ML061180078

Unit 3 Calculations for Oconee TAC MC8127

ML061180078
Person / Time
Site:	Oconee
Issue date:	03/27/2006
From:	Oakley R Duke Energy Corp
To:	Olshan L Office of Nuclear Reactor Regulation
References
TAC MC8127
Download: ML061180078 (79)

Text

J Leonard 01shan - TSC Number 2004-08 a sPage 1 From: Russell L Oakley <rloakley~duke-energy.com>

To: <Ino@nrc.gov>

Date: 3/27/06 9:43AM

Subject:

TSC Number 2004-08 Attached are the vendor calculations prepared by Stone & Webster for jet impingement and missile protection for Oconee Nuclear Station (ONS) Unit 3.

The calculations are now owner approved, and cover sheets showing owner approval signatures are also attached. These are the calculations which we discussed last week. These analyses were identified in the NRC's Safety Evaluat on dated November 1, 2005 as being needed for NRC review prior to approval of a license amendment request for Unit 3. The requested amendment changes TS Surveillances SR 3.5.2.6 and SR 3.5.3.6 to replace the words "trash racks and screens" with "strainers" to accurately describe the installed configuration of new sump strainers at ONS.

When approved, we will need to replace current TS pages 3.5.2-5a, 3.5.2-5b, 3.5.3-3a, and 3.5.3-3b with new pages 3.5.2-5 and 3.5.3-3. The new pages will read identically to 3.5.2-5a and 3.5.3-3a, but will have new amendment numbers. Your assistance with the requested changes is appreciated.

Please call me at 864-885-3829 if there are any questions.

(See attached file: Duke Signature Sheet of S&W Missile Evaluation.pdf)(See attached file: Calc S-005 R.1.pdf)(See attached file: Calc S-006.800 Rev 1

.pdf)(See attached file: Duke Signature Sheet of S&W Jet Imp - Pipe Whip Evaluation .pdf)

Graciela SanGabr el - GW}00001 .TMP Page 1 w Mail Envelope Properties (4427F9FC.477: 19: 38007)

Subject:

TSC Number 2004-08 Creation Date: 3/27/06 9:41 AM From: Russell L Oakley <rloakleydiduke-energy.com>

Created By: rloakley~duke-energy.com Recipients n c.gov TWGWPOOI.HQGWDOO0 LNO (Leonard Olshan)

Post Office Route TWGWPO01.HQGWDO0I nrc.gov Files Size Date & Time MESSAGE 1315 03/27/06 09:41AM Duke Signature Sheet of S&W Missile Evaluation.pdf 36770 Calc S-005 R.l.pdf 3509156 Calc S-006.800 Rev I .pdf 3649842 Duke Signature Sheet of S&W Jet Imp - Pipe Whip Evaluation .pdf 35486 Mirne.822 9899435 a ptions Expiration Datc: None Priority: Standard Reply Requested: No Return Notification: None Concealed

Subject:

No security: Standard

ShaW' Store &Webster, Ina CALCULATION TITLE PAGE CUENT & PROJECl' PAGE I OF 22 DUKE POWER COMPANY/OCONEE NUCLEAR STATION, UNIT 3 Total Pages (cinc. Attachmlents) = 36 CALCuLATnON TrrEE (Indicative of the Objective): QACATEGORY (X) 12I - NUCLEAR SAFETY i EATED Missile Evaluailon for the Emergency Sump Strainer.

OII 0III 0 _ -

THER CALCULATION IDENTFIcAToN NUMBER CURRENT WORK BREAKDOwN STnUCTURE OPTIONAL J.O. OR W.O. NrO. DIVISION & GROUP CALCULATION No. NO. WORK PACKAGE NO.

115524 STR S-005 800 N/A CoNFiRMATnoN

• APPROVAs-SIGN kTURE & DATE REV. NO. SUPERSEDES *REQUIRED (X)

OR NEW *CA(C. No.

CALc. NO. OR REV. No.

PREPARER(SyDATn(s) REVIEWER(S)/DATE(S) INDEPENDENT YES No REVIEWER(SyDATE(S)

Alex Cokonis Dennis Smith Dennis Smith 0 N/A X Z/X/o6' ~ -r /

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Al/ex Cokpil;, 4 ivvS fmT rml/r 3 5 8fl7I/

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GROUP NAME & LOCATON GROUP NAME & LOCATION Project File

SHaw stone &Wbster Ina CALCULATION SHEET

)N IDENTIFICATION NUMBER TABLE OF CONTENTS Title Page 1 Table of Contents 2 Review Statement 3 Revisicn Status Sheet 4 1.0 Objective 5 2.0 Methodology 6 3.0 Assumptions 8 4.0 Design Input 8 5.0 References 9 6.0 Calculation 11 7.0 Conclusion 22 Total Pages 22

[List Attachments] PIP 0-04-07314, CA #6 Total Pages 1 3LP-103 Valve Missile Total Pages 1 LP, CF Valve Cover Missile Trajectories Total Pages 1 RC Drain Valve Missiles Total Pages 2 System Diagrams Total Pages 9

Shaw stone &Webster, Inc CALCULATION SHEET Review of this calculation was based on the methods below:

1) Review of:

Initial Upon a) Inputs to ensure that they have been properly selected Completion and correctly used in the calculation. (Check One) i) Limited review (provide justification) ii) Line by line review 0El CD O Z:: I b) Assumptions to assure their validity and need for later confirmation.

c) Methodology to assure the appropriateness of the overall approach, its implementation, and the correctness of the specific equations utilized.

I) Limited review (provide justification) El ii) Line by line review 0 d) Results to ensure reasonableness and accuracy 0rm~~

e) If alternate calculation is performed to verify c) and d) check here and attach calculation as an appendix Co

2) Check of Calculation (Check One) a) Complete numerical check b) Numerical check of critical items 0 (state items and justification below)

3) Administrative check of format and content 0 _

4) Comments/Justification ReAviwMathods Selected as Indicated Above

.

Reviewerf 7,ate '

Independent RevieweF Date Lead concurrence Date

" - -

ShawW- Stone &VvMbster, k-r-CALCULATION SHEET REVISION STATUS SHEET Revision Affected Number Sections Description of Revision 0 ALL Original issue 1 Pages 7, 8, 10, 11, Incorporate Duke Power comments 18, 19, 22, Att. I

Sthaw -Stne &VWbstr, hc CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

).OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO. I WBS NO. PAGE 5 115524 l STR S-005 1 800 OF 22 1.0 OBJECTIVE The new emergency sump strainer, located inside containment, must be protected from missiles per NRC generic letter 2004-02 (Ref. 1) and Design Input Calculation Ref. 18. The missile protection methodology for Oconee has been defined in OSS-0254.00-00-4018 (Ref.

13) and the UFSAR section 3.5 (Ref. 14). The missiles are generated by high energy piping and equipment (P operating > 275 psi or T operating > 2001F) and missiles are generally l he following items per Ref. 13:

1. Valve stems

2. Valve bonnets

3. Instrument thimbles

4. Nuts and bolts The systems identified in Ref. 13 and 14 as potential sources of missiles in the Oconee units, are the Control Rod Drive, Core Flooding Line, L.P. Injection, R.V. Outlet Line to L.P., R.V. Inlet line from HP, S.G. Outlet line to pump inlet, Pressurizer to C.A. system line, primary pump seal water return to H.P. system line, Letdown cooler inlet & outlet lines, primary pump seal water inlet and outlet lines and primary pump vent and drain lines.

The purpose of this calculation is to identify the potential sources of missiles in the Emergency Sump area and determine if there is a design impact on the new emergency sump straine.

2.0 METHODOLOGY Missiles can be generated by the structural failure of various pressurized or rotating components. This evaluation assesses the change in missile target susceptibility due to extending the Reactor Building Emergency Sump (RBES) strainer above floor level. The RBES strainer is being modified to be larger as part of the resolution to GL 2004-02 (Reference 1).

None of the missile sources or other potential targets are being changed by this modification, except the RBES level instruments which are being relocated from one on each side of the sump to both on one end of the sump. The new locations for the instruments are shadowed by the strainers relative to missile trajectories, however, and do not require a separate evaluation.

Tne RBES strainers have a required safety related function only during LOCA events for pipe breaks of any size (References 20 & 21).

Missiles are postulated as an accident event due to a pressure boundary failure in two ways:

a. LOCA - If the generation of a missile causes a LOCA (e.g., ejection of a reactor coolant loop valve stem) and damages the RBES strainer, the results are unacceptable since the strainer can not be assured to perform its LOCA function.

SWWaw Stone &V~bsWer k-r-CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

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b. Non-LOCA - Conversely, if a missile is generated that does not create a LOCA, the plant can respond as currently designed even if the strainer is damaged and no provisions not already included in the plant design need to be addressed.

No new active or passive system failures are introduced by this modification, except if a n-iissile damages the RBES strainer during response to a LOCA event. The strainer was generally shielded from missiles in the original design by the sump walls since the strainer was below the floor level and floor decking above the strainer (screen).

The Oconee licensing basis for passive failures includes the following (Reference 3):

• For fluid systems, passive failures are only postulated in Emergency Core Cooling System (ECCS) systems, and only in the systems that must operate for an extended period of time during the long term response to Oconee events. The intent of postulating passive failures is to help ensure that long term event mitigation will be successful even with failures that could reasonably be postulated to occur with extended operation. Pressure boundary failures considered under the single failure criterion are therefore limited to leakage between flanges, gross valve or pump seal (or packing) leaks, etc., but shall not include pipe breaks or cracks.

• The ECCS consists of the High Pressure Injection (HPI) System, Low Pressure Injection (LPI) System, and the Core Flood (CF) System.

• The HPI System is not required to withstand passive failures because it does not provide "extended or indefinite" long term cooling.

• Since the Core Flood system is only required to operate during the short term, passive failures in the CF system are not postulated to interfere with its ability to mitigate events.

• If an event is initiated by the postulated failure of one of two or more redundant trains of a dual purpose fluid system (i.e., a system required to operate during normal unit operation as well as to shut the reactor down and mitigate the consequences of ar, event), and a single failure is required to be postulated when evaluating the event, the single failure shall not be assumed in the remaining train or trains of that system.

Missile generation due to passive failures during an accident event is not postulated since the licensing basis is limited to failures of "soft" parts (gaskets, seals, packing) and excludes gross structural failures. Since valve stem, bonnet or operator ejection is a gross structural failure it is not considered to be credible and is not postulated.

Therefore, only a missile resulting from an accident that creates a LOCA is of consequence to the function of the RBES strainer.

The methods used to identify missiles and assess their damage potential for the lines ident fied in the sump area, is based on the following criteria:

1. The line that creates missiles must be high energy (P operating > 275 psi or T operating >

2000F). The temperature and pressure are based on normal operating conditions.

Shaw -Stone &Vfbster, ha CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

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Conditions that exceed the 275 psi or 200OF limits but occur for less than 2% of the time are not considered high energy (Ref. 15).

2. Missiles have to be postulated for high energy lines, even if the lines were excluded for pipe l break by either SRP (Standard Review Plan) or LBB (Leak before Break) criteria.

3. The missiles are valve stems and bonnets, instrument thimbles, and nuts and bolts. Welded components, except instrument thimbles, are not considered for missiles. Welded components are considered for pipe break whipping effects and jet impingement.

4. The missile trajectory of nuts and bolts is in the same direction as the bolt axis with a trajectory tolerance of 50. The direction of the missile for a valve stem is the stem axis and the missile direction of valve bonnet is defined by the axis of the connecting bolts or normal to the opening where the bonnet is bolted.

5. Missile energy computation and damage effect on strainer will be based on Ref. 13 & 19 criteria. There are three methods (missile categories per Ref. 13), used to compute the velocity and energy of the missile. Category I is based on the strain energy and is used for bolt and nut missiles, Category II are small diameter objects, such as valve stems, that are accelerated by the pressure of the fluid and Category IlIl are large objects, such as valve bonnets, that are continuously accelerated by the fluid jet velocity. The resulting missile energy is used to determine the damage on the strainer or to design shields if required.

6. The missile is considered to lose the majority of its energy after it strikes the target. Credible secondary missiles as a result of impact with primary missiles and missiles due to gravitational effects need to be considered per Ref. 16.

7. The secondary shield building walls, column and floor (El. 797'-6"), surrounding the emergency sump strainer are considered as protection of the strainer from missiles from systems outside these barriers.

S8-W Sttne &Vvbster, ha CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

). OR V.ONO. lDIVISION & GROUP CALCULATION NO. REV NO. WBS NO. PAGE 8 115524 STR S-005 1 800 OF 22 3.0 ASSUMPTIONS None 4.0 DESIGN INPUTS

1. Walkdown of Unit 1 and Unit 2 emergency sump strainer was performed on Thursday April 21, 2005 and Tuesday November 1, 2005 respectively. The valves and other bolted piping components that are potential missiles, which were in the line of sight of the emergency sump strainer, were recorded. The Unit 1 and Unit 2 information which is similar to Unit 3 was used as reference for the Unit 3 missile evaluation.

2. Unit 3 piping layout drawings, Ref. 17a, b, c, d were used to identify locations of potential missiles. Sections from these drawings in the area of emergency sump are shown in Figures 1, 2, 3 & 4 and Attachments 2 & 4.

3. Photos were obtained for Unit 3 sump on 12/16/04. These photos were limited and identified the stored flanges, the level instruments, the CS piping and the overhead CF and LP piping. The Unit 3 LP and CF valves are shown in Attachment 3 to identify potential missiles from these valves.

4. Unit 3 system flow diagrams, References 2, 4 through 12, 23 and 25, were listed in Table A which identifies all systems in emergency sump area. The portion of the piping system in the area of the sump that is pertinent to this missile evaluation was also identified by a balloon on a copy of the referenced system flow diagrams and is shown in Attachment 5.

Shaw- Stone & VWbster. Ic CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

).OR .O NO. DIVISION & GROUP CALCULATION NO. REV NO. I WBS NO. PAGE 9 115524 STR S-005 1 800 OF 22 _

5.0 REFERENCES

1. Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, 9/13/04.

2. Oconee Drawing OFD-102A-3.1, Rev. 50, Flow Diagram of Low Pressure Injection System (Borated Water Supply and LPI Pump Suction).

3. Oconee Specification OSS-0254.00-00-4013, Rev. 2, Design Basis Specification for the Oconee Single Failure Criterion.

4. Oconee Drawing OFD-104A-3.1, Rev. 26B, Flow Diagram of Spent Fuel Cooling Sys4:em.

5. Oconee Drawing OFD-102A-3.3, Rev. 19, Flow Diagram of Low Pressure Injection System (Core Flood).

6. Oconee Drawing OFD-121 B-3.5, Rev. 26, Flow Diagram of Feedwater System (Steam Generator Drain & Recirculation System).

7. Oconee Drawing OFD-107A-3.2, Rev. 9, Flow Diagram of Coolant Storage System (Component Drain Pump).

8. Oconee Drawing OFD-1 01A-3.5, Rev. 20, Flow Diagram of High Pressure Injection System (SSF Portion).

9. Oconee Drawing OFD-101A-3.3, Rev. 19, Flow Diagram of High Pressure Injection System (Charging Section).

10. Oconee Drawing OFD-102A-3.2, Rev. 30, Flow Diagram of Low Pressure Injection System (LPI Pump Discharge).

11. Oconee Drawing OFD-1OOA-3.1, Rev. 31, Flow Diagram of Reactor Coolant System.

12. Oconee Drawing OFD-101A-3.4, Rev. 34, Flow Diagram of High Pressure Injection System (Charging Section).

1:3. OSS-0254.00-00-4018, Design Basis Specification for the Missile Protection, Rev. 1

14. UFSAR (Updated Final Safety Analysis Report) as off 6/30/04, Section 3.5

15. NUREG-0800, Standard Review Plan (SRP), Section 3.6.2, "Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", Rev. I

-July 1981.

113. NUREG-0800, Standard Review Plan (SRP), Section 3.5.1.2, Internally Generated Missiles (Inside Containment), Rev. 2 - July 1981.

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5524 STR S-005 1 800 OF 22 _

17. Drawings

a. O-2478A, Rev. 42. "Piping Layout Basement Floor Plan Reactor Building, Oconee NS Unit 3.

b. O-2478E, Rev. 15. "Piping Layout Basement Floor Partial Plan & Sections, O.T.S.G.

Recirculation System, Reactor Building, Oconee NS Unit 3.

c. 0-2478F, Rev. 12. "Piping Layout Basement Floor Partial Plan & Sections, SSF System, Reactor Building, Oconee NS Unit 3.

d. 0-2478D, Rev. 36. "Piping Layout Basement Floor Partial Plan & Sections, Reactor Building, Oconee NS Unit 3".

18. OSC-8739, Design Input Calculation for Units 2 & 3 Reactor Building Emergency Surip Strainer Replacement, Rev. 3.

19. ORNL-NSIC-22, Sept 1968, "Missile Generation and Protection in Light-Water-Cooled Power Reactor Plants".

20. Oconee Calculation OSC-6182, Rev. 15, Oconee Nuclear Station Units 1-3 Main Steam Line Break (MSLB) Event Mitigation Requirements.

21. Oconee Calculation OSC-7346, Rev. 1, Main Feedwater Line Break Event Mitigation Requirements.

22. Oconee Drawing OFD-121D-3.1, Rev. 34, Flow Diagram of Emergency Feedwater System.

23. 0-2422-X-061, Rev. 0, Instrument Detail, LPI Cross-Tie Flow Restrictor, Train A 3LPIFE-0006.

24. PIP 0-04-07314 2.5. Oconee Drawing OFD-1 10A-3.4, Rev. 7, Flow Diagram of Chemical Addition System (Post l Accident Liquid Sampling)

SCALwCUStone & bsterEk CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W1.O NO. DIVISION & GROUP CALCULATION NO. REV NO.

115524 STR S-005 I 6.0 CALCULATION SYSTEMS IN EMERGENCY SUMP AREA:

The emergency sump strainer is located in an enclosure in the basement floor of the containment surrounded from three sides, (north, east and west) by a 4 ft wall and partially protected on the south side by a 4'x6' column (see Figures 1, 2, 3 & 4). The missiles from the major lines in the reactor building, such as Reactor Coolant, Mainsteam and Feedwater are outside the enclosure (at higher elevation), and do not affect the subject strainer. There are, however, high energy systems located in the vicinity and inside of the emergency sump enclosure which were identified by Unit 3 photos, Unit 3 composite drawings and Unit I and Unit 2 walkdowns. Additional valve mis<;iles were identified by walkdown in PIP 0-04-07314, CA #6, Ref. 24 for Unit 2 (see Attachment 1), which were also identified by similarity for Unit 3 using Unit 3 composite drawings.

The pertinent systems are listed in Table A below. Potential missiles from these systems are evaluated in Table B.

Table A. - Piping Systems in Emer ency Sump Area (See Fig. 1, 2, 3 & 4)

Plant Normal Duke Pipe Design Design Operating Operating Unit 3 ID SYSTEM Class OD Pressure Temp 2 2 Press 2 Temp 2 Flow Diagram' (in) (psig) (F) (psig) (OF)

CS Coolant Storage E 4.5 65 300 Not used Not used OFD-107A-3.2 LWD CF Core Flooding A 14 2500 300 600 250 OFD-102A-3.3 CF Core Flooding Drain A 1.313 2500 300 600 250 OFD-102A-3.3 LP Low Pressure Injection A 10.75 2500 300 600 250 OFD-102A-3.3 LP LO'N Pressure Injection A 1.313 2500 300 600 250 OF11-102A-3.3

____ Drain l LP Low Pressure Injection A 12.75 2500 650 2185 603 OFD-102A-3.1

- Decay Heat l LPI Instrument LPI Cross-tie B 0.84 2500 300 600 250 0-2-422-X-061 CCW Component Cooling B 6.675 1275/1050 475/600 Not used Not used OFD-121D-3.1 HP Hich Pressure Injection B 2.375 2790 220 Not used Not used OFD-101A-3.5 FDW Feedwater- Steam D 4.5 1050, 275 600, 250 Not used Not used OFD-121B-3.5 Generator Drain &

Recirc SF Spent Fuel Cooling C 4.5 100 200 Not used Not used OFD-104A-3.1 RC RC - Cold Leg A 28 2500 650 2185 650 OFD-100A-3.1 3A2, 3B2 RC RC - Cold Leg Drain & A 1.9" 2500 650 2185 650 OFD-100A-3.1 Sample line 1.315" _ IOFD-1I OA-3.4

-Shv Stone &Webster, ha CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

OR V.O NO. DIVISION & GROUP CALCULATION NO. REV NO. l WBS NO. PAGE 1:'

115524 STR S-005 1 800 OF 22 Notes:

1. See the flow diagrams in Attachment 5 and Figures 1 through 4 for location of piping in the general area of the emergency sump. Table B is a specific subset of components found in

• hese piping systems which have been identified as having a potential clear path to the emergency sump/strainer area.

2. Design pressure and temperature are obtained from corresponding flow diagrams referenced in table above. Operating temperatures and pressures obtained from UFSAR, Ref. 14, Fig. 3-10 (CF, LPI) and Table 5-16 (RCS).

Shaw Stne & bster, hc CALCULATION SHEET c'o

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ShaW stone &Vkbste, ha CALCULATION SHEET Table B. - Missile Evaluation for Piping Systems in Emergency Sump Area Evaluation Criteria - For the missiles to be a concern, it must both create a LOCA and damage the RBES strainer.

Missile Does Missile Could Strainer Missil Energy Generation Be Impacted by SYSTEM Type (Note 14) Create a LOCA? Missile Conclusion CS Coolant Storage None N/A No No Not high energy on basis of low design pressure.

LWD (Note 1) (Note 6) No missile concern.

CF Core Flooding 14" Check Valve 2.00E6 fl-bs No YES Missile event is non-LOCA, (Note 4). No missile (3CF-13) Cover (Note 4) by 3CF-13 concern.

(Note 11)

CF Core Flooding Drain I " Gate Valve 200 fl-lbs No No Valve missile will not strike strainer and will not (3CF-31) bonnet & (Note 4) (Note 15) cause a LOCA. No missile concern.

hand wheel assembly LP Low Pressure Injection 10" Check Valve I.01E6 fl-lbs No YES Missile will not cause a LOCA, (Note 4). No (3LP-177) and flow (Note 4) by 3LP-177 and missile concern.

restrictor (3LPIFE- 3LPIFE-006 006) Cover (Note I 1)

LP Low Pressure Injection I" Gate Valve 200 fl-lbs No YES LP Drain missile event will not cause a LOCA, Drain (3LP-1 81) bonnet & (Note 4) by 3LP-1 81 (Note 4 & 16). No missile concern.

hand wheel (Note 16) assembly LP Low Pressure Injection 3" MOV (3LP-103) I.01E4 fl-lbs Yes for valve No Valve pointing away from strainer (Note 12). No

- Decay Heat bonnet & operator 3LP-103 (Note 12) missile concern.

assembly LPI Instrument LPI Cross- 'A" Instrument valve 100 fl-lbs No YES Missile event will not cause a LOCA (Note 3). No Itie l (3LP-190) stem j (Note 3) l by 3LPI-190 Imissile concern. l

Shaw -stone &VMbste, ha CALCULATION SHEET C'AI ('II AC"i'nl ifr-NClTItC.^ATl>tfA Kil IrAD~e:

I J.O. OR W.O NO. DIVISION & GROUP CALCULATION NO.

115524 STR S-005 Table B. - Missile Evaluation for Piping Systems in Emergency Sump Area Evaluation Criteria - For the missiles to be a concern, it must both create a LOCA and damage the RBES strainer.

Missile Does Missile Could Straincr Missile Energy Generation Be Impacted by SYSTEM Type (Note 14) Create a LOCA? Missile Conclusion LPI Instrument LPI Cross- 1/2"Instrument block 100 ft-lbs No YES Missile event will not cause a LOCA (Note 3). No tie valve (3LPIIV0162) (Note 3) by 3LP11V0162 missile concern.

stem lIp Iligh Pressure I" drain valve 200 ft-lbs No Yes by 31 IP-423) Missiles in direction of strainer however event Injection (311P-423), Stem (Note 5) (Note 9) can not cause a LOCA. No missile concern.

and hand wheel FDW Feedwater - Steam 3" valve (3FDW- N/A No No Valve 3FDW-326 points towards the strainer Generator Drain & 326) bonnet & (Note 1) (Note 8) however not enough energy to generate missile, Recirc assembly since FDW system is isolated and not operating during plant operation (Note 8). No missile concern.

FDW Feedwater - Steam 3" valve (3FDW- N/A No No Valve 3FDW-429 points towards the strainer Generator Drain & 429) bonnet & (Note I) (Note 8) however not enough energy to generate missile, Recirc assembly since FDW system is isolated and not operating during plant operation (Note 8). No missile concern.

FDW Feedwater - Steam 3" valve (3FDW- N/A No No Valve 3FDW-539 points towards the strainer Generator Drain & 539) bonnet & (Note 1) (Note 8) however not enough energy to generate missile, Recirc assembly since FDW system is isolated and not operating during plant operation (Note 8). No missile concern.

FDW Feedwater - Steam 3/4" drain valve N/A No No Valve 3FDW-540 points towards the strainer Generator Drain & (3FDW-540) stem (Note I) (Note 8) however not enough energy to generate missile, Reci assembly since FDW system is isolated and not operating during plant operation (Note 8). No missile concern.

ShiaW Stone &WbvSte, kn CALCULATION SHEET Table B. - Missile Evaluation for Piping Systems in Emergency Sump Area Evaluation Criteria - For the missiles to be a concern, it must both create a LOCA and damage the RBES strainer.

Missile Does Missile Could Strainer Missile Energy Gcneration Be Impacted by SYSTEM Type (Note 14) Create a LOCA? Missile Conclusion CCW Component Cooling 4" MOV (3CCW- 2.9E5 fl-lbs No YES by 3CCWV- Missile event can not cause a LOCA, (Note 1).

269) bonnet & (Note 1) 269 (Note 10) No missile concern.

operator assembly SF Spent Fuel Cooling None N/A No No No missile generating components in strainer (Note 1) (Note 7) area. No missile concern.

RC RC - Cold Leg- 3A2, None N/A No No No missile generating components in strainer 3B2 (Note 7) area. No missile concern.

RC RC - Cold Leg- 1.5" ," Valve stem 154 fl-lbs Yes No 3RC-84 points up and will strike the RC cold leg Drain & Sample & wheel (3RC-29, (Note 13) and other components and will not reach the 3RC-46,3RC-84) strainer. Other valves point away from strainer (Note 13). No missile concern.

RC RC - Cold Leg - 1.5" Valve stem & 154 fl-lbs No No Valves point up and away from strainer. No Drains wheel (3RC-47, missile concern.

3RC-30)

A Notes

1. Not part of reactor coolant pressure boundary.

2. No missile generating components in area of strainer.

3. Isolated from reactor coolant pressure boundary by 2 redundant check valves (3LP-177 and 3CF-14, Ref. 5) and 2 redundant normally closed block valves (3RC-45 and 3RC-43, Ref. 11). Event therefore can not cause a LOCA.

IA

Shaw-stone &Vbster, nc CALCULATION SHEET ICAl CI' 11ATi(nki lr-N1TIrClI'TArniK 1 IIwADR MCI J.0. OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO. WBS NO. PAGE 20 115524 STR S-005 1 800 OF 22

4. Isolated from reactor coolant pressure boundary by I check valve (3CF-14).

5. Portion of system in this area is actually SSF, which is normally isolated and not directly connected to reactor coolant pressure boundary.

6. Portion of system in this area does not contain high pressure.

7. Portion of system in this area has no missile generating components.

8. Portion of system in this area is normally isolated and used only during unit shutdown. Any pressure contained in the system during normal operation is due to valve leakage and is not of sufficient energy to generate a missile.

9. Valve stem is pointing 450 down. The potential missile will strike the floor first and then rebound into the strainer.

10. The valve (3CCW-269) missile will be 10 ft above the strainer in the horizontal direction due north and will strike the secondary shield wall 10 ft away. The missile will dissipate its kinetic energy on impacting the wall and will then drop on the strainer by gravitational inertia. See Figure 3.

11. The 3CF-13 and 3LP-177 valves and 3LPIFE-0006 flow restrictor are above the strainer and point up. The missile (valve cover plate) will be initially traveling upward and after striking the ceiling or other piping, will free fall and possibly strike the top of the strainer. See Attachment 3.

12. Valve 3LP-103 is pointing up and is located at elevation 797'-9". The secondary shield wall (east side of pool), protects the emergency strainer from the missile free fall. See Attachment 2.

13. The 1.5" drain valves 3RC-29 and 3RC-46 are pointing up and are 25 ft away from emergency sump hence missiles will not strike the strainer. The 1" sample line valve 3RC-84 is pointing up and will strike the RC cold leg and may rebound into the FDW recirc piping and puiip anid will dissipate a significant porlion of energy and may not reach the sump area. The other sample line valves, RC-462, 163, 174, 207, 208 and 209 are not a missile concern for the strainer since they are pointing away from the sump. See plan drawings, Attachment 4.

Shw-stone &Vmmter, hc CALCULATION SHEET CALCU'LTIMON IDENT!F!C^r'TIONi NU.'ADER J.O. OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO. l WBS NO. PAGE 21 115524 STR l S-005 1 800F 22

14. Missile energy values are approximate and are obtained from Ref. 13 by comparison to the energies of similar components and system pressure and temperature conditions. These approximate missile energies are used to assess by engineering judgment the missile damage potential to the new strainer.

15. Drain valve 3CF-3 1, pointing due East and away from strainer. Missile will exit the sump area through the East side opening of the emergency sump cubicle.

16. Drain valve 3LP-1 81 will travel horizontally and strike the East wall at EL. 790'-8" which will loose its energy and fall on the strainer by gravitational inertia. The energy may not be sufficient to pierce the strainer plate.

ShAw~ Ston &Vmoster' hac CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER OR W.O NO. l DIVISION & GROUP l CALCULATION NO. I REV NO. WBS NO. PAGE 22 j 15 2 4 STR S-005 1 800 OF 22

7.0 CONCLUSION

S Based on this evaluation the emergency strainer operability will not be compromised during E LOCA by missiles. The primary reasons are:

1. Systems CS, SF and LP(Decay Heat) have no missile producing components that face the strainer.

2. The main RC(28") cold leg has no missile producing components in vicinity of strainer.

3. FDW has valves (3FDW-326, 3FDW-429, 3FDW-539, 3FDW-540) pointing towards the*lA strainer but system is isolated during normal plant operation and considered low energy and does not generate a missile.

A

4. Missile events of HP, CF, LP, LP drain, LP instrument and CCW will not cause a LOCA and therefore strainer is not required to function.

5. CF drain missile does not strike strainer and also strainer is not required to function during these accidents.

6. RC drain valves and RC sample line valves point up or away from strainer.

SIOwStoneE&WTteAHkEN1 CALCULATION SHEET - ATTACHMENT I A1 PIP No: 0-04-07314, CA #6 Problemn Invrmtdgailkn Procesv Ocunee Nudlear S~a~kpn CA Seq. No:_6__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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Code .. I M(:E _ Closed MCF 01 13 R Protiosci Corrective Actaon-Perforn confirnatory walk-down of Unit 2 Reactor Building during 2EOC21 refueling outage to identify potential missiles which could impact the modified RBES strainer. Document the results of missile protection walkdown in this PIP CA. This CA has been discussed with (and accepted by) Bob Hcstcr of MCE/Civil.

Originated By: RL08372: OAKLEY, RUSSELl. 1, ream: RJI'21 11 RcdyPor Ap roval: RLS0i372 RJF21 II Approvil Assigned Io: RJF2 1ll RJF21 11 Genrl:Outage: N/A Mode: N/A Other rracking Processes Tl~n- h~inbcr icTsd Actual 'orreefise Action-Priority' 12b Actual CAC: B3 Status: Clnscd Due Datc: 1027/2005 Purposc -f this conrrctive action was to identify potential missilc sourccs that could impact the proposed rnodification to the Reactor Boui'ding Emcrgcncy Sump. During 2EOC21 refucling outage 1.R V Hcster, conducted a walkdown of the Elmergency Sump area to identify any patential missiles auch as nre described in the UI'SAR (valve stems, valve boinnets, valve bonnets studs, reactor coolant temperature sensors, manvway or inspection covers or manway or inspection cover studs) which could impact thc Fmcrgcncy Sunp or any of its components, particularly considering the new and snon to be installed sump screen. Valves 2FDW539, 2FDW429, 2HP423, 2CS 199 and 2LP162 IV, and another small valve on the bottom ofthe new LPI Cross Over Piping all have stems oriented tcwmard the sulop.

Proposed Corrective Action 17 is created to have ihe vrendor CadlCrcoicwd to verify thcsc valves were cvaluated a potential missiles.

Originauxd By: RVHI4032: IIESTER. ROBERT V Tcam: PAW4981 Group: MCE Date: 10/27/2005

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Shaw -Stone &Wetet Inha CALCULATION TITLE PAGE CLIENT & PROJECr PAGE I OF 20 DUKE POWER COMPANY/OCONEE NUCLEAR STATION, UNIT 3

_ __ __ __ ______ __ __ _ _ _ _ __ __ _ _______ _ __ _ _ _ _ _____ _____ ____ _____ _ _ _T _P _ Total_ ges ( cI. ttach aents = 44 tal Pages_

CALCULATION TnILE (Indicative of the Objective): QA CATEGORY (X) 0 I - NUCLEAR SAFETY RELATED Jet Impingement and Pipe Whip Evaluation for the Emergency Sump Strainer.

011 0111 0 OTHER CALCULATION IDENTIFICATION NUMBER CURRENT WORK BREAKDOWN STRUCTURE OPTIONAL J.0. OR W.O. No. DIvIsION & GROUP CALCULATION No. No. WORK PACKAGE No.

115524 STR S-006 800 N/A CONFIRMATION

• APPROVALS-SIGNATURE &DATE REV. No. SUPERSEDES *REC!UIRED (X)

OR NEW *CALC. No.

CALC. No. OR REV. No. _

PREPARER(S)/DATE(S) REVIEWER(SYDATE(S) INDEPENDENT YES No REVIEWER(S)yDATE(S)

Alex Coko is Dennis Smith Dennis Smith 0 N/A X

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DISTRIBUTION GROUI NAME & LOCATION GROUP NAME & LOCATION Project File

Shaw- stone &Webster, h CALCULATION SHEET TABLE OF CONTENTS Title Page 1 Table of Contents 2 Review Statement 3 Revision Status Sheet 4 1.0 Objective 5 2.0 Methodology 5 3.0 Assumptions 7 4.0 Design Input 7 5.0 References 8 6.0 Calculation 11 7.0 CDnclusion 20 Total Pages 20

[List Attachments] RCS Break Evaluation Total Pages 4 LP, CF Break Evaluation Total Pages 1 PIP 0-04-07314, CA #5 Total Pages 4 LP-Decay Heat Line Break Total Pages 1 RC Drain Jet Path Total Pages 1 E-Mail on RC Drain Line Jet Total Pages 1 System Diagrams Total Pages 9

ShawW Stone & Webster, ha CALCULATION SHEET Review of this calculation was based on the methods below:

1) Review of:

Initial Upon a) Inputs to ensure that they have been properly selected Completion and correctly used in the calculation. (Check One) i) Limited review (provide justification) El ii) Line by line review b) Assumptions to assure their validity and need for later confirmation.

c) Methodology to assure the appropriateness of the overall approach, its implementation, and the correctness of the specific equations utilized.

i) Limited review (provide justification) ii) Line by line review Cod d) Results to ensure reasonableness and accuracy e) If alternate calculation is performed to verify c) and d) check here and attach calculation as an appendix

2) Check of Calculation (Check One) a) Complete numerical check 45-yl b) Numerical check of critical items (state items and justification below)

3) Administrative check of format and content _ __:~!

4) Comments/Justification Raview Methnds Selected as Indicated Ahnve Reviewer Date 4-Independent Reviewer Date Lead concurrence Date

A' Shaw-Stone&WtsW,cr CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO.

115524 STR S-006 1 REVISION STATUS SHEET Revision Affected Number Sections Description of Revision 0 ALL Original issue 1 Pages 5, 7, 11, 12,17,18, Incorporate Duke Power 19, 20, Attachments 3, 6, 7 comments.

Shaw Stone &Vmmwste, har CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

D.OR VW.O NO. DIVISION & GROUP CALCULATION NO. REV NO. I WBS NO. PAGE '

115524 STR S-006 1 800 OF 20 1.0 OBJECTIVE The new emergency sump strainer, located inside containment, El. 774 ft, must be protected l from jets per NRC generic letter 2004-02 (Ref. 1) and Design Input Calculation Ref. 22. The intent of this evaluation is to identify all the piping systems in the vicinity of the rew emergency strainer and determine if there is the potential for damage to the strainer due to pipe rupture and jet impingement, using the pipe break design basis for the Oconee plant.

This evaluation will:

1. Determine if a rupture of any of the piping systems in the area can impact the design function of the strainer.

2. Determine if there is a need to design additional protection for the strainer such as pipe whip supports or jet impingement shields.

2.0 METHODOLOGY Postulated double ended ruptures in high energy piping can result in rapid movement in any unrestrained direction of the broken piping segments caused by escaping high energy fluid.

The movement of the pipe segments is generally called "pipe whip" and the escaping fluid are called "jets." Both the impact of whipping pipe and impingement of high energy jets can be very damaging. This evaluation assesses the change in jet impingement (JI) and pipe whip (PW) damage susceptibility due to extending the Reactor Building Emergency Sump (RBES) strainer above floor level. The RBES strainer is being modified to be larger as part of the resolution to GL 2004-02 (Reference 1).

None of the postulated pipe rupture locations or other potential targets are being changed by this modification, except the RBES level instruments are being relocated from one on each side of the sump to both on one end of the sump. The new locations for the instruments are shadowed by the strainers relative to PW and JI, however, and do not require a separate evaluation.

The RBES strainers have a required safety related function only during LOCA events for pipe breaks of any size (References 2 and 3).

Pipe ruptures are postulated as an accident event due to a pressure boundary failure in two ways:

a. LOCA - If the pipe rupture causes a LOCA and a jet damages the RBES strainer, tWe results are unacceptable since the strainer can not be assured to perform its LOCA function.

b. Non-LOCA - Conversely, if a pipe rupture does not create a LOCA, the plant can respond as currently designed even if the strainer is damaged and no provisions not already included in the plant design need to be addressed.

Shaw- Stone &Vkbster, ha CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

TT.O NO. l DIVISION & GROUP l CALCULATION NO. I REV NO. l WBS NO. PAGE e 15524 STR S-006 1 800 OF 20 No new active or passive system failures are introduced by this modification, except if a pipe rupture or jet damages the RBES strainer during response to a LOCA event. The strainer vias generally shielded from jets in the original design by the sump walls since the strainer was below the floor level and there is floor decking above the strainer (screen).

The Oconee licensing basis for passive failures includes the following (Reference 3):

• For fluid systems, passive failures are only postulated in Emergency Core Cooling System (ECCS) systems, and only in the systems that must operate for an extended period of time during the long term response to Oconee events.

• The ECCS consists of the High Pressure Injection (HPI) System, Low Pressure Injection (LPI) System, and the Core Flood (CF) System.

• Pressure boundary failures considered under the single failure criterion are limited to leakage between flanges, gross valve or pump seal (or packing) leaks, etc., but shall not include pipe breaks or cracks.

Pipe rupture due to passive failures during an accident event is not postulated since the licensing basis is limited to failures of "soft" parts (gaskets, seals, packing) and excludes gross structural failures.

The methods used to determine pipe breaks and resulting jets and assess their damage potential for the lines identified in the sump area is based on the following criteria:

1. Breaks and jets are postulated for high energy piping (normally operating at pressures higher than 275 psig or temperatures higher than 200 0F). Piping that sometimes contains high energy fluid can be classified as moderate energy based on the percent of time the system contains high energy fluid. Per Standard Review plan (SRP) 3.6.2, Ref. 17, a system operating as high energy less than 2% of the time is classified as moderate energy. A moderate energy system is not postulated to have significant jet effects.

2. No intermediate pipe breaks and jets are postulated for piping that has been evaluated and it was determined stresses are below the break postulation criteria of SRP 3.6.2, Ref.

17.

3. No pipe breaks and jets are postulated for piping excluded by application of Leak Before Break (LBB) methodology per NUREG 1061 (Ref. 29) requirements.

4. There are two types of jets per Ref. 24. The longitudinal pipe break causes a jet in the axial direction of the pipe and the path is defined by the trajectory of the whipping pipe. A circumferential break causes a radial break that is a function of the separation of the two ends of the pipe. Jet pressures are based on distance criteria using the standard 100 attenuation method or the two phase flow data of NUREG/CR-2913, Ref. 18. If the distance of the jet source from a target is significant or the break area is small, the potential target damage can be eliminated.

SEW- Stone &Vbbster, ha CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

. OR V.O NO. DIVISION & GROUP CALCULATION NO. REV NO. l WBS NO. PAGE 7 115524 STR S-006 1 800 OF 20

5. Potential damage can also occur from whipping pipes following a rupture. If there is adequate distance from a broken whipping pipe, damage from pipe whip can be avoided.

Generally if a target is not affected by the jet of a broken pipe, pipe whip damage is nol a concern.

6. The secondary shield walls and floor of the Reactor Building (El. 797'-6") surrounding the emergency sump strainer are considered as protection of the strainer from jet impingement and pipe whip from systems outside these barriers.

7. The break postulation methodology defined in Vol. 1 of Ref. 23 that is used for the generation of debris for resolution of GL 2004-02 (Ref 1), is not applicable for the generation of dynamic loads on the strainer. Per Vol.2, SER of Ref. 23, Section 7.1, theA requirements for the evaluation of the strainer for dynamic effects of pipe break is based on the current plant licensing basis for break postulation which is used in this evaluatiol.

3.0 ASSUMPTIONS None 4.0 DESIGN INPUTS

1. Walkdown of Unit 1 and Unit 2 emergency sump strainer was performed on Thursday April 21, 2005 and Tuesday November 1, 2005 respectively. The piping which were in the line of sight of the emergency sump strainer, were recorded. The Unit 1 and Unit 2 information which is similar to Unit 3 was used as reference for the Unit 3 jet and pipe break evaluation.

2. Photos were obtained for Unit 3 sump on 12/16/04. These photos were limited and identified the stored flanges, the level instruments, the CS piping and the overhead CF and LP piping.

3. Unit 3 piping layout drawings, Ref. 21a, b, c, d were used to identify piping that are potential jet impingement sources in the sump area. Sections from these drawings in the area of emergency sump are shown in Figures 1, 2, 3 and 4 and Attachments 4 and 5,.

4. UFSAR (Ref. 20) Section 5.4.8.6, excerpted in Attachment 1, was used to present the pipe break postulation criteria of the RC system.

5. Unit 3 system flow diagrams and drawings, References 5 through 13 and 28, as listed in Table A, which identify all systems in emergency sump area. The location of the piping in the area of the sump that is pertinent to this jet impingement evaluation was also identified by a balloon on a copy of the referenced system flow diagrams and is shown in Attachment 7.

SIw- st&

CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

J.O. OR V.O NO. l DIVISION & GROUP l CALCULATION NO. l REV NO. l WBS NO. PAGE E 11 5'24 STR S-006 1 800 OF 20

5.0 REFERENCES

1. Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, 9/13/04.

2. Oconee Calculation OSC-6182, Rev. 15, Oconee Nuclear Station Units 1-3 Main Stearn Line Break (MSLB) Event Mitigation Requirements.

3. Oconee Calculation OSC-7346, Rev. 1, Main Feedwater Line Break Event Mitigation Requirements.

4. Oconee Specification OSS-0254.00-00-4013, Rev. 2, Design Basis Specification for the Oconee Single Failure Criterion.

5. Oconee Drawing OFD-104A-3.1, Rev. 26B, Flow Diagram of Spent Fuel Cooling System.

6. Oconee Drawing OFD-102A-3.3, Rev. 19, Flow Diagram of Low Pressure Injection System (Core Flood).

7. Oconee Drawing OFD-121B-3.5, Rev. 26, Flow Diagram of Feedwater System (Steam Generator Drain & Recirculation System).

8. Oconee Drawing OFD-107A-3.2, Rev. 9, Flow Diagram of Coolant Storage System (Component Drain Pump).

9. Oconee Drawing OFD-101A-3.5, Rev. 20, Flow Diagram of High Pressure Injection System (SSF Portion).

10. Oconee Drawing OFD-101A-3.3, Rev. 19, Flow Diagram of High Pressure Injection System (Charging Section).

11. Oconee Drawing OFD-1 02A-3.2, Rev. 30, Flow Diagram of Low Pressure Injection System (LPI Pump Discharge).

12. Oconee Drawing OFD-100A-3.1, Rev. 31, Flow Diagram of Reactor Coolant System.

13. Oconee Drawing OFD-101A-3.4, Rev. 34, Flow Diagram of High Pressure Injection System (Charging Section).

14. Oconee Specification OSS-0254.00-00-1028, Rev. 23, Design Basis Specification for the Low Pressure Injection and Core Flood System (LPI).

15. Oconee Specification OSS-0254.00-00-1004, Rev. 22, Design Basis Specification for the SSF RC Makeup System.

Sh -Stone &Vebster, hrra CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR V/.O NO. l DIVISION & GROUP I CALCULATION NO. REV NO. WBS NO. PAGE 9 115524 STR S-006 1 800 OF 20 16;. NUREG-0800, Mechanical Engineering Branch (MEB) 3-1, 'Protection Against Postulated Piping Failures in Fluid System Piping Inside and Outside Containment", Rev.

1-July 1981.

17. NUREG-0800, Standard Review Plan (SRP), Section 3.6.2, "Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", Rev. 1

-July 1981.

18. NUREG/CR-2913, "Two Phase Jet Loads", January 1983.

19. Spec. OSS-0254.00-00-1028, "Low Pressure Injection and Core Flood System (LPI)",

Rev. 23.

20. Oconee UFSAR (Updated Final Safety Analysis Report) as of 6/30/04.

21. Drawings

a. O-2478A, Rev. 42. "Piping Layout Basement Floor Plan Reactor Building, Oconee NS Unit 3.

b. O-2478E, Rev. 15. "Piping Layout Basement Floor Partial Plan & Sections, O.T.S.G. Recirculation System, Reactor Building, Oconee NS Unit 3.

c. 0-2478F, Rev. 12. "Piping Layout Basement Floor Partial Plan & Sections, SSF System, Reactor Building, Oconee NS Unit 3".

d. O-2478D, Rev. 36. "Piping Layout Basement Floor Partial Plan & Sections, Reactor Building, Oconee NS Unit 3".

22. OSC-8739, Design Input Calculation for Units 2 & 3 Reactor Building Emergency Sump Strainer Replacement, Rev. 3.

23. Nuclear Energy Institute (NEI) 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, Rev. 0.

24. Stone & Webster Guideline, EMTR-3-0, Nov. 15, 1976. "Jet Impingement Forces on Essential Structures, Systems and Components".

25. Oconee Specification OSS-0254.00-00-4017, Rev. 4, Design Basis Specification for the Pipe Rupture.

26. Oconee Drawing OFD-102A-3.1, Rev. 50, Flow Diagram of Low Pressure Injection System (Borated Water Supply and LPI Pump Suction).

27. Oconee Drawing OFD-121D-3.1, Rev. 34, Flow Diagram of Emergency Feedwater System.

28. 0-2422-X-061, Rev. 0, Instrument Detail, LPI Cross-Tie Flow Restrictor, Train A 3LPIFII-0006.

ALCULAStone & steET k

CALCULATION SHEET

29. NUREG 1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks", November 1984.

30. PIP 0-04-07314

Shaw- Stone &Vmobste hra CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR VW.O NO. DIVISION & GROUP CALCULATION NO. REV NO.

115524 STR S-006 1 6.0 CALCULATION PIPING SYSTEMS IDENTIFIED IN EMERGENCY SUMP AREA:

At walkdown was performed (on Unit 1 and Unit 2 which are similar to Unit 3), to identify all the piping that potentially could impact the strainer by jet or pipe whip. The systems identified have been verified by reviewing Unit 3 composite drawings (see Figures 1-4 and flow diagrams Attachment 7). The portion of systems listed in Table A are the ones that directly face the strainer and emergency sump only and do not include portion of systems shielded by the secondary shield walls or floors or other barriers or where their distance from the sump is more than 40 pipe diameters (see Note 3 of Table A). A jet impingement and pipe break review for Unit 2 which is similar to Unit 3, shown in PIP, CA #5, Ref. 30, included portions of systems beyond the shield walls and other major barriers for evaluation 1

completeness and is included as Attachment 3 for reference. The jet impingement and pipe break potential effects on the strainer from these systems are evaluated in Table B.

Table A. .- Piping Systems in Emergency Sump Area (See Fig. 1, 2,3 & 4)

Plant Normal Duke Nominal Design Design Operating Unit 3 Pipe ID SYSTEM Class Size Location Valve Press 2 Temp 2 Press 2 Temp 2 Flcw Diagram'

_ I (psig) (F) (pSig) (OF)

CS Coolant Storage E 4" 2 ft east of 3CS-5 65 300 OF D-107A-3.2 LWD strainer Not used Not used CF Core Flooding A 14" 15 ft above 3CF-13 2500 300 600 250/100 OFD-102A-3.3 strainer CF Core Flooding A 1" 15 ft above 3CF-31 2500 300 600 250/100 OF:D-102A-3.3 Drain strainer LP Low Pressure A 10" 15 ft above 3LP-177 2500 300 600 250/100 OFD-102A-3.3 Injection strainer LP Low Pressure A 12" 15 ft above 3LP-1, 2, 2500 650 2185 603 OF:D-102A-3.1 Injection - 3" strainer & 103,104 Decay Heat 4" 15 ft SE _-

LP Low Pressure A 1" 15 ft above 3LP-181 2500 300 600 2501100 OF:D-102A-3.3 Inrection Drain strainer LPI LTl Cross-tie B 1/W 2 ft east of 3LPI- 2500 300 600 250/100 O-2422-X-061 flow restrictor - strainer 154,155

_ nstrument_

SF Spent Fuel C 4" 19 ft above N/A 100 200 OF:D-104A-3.1 Cooling strainer Not used Not used CCW component B 6" 10 to 19 ft 3CCW-269 1050/ 600/ OFD-121D-3.1 Cooling above 1275 475 Not used Not used strainer _

HP igh Pressure B 2" 10 to 16 ft 3HP-398 2790 220 OF!D-101A-3.5 Injection above & 10 Not used Not used ft south of strainer

Shaw' Stone &Vfbster, khac CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER _

J.O. OR W.O NO. lDIVISION & GROUP CALCULATION NO. I REV NO. l WBS NO. PAGE 1:2 115'24 STR S-006 1 800 OF 20 Table A. - Piping Systems in Emergency Sump Area (See Fig. 1, 2,3 & 4)

Plant Normal Duke Nominal Design Design Operating Unit 3 Pipe ID SYSTEM Class Size Location Valve Press 2 Temp 2 Press 2 Temp 2 Flow Diagram' IIII_ I _ I__(psig) (F) I P( ig) (_F) I FDW Feedwater- D 4",3" 2 to 10 ft 3FDW- 1050,275 600,250 O D-121B-3.5 Steam above &5 ft 326,336 Not used Not used Ge nerator Drain south west

&Recirc of strainer RC RC - Cold Leg - A 28" 20 ft SE & STG nozzle 2500 650 2185 650 OFD-100A-3.1 3A2,33B2 20ftSW RC RC - Cold Leg - A 1.5" 20 ft SE & 3RC- 2500 650 2185 650 OFD-100A-3.1 drains 20ftSW 29,30,31,46,

____ _47,48 _-

I Notes:

1. See the flow diagrams in Attachment 7 and Figures 1 through 4 to identify the piping near emergency sump area.

2. Design pressure and temperature are obtained from corresponding flow diagrams referenced in table above. Operating temperatures and pressures obtained from the UFSAR, Ref. 14, Fig. 3-10 (CF, LPI) and Table 5-16 (RCS).

3. Ref. 18 indicates that for diameter ratios greater than 10, (LD>1 0), the jet pressures are insignificant. For conservatism an LID>40 is used as a criterion in this evaluation to ensure t:hat the jet pressure will be insignificant when it reaches the strainer.

A

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CALCULATION SHEET I CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NO. DIVISION & GROUP PAGE 14 CALCULATION NO. REV NO. WBS NO.

115524 STR S-006 1 800 (OF 9n

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Shww stone& bbsar, ha CALCULATION SHEET

Shww- stone & Vmx~w,st hcr CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.0 NO. DIVISION & GROUP CALCULATION NO.

115524 STR I S-006 I

m X 0  ?=o I-- r "

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Shaw Stone &Webst, kha CALCULATION SHEET Table B. - Jet Impingement/Pipe Whip Evaluation for Piping Systems in Emergency Sump Area Evaluation Criteria - For the JI and PW from a pipe break to be a concern, it must both create a LOCA and damage the RBES strainer.

Could Strainer Use Does Pipe Break During a LOCA Be Generation Create Impacted by Pipe SYSTEM a LOCA? Break Conclusion CS Coolant Storage No No Not high energy on basis of low pressure and pipe LWD (Note I) break will not cause a LOCA. No JA (Jet Impingement) or PW (Pipe Whip) concern CF Core Flooding No No No pipe breaks based on MEB 3-1 and LBB. No JI I (Note 2) or PW concern.

CF Core Flooding Drain No No Line isolated from primary water by check valve.

(Note 8) Pipe break will not cause a LOCA. No JI or PW concern.

LP Low Pressure Injection No No No pipe breaks based on MEB 3-1 and LBB. No JI (Note 2) or PW concern.

LP Low Pressure Injection - Decay Heat Yes No Secondary Shield wall protects sump from decay (Note 7) heat pipe rupture. (Note 7). No JI or PW concern.

LP Low Pressure Injection Drain No No Line isolated from primary water by check valve.

(Note 8) Pipe break will not cause a LOCA. No JI or PW concern.

LPI LPI Cross-tie Flow Restrictor Instrument No A

No Line isolated from primary water by double check (Note 3) valves and normally closed block valves. Pipe break will not cause a LOCA. No JI or PW concern.

SF Spent Fuel Cooling No No Moderate energy system based on low pressure. No (lNote i, 6) A or rw concern.

ccW opnnonent oln No No Pipe break will not cause a LOCA. No J1 or PW Cooling M concern.

,I(Note I

Shaw stone & Mbster,ha CALCULATION SHEET Table B. - Jet Impingement/Pipe Whip Evaluation for Piping Systems in Emergency Sump Area Evaluation Criteria - For the JI and PW from a pipe break to be a concern, it must both create a LOCA and damage the RBES strainer.

l Could Strainer Use Does Pipe Break During a LOCA Be Generation Create Impacted by Pipe SYSTEM a LOCA? Break Conclusion lIp High Pressure Injection No No System considered moderate energy based on being (Note 4) active less than 2% of plant operating time. No JA or PW concern.

FD\V Feedwater - Steam Generator Drain & No No Pipe break will not cause a LOCA. No JI or PW Recirc (Note 1) concern.

RC RC - Cold Leg - 3A2, 3B2 No No No pipe breaks due to LBB criteria. No JI or PW (Note 5) concern.

RC RC - Cold Leg - Drains Yes No Jet pressure insignificant based on distance (> 200 (Note 9) pipe diameters), Note 9.

Notes:

1. Not part of reactor coolant pressure boundary. The CS, LWD, SF, CCW and FDW are Duke Class E, B, C, B and D respectively, hence these lines will not cause a LOCA.

2. Pipe breaks have been eliminated for CF and LP piping in this area on the basis of MEB 3-1 and LBB criteria. (References 13, 19 and 25). See Attachment 2.

3. Tlue LPI iisiiuieulllii iliu isisateu iiuiii ieaetoi coo;'lit )IeSSUIre buUUary by 2 redundant check valves (3LP-i 7/i, 3Fr-14, 1e1.

6) and 2 redundant normally closed block valves (3RC-45 and 3RC-43, Ref. 12). l

Shiw Stn&Vbstehc CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER

). OR W.O NO. DIVISION & GROUP CALCULATION NO. l REV NO. WBS NO. PAGE 19 115524 STR S-006 1 800 OF20

4. A portion of the system in this area is actually the SSF backup to the RC pump seal injection and other piping that is normally isolated and not directly connected to the reactor coolant pressure boundary. The high pressure operating modes are less than 2%

A of the normal operating time; therefore the piping is considered moderate energy (Ref. 17).

5. There are no pipe breaks or jets postulated for RC cold leg per UFSAR Section 5 (Ref. 20), based on LBB criteria (Ref. 20 & 25).

See Attachment I & 3.

6. SF is a moderate energy system.

7. Secondary Shield Wall protects the emergency strainer from 12" decay heat (LP) line LOCA break and jet. See Attachment 3, (LPI Drop line section) and drawing Attachment 4.

8. The CF and LP drain lines (Valves 3CF-31 and 3LP-181) are isolated from reactor coolant pressure boundary by 3CF-14 and 3RC-45, 3RC-43.

9. The flow path of the LOCA jet of the 1.5" RCS drain lines (3A2 RC cold leg to 3RC-29 and 3B2 RC cold leg to 3RC-46) is towards the emergency strainer sump (See drawing Attachment 5). The distance however of the drain to the strainer is 25 fts which is greater than 200 pipe diameters (L/D). Based on NUREG CR 2913, Refd 18, the pressures for subcooled flow reduce significantly after L/D > 10 (see Attachment 6). Based on distance, the RCS drain line jet pressure will be insignificant for the A

emergency strainer.

SIw- Stone &VWbsher Ircr CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER D.OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO. WBS NO. PAGE 2C 115524 STR S-006 800 OF 20 7.0 (:ONCLUSIONS Based on this evaluation, the design function of the U3 emergency strainer will not be compromised by jet impingement or pipe whip by any of the lines in the vicinity of the Emergency sump. The reasons are:

1. HPI and SF are moderate energy lines, therefore there are no breaks postulated for the portion of these lines that are in the sump area.

2. Systems CS, LDW, FDW and CCW are not part of the primary coolant boundary. A pipe beak in these systems does not impair the operability of the strainer to function during a LOCA event.

3. There are no pipe breaks postulated in the area of the strainer for the RC cold leg, CF and LPI lines.

4. The RC drain line jet is not significant for the strainer based on distance > 200 LID.

5. The LPI instrument line is isolated with 4 isolation valves from the primary coolant. A

6. The LP and CF drain lines are isolated from primary coolant boundary by a check and block valves.

7. The LP decay heat line pipe break and jet are obstructed from the emergency strainer by a secondary shield wall.

8. The surrounding secondary shield walls, floors and major equipment (RPV, STG) protect the strainer from other Reactor Building LOCA breaks and jets.

Since the strainer is not directly impacted by jets or ruptured pipes, there is no need for additional jet shielding or pipe rupture restraints to protect the strainer.

Shaw- Stone &v\obstw, kh CALCULATION SHEET - ATTACHMENT 1 CALCULATION IDENTIFICATION NUMBER

0. OR W.O NO. I DIVISION & GROUP I CALCULATION NO. REV NO. WBS NO. FAGE 115524 STR S-006 1 800 OF RCS BREAK EVALUATION Excerpt from:

UFSAR Chapter 5 Oconee Nuclear Station Page M31 DEC 2003) 5.4 - 17 UFSAR Chapter 5 Oconee Nuclear Station 5.4.8.6 LOCA Restraints The original design of the Oconee Reactor Coolant System included LOCA restraints on the RCS hot leg and cc-ld leg piping and on the Reactor Coolant Pumps (RCPs). Their original design function was to limit the hot and cold leg RCS piping movement in the event of a guillotine break of the RCS piping.

The B&W Owners Group Topical Report BAW-1847, Revision 1 (September 1985), demonstrated, with a fracture mechanics evaluation of the RCS piping, that such postulated RCS piping breaks had an extremely low probability of occurrence. This fracture mechanics evaluation of the RCS piping is known as Leak-Before-Break (LBB).

The NRC approved the B&W Owners Group Topical Report in a Safety Evaluation Report dated December 12, 1985. This SER and the subsequent February 18, 1986 letter to Duke provided the NCR'.,

authorization for the implementation of LBB for the Oconee Reactor Coolant System. With the implementation of LBB, the RCS piping large break LOCA's are no longer required to be postulated for the dynamic effects on the RCS piping and components, thus eliminating the need for the RCS piping and RCP L.OCA restraints.

As a result of steam generator replacement for Oconee, some of the RCS piping LOCA restraints were modified or deleted. The LOCA restraint attached to the RCS hot leg elbow located at elevation 809' was partially deleted. The LOCA restraints attached to the RCS cold leg piping located at elevation 794' were compltely deleted. The LOCA restraints attached to the RC pumps were completely deleted. Reference Figure 5-29 and Figure 5-30.

S Stw &Webster; kr CALCULATION SHEET- ATTACHMENT 1 CALCULATION IDENTIFICATION NUMBER DIVISION&GROUP CALCULATION NO. l REVNO. WBS NO. PAGE STR S-006 1 800 OF OF, HOT LEG LOCA ' X I RESTRAINTS X PUMP LOCA RESTRAINT (DELETED DURING l STEAM GENERATOR REPLACEMENT) >

92

-PARTIALLY DELETED DURING STEAM GENERATOR REPLACEMENT)

Sh -stone & vmster k-r-CALCULATION SHEET- ATTACHMENT I Figure 5-30. Reactor Coolant System Arrangement - Plan (Typical)

Shaw -Stone &Vkbster, k-r CALCULATION SHEET- ATTACHMENT 1 CALCULATION IDENTIFICATION NUMBER

.0. OR W.O NO. DIVISION & GROUP CALCULATION NO. l REV NO. WBS NO. PA 115524 STR S-006 1 800 C 5.4.8.6.1 Replacement Steam Generator LOCA Analysis For th: replacement steam generator RCS structural analysis, there are ten high energy line breaks considered.

1. Single Main Steam Line Break

2. Double Main Steam Line Break

3. Single Main Feedwater Line Break

4. Double Main Feedwater Line Break

5. Surge Line Break at the Hot Leg Nozzle

6. Surge Line Break at the Pressurizer Nozzle

7. Surge Line Break at the Intermediate Surge Line Drain, North Direction Thrust

8. Surge Line Break at the intermediate Surge Line, East Direction Thrust

9. Decay Heat Line Break at the Hot Leg Nozzle

10. Core Flood Line Break For each of the ten breaks the replacement steam generator and primary piping whip restraints are considered inactive because the component displacements are not large enough to cause a contact between the restraint and the component. The restraints on the reactor coolant pumps were not included in the analysis because they are to be removed during the steam generator replacement.

Each of the above high energy line breaks were analyzed using the proprietary Framatome ANP computer program BWSPAN to calculate the loads incurred through out the reactor coolant system due to the effects ofjet impingement and asymmetric cavity pressure. All of the reactor coolant system piping, componerns and supports have been shown to be acceptable for the loading applied by each of the above high energy line breaks.

CALCULATION SHEET - ATTACHMENT 2 CALCULATION IDENTIFICATION NUMBER

0. OR W.O NO. I DIVISION & GROUP I CALCULATION NO. REV NO. WBS NO. PAGE 1 115524 STR S-006 1 800 OF 1 LP CF BREAK EVALUATION:

Excerpt from:

Spec. Spec. OSS-0254.00-00-1028 Date: November27, 1995 Rev. 23 Page 23, 24 Secti:)n:

2.3.7 PIPE RUPTURE/SUBCOMPARTMENT PRESSURIZATION The piping upstream of check valves LP-176, and LP-177, including the cross connect piping, has been evaluated for high energy line break interactions by employing Standard Review Plan (SR.P)

Mechanical Engineering Branch (MEB) 3-1. Employing MEB 3-1 allows the postulation of break and crack locations based on the stress levels in the piping system which assumed the piping to be at full operating pressure. Stress levels in the piping have been kept below the break and crack stress thresholds given in MEB 3-1 (See PAEM-0 10). Thus no breaks or cracks are postulated (upstream of the new check valves) (References 2.5.1.9.18, 2.5.1.9.19 and 2.5.2.1.31, 2.5.2.1.34). Also, adjacent pipirg susceptible to HELB, as identified in PAEM-0 10, was evaluated for impact to LPI piping.

Additionally, piping downstream of check valves LP-176, and LP-177, to the Core Flood Tani:

Nozzles, and up to but not including the Core Flood Nozzles on the Reactor Vessel, has been evaluated for high energy line break interactions by employing Leak Before Break (LBB) Technology per NUREG 1061 (References 2.5.1.9.18, 2.5.1.9.19 and 2.5.2.1.32, 2.5.2.1.34). The successfal LBB evaluation assures that plant leak detection systems can detect as 1 GPM leak before, including the time required to safely bring the plant to a safe state, the piping system could develop a catastrophic break (References 2.5.1.9.18 and 2.5.1.9.19). This allows the dynamic effects of a particular HELB to be neglected except for Alloy 600 weld materials due to recent Primary Water Stress Corrosion Crack.ing (PWSCC) concerns. HELBs, then are postulated at the Core Flood Reactor Vessel Nozzles which contain Alloy 600 weld materials (References 2.5.1.9.18 and 2.5.1.9.19). At the primary and secondary shield wall Core Flood Penetrations and along the vertical CF line near valves 1CF- 13 and 1CF- 14, rupture restraints and rigid guides are in place. No rupture restraints and rigid guides are required at the Core Flood Tank end of the piping (References 2.5.1.9.18 and 2.5.1.9.19). Duke has successfully argued to the NRC that LBB technology could be used at the Core Flood Tank Nozzles, which also contain Alloy 600 weld materials, based on the relative low temperature of the fluid at these locations. Onset of PWSCC is known to occur with temperatures greater than 500 degrees F.

Under normal operation, the Core Flood Tank Nozzles experience ambient Reactor Building temperatures (- 125 degrees F).

SCAIW Stone &WAbster, k-3 CALCULATION SHEET - ATTACHMENT 3 JION IDENTIFICATION NUMBER Excepts from PIP 0-04-07314, CA #5 (Attachment 3 is for Unit 2, however, Unit 2 is similar to Unit 3 arrangement) A CA Seq. No: 5 ltRes Grou s>-.'rSt~st it..'".' .,Orit.Groo 2*>+:i~od:4l'hc G>e Oc.

MCE Closed MCE 01 B3 R Propomed Corrective Action:

Perform and document review of break locations in Unit 2 RCS system piping (including ECCS attached piping, decay hcat drop line, surge line, :HPI injection lines, etc) which could create a jet impingement load on the proposed (modified) Reactor Building Emergency Sumpstrainer. If such locations areidentified, provide analysis ofjet loading for design inputtothesump modification. Results of this rcview are needed by 7/1/05 to support mod implementation. This CA has been discussed with Tim Brown of MCE/Civil.

Originated By: RL08372: OAKLEY, RUSSELL L Team: RJF2 111 Grogp: MCE Date: 05/05/2005 Si awreT'pe '.9 s ivi ; .  :' .a T.,Zr ': 'axe Nr Ready For Approval: RL08372 RJF21 II MCE 05/05/2005 Aprrovul A.ssigned To: RJF211 _ RJF21 I1 MCE 05/0512005 A orove1d3y; RJr21il _ RJF21 1 MCE 05/16/2005 General:Ovtage: N/A Mode: N/A Other TrackIng Processes Tvnc ;Numbcr Text Actual Corrective Action:

Priority I2b Actual CAC: B3 Status: Closed DucDate: 06123/2005 Last Updatoi By: TDB2719: 3ROWN, 1]MOTHYYD Team: PANNW4981 Group: MCE Date: 06/21/2005 PIP No: 0-04-07314 11/e1/2005 12:14 Pagc IZ:14 11/01/2005 I'age 66 PIPNo: 0-04-07314

SOW Stone &Vcmb, Icr CALCULATION SHEET - ATTACHMENT 3 CALCULATION IDENTIFICATION NUMBER OR W.O NO. l DIVISION & GROUP CALCULATION NO. REV NO. l WBS NO.

11;524 STR S-006 1 800 Problem Insvatlegaton Process Oconec Nucar Sl tiwn A ompreiernsive riew of potential high energy line break (HELB) interactions with the the new Reactor Building Emergercy Sump (RBES)

• cc ens to be tnsrallcd via design change 200049C has been completed. The review considered all high energy lines that could potentially damage thc screens and allow debris to flow ilirougli theicreens. Since the emrngcncy sunp is credited for ntiligation of loss ofcoolant type accidcnts only potential interactior.s with those high energy lines associatld with the Reactor Coolant system were evaluated.

Tho fbllowing high energy lina associated with the Reactor Coolant system were considered in this evalujtion:

Rctcsor Coolant (RC) Hot Leg to Steam Generalur(SO) 2A RC Hot Leg to SO 2B RC Cold Leg from SG 2A to Reactor Coolant Purmp (KCP) 2AI RC Cold Leg from SG 2A tn RCP 2A2 RC Cold Leg fromtSG 28 to RCP 2BI RC Cold Leg from SO 28 to RCP 2B2 RC Cold leg from RCP 2AI to Reactor Vessel (RV)

RC Cold Leg from RCP 2A2 to RV RC Cold leg from RCP 2BI to RV RC Cold Leg from RCP 2B2 to RV Hit h Pressure Intiction (HPI) Normn.l Makeup l.ine 2AI (fivn valve 2HP-1 27 to 2AI Reactor Coolant cold leg)

HPI (HPI) Normal Malkup l.ine 2A2 (from valve 2HP-126 to 2A2 Recctor Coolant cold leg)

IIPI (IIPI) Emergency Makeup Line 2BI (from valve 2HP.153 to 2RI Reactor Coolant cold leg)

HPI (HPI) Emergency Makeup Line 2132 (from valve 2HP-152 to 202 Reactor Coolant cold ICg)

HPI Letdown Line (to valves 2Ir-1 & 2HP-2)

Liquid Waste Disposal (LW D) line (to valve 2RC-24)

LWD line (to valve 21tC-29)

LWD line (to valve 2RC-43)

LWD line (to V3lve 2RC-46)

Low Presuaue lnjcxtion (LPlY Core Flood (CF) line toRV (frormvalve U21P-I 76tt CF-12 and RV, and from CF tank 2A)

LWPi /CF line to RV(frum valve 2LP-177 to CF-I14 and RV, and from CF tank 2B)

LP I line (frnmeReactor Buildingk (RB) penetration 16 to valvc 2LP-176)

L I line (from R13 penetration 15 in valve 21.P- 177)

U .Crossover Line LP Drop Line (fror RC 2A Hot Leg to valvc 2LP-2)

Doaon Dilution Line (from LPI Drop line to valve 2LP-103)

Prcasurizcr Surge Line Pressurizar Spray Line Those lines covered by Leak Before Break (LUdB) analysis were screncd out asnot applicable. An l.BB analysis csscatially confirrns that should a piping *ysntlm decvelop a leak. the leak will be detected and ar orderly shutdown of the plant will occur in time to prevent a castrophie rupture of tle pipaig. Thus the dynamic efects ofti potulutcd IIELB need not be considered for those lincs that posasan lBB analysis. 1.1Banalysis have bccn completed for tho RC hot and cold lega. The RC bhoand cold le LLB8 analysis iscontained inRAW- 1t47. Rev. I. entitlod ThMe&W Oners OroupLcak Bcfore trek Evaluationuf.N argins Against rull Break forRCS Primary PipingofB&W designed NSS. The NRC staf accjptcd thmeMW in a Sacty Evaluation Report (SER)dated l212/85 addresed to Mr. LC Oakes, Clairmai. B&WOwncrs Group Leak Bcfare Break Taisk Force.

In iddition to the RC hot and cold legs, the LPI / CF lines between the valvs 21.P-176 & 177. the 2A and 28 CF Tanks, and up to bt rot including thc CF RV r.olzes have also recived a LItS analysis. These analysis are documented in Oconoe calculation OSC-8523 Oconicc Nuclear Station.

Uns 1.2. & 3. Leak Before Break Analysis of the Core Flood and Low Pressure Injection IDacsy Heat Removal Piping Systcams Tte NRC stafT acctpted the results of the LPI ICF LBB analysis in SERs dated 9/29103 (Unit 1), 2/504 (Unit 2), and 9/2/04 (Unit 3) respectively.

As noted, the CF RV nozzles wore not included into Uie scope of the LPI /l: LBS analysis due to concerns regarding, Alloy 600 and Primary Water Stress Corrosion Cracking (PWSCC). In order to protect the LPI crTOSvSrn from a postulated break at die CF RV noi"l.c rupture restraints wcr inmiallod. These rupture restraint prevent pipe whip and jct impingement concerns from a postulatcd break at the CF RV n=71c. They alqo will protect the new RULES screens. Reference drawings 2.53A.1479A-PRI 000 through PR 1005 for the Unit 2 CF RV nonkles rupture restraints.

lh: SERs rlfcrencud above for the LPI I CF LBB analysis also approved the implcmicr.tation of Branch Technical Position Mechanical Engineerir.g Urnnch (!E13) 3-1 'Postulated Rupture Locations in Fluid Systemr Piping Inside and Outside Containment' for portions of the Unit 2 LI' system inside wctainmnnet (rromn penetration 16to vatve 21.r-176, ao torn penetration I5 tn valve 2LP-177, including fse LPI crossover line). The imnrlenenwtionoWMERt 3.1 allowed h ueeof tiess rtiteriato postulate break and crack locstions. lhe piping stress and supporttfessrzint design orfhe Unit 2 LIr syste m is audi that the strcss criteria lor breaks and cracks isnet exceeded and thus no btr.es or cracks weepostaultod in this portion of the LPI sys:cm. The unit 2 piping stress analysis is deocuennted in calculation OSC-8228. 'Piping Ana!ysis ror Low Pressure Injction Systcm. Probbkm 2-53-22.-

Thus the following lines are exempt from the evaluation due to the existence of an LOUanalysis:

IA0MI245 12:14 Page 7 PIP No: 0-04-07JI4

Shaw' Stone &V\/ster, hc CALCULATION SHEET - ATTACHMENT 3 CALCULATION IDENTIFICATION NUMBER OR W.O NO. DIVISION & GROUP I CALCULATION NO. REV NO. WBS NO.

115524 STR S-006 1 800 roem Int tigatialon Process Oconee Ax~learSwieft Reacor Coolant (RC) Hot leg to Steam Gncerxwr(SG) 2A RC llot Leg to SG 2B RC Cold Leg from SG 2A to Reactor Coolant pump (RCP) 2AI RC Cold Lei from SG 2A to RCP 2A2 RC C'ld Leg from SG 20 to RCP 2B I RC told Leg fronm SC 2B to RCP 2B2 RC Cold Leg tram RCP 2AI to Reactor Vessel (KV)

RC Cold Leg from RCP 2A2 to RV RC Cold Leg from RCP 281 to RV RC Cold Leg from RCP 2B2 to RV Low PNu-ure Injection (LPIY Core Flood (CI) line to RV (fron valve 2LP-176 to CF-12 and RV, and from CF tank 2A)

IPI I CF tbtc to RV (from valve 2LP-177 to CI:14 and RV. nd from CUtank 2B)

Tnhe lllowing lines are exempt from the evaluation due to the implementation orML1 3.1 to eliminate break a-id crack locations LPI I nc (from Reactor Building (RB) penetration 16 to valve 2LP-176)

I.PI I nc(from RB penctration I5 to v!vc 2LP-177)

I ll crossover lrinc The rcmnainding lina arc evaluatcd below. Note tb. ie Unil 2 RULIS isloceu north ofthe RV with the floor of the sump at elvation 774'-0".

fhe p orttie new screens are at elevation 79n'-7.

• '+lligh Pressure Injection (11P1) Normal Makeup Line 2AI (ftrom va!ve 2HP-127 to ZAI Reactor Coolant cold Ieg)'",^

This :1-1/2pipe exitsofTthe2AI cold let at Sl 1-3"(See drawing O-147')A). Ibispiping ii allloutheast of the RV. Me pipe location prohibits interietion with the RBES screen components located on the north side of the RV.

"'lIPI (ll'l) Normal Mikeup line 2A2 (from valve 2HP-126 to 2A2 Reactor Coolant cold lc;)'***

Ih eipe run eonsists of a short straight pkece In the northeatt quandrant nfthe Reactor Buildin- (Rl) at elevation 8E1'.0' (SCedrawing 0-1479A).

Its clovatiVn. size (2.1/2). amid geometry prohibit interaction with the R1ES screen components.

• "IIPI (tPI) Emergency Makeup Line 2F1 (fimm valve 2HP-1 53 to 2BI Reactor Coolant cold leg)'

All ttepipe is in the aooutbest qadrant of dte RB aleevation 811'-3" (Seedrawing O-1479A). ILSlocaton prohibits interaction with thelWIS serec components.

• "'111i'l (HPI) Enragcncy Makeup Line 2R2 (firom valve 2HP-152 to 282 Reactor Coolant cold ler)0'4 This shor nrun of 2-.1' pipe is in the nothwest quandrant of the RB attelvation giI '-3" (Sec drawing O-1479A). lIhe pipe ize (2-1/2). elevntion.

atidg ometry prohbbit interaction withthe RBES scrn cmponns.

• 0"I1PILetdown Lint (to valves 2HP-l & 2HP-2)*"

lTe 1 tdown linecexits ofthe RB drain heaider. vhich in tumsexits otTthc 2BI cold le. (Sct draning O-1478A). All of this piping islocated wet and southwest of the RV. Its location prohibits inteaction with the RBES screen cormporents.

""Liquid Waste Disposal (LWD) line (to valve 2RC-24)"-'

This L.%%' line exits oufthe 2AI cald Ieg (See drtwing O-147SA). All ofthis piping is located in tho southeast quandnutt ofthe RB. Its lcation prohbnts interaction with the RIoES screen cornponents.

'"'1W\VD line (to valve 2RC-29)""

All othis piping islocated northeast oftho RV offthe 2A2 cold leg. The 2A hot leg blocks any potential breaks from approaching the RRFS

,ereei components.

"0*1"WD Ine (to valve 2RC-t3)"-'

All oi'this piping Ihlocated southwest of the RV off the 2131 eo!d leg. Its location prohibits interaction with the RR ES scroer etomponents.

"-I.WD line (to valve 2RC.46)'"'

11/91/2005 1214 Page 8 PIP No: 04-04.7314

Shaw -Stone& wVbsterInia CALCULATION SHEET - ATTACHMENT 3 CALCULATION IDENTIFICATION NUMBER OR W.O NO. DIVISION & GROUP CALCULATION NO. REV NO. WBS NO.

115524 STR S-006 1 800 Prolcen Iavesfigaslln Process Ocanee Nuclear Stadons lbis ,iping is located northwcstofttheRVotfic22 cold lcg. Thc2B hot lcghlocks any pntential hrckc from approaching thcRBES scrccn components.

J"JYLDrop Line (from RC 2A Hot Lcg to valve 2LP-2)+ +

Ihis piping at valvce2LP-2 is closest to thc RiBES. Since valves 2LP-1 and 2tP-2 arc nornatly closed (Rcf. Calc. OSC-8385), and the piping dowrstream oufthe valves in nolnorrally hsservicedurirngpoweroperations. there iv no rcvcrsc low whiporjdt. At a postulated brak location at the lbow at elevation 795-3' (ree O 147gtD Section G-CG). he forward whip would pivot about the uppercelbvw at elevation 8G1-6'6propellir.g thopping south.ti:rtherinsidethc2A cavity, awayfromthcRBES. tlowever thcjctwouldbetdirectedtowardtheRlBES. ThepipesizCofthc drop line Is 12'. The normal operating temperaturc and pressure ofthe drop line tre the same as those for the RC hot leg. The normal operating cond tions for the hot leg are approximately 600 degrees F and 2200 psig. thbis equates to apnroximately 50 degrees F subeooling. With thesa partecra known, thcjet length can be tetcrynint from NUREC / CR-2913, 'Two Ph JLt Loads'. For subcooling or27 dcgrets ror greater thc jct Icngth isdetined by NUREG / CR-2913 as 10 times ie diameter of the pipe or in this case. 10'. nThradius of tlejet islikcwise derind as 4.5timcstliccdianmtcror4l.5'.

As noxted above. the cl.vation ofthc drop line at the postulated break location is79S'-3'. The location oftthc top ofthe screens is 780'-7' (lIhe information for the top of screen clevation was given on an unofficial manufacturers5 drawng E-A2-103.129.596). bhis equates to adifterence of 14-S. Thcpostulated break location islocated 17'-9 Eastofthc centeriine ofthe RV. hIecenterline ofthe RilES coincides with thecer.terlinc of the RV. The RHF incasurcs I fRet in length (E-W) direction, and extends 9' East ofthe centerlin ofthe RV (See drawing 0-1067-A Sections A-A. and B-B). This equates to a ditfercico of S'-9 from the postulated break location and the edge of the sump in the Etst-West direction. In the North-South direction, the postulated break location is approximately 7-0" south of the edge ofthe sump. Thus the distance from the postulated brnal location from the cdgc of thc sump screen is the suart root of the sum of the .quares of he individual measurementt. or approximately 1S-6".

I1Fhis Jistancc exceeds both thejdt lenigth and jet rdlius, and thuc no interaction will occur with the RBES screen components.

4 1ttoron Dilution Line (rrom T.PI Dmrp line to valvc2LP-1 03)**.*

Ihe Uoron dilution line is a 3- line connected to the drop lincjust vertically up ((o elevation 797'.9 --se 0-1478D Section 0-Cl) from the postulated break location evaluated fOr the drop line. The small size oftic line and its location faither away frnm the sump prohibits ir.teraction with :he K3ES screen components.

l0rscsurizer Surge Line* **

thc tI" 'Prcarizcr Surge l.ine ishcatc offthc 283hot Icg (I dcv S03t-3' - see 0-1479A). west ofthe RV, and extends aouth to Pressurizer (soutiwevt quandrant). IU location prevent intcraction with the RBES screen components.

• • l'ressurizeir Spray Line"4 The .-1/2" Pressurizer Spray Line exits off the 21B I cold leg in the southwest quandrant orthc RB. Its location relative to the sump prevents intertetion with the sunsp screen components.

To cc nclude. therc arc no brach lines offthe RCS, shoul thcy cxperience n HELB, that cculd intcract with thc new RBES screns. Th RCS and CF lhes wercexeoptled due to theexiscence of LtB analystsi and the LPI lines, including the r.mcrossover lin, have tress low cnough to tbe bclovw the break and rrack trsholds given It MEEt 3-1 such that no breaks need be postulated.

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Patsy J. Earnhardt Project Manager Oconee Nuclear Station 864-885-3423 Timothy D Brown'GenlDukePower To Bruce H Justicc/Gcn/DukcPowcr@DukePowcr 07/26/2005 10:10 AM cc Patsy I Earnhardt/GcrlJDukcPowcr@DukcPowcr, George K Mc Aninch/Gcn/DukcPowcr@DukcPowcr Subject Re: Fw: U2 Emergency Sump - Review of S&W Jet Impingement Calc

Bruce, I have not finished reviewing the SAW calculation, however, I thought I might respond to your latest email regarding the potential interaction from piping containing 2RC-29 & 46 with the emergency sump The original justification in PIP 04-7314 CA-05 was for upstream of the valves not downstream. Normally the valves 2RC-29 & 46 are closed, thus the piping downstream of these valves is not pressurized to RCS pressure and thus does not qualify as a high energy line. Also, since this downstream piping is isolated from the RCS, then a break in this piping would not be considered as a LOCA and thus protecting the sump from these breaks would not be required.

Regarding the piping upstream of the two valves, the piping is normally at RCS cold leg conditions, i.e. 557 degrees F and 2150 psig. At these conditions, the RCS fluid has a subcooling of approximately 90 degrees F. Using NUREG CR 2913, for piping with subcooling above 27 degrees F, the length of the jet from a break is 10 x De, where De is the inside diameter of the pipe. The piping in question is 1-1/2" sch 160. The inside diameter for this piping is 1.338". 10 times 1.338" is 13.38 inches. The location of the sump is a little less than 20 feet from these drain lines. Upon comparison, it is obvious that postulated jets from the lines upstream of RC-29 & 46 will not impact the emergency sump.

If you have any questions, please let me know.

Timothy D. Brown HELB Project Manager Design Basis Group Oconee Nuclear Station Duke Energy 864-885-3952

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