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| issue date = 09/27/1983
| issue date = 09/27/1983
| title = Forwards Design Reverification Program, Vols 1 & 2,final Assessment Rept.Results of Program Will Be Presented to NRC in Late Oct 1983
| title = Forwards Design Reverification Program, Vols 1 & 2,final Assessment Rept.Results of Program Will Be Presented to NRC in Late Oct 1983
| author name = SORENSEN G C
| author name = Sorensen G
| author affiliation = WASHINGTON PUBLIC POWER SUPPLY SYSTEM
| author affiliation = WASHINGTON PUBLIC POWER SUPPLY SYSTEM
| addressee name = DENTON H R
| addressee name = Denton H
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000397
| docket = 05000397
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:REGULATORY:1ORNATION DISTRIBUTION SYSQI (RIDE)AOCE'SS ION NBR;8310070302 DOC~DATE: 83/09/27NOTARIZED:
{{#Wiki_filter:REGULATORY:1ORNATION           DISTRIBUTION SYSQI (RIDE)
NO DOCKET'" FAGIL!50~397 NPPSS Nuclear=Projects Unit 2i Nashin'gton IPublic Powe 05000397!'AUTH~NAME AUTHOR AFFILIATION SORENSENgGB'0
DOC ~ DATE: 83/09/27 NOTARIZED: NO
~Nashington Public'Power.SUpply System''RECIP",NAME RECIPIENT AFFIL'IATION DENTONiH~RE Office of Nuclear Reactor Regulationi Director>>"
    '"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>>


==SUBJECT:==
==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
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:. '
',.ENCL,[SIZE:.'L.$0.TITLE: Licensing"Submittali PSAR/FSAR Amdts--8 Related Correspondence'OTES:
L .$ 0
L>>i+>>1 0>>et'INTER RECIPIENT ID CODE/NAME NRR/DL/ADL NRR LB2 LA NAL: ELD/HDS2 IE/DEPER/EPB 36 IE/DEQA/QAB 21 NRR/DE/CEB 11 ,NRR/DE/EQB 13 NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/DHFS/HFEBOO NRR/DHFS/PSRB NRR/DSI/AEB
        .TITLE: Licensing         "Submittali PSAR/FSAR Amdts--8 Related                           Correspondence'OTES:
'26 NRR/DSI/CPB 10 NRR/DS I/ICSB 16 NRR/DSI/PS B 19.:.NRA/DS I/RSB 23 RGNS~EXTERNALe ACRS 01 ,OMB/DSS (AMDTS)LPDR 03" NSXC 05>>~COPIES LETTR'ENCL>>
L>>     i+>>1 0>>et' RECIPIENT           ~COPIES              RECIPIENT                                            ~COPIES ID CODE/NAME           LETTR'ENCL>>        ID CODE/NAME                                          LTTR ENCL NRR/DL/ADL                 1    "0'0    -NRR   LB2'C                                                1    0 NRR  LB2 LA                             AULUCKiR~            Oi                                    1 INTER NAL: ELD/HDS2                                         IE IE/DEPER/EPB 36                               FILE'E/DEPER/IRB 35                                    1 IE/DEQA/QAB 21                           NRR/OE/AEAB                                              .1 NRR/DE/CEB       11     .1              NRR/DE/EHEB
1"0'0.1 2 1 ,1~1 1 1~1 1 1 1 3>>RECIPIENT ID CODE/NAME-NRR LB2'C AULUCKiR~Oi IE FILE'E/DEPER/IRB 35 NRR/OE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/OHFS/L'QB 32'RR/DL/SSPB NRR/DSI/ASB NRR/DSI/CSB 09 NRR/DSI/METB 12'AB 22'G'FIL'4 R AHI/MI 8 BNL(AMDTS ONLY)FEMA REP DIV'39 NRC PDR 02'TIS~COPIES LTTR ENCL 1 0 1 1.1 2 1 1 1 1 1 1 1 1 1 0.1 1 1 1 1 TOTAL NUMBER OF~COPIES REQUIRED: L'TTR 53 ENCL' f'I IHI HI Hi 4 4 Hi~rg)HH I PH f k H fH4fif6 i i<e~g 1>>>1<<ar'<<iraqi.fi",>>4'I
                    ,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 ~
~b~i0">ii,'"1~rfH l fl ,>f f>,i'I Hf, f, l'f fH>c Ii'l'II I''14I 14'i I'1 IH41 J l HI H f IrlI jli 8<<NHH r, sf 744 il~'11 l I I>4f 1 I , HSi S A ay,e II i~'I 1(f f f f~fw/I~411 C$4,-i f 7 I'I'iV'1~I 11'4 HI 4 I''II>>w IV.'I C'H,'1,A">j'1>'I444P9P4H off'1 il1 Hq 1,1 lyifllP 4 ic, 1 1 11 f~i H4 l', I I'i l'I 4'".I h.iy~,4,fg, fr>If cifI'4 1r)4 Ilv I I t f Ir 4f I f I<IX HI, I'I y J I f I 4\4r HI fpL 4', HI@)r P>>IXII<3 ,)L f!.4 Washington Public Power Supply System'.O.
NRR/DE/SGEB 25 NRR/DHFS/HFEBOO           1              NRR/OHFS/L'QB                                              1 NRR/DHFS/PSRB             1                                  32'RR/DL/SSPB 1
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  
NRR/DSI/AEB '26            1 ~            NRR/DSI/ASB                                                1
                                                                            '4 NRR/DSI/CPB 10             1              NRR/DSI/CSB 09                                            1 NRR/DS I/ICSB 16           1              NRR/DSI/METB 12'                                          1 NRR/DSI/PS   B   19.     1                            AB                                          1
                                                                    'FIL 22'G
:.NRA/DS I/RSB     23       1                                                                              1 RGNS                       3>>            R      AHI/MI8                                            1    0
  ~
EXTERNALe ACRS                     01                     BNL(AMDTS ONLY)                                           .1
                    ,OMB/DSS (AMDTS)                           FEMA REP       DIV '39 LPDR              03"                    NRC PDR                                                   1    1 02'TIS NSXC              05>>                                                                                1     1 TOTAL NUMBER OF ~COPIES REQUIRED: L'TTR                   53   ENCL'
 
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                                                                                                                                                                              ,) L f!.4
 
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:==
==Subject:==
WNP-2 Design Reverification Program  
WNP-2 Design Reverification Program


==References:==
==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
LR 3 OOOC 880 xoG .OA I
  ~,
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Al N
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  &#xb9;l.
238X241AD, Rev. 9, "Feedwater      Control System - Master Parts  List".
A-13


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.
DL807E160TC, Rev. 0, June 15, 1978, "Device List    and System Elementary Diagram  Feedwater Control System".
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.
22A3041, Rev. 1, March 14, 1972,     "Essential Components", Design Specification .
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.
239X241AD, Rev. 9, ."Feedwater Control System (Turbine)" Master Parts List.
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.
PL368X482, Rev. 7, "Reactor Feedwater      Document List".
+e G.C.Sorensen, Acting Manager Nuclear Safety and Regulatory Programs GDB:awh Distribution attached oo<83l0070342 830927 PDR ADOCK 05000397 A LR 3 OOOC 880 xoG.OA~, h I 1 g\Al N DISTRIBUTION USNRC INTERNAL DISTRIBUTION T.NOVAK A.SCHWENCER R.AULUCK J.MARTIN (Region V)A.TOTH TAA 5 CONSULTANTS RV LANEY LH RODDIS, JR HE SflEETS FB JEWETT, JR S.LEVY CQ MILLER JR HONEKAMP CS CARLISLE LT HARROLD ,BA HOLMBERG J YATABE DL WHITCOMB DM BOSI JD MARTIN PL POWELL WW WADDEL SI STEVENS Docket fi,le kt/f il e PL 2/1 b GCS/lb GDB/lb sf (2)WNP-2 files 982A 982A 994E 410 420 410 927M: 956B 400 750 956B.99.4E 956B 340 387 917Y FINDINGS REVIEW COMMITTEE RJ BARBEE 927M EXTERNAL DISTRIBUTION JG TELLEFSON 901D CH McGILTON 956B LC OAKES 823 AJ FORREST-BSR-RO WG CONN-BAR-RO F.McCLEAN-GE AG HOSLER.JH BAKER NS PORTER 956B 956B 387 WS CHINN, BPA JR LEWIS, BPA NS REYNOLDS-D&L 399 399 0
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.
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~
22A3059, Rev. 1, November 6, 1972,     "Definition of Piping Interfaces
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
- 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


DIX 1 LIST OF POT.FINDING REPORTS (Page 1 of 13)PFR No.Classification**
22A3067, Rev. 2, October 12, 1972; "Mechanical Equipment Separation", Paragraph 4.5.
Review Area*Descri tion HPCS-1 HPCS-2 HPCS-3 HPCS-4 HPCS-5 HPCS-6 HPCS-7 HPCS-8 HPCS-9 HPCS-10 HPCS-11 HPCS-12 3.1 3.2.3.1.D 3.2.3.6.A 3.2.3.10.A 3.2.3.10.B 3.1 3.2.3.2 3.2.3.5.B 3.1 3.1 3.2.3.6.B 3.1 The MR criteria document does not include requirements for all design input areas identified on the requirements reverification checklist.
22A2271AS, Rev. 1, November 30, 1978,   "Preoperational Test Program",
The equipment piece number for diesel engine cooling water heat exchanger is not consistent on all drawings.The diesel air start system is not totally redundant as described on the Flow Diagram.Current calculation revisions were not used as the basis for subsequent calculations.
Pre-op Test Specifications.
MR and alternate calculations do not agree on the diesel exhaust pressure drop.Cold working of instrument tubing.Detail 8 showing HPCS Instrumentation is missing from Flow Diagram.HPCS/RCIC condensate storage level instrumentation separation is questioned.
22A3838, Rev. 1, March 8, 1976, "Recommended   Prerequisites for Pre-Operational Testing". Preoperational Test Specification.
FSAR does not state the correct ASHE Code Classification for the HPCS diesel cooling water heat excnanger.
A-15
FSAR states that all fuel oil piping is ASME III whereas some is 831.1.Calculations that justify condensate storage level transfer setpoint not found.FSAR does not state the piping material requirements specified in the ECD.**F-Finding 0-Observation NV-Not Valid*Corresponds to Report Section Number LIST OF POTENTIAL FINDING REPORTS (Page 2 of 13)o.assi ica ion eview Area*Oescri tion HPCS-13 HPCS-14 X HPCS-15 X HPCS-16 HPCS-17 3.1 3.1 3.2.3.6.A 3.2.3.7.A 3.2.4.3 Different sections of the engineering criteria document do not agree on piping corrosion allowance.
FSAR does not agree with ASNE piping code effective date specified in tne ECD.85R calculation on emergency water volume for HPCS pump suction is inconsistent with other calculations and the design events.HPCS relief valve design does not incorporate GE design specifications for double flange gaskets.The DSA diesel engine exhaust system line size does not correspond to manufacturers recommendations.
HPCS-18 HPCS-19 HPCS-20 HPCS-21 3.2.4.3 3.2.4.3 3.1 3.2.6.4.8 The diesel fuel oil system does not meet NFPA Std.37 requirements.
No air box drain collection tank is provided for the HPCS diesel.Design requirement documents and FSAR values for vital piping damping coefficient do not agree.There is clearance between the attached parts of two snubbers where gaps are not allowed.HPCS-22 HPCS-23 HPCS-'24 HPCS-25 3.2.5.3 3.2.5.1.D 3.2.5.1.0 3.2.5.1.D No design calculations traceable to the HPCS pump support anchor bolts were found.Design procedures covering aspects of tne Instrumentation Installation Contractor design process were considered inadequate.
Improper stress intensification factors were used in the analysis of PI Line X-73a.Evaluation of local stresses caused by weld attachments for Pl Line X-73a was considered to be inadequate.
**.Finding O'-Ooservation
*Corresponds to Report Se.on Number LIST OF POTE AL FINDING REPORTS (P of 13)P R No.ass>>cat>on
" evsew Area+F NV Descr>tron HPCS-26 HPCS-27 3.2.5.1.0 3.2.5.1.A No faulted conditions stress evaluation was found for PI Line X-73a.Loads used in the design of pipe supports for N2UO-2 piping system are not current.HPCS-28 HPCS-29 HPCS-30 HPCS-31 HPCS~32 HPCS-33 HPCS-34 HPCS-35 HPCS-36 3.2.6.4.A 3.2.4.5 3.2.3.4 3.2.3.5.8 3.2.4.6 3.2.4.4 3.2.4.5 3.2.4.6 3.2.6.2 Potential restraint io thermal expansion of PI Line X-73a was identified.
HPCS-FE-7 is installed with pressure taps and attached instruments located at the top rather tnan horizontally as suggested by good engineering practice.There are ambiguities in the piping code specifications for the CST to HPCS pump suction piping.Discrepancies in separation criteria were noted in the BRI documentation.
GE specifications for instrument setpoint, accuracy, drift and range are not consistent.
Instrument tubing match line elevations disagree between two isometric drawings.There is a discrepancy between the flange bore and pipe ID for HPCS-FE-7.
The nameplate and ranges specified in the instrument data sheet for OPIS-9 do not agree.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 HPCS-39 3.2.6.2.B 3.2.3.7.B HPCS-LS-2A was tagged with a tag identifying the level switch as HPCS-LS-28.
The instrument line for the suppression pool level switch is not orificed to provide containment isolation per RG 1.11.**F-Finding 0-Observation NV-Not Valid*Corresponds to Report Section Number LIST OF POTENTIAL FINDING REPORTS (Page 4 of 13)No~assi 1 cation evlew Area*Oescr i tion HPCS-40 HPCS-41 HPCS-42 HPCS-43 3.2.6.2 3.2.3.2 3.2.3.4 3.2.4.3 The specified and nameplate ranges of HPCS-PS-12 do not agree.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.The seismic classification of HPCS suction piping from the CST is incorrect.
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 HPCS-46 HPCS-47 HPCS-48 HPCS-49 HPCS-50 HPCS-51 HPCS-52 HPCS-53 3.2.3.4 3.2.4.7 3.2.4.7 3.2.3.3 3.2.4.7 3.2.3.3 3.2.3.6.C 3.2.3.3'.2.3.3 There are ambiguities in FSAR Table"3.2-1 on code class groups for the HPCS system.The adjustable range for Breaker 4-41 short circuit tripping does not meet the GE specification.
The relay element connected to Breaker 4-41 does not permit proper coordination.
As-built data was not used in BRI voltage drop calculation 2.06.03 for TR4-41.Ground fault alarm relays on Bus SN-4 will not function reliably.The effect of simultaneous starting of 480V and 4KV motors was not considered in BRI Voltage Drop Calculation 2.06.03, Rev.5.The present design does not include the required degraded voltage protection and auto return to standby.The vendor print file for TR 4-41 contains two contradicting drawings.No fault duty calculation was provided for NC-4A.**.rin4ing 0-Observation
*Corresponds to Report Se ,on Number LIST OF POTE AL FINDING REPORTS (P of 13)o.ass>>ca son evsew rea Description HPCS-54 HPCS-55 HPCS-56 HPCS-57 HPCS-58 HPCS-59 HPCS-60 3.2.6.3 3.2.3.7.A 3.2.5.2.B 3.2.5.2.A 3.2.5.1.8 3.2.5.1.8 3.2.5.1.B One of the bolts is missing from the HPCS pump grounding lug connection.
There is an equipment piece number discrepancy netween FSAR Table 6.2-16 and BKR Orawing t620 for several valves.Local pipe stress from a welded attachment lug for pipe support 910-N was not calculated adequately.
Miscellaneous errors exist in the design calculation for pipe support HPCS-66.There is an error in the piping design guide.Calculation 8.14.64A does not correctly calculate the functional capability stress of tne piping system.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 HPCS-63 3.2.5.1.B 3.2.5.1.8 The load data source for chugging, SRY and LOCA jet direct loads are not referenced in BRI calculation 8.14.64A.Tnere are documentation problems with seismic analysis input calculations for BRI Calculation 8.14.64A.HPCS-64-HPCS-65 HPCS-66 3.2.5.1.B 3.2.5.1.B 3.2.5.1.B Improper revisions were made to support load tables in BRI Calculation 8.14.64A.Tne thermal displacements at branch connections were not correctly sumnarized.
Support design loads were incorrectly reported for HPCS-910N in BRI Calculation 8.14.64A.**F-Finding 0-Observation NY-Not Yalid*Corresponds to Report Section Number 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 HPCS-70 HP CS-71 3.2.5.1.B 3.2.5.1.B 3.2.5.1.B An incorrect mass was used in the computer model of valve HPCS-V-15.
1'he physical properties of HPCS-V-15 used in the computer model did not come from tne referenced drawings.Various errors were made in the thermal expansion analysis in BRI Calculation 8.14.64A.HPCS-72 HPCS-73 HPCS-74 HPCS-75 HPCS-76 HPCS-77 3.2.5.1.B N.A.3.2.5.1.A N.A.N.A.X 3.2.5.1.A Emergency condition temperatures were not considered in the thermal expansion analysis in BRI Calculation 8.14.64A.This number was not used.Valve nozzle end loads and accelerations are not evaluated per requirements of the ECO.This number was not used.This number was not used.The SSE response spectra for mass point 40 (BRI Calculation 8.14.82)is not included in referenced document.HPCS-78 HPCS-79 HPCS-80 3.2.5.1.A 3.2.5.1.A 3.2.5.1.A The stress index, C2, used for the 3/4" elbowlet is lower than that required by ASrK Section III.An additional weight of 1047 pounds was added to the 12" HPCS-V-5.HPCS-V-76 was modeled using a weight 400 pounds less than the drawings indicate.**~inding 0-Observation Nv-Not Valid*Corresponds to Report S n Number LIST OF POTE AL FINDING REPORTS (P of 13)No.ass>cat>on evsew Area Descri tion HPCS-81 HPCS-82 HPCS-83 3.2.5.1.A 3.2.5.2.D 3.2.5.1.C Incorrect scales were used for ADLPIPE response spectra input.Thermal loads used for design of HPCS-52 do not match those in the applicable pipe calculation.
Elbow dimensions used in the analysis of small bore line DE-1738-1 are in error.RFW-1 RFM-2 RFW-3 RFW-4 RFM-5 RFW-6 RFW-7 RFW-8 RFM-9 RFM-10 RFW-11**F-Finding 0-Observation NV-Not Valid 3.4.6.3 3.4.4.3.8 3.4.6.3 3.4.4.3.D 3.4.4.1 3.4.4.1.8 3.4.4.2.A 3.4.6.3 3.4.4.3.C 3.4.4.3.0 3.4.4.3.0 RFW-TE-41A had been improperly terminated in the field.RFW line"A" temperature element installed orientation does not correspond to orientation shown on pipe isometric.
The signal.cable for RFW-TE-41A was incorrectly labeled.The wrong type of flow element was selected for RFW-FE-15.
RFW-V-32A was not specified to oe testable with low pressure air as required by 10CFR50 Appendix J.The feedwater heater relief valve capacity is not sufficient to provide relief for all hypothetical events.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.The air operator extension shaft of RFW-V-32A interferes with RWCU inlet line to header"A".Inconsistencies are noted on the elementary and other electrical drawings for RFW-V-32A.
Upstream straight piping section length for RFM-FE-1A is insconsistent with ECO requirements.
Downstream straight piping length requirements for RFW-FE-1A is inconsistent with the ECO.*Corresponds to Report Section Number 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 RFM-15 RFM-16 RFM-17 RFW-18 3.4.4.3 3.4.6.3 3.4.4.3.D 3.4.6.3 3.4.3.4 System flushing and protection screening for RFM-FE-lA is not installed.
The RFM-FE-1A pressure tap configuration and connections are not installed per manufacturers recollmendations.
RFW-FE-1A calibration curve anomolies.
RFM-DPT-803A signal loop wiring and instrument rack tubing runs are not labeled in accordance with contractor requirements.
Documentation inconsistencies were found in the review of RFW-V-32A containment isolation requirements.
RFM-19 RFM-20 3.4.3.4 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.
The BRI elementary diagram does not show the required interlock between V-1128 and DPS-4.RFW-21 RFM-22 3.4.4.1.8 3.4.4.1.8 Control valve cavitation problems exist witn some valves.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 RHR-2 3.3.4.3.C 3.3.3.1.C All required cable types were not listed in Class lE list.P The BRI wiring design for several RHR valves did not follow GE requirements.
~*finding 0-Observation NY-Not Ya3id*Corresponds to Report S.~n Number LIST OF POTE L FINDING REPORTS (P of 13)-o.ass>>cat>on**
Review Area*Descri tion RHR-3 RHR-4 RHR-5 RHR-6 RHR-7 RHR-8 RHR-9 RHR-10 3.3.3.5 3.1 3.1 3.3.3.4 3.3.3.4 3.3.3.3 3.3.3.1.C 3.3.3.1.D~Containment isolation valve limit switches prematurely indicate valve closure.FSAR incorrectly states that seismic reevaluation is supplemented by NUREG-0800.
I No design requirement was found to match FSAR comnitment for vertical cable tray run fire breaks.RHR-FC-64B was not included in the remote shutdown system design as required by specification 22A3085.Remote shutdown system design specification 22A3085, Para.4.1.1 is not met in that a new common point was created.BRI drawing E503-8, Rev.23 shows RHR-P-3 in Division B instead of Division 2.The GE documentation for RHR-V-38 tnrottling are contradictory.
The second level undervoltage relays will cause bypass of the 115 kV source and will lockout the shed ESF loads.RHR-11 RHR-12 RHR-13 RHR-14 RHR-15 RHR-16 3.3.3.1.D 3.3.3.1.D 3'-4.1.A 3.3.4.2.A 3.3.3.1.D 3.3.4.3.C Feeder loads for HC-7BB and 7BA are missing from the NC-78 load calculation.
Feeder circuit breaker for MC-7BB may be set too low.There is a discrepancy in the RHR-FCV-64 operating time specifications.
RHR-F IS-108 is overranged.
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.
The voltage drop from E-SL-81 to NC-BBB is larger than the 3X recomnended by BRI criteria.**F-Finding 0-Observation NV-Not Valid*Corresponds to Report Section Number LIST OF POTENTIAL FINOING REPORTS (Page 10 of 13)P R No.C ass>cat>on" ev>ew Area*Oescrs t>on RHR-17 RHR-18 RHR-19 RHR-20 RHR-21 RHR-22 RHR-23 RHR-24 RHR-25 RHR-26 X 3.3.4.2.A 3.3.4.2.8 3.3.4.B 3.3.4.1.C 3.3.4.1.C 3.3.4.3.C 3.3.5.2.B 3.3.5.2.B 3.3.5.2.8 3.3.5.2.A RHR-FT-1 impulse lines are not routed as shown with the flow diagram.The documentation (GE)for.RHR-FI-5 does not agree with the installed instrument indicating scales.RHR-MO-24B and 64B were ordered specifying the wrong environmental class.A cavitation check was not included in BRI Calculation 5.17.13 for RHR-R0-18.
A cavitation check was not included in BRI Calculation 5.17.26 for RHR-R0-3B.
Cable 2MBBA-20 is not sized for derated conditions.
Heat exchanger drawings do not match the calculations.
Heat exchanger installation does not reflect the calculation and installation specification requirements.
Oue to increased loadings, the anchor bolt analysis is incomplete.
The original calculations were not updated or referenced to supporting calcu 1 ations.RHR-27 RHR-28~RHR-29 RHR-30 RHR-31 3.3.5.2.A 3.3.5.2.A 3.3.5.2.A 3.3.4.3.A 3.3.4.3.B A buckling analysis was not performed as required by design criteria.The anchor bolt analysis for the upper lateral supports is incomplete.
Assumed future (design)hanger loads must be verified against the actual hanger loads.Motor starters and TR-8-81 are subjected to over voltages (SM-8 side of the 480 V system).Oocumentation discrepancies for the fuse and overload heater sizes for three valves were noted.RH**inding 0-Observation NV Nnt Vol irl N.A.To be inclu'n Pipe and Support Addendum.*Corresponds to Report Se~n Number LIST OF POTE AL FINDING REPORTS (P of 13)o.ass~scat>on evsew Area*Descri tion RHR-33 RHR-34 RHR-35 3.3.6 N.A.3.3.4.3.A 3.5.5.2 3.5.5.2 Lugs on the heat exchanger are not shimmed per the GE specifications.
This number was not used.Fuse/circuit breaker coordination information is missing.HPCS-HO-4 is not listed in QID file identified on the Class lE list.The QID file referenced for HPCS-RO-4 did not contain the required design certification documentation.
EQ-3 EQ-4 EQ-5 EQ-6 EQ-7 EQ-8 EQ-9 EQ-10 3.5.5.1 3.5.5.2 N.A.N.A.3.5.5.6 N.A.3.5.5.2 3.5.5.6 QID file for HPCS-42-4A7C does not include required qualification data.There is no in-situ pull/deflection operability test record-for valve RHR-FCV-64B in the QID file.Number not used.Number not used.Confirmation is required for existence of low pressure isolation alarm and procedure to isolate auxi liary steam system.Number not used.The dynamic qualification levels identified in the QID fi le for HPCS-LS-2A are less than the required inputs.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 0-Observation NV-Not Valid,*Corresponds to Report Section Number 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 EQ-13 3.5.5.6 3.5.5.6 Subcompartment pressure analysis does not consider a door in Room R408.Non-conservative isolation valve closure characteristics assumed in RCIC line break analysis.EQ-14 EQ-15 3.5.5.6 3.5.5.6 A non-conservative time delay was'assumed for generating RWCU oreak isolation signal.HELB calculations for EQ profiles did not specifically address single failure criteria.EQ-16 EQ-17 FP-1 FP-2 FP-3 FP-4 WL-1 WL-2 WL-3 3.5.5.6 3.5.5.1 3.5.3.3 3.5.3 3.5.3 3.5.3.2 3.5.6.2 3.5.6.2 3.5.6.1 Normal HVAC ductwork may not retain its integrity to support post-HELB cooldown.There are discrepancies between the model numbers on the Class lE/SRH lists and the installed components.
Several dedicated cables that require protection were not listed in the E-948 cable tray node su+varies.
Thermolag fire barrier is applied to an empty tray that is not required to be lagged.Cable spreading room pentration curbs shown on N-576 are not shown on 5-906.Note 7 on N521 SH2 should not apply to RHR-V-40.Hain steam tunnel north wall load combinations are not verified.FSAR criteria incorrectly applied to the main steam tunnel north wall deflection calculation.
Attachment loads were not considered in BRI design calculation for the main steam tunnel north wall.**Finding 0-Observation a>ll=.-Hnt llil ih*Corresponds to Report S on Number LIST OF POT L FINDING REPORTS (P 3 of 13)~ass 1 1ca ion eview rea Descri tion WL-4 WL-5 PB-1 PB-2 PB-3 3.5.6.2 3.5.6.2 3.5.4.1.B 3.5.4.1.C 3.5.4.1.0 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.
Jet impingement load factors were not properly considered in calculating the dynamic loading of the main steam tunnel north wall.Haterial allowables used for approval of loads and/or stresses for PWS-2-1 are not traceable.
Field walkdown of HPCS pipe break location identified more potential targets than those cited in the B&R calculation.
Post-accident damage sequence differs from that postulated in the original B&R calculation.
PB-4 PB-5 PB-6 PB-7 3.5.4.1.B N.A.3.5.4.2.B 3.5.4.1.E As-built strut size is smaller than the size specified in BRI calculation 8.01.52.Number not used.Field walkdown of RWCU pipe break location identified more potential targets than cited in the BRI calculation.
Process deficiencies in potential target resolution were noted.**F-Finding 0-Observation NV-Not Calid*Corresponds to Report Section Number 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-105K Rated.22A2887, Nuclear Boiler System.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, for Safeguards Systems".4.3.3.1.1, 4.3.3.1.2, SHT SHT 17, Table 3.1975,"Electrical Equipment; Separation P 1 ant Requirements.
P ar agraphs: 4.3.3.1, 10 Table IV, 4.4.1, 4.4.3, 4.4.3.4, 4.4.4, 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 ices Class i f 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 Specification-Local Instrument Panels".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&#xb9;l.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:
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".
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-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&#xb9;2-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
TM-330, Rev. N/A, June 28, 1972, "Medium Voltage Switchgear Basis".
&#xb9;72, Rev.0, September 25, 1975,"Feedwater System".EM-79-006, Rev.N/A, January 2, 1979,"MCC Master List".A-16 General Electric Documents:
TM-427, Rev. 1, February 21, 1973,             "Control and Secondary   Wiring Internal to Switchgear, Panels,           and Similar Enclosures".
21A8658, Rev.1, May 17, 1971,"General Requirements for Motor Operated Valve Actuators-Purchase Specification".
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".
21A9222, Rev.2, January ll, 1974,"Electric Motors, General-Purch ase Speci f i cation".21A92220M, Rev.5, December 14, 1979,"Motors, Vertical (RHR)-Purchase Specification".
TM-526, Rev. A, June 28, 1973, System             Description, Residual Heat Removal   System".
22A1483, Rev.4, February 19, 1974,"HPCS System-Design Specification".
TM-671, Rev. N/A,           July 5, 1974, "Contract &#xb9;2 -   PVC Cables".
22A1483AU, Rev.4, August 13, 1979,"HPCS System-Data Sheet".22A2710A, Rev.7, September 9, 1974,"Standby AC Power-BWR Requirements".
TM-990, Rev. 1, March           ll, 1977, "MCC - PCU Insulated Control Wiring".
22A2711, Rev.3, January 9, 1974,"Plant OC Power-Design Speci f ication".22A2817, Rev.3, November 27, 1973,"RHR System-Design Specification".
TM-1129, Rev. N/A, August           ll, 1978, "Class lE Motor Operated Valves".
22A2817AY, Rev.2, October 31, 1974,"RHR System-Data Sheet".22A3008, Rev.5, April 8, 1977,"BWR Equipment Environmental Interface Data-Design Specification".
System Description &#xb9;72, Rev. 0, September 25, 1975, "Feedwater System".
22A3038, Rev.6, February 5, 1979,"Motor List, Electric-Design Specification".
EM-79-006, Rev. N/A, January 2, 1979,             "MCC Master   List".
A-17 22A3061, Rev.0, September 3, 1971,"Electrical Codes and Standards-Design Specification".
A-16
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".
General Electric Documents:
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".
21A8658, Rev. 1, May 17, 1971, "General Requirements           for Motor Operated Valve Actuators   - Purchase Specification".
A.6.3 Contracts Contract&#xb9;35, Sect.15A,"Miscellaneous Pumps and Motors".Contract&#xb9;41A, Sect.15A,"Nuclear Valves".Contract&#xb9;41B, Sect.15A,"Nuclear Valves".Contract&#xb9;47A, Sect.16A,"Medium Voltage Switchgear".
21A9222, Rev. 2, January   ll, 1974, "Electric Motors, General-f Purch ase Speci i cation".
Contract&#xb9;49, Sect.16A,"Motor Control Centers".Contract&#xb9;62A, Sect.16A,"Electrical Cable".A-18 Contract&#xb9;62B, Sect.16A,"Electrical Cable".Contract&#xb9;62C, Sect.TP,"Electrical Cable".Contract 218, Sect.16A,"Electrical Installation".
21A92220M, Rev. 5, December 14, 1979, "Motors, Vertical (RHR)-
A-19 A.7 En ineerin Mechanics AJ.1~5 BR I Documents:
Purchase Specification".
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
22A1483, Rev. 4, February 19, 1974, "HPCS System         -   Design Specification".
-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:
22A1483AU, Rev. 4, August 13, 1979, "HPCS System   -   Data Sheet".
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
22A2710A, Rev. 7, September   9, 1974, "Standby   AC   Power   - BWR Requirements".
-Protection Against Pipe Breaks Outside Con tainment.Report WPPSS-74-2-Rl
22A2711, Rev. 3, January 9, 1974,   "Plant OC Power     - Design f
-Protection Against Pipe Breaks Inside Con tainment.A.7.3 Contr act S ecifications C215-Specification 2808-&#xb9;215,"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-&#xb9;220,"Instrumentation Installation", Contract No.220, Section 15E,"Piping and Tubing Supports".
Speci ication".
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"..
22A2817, Rev. 3, November 27, 1973, "RHR System       -   Design Specification".
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
22A2817AY, Rev. 2, October 31, 1974, "RHR System       -   Data Sheet".
-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
22A3008, Rev. 5,   April 8, 1977, "BWR Equipment Environmental Interface Data   - Design Specification".
-Mechanical Equipment Separation", MPL8 A62-4350, dated August 21, 1975.21A1740, Rev.3,"Purchase Specification
22A3038, Rev. 6, February 5, 1979, "Motor     List, Electric -       Design Specification".
-Valve, Gate", MPL8 E22-F004, dated 1/13/72.21A1884, Rev.2,"Purchase Specification
A-17
-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.
22A3061, Rev. 0, September     3, 1971, "Electrical Codes and Standards   - Design Specification".
21A9243DE, Rev.2,"Purchase Specification Data Sheet-High)Pressure Core Spr ay Pump", NPL&#xb9;E22-C001, dated October 29, 1973.a k 21AI880, Rev.1,"Purchase Specification
22A5267, Rev. 1, May 2, 1979,       "Regulatory Requirements   and Industrial Standards - Design Bases".
-Valve Data Specification, Valve Gate", NPL&#xb9;E22-F012, Dated April 18, 1972.21A1736, Rev.3,"Purchase Specification
22A7416, Rev. 0, February 19, 1981,       "Electrical Equipment, Separations   for Safeguards Systems - Plant Requirement".
-Valve Data Specification
22A2907, Rev. 3, March 28, 1974, "Feedwater Control System         - Design Specification".
-Valve, Gate", NPL&#xb9;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
22A2907AB, Rev. 1, August 16, 1971, "Feedwater       Control System - Data Sheet".
-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:
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".
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" NPL&#xb9;E22-1010, dated July 30, 1974.73lf931 Rev.7,"P&ID-HPCS System" MPL&#xb9;E22-1010, dated May 22, 1974.731E932AD Rev.3,"Process Diagram-HPCS System", MPL&#xb9;E22-1020, dated October 22, 1978.B&R Dr awings: M520 Rev.33 N711 Rev.22 N712 Rev.29 N713 Rev.29 M714 Rev.25 N715 Rev.27 M716 Rev.27 N718 Rev.33 S798 Rev.31 S796 Rev.14 S795 Rev.41 N732, Rev.18 M626, Rev.5 SN197, Rev.D SM193,.Rev.E SM191, Rev.E SM183, Rev.E SM136, Rev.D SM135 Rev.D N567 Rev.6 M200, Sht.132 Rev.5 M200, Sht.100 Rev.7A B-4 N744 Rev.7 N527 Rev.37 M569 M200, Sht.101 Rev.9 M200, Sht.2 Rev.5 PDM Drawings: E37 Rev.A2 E61, Rev.B CB&I Drawings: 72-4396-2 Rev.6, 72-2647-123 Rev.7 72-4396-lA Rev.7 72-2647-1 Rev.8 Zurn Drawings: I-80120-A (BE R Transmi ttal 213B-12318)
A.6.3 Contracts Contract &#xb9;35, Sect. 15A, "Miscellaneous       Pumps and Motors".
" 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.
Contract &#xb9;41A, Sect. 15A, "Nuclear Valves".
4 HPCS-630-5.6 HPCS-630-13.19 HPCS-630-26.
Contract &#xb9;41B, Sect. 15A, "Nuclear Valves".
28 COND-351-1.9 HPCS-633-1.
Contract &#xb9;47A, Sect.     16A, "Medium Voltage Switchgear".
2 HPCS-1458-3 HPCS-1459-2 HPCS-2569-1 HPCS-2570-1 D-220-X-78 HPCS-629-5.
Contract &#xb9;49, Sect. 16A, "Motor Control Centers".
7 HP CS-630-7.10 HPCS-630-20.
Contract &#xb9;62A, Sect. 16A, "Electrical Cable".
23 HPCS-630-29.
A-18
30 COND-351-10.
 
15 H PCS-1458-1 HPCS-1458-4 HPCS-1460-1 HPCS-1644-1 HPCS-1958-1 HPCS-630-1.4 PCS-630-11
Contract &#xb9;62B, Sect. 16A, "Electrical Cable".
.12 HPCS-630-24.
Contract &#xb9;62C, Sect. TP, "Electrical Cable".
25 HPCS-630-31.
Contract 218, Sect. 16A, "Electrical Installation".
33 HPCS-632-1.3 HPCS-1458-2 HPCS-1459-1 HP CS-1461-1 HPCS-2568-1 HP CS-2571-1 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  
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-&#xb9;215,           "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-&#xb9;220, "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", NPL&#xb9; E22-C001,   dated October 29, 1973.
a )
k 21AI880, Rev. 1, "Purchase Specification - Valve Data Specification, Valve Gate", NPL&#xb9; E22-F012, Dated April 18, 1972.
21A1736, Rev. 3, "Purchase Specification - Valve Data         Specification
        - Valve, Gate", NPL&#xb9; 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" NPL&#xb9; E22-1010, dated July 30,   1974.
73lf931 Rev. 7, "P&ID   - HPCS System" MPL&#xb9; E22-1010, dated May 22, 1974.
731E932AD Rev. 3, "Process   Diagram-HPCS System",   MPL&#xb9; 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 &#xb9;2, "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 &#xb9;13  -  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.2 Mechanical Diesel Disci line References 22A1483, Rev.4, High Pressure Core Spray Systems Design Specifications.
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.
22A1483AU, Rev.4, HPCS Design Specification Data Sheet.21A1848AB, Engine Generator for HPCS Purchase Specification Data Sheet.21A1848, Engine Generator for HPCS Purchase Specification
ASME B5PV    Code, Section  III, 1971  Edition, including  Addenda through Winter, 1973.
.21A1776, Rev.1, HPCS Diesel Service Water Pump Purchase Specification.
BER  Piping Design Guide, Rev. 3.
21A1776AD, Rev.1, HPCS Diesel Service Water Pump Purchase Specification Data Sheet.A990 GEAPPD, Thermxchanger Exchanger Specification Sheet.Contract 215, Material Specifications.
WNP-2 FSAR.
BER Engineering Criteria Document.B.2.2 Cal culations BIWR Nuclear Calculations:
U.S. NRC  Standard Review Plan 3.6.2.
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:
G.E. Design      Specification  22A2887, Rev. 6,   for the Nuclear Boiler System.
TM-0558 TM-0586 TM-0775 TN-0746 TM-0608 TM-1053 1M-1066 TN-0817 TM-0443 (Rev.A)O.G.Synchro-Check Relays Emergency D.G.Operation Diesel Generator Loading Interlocking for Diesel Generator Diesel Generator Cooling Water System Standby D.G.Light Load Operation Gas Disper.8 Met Analysis System Description, D.G.Systems 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 OE-789-1.3 OE-1738-1 OE-2836-1 OSA-4275-1 DS A-4396-1 DSA-4396-2 DSA-2536-1 Rev.7 Rev.5 Rev.7 Rev.5 Rev.5 Rev.4 Rev.5 Rev.1 1/25/83 10/12/82 12/2/82 12/1/82 12/10/82 10/12/82 9-21-82 11/19/82 8-9 Isometric Drawings (Contd.)OSA-2537-1 DSA-2537-2 DS A-253 7-3 DSA-2537-4 DSA-2537-5 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 Rev.4 Rev.5 Rev.5 Rev.5 Rev.4 8/6/82 7/24/82 8/2/82 11/29/82 8/19/82 MR Flow Diagram M-512 GE Piping Diagr ams for HPCS Diesel Engine GeneratorA990D08001 A990009001 HPCS Diesel Engine Generator Air Intake Piping Schematic HPCS Diesel Engine Generator Exhaust System Piping Schematic A990F03001 A990F04001 A990C06002 A990F06001 A990F 07001 A990F 02001 HPCS Diesel Engine Generator DLO Schematic Diagram HPCS Diesel Engine Generator Jacket Water System with Heat Exchanger Schematic Diag<am Fuel Oil Schematic Lister SR1A Diesel Engine HPCS Diesel Engine Generator F.O.Schematic Diagram HPCS Diesel Engine Generator Air Start System Schematic Diagram 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, Potential D.G.Turbocharger Problem Degradation of Fuel Oil Flow to the Emergency D.G.Emergency D.G.Lube Oil Cooler Failures 80-04 79-12 77-15 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 22A1483AU 22A2988 22A3008 22A3039 22A3067 22A3095 22A3 746 22A7416 Rev.4 Rev.4 Rev.6 Rev.5 Rev.1 Rev.3 Rev.0 Rev.1 Rev.0 Electrical Equipment Separ ation Design Specification, HPCS System HPCS System Data Sheet Electrical Separation (See 22A7416)BWR Equipment Environmental Interface Data Process Instrumentation Mechanical Equipment Separation Process Integrity of Piping and Equipment Local Instrument Panels Burns and Roe: Design Criteria Sections F and G B.3.2 Calculations Burns and Roe: 5.51.051 8.01.203 Target Determination Pipe Break Outside Containment 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 GEK 71337 July 1978 June 1978 High Pressure Core Spray System, 0&M 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 163C1043T C 731E931AD 731E950AD 234A9309TC 807E172TC Rev.2 Rev.1 Rev.7 Rev.Rev.3 Rev.19 HPCS Instrument Panel Arr angement HPCS Instrument Panel Piping Diagram HPCS P&ID HPCS FCD Instrument Data Sheets Elementary Diagrams HPCS System Burns and Roe 7E015 7E016 7E017 7E018," 7E019 7E020 Rev.2 Rev.2 Rev.1 Rev.1 Rev.2 Rev.2 Electrical Electrical Electrical Electrical Electrical Electrical Wiring Diagram HPCS-V-1 Wiring Diagram HPCS-V-4 Wiring Diagram HPCS-V-10 Wiring Diagr am HPCS-V-ll Wiring Diagram HPCS-V-12 Wiring Diagram HPCS-V-15 8-13 7E021 7E025 7E02 S 709 f 522 E 535-18A-18B E 536-2C-5B E537-I V-3A Rev.1 Rev.1 Rev.1 Rev.26 Rev.16 Rev.9 Rev.7 Rev.12 Rev.12 Rev.3 Rev.9 Electr i cal Electrical Electrical Structural Elem.Diag Connection Connection Connection Connection Connection Connection Cab.Wiring Diagram HPCS-V-23 Wiring Diagram Controls Sheet 1 Wiring Diagram Controls Sheet 1 , Reactor Building.Isolation Valves Wiring Diag.MC4A Wiring Diag.MC4A Wiring Diag.Term.Box and Misc.Wiring Diag.Term.Box and Misc.Wiring Diag.Term.Box and Misc.Wiring Diag.Control Room Term.-4B Rev.8-26A Rev.7 E539-2-14-21 E 540-4-6 N520 M527 M530 Rev.10 Rev.10 Rev.8 Rev.5 Rev.10 Rev.32 Rev.44 Rev.32 M543 Rev.33 M567 Rev.6 M568 Rev.22 M609 M619-6-19 Rev.10 Rev.4 Rev.4 M619-VI Rev.5 Connection Wiring Diag.Control Room Term.Cab.Connection Wiring Diag.Control Room Term.Cab.Connection Wiring Diag.Reactor IEC Connection Wiring Diag.Reactor ILC Connection Wiring Diag.Reactor IEC Connection Wiring Diag.Motor Op.Valves Connection Wiring Diag.Motor Op.Valves Flow Diagram, HPCS and LPCS Systems Flow Diagram, Condensate Supply System Flow Diagram, Nuclear Boiler Recirculation System Flow Diagram, Containment Cooling and Purging General Arrangement Reactor Bldg.El 422'and 441'eneral Arr angement Reactor Bldg.El 471'nd 501'nstrumentation Contract 220, General Notes Instrument Process AZ 0 to 180 0 0 Instrument Conn.Diag.H22-P009 Instrument Conn.Diag.H22-P024 N623 M624 M625 N626 N627 N628 N629 N734 Rev.8 Rev.8 Rev.11 Rev.7 Rev.13 Rev.17 Rev.7 Rev.24 Instrument Process Plan El 501'nstrument Process Plan El 512'nstrument Process Plan El 522'nstrument Process Plan El 541'nstrument Process Plan El 560'nstrumentation Process Penetration Schedule Instrumentation Process Partial Plans and Sections Miscellaneous Piping Plan and Sections at El 471')Johnson Controls B-220-007.0-H22P024 Rev.B-220-X-73
G.E. Document 0 22A1483, Rev. 4 (HPCS System Design         Specification).
-86A-86B Rev.Rev.Rev.0-220-X-73 Rev.,-86A Rev.-86B Rev.-87A Rev.-87B Rev.D-220-3500-250-CMS-LT-1&2, E-220-5500-RB-441 Rev.6-471 Rev.-501 Rev.-522 Rev.-548 Rev.-87A Rev.-87B Rev.0-220-007.0-H22P024 Rev.0-220-7.1-X-732-1 Rev.-lA Rev.-1B Rev.-1C Rev.Line Identification List (20 sheets)Line Identification List X-73 Line Identification List X-86A Line Identification List X-86B Line Identification List X-87A Line Identification List X-878 Tube Erection Isometrics (20 Sheets)Process Instrument Line X-73a Process Instrument Line X-73a Process Instrument Line X-73a Process Instrument Line X-73a Process Instrument Line X-73 Process Instrument Line X-86A Process Instrument Line X-86B Process Instrument Line X-87A Process Instrument Line X878 Rev.1, Local Instrument Installation Tube Routing React.Bldg.El.441'ube Routing React.Bldg.El.447'ube Routing React.Bldg.El.501'ube Routing React.Bldg.El.522'ube 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 w 3 Rev.3 Rev.2 Rev.3 Rev.3 Rev.3 Rev.3 RCIS-HPCS Switchgear Standpipe RCIS-HPCS Switchgear Standpi pe RCIS-HPCS Switchgear Standpipe RCIS-HPCS Switchgear Standpipe RCIS-HPCS Switchgear Standpipe RCIS-HPCS Switchgear Standpipe Dragon Valves, Inc.C-12583 Rev.E Excess Flow Check Valve~Daniel Industries C-2629 C-2630 ANS Orifice Flange, Upstream ANS Orifice Flange, Downstream B.3.6 Contract S ecifications 21A9376 21A9376A J 21A9417AB 2808-220 Rev.1 Rev.1 Rev.0 Flow Orifice Assembly Flow Orifice Assembly Data Sheet I.D.S.HPCS System 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".
ASME B5PV    Code Case N-122(1745)    03/01/76.
21A8658, Rev.1, May 17, 1971,"General Requirements for Motor.Operated Valve Actuators".
Westinghouse      Structural Analysis Program    PIPSAN.
21A9222, Rev.2, January ll, 1974,"Electrical Motors, General-Purchase Specification".
Contract 208 Specification, "Small-Diameter Piping        and Pipe Support Criteria" Section .
21A9222DI, Rev.3, July 18, 1980,"HPCS Pump Vertical Motor Data Sheet-Purch ase Speci f i cation".21A9300, Rev.3, July 25, 2978,"Switchgear Electrical Metal Enclosed for HPCS-Purchase Specification".
EhR Memo     EM-82-322, 6/28/82.
21A9300AO, Rev.3, August 17, 1974,"Metal Enclosed Electrical Switchgear
WRC  Bulletin    107, March 1979, Revision.
-Purchase Specification
ASME  Code Case N-318, Rev. 0.
-Data Sheet".21A9301AJ, Rev.3, November 26, 1974,"HPCS Motor Control Center-Purchase Specification
B-25
-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 f ication".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, Rev.0,"Cable Sizing-4.16 and 6.9 KV-Short Circuit Capacity".
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, E WD-72-0016 E 502-2 E 503-9 E 514-6 E 517-9 E 550 E 553-1 E 558-2 E 662-1 Standby Water Leg PP.HPCS-P-3 (E22-C003)
M.O.V.HPCS-V-4 (E22-F004)
Main One Line Diagram Aux.One Line Digaram Relay Settings 4.16 KV Switch-gear SH-4 4160 V SWGR, Elem.Diag.Cable Schedule-Power Class lE Electrical Equip List Turb.Gen.Bldg.Grounding Plans and Details Reactor Bldg.Grounding Plans and Details Sh.1 Rev.1, 05/17/82 Rev.2, Rev.19 Rev.16 Rev.6 08/31/82 01/19/83 12/18/82 12/18/82 Rev.18 Rev.26 Rev.4 Rev.4 02/02/83 03/17/83 03/28/83 04/12/82 Rev.11 10/07/82 B-20 E 680 E 785 Reactor Bldg.El 422'3" Power Conduit and Tray Plan Diesel Gen.Bldg.El.441'-0" Power and Tray Plans Rev.17 12/30/82 Rev.32 01/21/83 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&#xb9;2,"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&#xb9;13-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).
TPIPE User's Manual, Version G/C       4.3.
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.
G/C Engineering Handbook, User's Manual, Rev. 7.
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).
BER Project Engineering Directive, PEO-C0208-0689.
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
Johnson   Controls, Inc. Design Guide for     WNP-2.
&#xb9;00010, 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&#xb9;20.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&#xb9;1253 (SRV Displacements).
AISC Steel   Construction Manual, 7th Edition.
Tech.Memo&#xb9;1181, Rev.1 (SRV Response Spectra).Tech.Memo&#xb9;1257, Rev.2 (Seismic and Hydrodynamic Response Spectra).Tech.Memo&#xb9;1283 (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&#xb9;16,"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).
General   Electric Document 21A9243 DE, Rev. 1,     "Specification and Data Sheet - HPCS Pump".
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.
General   Electric "Operating Instructions for     HPCS Pump", 3/19/75.
POM, 0-101, Rev.K, Penetration X-31 Details.Anchor/Darling, 2621-3, Rev.B, Valve HPCS-V-16.
Buil.ding Code Requirements   for Reinforced Concrete,   ACI 318-71.
Anchor/Darling, 94-13473, Rev.A, Valve HPCS-V-15.
ANSI 831.1 Power   Piping Code,   1973 Edition, W73 Addenda.
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.
Contract 208 Specification, "Small-Diameter Piping       and Pipe Support Criteria" section.
B-29 BER Support Detail Drawing, HPCS-900N.
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 &#xb9;00010, 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 &#xb9;20.
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 &#xb9;1253 (SRV Displacements).
Tech. Memo &#xb9;1181, Rev. 1 (SRV Response   Spectra).
Tech. Memo &#xb9;1257, Rev. 2 (Seismic and Hydrodynamic Response     Spectra).
Tech. Memo &#xb9;1283 (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   &#xb9;16, "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.
BhR Support Detail Drawing, HPCS-901N.
IKR Support Detail Drawing, HPCS-52.B&R Project Engineering Directive, PED-C0208-0689.
IKR Support   Detail Drawing,   HPCS-52.
M200-SHT2-1., Rev.C (HPCS Isometric).
B&R Project Engineering Directive, PED-C0208-0689.
H200-SHT2-2, Rev.A (HPCS Supports Orientation).
M200-SHT2-1., Rev. C (HPCS Isometric).
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
H200-SHT2-2, Rev. A (HPCS Supports Orientation).
&#xb9;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.
S795 (X-6 Penetration Detail ).
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).
M601, Rev. 20 (Valve   List).
731E932AD, Rev.3 (HPCS Thermal Modes).761E716 (RPV Nozzle Allowable Loads).Bovee and Grail HPCS-630-31.33, Rev.9 (Construction Drawings).
HPCS-63, Rev. 3 (Support   Detail).
HPCS-630-29.30, Rev.8 (Construction Drawings).
HPCS-912N, Rev. 1 (Support Detail).
HPCS-630-26.28, Rev.9FO (Construction Drawings).
HPCS-911N,   (Support Detail).
Velan Dwg.PP2-2767-N-2, Rev.L (Valve HPCS-V-5).
HPCS-910N, Rev. 1 (Support Detail).
B-32 Velan Dwg.8P2-3311-N-16, Rev.F (Valve HPCS-V-51).
4 HPCS-918N, Rev. 2 (Support   Detail).
Anchor/Darling Dwg.82652-3, Rev.B (Valve HPCS-V-76).
HPCS-919N, Rev. 2 (Support   Detail).
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  
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   &#xb9; 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  &#xb9;BC-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 &#xb9;420-00-0015 (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 &#xb9;RFW-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 &#xb9;RFW-418-1.2, Rev. 11, "From Flow Meter to Reactor Vessel (Line 'A'), Date 7/15''75.
Jelco Drawing &#xb9;757-D-622, Rev. C,  "Tubing Arrangement IR-12", shows C34-N002A rack interconnections    and rack connections.
Jelco Dr awing &#xb9;757-E-675, Rev. 0, "Electrical Wiring Diagram, Instrument Rack IR-12", shows wiring.
Jelco Drawing &#xb9;757-E-538, Rev. 0, "Instrument Assembly IR-12", shows rack placement of C34-N002A.
Jelco Drawing &#xb9;757-E-535, Rev. 0, "Instrument Assembly IR-12", shows rack side views.
Circle  A.W. Products  Drawing &#xb9;757-E-532, Rev. D, "Instrument Assembly IR-64".
Bovee and  Grail Drawing &#xb9;RFW-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 &#xb9;RFW-418-3, "Reactor 1
FW  from Flow Meter to Reactor    Vessel  (Line "A"), Rev. 5.
Anchor Darling Valve Company Drawing &#xb9;3084-3, Rev. B, "24      in. - 900&#xb9; swing check valve, RFW-V-32A (B223-F032) ".
Jelco Controls Inc. Drawing &#xb9;757-E-703, Rev.      B, "Electrical Wiring Diagram IR-62".
0-27
 
Circle  A.W. Products Co. Drawing &#xb9;757-E-544,  Rev. C, "Instrument Assembly, IR-9".
Jelco Controls Drawing &#xb9;757-C-619, Rev. C,  "Tubing Arrangement; Instrument Rack IR-9".
Johnson  Controls Drawing &#xb9;D-220-072.0 - RFT-18/IR-18, Rev. I, Line Identification List".
Johnson  Controls Draw'ing &#xb9;D-200-245-TG-441, Rev. 0, "Tubing Routing (As-Built) ".
Jelco Controls Drawing &#xb9;757-E-506, Rev. 8, "Instrument Assembly, Rack 18".
Jelco Controls Drawing &#xb9;757-E-611, Rev. C, "Tubing Arrangement, Rack 18" Jelco Controls Drawing &#xb9;757-E-664, Rev. 8, "Electrical Wiring Diagram, Rack 18".
Circle  A.W. Products Drawing &#xb9;757-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


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.
D.5.4    En  ineerin  Mechanics References
21A3757AD, Rev.2, GE Purchase Specification Data Sheet for Shell Side Relief Valves on RHR Heat Exchangers.
                                                                              '0 D.5.4.1    Desi n  Re uirement References M400-3 Engineering    Criteria  Document Appendix 2 Pipe Support Design Guide.
21A8657, Rev.3, GE Purchase Specification, General Requirements for Valves.21A8658, Rev.1, GE Purchase Specification, General Requirements for Motor Operated Valve Actuators.
Technical    Memorandum 1271,   (/II Equipment Nozzle Allowable Loads 6/14/82.
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.
D.5 .4.2    Calculations 8.42.8000    Revision  1 Pipe Stress Code .
21A9347AF, Rev.1, GE Purchase Specification Data Sheet, General Requirements for Instrumentation and Electric Equipment.
8.16.2013 Hanger Design Calculation for RFW-24.
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.
8.16.4983 Hanger Design Calculation      for RFW-944N.
21A9425AB, Rev.1, GE Purchase Specification Data Sheet for RHR Heat Exchangers.
8.16.72.1 Hanger Design Calculation for      RFW-943N, RFW-21, RFW-17.
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
0-32
.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.
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.
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.
APCSB  9.5-1, Appendix  A, Guidelines for Fire Protection for Nuclear Power Plants    Docketed Prior to July 1,  1976.
22A3061, Rev.0, GE System Design Specification, Electrical Codes and Standards.
E.l.2 Calculations 2.06.04, Rev. 1, Radwaste Bldg. /Control Bldg. Feeder    and  Voltage Drop  Calculations.
22A3062, Rev.2, GE System Design Specification, Mechanical Codes.22A3067, Rev.3, GE System Design Specification for Mechanical Equi pment Separation.
2.06.05, Rev. 3, Reactor Bldg. Feeder    and Voltage Drop Calculations.
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.
2.07.01, Rev. 2, High Voltage Cable Sizing - Ampocities      and Conduits.
22A3730AB, Rev.0, GE System Design Data Sheet for RHR Heat Exchangers.
2.07.03, Rev. 1,. A.C. Motor Control Center    Bus and Cable  Sizing.
C-3 22A3746, Rev.1, GE System Design Specification for Local Instrument Panels.22A5233, Rev.0, GE Installation Specification for RHR Heat Exchangers.
E.1.2   Technical Memorandum TM  1227, Rev. 3, Fire Protection Study, 4/22/82.
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 Section Section Section Section Section D Electrical Engineering Criteria E Mechanical Engineering Criteria F Chemical and Nuclear Engineering Criteria G Instrumentation and Control Engineering Criteria H Technical Standards Applicability List 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.
TM  1272, Rev. 2, Thermo-lag  Fire Barriers for Electrical Cables, Cable Ampocity Derating, 10/22/82.
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
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.
.2.03.07, Rev.2, 480 V Switchgear Short Circuit Calculation.
22A3046, Rev. 1, Core Standby Cooling System Network Design Specifications, 7/14/77.
2.03.09, Rev.0, Motor Control Center Short Circuit Calculations.
22A2802, Rev. 2, GE Reactor System Heat Balance 105K Rated Power.
2.03.11, Rev.0, Fault Calculations for Paralleling DG1 and OG2.2.06.03, Rev.5, Main Oneline Voltage Drop Calculations.
BRI  Engineering Criteria  Document.
2.06.05, Rev.3, Reactor Building Feeder Voltage Drop Calculation.
E.2.2    Calculations 5.49.050, Rev. 1, Pipe Break Analysis, Inside Containment.
2.06.10, Rev.1, Service and Diesel Generator Building Feeder Voltage Drop Calculations.
5.49.051, Rev. 1, Target Determination, Pipe Breaks Inside Containment, 12/17/82.
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.
5.49.052, Rev. 1, Shutdown Analysis for Pipe Breaks Inside Containment.
5.17.20, Rev.0, Effectiveness Calculation for RHR Heat Exchanger.
5.51.050, Rev. 1, Pipe Break Analysis, Outside Containment 5.51.051, Rev. 1, Target Resolution     for Postulated Targets Outside Containment.
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 Inves ti 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,''Speed-Torque Characteristic of Centrifugal 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, Limi torque 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 f i cati ons Cont.No.Award Date Item 2808-10 2808-11 A 2808-41 A 2808-418 2808-42A 2808-215 1-14-72 2-18-72 12-3-73 12-3-73 5-13-74 5-13-74 Feedwater Heaters Reactor Feed Pumps Nuclear Valves Nuclear Valves Misc.Control Valves, Controllers and Acc.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.Limi torque 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)Contract Specification 2808-49, Motor Control Centers, Division 16, Section 16A.v)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 Codes and Standards NEMA 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&#xb9;BC-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, Sht.64B, Rev.4, Sht.64C, Rev.7, Sht.64D, Rev.4, Sht.12A, Rev.6, Sht.12B, Rev.4, Sht.12C, Rev.3, Sht.12E, Rev.2, Sht.12F, Rev.4, Connections""IR-64 Legend""Connection Diagram IR-64""IR-64 Arr angement""Connection Diagr am IR-64""Inst.Rack IR-12 Legend""Inst.Rack IR-12 Arrangement""Inst.Rack IR-12 Tubing Arrangement""Inst.Rack IR-12 Wiring""Inst.Rack IR-12 External Electrical Sht.12G, Rev.0,"Inst.Rack IR-12 External Electrical Connections" Sht.12D, Rev.5,"Inst.Rack IR-12 Tubing Arrangement Cont." M619, Sht.Sht.Sht.Sht.Sht.85, Rev.5,"Inst.Rack IR-18 Connection Diagram" 110, Rev.4,"IR-12 Instrument Connection Diagram" 112, Rev.6,"IR-12 Instrument Connection Diagram" 142, Rev.9,"IR-64 Reactor Building Inst.Rack" 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&#xb9;420-00-0015 (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&#xb9;RFW-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&#xb9;RFW-418-1.2, Rev.11,"From Flow Meter to Reactor Vessel (Line'A'), Date 7/15''75.Jelco Drawing&#xb9;757-D-622, Rev.C,"Tubing Arrangement IR-12", shows C34-N002A rack interconnections and rack connections.
Jelco Dr awing&#xb9;757-E-675, Rev.0,"Electrical Wiring Diagram, Instrument Rack IR-12", shows wiring.Jelco Drawing&#xb9;757-E-538, Rev.0,"Instrument Assembly IR-12", shows rack placement of C34-N002A.
Jelco Drawing&#xb9;757-E-535, Rev.0,"Instrument Assembly IR-12", shows rack side views.Circle A.W.Products Drawing&#xb9;757-E-532, Rev.D,"Instrument Assembly IR-64".Bovee and Grail Drawing&#xb9;RFW-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&#xb9;RFW-418-3,"Reactor FW from 1 Flow Meter to Reactor Vessel (Line"A"), Rev.5.Anchor Darling Valve Company Drawing&#xb9;3084-3, Rev.B,"24 in.-900&#xb9;swing check valve, RFW-V-32A (B223-F032)
".Jelco Controls Inc.Drawing&#xb9;757-E-703, Rev.B,"Electrical Wiring Diagram IR-62".0-27 Circle A.W.Products Co.Drawing&#xb9;757-E-544, Rev.C,"Instrument Assembly, IR-9".Jelco Controls Drawing&#xb9;757-C-619, Rev.C,"Tubing Arrangement; Instrument Rack IR-9".Johnson Controls Drawing&#xb9;D-220-072.0
-RFT-18/IR-18, Rev.I, Line Identification List".Johnson Controls Draw'ing&#xb9;D-200-245-TG-441, Rev.(As-Built)
".0,"Tubing Routing Jelco Controls Drawing&#xb9;757-E-506, Rev.8,"Instrument Assembly, Rack 18".Jelco Controls Drawing&#xb9;757-E-611, Rev.C,"Tubing Arrangement, Rack 18" Jelco Controls Drawing&#xb9;757-E-664, Rev.8,"Electrical Wiring Diagram, Rack 18".Circle A.W.Products Drawing&#xb9;757-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 D.5.4.1 Desi n Re uirement References
'0 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.
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.
1 8.01.51, Rev. 0, WPPSS N.P. No. 2, LPCS Pipe Whip Analysis.
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
5.49.056', Rev. 3, Target Resolution   for Postulated Targets Inside Containment, Draft.
-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 Narver, Inc., September 1975.0)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
~<,
WNP-2 AMENDMENT NO.9 April 1980 8310170212 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~i o~~~g~p~7 LPP+<M~PER.og.L~q P, FKd 0 pT%S WS T4<<~.4 5 A.W DzT s5 7~'P~oic.v pp~~~g)~p~g (+~T)$5(uO~vs40'%P+~4 AT'4Q gQ~p AS s~c~~>~~<~~~~g,g qqg.WNP-2 AMENDMENT NO.9 April 1980 The values of pr should be less than the allowable ductility ratios, p, given in Table 3.6-1.3.6.1.6.3.3 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.
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.
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.
E.2. 3   Technical Memorandum TN  1020, Rev. 1, Regulatory Guide 1.46, Recommendation  Concerning Implementation, 10/28/77.
Additionally, the most damaging single random active compo-nent failure in a system not effected by the postulated pas-siv component failure is postulated.
TM  1085, Rev. 1, Pipe Break Outside of Containment - Structural Effects, 10/6/78.
jf the direct conse-quences of the pasive component failure results in a turbine or reactor trip, then of fsite power is assumed unavailable.
TM  1151, Criteria for the Pipe Break and Missile Redundancy Evaluation Outside Primary Containment, 6/27/79.
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.
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
d.e.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 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.
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)
Design drawings are used to supplement the study in cases where piping or equipment have not been installed.
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
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.
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.
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.
NUREG  75/087, Standard Review Plan, Sections 3.5.1, 3.5.2, 3.6.1, 3.6.2.
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.
Regulatory Guide 1.46.
b.Isolated from the general area environment in compartments capable of maintaining normal equipment operating conditions.
Regulatory Guide 1.70.
Certain rotating equipment cannot be designed to function in the more severe, Local steam environment.
BTP NEB  3-1 and   APCSB  3-1, Section B.3, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment .
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.
      "Proposed  ASNE  Non-Mandatory Appendix  - Design Rules for Pipe Whip Restraints" Article L-1000,    NF  54, N/0 77-66 N76-6 January 1980.
Required redundancy is thus maintained for safety equipment.
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.
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.
: 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.
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,u&#xc3;s+VX'PMv locO For esults of the steam system study, alvsis indicates see 3.6.l.LL.4.n--<ar~eeauiL'ed reactor sa d ceilings act as jet shutdown.Some room walls, floors, an impingement
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.
: barriers,  
3.6.1.13.2       Jet Impingement Barriers AcT'~>acc<~T bAR,R,u&#xc3;s +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
                                                                                  ~ I~
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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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                                                                                      )!
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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
    ~
  ~
                                                '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 '
    ~  I
                                                                                      ~      I i
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    ~  I bl cQ le ct
          ~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


WNP-2 AMENDMENT NO.2 5 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.
L                                N,b 2 2'-Q 0Ol CV WASHINGTON PUBLIC POWER SUPPLY SYSTEM MODERATE ENERGY FLUIO PIPING SYSTEM FIGUf, )
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 f o1 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
NUCLEAR PROJECT  NO. 2         RUPTURE LOC. PLAN 9 EL. 501'       . 6-4p((
I


SUIIHARY DP SUBCOIIPARTHENT PRESSURE ANAI,YS IS (Page 1 oE 2 Compartment I)here Break Occurs piping System Eleva-tion fft.)442 Room Number n s-r~ic i on I inc R14/Rll3 RHR Pump Rooms 4 RCIC (l3)-4 Hax imum DiEfer-ential imari i 0.33 0.33 0.33.Design Pressure Jpsi)Differential Pressure Time Differential oE the Between the Peak Rooms/sec)R14, R113/R206 0.33 0.50 R14, R113/R12, 0.33 0.50 R114 R14g R113/R15p 0 33 0 50 R112 R15/R112 RCiC Pump Room 4" RCIC (13)-14 0.51 0.51 0.51 R15e R112/R205 R15, Rl)2/R14g R113 R15g R112/R6, R116 0.53 0.53 0.53 0.76 0.76 0.76 Ql N 471 501 50)I 501 R206 R 308 R 308 R3)3 EI.471'pen 4 AS (ll)-2 I'Ioor Area 0.05 0.05 TIP Room 0.05 4 RC?C (13)-4 0.32 PIP Room 6 RLICU (2)-4 0.41 EI.510'alve 6 RNCU (2)-4 Room R206/R103, R105, 0.35 R106, R305, R308 R310, R306, R315.R206/R114, R113, 0.35 R).12 R206/R116, R115 R308/R305, R206, 0.03 R313 R308/R305, R206, 0.35 R313 R313/R308, R400 0.35 0.08 0.08 0.50 0.60 R404 EI.522'pen 8'RD (l2)-3 P)oor Area 0.03 R404/R305, R504, 0.04 R508 0.05 (a)Tahi.appli..s to reactor builiiing seconiiary containment, exclusive of the main steam tunnel, tun-nel ventway, an)tunnel extension.
t7-'I"              22 '-Ca"     22'-6"      23'- 2" 4
C Pl 0 z M PJ CO'Z M r9 0
0 CU
    ~10
    \  ~
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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..
i 1


TABLE 3.6-11 DFSIGH LOAD IH AREAS ((HERE P IPIHG PAI LURE&OCCUR Ao-8 Pipe Brea~4 Hos.Room El ev.(ft.)R)S 422 Di f ferent ia 1 Pressure (esi)0.51 0 40 Differential Temperature op Int.to Int.int.to Ext.Live Load~(sf)llunp Loals (psf)From From Ploor C).~il in 59 Equip.l,o ad s s 1~4k Pump iso-4)XO 5)4)7~LZ.O-1~P 1W(-S)~1~9-Q I+8 yzr-1 gaze-49)W))+ti<++)
22'- Co" 0
R+-~i 1Z9-1$&'(~-''7)M i~m IC-52.&a.c-c L~)t5 1)gP~Ll(12.-~'t lqg-2g,q 1)'ze-9)39-1 I)B-XP,)>)~)b)7) 8)9 1%)1<)1 t R 113 441 R 112 R 206 441 471 0.51 n.n5 R 400 R 4n6&407 R 409 522 522 535 1.0 15.0 11.0 R 511 540 4.4 R 510 540 1.0 R 509 540 R 604 572 2.l 0.03 R 313 510'-6" 0.48 po po 0 0'po po 20O 20 20 0'0-40 40 40 40 250 59 250 59 250 32-250 40 60 6&34 30 250 41 250 126 08 0 400 00 55 400 65 51 400 08 250 15 50 36 250 40 00 Hone Hone Hone Hone Hone 1.5 Pump Hone Hone lleat Exchs.16.2&29.5 Hone lleat 6 Vent Unit 51K Steam T1)no el R 308 R 310 501 501 0.41 20.'0 0 20 4no 1000 63 1000 277 55 41 Hone None 4>C'X PJ m V'W Vl CO Z HO rt".:~ASEC 1.Por loc.)t ion of pipe bceak nos., see 2.P.>r verrical an 1 horlaont.)l seismi" factocs, see 3.7.Z 0 MKNDMENT NO.25 June 1982 l7'-1" 23'-2" l7)0 C4 CQ~I~I I f I CQ I IT)lA cz 4 g~WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO~2 HIGH ENERGY FLUID P IPING SYSTEM RUPTURE LOC PLAN 8 EL.572'.6-'  
OJ
                                                                /
8 CQ CQ, 04 g ~
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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'


AMENDMENT NO.25 June 1982 2 2'-(u" 23'-2" 24'-l 0".C)0Q cu v I CV cj rt W~C uk'ASHINGTON PUBLIC POWER SUPPLY SYSTEM HIGH ENERGY FLUIO PIPING SYS.RUPTURE LOC.NUCLEAR PROJECT NO.2 PLAN 9 EL.548'IGtlV,:
l7'- 9"   I I'3"     22 '-4s"     22'- &"      23'2"        24'-lO" 0
3.6-47' I
CV MASHIHGTOH PUBLIC POMER SUPPLY SYSTEH NUCLEAR PROJECT NO. 2 572'IGU:"
AMENDMENT NO.25 June 1982 Cl)0 C4 0QQ tY)rA N CU 0 LOG.l~i~F ht WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 HIGH ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 8 EL.522' AMENDMENT NO.25 June 1982 l7-9" Q I gll a.5 0%~ff)Cv 0 Q3 (Q)!I CV~~CV lY)cA CQ WASHINGTON PUSLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 HIGH ENERGY FLUID PIPING SYS.RUPTURE LOC.PLAN 8 EL.501'I('."'": 3.6-,'ld AbKNDMENT NO.25 June 1982 H.3 q I g<g 24'-l 0" 0 Al (Q 4.C4~I WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 HIGH ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 9 EL.471'i GU:.":~0', lc II Al1ENDNENT NO.25 June l982 H.B j7t qrp I l'-3" 22'o" 22'-Cu" 0~~cQ Cu rn HIGH ENERGY FLUIO PIPING SYS.RUPTURE LOC.NUCLEAR PROJECT NO.2 PLAN 8 EL.441'i Gti)')E 3.6-~.''
HODERATE EHERGY FLUID PIPiNG SYSTEM RUPTURE LOC. PLAH 9 EL.
I AMENDMENT NO.25 June l982 H3 Z K l7-q" Z2'-6" 2 2'-Cn" Z4'O h'~I 0~i I I OQ (Q C4~I bl cQ le ct~c=~~Wi C uaE WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 HIGH ENERGY FLUID PIPING SYS.RUPTURE LOG.PLAN 8 EL.422'-3" FIG Ui!;"..6.-~~~
i


H.3 J7'-t" lq'-3" 22-6" 2 2'-Co" Z4'-iO" s~tf)cQ CQ~~CQ 4 Qa~F iF WASHINGTON PUBLIC POWER SUPPLY SYSTEH NUCLEAR PROJECT NO.2 MOOERATE ENERGY FLUIO PIPING SYSTEH RUPTURE LOC.PLAN 9 EL.422'-3" FIGUjli.I 3.6-4'.'l
y Jph HB          Z      K        L                        N              N.s 2A '- lO" 0
04 0
tf) 04 0
Ct C4
          ~~
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 ~ .


'7'-9" 22'-Co" 22'-G" 23-'" 2'-10" 0 0 tA 04 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT HO.2 MODERATE ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 9 EL.441'FIGUk"" 3.6-0.'I l
C  '0 I
H.3 l7'" 22'-Co" 2 2'-Q" 23'-2" hl.8 0 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 MODERATE ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 9 EL.471'I GUR'c.';I.6-42ci I L 2 2'-Q N,b 0 Ol CV WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 MODERATE ENERGY FLUIO PIPING SYSTEM RUPTURE LOC.PLAN 9 EL.501'FIGUf,).6-4p((I
y/
C i 'f t
I ~
I
                          ~
                            /'3)
Jet Prom Qrcamferenthl Bros eath Eads Restrained (Flu      asr3..
                        > ~C~O~~


t7-'I" 22'-Ca" 22'-6" 23'-2" 4 0 CU~10\~eo cQ EL~MF F lG'OV.F WASHINGTON PUBLIC POMER SUPPLY SYSTEM NUCLEAR PROJECT HO.2 MODERATE ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 9 EL.522'IGURI i.6-42..1 22'-Co" 0 OJ/8 CQ CQ, 04 g~hl e ff)CA CO WASHIHGTOH PUBLIC POWER SUPPLY SYSTEM HUCLEAR PROJECT HO.2 MODERATE EHERGY FLUID PIPING SYSTEM RUPTURE LOG.PLAN 8 EL.548'IGURE 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 HODERATE EHERGY FLUID PIPiNG SYSTEM RUPTURE LOC.PLAH 9 EL.572'IGU:" i y Jph HB Z K L N N.s 2A'-lO" 0 04 0 tf)04 0 Ct C4~~a ffl dl CV M~cC i E~>Gu~i WASHINGTON PUBLIC POWER SUPPLY SYSTEH NUCLEAR PROJECT HO.2 HODERATE ENERGY FLUID PIPING SYSTEM RUPTURE LOC.PLAN 8 EL.606'-10 1/2" FIGUR';.6-42~.
AMENDMENTNO. 31 June 1983 270o                                 RC26
C'0 I y/C i'f t I~I~/'3)Jet Prom Qrcamferenthl Bros eath Eads Restrained (Flu asr3..>~C~O~~
                                                                ~RCR16 RCZ4
AMENDMENT NO.31 June 1983 270o RC26~RCR16 RCZ4~RCR15RC1 RC23~RCR17 RC15 RC20 RCR14 RCR1~RCR21~RHR SHUTDOWN SUCTION RCR20 RCR19 RCR18 RC3'RC16 RCR10 RCR11~RCR8 42II,Z 8MI gg~~Fi0 RC21 RC21LL gg~RCR13 RC12 RCR12 RHR SHUTDOWN~RETUFIN RC~RCR2lI'EY:
                                                                  ~   RCR15 RC1                                    RC23         ~   RCR17 RC15           RC20
~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 MAY ACT DURING THE POSTULATED BREAKS ARE SHOWN.WASHINGTON PUBLIC POWFR SUPPLY SYSTEM HVCLEAR PRQJF";;IO.2 REACTOR RECIRCULATION PIPING SYSTEM~,~FIGURE 3.6-35a
        ~
,l BRSCN No.BURNS AND ROE S A R CHANGE NOTICE (BRSVi4)~/~jgz Part I SAR sect on(s)af f ected: Part XI Description of Need for Amendment:
RCR14 RC21 RCR1 RCR18                  RC21LL RCR19 RCR20 gg ~RCR13
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
RC12 RCR12 RCR21 RHR SHUTDOWN
*!ANL P-" Approved Deviation  
                                                          ~RETUFIN RC3'                   RC16 RC~
~~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:  
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:==
==Subject:==
CHANGES DUE TO NEh LOADS PIPE BREAK""tA'SIS (SCN 82-175 ATTACt',ED)
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~5K+COnQg;.'I Cygelman AN Kugler TA Hangelsdorf-N Poivel 1 JJ Verderber MttP-2 Fi 1 es M Ouer BPA 8tt'R"BE..:4'"uR Site BAR Site BFCH BECH BKR HAPO  
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)
MilP-2                                                                           SCil " -/7 SAR CHAi'(GAL i'tOTKCE SAR   Section(s) Affected:
Affected: Description of Change: Reasons for Change: This SCH satisfies OCI Log Coavnitrrent Ho.: Th'is SCil con+its to the following:
Description of Change:
This SCtl'wi11 be incorporated into Amendment No.: See attached pages for original and/or revised SAR Section(s).
Reasons     for   Change:
Approvals:
This   SCH   satisfies     OCI Log Coavnitrrent     Ho.:
Signature indicates authorization to file the subject change into an amendment.
Th'is SCil   con+its to the following:
Lead Technical Reviewer(s)
This   SCtl 'wi11 be   incorporated into     Amendment     No.:
LTRs Si cnatur e Date>/~pe Remarks angry project~~)tanager Plant Operations f<anager f i~hg gn Proj ec.'iManager Project QA Manager" i~Yzwti/i>)<'3~,~>5.",App icable only for changes affecting Quality Assurance.
See   attached pages       for original and/or revised         SAR Section(s).
4.'si<,  
Approvals:         Signature indicates authorization to           file the subject change into an amendment.
)I GENERAL ELECTRIC CO.NUCLEAR POWER SYSTEHS OIYISION LICEHSIHG ACTIOH NOTICE WPPSS NUCLEAR PROJECT NO.2 Notice 4 Transmittal Date: Responds to:  
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:==
==SUBJECT:==
N, v FSAR: '.P.2 (~JZ r r V Gg'r n'7)HRC question 0:;;l.:, I" LSON MANAQaR ppqp is~iJceNsihG ACTION RE(VIREO:~+u 4ZW 4~~~4c'~M 0 Pir~4 Submitted by~.+~....I Date P.B.Kingston (Licensing ngsneer)Reviewed by A: F.DeYault rogect rng>neer Approved bv~(.o: Oate F.A.HacLean ro]ect Manager)c'~A" Distribution:
N,     v                       r    n FSAR:       '. P . 2   (~ JZ HRC question 0:                                                               ;;l.:, I"LSON MANAQaR ACTION RE(VIREO:                                                         p pqp is~ iJceNsihG
1.Licensing Eng.682 2.Projects 394 3.WPPSS (Original) 4.Burns~~Roe (R.O.}PK:cal/K'"~89
                                                                  ~+u 4ZW 4~ ~~4c'~ M 0
'~/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 r~L.E.Santos (l.icensing Engineer)Reviewed by A.;".OeYault (Ptojec Engineer)~".'.re'';r Approved by r~~r~X~<i'.~C%/.r'.
4 Pir~
MacLean (Project Manager).Date Oate Oa e Pi stributi on: L i cens i ng=ng.68.P ajac s c 3.'w'PPSS (Original)
Submitted by ~.+         ~   .... I P. B. Kingston (Licensing ngsneer)
Burns 8 Roe (R.O.)LS: hmc/1815 8/13/81 jLJ IrI.,E.  
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 pr''mary 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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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 g and components relied upon'or self e<<sector Shut 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}
l '
axe maintained pressurized under conditions
f'
+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~alan" conditions are pressuri"ed under both o>>Qe foU.exing coadi'oas: a.MME@temperature is 200 P or less.o 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 circumf Qronti@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'.2.1~1~1 a0 Postulated Pipe Break Locations in ASM Section III Class t Pipinq Runs 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 Ce (c), S.exceeds 3Sm but Se and Sr are (e)<f)eRch less than 2.4S , and U does not exceed O.l 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)Se is the naminal value of expansion stress as calcuLated 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 branch of the main pipe run, except that branch lines which are included with main~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 Ail h: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 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-, tulated~~-
~I l 4~~In.'.....moderate energy piping systems cons'sting of ASi~)E Code Section III Class" and 3 oioing and moderate energy non-nuc'ear piping, includi..g fluid system piping between pr''mary 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 f o13.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 go h.(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.
Por a Longitudinal split, the break area is assumed to be equal to cross-sectional flow area of the pipe.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&#x17d;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 3 6-3 l 4MOUIVl i<&.70 AT'E45i A PIPe ajA+Z i Eg gAr~>is pz Ae EhEnr7<>>7 uzEz Pawl ssc77ows


3.ongitudina3.
fib 0 t=ak.l3          2 F IGOEE        8.6.95k Able) /EST'RAIN TS IIICIHCHlhTlOH I'1PIHG SVSTEII Of'EAhTlHG STBESSES~hT IIHEhK LOChTlOHS eC, (i)
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.
Acl;ny                      ! ITltl.SS  IIATIO I'EII hf4E F(Illa'.
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~3.6.2.3.5 The flow from the pastu3.ated c=ack is assumed to resu3.t in an environment that wets all unprotected components wi~the competent,'ith subsequent.
j4$ tfa.yq      EQ(IO)                    I.Q(12)              I rj(13)
flaading in the c~~~ent I**\*~.I are detexmined an the basis of a conservatively estimated time period required to affect cor-rective action.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~.P atectian f~pipe whip need nct be provided if 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'
QIKhK        pdd .f        So                          So                                      USAGE                              IIMM( BASES 18Elff (l            I)     3Sa>                      3Sm                  :ISn>                FhCTCill                          SECTIOH HO.
me,u4 (1)-The transient forcing functions>
RCllLg f<ceg                                        0  19
at points along the pipe~aaae4W fram the propagation af waves'wave
            ~C,Z 'l CHILL cc,<      2 Xl    O. Q7                  Q,  gQ                                                        L onl4  ()) 3.G.L.Ll,l. <
~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.
AJo)eg:    cl)      ff'p        s'w
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
                                          $  e sn +I'im      ii    gg          Lc o j  p      a~4    <S                      47+)    3  o Qrtgg~         QCQ      L  L      aP  t>ee g      ~     I oo
                                                                                          )  P    o~~q        SC,a. F~)~     Q ~$ ~
B~ai~~                       ~   g      P iver          f   ~o (L)         Our    ~ 0      $
                                              . 1o~         Ar y,i (x)          ft<~           0<      aeewcchgg            yf    MQQvC          wgQL          II
                          ~'v'c ~Ccmv k eJL ~Q                                              ~(gl.
                ~o) 5cC.
L a       c.W Angita    J:  W    +c
: 5)       QI~~               ~       k <a~>.S        wL( A QJ


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.
V ~
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, RKV 3 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.
fr
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.2 S 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.
c~/2 Mcr d (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)(2)(3)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 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.Pipe whip.xest=aints,.
TYFICAL,FORCf DfF L.ECTION CLIRVf FOR A MtAR,244'. F@% RfSTHAINT t.OADfD AT 0 1100 F <<6{102M(5-dJXW              {
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', a e P.esianed by the architect/engine f he met.des-cribed, in c., (below)z.utilized"or this 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.
r  I  I I        f O
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 (i)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).
CC O
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.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.
ill LINfAR EAl.IZATION C
W,th raga~ta rebound, it should be noted that.if 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 f Br 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.
TOTAL. Ct.fARA      <<LSS in.
T.t should also be noted, that the assump-tion of a suddenly applied, constantly'aintained force, as used in the equation 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.
Alt              I      I 0   5     7      $     0      10    'l  l    12    12 14      15 DKFLKCTION,B Iln I (AT        CTQ.AIMT          I  ooeeiOH)
3..6.2.2.3 Naterial Properties Under Dynamic Loads 3.6.2.2.3.
T AI. Ct.fARANCf<<
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.
INITIAL.                       fFF ECTIVf Ct.fARANCf                    ClfARANCf 4 Ind                          1188inJ  .
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 co."itute px whip limi stops, ar permitte to deCo.a maxi..of 50%the min'm uniform"rain, du.'ag ner>absorp'.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.
MASHIHGTOH PUBLIC POWER SUPPLY SYSTEM                TYPICAL RESTRAIHT FORCE-OEFLECTIOH      FIGURE HUCLEAR PROJECT HO. 2                                               CURVE            3. 12-5
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.the ru.ne b ea event is considered to be a aulted t i r e con ition<.-".,'s rest aints, and structure 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 stat'loads, if the consequences of such'nstability
'result in 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.
~ I THIS FIGURE  HAS BEEN INTENTIONALLY OELETEO
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.
Ref e~ed t:o Figure 3.6-35a
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,~
    'WASHINGTON PUBLIC POiER SUPPLY SYSTEH BREA'OCATIONS %D          FIGURE RESTRAINTS ANAL'IZED, PDA 3.12-6 NUCLEAR PROJECT NO. Z                    'IERIFICATION PROGRPH
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
/AT QAlNjs CF>soge8.4-43 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~~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.b.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.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: (b)70K of Su, where Su=minimum ultimate strength by tests or ASTM specification;
TADLE 3. 12-3 BESTIIAltIT I'BOPL'IITiES USED lN  *IIALVS80~
+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.
Uuncral Auutr aint Datd fSr      1 Bar of   a Bustraint F  C 2
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
(h  restraint)
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.
Hhuru h restraint        d  pipe  - Total clearance C.~     P-a g  Q. l4 -g)
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 sta-ture, other piping systems, cable trays and canduits.3.6.2.3.4.1 ae Pipe Displacement Effects on Components in Same Piping Ran 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&#x17d;s for faulted conditions and l~ts ta ensure operability, if equired, are met.3'-54
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.
$ -4ppi 2I .
              ~ aerain~e~~~


iPlP-2 AiMENDi4ENT NO.25 June 1982 a.Assurance nf primary containment leak tightness.
O.
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
Ehbl.E 3. 12-3 (Cont Inuedl CONPAAISON OP PDA AHD HSC CODE 1 OC    Design break           Restraint                                                          Restraint            Restraint                    Pipe Indent                Indent PI ur'e 3. 12-6 Ho. or bere A)~QM                                      De!lection ~ln.
~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.
PD  =fPR De!lection          Deflection      ln.
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.
EL 1
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.
ACIES ACR1            5        5         003.2        708. 3   6.57      7.92C    79+9'6.4              1  17.72        15.Sb BC 2~              ACR1            5        5        766.4        458.4    14.99        7.495  12S      1  Cl.C      1  35. ~ 3       24.52 AC3LL              BCR2            6       6        747.0    -
lines may be completely d.The entire eactor recirculation cool'ng system'n 3.6.2.5.4.l4 and in Pigures 3.6-35a 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.
639.7      2.27      3.73    27.65'      4$ .351      17.14        lb. 11 BCILL              IECA2          6        6        796.6          780.3     lb.22      10. 54    S7.8    ~  S9.C      1 41.4 ~        43,0 RC4LL              BCA3                                            838.4      7.64      8.05    92.951      97.981        lb.b7        16 43 AC4LL RC4C RCCA RCR3 RCAI RCR3 8
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
8 +~C 8
.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 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'n individual basis to determine compliance with core cooling quirements.
1319. 0 1 lEEE.
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:
92b.5 E
a.Leak tightness of the containment f ission product barrier is assured throughout any LOCA.b.c 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.?ipe'whip supports are provided n the vicinity of normally open isolation valves inside and out-side primary containment for high energy systems, to assure that oper ab il 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 a.'ystem Arrangement russo mops'A"~'8" oF rHE~>PRE+The>recirculation 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 to 330).These five lines Mop vertically alongside the sac ificial shield~all<from ele-O vatxon 36'o a 16-inch diameter heade at centerline e evation of 528'A single 24-inch diameter line then drops vertically~o~mg7/>
1073.
the center of the header to eLevation 506'here it is routed into the discharge nozzle of the recirculation pump.
I 72~
t I~b MNP-2 AMENDMENT NO.9 April l980%1 8//~~+tsA II n i i y oriented R~p<<T;>v pg.the g'nd 38Q'ximuMs, with respect to 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
9    5 43 4.49 1.22 4.62
'hield mll to elevation 502'-6 T/8'here it is tatted to the euatiaa uatalg af the teait culation pump.Pipe Whip Protection For the recirculation pump s ction and discharge systems, the location of pos=ulated pipe""aaks L and pipe whip restraints are shown on Pigure 3.@-9~~~38 w ich is representative of both recir-cu ation oop.Where pipe'"reaks are postulated gogF<RgpNqg
                                                                                            $ .58 1.77 99 ~ 211 80 ~ 37 '9
<+~~~inside primary containment>
: 22. ~ 6'1 76 ~ 8S ~
the reci'rcula
F 31 7 ~
'ystem piping is restrained
891 23 ~ 38 22 F 54 23.68 17 lb ~ 73 9S 39 8
~prevent unaccep-~~~~~i table motion These restraints are generally 7oNS'J1tlg mK CQ~mounted on the side of the sacrificial shield 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 ResrRzw T 5 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.
AC7d              IECRI                              9$ 3. 3         ~l                5.76    76.4    1  lb.121      16.46        21 Cl AC8LL              RCA6                              S99~            0            i        0    ill+4Cl                  2C ~ 7C ACR7                              8S'S.~          0                      0    110. 741                  29. 316        8 39 RC9C                                                  5'7$ . ~    +lb.I6        ~ +16              50.631      C7. 331      13.2         34+SC BC 9LL            RCAb                              830. 2      5~~44. ~  11. 408          1$  9S.291      S6.9    1    36.C12        2C F 24 BC1 I A            IECA8                              81 ~ . '1    4934      1  . 8      5. 99    91.72  '0.07            31.404        23 ~ 71    o W
Pipe 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
K lb.i4 BCI 3             IECIE I0                          468.4        478.0      5.87      3.6C            1  58.391      13.37                    ~ e RCI6              ACAI 1                            687. 4    5I8    4    6, 59      i. 38  105      1  49. 8Cl      15. 37        lb. 22 0
'"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 v~x q C/l~Q ro 7 INSULATION CLEARANCE CLIP GUIDE ROD al ml I o O 0 x m g gj w C: 0 gm w X O e-a a BRACKET CLEVIS PXH-MELD STRUCT.4J c m z~c=RCazo Pc@~i RSP, sit'll'T'LOMB SU%,'TcoA.+
CO a
3L R 1 jP~ROC>>~, RCR'LCL RCPT'I goH: 5 QC1L'L'Lu Thaw w R.f'T v 4.e4 khan: "o~~Uoc~~--J o 4 Klihyl<srCa~)N(7 I cl'.(I)'f44 s~'I I A g4~I Q S e<~f~p~i c.A.(~)5 t's pm~I oc.aRM terr meta+:~<4 Pg Pc S.0~I 4 0 R.(=Qfc.tRcu uciad P ('4 g5 a.'5g l7 5
RC14C              BCAIO                              28S.O        309.6       2.83      $ .88    46.3   \  9$ .921      1S.4$        13. 96          O BCIiI.L            IECR20                              114. 3     129.9      0.94      3.36    10 5 1      37 1 1      22 13        23.56 Pipe Aupturs Novessnt    by Genersl Electric ccsEpany.
~)I I y i~~~~g~/I~Q~1 4~0 i e i 1 4 1 e~r~ia y I~~e~0 I i l r S-~J j l'f' fib 0 t=ak.l3 2 F IGOEE 8.6.95k Able)/EST'RAIN TS III CIHCHlhTlOH I'1PIHG SVSTEII Of'EAhTlHG STBESSES~hT IIHEhK LOChTlOHS (i)eC, QIKhK 18Elff (l Acl;ny j4$tfa.yq pdd.f I)!ITltl.EQ(IO)So 3Sa>I.Q(12)So 3Sm I rj(13):ISn>SS IIATIO I'EII hf4E F(Illa'.USAGE FhCTCill IIMM(BASES SECTIOH HO.RCllLg f<ceg 0 19 CHILL~C,Z'l cc,<2 Xl O.Q7 Q, gQ L onl4 ())3.G.L.Ll,l.
0
<AJo)eg: cl)ff'p s'w$e sn+I'im i i gg Lc o j p a~4 Qrtgg~QCQ L L aP t>ee g~I oo)P o~~q B~ai~~~g P iver f~o<S 47+)3 o SC,a.F~)~Q~$~(L)(x)~o)Our~0$.1o~Ar y,i ft<~0<aeewcchgg yf MQQvC wgQL~'v'c~Ccmv k eJL~Q L a c.W~(gl.5cC.II Angita J: W+c QJ 5)QI~~~k<a~>.S wL(A V~fr 1100 TYFICAL, FORCf DfF L.ECTION CLIRVf FOR A MtAR,244'.
 
F@%RfSTHAINT t.OADf D AT 0 F<<6{102M(5-dJXW
  )
{I r I f I O CC O ill C LINfAR EAl.IZATION TOTAL.Ct.fARA<<LSS in.Alt I I 0 5 7$0 10'l l 12 12 14 15 DKFLKCTION,B Iln I (AT CTQ.AIM T I ooeeiOH)T AI.Ct.fARANCf
4
<<INITIAL.Ct.fARANCf 4 Ind f FF ECTIVf Clf ARANCf 1188inJ.MASHIHGTOH PUBLIC POWER SUPPLY SYSTEM HUCLEAR PROJECT HO.2 TYPICAL RESTRAIHT FORCE-OEFLECTIOH CURVE FIGURE 3.12-5
 
~I THIS FIGURE HAS BEEN INTENTIONALLY OELETEO/Ref e~ed t:o Figure 3.6-35a'WASHINGTON PUBLIC POiER SUPPLY SYSTEH NUCLEAR PROJECT NO.Z BREA'OCATIONS
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
%D RESTRAINTS ANAL'IZED, PDA'IERIFICATION PROGRPH FIGURE 3.12-6
(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


TADLE 3.12-3 BESTIIAltIT I'BOPL'IITiES USED lN*IIALVS80~
Uuncral Auutr aint Datd fSr 1 Bar of a Bustraint F C (h restraint) 2 Hhuru h restraint d pipe P-a g Q.l4-g)-Total clearance C.~I'lpe Size I Inl Best Load Direction C2~mi t Initial Ef f ective 6: Restraint Clearance Clearance Total Clearance 12 Oo 27,733 0.24 6.12 4 1.941 16 16 24 24 24 Oo 90 Oo 90 38 Ppg 0.24~1 109,265 62, 599 1O2,228 O.24 6.2 8 8.978 8.222 55,531 0.375 C.972 4 109,888 109,835 4 0.2~5&i 1.934 12.187 1.984 13.685 5.698 12 90 14,795 0.401 9 0'4 12.247 5.941 16.247 5.934 16.187 5.4 17~685 9.698 12.462 it)ttsc dunoteu ttucluar services corporation, and PDA denotes Pipe Dynamic Analysis Pro9rum for pipe Sruak Hovumcnt" by General Eluctric Company.$2 I.-4ppi~aerain~e~~~
O.~~Ehbl.E 3.12-3 (Cont Inuedl CONPAAISON OP PDA AHD HSC CODE break Indent ACIES BC 2~AC3LL BCILL RC4LL AC4LL RC4C RCCA AC7d AC8LL RC9C BC 9LL BC1 I A BCI 3 RCI6 RC14C BCIiI.L Restraint Indent PI ur'e 3.12-6 ACR1 ACR1 BCR2 IECA2 BCA3 RCR3 RCAI RCR3 IECRI RCA6 ACR7 RCAb IECA8 IECIE I 0 ACAI 1 BCAIO IECR20 Ho.or bere A)~QM 5 5 5 5 6 6 6 6 8 8 8+~C 1 OC Design Restraint De!lection PD=fPR Restraint De!lection
~ln.Pipe Deflection ln.1 EL 17.72 15.Sb 003.2 708.3 6.57 766.4 458.4 14.99 747.0-639.7 2.27 796.6 780.3 lb.22 838.4 7.64 5 43 1319.0 1073.9 1 lEEE.E I 92b.5 72~8 9$3.3~l 4.49 1.22 S99~0 i 8S'S.~0 5'7$.~+lb.I6~+16 5~~44.~11.408 4934 1.8 830.2 81~.'1 468.4 478.0 5.87 687.4 5 I 8 4 6, 59 28S.O 309.6 2.83 114.3 129.9 0.94 79+9'6.4 1 7.92C 7.495 12S 1 Cl.C 1 35.~3 24.52 lb.11 27.65'4$.351 17.14 S7.8~S9.C 1 41.4~92.951 97.981 lb.b7 99~211 76~8S~23~38 80~37'9 F 891 22 F 54 22.~6'1 31 7~23.68 76.4 1 lb.121 16.46 3.73 10.54 43,0 8.05 16 43 17'$lb~73 9S 39 21 Cl 4.62$.58 1.77 5.76 ill+4Cl 110.741 2C~7C 29.316 0 0 8 39 34+SC 50.631 C7.331 13.2 1$9S.291 S6.9 1 36.C12 2C F 24 5.99 3.6C i.38$.88 3.36 91.72'0.07''1 58.391 23~71 lb.i4 lb.22 0 13.96 23.56 31.404 13.37 105 1 49.8Cl 15.37 46.3\9$.921 1S.4$10 5 1 37 1 1 22 13 o K W~e CO a O Pipe Aupturs Novessnt by Genersl Electric ccsEpany.0
)4 TABLE 3.6-6 Page 1 of 7 DESIGN'ASIS BREAK LOCATIONS OUTS IDE PRINARY COHTAINHHNT Hax.Porc (k ips)o Thrust vs Time F iilU Isometri No.(H200)f ine Des iona ion W<<CCC(13)-4 2'CICfi13)-4 8 RClC(13)-4 RCIP(13)-4 RC3jC(13)-4 RC(C(13)-4 R C(13)-4 R IC(13)-4 Diameter (Inches)1?0-1 120-2 120-3 120-4 120-5 120-6 l.2 0-7 120-8 Later 3.6-6 , 70 3.6-5, 66 Lat r*er*ater*f ater*f.ater*3.5-63, 64-3.6-79, 80 3.6-75, 76 Later*hat r.3 5-81, 82 4 4 4 120-10 126-1 126-2 l.26-3 1'P'R 11 R lQ Rf l.Xe(-1e)-4 CU(1)-4 CU(1)-4 CU(1)-4 125-5 126-6 128-R CCU(1)-4 Rl CU(1)-4 R'l U(2)-4 8 ,9 0 1 2 2 2 3.6-67, 68 RH (2)-4 1-8 f.a6e r*RNCU-4 128-9 6 Later*RHCU(2)-4.1 28-10 Later*RvfCU(2)-4
'28-11 Later~~RtfCU H-)-4---1 28---q---"-6----"-fia tee$--~WOO~)1~2~~-5 fat-c ,~HCU~Q- ----.l.-4-1-~--6------La t e r~-Plan~pocation'F i g 0 I t.3.6-49 3.6-49 3.6-49 3.6-48 3.6-48 3.6-48 3.6-48 3.6-47 3.6-47 3;6-51 3.6-50 3.6.=50 3.6-50 3.6-51 3.6-51 3.6-51 3.6-51 3.6-50 3.6-49 3;-6-4 9-3.6.&0 3.6-50 3.6=50-TABLE 3.6-6 Page 2 of 7 DES IGH BASIS BRBAK LOCATIOHS OUTSIDE PRIHARY COHTAIHHMT g6 2!7 2'!2)31 3)3,~4 Break t4o.Line Devil nation RHCU(3)-4 RHCU(3)-4 RWCU(3)-4 RHCU(3)-4 RHCU(3)-4 RHCU(3)-4 RHCU(3)-4 HS(20)-4 HS(20)-4 HS(20)-4 HS(20)-4 Isometric t$o.(M200)129-42 129-43 1.29-4 4 129-45 129-47 129-48 129-50 134-1 134-2 134-3 134-4 Diameter (Inches)5 4 r.a t Lat La e La e r*r*I*r*L t er*f ter*ater*Later*f ater*fater*Later*H x.Force (k ips)or Thr st vs.Time Figure Plan Iocation Figure 3.6-50 3.6-50 3.6-50 3.5-50 3.6-50 3.5-50 3.6-50 3.6-50 3.6-50 3.6-44 3.6-44 3.6-44 3.6-44 0 AS(11)-2 2 AS(11)-2 3 AS(11)-2 1.3 9-.1.1 3 9-3 139-4 3.6-97, 98 3.6-93, 94 Later*3.6-43 3.6-43 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 1979/2'f~p (RC2LL p i iRc@i~RCR20 RC13 RC16 RCRI I I I RC12 RCR9 RCR8 RC3LL RC11A RCR3A J RCOLL RCOCV RC4LI.RC 1CV RCR7 RCSLL RCR6 CEY'C4CV~T iPICAL 8REAK LOCATION'RR3A~TYI'IcAL RGSTRAINT OCSIGNATION Rr i1 S'il RC7y RCGA V/()C WASHINGTON PUBLIC POMER SUPPLY SYSTEM NUCLEAR PROJECT l(0.2 REAC:OR RECIRCULATIOH COOLING SYSTEM FIGURE 3.6-35 270''CR16 RC24 RCI RC23 RCR IS RC15 RC20 RCR17 RCR14 RCR\CR20 RCRIB F.2'I RC21LL RCR13 RCR'>I'HR SHUT DOWN SUCTION RC3@RC13 RCR11 RCR 10 RCR12, RC12 RHR SHUTOOWN RETURN RC1 RCRB RCR2II'EY ABACI TYPICAL AIIAAA LGCATIGN RCRI TYPICAL RES~RAIN T OESIGNATION SUFFIX"LL" INOICA 3 LONGITUOINAI.
Amendment No. 5 August 1979
BRFAK INOIC*TFS'OP A ONLY iNOTcS: I, i HIS FIGIJRE REPRESENTS LOOP A, LOOP B is SIMII.AR EXCEPT AS NOTED.2.SEE FIGURE 3,6 36h FQR RESTRAINT~BREAK LOCATION CORRELATION ANO BRcAK TYPES 3, ONLY THOSE RESTRAINTS THAT MAY ACT OURING THE POSTULATFO BREAKS ARE SHOWN.WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 BR'EAK LOCATIOiNS AND RESTRAINTS ANALYZED, POA VER IF I CATION PROG RAh1 FIGURE 3.6.35a 5 v/'gL g C" Fu>rn BR8002A[9/82)200M le O n m a r O O m m CA Q 0 O C 4/l C/l m 42 4i)2 PiPE SLEEVE IO PiPE SiEEVE 40 l5 LOOP l/z m 4(p 45~~O C qS+32 LoOP 5 TABLE 3.6-5 Page 3 oE 7 DES IGH BASIS BREAK LOCA'f'IOHS OUTSIDE P)/IMARY COHTAIHMEHT Break Ho.5 5 5 5 5 6 7 Lin u~esi nation AS(10)-2 AS(10)-2 RHCU(f)-4 RWCU(l)-4 RHCU(l)-4 RHCU (1)-4 RHCU(l)-3 RWCU(l)-3 RHCU(1)-3 RHCU(1)-3 RHCU(1)-3 RHCU(2)-3 RWCU(2)-3 Isometric Ho.((1700)14 1-P4 141-12 142-20 142-21 142-22 142-23 144-24 144-26 144-27 144-28 144-29 144-31 144-32 4 Max.Force (kips)or Thrust vs.Time Picture Later*l.ater*7,~g Later>>Fi(g.k8>Later" Later*Later*Later*Later*Later*f ater*Later*Later*f ater*Plan Location riciure 3.6-43 3.6-43 3.6-51 3.6-51 3.6-51 3.6-51 3.6-53 3.6-51 3.6-51 3.5-51 3.6-51 3.6-51 3.6-51 RWCU(2)-3 144-34 (fater*I 3.6-51 69 70 71 RWCU(2)-3 flS(9)-2 HS(1)-2 144-36 148-1 148-2 fater*3.6-53 3.5-112, 113 3.5-4'3 Lqter*3.6-43 73 74 75'fs(5)-2 HS(5)-2 Ifs(5)-2 148-148-6 148-7 Lager>>Laker>>Later*3.6-43 3.6-43 3.6-43 O TAAf E 3.6-5 Page 4 of 7 DESIGH BASIS BREAK LOCATIOHS OUTSIDE PRIHARY COHTAIHHEHT Break Ho.76 77 78 79 80 Line oeaictnation HS(5)-2 HS(5)-2 HS(5)-2 HS(5)-2 HS(5)-2 Isometric Ho., (H200)148-8 148-9 148-1 148-148-Diameter (Inches)2 2 2 2 2 tlat.Force (kips)or.Thrust vs.Time F~iure Later*e Later~Later*at r*Plan Location Figure 3.6-43 3.5-43 3.6-43 3.5-43 3.6-43 HCO(11)-1 149I'2 I I*I Later*3.6-62 i a HCO(ll)-2 149-5 3.6-99 I/3.6-58 95 96 97 98 99 100 RFtf (1)-4 RFH(1)-4 RFvl(1)-4 RFH(l)-4 AS(9)-2.AS(9)-2 5-1 24-Later*335 24 Later*335-Later*335-4 24 342-13 6 Later*342-14 Later~3.6-49 3.6-49 3.6-49 3.6-49 3;5-43 3.6-43'K O TABLE 3.6-5 Page 5 of 7 Dt'SIGN BASIS BREAK LOCATIONS OUTSIDE PRIHARY COHTAIHHENT Line Des iwnat ion AS Isometric Ho.(H200)Diameter (Inches)Hax.Force (kips)or Thrust vs.Time FiIaure Plan Location Figure HS(l)-4 400-8 Later*/3.6-44 HS(1)-4 400-11, 26 i Later*"/I 3.6-44 HS(1)-4 400-'l4 26 Later*/3.6-44 4 HS(1)-4 1.400-18'6 Later*3.6-44 CO(3)-2 CO(3)-2 CO(3)-2 llS(5)-i')S(5)-1 HS(5)-1 HS(5)-l HS(1)-1 HS(l)-I',440-1 440-2 440-3 4'47-19 447-2'5 4 47-26 447-27 448-15 448-16-.2.5 2.5 2.5 6 6 5 6 6 C Later~Later*Late'r'atej*
                                                                      'f~
Later'ater Later*Later*Later*H/A H/A H/A H/A W/A N/A N/A N/A N/A 0 0 DESIGN BASIS TAI3LE BREAK LOCATIONS 3.6-6 OUTSIDB Page 6 PRI jjARY COHTA,IHtjEHT of Break Ho.126 127 128 129 130 131 132 133 134 135 136 137 138, 139 140 141 142 143 144 145 146 147 148 149 150 Line D~esi nation HS(1)-1 jis(1)-1 jiS(1)-1 HS(l)-1 HS(1)-j HS(1)-1 HS(l)-1 jiS(1)-1 fiCO(5)-1 Hco(5)-1 HCO(5)-1 jiCO(5)-1 Hco(5)-1 jjCO(5)-1 jiCO(5)-1 HCO(5)-1 jjCO(5)-1 HCO(5)-1 Hco(5)-1 HCO(5)-1 Hco(5)-1 HCO(5)-1 jico(5)-1 HCO(5)-1 tjS(9)-4 Isometric No.(H200)448-17 448-18 448-19 448-20 448-21 448-22 448-23 448-24 449-13 449-14 449-15 449-16 449-17 449-18 449-19 449-20 449-21 449-22 450-33 450-24 450-25 450--26 450-27 449-28 451-6 Diameter (Inches)6 6 6 6 5 5 4 3 3 3 3 3 3 3 3 3 3 3 3 3 2.5 3 3 3 Ha@.Force (kigs)or Thrust vs.Time P~iure Later*Later*Later*Later" Later>>Later*Later>>Later*Later'ater*
                                            /2              (
Later>>tater" Later'ater*
p RC2LL                                   p           i iRc @i ~
Later'ater*
RCR20 RC13               RCRI I         I I
Later'ater*
RC16 RC12 RCR9 RCR8 RC3LL                                                     RC11A RCOLL RCR3A J                                                         RCOCV RCR7 RC4LI.                                                             RCSLL RCR6 RC 1CV RC7y CEY'C4CV
Later*j.a ter" Later*Later*Later*Later*Later*-Plan=Location F tgll te H/A H/A H/A H/A N/A N/A N/A H/A N/A H/A N/A H/A H/A N/A H/A H/A H/A N/A N/A H/A H/A H/A H/A H/A iV/A O TABLE 3.6-6 Page 7 of 7 DESIGH OASIS BREAK LOCATIOHS OU'L'E PRIHARY COHTAIHHEHT Brea)'o.151<.1 5"2-=" Line nestcSnation HS(9)-4 CR D (-1-2.)>>3---
                  ~ T iPICAL 8REAK LOCATION
Isometric Ho.(I<200.i 4g'1-7.H/.A D meter Inches)Hax.Force.(kips)or Thrust vs.Time Fi ure Plan Location 2 l.flite 3 Later*8;H/A-~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 m z a~W CO&o WNP-2.AiMENDMENT NO.25 June 1982 TABLE 3.6-7 Line Desi nation E Ei I/IE ET Ei ET I/I r-L SEISMIC AND QUALITY CLASSIFICATION Page 1 of'2 r Classification Diamete Seismic~~Qualit rc RCIC (13)-4 I E RWCU (1)-4 4/6 I RWCU (2)-4 4/6 I I r RWCU (1)-3 4 E RWCU (2)-3 I RWCU (3)-4 4/6 I RWCU'(5)-3 I RWCU (6)-4 I RWCU (7)-3 6 E AS (1)-2 4,8 1/E I AS (3)-2 2 I Ei AS (10)-2 6,8 AS (11)-2/2,3,4 i I.S (16)-2 2.5,3';I CO (3)-2 2,2.5 ii HCO (5)-1 2.5,3 EI HCO (5)-2/'/2~3/3 EI HCO (9)-2 2 HCO (12)-1 2.5,4,6 I EE HCO 1)-2 3 E HS~1)-1 4,6 1/Z1 HS (1)-2 2,4 EE S'(5)-1 6 HS (5)-2 4~3 Ei 3.6-86 W.Or 8y f7a-a o>so-/3&r="Pl<, Llama E, DESI(red pro g+/~~8WS/5'(PA.KMW C OC.&7/omS/r/E DzH, O~>75 g~p piZ/n~/<Cd WIT~/~a~ivr~
      'RR3A     ~ TYI'IcAL RGSTRAINT           RCGA V OCSIGNATION
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()C WASHINGTON PUBLIC POMER SUPPLY SYSTEM                                                               FIGURE REAC:OR RECIRCULATIOH COOLING SYSTEM               3.6-35 NUCLEAR PROJECT        l(0. 2
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'IN.(DA'r By.~~ra a arg Rpg e Q JrhS~E DE'+8+-i'-8~E: l.g~aC.yVO 2 DV&i4<''-I i-ic'~a iso Qrz.c.is K 4)ArE OE~ltr A&i7GAJ A/PE Dld.WAX FrrrcCE Pi6jrr r OIZ 7fjAu57 v, 7h.F p IP 3" I~W.O.Oraw By Tibiae.7Ac6 3.6-d.Ta~T ppr<-)J err+s 8n.smf A~OQ/QQ/75~)lgi&#xc3;kli'8/v docs>r-&4 J=)'~PE/~g 5 f~))~r-i'o5'WO.ohio zPP-rig~/<,-C'g d'S 4'.rS RE/'.cToR VE55EL IA5.NOZZLE NSP, 29j 2BZ 23I~Pho~P.WS Bl.l c e 2.WSOZ)-+I 2t tAS(l)-+e m-Rv-2.A IA5-RV-IA eve'22c'L'L5~+30 MS-RV"+A'45 RV-3A C32g PWS 31-Z, I.224 PM5 sl 1 PW5 31-+Cubi PWS BI-S C'3l 6%4 227.22'l-.228 225 O'L I I!.X-IRK PLOYS',g.CITRIC;T~OR
270''CR16 RC24 RCR IS RCI                                               RC23 RCR17 RC20 RC15 RCR14 RCRIB           F .2'I RCR\                  CR20                                              RC21LL RCR13 RCR'> I                                                         RC12 RCR11
~~~~QS-V-2.2 A.22K.~.".S'Bl-5.I j i.~<.$g,$PWS Bl"Co 220 2fl 213.via MASIII" OTN POOL 1 C POMER SUPPLy STSTEg IIUCLEAR PROJECT IIO I'IAIII STEAH LOOP A 15ONETRIC FIGURE 3-CN tRIP-2 Amendment No.~gQ SUHHARY OF POSTULATED PIPE BREAK LOCATIONS CIRCtiHFERENTIAL BREAKS LONGITUDINAL BREAKS Node 216~'t~1ode 219 Node 221'ode 2.24 Node 226 Node 227-Node 229 Node 230.Node 232.Node 291 Node 220'ode 225'ode 228 Node 231'or 493~os<<>>ypDE 6A7 H2lSHIHGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT~2'AIN')En''l LOOP A FIGURE 3.6-12 N~~)~24't+PW5 3'L-I aSI Rf ACTOR VE5SEL 250 MS.NOZ'XLE.NBS~2.lMS(l)-+~~HZ 24I 24og PW5 M=I I%9.:..64l~'i N5-RV-2.B..hhS-RV-+5 Ih5.RV-IB thS-RV 3B-to>fg A8;.: logos 435$'1 PWS 32.-+RI4 6 MS-RV-5B 642..6 44 443%46.241 24'>>PW5'M:2.IO RCICIIS-+:/OR CoNX SEE FIMI: 2I4A..-PW5 52.-3 243 I 244'245 I I i X-I SB'tA5-V-2.'LS PW5 SL-Co I I'55 FI.oV4 P,GSTH.IC fOR PW5 32;5 237.234 lip:~l,t I.llII STEAI LOOP S ISK=iRIC~g g~~~4,~~c f IGURE 3.6-13' Amendment ND.WD@NNP-2 SUMMARv OF POSTULATED PIPE BREAK LOCATIONS CIRCUMFERENTIAL BREAKS LONGITUDINAL BREAKS Node 235~Node 237.Node 240 i Node 242'ode 246'ode 248 Herl e 249.Node 251~Hode 292 Node 636'~~Node 640'ode 644'oE~llew Node-236'ode 241'ode 247'ode 250-e'o~CYz P>E 4'0 QOOE g3 y~~<<(z7~ouS'3 4 I WASHINGTON PUBLIC POWER SUPPLY SYSTEM I NUCLZAR PBOZECT NO 2 I'Ali'TEAN LOOP 9 ZZCURE 3.6-13' PWS 33-1 2t/C7 e 2@i PW5 33-+REACTCIR VES5EL Ihs HOiX,LE.N3C L'I3"2TO.r'ltoS~PttVS 3'3-I I 2.6 Iso)-+I I 7.I p'.)IF i~I1/'j l)p ,', f'$,7k t'p r I/7 g-Ib C FE 5 258" F LOU FI,SSf P.ICToR~a 1 4/52 G5I tA5.RV-IC C77 Q (77~5.I7AS-RV-K&8).c 4T~..Ms RV-Bc'-7~~1/~(5Q,~fAS-RV-~I 7~'I MS-RV-SC.l PWS 33-2.~2(7'BpCt F/2t5'-t;!, w~'(s-.MS"V 2:LC Pb;~1,17).t2'5 4 PV/5 33-S PW5 33 3-'2C2.2v4 2E 3.AHBIIDHENT HO, 9 april 1900'flak g 1~./37" 7 j7/7,.~7'55 MA lf6TON PNLIC PORE)SUPPLY SYSTBI (j~ltl STEg p j IIUCLEAR PROJECT tt0 2 HAltl STEAtl LOOP C ISONETRIC FIGURE.6-14a WNP-2 Amendment No,&DR  
                            '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
 
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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
Classification Line Desi nation            Diamete                Seismic      ~      ~Qualit rc RCIC    (13)-4                                      I                    E RWCU    (1) -4                4/6                                        I RWCU    ( 2) -4                4/6                  I                    I r
RWCU    (1) -3                  4                                        E RWCU    (2)-3                                                            I RWCU    ( 3 ) -4              4/6                    I RWCU'(5)-3                                          I RWCU    (6)-4                                        E                    I RWCU    (7) -3                  6                                         E AS  (1) -2                  4,8                    1/E I                Ei AS  (3)-2                      2                   I                    Ei AS (10)-2                    6,8                    I/IE AS (11)-2
  .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)
(9)-2
              -2/'/2          2.5,3 2
                                      ~ 3/3 EI EI HCO  (12 ) -1              2.5,4,6                I                    EE HCO      1) -2                3                  E HS  ~1)-1                    4,6                  1/Z1                  ET HS  (1)-2                    2,4                                        EE S '(5)-1                      6                  I/   I                r-L HS  (5) -2                    4 ~  3                                      Ei 3.6-86
 
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                                                                                                        /=/<'
                                                                                                      ,~/y Z,5 -yP,
                                                                                                      'W// <l-f7
                                                                                                        </a ~ 6-PW 6-/<8.
J, g  -)4-       /2 Cl<'r ~) </              cy                        7    (c-6~                          S. C'~J'7 P>3)   -'/                                                      Ip/(       3. 6  F7 4.wc u (z)
Y                      / /g 3.        6    -8'0
                                                          //
    /o c'            P.vtrCg (~)              t                                3,   <<.    ->s g-'krCIJ (i)         - il                  '//<'
                                                                              $ .4      - 7'2
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  /'( -k            /?.'wan      g / j) -4/        /              r  i(l-   Z. 6         -P'/        /=/g z.6        -5              )
  /> 6-Sl          g.~i~-i gr) - /                                               / ~ 7/             i(-/(~
l-wed(l)              -  l                                                                  3  <l-SJ p wctu            gJ) --'~     6                            g~.      6        Zo  /-g~ >.~-SJ
/28'-7              ./2~cc            (  Q cl  '/                J=J/      Z 0        -Z'/
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CS'/
            ~                    <               g ci G      3 ~  ri              FA. 3.4-s/
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  /wc //              R ~C        la  (X) -~r                     /=(r~       3    <'.
V.S 6I              a. c uri'- ~/
 
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                                                                                                                                        /  'o~~/A ~i=/ >7- <
                                                                                                                  /O~/IZ (kg)      IO4  r+nJ        z  CC~17~
o/z.
0-/SO            '/Arg      L7E.'-./&AHTlo/u                  Pits die. 7wvs/           vs T'/~6    gp~k            /=/( u/ZI=
gQ/-3f                            A.<htC <t    (33 I                                          Fi/      3. 6 7P gag          -Y/                e~cu        ('z) -y                                      /-/G. 3,$        - /7                    5  ~  6  -50 Cj          8-VYCu PS/         -'/                                     Pg~    Z g      77~                              5'Q i2'~dv                                        //h  lt P/G-    3 6     -73 I'
P. mc u P "g - L/                           y                      E.6    -73
      /I ~                                      c ~
(r)-y                                          P/c 3.6 -73
      /   '\
(~    7          g~CQ        /Jg                            3"            F/g 3< 6 )z rz) -y                                                    3,4 -73
                                                          -I/'2.<ucu J~        /
      /      h  f    g    Q          /Z. kv'   u P 3)              ~/                                                        Fi'/         3.Q        -So
      /.>        -  - g (/             4 ~cu (<)-p                                                  F/C-      t 6-7'Z                      ~~6      -j3 R~CV//'g) -l/                                               tl- J Q    3. 6 -7 3                              )~g
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      /   ~ I/                         w s ('~ >) - q                                                    2. S.S            KI/f~        3. 4      /r/W
: n. f        c)-y                                                                          //I
    /  /,   (/                        W. /'-.h-)                                                              gC                  /t l 3 '/''3~~                        W ~(Wg)          -y                                            P. 85~ b. F3          f=/G
      ) 3'J-                             rh5(p.a)                                    3"                        S.a            //C            3  ~   4    /t hf I
      /   <tC                                                                      3"          P/b.     3, $    -9~                                       )'gz l                                                                              3"            /=/C-    3. 6    -93 I a "/-~/I                                                                                  F/6- 3-6          -8/
3 0      87        /   /g       g.g        -yCz t
I/ 3 ci                                                                                                                                 5t o
                      //c'1/
                                                                                                              ~    4 -8'7
      /3~/         - ~,~                                                                                       3. < -8'7
                                                                                                                        -P7,
 
BURNS l1      L ANC'Z.
                                                                  < 4                            p8.bE        3  o7"-        7 I'7Am -r~o          Cga'         DE=-/&ed-rv osg    ApE p/8.      nIIX Wncc PgPS)                      /CWng      8 ac~ vw 8n Es~K                                                        yn.gg ys'7rlE >PPY$                        I l/g gA./;
    /r// -/0          H~     Cio)-a.                                    7. aF                                3  '-
                                                                                                                          -k'A
    / j('/ ~ //       +5 Q/v) -crt                                        7i cled                                            /
J <]l  -/g                C/o                                          7. A'E                              3. g  -        ".
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    )'g/                 m(       iog -@                                aa, 5/                         //~~     3.g
    /'// -Ql                                                              gz. bi                                  7.g -C/3
  //a      3  v        P.~c, u      pig-y             y" II r,s -8w                            p. g-5-/
i/'al                  4 ~cv Ps) -'j                  Cg II P/6 z. 5 -g5                              z z/
                                                                                                                          - 5"/
l2.~c< (I) -Y                  ~                                               H/6      5,  6 P w c u C 1)      - <I                     .Hfi    >.6      -Z~:
    / r/v -3'/            Wi~c~Ci)-Z                                       Js. >'/
  /9'i'- c"          0    /2 mc'.u(rQ-2                                    /3. Zg                              3'4    -5/
                              ~ca(zg -~
g-WCVglg                      g II
                                                                            /3  ~ 3 <r                          7.4    ->/
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II 4h/Cu(gg-         7                             3o.a        S'3 3,6   -5/
                            /Z K/4'    ( Q) 2                                        3</
lr'/'/ -   ? ~          f~                                                3o as f (/</                  W ~ c (~ g~ ) -     >~                            )3                      l   /CI-        I
  / v'/- s.b          '     1~+ ~ (g)                                     )5      3V                I
                                                                                                          /(      3.6    -5 j
  //''/ " 5 /             gwc u (Xg -3                                       3o iQ ~
I I P/6        3'-:/
  //y- 5-8                j2 ~cv (rg 3                  y II i'3. 2                  i  /C.      Z. 4  -5/
  ]<jr/ 5                R m< u (a.g-Z                                              i0 9'o.
do 'Fp-8~@ u $ 5$ ->                  It 3o. /B o 3 PH/E    3. 6
 
Rl IRNR  4g~
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4)~E  GE5'>&N87IOW    PIPE'j~                  ~O~i= CWII S)                        PHg Pl6v46'ac+T7oau 7+@.vsT vs'J~E                  /=i( un -F
                                                            $ 4 -9'7
: 3. 4  -'77 3<6  N7                              o
                                                              ~ 6-~i r                -. o -$ ~
                                                              ~ 4-9T 3.6 ~'7
: 3. 6-gv 3'i
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                                                            "~ 6  -07                  C7  ~Q I
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                                                                              /IC            5,.c I-!(~
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                                                        ~33. Jz
                                    ~ II gal-  C                            p c/
3~5 ~                                     I'/
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c e 2BZ RE/'.cToR VE55EL                29j          23I IA5. NOZZLE NSP,
                                                                                                                                            ~~P.WSPho Bl.l
: 2. WSOZ)-+
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e m-Rv-2.A                                    MS-RV IA5-RV-IA                                            "+A'45 eve                                        RV-3A
                                                          '22c                          ~+30                  C32g                            PWS 31-Z,
                                                    'L'L5                                                                                                        I
                                                    .224                                                      Cubi PM5 sl 1            PW5 31-+            PWS BI-S C'3l                .22'l 6%4 227
                                                                                                                                        -. 228 225 O'L
                                                                ',g.CITRIC;T~OR I                                                PLOYS I
                  !. X-IRK QS-V-2.2 A                            .22K
~ ~                                                                 . ~ .".
S  'Bl-5
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          .I j    i
  . ~<.$ g,$                      PWS Bl"Co                  220 2fl                            MASIII"OTN via  213                            POOL 1 C POMER SUPPLy STSTEg
                                              .                                   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 ')En''l                        3.6-12 NUCLEAR PROJECT  ~       2 LOOP A
 
N~                                     ~ ) ~
aSI Rf ACTOR VE5SEL MS. NOZ'XLE. NBS
                                                                                                        ~           +PW5 250 24't 3'L-I 2.l MS(l)-+
~ ~                                         N5-RV-2. B              .. hhS-RV-+5                        RI4 6 Ih5. RV- IB                thS-RV      3B          MS-RV-5B                        PW5 'M:2.
HZ
                                                  -to>fg      A8;.:                            642 24I                        435 $  '1        logos                ..6 44                        IO RCICIIS-+
:/OR CoNX SEE      FIMI PW5 24og M= I PWS 32.-+                   I %9.:.
                                                                                        .64l ~
443
: 2I4A..
                                                                                                                            -  PW5 52.-3 243
                                                                                                              %46                  I
                                                                                                              .241                      244
                                                                  '                                               24'>>            '245 I
i I
i X-ISB FI.oV4 P,GSTH.IC fOR
        'tA5-V-2.'LS lip: ~ l,t I
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
                                                                                                          . Ms RV-Bc 7                                                                                    PWS 33-2.
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==
==SUMMARY==
OF POSTULATED PIPE BREAK LOCATIONS CIRCUMFERENTIAL BREAKS LONGITUDINAL BREAKS Vode 254'ode 256'ode'59'ode 261 Node 265'ode 267>>Node 268'ode 270~dO-Vode 654'ode 255" Node 260-bl Node 266 Node 269~uaoE 4$~uonedS/.4odE 4'Jg fpoDE 4/7:~rvo<E.4~~'r1ASHZNGTON PUBLZC PCNER SUPPLY SYSTEM NUCLEAR PRQKKT NO 2 HAIN STE'Atl LOOP C PZGURE 3.6-14b REACTOR, yE55EI lAS NOZZ,I g g3P use, PINS 3+-!l(-ISO+'eS-V-22P PWS 3+-4.%13~~14 275',l28 I p 282, pMS-RV-+O (P~S3~+IAS-RV-aO.c 58.451 lg f MS-RV-Z,O FLDW ZSTRIW Gil MS-RV-!O r S.P'%MS 3+3's~CM.w" S.r,)78 GGl gf:-"~k!r'283 re.>~s'l84+i-,.;.g DMSO)g PVIS 3g-2,<285~~<%1C r e rr Jrrrrlrr<4
OF POSTULATED             PIPE BREAK LOCATIONS CIRCUMFERENTIAL BREAKS                                 LONGITUDINAL BREAKS 255" Node 260-Vode 254                                                  bl 256    'ode Node 266 Node 269 ~
<<4a u 0)--)'g$IIIIIOTL'IPIILIC POWER SUPPLy SZSlgI~'I~.t35'I FAR.PrrAJFf T AA.ical LOOP A!~OIrr:ynlg 1
uaoE 4$ ~
WNP-2 Amendment NO.~89,  
                                      '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==
==SUMMARY==
OP POSTULATED PIPE BREAK LOCATIONS C'IRCUMPERENTIAL BREAKS LONGITUDINAL BREAKS Vode 272 Node 275'ode 277 Node 280-Node 282.Node 283'ode 285'ode 286'ode 288'ode 290~~60 Node 276-Node 281'ode 284'ode 287'aoE 4s P.AtodE SS7 mom~~WASHINGTON PUBLIC POWER SUPPLY SYPH NUCLEAR PM'~NO 2 Hjtn,li"l".i"=AH LOOP 0 FIGURE 3.6-1Sb I21-~l Pv/S 21-(12.RFW(A-6-I26~129 I3(a 132 137 I>I V~S PWS 21-p 12-RFW(ll-0-135/-13<PWS 21-15>Xl~At-123 II I 114-y g II 3<112-133 PITS D-5 X-11A RFW-V-IOA 85 PWS 2.l-I+-'97R+PWS 21-16-IS-C9, PWS 21-IS-RF W-V-II A h/1 TEST CONN,~Pwh 1.1.11 PWS 21-13 65 61 PW5 21-2.I 2+<<18 Rf D.12&125 IS RFWII).+PW5 U-+12o-12.RFW(l~-+PWS 21-G l'AS 2.1 IO.-Ilo PWS?.l-~PWS 21-11 lo I q IOO-lo5 lu~Iol PWS G-8 104.Io'I P~IS 21-q IO8 g~lo2 IS<12.RF.D PWS D-I2.~k I,.IP P@E$,/PLY SYSTEII ggg0$'I2ji<-" FIGURE REACTOR FEEOI!ATER
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.
(<IIIE'A)ISOPETBIC BURNS AND P w.o.r Orawii h/N P-g rf~t tutor tu 7 i5'Fz/OS(</-87'->lI'2FH
Node 282.                                             AtodE SS7 Node 283
/OCr'7iOIuS
                              'ode                                   mom  ~~
(/ZC ul"IIPDIZ.
286 285'ode
/=~7 IR-(8~iS gS 4.o~d l 7 uDirva/PrZ Z@gC'edES/>'i/JJ 5/w(~//7 la v 1/(Z Qo/'34 (t 3v IOO/3w/0 J/37 IOv~jg'7/Oy'l oW/o7,/C(.lie/33'7>~mES g7)l3/Qg IO</o-/3y//7/c/9'78-.-JQG A r QZ/t~(/<l(~g1V~I ug C((.NO~i=~5ur(rl~L'/
                                'ode 290 288'ode
>y (""'OC (.(i~rr PER<r0nfi rv~i~r&#xc3;z.t
                ~~60 WASHINGTON PUBLIC POWER SUPPLY         SYPH                                                 FIGURE Hjtn,li"l ".i"=AH LOOP 0                 3.6-1Sb NUCLEAR PM'~         NO        2
/Lir A)./IC<gEr r~O.Form BR8002A (9/82)200M  
 
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
                            /33'7>
  ~(/ <l(~g1V~
5ur(rl~L'/ >y I ug C((.
("" 'OC (.(
NO~i=~
i~rr PER<r0nfi rv~i~r&#xc3;z.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
            ~~a,um/-~rz,            e~zwl 8~/JfC-L,mrpgyvarmp/            ++Sp~
                                                                              ~a    ~~
15    /
                          / yr>>                                      /V7                      ))3    A
                          /5G                                          /6g                      Qr
                                                                      ]jr J
            //v                                                                              /7F i"-
            //5            /7o J L/g            /7~
                            /73                                        /7J I "7$
                                                                        /  7 c/
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                              /zs)                                      y4/    =
          /67                /12
          /rg>>,                /54                                      /9'9
                              /5o J~w <                    '3
                              /C
          / 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. ~
7"  ~    O'Vr.'C.r': NJC Form BRSM2A (9/82) 200M
 
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
                                                                                                                                        -42l I
~  f.'-;.'                                                                                                                                      422
                                        . dgwcu(s)-+
                                                                                                                                                                >>I h                h t                  J SADDLE.
EhugnPFR.. -.
I.:.
FOIII +HT..QP-h
                                                                                                      't th
  . 1'. <<(H
          'i:.L',I'.
I'Fl..h<<<<"';.
I. ~
                                  ~I'g-,"':.'. '~
j 'tP J,<<>>'t>>h "hah'i>>t>>> (jjr htgtt tt'jt
                                                              ~                  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.
(%5    3&-.S s ~
PWS    M-9                                                                                                                                                2W RRC(Zl-qs s
              )th      ~      347
                                                                                                                                                              ~              j
                                                                                                                                                                              /'-~~-=''-- ..
I3:
345
                                                                                                                                                                              '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
                                                                                                        ~
3+
                                                                                                                                                                                                        +~323 s
I .
                                                                                                                                                                          ~
                                    ~~~54                                                                    '                                        + EOR l).I        3IS                              3H I'
N  ~
                                                                                                                                                                                                        'EI.
                                                                                                              ~,I
                                                                                                                                                                                            . 2. ARC(C)-AS r                                  . 353
                                                                                                                                                                    ~      -  ~    317 t) Il~ ~    I)-
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


185-PWS 28-1 16'i'FW(ll
NOJP-2 Amendment No.
+-lel~Ieo 92 I EPWS 28-8 190 NOD t II 189~PWS 28-$17a-PWS 28-2.I82 ITT s)C I CS-/PWS26-ts 184 12.RFWAl 4-pws'te Ib Isg~~It 9~Ill~tt 7 199~150~Zl AF wo)-$~146 ISie EPWS 28-+~17S-174 ISa4 113'72 IL F W-II-I La l/~ta51 COIIH L~~-153 pA 111 170 PWS 28-11 XL75 L3'3 FIF~-V-IoB EPWS 2.8-'I 15'c PWS 28.12,~Lll+18 AI 155 151 IS ISIS.Isiti REtl I~ICs I It 2 I5'j It.e-PWS 2II-I7 PHS 28-Ih?+its Rf.L3 PNs 78-to-PWS 28-l4 14l 14l 12.RFWtl).+PWS 28-8 IC,S~ICQ IC 4 PWS 28-1-Pws i8-8 I.'ASI;ig ji'l'lyg",supviT sTsTER";JNW,',~IO-.:,2..
APJ'i1    1980
REACTOR FEEO'sER (I.lllf 8)ISRLETRIC rICuRE 3.E-IFI:
                                                                                            +y w~   ~
MNI SunnarZ/OS/a--~+y g P>p)=r rZ c-X~/-.-a c.)>Wgm.S~~a,um/-~rz, e~zwl-8~/JfC L,mrpgyvarmp/
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
++Sp~//v//5 J L/g/67/rg>>, J~w</QlJ/F7 15//yr>>/5G/7o/7~/73 I"7$J~/zs)/12/54/5o/C'3/ky/gp/~ra/9'/~a~~/V7/6g]jr/7J/7 c//8'a y4/=/9'9))3 A J Qr/7F i"-)re Sita Cr re/~ulnar C PC err,~7"~O'Vr.'C.r':
  'WASHINGTON PUBLIC POWER SUPPLY SYSTEM                                                    FIGURE REACTOR HATER CLEANUP                      3. 6-18c I         NUCLEAR PROJECT Bo~ 2
NJC/2 fd<TOf//E L=O~+TFR (sJ Ju E/2)j/;.z/c.~Form BRSM2A (9/82)200M PW5'5Co-I9 9+5 BG 2.0 PN5 3Ej-IB t-PW5 aG-Iq h PW5 5co-IQ 425 Qg4~FE 3 AIIEIIDjILHT hO 4 Scg!Apt:Ll 1MO I*L 4ZS RWCLI-V-I+I9.PWS~S-IS'.42o.PIjAjth 36-IIt't I~f.'-;.'.dgwcu(s)-+
 
-42l 422>>I t J hh SADDLE.EhugnPFR..
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
-.I.:.''.!"-FOIII+HT..QP-.1'.'i:.L',I'.
                @jog'33 Nod e  3@A      ~              Noae 379 Node 391" Node 394 Node 395>
"';.~I'g-,"':.'.
Node 396~
'~<<(H I'Fl..h<<<<I.~j'tP J,<<>>'t>>h"hah'i>>t>>>(jjr~htgtt tt'jt j<<hJ<<t.t it h't th I IhlASIIIHGTOH PUULIC PNER SUPPLY SYSTEH HUCLEAR PRMECT HO.2 REACTOR MATER CLENIUP ISOHETRIC FIGURE 3.6-18
Node 366                                                    ~Le~
.C RWCu(3)-+349>So PCS 3II'I-IO PWS 3S-Z.398 39 I PWS 3r-3 4oo+CHEN'.CLEANOUT og~~36IIS 39l 3894, 4 fOR CONT.SEE 3 g 3p plwcu-y-IoFo VAl VE SAFE END$74 373 e.9ICL1+3uI 332.Il+R.RC(+)-45 388.387',+RRC(+)'-'ts
Node 367~                      L~
)83 31 Fo 377 gylC'U=4-RRC-4 WS.5F,5 P PWS 3I-R4 672 PiV5 86-I 3'les<-FOR CONT.SEE FIq 3 g 3'I4'RWCLI-V-gO RWCU-y-IC2.
Node    368'!
~C"='A.CLE<<<<T 3qZ~-3SI 8 PW$3G i.=~I(, VALVE 5P,FE Egg~51l RC-g5 3ll Sle...PWS 36-6 PW5 Mr26 PWS 36-1.3ISA (%5 3&-.S s~?ARRf (2Q-45.3I'3 PWS M-9 s)th~347 345 4~2.RED.344-=I.2.RRC(Q-WS%c.b'35'I.3c'X.".~358-.-3go RRC-V-Sl B 331'55 RRC.V-52.B
6
~~~54~~~Is i~s l I''N~~,I.~: I''.~" 325.'32o-..3~~.324~3+I.~323 3IS+3H'EI..2.ARC(C)-AS~-~317+EOR l).I PWS 3II I2.2W RRC(Zl-qs/'-~~-=''--
                              ~  Zr              O'T 7
.....84 5~.j y~P"g EDRRC-V-5I+
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
33I.PwS N I3:.33(3:-33o 331~'..92II~gee-v.ps 355.'27 334,-..933 r-.I)+'EOR(41)-I
 
.353 t)Il~~I)-~'QKIIOHENT NO, 9'I Apri1 1980 E F IFsOR1.6-IRI IllIlPIIII flilsis;[gsll ssi assi-lsistss
Amendment Na. 9 Apr>1 1980 RE,ACTOR V E.S SE L.
)~a.,]~Q~ILlKCJiIILO~~~I,:,'~j
S.L.,t". N.QXX,LF WII 2" SC.H.,BO PiPE ,
.,~IIggOtI VhTfR a~aIRW Iri~t.IRIC NOJP-2 Amendment No.+y APJ'i1 1980 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 I NUCLEAR PROJECT Bo~2 REACTOR HATER CLEANUP FIGURE 3.6-18c WNP-2 Amendment No.PQ A pri 1 1980 SUHi'~ARY OF POSTULATED P Z P E R RFAK LOCAT TONS C:RCU!!F.".R.".NT::.~
(SY l'-'.E3 r
~REAKS@jog'33 Nod e 3@A~Noae 379 Node 391" Node 366 Node 367~Node 368'!~Zr L~6 O'T 7 Node 394 Node 395>Node 396~~Le~8/@05 pS7 LONGITUDINAL BREAKS XT (D RASH&lGTON PUBLIC POWER SUPPLX SXSTKN NUCLEAR PBC:.i!~NQ 2 REACTOR WATER CLEANUP PXGURE 3.6-18d 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-, (SYl'-'.E3 3Ic r ll P l (l)'-+2.EIQ~3l&Gl2~.hlO!//~l'L())-4G~3og 505-30l 300 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 STANDBY LIQUID CONTROL ISOMETRiC FIGURE 3.6-19a W.O Orav~By Yi(t(5~~'t%1 y A 4 I rsUB BOC ply (, (5+38$i~I yr(r)-VS 2=7 Iy)q7 2'~y'SC c(IZPy6 3o!100 (~<<p 4/gIy'I/2-ulrr C'/"-~)5r c-/'y 4 g/C C/./~z.'<g/Pl 7'form 8R8002A (9/82)200@i g/.E Or ai iaMc hlA QQ Qr I g g5 dQ~~QJO 3Q 3>l 0~///A!i~-fb/u I'4'l'.6/c j'o&#x17d;~5'1/',.~/, y+y'.-<i/~LICE.EA/l
3Ic ll P l (l)'-+2
(~g/upI2~/4 g qg/~~~77/r,i Form BR8002A (9/82)200h1 F H R,-V"+l A RHR-V-Ill A 39 27 3P$'22 Xl.l9-25 24 Pw5 5-2.I+RHR,(l)-+RKACTOR.VKS5E.L, RHR.NOZ.jL.E iU G WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 RESIDULAL HEAT REMOVAL LPCI MODE (LOOP A)ISOMETRIC FIGURE 3.6-20
                              ~
&#xc3;NP-2 Amendment Ho.P V4 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'ode 18 Node 20.1a,~Node 27 4Ie4t4l~Node 26''Iod e 28'ode 31~HASHXNGTON PUBLIC PCNER SUPPLY SYSTZX NUCLEAR PROJECT?K)2 RESiOuAI.HEAT REMOVAL LPCr NOOE LOOP A FXGURE 3.6-20 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 NUCLEAR PROJECT NO.2 RESIDUAL HEAT REMOVAL LPCI MODE (LOOP 8)ISOMETRIC FIGURE 3.6 tRJP-2 Amendment No.~+@, STJVJ<APv OF POSTUTATEO PIPF.RRFAK T.OCATEOPS
EIQ 3l&
".Ft'.RE:!TTI XT RP,F'AKS Node 33 Node 34 Node 36 T,O'".,r,.Tl J,)T.NAT, RP.~:.K.'h de 35'ovie 43~Node 42~'lode 44~Node 47~!MSHZN~iN PUBLIC POWER SUPPLY SYSTEM e'UCLEAR PROJECT 80 2 I RES/DUAL}IEAT REt10'jtAL LPCI MODE, LOOP 8.FIGURE 3.6-21b 1 Pl(l)-+9 RHR," V-111C PWS 3-l S5 SZ 50 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 RESIDUAL HEAT REMOVAL LPCI MODE (LOOP C)ISOMETRIC~FIGURE 3.6 I 0 NNP-2 Amendment No.~+SUfft!ARY OF POSTULATED PZPF.RRFAK LOCATIONS c"'<1P~bc'NIT>
Gl2~.
~T bRP I x'tp d e~>9~Node 50~Node 52~I l',ONClZTUDENAL RRFAKS Node 51 8a&~Node 59'.lode 58~Node 60 Node 63 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO 2 RES IOUAL HEAT REMOVAL LPC I HOOE LOOP C FIGURE'.6-22b Amendmen Na.9 APril 1980 Q3q (o7 l2.RHR(l)-+5 l i/ALVF SAFE END RHR,-V-BOA 72 (uSA P.4 RKClRC.Pe+0 Dl&CHMGS WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 RESIDUAL HEAT REMOVAL SHUTDOWN COOLING (LOOP A)ISOMETRIC FIGURE 3.6-23a 0 l l , A
hlO
!JHP-2 Amendment No.+D+0>OS<UL-'.TED
                          //
>>PF.P.REEK LOC'.~O'1S C"'!FERFVT~AL RREiXKS LOF1GZTUDI 1AL RREAKS~Jode 65~'1ode 65K~.'1ode 65C'i'bande 65B'ode 68~Sod e 65G~!Inde 66~Node 67~.'1ode 69~3atxe-~A/ODE-To ilASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLZAR PBOJECT HO RESIDUAL HEAT RB<0'lAL SHUTDOWN COOLING LOOP A FIGURE 3.6-23b 24 Zec.ia.c.~VMS OlscvA~6'tie 7<C 12" RARE)-CS 74 iz HR+(IZ.&72+-78~vv'H R-V-50 6 VALVE~F2 EHa-WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO~2 RESIDUAL HEAT REMOYAL SHUTDOWN COOLING (LOOP 8)iSOMETRIC FIGURE 3,6-24a I  
                                          ~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
                                            &#xc3;NP-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 26''Iod
                                '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==
==SUMMARY==
OF POSTULATED PIPE BRFAK LOCATIONS P CU.'5FE RENT E AL BREAKS LOI'lQ XTlJD CHAL B REAKS Node 72A~Node 72B~Node 72De Node 72E~8ode 72G".'lode 75 Node 72C<llode'2F~Vod e 76 WASHINGTON PVBLZC POWER SUPPLY SYSTEM NUCLEAR BOJEl.'RES IOUAL HEAT REHOYAL SHUTOOkltl COOLING LOOP B FIGURE 3.6-2<b al R,HR-V"9 VAl-V6 5AI'"E END I~~RHR-V" ll3 24 RE.CiRC.PUMP GUCTlOh4 796 79K'79 D jar.79@COhPI->Nu<TlOm OP~576M 79K'W 791'-i WASHINGTON PUBLIC POWER SUPPLY SYSTEM'UCLEAR PROJECT NO.2 RESIDUAL HEAT REMOYAL SHUTDOWN COOLING SUPPLY ISOMETRIC FIGURE 3.6-25 Amendment HU.~84 LPiTP-2 SUhifQRY OF POSTULATED PIPE BREAK LOCATIONS C I RCUHFEREHTIAL BREAKS Node 79A~Mode 79E+Hode 79G~Node 79H Node 79J~Re-SQ e Node 82 LONGITUDINAL BREAKS.far'h Hode 79F i Hode 79Ii WASHINGTON PUBLIC PCWER SUPPLY SYSTEM NUCLEAR PROJECT MO 2 RESIDUAL HEAT REHOVAL SHUTDOMN COOL IHG SUPPLY FIGURE"'.6-25b REACTOR.VE,55+.R.t lC NOT:LEL N f C2$~G'900 T$5 SMALLl6N&uE l 1.8, WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 RCIC RPV HEAD SPRAY ISOMETRIC FIGURE 3.6-26a
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                   3.6-2<b NUCLEAR   BOJEl.


Amendment No.~WNP-2  
al R,HR-V"9 VAl-V6 5AI'"E END I      ~  ~
RHR - V" ll3 24 RE.CiRC.                        COhPI->Nu<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==
==SUMMARY==
OF POSTULATED PIPE BREAK LOCATIONS CIRCUlIPERENTIAL BREAKS Node 622'ode 623'ode 625 Node 626~LONGITUDINAL BREAKS Node 624 g py yE/PO Spacey~WASHZNGTCN PUBLZC POlfER SUPPLY SYSTEM NUCLEAR PROJECT EO 2 UTDONN PZGQRE.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 NUCLEAR PROJECT NO.2 LOW PRESSURE CORE SPRAY ISOMETRIC.
OF POSTULATED     PIPE BREAK LOCATIONS CIRCUlIPERENTIAL BREAKS                     LONGITUDINAL BREAKS Node                                        Node 624 623 622'ode 625    'ode Node 626         ~
FIGURE 3.6-27a 0  
g py yE/PO Spacey   ~
'WNP-2 Amendment No.April.1980 e SUHtIARY OF POSTULATED PIPE BREAK LOCATIONS CIRCUllFERENTIAL BREAKS LOl'7G IT U D INAL B REAKSllode Node lode Node Node Node 1>>2>>4>>5 0 6i 7~Node 3 WASHXNGTGN PUBLXC POWER SUPPLY SYSTEM(NUCLEAR PRO3ECT 5G~2~LOW PRESSURE CORE SPRAY FIGURE 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 lE HPCS (l)-+t4 WASHINGTON PUBLIC PONGY SUPPl Y SYSTEM NUCLEAR PROJECT NO.2 HIGH PRESSURE CORE SPRAY ISOMETRIC FIGURE 3.6-28a NNP-2 AMENDMENT NQ..MD A  
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==
==SUMMARY==
OF POSTULATED PIPE BREAK LOCATIONS CTRCUMFERENTEAL BREAKS Node 9~Node 10'ode 12'ode 13 Node 14~Node 15'oge ill LONGTTUDINAL BREAKS Node ll'ASHINGTON PUBLIC POWER SUPPLY SYSTEN NUCLEAR PROJECT 80 2 HIGH PRESSURE CORE SPRAY FIGURE 3.6-28  
OF POSTULATED PIPE BREAK LOCATIONS CTRCUMFERENTEAL BREAKS                        LONGTTUDINAL BREAKS Node 9 ~                                       Node Node 10 12    'ode 13  'ode Node 14~
~~r NO M~2 ISG2 PV45 SO g~l9l..PWS SO-g." 2', 2'PWSSO C2 jp'2 I f2 t2 o I~)g 4~~.'eo.".;..
Node 15 ill    'oge ll'ASHINGTON PUBLIC POWER SUPPLY SYSTEN                                               FIGURE HIGH PRESSURE  CORE  SPRAY NUCLEAR PROJECT 80         2                                               3.6-28
t tel, l(k]f jt r o~4 RCIC 03)-'t 24 fV DIU5 24'RAOluS M lA5.214 6 r I~',~~2A'70l A-..Rclc+@..r,~"''2IIA" O-I 2.21I"-210~'214k I4 RCIC(lb)4r 203-'214 f Zog a J-.2o iZ--21.+02NN.2ol:.Z4G'.>~PIN-+Tt'P.+PLACES 2CiS.PWS y 3,.~py.g i I(GIOII PNLIC POUER SUPPLY SYSIEII RIIR CO.BEIISltlG NOOE.",:IIIUCLEAR PROJECT IIO.2 RCIC TUROlttE STEAtl ISONEIRIC FIG UP<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 NUCLEAR PRCLTECT 80 RHR CONOEi'ISING 1'IOOE)CIC TURB Ii'lE STEAM FIGURE 6~ob I~X-22.*~$4 g 54l l."1~': bjgkm,,,'le)!IlFf'5R l~52o.'<,lcr 8555-15::-
 
IS....:.1.-5'NSB)-.+,.'S SS=JS;5'ToA t WS 55-I t.'Sa2.RE,EL'555*r)$oA 544 ASS-I~I 15QQ li I'I SC 2.pws 55-$1 1 59 1~",ZIA50).+'.
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                                                                                                                                  ~ .PWS l9l.
2.IAS(9 L:" 517.W5 53-..6\Qw)v.,1I.;--.a.g]5'-,FVIS 55-.8 89+.;.555.5'ZS(9)~~.'.',pWSSS IS.-'=54K r*560 ,PW5 55.-IE3,.'5&'T ,...PWS 55-.II..'5E)5: '5'I)g h'.fall~'PWS 55'5 572.-~58%.~564~555.l WS 53-2-58r 58'I)~!)GTOI PUSLIC P0%II SUPPLY SYSTEH)'Lji.lit@LEAR PRQECT1 Il0.R NL'~Uk~r~lll STEAN VALVES DRAlhAGEl FIFIIC tsa~nGIC*!1 FIGURE.6-30' AMENDMENT NO.M~
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                                                                                            '214  f                Zog a  J
                                                                                                                    .2o
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                                                                                                            "-210~
                  .'eo.".;..                                                                                                                      2CiS .
t PWS tel, l( k            24 'RAOluS                        ~ ~
                  ]fjt r o
                            ~
            ~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.
                                                                                  .'Sa2. RE,EL NS-.V- )) XS
                                                                                                '555 I65(9I-+ '"
1 54l                5'ZS(9) ~
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                                                                                                                    ",ZIA50).+'. ~    ",-                     ~
                                                                                                                                                                  '\',
f'WS 5b-IL
                            ~
                                $4 l.
                                                    '=
                                                      ~ .'.
54K 560
                                                                                                                                                                                                  ,PW5 55.-IE3,.
544                                                                                                                            5&'T IS....:. 1.- r ASS-I~I 41 L       g''                                                                                               ,...PWS 55-.II..
5'8
                                                                                            ~
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54$ )
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                                                                                                                                                            ~  T SS=JS;                    L-1)   -,...541;M~""-                                       C
                                                                                                                                                          ~Fh>5I                   '5' 5'ToA                                                                                                                                    5S2 ..-
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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                                                                                .W5 53-..6                           .,1I.;-
                                                                                                            -. a. g]5'-
                                                                        . fall~
                                                                'PWS 55'5 572.-
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                                                                      ~~
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                                                                                                                                !)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==
==SUMMARY==
OF POSTULATED PIPE BREAK LOCATIONS CZRCUMFERE."1T
OF POSTULATED     PIPE BREAK LOCATIONS CZRCUMFERE."1T AL BREAKS
~AL BREAKS jane~Node 543 He+~K Node 550'ode 554'ode 556'ode 557~i3~55ft-So&~99 Node 560'ode 562'ode 565~Node 566'ode 568>>Node 569~Node 570~S~k~e~~Node Node Node Node."1od e Node Node 580'82 583~585~586 588'89>>Node 574'ode 576>>Node 577 Wc&#xc3;b~zZK Ne4e-5R'WASHINGTON PUBX IC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO 2 HAiN STEAN VAL'JES ORAINAGE PIPING PXGURE 3.0-30b h=P 655 PW5 52:II C(sS.+s2 RED...-900 Ti8, ,SQUALL tbWC uE Fl.t., REAClOR yE55EL IAS hlOt.gl.g Ng PWS 52:IO tI~!,i:>I;',-.'r!i(}r~.'jll~Cotpb~bblsA PWS 52,-;I 489~;5 2.MS/2)+I (,~((r'2,(I~((~~r'-r r l.4,.p I-~h~,+.~..~: 'g((PW5 52.-8...t&G>tOe(..PW5.52:.l.
                                                  ~
r I:.PQ8.l(t(I5'P BHD.PL'ATE C1~-!-.4CB(~~...FPc.'<<pP(-'~FQ jc(pets-s2-t'-'.-~: aal~~kbg.-t~F(V';FC FP v(~~N p Cr'g.AV5S2 Ir.'.-'.=;:.z'.Ms[n.)-
jane~
~~S&2M~h'f78 r~h~~2.'MS(5o)-y MS-V-2.'('~.(~h'(~~~,2G MS(l)'-q.pg(.,'.~'i'I.,"I 7~~h 1 ht(.','(((.~t I>>pl (..r.'h~\l','tt'r+~-(*hr~t+Ih>>(h I;l (~~4(g&..>>(, 1 CKQi>3:9+EOR(4l)-I-"f,~h r.85 r.~g)gt=HD 4o84 gag'>>~A~&t8.I 48'SB'"~'-..X IA502)-+Ch3y t Nk'
Node 543 He+~K
'I NNP-2 AMENDMENT NO.~
                                'ode         So&~99 Node 557 ~
i3~55ft-560'ode 562'ode 565~
Node Node 577 Wc&#xc3;b~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
              'jll                                                   PWS 52,-;I             ~  Cotpb r      .
r.85
                                                                                                                                                              ~g ) gt=HD
                                                                                              ~bblsA                                    489                           4o84 I
                                                                              +                                                                                                gag'>>
I  -~h ~,
: 2. MS/2)
                                                                                                                                    ~;5                                      ~A~&t8.
(, ~(                                                                                     PW5    52.-8...t                                                    I
( r'2,(I   +.~ .. ~:                                                                         &G>tOe( ..                                                                             48'SB'"~    '-.
((                                                                C1
~
~
                                                                          ~  -!-.          PW5.52:.l.
X IA502)-+
    ~
r Ch3y 4CB(~                                                :. PQ8       l(t(I5'P I            ~ ...
r                                                                                    BHD . PL'ATE FPc
                                                                                                                                                              '<<pP(-'~FQ r                                  g((
jc(
l.r                                                                                pets-s2-t     '-'.   - ~: aal~~       kbg.       - t       FP v(
4,
          .p                                                                                            ~ F(V';         N                            FC                                       ~     ~
p
                                            '(~
                                                .( '                                   .AV5S2   Ir .
Cr'g pl                                        '(
                                                                            '.-'.=;: .z'.Ms[n.)-                                                                                             h
                                                                                                                                                                                        ~                                ~ ~
                                                                                                                                                                                                            ~
(
                                                                      ~~S&2M
                                                                                                            ~h
                                                                                            'f78                                                                                   ,2G  MS(l)'-q .
l', 'tt'     r+  ~ -(                                               r ~
h
                                                                                                                                ;l  ( ~   ~   4(g& ..>>(,                        pg( .,             '. ~
t+Ih>>(
                                                                                                                                                                                                                'i' hr                                              ~          ~
                                ~                   h 2.'MS(5o)-y
    ..       r.'h                                                                                                                                                                                         I
                                                                                                                                                                                                                  .,"I
        ~       \                 I                            MS-V-2.                                                   1 CKQi > 3:9       + EOR(4l)-I                                                                                                                   ht(.  ','(((      7  ~ ~
h 1
                                                                                                                                                                                                                . ~ t I>>
                                                                                                                                                                  "f,~h Nk        '
t
 
'I NNP-2                     AMENDMENT NO.         ~


==SUMMARY==
==SUMMARY==
OP POSTULATEC'lPE BREAK LOCATIONS CIRCUMPERRNTZAL BREAKS<<lode 663~e Node 676."lode 677>>Node 678 Node 679>>Ne4e-~8 lode 668~&e4~r&9.Ne4a-6-2Q Node 683~e9-Node 690'Inde 675A>><<lode 675B>>Node 695>>moo~4 f/'gDD 4 gW C~ye)E dyad moaE Cga F goose 48~C mone gg~nrOOF A/3 Q~o oi--gF8 C WASHDCGTOH PUBLXC PCRfER SUPPLY SYSTEM NUCLEAR PROJI~RO 2 tlAIN STEAN RPV HEAD VEi<<tT FZGURE 3.6-3kb  
OP   POSTULATEC'lPE BREAK LOCATIONS CIRCUMPERRNTZAL BREAKS
.7E hlOKl ll'q'110.2.SCH.Ca RPV E3RAIl4 CONNt 4 t h-4II4.%lb=....RWCU-V" lOl 41'2" 41ht/)~~I l..--.:io~P U-V t~.j".:WQ: 1 r or'r t I Il v".'t."-.'-.lib;2.RRC(SI3=q
                <<lode 663 ~                                   Node 676
...,'-,::" i t'.I-'-1't.".g P.-: '!>-I a~f~th I,~kg'~tg~ip 4I2A-.405 PttS t-'tt 40ls-.'os.-..1 4 Rv4cU 8)-4 4't i~t, t)IC.lK." 1't QUQp~',!.-.G4 RED A~".;'-':.Qijr'',::
                                                                ."lode 677>>
j+~4'RMby."i Ii~418A-xaac@$-~410 41!QOQ'ril 4'~4'VA.I:.:SRG(5'q41S.(Pi(l'1.-AS RISC,.Q-t 5 Sl.l~'/RED og I.Co"PIN SNGvC';I: 'Fo:.9 E.T.'9 t.gl-le,~$1R'H.]SVSTElI s'PICEEAII g@PEq.Ig..j t~\'i f t h XL iS 11.~.-"1ol r.'aM*'(4-+g~=;~TO&t C~?RRC REACTOR PRESSURE VESSEL ORAIR lsuIETRlc'r h FIGNE.6-32a
Node 678 e                                          Node 679>>
%.0.Oraq'y Thfe.~f, INC.adell, N J.ok No.Sheet ppro ed ge No.Con on S t-~O.&~n4urc Sv'/i"/WN Y 0/->>r>>'>><8%x'2"w 4 Or"+rigwi r r/C,C.ui~(SFW2.
Ne4e-~8 lode 668   ~                                   Node 683 ~
S~~~~5 g,E 3/oOE+00E AoD6 ai oo/=o'OOr=hJOOr." gp'OOE 717 7/3//o/rod IIZ v/0 tVA-'f)ill(r 7V&/"ri'-l (.f2'.C gE>caiir PrL/;>>vXi/=/t-></=.
                &e4~r&9.
A'riC~l"r fit~)l t r Form BR8002A (9i82)200M mZNDmmT Zp.9 April 1980 e gv i/2 qp..0 179 pgA 174 qS,)29 1BO ih i3$15 7A 7 8+r q p4 p 77 76'(7&~13 r 14 l0 1 q+~lo 1Z4 70 71 0~.b t~imp, IR7 y9)74 iP pS I77-l 2'20 LOOP C 170 11S ll t LOOF 0 LOOP A.LOOP 6, WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.2 iNAIN STEAM PIPING (LOOP A,B,C 8(0)INSIDE MAIN STEAM TUNNEL FIGURE 3.6-33a Cc  
Ne4a-6-2Q                                                       e9-Node Node 695>>
~'NP-2 AMENDMENT NO.P cP SU."1f&RY OF POSTULATED P'ZPR>PEAK LOCATEONS:.C::" FEB"T XAL R?EAKS: O."lCi i Tt JD 1.'1AL,~RR.":KS Node 7A w;gp',r 8llode'0 Noae 123 Node 124 Node 70A lode 123A Node 173A NI e(o a Node 70 NOQQ 71 4K3E4~lode 171 Node 173 Node 174 4 WASHINGTON PUBLZC POWER SUPPLY SYSTEH NUCLEAR PROJECT NO~2 HAIN STEAH LOOP A, 8, C, 8 0 INSIDE HAIN STEAH TUNNEL FIGURE-3.6-33}}
690'Inde 675A>>
                <<lode 675B>>
moo~ 4 4 gW C f/'gDD
                                                                ~ye) E dyad moaE Cga F goose 48~ C mone gg~
nrOOF             A/3 Q
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                                                                                        - 4II4.               1                                  or'r                                                    '. I
                                                                                  %lb= ....                   I
                                                                                                                ~          ~
t RWCU-V"lOl 41'2" 41ht/) l..- -.                                                               ...,'-,::"
4 t                                                                                                   P      U-V:io~                                      ; 2.RRC(SI3=q h
4I2A-.
405                              '~ 4 'VA.I:.
PttS t -'tt                                                                                                                                          '
:SRG(5'                                                                          t                 r 40ls .                                                                                                                         C  ~
t
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5' Sl.l 418A-
        't." .g P. -: '! - I ~                   1                                                                  ~
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                                                                                                                    /RED 4 Rv4cU 8)-4
          ~
          ~kg'~
f                                                                                                              og tg~ip                               4    't i ~
I 410                                          Co"PIN                                                                                h (Pi(l'1.-AS                  .
t,                                                                                            SNGvC 41!          RISC,.Q-t   )
IC.                                                                           ';I:
lK."                                                                                     "1ol
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A~".;
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                                                                                                                                                                                                                  ?
T.'9 t.gl-le,~           $ 1R'H.] SVSTElI           RRC REACTOR PRESSURE  VESSEL    FIGNE s'PICEEAII g@PEq. Ig..         j                     ORAIR lsuIETRlc                  .6-32a
                                                                                      'ril                                                    h
                                          +~4'RMby."iIi~
4                                                                    XL    iS 11.
 
                                                        ~f, INC.
adell, N J.
%.0.
Oraq  'y ok No.
ppro ed Sheet ge No.
Con on S t Thfe.
                                                                                            ~O.
                                                                          &~n4urc Sv'/   i"/WN Y   0/-   >>r>>'>> <       8%x'2"w 4   Or" +rigwi r r   /C,C.ui~(SFW2. S~~~~ 5 g,E 3/oOE       717
                          +00E AoD6         7/3 ai oo/=       //o o'OOr=       /rod hJOOr."       IIZ gp'OOE       v/0 tVA -'f) ill(r 7V&   / "ri'-l(. f2'.C gE>caiir     PrL/;>>vXi           /=/t-></=.
A'riC~l "r fit~)l t r Form BR8002A (9i82) 200M
 
mZNDmmT Zp. 9 April 1980 e
179                    gv       i/2       1BO qp.
pgA     174
                                              )29 qS, .0 1Z4 70 7A         71 2'20 7
8   +r       0 ~.b t~                                           170 y9   )74 iP pS   I77-                     11S ih        77                                                                  ll t i3$
15 76'(
7&~
q p4 p  imp, IR7 l
LOOP 14      l0      13                                                                            6, r                                                                        LOOP q+ ~
1 A.
LOOF lo 0
LOOP C
WASHINGTON PUBLIC POWER SUPPLY SYSTEM                                                                   FIGURE iNAIN STEAM PIPING (LOOP A,B,C       8( 0)
NUCLEAR PROJECT    NO. 2                        INSIDE MAIN STEAM TUNNEL                     3.6-33a
 
Cc
~
  'NP-2                                                                   AMENDMENT NO.       P cP SU."1f&RY OF POSTULATED P'ZPR >PEAK LOCATEONS
:. C::" FEB "T XAL   R? EAKS                           i
: O."lCi Tt JD 1.'1AL,~ RR.":KS Node       7A llode  '0                              Node      70A Noae 123 w;gp',r   8        Node 124 lode 123A Node 173A NI e(o a Node 70               lode 171 NOQQ 71             Node 173 4K3E4~             Node 174 4
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}}

Latest revision as of 08:50, 4 February 2020

Forwards Design Reverification Program, Vols 1 & 2,final Assessment Rept.Results of Program Will Be Presented to NRC in Late Oct 1983
ML17277A868
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Site: Columbia Energy Northwest icon.png
Issue date: 09/27/1983
From: Sorensen G
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To: Harold Denton
Office of Nuclear Reactor Regulation
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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>>

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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,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

LR 3 OOOC 880 xoG .OA I

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Al N

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

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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

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~ M~M '7%1~ h5 +~79,+~~@ ~Q Qg~ $=PP,Melo< 15 LcA'ITALO o~~ ~g~p ~i

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~<~~~~ 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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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.

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3.6-47'

I AMENDMENT NO. 25 June 1982 Cl) 0 C4 0

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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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AbKNDMENT NO. 25 June 1982 H.3 24 '- l 0" I g<g q

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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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NUCLEAR PROJECT NO. 2 PLAN 8 EL. 422'-3" .6.-~ ~~

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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

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MODERATE ENERGY FLUID PIPING SYSTEM NUCLEAR PROJECT HO. 2 RUPTURE LOC. PLAN 9 EL. 441' 3.6-0.'

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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, )

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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

/

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~

Uuncral Auutr aint Datd fSr 1 Bar of a Bustraint F C 2

(h restraint)

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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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

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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

Classification Line Desi nation Diamete Seismic ~ ~Qualit rc RCIC (13)-4 I E RWCU (1) -4 4/6 I RWCU ( 2) -4 4/6 I I 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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.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

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'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/.

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267>> rvo<E. 4 ~~

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268'ode dO-Vode

'r1ASHZNGTON PUBLZC PCNER SUPPLY SYSTEM PZGURE HAIN STE'Atl LOOP C 3.6-14b NUCLEAR PRQKKT NO 2

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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

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~k ggg0$ 'I2ji<-" REACTOR FEEOI!ATER

(<IIIE 'A) ISOPETBIC FIGURE

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)~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

~

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~ 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

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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 ~

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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+

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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$

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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 ~

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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.

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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

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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 +

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AMENDMENT NO. M~

SUMMARY

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~

jane~

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WASHINGTON PUBLIC POWER SUPPLY SYSTEM FIGURE iNAIN STEAM PIPING (LOOP A,B,C 8( 0)

NUCLEAR PROJECT NO. 2 INSIDE MAIN STEAM TUNNEL 3.6-33a

Cc

~

'NP-2 AMENDMENT NO. P cP SU."1f&RY OF POSTULATED P'ZPR >PEAK LOCATEONS

. C::" FEB "T XAL R? EAKS i
O."lCi Tt JD 1.'1AL,~ RR.":KS Node 7A llode '0 Node 70A Noae 123 w;gp',r 8 Node 124 lode 123A Node 173A NI e(o a Node 70 lode 171 NOQQ 71 Node 173 4K3E4~ Node 174 4

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

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