A-39b sh 4 Remove E51 -AOVF066 from table LBDCR 05.04-142 USAR Rev 12 16 USAR Table 3.6A-42 Delete Table LBDCR 05.04-142 USAR Rev 12 17 USAR Table 3.6A-43b sh 1 Add Break Point 100 to table LBDCR 05.04-142 USAR Rev 12 USAR Notes for Tables 3.6A-25 Add notes LBDCR 05.04-142 USAR Rev 12 18 through 3.6A-51, sh 7 USAR Figure 3.6A-14a Remove RCIC injection point on vessel LBDCR 05.04-142 USAR Rev 12 19 and add blind flange

RBG-47823 Page 5 of 10 20 USAR Figure 3.6A-20 21 USAR Figure 3.6A-25a 22 USAR Figure 3.6A-38 23 USAR Figure 3.6A-40 24 USAR Figure 3.6A-43 25 USAR Table 3.9A-11 sh 1 26 USAR Table 3.9A-11 sh 3 27 USAR Table 3.9A-15 28 USAR Page 3.9B-85 29 USAR Figure 3A.24-1 30 USAR Page 3C.2-14 31 USAR Page 3C.2-15 32 USAR Page 3C.2-22 33 USAR Page 3C.2-23 34 USAR Page 3D-1 35 USAR Page 5.4-18 36 USAR Page 5.4-19 37 USAR Page 5.4-20 38 USAR Page 5.4-24 39 USAR Page 5.4-28 40 USAR Page 5.4-31 41 USAR Page 5.4-32 42 USAR Figure 5.4-8 43 USAR Figure 5.4-12a 44 USAR Page 6.2-43 45 USAR Page 6.2-88 46 USAR Table 6.2-13 47 USAR Table 6.2-23 48 USAR Table 6.2-24 49 USAR Table 6.2-25 50 USAR Table 6.2-35 sh 1 51 USAR Table 6.2-35 sh 2 52 USAR Table 6.2-35 sh 3 53 USAR Table 6.2-40 sh 2 54 USAR Table 6.2-40 sh 5 55 USAR Figure 6.2-47 56 USAR Figure 6.2-48 57 USAR Figure 6.2-49 58 USAR Figure 6.2-63 59 USAR Figure 7.3-2 sh 3 60 USAR Page 7.4-3 61 USAR Page 7.4-12 62 USAR Page 7.4-26 63 USAR Figure 7.4-1 sh 4 64 USAR Figure 7.4-1 sh 5 65 USAR Figure 7.4-4 sh 1 Show HELB Point LBDCR 05.04-142 USAR Rev 12 Add line ICS-006-101 -2 to drawing LBDCR 05.04-142 USAR Rev 12 Revise RCIC pipe routing LBDCR 05.04-142 USAR Rev 12 Remove Head Spray line LBDCR 05.04-142 USAR Rev 12 Add line ICS-006-1 01 -2 to drawing LBDCR 05.04-142 USAR Rev 12 Remove E12-MOVF023 from table LBDCR 05.04-142 USAR Rev 12 Remove E51-AOVF065 and E51 - LBDCR 05.04-142 USAR Rev 12 AOVF066 from table Remove E51 -AOVF065 and E51-LBDCR 05.04-142 USAR Rev 12 MOVF013 from table Revise section to remove reference to LBDCR 03.09B-USAR Rev 24 head spray 024 Revise RCIC pipe routing LBDCR 05.04-142 USAR Rev 12 Add insert referencing injection line is LBDCR 05.04-142 USAR Rev 12 routed to the feedwater loop Add insert discussing circumferential LBDCR 05.04-142 USAR Rev 12 break and potential HELB targets Remove reference to Head Spray LBDCR 05.04-142 USAR Rev 12 Remove information about Head Spray LBDCR 05.04-142 USAR Rev 12 Remove reference to Head Spray as LBDCR 05.04-142 USAR Rev 12 having a guard pipe Remove reference to the cool RCIC LBDCR 05.04-160 USAR Rev 14 water mixing with hot steam Removing reference to LLRT Type A LBDCR 05.04-142 USAR Rev 12 testing Remove information about Head Spray LBDCR 05.04-142 USAR Rev 12 Change injection point from head spray LBDCR 05.04-142 USAR Rev 12 to feedwater system Remove E51-AOVF066 valve list LBDCR 05.04-142 USAR Rev 12 Remove reference to head spray nozzle LBDCR 05.04-160 USAR Rev 14 Remove reference to Head Spray LBDCR 05.04-142 USAR Rev 12 Revise RCIC pipe routing LBDCR 05.04-142 USAR Rev 12 Revise RCIC pipe routing LBDCR 05.04-142 USAR Rev 12 Remove reference to Head SPlay. LBDCR 05.04-142 USAR Rev 12 Remove penetration KJB-Z19 LBDCR 05.04-142 USAR Rev 12 Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Remove reference to RCIC head spray LBDCR 05.04-142 USAR Rev 12 Remove reference to RCIC head spray LBDCR 05.04-142 USAR Rev 12 Remove reference to RCIC head spray LBDCR 05.04-142 USAR Rev 12 Add E51 -MOVF013 and E12-MOVF022 LBDCR 05.04-142 USAR Rev 12 to table Remove valve associated with LBDCR 05.04-142 USAR Rev 12 Pentration KJB-Z19 from table Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Add note to table that RCIC head spray LBDCR 05.04-142 USAR Rev 12 line has been deleted Revise RCIC pipe routing LBDCR 05.04-142 USAR Rev 12 Remove reference to E12-MOVF023 LBDCR 05.04-142 USAR Rev 12 Remove reference to main turbine trip LBDCR 05.04-160 USAR Rev 14 on RCIC injection Remove E12-MOVF023 from the list LBDCR 05.04-142 USAR Rev 12 Remove reference to RCIC and safe LBDCR 05.04-160 USAR Rev 14 shutdown Remove RCIC initiation of turbine trip LBDCR 05.04-142 USAR Rev 12 Remove E51-AOVF066 LBDCR 05.04-142 USAR Rev 12 Remove E12-MOVF023 LBDCR 05.04-142 USAR Rev 12

RBG-47823 Page 6 of 10 66 USAR FiQure 9.3-1C 67 USAR Figure 9.3-7e 68 USAR Page 10.2-10 69 USAR Figure 10.3-1a 70 USAR Figure 10.4-7b

3. TECHNICAL EVALUATION Revise RCIC pipe routine LBDCR OS.04-142 USAR Rev 12 Remove piping from ES1 -AOVF066 LBDCR 09.03-263 USAR Rev 14 RHR/RCIC Head Spray Remove reference to main turbine trip LBDCR OS.04-160 USAR Rev 14 on RCIC injection Remove RC IC injection point to vessel LBDCR 10.03-113 USAR Rev 1S head spray Revise RCIC pipe routing LBDCR OS.04-142 USAR Rev 12 This amendment provides a technical evaluation for a spurious RCIC actuation injecting into the feedwater line and documents that the change does not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident.

USAR section 15.1 (Decrease in Reactor Coolant Temperature) corresponds with the "Increase in Heat Removal by the Secondary System", category of initiating events as described in Chapter 15 of Regulatory Guide 1.70 revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants - LWR Edition. The Regulatory Guide proposed the following initiating event categories:

1. Increase in heat removal by the secondary system (turbine plant)

2. Decrease in heat removal by the secondary system (turbine plant)

3. Decrease in reactor coolant system flow rate

4. Reactivity and power distribution anomalies

5. Increase in reactor coolant inventory

6. Decrease in reactor coolant inventory

7. Radioactive release form a subsystem or component

8. Anticipate transients without scram These initiating events directly correlate with section 15.1 through 15.8 of the RBS USAR, however the titles of the initiating events were changed during the analysis for clarity and consistency with other GE plants.

The basis for USAR section 15.1, Decrease in Reactor Coolant Temperature, is contained in the GE GESTAR II and GESSAR documents. GESTAR II Appendix A.15 develops a generic nuclear safety operational analysis. This document assigns events to categories in accordance with Regulatory Guide 1.70 Chapter 15 guidance. RG 1.70 Chapter 15 states that initiating events should be categorized by type and expected frequency of occurrence so that only the limiting cases in each group need to be quantitatively analyzed. GESTAR II describes the limiting Decrease in Core Coolant Temperature Event as a Loss of Feedwater Heating (manual control). As can be seen in the River Bend Transient Safety Analysis Design Report, 6224.302-000-035A, three systems were considered that could introduce a cold water perturbation: RCIC, HPCS, and the feedwater system. This report qualifies improper startup of HPCS or RCIC as events that would produce no significant power transients. This analysis demonstrates that these events were considered as USAR chapter 15.1 events but are bounded by the loss of feedwater heating event, so they are not evaluated as chapter 15.1 events due to the limited power transient. ABD 7.1

RBG-47823 Page 7 of 10 states that Inadvertent HPCS Pump Startup (Event 9 described in RBS USAR section 15.5.1 and appendix 15A) is classified under the category "Increase in Reactor Coolant Inventory" because this is the initial effect of this event. This classification is in accordance with Reg Guide 1.70 revision 3 guidance. (According to Reg Guide 1.70 revision 3, inadvertent operation of ECCS during power operation should be categorized as an Increase in Reactor Coolant Inventory Event.) RBS USAR Appendix 15A Event 9 states that Inadvertent HPCS Pump Startup is analyzed for this event rather than inadvertent operation of other NSSS pumps (RCIC, RHR, LPCS) as it delivers the greatest amount of cold water to the vessel and may operate over nearly all operating states. Inadvertent initiation of RCIC has been previously evaluated and is bounded by the Chapter 15.5.1 event, Event 9, in the RBS USAR. Changing the injection point of RCIC does not increase the probability or consequences of an inadvertent RCIC injection. All affected piping, fittings, and valve pressure boundaries are qualified to the appropriate fluid transients and operational conditions in accordance with the design and licensing basis. No instrument setpoints were changed as a result of this modification. The RCIC system's modes of operation are not changed or affected by this modification. Therefore there is no change in the probability of an inadvertent initiation of RCIC by this modification, so there is no impact to the probability or consequences of any previously evaluated accident.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the proposed change.

The original safety functions of the check valves E51-AOVF065 and 66 are as follows:

1.

Provide containment isolation in accordance with NUREG-0800, Section 6.2.4, Containment Isolation System, as discussed in the RBS Safety Evaluation Report (SER), Section 6.2.4, Containment Isolation System and 10CFR50. appendix A, GOC 56, as established in USAR Table 6.. 2-40, Containment Isolation Provisions for Fluid Lines, Page 7 of 18.

2.

Provide RCIC system piping pressure boundary requirements in accordance with ASME Code, Section III Class 2. Seismic Category I requirements per Regulatory Guide 1.29 and the requirements of 10CFR50, Appendix A, GOC 2, design basis for protection against natural phenomena, including flooding, tornadoes, hurricanes, and other natural phenomena including missiles and pipe whip as discussed in the RBS Safety Evaluate Report (SER) Section 5.4.8. Reactor Core Isolation Cooling System. The effects of the proposed change to these safety functions are as follows:

1.

Provisions for containment isolation are being provided in place of the valves by plugging and capping the RCIC piping. The closed pipe provides the pressure boundary integrity and the containment isolation boundary and exceeds the requirements of 10CFR50, Appendix A, GOC 56, because containment isolation is being provided by welded-closures which are leak tight boundaries in place of check valve seats. The welded-in fittings meet the same system and containment isolation

RBG-47823 Page 8 of 10 requirements as the valves which are ASME Code, Section III, Class I and 2, Seismic Category I. Therefore, the pipe is designed to and receives the same protection from natural phenomena and internally or externally generated missile, jet impingement, and pipe whip in accordance with 10CFR50, Appendix A, GDC 1, Quality Standards and Records; GDC 2, Design Basis for Protection Against Natural Phenomena; GDC 4, Environmental and Missile Design Basis; and GDC 16, containment Design, as discussed in NUREG 0800, Section 6.2.4, Containment Isolation System. Therefore, the RCIC abandoned piping including the closure fittings are consistent with the plant design and licensing basis for containment isolation.

2.

The RCIC system pressure boundary requirements are met because the replacement welded-in fittings, meet the same plant design and licensing criteria as established for the valves as discussed in Item 1 above. Therefore, the replacement closure fittings are consistent with RCIC system pipe pressure boundary design and licensing basis. The head spray line being removed, eliminates the function of EI2-MOVF023 as an MOV. Since this valve will no longer function as an MOV the control switches associated with this valve are being removed from the Main Control Room and the Remote Shutdown Panel. Therefore the requirement for operability of the valve controls is removed from the Remote Shutdown System Controls Table in the TRM. In conclusion, based on the considerations discussed above (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment with not be inimical to the common defense and security or 0 the health and safety of the public. 4.2 No Significant Hazards Consideration Determination Analysis In accordance with the provisions of 10 CFR 50.90, Entergy requests review and approval for a change to the River Bend Updated Safety Analysis Report (USAR), Technical Specification Bases (TS Bases) and Technical Requirements Manual (TRM). This change concerns a response to a Notice of Violation related to relocation of the Reactor Core Isolation Cooling (RCIC) injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' Residual Heat Removal (RHR) shutdown cooling return line. Based on the previous, inadequate 50.59, the NRC could not conclude that a spurious RCIC actuation into the feedwater line would not result in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the USAR. Entergy has evaluated whether or not a Significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or conseguences of an accident previously evaluated?

Response: No BASIS: The relocation of the RCIC injection point from the reactor vessel head spray nozzle to the 'A' Feedwater line via the 'A' RHR shutdown cooling return line does not adversely

RBG-47823 Page 9 of 10 affect the design function of an System, Structure, or Component (SSC) or a method of performing or controlling a design function of an SSC as described in the USAR so there is no change to the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the USAR. There is no impact to the likelihood of occurrence of a malfunction of a structure, system, or component because there are no structures systems or components changed or affected by the scope of this evaluation. Inadvertent initiation of RCIC may be categorized as either a Decrease in Reactor Coolant Temperature event or an Increase in Reactor Coolant Inventory event. River Bend Transient Safety Analysis Design Report, 6224.302-000-035A, states that three systems were considered that could introduce a cold water perturbation (Decrease in Reactor Coolant Temperature Event) at operating pressures: RCIC, High Pressure Core Spray (HPCS), and the feedwater system. This report qualifies improper startup of HPCS or RCIC as events that would produce no significant power transients. The proposed change relocated the injection point of the RCIC flow from the reactor head (RPV) to the feedwater line (FWS). This change will reduce the effects of steam quenching. However, the effect of steam quenching is not credited in any of the safety analysis. The only portion of the RCIC system operation that is credited is water injection at the required flow rate, and the design function as described in the USAR of the RCIC system is to maintain or supplement the reactor vessel water inventory. The source of water for the Inadvertent RCIC injection remains the same. The destination of the water for the Inadvertent RCIC injection is still the RPV. The ability of the rerouted equipment to satisfy the RCIC design function is not reduced from the original design requirement to inject 600 gpm into the RPV. This is maintained by the RCIC flow controller. The entry location from the RPV head spray to the feedwater line has no impact to the consequences of an inadvertent initiation of RCIC. As the consequences of an inadvertent initiation of RCIC are unchanged, the consequences of this event remain quantitatively bounded by the Loss of Feedwater Heating event described in section 15.1.1 of the USAR for the Decrease in Reactor Coolant Temperature category and bounded by the Inadvertent HPCS Startup for the Increase in Reactor Coolant Inventory category. Changing the injection point of RCIC does not increase the probability or consequences of an inadvertent RCIC injection. All affected piping, fittings, and valve pressure boundaries are qualified to the appropriate fluid transients and operational conditions in accordance with the design and licensing basis. No instrument setpoints were changed as a result of this modification. The RCIC system's modes of operation are not changed or affected by this modification. Therefore there is no change in the frequency of an inadvertent initiation of RCIC event. There is no change in the frequency of inadvertent initiation of RCIC by this modification, so there is no impact to the probability of any previously evaluated accident. Therefore, it is concluded that this change does not significantly increase the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

RBG-47823 Page 10 of 10 BASIS: The spurious start of RCIC accident is evaluated in the USAR as Event 9 "Inadvertent HPCS Pump Start (Moderator Temperature Decrease) as shown in USAR Appendix 15A. The Inadvertent HPCS Pump Start event bounds the inadvertent operation of RCIC event and is quantitatively analyzed in accordance with Reg Guide 1.70 rev. 3. This event may be classified as either a Decrease in Core Coolant Temperature event or an Increase in Reactor Coolant Inventory Event, however was categorized as an Increase in Reactor Coolant Inventory Event in the RBS USAR as this is the initial effect of this event. No new accident is created by the scope of this modification because all aspects of the existing Decrease in Core Coolant Temperature and Increase in Reactor Coolant Inventory events and their relationship to the spurious start of RCIC remain applicable Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No BASIS: The proposed change does not change any accident analyses. The proposed change does not exceed or alter a design basis or safety limit; therefore it does not significantly reduce the margin of safety. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety. Based on the preceding analysis, it is concluded that operation of the River Bend Station in accordance with the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92. 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

RBG-47823 Page 1 of 7 RBG-47823 Technical Requirements Manual (TRM) Page Markups

Remote Shutdown System TR 3.3.3.2 TABLE 3.3.3.2-2 page 2 of 3 REMOTE SHUTDOWN SYSTEM CONTROLS MINIMUM CHANNELS OPERABLE RSPl RSP2

15.

RCIC Vaccum Breaker Inboard Isolation MOV 1 NA (1ES1*MOVF078 )

16.

RCIC Turbine Flow Controller 1 NA-(lC61 *FICROO1)

17.

RCIC Turbine Trip, Throttling MOV 1 NA (lES1

• MOVC 0 02 )

18.

RCIC Turbine Local Control Select switch 1 NA (lC6lA-Sll)

19.

RHR Pump 1 2 ("1 ( lE12*PCOO2A, 2B, 2C)

20.

RHR Hx Shell Side Outlet MOV 1 1 (lE12*MOVl'003A, B) 21-RHR Pump Suction MOV 1 2' ( 1E12*MOVFOO4A, B; lE12*MOVFl05)

22.

RHR Shutdown Cooling MOV 2 'al NA ( lE12*MOVFOO6A, 6B)

23.

RHR Outboard Shutdown Isolation MOV 1 NA (lE12*MOVF008 )

24.

RHR Inboard Shutdown Isolation HOV 1 NA ( lE12*MOVFOO9)

25.

RHR Hx Flow to Suppression Pool HOV 1 1 (1E12*MOVFOllA, BI ~~

27.

RHR Test Line HOV 1 1 (1E12*HOVF02~, B)

28.

RHR Injection Shutoff MOV (lE12*MOVF027A, B) <a) One per control equipment RIVER BEND TR 3.3-14 (24111) 1 1 (continued) Revision 5

Reactor Coolant system r sure Isolation Valves TR 3.4.6 TR 3.4.6 Reactor Coolant System Pressure Isolation Valves

---

NOTE-------------------------------

The followinq surveillance Note applies to the identified SR of Technical Specification LCO 3.4.6. SURVEILLANCE REQUIREMENTS SR 3.4.6.1 SYSTEM

1. LPCS

2. HPCS SURVEILLANCE fREQUENCY

---

NOTE-----------------

18 months Technical Specification SR 3.4.6.1 is performed at an 18 month frequency. TABLE 3.4.6-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER 1E21*AOVF006 lE21*MOVFOOS lE22*AOVFOOS lE22*MOVF004 ruNCTION LPCS Injection LPCS Injection HPCS Injection HPCS Injection

4. RHR 41

• LPC n ec on RIVER BEND 1E12*MOVF042A LPCI A Injection lE12*AOVF041B LPCI B Injection 1!12*MOVF042B LPCI B Injection 1E12*AOVF041C LPCI C Inject.ion lE12*MOVF042C LPCI C Injection lE12*MOVF009 Shutdown Coolinq A, B Suction lE12*MOVF008 Shutdown Coolinq A.' B Suction TR 3.4-6 (16i)

REVISION 5

Reactor ,olant System Pre~sure Isolatiol alve Pressure Monitors TR 3.4.6.1 SURVEILLANC&~ REQUIREMENTS TSR 3.4.6.1. 1 TSR 3.4.6.1. 2 INSTRUMENT NUMBER lE21*~TN054 5 IE 2-lE12*PTNOS3a l E12*PTNOS3C 1E12*PTN057 RIVER BEND SURVEILLANCE FREQUENCY ~erform CHANNEL FUNCTIONAL TEST on the high/low 31 days pressure interface valves leakage pressure monitor alarm setpoints. Perform CHANNEL CALIBRATION on the high/low 18 months pressure interface valves leakage pressu~e monitor setpoints per Table 3.4.6.1-1. TABLE 3.4.6.1-1 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE ~RESSURE MONITORS FUNCTION Pump Discharge Pressure High ction Pressu Hi h A Pressure 19h RHR B Pressure High RHR C Pump Discharge Pressure High RHR Pump Shutdown Cooling Suction P,ressure High TR 3.4-8 (16iii) NOMINAL ALARM SETPOINT 580 psig 80 474 pS1g 474 psig 474 psig 174 psig ReVision 15

TABLE 3.6.1.3'1 (page 1 of 6) PRIMARY CONTAINMENT ISOLATION VALVES PENETltA TI 011 snI£Il VAl VF MIIMAER(a) WIIlfl

a.

aut~tic 'sol.fion Valyes MSIY lB21*Aav~022A(b)(g) lICJS*Z1A MSIV lB21*AOVF022SCb)(g) lKJS*Z1S MSIY lB21*AOVF022CCb) (g) lICJS*ZlC: MSIY lB21*AOVF022DCbHg*) 1I(JS*Z1D MSIY lB21*AOVF028ACg) lICJS*Z1A "'SIV lB21*AOVF0288Cg) lj:JS*ZlS "'SlY lB21*AOVF028C(g) lICJB*ZlC "'SlY lB21*AOYF02SO(g) 1I(JB*ZlD MAXI,.,.. lSOLATIOII VALVl! TIM! GII.a.I2 ( I ) CSecgnda) 6 5 6 5 6 5 6 5 6 5 6 5 6 5 6 5 I?CIVs TR 3.6.1.3 SECONDARY COIITAI NNI!IIT BYPASS PATM (Y~s/llo) No No No No No No No No Turbine Plane Mise. Drains 1821*MOVF067A(g) lICJS*ZlA 6 17.' No Turbine Plant Mise. Drains lB21*MOVF067B(g) lICJS*Z1B 6 16.1 No Turbine Plant Misc. Drains 1821*MOVF067C(g) lICJ8*Z1C 6 15.9 No ' Turbine Plant Mise. Drains lB21*MOVF0671)(g) lICJB*ZlD 6 19.' No Turbine Plant Misc. Oral". 1821*MOVF016(bHg) lICJB*Z2 6 16.5 No Turbine Plane Misc. Drai". lB21*MOVF019(g) lICJB*Z2 6 17.6 ' No RHR Return to FW lE12*MOVF053A~/II~ lICJB*Z3A 5 18.7 No ~~rn to FW lE12*MOVF05~ ~ ~ 18.7 No ~~~~ IIHR ShutdOWi Cool ing 5,-",ly II"R Shutdown Cooling $,-",ly LPCI A to Reactor LPCI 8 to Reactor MS'PLCS Line RWCU Disch. to Condenser RWCU Return to FW RWCU ~ Suction RWCU Pulp 0 I sch. R~ ~isch. to Condenser R~ Return to FW RWCU Ptalp Suet'_ R~ Pulp 0 I sdtl. RWCU Backwash DfRtr; RWCtJ Seckwalb Dflch. HPCS THt letum-Supp. Pool RIVER BEND lE12*MOVFOOS 1E1Z*MOVF009(b) 1E12*MOV~037A(m) lE1Z*MOYF037BC.) 1E33*MOVFOoaCd)Ck) lG33*MO'VF028 lG33*MOVF040 1G33*MOV,001 (b) lG33*MOVF053 1G33*MOVF034 1 G33*MOVF039 1G33*MOV~004 lG33*MOVF054 1WCS*MOVl7a lWCS*MOV1n lE22*MOVF023<l ) lICJB*Z20 11CJB*Z20 lICJS*Z2" 11CJ8*Z21' 1ICJB*Z1A,S,C. 0 lICJS*Z4 l1CJ8*Z6 1qa*Z7 11CJ8*Z129 l1CJ8*Z4 1ICJS*Z6 11(J8*Z1 1I(J8*Z1Z9 l1CJ8*Z5 .1ICJa*Z5 11(JS*Z11 TR 3.6-6 (20H) 5 29.7 No 5 25.3 No 14 73.1 No 14 74.S No 4 14.5 No 15 20.9 Yes 15 24.2 No 16 19.1 No 15 5.5 No 15 20.9 hs 15 24.2 No 7 6.6 No 15 5.5 No 1 12.1 Yel 1 12.6 Yes 50 No continued Revision 35

TABLE 3.6.1.3-1 Cpag.3 of 6) PRIMARY CONTAINMENT ISOLATION VALVES

b.

Maall' (SOlatjon Valves LPCI A to Reactor LPCI I to R.actor Reactor Plant Vent. DP Trana. Reactor Plant Vent. DP Trana. PVLCS Preslur. Tr~itter R.actor Plant Vent. DP Trana. Cont. Leakag. Monitor Pres ** Cont. Leakag. Monitor Pr.... Cont. L.akag *. Monitor Press. Cont. Leakag. Monitor Press. Cont. Monitor Pr.... Sensing Cont. Monitor Pres ** Sensil19 Reactor Plant Vent. DP Trans. Reactor Plant vent. OP Trans. Cont_ Monitor Press. Sensing Cont. Monitor Pres ** Sensil19 pVlCS Pressure Tr~ltt.r Reactor Plant Vent. DP* Trans. LPCI A to R.actor LPCI

• to Reactor SW Rtn VacUUl R.l.as.

SW Rtn VacUUM R.leas. SW Rtn Vac~ R.leas. SW Rtn Vacuum Releas. Feedwliter Line FHClwat.r Line KPCS P~ Suction frOll! Supp. Pool HPCS to Reactor KPeS lein. Flow 8ypau Supp. Pool PIotIIPbKIt Un. LPCS Suction fr. SUpp. Pool LPCS to R.actor LPCI A to R.actor lPCI A to Reactor LPCl a to R.actor LPCl

• to ** lCtor VAl Vf NIIHBERCa) lE1Z*F099A JE1Z'"FOm lHVR'"va(k) 1 HVR-Vl"oCk).

1 LSV*V64(k) lHVR'"V12Ck) lLMS*V14 lLMS*V1Z lLMS*V7 lLMS'"V16 lCMS*VZCk) lCMS*V3Ck) lHVR.Vl4Ck) lHVR.V16Ck) lCHS.V16Ck) lCMS"V15 Ck) 1 LSV.V65 Ck) lHVR'"V18(k) lE1Z*VF044A lE1Z*VF044B lSWP-SOVS2ZA(') lswp.SOVS2ZSCe) lSWP"SOVS22C") lSwp.SOV5220C,) lFWS.MOV7ACe) 1 FWS*MOV7I (.) lEZZ'"MOVF015(*)(j) 1 EZZ'"MOYFOO4Cb)(.) lE22*MOVF01Z(*)(j) 10FR*MOV146C.)(j) lE21*MOVFOO1,*)(j) 1E21*MOVF005(b)(.) leSl*MOVF068(*)(q) lE51'"MOVF01~j)(P) lE51*MOVF01~) PENETRATION N1J!!l!EI.. lKJS*ZZ1A lKJB*Z21S lICJS'"Z60ZA lKJS'"Z602S lICJS*Z602D lKJS*Z602F lKJS*Z60lA 1ICJS*Z603A lKJS'"Z603C lKJS*Z603C - 1ICJS'"Z605A lKJS*Z60511

ICJS*Z606A lICJB*Z606.

lICJS*Z606C lKJS'"Z6060 lKJ8'"Z606E lKJS'"Z606F lKJS'"Z21A lICJS*Z218 lICJS*Z53A lICJS*Z53B lICJS*Z53A lICJS'"ZS38 lKJS'"Z3A lKJS*Z38 lKJS'"Z8 lKJS'"Z9, 10U*Zl0 1KJS*Z11 1ICJS*Z11 11CJS-Z12 lKJ8-Z13, 10R8*Z1. lKJ8*Z17 lKJ8*Z18A 1elZ'"MOVFOZ7AC,) 1E1Z*MOVf04ZA(') lE1Z*MOVf0271(') lflZ*MOVf0421(') ~ JI Z lKJ.'"ZZ1A lKJI-ZZ18 1KJ8"Z21. MA)(II<<JM ISOLATION TIME (Secgnds) PCIVs TR 3.6.1. 3 SECONDAlty CONTAINMENT BXPASS PATH (Yes/No) NO No No No No No No No No No

• No No No No No No No No No No No No No No Yes Yes No No NO No No No No No No NO NO No No

~/~~~~:~~~ RIVER BEND TR 3.6-8 (20iv) Revi~ion 35

c. Other Isolltion Valves Feedwater LIne FeedNater Line Fee<Nater Line FHdwater Line RWCU Disch. to Condenser RWCU Sack~a.h Disch.

HPCS to Reactor TABLE 3.6.1.3-1 Cpage 5 of 6) PRIMARY CONTAINMENT ISOLATION VALVES YAIYE NIIHBER(II) lB21*AOVF032ACc) lB21*VF010A(b) lS21*AOVF032S(C) lS21*VF010B(b) l11CS*RV144 lWCS*RV1S4 lEZZ*AOVFOOS(b}(c) PENETRATION ~ lICJB*Z3A lKJB*Z3A HCJB*Z38 1J(J8*Z3S lKJS*Z4 lKJS*Z5 lKJB*Z9,1DRB* ztO VALVE G&aJ.eC l) PCIVs TR 3.6.1.3 MAXI,... ISOLATION TIME (SecondS) . SE COIIDAItY CONTAINMEII T ~ 2lllt (Yes/No) Yes Yes Yes Yes Yes Yes No Supp. Pool Pur.,-Blck Return Line lDFR*V181(j) lICJB*Z11 No Supp. Pool Pur.,-Sack Return Line lDFR*V162C j) lICJS*Z11 No HPCS Th. Relief to Supp. Pool leZ2*RVF014(h) lKJS*Z11 No HPCS Th. Relief to Supp. Pool 1E22*RVF03S(h) lICJS*Z11 No HPCS Th. Relief to Supp. Pool lE22*RVF039(h) lKJB*Z11 No LPCS to Reactor lE21*AOVF006(b)(c) 1ICJS*Z13,lDRS No ~~~~ ~~~ RHR Shutdown Cooling Sup. lRMS*V240 lKJ ZZO No LPCI C to Reactor. lE12*AOVF041CCb)(c) lKJS*Z21C,lDR No S*Z22C RCIC/RHIt Isolation RHR A Thermal Relief to Supp Pool RHR A Hx ViR to Supp. Pool RHR A Hx VIdt to Supp. Pool RHR A Hx VIR to Supp. Pool lPCS Th. Relief to Supp. Pool LPCS Th. Relief to Supp. Pool RMR St. Condenling Th. Relief to Supp. Pool RHR

• TheMall Relief to Supp Pool RHR
• Hx viR to~. Pool RHR
• Hx VU to:...... Pool RHR
• HI( V&I to "iii. Pool RHR
• Mx vU t.,- ~

Pool RHIt 8 Hx VIdt to Supp. Pool SPC Disch. to RMIt Th. Relief SPC Suction fro. RHR Th. Relief Fuel Pool C&C Disch. Fuel Pool C&C Suction Fuel Pool Purif. S~tion CRD Hyd. SV-. Sup. RIVER BEND lE12-VF102(q) lKJS*Zl8C RHS-RV67A(h)(q) lKJS*Z23A 1!lZ*RVF025A(h}(q) lICJS*Z23A lE1Z*RVF017A(h)(q} lKJS*Z23A lE1Z*RVFOO5(h)(q) lKJB*Z23A lE21*RVf01a(h)(q) lKJB*ZZ3A lE21.RVF031 (h)(q) - lKJB*Z23A lE1Z*RVF036(h)(q) lKJS*ZZ3A RHS'671(h)(q) lKJS*Z231 lE1Z*RVFOZ5C(h)(q) lICJB*Z23S lE12*RVFOZ51(h)(Q) lKJS*ZZ38 lE12*RVF030(h)(q) lKJS*Z238 lE1Z*RVF101(h)(q) lKJB*Z231 1E1Z*RVF?'rsCh )CQ) lKJS*Z231 RHS-RV66 ~ lKJS*ZZ4C RHS'RV6S(J) lKJB*Z25C 1SFC*Vl01 lKJS*Z26 1SFC*V350 l1CJS*127 1SFC*V351 lKJS*lZS lC"*VF1ZZ lKJS*Z29 TR 3.6-10 (20vi) No No No No No No No No No No No No No No No NO No No No No continued Revision 35

RBG-47823 Page 1 of 2 RBG-47823 Technical Specification Bases (TS Bases) Page Markups

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC System BASES BACKGROUND The RCIe System is not part of the EeeS; however, the RCIC System is included with the ECeS section because of their similar functions. The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of RPV.water level. Under these conditions,.the High Pressure Core Spray (HPeS) and RCIC systems perform similar functions. The RCIe System design requirement.s ensure that the criteria of Reference 1 are satisfied. The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and va1ves *to provide steam to the turbine,.as well as piping and valves ~ ~ ~ the S~l on source to the core Vla the i_ = fr~ ~~,f;E Suction piping is provided.frocnensa e. RIVER BEND s o.age tank (CST) and the suppreSS10n pool. Pump suctlon is normally aligned to 'the CST to minimize injection of - suppression pool water into the RPV. However, if the CST water supply is low, or the suppression pool level is high, an auto~atic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from main steam line A, upstream of the inboard main steam line isolation valve. The ReIe System is designed to provide core cooling for a wide range of reactor pressures, 150 pSlg to 1177 psig. Upon receipt of an initiation signal, the RCIe turbine accelerates to a specified speed. As the ReIC flow increases, the turbine control valve is automatically adjusted to maintiin design flow. Exhiust steam from the RCIC turbine is dischirged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV. (continued) B 3.5-20 Revision No. 0

RBG-47823 Page 1 of 68 RBG-47823 Updated Safety Analysis Report (USAR) Page Markups

BDS USI~ T1BLE 3.5-28 lNALYSIS OP Y1LVB BONNETS IND VALVB StERS 1S POTEHTIAL ftISSILES DiL£ 'It ttl. LCAI " *..!. tift Valve Size J.iD.l !:ll!f. ~l~ Nu.h~ PS-SC 10 BB-SC 6 1 G33*/lOVF039 lG33*ItOTPOIi0 10 PS-Gate lB22*YP036 20 lB21.nOlll 1B21."Oll065A 11I1S*"0'71 6 lG33*!OVFOO1 1G33*ftovrOOIl 8 1E51*!OVF063 BB-Gate 1ES1*ftOvr064 10 lE12*"039B tE21.VP001 18 1E12.Vr010 1E12.S0VP009 6 SB-Globe 1G33* IIOU 102 PS s ?ressure-sealed bonnet BE ? Bolted bODnet SC c Svinq cbeck 11 ~ lIot applicable Line lCu.ber 885-10 11 lIfCS-006-136-2 lI1CS-006-139-2 1CSB-010-45-1 1FIfS-02'O-66-1 1PIIS-020-62-2 lY1fS-020-62-2 llICS-006-5-1 liCS-006-4-1 lICS-008-1-1 lICS-008-3-1 lRHS-Ol0-16-1 lCSL-Ol0-1l)-1 lRKS-O 18-53-1 lRHS-018-53-1 liCS-006-5-1

---

~S!!!L.2L§~~L1l§.L

!ill~L~~~ ________ Bonnet _____ !1!.£!!§.L!i!!S _____ Critical Valve Body at Valve !!lic!.!!~§.§ ili~~ B~aring .!!~!lling Bing Inter'llsoe ~ l!~! .0 If A A 30.0 1f1 Ifl VA lU IU HA 12.. 0 15.0 16.0 10.0 32.0 11.0 10.* 0 6.0 8.0 6.0 21.0 HI. 5.0 16.0 8.0 9.0 to.O 23.0 ".0 50.0 9.0 5.0 54.0 III III HA If1 4.0 4.0 43.0 6.0 4.* 0 35.0 8.0 3.0 52.0 IIA Ifl III Kl 8.0 5.0 ~ ~~ ~~ t'l)a f~ 1 of 1 August: 1981

RBS OSAR TABLE 3. U-9a SO!81PI or STRESSES I~ HIGH-RIEPGI IS!B CLASS 1 PIPIWG Location Break -~-if---------ii---------r----- !B~DV1TEB SISTE! - IISIDE COITIIIKEIT Pipe Bl:eak stress Liait Description of !9in! _Jft=!!l_ _'degl_ jft-inl ~!Ai!~~~~-!~nqe(l) Bq. 10 Eq. 12 Eq. 13 ~ 1I!§!L jpsiL llliL CuauiatiY8 Osaqe _l~£t~!.!.!. 1a.Um-.temJ. Break Po!nts_ Br!!k type 111 60U 601111 60CII 701 100 71~ 13011 1112-3 1/2 122-0 122-0 122-0 123-9 1112-3 1/2 122-0 122-0 135 32.08 21.75 25.4q 24.59 liS 26.5 7.5 12-0 3/16 20-" 1/16 19-6 1/16 21-0 1/2 17-9 11/16 12-0 3/16 18-3 26-" 13/,6 66,979 27,002 66,979 27,002 66,979 27,002 - 68,643 40,358 28,336 28,336 28,336 18,663 20,035 0.1603 0.1603 0.1603 0.2362 "' r lt12 47 r 472 47 r472 54,336 "r 472 (I)Slresses were calculated in accordance with Equations 12 and 13 of ASK! Section III r paraqrapbs 18-3653.6(a) and IS-3653.6(b)r respectiyely. Cu.ulatiYe usage factors were calculated in accordance with lS8! Section IIIr subarticle 18-3650. !s%: IP z Inter.edlate point fP = Ter.inal pOint CII ~ Circu.ferential break LB = Longitudinal -break fOTES: See Fig. 3.61-17 for break locations. The data presentea in this tabla were use~ in conjunction with Section 3.6.21 to deter.ine TO IP IP IP IP 'I'P IP TP the break locations sbown io 1iq. 3.6A-17. The reference coordinate syste. is shown in Table 3.61-1. 1 of 1 CB CB I) -LB CB & LB CB & L8 CD & L8 CD CB & L8 CB ~ "'~ ~ !(~ ~~ ~~ ..... -...0 august 1981

BBS USAB TABLE 3.61-10. SOKB1!' OF STRESSES 1181GB-BJERG' lSB! CL1SSES 2 liD 3 PIPIIG rE!DW1TBB 51ST!! - OOTSIDE COMfiI.BEIT (FVS-020-62-2, P85-020-32-4, ---ii-- I-2!SI~~SZlL-Bxed: I ~g!g! .tlk!!U. 'U-~D) 1 121-8 1/16 101-9 1/2 3 119-0 5/8 119-5 6 108-8 119-5 7 108-3 43/64 1It9-5 Key: Ip* Inter.ediate poiDt TP * !erainal point CB = Circuaferential break Z, .il!::!!!l 3-6 3-6 35-8 38-2 -. Stres§(l) !q. 9 II). 10 ll!!l. .!l!m.L 14,137 9,761 !'ota! Additbe Stress __ jJruL. 21,157 16,573 Pipe Break Stress Li.it 0.8 (1.2 Sh. sA' fed) Description of Break eQiIl!L-TP IP IP TP (')Stresse. were calculated in accordance with Eqaations 9 and 10 of IS!! Section III, paragraph JC-365~ 10TES: See Pig. 3.61-18 for break locations. The data presented in this table were used io conjunction with Section 3.6.21 to deteraine the break locations sho.Q in rig. 3.61-15. The roferenco coordiD~te sJste. is sbowu ~ Teble 3.61-5. 1 of 1 Break ~ CB CB* ca CB . ~~ ~~ ~~ ~~ lugust 1981

Break !oint 1 2 3 12 J,g~!! Bl &2 _J!!-i~l_ ~ 146-3 3/4 146-1 1/8 145-1 1/8 122-6 3/1t 61.5 3.5 o 2Q Key: IP = Interaediate point ~~ E Tar.inal pOint CD = Circnaferentia break LB ~ Longitudina reak RBS usn 'UBL! 3.61-11a SU!!IRI OF STRBSSES 1M BIGB-EMERG! IS!! CLASS 1 PIPING aCIC S!S~E! - IRSIDB CONTIIISBNT !4xl.UI Stress Ragqe(&! Bg. 10 Eg. 12 Eg. 13 ,.m!!l_ llliL ~L 49,639 '19.974 2,814 3,893 3Q,149 35,148 Cuaulatlye Usage -I!lct0C!!. 0.0808 0.0837 TP IP IP TP (l)Stresses were c cQlated in accordance witb equations 10, 12, aDd 13 of ISS! section III, paragraphs IB 653.1(a" MB-3653.6(a" and IB-l653.6(b), respectively. cuaulatiye usage factors were calcul 84 in accordance with lSftE Section III, subarticle RB-3650. 10TI:S: l!J:eak Ty~ CB CB CB CB OiLi1t: ~O. LCN 'S.'t-l'lL -r- ~~ ~~ ~~ ~~ ....t) data presented in this table were Qsed in conjunction with section 3_6.21 to deter.ine break locations sbown in Pig. 3.61-12. ~he reference coordinate sJste. is shown in Table 3.61-1.* 1 of 1 11lgust 1987

Break foint 1 2

---

___ ~9.£!!!2!!_

JU 12 Jll::!Dl ~ 172-11 1/11 172-11 1/4 210 270 4-0 RBS DSlB TABLE 3.6&-11b SUftftl~T or STRESSES II HIGB-EIZ9G! IS!E CL1SS 1 PIPING RCtC BE1D SPB~I SISTER - INSIDi COIT1IIKBBT ~xi.ua St~ess_~~ngeCl! I Eq. 10 Eg. 12 Eg. 13 JEsil 1eill ~l Cuaulat.iTe Osage -lac12f-. (10 stress needed since there are only two terainal points) Description of Bretk Points TP TP §reak Type ell ell DiLf.Ti.. fEll.. LeN 5,'4- \\L{t Itey: IP = Interaedlate poi TP = Tarainal point CB = Clrcuaferent LB = tongit.udln (&>St.resses vere paragraphs

• calcula t."An in accordance wit.h Bq~atioDs 10, 12, and 13 of 15ft! Section III, 18-3653.6(&), and MB-3653.6(b). respectiYely.

Cuaulatiye usage factors with lSft! Section III, subarticle 88-3650. IOTES: 3.61-20 for break locations ~ The data presented i~ this t~ble were used in conjunction with Section 3_6_21 to deter.iDe the break locations shown in Fig. 3.61-20. The reference coordinate sJste. is shown in Table 3.61-1. 1 of 1 ~ ~~ I~~ ~~ ~--o luqust 1981

lt11Z9~ -t){)Ct;9 RBS USAR faJ '~ ~ TABLE 3.6A-21 HIGH-ENERGY PIPING INSIDE CONTAINMENT Piping Systems Main steam system Main stearn drains Reactor core isolation cooling (RCIC) system - steam ~~ Feedwater system ?£R UtJ!S'.~~ll{t Recirculation system High pressure core spray system (HPCS) (reactor pressure vessel (RPV) to first check valve) Low pressure core spray system (LPCS) (reactor pressure vessel to first check valve) Reactor water cleanup (RWCU) system RPV vent line Residual heat removal (RHR) system (shutdown suction) Residual heat removal system - low pressure core injection (LPCI) Control rod drive (CRD) system Standby liqUid control system NOTE: High-energy piping (postulated to break) is defined in Section 3.6.2A. Refer to Fig. 3.6A-12 through 3.6A-32 for further definition of boundaries. 1 of 1 August 1987

RBS USAP TABLE 3.6A-30b SO""ARY OF PIPING FAILOFE ANALYSIS Piping System: Reactor Core Isolation Cooling Syste on ainment) Pipinq tine Nullers: l-ICS-008-31-1 r l-ICS-QOS, 1-1.($- 00"'-101 Consequence of Piping Pailure: Pipe wbip t'\\tk) f?~ L.c.tJ ~.a{-\\-tz.. Break LQCa!ion(6' Break Break -El-I -i-- Types i2!U H1:!!!l jH.:isl j{t.-in) (1) 1 121-8 1/16 107-9 1/2 10-6 C 17 117-0 5/8 110-3 1/2 10-6 C too t'l"-'\\ ut-l.. '\\-s

c.

Blowdown Source (2) R D R _______ -1~£g~~ _____ _ 1)PenetratioD lKJB*Z15

2) lICS-ftOlP064 lione l)Penetration lKJB*Z15

2) lICS-ltOYF064

3) '"S5-024-8-2 4)1-iCS-004-1,6-Q Protection fteasures

( ~) PBR-801 PRR-814 -807 o Mone R tl) 1.0" fll4' L.,~ MI1JH'- TE( t1.) "" '"'~ vUt ~"'Iol~~ l3) 1,- Wc.~ ~llk $)t~~ I..l~(. lil \\ ~)~ tt\\o" 3~ lS) 1," M~5 LIN£. Evaluation (.) PRR PRR !~!arks ADO P{ft LeN 6.'1-\\41. ~ ~ , of 4 v~ <46 ~~ -t:::-.'-.() v August 1987

fiBS USAf' TABLE 3.6A-39 scr~nAFY OF PIPING FAILUBE ANALYSIS Piping System: "ain 5teaa Vent Line (Inside Contain.ent) Piping Line Numbers: 1"55-004-3-1 Conseq~ence of Piping Pailure: Jet Impingement Break Location Break Break - -E"l---------Az----r--- Types fg!n! jt~~!Bl jg~l j!!=!nl (1) 1 173-0 5/16 90 1-] 15/16 C 31 112-9 90 6-10 c Protection Essentia1 "easures -IAfg~t§ ~ Hone ~ _LetHA-lolL 1 of 5 E'Ia1uation (.) 1, ACI lQgust 1981 Rellarks ~ ~ Ll~ ~§ --.J-..C)

IIBS USAF TABLE 3.6A-39 (Cont) pipinq System: "ain Steam Vent Line (Inside Contain.ent) pipinq Line Kuabers: 1nSS-002-2-1, 1~SS-002-1-1 Consequence of Piping Failure: Jet Impinge.ent Break ~g!!!! 8 9 9A 101 19 -l!.I~altJ&£!!!iQn __ _ EI Az r. j~!:!Bt j~~l jf!~!~ 156-8 90 15-5 1/2 156-8 90 15-10 1/2 156-5 90 15-1 1/2 156-2 90 15-3 7/8 150-8 7/8 106 19-3 Break Types ( I) c c C C C Essential !~fg~l§ ____ _

1) 1CX507RC1 conduit for C~S9TD43A,43C,43F

2) lCX507RC2 condcit for CnSRTD43A

3) lCX507RC3 conduit for C~ SRTlJ43C

4) 1CI507RC4 conduit for CftSPT043E S)lXJBSQ40 junction for_cIlSPTD411.43

,,'CX507FC1 conduit for CnSRTD43A,43C.43E 2,'CI507RC2 conduit for Cl!SRTDQ31 3)lC1507BC3 conduit for CftSPTD43C Q)lCXS07BC4 condcit for CftSPTD43E 5,'IJ9S"QO junction box PTD4lA. 43C r It3E 1)lCCS02Q01 conduit for ADS yalve 1B21P,,0411 2,'CC502BD3 conduit for ADS valve 1B21RVP051G 1)1CI501RC2 conduit for C~SlI'l'D43A 2)1CXS01PC3 conduit for CftSRTD43C

3) 1CX501RC4 conduit for SRTD43 "one 3 of 5 Protection "easures (3)

Eval.uation (*u

1. Mote A

2) Kote A

3) Mote A

4) Hote A

5) !fote A Mote

2) Note

3) Mote Il) Hote

5) !lote

1) Kote S

2) DS!

1) Kote A

2) Kote A

3) Note Po August 1987 m~rks L.tllPU.

L.eH ".&l-I-{Z, :It - ~ Q~ ~ I ~e. ~~ ~~ ~

RBS OSAR TABLE 3.6A-39 (Cont) Piping System: ~aln Steaa Vent line (Inside Containment) Piping Line Nuabers: 1~SS-002-2-1, 1!SS-002-1-1 Conseguence of Piping Pailure: Jet I~pingeBent . __________ ~!;~!\\L1.Q£~!!Q!l _____.!_- Break Break El Az r Typps £.Qin! jn=!nl H~9l J~!=.!!lt ( 1) 20 150-6 1/2 105 19-6 C 33 146-6 1/2 76 16-6 3/4 C 34 156-8 90 14-8 C Essential ____ -1~gets ____ _ HODe None Protection fteasures en Evaluation ( 4)

1) lCX507RC2 conduit for

1) Ko.te A CI!SIlTOQ3A

2) 1CX507J1C3 conduit for

2) Kote A CI'ISRTOQ3C 3)lCXS01RC4 conduit for 3)Kote A G~

OiJ...£.Tf Pin Lc::t.l5"A -,~t 4 of 5 August 1987 ~arks ~ . 2~ ~~. ~~ ~'-O

RBS USAF. TABtE 3.6A-42 SU""AP! OF PI~ING FAILURE ANALYSIS Piping Systep: Ileac Piping Line Nuabers: Core Isolation Cooling Read Spray (Inside Containment) CS-006-6-1 CODsegueDce of Piping Jet Impinge.ent _________ ~I~A!_1~~!i Break Break El Az Types g9iB! .c!!::!~t jg~gt (I) _ Ess!mlliU~~ 1 172-11 1/" 270 C 1)1-ICS-1151*AOYf066 2 112-11 1/1l 270 4-0 C Table 3.61-51. 1 of 1 Evaluation (.)

1) ACI, ilL

" ACI, RL tie. arlee l)£("(fE() ?fIZ. LeN 5.1/ -I ~z. a~ ~r\\) ~~ ~I ~~ .81 -.n

P.BS USAP TABLE 3.61-43b SUftK1BY Of PIPING FAIIORE ANALYSIS Pi~inq Slste.: Reactor Core Isola~~~~ ~u~~e Pipinq line Nu.ber~: 1-ICS-008-if- ~j§ ~~ Consequence of PiP1D9 Failure: Je 1 ae AnD PiA. LeN SA -l'U.. Br~ak J&!:~tion ______ Break --ii--- I - z Break Types R9!!lli j~.!::l!!l lt1:!!!l l.(!:!Bl. OJ I§§!milll Tar~ts 1 121-8 1/16 107-9 1/2 10-6 C Mone 17 111-0 5/8 110-3 1/2 10-6 C Ifone 'DO 11."-'\\ H&-1. ~~~ C. <! ) to" Fw ~t\\UJ( - fWS-18 u,).. ".~ v~-iU~,$,,"Of.~ l)) Pf.tt.1f9""-/6<<Jt1r"'& it.. \\1.,,'.. 4' ~\\ ~ ~~ - Vf 308 ~~ ",1.£1" t.u.&£. ( sJ (4)$ ~ ~1'~ 1", ~o 'LeJ lc.~ CJtI~) e."'l~T FPt. fW'l& Protection lie as ares (~J fri>D pu. L'N ~A-L~t , ot If Evalua.tion (.) (~\\.s-n. (1) ~o;( 1.B (.,) -"Qi( X£. (,,) NDi{ l-D Re*ru~ ~~ '--~ ~ ~~ t\\:), ~ B ~ -..D August 1987

Note V: Note W: Note X: ~ Note XA Note XB Note XC Note XD RBS USAR NOTES (Cont) This radiation monitor is not required for this break since it monitors radiation in the drywell during post-LOCA and this break is not a LOCA. The effects of pipe whip or pipe rupture on the demineralizer and its supports is acceptable for this break. Since the ruptured line enters the demineralizer at the

bottom, the base support bolts would be loaded in shear and, even if the bolts fail locally, the tank would still be held in place by the pl'atform halfway up the tank, and by the piping connected to the top of the tank.

A failure of the hydrogen of this break is acceptable failure of one division from any given point in unaffected hydrogen ignitor ignitor(s) as a result since, even with the of power, the distance the drywell to an does not exceed 30 ft. These are non essential items whose failure will automatically trip the reactor protection system (fail-safe). They are not required to isolate the subject break (reference LOOP ILDS*2). A failure of these valves, pipe supports, and/or associated pIping is acceptable because the affected system is non essential and not required for safe shutdown, the combined HELB flow area of this piping, plus the subject RCIC HELB is bounded by the postulated 20 inch feedwater pipe double ended rupture in this volume. These piping failures are automatically isolated by the G33 motor operated isolation valves which are unaffected by the subject failures. A failure of this line is acceptable because it is non-essential and no~ required for safe shutdown. It is moderate energy and the flooding/spray are bounded by the postulated 20 inch feedwater pipe double ended rupture. This portion of the essential system is not required for safe shutdown or to isolate the break. Failure consequence is acceptable since target is non-essential and its failure will not adversely affect any other essential systems. 7 OF 7 August 1987

<:: 'tIJJ i-mm 2~Q ~JJ Ie-% ~m c ZIII mcnz !!1 m n oo4111 0 111 :11 <Z z~~ ~c..... <." ~cn ~m-zz'U <~ 1:..,'" ~:> mc: 1II rn~ zz~

1_ ':m:.

~O gz ~ Y (VERn SD 5D _____ ~

i. a. 150'... 112-

~( "".,... "~ " ~ ! w ~ ORYWEL1. HEAD ,~ (,~ -.. k __ / TOP OF REFUELING SUI. SUPPORT .~ ~ a.1I'*... J/.* @ ~PEH. 10RII*Zl.1 ~'II*PRRI4 EL 1ST*,," SD NOTES:

1. 811~ LOCATIONI AR£ DESIGNATED

2. ::'~REAI( RESTRAINTS ARE DESIGNATED IYIUX*"RfUtXX J. InlESI RESULTS AR! alV£H IN T AIIU!

UA... CORRESPONDING TO '04£ HUMUUc.t.t. 8IIUI( "OINTS SHOWN HERE

• . SYMIIOLS FOR '04£ TYPES OF,RUK':

I( ClRCUM"UI£HTlAL IIRUI( ONLY

• ClRCUMnR£HTlAL AND LONGITUDINAL

"'UI(

L N H.E.L.B. POINT HIGH <J MODERA IT ENERGY ct. EL. 124' g" / d_rCS-006-101-2 ~ICS-V3004 NOTES:

1.

BREAK LOCA nONS ARE DESIGNATED 8Y Q

2.

PIPE BREAK RESTRAINTS ARE DESIGNATED BY l-XXX-PRR-XXX .3. INTERMEDIATE BREAK LOCATION ARE DETERMINED BY BY THE CRITERIA PRESENTED IN SECTION 3.6-2A

4.

NO STRESS NEEDEDSINCE ONLY ONE TERMINAL POINT EXISTS.

5.

SYMBOLS FOR THE TYPES OF BREAKS

• CIRCUMFERENTIAL BREAK ONLY CIRCUMFERENTIAL AND LOGITUDINAL BREAK FIGURE 3.6A-20 HIGH ENERGY PIPE BREAK RCIC/RHR INTERCONNECTION Entergy Operations, Inc.

RIVER BEND STATION lJIOATEO SAFETY ANALYSIS fEPOAT ~~----------------~---------------~

IE12"VF050a ~ ...,1) ____ 1*RHS*Ol0-14-a I -WCI*Oo.*172-2 i'- --~ 1*FWS*020*13-2 ~VO~L(N a)~-141-IEI2"VFOIOA I-WCS.QIM-22-2 f~( ,~~. 1-4 1*RHI*010*16-2 UP C'LLED~*' DOWN NOTES: 1. BREAK LOCATIONS ARE DESIGNATED BY e

2. PIPE BREAK RESTRAINTS ARE DESIGNATED BY IIOCX " PRRXXX

3. STRESS RESULTS ARE GIVEN IN TABLE 3.61.-16.

CORRESPONDING TO THE HUMEfUCAL BREAK POINTS SHOWN HERE

4. SYMBOLS FOR THE TYPES OF BREAKS

)(. CIRCUMFERENTIAL BREAK ONLY

• • CIRCUMFERENTIAL & LOHGITUDINAL BREAK

~ fIGURE 3.6A-25a c: r::s '-~ ~8 ~~ ~-..D ..r:::. HIGH ENERGY PIPE BREAK RESIDUAL HEAT REMOVAL PIPING OuTSIDE CONTAINMENT RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

,W" -10."4'- 1 IL " *.* * ",- 'WCI-teo* ** *, Ie. I'.**. " *. R~'~ L L N ~.q -l't'Z. II"ANs.. **'L.. '1ICI-OGI - tg4 * *

• t. *** *-U*

.<<I-eIU*.I*. II..,(-.- ~, "of AU. "I)C". J Z I"'..u: )5& 401 ,.\\ AN I. II." ~. ~ OG.t - 4l-. '.CI

• Q.4 -U -1
  • CI - OCH *"

-) Ii. I......

• Iwel-OOI-U -'

I*'I -004 1 114*-.*

• • c. -oo........
• lll.*.. r "I*

.* e. *oo.. *.. -. IL IIQ**I"'"' ~I

• u... ~

'1lnll ~

§!

§ -..!) 1:, rl f~tG':"'UR~E-3-.6-A-.3-8-------,-II,a HIGH ENERGY PIPING COMPOSITE REACTOR BLDG. El 162'

• 3"

~ r---R-'V-E-R-S-E-N-O-S-T-A-T-IO-N--;ILf Uf'OA TED SAfET,( ANAL '(SIS REPORT

A, _ _ f_. II... *l .,.a_ fl_ ~~-::':' " 1.. *,;1 ! .L_.,'. j l III - " -~ I ~ ~_. ~I(.~ I -~. ..!.:::J ~ .t"-"'-r-**:x:* :.- t i I ; '. ~ I .1 j:... __ ~. .~"" "1'. U""'; ~ ~. . a /-- .f j_ -. 'I l,,~.. R(,v15C PUl- .Lst:Zd til C. N t5, i-/~ 1-n!lLf_ ....... u... * =.:::.... ~ :, 1/'.--. .~ II \\ _ ~ ' (( "":1.flfFPl', _ .. ~ r.u_o"'**r'*. 1

..... J Jf ft ""d I "r FIGURE 3.6A-CO I -I HIGH ENERGY PIPHG COMPOSITE REACTOR BULOHG SECTION 2-2

~--------------~l~ RIVER BEf'D STA TION~ UPDATED SAFETY ANALYSIS REPORT IS' 'v~ ~"-{) ~ ~

oC1\\ \\;J UI(hlClo.o. I""~.TC'" .1Ill.1l9'.~lI'* (11011' '.i:...I:.:l ~ 4 1 I I --1 _______ ::,~!'"-II.., 1IJt)I~* K/II-ZI'" a Oo.... *n. L~r ( (F,e...o.u-4.11 L ~----....., f~ ~ ~ &,~ CI 'co __... . /..A>'-S ,L~/"'(. ~.~;t1'J,zI.l'" L C/'- _ *. J,.. ~ tII.SI-GI..... )..4.. J ty- ~~ ~~*k ___ ~~~~ @1 1... ~C~_~6-/O/_2 ,£. II."'~" ,a~. _.,,,, (F..... E'.. *.5) 2 L e n ___ .fIlCI..... tJ.IIl... .,,(,.,(-01 P1.AIt EL ".4' tL&l.'4* ~1l~.W 1DI.*~

• a.M 0..12,....
• f~a:t.~p*uc~

.;-I-III-f--l---l--::, ~i ~~~;'I~; 11 ~\\:s (IIGU~O.6A-'" ~ II ~ fiGURE 3.6A-43

0 J;;'

I HIGH ENERGY PIPING I... ~ I ~'~f~' COMPOSITE AUXILIARY BUILDING ~ ~ J ELEVATION 114'-0" ~..c RIVER BEND STATION I~ UPDATED SAFETY ANALYSIS REPORT

S,.t... am.

• -+1

.uoleu bon. ayatem tadlla. d.rabw (&:l1) lIalA.taam*poelU.. ...... oollbol (831' BucleuboU. qat_.... _t_ (1121' RnlclaalH.. t 7... --+4

• -+7 7.........

Rnlaloll I lIull.umb_ IB:ll*110VJ'016 IB:lI*llOVJ'Ol' lIQI*MOVJ'067A,B,C,D IB:lI"OVJ'O:l7A.,11,C,D

• IB:lI"OVJ'OlS lUl"OVJ'09IA.,11,C.D 182I"OVJ'06SA,8 1&12*110Vl'OOI 1&12*1I0\\T037A,8 1& 12*110VJ"03TA,a 1&12*110Vl'04O IB12*110VJ'042A,a,c 1&12*110VJ'049 181a"OYrONA,a 1812*110VJ'064A,J1,C 1812*MOVJ'06lA,8 1812*110VIOS 1812*A.0VJ098 1812"'VJ'075A,B 1812-.oVJ'0911 1&12*8OVJ'CN5OA,8 1812*A.0VJ'04IA,J1 1812*A0VJ'04IC 1812*aVJ'OOS 1812*aVJ'OI7A.

1812"VJ'Ol78 1&12*.VJ"0211A. la12*.VJ'02eB v ** IIbe 3 3 11/2 3/4 3 24 20 II 10 10 6 10 6 10 4 II 20 10 3/4 3/4 3/4 10 10 3/4.1 3/4.1 3/4al 11/2d II/3d raAR TABLa 3.9A*11 ACTIVS VA.l.VD (BOP) Presnre TYDe Gate-IIO Gat.. IIO GIobe-IlO OIaM-IIO Gat.. IIO Gat... o Gate-.O Gate*.O Gat.IIO 0.... 0 0.... 0 Oat.IIO Gat ** o GIobe*IIO o.ta-IIO B1ltt~ Oat. T-w.laCbaok 1'patt... 1'patt... 1'patt.. Cbeck Check aav aav sav IRV SRV 10r9 ASII& RatIN(60) 600 600 1Il00 1Il00 600 900 900 600 300 300 300 600 3DO 900 300 ISO lao 300 300 3DO 300 600 600 3OO/ISO 3DO/lao 300/100 300/100 3DO/I50 CI *** 1 1 1 2 2 2 2 1 2 2 2 1 2

I

I 3

2 2 2 2 2 1 1

1

1 2

1

I Mfr.

1 1 1 1 1 1 I 1 I I 1 1 I 1 I 3 1 6 6 6 4 8 I I

• v.....

OiMnltor 8B.oo-~5(1) 8a.OO-~IS(I) 88-00-15(1. 8118-000-2(1) 811J1-OO..2CS( I, 8l1li-4*2110(1, 8118-4-2110(1' 8B*2-6O(1, 8J11B.O.4O(1, 1.... 1*25(1' ......005*10(1' 88-2-60(1' IB-DOS-lS(I) 1118-3-10(1, 18-005*10(1, 11IJI-OO..7.5(l, IlIa.o..2S(I, (3t .. O<<el 77KK-OOlI4, .00el 77KK*OOI14, 1I00el 77KK-001(4) (3' (3' 1I0De 1I01l. Bo ** Ro ** 1I0D. ActlYe hacUOD 27 27 27 1 1,2 1,2 27 27 27 6 7 26 7 I 27 21 27 29 7 30 '7 31 27 89 89 59 59 59 ()~ ~~ ~~ ~'-C AUlUat 1996

RBS USAR TABLE 3.9A-11 (Cant) system Mark Valve Pres:!ure ASHE Valve Active Name Number Size ~ _Ratinqy (60) Class Mfr. Operator Function 2 "'6'-- Q(£'"". ow .. gee~P~-==r:t==='13) ow o{-_. ~bP-".. - ~ .. --- = + = -t

• 1& -

DiU: Tt p~ LC."'.,.,\\- \\" L ll!;:> 1

• AOV'I'trol -

o el 667NS~ES (6) 35 lESl *AOVr005 1 Globe 150 2 7 Model 667NS-~5(6) 35 IE51*AOVF02S 1 Globe 1500 2 7 Model 667NS-DBQNS(6) 36 lE51*AOVF026 1 Globe 1500 2 7 Model 667NS-DBQNS(6) 36 1E51*AOVFOS4 1 Globe 1500 2 7 Model 667NS-DBQNS(6) 37 l!:Sl*PCVF015 1 Globe 900 2 6 Model 71010-5(4) 60 lE51*RVF017 3/4xl SRV 150/150 2 8 None 59 l!:Sl*RVr018 1 1/2x 5RV 150/150 2 8 None S9 2 1/2 lE51*RVF090 3/4xl SRV 1500/150 2 8 None 59 lICS*RV130 3/4xl SRV 1500/150 2 8 None 59 Fire protection IFPW*MOV121 6 Gate-MO 150 2 1 5MB-00-10(1) 27 water (FPW) 1FPW*MOV122 6 Gate-MO 150 2 1 5MB-OO-IO(l) 14 Feedwater (FWS) IFWS*MOV7A,B 20 Gate-MO 900 2 1 5MB-4-200(1) 27 Reactor water 1G33*MOVFOOI 6 Gate-MO 600 1 1 58-1-25 (1) 27 cleanup IG33*MOVFOO4 6 Gate-MO 600 1 1 58-1-40 (1) 27 . (G33) (WCS) IG33*MOVF028 4 Gate-MO 600 2 1 SB- 005-l0(1) 27 1G33*MOVF034 4 Gate-MO 600 2 1 58-005-10(1) 27 lG33*MOVF039 6 Gate-MO 900 2 1 58 25(1) 27 1G33*MOVF040 6 Gate-MO 900 2 1 58-0-25 (1) 27 .~7 1G33*MOVFOSJ 4 Gate-MO 900 2 1 5B-0-25 (1) 27 lG33*MOVfOS4 4 Gate-MO 900 2 1 5B-0-25 (1) 27 7.- IG3J*SOVF041 3/4 Y Pattern 900 3 6 Model 77KK-OOl(4) 30 1WCS*RV144 J/4x1 5RV 1500/150 2 8 None 59 lWCS*RVI54 J/4x1 SRV ISO/ISO 2 8 None 59 1WCS*MOV111 4 Gate-MO 600 2 1 5MB-0-7 1/2 (1) 14 lWCS*MOVl72 2 1/2 Gate-MO 150 2 1 5MB-OOO-5(1) 27 I1fCS*MOV173 2 1/2 Gate-MO 150 2 1 5MB-000-5(1) 14 lWCS *MOV17 8 2 1/2 Gate-MO 150 2 1 5MB-OOO-5(1) 27 Diesel Generator lEGF*PCV25A,B 3/4 Y Pattern 900 3 6 Model 77KK-007(4) 33

-0-..

Fuel Oil Storage 'b~ and Transfer Sy~tem ~tJ (EGF) ~, Reactor plant lCCP*MOV138 10 Gate-MO 150 2 1 5MB 25 (1) 27 ~8 component cooling 1CCP*MOV158 10 Gate-MO 150 2 1 5MB-0-25 (1) 27 -...t~ 150 2 1 5MB-O-25 (1) 27 ~~ (CCP) lCCP*MOV159 10 Gate-MO lCCP*MOV163 2 Globe-MO 600 3 1 5MB-OOO-2(1) 2 lCCP*MOV169 2 Globe-MO 600 3 1 SHB-OOO-2(1) 2, 3 lCCP*MOVI29 12 Butterfly 150 3 3 SMa-ODO-5Il) 2, 3 Revision 7 3 of 9 January 1995

f RBS USAR TABLE 3.9A-1S PRESSURE ISOLATION VALVES UNDER ASME XI INSERVICE TESTING PROGRAM High Pressure Core Spray lE22*AOVFOOS lE22*MOVF004 Low Pre sure Core Spray lE21*AOVF006 lE21*MOVFOOS Reactor Core Isolation Cooling )~ DtL~1[P£L/..CN!~'I~'L ~~ Residual Heat Removal lE12*AOVF041A lE12*AOVF041B lE12*AOVF041C lE12*MOVF042A lE12*MOVF042B lE12*MOVF042C lE12*MOVF023 lE12*MOVF008 lE12*MOVF009 lRHS*V240 1 of 1 August 1987

3.9.5.1.1.128 Vent and Head Spray Nozzle This component is not a core support structure. It is included here to describe a safety class feature in the reactor pressure vessel. The vent and head spray assembly performs ~the function of providing a vent for the noncondensible gases in the vessel head~, and providing a spray nozzle for the injestion of cooling water into the upper areas of the vessel. When reastOF coolant is returned to the roastor vessel, part of the flow san be diverted to a spray nozzle in the roastor head. This spray maintains saturated conditions in the reastor vessel head volume by sondensing steam being genorated by the hot roastor vessel walls and internals. The spray also desreases thermal stratifisation in the reastor vessel soolant. This ensures that the water level in the reactor vessel can rise. The higher water level provides conduction cooling to more of the mass of metal of the reastor vessel and, therefore, helps to maintain the Gooldown fate... The vent and head spray assembly is bolted to a mating flange on the reactor vessel head nozzle. External piping is connected to the assembly by means of standard flanges. (See Section 5.4.7, Residual Heat Removal System.)

3, qB-85

r I, r' - ~ '..... _. ~. FIGURE 3A.24-1 Il-IR. 9r, -(jO& 9 j-}4Je. 2</1 R#15 ~tLt LeN t{.~. (~~ RCIC PIPING SYSTEM USED IN THE ANALYSIS OF WATER HAMMER LOAD~ RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

RBS TJSAR t 'o the RHR heat exchangers, they are classified ' as moderate energy. A 4-in line supplying the RCIC turbine branches off the 8-in line running west, drops down through the tunnel floor (el 114 ft-O in) "and again through floor el 9S ft-9 in to an eYevation of 73 ft-9 1/4 in and then runs horizontally to the normally closed valve F04S. Up to this valve the line is classified as high energy while from the valve to the RCIC turbine, it is classified as moderate energy. From the moment-limiting device adjacent to the inner isolation valve F063, through the guard pipe and outer isolation valve

F064, up to the moment-limiting device adjacent to the jet impingement wall, the piping meets the stress criteria for no postulated breaks, as discussed in Section 3.6A.

PJSEIlT A ~ L,c.AJ ~'f-II{Z. Inside the Drywell For stress analyses, the RCIC steam line within the drywell was modeled as a ~ranch of the main steam piping. Due to the postulated breaks, the line could potentially whip into several targets, including the unit cooler,

MSS, FWS, and lCS restraints, FWS and rcs lines and supports, and the shield wall.

To preclude the damage that could be caused by the whipping RCIC line, a total of six restraints have been installed along the RCIe lines inside the drywell. All of these restraints are omnidirectional except for restraints PRR-80S and PRR-806 which are moment-limiting (zero gap) restraints to keep the stress within acceptable limits in the isolation valve and the break exclusion zone. Essential jet impingement targets for the leS piping system include conduits for the RCIC isolation valve and main stearn safety relief valves. Since the conduits are capable of withstanding jet impingement loads, the safety function of these targets would not be affected. A conduit for a hydrogen ignitor is also impinged by a jet, but the failure of one igni,tor is acceptable since this failure and a worst single active or passive failure will not affect the safety function of this system. For the RPV level and pressure instrument line impinged by the jet, jet impingement restraints have been incorporated to ensure that the line could withstand the load and perform its safety function. 3C.2-14 August 1987

Also the 6-inch Rele injection line is routed to the 20-inch Feedwater A loop by tapping into the 10-inch RHS shutdown cooling mode return line to Feedwater inside the Main Steam Tunnel just south of the Jet Impingement Wall. Insert B For the 6-inch RCIC injection line, a circumferential break is postulated at the high/moderate energy line interface at the RCIC check valve interface with the high energy RHS line. The following are the potential HELB targets from this pipe break and their evaluations: (1) 24" MSS piping, 20" FWS piping and valve-- These targets are acceptable because the targeted piping sizes are larger than the ruptured pipe, (2) 4" WCS piping, valve, and supports, 2" WCS tank drain line-- These targets are acceptable since they are non-essential for plant shutdown and the combined HELB flow area of the failed piping and the subject RCIC HELB is bounded by the 20" Feedwater double ended rupture in this volume, (3) steel platform/grating at el 124'- 9"-- Failure of this potential target is acceptable because no essential systems can be adversely. affected by the failure of this platform, (4) Main Steam Isolation Valve Seal. System valve E33-VF303B and inlet line, and cables and conduits for FWS valve-- These targets are not required for safe shutdown or to isolate the break, (5) Leak Detection system thermocouples 1 LDS*RTD2A & 2B-- These are non-essential items whose failure will automatically trip the reactor protection system. ~~~~~,-..-

RBS USAR Inside the Steam Tunnel All RCIC piping, from inboard of the first moment-limiting (zero gap) restraint in the drywell to outboard of the seond moment-limiting, (zero gap) restraint in the auxiliary

building, meets the stress criteria for no postulated breaks, ' as discussed in Section 3.6A.

Inside the Auxiliary Building For stress analyses, the RHS line between the two normally closed valves F52A and F52B was modeled together with the ReIe piping. The only auxiliary whipping each leg. postulated breaks for the 4-in line entering the building are at each of the terminal ends, and any of the pipe is precluded by the two restraints on Due to the postulated breaks, the 8-in line could potentially whip into the res and WCS lines and the jet impingement wa*ll, and this is precluded by Restraint PRR-914. For the 4-in line, which branches vertically downwards from the a-in line, apart from the terminal end breaks, three other breaks are postulated at the elbows. In this instance, however, since the potential structural and piping targets have been designed to withstand the pipe whip loading, restraints are n?t required. The essential conduit targets included conduits associated with an Ies containment isolation valve which was impinged by a jet discharging through a penetration hole in the jet impingement wall. The conduit was protected by providing a shield at the jet impingement wall. LPCS/HPCS System "' T (" , I'lt <;" t," ~.:: f~ '-; J _ ' J

• ',,)

~, These systems do and hence only a normally exposed energy~ not operate during normal plant operation, small portion of the piping which is to reactor pressure is classified as high The postulated pipe break locations and restraints for the LPCS and HPCS systems are shown on Fig. 3.6A-22 and 3.6A-21, respectively. The results of the associated stress analyses are summarized in Tables 3.. 6A-14 and 3. 6A-13, respecti vely. 3C.2-1S August 1987 (

Also the 6~inch RCIC injection line is routed to the 20~inch Feedwater A loop by tapping into the 1 O~inch RHS shutdown cooling mode return line to Feedwater inside the Main Steam Tunnel just south of the Jet Impingement Wall. Insert B For the 6-inch RCIC injection line, a circumferential break is postulated at the high/moderate energy line interface at the RCIC check valve interface with the high energy RHS line. The following are the potential HELB targets from this pipe break and their evaluations: (1) 24" MSS piping, 20" FWS piping and valve-- These targets are acceptable because the targeted piping sizes are larger than the ruptured pipe, (2) 4" WCS piping, valve, and supports, 2" WCS tank drain line-- These targets are acceptable since they are non-essential for plant shutdown and the combined HELB flow area of the failed piping and the subject RCIC HELB is bounded by the 20" Feedwater double ended rupture in this volume, (3) steel platform/grating at el 124'- 9"-- Failure of this potential target is acceptable because no essential systems can be adversely. affected by the failure of this platform, (4) Main Steam Isolation Valve Seal. System valve E33-VF303B and inlet line, and cables and conduits for FWS valve-- These targets are not required for safe shutdown or to isolate the break, (5) Leak Detection system thermocouples 1 LDS*RTD2A & 2B-~ These are non-essential items whose failure will automatically trip the reactor protection system. A...._~ ___ ~--'

RBS USAR of the isolation valve targets is acceptable because the break isolation can be provided by other redundant check valves.

However, if break isol~tion is provided by valve lG33*MOVE004, drywell flooding may occur; this is acceptable since it will not degrade' safe shutdown.

Other j et impingement targets required for containment isolation include WCS lines and supports, SVV lines, and RES valves; they are acceptable since an analysis of these systems indicates that the system requirements could still be met after the rupture event. Outside the Containment Three zero-gap restraints have been installed outside the containment, adjacent to the penetrations, to protect the break exclusion area from the consequences of a ruptured pipe. Targets, that could be impacted by a whipping line due to an RWCU piping break include RHS, lCS, and FWS lines and ICS restraints. However, in all the above instances, since the whipping line is smaller than the target line, the target cannot be damaged. Other pipe whip targets include various walls and

floors, all of which have been structurally designed to withstand the pipe whip

loading, and a

ventilation duct that is not required for safe shutdown. Even though targets impinged by a jet from a ruptured RWCU line include essential conduits leading to an ReIe fill pump motor and various valves of essential

systems, a more detailed review revealed that these particular portions of the essential systems were not required for safe shutdown.

Other targets affected by the jet include conduits for flow transmitters used to detect leakage. However, once a break occurs in a particular volume, the flow transmitters will not be required since area temperature monitors will detect \\ and isolate the break. Bence, the failure of these targets is acceptable. The jet impingement targets also include conduits from area ambient temperature elements required for breakpoint isolation, but this is acceptable since, in this instance, the ~lements are' not in the postulated breakpoint volume. 3C.2.9 ReI flle postulated pipe break locations and res~raift~s are 3hown eft F1~. a.6~ 20, The Easyl.. sf ~aa assQ_iated stress eaietliat:iena are 9~ari~eel in 'tele 3.6h llb~ 6 1ft diBeftar~e pipe pefte~ra.e& tAa QeR~aiRMeR~ C.2-22 August 1987

RBS USAR l"te£'fttally eleseei !llve (F066), e~ter:3 the RPr; head at el 112 ft:- 11 1/4 i R

• ORly tAe 81=1.0.1: p.e8&u.i:a:ed sectiol=l of pipiR9 from tiRe Ro~lly cl.o~ed ITalITe rOEiEi to tbe aev Read is clil$sifl.Qd ail ni<Jh ef\\el'ElY*

eire~ereftti!ll breaks na?e beeR ~egE~lateQ at the' teill'miRal QRda et ~ipiFuJ' The eRly tarljet tAe EliPt~EeQ peEtteR of CAO pipQ eel:lle whi~ Ei~e to j et: tMY*s~ 'Fe.. tRe RPV.,,91;11" liIe tAo pup i~91:l1atiof\\ fpame, wnieh Rae seeR e?val~ateQ to 9ASYE8 its gt£~ot~£al iRteE}rity. ~he efteet9 ef pipe whip eft the oeAtaifUfteRt peRetratieR (lK.;a*Z19", w~pe al09 61Qftsiderelii, aRa the. PQRo1:.atieR \\1a8 QI.la1... to" to 8"8".0" i.ta flarac:tioRal. c:apabil.itwy.;

heftee, there is fte

~ee&:i Eer pi~e 'Wl'l:i~ restraifl:1!s. Pc jet ~U.&~Aa.'iiR9 f~om 1:1::10 post'll ated pipe bJ:Qaks could also ~in~e 9ft ftonnally 61190eEi ')alve AQJliGii, wAieh has seeR erJ'all:lstee ~ eRByre ita at*w~t.. *a. iRtQ'i5ity. 3C.2.10 3-In and Smaller High-Energy Piping 3C.2.10.1 Control Rod Drive Hydraulic System The piping and break locations are shown on Figures 3.6A-24b and 3.6A-24c. The stress analyses used to determine the ~reak locations are summarized in Tables 3.6A-19a and 3.6A-19b. General From the condensate storage tank, the 'cao lines enter the fuel building and, after pass~ng through two filters, connect to two drive water pumps. No breaks are postulated in this portion of the piping since it is not considered high-energy piping. 'the pressurized lines that leave the pumps go through two more filters before entering the containment building and the flow control station. The high-energy portion of the supply piping that leaves the control station is comprised of the following four lines:

1.

The charging line, which provides a constant flow of pressurized water to charge the scram accumulators in the hydraulic control units (HCUs). " 2. The qoolinq line, which maintains proper cooling of the '<#:i ve mechanisms by providing a bypass flow of water to each of the drives, via the RCUs, during normal operation periods when rod drive movement is not required. c" -I'I'Z.. No high energy line breaks as:snciated with RCIC inside containment IR postulated, and coosidcraaioDl of RCIC line rupb1re. pipewhip or wgeq arc 00 longer applicable, Howevez, the failure of the IUMM:d 6" RCIC bead spray line and information that is part of the original design will be maintaiDed as the bounding conditions. 3C.2-23 August 1987

( RBS USAR APPENDIX 3D DESIGN CRITERIA FOR PROCESS AND GUARD PIPES IN MARK III BWR CONTAINMENTS 3D.1 GENERAL A guard pipe is defined as a process pipe enclosure used to direct high-energy fluids which may escape from the process pipe as it passes through the containment back into the drywell. Guard pipes are installed on part of the reactor coolant high-energy fluids through Guard pipes are provided when to the process line: those process lines which form pressure boundary and carry the drywell and containment. the following criteria apply

1.

A crack or leak occurring in the process pipe within the containment along with a single active failure of the drywell isolation valve would otherwise r'esul t in overpressurization of the containment vessel.

2.

Either the temperature of the process fluid exceeds 200 0 F or the pressure of the fluid exceeds 275 psig. Guard pipes are installed on the following process lines: 4 in diameter main stearn lines 2 in diameter feedwater lines 1 - S-in diameter steam drain line 1 - a-in diameter ReIe steam sUi~l~ d J C I£:~rl II. ~.... I. ,~ LeN ",-112- -~~n d1ame~er ~u'aown suction line 1 - 6-in diameter RWeU suction line A typical guard pipe assembly is shown in Fig. 3.8-4. The guard pipe is joined to the process pipe by a flued head forgad extension. The guard pipe is anchored to the reinforced concrete drywell wall. A metal bellows is used to provide a flexible seal between the guard pipe and the containment vessel. A nonmetallic seal is also installed around the guard pipe where it penetrates the shield building. A lateral restraint is provided between the shield building and guard pipe to restrict the lateral movement of the guard pipe due to dynamic loads and also to 30-1 August 1987

LtN o5.tJ/f-J~() P~e I ~f" RBS USAR 5.4.6.1.1 Residual Heat and Isolation 5.4. 6.1.1. 1 Residual Heat The RCIC system initiates and discharges, within 30 sec, a constant flow of 600 gpm into the reactor vessel over a specified pressure range. The RCIC water discharged into the reactor vessel varies between a temperatu~re of 40°F 'u ,to and including a te er ture of 140°F. The ~ cool RCIC water aBd the fie~ 3tQ~~ does the following:

1.

Q\\olQ14ches the steam be-lete. :pe.'

2.

Removes reactor residual heat L.ttJ Ds.o~-IIoD

3.

Replenishes reactor vessel inventory. Redundantly, the HPCS system performs the same function, hence, providing single failure protection. Both systems use different electrical power sources of high reliability, which permit operation wi th either onsite power or offsite power. Additionally, the RHR system performs a residual heat removal function. The RCIC system design includes interfaces with redundant leak detection devices, namely: 1. A high pressure drop across a flow device in the steam supply line equivalent to 300 percent of the steady state steam flow at 1,192 psia.

2.

A high area temperature, utilizing temperature switches as described in the leak detection system. High area temperature is alarmed in the main control room.

3.

A low reactor pressure of 50 psig minimum.

4.

A high pressure between the turbine exhaust rupture diaphragms. These

devices, activated by the redundant power supplies, automatically isolate the steam supply to the RCIC turbine.

Other isolation bases are defined in the following section. HPCS provides redundancy for RCIC should RCIC become isolated, hence, providing single failure protection. 5.4-18 August 1987

RBS USAR 5.4.6.1.1.2 Isolation Isolation valve arrangements include the following:

1.

The ReIe steam line which branches off one of the main steam lines between the reactor vessel and the MSIV has two automatic motor-operated isolation valves. . One is located inside and the other outside the drywell. An automatic motor-operated inboard ReIe isolation bypass valve is used. The isolation signals noted earlier close these valves. -~- " t 2. The abandon RCIC pump discharge line penetrates the containment; however, the line * ~ is capped in the Auxiliary building and plugged at the refueling cavity floor. This ~ ~ ~wiD be tested as p8ft efthe LLRT 'Fype A test f ~ 4+-e 3. ~ Rwl5( rut Lc...N 5:q-,,/t-The ReIe turbine exhaust line vacuum breaker system line has two automatic motor-operated valves and two check valves. This line runs between the suppression pool air space and the turbine exhaust line downstream of the exhaust line check valve. Positive isolation is automatic via a combination of low reactor pressure and high drywell pressure. The vacuum breaker valve complex is placed outside primary containment due to a more desirable environment. In addition, the valves are readily accessible for maintenance and testing.

4.

The ReIe pump suction line, minimum flow pump discharge

line, and turbine exhaust line all penetrate the containment and are submerged in the suppression pool.

The isolation valves for these lines are all outside the containment and requir.e remote-manual operation. 5.4.6.1.2 Reliability, Operabili'ty, and Manual Operation 5.4.6.1.2.1 Reliability and Operability Th~ RCIC system, as noted in Table 3.2-1, is designed commensurate with the safety importance of the system and its equipment. Each component is individually tested to confirm compliance with system requirements. The system as a whole is tested during both the startup and pre-operational phases of the plant to set a base mark for system reliability. To confirm ~at the system maintains Revision 4 5.4-19 August 1991

RBS USAR Ml<.qv"'OiJ(p9 ~f? 'L91 this mark, functional and operability testing is performed at predetermined intervals throughout the life of the reactor plant (Section 5.4.6.2.4). LCN~ \\ A design flow functional test of the RCIC system may be performed during normal plant operation by drawing suction from the condensate storage tank and discharging through fu'll the de to The remains close ur~ rema1ns undisturbed. All components of system are capable of individual functional testing normal plant operation. System control provides automatic return from test to operating mode if system initiation is required. There are three exceptions:

1)

Auto/manual initiation on the flow controller is required for operator flexibility ' during ~ystem operation.

2) Closure of either or both of the steam inboard/outboard isolation valves requires operator action to properly sequence their opening.

An alarm sounds when either of these valves leaves the fully open position.

3) Other bypassed or otherwise deliberately rendered inoperable parts of the system are autom~tically indicated in the main control room at the system level.

5.4.6.1.2.2 Manual Operation In addition to the automatic operational provisions operation, initiation are included for remote-manual and shutdown of the RCIC

system, or shutdown siqnals do not exist.

I

features, startup, provided Revision 4 5.4-20 August 1991

1 \\ RBS tlSAR 5.4.6.2.2 Equipment and Component Description 5.4.6.2.2.1 Desiqn Conditions Operating parameters for the components of the acre

system, defined as

follows, are shown on Fiq. 5.4-9.

The RCte components are:

1.

One 100 percent capacity turbine and accessories

2.

One 100 percent accessories capacity pump assembly and

3.

Piping, valves, and instrumentation for:

a.

Steam supply to the turbine

b.

DELETED

c.

Turbine exhaust to the suppression pool

d.

Makeup supply from the condensate storage tank to the pump suction

e.

Makeup supply from the suppression pool to the pump suction

f.

g.

DELETj:p/i LCN ~>f-J~l. ..:.oJ

)

~... FEE.DWAT'E~ SYSTEI-'l Pump discharge to the 1\\ ;, ..:I'~ '_ _~.~ **

fteeele, including a test 11ne

~o tn~ condensate storage tank, a minimum flow bypas~ line to the suppression pool, and a cool.ant water supply to accessory equipment. The basis for the design conditions is ASME Section III, Nuclear Power Plant Components. 5.4.6.2.2.2 Design Parameters Design parameters for the RCIC system components are listed as ~ follows (BFfJ Fiq. 5.~-8 for cross-reference of compollent numbers): I I Revision 4 5.4-24 August 1991

.......e-Pump test return valve (MPL No. E51-F022) Pump suction valve, suppression pool (MPL No. E51-F031) (MPL No. 851 ~066) Ol:lteeal'ei cPtecl( ':alry'e (UE'L No. 851 E'065) Turbine exhaust isolation valve (MPL No. ESI-F068) Isolation valve, steam warmup line (MPL No. ESI-F076) Vacuum breaker isolation valves (MPL No. ES1-FO?7 & F01B) Vacuum breaker check valves (MPL No. E51-F019 FOBl) Revision 8 RBS USAR Capable of throttling control aga'1nst differential pressures up to 1,100 psi, and closure against differential pressure at 1,400 psi Capable of opening and c~osing a~ainst 75 psi fU.lC~,"'-It/I-dl.fferentl.al pressure ~t4Jl5(. this valve, and its functional ea~aeility is QelQ~nstrjjteQ separa-te.ly. This valve is tested leoally QuriR9 ~efl:lelin~ and eel~ sh~tdowR conditions. The valve is ~apable gf proper !I:lftctienal operation aurin~ maximum aMBieat ~oRdit;QAS Accessible dl:lrinq plaft~ eper a~ion aft~ capable of local Opens an or closes agal.nst 30 psi differential pressure at temperature of 330 o F. Physically located in the line on a horizontal run, as close to the containment as practical Opens and/or closes against differential pressure of 1,111 psi with minimum travel of 4 in/min Opens and/or closes against differential pressure of 30 psi at a minimum rate of 12 in/min Full flow and open with a minimum pressure drop (less than 0.5 psi) across them 5.4-28 August 1996

L cAl os. ()9-/("() P",-se J.. of & RBSUSAR 5.4.6.2.3 Applicable Codes and Classifications Dele-k.. per L ot oS-.Q~ -,too The RCrC system components within the drywell up to and including the outer isolation valve are designed in accordance with ASME Code, Section III, Class I, Nuclear Power Plant Components. The RCIC system is also designed as Seismic Category I. The RCIC system component classifications and those for the condensate storage system are given in Table 3.2-1. 5.4.6.2.4 System Reliability Considerations To assure that the RCIC operates when necessary and in time to prevent inadequate core cooling, the power supply for the system is taken from immediately available energy sources of high reliability. Added assurance is given in the capability for periodic testing during station operation. Evaluation of reliability of the instrumentation for the RCIC shows that no failure of a single initiating sensor either prevents or falsely starts the system. The most limiting operating condition for the RCIC pump occurs when the pump takes suction from the suppression pool and discharges at its rated flow of 625 gpm. This represents the limiting operating condition because of the minimum static suction head (16.5 ft) and the maximum temperature/vapor pressure (170cF/6.0 psia) of the water th~t might exist during RCIC system operation. The NPSH margin during this condition is 1.8 ft (NPSH available = 22.8 ft, NPSH required = 21 ft). The RCIC system meets the requirements of Regulatory Guide 1.1 since the calculation ofNPSH available takes no credit for increased containment atmospheric pressure accompanied by a LOCA and is computed using the maximum anticipated water temperature of 170°F. In order to assure HPCS or RCIC availability for the operational events noted previously, certain design considerations are utilized in design of both systems. 5.4*31 August 1987s

RBS USAR Physical Independence The two systems are located in separate areas of the auxiliary building. Piping runs are separated, and the water delivered from each system enters the reactor vessel via different nozzles. Prime Mover Diversity and Independence Prime mover independence is achieved by using a steam turbine to drive the RCIC pump and an electric motor driven pump for the HPCS system. The HPCS motor is supplied from either normal ac power or a separate diesel generator. Control Independence Control independence is secured by using different ba~tery systems to provide control power to each unit. Separate detection initiation logics are also used for each system. Environmental Independence Both systems are designed to meet Safety Class 1 or Safety Class 2 requirements, as applicable. Environment in the equipment rooms is maintained by separate auxiliary systems. Periodic Testing A design flow functional test of the RCIC can be performed during plant operation by taking suction from the condensate . storage tank and discharging through the full flow test ~~~!i~~~tEtb~a;c~k~~t~~~tih3e~~c~o~n~d5!e~n~s~a~:l' storage tank. The re~alns closed during n opera lon s undisturbed. All of the system are capable of individual testing during normal plant operation. Control system design provides automatic return from test to operating mode if system initiation is required. The three exceptions are as follows:

1.

The auto/manual station on " This feature is required for during system operation. the flow controller. operator flexibility

2.

Steam inboard/outboard ' isolation. valves. Closure of either or both of these valves requires operator action to properly sequence their opening. An alarm sounds when either of these valves leaves the fully open position. 5.4-32 August 1987

II .~ liif T .~ ~iit ~1ooCJ 1 i11h (f(- tJ ....... ~, ~rl...... --=---M-l L a co c I r' ~I ; 'i B ~I ~ ~l I_ i; i ~.§ i l f f$1 i~ ~ f! t i I I' c:: Ii 11.1 .§ flU 'ta III I I nl~l3 I ,~ i i~ iI;; Ii )~ l~ (~--- ----------------------------- -- ------ -- --~, \\ I,,,, l"

T I !I I + I I ! i i ~ ~ ~ .f '~ ru ~ ~ i L-- 2 F--+ o u c I ~ ;~ -~ ~ '-I"

RBS USAR 6.2.1.2. 2 Design Features The containment includes the following four subcompartments:

1.

Reactor Pressure Vessel-Shield Wall Annulus - The 2 ft thick cylindrical primary shield wall which surrounds the RPV has an outside diameter of 29 ft 10 in and extends from the vessel pedestal to el 147 ft 5 in. Breaks in the recirculation water outlet piping and feedwater piping are analyzed.

2.

Drywell Head The drywell head is located above the RPV head and surrounds the RPV head, connecting to the drywell bulkhead at el 162 ft 3 in. Five normally open ventilation exhaust hatches are located in the bulkhead at azimuths 30, 75, 165, P*"- LC ~\\ 51-/fl.." 225, and 345 deg venting into the drywell. (Theae ~lSl ~ hatches are ~ closed only during refueling ~ -- --~ J ~ .~... ~ ( 3. RWCU Heat Exchanger Room - The RWCU heat exchanger,

room, located at el 147 ft 3 inches in the containment, vents through the wire door in the south wall and through two 13 ft x 2 ft 2 in openings in the north wall into the containment.

RWCU line breaks are analyzed in this room.

4.

RWCU FilterjDemineralizer Rooms The RWCU filter/demineralizer rooms are located at azimuth 270 deg and el 162 ft 3 in. Piping penetration sleeves provide the only vents from the filter/demi neralizer rooms. RWCU piping is routed to and from the demi neralizers through the east wall of the cubicles which separates them from the holding pump room and valve nest area. Complete circumferential DER of the a-in diameter RWCU line connected to the bottom of the demineralizer is analyzed in this subcompartment. Drawings depicting

piping, equipment, and compartment/venting locations are provided in Section 3. 6.

The volumes and vent areas are discussed in Sect~on 6.2.l.2:~. The subcompartments described do not incorporate blowout panels. No credit is taken for vent areas that become available after the pipe break occurs. 6. 2. 1.2.3 Design Evaluation The breaks utilized in the design evaluation of the containment subcompartments are listed in Table 6.2-13. The 6.2-43 August 1987

RSS USAR steam isolation valves open for this function during postulated small leaks or breaks. Therefore, high drywell pressure has been deliberately omitted from isolation of main steam lines. These lines do isolate on the parameters identified as Group 6 in Section 6.2.4.3.7 and via remote manual operation.

2.

Residual Heat Removal System The Residual Heat Removal System complies with all applicable portions of NUREG-0737, Item 1I.E.4.2, except:

a.

Position 1: The RHR automatic isolation valves in the nonessential lines listed below do not receive the high dr~ell pressure signal. Penetration Title Valve No. lKJB*Z3A lKJB*Z3B D£L£re~B.= PtfL LC N 5.'1-1'11.- lKJB*Z20 lKJB*Z21A .lKJB*Z21B RHR Shutdown Cooling Injection lE12*MOVF053A RHR Shutdown lE12*MOVF053B Cooling Injection ~~~~:---lS12::Q2a-, ~~~-R~~ ___.~ RHR Shutdown Cooling lE12*MOVF008 Supply lE12*MOVF009 Coolant to Sparger in Reactor Vessel Coolant to Sparger in Reactor Vessel lE12*MOVF037A lE12*MOVF037B The high drywell pressure (Signal K) deliberately omitted from the isolation logic for these line valves to avoid the RHR shutdown cooling mode for small leaks. has been initiation the loss of breaks or 6.2-88 August 1987

fiBS USAo TABLE 6.2-13 COITAIHftEHT SOBCOftPART!EKT AIALYSIS SO""l~Y Design Tables ________ I!gqres

• Basis

'ent 'odal Line M04al Path Blowdown !lodalizat ion "odal Pressure

i9~~Q!2!!:!!~!!.~

.!!Q!\\~ J!~~!!_ De§£I!P!:!Q!!. ~~ru:tlill9n _!!!!L- ---'ll~!_ ~~sgres ~!f~entlals PPl' - ShiElld 27 Hode( I, Peed"ater 6.2-14 6.2-15 6.2-16 6.2-38 6.2-39 Vall AnDulus t\\py - Shield Vall ADnolus 25 Rode'l) Feed.ater 6.2-17 6.2-18 6.2-19 6.2-41 6.2-'2 Bn - Shield lall Annulus 26 Node' Z) Recirco latioD 6.2-20 6.2-21 6.2-22 6.2-44 6.2-45 Drywell Head 2 lode water o~ tlCIC be (3) 6.2-23 6.2-24 6.2-25 6.. 2-47 6.2-48 spraJ RiCO Heat Ezchanger 1'00. 2 Ifode nCD 6.2-26 6.2-27 6.2-213 6.2-50 6.2-51 RiCU PiltE'rl Deaineralizer Roo.s q !lode RWCO 6.2-29 6.2-30 6.2-31 6.2-53 6.2-511 ADO?ff LtN. {.~-I~L Aon 9iR \\.c.1'l !I.'{,\\'\\1.. (1)"o4el of co.plete (3600) aDllulos (t04e1 of half (180 0 ) of anDulus due to sa ** ar .J ) The RCIC bead spray [nihis been aeleted and-the wociatoo nigh energy line-breaks are no longer possible. However this failure and infonnation is being provided as the bounding (Gnditions tbat were established as part of the original plant design and licensing basis. '-"-""-.J"-'--'-__ - __ ....... J~......., ** ~ ... ~ ........ ~ ~ __ ~ I of 1 6.2-.0 6.2-43 6.2-46 6.2-49 6.2-52 6.2-55 August 1987 ~ L~~ ~8 ~~ vt--o

l'oluae Volllae Teap. _-1!2.s... l£lL{!l. j~ll 1 4629 150 2 231561 150 RBS OSAR 'Us!.! e. 2-2~ I')OD pu. lLl'l'A-ll\\1.. SUBCOftPARTftEMT NODAL DESCRIPTION 6-1",crc READ SP~AY III! B~EAK DRYWELL BEAD SOBCOftPARTftEIT

---

.ru!~ (lr~ConaitiODs ___

Initial Conditio~ ___ Break pressur~ Do.iaity " Break Break Area Break -lJ!~!.!l ~L !g!..!._No.!. liruL J§~LtU An!L lq.1 50 100 PCIC 0.181 DER 14.1 50 0 - Calculated

• Peak Pressure Difference

-ill.!!L-7.96 0.0 -'AOD fU-Lc..l-1~A-l42.. ( I ) The Rele head spray line has been deleted and the associated high energy line breaks are no longer possible. However this failure and infonnation is being provided as the bounding conditions that were established as part of the original plant design and licensing basis.

---

~-------------------------

• Mcdal peak pressure ainus pressure in Node 2 iPi-PZ) 1 of 1

~ '-> ~ ~~ ~~ ~! Aaqllst 1987

Vent Path !2.:, 1 2 3 4 5 Fro. '1'0 Description Yo1. Vol. of Tent lIode lIode Path Flow 'fQ.:, !Q.:, j£~~~dlOncbQ~£~r 1 2 UDchoket! 1 2 Unchoked 1 2 UnchoJted 1 2 onchoked 1 2 Onchoked PBS USAf> TABLe 6.2-2&-- A-DD Pit Lc.N ~q~I'lL SUBCO~~ARTftENT VENT PATH DESC~IPTIO" 6-IH aCIC ijeAD SPBAY LI"E BR!A~ DPYWE1L HElD SUBCO"PA~TftEKT Vent Area L/A !l~!Lt.o ~~L~Hicie.B! ________ j~f!l ill:~ Fri£!1QB Turning Ezgansion £ootractiQn IQ!!! 1.07 0.1267 0.001 1.29 0.99~ 0.492 2.18 2.6/& 0.1160 0.001 1.21 0.998 0.464 2.14 2.64 0.09 0.001 1.29 0.998 0.416 2.77 2.64 \\).0848 0.001 1.29 0.998 0./&78 2.17 2.64 0.071 0.0;)1 1.29 0.998 0./&82 2.17 .%0 PU LCN6A-I'/L ( I ) The RCIC head spray line bas been deleted and the lluociated high energy line breaks are no longer possible. However this failure and infonnation is being provided liS the bounding conditions tbat were established as part of tbe original plant design and licensing basis. ~ (j;:; ~~ ~Q "-l~ 1 of 1 AIlCJust 1987

Time (sec) 0.0 10.0 RBS USAR TABLE BLOWDOWN DATA 6-IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMENT Blowdown Blowdown Mass Blowdown Energy Flow Rate Enthalpy Release Rate (Ibm/sec) (Btu/lbm) (Btu/sec) 407.25 1,191.0 4.85 x 10 5 407.25 1,191.0 4.85 x 10 5 For this case, the mass and energy release is assumed constant until after the occurrence of the peak pressure difference between Nodes 1 and 2. Note 2 The RCIC bead spray liDe bas been deleted and the awx:iated high energy line breaks are DO longer possible. However, this failure and information is being provided as the bounding __ conditions that were estab1.isbed as part of the original plant design and licensing bases. 1 of 1 August 1987

(

l.

2.

3.

4.

RBS USAR TABLE 6. 2-35 CRITERION 55 - INFLUENT LINES, REACTOR COOLANT PRESSURE BOUNDARy<l) Influent Lines Feedwater HPCS RHR return to feedwater LPCS and LPCI

a. LPCI A line

b. LPCI B line

c. LPCI Cline

d. LPCS line RD tem su (llRefer to Fig.

6.~-63 and Table 6.2-40. Paragraph 55.1 55.2 55.3 55.4 55.4 55.4 55.4

55.

55.1 - The feedwater line is part of the reactor coolant pressure boundary as it penetrates both the primary containment and the drywell to connect with the reactor pressure vessel. It has three isolation valves and one guard pipe. The isolation valve inside the drywell is a simple check valve. The quard pipe protects the primary containment from overpressurization in the event of a pipe failure between the drywell and primary containment walls. Outside the primary containment is a testable (air-assisted to close) check valve, and farther away from the primary containment is a motor-operated gate valve with provisions to utilize the PVLCS. Should a break occur in the feedwater line, the check valves prevent significant loss of inventory and offer immediate isolation. During the postulated LOCA, it is desirable to maintain reactor coolant makeup from all sources of supply. For this reason, the motor-operated valve does not automatically isolate upon signal from the reactor protection system. However, this valve is capable of being remotely closed from the main control room to provide lonq-t~rm leakage protection upon operator judqment that continubd makeup from the feedwater. source is unnecessary.

• ... 3 55.2 - The aPCS line penetrates both the primary containment and the drywell to inject water directly into the RPV.

Isolation is provided by a manually testable check valve located inside the drywell and a motor operated gate valve 3"* Revision 3 1 of 3 August 1990

( RBS USAR TABLE 6.2-35 (Cont) /l/1<.1(p ~ ~ 9 . pay 21, with remote manual actuation which is located as close as possible to the exterior wall of the shield building. Long-term leakage control is maintained by this block valve. If a LOCA occurred, this motor-operated block valve would receive an automatic signal to open. During operation of the HPCS system, the aPCS liquid temperature is low enough that primary containment overpressurization cannot exist if the line breaks between primary containment and drywell. 55.3 - The RHR return to feedwater is joined by the RWCU AIJ\\)~(-IC. return line outside the primary containment to form a common line with feedwater which penetrates the primary containment and the drywell to discharge into the RPV. Inside the drywell is a simple check valve. For a line break outside the primary containment, isolation is provided by this check valve and an automatically actuated motor-operated block valve located outside and as close to the primary containment as possible. Long-term leakage control is ensured by this block valve. A guard pipe between the primary containment and the drywell protects the primary containment from overpressurizing.

• .... 3 55.4 - Satisfaction of isolation criteria for the LPCI mode of the RHR system and the LPCS system is accomplished by use of an automatic or remote manual motor-operated block valves and manually testable check valves where applicable.

These valves are normally closed with the block valve located outside the ' primary containment receiving an automatic signal to open at the appropriate time to ensure that acceptable fuel design limits are not exceeded in the event of a LOCA. Lines LPeI A and B each have a normally closed motor-operated valve between the drywell and primary containment going to the refueling cavity sparger, and two normally closed manual valves on branch lines, one going to the fuel pool cooling and purification system, and,the other coming from the condensate makeup and drawoff system. 3~*, 55.5 - Isolation in the control rod drive supply line is provided by a simple check valve on the line inside the primary cont~inment and a remote manual block valve (motor-operated valve) ' on the line outside the primary containment. The supply line is not a part of the RCPS. 2 of 3 August 1990

( RBS USAR TABLE 6.2-35 (Cont) /l/1<.1(p ~~ 9 . pa..y 21, with remote manual actuation which is located as close as possible to the exterior wall of the shield building. Long-term leakage control is maintained by this block valve. If a LOCA occurred, this motor-operated block valve would receive an automatic signal to open. During operation of the HPCS system, the HPCS liquid temperature is low enough that primary containment overpressurization cannot exist if the line breaks between primary containment and drywell. 55.3 - The RHR return to feedwater is joined by the RWCU A~DI1..(-IC. return line outside the primary containment to form a common line with feedwater which penetrates the primary containment and the drywell to discharge into the RPV. Inside the drywell is a simple check valve. For a line break outside the primary containment, isolation is provided by this check valve and an automatically actuated motor-operated block valve located outside and as close to the primary containment as possible. Long-term leakage control is ensured by this block valve. A guard pipe between the primary containment and the drywell protects the primary containment from overpressurizing.

• -+3 55.4 - Satisfaction of isolation criteria for the LPCI mode of the RHR system and the LPCS system is accomplished by use of an automatic or remote manual motor-operated block valves and manually testable check valves where applicable.

These valves are normally closed with the block valve located outside the ' primary containment rece~v1ng an automatic signal to open at the appropriate time to ensure that acceptable fuel design limits are not exceeded in the event of a LOCA. Lines LPCI A and B each have a normally closed motor-operated valve between the drywell and primary containment qoin9 to the refueling cavity sparger, and two normally closed manual valves on branch lines, one going to the fuel pool coolinq and purification system, and the other coming from the condensate makeup and drawoff system. 3"*. 55. 5 - Isolation in the control rod drive supply line is provided by a simple check valve on the line inside the primary cont~inment and a remote manual block valve (motor-operateA valve) ' on the line outside the primary containment. The supply line is not a part of the RepS. 2 of 3 August 1990

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() RBS USAR TABLE 6.2 - 40 (CONTJ C(l~'Al lH10VQH LO::. lE.NOIH Of VALVE ~Cftt. OE&I"," PIPE faDM CIUllA,. LIN!. ($SOV LINE 'SAN lSOl..., lUff TYP( lYP£ c: ,EH. 5'$1£.. KAkI£ fLUID

• SUE ESF
• t56 lEAKAOE

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CD DRYWELL HEAD DRYWELL BELOW REFUELING SEAL ~ R.9 (.p ~ CfJ~ 9 fXtJ~ 2S-1 rc. Note The RCle head spray line bas bceD dc1eCeci aDd the associated high energy line breab are DO . longer pocsaolc. However, this failure aud information is being provided as the bounding alIIditioPs that..... CSlabtisbod.. put.. til< :pao1 plaut desilP' and \\ioeosj.g....... I j ~. ( LC." 5, Lj -/,-/z, ~bJl S (. PUL-FIGURE 6.2-47 . NODALIZA TlON DIAGRAM 6 - IN RCIC HEAD SPRAY LINE BREAI< DRYWELL HEAD SUBCOMPARTMENT RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT

28 26 24 ~ ~ u; 22

0. -

L I 14 12 0.00 /DRYWELL HEAD P, - ~ ,DR~ ~ 1.00 2.00 3.00 4.00 TIME AFTER BREAK (SEC) Note Tbc R.CIC bead spray liDc bas been deleted and the associated high c:ncrgy liDc breaks are DO longer possable. However, this failure and information is being provided as the bounding . CODditiona that were established as part of the original plant design and lia:nsing bases. FIGURE 6.2-48 NODAL PRESSURES 6 ~ IN RCIC HEAD SPRAY LINE BREAK DRYWELL HEAD SUBCOMPARTMEN' RIVER BEND STATION UPDATED SAFETY ANALYSIS REPOR

Q vt A. - u.I ~ vt vt UoI ~ A. ~ -... Z UoI ~ UoI u.. u.. 0 10.00 8.00 ( 6.00 I 4.00 2.00 0.00 0.00 ........ ~ Pl-P2 1.00 2.00 3.00 4.00 TIME AFTER BREAK (SEC) The ROC bead spray line has been dcletcd and the associated high energy line breaks are DO lODger po6sible. However, this failure and information is being provided as the bounding conditions that were established as pan of the original plant design and licensing bases. FIGURE 6.2-49 NODAL PRESSURE DIFFERENTIALS 6-IN RCIC HEAD SPRAY LINE BREA~ DRYWELL HEAD SUBCOMPARTMENl RIVER BEND STATtON UPDATED SAFETY ANALYSIS REPORT

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LeN OS,d'i-lfcD Pa.je 'i of 0 RBS USAR System Operation Schematic arrangements of system mechanical equipment operator information display are shown in Fig. 5.4-8. system component control logic is shown in Fig. 7.4-1. and RCIC Instrument location

drawings, and elementary diagrams are identified in Section 1.7.

e~12 The RCIC system can be initiated either manually or automatically. The initiation signal logic is then sealed-in for a fixed time duration to allow system startup. The MCC circuitry maintains the components sealed in after initiation. The main control room operator can initiate RCIC by operating the manual initiation switch which simulates an automatic initiation or by activating each piece of equipment sequentially as required. The ini tiation signal for steam flow to the RCIC turbine and water injection to the reactor vessel is sealed in for a specified time period and then drops ou~is i itiation signal to the gland seal compressor ~.~ is sealed in and remains sealed in until m nu res. Delete.. fer LCJJ CJS.()I./-I{'o 12+-e RCIC is automatically initiated by four redundant differential pressure transmitters/trip relay

contacts, arranged in a

one-out-of-two-twice logic configuration, which sense reactor vessel low water level (trip level 2). The RCIC steam line isolation and the turbine steam exhaust motor-operated (MO) valve control switches are keylocked in the open position. The turbine trip and throttle valve is normally open aDd requires no change of position for automatic system initiation. The RCIC system responds to an automatic initiation signal and reaches design flow rate within 30 sec as follows (actions are simultaneous unless stated otherwise) :

1.

The pump suction from the condensate storage tank valve MO FOIO is signaled open.

2.

~o ensure pump discharge flow is directed to the reactor vessel only, the test return lines to the condensate storage tank valves MO F022 and MO F059 are signaled closed. Revision. 12 7.4-3 December 1999l'\\

RBS USAR Ill£.. 9tp "1J()tR9 IJ4je.. 29( lE12*MOVF006B - Motor-operatea valve (shutdown ,cooling) lE12*MOVF008 - Motor-operated valve (outboard shutdown isolation) lE12*MOVF009 - Motor-operated valve (inboard suction isolation) lE12*MOVFOllA - Motor-operated valve (RHR heat ~ PIfl... l.CN t).~.. Nl exchanger flow to suppression pool) Me~er epera~ed walve (~eae~er head spray) Motor-operated valve. (RHR test line) lE12*MOVF027A - Motor-operated valve (injection shutoff) lE12*MOVF037A - Motor-operated valve (shutoff upper pool cooling) lE12*MOVF042A - Motor-operated valve (RHR injection) lE12*MOVF047A - Motor-operated valve (heat exchanger shell side inlet) 1E12*MOVF048A - Motor-operated valve (heat exchanger shell side-bypass) lE12*MOVF040 - Motor-operated valve (discharge to radwaste) lE12*MOVF053A - Motor-operated valve (RHR injection) lE12*MOVF064A - Motor-operated valve (RHR pump, minimum flow) 1E12*MOVF068A - Motor-operated valve (heat exchanger I water discharge valve) Revision 4 7.4-12 Auc;ust 1991

L. eN ~St ()II-I 100 Pa.~e s of "

f.

RBS USAR Channel Independence {IEEE 279-1971, Paragraph 4.6) Channel independence is maintained as described in Section 8.3.1.4. The and Protection Interaction (IEEE ~~s;L.I.-I~raph 4.7) has no interaction with plant ~ele.k fer-LCN (:S,Dl./ -1100 'The RGIG inter=aotion \\,'itfi tfie main turbine 'oFltrels i s t.hre\\ol:~l:l an i solation device, ~a~~~dancy i~ ~aiHtainQd hy the use o£ ~iver=se eq'lipr:urnt to raac~.afa.11101:tdo';!n as QQscrisoa o r er di cuss~on on to G C 24, see Section 3.1. h. Derivati on of System Inputs (IEEE 279-1971, Paragraph 4. 8) All inputs to the RCIC system that are essential to its operation are direct measures of appropriate variables. SLCS display instrument ation in the main control room provides the operator with directly measured information on reactor vessel water

level, pressure, neutron flux

level, and control rod position and scram valve status.

Based on this information the operator can assess the need for SLCS.

i.

Capability for Sensor Checks (IEEE 279-1971, Paragraph 4.9) Refer to Section 7.4.2.3, Regulatory _Guide 1.22.

j.

Capability for Test and Calibration (IEEE 279-1971, Paragraph 4.10) Refer to Section 7. 4.2.3, Regulatory Guide 1.22.

k.

Channel Bypass or Removal from Operation (IEEE 279-1971, Paragraph 4.11) 7.4-26 August 1987

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LICENSINu CHANGE NOTIl:.£ --- ENTERGY EQUIPMENT NO. REFUELING se:AL SUPPORT NOTE 5 E31-vr004 E2 F1G. NO. LeN NO. oq,63-tX'w3 PAGE t OF 3 FROI.! CCP.RV151 9.'3-7e

LCN 05'.0'-/-//:,0 fa5e ~ of t:, RBS USAR --+7

11.

Both turbine trip pushbutton controls in the main control room are actuated

12.

Low emergency trip system pressure with main generator offline and no mechanical trip valve output pressure

13.

Loss ofEHC 125 VDC trip power with turbine speed below a certain level.

14.

Reactor vessel water level high - level 8 (two-out-of-three logic) from reactor feedwater control system. Refer to Sections 7.7.1.3 and 10.4.7.5.1 for further discussion on feedwater system. IS. ~~I\\Kb~ ~ be,lek per-7~- lttjll5',D4-/~o The *turbine trip signals are monitored by the plant computer. Alanns are provided in the main control room for loss of turbine 125 VDC power, excessive turbine thrust bearing wear, and condenser low vacuum. An indication of the cause of a turbine trip is provided on the EHC cabinet (GE first hit detection annunicator system). 10.2.6.2 Turbine Supervisory Instruments The following turbine supervisory instruments (TSI) are provided in the main control room: I. Turbine eccentricity, speed, and control valve position recorder

2.

Turbine vibration recorder

3.

Differential expansion (difference in expansion between turbine rotor and shell), turbine rotor expansion, and temperature (exhaust hood, first stage shell, valve chest, and crossaround pipe) recorder. Alanns are provided in the main control room for the following:

1.

High turbine differential expansion

2.

Low turbine differential expansion --+8

3.

Card out of file or TSI power supply failure

4.

TSllow voltage failure

5.

High turbine vibration

6.

Reactor vessel water level high

7.

~J)e.\\e+e per lLN. {)5( (;tl - /~O

8.

Turbine Vibration Trip Enabled 8~- Revision 8 10.2-10 August 1996

LICE:NSINu CHANGE NOTICLI LeN NO. 10.3-113 --- ENTERGY o I I') CJ) EQUIPMENT NO. LeN No. 10.3-113 (REF ER01 - 0218- 000) REACTOR PRESSURE VESSEL B13*REV D003(Z- ) FIG. NO. PAGE __ OF __ 10.3-1a

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