View source for CNS-18-008, License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to Exempt These CIVs from Type-C Local Leak Rate Testing

{{Adams
| number = ML18200A252
| issue date = 07/19/2018
| title = License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to Exempt These CIVs from Type-C Local Leak Rate Testing
| author name = Simril T
| author affiliation = Duke Energy Carolinas, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000413, 05000414
| license number = NPF-035, NPF-052
| contact person = 
| case reference number = CNS-18-008
| document type = Letter type:CNS, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Bases Change
| page count = 55
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Tom Simril
( ., DUKE                                                                                 Vice President ENERGY                                                                   Catawba Nuclear Station Duke Energy CN01VP I 4800 Concord Road York, SC 29745 o: 803 .701 .3340 f: 803 .701 .3221 tom.simril@duke-energy.com CNS-18-008                                                                             10 CFR 50 .90 July 19, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==Subject:==
Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Renewed Facility Operating License Numbers NPF-35 and NPF-52 License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to exempt these CIVs from Type-C Local Leak Rate Testing Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) requests an amendment to the Catawba Unit 1 and 2 Renewed Facility Operating Licenses to revise the UFSAR Section 6.2.4.2.2 to remove the Containment Valve Injection Water System (NW or CVIWS) Supply from specified Containment Isolation Valves (CIVs) and to exempt these CIVs from Type-C Local Leak Rate Testing (LLRT)
The Enclosure provides the description and assessment of the proposed changes. Attachment 1 contains the existing Updated Final Safety Analysis Report (UFSAR) pages marked to show the proposed changes. Attachment 2 provides the existing Technical Specification (TS) Bases pages marked-up to show the proposed changes, for information only. Changes to the existing TS Bases will be implemented under the TS Bases Control Program .
In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this amendment request has been previously reviewed and approved by the Catawba On-Site Review Committee The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The bases for these determinations are included in the Enclosure.
There are no regulatory commitments contained in this letter or enclosure.
Duke Energy is requesting that the NRC review and approve this LAR within one year from the date of submittal. Duke Energy is also requesting the standard 120-day implementation period in conjunction with this LAR.
WWW.duke-energy .com

U.S. Nuclear Regulatory Commission July 19, 2018 CNS-18-008 Page 2 In accordance with 10 CFR 50.91, Duke Energy is notifying the State of South Carolina of this license amendment request by transmitting a copy of this letter and enclosures to the designated State Officials.
Please direct any questions or concerns to Carrie Wilson, Senior Engineer, Catawba Regulatory Affairs, at (803) 701-3014.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on July 19, 2018.
Sincerely, Tom Simril Vice President, Catawba Nuclear Station

==Enclosure:==
Description and Assessment of the Proposed Change Attachments:
: 1. UFSAR Pages (Mark-up)
: 2. Technical Specification Bases Pages (Mark-up, For Information Only)

U.S. Nuclear Regulatory Commission July 19, 2018 CNS-18-008 Page 3 cc:     (with Enclosure/Attachment)
C. Haney, USNRC Region II - Regional Administrator J. Austin, USNRC Senior Resident Inspector - CNS M. Mahoney, NRR Project Manager A. Wilson, Attorney General (SC)
Lynne Garner, Manager, Radioactive and Infectious Waste Management (SC)

ENCLOSURE Description and Assessment of the Proposed Change

==Subject:==
License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to exempt these CIVs from Type-C Local Leak Rate Testing
: 1. 

==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION 2.1   System Design and Operation 2.2   Current Licensing Basis 2.3   Reason for the Proposed Change 2.4   Description of the Proposed Change
: 3. TECHNICAL EVALUATION 3.1 NW System Design Basis 3.2 NI System Design Basis 3.3 NS System Design Basis
: 4. REGULATORY EVALUATION 4.1   Applicable Regulatory Requirements/Criteria 4.2   Precedent 4.3   No Significant Hazards Consideration Analysis 4.4   Conclusions
: 5. ENVIRONMENTAL CONSIDERATION
: 6. REFERENCES ATTACHMENTS:
: 1. UFSAR Pages (Mark-up)
: 2. Technical Specification Bases Pages (Mark-up, For Information Only)

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 2
: 1. 

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) requests an amendment to the Catawba Unit 1 and 2 Renewed Facility Operating Licenses to revise the UFSAR Section 6.2.4.2.2 to remove the Containment Valve Injection Water System (NW or CVIWS) Supply from the following Safety Injection (NI) and Containment Spray (NS) Containment Isolation Valves (CIVs) and to exempt these CIVs from Type-C Local Leak Rate Testing (LLRT):
* 1(2)Nl-162A: NI Pumps to Cold Leg Injection Header Isolation Valve
* 1(2)Nl-173A: Residual Heat Removal (ND) Header A to Reactor Coolant (NC) Loops C and D Cold Legs Isolation Valve
* 1(2)Nl-178B: ND Header B to NC Loops A and B Cold Legs Isolation Valve
* 1(2)Nl-1838: ND Header A and B Hot Leg Injection Isolation
* 1(2)NS-29A: NS Pump A to NS Header A Containment Isolation Valve
* 1(2)NS-32A: NS Pump A to NS Header A Containment Isolation Valve
* 1(2)NS-43A: ND Pump A to NS Spray Header Containment Isolation Valve
* 1(2)NS-12B: NS Pump B to NS Header B Containment Isolation Valve
* 1(2)NS-15B: NS Pump B to NS Header B Containment Isolation Valve
* 1(2)NS-388: ND Pump B to NS Spray Header Containment Isolation Valve The proposed change is acceptable because the affected valves do not constitute a potential primary containment atmospheric pathway during and following a Loss of Coolant Accident (LOCA). Furthermore, a program is in place that monitors and reduces leakage from those portions of systems outside containment to ensure these systems will maintain containment isolation functions.
The CNS Updated Final Safety Analysis Report (UFSAR) will be updated, as required, to reflect this License Amendment Request (LAR) to abandon the NW system to specified CIVs after the change has been approved. See Attachment 1 for the markup of proposed changes to Section 6 of the CNS UFSAR.
: 2. DETAILED DESCRIPTION 2.1 System Design and Operation The Containment Valve Injection Water System (NW) prevents leakage of containment atmosphere past certain gate valves used for containment isolation following a LOCA by injecting seal water at a pressure exceeding containment accident pressure between the two seating surfaces of the flex wedge valves. The system consists of two independent, redundant trains; one supplying gate valves that are powered by the A train diesel and the other supplying gate valves powered by the B train diesel. This separation of trains prevents the possibility of both containment isolation valves not sealing due to a single failure.
Each train consists of a surge chamber, which is filled with water and pressurized with nitrogen.
One main header exits the chamber and splits into several headers. A solenoid valve is located in the main header before any of the branch headers, which will open after 60-second delay on ST signal (Phase A Containment Isolation initiated by means of Containment Pressure - High Signal or upon receipt of a Safety Injection Signal). Each of the headers supply injection water to containment isolation valves located in the same general location, and close on the same engineered safety signal. A solenoid valve is located in each header, which supplies seal water to valves closing on SP signal (Phase B Containment Isolation on receipt of Containment

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 3 Pressure - High-High Signal). These solenoid valves open after a 60 second delay on SP signal.
Since a ST signal occurs before a SP signal, the solenoid valve located in the main header will already be injecting water to Containment isolation valves closing on ST signal. This leaves an open path to the headers supplying injection water on SP signal. The delay for the solenoid valves opening is to allow adequate time for the slowest gate valve to close, before water is injected into the valve seat.
Individual solenoid valves that are provided for the Safety Injection (NI) and Containment Spray (NS) containment isolation valves receiving seal water injection. Following a postulated accident, one or more of the NI or NS containment isolation valves may be open. These solenoid valves will receive a signal to open, following a ST signal and once its associated containment isolation valve has closed. If the containment isolation valve subsequently opens, its associated NW solenoid valve closes. Thus after a ST signal, the NW solenoid valve will be open when its containment isolation valve is closed and vice versa.
One header for each train penetrates the Containment. The NW Containment isolation valve on the outside of the Containment opens on ST signal, allowing seal water to be injected to those containment isolation valves located inside the Containment. Inside Containment, solenoid valves isolate the headers that supply injection water to those valves closing on SP signal. The solenoid valves open after a 60 second delay on SP signal.
Makeup water is provided from the Demineralized Water Storage Tank for testing and adding water to the surge chamber during normal plant operation. Assured water is provided from the essential header of the Nuclear Service Water System. This supply is assured for at least 30 days following a postulated accident. If the water level in the surge chamber drops below the low-low level or if the surge chamber nitrogen pressure drops below the low-low pressure after a ST signal, a solenoid valve in the supply line from the Nuclear Service Water System will automatically open and remains open, assuring makeup to the NW system at a pressure greater than 110% of peak Containment accident pressure.
The NW system is designed to meet all Regulatory and Testing requirements set forth in 10 CFR 50, Appendix J and ASME Code Section IX.
Where the potential exists to overpressurize containment penetration piping due to thermal expansion of the fluid trapped in the penetration piping, overpressure protection is provided.
Overpressure protection is provided to relieve the pressure buildup caused by the heatup of a trapped volume of incompressible fluid between two positively closing valves (due to containment temperature transient) back into containment where an open relief path exists. This open relief path could be the relief valve on any normally aligned component, or to the Reactor Coolant System (NC) itself.
2.2 Current Licensing Basis Catawba Units 1 and 2 are presently licensed with the Containment Valve Injection Water System (NW) that prevents leakage of containment atmosphere past certain gate valves used for containment isolation following a LOCA by injecting seal water at a pressure exceeding containment accident pressure between the two seating surfaces of the flex wedge valves.
The subject CIVs are currently NW Leak Rate Tested (Reference 6.39) consistent with the original licensing basis which was under 10 CFR 50, Appendix J, prior to availability of Option B.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 4 UFSAR Figure 6-113 - Valve Arrangement (VAL ARR)
* g 85   *              ~                 *                                &#xa3; g
88   *             ~
* g 89   -~--~--~*                            --+----~~-----
INSIDE    OUTSIDE
                                                                                          ...[_ SEAL vi ATE~
CONTAINMENT
* OPEN TO CONTAINMENT Provisions to detect possible leakage from the lines outside containment during normal operation are required by Technical Specification Program 5.5.3, Primary Coolant Sources Outside Containment and are governed for NI and NS Systems by the following periodic test procedures to ensure that the closed systems outside containment provide an acceptable secondary containment boundary during and after a LOCA:
* PT/1 (2)/A/4150/002 - Visual Inspection of Radioactive Systems Outside Containment (Reference 6.20)
* PT/1 (2)/A/4201/001 - Leak Rate Determination for the NI System (Reference 6.21)
Test Methodology: The NI system is aligned to recirculate the Refueling Water Storage Tank (FWST) to pressurize the system and a visual of system components is performed to identify leakage. Components with identified leakage are required to have a Work Request/Order submitted to repair the leak, an action initiated to track the status of the leak, and boric acid corrosion concerns reported to Fluid Leak Management.
* PT/1 (2)/A/4203/003 - Leak Rate Determination for NS System Outside of Containment (Reference 6.22)
Test Methodology: The NS system is aligned to recirculate the FWST to pressurize the system, and a visual inspection of system components performed to identify leakage. Components with identified leakage are required to have a Work Request/Order submitted to repair the leak, an action initiated to track the status of the leak, and boric acid corrosion concerns reported to Fluid Leak Management.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 5 The following table is excerpted from the system Design Basis Documents and FSAR Tables 6-74 and 6-77 to provide details of the valve service and testing with the following applicable notes:
UFSAR Table 6-74 Notes (in part):
lli1IES..!_
: 2. AN ATTACHED CLOSED SYSTEM IS CONSIDERED AS A BOUNDARY WHICH PRECLUDES BYPASS LEAKAGE IF THE SYSTEM:
A. EITHER <UOOES NOT DIRECTLY COMMUNICATE, I.E., PRESENT A CL OSED PRESSURE BOUNDARY, WITH THE CONTAINMENT ATMOSPHERE, OR IZIDOES NOT DIRECTLY COMMUNICATE, I.E., PRESENT A CLOSED PRESSURE BOUNDARY WITH THE ENVIROMENT, FOLLOWING A LOCA.
B. CLOSED PRESSURE BOUNDARY IS DESIGNED TO DUALITY GROUP B OR C STANDARDS.
C. SYSTEM CLOSED PRESSURE BOUNDARY IS SEISMIC CATEGORY 1.
D. IF THE SYSTEM CLOSED PRESSURE BOUNDARY IS INSIDE CONTAINMENT, ITS DESIGN PRESSURE AND TEMPERATURE EXCEEDS OR IS EOUAL TO CONTAINMENT DESIGN PRESSURE AND TEMPERATURE. IF THE SYSTEM CLOSED PRESSURE BOUNDARY IS OUTSIDE CONTAINMENT, ITS DESIGN PRESSURE EXCEEDS OR IS EQUAL TO CONTAINMENT DESIGN PRESSURE.
E. SYSTEM CLOSED PRESSURE BOUNDARY IS DESIGNED FOR PROTECTION FROM THE EFFECTS OF PIPE WHIP, MISSILES, AND ANY JET FORCES RESULTING FROM THE LOCA.
F. SYSTEM CLOSED PRESSURE BOUNDARY IS MAINTAINED DURING NORMAL PLANT OPERATION.
: 6. ALTHOUGH THE CONTAINMENT ISOLATION SYSTEM IS ITSELF AN ENGINEERED-SAFETY-FEATURE, THESE LINES BELONG TO SYSTE~S WHICH PERFORM AS ENGINEERED-SAFETY-FEATURE AFTER LOCA.
UFSAR Table 6-77 Notes (in part):
CATAWBA NUCLEAR STATION                                          APPENDIX 6. CHAPTER 6 TABLES AND FIGURES NOTES TD TABLE 6-77 PAGE 1 OF 4 CONTAINMENT ISOLATION VALVE AND ACTUATION DATA NOTES1
: 2. DEFINITION OF ACTUATION SIGNALS S - SAFETY INJECTION SIGNAL CT SIGNAL ALSO ACTIVATED BY S SIGNAL>
T - CONTAINMENT ISOLATION SlGNAL (PHASE A CONTAINMENT ISOLATION>
P - CONTAINMENT HIGH-HIGH PRESSURE SIGNAL (CAUSES MAIN STEAM LINE ISOLATION AND PHASE 8 CONT
* 1SOL. >
Abbreviations not previously defined:
* Pen No. - Mechanical Penetration Number
* SOV - Solenoid Valve
* MAN - Manual Valve
* ACT SIG -Actuation Signal
* CIV POS - CIV Position
* FS - CIV Fail Safe Position
* SD - CIV Shut Down Position
* PA - CIV Post Accident Position
* ESF - Engineered Safety Features o POS - Position

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 6 UFSAR Table 6-74, 6-77, and Desi Jn Basis Document (DBD) Information NW                                                    CIV T 6-74, 6-77         TG-77                     T6-77                 TG-74 VLV                                                   DBD SEISMIC CAT sov 1 CLOSED CIV CIV         ACT               PRESS ITEM NO./    PEN   GD   VAL                        POS-              REMARKS / POTENTIAL   ESF /
SIG            BOUND TO SERVICE     NO. C   ARR                         FS-              BYPASS LEAKAGE PATH   POS MAN                   ENVIRON SD-PA AFTER LOCA (NOTE 2) 1(2)NW                               ISOLATION VALVES 29 / NI PUMPS                   1(2)
                                        -185A         O-FAI-       X     OPEN DURING LOCA TO A & 8 DISCHG   352  55    BS    NI-          NA                                              NO 1(2)NW          C-O                   ALLOW COLD LEG TO COLD LEGS                    162A
                                         -186                              RECIRC-NOTE 6 / NO 1(2)NW                               ISOLATION VALVES 27 / ND HX A                   1(2)
                                        -175A         O-FAI-       X     OPEN DURING LOCA TO DISCHG TO     336  55    BS    NI-          NA                                              NO 1(2)NW          0-0                   ALLOW COLD LEG COLD LEGS                     173A
                                         -176                              RECIRC-NOTE 6 / NO 1(2)NW                               ISOLATION VALVES 28 / ND HX B                   1(2)
                                        -2378         O-FAI-       X     OPEN DURING LOCA TO DISCHG TO     307  55    85    NI-          NA                                              NO 1(2)NW          0-0                   ALLOW COLD LEG COLD LEGS                     178B
                                         -238                              RECIRC-NOTE 6 / NO 24/ND                            1(2}NW                               ISOLATION VALVES 1(2)
CROSSOVER                           -2428         C-FAI-       X     OPEN DURING LOCA TO 207  55    88    NI-          NA                                              NO DISCHG TO                          1(2)NW         C-C/O                   ALLOW HOT LEG 1838 HOT LEGS                            -243                             RECIRC (NOTE 6) / NO 1(2}NW                               ISOLATION VALVES 38 /                      1(2)  -2178                      X     OPEN DURING LOCA TO   YES/
C-FAI-CONTAINMEN     387  56    89    NS-           p                       ALLOW CONTAINMENT     OPE 1(2)NW          C-O T SPRAY LINE                    128                                   SPRAY FLOW (NOTE 6) /   N
                                         -220 NO 1(2}NW                               ISOLATION VALVES 37 /                      1(2)  -2228                      X     OPEN DURING LOCA TO   YES/
CONTAINMEN                                    p C-FAI-380  56    89    NS-                                   ALLOW CONTAINMENT     OPE 1(2)NW          C-O T SPRAY LINE                    158                                   SPRAY FLOW (NOTE 6) /   N
                                         -225 NO 1(2)NW                               ISOLATION VALVES 36 /                      1(2)  -200A                      X     OPEN DURING LOCA TO   YES/
C-FAI-CONTAIN MEN    370  56    89    NS-           p                       ALLOW CONTAINMENT     OPE 1(2)NW          C-O T SPRAY LINE                    29A                                   SPRAY FLOW (NOTE 6} /   N
                                         -203 NO 1(2}NW                               ISOLATION VALVES 35 /                      1(2)  -19SA                      X     OPEN DURING LOCA TO   YES/
CONTAIN MEN                                   p C-FAI-362  56    89    NS-                                   ALLOW CONTAINMENT     OPE 1(2}NW          C-O T SPRAY LINE                   32A                                    SPRAY FLOW (NOTE 6) /   N
                                         -198 NO 1(2)NW                               ISOLATION VALVES 40/ND                       1(2)  -2278                      X     OPEN DURING LOCA TO CONTAIN MEN                                        C-FAI-369  56    89    NS-          NA                       ALLOW CONTAINMENT      NO 1(2)NW         C-C/O T SPRAY LINE B                  388                                    SPRAY FLOW (NOTE 6} /
                                         -228 NO 1(2)NW                               ISOLATION VALVES 39/ ND                      1(2}  -180A                      X     OPEN DURING LOCA TO CONTAIN MEN                                        C-FAI-381  56    89    NS-          NA                       ALLOW CONTAINMENT      NO 1(2)NW         C-C/O T SPRAY LINE A                  43A                                    SPRAY FLOW (NOTE 6) /
                                         -181 NO

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 7 2.3 Reason for the Proposed Change Elimination of NW from the specified CIVs is desired primarily due to reliability issues with the NW solenoids that supply the associated CIVs. NW solenoid valve failures render the train of NW inoperable due to possible flow diversion paths exceeding makeup capability. This change will improve NW system reliability, reduce personnel dose associated with solenoid valve troubleshooting and maintenance, and reduce the rate of TS 3.6.17, Condition A. One CVIWS train inoperable, 7 day LCO action statements and associated organizational response. Removing NW from these CIVs will not require any additional testing. (e.g., LLRT) 2.4 Description of the Proposed change Duke Energy proposed to revise the CNS UFSAR Section 6.2.4.2.2 Containment Valve Injection Water System to remove the reference to specific solenoid valves that supply seal water to valves closing on an SP signal. These solenoid valves are for the NI and NS containment isolation valves receiving seal water injection. UFSAR Table 6-77 will also be modified to remove the NW system designator for LRT type. See Attachment 1 for UFSAR mark-up pages.
: 3. TECHNICAL EVALUATION The specified CIVs, per CNS UFSAR, Table 6-77, (1(2)Nl-162A, 1(2)Nl-173A, 1(2)Nl-178B, 1(2)Nl-183B, 1(2)NS-29A, 1(2)NS-32A, 1(2)NS-43A, 1(2)NS-12B, 1(2)NS-15B, 1(2)NS-38B) are currently required to be NW leak tested on an 18 month frequency along with respective Unit/Train CIVs supplied by NW under SR 3.6.17.2, "Verify valve injection flow rate is< 1.29 gpm (Unit 1)
 < 1.21 gpm (Unit 2) for Train A and < 1.16 gpm for Train B with a surge tank pressure ~ 36.4 psig." This surveillance requirement verifies that each NW train can perform its required function when needed by measuring the existing conditions for the valves being injected. The CIVs served by the NW system may be tested individually or simultaneously.
CNS was approved for 10 CFR 50, Appendix J, Option B and Regulatory Guide 1.163 under Technical Specification Amendment Number 192/184 on July 31, 2001. Option B endorses Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program" (Reference 6.12) as acceptable guidance for implementing Option B. Regulatory Guide 1.163 further endorses NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J." NEI 94-01 is currently Rev. 3-A incorporating regulatory positions stated in Regulatory Guide 1.163 (September 1995) as modified by NRC Safety Evaluations of June 25, 2008 and June 8, 2012. NEI 94-01, Rev. 3-A establishes criteria for when an LLRT is not required on a Containment Isolation Valve:
         "An LLRT is a test performed on Type B and Type C components. An LLRT is not required for the following cases:
* Primary containment boundaries that do not constitute potential primary containment atmospheric pathways during and following a Design Basis Accident (OBA);
* Boundaries sealed with a qualified seal system; or,
* Test connection vents and drains between primary containment isolation valves which are one inch or less in size, administratively secured closed and consist of a double barrier."
The exemption of the valves from the LLRT program is based on the guidance provided by 10 CFR 50, Appendix J, Option B, NEI 94-01, Rev. 3A and ANSI 56.8-2002.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 8 The valves do not constitute a potential primary containment atmospheric pathway during and following a Design Basis Accident (OBA), as noted in UFSAR Table 6-74.
3.1 NW System Design Basis The NW System is designed to inject water between the two seating surfaces of gate valves used for Containment isolation. The injection pressure is higher than Containment design peak pressure during a LOCA. This will prevent leakage of the Containment atmosphere through the gate valves, thereby reducing potential offsite dose below the values specified by 10CFR100 limits following the postulated accident.
Per Reference 6.33, "Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits."
The Containment Valve Injection Water System (NW) is designed to provide a water seal to a specific class of containment isolation valves (gate valves) during a LOCA, to prevent leakage of containment atmosphere through the gate valves. During normal power operation, the system is in a standby mode and does not perform any function.
The NW System shall be activated by an ST signal to perform its safety related function. That function is limiting the release of containment atmosphere past specific containment isolation valves, in order to mitigate the consequences of a LOCA. Containment Isolation Valves, for systems which are not used to mitigate the consequences of an accident, will be supplied with NW System seal water upon receipt of an ST signal. Containment Isolation Valves, for accident mitigating systems which are supplied with seal water from the NW System, have their seal water supplies actuated by an SP signal.
The NW System containment piping penetrations shall be classified as Duke Class B in accordance with Reference 6.34. Containment penetration isolation valves shall be provided in accordance with Reference 6.2.
Reference 6.35, provides design information associated with the NW System containment penetrations. This table identifies NW System valves provided in accordance with containment penetration design criteria. NW System containment penetrations M243 and M253 are classified as Overpressure Protection Category 2.a per Section 3.2.13 of Reference 6.34.
Justification of the classification is documented in Reference 6.36.
The NW penetration valves are not required to be type C tested. The NW System penetrations are in use during and following an accident to provide Containment Valve Injection Water System flow to certain containment isolation valves. In the event that the containment isolation valves on these NW System penetrations should fail open, an effective water seal would be maintained at a pressure greater than peak Containment pressure by the containment valve injection water surge chamber.
The NW System design shall meet the requirements set forth in Reference 6.37. Specific requirements are as follows:

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 9
: 1. Section 111.C.2.B. Valves which are sealed with fluid from a seal system shall be pressurized with that fluid to a pressure not less than 1.10 times peak containment internal pressure.
: 2. Section 111.C.3.B. The installed isolation valve seal water system fluid inventory shall be sufficient to assure the sealing function for at least 30 days at a pressure of 1.10 times peak containment internal pressure.
3.2 NI System Design Basis The Catawba Safety Injection (NI) System constitutes a major portion of the Emergency Core Cooling System (ECCS). The NI System, along with the Refueling Water (FW) System, Residual Heat Removal (ND) System, and Chemical and Volume Control (NV) System, provides emergency cooling to the reactor core in the event of a break in either the primary (reactor coolant) system or the secondary (steam) systems. An incident which causes a loss of fluid from the Reactor Coolant (NC) System is referred to as a Loss of Coolant Accident (LOCA). In the event of a Steam Line Break (SLB) or a Feedwater Line Break (FLB), the NI System provides borated water to the core to control core reactivity.
The NI System is used to make up to the Cold Leg Accumulators (CLAs) during Normal Operation in order to meet the level, pressure and boron concentration Technical Specification requirements for the CLAs.
The safety related function of the Safety Injection (NI) System is to provide emergency core cooling in order to prevent unacceptable fuel damage and to assure that the core remains intact during all phases of a Design Basis Event (DBE). To avoid excessive core damage, it is necessary to limit fuel temperatures, particularly the cladding temperature, and limit any zirconium-water reaction to an insignificant amount. These temperature and reaction limits are documented in Section 2.1.2.29, "Loss of Coolant Accidents."
The NI System is normally aligned to function as a part of the Emergency Core Cooling System (ECCS). During normal operation, NI is aligned to inject borated water from the Refueling Water Storage Tank (FWST) to the NC cold legs upon receipt of a Safety Injection signal. Refer to Section 2.1.2.1, for additional information.
The NI System, along with the Residual Heat Removal (ND) System, Chemical and Volume Control (NV) System, and Refueling Water (FW) System, is a part of the ECCS at Catawba. As required by Reference 6.29, a system to provide abundant emergency core cooling shall be provided. The ECCS components are designed such that a minimum of one NV Pump, one NI Pump, one ND Pump, and three low pressure CLAs together with their associated valves and piping will assure adequate core cooling in the event of a Design Basis Loss of Coolant Accident. The system safety function shall be to inject borated water to the core and assist the ND System in transferring heat from the core following any loss of reactor coolant at a rate such that (1) the core remains intact and (2) reactivity is controlled. The Design Basis Events which impose requirements on the design of the Safety Injection (NI) System are:
Main Feedwater Line Rupture Main Steamline Break (including spurious steam line valve openings)
* Steam Generator Tube Rupture Loss of Coolant Accident - Small Break (including spurious pressurizer valve openings)
* Loss of Coolant Accident - Large Break
* Control Rod Drive Mechanism Rupture

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 10 There are three primary functions of the ECCS. These functions are: (1) removing stored and fission product decay heat, (2) controlling reactivity, and (3) precluding reactor vessel boron precipitation. Following is a description of these functions and how the NI System supports each function.
A Loss of Offsite Power (LOOP) may occur by itself or concurrent with other Design Basis Events such as a Loss of Coolant Accident (LOCA). During a LOCA/LOOP event, the NI System is required to serve as part of the ECCS by transferring decay heat and other residual heat from the NC System. The LOCA/LOOP event represents the worst case scenario and bounds operational requirements of the NI System during other Design Basis Events.
Valve 1(2)Nl-I62A is a motor operated gate valve located on the cold leg injection header from the NI Pump discharge. This valve is normally open and remains open with power removed to preclude inadvertent closure during cold leg injection and recirculation.
Valve 1(2)Nl-I73A is a motor operated gate valve located on the ND Pump A discharge line to NC Loops C and D cold legs. This valve is normally open with power removed to preclude inadvertent closure and remains open during cold leg injection and recirculation.
Valve 1(2)Nl-I78B is a motor operated gate valve located on the ND Pump B discharge line to NC Loops A and B cold legs. This valve is normally open with power removed to preclude inadvertent closure and remains open during cold leg injection and recirculation.
Valve 1(2)Nl-183B is a motor operated gate valve located on the ND Pump discharge header to NC Loops B and C hot legs. This valve is normally closed with power removed to preclude inadvertent opening and remains closed during cold leg injection and recirculation.
For Valves 1(2)Nl-162A, 1(2)Nl-173A, & 1(2)Nl-178B:
Following a Loss of Coolant Accident, power is restored, and these valves are closed when realigning the NI System to the hot leg recirculation mode of operation. Each valve serves as an outside containment isolation valve for its respective penetration; however, it does not perform an automatic containment isolation function since the penetration is sealed against leakage by a water seal provided by the NW System when the valve is closed during hot leg recirculation and open during cold leg injection and recirculation. The control circuit for the NI MOV provides interlocks for the control of its associated NW SOV, which controls the injection of water between the disks of the NI MOV to aid in containment isolation.
For Valve 1(2)Nl-183B:
In the event that neither NI pump is able to provide hot leg recirculation flow following a Loss of Coolant Accident, power to the valve can be restored and the valve opened to align the ND pump discharge to the NC system Hot Legs. It should be noted that this alignment is not credited for the mitigation of any Design Basis Event and is only available as an added level of defense. 1(2)Nl-183B serves as an outside containment isolation valve for penetration M207; however, it does not perform an automatic containment isolation function since the penetration is sealed against leakage by a water seal provided by the NW System when the valve is closed during cold leg injection and recirculation and open during hot leg recirculation. The control circuit for 1(2)Nl-I83B provides interlocks for the control of valve 1(2)NW-242B, which controls the injection of water between the disks of valve 1(2)Nl-183B to aid in containment isolation.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 11 Taibl, I~ Ya.In Pos;idol,$ for tb~ Saftty hj<<rioa       tf'lll LOCA Pha5H Y.dnNo.                                  Co!dL      hj.         Cold Lq: Rtt.irt.        BoruaRtdrt.
fN19A . 10B                                  Optu                     Opm                     Optn iNI47A                                      Clofflt                   C1oted                  Clowd
 ~NJS.4~ 6SB. 76A. SSB 1                      Optn                     Opm                     Optn
 !~1195A                                     Closed                    Oowd                   Cloud iNI96B                                      Clowd                     Closed                  Closed SNilOOB S~lO)As 13)B 1"Sl115~ U4A lNlllSk !SOB OpfA Opm Optn Opfn Cl0cted Open Clo$ed Opell
                                                                                               ~-
Clowd Cloud 0-ostd INU20B                                     Ciowd                     C10kd                   Clo-std INU21A.152B                                c~                        Closed                   Optn
 -l ~tUB. lS3A                               c~                        Closed                 *c~
1
     'Vlh ~ GJ.Jy bt do~ m tbt ~  of :t Smlll-Bn:u. LOCA to ~~t~ dill CL.Ai~~            ~-~ of mtro        m1olh.,NCSy:
S~       C~"S-lUl.!\1-00-0001 Dllt.            ...\pril l!., lHJ Re\*. 60 PJl't l0.9'of142 ColdL~lnj.                                       Hor UC Rrcirt.
J~U6B 1NI1478                                                             Clo'ted                 Ctostd lNUS4B. l8SA l}ti1332A. lHB JNIU4B

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 12 NI System Sketch SAFETY INJECTION SYSTEM (NI)
Nl-332A
              ''Er'
                                                                                             ~ ND I A CLA Hrt L:eg;                    NDPwnp                          FtllLme G-o!i!> Connect "C'
FWST Cokl Leg "D" er \
                          ~~f.-----<'.'      ND I A Nl- l 18A ColdLeg"C" Nl-115A
              'I*\        650 GPM ct 7 ,; pw,:
run h'1- 162A Cold Leg, "B" (f.A                                                 Nl-1508                  NI- J-i7B
                          ~                 ND I B CoklLeg"A" Nl- 100B tl..-\
                "r\"
Hot Let~
run                       **      < ND 1B "D"                                                                                     Nl- 13613

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 13 ECCS System Sketch EMERGENCY CORE COOLING SYSTEM (ECCS)
FWST
                                                                                                                                                   ' , ha,~?lll.!
NV-252A                                             NV-294            NV-3 1:!       NV-314
                                                                                                'Nl -9                                                                     All NV-253B                                                                                                                           Ctlld Legs VCT                                                                          Nl- 10 NV- 188A         NV- 1898 l\'D-28A C
Cold Le~
Nl- 18SA 1'.'D-26             l\'D-32A          Nl: 173A     I     -~ o B
Conta inment Sump                                                                                                                         Hotu:-gi.
                                                                                                                            "l\'I- I R3B              '-- C ND-59B Nl - 184 B                                                                                                                    B
                                                                                                                                         ~
ND-0SB Cold u:-gs FW-55B                                                                         1'.'D-60 A
KC NJ.'.J J2A NI~134B                                                                                                                             r--- B Hot u:-g, NI~3J3B NI- II RA                           L--_.. c I\.'l - 103A All
                                                                                                                ~ Cokl Legs
                                                                               !'l.'1- 1478 Nl- 100B
                                                                                                                                                     .--- A Hot Legs
                                            'Nl - 1358                                                                                   Nl -1.528   .____.._ D

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 14 3.3 NS System Design Basis The Catawba Containment Spray (NS) System is an engineered safeguard feature which serves to remove thermal energy from the containment atmosphere in the event of a Loss Of Coolant Accident (LOCA). It performs this function in conjunction with the Emergency Core Cooling System (ECCS), which cools the reactor during injection and recirculation modes of operation. However, the NS System is not part of the ECCS. The heat removal capability of the spray system maintains containment pressure below the design pressure value after the ice in the Ice Condenser (NF) has been depleted, and steam generated in the core continues to enter containment. The NS System also serves to remove fission product iodine from the post-accident containment atmosphere. In addition, the NS System is designed for the suppression of steam partial pressure in the upper containment volume due to operating deck leakage from a LOCA.
The primary safety related function of the Containment Spray (NS) System is to remove thermal energy from the containment atmosphere in the event of a Loss Of Coolant Accident (LOCA).
Pressure suppression is provided for all break sizes up to and including the double-ended rupture of the largest pipe in the Reactor Coolant (NC) System. The NS design is based on the conservative assumption that reactor core residual heat is continuously released to the containment as steam, eventually depleting the ice in the Ice Condenser. The Containment Spray (NS) System is sized to maintain the containment pressure below the design pressure (15 psig). In addition, the NS System is capable of rapidly reducing the containment pressure and temperature following a LOCA and maintaining them at acceptably low levels.
In addition to the energy input to the containment assumed for the Design Basis Event (DBE) described above, the NS System is also designed, assuming the energy release from a metal-water reaction of 33 percent of the Zircaloy fuel cladding, with hydrogen burned as it is produced, to limit containment internal pressure to below design pressure. The NS System also serves to remove fission product iodine from the post-accident containment atmosphere.
In addition, the NS System is designed for the suppression of steam partial pressure in the upper containment volume due to operating deck leakage from a LOCA. This capability is inherent in the system if the primary design requirement is met. The requirement for the spray is to absorb steam leakage through the operating deck at the maximum possible long term deck differential pressure of one pound per square foot, which is equivalent to the Ice Condenser door opening differential pressure.
During normal unit operation, the Containment Spray (NS) System is in a standby ready mode and does not perform any specific function.
The Design Basis Events which impose requirements on the design of the Containment Spray (NS) System are:
* Loss of Coolant Accident - Small Break (includes spurious pressurizer valve openings)
* Loss of Coolant Accident - Large Break Valve 1(2)NS-12B is a motor operated gate valve located on the NS Pump B discharge line to NS spray header B. This valve, along with 1(2)NS-15B, 1(2)NS-29A, 1(2)NS-32A, is normally closed during unit operation. 1(2)NS-12B serves as an outside containment isolation valve for penetration M387.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 15 Valve 1(2)NS-15B is a motor operated gate valve located on the NS Pump B discharge line to NS spray header B. This valve, along with 1(2)NS-12B, 1(2)NS-29A, 1(2)NS-32A, is normally closed during unit operation. 1(2)NS-15B serves as an outside containment isolation valve for penetration M380.
Valve 1(2)NS-29A is a motor operated gate valve located on the NS Pump A discharge line to NS spray header A. This valve, along with 1(2)NS-12B, 1(2)NS-15B, 1(2)NS-32A, is normally closed during unit operation. 1(2)NS-29A serves as an outside containment isolation valve for penetration M370.
Valve 1(2)NS-32A is a motor operated gate valve located on the NS Pump A discharge line to NS spray header A. This valve, along with 1(2)NS-12B, 1(2)NS-15B, 1(2)NS-29A, is normally closed during unit operation. 1(2)NS-32A serves as an outside containment isolation valve for penetration M362.
Valve 1(2)NS-38B is a motor operated gate valve located on the ND Pump B discharge line to its respective ND Auxiliary Containment Spray Header. This valve, along with 1(2)NS-43A, is normally closed during unit operation. 1(2)NS-38B serves as an outside containment isolation valve for penetration M381.
Valve 1(2)NS-43A is a motor operated gate valve located on the ND Pump A discharge line to its respective ND Auxiliary Containment Spray Header. This valve, along with 1(2)NS-38B, is normally closed during unit operation. 1(2)NS-43A serves as an outside containment isolation valve for penetration M369.
The following functional description for 1(2)NS-12B is the equivalent for 1(2)NS-15B 1 1(2)NS-29A and 1(2)NS-32A (except 1(2)NS-29A & 1(2)NS-32A are Train A related):
1(2)NS-12B is interlocked with the Containment Pressure Control System (CPCS).
1(2)NS-12B is manually operated from the control room, concurrent with the CPCS permissive (containment pressure greater than or equal to 1.00 psig), to provide a flow path from NS Pump B to its respective spray nozzles. 1(2)NS-12B is closed and prevented from opening by the CPCS when containment atmosphere pressure is less than or equal to 0.25 psig.
The CPCS controls for the NS valves are independent of those for the NS Pumps. If a single failure allowed NS Pump B to continue to operate with containment pressure being less than or equal to 0.25 psig, 1(2)NS-128 would not open, thereby preventing an erroneous spray inside containment. Closure of 1(2)NS-128 also serves to minimize the effects of a potential waterhammer due to NS Pump restart following the subsequent trip at the CPCS setpoint. This valve serves as a second line of defense against potential waterhammer in addition to check valve 1(2)NS-99 which is intended to prevent backleakage of the water in the vertical piping to the spray header through NS Pump B.
With containment pressure, either actual or simulated, at or above 1.00 psig, 1(2)NS-12B may be manually operated via Open/Close pushbuttons located on main control board MC11 with "red-green" position indicating lights.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 16 The following functional description for 1(2)NS-38B is the same for 1(2)NS-43A (< > reflects 43A related):
As a backup provision for the NS System, the ND System can be manually aligned through valves 1(2)NS-38B and 1(2)NS-43A to provide flow to two separate sets of spray nozzles.
The ND Auxiliary Containment Spray headers can be aligned for beyond design basis events where containment pressure could exceed the design pressure.
1(2)NS-38B <1 (2)NS-43A> is interlocked with the Containment Pressure Control System (CPCS). 1(2)NS-38B <1 (2)NS-43A> is closed and prevented from opening by the CPCS when containment atmosphere pressure is less than or equal to 0.25 psig . Closure of 1(2)NS-38B <1 (2)NS-43A>                  also serves to minimize the effects of a potential waterhammer due to ND Pump operation following the first actuation of ND Auxiliary Spray and automatic isolation at the CPCS setpoint. This valve serves as a second line of defense against potential waterhammer in addition to check valve 1(2)ND-44 <1 (2)ND-10> which is intended to prevent backleakage of the water in the vertical piping to the spray header through ND Pump B <A>. Valve interlocks are provided such that 1(2)NS-38B <1 (2)NS-43A> can only be opened if containment isolation valve 1(2)Nl-184B
       <1 (2)Nl-185A> is fully opened, ND isolation valve (1 (2)ND-368 or 1(2)ND-37A) <1 (2)ND-1B or 1(2)ND-2A> is closed, and containment pressure is greater than or equal to 1.00 psig .
1(2)NS-38B <1 (2)NS-43A> may be manually operated, subject to the above interlocks, via Open/Close pushbuttons located on main control board MC 11 with "red-green" position indicating lights.
Spec . C7'"'S-1563.:'\$-00-0O01 Date:       No,ember 1, 1993 Rev. 34 Page 43 of70 Table 1. Yal.e Posi tions for the Containment Spra)- S)*,stem and LOCA Phases Yah-e ~o.                               Cold Leg Inj.        Cold Leg Red rc.      Hot Lfg Recirc.
LNI184B, 185A                              Closed                 Open                     Open lNSlB, 18A                                 Closed                 Open                     Open l~S3B, 20A                                  Open                  Closed                  Closed l~SS, 25                                   Closed                 Closed                  Closed l~S12B, 15B, 29A 32.A 3
1NS38B 43A 1NS70                                      Closed                                         Closed 1 Valv es openuponreceipt ofa High-High Containment Pressure Signal,. using the CPCS asa pennissive.
2 Valves  open if ND Au.-xili.ary Containment Spray is desired.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 17 NS System Sketch CONTAINMENT SPRAY SYSTEM (NS)
ND ~mp )
                                                                                                      ~
NS-43A 1\J-----C><]
LO.
ZS ZS ZS: ZS .6 ND~p )
                                                                                                      ~
NS-38B l\J-----{X]
LO ZS ZS ZS ZS .6 FWST I                                                    L.C RHT    ~                    UIRN-14:
I I          I  NS~ A           LO NS l-lx NS - 20A                 I     \   I    \      A
                                      -T Gl NS- I SA H       I~
RN- 144A I         NS -32A         r_o ND Pump A
Nl- 185A L C.
Conrauunent I                                 RN Sump ND Pump
                                                           , RNnl:
B NI-184BJ NS-18 NS- 128 I      I                  LO.
NSHx B
                                             ~        \___
NS-38
_J
                                                                       /
RHT I    ~
NS- l SB        LO n r RN-2298

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 18 Technical Specification Information 1(2)Nl-162A, 1(2)Nl-173A, 1(2)Nl-178B, 1(2)Nl-183B, 1(2)NS-43A and 1(2)NS-38B are located in ECCS Systems required by TS 3.5.2, ECCS Operating, TS 3.5.3 ECCS Shutdown for accident mitigation. 1(2)Nl-173A and 1(2)Nl-178B are also in the flow path of Residual Heat Removal (RHR) system during shutdown operation and therefore support RHR operability per Technical Specification 3.4.6, 3.4.7, 3.4.8, 3.9.4 and 3.9.5. 1(2)NS-29A, 1(2)NS-32A, 1(2)NS-12B and 1(2)NS-15B, are required operable per TS 3.6.6, Containment Spray. The ECCS, RHR and NS functions of these valves are not affected by removal of NW from the valves.
Technical Specification 3.6.17, Containment Valve Injection Water System (CVIWS), requires two operable trains. The surveillance requirements establish a minimum accumulator tank pressure, the maximum total leakage of all NW supplied valves per train and verification that each automatic valve actuates to the correct position to maintain operability. The technical specification does not list the valves supplied by NW or the automatically actuated NW valves. The Bases for TS 3.6.17, mentions that some valves required for accident mitigation are supplied by NW. The statement in the bases will be changed under the provisions of the TS Bases Control Program. A mark-up of TSB 3.6.17 has been provided in Attachment 2 for information only.
1(2)Nl-162A, 1(2)Nl-173A, 1(2)Nl-1788, 1(2)Nl-1838, 1(2)NS-29A, 1(2)NS-32A, 1(2)NS-128, 1(2)NS-158, 1(2)NS-43A and 1(2)NS-38B are containment isolation valves required per Technical Specification 3.6.3, Containment Isolation Valves. Removal of NW from the listed NI and NS valves would not impact Technical Specification 3.6.3, since the penetrations isolated by these valves are not bypass leakage paths per UFSAR Table 6-74 and therefore do not require Type C leak rate testing.
From 10 CFR 50.36 (c)(2)(ii):
      "A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:
(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety."
With respect to Criterion 1, the removal of NW from the subject CIVs does not involve or remove installed instrumentation that would indicate degradation of the reactor coolant pressure boundary.
With respect to Criterion 2, the NW system is designed to limit primary containment leakage by application of seal water 1.1 times the containment design pressure between the gate valve seats.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 19 The subject NI and NS CIVs are not bypass leakage paths based on UFSAR Table 6-74, because they are connected to a closed, seismic and missile protected, Safety Class 2 systems outside containment used for accident mitigation.
With respect to Criterion 3 and 4, the subject NI and NS valves remain containment isolation valves required by Technical Specification 3.6.3, even if NW supply to the individual valves is removed. The subject NI and NS CIVs are not bypass leakage paths because they are connected to a closed class 2 system outside containment that is used for accident mitigation. Therefore, the valves do not require Type C Leak rate testing.
The four criteria in 10 CFR 50.36 (c)(2)(ii) are not affected by this change, therefore no additional Technical Specifications are required. The removal of NW does not require a change to Technical Specification 3.6.3 or 3.6.17.
Summary 1(2)Nl-162A, 1(2)Nl-173A, 1(2)Nl-178B, 1(2)Nl-183B, 1(2)NS-29A, 1(2)NS-32A, 1(2)NS-43A, 1(2)NS-12B, 1(2)NS-15B, 1(2)NS-38B valves are containment isolation valves listed in Table 6-77 of the UFSAR. The Safety Injection (NI) system valves are isolation valves located in penetrations classified as General Design Criterion (GDC) 55, Reactor Coolant Pressure Boundary Penetrating Containment, because they connect to the Reactor Coolant (NC) System.
The Containment Spray (NS) system valves are classified GDC 56, Primary Containment Isolation, because these penetrations supply the NS spray rings which communicate with the containment atmosphere. These penetrations also perform Engineered Safety Functions (ESF) in support of containment heat removal during an accident. The design requirements for containment penetrations are established in ANSI N271/ANS-56.2, Containment Isolation Provisions for Fluid Systems. GDC 55 and GDC 56 require a normally closed or automatic containment isolation valve inside and outside containment unless it can be demonstrated that the design is acceptable on some other defined basis.
ANSI N271/ANS-56.2, Section 3.6 describes guidelines for "Other Defined Basis." The "other defined basis" for these penetrations is described in ANSI N271/ANS-56.2 Section 3.6.3 and allows remote manual valves for systems associated with Engineered Safety Features in lieu of an automatically closing isolation valve. In this case, the penetration is isolated by means of a check valve inside containment and a remotely operated manual gate valve outside containment.
Per UFSAR Table 6-74, Potential Bypass Leakage Paths through Containment Isolation Valves, these penetrations are not classified as "Potential Bypass Leakage" pathways. Per Note 2 of Table 6-74, the penetrations are attached to a closed system boundary. The closed system boundary design requirements are further defined in Note 2. Since these penetrations are connected to closed loop systems outside containment any leakage past the inner or outer isolation would be retained by the Class B, missile protected, closed systems outside containment. The check valves inside containment on each of these penetrations do not require type C leak rate testing (LLRT) based on UFSAR Table 6.74 and Table 6-77. The subject manual gate valves outside containment are currently supplied with NW per notation in UFSAR Table 6-77, and they can be exempted from the LLRT program based on the guidance provided by 10 CFR 50, Appendix J, Option B, NEI 94-01, Rev. 3A and ANSI 56.8-2002.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 20 The following Failure Modes and Effects Analysis is provided to further support the conclusions covered in this Section:
Note: An additional boundary is being maintained by the closed system outside containment.
Valve No. 1(2)Nl-162A Single Active Failure          ECCS Operation            Effects Phase None (base case)               Cold Leg Injection        None. Valve is normally open with power removed.
Cold Leg Recirculation    None. Valve is normally open with (CLR)                     power removed.
Hot Leg Recirculation     Valve is closed via control board (HLR)                     switch. After HLR is aligned, 1(2)NI-162A is isolated by closing 1(2)NI-118A and 1(2)Nl-150B. NI Pump(s) are started which pressurizes boundary which would prevent any seat leakage from Containment.
Loss of A-Train AC Power       Cold Leg Injection        None. Valve is normally open with power removed.
Cold Leg Recirculation    None. Valve is normally open with power removed.
Hot Leg Recirculation     With no A-Train AC power, valve cannot be closed to align to HLR.
1(2)Nl-150B is closed to provide train separation and B-Train HLR. Check valve downstream of 1(2)Nl-162A will provide isolation from Reactor Coolant System (RCS) for A-Train.
Loss of B-Train AC Power      Cold Leg Injection        None. Valve is normally open with power removed.
Cold Leg Recirculation   None. Valve is normally open with power removed.
Hot Leg Recirculation    With no B-Train AC power, valve is closed to align to HLR. 1(2)Nl-118A is closed to provide train separation and A-Train HLR. Check valve downstream of 1(2)Nl-162A will provide additional isolation from RCS.
Failure of Associated Pump     Cold Leg Injection        None. Valve is normally open with to start                                                 power  removed.
Cold Leg Recirculation    None. Valve is normally open with power  removed.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 21 Valve No. 1(2)Nl-162A Single Active Failure        ECCS Operation         Effects Phase Hot Leg Recirculation  Valve is closed via control board switch. After HLR is aligned, 1(2)NI-162A is isolated by closing 1(2)NI-118A and 1(2)Nl-1508. The available ND Pump may be started. Check valve downstream of 1(2)Nl-162A will provide additional isolation from RCS.
Failure of valve to open     Cold Leg Injection     None. Valve is normally open with power   removed.
Cold Leg Recirculation None. Valve is normally open with power   removed.
Hot Leg Recirculation  None. Valve is normally open with power   removed.
Failure of valve to close    Cold Leg Injection     None. Valve is normally open with power removed.
Cold Leg Recirculation None. Valve is normally open with power removed.
Hot Leg Recirculation  After HLR is aligned, 1(2)Nl-162A is isolated by closing 1(2)Nl-118A and 1(2) N 1-1508. Check valve downstream of 1(2)Nl-162A will provide additional isolation from RCS.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 22 Valve No. 1(2)Nl-1838 Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. Valve is normally closed with power removed. Check valve downstream of 1(2)Nl-183B will provide isolation from RCS as well.
Cold Leg Recirculation  None. Valve is normally closed with power removed. Check valve downstream of 1(2)Nl-183B will provide isolation from RCS as well.
Hot Leg Recirculation   Valve is opened via control board switch. After HLR is aligned, 1(2)Nl-173A and 1(2)Nl-178B are closed to isolate CLR. ND Pump(s) are started to establish HLR.
Loss of A-Train AC Power         Cold Leg Injection     None. Valve is normally closed with power removed.
Cold Leg Recirculation None. Valve is normally closed with power removed.
Hot Leg Recirculation  With no A-Train AC power, valve is opened to align to HLR.
1(2)ND-65B is closed to provide train separation and HLR will not be available through the ND pumps.
Loss of B-Train AC Power         Cold Leg Injection     None. Valve is normally closed with power removed.
Cold Leg Recirculation None. Valve is normally closed with power removed.
Hot Leg Recirculation  With no B-Train AC power, 1(2)ND-32A is closed to provide train separation and HLR will not be available through the ND pumps. Check valve downstream of 1(2)Nl-183B will provide additional isolation from RCS.
Failure of Associated Pump to   Cold Leg Injection      None. Valve is normally closed start                                                   with power removed.
Cold Leg Recirculation  None. Valve is normally closed with power removed.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 23 Valve No. 1(2)Nl-1838 Single Active Failure           ECCS Operation          Effects Phase Hot Leg Recirculation   Valve is opened via control board switch. 1(2)Nl-173A and 1(2)NI-1788 are closed to isolate CLR.
1(2)ND-32A or 1(2)ND-658 is closed to provide train separation.
A or 8-Train ND Pump is placed in service.
Failure of valve to open        Cold Leg Injection      None. Valve is normally closed with power removed.
Cold Leg Recirculation  None. Valve is normally closed with power removed.
Hot Leg Recirculation   None. Valve is normally closed with power removed. HLR will not be established through ND pumps. Check valve downstream of 1(2)Nl-1838 will provide additional isolation from RCS.
Failure of valve to close        Cold Leg Injection     None. Valve is normally closed with power removed.
Cold Leg Recirculation None. Valve is normally closed with power removed.
Hot Leg Recirculation  None. After HLR is aligned, 1(2)Nl-1838 is open to provide HLR with 1(2)Nl-173A and 1(2)NI-1788 close.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 24 Valve No. 1(2)Nl-173A Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   Valve is closed via control board switch. HLR is aligned via one train of ND. ND Pump is started to establish HLR. Check valve downstream of 1(2)Nl-173A will provide additional isolation from RCS.
Loss of A-Train AC Power        Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation  With no A-Train AC power, 1(2)ND-65B is closed to provide train separation. Valve is closed via control board switch. With no A-Train AC power, HLR will not be available through the ND pumps.
Check valve downstream of-1(2)Nl-173A will provide additional isolation from RCS.
Loss of 8-Train AC Power         Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   With no 8-Train AC power, 1(2)ND-32A is closed to provide train separation. Valve is closed via control board switch. With no 8-Train AC power, HLR will not be available through the ND pumps.
Check valve downstream of 1(2)Nl-173A will provide additional isolation from RCS.
Failure of Associated Pump to   Cold Leg Injection       None. Valve is normally open with start                                                    power  removed.
Cold Leg Recirculation   None. Valve is normally open with power  removed.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 25 Valve No. 1(2)Nl-173A Single Active Failure           ECCS Operation          Effects Phase Hot Leg Recirculation   Valve is closed via control board switch. HLR is aligned via one train of ND. ND Pump is started to establish HLR. Check valve downstream of 1(2)Nl-173A will provide additional isolation from RCS.
Failure of valve to open        Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   None. Valve is normally open with power removed.
Failure of valve to close       Cold Leg Injection       None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   HLR is aligned via B-Train of ND.
1(2)ND-32A is closed to provide train separation. ND Pump 8 is started to establish HLR. Check valve downstream of 1(2)Nl-173A will provide additional isolation from RCS.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 26 Valve No. 1(2)Nl-1788 Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   Valve is closed via control board switch. HLR is aligned via one train of ND. ND Pump is started to establish HLR. Check valve downstream of 1(2)Nl-178B will provide additional isolation from RCS.
Loss of A-Train AC Power        Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   With no A-Train AC power, 1(2)
ND-65B is closed to provide train separation. Valve is closed via control board switch. With no A-Train AC power, HLR will not be available through the ND pumps.
Check valve downstream of 1(2)Nl-178B will provide additional isolation from RCS.
Loss of B-Train AC Power         Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation   With no B-Train AC power, 1(2)ND-32A is closed to provide train separation. Valves is closed via control board switch. With no B-Train AC power, HLR will not be available through the ND pumps.
Check valve downstream of 1(2)Nl-178B will provide additional isolation from RCS.
Failure of Associated Pump to   Cold Leg Injection       None. Valve is normally open with start                                                    power  removed.
Cold Leg Recirculation   None. Valve is normally open with power  removed.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 27 Valve No. 1(2)Nl-1788 Single Active Failure           ECCS Operation         Effects Phase Hot Leg Recirculation  Valve is closed via control board switch. HLR is aligned via one train of ND. ND Pump is started to establish HLR. Check valve downstream of 1(2)Nl-178B will provide additional isolation from RCS.
Failure of valve to open        Cold Leg Injection     None. Valve is normally open with power    removed.
Cold Leg Recirculation None. Valve is normally open with power    removed.
Hot Leg Recirculation  None. Valve is normally open with power    removed.
Failure of valve to close       Cold Leg Injection      None. Valve is normally open with power removed.
Cold Leg Recirculation  None. Valve is normally open with power removed.
Hot Leg Recirculation  HLR is aligned via A-Train of ND.
1(2)ND-65B is closed to provide train separation. ND Pump A is started to establish HLR. Check valve downstream of 1(2)Nl-178B will provide additional isolation from RCS.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 28 Valve No. 1(2)NS-128, 1(2)NS-158 Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       None. Valve is opened via control room switch.
Loss of A-Train AC Power        Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       None. Valve is opened via control room switch.
Loss of B-Train AC Power        Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       Valve(s) will not open.
Failure of Associated Pump to   Cold Leg Injection      None. Valve is normally closed.
start                            Cold Leg Recirculation Check valve(s) downstream will Hot Leg Recirculation  provide additional isolation from containment.
Containment Spray      Containment Spray will not be available.
Failure of valve to open         Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation Hot Leg Recirculation Containment Spray       Containment Spray will not be available.
Failure of valve to close        Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation Hot LeQ Recirculation Containment Spray       Check valve(s) downstream will provide additional isolation from containment.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 29 Valve No. 1(2)NS-29A, 1(2)NS-32A Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       None. Valve is opened via control room switch.
Loss of A-Train AC Power        Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       Valve(s) will not open.
Loss of B-Train AC Power        Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation  Check valve(s) downstream will Hot Leg Recirculation   provide additional isolation from containment.
Containment Spray       None. Valve is normally closed.
Check valve(s) downstream will provide additional isolation from containment.
Failure of Associated Pump to    Cold Leg Injection     None. Valve is normally closed.
start                            Cold Leg Recirculation Check valve(s) downstream will Hot Leg Recirculation  provide additional isolation from containment.
Containment Spray      Containment Spray will not be available.
Failure of valve to open         Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation Hot Leg Recirculation Containment Spray      Containment Spray will not be available.
Failure of valve to close       Cold Leg Injection      None. Valve is normally closed.
Cold Leg Recirculation Hot Leg Recirculation Containment Spray       Check valve(s) downstream will provide additional isolation from containment.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 30 Valve No. 1(2)NS-43A Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  None. Valve is normally closed.
Check valve(s) downstream will provide additional isolation from containment.
Hot Leg Recirculation   None. Valve is normally closed.
Check valve(s) downstream will provide additional isolation from containment.
Loss of A-Train AC Power        Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  Auxiliary Containment Spray will not be available through A-Train ND Pump.
Hot Leg Recirculation  Auxiliary Containment Spray will not be available through A-Train ND Pump.
Loss of B-Train AC Power         Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  None. Auxiliary Containment Spray will be available through A-Train ND Pump.
Hot Leg Recirculation   None. Auxiliary Containment Spray will be available through A-Train ND Pump.
Failure of Associated Pump to    Cold Leg Injection      None. NS Spray is only used start                                                    when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  Auxiliary Containment Spray will not be available through A-Train ND Pump.
Hot Leg Recirculation   Auxiliary Containment Spray will not be available through A-Train ND Pump.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 31 Valve No. 1(2)NS-43A Single Active Failure           ECCS Operation         Effects Phase Failure of valve to open        Cold Leg Injection     None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation Auxiliary Containment Spray will not be available through A-Train ND Pump.
Hot Leg Recirculation  Auxiliary Containment Spray will not be available through A-Train ND Pump.
Failure of valve to close       Cold Leg Injection     None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation Check valve(s) downstream will provide additional isolation from containment.
Hot Leg Recirculation Check valve(s) downstream will provide additional isolation from containment.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 32 Valve No. 1(2)NS-388 Single Active Failure           ECCS Operation          Effects Phase None (base case)                Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  None. Valve is normally closed.
Check valve(s) downstream will provide additional isolation from containment.
Hot Leg Recirculation   None. Valve is normally closed.
Check valve(s) downstream will provide additional isolation from containment.
Loss of A-Train AC Power        Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  None. Auxiliary Containment Spray will be available through 8-Train ND Pump.
Hot Leg Recirculation   None. Auxiliary Containment Spray will be available through 8-Train ND Pump.
Loss of 8-Train AC Power         Cold Leg Injection      None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  Auxiliary Containment Spray will not be available through 8-Train ND Pump.
Hot Leg Recirculation   Auxiliary Containment Spray will not be available through 8-Train ND Pump.
Failure of Associated Pump to   Cold Leg Injection      None. NS Spray is only used start                                                    when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation  Auxiliary Containment Spray will not be available through 8-Train ND Pump.
Hot Leg Recirculation   Auxiliary Containment Spray will not be available through 8-Train ND Pump.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 33 Valve No. 1(2)NS-388 Single Active Failure           ECCS Operation         Effects Phase Failure of valve to open        Cold Leg Injection     None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation Auxiliary Containment Spray will not be available through B-Train ND Pump.
Hot Leg Recirculation  Auxiliary Containment Spray will not be available through B-Train ND Pump.
Failure of valve to close       Cold Leg Injection     None. NS Spray is only used when aligned to recirculation, at least 50 minutes following the reactor trip.
Cold Leg Recirculation Check valve(s) downstream will provide additional isolation from containment.
Hot Leg Recirculation Check valve(s) downstream will provide additional isolation from containment.

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 34
: 4. REGULATORY EVALUATION 4.1    Applicable Regulatory Requirements/Criteria The primary regulatory documents that are relevant when considering removal of NW from NI and NS CIVs and exempting the CIVs from LLRT are as follows:
* 10 CFR 50, Appendix A, General Design Criteria: GDC 55, Reactor coolant pressure boundary penetrating containment
* 10 CFR 50, Appendix A, General Design Criteria: GDC 56, Primary containment isolation
* 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors
* NUREG-0800, Standard Review Plan, Section 6.2.4, Containment Isolation Systems, Rev. 3 GDC 55 and 56 are defined as follows:
GDC 55, Reactor coolant pressure boundary penetrating containment:
Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
( 1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.
Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.
GDC 56, Primary containment isolation:
Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 35 can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
(1) One Jocked closed isolation valve inside and one Jocked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One Jocked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.
10 CFR 50, Appendix J, Option B introduction:
One of the conditions required of all operating licenses and combined licenses for Jight-water-cooled power reactors as specified in&sect; 50.54(0) is that primary reactor containments meet the leakage-rate test requirements in either Option A or B of this appendix. These test requirements ensure that (a) leakage through these containments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the technical specifications; and (b) integrity of the containment structure is maintained during its service life. Option B of this appendix identifies the performance-based requirements and criteria for preoperational and subsequent periodic leakage-rate testing.
CNS was approved for 10 CFR 50, Appendix J, Option B and Regulatory Guide 1.163 under Technical Specification Amendment Number 192/184 on July 31, 2001. Option B endorses Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program" as acceptable guidance for implementing Option B. Regulatory Guide 1.163 further endorses NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J." NEI 94-01 is currently Rev. 3-A incorporating regulatory positions stated in Regulatory Guide 1.163 (September 1995) as modified by NRC Safety Evaluations of June 25, 2008 and June 8, 2012. NEI 94-01, Rev. 3-A establishes criteria for when an LLRT is not required on a Containment Isolation Valve:
        "An LLRT is a test performed on Type B and Type C components. An LLRT is not required for the following cases:
* Primary containment boundaries that do not constitute potential primary containment atmospheric pathways during and following a Design Basis Accident (OBA);
* Boundaries sealed with a qualified seal system; or,
* Test connection vents and drains between primary containment isolation valves which are one inch or less in size, administratively secured closed and consist of a double barrier."

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 36 Under Option B, valves may be exempted from LLRT Type C testing if they are not a potential containment atmosphere leakage path. The valves do not constitute a potential primary containment atmospheric pathway during and following a OBA, as noted in UFSAR Table 6-74.
The exemption of the valves from the LLRT program is based on the guidance provided by 10 CFR 50, Appendix J, Option B, NEI 94-01, Rev. 3A and ANSI 56.8-2002.
The proposed change does not affect compliance with the noted regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met 4.2 Precedent CNS Unit 1 and Unit 2 are four loop Westinghouse design. McGuire Nuclear Station (MNS) Unit 1 and Unit 2 is also a four loop Westinghouse design. MNS does not have a NW system, and MNS does not perform LLRTs for the equivalent NI and NS CIVs. (Reference 6.40)
Comanche Peak is a four loop Westinghouse design, as are the Byron/Braidwood Stations, Diablo Canyon Station, and Wolf Creek/Callaway Stations. TXU Generation Company LP stated in their LAR (Reference 6.26) that they contacted a number of plants with system design essentially equivalent to Comanche Peak (Westinghouse 4 Loop plants), and these stations do not perform LLRTs for these Residual Heat Removal (RHR) and Containment Spray Valves. The technical justification for not performing local leak rate testing of these valves is described in their respective UFSARs. A similar discussion forms the basis for Catawba's request for elimination of NW with exemption from LLRT.
VC Summer Unit 1 is a three loop Westinghouse design. Farley Units 1 and 2 and Shearon Harris are other domestic plants which have the same SI System configuration as VC Summer. Both units were contacted by VC Summer, per their LAR (Reference 6.25), and these other plants confirmed that they do not perform LLRT on the High Head Safety Injection (HHSI) CIVs. This was a part of their original licensing basis and the valves are not in the LLRT program.
A recent request to exempt LLRT using a closed system outside containment was submitted by VC Summer on October 12, 2011 for HHSI CIVs (Reference 6.25) and Comanche Peak on March 18, 2003 for RHR and Containment Spray CIVs (Reference 6.26). Based on information submitted, the NRC granted the exception based on closed loop outside of containment and issued a SER on March 5, 2004 for Comanche Peak (Reference 6.31) and an SER on July 9, 2012 for VC Summer. (Reference 6.32) 4.3 No Significant Hazards Consideration Analysis Duke Energy requests an amendment to the Catawba Unit 1 and 2 Renewed Facility Operating Licenses to revise the UFSAR Section 6.2.4.2.2 to remove the Containment Valve Injection Water System (NW) Supply from specified Containment Isolation Valves (CIVs) and exempt these CIVs from Type-C Local Leak Rate Testing (LLRT).
Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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 consequences of an accident previously evaluated?

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 37 Response: No.
The amendment request is to remove select Containment Isolation Valves from the Local Leak Rate Test (LLRT) program. These valves were originally included in the LLRT under 10 CFR 50, Appendix J, in what is now Option A. CNS has been approved for 10 CFR 50, Appendix J, Option B under License Amendment No. 192/184. Under Option B, valves may be exempted from LLRT Type C testing if they are not a potential containment atmosphere leakage path. Based on the design and operation of the NI and NS Systems, the valves do not constitute a containment atmospheric leakage path as covered in the Safety Evaluation. Since the valves are not a leakage path, there is no impact on the consequence of an accident. Moreover, the valves are not a part of the Reactor Coolant Pressure Boundary, thus they do not affect the probability of an accident.
Therefore, the proposed change does not result in a significant increase in 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.
The systems design and operation are not changing. This test exemption does not change the way the valves are used as a part of the NI and NS Systems. A detailed Failure Modes and Effects Analysis was completed to confirm the system operation would meet the containment isolation design function.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The test exemption is within existing regulatory requirements. The application of a closed loop outside of containment is appropriate and consistent with regulatory positions. With containment integrity maintained within the allowable regulatory framework, there is no reduction in the margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
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,

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 38 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 A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, 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.
: 6. REFERENCES 6.1    10 CFR 50, Appendix A, General Design Criteria 55, Reactor Coolant Pressure Boundary Penetrating Containment 6.2    10 CFR 50, Appendix A, General Design Criteria 56, Primary Containment Isolation 6.3     10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors 6.4   NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part.
50, Appendix J, Rev. 3A 6.5   ANSI 56.8-1994, Containment System Leakage Testing Requirements 6.6   ANSI 56.8-2002, Containment System Leakage Testing Requirements 6.7   C. Patel (NRC) to G. Peterson (Duke Energy), Catawba Nuclear Station, Units 1 and 2 RE:
Issuance of Amendments (TAC NOS. MB1383 and BM1384) - 192/184 for implementation of 10 CFR Part 50, Appendix J, Option Band reference Regulatory Guide 1.163, July 31, 2001 (ADAMS Accession Number ML012290327) 6.8    CNS-1465.00-00-0005, Rev. 2, Plant Design Basis Specification for Design Basis Events 6.9    CNS-1562.Nl-00-0001, Rev. 50, Design Basis Specification for the Safety Injection (NI)
System 6.10   CNS-1563.NS-00-0001, Rev. 34, Containment Spray System (NS) Design Basis Specification 6.11   CNS-1569.NW-00.0001, Rev. 19, Containment Valve Injection Water System (NW) Design Basis Document 6.12   Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program 6.13   ANS-56.2/ANSI N271-1976, Containment Isolation Provisions for Fluid Systems 6.14   CN-1562-01.03, Rev. 19, Flow Diagram NI System 6.15   CN-2562-01.03, Rev. 26, Flow Diagram of Safety Injection System (NI) 6.16   CN-1563-01.00, Rev. 40, Flow Diagram NS System 6.17   CN-2563-01.00, Rev. 36, Flow Diagram of Containment Spray System (NS) 6.18   CN-1569-01.00, Rev. 27, Flow Diagram NW System 6.19   CN-2569-01.00, Rev. 28, Flow Diagram of Containment Valve Injection Water System (NW) 6.20    PT/1 (2)/A/4150/002 - Visual Inspection of Radioactive Systems Outside Containment 6.21    PT/1 (2)/A/4201/001 - Leak Rate Determination for the NI System 6.22    PT/1 (2)/A/4203/003 - Leak Rate Determination for NS System Outside of Containment 6.23    NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980 6.24    NUREG-0800, Standard Review Plan, Section 6.2.4, Containment Isolation Systems, Rev.
3

U.S. Nuclear Regulatory Commission CNS-18-008 Enclosure, Page 39 6.25 T. Gatlin (SCE&G) to R. Martin (NRC), Virgil C. Summer Nuclear Station (VCSNS) Unit 1, Docket NO 50-395, Operating License No. NPF-12, LAR 11-04821, Request for Prior NRC Approval to Exempt Five High-Head Safety Injection (HHSI) Containment Isolation Valves (CIVs) from Local Leak Rate Testing (LLRT) Requirements, October 12, 2011. (ADAMS Accession Number ML11286A318) 6.26 C. Terry (TXU Energy) to U.S. NRC, Comanche Peak Steam Electric Station (CPSES),
Docket NOs. 50-445 and 50-446, License Amendment Request (LAR) 02-10, Revision to Technical Specification (TS) 3.6.3, Containment Isolation Valves, March 18, 2003. (ADAMS Accession Number ML030860407) 6.27 CNS TS 6.28 CNS TS Basis 6.29 10 CFR 50, Appendix A, General Design Criterion 35, Emergency Core Cooling 6.30 Catawba FSAR Table 6-74, Potential Bypass Leakage Paths through Containment Isolation Valves 6.31 M. Thadani (NRC) to M. Blevins (TXU Energy), Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, - Issuance of Amendments RE: Revision to Technical Specification for Containment Isolation Valves (TAC NOS. MB8185 and MB8186), March 5, 2004.
(ADAMS Accession Number ML040690358) 6.32 R. Martin (NRC) to T. Gatlin (SCE&G), Virgil C. Summer Nuclear Station, Unit 1, Issuance of Amendment (TAC NO. ME7343), July 9, 2012.                 (ADAMS Accession Number ML12184A135) 6.33 10 CFR 50, Appendix A, General Design Criteria 54, Piping Systems Penetrating Containment 6.34 CNS-1465.00-00-0003, Plant Design Basis Specification for Containment Process Penetrations 6.35 Catawba UFSAR Table 6-77, Containment Isolation Valve and Actuator Data 6.36 Calculation CNC-1223.02-00-0016 Evaluation of Containment Isolation Overpressure Protection Features in Response to PIR 0-C90-0273 6.37 CNS-1465.00-00-0007, Plant Design Basis Specification for Seismic Design 6.38 CNS-1553.NC-00-0001, Design Basis Specification for the Reactor Coolant (NC) System 6.39 PT/1 (2)/A/4200/001 T, Containment Penetration Valve Injection Water System Performance Test 6.40 McGuire UFSAR Table 6-112, Containment Isolation Valve Test Data UFSAR Pages (Mark-up)

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 2 Catawba Nuclear Station                                                                            UFSAR Chapter 6 The supports and restraints are app lied such that integrity 1s assur,ed and pape stresses and support reactions are within allowable limits . Valves, an nonsafety-related systems where function pem1its, are nom1alffy positioned closed to i:nnnuze any release following a design basis event. Those valves th at are required to change posotion following a design basis event are equipped with valve operators to move the *,-*alve rapidly.
Containment isolation valves and operators inside the Containment are designed to withstand a maximum integrated rad iation dose of 2 x 108 rads during e life of the plant Containment isolauon va ves at are located inside the Contam ment are designed to function under e pressure-temperature conditions of IJo normai operation and that during the design basis event The pressure-temperature condition used for valve desagn under normal operation 1s 14-7 psja and 120&deg;F. The pressure-lempernturn condition used for val ve external design under accident conditions is 15 psig and 300"'F .
6.2.4.2.2             Conta inment Va lve Injection Water System The Contamment Valve Injection Water Sy*ste m (N\/V i is shown in Fig ure 6-11 6. H prevents leakage of containment atmosphere past certa n gate valves used for containm ent iso a -on fo Uowing a LOCA by 1njecung seal water at a pressure e;:ceedmg containment accident pressure between the two seating surfaces of the ne:e: wedge val\res The syste m consists of two independent, redundant trains; one supplying gate valves that are powered by the A train diesel and the other supplyiing gate val*i,*es powered by the 8 train diesel. This separa -on of trains prevents the possibinty of both containment iso1at on val.,.*es not sealing due to a single failure.
Each train cons1sts of a surge chamlJer which is filled with water and pressufiized with nitrogen .
One main header e:<its the chamber and splfits *nto several headers A solenoid valve is located in the main header before any of the branch headers which will open a er 60 second de1ay on ST signat Each of the headeirs supply injecti*on water to containment isolation valves locate-d in the same general locauon, and close on the same engineered safety signal. A so~e.Reiel valve i&
l0&a~0e in eaeR homier ~.1!:lisR suppl~e&* &NI w,der te *;awes elsa~F1t SA SP 5i~A-BL Tl1ese seleneiEI valves 9J)0R au.er a eQ ,soe&Re Elelo,i &A *s p ShjFlai. &iRee a ST oigAol eeettFS &efefe e SI? siQr:ial, tt:l e solenoitd ua,we locat.eg ~R-ll;ie JRai~ b:ueador w1U airoady ee ~RjoeU,;iB wat<er te
    ,er-:aa~RfllQRt ~&ela~ieR VQR,SQS &.EMliRS QR 8+ ShjR8L +l;is,~e&\'es 8fl epeA ~8Yt le tfle lteaders sttpplying iJ;jeeUeA &#xa5;::-ater 9A iP sisnaL. The delay for the soteno*ct valves opening is to allow adequate time for th*e slowest gate valv'e to close, before water is mjected into the valve sea .
lf11i1, 1iau0, ssleneie *11e l, 1es l!lFe fJF@\lidee fer4ft&Nl-aoo,~~8 &#xa3;i9f)tainmeRl is0l-aUeR *ta~ve& r0Go i>;11R~
68al water iR:joctiQR. f:ellowiAi a post1:1 latod asciGer;R one or R~o,:e ot the ~II gr r>JS coAtainiilent isslofheA &#xa5;alt,*eo ftl8)' ae al'eft_ +Ftese seloAeie , *al\les will Feooi*.;19 a signal *&e e~oR, f-Olk:lwHi!,i a ST Si&sect;:M I 8AS Mee i~a 8996ei0le8 eeAl-8iAM&Al-iN~9Q9FI , *a lvo l:las Gl868d_ If Vlo &9Rt3ifl~OAt iaeMlHeA *re 1,,,e .sttl3 9e fl 1:1eA~ly e13eA!i; ia esseeialee NW eo~eRoie \ltaP,e steises. Tt:i bis after a ST si!:)Rei, t:A:e NW soleAoia *;0l1;1e wnl &e e~eft-Yfhen-ifis &9AIDiRMeA~ 1selaaeR '+1alJ.,o is siesee aRa 1iee *,efJe.
One headeir for each train penetrates the Containment. Hie NVV Containment isolation valve on the outside of the Containment opens on ST signal, aHovting seal water to l)e injected to those con ainment isolation valves located inside the Contairnnent lnstde GonkiliH!nefl~ solefloia oatues isel~                          tRat stir,pi:, inieetioft m11ter-te ~eee .aluc3 elesiA@ ert SP 3fgAeL Hte seloReie vai,1oe ~R aAor a eQ eeeeAEl ae~ eR SP sigflel.
Makeup water is provided from the Demineralized Water Storage Tank for testing and adding water to the surge cham ber during normal plan
* operation _ Assu red water is provided from the (09 OCT 2016)                                                                                             6.2 -65

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 3 Catawba Nuclear Station                                                                                                                                UFSAR Table 6-77 (Page 8 of 53)
APPENDIX A FROM TABLE 6-77,
                                                                                                                                                                                                                 <PAGE 05 OF 23)
UNIT l CONTAINMENT ISOLATION VALVE DATA
                                                                                                                                                                                                       ~21 NOM,         m                                 101                                                     l4>   [0     141         VENl l     (19)
TS J.6.J          Cl)  LINE  Cl5l SEl~IC                     C2)      150            1191   CJ>
* l I   C4)  14)    FAIL   SIIUT  POST   FSAR   DRAJN    nPE ITEM          SEFIVICE     /JPP  PEN  Al'Pt.lCIIBLE CDC VALVE SIZE FLOW  EDUJP,      141'116)   ACTUATJON   TJIII(  LRT    VALVE  TYPE   SIZE. TYPE  ACT  ~       SIIFE  DOWN  ACCIO. FIG,  TYl'f. A     C   RELEASE  JNT
        "&deg;*
2J       Ill lESl LIi&#xa3; J
YES NO.
                                       "'322 CONDITION A,8 HD.
S6 ARR.
AIH JN.
J/4 DIR.
I 1/D ll/N VALVE NO.
INJ"95A SH,NAL T
CSEC>
1' IB TYP&#xa3; C
LDC, l
VALVE GT IN.
J/
* ACT.
[
TYP(
I\RH PCS.
C POS.
fAI PCS.
C PDS, C
NO.
6*129 T[ST YE~
TEST YES LOC.
AU   NONE INl-968          T    s Ill       C      0     r.l.   :11*    t   ARM   C:      FAI     C:    C.
INl*l2m           T    " IB        C      0     r.l    J/
* r   IIRH  C       Fnl     C     C    6-131)
INJ*Hl          NA        NA      C       I    CK     31*   NIA   Pl/A INI-JGl CVI        NA        HA             0 INl*362 <DI         NII       HA             0
                                                                                         ]N,~,~r;*M        N<I       NA             u IN1PG522B         NA        NA             0 24       ND CROSSOVER      ND  M2tl7      A,8      55    88    12   I    N/Y      lNl-1!>48         T    s: IB    Nor,e       I   Ol     3/4     E   ARM   C       FAJ     C     C    6-1:.l    ND       ND    NONE     39
* ISCHO   TO HOT LEOS                                                         JNl*l8:!0        NA       NA    ~         0     Cit      12    [    RM   C       FA!     C    C/'D INJ-125         NA       HA    POE        I    CK       8   NIA   H/A   C       N/A     C    C/0 1Nl* 129        NA       HA    NCH:       1    CK       8   NIA   N/A   C       k/A     C    C/0 1,11-216 CY>        NI\      NA              D 25         NI PUMP 8 JISCI-(; TO HOT LEGS ND  11320       NA      55    88
* I    YIY      1Nl*IS28
                                                                                          !NI*!~
NA 1
NII s IB IO.E PlClllE 0
l CiT Ol    31*
4     [
E RM ARM C
C FAI FA)
C C
C/0 C
6* 130    NO      NO     HONE     51 INJ-156         NA       NA    NCf,&#xa3;      I    CK       z   NIA  NIA    C      k/A      C    CIO LNl* 159       NA        NA    NOf&#xa3;       I    CIC      z   NIA  N/A    C      N/A      C    CID INl*Zll lV>        NA        NA              0 INl-400 CVI        NA        NA              1 JNl*-401 tVI       NA        NA              I INJ-*29 COi        NA        NA               l INl*4::III (DI     NA        NA              I INl--402 tYI       NA        NA               l INl*J55        NA        NA              I          I 1/2 INJ*ISB        NA        NA              I          l l/2 IM1q55ll,l INJIT54BB         ""'
NA
                                                                                                                    ~"
NA              I I             2 2
26         NI PUMI' A     ND   M317        NA      55   Q9     4    [    Y/Y     INl* lllA        NA        NA    Net.&#xa3;     D    CiT       4    r    RM    C      FAI     C     C/0   6-130    ND       NO     HONE    51 s 11, DISCHG TO f<JT LEGS                                                          IN!*l22B
                                                                                          !Nl-12-4 INl-12B I    T NA NA NA NA NOf&#xa3; NOIE NOIE J
l I
CiL CK CIC JI*
2 2
[
NIA NIA I\RH N/A NIA C
C C
FA!
IUA IVA C
C C
C CID C/0 INl-209 tVI        NA        NA              0
                                                                                          !Nl-123         NA        NA              l          I 1/2 INJ-IZ7         ,...      NA              I          I 1/"Z INIFE546B         NA        NA              l             2 lNIF&#xa3;547H         NA        NA              1             2 EFIH: CN111r,19, LIT I                                                                                                                                           I owo:    CN-FSAR-btl. 01-fi'lFi              I REV:   9

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 4 Catawba Nuclear Station                                                                                                                         UFSAR Table 6-77 (Page 9 of 53)
APPENDIX A FROM TABLE 6-77, (PAGE 06 OF 23)
UNIT I CONTAINMENT ISOLATION VALVE DATA 1!)21 N(JM,           (71                              (471                                                  U)    w     !* I        YENT l    118)
TS 3,6,J       Cl)  LINE   Cl5l SEISMIC                        C2l   150           (19)   CJJ         I *)   C-'l   14]  FAIL  SIIU1 POST    fSAR   DRAIN   T't'PE ITEM        SEFMCE     t,pp  F'EN APPL[CABLE CDC VALVE SIZE  FLOW   EOUIP.        1~71161    ACTUATION TIii:   lllT  YALYE TYPE   SIZE, TYPE   ACT   ~      SAFE  DOWN ACCJD. FIC. TYPE A     C     RELEASc'. JNT Ill),                   J   P<<J.
* CONOJTl N NO. ARR. IN. OIR. 110     YALYE NO.       SIGNAL   ISECI   TYPE   LOC, VAlYE    IN. ACT. TYP(    POS,  POS. POS,  POS,    NO. T(ST    TEST      1.oc.
27   M)  HK A DISCHG  ND  MJJ6      A;B    5'5   es    8      I     YIY       lNl--l7JA        NA      NA  r-,nr-     D    GT      8    E     RM      0    FA!     0     0   6* lJl    >Ill    ND      NONE     39 TO COLO L&#xa3;GS                                                            lid* 17,        NA      NA    NCM:      I   GI.. 31*    0     RM      C     C      C     C NA      NII  IO&#xa3;       l    (I(         NIA   NIA     C     NIA     0     0 INl-176 lNl* 176 INl-21 CD>*
NA NII NA NA NW 0
J    Cit     '
6  NIA    NIA      C    NIA    0      0 INl* l9 IYl         NA      NA              l INl-42\!l CYI       PjA     NA             l INl-421 CYI         NA      NA             I INl- 26 COi         NA      NA             I IN(-418 CYI         NA      NA             l INl*
* 78       NA      NA             l          J/4 INJFE59911         NA      NA             I            6 INJFE59BB          NA      NA             l            6 IIH- 99          NA     NA                  GI. 1/2 INl-&deg;"1!111 CV
* WA     NA                   Cl. 11:1 1Nl-51ll          NA     NA       C      I   CK     1/2   NA    ~     Cl
* NA   Cl
* Cl
* YES ltlH'iB2 IOI          NA     NA                  GI.. 1/2 INl-531 <VJ           NA     NA                  r,1 INJ-532          NA     NA       C      I   CK     112   NA    NA    CID     NA   C/0   C/0                    YES INl-53:l IOI          NA     NA                  GI..
28    M)  HX B OISCHO TO COLO LEGS NO   Mlll7    A,B     55   BS    8      I     'f/Y      lNl*l71l8 INl-179 1111-ll!!B INl-181 NA NII NA NA NA NII NA NA
                                                                                                                     -P-.o,E NJt<
tOE 0
I I
1 GT a.
Cl(
oc 8
3/4 G
6 E
D N~A NIA RM RM 11/A H/A 0
C C
C FA!
C NIA 0
C 0
0 0
C 0
0 6* 131   NO      HO      NONE     39
                                                                                        !NI* 16         NA      NA             I INl*2l'5 10)        NA      NA             0 INl-~ll         NII     NO             I          ,.,4 INffE69 llil       NA      NA             I            6 INIFE!SS2B         11111   NA             I            6 l"ll*'ls-4 IV)       NA      NA                  GL    1/2 JNl-195          NA      NA      C      I    CK    1/2   NA     ~     Cl
* NII  CID   CID                    YES INl-~96 c I
* NA      NA                  GI.. 1/2 l'-11*'53& (VJ       NA      NA                  r.l.
INJ-!131         NA      NA      C      I    CK    IIZ   NA     NA    r:tn    NA   r:tn  r:tn                   YES INl-!:iJB IOI        HA      NA                  Ct.
ERN: CN009 /Hil                                                                                                                                             owo: CN-FSAR-677. ~1-fl16                  I REV:     9

CW)
LC)
    ~
0 0
    ~
C1) 0)
C'O Q.
r--.
r--.I
    <O
    ~                     ~ i<.H.J lu u
    .0 C'O l-o::                                  ..,..,
    <(
LL
                  > C z.
C z;
C z
                             ~
I.,
Ill I:!
C                             ICE C
                                         "'c 0
    ...,    Zi."'                  1..
C'O    ._2 ~                  1::z:
I..
C'O C1)
LO a,  u C) co  :::,
a.. z C'O
    .0 3
    ...,C'O C'O u

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 6 Catawba Nuclear Station                                                                                                                                            UFSAR Table 6-77 (Page 11 of 53)
APPENDIX ti FROM TABLE 6-77,
                                                                                                                                                                                                                                   !PAGE 08 OF 23)
UNJT   1 CONTAINMENT ISOLATION VAL VE DATA 152)
NOM.             171                                   (471                                                         14)
* CI     14)        VENT I     1181 TS 3.6.3             (II  LIi    C1Sl SEIS"'IIC                         (2)      ISO          119l       131         w       f *)   w      FAIL   SHUT    POST  FSM    ORIII N  TYi'&#xa3; ITEM          SER\llt&#xa3;      />PP   PEN QPPl..11:"'au: G0C   V~ YE SIZE   FLOW   EOUIP.         ,.,, (61     QCluATJON   1111(    1.AT YAI. YE   TYPE  SIZE,  Tl'!'(  ACT   N~I.     $Qr[   DOWN ACCIO,   FIG. TYP( A      C     RE~(QSE   J NT J   N:),  CQ,IOITJ()>I        QAA,          DIR. J O     YAL'IE ,C.                           1YP&#xa3;   LOC,    VALVE    IN. ACT. 1YP&#xa3;    POS. POS. POS. POS. 110,   TEST    TEST       LOC.
NO,                                                  11(),        Ill.                                      SICNAL     ISEO 38       CONTAJM14Ud       110  M387      A, B       56    99     B        I   Y/Y          11115-129           p        NA            0       GT       8     E    ARM      C      FAI      C      0  6-IB9     NJ       NO       NONE     ~5 SPRAY Lllif                                                                        1115-ll           NA        l;A  ljON&#xa3;     I      oc      B   ljlQ    NIA      C      NIA      C      0 1115*1 8 7 IVl         NA        NA             I 11115*37 1\1)         NA        NA            0 INS*'.16 CPXI         NA        l;A           0                2 39    NO CCNT AINMENT       110 M:!69     A,8        56     99     B       I    YIY          INS*43A           NA        NA   _,_      0        GT      8     E      RM     C       FIii    C     C,Q  6-109     II()    NO       !ICON(    45 SP1'1AY Lltf: A                                                                        JNS- 46          NA        NA    NONE     I       CIC     8   NIA     NIA     C      NIA      C     C/0 1115* 112 (VI         NA        NA             I JP,5*71 IVl          NA        NII           0 1115*7 Ill           NA        NI\           0 INS*    **  IP XI     111A      NA            0                2 41l   Ill)  CCNTAJ NM[NT SPRAY I.I I'<&#xa3; B NO  M381      A,8        56     89     8       I    YIY          JNS-388 INS** I INS*lll lYI NA NA 111A NA NA l;A
                                                                                                                                       ~
0 I
I GT
(;K a
B E
NIA RM NIA C
C FAI Ill/A C
C t/0 C,Q 6-189     I,()    NO       NONE      45
                                                                                                   ]IQS*68 1\1)         NA        NA            0 INS
* 39 IPXI         NA        NA            0                2 1NS*7J m             NA        NA            0 41   REACT DR CXXIL ANl   YES   M348      A, 9       56     Al   314     J/0    NIN        IWL- 458 A            r   "Ill        C      I       GI. l/4      E    ARM      0      FM       0      C   11*11    YES     YES         AB   ~NE CRAIN TANI( GAS
* 1'1/L- !518           T   "Ill        C     0        G1. 3/4      E    ARM      0      FAI      0      C SP--CE TO W'li s"s
         *2    R&#xa3;1\CTOR CDOLAIIT   YES
* M3 5      A, B       56     119   3       0     N/1        IWL* BBSA            T    .: 18      NW      I      GT       3     E    ARM      0      FIil     D      C   1!*11        *
* NO 1 31 YES 1 2l      AB   NJNE DRAI N    T-   11X                                                                  1\IL* BB78           T    s 10       NW     0       GT       3     E    ARM      0      FAl      0      C OISCHC                                                                        IWL *tlaG          NA       NA       C      I      ex     112   NIA    NIA      C      NIA      C      C
         *J         YENl UNll      YES   Mlll      A, 8       56     A9    6       D     NI~       1WL*867A              *
(l IT   "1111      NW      I      GT      4      E    AR.. D      FAI      D      C   11-12  NO 1431 YES 142')     AB   l,J)Nf CONDENSATE ORA lN HEADER IWL*069B lliL*l!68 11* 1 r NA
                                                                                                                              ~ Iii NA w
C 0
I GT C(
                                                                                                                                                                *I     E NIA AA..
NIA D
C FAI NIA 0
C C
C 4(   CCIIIT FLCOl SU~     YES   M37 4     A, B       56     A9
* 0     NIN       III L-82f>A
* ll IT   s 10       NW     I       GT
* E    ARM      0      FAI      0      C   11-15
* NO 1 31 YES 1421      AB   Pllll'<E a 11\t.:ORE INST                                                                   1111.-8279
* 11 1 T  s 10       NW     0       GT      4      E    ARM      0      FA!      0      C Sl,NP PUMP                                                                       lltl.-321                                                J/4    NIQ    Ill/A           NIil     C DISC~                                                                     [IIL*A8 2 IV)
NA NA NA l;A C     I 0
i;c
                                                                                                                                                                                     '                      C
         *S          SC DRIIIN     YES   MJ59     1'1,9       56     Q9    3       D     NIN         lWL*A21            NA       NA      NW     I       GT      l    HW       M     LC      NIA      C      C   11* 17
* NO 1 31 YES 1421      AB   IOIE Pt.MP DISCH&                                                                       I WL -A22          NA       l;A      C     I       CK    3/4    NIA     NA      C      NIA      C      C LWL -A2
* NA       NA      NW     0       GT      l    H\I      M     LC      NIA      C      C I W\. *AZJ IVI         NA       NA             I
        *6        EDUI P 0&#xa3;CCN     YES   MJe6     A, B        56     A*     I      l     NIN          IIIE-2B           NA       NA       C     0       G1.      1   IN
                                                                                                                                                                     .,..,    M     LC      NI A     C      C   NONE     YES     YES         AB   NJNE LJIE 1131                                                                        III E* ZZ         NA       l;A      C     I       Gl       I            M     LC      NIA      C      C l WE*661V)           NA       NA             0 47   FL&#xa3;L TRANSFER        YES   CJB
* 48        5B     C2   21!   IOE      YIY       FLANCrn             NA        NA       B     I     OS/FC   NIA    NIA     NA      C      NIA      D      C   NONE     N)      110       NOi;&#xa3;   21 lUBE ERN: 1                                                                                                                                                                            I DWG: LN-~..., u M-h//. 0 1- Lll H             I REV:    ~

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 7 Catawba Nuclear Station                                                                                                                                           UFSAR Table 6-77 (Page 33 of 53)
APPENDIX A FROM TABLE 6-77.
IPAGE 06 OF 26l UNIT 2 CONTAJIIMENT ISOLATION VAl.YE DATA
                                                                                                                                                                                                                  ,.,          *~z, MJM,            171                                  (* 7)
* l I    (*)                VENT l    (181 T'5 J.6.J           II)  lit.I:  IH11 SEl!>'IIC:                     (l')      150            CIS>    (Jl          141   (4)       (*)      FAil   SHUT    PO<;"T  FSAR  tFIAIN  TYl'f ITEM     SEIMC[                  APP  PEN  APl'l..lCAIIL[ GOC  YALYE SIZE   FLOtf  EOUJP.
* 1 7) 161    ACT~TIO'I   TIME      UH    VAi.YE TYPE   SIZE, lVP[   ACT    IIOflJ,4111. S#"[   001111 ACCJD,   FJG. TYPE A      C    IIELE6SE JNT NO.                              J   NO. CtJNOIIIOII   I<>, ARR. IN. DIR. 1/0       v-.i.vt MJ. SICHN.. ISEC!    t'l'PE  LCX:. VN,,~    IN. a.ct. IYPE       PDS,      POS. PDS. 1'05,    NO. 1&#xa3;51   TESI       LCX:,
2
* NU   .tll..Jl',I-.Jllfo:.N DISClfJ TD HOT LECS ND  MZlll      A,11
                                                                   ""   118    12       I   tUY 2Nl*la:JS 2Nl*l25 2Nl-1'!1 11         I H"'
C IB hA NIOo tr..NE D
J I
ca.
OT c~
314 12 9
E.
E llRH RM C
C FIil FIii C
C     CID C            NO     MJ       NIJ'E  l9 2Nl*l211 N"
N,_        IIA  NCNE       I    c:r      B   "'"'
w .. w"""'..      C C
                                                                                                                                                                                                ,VA wa.
C C
CID 1:-/0 21<1-21G CYl        N"        M               0 2!l    NI PUMP B DISCt<l TO ltOT NO  "321!       ...      !!!I  88
* I    Y/Y        ;:Nl-1!12'0 2Nl* 153A NA T   " IB IIA   NCNE NCHE 0
l c;T CL     314 E
E RM AIIM C
C FAI FAI C
C C/0 C
NO      MJ      NIJ',E  ~1 LEUS                                                                                  2Nl-l!l6         NA        NA    ~* E       I    c:~      2  WA    WA           C:      NIA       C     i;..,a 21w1,,
2"1-211 IV!
2'i]- l!IB 11/l NII N..
N..
NA
                                                                                                                                   ,.A Nellf 0
l I
Cl(      2
                                                                                                                                                                     "'" ......         C       !II/A     C    CJD 2'NJ- IIII CVI      NA       ....             I l'IJ-
* Z, CDJ      NA        'IA              I
                                                                                                           * (OJ 2Nl* lli!l          N/1                       1 2N1-*112    CY)     NA       ""NA             1 2Nl-* 67 CYl                  NII             I ZNI**~ CVI N"'
N,_      ....             1 2NIFE6538          NA                        I              2 2NIFE!!>4B11       N..      ""NA              I             2 21<1-166                                   J 2NJ-l!M!I NII NA        ""'
HA               I I 112 I 1/2 2'     NI f'UMP A 015D<l TO NOT NO  14317      NA        !16   ee
* I    YIY        2Nl-121A 2Nl-12.18 NI\
T    ~
N"111  NCI*&#xa3; NCHE 0
1 CT GI.
J/*
E E
                                                                                                                                                                            ~
ARM C
C FA!
FAI C
C C/0 C
NO      ~       NC),&#xa3;  ~l LECS                                                                                 2"1-12*         NA        Hl't   -&#xa3;        J     CIC     2   NIA   WA          C        NIA      C     C/D "U-ll8           NA        HA     -t        l     Ct      l   NIA   WA          C        WI\      C     t/CI 2'111*2991Vl       NA                         0 2"1-117 NA        '"'
NA               I          I 1/2 2NlrE6 Gil                   NII              J             2 7Nrr&#xa3;5-178         NA        NA               I             2 2Nl* 123         ,.,..                      I          I 112 EFIN:CNIZl0S7R01                                                                                                                                                                      I D'w0: CN-FSAR-67 >'. 02 - 06                I   FIE"V; s

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 8 Catawba Nuclear Station                                                                                                                                           UFSAR Table 6-77 (Page 34 of 53) nPPENOIX A FROM TABLE 6-77.
                                                                                                                                                                                                                                  !PAGE 07 OF 26l UNIT 2 CONTAI~ENT ISOLATION 1/Al.'IE DATA 152>
rs J.6.J             IL>
l(]W, WE     1151 171 SEISMIC                      C2l 1471 ISO                 CJSO    (JI         141    ,.,     141 I.,
FAil m
SHUT l"I POSl  FSIIR V[Hl 1 t;llAIN 1181 TTPE:
ITEI<     SERYJCC         I\PP  PEN  Al'l'l..lCABLC CDC  YALYE  SIZE  FLOlil  EWJP.       (41)161   IICl\,WITIO'I 11~            LRT    VAi.YE TYPE  sm:. l'l'PE  OCT   '&deg;"10j_   S~C    OOl/11 ACCIO. FIG. TYPE A       C   llE:t.EME JNT PIO.                       J   NO. CONDlfl~      PO,   AAR. m. DIA. 1/D     VAi.VE Ill. SIi.NOi. ISECI         HP&#xa3;      LO:,  VO(.~   IN. AC1. IYI'(    PCS,    POS. PCS,    POS. NO. TES1    1ES1      LO:,
u    NU HX  ~ UJ::IILFTU  MJ  "".!b        A,~     !>!>   .. ~   e      I     Y/Y     2N1*113'>          NA NA NA IIA
                                                                                                                                      ,_..._      a
                                                                                                                                                         "'     9 JI' E    RM        0      FAI     0       D             NO      Ill
                                                                                                                                                                                                                                          """"    3!I TD CDI.D ~EOS                                                                    2Nl*l74                                 !Qj[       I     01.          D    RM        C        C     C       C 2"]-175           NII                   NONE       J     CK     r.                   C                      0
                                                                                                ~1*176            N,.        ""                          CK     6 NIA NIA NIii NIA NI" N/A 0
0       0 21111-21* IOI                   ""'"411    NCNE 0
I                                 C 2N]*4 IS "'1 2Nl*42\II CYl NA NII        ..,.
                                                                                                                             ""                    1 J
:INl*42J CYl         NA         NA                    J 2NJ*426 COi          NA         NII                   J 2Nl*4 IB           NA         NA                    J
                                                                                              ;mff[~!IIID                                                       g 2Nlf&#xa3;589e           """"
NA NA J
J            r.
2'1]*08 2Nl!!llll IOI       ~          "'"
NO I
GI.
314
:.,;111111
                                                                                              ;,11,~ (VI         ...
kA NA NA C       I     0(
GL in    NIA
                                                                                                                                                                          """      C/0      NIA   CIC     0/C lllf!IJI NI
                                                                                                 ;,111,;:i,      ""                       C Cl (l(  J/2   >IIA            CID      N/A   CIC     C/0 2Nf5J3 IOI NA NA NA NA I
ul.               '""                                                   YES ZII  NO HX B DISCHG      NO   H397        A,B      ~     u~      e     J      Y/Y     2N!-17BEI         NA         NA       .......,._   a     Cil     B    E     fr.-,    0       FAl    0       a              NII    M:I     ND-&#xa3;     3!I TD COLD LEGS                                                                     2NJ*l79          NA         1111       -E         J     OL   314     D     A,,1     C        C     C       C 2NJ-1ea          NA NA         ""         NCNE IQ[
J     CK CK i; NIA II/A NIA       C N/A     0       0 2"1*181 2Nl*416 t'Nl*215 C[JI N*
NII NA I
I 0
                                                                                                                                                                '         WA        C              0       C 2f,;J-459 CYl 2N[F!;5,1e NA        ....
HA I
I NA                                             6 2NlfE592~          Nii        1111                  J             6 2Nl*07           NA                               I          3/4 2Nl49'< a>>          NA       ""
NA                           u     112 2Nl*9!i 2Nl496 <VI 2Nl536 m NA NA NA NA C       I     C-*
Gt.
UL 112 Ill 112 NIA    NIA      CIO      NIA   CID     CIO                   YlS
                                                                                                 ..,,,!137 2Nl!IJB IOI NA NA        .......       C        I     C.c
[;!.
112 112 NIA    NIA      C/0     NIA    CID     CID                   YE!i E"t..:CN0097R 11                                                                                                                                                                 I D'WO: CN-FSAR-677. 02-07                    I "EV:     9

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 9 Catawba Nuclear Station                                                                                                                                         UFSAR Table 6-77 (Page 35 of 53) 11PP1.'0JX n FROM TA8LE 6-77.
                                                                                                                                                                                                                                    !PAGE 08 OF 261 UNIT 2 CONTAl~ENT ISOLATION V"'-.YE DATA
                                                                                                                                                                                   ~--      ,., ,., ,.,                   l!IZl
(* 71 TS J.~.J            Ill
                                                                       ..OM, LIi&#xa3;   Cl!.ll 171 SEISMIC                      (21     ISO
* HP      (J)             UI
* Cl     ,~,    FAil   SHUT     POST  FSAII VElll
* IJIAIN (181 TYPE coc                                                                                                                                    oo.....
ITEM       SERVICE       API'  PEl\l OPl'LICA!lt.E       VALVE SIZE  FLO\c    EOJJP. l 7l 161   llCTUIT[()j lli't:    LRT     ~bL'ff. TYPf    S[ZE,   lYPE     ilCT           SIIFE          ACCIO, FJG,  TYPE IO       C   IIELEIO!IE JNT NO.                       J   NO. CtlNOITI~     PCJ. ARR. IN. DIA.       110   VAl.'tt t,J,    Slt.NAI.. IS(CI     nPE      LDC,   VAi.Ye;    IN. IICf. TYPE     POS. PCS. POS. PCS. ND. TEST     TESf      LDC.
Zll  NJ~-*~ A. B        NO   "~        A .ll      !lb    Hb           I      YIY     ;ffl!-laZO          NA     NA    ~~           IJ      ul       *              ""'     u      H\I     C        u             NU       OIJ     ~        3'!f CISCH(i TC C:ll.C                                                                '111*11.'J         tiA    111A   NCNE         I      DI. JI
* D      fl!,!   C       t      C        C L[I:;!;                                                                    2Nl-t&5            NA     NA     NCI<[        I      CK       2     NIA     NIii     C     NIA      C       OIC 2"1*11>7           NA     NA     NCNE         I      CK       l     NIA    NI"       C     NIA      C       Dlt:
2NI-IG9           NII     111A   IOI[         I      CK       2     ,VI>,   NII\     C     ,VA     C        DIC 2Nl* 171          NA      111A   NCNE         I     0(        z     NIA    NIA       C     NIA     C        DIC 2N1*
* 95         NA      NA       C          I     CK         1    NIA    NII\      C     ,VA     C         C                     YES 2""1- 1* 1Y1         NA      NI>                 I 2N1** 8l cYl         NA      .....               I 2'Nl-* 171Yl         NA      NA                  I
                                                                                             ~1** 37 IY1 2'1111*213 CVJ N-NI\
NA                 0 I
21'11* 23
                                                                                                                  ,.,. NA                 I                 I 2NJ - 1'4                NA                  J             I 1/2 2Nl* t65          NA     111A                I             I 112 2Nl
* IGl!I       Nt\    NA                  I             I 112 Ziil* 1711        NA     NA                  I             1 1/2 2NlrE6* 911          NI\    'Ill                I 2NIFU,,H            tiA     NA                  I NII 2NffE65 lil 2NlFE~~2II          NA      ""
                                                                                                                         "A I
I
                                                                                                                                *l!EFE  ~  TO DP 2'/A/&ZI etv&
311    CCNTAI-NT        Ml   MJIIJ       48        G&    Cl     1B    0       YIY    2'Nl*IB!lA         NA      NA      NDIE        D      GT       11       &#xa3;     RH       C     FAI     C       CID            NO       NO      NJN!:    22 SUMP RECIRC UN&#xa3;                               68    Cl                           2111*5l5           NA     NA     N(J),f'      0      CL         l    MIA    l;/A      r. MIA     C         C                    \'[~       A8 A                                     6C    CT                           2Nt-51&            NA      NA    NOi&#xa3;         0      CL        I     MIA    NIA       C     MIi*    C         C                    ~~        OB
                                                            !;II  ti                           2Nl*517           NA      ljA     N()hf;     0       ~         I     NIA    ti/A      t     NIA     C         t                    Yl:S      All 68    Cl                           2N1*5l&           NA      NA      NCII,&#xa3;      0      Cl         l    MIii   NIA       C     NIii    C         C                    YES        AR
                                                            ~     Cl                           1Nl-!lt9          NA      NA     NONE         D      CL        I     ti/A  NIA        C     NIA     C         C                   ll:S       "8 6G    Cl                           2Nt-5211          NA      NA     NOi&#xa3;         D      Cl        I     NIA   N/11       C     NIA     C         C                    Y[S       "8 l)6    Cl                           lMl-l!ZI          NA      NA     !<ONE       0       CL        t    NIA    NIA        C     MIi>    C:        C                    Y&#xa3;S       A6 6G     Cl                           2Nl-522           NA      NA     NOi&#xa3;        D       GL        l    NIA    N/A       C      NIA     C         C                   >'ES       OB JI     CrNTAl""!NT SUMP RECIRt LIM!:
I\O  IQIIJ       411       N, 66 CI Cl lll   D       YI-(   2'11* ISALI 2Nt-513 NA NA NA NA "0""
tilll&#xa3; D
0
(;l CL Ill l
f NIA     ""'
NIA       C C     ~Al NIA C
C C/0 C
NO       NO
                                                                                                                                                                                                                                 \"ES NJN!'
A6 22 B                                    56     Cl                           ml*52*           NA       NA     NO>E        D       GL        I    NIA    till>      C     NIA     C         C                   'ftS       1113 6C 5i Cl
                                                                  ,I 2Nl--52'5 2'1[*526 NA
                                                                                                                 '!A NA     N0t&#xa3;
                                                                                                                                'iQI&#xa3; 0
0 CL Cl.
l t
NIA l.jl&
NII>
1;1.a     C C     NIA
                                                                                                                                                                                           >jlA C
c-C C
YES Vii llB All 6'     Cl                           2Nl*527           NA      NA     NCI&#xa3;        0       GL        I    NII>   NIA       C      NIA     C         C                   'ftS       AO
                                                           !Wi    t;J                          2Nl-,28           NA      NA     NONE        0       CL        I    NIA    NII>      C      NIA     C         C                   l'ES       All 61:    Cl                           2'11-52!1         NA      NA     NOi,&#xa3;       D       Cl        I    NIii   I'll-"    C      NIii    C         C                   YES        llB
:lo    Cl                           2NJ-,:,a         NA       NA     r.ONE       0       GL        I    NIA    NIA       C      NIA     C         C                   YES        OB EA~CN0097R31                                                                                                                                                                     I a.a: CN-FSAA-677. 02-08                       I PIEV:     s

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 10 Catawba Nuclear Station                                                                                                                                                      UFSAR Table 6-77 (Page 36 of 53)
APPENOJX A FROM TABL.E 6-77.
                                                                                                                                                                                                                                               !PAGE 09 OF 26>
UNIT 2 CONTAI~ENT ISOLATlON I/At.VE DATA IIOH,            171                                   ,n,                                                         ,.,       141   (*)    lfl l~Zl VDIT l   (101 T', J.6.J           Ill    LlloE  11::ll SEISMIC                       (Z)         150            1150      C;JI             I'!      w               FAil   SHUT   POST   F5AR  CAAIN  IYPf ITEM            SER\'lC&#xa3;      APP    PEN  .-PPL!Ctllk.&#xa3; GO::   YALV[   SIZE  FLO,,,   ED.JlP,      1* 7) 161   QCTUIHICJI     lit<      LRT  VIII.YE  TYPE      SIZE,  TYPE      oer  ..ORMIII.. SAFE   001/N Act! *, FIG. TYPE A      C   RELEASE    JNT NO.                            J    NO. CDNDIII~     tr.l.  "RR.      IN. DIR.      1/0     VAi,'([ M:I. SJGNjll,     ISECI     TYPE   LCC,   Ylll..'lf;  IN. ACT. TYPE     PIJS. PIJS. PDS. PIJS,   NO,  TEST   TEST      LCC, 3Z               SPARE      YES   Mll!l        ,a      !!,I)   t*      211  NIA       NIA         M]t,E           NA         IIA        a   N/A    Fl""J.E      29    Nie. NIA        C      NIA    NIA    NIA           NO      ND     ,ONE      27 33        o..NIT l ONI.YI 3
* 51"1\11&#xa3;    Y[!:  Mn*          *B      58      C*      21!  NIA       NIA         Ml!&#xa3;           NA          NA         9   NIA    ~UINr&#xa3;       2111  NIA      NIA        C      NIA    NIA    NIA           NO      ND     N'.INE    27 J6          CCIHAJNMfNT SPRAY LINE NO   l<l62       A,B      56     B9        8      I     Yh        2MS*32A 2NS*3J 2N!l* I IB <YI p
NA M
NA
                                                                                                                                             ----*      0 I
I OT Ck          8 8     E fol/A ARM NIA C
C FAI N/A C
C 0
D NO     N'.I    NIN::     *5 2NS*6!1 IYI         NA          ""'NA            0 2115*~ IP'll        IIA
                                                                                                                                       ""              0                   l J6          ~TAJNMftlT SPlti>Y LIN[
ND   ~7B        A,B       56     e, I
8     I      VIV       ZMS-~
2NS-l3 llj!j*UIS IYI p
N..
NA NA NA NII
                                                                                                                                             --MCfl[
D I
I or Cl 8
9   NIA E     "RM NIA C
C:
FAI NIA C
C 0
0            "&deg;     N'.I    Nc,,.E    *5 2NS*G2 IYI         NA           NA              0 2N5*6I IPll                      NA              D                   l
                     ~TAJNHENT              ~lie       A,B              es                     V/Y       i'NS*l:50          p          NA                      ul              ...,,..
J7 SPAAV LINE NO                        '$                8     I 2NS*l6 2115* UIS CYI NA NA Nfll NA NONE 0
I 1
C(
6 6
E      ARH NIA C
C FA[
C C
0 D            "&deg;     t.lJ    NCl"ot:
211S* B 1\11       NA           Hit             0 2115*4!1 IPXI       NA           HA              0                   z JO           *~*a    NM[ICT SPRAY lllE t<<l  H:187       ,. 8     56      BS       II    1       Y/Y       21'!5*126 ZNS-13 p
NA NA NA
                                                                                                                                             ~-
M:JNI:
0 I
GT Cl(
II 8
f NIA ARM NIA C
C r111 NIA r.
C 0
D NJ
                                                                                                                                                                                                                                             "&deg;      Nil'<F    *5 l'NS* 11J7 1\11     NA          NA                I NA                           g
                                                                                                       *Nli*J7 1n                     WI
         ~ ..    .,, CDNTA,-*Nl SPRAY LINC a I<!) M36!1       AR       !Iii    R!>      8     l       y;y 7115-36 IPXI
                                                                                                         "'5-43" 211$* 6 71IS*J l l 1\11 NA NA NA NA Nt, NA NA NA N'.IMC n
n I
I r:r D<
A 8
Hl,IO RM N.l'A C'.
C FAl NIA I'
C r.1n C/0
                                                                                                                                                                                                                                      ..,    ..,     u~c 2NS*71 IVI         NA                           0 2H5*74 ITI Zl-6*  **   (PXI NA NA NA NII 0                 2
        *B       toll CDNTAIUWFh~    ~                          56      89             I      Y/V        2',5-3116                    NII    1-Mw-     a       GT                       RM                                                           No.I'     *5 SPRAY LlN[ 9 M:181       A.B                       II 21&- l *
                                                                                                      ,N!;*lll lVI NA NA NA t()ll[   I I
D(
II 8
E NIA C
t FAl NIA C
C C/D CID            "&deg;     Nl 2t&*l:-I                                        0
                                                                                                      ?NS-39 (1'1(1 2NG-7J Ill IVI    ti'\
NA NA NA Nllo D
0 l
        *l     RUCTDR COOLANT        YES   H:}48      A.8       56      Al     31
* 110      N/111    2WL-
* l:JIM        T      ,,; IB        C      [      GL        31
* r     IIRM       0       fll]    0      C           1ES    YES        06     NONE
* RAIN lAI\K DAS                                                                       :lWL**!llR          T      s Ill         C      D      12.          I     E     IIRM       D       FAI     D      C 1il'ACE 10 WG sv, e:RtGCN0~197R51                                                                                                                                                                                 I O\oro; CN-FSAR-677. 02-09                 I REV.      9

Attachment 2 Technical Specification Bases Pages (Mark-up, For Information Only)

U.S. Nuclear Regulatory Commission CNS-18-008 , Page 2 CVIWS B 3.6.17 B 3.6 CONTAJNME          SYSTEMS B 3.6 17 Containment Va ve Injection Water System (CVrwS)
BASES BACKGROUND           Toe CVIWS is required by 10 CFR 50, Appendix A, GDC 54 , "Piping Systems Penetrating Containment" {Ref. 1), to ensure a water seal to a specific class of containment isolation valves {double disc gate vcrlves) during a LOG-A. to prevent leakage of containment atmosphere through the gate valves.
The CVIWS as designed to inject water between the two seating surfaces of double disc gate valves used for Contamment isolation. Toe injection pressure 1s higher than Contamment design peak pressure during a LOCA Thls will prevent reakage of the Containment atmosphere through the gate valves, therelY; reducing potential offsite dose below the va ues specified by 10 CFR 50 67 limits following the postulated accident During noffilal pmver operation , the system is in a standby mode and does not perform any function. During accident situations the CV1WS 1s activated to perform its safety related function, thus limiting the release of containment atmosphere past specific containment isolation valves, in order lo mitigate the consequences of a L0CA. Containment isolation vatves, for systems which are not used to mitigate the consequences of an accident, will be supphed with CVIWS seal water upon receipt of a Phase A isotation signaL COfttsri 1me1 1t ;.set:at* A '18l1t1es, fer 8eeiaeftt R'liti~aM9 sy&ie~s "i1Ri6h are &bl~liea ,... se~ wa&er ffGFR &Jle bHr,&fg, M 1,e their 3eft Y-18ler 3ttJ)p!Jes 8:.tl-Jatea e-y e Geft18inffleflt PFeSet:fFe I ligh i ti!jA si!jMI.
The system consists of two independent, redundant trains; one supplying I
gate valves that are powered by the A train diesel and the other supplying gate valves po,,vered by the B train diesel. This separation of trains prevents tile possibility of both contamment is~ ation valves not sealing due to a srngre failure.
Each train consists of a surge chamber which is filled with water and pressunz,ed with nitrogen. One main header exits the chamber and splits into several headers . A solenoid vatve is located in the main header before any of th e branch headers which will open after a 60 second delay on a Phase A isolation signal. Each of the headers supply injection water Gatawba Units 1 and 2                       B 3.6.17-1                             Revision No. 1

U.S . Nuclear Regulatory Commission CNS-18-008 , Page 3 CVIWS B 3.6 .17 BASES BAC GROUND (continued) to containment isolation valves located in the same general location, and close on the same eng*neered safety signal. ,~. seleflete *,!ftlve tS *eeetea iR ead:I hoaEler wRi&h sup~ieG seal water te vaJ1i1es slll&i~ BA a GeftteirtffiCAt PFeS!Sl:IFe J l~ft I l~A aigflaf. These selet1Bi8 ffl'aies 8f)eA aflar a eQ seeo,:ie eefay eR e GertteiFtlfleflt PFS3Stire Ui!lh I Jigh ~
Siftee e P~ase A iaaletten a~naf eecttf9 aefere e Gefttfliftffleftt ~ F e
                                 -      *    *     ~           -             4   *        -
* Makeup water is provided from the Oemineralized Water Storage Tank for testing and adding water to the surge chamber during normal plant OJ)eration . Assured water is provided from the essential header of the Nuclear Service Water System (NSWS). This supply is assured for at least 30 days foUowing a postulated accident. If the water level in the surge chamber drops below the low-low level or rf the surge chamber nitrogen pressure drops below the tow-row pressure after a Phase A isolation signal , a solenoid valve in the supply line from the NSWS will automatically open and remains open, assunng makeup to the CVIWS at a pressure greater than 110% of peal< Containment accident pressure.
Overpressure proteclion 1s provided to relieve the pressure buil-dup caused by the heatup of a trapped volume of incompressible fluid between two positively closing valves (due to contarnment temperature transient) back into contrunment where an open relief path exists.
APPLICABLE            The CVIWS design basis is established by the consequences of the SAFETY ANALYSES limiting OBA, which is a LOCA. The accident analysis {Ref. 2) assumes lhat only one train of the CVIWS is functional due to a single failure that cfisables the other train. Makeup water can be assured from the NSWS for 30 days following a postulated LOCA.
The CV1WS satisfies Criterion 3 of 10 CFR 50 36 (Ref. 3).
GataW'ba Unrts 1 and 2                    B 3.6.17-2                            Revision No. O}}