ML17277A868

Forwards Design Reverification Program, Vols 1 & 2,final Assessment Rept.Results of Program Will Be Presented to NRC in Late Oct 1983

ML17277A868
Person / Time
Site:	Columbia
Issue date:	09/27/1983
From:	Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML17277A869	List: ML17277A868 ML17277A870
References
NUDOCS 8310070342
Download: ML17277A868 (340)

Text

REGULATORY:1ORNATION DISTRIBUTION SYSQI (RIDE)

DOC ~ DATE: 83/09/27 NOTARIZED: NO

'"AOCE'SS ION NBR; 8310070302 DOCKET FAGIL!50~397 NPPSS Nuclear= Projects Unit 2i Nashin'gton IPublic Powe 05000397

'AUTH ~ NAME AUTHOR AFFILIATION

! SORENSENgGB'0 ~ Nashington Public 'Power. SUpply System

RECIP",NAME RECIPIENT AFFIL'IATION DENTONiH ~ RE Office of Nuclear Reactor Regulationi Director>>

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SUBJECT:

Forwards "Des> eve ication Programs," Vols 1 8 2"<final assessment rept, Results- of program will be, presented to ARC in late Oct 1983, DISTRIBUTION,CODE: B001S iCOPIES iRECEIVED:LTR,J ',.ENCL, [ SIZE:. '

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.TITLE: Licensing "Submittali PSAR/FSAR Amdts--8 Related Correspondence'OTES:

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,NRR/DE/EQB 13 2 NRR/DE/GB 28 2 NRR/DE/MEB 18 1 NRR/DE/MTEB 17 1 NRR/DE/SAB 24 ,1 ~

NRR/DE/SGEB 25 NRR/DHFS/HFEBOO 1 NRR/OHFS/L'QB 1 NRR/DHFS/PSRB 1 32'RR/DL/SSPB 1

NRR/DSI/AEB '26 1 ~ NRR/DSI/ASB 1

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Washington Public Power Supply System'.O.

Box 96B 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 September 27, 1983 Mr. Harrold R. Denton, Director Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

WNP-2 Design Reverification Program

References:

a) Letter, G.D. Bouchey to H.R. Denton, "Nuclear Project No. 2 - Verification of Design and Construction Adequacy,"

dated October 22, 1982.

b) Letter, R. L. Ferguson to W.J. Dircks, "WNP-2 Plant Verification Program for WNP-2," dated November 24, 1982.

c) Letter, H.R. Denton to R.L. Ferguson, "Design Verifica-tion Program for WNP-2," dated December 28, 1982.

d) Letter, G.D. Bouchey to A. Schwencer, "Nuclear Project No. 2 - qualification of Engineers Assigned to the WNP-2 Reverification Reviews," dated January 13, 1983.

References (a) and (b) described the Supply System programs for assuring that WNP-2 is designed and constructed in accordance with our commitments. One element of that overall program was an in-depth design reverification review of three reactor systems to provide added assurance of WNP-2 design 'adequacy.

Reference (c) indicated your acceptance of the program proposed by, the Supply System and requested additional information regarding the qualifications and independence of the engineers assigned to perform the design reviews. Refer-ence (d) supplied the requested resumes and independence certifications.

Enclosed are copies of the final assessment report which provides the results of the WNP-2 Design Reverification Program. A meeting is being scheduled with NRC staff in late October, 1983, to present the results of the program.

If questions arise regarding the WNP-2 Design Reverifi'cation Program, you.

may contact Dr. G. D. Bouchey, (509)372-5359.

+e G. C. Sorensen, Acting Manager Nuclear Safety and Regulatory Programs GDB:awh oo<

Distribution attached 830927 83l0070342 05000397 PDR ADOCK A

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DISTRIBUTION USNRC INTERNAL DISTRIBUTION T. NOVAK CS CARLISLE 982A A. SCHWENCER LT HARROLD 982A R. AULUCK ,BA HOLMBERG 994E J. MARTIN (Region V) J YATABE 410 A. TOTH DL WHITCOMB 420 DM BOSI 410 JD MARTIN 927M:

TAA 5 CONSULTANTS PL POWELL 956B RV LANEY WW WADDEL 400 LH RODDIS, JR SI STEVENS 750 HE SflEETS Docket fi,le 956B.

FB JEWETT, JR kt/file 99.4E S. LEVY PL 2/1 b 956B CQ MILLER GCS/lb 340 GDB/lb 387 JR HONEKAMP sf (2)

WNP-2 files 917Y FINDINGS REVIEW COMMITTEE EXTERNAL DISTRIBUTION RJ BARBEE 927M AJ FORREST BSR-RO JG TELLEFSON 901D WG CONN - BAR-RO CH McGILTON 956B F. McCLEAN - GE LC OAKES 823 WS CHINN, BPA 399 AG HOSLER . 956B JR LEWIS, BPA 399 JH BAKER 956B NS REYNOLDS - D&L NS PORTER 387

0 WASHINGTON NUCLEAR PLANT 2 DESIGN REVERIFICATION PROGRAM Volume II:

Appendices to Final Assessment Report September 1983 Washington Public Power Supply System Richland, Washington 99352 DOCI,et'-3'l~

Control @88~~" ~""

Date of Docomeub REG~TDR DDCKf M

APPENDIX 1 WNP-2 Requirements and Design Reverification Final Assessment Report List of Potential Finding Reports

DIX 1 LIST OF POT . FINDING REPORTS (Page 1 of 13)

PFR No. Classification** Review Area* Descri tion HPCS-1 3.1 The MR criteria document does not include requirements for all design input areas identified on the requirements reverification checklist.

HPCS-2 3.2.3.1. D The equipment piece number for diesel engine cooling water heat exchanger is not consistent on all drawings.

HPCS-3 3.2.3.6.A The diesel air start system is not totally redundant as described on the Flow Diagram.

HPCS-4 3.2.3.10.A Current calculation revisions were not used as the basis for subsequent calculations.

HPCS-5 3.2.3.10.B MR and alternate calculations do not agree on the diesel exhaust pressure drop.

HPCS-6 3.1 Cold working of instrument tubing.

HPCS-7 3.2.3.2 Detail 8 showing HPCS Instrumentation is missing from Flow Diagram.

HPCS-8 3.2.3.5.B HPCS/RCIC condensate storage level instrumentation separation is questioned.

HPCS-9 3.1 FSAR does not state the correct ASHE Code Classification for the HPCS diesel cooling water heat excnanger.

HPCS-10 3.1 FSAR states that all fuel oil piping is ASME III whereas some is 831.1.

HPCS-11 3.2.3.6.B Calculations that justify condensate storage level transfer setpoint not found.

HPCS-12 3.1 FSAR does not state the piping material requirements specified in the ECD.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Valid

LIST OF POTENTIAL FINDING REPORTS (Page 2 of 13)

o. assi ica ion eview Area* Oescri tion HPCS-13 3.1 Different sections of the engineering criteria document do not agree on piping corrosion allowance.

HPCS-14 X 3.1 FSAR does not agree with ASNE piping code effective date specified in tne ECD.

HPCS-15 X 3.2.3.6.A 85R calculation on emergency water volume for HPCS pump suction is inconsistent with other calculations and the design events.

HPCS-16 3.2.3.7.A HPCS relief valve design does not incorporate GE design specifications for double flange gaskets.

HPCS-17 3.2.4.3 The DSA diesel engine exhaust system line size does not correspond to manufacturers recommendations.

HPCS-18 3.2.4.3 The diesel fuel oil system does not meet NFPA Std. 37 requirements.

HPCS-19 3.2.4.3 No air box drain collection tank is provided for the HPCS diesel.

HPCS-20 3.1 Design requirement documents and FSAR values for vital piping damping coefficient do not agree.

HPCS-21 3.2.6.4.8 There is clearance between the attached parts of two snubbers where gaps are not allowed.

HPCS-22 3.2.5.3 No design calculations traceable to the HPCS pump support anchor bolts were found.

HPCS-23 3.2.5.1. D Design procedures covering aspects of tne Instrumentation Installation Contractor design process were considered inadequate.

HPCS-'24 3.2.5.1.0 Improper stress intensification factors were used in the analysis of PI Line X-73a.

HPCS-25 3.2.5.1. D Evaluation of local stresses caused by weld attachments for Pl Line X-73a was considered to be inadequate.

• • . Finding *Corresponds to Report Se .on Number O'- Ooservation

LIST OF POTE AL FINDING REPORTS (P of 13)

P R No. ass>>cat>on " evsew Area+ Descr> tron F NV HPCS-26 3.2.5.1.0 No faulted conditions stress evaluation was found for PI Line X-73a.

HPCS-27 3.2.5.1.A Loads used in the design of pipe supports for N2UO-2 piping system are not current.

HPCS-28 3.2.6.4.A Potential restraint io thermal expansion of PI Line X-73a was identified.

HPCS-29 3.2.4.5 HPCS-FE-7 is installed with pressure taps and attached instruments located at the top rather tnan horizontally as suggested by good engineering practice.

HPCS-30 3.2.3.4 There are ambiguities in the piping code specifications for the CST to HPCS pump suction piping.

HPCS-31 3.2.3.5.8 Discrepancies in separation criteria were noted in the BRI documentation.

HPCS~32 3.2.4.6 GE specifications for instrument setpoint, accuracy, drift and range are not consistent.

HPCS-33 3.2.4.4 Instrument tubing match line elevations disagree between two isometric drawings.

HPCS-34 3.2.4.5 There is a discrepancy between the flange bore and pipe ID for HPCS-FE-7.

HPCS-35 3.2.4.6 The nameplate and ranges specified in the instrument data sheet for OPIS-9 do not agree.

HPCS-36 3.2.6.2 There is a discrepancy between GE and BRI recommendations upstream and downstream straight pipe run for orifice flowmeters.

HPCS-37 3.2.6.2.C Discrepancies between the GE and BRI requirements for impulse line slope and instrument elevation are noted.

HPCS-38 3.2.6.2.B HPCS-LS-2A was tagged with a tag identifying the level switch as HPCS-LS-28.

HPCS-39 3.2.3.7.B The instrument line for the suppression pool level switch is not orificed to provide containment isolation per RG 1.11.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Valid

LIST OF POTENTIAL FINDING REPORTS (Page 4 of 13)

No ~ assi 1 cation evlew Area* Oescr i tion HPCS-40 3.2.6.2 The specified and nameplate ranges of HPCS-PS-12 do not agree.

HPCS-41 3.2.3.2 The valve interlock control function for HPCS-LS-2A is correctly shown in the GE specifications and FCO but not shown on tne GE PAID or BRI Flow Diagram.

HPCS-42 3.2.3.4 The seismic classification of HPCS suction piping from the CST is incorrect.

HPCS-43 3.2.4.3 There are discrepancies in the BRI calculations which sized restrictive orifice HPCS-R0-4.

HPCS-44 3.2.3. 1.C The calculated pressure drop for HPCS diesel starting air system exceeds manufacturers recommendations.

HPCS-45 3.2.3.4 There are ambiguities in FSAR Table "3.2-1 on code class groups for the HPCS system.

HPCS-46 3.2.4.7 The adjustable range for Breaker 4-41 short circuit tripping does not meet the GE specification.

HPCS-47 3.2.4.7 The relay element connected to Breaker 4-41 does not permit proper coordination.

HPCS-48 3.2.3.3 As-built data was not used in BRI voltage drop calculation 2.06.03 for TR4-41.

HPCS-49 3.2.4.7 Ground fault alarm relays on Bus SN-4 will not function reliably.

HPCS-50 3.2.3.3 The effect of simultaneous starting of 480V and 4KV motors was not considered in BRI Voltage Drop Calculation 2.06.03, Rev. 5.

HPCS-51 3.2.3.6.C The present design does not include the required degraded voltage protection and auto return to standby.

HPCS-52 3.2.3.3 The vendor print file for TR 4-41 contains two contradicting drawings.

HPCS-53 '.2.3.3 No fault duty calculation was provided for NC-4A.

• • . rin4ing *Corresponds to Report Se ,on Number 0 - Observation

LIST OF POTE AL FINDING REPORTS (P of 13)

o. ass> >ca son evsew rea Description HPCS-54 3.2.6.3 One of the bolts is missing from the HPCS pump grounding lug connection.

HPCS-55 3.2.3.7.A There is an equipment piece number discrepancy netween FSAR Table 6.2-16 and BKR Orawing t620 for several valves.

HPCS-56 3.2.5.2.B Local pipe stress from a welded attachment lug for pipe support 910-N was not calculated adequately.

HPCS-57 3.2.5.2.A Miscellaneous errors exist in the design calculation for pipe support HPCS-66.

HPCS-58 3.2.5.1.8 There is an error in the piping design guide.

HPCS-59 3.2.5.1. 8 Calculation 8.14.64A does not correctly calculate the functional capability stress of tne piping system.

HPCS-60 3.2.5.1. B The pipe crack evaluation appears to be incomplete for BRI Calculation 8.14.64A.

HPCS-61 3.2.5.I.B Tne displacement summaries for branch pipe connections do not include rotations.

HPCS-62 3.2.5.1. B The load data source for chugging, SRY and LOCA jet direct loads are not referenced in BRI calculation 8.14.64A.

HPCS-63 3.2.5.1.8 Tnere are documentation problems with seismic analysis input calculations for BRI Calculation 8.14.64A.

HPCS-64 3.2.5.1.B Improper revisions were made to support load tables in BRI Calculation

8. 14.64A.

-

HPCS-65 3.2.5.1.B Tne thermal displacements at branch connections were not correctly sumnarized.

HPCS-66 3.2.5.1.B Support design loads were incorrectly reported for HPCS-910N in BRI Calculation 8.14.64A.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NY - Not Yalid

LIST OF POTENTIAL FINDING REPORTS (Page 6 of 13}

o. ass1 lcat1on"* Review Area* Oescri tion HPCS-67 3.2.5.1.B Errors were found in revised thermal expansion computer runs in BRI Calculation 8.14.64A.

HPCS-68 3.2.5.1. B Some stress intensification factors were not included in the stress analysis in BRI Calculation 8.14.64A.

HPCS-69 3.2.5.1.B An incorrect mass was used in the computer model of valve HPCS-V-15.

HPCS-70 3.2.5.1. B 1'he physical properties of HPCS-V-15 used in the computer model did not come from tne referenced drawings.

HP CS-71 3.2.5.1.B Various errors were made in the thermal expansion analysis in BRI Calculation

8. 14. 64A.

HPCS-72 3.2.5. 1.B Emergency condition temperatures were not considered in the thermal expansion analysis in BRI Calculation 8. 14.64A.

HPCS-73 N.A. This number was not used.

HPCS-74 3.2.5.1.A Valve nozzle end loads and accelerations are not evaluated per requirements of the ECO.

HPCS-75 N.A. This number was not used.

HPCS-76 N.A. This number was not used.

HPCS-77 X 3.2.5.1.A The SSE response spectra for mass point 40 (BRI Calculation 8.14.82) is not included in referenced document.

HPCS-78 3.2.5.1.A The stress index, C2, used for the 3/4" elbowlet is lower than that required by ASrK Section III.

HPCS-79 3.2.5. 1.A An additional weight of 1047 pounds was added to the 12" HPCS-V-5.

HPCS-80 3.2.5.1.A HPCS-V-76 was modeled using a weight 400 pounds less than the drawings indicate.

• • ~inding *Corresponds to Report S n Number 0 - Observation

-

LIST OF POTE AL FINDING REPORTS (P of 13)

No. ass> cat>on evsew Area Descri tion HPCS-81 3.2.5.1.A Incorrect scales were used for ADLPIPE response spectra input.

HPCS-82 3.2.5.2.D Thermal loads used for design of HPCS-52 do not match those in the applicable pipe calculation.

HPCS-83 3.2.5.1.C Elbow dimensions used in the analysis of small bore line DE-1738-1 are in error.

RFW-1 3.4.6.3 RFW-TE-41A had been improperly terminated in the field.

RFM-2 3.4.4.3.8 RFW line "A" temperature element installed orientation does not correspond to orientation shown on pipe isometric.

RFW-3 3.4.6.3 The signal .cable for RFW-TE-41A was incorrectly labeled.

RFW-4 3.4.4.3.D The wrong type of flow element was selected for RFW-FE-15.

RFM-5 3.4.4.1 RFW-V-32A was not specified to oe testable with low pressure air as required by 10CFR50 Appendix J.

RFW-6 3.4.4.1. 8 The feedwater heater relief valve capacity is not sufficient to provide relief for all hypothetical events.

RFW-7 3.4.4.2.A Motor operator for RFW-V-65 is supplied with Class lE power per PED 218-E-2858 but Drawing E-528, Sheet 27 has not been updated.

RFW-8 3.4.6.3 The air operator extension shaft of RFW-V-32A interferes with RWCU inlet line to header "A".

RFM-9 3.4.4.3.C Inconsistencies are noted on the elementary and other electrical drawings for RFW-V-32A.

RFM-10 3.4.4.3.0 Upstream straight piping section length for RFM-FE-1A is insconsistent with ECO requirements.

RFW-11 3.4.4.3.0 Downstream straight piping length requirements for RFW-FE-1A is inconsistent with the ECO.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Valid

LIST OF POTENTIAL FINDING REPORTS (Page 8 of 13) 0~ ass 1 1ca loll ev1ew rea eSCr1 t1On RFW-12 3.4.4.3 Connecting pipe size and pressure loss documentation inconsistencies are noted for RFM-FE-lA.

RFW>>13 3.4.4.3 RFM-FE-lA is not installed as shown on GE draQings.

RFW-14 3.4.4.3 System flushing and protection screening for RFM-FE-lA is not installed.

RFM-15 3.4.6.3 The RFM-FE-1A pressure tap configuration and connections are not installed per manufacturers recollmendations.

RFM-16 3.4.4.3. D RFW-FE-1A calibration curve anomolies.

RFM-17 3.4.6.3 RFM-DPT-803A signal loop wiring and instrument rack tubing runs are not labeled in accordance with contractor requirements.

RFW-18 3.4.3.4 Documentation inconsistencies were found in the review of RFW-V-32A containment isolation requirements.

RFM-19 3.4.3.4 Loss of signal lock-up interlocks for RFW-DT-lA, DT-18 and-FCV-10 have not been implemented in accordance with GE recommendations.

RFM-20 3.4.3.4 The BRI elementary diagram does not show the required interlock between V-1128 and DPS-4.

RFW-21 3.4.4.1. 8 Control valve cavitation problems exist witn some valves.

RFM-22 3.4.4.1.8 There are inconsistencies and design input errors in the sizing calculation for RFW-FCV-15.

NL-1 3.4.5.3 Vendor approved nozzle loads did not include flange deadweights for RFM-p-1A and 18.

RHR-1 3.3.4.3.C All required cable types were not listed in Class lE P

list.

RHR-2 3.3.3.1.C The BRI wiring design for several RHR valves did not follow GE requirements.

~* finding *Corresponds to Report S .~n Number 0 - Observation

- Ya3id

LIST OF POTE L FINDING REPORTS (P of 13)-

o. ass>>cat>on** Review Area* Descri tion RHR-3 3.3.3.5 Containment isolation valve limit switches prematurely indicate valve closure.

RHR-4 3.1 FSAR incorrectly states that seismic reevaluation is supplemented by NUREG-0800. I RHR-5 3.1 No design requirement was found to match FSAR comnitment for vertical cable tray run fire breaks.

RHR-6 3.3.3.4 RHR-FC -64B was not included in the remote shutdown system design as required by specification 22A3085.

RHR-7 3.3.3.4 Remote shutdown system design specification 22A3085, Para. 4.1.1 is not met in that a new common point was created.

RHR-8 3.3.3.3 BRI drawing E503-8, Rev. 23 shows RHR-P-3 in Division B instead of Division 2.

RHR-9 3.3.3. 1.C The GE documentation for RHR-V-38 tnrottling are contradictory.

RHR-10 3.3.3.1.D ~ The second level undervoltage relays will cause bypass of the 115 kV source and will lockout the shed ESF loads.

RHR-11 3.3.3.1.D Feeder loads for HC-7BB and 7BA are missing from the NC-78 load calculation.

RHR-12 3.3.3.1. D Feeder circuit breaker for MC-7BB may be set too low.

RHR-13 3 '-4.1.A There is a discrepancy in the RHR-FCV-64 operating time specifications.

RHR-14 3.3.4.2.A RHR-F IS-108 is overranged.

RHR-15 3.3.3.1.D V-4B is missing from Drawing E528-36; V47B is missing from E528-37. Fuse and thermal overload sizes are not included on the E-528 drawing for RHR-V-4B and RHR-V-47B.

RHR-16 3.3.4.3.C The voltage drop from E-SL-81 to NC-BBB is larger than the 3X recomnended by BRI criteria.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Valid

LIST OF POTENTIAL FINOING REPORTS (Page 10 of 13)

P R No. C ass >cat>on " ev>ew Area* Oescrs t>on RHR-17 3.3.4.2.A RHR-FT-1 impulse lines are not routed as shown with the flow diagram.

RHR-18 X 3.3.4.2.8 The documentation (GE) for. RHR-FI-5 does not agree with the installed instrument indicating scales.

RHR-19 3.3.4.B RHR-MO-24B and 64B were ordered specifying the wrong environmental class.

RHR-20 3.3.4.1.C A cavitation check was not included in BRI Calculation 5.17.13 for RHR-R0-18.

RHR-21 3.3.4.1.C A cavitation check was not included in BRI Calculation 5.17.26 for RHR-R0-3B.

RHR-22 3.3.4.3.C Cable 2MBBA-20 is not sized for derated conditions.

RHR-23 3.3.5.2.B Heat exchanger drawings do not match the calculations.

RHR-24 3.3.5.2.B Heat exchanger installation does not reflect the calculation and installation specification requirements.

RHR-25 3.3.5.2.8 Oue to increased loadings, the anchor bolt analysis is incomplete.

RHR-26 3.3.5.2.A The original calculations were not updated or referenced to supporting calcu 1 ations.

RHR-27 3.3.5.2.A A buckling analysis was not performed as required by design criteria.

RHR-28 3.3.5.2.A The anchor bolt analysis for the upper lateral supports is incomplete.

~ RHR-29 3.3.5.2.A Assumed future (design) hanger loads must be verified against the actual hanger loads.

RHR-30 3.3.4.3.A Motor starters and TR-8-81 are subjected to over voltages (SM-8 side of the 480 V system).

RHR-31 3.3.4.3.B Oocumentation discrepancies for the fuse and overload heater sizes for three valves were noted.

RH N.A. To be inclu 'n Pipe and Support Addendum.

• • inding *Corresponds to Report Se ~n Number 0 - Observation

LIST OF POTE AL FINDING REPORTS (P of 13)

o. ass~ scat>on evsew Area* Descri tion RHR-33 3.3.6 Lugs on the heat exchanger are not shimmed per the GE specifications.

RHR-34 N.A. This number was not used.

RHR-35 3.3.4.3.A Fuse/circuit breaker coordination information is missing.

3.5.5.2 HPCS-HO-4 is not listed in QID file identified on the Class lE list.

3.5.5.2 The QID file referenced for HPCS-RO-4 did not contain the required design certification documentation.

EQ-3 3.5.5.1 QID file for HPCS-42-4A7C does not include required qualification data.

EQ-4 3.5.5.2 There is no in-situ pull/deflection operability test record-for valve RHR-FCV-64B in the QID file.

EQ-5 N.A. Number not used.

EQ-6 N.A. Number not used.

EQ-7 3.5.5.6 Confirmation is required for existence of low pressure isolation alarm and procedure to isolate auxi liary steam system.

EQ-8 N.A. Number not used.

EQ-9 3.5.5.2 The dynamic qualification levels identified in the QID file for HPCS-LS-2A are less than the required inputs.

EQ-10 3.5.5.6 Computer runs for the HVAC cooldown phase of HELB environments are not

'documented in the calculation file.

EQ-11 3.5.5.6 EQ environment calculation predicts peak pressures across RNCU heat exchanger room (R510) walls exceeding FSAR design values.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Valid,

LIST OF POTENTIAL F INOING REPORTS

{Page 12 of 13)

P R No. C ass> icat>on ev>ew Area Oescl 1 t1 on EQ-12 3.5.5.6 Subcompartment pressure analysis does not consider a door in Room R408.

EQ-13 3.5.5.6 Non-conservative isolation valve closure characteristics assumed in RCIC line break analysis.

EQ-14 3.5.5.6 A non-conservative time delay was 'assumed for generating RWCU oreak isolation signal.

EQ-15 3.5.5.6 HELB calculations for EQ profiles did not specifically address single failure criteria.

EQ-16 3.5.5.6 Normal HVAC ductwork may not retain its integrity to support post-HELB cooldown.

EQ-17 3.5.5.1 There are discrepancies between the model numbers on the Class lE/SRH lists and the installed components.

FP-1 3.5.3.3 Several dedicated cables that require protection were not listed in the E-948 cable tray node su+varies.

FP-2 3.5.3 Thermolag fire barrier is applied to an empty tray that is not required to be lagged.

FP-3 3.5.3 Cable spreading room pentration curbs shown on N-576 are not shown on 5-906.

FP-4 3.5.3.2 Note 7 on N521 SH2 should not apply to RHR-V-40.

WL-1 3.5.6.2 Hain steam tunnel north wall load combinations are not verified.

WL-2 3.5.6.2 FSAR criteria incorrectly applied to the main steam tunnel north wall deflection calculation.

WL-3 3.5.6.1 Attachment loads were not considered in BRI design calculation for the main steam tunnel north wall.

• • Finding *Corresponds to Report S on Number 0 - Observation

= .-Hnt

LIST OF POT L FINDING REPORTS (P 3 of 13)

~ ass 1 1ca ion eview rea Descri tion WL-4 3.5.6.2 Hain steam tunnel north wall minimum reinforcing steel inconsistent with FSAR. The minimum reinforcing steel ratios used in the main steam tunnel are not consistent with FSAR descriptions but do meet ACI 318-1971 requirements.

WL-5 3.5.6.2 Jet impingement load factors were not properly considered in calculating the dynamic loading of the main steam tunnel north wall.

PB-1 3.5.4.1. B Haterial allowables used for approval of loads and/or stresses for PWS-2-1 are not traceable.

PB-2 3.5.4.1.C Field walkdown of HPCS pipe break location identified more potential targets than those cited in the B&R calculation.

PB-3 3.5.4.1.0 Post-accident damage sequence differs from that postulated in the original B&R calculation.

PB-4 3.5.4.1. B As-built strut size is smaller than the size specified in BRI calculation 8.01.52.

PB-5 N.A. Number not used.

PB-6 3.5.4.2. B Field walkdown of RWCU pipe break location identified more potential targets than cited in the BRI calculation.

PB-7 3.5.4.1. E Process deficiencies in potential target resolution were noted.

• • F - Finding *Corresponds to Report Section Number 0 - Observation NV - Not Calid

SECTION A - RE(UIREMENTS REVERIFICATION A. 1 Mech ani cal

~5ifi BRI Documents:

B 8 R Engineering Criteria Document, Rev. 11.

B 8 R Tech. Memos 443, Rev. A; 526, Rev. A; 308, 667, 1010, 148, 156, 653, 776, 785, 845.

General Electric Documents:

22A1843, HPCS System Design Specification, Revision 4.

22A1843AU, HPCS System Design Specification Data Sheets, Revision 4.

731E931, PAID - HPCS System, Revision 7.

731E932AD, Process Diagram - HPCS System, Revision 3.

731E950AD, Flow Control Diagram - HPCS System, Revision 2.

GEK-71334, Hanford 2 Operation and Maintenance Instruction HPCS System, July 1978.

22A3095, Pressure Integrity of Piping Design Specification.

22A3095AD, Pressure Integrity of Piping Design Specification Data Sheet.

22A3790, System Design Pressures Design Specification.

22A3062, Mechanical Codes and Standards Design Specification.

22A2625, System Criteria and Applications for Protection Against Dynamic Effects of Pipe Break Design Specification.

22A2988, Separation Criteria, Revision 6.

22A7416, Separation Criteria, February 1981.

3316-031, Instruction Manual - HPCS Diesel Generator.

21A8657, Rev. 3, Valves.

21A8658, Rev. 1, Electric valve actuaters.

21A9347AF, Rev. 1, Instrumentation and Electric equipment.

22A2625, Rev. 1, Protection against pipe whip.

22A2702AB, Rev. 1, Seismic design.

22A2817, Rev. 3, Residual heat removal.

22A2817AY, Rev. 0, Data sheet for 22A2817.

22A3007, Rev. 1, Testability of instrumentation and controls.

'I 22A3008, Rev. 5, Equipment environmental data.

22A3039, Rev. 1, Process instrumentation.

22A3062, Rev. 2, Mechanical codes and standards.

A-2

22A3095AD, Rev. 1, Data sheet for 22A3095.

22A3730, Rev. 0, RHR heat exchanger.

22A3730AB, Rev. 0, Data sheet for 22A03730.

22A3797, Floor response spectra.

22A5267, Rev. 1, Regulatory requirements.

22A7416, Rev. 1, Electrical separation.

21A8658, General Requirements NOV Actuation.

22A2703E, Radiation Sources.

22A2703F, Radiation Sources.

22A2707, Water Quality.

22A2708, Mater Sampling.

22A2710, Standby AC Power.

22A2711, Plant DC Power.

22A2719AB, RFP Turbine Responses .

22A2719, FW Flow Neasurement and Control.

22A2800, Rated Steam Output Curve.

22A2801, GE Reactor System Heat Balance Rated.

A-3

22A2802, GE Reactor System Heat Balance 22A2887, Nuclear Boiler System.

- 105K Rated.

0 22A2907, Feedwater Control System.

22A3061, Rev. 0, Electrical Codes and Standards.

22A3790, Feedwater System Description.

22A3046, Rev. 1, Core Standby Cooling System Network.

A.1.2 Mestin house Thermal Performance Oata AB095-1554, 1205849 KW, Maximum Calculated Not Guaranteed AB095-1555, 115745 KW, Maximum Guaranteed AF111-0330, No. 5 Extraction AF111-0331, No. 6 Extraction AE111-0572, Nos . 4 and 5 Extraction Zone Enthalpy AE111-0573, No. 6 Extraction Zone Enthalpy A.l.3 Codes and Standards ASNE Boiler and Pressure Vessel Code, 1971 Edition with Addenda through Winter 1973.

ANSI-B. 31.1, Power Piping Code, 1973 Edition with Addenda through Minter 1973.

A-4

AISC Manual of Steel Construction, Seventh Edition, 1970.

WNP-2 FSAR with Amendments through 26, November 1982, Sections 1.2, 3.1, 3.2, 3.5, 3.11; 5.2, 6.1, 6.2, 6.3, 9.5, Appendix F, 14.2.

A-5

0-A.2 Instruments and Controls (Generic Design Requirements Applicable to HPCS, RHR and RFW Systems)

.2.1 ~Rifi BR I Documents:

BRI Design Criteria, Section G Instrumentation and Control".

Paragraphs 4.0, 4.4, 6.0, 7.4.2, Page G-45, Paragraph 2, Paragraph 7.4.1 General Electric Documents:

22A3039, Rev. 1, March 26, 1973, "Process Instrumentation".

Sections: Paragraph 4.3.4.2.

22A3061, Rev. 0, September 3, 1971, "Electrical Codes and Standards".

22A3062, Rev. 0, March 10, 1971, "Mechanical Codes and Standards".

22A3095, Rev. 0, July 17, 1972, "Pressure Integrity of Piping and Equipment Pressure Parts". Sections: Paragraph O'A3.3 22A3790, Rev. 0, May 31, 1973, "System Design Pressures".

22A3059, Rev. 1, November 6, 1972, "Definition of Piping Interfaces

- Reactor Coolant Pressure Boundary".

22A2702A, Rev. 1, January 7, 1971, "Seismic Design" Design Specification.

21A8696, Rev. 0, May 10, 1971, "Seismic Requirement for Class I Instrumentation".

A-6

21A8658, Rev. 1, May 17, 1971, "General Requirements for Motor Operated Valve Actuators". Purchase Requisition.

22A3008, Rev. 5, April 8, 1977, "BWR Equipment Environmental Interface Data". Sections: Paragraph 3.1, 3.2, 4.1, 4.2, and 4.5.

22A3095 AD, Rev. 0, September 26, 1973, "Design Requirements for Pressure Integrity of Piping and Equipment Pressure Parts - Data Sheet".

22A2718, Rev. 5, April 10, 1974, "Special Wire and Cable".

22A3067, Rev. 2, October 12, 1972, "Mechanical Equipment Separation". Paragraph 4.5 22A7416, Rev. 0, "Electrical Equipment, Separation for Safeguards System". Specification February 19, 1982. ~

22A2988, Rev. 6, June 20, 1975, "Electrical Equipment; Separation for Safeguards Systems". P 1 ant Requirements. P ar agraphs: 4.3.3.1, 4.3.3.1.1, 4.3.3.1.2, SHT 10 Table IV, 4.4.1, 4.4.3, 4.4.3.4, 4.4.4, SHT 17, Table 3.

22A2625, Rev. 2, March 9, 1973, "Dynamic Effects/Pipe Break".

Design Guide.

A.2.3 Contracts Contract 42 Tech. Spec. Div. 15 Contract 215 Tech. Spec. Div. 50 Contract 220 Tech. Spec. Div. 50 Page 50A-16, Page 50A-34A, Page 50A-37, 38 A-7

j, A.3 RHR S stem - Desi n Re oirements i&C Section

.3.1 ~5 BR I Documents:

Engineering Design Criteria, Section G General Electric Documents:

22A2817, Rev. 3, November 27, 1973, "Residual Heat Removal System-System Design Specification", Section 4.3, 4.1.2, 4.1.2.4, 4.5.

22A2817AY, Rev. 0, October 31, 1974, "Residual Heat Removal System-System Design Specification - Data Sheet", Sections 2.1, 4.4, and 4.6.

22A3008, Rev. 5, April 8, 1977, "BWR Equipment Environmental Interface Data".

22A3041, Rev. 1, March 14, 1971, "Essential Components".

22A3185, Rev. 1, Febru'ary 4, 1975, "Piping Interfaces".

22A2711, Rev. 3, January 9, 1974, "Plant D-C Power".

22A2718, Rev. 5, April 10, 1974, "Special Wire and Cable".

22A7416, Rev. 0, March 3, 1982, "Electrical Equipment, Separation for Safeguards System".

22A3007, Rev. 1, December 1, 1971, "Engineering Safeguards Systems, Criterion for Testability of Instrumentation and Controls".

A-8

22A3061, Rev. 0, September 3, 1971, "Electrical Codes and Standards".

22A3067, Rev. 2, October 12, 1972, "Mechanical Equipment Separation".

22A2710A, Rev. 7, September 9, 1974, "Standby A-C Power".

22A3095, Rev. 0, July 17, 1972, "Pressure Integrity of Piping and Equipment Pressure Parts".

22A3095AD, Rev. 0, September 26, 1973, "Design Requirements for Pressure Integrity of Piping and Equipment Pressure Parts - Data Sheet".

20A4756, Rev. 1, December 28, 1970, "Logic Symbols ".

22A3059, Rev. 1, November 6, 1972, "Definition of Piping Interfaces Reactor Coolant Pressure Boundary".

.22A2707, Rev. 5, May 28, 1974, "Water guality".

22A2749, Rev. 1, June 24, 1975, "Cleaning of Piping and Equipment".

22A3790, Rev. 0, May 31, 1973, "System Design Pressures".

22A3039, Rev. 1, Mar ch 26, 1973, "Process Instrumentation".

MPL A62-4310, "gualification Testing of Instrument and Control Oev f ices Class i i ed as Essen ti al .

21A8696, Rev. 0, May 10, 1971, "Seismic Requirements for Class I Instrumentation ". Sections SHT 2, 3.

22A3062, Rev. 2, March 10, 1971, "Mechanical Codes and Industrial Stan dar ds".

A-9

i 22A3746, Rev. 1, January 21, 1974, "System Design Local Instrument Panels".

Specification-22A2702A.

A.3.2 Contracts Contract 42, Division 15, Sections 15A, 8, and C Contract 58, Division 50 Contract 59, Division 16, Section 16A Contract 59, Division 50 Contract 215, Division 50 Contract 218, Division 50 Contract 220, Division 50 A-10

>

0

A.4 HPCS S stem - Desi n Re uirements I 8 C Section BR I Documents:

Engineering Design Criteria, Section G, Paragraph 4.0, 4 General Electric Documents:

22A1483, Rev. 4, February 19, 1974, "High Pressure Core Spray System", Sections 3.1, 3.2, 3.3, 4.3.1, 4.3.1.2, 4.3.1.3, 4.3.1.5, 4.5.

731E932AD ll P&ID, HPCS System", SHTS 1 and 2.

22A3039, Rev. 1, March 26, 1973, "Process Instrumentation" System Design Specification .

22A3061, Rev. 0, September 3, 1971, "Electrical Codes and Standards".

22A3062, Rev. 2, March 10, 1971, "Mechanical Codes and Standards".

22A3095, Rev. 0, July 17, 1972, "Pressure Integrity of Piping and Equipment Pressure Parts", Section 4, Table A.

22A3790, Rev. 0, May 31, 1973, "System Design Pressures".

22A3059, Rev. 1, June 24, 1975, "Cleaning of Piping and Equipment".

22A1483AU, Rev. 4,,August 13, 1979, "High Pressure Core Spray System", Design Specification Data Sheet.

22A8696, Rev. 0, May 10, 1971, "Seismic Requirements for Class I Instrumentation", Sections: SHTS 2, 3.

A.4. 2 Contracts:

Contract 42 Tech. Spec. Div. 15 Contract 215 Tech. Spec. Div. 50 Contract 220 Tech. Spec. Div. 50 A-12

A.5 RFW S stem - S ecific Desi n Re uirements IEC Section BRI Documents:

Engineering Design Criteria, Section G General Electric Documents:

22A2907, Rev. 3, March 28, 1974, "Feedwater Control System (Steam Turbine Driven Reactor Feed Pumps) ", System Design Specification, Sections 5.3, 4.3.2.2, 3.1.3.2, 3.3, 4.3.2.

22A2907AB, Rev. 1, August 16, 1971, "Feedwater Control System (Steam Turbine Driven Feed Pumps)" Design Specification, Section 4. 1.3.

22A2719, Rev. 2, June 15, 1973, "Feedwater Flow Measurement and Control" Specification, Section 4.4. 1.1.

22A2719AB, Rev. 0, July 26, 1971, "Feedwater Flow Measurement and Control" BWR Plant Requirements, Section 2.3.

22A3790, Rev. 0, May 31, 1973, "System Design Pressures".

22A2887, Rev. 6, January 29, 1979, "Nuclear Boiler System", Design Specification.

22A3095, Rev. 0, July 17, 1972, "Pressure Integrity of Piping and Equipment Pressure Parts", Sections: SHT 10, D2, SHT 95, SHT 90, 91; Table I, SHT 98 Comment ¹l.

238X241AD, Rev. 9, "Feedwater Control System - Master Parts List".

A-13

DL807E160TC, Rev. 0, June 15, 1978, "Device List and System Elementary Diagram Feedwater Control System".

22A3041, Rev. 1, March 14, 1972, "Essential Components", Design Specification .

239X241AD, Rev. 9, ."Feedwater Control System (Turbine)" Master Parts List.

PL368X482, Rev. 7, "Reactor Feedwater Document List".

22A3095AD, Rev. 0, September 26, 1973, "Design Requirements for Pressure Integrity of Piping and Equipment Pressure Parts - Data Sheet", Sections: SHT 20 A2.1, SHT 98 Paragraph C.

22A3059, Rev. 1, November 6, 1972, "Definition of Piping Interfaces

- Reactor Coolant Pressure Boundary".

22A2707, Rev. 5, May 28, 1974, "Water Quality.

22A2887AB, Rev. 4, "Nuclear Boiler System REVAB" System Design Specification.

22A86796, Rev. 1, March 7, 1978, "Seismic Requirements for Essential Instrumentation", Purchase Specification, Sections: SHT's 2, 3.

21A8657, Rev. 3, May 20, 1975, "General Requirements for Valves".

22A2988, Rev. 6, June 20, 1975, "Electrical Equipment, Separation for Safeguards Systems". Plant Requirements, Paragraphs: 4.3.3.1, 4.3.3.1.1, 4.3.3.1.2, SHT 10 Table IV, 4.4.1, 4.4.3, 4.4.3.4, 4.4.4, SHT 17 Table 3.

A-'14

22A3067, Rev. 2, October 12, 1972; "Mechanical Equipment Separation", Paragraph 4.5.

22A2271AS, Rev. 1, November 30, 1978, "Preoperational Test Program",

Pre-op Test Specifications.

22A3838, Rev. 1, March 8, 1976, "Recommended Prerequisites for Pre-Operational Testing". Preoperational Test Specification.

A-15

BR I Documents:

BhR Engineering Criteria Document, Rev. 11, March 16, 1982, Plus Project Criteria Advance Changes dated up to November 1, 1982, Sections D and F.

TM-330, Rev. N/A, June 28, 1972, "Medium Voltage Switchgear Basis".

TM-427, Rev. 1, February 21, 1973, "Control and Secondary Wiring Internal to Switchgear, Panels, and Similar Enclosures".

TM-443, Rev. A, March 29, 1973, "Systems Description, High Pressure Core Spray System".'N-510, Rev. N/A, May 3, 1973, "Motor Control Center Basis".

TM-526, Rev. A, June 28, 1973, System Description, Residual Heat Removal System".

TM-671, Rev. N/A, July 5, 1974, "Contract ¹2 - PVC Cables".

TM-990, Rev. 1, March ll, 1977, "MCC - PCU Insulated Control Wiring".

TM-1129, Rev. N/A, August ll, 1978, "Class lE Motor Operated Valves".

System Description ¹72, Rev. 0, September 25, 1975, "Feedwater System".

EM-79-006, Rev. N/A, January 2, 1979, "MCC Master List".

A-16

General Electric Documents:

21A8658, Rev. 1, May 17, 1971, "General Requirements for Motor Operated Valve Actuators - Purchase Specification".

21A9222, Rev. 2, January ll, 1974, "Electric Motors, General-f Purch ase Speci i cation".

21A92220M, Rev. 5, December 14, 1979, "Motors, Vertical (RHR)-

Purchase Specification".

22A1483, Rev. 4, February 19, 1974, "HPCS System - Design Specification".

22A1483AU, Rev. 4, August 13, 1979, "HPCS System - Data Sheet".

22A2710A, Rev. 7, September 9, 1974, "Standby AC Power - BWR Requirements".

22A2711, Rev. 3, January 9, 1974, "Plant OC Power - Design f

Speci ication".

22A2817, Rev. 3, November 27, 1973, "RHR System - Design Specification".

22A2817AY, Rev. 2, October 31, 1974, "RHR System - Data Sheet".

22A3008, Rev. 5, April 8, 1977, "BWR Equipment Environmental Interface Data - Design Specification".

22A3038, Rev. 6, February 5, 1979, "Motor List, Electric - Design Specification".

A-17

22A3061, Rev. 0, September 3, 1971, "Electrical Codes and Standards - Design Specification".

22A5267, Rev. 1, May 2, 1979, "Regulatory Requirements and Industrial Standards - Design Bases".

22A7416, Rev. 0, February 19, 1981, "Electrical Equipment, Separations for Safeguards Systems - Plant Requirement".

22A2907, Rev. 3, March 28, 1974, "Feedwater Control System - Design Specification".

22A2907AB, Rev. 1, August 16, 1971, "Feedwater Control System - Data Sheet".

A.6.2 Su 1 S stem Documents Supply System EDI-4.8, Rev. 0, September 22, 1981, "Acceptance Criteria for WNP-2 Safety Related Equipment gualification".

A.6.3 Contracts Contract ¹35, Sect. 15A, "Miscellaneous Pumps and Motors".

Contract ¹41A, Sect. 15A, "Nuclear Valves".

Contract ¹41B, Sect. 15A, "Nuclear Valves".

Contract ¹47A, Sect. 16A, "Medium Voltage Switchgear".

Contract ¹49, Sect. 16A, "Motor Control Centers".

Contract ¹62A, Sect. 16A, "Electrical Cable".

A-18

Contract ¹62B, Sect. 16A, "Electrical Cable".

Contract ¹62C, Sect. TP, "Electrical Cable".

Contract 218, Sect. 16A, "Electrical Installation".

A-19

A.7 En ineerin Mechanics AJ.1 ~5 BR I Documents:

PSDG M400 through M411 - "Pipe Support Design Guide and Work Procedures" for WNP-2, Sections M400 through M411, Rev. 7, 9/16/82.

Burns and Roe, Inc. Design Guide, Rev. 0 (For piping stress analysis only, WNP-2).

TM 429 - BE R, Inc. Technical Memorandum No. 429, "Piping Loads on Equi pment", 12/19/72.

TM 443 - BER, Inc. Technical Memorandum No. 443, "System Descrip-tion High Pressure Core Spray System", Rev. A, 5/4/73.

TM 482 - BER, Inc. Technical Memorandum No. 482, "Seismic Loading for Class II Seismic Piping", 3/23/73 .

TM 1181 - BER, Inc. Technical Memorandum No. 1181, "SRV Discharge Loads: Drywell", 9/17/80 .

TM 1223 - BSR, Inc. Technical Memorandum No . 1223, "Annulus Pressur izaCion - Building Response", 2/17/81.

TM 1226 - B5R, Inc. Technical Memorandum No . 1226, "Piping System Evaluation for Hydrodynamic Loads", Rev. 2, 10/30/81.

TM 1237 - BE R, Inc. Technical Memorandum No. 1237, "Chugging Loads",

7/1/81.

Engineering Criteria Document, Rev. ll, 3/16/82.

A-20

TM 1240 - B&R, Inc. Technical Memorandum No. 1240, "Functional Capability Criteria for WNP-2 Piping", Rev. 1, 2/2/82. 0 TM 1248 - B&R, Inc. Technical Memorandum No. 1248, "LOCA Chugging Loads on WNP-2 Submerged Structures", 11/25/81.

TM 1253 - B&R, Inc. Technical Memorandum No. 1253, "SRV Loads:

Displacements", 1/13/82.

TM 1254 - B&R, Inc. Technical Memorandum No. 1254, "SRV Discharge Loads Wetwell".

TM 1257 - B&R, Inc. Technical Memorandum No. 1257, "Structur al Response Spectra", 3/5/82.

TM 1263 - B&R, Inc. Technical Memorandum No. 1263, "Hydrodynamic Loads to be Used for the DAR, Rev. 3 Assessment", 4/20/82.

TM 1059 - B&R, Inc. Technical Memorandum No. 1059, "Load Capacity of Primary Containment Weld Pads", Rev. 1, 1/31/78.

TM 1085 - B&R, Inc. Technical Memorandum No. 1085, "Pipe Break Outside of Containment - Structural Effects", 4/6/78.

TM 1020 - B&R, Inc. Technical Memorandum No. 1020, "Regulatory Guide 1.46; Recommendation Concerning Implementation", Rev. 1, 10/19/77.

TM 1151 - B&R, Inc. Technical Memorandum No. 1151, "Criteria for Pipe Break and Missile Redundancy Evaluation Outside Primary Con tainmen t", 6/27/79.

TN 1210 - B&R, Inc. Technical Memorandum No. 1210, "Statistically Derived Allowables for Expansion Bolts", 10/17/80.

0

TM 1271 - B&R, Inc. Technical Memorandum No. 1271, "gC II Equipment Nozzle Allowable Loads", 6/14/82.

DWG M520 - B&R, Inc. Drawing No. M520, "Flow Diagram, HPCS and LPCS Systems, Reactor Building", Rev. 27.

DWG M521 - B&R, Inc. Drawing No. M521, "Flow Diagram, Residual Heat Removal System", Rev. 35.

t OWG M200-112 - Drawing "Residual Heat Removal System", Rev. 4.

DWG M200-150 - Drawing "Residual Heat Removal System", Rev. 7.

General Electric Documents:

22A1483 - General Electric Design Specification, "High Pressure Core Spray System", Rev. 4, 2/19/74.

22A2817 - General Electric Design Specification, "Residual Heat Removal System", Rev. 3, ll/27/73.

22A2887 - General Electric Design Specification, "Nuclear Boiler Sys tern", Rev. 6.

22A3790 - General Electric System Design Specification, "System Design Pressures", Rev. 0, 5/31/73.

22A3797 - General Electric Design Analysis, "Floor Response Spectra, Primary Containment", Rev. 1, 5/22/75.

761E428 - Heat Exchanger Outline Drawing, Rev. 2.

NEDO 21061 - General Electric Report, "Dynamic Forcing Functions Information Report", Rev. 3.

A-22

22A3095AO - General Electric Data Sheet, "Pressure Integrity of Piping and equipment Pressur e Par ts", Rev. 0.

22A3170 - General Electric Certified Design Specification, "Piping, Main Steam and Recirculation", Rev. 0.

731E932 Drawing - Process Diagram and Data Sheet for HPCS System, Rev. 3.

731E966 Drawing - Process Diagram and Data Sheet for RHR System, Rev. F.

A.7.2 Su 1 S stem Documents Report WPPSS-74-2-R3 - Protection Against Pipe Breaks Outside Con tainment.

Report WPPSS-74-2-Rl - Protection Against Pipe Breaks Inside Con tainment.

A.7.3 Contr act S ecifications C215 - Specification 2808-¹215, "Mechanical Equipment Installation and Piping", Contract No. 215.

C215 158 - Section 15B, "Piping Systems", of C215 Spec.

C215 15Q - Section 15Q, "Pipe Supports", of C215 Spec.

C220 15E - Specification 2808-¹220, "Instrumentation Installation",

Contract No. 220, Section 15E, "Piping and Tubing Supports".

C208 - Specification C-0208, "Small Diameter Piping and Pipe Support Criteria", Rev. 1, Modification 4, 5/29/81.

0

A.7.4

~ ~ Codes and Standards ASME Sec. III - ASME Boiler and Pressure Vessel Code, Section III, Div. 1, 1971 Edition through Winter, 1973 Addenda.

ASME NB-3000 - Article NB-3000, "Design", of ASME Sec. III.

ASME NC-3000 - Article NC-3000, "Design", of ASME Sec. III.

ASME ND-3000 - Article ND-3000, "Design", of ASME Sec. III.

ASME NF-3000 - Article NF-3000, "Design", of ASME Sec. III.

ANSI 831.1 - American National Standard Code for Pressure Piping, "Power Piping", 1973 Edition through Winter, 1973 Addenda.

ANSI 831.1 - 101 - Section 101, "Design Conditions", of ANSI 831.1.

ANSI 831.1 - 102 - Section 101, "Design Criteria", of ANSI 831.1.

ANSI 831.1 - 104 - Section 101, "Pressure Design of Components", of ANSI 831.1.

AISC Manual - Amer ican Institute of Steel Construction, Inc. "Manual of Steel Construction", 7th Edition, 1970.

AISC Spec. - AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings", 2/12/69.

ANS-58.2 - ANSI N176, "Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture",

Dec. 1979.

A-24

A.7.5 NRC Documents NRC Topical Report 7/17/80 - NRC Topical Report, "Evaluation of Topical Report - Piping Functional Capability criteria", 7/17/80.

NRC RG 1.29 - NRC Regulatory Guide 1.29, "Seismic Design Classification", Rev. 3.

NRC RG 1.46 - NRC Regulatory Guide 1.46, "Protection Against Pipe Whi p Ins i de Con tainment ", Rev. 0.

NRC RG 1.48 - NRC Regulatory Guide 1.48, "Design Limits and Loading Combinations for. Seismic Category I Fluid System Components", Rev. 0.

NRC RG 1.60 - NRC Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants", Rev. l.

NRC RG'1.61 - NRC Regulatory Guide 1.61, "Damping for Seismic Design of Nuclear Power Plants", Rev. 0.

NRC RG 1.92 - NRC Regulatory Guide 1.92, "Combining Modal Response and Spatial Components in Seismic Response Analysis", Rev. l.

10 CFR 50 - Title 10, Chapter 1, "Code of Federal Regulations-Energy", Part 50.

NRC IKE 79-02 - NRC Inspection and Enforcement Bulletin No. 79-02 "Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts", Rev. 2, ll/8/79.

NRC SRP 3.6.1 - NRC Standard Review Plan, Section 3.6.1, "Plant Design .for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment"..

A-25

NRC SRP 3.6.2 - NRC Standard Review Plan, Section 3.6.2 "Determination of Break Locations and Dynamic Effects Associated with the Postulated Piping Failures.

A-26

SECTION B HPCS SYSTEM REVIEW REFERENCES B. 1 Mechanical Disci line References General Electric:

22A1483, Rev. 4, "Design Specification - High Pressure Core Spray System", MPLt E22-4010 dated February 19, 1974.

22A1483AU, Rev. 4, "Design Specification Data Sheet - High Pressure Core Spray System", MPL8 E22-4010, dated August 13, 1979.

22A3095AD, Rev. 1, "Design Specification Data Sheet - Pressure Integrity of Piping and Equipment Pressure Parts", MPl 8 A62-4030 dated September 26, 1973.

22A2702AB, Rev. 1, "Design Specification Data Sheet 2, Seismic Design", MPL8 A62-4090, dated January ll, 1972.

22A3067, Rev. 3, "Design Specification - Mechanical Equipment Separation", MPL8 A62-4350, dated August 21, 1975.

21A1740, Rev. 3, "Purchase Specification - Valve, Gate", MPL8 E22-F004, dated 1/13/72.

21A1884, Rev. 2, "Purchase Specification - Valve Data Specification, Valve, Gate", MPL8 E22-F004, dated 1/14/75.

21A9243, Rev. 0, "Purchase Specification - Auxiliary Pumps for Boiling Water Reactors", MPL8 E22-C001, dated May 1, 1973.

21A9243DE, Rev. 2, "Purchase Specification Data Sheet - High Pressure Core Spr ay Pump", NPL¹ E22-C001, dated October 29, 1973.

a )

k 21AI880, Rev. 1, "Purchase Specification - Valve Data Specification, Valve Gate", NPL¹ E22-F012, Dated April 18, 1972.

21A1736, Rev. 3, "Purchase Specification - Valve Data Specification

- Valve, Gate", NPL¹ E22-F012, dated January 25, 1972.

MR Contract Specifications 2808-215, Section 15A, 15B, 15F, 15G 2808-69, Section 15A 2808-213 PDN Contract Purchase Specification, PUSP-16713-3 Rev. B, dated August 7, 1980.

BLR Engineering Criteria Document, Sections E, F and I B. 1. 2 Cal cul ations BSR Cal cul ations:

5.19.01, Rev. 0, "HPCS Pipe Sizing", June 15, 1971.

5.19.02, Rev. 0, "HPCS System - Preliminary Line Sizing" December 12, 1971.

5.19.07, Rev. 0, "HPCS Piping Schedule", April 18, 1973.

5. 19.08, Rev. 0, "Restrictors - HPCS System", September 12, 1975.

B-2

5.19.10, Rev. 0, "ECCS Minimum NPSH Calculations; R.G. 1.1",

Rev. 0, November 10, 1976.

5.19.11, Rev. 4, "Pressure Drop Calculation HPCS System", August 20, 1981.

5.19.12, Rev. 0, "HPCS System - Water Leg Low Pressure Alarm",

September 6, 1979.

5.19.13, Rev. 1, "Sizing of HPCS Emergency Water Volume",

September 15, 1981.

5.19.14, Rev. 0, "NPSH of HPCS Pump - Maximum Allowable Suppression Pool Temperature", September 10, 1980.

10.04.71, Rev. 0, "WPPSS Hanford No . 2 Condensate Tank NRC guestion 211.61", June 10, 1979.

10.04.72, Rev. 0, "WPPSS NP82 - Analyze'ortex Formation at the HPCS/RCIC Suction Inlet in the CST", August 25, 1980.

B. 1.3 Technical Memoranda B@R Technical Memorandum 443 Rev. A "System Descr iption High Pressure Core Spray System", March 12, 1973.

B. 1.4 Manual s General Electric Operation and Maintenance Instructions:

GEK 71334, "High Pressure Core Spray System", July 1978.

B-3

VPF 3238-842-2, Rev. B, "Instruction Manual - Motor-Operated Gate Valves, GE Order 205-AE204, Darling E-5310" dated Jurie 21, 1980.

VEL-HO-1 "Velan Operation and Maintenance Manual - Check Valves".

VPF 3069-30-3, "Ingersoll-Rand 0&N Manual - High Pressure Core Spray Pump", March 19, 1975.

B.1.5 ~Drawin s General Electric Drawings:

731E931AD Rev. 0, "P&ID-HPCS System" NPL¹ E22-1010, dated July 30, 1974.

73lf931 Rev. 7, "P&ID - HPCS System" MPL¹ E22-1010, dated May 22, 1974.

731E932AD Rev. 3, "Process Diagram-HPCS System", MPL¹ E22-1020, dated October 22, 1978.

B&R Dr awings:

M520 Rev. 33 N732, Rev. 18 N711 Rev. 22 M626, Rev. 5 N712 Rev. 29 SN197, Rev. D N713 Rev. 29 SM193,. Rev. E M714 Rev. 25 SM191, Rev. E N715 Rev. 27 SM183, Rev. E M716 Rev. 27 SM136, Rev. D N718 Rev. 33 SM135 Rev. D S798 Rev. 31 N567 Rev. 6 S796 Rev. 14 M200, Sht. 132 Rev. 5 S795 Rev. 41 M200, Sht. 100 Rev. 7A B-4

N744 Rev. 7 M200, Sht. 101 Rev. 9 N527 Rev. 37 M200, Sht. 2 Rev. 5 M569 PDM Drawings:

E37 Rev. A2 E61, Rev. B CB&I Drawings:

72-4396-2 Rev. 6, 72-4396-lA Rev. 7 72-2647-1 Rev. 8 72-2647-123 Rev. 7 Zurn Drawings:

I-80120-A (BE R Transmi ttal 213B-12318)

"

Anchor-Darling Drawings:

94-13262 Rev. E 94-13306 Rev. C 94-13401 Rev. B Velan Drawing:

P2-2767-N-2 Rev. L J. E. Lonergan Drawing:

A-2647 Rev. A Ingersoll-Rand Drawings:

0-12X20KD86XEZ 0-12X20K00321XZC 8-5

Permulit Drawing:

556-30530 Rev. 9 Isometric Drawings:

HPCS-629-1. 4 HPCS-629-5. 7 HPCS-630-1.4 HPCS-630-5.6 HP CS-630-7 . 10 PCS-630-11 . 12 HPCS-630-13.19 HPCS-630-20. 23 HPCS-630-24. 25 HPCS-630-26. 28 HPCS-630-29. 30 HPCS-630-31. 33 COND-351-1.9 COND-351-10. 15 HPCS-632-1.3 HPCS-633-1. 2 H PCS-1458-1 HPCS-1458-2 HPCS-1458-3 HPCS-1458-4 HPCS-1459-1 HPCS-1459-2 HPCS-1460-1 HP CS-1461-1 HPCS-2569-1 HPCS-1644-1 HPCS-2568-1 HPCS-2570-1 HPCS-1958-1 HP CS-2571-1 D-220-X- 78 8.1.6 WNP-2 FSAR Sections 3.2, 6.1, 6.2, 6.3, 15 B.l.7 Other References "High Pressure Core Spray System Design Reverification Plan", Revision 1, System design Engineering, WPPSS, dated February 20, 1983.

SDEI-3.5 "Design Reverification", Revision 3, System Design Engineering Instruction 3.5, MPPSS, dated December 8, 1982.

TDP 3.4 "Preparation, Verification, and Control of Calculations",

June 8, 1982 "MNP-2 Plant Verification Report" MPPSS, dated June 1982 VPF-3069-91-1, "Certified Test Report for HPCS-P-1", April 10, 1975 B-6

Form N-5, Data Reports for Field Installation of Nuclear Power Plant Components, Component Supports and Appertenances (by HPCS System Line and Code Class)

Crane Technical Paper No . 410, "Flow of fluids Through Valves, Fittings, and Pipe", Crane Co., Twentieth Printing - 1981.

NEDM-20363-13, Hydraulic Analysis Procedures for BWR Piping Systems",

GE, September 1975.

AEC-TR-6630, "Handling of Hydraulic Resistance, Coefficients of Local Resistance and of Friction", I.E. Idel'chik, 1960.

B-7

B.2 Mechanical Diesel Disci line References 22A1483, Rev. 4, High Pressure Core Spray Systems Design Specifications.

22A1483AU, Rev. 4, HPCS Design Specification Data Sheet.

21A1848AB, Engine Generator for HPCS Purchase Specification Data Sheet.

21A1848, Engine Generator for HPCS Purchase Specification .

21A1776, Rev. 1, HPCS Diesel Service Water Pump Purchase Specification.

21A1776AD, Rev. 1, HPCS Diesel Service Water Pump Purchase Specification Data Sheet.

A990 GEAPPD, Thermxchanger Exchanger Specification Sheet.

Contract 215, Material Specifications.

BER Engineering Criteria Document.

B.2.2 Cal culations BIWR Nuclear Calculations:

5.43.01, "Diesel Engine System Calculations" Rev. 0, 2/19/74.

5.43.02, "Diesel Oil Tanks (Storage and Day Tanks) Capacity Verification", Rev. 0, 8/9/79.

B-B

8.2.3 Technical Memoranda BER HPCS Diesel Generator Technical Memorandum:

TM-0558 O.G. Synchro - Check Relays TM-0586 Emergency D.G. Operation TM-0775 Diesel Generator Loading TN-0746 Interlocking for Diesel Generator TM-0608 Diesel Generator Cooling Water System TM-1053 Standby D.G. Light Load Operation 1M-1066 Gas Disper. 8 Met Analysis TN-0817 System Description, D.G. Systems TM-0443 (Rev. A) System Description B. 2. 4 Manual s Instructions/Parts Manual for Hanford II, Diesel Generator, Contract No.

205-AD583, PSD IWO No. A-990, by Power Systems Division, Book,0ne, Sections 8, 10, 14 NI 1748 Rev. B, Doc. No. 3316-031 Pacific Pumps Instruction Manual CV I 2-2E22-13-11 B.2.5 ~Drawin s Isometric Drawings:

OE-797-1.5 Rev. 7 1/25/83 OE-789-1. 3 Rev. 5 10/12/82 OE-1738-1 Rev. 7 12/2/82 OE-2836-1 Rev. 5 12/1/82 OSA-4275-1 Rev. 5 12/10/82 DS A-4396-1 Rev. 4 10/12/82 DSA-4396-2 Rev. 5 9-21-82 DSA-2536-1 Rev. 1 11/19/82 8-9

Isometric Drawings (Contd.)

OSA-2537-1 Rev. 4 8/6/82 DSA-2537-2 Rev. 5 7/24/82 DS A-253 7-3 Rev. 5 8/2/82 DSA-2537-4 Rev. 5 11/29/82 DSA-2537-5 Rev. 4 8/19/82 DO-448-1B D0-1620-1 DO-1620-2 DO-2530-3 D0-2531-1 DO-2531-3 DO-2532-1 DO-2532-2 D0-2532-3 D0-2533-1 DO-2533-2 D0-2533-3 DO-2675-1 D0-2797-1 DO-4328-1 DCW-2510-1 DCW-2510-2 DW-1965-11 MR Flow Diagram M-512 GE Piping Diagr ams for HPCS Diesel Engine Generator A990D08001 HPCS Diesel Engine Generator Air Intake Piping Schematic A990009001 HPCS Diesel Engine Generator Exhaust System Piping Schematic

A990F03001 HPCS Diesel Engine Generator DLO Schematic Diagram A990F04001 HPCS Diesel Engine Generator Jacket Water System with Heat Exchanger Schematic Diag< am A990C06002 Fuel Oil Schematic Lister SR1A Diesel Engine A990F06001 HPCS Diesel Engine Generator F.O. Schematic Diagram A990F 07001 HPCS Diesel Engine Generator Air Start System Schematic Diagram A990F 02001 HPCS Diesel Engine-Generator Assembly B.2.6 Other References I&E Bulletins - HPCS Diesel Generator Emergency D.G. Lube Oil Addition and Onsite Supply, 80-04 Potential D.G. Turbocharger Problem 79-12 Degradation of Fuel Oil Flow to the Emergency D.G. 77-15 Emergency D.G. Lube Oil Cooler Failures 80-11 Standards of Tubular Exchanger Manufacturers Association DEMA Standard Practices for Low and Medium Speed Stationary Diesel Engines ASTM Standards, Part 17, Classification of Diesel Fuel Oils

'ational Fire Protection'ssociation Standar ds 30, 37, and 70 SLT-57.2-5 (Rev. 0) HPCS Diesel Engine Jacket Cooling Water Flush and Fill

General Electric:

22A1483 Rev. 4 Design Specification, HPCS System 22A1483AU Rev. 4 HPCS System Data Sheet 22A2988 Rev. 6 Electrical Separation (See 22A7416) 22A3008 Rev. 5 BWR Equipment Environmental Interface Data 22A3039 Rev. 1 Process Instrumentation 22A3067 Rev. 3 Mechanical Equipment Separation 22A3095 Rev. 0 Process Integrity of Piping and Equipment 22A3 746 Rev. 1 Local Instrument Panels 22A7416 Rev. 0 Electrical Equipment Separ ation Burns and Roe:

Design Criteria Sections F and G B.3.2 Calculations Burns and Roe:

5.51.051 Target Determination Pipe Break Outside Containment 8.01.203 Pipe Break Locations B. 3.3 Technical Memorandum Burns and Roe:

TM-1151, Criteria for Pipe Break and Missile Redundancy Evaluation B-12

Letter BRBEC-F-82-3752, dated October 21, 1982 Letter BRBEC-F-83-2174, dated March 22, 1983 B.3. 4 Manual s General Electric GEK 71334 July 1978 High Pressure Core Spray System, 0&M GEK 71337 June 1978 Vendor Suppl ied Instruments,, OEM Dragon Valves, Inc.

12583 Rev. 0 Excess Flow Check Valve Instrument Manual 8.3.5 ~0r awin s General Electr ic 127D1840TC Rev. 2 HPCS Instrument Panel Arr angement 163C1043T C Rev. 1 HPCS Instrument Panel Piping Diagram 731E931AD Rev. 7 HPCS P&ID 731E950AD Rev. HPCS FCD 234A9309TC Rev. 3 Instrument Data Sheets 807E172TC Rev. 19 Elementary Diagrams HPCS System Burns and Roe 7E015 Rev. 2 Electrical Wiring Diagram HPCS-V-1 7E016 Rev. 2 Electrical Wiring Diagram HPCS-V-4 7E017 Rev. 1 Electrical Wiring Diagram HPCS-V-10 7E018," Rev. 1 Electrical Wiring Diagr am HPCS-V-ll 7E019 Rev. 2 Electrical Wiring Diagram HPCS-V-12 7E020 Rev. 2 Electrical Wiring Diagram HPCS-V-15 8-13

7E021 Rev. 1 Electr i cal Wiring Diagram HPCS-V-23 7E025 Rev. 1 Electrical Wiring Diagram Controls Sheet 1 7E02 Rev. 1 Electrical Wiring Diagram Controls Sheet 1 S 709 Rev. 26 Structural , Reactor Building f522 Rev. 16 Elem. Diag . Isolation Valves E 535-18A Rev. 9 Connection Wiring Diag. MC4A

-18B Rev. 7 Connection Wiring Diag. MC4A E 536-2C Rev. 12 Connection Wiring Diag. Term. Box and Misc.

-5B Rev. 12 Connection Wiring Diag. Term. Box and Misc.

E537- I V Rev. 3 Connection Wiring Diag. Term. Box and Misc.

-3A Rev. 9 Connection Wiring Diag. Control Room Term.

Cab.

-4B Rev. 8 Connection Wiring Diag. Control Room Term.

Cab.

-26A Rev. 7 Connection Wiring Diag. Control Room Term.

Cab.

E539-2 Rev. 10 Connection Wiring Diag. Reactor IEC

-14 Rev. 10 Connection Wiring Diag. Reactor ILC

-21 Rev. 8 Connection Wiring Diag. Reactor IEC E 540-4 Rev. 5 Connection Wiring Diag. Motor Op. Valves

-6 Rev. 10 Connection Wiring Diag. Motor Op. Valves N520 Rev. 32 Flow Diagram, HPCS and LPCS Systems M527 Rev. 44 Flow Diagram, Condensate Supply System M530 Rev. 32 Flow Diagram, Nuclear Boiler Recirculation System M543 Rev. 33 Flow Diagram, Containment Cooling and Purging M567 Rev. 6 General Arrangement Reactor Bldg. El 422'and Arr angement Reactor Bldg. El 441'eneral M568 Rev. 22 471'nd 501'nstrumentation Contract 220, General Notes M619-VI Rev. 5 0 0 M609 Rev. 10 Instrument Process AZ 0 to 180 M619-6 Rev. 4 Instrument Conn. Diag. H22-P009

-19 Rev. 4 Instrument Conn. Diag. H22-P024

N623 Rev. 8 Instrument Process Plan El Rev. 8 Process Plan El 501'nstrument M624 M625 Rev. 11 Process Plan El 512'nstrument 471')

Rev. 7 Process Plan El 522'nstrument N626 Rev. 13 Process Plan El 541'nstrument N627 N628 Rev. 17 Process Penetration Schedule 560'nstrumentation N629 Rev. 7 Instrumentation Process Partial Plans and Sections N734 Rev. 24 Miscellaneous Piping Plan and Sections at El Johnson Controls B-220-007.0-H22P024 Rev. Line Identification List (20 sheets)

B-220-X-73 Rev. Line Identification List X-73

-86A Rev. Line Identification List X-86A

-86B Rev. Line Identification List X-86B

-87A Rev. Line Identification List X-87A

-87B Rev. Line Identification List X-878 0-220-007.0-H22P024 Rev. Tube Erection Isometrics (20 Sheets) 0-220-7.1-X-732-1 Rev. Process Instrument Line X-73a

-lA Rev. Process Instrument Line X-73a

-1B Rev. Process Instrument Line X-73a

-1C Rev. Process Instrument Line X-73a 0-220-X-73 Rev. Process Instrument Line X-73

,

-86A Rev. Process Instrument Line X-86A

-86B Rev. Process Instrument Line X-86B

-87A Rev. Process Instrument Line X-87A

-87B Rev. Process Instrument Line X878 D-220-3500-250-CMS-LT-1&2, Rev. 1, Local Instrument Installation E-220-5500-RB-441 Rev. 6 Tube Routing React. Bldg. El.

441'ube

-471 Rev. Routing React. Bldg. El.

447'ube

-501 Rev. Routing React. Bldg. El.

501'ube

-522 Rev. Routing React. Bldg. El.

522'ube

-548 Rev. Routing React. Bldg. El. 548'

Bovee and Crail HPCS-630-7.10 Rev. 8 Discharge from HPCS-P-1 to RPV G i lber t/Commonweal th COND-4631-1 Rev. 3 RCIS-HPCS Switchgear Standpipe Rev. 2 RCIS-HPCS Switchgear Standpi pe w3 Rev. 3 RCIS-HPCS Switchgear Standpipe Rev. 3 RCIS-HPCS Switchgear Standpipe Rev. 3 RCIS-HPCS Switchgear Standpipe Rev. 3 RCIS-HPCS Switchgear Standpipe Dragon Valves, Inc.

C-12583 Rev. E Excess Flow Check Valve

~

Daniel Industries C-2629 ANS Orifice Flange, Upstream C-2630 ANS Orifice Flange, Downstream B.3.6 Contract S ecifications 21A9376 Rev. 1 Flow Orifice Assembly 21A9376A J Rev. 1 Flow Orifice Assembly Data Sheet 21A9417AB Rev. 0 I.D.S. HPCS System 2808-220 Johnson Controls B.3.7 Other References Drawing Control Log, Dated 2-13-83 WNP-2 NUREG-0588, Environmental Equipment (}ualification Report 8-16

B.4 Electrical Disci line References B&R Engineering Criteria Document, Rev. 11, March 16, 1982, plus Project Criteria Advance Changes dated up to April 6, 1983, Section D and Appendix 3, "WNP-2 Electrical Separation Practices",

Rev. 2, March 21, 1983.

21A1884, Rev. 2, April 23, 1975, "HPCS Gate Valve Data-Purchase Specification".

21A8658, Rev. 1, May 17, 1971, "General Requirements for Motor.

Operated Valve Actuators".

21A9222, Rev. 2, January ll, 1974, "Electrical Motors, General-Purchase Specification".

21A9222DI, Rev. 3, July 18, 1980, "HPCS Pump Vertical Motor Data f

Sheet - Purch ase Speci i cation".

21A9300, Rev. 3, July 25, 2978, "Switchgear Electrical Metal Enclosed for HPCS - Purchase Specification".

21A9300AO, Rev. 3, August 17, 1974, "Metal Enclosed Electrical Switchgear - Purchase Specification - Data Sheet".

21A9301AJ, Rev. 3, November 26, 1974, "HPCS Motor Control Center-Purchase Specification - Data Sheet".

22A1483, Rev. 4, August 7, 1974, "HPCS Design Specification".

22A1483AU, Rev. 2, August 13, 1979, "HPCS Design Specification-Data Sheet".

22A2710A, Rev. 7, September 9, 1974, "Standby AC Power - BWR Requiremen ts".

22A7416, Rev. 0, March 5, 1981, "Electrical Equipment Separation for Safeguards System BWR Plant Requirements Specification".

22A3008, Rev. 5, Apr il 8, 1977, "BWR Equipment Environmental Interface Data - Design Specification".

22A3061, Rev. 0, September 3, 1971, "Electrical Codes and Standards - Sys tern Desi gn Speci fication".

22A7416, Rev. 0, March 5, 1981, "Electrical Equipment, Separation for Safeguards Systems".

238X185AD, Rev. 13, December 29, 1981, "HPCS Parts List".

22A3038, Rev.6, "List of Electric Motors - Design Specification".

B.4.2 Calculations 2.02.02, Rev. 1, "Main Plant One-Line Auxiliary Load Calculations".

2.02.07, Rev. 2, "Motor Control Centers - Load Calculations".

2.02.16, Rev. 1, "Load Summary - Major Plant Operating Modes".

2. 02. 18, Rev. 0, "Volt. Switchgear Load Study".

2.03.02, Rev. 5, "Main One-Line Short Circuit Calculation".

2.03.07, Rev. 2, "480 V Switchgear Short Circuit Calculations".

2.03.09, Rev. 0, "Motor Control Center Short Circuit Calculations".

2.05.01, Rev. 3, Battery and Battery Charger Calculation 250 VDC, 125 VDC and 24 VDC Systems".

2.06.03, Rev. 5, "Main One-Line Voltage Drop Calculations".

2.06.07, Rev. 1, "Service and Diesel Generator Building Feeder and Voltage Drop Calculation".

2.06.10, Rev. 1, "Service and Diesel Generator Building Feeder and Voltage Drop Calculation".

2.06.17, Rev. 0, "4160/6900 V. Motor Feeder Cable - Voltage Drop".

2.07.01, Rev. 2, "High Voltage Cable Sizing - Ampacities and Conduits".

'.07.05, Capacity".

Rev. 0, "Cable Sizing - 4.16 and 6.9 KV - Short Circuit 2.07.09, Rev. 3, "125 VDC System Cable Sizing for Circuit Breakers".

2.07.10, Rev. 0, "D.C. System Cable Sizing for Voltage Drop Calculation".

9.21.02, Rev. 0, "Reactor Building - Emergency Cooling and Critical Area Cooling System".

9.24.00, Rev. 6, "HVAC - Diesel Generator Building".

8.4.3 Technical Memoranda B5R Technical Memo $ 1060, Rev. 3, January 22, 1980, "Voltage Drop Study".

8-19

B.4.4 Manuals CVI 49-00,25, Issue 1, "ITE Instruction Manual" for Motor Control Centers.

VPF 3395-27, "Installation Oper ation and Maintenance and Instruction Manual for HPCS Metal Clad Switchgear ".

VPF 3390-12, "Switchgear Equipment Instruction Manual".

CVI 2-02E22-09, 10, Issue 1, "ICS Manual".

AEF 62B-00-0112, "Instrumentation Control Power Coaxial and Triaxial Cables Installation Instruction Manual ".

B.4.5 ~Dnawin s.

Burns and Roe E WD-7E-003, Standby Water Leg PP. Rev. 1, 05/17/82 HPCS-P-3 (E22-C003)

E WD-72-0016 M.O.V. HPCS-V-4 (E22-F004) Rev. 2, 08/31/82 E 502-2 Main One Line Diagram Rev. 19 01/19/83 E 503-9 Aux. One Line Digaram Rev. 16 12/18/82 E 514-6 Relay Settings 4.16 KV Switch- Rev. 6 12/18/82 gear SH-4 E 517-9 4160 V SWGR, Elem. Diag. Rev. 18 02/02/83 E 550 Cable Schedule - Power Rev. 26 03/17/83 E 553-1 Class lE Electrical Equip List Rev. 4 03/28/83 E 558-2 Turb. Gen. Bldg. Grounding Rev. 4 04/12/82 Plans and Details E 662-1 Reactor Bldg. Grounding Plans Rev. 11 10/07/82 and Details Sh. 1 B-20

E 680 Reactor Bldg. El 422'3" Power Rev. 17 12/30/82 Conduit and Tray Plan E 785 Diesel Gen. Bldg. El. 441'-0" Rev. 32 01/21/83 Power and Tray Plans PED 218-E-4533 807E183TC, Sheets 1 to 6, overall Rev. "HPCS Power Supply Elementary Diagram".

2 807E172TC, Sheets 1 to 8, overall Rev. "HPCS Elementary Diagram".

992C349BC, Rev. 4, "HPCS Pump Motor Outline".

731E302AD, Sheets 1 to 3, overall Rev. "HPCS Power Supply One Line Diagram".

VPF 3395-9, Rev. , "HPCS Motor Control Unit Wiring Diagrams ".

VPF 3395-10, Rev. , "HPCS Motor Control Unit Wiring Diagrams".

VPF 3395-11, Rev. , "HPCS Motor Control Unit Wiring Diagrams".

VPF 3395-2, Rev. , "480V Motor Control Center for HPCS".

VPF 3395-2, Rev. ', "480V Motor Control Center for HPCS Bill of Material".

147C1614, Rev. 1, "HPCS Transformer Outline".

5528, Rev. , "HPCS Transformer Nameplate Detail".

0123D3805, Rev. , "HPCS Metal Clad Switchgear Interconnection".

B-21

B.4.6 Memoranda NED0-10905, Rev. 3, "Topical Report HPCS Power Supply Unit and Amendments".

GEWP-2-81-189, HPCS Relay Settings.

EM-79-006, Rev. 0, January 2, 1979, "MCC Master List".

8.4.7 Contract S ecifications Contract 2, Division 2, Section 2A, "Nuclear Steam Supply System".

Contract 35, "Miscellaneous Pumps and Motors", Division 15, Section 15A.

Contract 49, Division 16, Section 16A, "Motor Control Centers".

Contract 62A, Division 16, Section 16A, "Electrical Cable".

Contract 62B, Division 16, Section 16A, "Electrical Cable".

B.4.8 Other References IEEE 141-1976, "The Red Book - Recommended Practice for Electric Power Distr ibution for Industrial Plants".

IEEE 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations".

IEEE 308-1974, "Criteria for Class lE Power Systems for Nuclear Power Generating Stations".

B-22

IEEE 308-1971, "Criteria. for Class lE Power Systems for Nuclear Power Generating Stations".

NEMA-MG-1-1978, "Motor and Generator Standards".

NEMA-ICS-1-108, "Service and Installation Conditions".

NEMA-ICS-2-321, "AC General-Purpose Class A Magnetic Controllers for Induction Motors Rated in Horsepower 600 Volts and Less, 50 and 60 Hertz".

NEMA-ICS-2-322, "AC General-Purpose Motor Control Centers".

NEMA-ICS-2-327, AC General-Purpose Class A Magnetic Controllers for Induction Motors, Rated in Full-Load and Locked-Rotor Current, 600 Volts and Less, 50 and 60 Hertz".

IPCEA Pub . No . S-68-516, Interim ¹2, "Cables Rated 5000 Volts and Less and Having Ozone-Resistant Ethylene-Propylene-Rubber Integral Insulation and Jacket".

ICEA Pub . No . P-54-440 (Second Edition), "Ampacities - Cables in Open-Top Cable Trays".

NFPA-70-1981, "National Electrical Code".

IEEE Conference Paper C72-121-7, "IEEE Flame Test Report".

Washington Public Power Supply System WNP-2 Class lE Equipment List (WNP-2'ClE List) dated 3/2/83.

Crane-Oeming Pumps - Test No. T-552B - Item ¹13 - HPCS Water Leg Pump.

B-23

Westinghouse "Report of Test Form for Induction Motors", Form A-2, Induction Motor L-980864-03-A dated 5/4/76.

Gould/Shawmut Bulletin AT-620R, "UL Class RK-5 Current Limiting Fuses".

WNP-2 Master Equipment List (MEL).

Okonite Bulletin 721.1 "Engineering Data for Copper and Aluminum Conductor Electrical Cables".

Okonite Bulletin 776, "Okonite Cable".

Cable Pull Slips - Fishback/Lord - for Cables 3HPCS-0030, 3HPCS-0080, 3HPCS-0340.

WNP-2 FSAR Appendix F.

Darling Valve and Manufacturing Company, "Test Data Report", Shop Order No. E5310-4-1, Customer P. 0. 205-AE204, Valve Tag No.

E22-F004, dated 9-20-74.

Anchor-Darling Valve Co'. letter "Certification", P.O. No. 205-AE204, MPL No. E22-F004. Dated 10-7-74.

B-24

B.5 En ineerin Mechanics Disci line References B.5.1 S eci fications/Codes/Guides BER Engineering Criteria Document, Sections E and I, Appendies A and 2 Rev. 3.

ASME B5PV Code, Section III, 1971 Edition, including Addenda through Winter, 1973.

BER Piping Design Guide, Rev. 3.

WNP-2 FSAR.

U.S. NRC Standard Review Plan 3.6.2.

G.E. Design Specification 22A2887, Rev. 6, for the Nuclear Boiler System.

G.E. Document 0 22A1483, Rev. 4 (HPCS System Design Specification).

ASME B5PV Code Case N-122(1745) 03/01/76.

Westinghouse Structural Analysis Program PIPSAN.

Contract 208 Specification, "Small-Diameter Piping and Pipe Support Criteria" Section .

EhR Memo EM-82-322, 6/28/82.

WRC Bulletin 107, March 1979, Revision.

ASME Code Case N-318, Rev. 0.

B-25

TPIPE User's Manual, Version G/C 4.3.

G/C Engineering Handbook, User's Manual, Rev. 7.

BER Project Engineering Directive, PEO-C0208-0689.

Johnson Controls, Inc. Design Guide for WNP-2.

AISC Steel Construction Manual, 7th Edition.

General Electric Document 21A9243 DE, Rev. 1, "Specification and Data Sheet - HPCS Pump".

General Electric "Operating Instructions for HPCS Pump", 3/19/75.

Buil.ding Code Requirements for Reinforced Concrete, ACI 318-71.

ANSI 831.1 Power Piping Code, 1973 Edition, W73 Addenda.

Contract 208 Specification, "Small-Diameter Piping and Pipe Support Criteria" section.

B.5.2 Calculations BSR Calculation 8.14.64A, Rev. 0.

Burns and Roe, Inc. Calculation 8 8-70-02 (Thermal Expansion).

85R McDonald Douglas STRUOL run 81219, 68 pages, November 1, 1982.

B&R McDonald Douglas STRUDL run 82016, 65 pages, October 29, 1982.

B5R Calculation 8.15.65 for HPCS-52.

8-26

B&R Calculation 8.15.225 for HPCS 901N.

Gilbert/Commonwealth Calculation No. OE-1738-1, Rev. 7.

G/C Calculation ¹00010, Rev. 1, Design Guide for Shear Lugs.

Pipe Support Calculation No. JCI-220-CLC-961, Rev. l.

U-Bolt Calculation No. JCI-220-CLC-529, Rev. 2.

Pipe Stress Calculation NUPIPE Run X-73AIN, T-20, 83/03/24.

Base Plate Calculation JCI-220-CLC-997, Rev. l.

Piping Analysis Program Summary, X-73a IN, 3/28/83.

Pipe Support Calculation, OE-1738-11 and llA, Rev. 8.

Pipe Stress Analysis, TPIPE run BC2RFWC, Rev. 7, Run 5.

G/C Calculation No. 0000-12, Rev. 0.

Pipe Support Calculation No. 8.15.1133, Rev. 3.

NUPIPE Printout "X73AIN AS-BUILT CONF IG", 82/09/08.

JCI Calculation 220-CLC-4119, Rev. 0.

JCI Piping Analysis Program Summary, File X-73a, Trial ¹20.

B&R Cal cul ation 8.14.82 11/10/82.

Pipe Support Calculation No. 8.15.1076, Rev. 2.

B-27

Pipe Stress Calculation No. 8.14.82, Support Load Summary Sheets, Rev. 9.

BER Calculation No. 6.17.22, Book SV-72.

B.5.3 Technical Memoranda T.M. 1226, Rev. 3, "Piping System Evaluation for Hydrodynamic Loads".

T.M. 1240, Rev. 1, "Functional Capability Criteria for WNP-2 Piping".

Tech. Memo ¹1253 (SRV Displacements).

Tech. Memo ¹1181, Rev. 1 (SRV Response Spectra).

Tech. Memo ¹1257, Rev. 2 (Seismic and Hydrodynamic Response Spectra).

Tech. Memo ¹1283 (Reduction of SRV Loading).

B.5.4 Manuals "Weldolet Stress Intensification Factors", Bonney Forge, 1976.

"NAVCO Piping Datalog", Edition No. 10, 1974, National Valve and Manufacturing Company, Pittsburgh, PA.

"ANSYS Rev. 4 User's Manual", Rev. A, 2/1/82, Swanson Analysis Systems, Inc.

ADLPIPE User Manual, Rev. J, ll/18/82, issued by CDC.

TPIPE User's Manual, Version G/C 4.3.

B-28

Project Engineering Directive 220-M-0853, 07/23/82.

ADLPIPE Input Prepar ation Manual, May 1981 Revision.

ADLPIPE Reference ¹16, "Lumped Mass Location", March, 1975.

B. 5.6 ~0rawin s B&R Stress Isometric, M200-Sh. 100, Rev. 7A.

Bovee and Grail Construction Isometric, HPCS-629-1.4, Rev. 9FO (as-built).

Pittsburgh-Des Moines Steel Co., AB-E68, Rev. N, 6/9/82, Wetwell Piping.

B&R Flow Diagram, M-520, Rev. 33.

G.E., 731E932AO, Rev. 3, HPCS Operating Conditions.

POM, 0-101, Rev. K, Penetration X-31 Details .

Anchor/Darling, 2621-3, Rev. B, Valve HPCS-V-16.

Anchor/Darling, 94-13473, Rev. A, Valve HPCS-V-15.

PDM, AB-E150, Wetwell Piping and Support Details.

POM, AB-E118-31, Wetwell Piping and Support Details.

POM, AB-E12, Wetwell Piping and Support Details.

B&R, S-795, Penetration Dimensions.

B-29

BER Support Detail Drawing, HPCS-900N.

BhR Support Detail Drawing, HPCS-901N.

IKR Support Detail Drawing, HPCS-52.

B&R Project Engineering Directive, PED-C0208-0689.

M200-SHT2-1., Rev. C (HPCS Isometric).

H200-SHT2-2, Rev. A (HPCS Supports Orientation).

S795 (X-6 Penetration Detail ).

M601, Rev. 20 (Valve List).

HPCS-63, Rev. 3 (Support Detail).

HPCS-912N, Rev. 1 (Support Detail).

HPCS-911N, (Support Detail).

HPCS-910N, Rev. 1 (Support Detail).

4 HPCS-918N, Rev. 2 (Support Detail).

HPCS-919N, Rev. 2 (Support Detail).

HPCS-66; Rev. 3 (Support Detail).

HPCS-904N, Rev. 2 (Support Detail).

HPCS-906N, Rev. 1 (Support Detail).

B-30

HPCS-64, Rev. 2 (Support Detail).

HPCS-907N, Rev. 1 (Support Detail).

HPCS-908N, Rev. 2 (Support Detail).

Pipe Isometrics, BER M200-606, Rev. 4.

Pipe Isometric, G/C DE-1738-1, Rev. 8.

Flow Diagram, EhR M512, Rev. 27.

Valve drawing, Borg-Warner ¹ 38020, Rev. f.

Pipe Support Drawing, B-220-670-35, Rev. l.

Pipe Fabrication Isometric, D-220-7.1-X-73a, Rev. 2.

Pipe Isometric and Support Drawings, DE-1738-1, 3 sheets.

Pipe Support Drawing, HPCS-910N, Rev. 3Fo.

Pipe Fabrication Isometric, HPCS-630-26.28, Rev. 8.

JCI 0-220-7.1-X-73a, As-Built.

JCI Pipe Support Drawings, as listed in D-220-7.1-X-73a.

MR Blow Diagram, M520, Rev. 27.

Dragon Valve Drawing C-10580, Rev. 0.

Pipe Fabrication Drawing HPCS-630-29.30, Rev. 7.

B-31

Pipe Support Standard Drawings H501 (5 sheets), H502 Rev. 0, H503 Rev. 0.

BER Drawing S-701, Rev. 9.

BKR Drawing S-702, Rev. 6.

BER Drawing S-749, Rev. 17.

B&R Drawing S-750, Rev. 21.

EhR Drawing S-660, Rev. 28.

Ingersoll-Rand Pump Drawing C-12X20KO86X2-H, Rev. 6.

General Electric 167E2054, Rev. 0 (Nozzle Thermal Transients).

731E932AD, Rev. 3 (HPCS Thermal Modes).

761E716 (RPV Nozzle Allowable Loads).

Bovee and Grail HPCS-630-31.33, Rev. 9 (Construction Drawings).

HPCS-630-29.30, Rev. 8 (Construction Drawings).

HPCS-630-26.28, Rev. 9FO (Construction Drawings).

Velan Dwg. PP2-2767-N-2, Rev. L (Valve HPCS-V-5).

B-32

Velan Dwg. 8P2-3311-N-16, Rev. F (Valve HPCS-V-51).

Anchor/Darling Dwg. 82652-3, Rev. B (Valve HPCS-V-76).

B.5.7 Memoranda U.S. NRC Memo, "Evaluation of Topical Report - Piping Functional Capability Criteria", R. L. Tedesco from J. P. Knight, 7/17/80.

Memo from J. Braverman to R. E. Snaith on 2/10/81 (Seismic Anchor Motions).

Memo from D. Bagehi to R. E. Snaith on 2/1/82 (Seismic Anchor Motions).

5.5.6 Other Velan Valve Stress Report 0 SR-6335.

B-33

SECTION C RHR SYSTEM REVIEW REFERENCES C.1 ~5il'i 21A3757, Rev. 0, GE Purchase Specification for Relief Valves on RHR Heat Exch angers.

21A3757AA, Rev. 2, GE Purchase Specification Data Sheet for Tube Side Relief Valves on RHR Heat Exchangers.

21A3757AD, Rev. 2, GE Purchase Specification Data Sheet for Shell Side Relief Valves on RHR Heat Exchangers.

21A8657, Rev. 3, GE Purchase Specification, General Requirements for Valves.

21A8658, Rev. 1, GE Purchase Specification, General Requirements for Motor Operated Valve Actuators.

21A8706, Rev. 3, GE Purchase Specification, Heat Exchanger Materials for General Electric Design.

21A9222, Rev. 2, GE Purchase Specification, General Requirements for Electric Motors.

21A9222DM, Rev. 5, GE Purchase Specification Data Sheet for Vertically Mounted RHR System Motor.

21A9243, Rev. 0, GE Purchase Specification for Auxiliary Pumps for Boiling Water Reactors.

21A9243DJ, Rev. 3, GE Purchase Specification Data Sheet for RHR Pumps.

21A9347AF, Rev. 1, GE Purchase Specification Data Sheet, General Requirements for Instrumentation and Electric Equipment.

21A9376, Rev. 1, GE Purchase Specification for Flow Orifice Assembly.

21A9388AB, Rev. 0, GE Purchase Specification, Instrument Data Sheet for the RHR System.

21A9425, Rev. 1, GE Purchase Specification for RHR Heat Exchangers.

21A9425AB, Rev. 1, GE Purchase Specification Data Sheet for RHR Heat Exchangers.

22A2707, Rev. 5, GE Design Specification, BWR Plant Requirements for Water Quality.

22A2710A, Rev. 7, GE Design Specification, BWR Plant Requirements for Standby AC Power.

22A2711, Rev. 3, GE Design Specification, BWR Plant Requirements for OC Power.

22A2714AB, Rev. 1, GE Design Specification, BWR Plant Requirements for Ventilating, Cooling and Heating.

22A2750, Rev. 4, GE Design Specification for Inservice Inspection .

22A2750AO, Rev. 1, GE Quality Assurance Data Sheet for Inservice Inspection .

22A2817, Rev. 3, GE System Design Specification for the RHR System.

22A2817AY, Rev. 0. GE System Design Data Sheet for the RHR System.

C-2

22A2988, Rev. 6, GE BWR Plant Requirements, Separation of Electric Equipment for Engineered Safeguard Systems (see also 22A7416 Rev. 0).

22A3007, Rev. 1, GE System Design Specification, Testability Criterion for Instrumentation and Controls in Engineered Safeguard System.

22A3008, Rev. 5, GE Design Specification, Environmental Interface Data for BWR Equipment.

22A3038, Rev. 6, GE Design Specification, Data Listing for Electric Motors to be Supplied by the APED of GE (see also 21A9222).

22A3039, Rev. 1, GE System Design Specification for Process Instrumentation.

22A3061, Rev. 0, GE System Design Specification, Electrical Codes and Standards.

22A3062, Rev. 2, GE System Design Specification, Mechanical Codes .

22A3067, Rev. 3, GE System Design Specification for Mechanical Equi pment Separation.

22A3085, Rev. 3, GE Design Specification for the Remote Shutdown System.

22A3095, Rev. 0, GE System Design Specification, Pressure Integrity of Piping and Equipment Pressure Parts.

22A3095AO, Rev. 1, GE System Design Data Sheet, Pressure Integrity of Piping and Equipment Pressure Parts.

22A3730, Rev. 0, GE System Design Specification for RHR Heat Exchangers.

22A3730AB, Rev. 0, GE System Design Data Sheet for RHR Heat Exchangers.

C-3

22A3746, Rev. 1, GE System Design Specification for Local Instrument Panels.

22A5233, Rev. 0, GE Installation Specification for RHR Heat Exchangers.

22A5267, Rev. 1, GE System Specification on Regulatory Requirements, Industrial Standards and Design Bases .

22A7416, Rev. 0, GE BWR Requirements for Separation of Electrical Equipment in Engineered Safeguard Systems (see also 22A2988 Rev. 6).

234A9407TC, Rev. 4, GE Instrument Data Sheets for the RHR System.

249A1401TC, Rev. 1, GE Instrument Data Sheets for the Remote Shutdown System.

B&R Engineering Criteria Document:

Section D Electrical Engineering Criteria Section E Mechanical Engineering Criteria Section F Chemical and Nuclear Engineering Criteria Section G Instrumentation and Control Engineering Criteria Section H Technical Standards Applicability List Section I Piping and Pipe Support Criteria C-4

C.2 BER Desi n Calculations 2.02.02, Rev. 1, Main Plant Oneline Auxiliary Load Calculations.

2.02.07, Rev. 2, Motor Control Center Load Calculations.

2.02.18, Rev. 0, 480 V Switchgear Load Study.

2.03.02, Rev. 5, Main Oneline Short Circuit Calculation .

2.03.07, Rev. 2, 480 V Switchgear Short Circuit Calculation.

2.03.09, Rev. 0, Motor Control Center Short Circuit Calculations.

2.03.11, Rev. 0, Fault Calculations for Paralleling DG1 and OG2.

2.06.03, Rev. 5, Main Oneline Voltage Drop Calculations.

2.06.05, Rev. 3, Reactor Building Feeder Voltage Drop Calculation.

2.06.10, Rev. 1, Service and Diesel Generator Building Feeder Voltage Drop Calculations.

2.06.17, Rev. 0, 4.16 KV and 6.9 KV Motor Feeder Cable Voltage Drop Cal cul ation.

2.07.01, Rev. 2, High Voltage Cable Sizing, Ampacities and Conduits.

2.07.05, Rev. 0, 4.16 KV and 6.9 KV Cable Sizing, Short Circuit Capacity.

2.07.09, Rev. 3, 125 V DC System Cable Sizing for Circuit Breakers.

2.07.10, Rev. 0, OC System Cable Sizing for Voltage Drop.

C-5

2.12.00, Rev. 5, Relay Setting Time Current Characteristic Curves.

2.12.14, Rev. 1, 4.16 KV Switchgear Relay Settings.

5.17.13, Rev. 0, Flow Restrictor Sizing, RHR System.

5.17.19, Rev. 1, RHR System Pressure Drop Calculations.

5.17.20, Rev. 0, Effectiveness Calculation for RHR Heat Exchanger.

5.17.26, Rev. 0, RHR Testline Orifice Sizing Calculation.

5.17.29, Rev. 1, RHR LPCI Line Orifice Sizing Calculation.

6.19.19, Rev. 0, Pages 38 through 56A, Structural Calculation, Reactor Building, Interior Walls at Elevation 572.0 ft.

6.19.34, Rev. 2, Sheets 1 through 9 (Pages 62 to 70C), Structural Calculation, Reactor Building, Equipment Foundations.

7.00.55, Rev. 3, Minimum Flow Control Valve Sizing Calculations.

8.14.127B, Rev. 6, Structural Design Calculation, Anchor Group 36.

8.15.213, Rev. 4, Review/Redesign Calculation for Piping Support RHR-HGR-184 (PS-l, Y), Node 163.

8.15.2341, Rev. 2, Review/Redesign Calculation for Piping Support RHR-HGR-436 (PS-6, Y) Node 1220.

9.21.02, Rev. 0, Reactor Building, Emergency Cooling and Critical Area Cooling System.

9.32.00, Rev. 3, HVAC for Control Room, Cable Spreading Room and Critical Switchgear Room.

C-.6

C.3 Technical Memoranda TM 151, Rev. 0, B&R Technical Memorandum, RHR Heat Exchanger Leak ti Inves gati on.

TM 181, Rev. 0, B&R Technical Memorandum, Shielding Requirements for the RHR System.

TM 194, Rev. 1, B&R Technical Memorandum, RHR Heat Exchanger Leak Inves ti gati on.

TN 327, Rev. 0, B&R Technical Memorandum, Shielding Requirements for the RHR Heat Exchanger Rooms.

TN 420, Rev. 4, B&R Technical Memorandum, Electric Cable - Listing of Outside Diameter, Weight, Pulling Tension and Bending Radius.

TN 526, Rev. A, B&R Technical Memorandum, System Description for the RHR System.

TN 563, Rev. 0, B&R Technical Memorandum, RHR Heat Exchanger Leakage Inves ti gation.

TM 610, Rev. 0, B&R Technical Memorandum, RHR System Relief Valve Siz ing.

TM 1000, Rev. 0, B&R Technical Memorandum, Actuation of RHR Heat Exchanger Relief Valves.

TM 1016, Rev. 0, B&R Technical Memorandum, Cavitation in the RHR System.

TN 1060, Rev. 3, B&R Technical Memorandum, Voltage Droop Study.

TM 1129, Rev. 0, B&R Technical Memorandum, Class lE Motor Operated Valves.

C-7

TM 1131, Rev. 0, BER Technical Memorandum, Design Changes for Line RH (16).

TM 1232, Rev. 0, NhR Technical Memorandum, Service Water Requirements.

C-8

C.4 Vendor Manuals GEK-71330, July 1978, Operation and Maintenance Instructions for the Remote Shutdown System.

GEK-71336, July 1978, Operation and Maintenance Instructions for the Residual Heat Removal System.

GEK-71337, June 1978, Operation and Maintenance Instructions for Vendor Supplied Instruments CVI 47A-OO, 131, Issue 1, Operation and Maintenance Instructions plus Parts Catalog for Medium Voltage Metal Clad Switchgear.

CVI 49-00, 25, Issue 1, ITE Instruction Manual for Motor Control Centers.

CVI 2-02E12-08, Sheet 10, Issue 1, Operation and Maintenance Manual for RHR Pumps (Ingersoll Rand).

C-9

C.5 ~0rawin s C.5.1 Mechanical and Nuclear M-151, Rev. 0, B&R General Arrangement Drawing, Ground Floor (Elevation 441.0 ft).

M-152, Rev. 0, B&R Gener al Arrangement Drawing, Mezzanine Floor (471.0 ft).

M-153, Rev. 0, B&R General Arrangement Drawing, Operating Floor (501.0 ft).

M-154, Rev. 0, B&R General Arrangement Drawing, Reactor Duilding Floor Plans at 422.25 ft, 510.5 ft, 522.0 ft, 548.0 ft, 572.0 ft, and 606.88 ft.

M-155, Rev. 0, B&R General Arrangement Drawing, Reactor Building Vertical Sections.

M-159, Rev. 0, B&R Equipment List for General Arrangement Drawings.

M-501, Rev. 21, B&R Chart of Flow Diagram Symbols.

M-521, Sheet 1 and 2, Rev. 39, B&R Flow Diagram of the Residual Heat Removal System.

M-524, Sheet 1 and 2, Rev. 37, B&R Flow Diagram of the Standby Service Water System.

197R567, Rev. 3, GE Piping and Instrument Symbols .

731E961AD, Sheet 1 and 2, Rev. 4, GE Piping and Instrumentation Diagram for the RHR System.

C-10

731E966, Rev. 6, GE Process Diagram for the Residual Heat Removal System.

731E966AO, Sheet 1, Rev. 2, Sheet 2, Rev. 0, GE Process Data Sheet for the Residual Heat Removal System.

762E481, Rev. 5, GE Assembly Drawing, RHR Heat Exchanger.

762E483, Rev. 3, GE Drawing of RHR Heat Exchanger Channel.

762E484, Rev. 4, GE Drawing of RHR Heat Exchanger Tube Bundle.

762E485, Rev. 3, GE Drawing of RHR Heat Exchanger Tube Sheet.

9210280, Rev. 0, GE Instrument Symbols.

105D4981, Rev. 2, GE Drawing of RHR Heat Exchanger Channel Cover.

10504984, Rev. 3, GE Drawing of RHR Heat Exchanger Baffle Plate.

137C7572, Rev. 0, GE Installation Drawing for RHR Heat Exchanger Relief Valve.

M-200, Sheet 106, Rev. 5, BER Isometric Diagram with RHR-V-4B, RHR-V-6A, RHR-P-2B Suction.

M-200, Sheet 107, Rev. 5, NhR Isometric Diagram with RHR-HX-18 Inlet, RHR-V-47B, RHR-V-48B, RHR-V-89, RHR-V-116) RHR-V-115, RHR-FCV-64B, RHR-R0-1B, RHR-V-188.

M-200, Sheet 112, Rev. 4, BSR Isometric Diagram with RHR-FE-14B, RHE-F IS-10B, RHR-V-3B.

M-200, Sheet 113, Rev. 4, 85R Isometric Diagram with RHR-V-428, RHR-V-538, RHR-V-178) RHR-V-168.

M-200, Sheet 150, Rev. 7, B&R Isometric Diagram with RHR-V-278, RHR-V-248, RHR-V-1728, RHR-R0-38.

M-701, Rev. 24, Reactor Building Layout at Elevation 422.25 ft..

M-702, Rev. 21, Reactor Building Layout at Elevations 441.0 ft and 444.0 ft.

M-703, Rev. 18, Reactor Building Layout at Elevation 471.0 ft.

N-704, Rev. 22, Reactor Building Layout at Elevation 501.0 ft.

M-705, Rev. 23, Reactor Building Layout at Elevation 522.0 ft.

N-706, Rev. 32, Reactor Building Layout at Elevation 548.0 ft.

N-707, Rev. 15, Reactor Building Layout at Elevation 572.0 ft.

M-708, Rev. 28, Reactor Building Layout Details at Various Elevations and Vertical Sections.

M-709, Rev. 31, Reactor Building, Vertical Sections.

C.5.2 Instrumentation and Control 197R567, Rev. 3, GE Piping and Instrument Symbols.

731E961AD, Rev. 4, 2 Sheets, GE Piping and Instrumentation on Diagram for the RHR System.

731E966, Rev. 6, GE Process Diagram for the RHR System.

C-12

731E999, Rev. 5, GE Functional Control Diagram for the RHR System.

I J

762E280AD Rev. 0, GE Functional Control Diagr am for the Remote Shutdown Panel.

807E170TC, Rev. 14, GE Elementary Diagram for the RHR System.

807E151TC, Rev. 10, GE Elementary Diagram for the Remote Shutdown Panel.

10504947AD, Rev. 1, GE IED for the Remote Shutdown Panel.

127D1812TC, Rev. 3, GE Tubing Diagram for Rack H22-P021.

127D1841TC, Rev. 3, GE Arrangement Drawing for Rack H22-P021.

828E191TC, Rev. 9, GE Connection Diagram for Rack H13-P618.

828E289TC, Rev. 5, GE Connection Diagram for Rack H22-P021.

828E466TC, Rev. 8, GE Arrangement Drawing for Remote Shutdown Panel.

J 828E482TC, Rev. 8, GE Connection Diagram for Remote Shutdown Panel.

9210280, Rev. 0, GE Instrument Symbols.

145C3008, Rev. 8, GE Differential Pressure Switch Diagram, Purchased Part.

145C3011, Rev. 8, GE Diagram for Differential Pressure Switch.

159C4540, Rev. 6, GE Diagram for Meter, Model 180.

163C1183 Rev. 5GE Diagram for Differential Pressure Transmitter.

C-13

9E003, Rev. 2, B&R Electr ic Wiring Diagram for RHR-P-2B.

9E004, Rev. 1, B&R Electric Wiring Diagram for RHR-P-28.

9E010, Rev. 1, B&R Electric Wiring Diagram for RHR-P-3.

9E017, Rev. 2, B&R Electric Wiring Diagram for RHR-V-3B.

9E019, Rev. 2, B&R Electric Wiring Diagram for RHR-V-4B.

9E022, Rev. 2, B&R Electric Wiring Diagram for RHR-V-6B.

9E034, Rev. 1, B&R Electric Wiring Diagram for RHR-V-24B.

9E038, Rev. 1, B&R Electric Wiring Diagram for RHR-V-27B.

9E047, Rev. 1, B&R Electric Wiring Diagram for RHR-V-47B.

9E049, Rev. 2, B&R Electric Wiring Diagram for RHR-V-48B.

9E057, Rev. 1, B&R Electric Wiring Diagram for RHR-FCV-64B.

E-522, Rev. 16, B&R Elementary Diagram for Isolation Valve Status Display Panel.

E537, Sheet 6C, Rev. 11,. B&R Connection Wiring Diagram for Control Boards.

E539, Sheet 16, Rev. 10, B&R Connection Wiring Diagram for RHR System.

E539, Sheet 20, Rev. 8, B&R Connection Wiring Diagr am for RHR System.

E-697, Rev. 32, I&C Conduit and Tray Diagram at Elevation 501.0 ft.

C-14

M-153, Rev. 0, B&R General Arrangement Drawing, Operating Floor (501.

ft).

N-154, Rev. 0, B&R General Arrangement Drawing, Reactor Building Floor Plans at 422.3 ft, 510.5 ft, 522.0 ft, 548.0 ft, 572.0 ft, 606.9 ft.

N-155, Rev. 0, B&R General Arrangement Drawing, Reactor Building Vertical Sections.

M-159, Rev. 0, B&R Equipment List for General Arrangement Drawings.

M-501, Rev. 21, B&R Chart of Flow Diagram Symbols.

M-521, Sheet 1 and 2, Rev. 39, B&R Flow Diagram of the RHR System.

M-568, Rev. 23, B&R Radiation Zone Drawing for Reactor Building at Elevations 471.0 ft and 501.0 ft.

N-706, Rev. 33, B&R Piping Plan, Reactor Building at Elevation 548.0 M-735, Rev. 26, B&R Piping Plan, Reactor Building at Elevation 501.0 ft.

M-807, Rev. 18, B&R HVAC Plans, Reactor Building at Elevation 501.0 ft.

M200, Sheet 107, Rev. 5, B&R Piping Diagram, Contract 215.

N200, Sheet 112, Rev. 4, B&R Piping Diagram, Contract 215.

M619, Sheet 15, Rev. 7, B&R Tubing Connection Diagram, Contract 220.

M619, Sheet 16, Rev. 7, B&R Tubing Connection Diagram, Contract 220.

D-220-0090-H22-P021, Rev. 1, JCI Diagram.

C-15

0-220-3500-5.0-RHR-FT-l, Rev. 1, JCI Diagram.

E-220-5500-RB-501, Rev. 5,. JCI Drawing.'2A8654, Rev. D, Fisher Controls Drawing of Limitorque Actuated Control Valve.

C.5.3 Electr ical 9E003, Rev. 2, B&R Electric Wiring Diagram for Pump RHR-P-2B.

9E017, Rev. 1, B&R Electric Wiring Diagram for RHR-V-3B.

9E034, Rev. 1, B&R Electric Wiring Diagram for RHR-V-24B.

9E057, Rev. 1, B&R Electric Wiring Diagram for RHR-FCV-64B.

E501, Rev. 9, B&R Electrical Symbol List.

E502, Sheet 2, Rev. 19, B&R Main Oneline Diagram, Emergency Buses.

E503, Sheet 7, Rev. 25, B&R Auxiliary Oneline Diagram, Motor Control Centers.

E503, Sheet 8, Rev. 23, B&R Auxiliary Oneline Diagram, Motor Control Centers.

E503, Sheet 12, Rev. 23, B&R Auxiliary Oneline Diagram, Motor Control Centers.

E514, Sheet 8, Rev. 2, B&R Diagram, Relay Settings for 4.16 KV Switchgear, SM-8.

E517, Sheet 3, Rev. 12, B&R Elementar y Diagram for 4.16 KV Switchgear.

C-16

E517, Sheet 4, Rev. 8, B&R Elementary Diagram for 4.16 KV Switchgear.

E517, Sheet 9, Rev. 17, B&R Elementary Diagram for 4.16 KV Switchgear.

E517, Sheet 10, Rev. 13, B&R Elementary Diagram for 4.16 KV Switchgear.

E517, Sheet 13, Rev. 9, B&R Elementary Diagram for 4.16 KV Switchgear.

E517, Sheet 18, Rev. 1, B&R Elementary Diagram for 4.16 KV Switchgear.

E518, Sheet 6, Rev. 11, B&R Elementary Diagram for 480V Switchgear.

E519, Sheet lA, Rev. 4, B&R Elementary Diagram for Valve Control.

E528, Sheet 25, Rev. 1, B&R MCC Equipment Overload Summary for MCC-MC-7B-A.

E528, Sheet 35, Rev. 3, B&R Overload Summary for MCC-MC-88.

E528, Sheet 36, Rev. 1, B&R Overload Summary for MCC-MC-BB-A.

E528, Sheet 37, Rev. 0, B&R Overload Summary for MCC-MC-BB-B.

E533-21VH-5, Rev. 2, Bill of Material for Electrical Devices, 4.16 KV Switchgear, SM-8.

E550, Rev. 35, Power Cable Schedule.

E551, Rev. 38, Control Cable Schedule.

E558, Sheet 2, Rev. 4, Turbine Generator Building, Grounding Plans and Details.

C-17

E680, Rev. 18, Reactor Building at Elevation 422.25 ft, Power Conduit and Tray Plan.

E681, Rev. 11, Reactor Building at Elevation 441.0 ft, Power Conduit and Tray Plan.

E682, Rev. 35, Reactor Building at Elevation 471.0 ft, Power Conduit and Tray Plan.

E684, Rev. 31, Reactor Building at Elevation 522.0 ft, Power Conduit and Tray Plan.

E685, Rev. 20, Reactor Building at Elevation 548.0 ft, Power Conduit and Tray Plan.

E686, Rev. 25, Reactor Building at Elevation 572.0 ft, Power Conduit and Tray Plan.

E745, Sheet 1, Rev. 18, Radwaste and Control Building at Elevation 437.0 ft, Power Conduit and Tray Plan.

E747, Sheet 1, Rev. 37, Radwaste and Control Building at Elevation 467.0 ft, Power Conduit and Tray Plan.

E915, Rev. 12, Reactor Building at Elevation 422.25 ft, Location Plan for Cable Tray Nodes.

E916, Rev. 5, Reactor Building at Elevation 441.0 ft, Location Plan for Cable Tr ay Nodes.

E917, Rev. 10, Reactor Building at Elevation 471.0 ft, Location Plan for Cable Tray Nodes.

C-18

E191, Rev. 7, Reactor Building at Elevation 522.0 ft, Location Plan fo Cable Tray Nodes .

E922, Sheet 2, Rev. 6, Reactor Building Sections, Location Plan for Cable Tray Nodes.

E922, Sheet 4, Rev. 7, Reactor Building Sections, Location Plan for Cable Tray Nodes.

E927, Sheet 1, Rev. 10, Radwaste and Control Building at Elevation 437.0 ft, Location Plan for Cable Tray Nodes.

E929, Rev. 9, Radwaste and Control Building at Elevation 467.0ft, Location Plan for Cable Tray Nodes.

4 E934, Sheet 2, Rev. 10, Cable Spreading Room in Radwaste and Control Building, Location Plan for Cable Tray Nodes.

E935, Sheet 4, Rev. 7, Section 4-4 of Radwaste and Control Building, Location Plan for Cable Tray Nodes.

M-521, Sheet 1 and 2, Rev. 39, BER Flow Diagram of the Residual Heat Removal System.

922C302FO, Rev. 6, Outline for Induction Motor RHR-M-28.

C.5.4 Structural M-151, Rev. 0, BER General Arrangement Drawing at Elevation 441.0 ft (Gr ound Floor).

M-152, Rev. 0, BER General Arrangement Drawing at Elevation 471.0 ft (Mezza Floor).

C-19

N-153, Rev. 0, B&R General Arrangement Drawing at Elevation 501.0 ft (Operating Floor ).

N-154, Rev. 0, B&R Reactor Building Floor Plans at Elevations 422.25 ft$ 510.5 ft, 522.0 ft, 548.0 ft, 572.0 ft, 606.88 ft.

M-155, Rev. 0, B&R General Arrangement Drawing, Reactor Building Vertical Sections.

M-159, Rev. 0, B&R Equipment List for General Arrangement Drawing.

M-501, Rev. 21, B&R Chart of Flow Diagram Symbols.

N-521, Sheet 1, Rev. 39, B&R Flow Diagram of the Residual Heat Removal Sys tern.

M-521, Sheet 2, Rev. 39, B&R Flow Diagram of the Residual Heat Removal Sys tern.

H-501, Sheets 1, 2, 3, all Rev. 0, B&R Construction Tolerances, Piping and Pipe Supports.

S-660, Rev. 28, B&R Drawing, Structural Anchor Bolt Schedule.

S-722, Rev.16, B&R Drawing, Reactor Building Details at Elevation 572.0 ft.

S-769, Rev. 7, B&R Drawing, Reactor Building Details.

S-772, Rev. 40, B&R Drawing, Reactor Building Equipment Foundations Sheet 2.

S-794, Sheet 1, Rev. 24, Structural Drawing of Primary Containment.

C-20

S-1000, Sheet 1, Rev. 19, List of Reactor Building Piping Restraints.

/

S-1062, Rev. 5, Load Table for Piping Supports in Primary Containment.

761E428, Rev. 2, GE Drawing, Residual Heat Removal System.

I 762E481, Rev. 5, GE Drawing of the RHR Heat Exchanger.

762E484, Rev. 4, GE Drawing of Tube Bundle for RHR Heat Exchanger.

762E485, Rev. 3, GE Drawing of Tube Sheet for RHR Heat Exchanger.

10504984, Rev. 3, GE Drawing of Baffle Plate for RHR Heat Exchanger.

N-200, Sheet 107, Rev. 5, BER Isometric Diagram with RHR-HX-1B Inlet.

M-200, Sheet 112-1, Rev. 5B, BER Isometric Diagram with RHR-FE-14B (at Elevation 565.5 ft).

N-200, Sheet 112-2, Rev. A, Data Sheet for N-200 Sheet 112-1.

M-200, Sheet 150, Rev. 7A, PAR Isometric Diagram with RHR-V-24B.

M-701, Rev. 19, BER Drawing, Reactor Building Floor Plans at Elevation 422.25 ft, Vertical Sections..

N-702, Rev. 21, BSR Drawing, Reactor Building Layout at Elevations 441.0 ft and 444.0 ft.

N-703, Rev. 18, B&R Drawing, Reactor Building Layout at Elevation 471.0 ft.

M-704, Rev. 22, B5R Drawing, Reactor Building Layout at Elevation 501.0 ft.

C-21

M-705, Rev. 23, BINR Drawing, Reactor Building Layout at Elevation 522.0 ft.

M-706, Rev. 32, BER Drawing, Reactor Building Layout at Elevation 548.0 ft.

M-707, Rev. 15, B5R Drawing, Reactor Building Layout at Elevation 572.0 ft.

M-708, Rev. 21, BER Drawing, Various Reactor Building Sections and Details.

RHR-184 S0068, Sheet 10F4, Rev. 4, B&R Drawing of Piping Support RHR-HGR-184.

C-22

0 C.6 Memoranda General EM-79-006, B&R Engineering Memorandum;- MCC Master List, January 2, 1979.

EM-79-238, B&R Engineering Memorandum, MCC Master List Revisions, March 22, 1979.

GEBR-2-81-182, GE Letter to B&R, on Increased Loads .

GEBR-2-81-189, GE Letter to B&R, on Increased Loads.

C-23

C.7 Contract S ecifications Contract 2, Division 2, Section 2A, Nuclear Steam Supply System.

Contract 41A, Division 15, Section 15A, Nuclear Valves.

Contract 41B, Division 15, Section 15A, Nuclear Valves.

Contract 42, Fisher Controls, Incd..

Contract 42A, Division 15, Section 15B, Control Valves, equality Class I.

Contract 47A, Division 16, Section 16A, Metal Clad Switchgear.

Contract 49, Division 16, Section 16A, Motor Control Centers.

Contract 62A, Division 16, Section 16A, Electrical Cable.

Contract 62B, Division 16, Section 16A, Electrical Cable.

Contract 215, Division 15, Section 158, Piping Systems, Section 15F, Valves, Section 15G, Specialties.

Contract 220, Johnson Controls, Incd.

C-24

C.8 Other Documentation Utilized for Investi ation C.8.1 Test Data 2993-112-1, Rev. 0, Ingersoll Rand Pump Test Data, Curve N-621, Pump Serial Number 047-3111, dated December'6, 1974.

2993-117-1, Rev. 1, Ingersoll Rand Pump Test Data, Curve N-155,

Characteristic of Centrifugal Speed-Torque Pump, Start With Open-Discharge.

2997-24, Rev. 1, Curve 388-AA-578, Speed-Torque-Current Curves for Induction Motor, RHR Pump Motors for Hanford II, B&R File 41A-00-0073 Rev. 3, Limitorque Corporation, Master Certification Sheet l.

lKR 41B-00-0108, Limitorque Motor Data.

WPPSS gA EEI-02-KNC-80-022, Test Repor t, Limitorque Valve Actuator gualification for Nuclear Power Station Services, Report B0058, Test per IEEE Standards 382-1972, 323-1974, 344-1975, by Limitor que Corporation, dated January 11, 1980.

Cable Pull Slips for Cables 2SM8-50 and 2MBBA-20.

BER 41A-00-8496, Motor Test Report for RHR-M0-3B.

gA Film 02-003-1254, Anchor Valve Company, Certified Operation Test Report for RHR-M0-24B.

gA Film 02-009-322, Report of Test Certification for RHR-M0-648.

gA Film 02-009-323, Fisher Conrol Company, Manufacturer Certification for RHR-M0-64B.

C-25

WPPSS SLT EDS-8, System Lineup Test for RHR-P-28, 4 Pages, dated May 28, 1981 and October 19, 1981.

WPPSS SLT EDS-l, System Lineup Test for RHR-P-28, dated January 8, 1982.

C.8.2 Standards and Re ulator Guides IEEE-141-1969, The Red Hook, Recommended Practice for Electric Power Distribution in Industrial Plants.

IEEE-279-1971, Criteria for Protection Systems in Nuclear Power Generating Stations.

IEEE-308-1974, Criteria for Class lE Power Systems in Nuclear Power Generating Stations.

IEEE-323-1971, gualifying Class lE Equipment for Nuclear Power Generating Stations.

IEEE-323-1974, gualifying Class 1E Equipment for Nuclear Power Generating Stations.

IEEE-382-1972, Type Test of Class lE Electric Valve Actuators for Nuclear Power Generating Stations.

IEEE-383-1974, Type Test of Class lE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations.

NEMA-MG-1-1978, Motor and Generator Standards.

NEMA-ICS-1-108, Service and Installation Conditions .

NEMA-ICS-2-321, AC General Purpose Class A Magnetic Controllers for Induction Motors, Rated in Horsepower, 600 V and less, 50 and 60 Hz.

C-26

NEMA-ICS-2-322, AC General Purpose Motor Control Centers.

NEMA-ICS-327, AC General Purpose Class A Magnetic Controllers for Induction Motors, Rated in Full Load and Locked Rotor Current, 600 V and less, 50 and 60 Hz.

IPCEA-S-68-516, Interim Publication 2, Cables Rated 5.0 KV and Less, Having Ozone Resistant Ethylene-Propylene-Rubber Integral Insulation and Jacket.

IPCEA-P-54-440, 2nd Ed., Ampacities of Cables in Open Cable Trays.

NFPA-70-1981, National Electrical Code.

ANSI-C37.04-1979, Rating Structure for AC High Voltage Circuit Breakers, Rated on a Symmetrical Current Basis.

ANSI-C37.06-1979, Preferred Ratings and Related Required Capabilities for AC High Voltage Circuit Breakers, Rated on a Symmetrical Current Bas is.

ANSI-C37.010-1972, Application Guide for AC High Voltage Circuit Breakers, Rated on a Symmetrical Current Basis.

ANSI-C37.010-1979, Application Guide for AC High Voltage Circuit Breakers, Rated on a Symmetrical Current Basis.

RG-1.131, Regulatory Guide, gualification Tests of Electric Cables, Field Splices and Connections for Nuclear Power Generating Stations.

C.8.3 Miscellaneous Other Documentation WPPSS WNP-2 Class lE Equipment List, dated March 2, 1983 and January 4, 1983.

C-27

GE ESM Book 3, GE Electrical Equipment Specification Manual, Application Guide for Systems and Utilization Equipment.

WX-AD-32-262, Westinghouse Application Data 32-262 for Type DHF Circuit Breakers.

PPM 10.25.13, WNP-2 Plant Procedure Manual, Electrical Maintenance Programs and Procedures, Westinghouse High Voltage Circuit Breakers.

238X184AD, Rev. 7, Par ts List for Residual Heat Removal System.

C-28.

SECTION 0 RFW SYSTEM REVIEW REFERENCES 0.1 Mechanical References D.l.l Desi n S ecifications General Electric Desi n S ecifications 22A719, Rev. 0, Feedwater Flow Measurement and Control.

22A2800, Rev. 1, Rated Steam Output Curve.

22A2801, Rev. 1, Reactor System Heat Balance - Rated.

22A2802, Rev. 1, Reactor System Heat Balance - 1055 of Rated.

22A2887, Rev. 6, Nuclear Boiler System.

22A3007, Rev. 5, BWR Equipment Envir onmental Interface Data.

22A3067, Rev. 3, Mechanical Equipment Separation.

22A3095AD, Rev. 1, Pressure Integrity of Piping and Equipment, Press.

Parts.

22A2907; Rev. 3, FW Control System (Steam Turbine Driven RFW Pumps).

22A2907AB, Rev. 1, Feedwater Control System.

Burns and Roe En ineer in Criteria Document Section E - Mechanical Engineering Criteria

Section F - Nuclear Power Engineering Design Criteria Section G - Instrumentation and Control Criteria Section I - Process Piping and Pipe Supports Westin house Thermal Performance Data Heat Balances .

AB095-1554-1205849 KW, Maximum Calculated, Not Guaranteed.

AB095-1555-1154745 KW, Maximum Guaranteed.

Industr Standards Heat Exchanger Institute Std. for Closed FW Htrs, 1st Ed., 1968.

American Petroleum Institute Std. RP-520.

D.1.2 Calculations 4.20.04 - Feedwater System - From Reactor Feed Pumps to the Reactor Vessel, 11-16-76.

4.25.01 - Reactor Feedwater System Pressure Drop Gale. , 3-13-78.

5.07.72 - Pressurization of M.S. Tunnel From an M.S. Line Break, 5-13-79.

5.07.73 - Pressurization of M.S. Tunnel From an F.W. Line Break, 8-14-79.

D-2

7.00.50, Sht. 5 - RFW-V-115A, B Flow Control Valve Sizing 5-18-72.

Sht. 6 - COND-V-149, Control Valve Sizing, 1-25-72.

Sht. 6A - RFW-FCV-15, Control Valve Sizing, 3-11-83.

w Sht. 5A, Rev. 1 - RFW Resizing of.RFW-PCU-15, 3-22-83.

0.1.3 Technical Memorandums TM 667 - Feedwater Delivery System 6-26-74.

TM 1010 - Oper ation of Feedwater Delivery System, 4-29-77.

D.1.4 Manuals Anchor Darling Valve Operation and Maintenance Manual, AVC-198.

Southwest Engineering Manual for Feedwater Heaters.

Velan Valve Instruction Manual.

Ingersoll-Rand Reactor Feedwater Pump Manual.

Delaval Reactor Feedpump Turbine Drive Instruction Book.

0.1.5 ~0rawin a Burns and Roe M504, Rev. 40, Condensate and Reactor Feedwater Flow Diagram.

M506, Rev. 40, Misc. Drains, Vents and Sealing Systems.

D-3

M529, Rev. 35, Nuclear Boiler, Main Steam Flow Diagram.

M645, Rev. 15, RFW and Cond. Piping Sections.

M200-27, Rev. 6, FW Piping In Containment: Line A.

M200-28, Rev. 5, FW Piping In Containment: Line B.

M200-334, Rev. 6, FW Piping, RFW Pumps to 86 Htr and Condenser.

M200-335, Rev. 7, FW Piping, RFW Pumps to Reactor.

M200-341, Rev. 3, Cond.; L.P. Htrs 5A and 5B to RFW Pumps.

Bovee and Grail Isometrics COND-385-1.4, Rev. 6, Seal Water to RFW Pumps lA and 1B

-385-5.6, Rev. 3, Seal Water to RFW Pumps lA and 1B RFW-413-1.5, Rev. 10, From FW Pump. 1A to Condenser

-6.8, Rev. 6, From FW Pump lA to Condenser

-414-1.5, Rev. 10, FW Pump 1B to Condenser

-6.8, Rev. 6, FW Pump 1B to Condenser

-415-1.5, Rev. 7, Recirc. Line, HP Htrs. to Condenser

-6 .7, Rev . 5, Recirc. Line, HP Htrs . to Condenser

-8.10, Rev. 6, Recirc. Line, HP Htrs. to Condenser

-11. 12, Rev. 7, Recirc. Line, HP Htrs. 'to Condenser

-13.14, Rev. 6, Recirc. Line, HP Htrs. to Condenser

-416-1.5, Rev. 5, From FW Pump lA and 1B to HP Htrs. 6A and 6B

-6.9, Rev. 5, FW Pumps lA and 1B to HP Htrs. 6A and 6B

-10.12, Rev. 9, FW Pumps to HP Htrs. 6A and 6B

-13.14, Rev. 7, FW Pump to HP Htrs. 6A and 6B D-4

Bovee and Cra i 1 Isometrics Cont 'd

-417-1.3, Rev. 5, HP Htr. 6A to Flow Meter

-4.5, Rev. 3, HP Htr. 6A and 6B to Flow Meters

-6.8, Rev. 3, HP Htr. 6A and 6B to flow Meters

-9.10, Rev. 2, HP Htr. 6A and 6B to Flow Meters

-ll.13, Rev. 2, HP Htr. 6A and 6B to Flow Meters

-418-1.2, Rev. 10, Flow Element to Cont. (Line A)

-3, Rev. 4, Flow Element to Cont. (Line A)

-4, Rev. 7, Cont. to Reactor Vessel (Line A)

-5.6, Rev. 5, Cont. to Reactor Vessel (Line A)

-7.8, Rev. 5, Cont. to Reactor Vessel (Line A)

-9.10, Rev. 7, Cont. to Reactor Vessel (Line A)

-11.12, Rev. 6, Cont. to Reactor Vessel (Line A)

-13, Rev. 6, Cont. to Reactor Vessel (Line A)

-419-1.2, Rev. 8, flow Meter to Cont. (Line B)

-3, Rev. 4, Flow Meter to Cont. (Line B)

-4, Rev. 4, Cont. to Reactor Vessel (Line B)

-5.7, Rev. 7, Cont. to Reactor Vessel (Line B)

'-8.9, Rev. 7, Cont. to Reactor Vessel (Line B)

-10.11, Rev. 5, Cont. to Reactor Vessel (Line B)

-12.13, Rev. 7, Cont. to Reactor Vessel (Line B)

-479-1.3, Rev. 2, FW Pump 1B to Hp Htrs. 6A and 6B

-480-1.4, Rev. 4, Bypass Line, RFW Pump Disch. to Hx6A Disch.

Vendor Drawin s CCI Control Valve, Dwg. 8921901077, Rev. H.

Anchor Darling Valve Owg. $ 3084-3, Rev. A.

Fisher Control Dwg. 852A8558, Rev. C.

I-R Pump Curve Dwg. 849413.

D-5

I-R Seal Injection Control Dwg. 82636-C-18C.

I-R CN Pump Owg. 8C-18X17CNGOOX4B.

I-R CN Pump Parts List, Owg. OC-18X17CN500X4.

Velan Owg. PP2-3319-N-33, Rev. J.

D.l.6 Memoranda WPBR-73-891, Containment Isolation Valves, 12-11-73.

BRWP-74-365, Containment Isolation Valves, 4-10-74.

WPBR-74-460, Containment Isolation Valves, 4-19-74.

EN-RLH-81-05, Containment Iso. and Testability Eval., 10-12-81.

'I 0.1.7 Contract S eci fi cati ons Cont. No. Award Date Item 2808-10 1-14-72 Feedwater Heaters 2808-11 A 2-18-72 Reactor Feed Pumps 2808-41 A 12-3-73 Nuclear Valves 2808-418 12-3-73 Nuclear Valves 2808-42A 5-13-74 Misc. Control Valves, Controllers and Acc.

2808-215 5-13-74 Mechanical Equipment Installation BEW Equipment Spec. 808-1004-352-00 (RFW-FCV-15) 0-6

0.1.8

~ ~ ~Re orts Anchor Darling Valve Design Report: 24"-9008 Check Valves.

Anchor Darling Material Certification Report for RFW-V-32A.

CCI Material Certification Report (RFW-FCV-15).

Velan Certificate of Compliance (RFW-V-65A).

D-7

0.2 Electrical References D.2.1 Desi n S ecifications BER Engineering Criteria Document, Section D, Electrical Engineering Criter ia.

BIIR Engineering Criteria Document, Appendix 3, Electrical Separation Practices, Rev. 1, 12-22-82.

D. 2.2 Calculations 2.02.02 (Main Plant Bus Load Calculations) Rev. 1, OL 6/15/81.

2.02.07 (Motor Control Centers Load Calculations), Rev. 1, DL 10-12-76.

2.03.07 (480 Volt Switchgear Short Circuit Calculations), Rev. 2, DL 1/20/77.

2.03.09, (MCC Short Circuit Calculations), Rev. 0, DL 1/24/78.

2.06.03, (Computer Run) - (Main One Line Voltage Drop Calculations),

Rev. 5, OL 1/18/80.

2.06.05 (Reactor Building. Feeder and Voltage Drop Calculations),

Rev. 3, OL 2/8/77.

2.06.06 (Turbine Generator Building, Feeder and Voltage Drop Calculations), Rev. 1, DL 12/16/74.

2.06.10 (480 Volt MCC Voltage Drop Calculation and Cable Sizing),

Rev. 1, OL 4/30/74.

D-8

2.12.00 (Relay Setting Time Curr ent Characteristic Curves), Rev. 5, DL 9/15/82.

2.12.12 (480 Volt Switchgear Relay Settings Motor Data), Rev. 1, DL 11/30/76.

D. 2.3 Technical Memorandum/En ineerin Memo EN-79-006, Rev. 0, 1/2/79, NCC Master List.

Tech. Nemo 1060, Rev. 2, Voltage Drop Study.

85R Engrg. Memo EM-79-239, Rev. 0, 3/22/79, MCC Master List Revision.

D.2.4 Manuals ITE Imperial Corporation, Rowan Controller Manual.

Reactor Feed Pump drive Turbine (Delaval), 2808-12.

Limitorque Manual, SNDI-170.

0.2.5 ~Drawin s The following fKR drawings with revision numbers listed were reviewed:

EWD-72E-001, NOV RFW-V-65A (B22-F065A), Rev. 1, 7/22/82.

EWD-72E-013, MOV RFW-V-109, Rev. 1, 2/3/83.

EWD-72E-015, MOV RFW-V-112A, Rev. 1, 7/22/82.

EWD-72E-037, Turb. RFW-DT-1A Turning Gear RFT-M-TNGA, Rev. 1, 7/22/82.

D-9

EWD-72E-039, Turb. RFW-DT-1A Main Oil Pump RFT-M-NOPA, Rev. 2, 8/31/82.

E502-2, Main One Line Diag., Rev. 19, 1/19/83.

E503-1, Aux. One Line Diag., Rev. 15, 3/21/83.

E503-6, Aux. One Line Diag., Rev. 26, 3/22/83.

E515-1, Breaker Setting 480V Swgr. SL-11 to SL-31, Rev. 1, 10/19/81.

f515-3, Breaker Setting 480V Swgr. SL-63 to SL-81, Rev. 2, 2/20/82.

E528-1, NCC Equip. Overload Summary NCC-NC-lA, Rev. 1, 12/17/82.

E528-2, NCC Equip. Overload Summary NCC-MC-lB, Rev. 2, 11/17/82.

E535-3A, Connection Wiring Diag. Motor Contr ol Center, Rev. 9, 12/07/82.

E535-3B, Connection Wiring Diag. Motor Control Center, Rev. 10, 2/1/83.

E535-10A, Connection Wiring Diag. Motor Control Center, Rev. 11, 4/13/82.

E535-10B, Connection Wiring Diag. Motor Control Center, Rev. 13, 2/1/83.

E528-27, MCC Equip. Overload Summary MCC-MC-7C, Rev. 0, 12/17/82.

E537-19A, Connection Wiring Diag. Control Room Term. Cabinet, Rev.

6, 4/4/83.

D-lo

E550, Cable Schedule - Power, Rev. 34, 12/7/82.

E558-2, Turb. Gen. Bldg. Grounding Plans and Details, Rev. 4, 4/12/82.

E902-3, Turb. Gen. Bldg. Grnd. Fl. El. 441'-0" Location Plan Cable Tray Nodes, Rev. 1, 7/16/75.

E918, Reactor Bldg. El. 501'-0" Location Plan Cable Tray Nodes, Rev.

11, 4/6/83.

E929, Radwaste and Control Bldg. El. 467'-0" Location Plan Cable

. Tray Nodes, Rev; 10, 4/6/83.

E933, Radwaste and Control Bldg. Misc. Elev's. Location Plan Cable Tray Nodes, Rev. 4, 4/6/83.

E935-4, Radwaste and Control Bldg. - Section "4-4" Locations Cable Tray Nodes, Rev. 8, 4/6/83.

Other Vendor Drawin s Reviewed B&R File No. 4900 0001, ITE Imperial Corp., MCC Layout for MCC-MC-lB.

B&R File No. 4900 0035, ITE Imperial Corp., MCC Layout for MCC-MC-7C.

B&R File No. 1200 0003, Console Oil Diagram (Delaval Turbine, Inc.).

B&R File No. 41A-00-0073, Limitorque Corp.

B&R File No. 43-00-0061, Walworth Co.

B&R File No. 43-00-0112, Walworth Co.

GE Motor for Turning Gear, DD-17271.

D.2.6 Memoranda Included in Section D.2.3 D.2.7 Contract S ecifications: BIIR i) Contract Specification 2808-12, Reactor Feed Pump Turbine - Bid Issue, BD-24.

ii ) Contract Specification 2808-41, Nuclear Valves, Division 15, Section 15A.

iii ) Contract Specification 2808-43, Standard Cast or Forged Steel Valves, Division 15, Section 15A.

'v) v)

Contract Specification 2808-49, Motor Control Centers, Division 16, Section 16A.

Contract Specification 2808-62A and 62B, Electrical Cable.

O.2.8 Others Ven dor Dr awin s Veelan Engrg. Co., Test Reports for RFW-M0-65A, (Veelan Order No.

P2-3313-N).

Walworth Co., Test Report for RFW-M0-109, RFW-M0-112A, (Walworth Co., P.O. PP 32500, 5/25/77).

Delaval Certificate of Conformance for RFT-M-MOPA, RFT-M-TNGA.

Bussman Fuse Manufacturing, Part III, Component Protection for Electrical Systems.

D-12

Industr NEMA Codes and Standards MG-1, Para. MG1-1.26 (Totally Enclosed Machine).

0 NEMA ICS-2-322.21 (Combination Motor Control Unit Ratings).

NEMA ICS-2-321.41 (Short Time Capability).

IPCEA - No. P-54-440, "Ampacities, Cables in Open Top Cable Trays".

NfPA 70-1981, "National Electric Code".

ANSI C37.04-1979 (American National Standard Rating Structure for AC High Voltage Circuit Breakers Rated on a Symmetrical Current Basis).

ANSI C37.010-1979 (American National Standard). IEEE Application Guide for AC High Voltage Circuit Breakers Rated on a Symmetrical Current Basis.

IEEE-279-1971 (Criteria for Protection Systems for Nuclear Power Generating Stations).

IEEE-308-1974 (Criteria for Class lE Power Systems for Nuclear Power Generating Stations).

IEEE-323-1974 (gualifying Class "lE Equipment for Nuclear Power Generating Stations).

IEEE-344-1975 (Recommended Practices for Seismic qualification of Class lE Equipment for Nuclear Power Generating Stations).

IEEE-382-1974 (Type Test of Class lE Electric Valve Operators for Nuclear Power Generating Stations).

D-13

IEEE-383-1974 (Type Test of Class lE Electric Cables, Field Splices and Connections for Nuclear Power Generating Stations.

IEEE-384-1977 (Criteria for Independence of Class 1E Equipment and Circuits).

R-G-1.75, Physical Independence of Electric Systems.

NUREG 0588, Category 2, (Environmental gualification of Class 1E Equi pment).

0-14

I 0.3 Instrumentation and Control References 0.3.1 S ecifications General Electric and Burns and. Roe Inc.)

22A2907, Rev. 3, "Feedwater Control System (Steam Driven Turbine Reactor Feed Pumps ", 3/28/74.

22A2907AB, Rev. 1, "Feedwater Control System" Data Sheet, 8/16/71..

22A2719, Rev. 2, "Feedwater Flow Measurement and Control" Design Specification, Dated 7/26/71.

22A2719AB, Rev. 0, "Feedwater Flow Measurement and Control" BWR Plant Requirements Specification, 7/26/71.

732E120AD, "IED - Feedwater Control System, Turbine Feed Pumps",

Rev. 3.

807E160TC, "Feedwater System" Elementary Diagram, Sheets 1, Rev. 12; 2, Rev. 12; 3, Rev . 10; 4, Rev . 12; 5, Rev . 8.

807E153TC, "Nuclear Boiler Process Instrumentation System" Elementary Diagram, Sheets: 1, Rev. 13; 1A, Rev. 10; 2, Rev. 11; 3, Rev. 3; 4, Rev. 12.

DL807E160TC, "Device List - System Elementary C34A", (6/15/78).

234A9304TC, "IDS - Feedwater Control System", Dated 7/6/73.

GEK-71337, "Instrumentation Manual for Vendor Supplied Instruments",

(Feedwater Control System Device CVI Data), Volumes I, II, III, IV, V and VI.

0-15

22A3067, Rev. 3, "Mechanical Equipment Separation" System Design Specification, Dated 8/31/75.

22A7416, Rev. 0, "Electr ical Equipment, Separation for Safeguards Systems" Design Specification, Dated 2/19/81.

22A3085, Rev. 3, "Remote Shutdown System" Design Specification, Dated 5/25/79.

22A3007, Rev. 1, "Engineering Safeguards Systems, Criteria for Testability of Instrumentation and Controls", 12/1/71.

22A8658, Rev. 1, "General Requirements for Motor Operated Valve Actuators", Dated 5/17/71.

GEK-71314, "Feedwater Control System, 0 and M Manual", Dated 9/78.

166B7135A, "Information Document - Feedwater Dynamic Analysis Data",

Sheets: 1, Rev . C; 2, Rev . C; 3, Rev. C; 4, Rev. C; 5, Rev. C; 6, Rev. C; 7, Rev. C; 8, Rev. C; 9, Rev. C; 10, Rev. C; 10A, Rev. C; ll, Rev . C; 12, Rev. C; 13, Rev. C; 14, Rev . C; 15, Rev. C; 16, Rev. C; 17, Rev. C; 18, Rev. C Burns and Roe Engineering Design Criteria, Section F, Table 7.4-3, Equipment Classifications.

22A3039, Rev. 1, "Process Instrumentation", 3/26/73, Design Specification Para. 4.2.2, 4.3.3, Figures 12, 1.8.10, Para. 4.2.4;

4. 2.5.

22A3041, Rev. 1, "Essential Components", 3/14/77.

22A3746, Rev. 1, "Local Instrument Panels" Design Specification, 1/21/74.

D-16

22A3008, Rev. 5, "BNR Equipment Environmental Interface Data",

(4/8/77), Design Specification.

239X241AO, "Feedwater Control System (Turbine Driven Reactor Feed Pumps) - Parts List", Rev. 10, Dated 6/4/80.

234A9301TC, Sheet 22, Rev. 1 (8/1/73), "IDS - Nuclear Boiler System".

22A3181AD, Rev. 0, "Flow Element (Main Steam Restrictor" System Design Specification and Data Sheet (11/13/73).

127D1835TC, Rev. 1 (7/19/73), "Main Steam Flow Instrument Panel A (H22-P015) .

21A9387AB, Rev. 0, "IDS - Feedwater Control System - Turbine Drive" (9/17/71), Sheet 5.

21A9430, Rev. 0, "Main Steam Flow Element", (ll/4/71).

22A2887AB, Rev. 4, Sheet 4, "Nuclear Boiler System Data Sheet" (1/10/75) .

163C1029TC, "Piping Diagram - Main Steam Flow Instrument Panel A (H22-P015), Rev. 2 (7/22/77).

12701845TC, Rev. 2 (7/22/77), "Connection Diagram - Main Steam Flow Instrument Panel A (H22-P015).

163C1183, Rev. 0, "Differential Pressure Transmitter Detail", 4/4/74.

12701826TC, Rev. 4, "Arr angement, Reactor Vessel Level and Pressure Instrument Panel A (H22-P004) ".

12701814TC, Rev. 3, "Piping Diagram, Reactor Vessel Level and Pressure Instrument Panel A (H22-P004)".

127D1827TC, Rev. 2, "Electrical Diagram, Reactor Vessel Level and Pressure Instrument Panel A (H22-P004)".

117C-4928, Rev. B, "Feedwater Flow Meter Section - Purchased Part" (Shows C34-N001A, B as a double section in which each section is double flanged (flanged at both ends), dated 2/16/71.

761E443, Rev. 1, "Primary Steam Piping Nuclear Boiler - Purchased Part", Dated 2/8/70 (shows C34-N001A, B Specifications).

131C7598, Sheet 1, Rev. 1, "Flow Meter Section - Feedwater Control System", Dated 6/1/71 (C34-N001A, B specification drawing), shows C34N001A, B as a double section in which the sections are flanged together only. The outer ends are for welding.

21A9414, Rev. 1, "Feedwater Flow Meter Section" - Purchase Specific 1/7/71 (has calibration procedures and materials, etc. specification for C34-N001A and B) entire document.

21A9414AB, Rev. 2, "Feedwater Flow Section" - Purchase Specification Data Sheet, Dated 8/24/73, entire document.

328X154TC, Section A, Rev . 11, "Shipping Group Parts List - Nuclear Boiler Local Instrumentation ".

238X178Al, Page 7, Rev. 22, "Nuclear Boiler System - Master Parts List" (shows B22-N041 temp. elements code, equipment and source classifications).

159C4520, Sheet 1, Rev. 6, "Temperature Element - Nuclear Boiler",

,(Details on 822-N041A or RFW-TE-41A).

159C4520, Sheet 2, Rev . 6, "Temperature Element - Nuclear Boiler",

(More B22-N041A details).

D-18

22A2887, Rev. 6, "Nuclear Boiler System", 1/29/79, Para. 4.11.3.3, Design Specification .

22A2718, Rev. 5, "Special Wire and Cable", 4/10/74, Para. 2.13.2, 2.13.4 (gives wiring type criteria and lead resistance criteria).

828E185TC, Rev. 4, "Arrangement, Nuclear Steam Supply Shutoff Temperature Recorder VB".

22A3041, Rev. 1, "Essential Components", 3/14/72, Design Specification.

22A8696, Rev. 1, "Seismic Requirements for Essential Class I Instrumentation", 3/7/78.

22A2702A, Rev. 1, "Seismic Design", 1/7/71, Design Specification.

22A3059, Rev. 1, Cleaning of Piping and Equipment", 6/24/75.

248A9393, Rev. 0, "General Use, Controller Assembly Data Sheet".

GE-l, Feedwater Control System."Preoperational Test Instruction" (12/12/77), Rev. 0.

STI-23X, Feedwater Control System Tune-Up Procedure, "Startup Test Instructions" (6/10/81), Rev. 2.

GEZ-6894, "Hanford 2 Nuclear Power Station Control Systems Design Report", R. W. Polomik, S. T. Chow (2/80), Chapter 7.

22A4152, Rev. B, "Startup Test Program", Sht. 53 (shows Feedwater Sys tern Control response performance cr i ter ia) .

22A2271AS, Rev. 1, "Preoperational Test Program" (shows Feedwater System).

D-19

22A2801, Rev. 1, "GE Reactor System Heat Balance - Rated" System Design Specification, Dated 1/24/72.

22A2802, Rev. 1, "GE Reactor System Heat Balance - 105K of Rated" System Design Specification, Dated 1/24/72.

22A2800, Rev. 2, "Rated Steam Output Curve" Design Specification, Dated 1/9/79.

22A3148, Rev. 1, "Heat Balance, Reactor System - 105K of Rated" Information Document, Dated 1/9/79.

22A3149, Rev. 1, "Heat Balance, Reactor System - Rated" Information Document, Dated 1/9/79.

P.O. 282-F9762, Rev. 0, "Temperature Element Product guality Checkl ist", Dated 9/17/74 Burns and Roe Engineering Criteria Document, Rev. 11, 3/16/82 Section G.

Instrumentation and Control, Section F Equipment Classification, Appendix 3, "WNP-2 Electrical Separation Practices", Rev. 1.

D; 3.2 Calculations 7.10.02, Rev. 3, "Flow Element Sizing Calculations", 10/26/76, Sheet 8.

Alden Research Laboratories Worchester Polytechnic Institute, "Calibration - Two 24" Flow Nozzle Assemblies, Serial Numbers N-1031, N-1032. The Peroatit Company Purchase Order Number L-58671-1565", Dated October, 1974, (Calibration Data for C34-N001A and C34-N001B).

Vickery - Simms ¹BC-N-1005-5, Orifice Bore Calculations.

D-20

D.3.3

~ ~ Technical Memorandum BRI Technical Memorandum 1010, "Operation of Feedwater Delivery System" (4/29/77), (with updated Exhibits and FE 8166B7135A drawings).

BRI Technical Memorandum 667, "Feedwater Delivery System" (6/26/74).

BRI Technical Memorandum 572, "Feedwater Control System" (9/21/73).

BRI Technical Memorandum 308, Rev. A, "System Description-Condensate/ Reactor Feed" (10/6/72).

0.3.4 Manuals Vendor Anchor Darling Valve Company, "Instrument Manual, Operator-Maintenance Instructions and Parts Catalog for WNP-2" (V-32A, 8, V-10A) B), WPPSS CVI 02518-00-75-1, 11/28/76.

Permutit Corporation Operating Instructions for C34-N001A and C34-N0018, Rev. 1, BRI AEF 02-11-0710.

Anchor Dar ling Co. Instruction Manual, Operator - Maintenance Instructions and Parts Catalog", CVI 02-41B-OO, Sht. 75, Issue l.

"Self Drag Flow Control Valve Operation and Maintenance Manual",

Babcock and Wilcox CVI 02-42D-OO, Sht. 12.

Woodward Governor Operation and Maintenance Manual Reactor Feedwater Turbines CVI 02-12-00, Sht. 16.

Fisher Technical Bulletin 62.1:546, dated 12/76, "Type 546, 546S and 546ST, Electro-Pneumatic Transducers.

0-21

0.3.5 ~0r awin s Burns and Roe Drawin s Mechanical M151, Rev. 0, "General Arrangement - Ground Floor Plan".

M152, Rev. 0, "General Arrangement - Mezzanine Floor Plan".

M153, Rev. 0, "General Arrangement - Operating Floor Plan".

M154, Rev. 0, "General Arrangement - Reactor Building and Miscellaneous Plans".

N502, Rev. 27, "Main and Exhaust Steam System, Turbine Generator Building".

M504, Rev. 36, "Flow Diagram, Condensate and Feedwater System".

N506, Rev. 28A, "Flow Diagram Miscellaneous Drains, Vents and Sealing Systems, Turbine Generator Building".

N509, Rev. 16, "Flow Diagram - Turbine Oil Purification and Transfer System, Turbine Generator Building".

N529, Rev. 28, "Nuclear Boiler System - Flow Diagram".

N610, Rev. 5, "Installation of Thermowells and Sample Probes".

M200, Sheet 335, Rev. 7, "Reactor Feedwater Piping, RFW Pumps to Reactor", 5/16/80.

D-22

N543, Rev. 25, "Flow Diagram - Reactor Building Primary Containment Cooling and Purging System".

N617, Sht. 64A, Rev. 6, "IR-64 Legend" Sht. 64B, Rev. 4, "Connection Diagram IR-64" Sht. 64C, Rev. 7, "IR-64 Arr angement" Sht. 64D, Rev. 4, "Connection Diagr am IR-64" Sht. 12A, Rev. 6, "Inst. Rack IR-12 Legend" Sht. 12B, Rev. 4, "Inst. Rack IR-12 Arrangement" Sht. 12C, Rev. 3, "Inst. Rack IR-12 Tubing Arrangement" Sht. 12E, Rev. 2, "Inst. Rack IR-12 Wiring" Sht. 12F, Rev. 4, "Inst. Rack IR-12 External Electrical Connections" Sht. 12G, Rev. 0, "Inst. Rack IR-12 External Electrical Connections" Sht. 12D, Rev. 5, "Inst. Rack IR-12 Tubing Arrangement Cont."

M619, Sht. 85, Rev. 5, "Inst. Rack IR-18 Connection Diagram" Sht. 110, Rev. 4, "IR-12 Instrument Connection Diagram" Sht. 112, Rev. 6, "IR-12 Instrument Connection Diagram" Sht. 142, Rev. 9, "IR-64 Reactor Building Inst. Rack" Sht. 104, Rev. 5, "Inst. Rack IR-9 Connection Diagram".

M621, Sht. 1, Rev. 5, "Panel/Console/Cabinet/Rack Classification List" Sht. 4, Rev. 2, "Panel/Console/Rack List".

M620, Sht . 504-17, Rev. 0, "H. P. Heater Outlet Line N.O. Valve Control Logic Diagram" Sht. 506-10, Rev. 1, "Reactor Feedwater Pump Turbine RFW-DT-1A Drain Valve Control Sch. and Logic Diagram".

N200-335, Rev. 7, "Reactor Feedwater Piping RFW Pumps to Reactors",

5/22/80.

D-23

M502, Rev. 27, "Flow Diagram - Main and Exhaust Steam System, T.G.

Buil ding", 2/25/83.

M504, Rev. 36, "Flow Diagram - Feedwater and Condensate System, T.G.

Buil ding", 1/14/83.

M506, Rev. 28A, "Flow Diagram - Misc. Drains, Vents and Sealing System T.G. Building", 1/28/83.

M509, Rev. 16, "Flow Diagram - Turbine Oil Purification and Transfer System T.G. Building", 12/10/82.

M529, Rev; 28, "Flow Diagram - Nuclear Blr. Main Steam System, Reactor Building", 3/4/83.

M610, Rev. 5, "Installation of Sample Probes and Thermowells",

10/25/82.

N617-12A, Rev. 6, "Instrument Rack IR-12 Legend", 5/26/82.

M617-12B, Rev. 4, "Dwg. Voided by PED 220-I-0772", 10/08/81.

M617-12C, Rev. 3, "Instrument Rack IR-12 Tubing Arrangement",

5/26/82.

M617-12D, Rev. 5, "Instrument Rack IR-12 Tubing Arrangement",

5/26/82.

M617-12E, Rev. 2, "Dwg. Voided by PED 220-I-0772", ll/13/81.

M617-12F, Rev. 4, "Owg. Voided by PED 220-I-0772 Electrical Connections" 10/12/79.

D-24

M617-12G, Rev. 0, "Instrument Rack IR-12 External Electrical Connections".

M617-64A, Rev. 6, "Instrument Rack IR-64 Legend", 2/3/83.

M617-64B, Rev. 4, "Owg. Voided by PED 220-I-0772", 12/18/81.

M617-64C, Rev. 7, "Instrument Rack IR-64 Tubing", 5/26/82.

M617-640, Rev. 4, "Dwg. Voided by PED 220-I-0772", 12/28/81.

M619-85, Rev. 5, "IR-1B Reactor feed Pump 1B Instrument Rack",

3/14/83.

M619-142, Rev. 9, "IR-64 Reactor Building Instrument Rack El.

501'-0", Div. II", 3/14/83.

M620-504-17, Rev. 0, "H.P. Htr. Outlet Line M.O. Valve Control Logic Diagram", 9/7/76.

M620-506-10, Rev. 1, "Reactor Feedwater Pump Turbine RFW-DT-lA Drain Valve Control Schematic and Logic Diagram", 3/1/76.

M621-1, Rev. 5, "PNL Console Cabinet Rack List", 6/12/82.

M621-4, Rev. 2, "PNL Console Cabinet Rack List", 4/14/77.

Various Vendor Or awin s Control Components Inc. Drawing No. 9225, Rev. 11, "Self Drag Element 12" x 12" Angle Body" (1/6/77), BRI AEF ¹420-00-0015 (R FW-F CV-10) .

0-25

Woodward Governor Co. Drawing 89930-333, Sheet 2, "Control - 2301 Panel" (11-23-73).

Delaval Turbine Inc. Drawing C-72374, Sheets 9, Rev. 9; 13, Rev. 10, "Woodward Governor Schematic".

Delaval Turbine Inc. SCCA-2561, Rev. 2, "Reactor Feedpump Drives by Delaval Turbine Inc." (5/5/72), shows performance curves.

Ingersoll-Rand Inc. 049056, "Reactor Feed Pump curves" (7/10/72).

Johnson Controls Drawing 88-220-063.0, H22-P015, Sheet 1, Rev. 3, Sheet 1, Rev. 5, "Line Identification List", Rack H22-P015.

Johnson Controls Drawing PB-220-063.0, H22-P015, Sheet 2, Rev. 2; Sheet 3, Rev. 2; Sheet 4, Rev. 2; Sheet 5, Rev. 2; Sheet 5A, Rev. 0; Sheet 5B, Rev. 0; Sheet 5C, Rev. 0.

Perwtit Corpor ation Drawing 556-27984, Rev. 6, "Outline and Assembly - Feedwater Flow Pipe Section, Size (24") 20.668" X 10.334" (directly references D-4 and C-1 and C-2), Dated ll/28/73.

Permutit Corporation Drawing 8556-28016, Rev. 1, "Tube Bends Layout

- For Feedwater Flow Element - Size 20.668" X 10.334 (24" - Sch.

120), (directly references C-1 and C-2), Dated 12/29/71.

Permutit Corporation Drawing 0555-26992, Rev. 1, "Flow Straightener for 24,",Sch. 120 Pipe - Project Hanford II", Dated 9/27/73.

Johnson Controls, Inc. Drawing PD-220-2000 - FX-6A, Rev. 0, "Local Flow Test Connection WPPSS Nuclear Project No. 2", Dated 5/16/79 (shows C34-N001A flow test connections and orientations).

D-26

Bovee and Grail Inc. Drawing ¹RFW-418-1.2, Rev. 11, "From Flow Meter to Reactor Vessel (Line "A"), (shows C34-N001A and mounted to piping

- shows pressure connection orientation and piping dimensions),

Dated 7/15/75.

Bovee and Grail Drawing ¹RFW-418-1.2, Rev. 11, "From Flow Meter to Reactor Vessel (Line 'A'), Date 7/1575.

Jelco Drawing ¹757-D-622, Rev. C, "Tubing Arrangement IR-12", shows C34-N002A rack interconnections and rack connections.

Jelco Dr awing ¹757-E-675, Rev. 0, "Electrical Wiring Diagram, Instrument Rack IR-12", shows wiring.

Jelco Drawing ¹757-E-538, Rev. 0, "Instrument Assembly IR-12", shows rack placement of C34-N002A.

Jelco Drawing ¹757-E-535, Rev. 0, "Instrument Assembly IR-12", shows rack side views.

Circle A.W. Products Drawing ¹757-E-532, Rev. D, "Instrument Assembly IR-64".

Bovee and Grail Drawing ¹RFW-415-8.10, Rev. 6, "Drain From 30" Reactor Feedwater Line to High Pressure Condenser HX-9", 3/25/80.

Bovee and crail Construction Drawing ¹RFW-418-3, "Reactor 1

FW from Flow Meter to Reactor Vessel (Line "A"), Rev. 5.

Anchor Darling Valve Company Drawing ¹3084-3, Rev. B, "24 in. - 900¹ swing check valve, RFW-V-32A (B223-F032) ".

Jelco Controls Inc. Drawing ¹757-E-703, Rev. B, "Electrical Wiring Diagram IR-62".

0-27

Circle A.W. Products Co. Drawing ¹757-E-544, Rev. C, "Instrument Assembly, IR-9".

Jelco Controls Drawing ¹757-C-619, Rev. C, "Tubing Arrangement; Instrument Rack IR-9".

Johnson Controls Drawing ¹D-220-072.0 - RFT-18/IR-18, Rev. I, Line Identification List".

Johnson Controls Draw'ing ¹D-200-245-TG-441, Rev. 0, "Tubing Routing (As-Built) ".

Jelco Controls Drawing ¹757-E-506, Rev. 8, "Instrument Assembly, Rack 18".

Jelco Controls Drawing ¹757-E-611, Rev. C, "Tubing Arrangement, Rack 18" Jelco Controls Drawing ¹757-E-664, Rev. 8, "Electrical Wiring Diagram, Rack 18".

Circle A.W. Products Drawing ¹757-A-506, Rev. C, "Material List, Rack 18".

Control Components Inc. Drawing 9225, Rev. 2, "Self Drag Element 12" X 12" Angle Body", Shows technical data on RFW-FCV-10 (required output of RFW-E/P-10).

Jelco Controls Drawing 757-E-705, Rev. 8, "Electrical Wiring Diagram IR-64".

Circle A.W. Products Co. Drawing 757-E-597, Rev. C, "Instrument Assembly IR-62".

D-28

D.3.~ 6

~ Memor an da

/

Letter dated 4/12/82, no number, "RETRAN Initialization of WNP-2 Model (Draft) ".

Letter dated 9/15/80 to G. L. Gelhaus from F. J. Markowski/S. F.

Deng, "WNP-2 RETRAN Plant Model, Addition of Plant Control Systems".

WPPSS IOM to R. J. Barbee, Plant Technical from C. A. Fu, G.E. Std.

and A WNP-2, "FW Flow Meter Calibration", Dated 1/26/83.

IOM EN-RLH-81-05, "Containment Isolation and Testability Evaluation", R. L. Heid, 10/12/81.

BRWP-R0-82-92, "Containment Isolation Review", 3/18/82.

BRWP-R0-82-153, "Same as G-3", 6/1/82.

BRAD-41B-82-002, "Contract 41B RFW-V-32A, B, "Valve Seat Modifications - guotation Request", 1/21/82.

BRAD-41B-77-014, 6/ll/77, "Revised Thermal Transient Data for RFW Valves RFW-V-10A, B and RFW-V-32A, B".

Rosemount Inc., "Material Report and Certification GE Purchase Order No. 282-F-9762", Dated 2/2/74.

Rosemount Inc., "Certificate of Compliance and System Calibration Data Sheet", Dated 8/22/74.

D. 3.7 Contract S ecifications Technical Specification 2808-59, "Instrumentation and Control Boards".

D-29

Specification 2808-215, "Mechanical Equipment, Installation and Piping", Section 15B.

Specification 2808-220, "Instr umentation Installation" Division 50.

BRI Contract Bid Specification 2808-41, Attach. 1, "Nuclear Valve List - Nuclear Boiler, Reactor Feedwater", Page 15A-35, Rev. 3, 3/9/76, Pages 15A-157, 158, 166, 167, 140, Bid Issue 7/17/73.

Anchor Darling Contract Specification 2808-41, Part V, "Valve Specification".

Specification 2808-1, "NSSS Equipment Specifications".

Contract 2808-62, "Electrical Cable" Section 16A, Page 16A-6, (Guies Type L2 Cable for RFW-TF-41A).

Specification 2808-218, Section 50A, "Instrumentation and Control Board Installation".

Specification 2808-58, "Local Instrument Racks".

Specification 2808-218, "Electrical Installation", Section 50A,

, "Instrumentation and Control Boards Installation".

Johnson Controls Contract 220, Tubing Isometric Drawings.

WPPSS Document Change Control "FJN" gWNP2WBG-215-F-78-1401 (Contract Modification - Reactor Feedwater Calibration Standard).

D-30

D.3.8

~ ~ Other Instrument Society of America Reprint, "Survey of 'Information Concerning the Effects of Nonstandard Approach Conditions Upon Orifice and Venture Meters", P. S. Starrett, H. B. Voltage, P. F.

Halfpermy, July 1980.

System Description No. 72, "Feedwater System", WPPSS Nuclear Project No. 2, Rev. 0, 9/25/75, pages 29, 30.

WPPSS Power Ascension Test 8.2.23.0, "Feedwater System Power Ascension Test Procedure", rough draft.

BWR Systems Analysis Course, Vol. II, Tab. 15, "Feedwater Level Control System" (6/6/81).

Instrument Society of America ISA-S26 (1968), "Dynamic Response Testing of Process Control Instrumentation".

WPPSS T/SU SPR-E-2156 (2/24/83), "RFW-FCV-10 Pressure Switch and Solenoid Valve".

WNP-2 FSAR, Para. 7.7.1.4, "Feedwater Control System"; 6.2.4, "Containment Isolation System"; 10.4.7.3,".

Code of Federal Regulations.10CFR50, Appendix A, Criterion 55, Page 402.

NRC NUREG-0800, "Standard Review Plan", Para. 6.2.4, "Containment Isolation System", Rev. 2 (7/81).

. D-31

D.5.4 En ineerin Mechanics References

'0 D.5.4.1 Desi n Re uirement References M400-3 Engineering Criteria Document Appendix 2 Pipe Support Design Guide.

Technical Memorandum 1271, (/II Equipment Nozzle Allowable Loads 6/14/82.

D.5 .4.2 Calculations 8.42.8000 Revision 1 Pipe Stress Code .

8.16.2013 Hanger Design Calculation for RFW-24.

8.16.4983 Hanger Design Calculation for RFW-944N.

8.16.72.1 Hanger Design Calculation for RFW-943N, RFW-21, RFW-17.

0-32

SECTION E - SYSTEMS INTERACTIVE REVIEW REFERENCES E.l Fire Protection 1.1.1 ~E WNP-2, Final Safety Analysis Report, Appendix F, Ammendment 26 10CFR50, Appendix R.

APCSB 9.5-1, Appendix A, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976.

E.l.2 Calculations 2.06.04, Rev. 1, Radwaste Bldg. /Control Bldg. Feeder and Voltage Drop Calculations.

2.06.05, Rev. 3, Reactor Bldg. Feeder and Voltage Drop Calculations.

2.07.01, Rev. 2, High Voltage Cable Sizing - Ampocities and Conduits.

2.07.03, Rev. 1,. A.C. Motor Control Center Bus and Cable Sizing.

E.1.2 Technical Memorandum TM 1227, Rev. 3, Fire Protection Study, 4/22/82.

TM 1272, Rev. 2, Thermo-lag Fire Barriers for Electrical Cables, Cable Ampocity Derating, 10/22/82.

E.2 Pi e Break/Missile Evaluation/Jet Im in ment Fallin Ob ects Floodin E.2.1 ~S 22A2625, System Criteria and Application for Protection Against the Effects of Pipe Breaks, June 15, 1973.

22A3046, Rev. 1, Core Standby Cooling System Network Design Specifications, 7/14/77.

22A2802, Rev. 2, GE Reactor System Heat Balance 105K Rated Power.

BRI Engineering Criteria Document.

E.2.2 Calculations 5.49.050, Rev. 1, Pipe Break Analysis, Inside Containment.

5.49.051, Rev. 1, Target Determination, Pipe Breaks Inside Containment, 12/17/82.

5.49.052, Rev. 1, Shutdown Analysis for Pipe Breaks Inside Containment.

5.51.050, Rev. 1, Pipe Break Analysis, Outside Containment 5.51.051, Rev. 1, Target Resolution for Postulated Targets Outside Containment.

5.51.052, Safe Shutdown Analysis Outside Containment.

1 8.01.51, Rev. 0, WPPSS N.P. No. 2, LPCS Pipe Whip Analysis.

5.49.056', Rev. 3, Target Resolution for Postulated Targets Inside Containment, Draft.

SVIII, Vol. 81, Radwaste Missile Barriers, NG Sets 1 and 2 5.50.51, Target Oetermination for Credible Missiles Outside Con tainment, 6/25/82.

E.2. 3 Technical Memorandum TN 1020, Rev. 1, Regulatory Guide 1.46, Recommendation Concerning Implementation, 10/28/77.

TM 1085, Rev. 1, Pipe Break Outside of Containment - Structural Effects, 10/6/78.

TM 1151, Criteria for the Pipe Break and Missile Redundancy Evaluation Outside Primary Containment, 6/27/79.

E.2.4 ~Drawin s Electrical E-550 E-551 Mech an i cal M-519 M-520 M-521 M-523 M-529 M-530 N-543 N-557 E-3

Stru ctur al S-794 S-918 S-1001, Rev. 10 S-1000, Rev. 21 S-783, Rev. 12 S-1024, Rev. 2 Isometric RWCU-895-8.12 RWCU-894-14. 21 RW CU-277-1. 3 RWCU-895-1. 7 D220-X-106 D 220-X-108 D-220-031.0-IR-68 CEP-625-11.12 M200 Sht. 129 RCIC-664-1.7 M200 Sheet 126 M200 Sheet 128 D220-7.1-X-78(e)

E DR-571-4. 5 HPCS-630-31. 33 HPCS-630-29. 30 ED-A-9 ED-A-16 ED-A-6 ED-A-5 M-200, Sheet 2 RHR-4434-1

Hanger RWCU-181 RWCU-928N RWCU-238 HPCS-64 HPCS-66 F20APKD500X4-C IR-RHR Pump Detail 238X178AD 239X527AD 239X 241 AD 238X201AD E.2.5 Other WNP-2, Final Safety Analysis Report.

NUREG 75/087, Standard Review Plan, Sections 3.5.1, 3.5.2, 3.6.1, 3.6.2.

Regulatory Guide 1.46.

Regulatory Guide 1.70.

BTP NEB 3-1 and APCSB 3-1, Section B.3, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment .

"Proposed ASNE Non-Mandatory Appendix - Design Rules for Pipe Whip Restraints" Article L-1000, NF 54, N/0 77-66 N76-6 January 1980.

Crane Technical Paper 8410, "Flow of Fluids Through Valves, Fittings, and Pipe".

E-5

ASME Boiler and Pressure Vessel Code, Section III, Appendix I.

AISC 7th Edition, "Manual of Steel Construct~on", June 1973.

American National Standard ANS-58.2, "Design Basis for Protection of Nuclear Power Plants Against Effects of Postulated Pipe Rupture",

ANS I-176.

Teledyne Engineering Services Technical Report TR-4536-1 Missile Impact Analysis, November 7, 1980.

Hexcel Manual TSB122 - Design Data for Preliminary Selection of Honeycomb Energy Absorption Systems Gwaltney, R. C., "Missile Generation and Protection in Light Water Cooled Power Reactor Plant", Oak Ridge National Laboratory.

R. P. Kennedy, "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects", Holmes 5 0)

Narver, Inc., September 1975.

BC-TOP-9A, "Design of Structures for Missile Impact", Bechtel Power Corporation, September 1974.

ANSI 177-1974, Plant Design Against Missiles.

E-6

E.3

~ /uglification of Safet

~ ~ ~

Related E ui ment for Environmental

~

Conditions and 0 namic Loads E.3.1 Calculations Supply System Calculations:

NE-02-81-06-0, August 13, 1982, "WNP-2 Subcompartment Temperature and Pressure Analysis for Postulated High Energy Pipe Breaks in the Reactor Building".

NE-02-81-07-0, September 10, 1982, "Postulated Pipe Break of 4" RCIC(13)-4 in RCIC Pump Room (R15) and Room (R112) Above RCIC Pump Room".

NE-02-81-08-0, September 8, 1982, "Postulated Pipe Break of 4" RCIC(13)-4 in Room (R113) Above RHR Pump 2C Room".

NE-02-81-09-0, September 10, 1982, "Postulated Pipe Break of 4" RCIC(13)-4 in TIP Room (R308) ".

NE-02-81-13-0, September 10, 1982, "Postulated Pipe Break of 6" RWCU(2)-4 in the Valve Room (R313) Above TIP Room".

NE-02-81-14-0, September 16, 1982, "Postulated Pipe Break of 6" RWCU(2)-4 in Valve Room (R408) North of Containment EL 522'.

NE-02-81-15-0, December 16, 1982, "Postulated Pipe Break of 4" RWCU(l)-4 in RWCU Pump Room (R406 or R407}".

NE-02-81-16-0, September 14, 1982, "Postulated Pipe Break of 6" RWCU(l}-4 in Valve Room (R409) Above RWCU Pump Rooms".

E-7

NE-02-81-17-0, December 16, 1982, "Postulated Pipe Break of 6" RWCU(2)-4 in Valve Room (R509) North of Containment EL 548'".

NE-P2-81-18-0, November 5, 1982, "Postulated Pipe Break of 6" RWCU(l)-4 in Valve Room (R511) South of Containment EL 548'".

NE-02-81-19-0, December 16, 1982, "Postulated Pipe Break of 6" RWCU(1)-4 in the RWCU Heat Exchanger Room (R510)".

NE-02-81-20-0, December 16, 1982, "Postulated Pipe Break of Auxiliary Steam Line".

NE-02-82-41-0, September 10, 1982, "Cooldown of Reactor Building Rooms Followng a Pipe Break - Computer Model".

BRI Calculations:

5.07.14.1, October 29, 1976, "Blowdown From 4" AS(ll)-2".

5.07.31, October 22, 1976, "Volume and Vent Area for Reactor Building".

5.07.32, October 25, 1976, "Pressurization of HPCS Rooms Rll/R106 (El. 422'3")".

5.07.62, September 21, 1979, "Pressurization of Rooms 509/510 at El.

545'.

5.07.59.2, September 20, 1979, "Modification of Valve Room 408 at El. 522'.

E-8

~

2.3.2

~ ~ Other ANCR-NUREG-1335, September 1976, "RELAP4/M005 A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems - User's Manual".

NUREG/CR-1185, Addendum 1, June 1980, "COMPARE-M001 Code Addendum" and LA-7199-MS, March 1978, "COMPARE-MODl: A Code for the Transient Analysis of Volumes with Heat Sinks, Flowing Vents, and Ooors".

NUREG-0800, July 1981, "US NRC Standard Review Plan".

NUREG-0588, Rev. 1, "Interim Staff Position on Environmental qualification of Safety-Related Electrical Equipment".

WNP-2 Environmental qualification Report for Safety Related Equipment, September 1982.

WNP-2 Oynamic qualification Report for Safety Related Equipment, September 1982.

0 E.4 Structur al Members

~Si f i WNP-2 Final Safety Analysis Report BRI Engineering Criteria Document E.4.2 Calculations SV-184, Pipe Break in Main Steam Tunnel.

SIII-18, Turbine Generator Building - Operating Floor.

SV-14, Reactor Building Elevation 441'-0" and 444'-0".

5.51.050, Rev. 1, Pipe Break Analysis Outside Containment.

E-10

E.5 Instrument Racks Contract 2808-58; Local Instrument Racks E.5.2 ~Drawin s BRI:

M 621, Rev. 5, Instrument Rack List M 62la, Rev. 2, Instrument Rack List M 567, Rev. 7, Reactor Building General Arrangement M 568, Rev. 23, Reactor Building General Arr angement M 569, Rev. 23, Reactor Building General Arrangement M 584, Rev. 6, Standby Service Water Pump House Arrangment S 540, Rev. 8, Pump House Instrument Rack Supports S 1083, Rev. 1, Reactor Building Instrument Rack Supports E538-15VF-1, Rev. 0, Arrangement Drawing For IR-21 E538-15VF-2, Rev. 0, -Arrangement Drawing For IR-22 E538-16VF-l, Rev. 0, Arranyaement Drawing For IR-24 E538-16VF-2, Rev. 0, Arrangmement Drawing For IR-25

E538-16VF-3, Rev. 0, Arrangmement Drawing For IR-26 E538-18YF-1, Rev. 0, Arrangmement Drawing For IR-61 E538-18VF-2, Rev. 1, Arrangmement Drawing For IR-62 E538-18VF-3, Rev. 2, Arrangement Drawing For IR-63 E538-19VF-1, Rev. 0, Arrangmement Drawing For IR-64 E538-19VF-2, Rev. 0, Arrangmement Drawing For IR-65 P

E538-19YF-3, Rev. 1, Arrangmement Drawing For IR-66 E538-20VF-l, Rev. 1, Arrangmement Drawing For IR-67 E538-20VF-2, Rev. 0, Arrangmement Drawing For IR-68 E538-21VF-l, Rev. 1, Arrangmement Drawing For IR-69 E538-21VF-2, Rev. 0, Arrangmement Drawing For IR-70 E538-22YF-l, Rev. 0, Arrangmement Drawing For IR-71 E538-22YF-2, Rev. 0, Arrangmement Drawing For IR-72 E538-23VF-l, Rev. 0, Arrangnement Drawing For IR-73 E538-23VF-2, Rev. 0, Arrangmement Drawing For IR-74 E-12

Vendor Drawings: Jelco (Circle AW), Rack Outline and Details IR-21 CVI 02-58-00-32-1; -2; -3 Rev. 2-24-78.

IR-22 CVI 02-58-00-33-1; -2; -3 Rev. 1-31-78 IR-24 CVI 02-58-00-40-1; -2; 03 Rev. 12-5-77 IR-25 CVI 02-58-00-41-1; -2; -3 Rev. 1-31-78 IR-26 CVI 02-58-00-4-1; -2; -3 Rev. 1-31-78 IR-61 CVI 02-58-00-9-1; -2; -3 Rev. 1-31-78 IR-62 CVI 02-58-00-25-1; -2; -3 Rev. 1-31-78 IR-63 CVI 02-58-00-20-1; -2; -3 Rev. 1-31-78 IR-64 CVI 02-58-00-22-1; -2; -3 Rev. 1-31-78 IR-65 CVI 02-58-00-10-1; -2; -3 Rev. 1-31-78 IR-66 CVI 02-58-00-13-1; -2; -3 Rev. 3-16-78 IR-67 CVI 02-58-00-24-1;. -2; -3 Rev. 1-31-78 IR-68 CVI 02-58-00-11-1; -2; -3 Rev. 1-31-78 IR-69 CVI 02-58-00-19-1; -2; -3 Rev. 1-31-78 IR-70 CVI 02-58-00-21-1; -2; -3 Rev. 1-31-78 IR-71 CVI 02-58-00-12-1; -2; -3 Rev. 1-31-78 E-13

IR-72 CVI 02-58-00-18-1; -2; -3 Rev. 3-16-78 IR-73 CVI 02-58-00-23-1; -2; -3 Rev. 1-31-78 IR-74 CVI 02-58-00-50-1; -2; -3 Rev. 2-28-78 E.5.3 Other Equipment Environmental and Seismic qualification Oocumentation File; (ID 185002, 10-13-82.

Circle AW Products Letter to Burns and Roe of 1-6-77; CAWBR-58-77-051.

Burns and Roe Letter to Circle AW Products of 11-16-77; BRCAW-58-77-083.

E-14

~ <,

8310170212 WNP-2 AMENDMENT NO. 9 April 1980 Pressure loads due to pipe break do not necessarily peak with pipe whip and jet impingement loads; however, in the analysis, they are considered to act simultaneously.

With regard to pipe break, when high energy pipes under pressure fail, a fluid jet is created. The associated jet impingement force on a target as well as the reaction force exerted on the piping by the fluid jet force have a time history qualitatively presented in Figure 3.6-118. This force is conservatively idealized as a step function load. For the fluid forces 'associated with these pipe failures, see Table 3.6-6.

To obtain a solution for the actual complex system, the struc-ture is idealized by:an equivalent single degree of freedom system (see Figure 3.6-119) following the procedures described by J. M. Biggs in Chapter 5 of "Introduction to Structural-Dynamics" (Reference 3.6-1). The response of this mathemati-cal idealization to a step function load (jet impingement) or to a step function load concurrently with an impact loading (due to whipping pipe) involves an energy transfer from the impacting object to the impacted structure. The following exposition on how this energy transfer is addressed makes use o f procedures that have been. presented by the Bechtpl Corporation in its report on missile impact, Topical Report BC-TOP-9A, Revision 2 (Reference 3.6-13) .

3.6.1.6.3.2 Structural Response to Whipping Pipe Missile Impact Load

a. Discussion A method of energy-balance procedures is utilized in order to evaluate the structural response, when a missile impacts a target. The method uti-lizes the strain energy of the target at maxi-mum response to counteract the residual kinetic energy of the target or target missile com-bination that results from the missile impact.

A missile of mass Mm is postulated to strike a spring-backed target mass, Me, with a velocity, Vs. Since the actual coupled mass during impact varies, an estimated average effective target mass Mel is used to evaluate the inertia effects during impact. The impact of the missile is con-sidered plastic. This assumes that the missile remains in contact with the target after impact.

3.6-6d

r

~ M~M '7%1~ h5 +~79,+~~@ ~Q Qg~ $=PP,Melo< 15 LcA'ITALO o~~ ~g~p ~i

~ 7 LPP +<M~PER. og. L~q P, FKd 0 pT WS T4<<~. 4 5 A.W DzT s5 7~'P~oic.v pp ~ ~~g)~p ~g (+~T)$ 5(uO ~vs40 '%P +~4 AT'

%S AS s~c~~ >~

4Q gQ~p

~<~~~~ g,g qqg. WNP-2 AMENDMENT NO. 9 April 1980 The values of pr ratios, ductility should be less than the allowable

3. 6-1.

p, given in Table 3.6.1.6.3.3 Jet Impingement Jet impingement loads are loads that emanate from a break in a high energy line. It is postulated that the characteristics of the jet are such that the jet exits from a break opening in the pipe equal in area to the cross sectional area of the pipe itself (see Figure 3 6-117). The jet is postulated to travel conforming to the configuration of the cross sectional area of the pipe for a distance of five pipe diameters and then to diverge at an angle of divergence of 10'. For e jet thrust forces at the postulated breaks, see Table 3.6-6. Jet loads impacting structures are treated as equivalent static.

loads. A dynamic load factor is applied to the jet force ema-nating from the pipe and the resulting load is modified by an appropriate load factor according to its use in combination with other loads.. The structure impacted is then evaluated for structural capability.

3.6.1.6.4 Allowable Design Stresses and Strains For allowable design stresses and strains for reinforced concrete and structural steel, see 3.8.4.5 and Tables 3.8-12 and 3.8-17, except as modified in 3.6.1.6.4.1 and 3.6.1.4.2.

3.6.1.6.4.1 Pipe Whip Loading With or Without Other Loads The acceptability of pipe whip loading with or without other loads is considered from two aspects:

a. The overall structural response of the impacted structural element

b. The local damage sustained by the impacted struc-tural element.

The overall structural response is considered acceptable if the ductility ratio resulting from the loading does not exceed the maximum allowable ductility ratios as given in Table 3.6-

1. The determination of ductility ratios utilizes the proce-dures set forth in 3.6. 1.6.3 and the loading combinations in 3.6.1.6.6. In using these procedures, the allowable limit on section strength,-M , used in the d termination of yield displacement Xe, ( 3. 6. 1. 6. 3. 2e, Tables 3. 6-9, 3. 6-10 and Figure 3.6-120) is computed in accordance 3.6-6j

WNP-2 AMENDMENT NO. 25 June 1982 electrical division to which the component belongs; what the function of the component is; the various references, such as the drawings, in which the component is found; devices inter-connecting the component and another system; and additional information of this type. This coding facilitates storage of the input for retrieval at any time.

Table 3.6-6 lists the high energy design basis break loca-tions outside containment, the piping subsystems involved, the ipe diameter, the plan figure showing the piping subsys-tem, he maximum blowdown thrust or the thrust versus time f igure~ Q I Figures 3.6-12 through 3.6-36 illustrate and list the high energy break locations inside containment.

Moderate energy crack locations are postulated in accordance with Standard Review Plans 3.6.1 and 3.6.2,.

3.6.1.11.2 Method of Analysis for Postulated High Energy Fluid System Ruptures 3.6.1.11.2.1 Effects of Postulated Passive Component Failures Postulated pipe breaks in high energy fluid systems are in-vestigated to determine their effects on the ability to bring the plant to a safe shutdown and to limit the of fsite radio-logical consequences to an acceptable level as stated in 10CFR50.

On a case-by-case basis, the effects of pipe whip, jet im-pingement, and the resulting environmental conditions on safety-related equipment are evaluated. The effects of the postulated pipe break are dependent on the fluid oroperties of the system, the location and orientation of the oipe break, the proximity to safety-related systems, components, and structures, and the individual design limits of the safety-related systems, components, and structures.

3.6-7

HNP-2 AMENDMENT NO. 25 June 1982 3.6.1.11.3 Method of Analysis for Postulated Moderate Energy Fluid System Ruptures 3.6.1.11.3.1 Approach postulated ruptures in moderate energy fluid systems do not generate pipe whip. The analysis investigates the effects of the environment which results from such a postulated rupture on safety-related equipment, including the effects of ~ater spray.

The 'effects of the postulated moderate energy pipe cracks are dependent on the fluid properties, available f1uid reservoir, drain systems, location of the safety-related equipment, com-ponents, and structures, and the individual design limits of the saf ety-related equipment, components, and structures.

Where moderate energy pipe cracks are postulated in close proximity to high energy systems, the environmental analysis compares the effects of both high and moderate energy pipe ruptures. The most limiting case is evaluated for safe cold shutdown.

Moderate energy pipe cracks are postulated according to the criteria in 3.6.2.1.

3.6.1.11.3.2 Method of Analysis The locations of all postulated ruptures, resulting in through wall leakage cracks, are identified for later retrie-val. The analysis assumes that the spray resulting from a postulated moderate energy rupture causes the malfunction of all equipment not enclosed by watertight compartments.

Additionally, the most damaging single random active compo-nent failure in a system not effected by the postulated pas-siv component failure is postulated. jf the direct conse-quences of the pasive component failure results in a turbine or reactor trip, then of fsite power is assumed unavailable.

3.6.1.11.4 Summary of Analysis c

gana~l'es discussed in 3.6. 1.11. 2 and 3.6.1.11. 3 ~~~

identify a~ location where a postulated passive component

3. 6-8

WNP-2 AMENDMENT NO. 25 June 1982 Impacted pipes of smaller nominal diameter than the impacting pipe are assumed to fail, regard-less of wall thickness of impacted pipe. Im-pacted pipes of both larger nominal diameter and thinner wall thickness than the impacting pipe are assumed to develop through wall leakage cracks.

c. Additionally, a single random active component not affected by a) and b) is assumed to malfunc-tion. Should a) or b) result in a turbine gen-erator or reactor trip, then offsite power is assumed unavailable.

d. After a), b), and c) above have been valuated, possible shutdown modes are analyzed. If shut-down is possible, the postulated passive compon-ent failure is not significant from a safety standpoint.

W

e. Should alternate shutdown modes not be available then:

1. Reroute or relocate cable,'ipe, 'or equip-ment to prevent loss of function.

2. 'If (1) is not feasible, shield the adversely affected component(s) to prevent loss of function.

f. The flooding and environmental effects of mode-rate energy failure are evaluated to determine whether, they are more severe than the high en-ergy breaks and are addressed in 3.6.1.15.

The area temperature is evaluated by determining the Limiting postulated pipe break and using RELAP4/NOD5 (Reference 3.6-21). The limiting pipe break for temperature analysis is that pipe break giving the highest energy release rate over the longest blowdown period.

The effects of flooding are evaluated by determining the lim-iting pipe break and calculating the effects of the ft.uid release. The limiting pipe break for flooding analysis is that pipe break with the highest mass flow rate over the longest blowdown period.

Peak differential pressure analysis results are provided in Table 3.6-12 and discussed in 3.6.1.20.

~>5i ~ ~15

3. 6-10

WNP-2 AMENDMENT NO. 25 June 1982

~~@~~+ MgQ QP ac@~ oK%~~~

~

% MPH %TKQ + +A~ R( LA~ ~ ~ p failure in a high or moderate energy syste ecluded t safe shutdown and cooling of the reactor This analysis by actual examination of the plant is under-taken to provide results based on as-built conditions.

Design drawings are used to supplement the study in cases where piping or equipment have not been installed. Prior to fuel load, a walkdown of the plant is performed to verify the results of the analysis and confirm that all design modifica-tions have been implemented.

Piping layouts for areas containing high and moderate energy lines, whose failure can af'feet the performance of safety-related equipment, are presented as Figure . 6-43 through 3.6-62, inclusive.

Section 3.6. 1. 11 discusses in,detai the methods used to dem-onstrate that no single postulated passive component failure, in conjunction with a single active component failure, pre-cludes safe shutdown of the plant.

The following should serve to further clarify the method of analysis:

a. The forces developed at each postulated high energy pipe break are determined by the methods of 3.6.2.2. The effects of the resultant pipe whip and jet impingement are evaluated. Credit is taken for automatic isolation and operator action to mitigate the consequences of the post-ulated pipe break, if the equipment required for this function is not affected by the break or included in 3.6.1.11.4(c) below.

b. As a first step, all equipment impacted by the whipping pipe or jet is assumed to fail. Kf the equipment is required for safe cold shutdown or accident mitigation, a detailed analysis is per-formed to determine if the equipment will ac-tually fail. Structures contacted by the whip-ping pipe or jet are evaluated for structural adequacy by the methods of 3.6.2.2.

3, 6-9

NNP-2 ANENDHENT NO. 25 June 1982 3.6.1.13 Electrical Equipment Pnvironmental Qualifications All electrical systems, necessary for safe shutdown and nec-essary to maintain the plant in a safe shutdown condition, are designed to remain functional in the general area envir-onment resulting from a high energy line br ak or from leak-age cracks in moderate energy piping. Specif ic equipment is either:

a. Designed to remain functional as long as neces-sary in the general area environment.

b. Isolated from the general area environment in compartments capable of maintaining normal equipment operating conditions.

Certain rotating equipment cannot be designed to function in the more severe, Local steam environment. However, due to physical separation, rotating equipment, of not more than one subsystem, is exposed to the local conditions which exceed the generaL area accident environment. Required redundancy is thus maintained for safety equipment.

Refer to 3.11 for a more complete description of environmen-tal design of electrical equipment.

3. 6. 1. 13. 1 Ident i f icat ion or Equipment Safety equipment required to mitigate the consequences of an accident and place the reactor in a cold shutdown condition i" Listed in Table 3.11-2. The table also indicates the ce-quired duration, following an accident, which equipment is required to ooerate.

3.6.1.13.2 Environmental Design Refer to 3.11 for a discussion of environmental presign and an analysis of safety-related electricaL components. The sec-tion identifies the safety-related equipment that must oper-ate in a hostile environment, and Table 3.11-2 indicates the postulated environmental envelop conditions'or both the gen-eral and local accident areas.

3.6.1.13.2 Jet Impingement Barriers AcT'~>acc<~T bAR,R,uÃs +VX

'PMv locO For esults of the steam system study, see 3.6. l. LL. 4.

alvsis indicates -<ar ~ eeauiL'ed reactor sa n-shutdown. Some room walls, floors, an d ceilings act as jet impingement barriers,

WNP-2 AMENDMENT NO. 25 June 1982 3.6.2.3.2 Jet Impingement Effect 3.6.2.3.2.1 Physical Separation The physical separation of different essential systems and components is used to ensure that the plant retains function of sufficient essential systems to assure safe shutdown in the event of a postulated LOCA, and subsequent generation of a jet stream together with an additional single random active component failure and the loss of offsite power.

Where physical separation cannot be used to protect systems, a detailed analysis is performed to determine the effects of jet impingement on their operability. If necessary, barriers are provided to protect structures, systems, and components required for a safe shutdown, to prevent offsite radiological consequences, and to mitigate the effects of. a LOCA.

3.6.2.3.2.2 Jet Impingement Evaluation The evaluation of the adequacy of physical separation in-cluded the inspection of all essential systems and their com-oonents that are necessary to start, operate, and control the essential systems required for safe shutdown. The evaluation i nc luded the fo1 low ing:

a. Review pipe break locations ' '. n-orientation and geometry.

b. Review effected equipment by both design drawing examination and plant walkdown.

c. Review signals that result in the actuation of essential systems.

d. Review s'gnals that are necessary to be returned to inside primarv containment, to ac" ivate the shutdown systems.

e. Review availability of power that is required inside primary containment to operate the essen-tial systems.

f. Review mechanical engineered safety systems re-quired for safe shutdown.

3.6-47

SUIIHARY DP SUBCOIIPARTHENT PRESSURE ANAI,YSIS ( Page 1 oE 2 Compartment I)here Break Occurs piping System Differential Pressure Hax imum Time Eleva- DiEfer- Differential oE the Design tion Room I inc ential Between the Peak Pressure fft. ) Number n s - r ~ic i on imari i Rooms /sec) Jpsi) 442 R14/Rll3 RHR Pump Rooms 4 RCIC (l3)-4 0.33 R14, R113/R206 0.33 0.50 0.33 R14, R113/R12, 0.33 0.50 0.33. R114 R14g R113/R15p 0 33 0 50 R112 R15/R112 RCiC Pump Room 4" RCIC (13)-14 0.51 R15e R112/R205 0.53 0.76 0.51 R15, Rl)2/R14g 0.53 0.76 R113 0.51 R15g R112/R6, 0.53 0.76 R116 471 R206 EI. 471'pen 4 AS (ll)-2 0.05 R206/R103, R105, 0.35 0.08 I'Ioor Area R106, R305, R308 R310, R306, R315 0.05 .R206/R114, R113, 0.35 0.08 R).12 Ql 0.05 R206/R116, R115 N 501 R 308 TIP Room 4 RC?C (13)-4 0.32 R308/R305, R206, 0.03 0.50 R313

50) R 308 PIP Room 6 RLICU ( 2) -4 R308/R305, R206, 0.35 R313 I 501 R3)3 EI. 510'alve 6 RNCU (2)-4 0.41 R313/R308, R400 0.35 0.60 Room R404 EI. 522'pen 8'RD (l2)-3 0.03 R404/R305, R504, 0.04 0.05 P)oor Area R508 (a) Tahi . appli..s to reactor builiiing seconiiary containment, exclusive of the main steam tunnel, tun- C Pl nel ventway, an) tunnel extension. 0 z M PJ CO 'Z M r9 0

TABLE 3. 6-11 DFSIGH LOAD IH AREAS ((HERE P IPIHG PAI LURE& OCCUR Differential llunp Loals Pipe Di f ferent ia 1 Temperature Live (psf ) Equip.

Brea~4 El ev. Pressure op Load From From l,o ad s Hos. Room (ft.) (esi) Int. to Int. int. to Ext. ~(sf ) Ploor C). ~il in s Ao -8 R )S 422 0.51 0 40 59 1 ~ 4k Pump iso -4 R 113 441 po 250 59 60 Hone

)XO 5 )4 )7 R 112 441 0.51 po 40 250 59 6& Hone R 206 471 n.n5 0 40 250 32- 34 Hone

~

LZ.O -1 R 313 510'-6" 0.48 0' 40 250 40 30 Hone

~P R 400 522 1.0 250 41 08 Hone 1W( -S )~ R 4n6 522 15.0 po 250 126 0 1.5 Pump 1~9 -Q I +8 & 407 yzr -1 R 409 535 11.0 po 250 40 00 Hone gaze-49)W))+ti<++) R+-~i

'(~-7)M 1Z9- 1$

IC-52.

a.c-c

) t5 1) gP

&

i~m

&

L~

R R

511 510 540 540 4.4 1.0 0'0 20O 20 400 400 00 65 55 51 Hone lleat

~ Ll(12.- ~ 't Exchs.

lqg -2g,q 1 16.2 & 29.5

)'ze -9 R 509 540 2. l 20 400 08 50 Hone

)39-1 R 604 572 0.03 40 250 15 36 lleat 6 Vent I )B -XP,)>)~)b)7) Unit 51K 8)9 1%)1< ) 1 t 4 >

C 'X R 308 501 0. 41 0 4no 1000 63 55 Hone m PJ Steam R 310 501 20.'0 20 1000 277 41 None V' W Vl T1)no el CO Z Z

~ASEC 0 HO rt".: 1 . Por loc.)t ion of pipe bceak nos., see

2. P.>r verrical an horlaont.)l seismi" factocs, see 3.7.

1

MKNDMENT NO. 25 June 1982 l7'- 1" 23'- 2" l7) 0C4

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f I

CQ CQ I

IT) lA4 cz g ~

WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO ~ 2 PLAN 8 EL. 572'.6-'

HIGH ENERGY FLUID P IPING SYSTEM RUPTURE LOC

AMENDMENT NO. 25 June 1982 2 2'- (u" 23'- 2" 24'- l 0".

C) 0Q cu v I CV cj rt W~C PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 uk'ASHINGTON PLAN 9 EL.

548'IGtlV,:

HIGH ENERGY FLUIO PIPING SYS. RUPTURE LOC.

3.6-47'

I AMENDMENT NO. 25 June 1982 Cl) 0 C4 0

QQ tY) rA N CU 0

LOG.l ~i~ F ht WASHINGTON PUBLIC POWER SUPPLY SYSTEM HIGH ENERGY FLUID PIPING SYSTEM NUCLEAR PROJECT NO. 2 RUPTURE LOC. PLAN 8 EL. 522'

AMENDMENT NO. 25 June 1982 a.5 l7-9" Q I gll 0

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ff)

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)!

I Q3 (Q

CV

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CV lY) cA CQ WASHINGTON PUSLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 PLAN 8 EL.

501'I(

HIGH ENERGY FLUID PIPING SYS. RUPTURE LOC.

'."'":

3. 6-,'ld

AbKNDMENT NO. 25 June 1982 H.3 24 '- l 0" I g<g q

0 Al (Q

4.

C4

~ I WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 RUPTURE LOC. PLAN 9 EL. 471

'i HIGH ENERGY FLUID PIPING SYSTEM

~

GU:.":

0', lc

II Al1ENDNENT NO. 25 June l982 H.B j7t qrp I l'-3" 22'o" 22'- Cu" 0

~

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'i cQ Cu rn Gti)')E HIGH ENERGY FLUIO PIPING SYS. RUPTURE LOC.

NUCLEAR PROJECT NO. 2 PLAN 8 EL. 441 3.6-~.

I AMENDMENT NO. 25 June l982 H3 Z K l7- q" Z2 '-6" 2 2'-Cn" Z4'O h '

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~c=~~ Wi C uaE WASHINGTON PUBLIC POWER SUPPLY SYSTEM RUPTURE LOG. FIG Ui!;".

HIGH ENERGY FLUID PIPING SYS.

NUCLEAR PROJECT NO. 2 PLAN 8 EL. 422'-3" .6.-~ ~~

H.3 J7'- t" lq '-3" 22 -6" 2 2'-Co" Z4'- iO" s ~

tf) cQ CQ

~ ~

CQ Qa~F 4

iF WASHINGTON PUBLIC POWER SUPPLY SYSTEH MOOERATE ENERGY FLUIO PIPING SYSTEH FIGUjli I.

NUCLEAR PROJECT NO. 2 RUPTURE LOC. PLAN 9 EL. 422'-3" 3.6-4'.'l

' 7'-9" 22'-Co" 22'-G" 23-'" 2 '-10" 0

0 tA 04 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGUk" "

MODERATE ENERGY FLUID PIPING SYSTEM NUCLEAR PROJECT HO. 2 RUPTURE LOC. PLAN 9 EL. 441' 3.6-0.'

I

l H.3 hl.8 l7'" 22'-Co" 2 2'-Q" 23'- 2" 0

WASHINGTON PUBLIC POWER SUPPLY SYSTEM I GUR'c. ';

MODERATE ENERGY FLUID PIPING SYSTEM I NUCLEAR PROJECT NO. 2 RUPTURE LOC. PLAN 9 EL. 471' .6-42ci I

L N,b 2 2'-Q 0Ol CV WASHINGTON PUBLIC POWER SUPPLY SYSTEM MODERATE ENERGY FLUIO PIPING SYSTEM FIGUf, )

NUCLEAR PROJECT NO. 2 RUPTURE LOC. PLAN 9 EL. 501' . 6-4p((

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t7-'I" 22 '-Ca" 22'-6" 23'- 2" 4

0 CU

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cQ EL~MF F lG'OV.F WASHINGTON PUBLIC POMER SUPPLY SYSTEM NUCLEAR PROJECT HO. 2 RUPTURE LOC. PLAN 9 EL.

522'IGURI MODERATE ENERGY FLUID PIPING SYSTEM

. 6-42..

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22'- Co" 0

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CA CO WASHIHGTOH PUBLIC POWER SUPPLY SYSTEM HUCLEAR PROJECT HO. 2 RUPTURE LOG . PLAN 8 EL.

548'IGURE MODERATE EHERGY FLUID PIPING SYSTEM 3.6-421'

l7'- 9" I I'3" 22 '-4s" 22'- &" 23'2" 24'-lO" 0

CV MASHIHGTOH PUBLIC POMER SUPPLY SYSTEH NUCLEAR PROJECT NO. 2 572'IGU:"

HODERATE EHERGY FLUID PIPiNG SYSTEM RUPTURE LOC. PLAH 9 EL.

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y Jph HB Z K L N N.s 2A '- lO" 0

04 0

tf) 04 0

Ct C4

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a ffl dl CV M~cC i E ~ >Gu~i WASHINGTON PUBLIC POWER SUPPLY SYSTEH FIGUR';

HODERATE ENERGY FLUID PIPING SYSTEM NUCLEAR PROJECT HO. 2 RUPTURE LOC. PLAN 8 EL. 606'-10 1/2" . 6-42 ~ .

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Jet Prom Qrcamferenthl Bros eath Eads Restrained (Flu asr3..

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AMENDMENTNO. 31 June 1983 270o RC26

~RCR16 RCZ4

~ RCR15 RC1 RC23 ~ RCR17 RC15 RC20

~

RCR14 RC21 RCR1 RCR18 RC21LL RCR19 RCR20 gg ~RCR13

~

RC12 RCR12 RCR21 RHR SHUTDOWN

~RETUFIN RC3' RC16 RC~

RCR10 RCR11 RHR SHUTDOWN SUCTION ~ RCR8 42II,Z 8M I

gg~~Fi0 RCR2lI'EY:

~RCI TYPICAL BREAK LQCATICN RCR1 = TYPICAL RESTRAINT DESIGNATION SUFFIX "LI ". INDICATES LONGITUDINAI BREAK

'NDICATES LOOP A ONLY NOTES: (1) THIS FIGURE REPRESENTS LOOP A. LOOP 8 IS SIMILAR EXCEPT AS NOTED.

(2) SEE FIGURE 3.6-35b FOR RESTRAINT-BREAK LOCATION CORRELATION AND BREAK TYPES.

(3) ONLY THOSE RESTRAINTS THAT MAYACT DURING THE POSTULATED BREAKS ARE SHOWN.

WASHINGTON PUBLIC POWFR FIGURE SUPPLY SYSTEM REACTOR RECIRCULATION PIPING SYSTEM 3.6-35a HVCLEAR PRQJF ";;IO. 2

~, ~

,l BRSCN No.

BURNS AND ROE S A R CHANGE NOTICE

~/~ jgz (BRSVi4)

Part I SAR sect on(s ) af fected:

Part XI Description of Need for Amendment: P, c Part XII Are there any new commitments in change: HS MO Identify:

See attached pages for proposed revisions Attach supporting documen-tation or information

- ar" -lV Approvals Approvals indicate authorization to submit the oroposed change to the client. Differing v'wpoints shou'd be resolved as much as possible be-ore sign-off. 'Resolution of conflict should be explained in remarks.

S 'ture Date Remarks Licensing Lead Othe Appro.seriate Licen-sing Eng. or Sup.

Supervisor

• !ANL P-" Approved Deviation

~~HO Washington Pubs'ic Power St.pp)y System P.O. Box 968 3000Gecrge Wasi~inqton'iVay Rich!and. iA'ashington B9352 i509) 372-5000 May 26, 2983 HPBR-RO-83-163 NS-L-02-JCA-83-060 Hr. J. A. Forrest Pl oject t"anagel Burns:nd Roe, Inc.

601 Nil liams Blvd.

Richland, MA 99352 Dea r t ir. Fo rres t:

Subject:

CHANGES DUE TO NEh LOADS PIPE BREAK ""tA'SIS (SCN 82-175 ATTACt',ED)

Please review arid concur;lith the attached SCN 82-17.; for-incorporation into the Supplv System's final amendment;nto che WNP-2 FSAR. The subject SCN also ',ncludes r.visions to the FSAR by General Flectric.

Please respond by June 8, 1983.

Very truly yours, L.T. Harrold Assistart Director, ltNP-2 Engineering JCA/mt Attachment cc: ktS Chin BPA

~

5K+ COnQg;.'I 8tt'R"BE..:4'"uR Cygelman Site AN Kugler BAR Site TA Hangelsdorf- BFCH N Poivel 1 BECH JJ Verderber MttP-2 Fi 1 es M Ouer BKR HAPO

MilP-2 SCil " -/7 SAR CHAi'(GAL i'tOTKCE SAR Section(s) Affected:

Description of Change:

Reasons for Change:

This SCH satisfies OCI Log Coavnitrrent Ho.:

Th'is SCil con+its to the following:

This SCtl 'wi11 be incorporated into Amendment No.:

See attached pages for original and/or revised SAR Section(s).

Approvals: Signature indicates authorization to file the subject change into an amendment.

Si cnatur e Date Remarks Lead Technical Reviewer(s) LTRs

>/~pe

~~

project angry Plant Operations f<anager

)tanager i~ Yzwti /i >)<'3 Proj ec.

fi~hg gn

'

iManager

~, Project QA Manager"

~>5 . ",App icable only for changes affecting Quality Assurance.

4.

'si <,

) I GENERAL ELECTRIC CO.

NUCLEAR POWER SYSTEHS OIYISION LICEHSIHG ACTIOH NOTICE WPPSS NUCLEAR PROJECT NO. 2 Notice 4

'

Transmittal Date: Gg Responds to: r V r ) '7

SUBJECT:

N, v r n FSAR: '. P . 2 (~ JZ HRC question 0:  ;;l.:, I"LSON MANAQaR ACTION RE(VIREO: p pqp is~ iJceNsihG

~+u 4ZW 4~ ~~4c'~ M 0

4 Pir~

Submitted by ~.+ ~ .... I P. B. Kingston (Licensing ngsneer)

Date c'~A" Distribution:

1. Licensing Eng. 682

2. Projects 394 Reviewed by 3. WPPSS (Original)

A: F. DeYault rogect rng>neer 4. Burns ~~ Roe (R.O.}

Approved bv ~ ( .o  : Oate F. A. HacLean ro]ect Manager)

PK:cal/K'"~89

'~/i9/82

GENERAL ELECTRIC CO.

NUCLEAR POWER SYSTEMS OIVISION LICENSING ACTION NOTICE MPPSS NUCLEAR PROJECT NO. 2 Notice 0 Rev.

1 Oate. August 12, 1982 Responds to:

SUB JECT New Loads Pi pe Break Anal ysi s FSAR. Sections 3.6, 3.12 NRC Question 8: N/A ACTION REQUIREO:

Attached are the recottrtended FSAR chanaes of Section 3.6 and 3.12 to reflect the New Loads Pipe Break Analysis.

Please Note: Ge recanmends that Burns 8 Roe review Section 3.6.2.5.3.6, items a,b,c for consistency with Section 6.2.4. It may be preferrable o replace tie write-up here wi;h a re-ference to Section 6.2.4.

Submit ed by r L. E. Santos (l.icensing Engineer) r~ Date Pi stributi on:

L i cens i ng =ng. 68.

P ajac s c Reviewed by Oate 3. 'w'PPSS (Original)

A. ;". OeYault (Ptojec Engineer) Burns 8 Roe (R.O.)

r~~ re;

~".'. r Approved by r~X~<i'.~C%/.r'.

Oa e MacLean (Project Manager) .

LS: hmc/1815 8/13/81 jLJ IrI

.,E.

ASBCCZA~ HXTH THE PQSTU~~ 3UPTUBE OP PIPXHG Information cancer:Mg postulated break and c ack location

~%aria and methods o" analysis for evaluating the dynamic effects associa~ Mth postulated breaks and cracks ia high and moderate oner@@ fluid system piping inside and outside of primary coaCaimaeat is praaeated in this section. The infoxmation presented m~

this section, and ia 3.6.3.< con-firnus that the requi smeats for the protection of structures, systems and components relied upon'or self e <<sector Shut g

dawn, or to mitigate the consequences of a postulated pipe break< have heea met.

3.6.2.l Criteria Uoed to Define Break aaP. C-ack Lacst9.on aad Coaff.guratioa Tha fo3.1euing section establishes the cMte ia or the loca-tion aad configuration of postulated breaks and cracks in high energy and moderate energy piping systems both inside aad outside of pr~~ coat-air usat.

High~orgy fluid systems aro defined as those systems, o portions of systems, that during normal plant conditions(a}

axe maintained pressurized under conditions +he e either oae or both of the foll@sing a e met:

a. Maximal temperature axceeds'200 P

b. M mi:mam pressure exceeds 275 ps'y Moderate energy fluid systems are defined ns those systems, or portions of systems, that du~mg no~

are pressuri"ed under both o>> Qe foU.exing coadi 'oas:

alan" conditions

a. MME@ temperature is 200 o P or less.

b. Mzu~ p essu e is 275 psig or less.

(a) M~m'3.ant conditions is defined as We plant opera~

conditions during reactor startup, operation at Pcwexr ~ax b b s

Piping. systems are c3.assif9.ed as moderate~ergy systems>

+hen they operate as high'nergy piping for only short perLQds in perf arming their, system function ~ Par the

'major aporatiaaal period they @xa1ify as moderate-energy fluid systems. Aa operational period is considered "short if tho total fraction af time that the system ope ates, within the pressure-temperature conditions specified far high-energy fluid system< is 3.ess than app~imate3.y percent af gee time period that the system oper@tea as a

~

moderate energy fluuid system, or less than aae percent of the narma3. opernting life span of the plant.

A postulated pipe break is defined as a sudden, g ass failure of the pressu"e bounds~ either in the faxa of a ccmp3.ete circumfQronti@X sever?Lace (gui3.late b ) ar ss deve].op~

ment of a sudden longit c ack (I.angitu-dinaL split) . These ere past ated or gh energy fluid

.,systems aaly..Par moderate energy fluid'systems,,oi e LEAkAGE ruptu e is confined to postulation a cracks in piping and branch nuns. These cracks af ect the surround-ing enviranmenta3. conditioas anly, and do not cause )et im-pingement or uncaaM~1Z.ed'hipping of the pipo.

A moderate energy piping system c ack is not postu3.ated simultane'ously vt,th a high energy pipiag system break, nor is any pipe break or c ack outside caatainmeat postulated concur eat'y with a pipe break ar c=ack inside conte&ment.

Postulated pipe break 3.ocations a=e selected as described herein; and axe based an the guidelines provided in Regu-latory Guide 1.46, Rav. 0; the U.S. Muclear Regulatory Commission (KC) Branch Technica3. Position APCSB 3-1, Appendi= B; and cs exyanded in MRC Branch Technical Position NEB 3-3. for. piping inside and outside pr~~ contain"vent.

3.6.2.1.1 Postulated Pipe Break Locations ~ H'qh ~ergy Pluid System Piping 'fot in the Cantsi~t Peaet=atioa A-ea .

Pipe breeks (nat including leakage c acks) aiba postulated at locations as Mdicated bo3.av:

3 ~ 6-24

AMENDMENT HO ~ 9 April 1980 3 ' .2.1 Postulated Pipe Break Locations in ASM Section

~ 1 ~ 1 III Class t Pipinq Runs a0 The terminal ends(a) of the pressuri ed portions of the run.

Intermediate locations of postulated pipe breaks are selected by application of one of the, follow-inq sets of rules:

( 1) Pipe break is postulated at each location of sicmificant change in flexibility, such as pipe fittings (elbows, tees and re-ducers), and circumferential connections to valves and fiances.

(2) Based on stress and fatigue analysis, as calculated according to ASNE Code Section III Sub-article HB-3600, no break is pos-tulated if any of the followinq applies:

(a) Sn(b) does not exceed 2.4Sm(c) o (b) Sn exceeds 2.4Sm but does not exceed 3Sm, and the Cumulative Usage Pactor (U)(d) does not exceed 0. 1 Terminal ends are extremities of piping .uns that can-nect to structures, equipment> or pipe anchors that act as rigid constraints to free thermal expansion of pipincr. A branch connection to a main piping run is a terminal end for a branch run, excep't when the nominal size of the branch is at least one half that of the main piping run< and the branch and main runs are modeled as a common pipincr system during the piping stress analysis.

. Sn is the primary, plus secondary stress intensity range, as calculated by use of Zquation (10) of ASHE Code

~

Section IXX Subsection HB, Par'agraph iM 3653.-1 between any two load sets (includincr the zero load set) for normal and upset plant conditions, including an OBE event tr'ansi en t.

Sm is the design stress intensity< as described'n ASME Code Section IXX Subsection HB Paracraph-NB 3229.

V is the Cumulative Usage Pactor that indicates the tota3.

Caticue damage as calculated'by the procedure in ASHE Code Section IXX Subsection BB, Paraqraph NB 3653.

3. 6-25

.exceeds but (e) and <f) are (c), S 3Sm Se Sr eRch less than 2.4S , and U does not exceed O.l Ce Shen twa or mars intermediate locations cannot be detsnained by stress ar usage factor limits as described abave, then intermediate locations of significant change in flexibility are chosen as postulated pipe rupture locations an a easonable basis for each piping run(a) or branch run(b) as necessary to provide pratsc-tian. A easonahle basis as used herein can-sidsrs the locations of highest camputed value of stress/ Sn Cumulative usage factor is also considered. As a minimum, ~so intermediate locations are chasen for each piping run .or branch run, except, fax a piping run having only ane change in direction in which case only one termediats break is postulated. Xntermediate breaks are not postulated in sec"ions of straight pipe, where there are no pipe fittings, valves, or flanges.

(e) is the naminal value of expansion stress as calcuLated Se by use of Ecpxation (12) of ASIDE Code Section XXX Sub-section LLB g Paragraph HB 3653 ~ 6 (a)

(f) Sr is the range of priory plus secandary membrane plus bending st-ess intensity, exlcuding thermal bending and thermal expansion stresses as calculated by use of Ecgxatian (13) af. ASHE Code Section XXX Subsection %3.

(a) A piping run is defined as piping which intercannects equipment such as pressure vessels, pumps, and aMer ecgxipment that act as rigid canstraints ta free thermal expansion of piping.

(b) A branch run is defined as differing f=om a pipe run only M that it originates at a piping intersect'n as a which are included with main ~

branch of the main pipe run, except that branch lines piping in the stress analysis computer mathematical model and are shown ta have significant effect on the main run behavior are considered pa~ of the main run.

3. 6-26 8a i7s-

~ ~ ~

w".4P- 2 Ailh:o D4f E;JT lO. 9 April 1980 Piping and electrical penetration details are discussed and shown in 3.8. 6.

The stress criteria for postulating breaks n containment penet ation pioing between isolat'on valves is given in

3. 6. 2. 1. 2. 1 and 3. 6. 2. 1. Z. 2.

Nelded attachments, for oipe suoports or other purposes, to these portions of piping are avoided except where detailed s "ress analyses, or tests, are oer formed to demonstrate comoliance with the Limits of 3. 6. 2. l. 2. In addit'on, the number of circum'erential and longitudinal piping welds and branch connections a'e minimized.

Any pipe anchors or. restraints (e.g., connections to con-tainment penetrations and pipe whip restraints) are designed such that they are not welded directly to the outer surface of the piping except where such welds are 100 oercent, vol-umetrically examinable while in service, and a detailed stress analysis is oerformed to demonstrate compliance with he limits of 3. 6. 2. l. 2.

Tunne'tructures surrounding th orimary containment pene-tration.-piping are des'gned for the thermal and oressu e loads of a through-wall Leakage crack regardless of c" ck postulat'on reauirements. Refer>>o 3.6.1.20 for further discussion Access for inservice inspection of welds in high energv {hot type) containment penetration assemblies is desc" ibed in 3 8 6. -'- l.

~ ~ Al.'ecuired inserv ice inspection locations are accessible.

3. 6.2. 1.3 Postulated~ eakage Crack i.ocations in H'gh and tulated~~ moderate Fnergy FLuid Systems Tn high energy piping systems consisting of ASHE Code Section I I C l ass l. p i p ing, ( inc iud ing flu ' system piping between primary containment isolation valves) cracks are not pos-,

~ I l

4~~

In of

.'.....

ASi~)E Code moderate energy piping systems cons'sting Section III Class " and 3 oioing and moderate energy non-nuc'ear piping, includi..g fluid system piping between prmary containment isolation valves, cracks are not 3 ~

6-"9

76%9 2 AMENDMENT NO 9 April 1980 pastulated provided the stress range of 0 4 (1.2Sh(aI + SA(b~)

is nat exceeded fo>> the load combination which includes the effects af pressure, weight, ather sustained loads and occasional laads such 'as the operating basis earthquake, and thermal expansion loads Since all piping in structures housing safety-related systems are supported and cont ol,led as Seismic Category I systems regardless of service, the criteri.a for postulated cracks is the same as above for all systems.

3. 6-2. 1-4 Types of Breaks and Cracks Postulated in High Energy and Moderate Energy Pluid System Piping

3. 6.2.1.4.1 Breaks in High Energy Pl.uid System Piping The following types of breaks are postulated in hi.gh energy fluid system piping:

a. No breaks need be postulated in pi.ping having a nominal diameter less than, or equa3. to one inch.

b. Circumferential breaks are postulated only in piping exceeding a one inch nominal pipe

,

diameter.

c. Gongitudinal splits are postulated on3.y in piping having a nominal pipe diameter equal to or greater than 4 inches.

d. Gongitudinal splits are not postulated at terminal ends

e. At each of" the postulated break Locations, consideration is given to the occurrence of either a longitudinal split or circumferential break. Both types of breaks are considered, iZ the maximum stress ranges in the circumfe ential and axial directions are not significant3.y dif erent Only one type break is considered as fo13.ass:

Sh is the allo~able st=ess at maximum (hot) temperatures defined in ASME Code Section IXX, Article NC 3613..2 SA is the al3.a~able stress range for thermal expansion, as defined in ASME Code Section XXX, Article NC 361'.2.

3.6 30

AHENDNENT NO 9 April 1980 (2) 3:f this type of analysis indicates that the maximum stress range< in the circumferentia3.

direction< is at Least l.5 t'mes that in the axial direction, only a 3ongitudinal 'split is postulated.

Mhere break locations are selected without the benefit of stress calculations, circumferential breaks are postulated at the piping welds to each fitting, valve or welded attachment. Postulated longitudinal splits are described in FSAR 3.6 2.1.4.1.i.

go Por a Longitudinal split, the break area is assumed to be equal to cross-sectional flow area of the pipe.

h. Por circumferent'al b'reaks, pipe"whipping is assumed to occur in the plane defined by the piping configuration, and is assumed to cause pipe movement in the .direction of the jet reaction.

. A 3.ongitudinal break is assumed to result in an axial split without severence and to be oriented ht any point about the circumference of the pipe< or alternately, at the point;(s) of highest stress as indicated by a detailed st ess analy-sis Xf a postulated break location is at a non-axisymmetric fitting, such as a tee or elbow/

the split is assumed to be oriented (but not concurrently) on each side of the fitting at its center, perpendicular to the plane of the fitting and is assumed to cause pipe movement in the direction of the get reaction.

Por a circumferential break, We dynamic force of the ]et discharge at the break Location is based upon the effective c'oss-sectional flow area of the pipe and on a calculated fluid pressure> as modified bv an analytically or experimentally determined thrust coef f 'ient.

A c'rcumferential break is assumed to result in pipe severence with fuLL mparation, except as limited by structuraL design features. The break is assumed to be or'ented perpendicular to the 4MOUIVl i< &. 70 AT' E45i A PIPe ajA+Z i Eg gAr~

>is pz Ae EhEnr7 3 6-3 l <>>7 uzEz Pawl ssc77ows

3.ongitudina3. ass of the pipe. Line res -ic-ticns, flow limiters, and the absence af energy xesexvoixs are accaunted fox', ~ the calculation of the design )et discharge.

3.6.2.3..4.2 Cracks in Kigh Energy and Moderate Energy Pluid System Piping The following controlled, thxough-wall leakage cracks, are postulated in high energy and madex'ate energy fluid systems (or portian of systems):

a. Cxacks axe postulated in fluid systems or por-tions of systems whose sixe exceeds a nominal pipe diameter of one inch.

b.'luid flow, fram the postulated crack, is based on a circular opmxing of axea equa3. to that of a rectangle one-ha3.f pipeMiameter in length and one-half pipe wall thickness in width.

C ~ The flow from the pastu3.ated c=ack is assumed

'ithI* * \ *~

to resu3.t in an environment that wets all unprotected components wi~ the competent, subsequent. flaading in the c~~~ent

. I are detexmined an the basis of a conservatively estimated time period required to affect cor-rective action.

3.6.2.3. 5 Protection Criteria "or the Ef acts of Pipe Break

~tact'on fram the effects cf a whipping pipe due to a pipe break is provided whexe necessa~.

need nct be provided if P atectian f~ pipe whip any.ane of the follcwing conditions 6K'.sts

a. The piping is classi ied as mcdex'ate ene gy p3.ping ~

b. Pallowing a single postulated pipe break, piping for which the unrestrained mrvement af either end af the ruptured pipe, in the d~ecticn of the jet reaction abaut a plast'c hinge, formed within the piping, cannot impac any stoic uxe, system ox'cmpcnent important ta safety.

3.6-32

B' (1) The

-

transient forcing functions>me,u4at points along the pipe~aaae4W fram the propagation af waves'wave ~xst) along the pipe, and, A~ 7+~ 8R~Al4~

f~ Me reaction force due ta Me momentum of Me f1uid leaving Me encL of Me pipe (hlawdawn ~est) .

(2) The waves cause various sec"ians of the pipe to be loaded wiM timeMegendent forces. Tt is assumed Mat the pipe is ane-diI:~nsional in that Mere is no attenuation or ref1ectian of Me pressure waves at bends, eMows, and the Like. Pollawing Me rupture, a decam-pression wave is assumed ta travel fram the break at a speed equal to Me local speed. af sound within Me fluid. Nave reflections M~M accur at the break end, and the pressure vessel Erma until a steady flow condition is es ablished.

baunda~ coFditions. The blawdawn t?xrust causes a reaction force perpend'cula ta the plane of Me pipe break~ gzAcpre6 A zpvac, SYRIA>Y ><A7<

V'AJ uE.

(3) The initial blawdawn farce an Me pipe 's taken as the sum af the wave and blawdawn thrusts and is equal to the vessel aressure (P<) times the break ar'ea (A) . After the in7tial decampressian period (i.e., the time it takes far a wave to reach the first change in dire'ctian), the arce is assumed ta drop off to the value af Me blowdown Mrust (i.e., O.'7 P~a).

(4) Time histaries of transient pressure, flaw rate, and other the~dynamic properties of the fluid can be used to calculate the blow-down force an the pipe using Me following equation:

P m (P~P a) whe e:

P ~ Blawdawn Parce P ~ Pressure at exit plane 3 '-34

Pa ~ Ambient pressure u ~ Velocity at ex" t plant Density at exit pLane A ~. Axea oZ break g ~ Qxavi.tational constant (S) Pollcving the transient pe"'iod, a steady-state period is assumed to ex"st. Steady-

<<~7- state blavdcwn forces are calculated, can-sidexing f ict'ona3. ef acts'. For these effects reduce the blovdcvn forces fxcm the theoretical ma~urn of 1-'26 P+- The method oC accounting for these ef acts is .presented in Reference 3.6 3. Por submooled vater, a reduction fxcm the theoretical maximum of 2.0 P A is found thxough the use of Bernoulli's and other standa d equations, such as Darcy's equation, which account for friction.

b. The foLLaving is an alternat method for calcu-La~g hl~dcwn forcing functions.

The computer coda RZLAP3 (Reference 3.6-9) is used to obtain exit plane thermodynamic states for postulated ruptures (see 3.12.ll for urthe discussion of HZLAP3} . SpeciQ.cally, RKV3 calculates exit pressure, specific volume and mass rate. Pram these data the bL~down reac-tion load is calculated using the foll~ing relation:

T~P~P+QV~

~c R>>- T xA where:

- th mt per unit brea3c are 3.6-35

P - receiver pressure 6> - exit mass flux, v~ - exit speci,fi.c volume

- grav'tational constant R - Reaction force on the pipe 3.6.2.2.2 AnalyticaL Methods to Define Response models 3.6.2.2.2.1 Gene al Desc-iption of Analytical Met3xcds The prediction of time-dependent and steady-thrust reaction loads caused by blcvdevn of sub>>cooled, saturated, and tvo-phase luid from a ruptured pipe, is used in the design of piping systems and in the evaluation of dynamic effects of pipe breaks.' detailed d9.scussion of the analytical methods employed to compute these blmdaom loads are given in 3.6.2.2.l. The analytical methods used to account for this loading are discmsed beL~.

3.6 2.2 2 2 Dynamic Analysis of the Mfects of Pipe Rupture a>> Cr iteria (1) Analysis is performed for each postulated pipe break.

(2) The analysis includes the dynamic response of a13. components of the sys em includinq

~ pipe> pipe +hip rest=aints and al3.

structures requized to t=ansmit Loading to foundation>> The st LTctures are analyzed for a suddenly applied force in conjunction Wth impact and rebound ef ects due to gapa between piping and pipe whip rest=aints.

3 6 36

NNP-2 AMENOMENT NO . 2S June l982 (3) The analytical model adequately represents the mass/inertia and stiffness prope ties of the system.

(4) Pipe whipping is assumed to occur in the plane defined by the piping geometry and configuration, and to cause pipe movement in the di=ection of the jet reaction.

(5) Piping contained within the broken loop, is no longer considered oart of the reactor coolant pressuce boundary (RCP8). Plastic deformation in the pipe is considered as a potent'al energy absorber. limits of strain

' i"* '-

-'r>>'ipi,ng systems are des'.gned so that 4~~sC.>.4:9."-

plastic instability not occur in the oipe at the design dv amic and static loads, unless damage studie are oerformed which show, that the conse ences not result in the direct damage o any es ential svstem or component. Mcr d c~/2 (6) Components, such as vessel safe ends anc valves, which are attached to the broken piping system and do not serve a safety function oc whose failuce would not fucthec excalate the consequences of the accident, ace not designed to meet ASME Code require-.

ments foc essential components under faulted loading. However, if these components ace requi;ced foc saf shutdown, or if hey serve .

a safety func"ion to protect the structural

.ntegcity of an essential component, then "hese components are designed to Code limits for faulted conditions and to ensu=e v ooeraoili tv, P . /

3.6-37

b. Analytical Models (l) t.umped-Parameter Analysis Madel: Lumped mass points axe inte connected by springs ta ta3ce inta account for the effec s af inertia and stiffness inherent in the system, and time histories of the responses axe camputed by numerical'ntegration to accaunt, faz:

gaps and- inelastic effects. This analyticaL method is discussed in detail in Reference 3.6 4 (2) N'nergy-Balance Analysis Madel: Kinetic energy, generated during the ff st cpxax er cycle movement of the ruptured pipe as im-parted to the piping/restraint system through impact, is canverted moto equivalent st ain enexgy. Defoxmatians of the pipe and the rest aint are compatible with the level af absorbed energy.

(3) Pipe whip .xest=aints,. for the reactar xeci"-

culatian system, are. designed by the HSSS supplier. The analytical method utilized fox this design is the camauter program PDA which is described in Refexenc .5-4 and further discussed in 3.1Z.33. Pipe whip rest=aints for all other piping systems, x~ixing such pxatec the architect/engine

',

f cribed, in c., (below) z. utilized "or this a e P.esianed by he met. des-pipe whip restraint design.

c. Simplified Dynamic Analysis

{1) Zn axdex,to simplify dynamic analysis the fallowing consexvative assumptians are utili"ed:

(a) The entire stature including pipe, restraint UJxkagec support beams azzd ma)ax stature to foundatian connections ahsaxh energy by elastic, elasta-plast'c, ar plastic defaxmatian. Xn cx"e to pxa<<

vide a simplified dynamic mathematical madel, ane member is generally con-sidered, to absorb all the energy.

This member is classified as an enexgy 3,6 38

Reference 3.6-6 provides the ductility ratio that correspands ta collapse (u ).

Par sutural steel, members, Chase values vary, v9.th upper limits in the order af 20 Ca 30 and up (for very ductile structures). Por MP-2, the '

nuudznuza permiss&~le ductility ratio 1imited ta 50% of (p ), except that energy absorbing memberS in Meet con-tact W th priory containment are limited to 5% of (g ) . Por WN2-2, only steel mem-bers are ut21iced as energy absorbing members< as defined in 3.6 2.3 3.2.d.

~ ~

Tha maximum values of (p ), for various structural camponents, a9e given in Table 3.6-1.

(i) The eqaation de 've'd in Pigure 3;6-2 accounts for a suddenly appl'ed, con-stantly maintained farce, in can-

)unction vith a kiaxetic energy of im-pact on the resisting member. Total transfer of energy is implied. This is cambined with the constantly main-ed force (fram ruptured piping blovdawn) on the estraint structure ..

This assumption is consistent vith a "era caef iciant of restitution (full plasticity), and is. a conservative assumption.

W,th raga~ ta rebound, noted that. if it should be a coefficient of re-stitution of unity is assumed (full rebound), Chere is "e o kinet'c energy t=ansfer to the rest=aint stature.

Xf a coefficient of restitution less than unity is assumed (partial re-bound), there is a partial amount Q f kL?letic energy trans fBr to the 78 st=aint st~ ure.

A coefficient of restitution of "ero, conservatively assumed in the appli-cacian of the ecpxation mencioned above,

3. 6-41

liNP-2 ANEMONE.'1T i'10. 9 April 1980 gives zero rebound with 100% kinetic energy transfer to the rest a'nt structure.

T.t should also be noted, that the assump-tion of a suddenly applied, force, as used in the equation constantly'aintained mentioned above is conservative with respect to rebound. Rebound implies a finite time of short duration contact with the restraint structure, in contrast to the infinite time assumed..

(3) Actual structural resistance, for the above structures, is determined by methods of limit analysis using a dynamic yield strength, as defined in 3.6.2.2.3.1.

3..6.2.2.3 Naterial Properties Under Dynamic Loads 3.6.2.2.3. 1 Dynamic Yield Strength To a=count for the rapid strain rate effects, dynamic vield strength is util'zed. Tnis phenomenon is documented in References 3.6-6 and 3.6-7. Naterial tests hav shown a con-sistent increase in yield strength under rapid loading. Under rapid strain rate, carbon steel yield strength consistently improves by more than 40%. High strength alloy steel displays a somewhat smaller improvement. Por WIP-2, a conservative dynamic yield strength of 1108 of minimum static yie'd strength, at the specified operating temperature, is utilized.

3.6.2.2.3.2 Naximum Strain of Tension Nembers ere tensi members, s 's U-Ba shown on Fig. 3.6 4 which maxi..

absorp '.

co. "itute px of 50%

whip limi stops, ar permitte to deCo the min'm uniform "rain, du. 'ag ner>

. a 3i0 20203

~ ~ 3 Nav indium DeCormation of Flexura'embers Deformat'ns of enercy absorbing flexural support members are generally limited to =0% o that deformation which corresponds to structural co lapse, except that deformation oC nergy absorbing members is cirect contact ~ith the primary contain-ment vessel 's l'mited to 5$ of that deCormation which corresponds to structural collapse.

3. 6-42

WHP" 2 Insert . 3. 6-42

3. 6.2.2. 3. 2 Maximum Strain of Tension Members Pure tension members, such as U-Bars shown on Figure 3.6-4 which act to limit pipe whip are permitted to deform during energy absorption, (a) a maximum of 50~ of the minimum uniform strain (at the maximum stress on an engineering stress-strain curve) based o8>rCktraint material tests, or (b) one"half of minimum percent elongation as specified in the applicable ASHK Code Section IIfor ASTH Specifications, if demonstrated to be ~~

The dynamic tensile and impact properties are specified to be not less than: (a) 70 of'he static percent elonga-tion, or (b) 80~ of the statically determined minimum total energy absorption.

LS:hjr/C07298 B/3/S2

c. Jet impingement 'oading on primary conta'nment penet ations is d'scussed in 3.8.6.

'.6.2.3.3 Pipe Nhip Restraints 3.6.2. 3.3. 'efinition of Function Pipe whip restraints, as difierentiated from piping supports, are designed to function and carxy load for an extremely low pxobability gross failure in a piping system ca"rying high energy fluid. The piping integrity does not depend on the pipe whip restraints for any loading combination. Xf the piping integrity 's compromised by a pipe break, the pipe whip restra'nt acts to limit the movement of the broken pipe to an acceptab' distance. The pipe whip restraints (i.e., those devices which serve only'o control the move-ment of a ruptured pipe following gross a'ure) will be

,subj.ected .to a once in a lifetime loading.

tthe ru.ne b stat' ea event is considered to be a aulted con ition< .

- "., i's rest aints, and r structuree to T

Plastic. deformation ox the pipe is cons'd red as a potential energ~ absorbe . Pioing systems are des'gned so tnat astxc instability not occur in "the p'e under de-sign dynamic and

'nstability '

result loads, in if the consequences of such the loss ox the prima y cont inment

'n" egrity loss of required plant shutdown capab'ity.

3.6.2.3.3.2 Pipe Nhip Restra'nt Features

a. The restraints are c'ose to the pipe to mini-mize the kinet'c energy of impact and yet are sufficiently removed from the pipe to permit unrest icted ther-.,l, pipe movement.

b. To facilitate in-se vice 'ns "ect'on of piping, the restraints are gene ally located a suit-able distance away from all c'umie ent'al welds and a e of bolted construction so as to be removab'e.

c. Pipe whip restraint st uctur~s all into cne of the zollowing two categories:

(1) "-nergy absorbing members these are modelled as clast'c, elasto-plas"ic or plastic springs 'n a dynamic analys's.

3.5-51

INSERT FSAR p. 3.6-53.

Section 3.6.2.3.3.1 The design and analysis of these components for this event are described later in this Section, and in Section 3.6.2.2. Piping is no longer considered to be a part of the RCPB following the break.

The required resistance (strength) of these structures is derived by apalication of the principles of structural dynamics.

(2) Load tzansmit~~g members - These aze relatively stiff components and are modelled as rigM members in We dynamic analysis Their function is to t azmmit loading from the source to foundation. The load due to the postulated pipe rupture is in the form of an ecpxivalent static load and is derived as a result of the dynamic analysis performed for the ener'gy absorb-ing members.

d. ~mergy absorbing members are ductile structures such as simple beams, f ames and ring, girders, (including the piping system itself}, havinq the capability to deflect significantly in absorbing the energy impar ed to them by a pos ulated broken pipe. Por loading conditions, inc3.uding-the effects of postulated p'pe rup-ture< these members are designed within the limits foz inelastic systems as stated in Table P1322'.2-1 of ASM Boiler and Pressure Vessel Code'ection lXZ Appendix P "Rules foz Evaluation of Paulted Conditions", adjusted to~account o rapid strain rate effects, as discussed in 3.6.2.2.3. These members are constwcted to meet the r~rements of Quality Class X st~ctuzes.

U-Bar straps, as shown in Pigure 3.6-4 and de- ~E<<R<V scribed in 3.6.2.2.3.2, ~ah4act asnon-1~ear,~ 4~<S~++a non-rebounding~plastic springs. The U-Bar straps are just-fied by empirical data, ~S DESCZl8EB zW 3.$ .<.<.A.M.MQ) And> ~ yg. ~9P .

e. Load trinsmit~g members are riqid components such as clevises, brackets or pins, r'gid pipe whip restraint weldments as shown ia Piguzes 9.4-P AAQ 3.6-5a through 3.6-5e, or similar components; as well as major st~ctures such as the drywell diaphragm floor, 'primary containment vessel, reactor pedestal, reactor building and foundation.

Por loadinq conditions, including the ef eats of postulated pipe ruptu=e, th~members are designed within the limits smted in Table P1322.2-1 of AS'ade Sec 'on -XXX Appendix "Rules for ZvaluaMg Paulted Condit'on" for 3 6-52

components and component supports; except that the members beyond thase included in the dynamic analytical madel (i.e. -xeactar pedestal, reactor build'ng, as well, as certain steel members assumed to be infinitely rigid) are designed ta AXSC, ACX and other appropriate structuxal component cx'itex'ia. All these members are constructed to the recuirements of Quality Cla 's I structures.

~~ >~~sRv. PAHZeqaPW

f. The recirculation pump discharge and suction piping utilizes the U-Bar strap pipe whip~ace

/AT QAlNjs CF >soge8.4-43

~ ~a~ while all othex'ystems listed in Table 3.6-2 utili.ze rigid types as 'shown in Figures 3.6-5a through 3.6-5e or similar configurations.

gi Typical installations of pipe whip rest aints are shown in Figures 3.6-6 through 3.6-10.

3.6.2.3.3.3 Pipe Whip Restraint Loading

a. Por the purpose of predicting the pipe rupture forces associated with the reactor blowdown, the local line pressures are assumed to be those noxmally associated with the reactor operating at 105 percent of rated power and with a vessel dome pressure of 1025 psig.

b. Xn calculating pipe reaction, full credit is taken for any line restriction and line "ric-tion between the break and the oressure reser-voir. The following represent typical restric-tions to flow which are specifically consider~ 2:

(1) Jet pump nozzles (2) Core spray nozzles (ins'de 'nte nals shroa d)

(3) Peedwater spaxger (4) Steamline flow limiter The hydraulic bases and calculat'anal techniq~.es for predicting unbalanced forces on a pipe as o-ciated with a postulated instantaneous pipe r zp-ture are as discussed in 3.6.2.2.1.

~:

3.6-53

WHP2 Insert Pa e 3.6-53 The design limits for connecting members such as cievises, brackets, and pins per Figure 3.6"4 are based on the following stress limits:

(1) Primary stresses (in accordance with definitions in ASME Section III) are limited to the higher of:

70K of Su, where Su = minimum ultimate strength by tests or ASTM specification; (b) + 1/3 (Su - Sy), where Sy = minimum yield strength by test ASTM specification; or (2) Recommended stress limits in accordance with ASME Code Section III, Subsection HF for faulted conditions, if applicable. The design limits for welds of connecting members to steel structures are based on the following stress limits: the maximum primary weld stress intensity (two times shear stress) is limited to three times AWS or AISC building allowable weld shear stress.

Sy LES: sem/807293 8/3/82

0 I ~

%P 2 e

c. The dyn-mac loading on the pipe 'whip restraint cammances at the effective time af impact af the pipe with the, rest aint. Zt includes the follow-ing i (1) Unbalanced farce on the pipe associated with a postulated instantaneous pipe rup-ture in the farm of a suddenly applied force.

(2) Dynamic inertia load of the maving sectian of pipe which is accelerated by the un-balanced force associated with the pipe rup-ture and collides with the restraint. This load is in the form of kinetic ene gy of impact.

3.6.2.3.4 Pipe Nhip Effects oa Safety Related .Components Pipe whip (displacement) effects on safety related st~ctures, systems and components can be placed in two categories:

(a) pipe displacement effects on components (no@"les, valves, tees, etc.) wh'ch are in the same piping run in which the break occur=ed'nd (b) cont=oiled pipe whip displacements as they apply to external components such as building ture, other piping systems, cable trays and canduits.

sta-3.6.2.3.4.1 Pipe Displacement Effects on Components in Same Piping Ran ae The criteria which ia used far detenxining the effects of pipe displacements on in-line compo-nents are as follows:

(1) Components such as vessel safe ends, and valves which are attached ta the broken piping system and do nat serve a safety function or whose fa'lure wau'd nat further escalate the consequences of the accident, need nat be designed to meet %%K Code Section XII imposed requirements for essential camponents under faulted loading.

(2). I these components a-e required far safe shutdown, or serve a safety function ta protect the st~tura3. int~ity of an essential ccmpanent, the Cade requiremen's for faulted conditions and ensure operability, if l~ts ta equired, are met.

3 '-54

iPlP-2 AiMENDi4ENT NO. 25 June 1982

a. Assurance nf primary containment leak tightness.

0 ~ Assurance tha" ootential for damage is such that tne maximum pipe break areas and/or combinations of pipe break areas do not exceed the values described in 3. 6. 2. 5. 3. 2 so that emergency core cooling system capability is not impaired.

c. Assurance that the cont<<ol rod drive system maintains sufficient function to assure reactor shutdown.

Assurance that there is sufficient capabi1ity to maintain the reactor in a safe shutdown conditions The criteria used to define pipe rupture locations for piping systems discussed in 3. 6. 2. 5. 4 follows 3. 6. 2.1. l. lb(!.) exceot for the following which follow 3.6.2.1.L.Lb(2):

u~J z.C.-".t-ii ~ i~ mi aA .Y~~

a. One elbow only, in each of tne two redundant reactor feedwater svstems inside primary con-tainment, in 3.6.2.5.4.2 and in P;gures 3.6-.16 and 3.6-17a.

b.;he entire standby liquid cont<<ol (SEC) system in 3.6.2.5.4.4 and in Figure 3.6-19a.

c. The entire RPV drain system in 3.6.2.5.4. 13 and in: igure 3. 6-32a.

Figures 3.6-12a through 3.6-35 show the oiping configurations for each high energy system ins'de primary containment and

-

include numerical i".entif ication of all signif icant points of

~ nterest in t'e piping system, Locations of oipe whip sup-po:"s and pos"ulated oipe break locations. The pipe whip supports are identified by the acronym PNS followed by an identification numbe. on 2'igures 3.6-}.2a through 3.6-34~nd as noted Qn ."-igure 3.6-35.

4.

3.6,2.5. 3 Sys"em Requirements Subsequent to Postulated Pipe Ruptur e 3,6.2.5.3-1 Control Rod insertion Capability o maintain the abili" I to insert the controL <<ods in the event of a pipe break, no more "han one in any array of nine controL rod "rive (CRD) withdrawal. lines may be completely

d. The entire eactor recirculation3.6-35a cool'ng system

'n 3.6.2.5.4.l4 and in Pigures and 3.6-35b.

3.6-57 cg ~ /7g

3.6.2.5.3.2 Core Cooling Requirements The designed ECCS capability can be mainta'ned provided that dynamic effects consequences do not exceed the following break area, break combination, and maintenance of minimum core cooling recuirements.

3.6.2.5.3.3 Maximum Allowable Break Areas For breaks involving reci culation piping, the total effective area of all broken pipes, in-cluding the effective area of the recirculation line break, does not exceed the total effective area of the design basis double-ended recircu-lation line break. By limiting the t'otal area of all broken pipes involving recirculation loops, to an area less than, or equal to that of the design basis accident (DBA) (circumferential break of reci c'ulation loop), no accident can be more severe than the,DBA.

b.

GG f JH8'7 3.6.2.5.3.4 Break Combinations Ia addition to the pipe break area restrictions, breaks involving one recirculation loop do not result in loss of function or damage to the other recirculation loop, or loss of coolant from the other loop in excess of that which can result rom a break of the attached cleanup connection on the suet'on side of the loop.

3.6.2.5.3.5 Required Cooling Sys-ems C

3 ~ 6-58

. INSERT. FSAR p. 3.6-58 Sect. 3.6.2.5.3.3 (b) For breaks not involving recirculation piping, the total effective area of all broken pipes for a given system shall not exceed the total effective area of the double-ended break of the maximum area pipe connected to the

, reactor boubdary for that system.

Sect. 3. 6. 2. 5.3. 5 To ensure compliance with Appendix A of 10 CFR Part 50. General Oesign Criteria for Nuclear Power Plants, the cooling system requirements after an additional single active safety system basis to determine compliance with core cooling

'n failure are defined in Table 6.3-7. Cases which o n'o't meet the requirements in Table 6.3-7 must be assessed individual quirements.

t AMENDMENT NO. 14 April 1981

a. Por breaks not involving recirculation pip'i g, at ast two KPCX pumps or one core spray sy em is av 'able for core cooling.

b. Por b aks involving recirculation aping, at least o core spray line and 2 CX pumps, or 2 core spar lines, are availabl for core cooling.

c. Por a LOCA wx

-

a total ef ctive break area less than 0.7 ft2, e'ther the CS or ADS 'is available for reactor depr suri ion.

d. Por liquid breaks, as cleanup suction or the combination of li id ~. steam breaks whose total break are is 'ess an 0.7 ft2 in which the ADS syste is required depressurization, at least 6 valves are avax ble.

e. For brea less than the equivalen flow area of one op ADS valve, at least 6 ADS v ves are avail le. However, the required numbe .of ADS val~ s is one less for each additional st m b ak area equivalent to the area of one op S valve.

3 . 6. 2. 5. 3. 6 Con tainment Sys tern Zn tegr i ty The following wer5 considered in addressing the LOCA dynamic effects with respect to containment system 'ntegrity:

a. Leak tightness of the containment f ission product barrier is assured throughout any LOCA.

b. For those lines which penetrate the containment and are closed during normal operation, the inboard isolation valves are as close as prac-ticable to the reactor pressure vessel. This arrangement reduces the length of pipe subject to a pipe break.

c ?ipe 'whip supports are provided n the vicinity of normally open isolation valves inside and out-side primary containment for high energy systems, il to assure that oper ab i ty of these valves remains unimpaired during a postulated oipe rup-ture event.

3. 6-59

",

AMENDMENT HQ 9

. Am~il 1980 support is also utilized as a rigid three-+ay support e 3.6.2.5.4.14 Reactor Recirculation Cooling System russo mops ~'8" a.'ystem 'A"Arrangement The>recirculation oF rHE ~ >PRE+

piping ~consist/ of the curn discharge and suction piping ~mmmm . e recir culation pump A and B" discharge lines are AZ/AAc"~ tiV A z>IAHz~- - in the C/cAu.g o/Po~~ northern and southern segments, or primary con-~

&4nlnlgg axnment. Th ines exit the reactor pressure

/ "A" -vessel in five> equally spaced, 12-inch diameter lines commencing at a=imuth 30 and endin at R It O vatxon 36'o to 330 ) . These five lines Mop vertically alongside the sac ificial shield ~all< from ele-a 16-inch diameter heade at centerline e evation of 528' 24-inch diameter line then drops vertically~o~mg7/>

A single the center of the header to eLevation it is routed into the discharge nozzle of the recirculation pump .

506'here

t I ~

MNP-2 AMENDMENT NO. 9 April l980

%1 8// ~~+ tsA II n i i y oriented the g'nd 38Q'ximuMs, with respect to R~ p<<T;>v pg.

the reactor pressure vessel. Each~~ consist of a single 24-inch diameter .line which exits the reactor pressure vessel at elevation 535'-3/4" and drops vertically alongside the sacrificial

'hield mll to elevation 502'-6 T/8'here it is tatted to the euatiaa uatalg af the teait culation pump .

b Pipe Whip Protection For the recirculation pump s ction and discharge systems, the location of pos=ulated pipe ""aaks and pipe whip restraints are shown on Pigure 3.@-9~

L

~~38 w ich is representative of both recir-cu ation oop . Where pipe '"reaks are postulated gogF<RgpNqg <+ ~~~inside primary containment> the reci'rcula

~~~~

7oNS 'J1tlg mK CQ~

~ i table piping is restrained ~ prevent unaccep-motion These restraints are generally mounted on the side of the sacrificial shield

'ystem wall structure or the reactor pressure vessel (RPV) pedestal> immediately below. Pour restraints, which are locate" near the diaphragm floor and are not near the sacrificial shield wa3.1 or the RPV pedestal, comist of saddle type st~ctures mounted on the diaphragm floor.

Ca Verification of Pipe Whip ?zctection Adequacy Sufficient pipe whip protect'on is proveded for

'the reactor reciculation co ling system piping to assure safety as defined 'a 3.6.2.5.2. Pipe ResrRzw T 5 whip are provided "" prevent impact with the diaphragm floor as well as to mitigate the consequences of a pipe ruptuz with respect to surrounding piping systems> ~ructures and com-ponents required for safe sh'-down.

The physical separat'on.of '"e rec'rculation system from the containment vessel precludes any damage that, could result as a zesult of postu-lated pipe break.

3 ~ 6-73

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MASHIHGTOH PUBLIC POWER SUPPLY SYSTEM TYPICAL RESTRAIHT FORCE-OEFLECTIOH FIGURE HUCLEAR PROJECT HO. 2 CURVE 3. 12-5

~ I THIS FIGURE HAS BEEN INTENTIONALLY OELETEO

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Ref e~ed t:o Figure 3.6-35a

'WASHINGTON PUBLIC POiER SUPPLY SYSTEH BREA'OCATIONS %D FIGURE RESTRAINTS ANAL'IZED, PDA 3.12-6 NUCLEAR PROJECT NO. Z 'IERIFICATION PROGRPH

TADLE 3. 12-3 BESTIIAltIT I'BOPL'IITiES USED lN *IIALVS80~

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Hhuru h restraint d pipe - Total clearance C.~ P-a g Q. l4 -g)

I'lpe Size Inl Best Load Direction

~mi t Initial Ef fective Total I C2 6: Restraint Clearance Clearance Clearance 12 Oo 27,733 0.24 6.12 4 1.941 5.941 12 90 14,795 0.401 9 0 '4 12.247 16.247 16 Oo 109,265 0. 6.2 24 8 1.934 5.934 16 90 Oo 62, 599 ~1 8.978 12. 187 16. 187 24 1O2,228 O.24 8.222 1.984 5. 4 24 90 55,531 0.375 C.972 4 13.685 17 ~ 685 24 38 109,888 5.698 9.698 Ppg 4 109,835 0.2~ 5&i 12.462 it) ttsc dunoteu ttucluar services corporation, and PDA denotes Pipe Dynamic Analysis Pro9rum for pipe Sruak Hovumcnt" by General Eluctric Company.

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PD =fPR De!lection Deflection ln.

EL 1

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0

)

4

TABLE 3. 6 -6 Page 1 of 7 DESIGN'ASIS BREAK LOCATIONS OUTS IDE PRINARY COHTAINHHNT Hax. Porc Isometri (k ips) o Plan f ine No. Diameter Thrust vs Time ~pocation Des iona ion W

(H200) (Inches) F iilU 'F i g 0 I t.

<<CCC(13)-4 1?0-1 Later 3.6-49 2 'CICfi13)-4 120-2 3.6-6 , 70 3.6-49 8 RClC(13)-4 120-3 3.6- 5, 66 3.6-49 RCIP(13)-4 120-4 4 Lat r* 3.6-48 RC3jC(13)-4 120-5 4 er* 3.6-48 RC(C(13)-4 120-6 ater* 3.6-48 R C(13)-4 l. 2 0-7 f ater* 3.6-48 1'P 11

'R R R

IC(13)-4 Xe(-1e ) -4 CU( 1) -4 120-8 120-10 126-1 4 f.ater*

3.5-63, 64-3.6-79, 80 3.6-47

3. 6-47 3; 6-51 3.6-50 lQ Rf CU(1)-4 126-2 3. 6-75, 76

l. CU(1)-4 l. 26-3 Later* 3.6.=50 R CCU( 1) -4 125-5 hat r. 3.6-50 Rl CU(1) -4 126-6 3 5-81, 82 3.6-51 R'l U( 2) -4 128- 3. 6-67, 68 3.6-51 8 RH (2) -4 1 -8 f.a6e r* 3.6-51

,9 RNCU -4 128-9 6 Later* 3.6-51 Later* 3.6-50 0 RHCU(2)-4 RvfCU(2)-4 '28-11 28-

.1 28-10 3.6-49

"-6 -"Later~~

1 RtfCU H-) 4 1 -

-q fia tee $

-- 3;-6-4 9-2 3. 6.&0

~2 ~~

2 2 ,~HCU~Q-

~WOO~) 1

-- .l.- 4 ~ - -- fat-ct r ~-

5 6 La e

3. 6-50

3. 6=50-

TABLE 3.6-6 Page 2 of 7 DES IGH BASIS BRBAK LOCATIOHS OUTSIDE PRIHARY COHTAIHHMT H x. Force Isometric (k ips) or Plan Break Line t$ o. Diameter Thr st vs. Time Iocation t4o. Devilnation (M200) (Inches) Figure Figure g6 RHCU(3) -4 129-42 5 r.a t r* 3.6-50 2!7 RHCU(3) -4 129-43 Lat r* 3.6-50 2'!

RWCU(3)-4 1.29-4 4 4 La e I* 3.6-50

2) RHCU(3)-4 129-45 La e r* 3.5-50 3.6-50 31 RHCU(3)-4 129-47 L t er* 3.5-50

3) RHCU(3)-4 129-48 f ter* 3.6-50 3.6-50 3,~4 RHCU(3)-4 129-50 ater* 3.6-50 HS(20)-4 134-1 Later* 3.6-44 HS(20)-4 134-2 f ater* 3.6-44 HS(20)-4 134-3 fater* 3.6-44 HS(20)-4 134-4 Later* 3.6-44 0 AS(11) -2 1.3 9-.1 3.6-97, 98 3.6-43 2 AS(11)-2 .1 3 9-3 3. 6-93, 94 3.6-43 3 AS(11)-2 139-4 Later* 3.6-43 AS(11) -2 139-7 fater* 3.6-43 141-1 f.ater* 3. 6-43

~st o

)

Amendment No. 5 August 1979

'f~

/2 (

p RC2LL p i iRc @i ~

RCR20 RC13 RCRI I I I

RC16 RC12 RCR9 RCR8 RC3LL RC11A RCOLL RCR3A J RCOCV RCR7 RC4LI. RCSLL RCR6 RC 1CV RC7y CEY'C4CV

~ T iPICAL 8REAK LOCATION

'RR3A ~ TYI'IcAL RGSTRAINT RCGA V OCSIGNATION

/

Rr i1 S'il

()C WASHINGTON PUBLIC POMER SUPPLY SYSTEM FIGURE REAC:OR RECIRCULATIOH COOLING SYSTEM 3.6-35 NUCLEAR PROJECT l(0. 2

270CR16 RC24 RCR IS RCI RC23 RCR17 RC20 RC15 RCR14 RCRIB F .2'I RCR\ CR20 RC21LL RCR13 RCR'> I RC12 RCR11

'HR RCR12, RHR SHUT DOWN SHUTOOWN SUCTION RETURN RC13 RCR 10 RC1 RC3@

RCRB RCR2II

'EY iNOTcS:

TYPICAL AIIAAALGCATIGN I, i HIS FIGIJRE REPRESENTS LOOP A, ABACI LOOP B is SIMII.AR EXCEPT AS NOTED.

RCRI TYPICAL RES RAIN T OESIGNATION

~

2. SEE FIGURE 3,6 36h FQR RESTRAINT ~

SUFFIX "LL" INOICA 3 LONGITUOINAI.BRFAK BREAK LOCATION CORRELATION ANO INOIC*TFS 'OP A ONLY BRcAK TYPES 3, ONLY THOSE RESTRAINTS THAT MAY ACT OURING THE POSTULATFO BREAKS ARE SHOWN.

BR'EAK LOCATIOiNS AND WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RESTRAINTS ANALYZED, POA NUCLEAR PROJECT NO. 2 VER IF I CATION PROG RAh1 3.6.35a

5 v/'gL g C" Fu>rn BR8002A [9/82) 200M

le O

n m a r

O )2 PiPE m O SLEEVE m

IO CA Q

0 O

C 4/l C/l m

4i 42 PiPE 40 LOOP l5 SiEEVE l/z m

+ 32

~~O LoOP 45 C 5

4(p qS

TABLE 3.6-5 Page 3 oE 7 DES IGH BASIS BREAK LOCA'f'IOHS OUTSIDE P)/IMARY COHTAIHMEHT Max. Force Isometric (kips) or Plan Break Lin Ho. Thrust vs. Time Location Ho. u~esi nation ( (1700) Picture riciure 5 AS( 10) -2 14 1-P4 Later* 3.6-43 5 AS(10)-2 141-12 l.ater* 7, ~g 3.6-43 5 RHCU(f)-4 142-20 Later>>Fi( g.k8> 3.6-51 5 RWCU(l)-4 142-21 Later" 3.6-51 5 RHCU(l)-4 142-22 4 Later* 3.6-51 6 RHCU ( 1) -4 142-23 Later* 3.6-51 7 RHCU(l)-3 144-24 Later* 3.6-53 RWCU(l)-3 144-26 Later* 3.6-51 RHCU(1)-3 144-27 Later* 3.6-51 RHCU(1)-3 144-28 f ater* 3.5-51 RHCU( 1 ) -3 144-29 Later* 3.6-51 RHCU(2)-3 144-31 Later* 3.6-51 RWCU(2)-3 144-32 f ater* 3.6-51 RWCU(2)-3 144-34 (fater*

I 3.6-51 69 RWCU(2)-3 144-36 fater* 3.6-53 70 flS(9)-2 148-1 3.5-112, 113 3. 5-4'3 71 HS(1)-2 148-2 Lqter* 3. 6-43 75'fs(5)-2 73 74 HS(5)-2 148-148-6 Lager>>

Laker>>

3.6-43 3.6-43 Ifs(5)-2 148-7 Later* 3.6-43 O

TAAf E 3. 6-5 Page 4 of 7 DESIGH BASIS BREAK LOCATIOHS OUTSIDE PRIHARY COHTAIHHEHT tlat. Force Isometric (kips) or. Plan Break Line Ho., Diameter Thrust vs. Time Location Ho. oeaictnation (H200) (Inches) F~iure Figure 76 HS(5)-2 148-8 2 Later* 3.6-43 77 HS(5)-2 148-9 2 e 3.5-43 78 HS(5)-2 148-1 2 Later~ 3.6-43 79 HS(5)-2 148- 2 Later* 3.5-43 80 HS(5)-2 148- 2 at r* 3.6-43 I I

• HCO( 11) -1 149I'2 Later*

I

3. 6-62 i

a HCO(ll)-2 149-5 3.6-99 3. 6-58 I/

95 RFtf ( 1) -4 5-1 24 - Later* 3.6-49 96 RFH(1)-4 335 24 Later* 3.6-49 97 RFvl(1)-4 335- Later* 3.6-49 98 RFH(l) -4 335-4 24 3.6-49 99 AS(9)-2 342-13 6 Later* 3;5-43 100 .

AS(9)-2 342-14 Later~ 3.6-43

'K O

TABLE 3.6-5 Page 5 of 7 Dt'SIGN BASIS BREAK LOCATIONS OUTSIDE PRIHARY COHTAIHHENT Hax. Force Isometric (kips) or Plan Line Ho. Diameter Thrust vs. Time Location Des iwnat ion (H200) ( Inches) FiIaure Figure AS HS(l)-4 400-8 Later* / 3.6-44 HS( 1) -4 400-11, 26 i Later*

"/ 3.6-44 I

HS(1) -4 400-'l4 26

/

Later* 3.6-44 4

HS(1) -4 1

.400-18 '6 Later* 3. 6-44 C

CO( 3) -2 ',440-1 .2. 5 Later~ H/A CO( 3) -2 440-2 2.5 Later* H/A CO( 3) -2 440-3 2.5 Late'r'atej* H/A llS(5)-i')S(5)-1 4'47-19 6 H/A 447-2'5 6 W/A Later'ater HS(5)-1 4 47-26 N/A HS(5)-l 447-27 5 Later* N/A HS(1)-1 448-15 6 Later* N/A HS(l)-I 448 6 Later* N/A 0

0

TAI3LE 3.6-6 Page 6 of DESIGN BASIS BREAK LOCATIONS OUTSIDB PRI jjARY COHTA,IHtjEHT Ha@. Force Isometric (kigs) or Plan =

Break Line No. Diameter Thrust vs. Time Location Ho. D~esi nation (H200) (Inches) P~iure F tgll te 126 HS(1) -1 448-17 6 Later* H/A 127 jis(1) -1 448-18 6 Later* H/A 128 jiS(1) -1 448-19 6 Later* H/A 129 HS( l)-1 448-20 6 Later" H/A 130 HS(1)-j 448-21 5 Later>> N/A 131 HS(1) -1 448-22 5 Later* N/A 132 HS(l)-1 448-23 4 Later>> N/A 133 jiS(1)-1 448-24 Later* H/A 134 fiCO(5)-1 449-13 3 Later'ater*

N/A 135 Hco(5)-1 449-14 3 H/A 136 HCO(5)-1 449-15 3 Later>> N/A 137 jiCO(5)-1 449-16 3 tater" ter" H/A 138, Hco(5)-1 449-17 3 Later'ater*

H/A 139 jjCO(5)-1 449-18 3 N/A 140 jiCO(5)-1 449-19 3 Later'ater*

H/A 141 HCO(5)-1 449-20 3 H/A 142 jjCO(5)-1 449-21 3 Later'ater*

H/A 143 HCO(5)-1 449-22 3 N/A 144 Hco(5)-1 450-33 3 Later* N/A 145 HCO(5)-1 450-24 3 j.a H/A 146 Hco(5)-1 450-25 3 Later* H/A 147 HCO(5)-1 450--26 2.5 Later* H/A 148 jico(5)-1 450-27 3 Later* H/A 149 HCO(5)-1 449-28 3 Later* H/A 150 tjS(9)-4 451-6 3 Later*- iV/A O

TABLE 3.6-6 Page 7 of 7 DESIGH OASIS BREAK LOCATIOHS OU'L'E PRIHARY COHTAIHHEHT Hax. Force Isometric . (kips) or Plan Line meter Thrust vs. Time Location

-~

Brea)'o. Ho. D nestcSnation (I<200.i Inches) Fi ure 2 l.flite 151 HS(9) -4 4g'1-7 3 Later* H/A

<.1 5"2-=" CR D (-1-2.)>>3--- .H/.A 8; Se 3 ..18.3.5

'

a, re-sponse t HRC Ql stion 010 14.

• Information is scheduled to be ready for Staff review in late 1982.

pl az m

~W CO &

o

WNP-2 . AiMENDMENT NO. 25 June 1982 TABLE 3.6-7 SEISMIC AND QUALITY CLASSIFICATION Page 1 of '2 r

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.S (16)-2

/ 2,3,4 2.5,3 i '; I I

ET Ei CO (3) -2 2,2.5 ii HCO (5) -1 HCO HCO

( 5)

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. IIUCLEAR PROJECT IIO I'IAIII STEAH LOOP A FIGURE 15ONETRIC 3-CN

tRIP-2 Amendment No. ~gQ SUHHARY OF POSTULATED PIPE BREAK LOCATIONS CIRCtiHFERENTIAL BREAKS LONGITUDINAL BREAKS Node

't 216 Node 220'ode

~

~1ode 219 228 225'ode Node Node 231 221'ode 493

'or 2.24 Node Node 226 227-

~os<<>> 6A7 Node 229 ypDE Node 230.

Node 232.

Node 291 H2lSHIHGTON PUBLIC POWER SUPPLY SYSTEM FIGURE

'AIN ')Enl 3.6-12 NUCLEAR PROJECT ~ 2 LOOP A

N~ ~ ) ~

aSI Rf ACTOR VE5SEL MS. NOZ'XLE. NBS

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PW5 32;5 PW5 SL- Co I

I 237 .

'55 234

.llII STEAI ISK=iRIC fIGURE

~g g~~~4, LOOP S

~~ c 3.6-13'

Amendment ND.WD@

NNP-2 SUMMARv OF POSTULATED PIPE BREAK LOCATIONS CIRCUMFERENTIAL BREAKS LONGITUDINAL BREAKS Node- 236'ode Node 235~ 241'ode Node 237.

250-247'ode i

e'o~ CYz Node 240 4'0 Node 242 P >E QOOE g3 y 246

'ode

~~<<(z7 248 Herl e 249.

'ode

~ouS'3 Node 251 ~

Hode 292 Node 636'~~

Node 640'ode 644'oE

~llew 4

I'Ali'TEAN I WASHINGTON I

PUBLIC POWER SUPPLY SYSTEM NUCLZAR PBOZECT NO 2 LOOP 9 ZZCURE 3.6-13'

REACTCIR VES5EL Ihs HOiX,LE. N3C L'I3 "2TO . r

'ltoS~ PttVS 3'3-I I

2t/C7 e 2@i PW5 33-+ 2.6 Iso)-+ I PWS 33-1 I 7

tA5. RV- IC

.I C77 Q (77 p 7

'.

)IF i ~5. ~ 'I I7AS- RV- K

~

I1

&8 )

/ 'j l)p .c 4T ~ .

'- l

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r ~ ~

I 258" 1 / ~(5Q,fAS-RV-~ ~

~2( 7 I

4/52 'BpCt

,', f'$,7k t'p FE MS-RV-SC.

F LOU G5I /F g-Ib C 5 FI,SSf P.ICToR 2t5'-

/ 7

~ a 1

w~'(s-. 2v4 MS"V 2:LC PW5 33 3

'55

-'2C2.

Pb; PV/5 33-S 2E 3.

t;!,

~ 1,17 AHBIIDHENT HO, 9

) .t 2'5 4 april 1900 7

1 37" MA lf6TON PNLIC PORE) SUPPLY SYSTBI ( j~ltl STEAtl HAltl

'flak g j7/7,. ~ 7 j IIUCLEAR PROJECT tt0 2 STEg LOOPp C ISONETRIC FIGURE

~. / .6-14a

WNP-2 Amendment No, &DR

SUMMARY

OF POSTULATED PIPE BREAK LOCATIONS CIRCUMFERENTIAL BREAKS LONGITUDINAL BREAKS 255" Node 260-Vode 254 bl 256 'ode Node 266 Node 269 ~

uaoE 4$ ~

'ode'59'ode uonedS/.

654'ode 261 4odE 4'Jg Node fpoDE 4 /7: ~

267>> rvo<E. 4 ~~

265'ode Node 270~

268'ode dO-Vode

'r1ASHZNGTON PUBLZC PCNER SUPPLY SYSTEM PZGURE HAIN STE'Atl LOOP C 3.6-14b NUCLEAR PRQKKT NO 2

REACTOR, yE55EI lAS NOZZ,I g g3P use, PINS 3+-!

',l 28 I

ppMS-RV-+O 282, DMSO) g

( P~S3~+

IAS-RV-aO

.c 58 l(-ISO .451 lg f

MS-RV-Z,O

+ 'eS-V-22P FLDW ZSTRIW Gil MS-RV-!O r

S . P'%MS 3+3

'

PVIS 3g-2, w" s~

S. CM.

r,)78 GGl PWS 3+-4

~

gf:-"~k!r re.>~ s '283

'l84 +

< 285

.%13

~14 275 i -, .;.g

~ ~ < )- -)

%1C 'g$ IIIIIOTL'IPIILICPOWER SUPPLy SZSlgI r

e rr Jrrrrlrr<4 <<4a u0 ~'I ~.t35'I FAR. PrrAJFf T AA. ical LOOP A !~OIrr:ynlg

1 Amendment NO.~89, WNP-2

SUMMARY

OP POSTULATED PIPE BREAK LOCATIONS C'IRCUMPERENTIAL BREAKS LONGITUDINAL BREAKS Vode 272 Node 276-Node 281 Node 'ode 277 275'ode 284'ode 287'aoE Node 280- 4s P.

Node 282. AtodE SS7 Node 283

'ode mom ~~

286 285'ode

'ode 290 288'ode

~~60 WASHINGTON PUBLIC POWER SUPPLY SYPH FIGURE Hjtn,li"l ".i"=AH LOOP 0 3.6-1Sb NUCLEAR PM'~ NO 2

I26 Pv/S 21-(

I21-~l ~ 129 132 I3(a 137 I>I V~S PWS 21-p

12. RFW(A-6 12- RFW(ll-0

- 135 114- / -13<

PWS 21-15 133

>Xl ~ y g II3 PITS D-5 At- <112 123 II I PWS 21-16

'97R

+ PW5 21-2.

-IS- I IS RFWII).+ .-Ilo l'AS 2.1 IO PWS l-~

?.

C9, 2+ <<18 Rf D.

PWS 21-IS 12&

PWS 21-11 125 PW5 U-+ g ~ lo2 lo I 12o-q IS< 12. RF.D PWS D-I2.

IOO lo5 X-11A RF W-V- IIA 12. RFW(l~-+ 104.

~ luIol G-8 PWS h/ 1 TEST CONN, RFW-V-IOA

~ PWS Pwh 1.1.11 21-13 85 65 PWS 2. l-I+ 61 PWS 21-G P~IS 21- q IO8 Io'I I,.IP P@E$ , /PLY SYSTEII

~k ggg0$ 'I2ji<-" REACTOR FEEOI!ATER

(<IIIE 'A) ISOPETBIC FIGURE

BURNS AND P w.o. r Orawii h/N P-g rf~t tutor tu 7 i5'Fz

/ OS( </-87'->lI'2FH /OCr'7iOIuS

( /ZC ul"IIPDIZ./=~7 IR-( 8~iS gS 4.o~d l 7 uDirva/ PrZ Z@

gC'edES

~mES

/> 'i/ JJ 5 g7 9'78-.-

/w(~ //7 la v 1/(Z )l3 JQG A Qo /Qg r QZ/t

/'34 (t 3v IOO /3w

/0 J /37 IO<

IOv~ jg'7 /o-

/Oy /3y

'l oW

/o7, //7

/C(. /c/

lie

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("" 'OC (.(

NO~i=~

i~rr PER<r0nfi rv~i~rÃz.t /Li r A ).

/IC<gEr r ~O.

Form BR8002A (9/82) 200M

185-PWS 28-1 16'i'FW(ll NOD

+ ~

Ieo lel t II I 92 190 189 EPWS 28-8

~

PWS 28-$

17a - PWS 28-2.

ITT s)C I82 184

12. RFWAl 4 I CS- 12. RFWtl).+

-pws'te

/ ~~It 9 Ib Isg PWS26-ts ~Ill 199 150 ~~ ~tt

~ 146 7

Zl AF wo)-$

ISie ISa4 EPWS 28-+

IL F W-II-I La 113 ~ 17S

-174

- 153

'72 XL75 PWS L~~

2II-I7 l/~ta51 COIIH pA 111 PWS 28-11 170 PHS 28-Ih EPWS 2.8-'I L3'3 FIF~-V- IoB  ?+its Rf.L3 PNs 78-to PWS 28-l4 14l 14l 18 AI Lll +

15'c 155 PWS 28-8 IS 151 ISIS PWS 28.12, ICs I

~ IC,S ICQ

~ IC 4 I ~

. Isiti REtl It 2 PWS 28-1 I5'j Pws i8-8 It.e-I.'ASI ;igji'l'lyg",supviTsTsTER

";JNW,',~IO-.:,2.. REACTOR ISRLETRIC FEEO'sER (I.lllf 8) rICuRE 3.E-IFI:

MNI SunnarZ / OS /a- - ~+y g P>p)= r rZ c-X~ /-. a c.)> Wgm.S

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/rg>>, /54 /9'9

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/ QlJ /ky

/gp

/F7 /~ra

/ 9'/

)re Sita Cr re /~ulnar C PC err,~ /2 fd<TOf/ /E L=O~+TFR (sJ Ju E /2) j/;. z/c. ~

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PW5 '5Co- I9 AIIEIIDjILHThO 4 Scg

! Apt:Ll 1MO

• h I PN5 3Ej-IB 9+5 BG 2.0 t-PW5 aG-Iq PW5 5co-IQ 425 L 4ZS Qg4~ RWCLI-V-I FE 3

PWS ~S-IS

+I9.

. 42o.

' 36- IIt't PIjAjth

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~ f.'-;.' 422

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~ j<<hJ<< t.t it I IhlASIIIHGTOH PUULIC PNER SUPPLY SYSTEH HUCLEAR PRMECT HO. 2 REACTOR MATER CLENIUP ISOHETRIC FIGURE 3.6-18

.C RWCu(3)-+

PiV5 86-I 349 3'les >So PWS 3S-Z.

< FOR CONT.SEE FIq 3 g PCS 3II'I-IO 398 3'I4 ' RWCLI-V-gO 39 I 3r-3 RWCU-y-IC2. ~C"='A.CLE<<<<T PWS 4oo

+ CHEN'. CLEANOUT 3qZ PW$ 3G i ~-3SI 8

. =~I(,

VALVE 5P,FE Egg fOR CONT. SEE 4 og ~~36IIS ~ 51l RC-g5 39l +

3 g 3p plwcu -y- IoFo 3894, e.9ICL1 VAl VE SAFE END

$ 74 3uI Il 373 332. +R.RC(+)-45

, +RRC(+)'-'ts 388 )83 31 Fo

.387' 377 PWS 3I -R4 gylC'U =4

-RRC-4 P

WS.5F,5 672 PWS 36-6 3ll PW5 Mr26 Sle... PWS 36-1

.

3ISA

?ARRf (2Q-45

.3I'3 PWS 3II I2.

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PWS M-9 2W RRC(Zl-qs s

)th ~ 347

~ j

/'-~~-=-- ..

I3:

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'27

...84 5 . y~P"g ED RRC-V-5I+

4~2. RED 33I .

I.

I' '. ~" PwS N . 33(3: - 33o

2. RRC(Q-WS

.344-= . ~:

331 ~ ' ..92II

%c.b '35'I .

355 .

~gee-v.ps 3c'X.

- ". ~ 358 - . ~

~~

Is i 334, ..933 3go RRC-V-Sl B 331 '55 '32o- .. 3~~ .

325.

324 RRC.V-52.B l

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. 2. ARC(C)-AS r . 353

~ - ~ 317 t) Il~ ~ I)-

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I)

'I'QKIIOHENT NO, 9

~

+'EOR(41)-I Apri1 1980 IllIlPIIIIflilsis; [gsll ssi assi-lsistss

)~a.,]~ Q~ILlKCJiIILO~~~I,:,'~j ., ~IIggOtI VhTfR a~aIRW Iri~t.IRIC F IFsOR1E

.6-IRI

NOJP-2 Amendment No.

APJ'i1 1980

+y w~ ~

SUHf!ARv OF POSTULATED PIPE RREAK LOCATXONS C: RCL'!F ERE."JT XAL BREAKS LO.'!GXTf!DZ"JAL -"ip f.. sf:.5 ilode 427 i

'WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE REACTOR HATER CLEANUP 3. 6-18c I NUCLEAR PROJECT Bo~ 2

Amendment No. PQ WNP-2 A pri 1 1980 SUHi'~ARY OF POSTULATED P Z P E R RFAK LOCAT TONS C:RCU!!F.".R.".NT::.~ ~ REAKS

@jog'33 Nod e 3@A ~ Noae 379 Node 391" Node 394 Node 395>

Node 396~

Node 366 ~Le~

Node 367~ L~

Node 368'!

6

~ Zr O'T 7

8/@05 pS7 LONGITUDINAL BREAKS XT ( D RASH&lGTON PUBLIC POWER SUPPLX SXSTKN PXGURE REACTOR WATER CLEANUP 3. 6-18d NUCLEAR PBC:.i!~ NQ 2

Amendment Na. 9 Apr>1 1980 RE,ACTOR V E.S SE L.

S.L.,t". N.QXX,LF WII 2" SC.H.,BO PiPE ,

(SY l'-'.E3 r

3Ic ll P l (l)'-+2

~

.

EIQ 3l&

Gl2~.

hlO

!

//

~l'L())- 4G

~ 3og 505 30l 300 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE NUCLEAR PROJECT STANDBY LIQUID CONTROL ISOMETRiC NO. 2 3.6-19a

5~ ~ 't% 1 y A 4 W.O I rsUB Orav

~ By Yi(t(

BOC ply

(, (5+

38$

i

~ I yr(r) -VS 2=7 Iy) q7

~y 'SC 2'

c(IZPy6 3o!

100

(~ <<p 4 / gIy'I/2

-ulrr C'/ "- ~)5r c-/'y 4 g/C C/. /~z.'<g/Pl 7'

form 8R8002A (9/82) 200@i

ai iaMc hlA QQ Qr g/.E I Or g g5 dQ~~

QJO 3Q 3>l 0

~// /A !i~-fb/u I'4'l'.6/c j'o'~

(~g/upI2 ~/ 4 g qg/~ ~~77/r,i 5'1 /',.~/, y + y'.- < i/ ~LICE.EA/l Form BR8002A (9/82) 200h1

F H R,-V" +l A RHR -V- Ill A 39 27 3P

$ 25 24 Pw5 5-2.

I+ RHR,(l)-+

'22 Xl .

l9 RKACTOR. VKS5E.L, RHR. NOZ.jL.E iU G WASHINGTON PUBLIC POWER SUPPLY SYSTEM RESIDULAL HEAT REMOVAL FIGURE LPCI MODE (LOOP A)

NUCLEAR PROJECT NO. 2 ISOMETRIC 3.6-20

Amendment Ho. P V4

ÃNP-2 8Ut!>!ARY OF POSTULATED P X PE RRFAK 'CATTO."IS O'F.F!E."JT 2 L BREAKS LOH(i ET'D I MAL R REAKS

'.Iod e 17 1a, ~

18 'ode Node 20. Node 27 4Ie4t4l~

Node e 28 26Iod

'ode 31 ~

RESiOuAI. HEAT REMOVAL LPCr FXGURE HASHXNGTON PUBLIC PCNER SUPPLY SYSTZX NOOE LOOP A 3.6-20 NUCLEAR PROJECT ?K) 2

39 Pws +- < F l(i)-WS l+ RHR.(l)-+ RHR-V- +l B REACTOR, VE.SSKL RHR XOZZ.l K Nl"o l+~ j2. RE.D.

5'7 WASHINGTON PUBLIC POWER SUPPLY SYSTEM RESIDUAL HEAT REMOVAL FIGURE NUCLEAR PROJECT NO. 2 LPCI MODE (LOOP 8)

ISOMETRIC 3.6 Amendment No.~+@,

tRJP-2 STJVJ<APv OF POSTUTATEO PIPF. RRFAK T.OCATEOPS

".Ft'.RE:!TTI XT RP,F'AKS T,O'".,r, . Tl J,) T.NAT, RP.~:.K.'h Node 33 de Node 34 35'ovie Node 36 43 ~

Node 42 ~

44~ 'lode Node 47 ~

MSHZN~iN PUBLIC FIGURE POWER SUPPLY SYSTEM RES/DUAL }IEAT REt10'jtAL LPCI 3.6-21b

!

e I

'UCLEAR PROJECT 80 2 MODE, LOOP 8 .

1 Pl(l)-+9 RHR," V-111C PWS 3-l S5 SZ 50 WASHINGTON PUBLIC POWER SUPPLY SYSTEM RESIDUAL HEAT REMOVAL FIGURE NUCLEAR PROJECT LPCI MODE (LOOP C) 3.6 NO. 2 ~

ISOMETRIC

I 0

NNP-2 Amendment No. ~+

SUfft!ARY OF POSTULATED PZPF. RRFAK LOCATIONS c"'<1P~bc'NIT> ~ T bRP l',ONClZTUDENAL RRFAKS I

x'tp d e ~>9 ~ Node 51 Node 50 ~

Node 52 ~

8a&~

I Node 59

'.lode 58 ~

Node 60 Node 63 WASHINGTON PUBLIC POWER SUPPLY SYSTEM RES IOUAL HEAT REMOVAL LPC I FIGURE HOOE LOOP C '.6-22b NUCLEAR PROJECT NO 2

Amendmen Na. 9 APril 1980 Q3q l2. RHR(l)-+5 (o7 l i/ALVF SAFE END RHR,-V- BOA 72 (uSA P. 4 RKClRC. Pe+0 Dl&CHMGS WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RESIDUAL HEAT REMOVAL NUCLEAR PROJECT NO. 2 SHUTDOWN COOLING (LOOP A) ISOMETRIC 3.6-23a

0

,

l l A

Amendment No. +D+

! JHP-2 0 >OS<UL-'.TED >>PF. P.REEK LOC'. ~O'1S C"'!FERFVT~AL RREiXKS LOF1GZTUDI 1AL RREAKS

~Jode 65 ~ i'bande

'1ode 65K ~ 65B'ode

.'1ode 65C' 68 ~

Sod e 65G ~

!Inde 66 ~

Node 67 ~

3atxe A/ODE-

~

.'1ode 69 ~

To ilASHINGTON PUBLIC POWER SUPPLY SYSTEM RESIDUAL HEAT RB<0'lAL SHUTDOWN FIGURE COOLING LOOP A 3.6-23b NUCLZAR PBOJECT HO

24 Zec.ia.c.

~VMS OlscvA~6 'tie 7<C 12" RARE) -CS iz HR+( IZ. &

74 72+

-78

~vv 'H R-V-50 6 VALVE ~F2 EHa WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RESIDUAL HEAT REMOYAL NUCLEAR PROJECT NO ~ 2 SHUTDOWN COOLING (LOOP 8) iSOMETRIC 3,6-24a

I

SUMMARY

OF POSTULATED PIPE BRFAK LOCATIONS P CU.'5FE RENT E AL BREAKS LOI'lQ XTlJD CHAL B REAKS Node 72A ~ Node 72C<

Node 72B~ llode'2F~

Node 72De Vod e 76 Node 72E ~

8ode 72G"

.'lode 75 WASHINGTON PVBLZC POWER SUPPLY SYSTEM RES IOUAL HEAT REHOYAL SHUTOOkltl FIGURE

' COOLING LOOP B 3.6-2Nu<TlOm OP ~576M PUMP GUCTlOh4 79@

796 79K

'79 D 79K'W 791

'

jar.

-i WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RESIDUAL HEAT REMOYAL

'UCLEAR PROJECT NO. 2 SHUTDOWN COOLING SUPPLY ISOMETRIC 3.6-25

Amendment HU.~84 LPiTP-2 SUhifQRY OF POSTULATED PIPE BREAK LOCATIONS C I RCUHFEREHTIAL BREAKS LONGITUDINAL BREAKS Node 79A~ .far'h Hode 79F i Hode 79Ii Mode 79E+

Hode 79G~

Node 79H Node 79J

~Re-SQ e

Node 82 FIGURE WASHINGTON PUBLIC PCWER SUPPLY SYSTEM RESIDUAL HEAT REHOVAL SHUTDOMN "'. 6-COOL IHG SUPPLY NUCLEAR PROJECT MO 2 25b

C2$

~G'900 T$ 5 SMALL l 6N&uE l 1.8, REACTOR. VE,55+.

R.t lC NOT:LEL N f WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RCIC RPV HEAD SPRAY ISOMETRIC NUCLEAR PROJECT NO. 2 3.6-26a

Amendment No.~

WNP-2

SUMMARY

OF POSTULATED PIPE BREAK LOCATIONS CIRCUlIPERENTIAL BREAKS LONGITUDINAL BREAKS Node Node 624 623 622'ode 625 'ode Node 626 ~

g py yE/PO Spacey ~

WASHZNGTCN PUBLZC POlfER SUPPLY SYSTEM PZGQRE UTDONN NUCLEAR PROJECT EO 2 .6-26b

12."xiO aE,O. ELL.

PW5 l-1 f 5 CPC S-V-5l 1> CPCG(l)-+

LPC5- Y-6 WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE LOW PRESSURE CORE SPRAY ISOMETRIC.

NUCLEAR PROJECT NO. 2 3.6-27a

0 Amendment No.

'WNP-2 April. 1980 e SUHtIARY OF POSTULATED PIPE BREAK LOCATIONS CIRCUllFERENTIAL BREAKS LOl'7G IT U D INAL B REAKS llode 1>> Node 3 Node 2>>

lode 4>>

Node 5 0 Node 6i Node 7 ~

WASHXNGTGN PUBLXC POWER SUPPLY SYSTEM( FIGURE LOW PRESSURE CORE SPRAY NUCLEAR PRO3ECT 5G~ 2 ~

3.6-27b

4 Pt(t)-+5 RE,AC'TOP VK55E.L HPC.5 NOZ.Z,LK 8 IG t~'~lO R,ED EL,L.

POI5 2."'i HPC."5 - V- 5 t t4 lE HPCS (l)-+

WASHINGTON PUBLIC PONGY SUPPl Y SYSTEM FIGURE HIGH PRESSURE CORE SPRAY ISOMETRIC NUCLEAR PROJECT NO. 2 3.6-28a

AMENDMENT NQ ..MD A NNP-2

SUMMARY

OF POSTULATED PIPE BREAK LOCATIONS CTRCUMFERENTEAL BREAKS LONGTTUDINAL BREAKS Node 9 ~ Node Node 10 12 'ode 13 'ode Node 14~

Node 15 ill 'oge ll'ASHINGTON PUBLIC POWER SUPPLY SYSTEN FIGURE HIGH PRESSURE CORE SPRAY NUCLEAR PROJECT 80 2 3.6-28

~~

r NO M~2 ISG2

~ .PWS l9l.

SO-g."

PV45 SO g 2',

2'PWSSO C2 214k M lA5.

I4 RCIC(lb) 4r 214 6 203

'214 f Zog a J

-

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- -21 2A . +02NN.

iZ 2ol

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jp r

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4 RCIC 03)-'t Rclc+@ ..

~

I r, ~

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',

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~py .g y 3,. i I(GIOII PNLIC POUER SUPPLY SYSIEII RIIR CO.BEIISltlG NOOE FIG UP<

.",:IIIUCLEAR PROJECT IIO. 2 RCIC TUROlttE STEAtl ISONEIRIC I.6-29'

NNP-2 AMENDMENT NO W~

SUMt&RY OP POSTULAT-D PIPE RRF'AK OCATZONS CZRC"')~~i.~NTKAL 3RI'.AKS I,O'1GETUDi'!AL PiPEAKS 484m 'L99 Node. 213 +

Sade-&&6 ianna i s

') e

'~nc1+ 212@ ~

ode 214 r WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE RHR CONOEi'ISING 1'IOOE )CIC 6 ~ob NUCLEAR PRCLTECT 80 TURB Ii'lE STEAM

I~

15QQ SC 2.

li I pws 55-$

'I t WS 55-I t X-22.

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'555 I65(9I-+ '"

1 54l 5'ZS(9) ~

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1

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~

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'=

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54K 560

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544 5&'T IS....:. 1.- r ASS-I~I 41 L g ,...PWS 55-.II..

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~

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54$ )

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~ T SS=JS; L-1) -,...541;M~""- C

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bjgkm,,, 'le

&45 52o. ,FVIS 55-.8

'<,lcr Ns-v-2.'N:

. 89+ .;.

555

)!IlFf'5R l~ 8555-15::- 2. IAS(9 L \

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.

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. fall~

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~~

~58%

564

.

555.

l WS 53 ) ~

58r 58'I

!)GTOI PUSLIC P0%II SUPPLY SYSTEH r~lll STEAN VALVES DRAlhAGE FIGURE

)'Lji.lit@LEAR PRQECT1 Il0. R FIFIIC tsa~nGIC *

!l .6-30' 1

NL '~Uk~

AMENDMENT NO. M~

SUMMARY

OF POSTULATED PIPE BREAK LOCATIONS CZRCUMFERE."1T AL BREAKS

~

jane~

Node 543 He+~K

'ode So&~99 Node 557 ~

i3~55ft-560'ode 562'ode 565~

Node Node 577 WcÃb~zZK Ne4e Node 580 Node 576>>

574'ode 5R

'82 Node 566 Node 583 ~

Node 568>> 'ode Node 585 ~

550'ode Node 569~ ."1od e 586 Node 570~ Node 588'89>>

S~k~e~~ Node 556 554'ode

'WASHINGTON PUBX IC POWER SUPPLY SYSTEM HAiN STEAN VAL'JES ORAINAGE PXGURE PIPING 3.0-30b NUCLEAR PROJECT NO 2

655 PW5 52:II .+s2 RED.

h =P

..-900 Ti8,

,SQUALL tbWC uE Fl.t.,

REAClOR yE55EL C(sS IAS hlOt.gl.g Ng tI~!,i:> I; PWS 52:IO

',- .

~ .

'r!i(}r

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r.85

~g ) gt=HD

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+ gag'>>

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( r'2,(I +.~ .. ~: &G>tOe( .. 48'SB'"~ '-.

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.

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.

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NUCLEAR PROJECT NO. 2 INSIDE MAIN STEAM TUNNEL 3.6-33a

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'NP-2 AMENDMENT NO. P cP SU."1f&RY OF POSTULATED P'ZPR >PEAK LOCATEONS

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WASHINGTON PUBLZC POWER SUPPLY SYSTEH FIGURE HAIN STEAH LOOP A, 8, C, 8 0 INSIDE NUCLEAR PROJECT NO~ 2 HAIN STEAH TUNNEL

-3.6-33