ML061730386

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Relief, Relief Request No. ISPT-09 for the First Ten-Year Inservice Inspection Interval
ML061730386
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 07/11/2006
From: Marshall M
NRC/NRR/ADRO/DORL/LPLD
To: Singer K
Tennessee Valley Authority
Marshall M. NRR/DLPM 415-2734
References
ISPT-09, TAC MC8305
Download: ML061730386 (15)
v • d • e


Text

July 11, 2006 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 RELIEF REQUEST NO. ISPT-09 FOR THE FIRST TEN-YEAR INSERVICE INSPECTION INTERVAL (TAC NO. MC8305)

Dear Mr. Singer:

By letter dated September 8, 2005, Tennessee Valley Authority (TVA), the licensee for the Watts Bar Nuclear Plant, Unit 1 (WBN), submitted Relief Request (RR) No. ISPT-09. In this submittal, TVA requested relief from the requirements of Paragraph (g)(4), Section 50.55a of Title 10, Code of Federal Regulations, and the system leakage test requirements that are specified in Table IWB-2500-1, Examination Category B-P, and Paragraph IWA-5222(a) of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). RR No. ISPT-09 is applicable for the First Ten-Year Inservice Inspection Interval for WBN, and pertains to the system leakage test requirements for ASME Code Class 1 portions of the safety injection and residual heat removal systems.

The Nuclear Regulatory Commission (NRC) staff has completed its review of RR No. ISPT-09, and has determined that the licensees request for alternative pressure tests is acceptable for implementation under the hardship provisions of 10 CFR 50.55a(a)(3)(ii). The NRC staffs safety evaluation is enclosed.

Sincerely,

/RA by L. Raghavan for/

Michael L. Marshall, Jr., Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

Safety Evaluation cc w/enclosure: See next page

ML061730386

  • No Legal Objection NRR-028 OFFICE LPL2-2/PM LPL2-2/LA CFEB/BC OGC LPL2-2/BC NAME DPickett RSola S. Sheng for SHamrick
  • L. Raghavan for K. Gruss (via memo) LMMarshall DATE 6/27/06 6/26/06 5/15/06 7/6/06 7/11/06 Mr. Karl W. Singer Tennessee Valley Authority WATTS BAR NUCLEAR PLANT cc:

Mr. Ashok S. Bhatnagar, Senior Vice President Mr. Paul L. Pace, Manager Nuclear Operations Licensing and Industry Affairs Tennessee Valley Authority Watts Bar Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Spring City, TN 37381 Mr. Larry S. Bryant, Vice President Mr. Jay Laughlin, Plant Manager Nuclear Engineering & Technical Services Watts Bar Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place P.O. Box 2000 1101 Market Street Spring City, TN 37381 Chattanooga, TN 37402-2801 Senior Resident Inspector Mr. Robert J. Beecken, Vice President Watts Bar Nuclear Plant Nuclear Support U.S. Nuclear Regulatory Commission Tennessee Valley Authority 1260 Nuclear Plant Road 6A Lookout Place Spring City, TN 37381 1101 Market Street Chattanooga, TN 37402-2801 County Executive 375 Church Street Mr. Michael D. Skaggs Suite 215 Site Vice President Dayton, TN 37321 Watts Bar Nuclear Plant Tennessee Valley Authority County Mayor P.O. Box 2000 P. O. Box 156 Spring City, TN 37381 Decatur, TN 37322 General Counsel Mr. Lawrence E. Nanney, Director Tennessee Valley Authority Division of Radiological Health ET 11A Dept. of Environment & Conservation 400 West Summit Hill Drive Third Floor, L and C Annex Knoxville, TN 37902 401 Church Street Nashville, TN 37243-1532 Mr. John C. Fornicola, Manager Nuclear Assurance and Licensing Ms. Ann P. Harris Tennessee Valley Authority 341 Swing Loop Road 6A Lookout Place Rockwood, Tennessee 37854 1101 Market Street Chattanooga, TN 37402-2801 Mr. Glenn W. Morris, Manager Corporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM TENNESSEE VALLEY AUTHORITY REQUEST FOR RELIEF NO. ISPT-09 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO: 50-390

1.0 INTRODUCTION

By letter dated September 8, 2005 (Reference 1), Tennessee Valley Authority (TVA), the licensee for the Watts Bar Nuclear Plant, Unit 1 (WBN), submitted Relief Request (RR)

No. ISPT-09. In this submittal, TVA requested relief from the requirements of Paragraph (g)(4),

Section 50.55a of Title 10, Code of Federal Regulations (10 CFR), and the system leakage test requirements that are specified in Table IWB-2500-1, Examination Category B-P, and Paragraph IWA-5222(a) of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). RR No. ISPT-09 is applicable to the First Ten-Year Inservice Inspection Interval (1st 10-Year ISI Interval) for WBN, and pertains to the system leakage test requirements for ASME Code Class 1 portions of the safety injection (SI), residual heat removal (RHR), and auxiliary spray (AUX-S) systems.

The licensee responded to the staffs Request for Additional Information (RAI) in a letter to the Nuclear Regulatory Commission (the Commission, NRC) dated March 31, 2006 (Reference 2).

Henceforth, all RAI responses referenced in this evaluation refer to this RAI response letter.

2.0 REGULATORY EVALUATION

Inservice Inspection (ISI) of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulation in 10 CFR 50.55a(a)(3) states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Enclosure

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions, and the pre-service examination requirements, set forth in Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals must comply with the requirements in the latest edition and addenda of Section XI incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, as subject to the limitations in the rule.

The ISI Code of Record for the 1st 10-year inspection interval for WBN is the 1989 Edition of the Section XI, no Addenda. Pursuant to 10 CFR 50.55a(g)(4)(iv), inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (b) of this section, subject to the limitations and modifications listed in paragraph (b) of this section, and subject to Commission approval. Portions of editions and addenda may be used provided that all related requirements of the respective editions or addenda are met.

3.0 TECHNICAL EVALUATION

3.1 Applicable Components The components identified in the attached table Component Descriptions for Relief Request No. ISPT-09, are based on the amended component descriptions in Reference 2. The component descriptions identify which valves define the boundaries of the applicable line segments. The table also provides the test pressures for performing the licensees alternative system leakage tests of the applicable SI and RHR lines segments.

The applicable SI, RHR, and AUX-S line segments are categorized as ASME Code Class 1 because the line segments are aligned to the reactor coolant pressure boundary (RCPB) out to and inclusive of the second containment isolation valve (CIV) for the systems. With the exception of CIVs in one of the applicable RHR line segments and one of the CIVs in the applicable AUX-S segment, each of the line segments is isolated from the RCPB by self-actuating check valves (CVs).

For one of the RHR piping segments and for one of the CIVs in the applicable AUX-S line segment, the CIVs are motor-operated flow control valves (FCVs). The FCVs are interlocked to ensure redundant isolation of the RCPB from the ASME Code Class 2 portion of the RHR which is subject to a lower design pressure (600 pounds per square inch gauge (psig)). The Technical Requirements Manual for WBN specifies that TVA should maintain the FCVs in the closed and de-energized configuration prior to raising the reactor coolant system (RCS) pressure above 425 psig. Plant operating instructions specify that the FCVs are to be maintained in the closed configuration prior to raising the RCS pressure above 370 psig or the RCS temperature above 350 degrees Fahrenheit (EF).

3.2 Applicable Code Edition and Requirements TVA requested relief from the following Section XI requirements, as invoked by reference in 10 CFR 50.55a(g)(4):

  • Section XI, Table IWB-2500-1, Examination Category B-P, All Pressure Retaining Components, Inspection Item B15.51, Piping, and Inspection Item B15.71, Valves:

Inspection Items B15.51 and B15.71 require that the licensee perform inservice hydrostatic pressure tests and VT-2 visual examinations of all ASME Code Class 1 pressure retaining piping and valves in accordance with the requirements of Section XI, Paragraph IWB-5222. The inspection items require that the inservice hydrostatic pressure test be performed at a frequency of once during every 10-Year ISI interval.

The nominal operating temperature of the RCPB at 100 percent rated power is above 500 EF. The nominal operating pressure of the RCPB at 100 percent rated power is 2235 psig. Thus, Table IWB-5222-1 requires that TVA perform the required hydrostatic pressure test at a test pressure of 2280 psig (i.e., 1.02

  • 2235 psig).
  • In response to RAI No. 1, TVA clarified that the 1989 Edition of Section XI is the Code of Record for WBN. TVA also clarified that Inspection Items B15.51 and B15.71 are the inspection items in Section XI, Table IWB-2500-1, Examination Category B-P that are applicable to RR No. ISPT-09.

3.3 Basis for the Relief Request The applicable RHR, SI, and AUX-S line segments are each designed with two system CIVs in order to comply with the requirements of 10 CFR Part 50, Appendix A, General Design Criteria, Criterion 55, Reactor Coolant Pressure Boundary Penetrating Containment. The RCPB extends to, and is inclusive of, the second CIV in the piping segments.

TVA is requesting relief from compliance with the requirements that are specified in:

(1)Section XI, Table IWB-2500-1, Examination Category B-P, Inspection Items B15.51 and B15.71; (2)Section XI, Paragraph IWB-5222(a); and (3)Section XI, Table IWB-5222-1.

TVA stated that the relief is being requested for the 1st 10-Year ISI Interval for WBN.

TVA stated that implementation of the required hydrostatic pressure tests on the applicable RHR, SI, and AUX-S line segments would require either: (1) opening up the first CIV in the line segments, (2) installing temporary bypasses around the first CIV, or (3) modifying the systems to bypass the first CIV. TVA stated that implementing any of these actions would require it to defeat the double containment isolation function of the inboard and outboard CIVs, which are required to be configured in the closed position when in operating MODES 1, 2, or 3.

TVA stated that this constitutes a hardship for the facility as defined in 10 CFR 50.55a(a)(3)(ii).

3.4 Proposed Alternative In lieu of compliance with the applicable hydrostatic pressure test requirements, TVA proposed to perform alternative system leakage tests of the applicable line segments at reduced test pressures.

  • The line segment from the high and intermediate pressure portions of the SI system and the SI accumulators will be pressurized using the SI pumps to approximately 1500 psig, which is the pressure of the SI pumps while running in recirculation flow mode.
  • The lines segments in the flow path from the cold leg accumulator outlet isolation valve to the RCPB cold leg piping will be pressurized to the 610!660 psig range using the accumulators, as required by plant Technical Specifications.
  • The lines segments in the RHR system will be pressurized to approximately 350 psig and visually examined while the RHR is providing shutdown cooling during startup following the refueling outage.
  • The AUX-S line segment will be examined at the pressure existing between the isolation valve 1-FCV-62-84 and 1-CKV-62-661. TVA stated that, although there are no test connections within this line segment that would permit direct measurement of the pressure between the valves, the line would be subject to a minimum test pressure of 325 psig, which would be the pressure when the AUX-S system is actuated during plant startup, prior to initiating a reactor coolant pump (RCP). The AUX-S system would be isolated from the RCPB following initiation of a RCP.

TVA stated that the alternative system leakage tests would be subject to the minimum hold time that is required in accordance with Section XI, Paragraph IWA-5213(d) and to the minimum test temperatures that are required in accordance with Section XI, Paragraph IWB-5230. These criteria for minimum hold times and minimum test temperatures are acceptable because they will be implemented in compliance with Section XI, as invoked by 10 CFR 50.55a.

3.5 Evaluation The licensees proposed alternative system leakage tests are intended to test the subject piping segments at the corresponding system test pressures and temperatures when the systems are configured for containment isolation in MODE 3, or in MODE 2, prior to initiation of a RCP.

TVAs proposed alternative system leakage tests are based on the leak rate (LP) that could be expected to result if the pressure test were performed at the lower system leakage test pressures (PP in psig). TVA used the following equation to estimate the leak rates that would result from a postulated through-wall crack in the line segments, as exposed to the lower system leakage test pressures:

Lp = LXI x ( Pp / PXI )1/2 Equation 1.

In Equation 1, LP represents the leak rate that is anticipated from a through-wall crack at the lower test pressure; LXI represents leak rate that would be expected to occur if the test were performed at (i.e., normalized to) the required hydrostatic test pressure; Pp is the alternative test

pressure (in psig); and PXI is the required hydrostatic test pressure of 2280 psig. TVA stated that Equation 1 is based on ASME Operation and Maintenance (OM) Standard 10, paragraph 4.2.2.3(b)(4). This is acceptable because Standard 10 of the ASME OM Code is the applicable OM Code that has been endorsed when applying the 1989 Edition of the ASME Code.

Using Equation (1), TVA estimated that, if the applicable line segments were to contain through-wall flaws, the alternative system leakage tests would result in the following leak rates:

  • SI accumulator line segments (PP = 610 ! 660 psig): LP will be 52 - 54 percent of LXI
  • SI line segments (PP = 1500 psig): LP will be 81 percent LXI
  • RHR line segments (PP = 350 psig): LP will be 39 percent of LXI
  • AUX-S line segment (minimum PP $ 325 psig): LP will be equal to or greater than 38 percent of LXI These leak rates are high enough to be detected by the licensees VT-2 examinations. Any evidence of pressure boundary leakage that is detected during implementation of the alternative system leakage tests will require the licensee to enter the degraded pressure boundary components into the Corrective Actions provisions of Section XI, Paragraph IWA-5250.

Pressurization of the subject piping segments in accordance with applicable hydrostatic test requirements would require the licensee to defeat the double containment isolation design requirement or implement significant plant modifications. Either approach would subject the licensee to an undue burden with no compensating increase in quality or safety. The licensees basis demonstrates that the alternative system leakage tests are capable of resulting in detectable leak rates from the applicable line segments. Therefore, the staff concludes that the licensees proposed alternative system leakage tests will provide sufficient monitoring of the structural integrity of the subject components, in lieu of the required hydrostatic pressure tests.

Furthermore, the licensee will perform additional examinations of the RCPB in accordance with other Examination Categories in Section XI, Table IWB-2500-1, and also perform augmented system walkdowns of the RCPB for evidence of boric acid leakage as part of its Generic Letter 88-01 inspections. These additional examinations provide additional monitoring of the structural integrity of the RCPB.

Based on this assessment, the staff concludes that compliance with the hydrostatic pressure test requirements during the 1st 10-Year ISI Interval will create a hardship for the licensee. The staff also concludes that TVA has proposed acceptable alternative system leakage tests for these line segments.

4.0 CONCLUSION

The staff has reviewed the request for relief in RR No. ISPT-09 along with the proposed alternative system leakage tests for the ASME Code Class 1 SI, RHR, and AUX-S line segments against the hardship provisions of 10 CFR 50.55a(a)(3)(ii). Based on its evaluation, the staff concludes that: (1) TVA provided a sufficient basis to demonstrate that compliance with the requirements of Section XI, Table IWB-2500-1, Examination Category B-P, Inspection Items B15.51 and B15.71, and the requirements of Section XI, Paragraph IWB-5222 and Section XI, Table IWB-5222-1 (as applied to these line segments) would create a hardship or unusual difficulty for WBN without a compensating increase in the level of quality and safety

and, (2) TVAs proposed alternative system leakage tests for these line segments are acceptable.

Therefore, based on the staffs assessment, the staff concludes that RR No. ISPT-09 may be granted for the 1st 10-Year ISI Interval pursuant to the hardship provisions of 10 CFR 50.55a(a)(3)(ii).

5.0 REFERENCES

1. Serial Letter from P. L. Pace (TVA) to the Nuclear Regulatory Commission Document Control Desk, Watts Bar Nuclear Plant (WBN), Unit 1 - American Society of Mechanical Engineers (ASME),Section XI Inservice Pressure Testing Program Request for Relief ISPT-09, dated September 8, 2005.
2. Serial Letter from P. L. Pace (TVA) to the Nuclear Regulatory Commission Document Control Desk, Watts Bar Nuclear Plant (WBN), Unit 1 - American Society of Mechanical Engineers (ASME),Section XI Inservice Pressure Testing Program Request for Relief ISPT-09, dated March 31, 2006.

Principal Contributor: James Medoff Date: July 11, 2006

Attachment:

As stated

Component Descriptions for Relief Request No. ISPT-09 Component Description Pipe Line Pipe Alternate Diameter Segment Design Test (Inches) Length Pressure Pressure (Inches) (psig) (psig)

Safety Injection Accumulator No. 1 Piping Segments from the Outlet Isolation Valve to the Reactor Coolant System, Including the Branch Connections from the RHR and SIS Safety Injection Accumulator No. 1 Outlet 10 23 2485 610 - 660 Isolation Valve to Outlet Check Valve (FCV-63-118 to CKV-63-622)

Safety Injection Accumulator No. 1 Outlet 10 18 2485 1500 Check Valve to Loop 1 Cold Leg (CKV-63-622 to CKV-63-560) 6-Inch Branch Connection from the 10-Inch 6 23 2485 1500 Safety Injection Accumulator No. 1 Outlet Piping to the Low Pressure Safety Injection (RHR) Check Valve CKV-63-633 2-Inch Branch Connection from the 6-Inch 2 10 2485 1500 RHR System Branch to the Safety Injection System Check Valve CKV-63-551 Safety Injection Accumulator No. 2 Piping Segments from the Outlet Isolation Valve to the Reactor Coolant System, Including the Branch Connections from the RHR and SIS Safety Injection Accumulator No. 2 Outlet 10 16 2485 610 - 660 Isolation Valve to Outlet Check Valve (FCV-63-98 to CKV-63-623)

Safety Injection Accumulator No. 2 Outlet 10 15 2485 1500 Check Valve to Loop 2 Cold Leg (CKV-63-623 to CKV-63-561) 6-Inch Branch Connection from the 10-Inch 6 12 2485 1500 Safety Injection Accumulator No. 2 Outlet Piping to the Low Pressure Safety Injection (RHR) Check Valve CKV-63-632 Attachment

Component Description Pipe Line Pipe Alternate Diameter Segment Design Test (Inches) Length Pressure Pressure (Inches) (psig) (psig)

Safety Injection Accumulator No. 2 Piping Segments (Continued) 2-Inch Branch Connection from the 6-Inch 2 21 2485 1500 RHR System Branch to the Safety Injection System Check Valve CKV-63-553 Safety Injection Accumulator No. 3 Piping Segments from the Outlet Isolation Valve to the Reactor Coolant System, Including the Branch Connections from the RHR and SIS Safety Injection Accumulator No. 3 Outlet 10 9 2485 610 - 660 Isolation Valve to Outlet Check Valve (FCV-63-80 to CKV-63-624)

Safety Injection Accumulator No. 3 Outlet 10 17 2485 1500 Check Valve to Loop 3 Cold Leg (CKV-63-624 to CKV-63-562) 6-Inch Branch Connection from the 10-Inch 6 17 2485 1500 Safety Injection Accumulator No. 3 Outlet Piping to the Low Pressure Safety Injection (RHR) Check Valve CKV-63-634 2-Inch Branch Connection from the 6-Inch 2 20 2485 1500 RHR System Branch to the Safety Injection System Check Valve CKV-63-555 Safety Injection Accumulator No. 4 Piping Segments from the Outlet Isolation Valve to the Reactor Coolant System, Including the Branch Connections from the RHR and SIS Safety Injection Accumulator No. 4 Outlet 10 22 2485 610 - 660 Isolation Valve to Outlet Check Valve (FCV-63-67 to CKV-63-625)

Safety Injection Accumulator No. 4 Outlet 10 24 2485 1500 Check Valve to Loop 4 Cold Leg (CKV-63-625 to CKV-63-563) 6-Inch Branch Connection from the 20-Inch 6 21 2485 1500 Safety Injection Accumulator No. 4 Outlet Piping to the Low Pressure Safety Injection (RHR) Check Valve CKV-63-635 2-Inch Branch Connection from the 6-Inch 2 7 2485 1500 RHR System Branch to the Safety Injection System Check Valve CKV-63-557

Component Description Pipe Line Pipe Alternate Diameter Segment Design Test (Inches) Length Pressure Pressure (Inches) (psig) (psig)

High Pressure [BIT Injection] Safety Injection System Piping Segments from Check Valve CKV-63-581 to the Reactor Coolant System High Pressure Safety Injection Piping from 3 20 2485 1500 CKV-63-581 to Loop 3 Cold Leg Injection Check Valve CKV-63-588 (includes three row 2.5 83 2485 1500 descriptions to account for the reductions in the diameter of the piping segment) 1.5 39 2485 1500 1.5-Inch Branch Connection from the 3-Inch 1.5 123 2485 1500 Common Header coming from CKV-63-581 to Loop 1 Cold Leg Injection Check Valve CKV-586 1.5-Inch Branch Connection from the 3-Inch 1.5 25 2485 1500 Common Header coming from CKV-63-581 to Loop 4 Cold Leg Injection Check Valve CKV-589 1.5-Inch Branch Connection from the 2.5- 1.5 107 2485 1500 Inch Common Header coming from CKV 581 to Loop 2 Cold Leg Injection Check Valve CKV-587 RHR Hot Leg Injection Piping Segments Aligned to Loop 1: From Check Valve CKV-63-640 to the Reactor Coolant System (Includes the Branch Connection from the Safety Injection System)

Low Pressure Safety Injection Piping from 8 29 2485 1500 RHR Check Valve CKV-63-640 to the Loop 1 Hot Leg Check Valve CKV-63-641 (includes two row descriptions to account for the 6 2 2485 1500 reduction in the diameter of the piping segment) 2-Inch Branch Connection from the 8-Inch 2 5 2485 1500 RHR Piping to Safety Injection System Check Valve CKV-63-543 RHR Hot Leg Injection Piping Segments Aligned to Loop 3: From Check Valve CKV-63-643 to the Reactor Coolant System (Includes the Branch Connection from the Safety Injection System)

Low Pressure Safety Injection Piping from 8 42 2485 1500 RHR Check Valve CKV-63-643 to the Loop 3 Hot Leg Check Valve CKV-63-644

2-Inch Branch Connection from the 8-Inch 2 7 2485 1500 RHR Piping to Safety Injection System Check Valve CKV-63-545 Component Description Pipe Line Pipe Alternate Diameter Segment Design Test (Inches) Length Pressure Pressure (Inches) (psig) (psig)

SI Hot Leg Injection Piping Segment from Check Valve CKV-63-647 to the Reactor Coolant System Safety Injection Pump Piping from Check 2 60 2485 1500 Valve CKV-63-547 to Check Valve CKV-63-559 (includes two row descriptions to account for the expansion of the diameter 6 0.5 2485 1500 of the piping segment)

SI Hot Leg Injection Piping Segment from Check Valve CKV-63-647 to the Reactor Coolant System Safety Injection Pump Piping from Check 2 44 2485 1500 Valve CKV-63-549 to Check Valve CKV-63-558 (includes two row descriptions to account for the expansion of the diameter 6 0.5 2485 1500 of the piping segment)

RHR Loop 4 Suction Piping Segment from Flow Control Valve FCV-74-2 and its Bypass Valve FCV-74-8 to the Reactor Coolant System RHR Piping Between Flow Control Valves 14 50 2485 1500 FCV-74-1 and FCV-74-2, including the branch connection in the piping segment leading to bypass valve FCV-74-9 and the branch connection in the piping segment leading to bypass valve FCV-74-8 10 22 2485 1500 (The first row entry corresponds to the main piping segment consisting of the 14-inch diameter piping. The second row entry corresponds to the total amount of 10-inch diameter piping in the branch connections)

Auxiliary Spray System Piping Auxiliary Spray System Piping from Flow 3 41 2485 325 - 2235 Control Valve FCV-62-84 to Check Valve CKV-62-661

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