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{{Adams
#REDIRECT [[TXX-9109, Annual Operating Rept & Annual 10CFR50.59 Rept for 1990]]
| number = ML20217A967
| issue date = 12/31/1990
| title = Annual Operating Rept & Annual 10CFR50.59 Rept for 1990
| author name = Cahill W
| author affiliation = TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000445
| license number =
| contact person =
| document report number = TXX-91090, NUDOCS 9103080007
| document type = ANNUAL OPERATING REPORT, TEXT-SAFETY REPORT
| page count = 108
}}
 
=Text=
{{#Wiki_filter:. .      --__ _ _ _ _ _ _ _ _ _ - _ _
M
                                                                              ~
                                                                              ===                      Log # TXX 91090
_            _,,        File # 10112 (clo & Att 2)
                                                                              =          =                      10250 (clo & Att 15 7UELECTRIC                        ll$$3[(b)(2) i wunne J. Cahat, Jr.
rucunw sw anar,u February 28, 1991 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
 
==SUBJECT:==
C0HAhCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NO. 50-445 ANNUAL OPERATING REPORT AND ANNUAL 10CFR50.59 REPORT Gentlemen:
Attachment 1 is the first Annual Operating Report prepared and submitted pursuant to Specification 6.9.1.2 of Appendix A (Technical Specifications) to the Comanche Peak Unit 1 Steam Electric Station Operating License. NPF-87.
                            ^
Attachment 2 is the annual report required by 10CFR50.59(b)(2) for 1990.                                                This report contains descriptions of the changes, tests and experiments completed on Comanche Peak Unit 1 under the provisions of 10CFR50.59(a), including a summary of the safety evaluation of each.                                    Items in this report are referenced by their 50.59 evaluation numbers. This report covers the period frcm February 8, 1990 (receipt of Low Power Operating License. NPF-28), through December 31, 1990.
If ybu have any questions, please contact Mr. Jorge L. Rodriguez at (214> 812 8323.
Sincerely,
                                                                                                                              ,/ / '
                                                                                                                                        /
William J. Cahill, Jr.
JLR/gj Attachments c  --Mr. R. D. Hartin, Region IV Resident Inspectors, CPSES (3)
'                    Mr. J. W. Clifford, NRR Hr. G. G. Benoit, ONRR (w/att 1) 9103080007 901231 PDR    ADOCK 05000415 R                      PDR
        - M U U U i If                                        400 North Olive Street L B. 81 Da!!as, Texas 75201
: l. Attachment 1 to TXX-91090 Page.1 of 16 COMANCHE PEAK STEAM ELECTRIC STATION ANNUAL OPERATING REPORT 1990 TEXAS UTILITIES ELECTRIC COMPANY
 
Attachment 1 to TXX-91090                                                      l Page 2 of 16                                                                    i 1
l IAllLE OF CQKil;XIl 1.0  Summa y of Operating Experience 2.0  Outages and Reduction in Power 3,0  Personnel Exposure and Honitoring Report 4.0  Report of Results of Specific Activity Analysis in which the Primary Coolant Exceeded the Limits of Technical Specification 3,4,7 I
 
I Attachment I to TXX-91090 Page 3 of 16                                                                        j I
1.0 EMMMARY OF OPERATING EXPERIEHL1 The Comanche Peak Steam Electric Station is a pressurized water reactor licansed at 3311 Hegawatt thermal (HWt). It is locatec in Somervell County in North Central Texas about 65 miles southwest of the Dallas Fort Worth Hetropolitan area. The nuclear steam supply system was purchased from Westinghouse Electric Corporation and is rated for a 3425 MWt output.
The Comanche Peak nuclear power plant achieved initial criticality on April 3, 1990. Initial power generation occurred on April 24, 1990, and the plant was declared commercial on August 13. 1990. Since being declared commercial, Comanche Peak Unit I has generated 2.513,514 MW hours of electricity as of December 31, 1990, with a net plant capacity f actor of 65.5 (using net HDC).
The unit and reactor availability was 84.9 and 94.7% respectively.
Figure 1 provides a histogram of the average daily electrical output 01 the unit for 1990. Table 2.1 is a compilation of the monthly summaries of the operating data and Table 2.2 contains the yearly- and total summaries of the operating data.
2.0 OUTAGES AND REDUCTION IN POWEB Table 2.3 describes plant shutdowns and provides explanations of significant d-ps in average power levels.
3.0 EERSONNEL EXPOSURE AND HONITORINA_ REPORT The personnel exposure and monitoring report is provided in Table 3.0.
 
n FIGURE 1                                          %5 n
i                      COMANCHE PEAK UNIT 1 (CPK1)                                          ai 5                                                                                            E~
GENERATION PROFILE
                      #                                                                                    o AVERAGE DAILY UNIT POWER LEVEL - STARTUP I.200 t
fl  I1      I g
          !,000 O
800 f
        $600                        TEST PLAlEAU
        =                                            -
D      D yp e.e,n w au.a n,.e, l
w 2*        . . .
m-o
                                                            ~
s                                            >
0                                                                            8/1 8/13 y              4/9      5/1                    6/1                7!I 1990
 
FIGURE 1                                    3E l                              COMANCHE PEAK UNIT 1 (CPK1)                                      [,i "
GENERATION PROFILE-                              ;;-
o  i f            AVERAGE DAILY UNIT POWEP LEVEL- COMMERCIAL OPERATIONS 1,200                                                                                            ;$
W h                    5
                                                                                          ._  .A f    8
                                                                                  '>fy' i
i.000                                                                          3; v
Q                  >                        Y 800 l
i,                                      l
                                              ~~
l                                          I                                f l
I w so 400                                      , , _
w .,                                                    <U' "Eh 'j l            !,
200                                      r~ .,
  ,                                          - w w us h        ,
f
                                          '-: r                        ,  i Sep                    Oct              Nov              Dec Aug 14 1990 l
 
TABLE 2.1 (PAT 10F 2)
                                                                                                                        -o, n n>
El.ELTRIC RMR (DERATICN DATA (1990)
                                                                                                                        <I' "a n
o -r KNHLY n
                                          -            Fw1            m            w          w      ne              :- !
n- i i
0              152.5    021.1  393.2              N Hmrs a wts critiml                M            0 7 '
0            0          122.9  218.4              3 a Mene 9ut&wi Hirs                M            0 S
0              123.8    530.1  272.4 Hatrs Gmerattr (h-Lire            M            0 M            0              0            0          0      0 thit Rseve Sutziwi Hxrs 0              127.134.8 748.810 518.035 Gruss Themal Ervrty Gmerataf OWO  M            0 0              14,840    197,108 156.970 Gmss Elec. Eneny Greratsi (MdB    M            0 M            0              0            1951      165.742 128. 8 6 Net Elec. Eherty Gmerated OWO M              M              M        M      M E Servim Facttr                    M M              M              M        M      M E kailsility Factcr                M M            M              M              M        M      M thit Servi Factrr M              M              M        M      M thit Aailability Factrr            M M            M              M              M        M thit C@acity Factor (usfrg ME ret)                                                                      t
                                                                                                                      ^
M              M              M        M      M thit C@acity Factrr (usiru ER ret) M M              M              M        M      M thit Fared (Ltsg! Rate            M 672.0          744.0          719.0    744.0  720.0 Hxrs in Ptnth                      744.0 1150.0        1150.0    1150.0  1150.0 1150.0      1150.0 Net MC (Ne) estinetaf
 
TMEE 2.1 (PME 2 0F 2)
                                                                                                                                                - o :>-
E11DMC RMR(BORTI(N DEA (1990)                                ([
K"3                                              o m g ..
R
                                            -+                            Mr          Aset          Setoter    om        Mvkar    Oe: star Htitrs a was critical                          609.5        SE            995.2      745      4 11.2    744                .
R Rsme 9utriwiitiurs                            134.5        216            124.8      0        111.1    0              h
;                        Rxrs Generator Ch-Une                            EB.5        494 3        557.2      745.0    415.5    744 i
thit Resme Sutriwi kurs'-                        0            0            0          0        0        0 Oms lhenel Bierg Gmerataf (Me0                  1.EB.199    1.515.3CE      1.480.101  1.911.524 1.073.Z37 2.374.056 Gus Dec. Eherw Gmerated (Me0                    59.414      495 3 31      474.111    622.541  348.0 0  795.409 kt Dec. Eherw Gmerated (Me0                      518,312      4 8 .242      440.564    587.620  321.121  78).5@
m Servim Factor                                E            91.7          T.7        100      13.9      10 0 RX hailtility Factor                      .
E            1CD.0          100.0      100.0    75 3      100 thit Servim Factor                              E            90.2          77.4        100      57.7      100 thit hailability Factor                          E            90.2          77.4        XX)      57.7      100 thit C@acity Facttr (using M)C net)
* 79.2          53.8      @.4        3.2      s.9 thit Capacity Factor (using [ER ret)            *-          78 3          53.2      8.6        3.8      8.9 thit Faud Qatage Rate                            E            9.8          21 3      0.0        9.7      0 Rxis in Month                                    744.0        744.0        720.0      745.0      720.0    744.0                s Net ME (Me) estiseted                            1150.0      1150.0        1150.0    1150.0    1150.0    1150.0
 
      ~
a EN TABLE 2.2                                        g-on $?
ELECTRICAL POWER GENERATION DATA                              o%
                                                    '1990                                          "'E o-YEAR              YEAR      CUMULATIVE (Pre-Commercial)      (Commercial)                    g-Hours,RX was critical-                            1893.3            2926.4        4819.7        g 927.7        $a RX Reserve Shutdown Hours                          654.8              272.9 1626.1            2865.7        4491.8 Hours Generator On-Line 0                  0              0 Unit Reserve Shutdown Hours Gross Thermal Energy Generated (MWH)          3.279.471          8.156.928      11.436.399 Gross, Electric Energy Gen. (HWH)              989.787          2.v {.000      3.654.787 Net Elec. Energy Generated (MWH)                881.263          2.513.514      3.394.777 NA                86.7          86.7 RX Service Factor RX Availability Factor                              NA                94.7          94.7 NA                84.9          84.9 Unit Service Factor bnit Availability Factor                            NA                84.9          84.9 NA                65.5          65.5            l Unit Capacity Factor (using MDC net)
NA                64.7          64.7 Unit Capacity Factor (using DER net)
Unit Forced Outage Rate                            NA                  7.8            7.8 5383                3377          3377            ;
Hours ~in Reporting Period                                                                            l l
i
                                                                            ~
 
h i
Table 2.3 -
I ouT mmnas a Run antnos                                              ;i t        ;
EDfD &                                                    eE TYE                                    StKTIMi(DM                                                  o 2:
F:fGCED  DRATI(M                          TE KETIR (R                                                'E          >
: 10.  [ RTE  S:SDe m    (HM6)              KATN      EDUQG RMR            00RETIVE ET10KMD9065                  7
* n 90[E01      'S    ' 6 72. 0              H                  4          thit in Starte Rese prior to initial criticality.                  M''
7 !
tav Fber (irratirg Licme remived          .3 - .;
Fdnay 8.1990. Initial Ftel load            1i '
and Entry into MIE 5 ampleted m                    ;
Fd n w y 17, 1990. A diallege to a                  ;
O RM (lav Teperatire Orspresure                    ;
Protectim) otamsd on G'lZ1/90 drirg E fill ard vet. ("prial                      i Report 1-SR4}0E)                                    l t
i 90001        5    744.0                  H                  4          thit in "lartp Rwre prior to initial criticality.                            ,
900411        F    29.0                  A                  1          Fber Ryge Detector twlaaset                        l t
,                                                                            900421        F    12.9                  A                  3          RK Trip on Hi@ Satne Rary Fita.                    [
thit ertend MIE 3. Trip de to                      j inadvete't tsiblockirg of Saisoe                    ;-
Rrry. Sensitive reset switdi                        i dististied drile clesnfrg cmtrol                    !
board. (LER-90-UB)                              t i.
(D1 90EfD        F    41.4                  G                  3          Reactor Trip m lat-lav Steen                      [
Generator level, thit entmd MEE 3.                  '
last level de to 1cr;s of Fedeter to me Stest Gerrator de to Tedriicisi l                                                                                                                                                      eTur wykirg m Presstre trarturitter              -
for Fedeter Contrui. (LER 904113)                [
r h
t
 
'                                                    Table 2.3
                                                                                                          .IN tNIT mmXW6 #0 fMR POETIWS                                    %"-
                                                                                                              ?
,                        ETMD OF                                                                          0g TYE                              SUITIMi DON                                          oE 4
F:FORID      DFATI(N              TE EKTCR m                                            --
i M). IMTE . 5:SO(DlilD      00RS)  KAS[N      IUNIIMi RMR      00FMETTE ACTIO6/CD99TS              *  , , ,
o N
(D2  90312        F          43.7    A                  1        Fmbeter Cmtrol Valve Di'A              7 Sgaratim. Unit etterd KIE 2            $r experiecirg prttleas with a Feksts      S' Cmtrol Valve. Shapetly determirvd tiet di'A had separated frm tte sten. Valve was rgeired.
0[D  90322        S          95.5    B                  3        Part of plared toss of Offsite Fuer Test per Initial Start @ Pruym.
(D4  90327        F          33.3    A                  2        Failtre of Fmieter Cmtrol Vale Solenid. Mnal Amctor Trip. thit eitered MIE 3. Trip in reprre to failai stut Fmieter Cmtrol Valve ad attidpated law Steen Gmeratar level. Failtre de to dort cri Fmieter Contrul Valve. (LER 9tH)17) 0[15 RX1iO6      5        447.6    8                  2        9utdwi in ardtsction with I4 mote 9utdwi Test tr,e1 to eiter pland autap. Mnel R5 actor Trip. thit eftered KIE 3. Plamed trip for Rstote 9utdwi Test ad straice to pland cutsy for wrt ei Msin Gmeratar Amiliay w=56 (prisery wate grurd. deft tiwsi lite oil pum).
006  90G707      5        0        8                  4        50 loid re,)ection test frun 75:            ,
N .T plateau.                                j 007  900724      5        0        8                  4        50% lod rejecticn +at fra 100 Iber Plat.~ -
[
 
Td>la 2.3 2N tnIT sumes m) m motcncus                                            1; ;;        .
                                                                                                                          ~9:          ,
wrMn oF                                                      -g.
wm                          sumM; tom -                                                9, a .
F:lGCED  OtRATICH            TE RK10R (R                                              . , _ , -
: 10. DATE , 5:SDELUD    (MUS)  RASN      EDlCDG TUER        (GHLTIVE ACTIOG/QMe(T5                  *  .,,
,                                                                                                                              O Y
008 90T25        S      134.5    8                  3        10% tms of Load Test. resulted in            7 i (mlame!) astourtic scrm. tirit -              ~
eitersi MEE 3. Started olaried                    .
                    .                                                            astap to replam tellos m                    'O Contairwnt Pressure instruarit 1-Pf R34. neilver of Caplianm to perfona repla&urnt in KDE 3 stained.
GB  90GD1        F      172.5    F                  4-        Omtiturtim of shtrtwi startal                          !
90726. Extauhd sdeislei                                ,
shhtwi to revier tradrertent SI evets.
010 90EKB        F      33.3      A                  3        Reactor trip on to-lo SiIEvel de to failed stut feedseter valve. . PX Trip,                ,
tilit etem! MEE 3. Trip de to loss
* of fedeter flow to Si #4 ad stbegnt to-to tevel. (ifR 904E3).
* Festat &lmel de to selare of                        ,;
Safety 01111 lister Ptsp just pricr to trip.
4
                                                  -COMMERCIAL OPERATION            --
i 011 90GD6        F      0        A                  4        Bearirg Failtre at Hester &ain Ptsp.
f%er misction for 29.6 tars to rmla .
i
 
Table 2.3
                                                                                                                  .IN uaT Samxws no awR parnoc                                                      2 ;'
1 ElMD T                                                                "2  1 wt                          sameam                                                                *n F:RRID  RFATI(N              TE ETOR (R                                                            --:
KA3N                                                                              *
: 10. DATE . S:SOE111D  (H0lRS)            FEtIDG F0iER    (IEELTM KTIO6/0MOUS N'
012 90(B:5      F    44.0      A                  3      lbrtar Trip m Hi-SG tevel de                            7 feetneter lhplatirg Vale studc gm.                      ~i m Trip, thit eitsed MIE 3. Trip                        8 i                                                            de to Hi-Si E level de feed thJ                            >
Valve stui qm. (LER 9045) 013 90WO7        F    30.4      A                  2      Mrual Smsn after Fmieter Cmtrul valve failed gm. Mrual RX trip.                          .;
thit etertd t0DE 3. RX tripped to                          t arcid Hi@ SG R level de to Festeter Cmtrol valve failirg gm. (LIR 9047) 014 90W(B        F    19.5      H                  3      Arantic Smsn m High Flm Rate                              ,
csral by Littnirg Strike. RX Trip, thit etsui MIE 3. Li@trrirg strike irrimi voltage @ikes ard gauratal Hi@ Flta Rate Trip. (ifR 904B) 015 90W10        F    73.7      A                  3      A2txzetic Smsn m Ttsbire Trip. PX Trip, thit e tard MIE 3. RX Trip on Ttstire Trip de to hi@ M)isture S@arator fWees (KR) level.
FaJlty level cmtrol caral flooded MR. (LIR 9045) 016 90W15        F    22.1      A                  2      Mnal Smsn de to loss of Fmbeter.
FX Trip. thit eiter=d MIE 3. Hrtel RX Trip to avoid to-lo Si 73 Level de Msin Freie+s Rap tripped m lov suctim pressure follovirg Hester 0 rain Rap Trip. (LER 9 CHID) i
 
Table 2.3                                        -e >
E" uuT strnNs no m marTIms                                    ag
                                                                                                - :r KT4D OF                                            "#
O, :n3 TYE                        9 m TI'G CO N
                                                                                                ~~
F:FO O L OlFATICN            TE RK.TCR (R
* ND.  (RTE . S:SOG1m    (KlRS) MXIN      RDLING FGER    WHITM ACTIN 3                          y 0
                                                                                                    -4 I
i 4      Rd1M Rwr Qeratim fcr 52.1 hars          7 F    0        B 017 90(B16                                                for Balanm of Plant Crumirg            3 (cmtrtis calitraticr. steen lek        8  j rgmir).
c)      A                  4      Thimi gwr for 17.4 hars to 018 RXB22        F irvstigste care of heter &ain systea trarsielt ard curvation of less otrM driru trasielt.
l 9.7      8                  4      Turbire off lire of 9.7 tars for M9  9(IBa        S reirteure m Feedeter Ccrtrol Valves.
7.4      A                  4      Turtire tri; ped de to witdyard GD  90093        F truiker pxtien. thit entemd KDE 2 to cmirt neinteure an Feieter Cmtrol Valves. Drisim to eite altar naie folicwiry teater tain systm irrired trarsist tfut revirsi ruimi pwr to irwestigrte ad comrt lais card ty tracimt.
T1  MIRE        S    0        F                  4      Qvratim at 73 Px Rwr for 112.5 tors thruffi ed of mmth fcr ftel arservatim ad cycle extensim pst Stz tr 1991.
022 90010H      S    0        F                  4      Qvratim at ruird Px Rwr thraxfut tre atrith far fiel anzvatim ad cycle exiesim past
                                                            #u ner 1991.
 
Td>1e 2.3 EN tNIT simnes no nun RRcTras                                              ;;
M3MD &                                                      E mt F:f0GD      OtFATI(N sumE aw TE KXim (R u
M).              DATE  . S:50fRLfD      (MIPS) WA7N            EDLIE RMR            00@fLTPE ACTIOG/CD9M3                  *
                                                                                                                                                    .N 00            9011(B            S        2B.5      8                      1          Plared outar for 8(P neirtmance ad          7 irure flux thint>1e clartirg prior to .    $
wirts peak Imd period.                      S o
T4            901114            F        13.0      G                      4            Rstart & layed follovirg sde11ed outage. Failure to pyfor1n sneillance test in MIE 4 rupind plait to cool dw to MIE 4 afts adiievirg MIE 3.
025          901114            F        0        B                      4            Rsijoed Iber ge atim for 15.3 turs for KP grumirg.
026          901119            F        32.0      A                      1          Prut& tim irwster falltre resultei in Tedi Src repired 1stdwi.
[ht>1e to resttre irwyter within 24 inss per Tedrifosi 3xrificatim 3.8.3.1. plant stutdm initiated ad notificatim of unustal ersit
                                                                                                        &claraf. (LIR 904)(1) 027          901212            F        0        8                      4            Rsirrd Rwr (teatim for 83.9 ints:
L,wstigste Ccrdmer trie lesar.
Trubleda:t Ttatnre DC systs.
MR ad Fdi r@ airs.
: 1)            KA7N                                                                      2)    KTMD A: EDJIRENT Fall 1K (ERAIN)          E: GGATE TIMINIE #O LIGM5E DUMDET104      1: MWW.
B: PMINT (R TEST                      F: A011NI5il%TIVE                        2: MWW. SUMM C: ERELDG                            G: 09ATIGEL DHR (EXPLAIN)                  3: ALED% TIC ST*t D: EIIAATUN E5TRICTIGl              H: OnfR (EXPLAIN)                          4: OTifR (E M AIN)
I
 
l Attachment'1 to TXX-91090-                                                                                                                                ]
    'Page;15'of 16-
                                                              ' TABLE 3'O J
Personnel Exposure-and Monitoring Report m
Work & Job Function-                  # Personnel b100 Mrom)                                          Total Man-Rom Station Utility Contract                                    Station    . Utilhv    Contract Reactor Operatens
                & Surveillance -
Maintenance & Construction            2          0                    0                  0.243        0.000        0.004 Operations - .                        0-          0                    0                  0.319        0.000        0.064 Health Physks & Lab                  0          0                    5                  0.506        0.000        1.056 Supervisory & Offke Staff .          O_          0                    0                  0.033        0.000        0.000              ,
Engineering Staff                    0          0                    0                  0.105        0.000        0.000
            - Routine Plant Maintenance Maintenance & Construction            0          0                    1                  0.430        0.000        0,884
                  - Operations --                          0          'O                    O                  0.057        0.000        0.388 Health Physics & Lab .                0          0                    0                  0.095        0.000        0.110 Supervisory & O!fice Staff            0          0                    0                  0.000        0.000        0.000-Engineering Staff                    0          0                    0                  0.036-      0.000        0.065 -
Inservice inspection Maintenance & Construction          -0          0                    0                  0.000        0.000        0.004 Operations ..          .
0          0                    0-                  0.000        0.000        0.272
                    - Health Physks & Lab --              0.          0                    0                  0.041        0.000        0.101 Supervisory & Office Staff            0          0                    0                  0.010        0.000        0.000 Engineering Staff-0          0                    0                  0.011        0.000-        0.021 Special Plant Maintenan:o Maintenance & Construction            0        .0                      0                  0.000      .0.000        0.004
                    -Operations . .                        0        .0-                    0                  0.000        0.000        0.110 Health Physics & Lab                  0          0                    _0                  0.000        0.000        0.000
                    . Supervisory & Office Staff            0          0-                  .0                    0.000      'O.000        0.000 Engineering Staff -                  0          0-'                  0            - 0:000            0.000        0.000-Waste Procesarig
                  . Maintenance & Construction          'O            O                    O                  0.000        0.000        0.004 Operations            .
0          0                    0                  0.000        0.000        0.008 Health Physics & Lab .              -0            0                    0-                  0,036        0.000        0.000 Supervisory & Offke Staff -          0-          0                    0                  0.000'-      0.000        0.000-Engineering Staff                    0          0                    0            - 0.000            0.000      -0.000 Refueling -                                              ..                                    .
Maintenance & Construction .          0          0                    0.                  0.000      -0.000        0.010 Operations -                          0          0                    0                  0.036-      0.000        0.014 Hoeith Physics & Lab '                0          0                    0                  0.019        0.000        0.000 Supervisory & Office Staff -          0          0                  ,0                    0.000-      0.000        0.000-Engineering . Staff                  0        -0                      0                  0.011-      0.000        0.000
            - 1 otals --
Maintenance & Construction            2        :0                      1                  0.673      - 0.000        0,910            '
                  . Operations .                          0-          0                    0                '0.413        -0.000        0.856 Health Physics'& Lab .                1-          0                    .7                  0,698        0.000 ~      1.266
                  ' Supervisory & Office Staff 0          0                    0                  0.052'      O.000        0.000
                    - Engineering Staff                    0          0-                    0                  0.163        0.000        0.086 Grand Totals                                3          0                    8                  1.999        0.000        3.118 P.O. Bot 1002 Glen Ro$e. Texas 7604341002 u,  . - - _.. - . _ _ _                          __                    . . , _ , - . _ _              ~, _                            _    .
 
Attachment 1 of TXX 91090 Page 16 of 16 4.0 A REPORT Of RESULTS Of SPECif1C ACTIVITY ANALYSIS IN WHICH THE PRIMARY COOLANT EXCEEDED fHE LlHITS OF TECHhlCAL SPECIFICATION 3.4.7.
Technical Specification 6.9.1.2.b requires the results of specific activity analyses in which the primary coolant exceeded the limits of Technical Specification 3.4.7 Unit I has not exceeded the limits of Technictil Specification 3.4.7 for the calendar year of 1990.
There has been no irradiated fuel examinations conducted during this reporting period.
l i
l
 
Attachment 2 to TXX 91090 COMANCHE PEAK UNIT 1 ANNUAL 10CFR50.59 REPORT This report contains a description ind a summary of the following 10CFR50.59 evaluations.
SE 89 066 REY. O                SE 90 043 REV. O          SE 90 099 REV. O SE 89-072 REV. 0 5            SE-90 044 REV. O          SE-90 100 REV. O SE-89 083 REV. 0 1            SE-90-045 REV. O          SE 90 105 REV. O SE-89-113 REV. O              SE 90 046 REV. O          SE-90 107 REV. O SE-89 136 REV. O              SE-90 048 REY. O          SE-90 108 REV. 0  ,
SE 90 003 REV. O              SE-90 049 REV. O          SE-90 201 REV. O SE-90 017 REV. O              SE 90 050 REV. 0          SE 90 202 REV. O SE 90-025 REY.                SE 90 051 REV. O          SE 90 204 REV. O SE-90 027 REV.                SE 90 053 REV. O          SE 90 205 REV. O SE 90 028 REV O              SE-90 054 REV. O          SE-90 206 REV O SE 90 029 REV-. O            SE-90 055 REV. O          SE-90 208 REV O SE-90-030 REV. O            SE 90 056 REY. O          SE-90-209 REV. O SE-90 032 REV. O            SE-90 057 REV. O          SE-90 211 REV. O SE-90 033 REV. O            SE-90 058 REY. O          SE 90-221 REV. O SE-90-035 REV. O            SE 90 059 REV. O          SE-90 222 REV. O SE-90-036 REV. O            SE-90 060 REV. O          SE-90 223 REV. O SE-90-937 REV. O          SE 90 061 REY. O          SE 90 225 REV. O SE 90 038 REV. O            SE 90 062 REY. O          SE-90 228 REV. O SE-90-039 REY. O          SE-90 063 REV. O          SE-90 229 REV. O SE-90 040 REV. O          SE-90 064 REV. O          SE-90-230 REY. O SE-90 041-REY O          SE 90-065 REV. O          SE-90 236 REV. O SE 90-042 REV. 0-1        SE 90-066 REV. O SE 90-067 REY. O SE-90 068 REV. O SE-90-069 REV. O SE 90 070 REV. O SE 90 071 REV. O              '
SE 90 072 REV. 0 SE 90 073 REV. O SE-90-075 REY. O SE-90 076 REV. 0-SE-90 077 REV. O SE-90-079 REV; O SE-90 080 REV. 0 3 SE-90-081 REV. O SE 90 084 REV. O SE-90 089 REV. O SE 90-090 REV. O SE-90-091 REV. O SE-90 093 REV. O SE 90 094 REV, O SE-90-097 REV O SE-90 098 REV. 0 l
 
Attachaent to TXX 91090                                              TV Electric Page 1 of 90                                                          Unit: IXN Evaluation Number SE 89 066 Activity
 
==Title:==
Replace Piping in the Surface Water Pretreatment System Description of Change (s):
The chemical feed piping for the Acid Feed Tank and Phoshete feed Tank of Water Treatment system is corroded and is being replaced with Chlorinated Poly Vinyl Chloride (CPCV) piping. Also suction strainers, pump internal check valves, and pulsation dampers are being added. This is a non safety system, as shown in FSAR Figure 9,2 4A sheet 3, and these changes do not affect plant safety.
Summary of Evaluation:
The piping in the surface water pre treatment system was replaced because the piping material was significantly corroded. The new CPCV piping material is not considered to be susceptible to acid and caustic degradation under the Water Treatment System's
    -conditions, The changes do not affect the design, operation and performanre of the system. The modification does not change the basic system logic for start-up, operation, or shutdown. The system is non-safety related and is located outside safety related areas, therefore no consideration of seismic tallure or system interaction is. required.
* Attachment to TXX 91090                                              TV Electric Page 2 of 90                                                        Unit: 1XN Evaluation Nueber SE 89 072 Revtsion 5 Activity
 
==Title:==
Addition of pressure and flow transmitters to the OAS.
Description of Change (s):
The data acquisition system (DAS) is a temporary non safety system installed to monitor various plant parameters. The OAS cabinets, raceways, junction boxes, and equipment are seismically supported and restrained. Power to the DAS is from a non safety UPS distribution panel. This change to the DAS and Rev. 5 to the Safety evaluation adds pressure transmitters to the Hein steam system (MS) and flow transmitters to the feedwater (FW) system in order to provide input to the DAS.
Summary of Evaluation:
The safety evaluation addresses DAS power supply adequacy and electrical separation requirements for both power and signal circuits.
The SE addresses cable separation for DAS cables, both soft ruri and in raceways, and the effects of DAS cable runs in trays on cable tray loadings, The Safety Evaluation addressed the effects of DAS on fire protection by evaluating the impact of cables on combustible loadings and Appendix R separation, and by evaluating the impact of the modifications _to safe shutdown components. The SE verified that DAS installed pressure transmitters in Hain-Steam system are seismically supported and meet the system pressure boundary requirements.
The SE concluded that thert were no new failure modes were introduced, and the probability of an accident was not increased by the installation of the modification.
 
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==Title:==
Peak Cladding Temperature (PCT) Changes for the LOCA Analysis Description of Change (s):
The analysis of peak cladding temperature (PCT) in the Final Safety Analysis Report (FSAR) for both the large and small break LOCAs was updated to reflect the results of a number of safety evaluations performed by Westinghouse to establish the relevant PCT penalties, The situations wh*:h caused the large break LOCA penalties are:
: 1. A steam generatcr bypass flow modification (to 90% full feedwater flow at full reactor power),
: 2. An error in the accumulat;r water volume (6 f t3 less than the 850 f t3 used in_ the original analysis),
3,  Incorporation of a delay timer in the closure circuit of he Residual Heat Removal pump miniflow valves,                                    ,
: 4. The correction .for errors in the WREFLOOD Code, as identified in Generic Letter 86-16,
: 5. The consideration of steam generator plugging (use of 1%), and
: 6. An error in steam generator flow a aa (modeled as 2,1% larger than design).
  . Summary af Evaluation:
Applying these penalties on an individual basis resulted in a total PCT increase of 47.8 degrees F and 205.2 degrees F for the large break and small break LOCAs, respectively, The new PCT for the large break LOCA is'2058.5 degrees F, whereas that for the small break is 1992.7 degrees F. These new values remain well below the PCT limiting value of 2200 degrees F defined in 10CFR50,46, so that there is no reduction in the margin of safety.
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Deferral Of Unit 1 Spent fuel Pool Pipe Support Rework Until Post Fuel Load Description of-_ Change (s):
                            -Implementation of two Spent Fuel Pool Cooling and Cleanup (SFPCC)' system pipe support rework packages (numbers-1-0868 and 10870) was deferred until af ter Unit 1 initial fuel load.                The SFPCC systemLfunction.is to provide cooling and cleanup support for the SFP's.while thermally. hot irradiated fuel assemblies exist in the SFP, The possibility of=having irradiated-fuel assemblies was only presented _after CPSES Unit I achieved initial criticality (i.e. Mode 2 operation). The'two SFPCC pipe support rework packages involved the modification of ten Unit 1 Spent- Fuel Pool (SFP) pipe supports and associated activities that support the satisfactory completion of required documentation.(1.e.. OC inspection. As built walkdown/evaluati.on, and NIS-2 completion for the SFPCC system).
The deferral of this activity was necessitated by the presence of 61 new' fuel assemblies, including two primary sources.,in.the SFP. The
                                      ~                              ~
Special Nuclear Material (SNM) license SNM-1912 requires 'that no
                                                                          ~
: construction activities are allowed that'could possibly cause damage to the fuel assemblies in_ storage,                in compliance with the SNH-license the pipe support rework was performed after fuel load so 4                      as not to;present a unnecessarytrisk of physical or radiological interaction .with the new fuel and primary- sources.
        .                LThe SpentiFuel Pool--_ Cooling and Cleanup system is classified as a
                                                  ~
Seismic, Category ! system as delineated by^FSAR section 3.2.1.1.2 g              _
and Table' 17A'-1. _ The _ pipe support 1 rework per_ formed ~ enabled _ the SFPCC systemLto comply with CPSES' Seismic-Category 1 design criteria as                                          ;
                            -presented by the-above stated sections of the:FSAR.
Summary of_ Evaluation:
Im01ementation of the pipe support rework on CPSES: SFP Number 1 is
                          -considered acce, table to perform during this time frame based upon system operabi'ity; requirements.- -SFPCC is not required prior to-
                          -H0DE 2. Completion-of the rework was a MODE 2 constraint.-
h p
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==Title:==
Addition of chemical addition filter /demineralizer skid to the Component Cooling Water (CCW) system.
Description of Change (s):
Added the filter /demineralizer chemical skid (FDS) to the CCW system as a permanent part of the CCW system by providing a two inch line.
    -The permanent plant installation of the FDS into the CCW system is a design improvement which provides filtration required to maintain purity levels of the system by reducing suspended solids to an acceptable level, by addition of hydrazine, and by providing ion exchange to maintain chloride level within specification.
Summary of Evaluation:
The F05 flow of 50 GPM through a 2 inch line is small as compared to the CCW system flow of approximately 4000 GPH, and will not significantly affect the flow distribution in the CCW system.
The 'non safety loop of the CCW system, where the FDS is tied in, does not provide a safety related function. The safety related system portion is isolated from the non safety loop by automatic isolation valves which actuate under various accident conditions. There it adequate volume in the CCW surge tank and adequate CCW pump NPSH to account for postulated system outleakage from the non-safety loop until the isolation valves close, The FDS supply and teturn pipe size is. bounded by the existing flooding analysis of the area. The piping and supports have been qualified for the seismic movements associated with the fuel building. The FDS is mounted Seismic Category !! and the FDS components have been qualified Seismic Category 11. No safety related components will be impacted by piping or FDS mounted components under seismic loadings.
Any postulated break in the CCW supply and return piping are bounded by existing analysis, No limiting condition for operation or surveillance requirement is affected by the addition of the FDS to the CCW system, o
 
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==Title:==
Addition of Heating Elements to Feedwater Isolation Valves                    j l
Description of Change (c):                                                      i l
In early 1990, TV Electric reported to the Nuclear Regulatory Commission (NRC), in accordance with 10CFR50.55(e), that the ferritic material of which the Feedwater Isolation Valves (FWlV) were manufactured had not been impact tested-as stated in the Final Safety Analysis Report (FSAR). Extensive testing and evaluation has been performed to establish the acceptability of the materials used. Based upon the results of impact testing, the lowest service temperature for the body and neck of the FWlV's, when the FWIV are pressurized, has been established at 90 degrees Farenheit. The fracture mechanics analysis yielded a lower service temperature, but the flaw size assumption was not readily supportable, To maintain the FWlV above the 90 degree limit, electric heaters have been installed. The heaters are required to be operating whenever the FWlV are below 90 degrees which would normally occur during cold feedwater system startup conditions.
Summary of Evaluation:
All of the modificatian is external to the FWIV's and has no interf ace with any safety related system. For this reason, the addition of the electric heaters does not increase the probability of a licensing basis accident and does not increase the probability of fo11ure of the feedwater system.
i i
 
o Attachment to TXX-91090                                              TU Electric
* Page 7 of 90                                                          Unit  1XN Evaluation Number SE-90 017 Activity Titlet Change FSAR figure 11.2 3 to show normal position for valve XWP 0166.
Description of Change (s):
Revise FSAR Figure 11.2-3 to show the correct normal operating position for valve XWP-0166-(changed from normally open to normally closed).
This valve is a boundary valve between the Waste Evaporator Condensate Storage Tank and the Reactor Hakeup Water Storage Tank (RHWST). The normal closed pcsition reduces the potential for inadvertent discharge of cont aminated water to the RHWST, Summary of Evaluation:
There are no safety considerations associated with this change. The original designation of normally open wa( a drafting error.
 
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Activity
 
==Title:==
Undervoltage Protection Time Delay Relays Description of Change (s):
TSAR Figure 040.109-1 and Table 040.109 1 required updating to reflect the as built /as designed time delay setpoint for time delay relay.
27AX/ST1, and corrects the time delay setting for the offsite source low voltage alarm relay, 62/ST2, respectively, in addition. Table 040.109 was revised to reflect the correct offsite source low voltage alarm relay setting from 339.5kV to 339kV.
Summary of Evaluation:
Device 27AX/ST1 is a time delay relay that times out after the offsite source availability undervoltage relays drop out. The time delay for this relay is 0.5 seconds. The tfmer is set to allow sufficient time
  ' delay, 0.4 seconds, to clear a bolted three phase fault by coordinated overcurrent relays.                                                                                              The 0.4 seconds include the maximum overcurrent relay operating time of 0.1 seconds plus the 6.9kV switchgear breaker trip time of 0.3 seconds. Therefore, a time delay of 0.5 seconds is sufficient for the 6.9kV switchgear breaker to open with a margin nf 0.1 seconds.                              Temporary excursions and momentary source voltage degradation caused by a fault could exist for the 0.S seconds until cleared by the 6.9kV switchgear breaker, but has been determined to have no adverse effect on the elect rical system.
Device 62/ST2-is a time delay relay for the offsite source icw voltage alarm relays. The 62/ST2 relay provides an 40 second delay tt allow system voltages to stabilize after remote grid disturbances, This setting is correctly specified on FSAR Figure 040.109-1, but incorrectly specified-in FSAR Table 040.109 Sheet 3. Therefore, the change to Table-040.109 62/ST2 timer setting is editorial in nature, such that both the FSAR Figure and Table are consistent and correct.
The of fsite source low voltage alarm relays described above are 27A-3/ST2 and 27 4/ST2 and provide an alarmed condition in the control room when an 345kV offsite undervoltage condition exists. These.
relays are set to drop out at the minimum expected operating voltage of approximately 340kV.                                                                                            The actual setting of the relays-is 339kV, which is based on the potential transformer drap out voltage of 113V, as reflected in FSAR Table 040.109. Sheet 2, but incorrectly specified as 339.5kV in FSAR Table 040.109, Sheets 3 and 4, The above changes reflect the as-built settings, as specified in the CPSES design basis documents and calculations, such that the device will perform their intended function when required.
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==Title:==
Operation of Unit I with a modified primary coolant chemistry pH Program during the first fuel cycle.
Description of Change (s):                                                      *"
A modified coolant chemistry control program having an upper bound of approximately 2.? ppm lithium (i.e., 2.05 plus or minus 0.15 ppm lithium) will be used to operate Unit I during the first fuel cycle.
During the past decade research has indicated that operating the reactor coolant system within an elevated lithium /pH reactor coolant chemistry regimen reduces the precipitation of nickel-ferrite corre>ien products on core cladding, thereby leadiog to reduction in ex core radiation fields and reduced man rem exposures. Therefore, from an ALARA viewpoint, operation with an elevated pH is beneficial.
Summa ry of Evaluation:
The operation of Unit I under the prescribed guidelines of a modified lithium control program is expected to have a negligible impact on the reactor coolant system materials, auxiliary system materials, and fuel cladding.
Primary-to secondary leakage of reactor coolant containing elevated lithium is determined not to contribute to secondary-side corrosion issues, as a result of the low rate of boric acid ingress and the known corrosion. inhibition capability of boric acid.
The single most important factor in initiation of primary water stress corrosion cracking (PWSCC) in steam generator tubing, namely-the residual tensile stress, has been addressed in the Unit I steam generators.
Operation within this program is not expexted to have an adverse effect on Zircaloy' fuel cladding performance based on satisfactory
  -observation for fuel average burnups-to 60 GWd/Htu by Westinghouse.
The effect of the subject program on post LOCA chemistry is to increase the pH of the sump solution by less than 0.1 pH and is therefore not expected to have any significant impact on radiological consequences, hydrogen production, or equipment                  1 protection.
 
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==Title:==
idition of valve tog numbers in the Instrument Air System.
Description of Change (s):
4
                                                              +.1  valve tag numbers to instrument air isolation valves to c          ,onents which are supplied with instrument air. The valve tag numbers are assigned in accordance with site procedures (l.A.W STA-618) which reduces the chance of operator error.
Summary of E"aluation:
Instrument Air valves provide isolation to accumulators in the Auxiliary Feedwater and Hain Steam System which are required to supply air to components during a loss of instrument air event.              The remaining valves are in instrument air lines which supply air to other safety and non-safety related components that do not rely on instrument air to fulfill a safety function.
Valve tag numbers have been assigned to plant design drawings as per site procedure (l.A.W STA 618) to reflect the-as built condition of the plant and to reduce the probability of operator error, f
This design-change activity will provide more information to operations to provide for more detailed procedures and reduce the probability of operator error and to ensure proper valve line-up in the instrument air system.
 
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==Title:==
Changes Door 217 to non security door and Door 218 to card reader controlled security door.
1 Description of Change (s):
In order to reduce the time and distance required by the Auxiliary Operator on making rounds during plant operations, Door 217 (S 290) was changed from a locked non routine security door to a non security door to open up ths 1 rain A Switchgear Room to the Auxiliary Building                                        ,
Hallway, Door 218 ($ 350) was changed from a non security door to a card reader controlled security door.                      The design modification changed the Security Plan figures which show which doors are lo;ked or have card readers and redefined the sub+ areas within the Vital Complex Area.
Summary of Evaluations _
The Auxiliary Operator on making his rounds during plant ope',ations bad to go to the elevation 032 to get to the Train A Switchgear Room (elevation 810) and the Diesel Generator Rooms. The implimentation of the design modification reduced the time and distance resuired for the operators.
The unlocking of Door S 29C (elevation 810) opened up the Train A Switchgear Room to the Auxiliary Building Hallway which is a vital to vital security area boundary. The Hot Shutdown Panel (elevation 832) is still within a vital to vital security area boundary controlled by the add 4 tion of new card readers on Door S 360. Train 8 Switchgear Room is unchanged except that one lesa card render is required to reach the room using Door S 290 (it still has its own card reader on the saatrwell).
The effect of this activity and the failure mode associated with the activity will not impact the probability of failure of the system.
Upon the loss of AC power, the Card Reader Doors could become locked, however, these doors can be operated by a key to enter and a thumb switch to egress (which is no different from all security operated doors),
The net effect to safety systems in this area is to proviue additional pathways and thus enhanced accessibility to the equipmeet if Operations personnel are required to attend to the equ:pment during any mode of operation.-
 
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==Title:==
thanges bior 018 to a card reader controlled security door.
Description of Change (s):
In order to reduce the time and distance required by the Auxiliary Operator on making rounds during plant operation Door 018 (E 3CX) was changed f rom a locked, non routine security door to a card reader controlled security door. This modification allows an operator to                              l I
access the Turbine Building without having to return to the elevatic' 810 hallway before entering the Turbine Building.              The design modification changed the Security Plan figures which show which doors are locked or have card readers.
Summary of Evaluation:
The Auxiliary Operator working in the mechanical araa in the Electrical and Control Building was required to return to the elevation 810 hallway befort entering the Turbine Building.
Door 018 (E 3CX) was changed from a locked non routine serurity door to a card reader controlled security door,            The imp'lementation of the modification reduced the time and distance required for the operator.
The effect of this activity and the failure mode associated with the activity will not impact the probability of failure of the system.
Upon the loss of AC power, the Card Reader Doors could become locked, however, these doors can be operated by a key to enter and a thumb switch to egress (which is no different from all security operated doors).
The net effect to safety systems in this area 11 to provide additional pathways and thus enhanced accessibility to the equipment if Operations personnel are required to attend to the equipment during any mode of operation,
                                                                        ,_  ____        .-- ._ _ - _ _ , .            ~  __  _- -        _ _ _ _          _ _,
 
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Evaluation Number SE 90 032 Activity
 
==Title:==
Existing Pipe Supports Considered During Pipe Rupture Event As Hitigating Hardware Description of Change (s):
high energy line break (HELB) pipe whip design criteria delineated in FSAR section 3.6B.2.3.3C is clarified to specifically allow the use of              t pipe support hardware in mitigating the consequences of a HELB.
FSAR figures 3.60 19, 3.6B 20. 3.6B 21 and 3.6B 22 have been revised to reflect those pipe supports credited with providing protection during a HELB event.
Summary of Evaluation:
This change will allow credit to be taken for pipe support hardware in providing protection and mitigating the advervse effects of a postulated HELB when their use as a pipe whip restraint is anelytically justified. These pipe supports are required to function and maintain pressure integrity of the piping system during normal plant conditions in accordance with ASME B&PV Code, . However, during a HELB event, pipe support hardware is analytically shown to be capable of providing protection in accordance with the design rules applicable to structures reqcited to perform an intended safety function during a HELD.
s
 
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==Title:==
Revision Of FSAR Tables 3.90 10 And 3.9N 10 On Active Valves Description of Change (s):
A portion of the change entails the deletion of four Hydrogen Parge system butterfly motor operated valves from FSAR Active Valve Table 3.9B 10. The Hydrogen Purge system is common to both units and is designed to function as e supplementary system to the electric hydrogen recombiners. The deletion of the four hydrogen purge valves is consistent with FSAR sections 3.90.3,1.1 definition of active valves. 6.2.5.1.3 and 6.2.5.2.2 discussion of the Hydrogen Purge system, and Figure 9.4 6.
The remainder of the changes involved several editorial corrections to valve numbering designations reflected by FSAR Active Valve Tables 3.9N-10 and-3.9B 10.
This change was made to complete declassification of the Hydrogen Purge System which was declassified in FSAR Amehdttnt 70 from ESF to non safety related.
Summary of Evaluation:
The removal of the four Hydrogen Purge Butterfly ukter operated valves, which do. not provide isolation, f rom the Active Valve Table is consistent with the system classification as non n9:\ter safety (NNS) related. .lsolation of this system from contairmert is provided by Containment Isolation Valves (CIV). Classifice-tion er NNS is commensurate with the Hydrogen Purge system func'; ion as a supplemental system to the Electric Hydrogen Recombiner and no credit is taken for Hydrogen Purge system operation during an sccident or achievement of safe shutdown.
The part of this safety evaluation that pertains to the incorporation of editorial corrections to valve tag numbers provided in Active Valve Tab'les 3.9N 10 and 3.9B 10 involves no failure modes or change to the design basis. Implementation of correct valve tag numbers on the Active Valve Tables provides consistency and accuracy with the as built plant configuration contain'ed in design bases documentation, drawings and procedures.
 
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Evaluation Number SE-90 035 Activity
 
==Title:==
Instrumentation and Control Circuits thtt Retnain Associated after the Isolation Device Description of Change (s):
FSAR Section 8.3.1.2.1.7b has been revised to provide a discussion regarding instrumentation and control cables that connect non Class IE devices to Class 1E systems, but are routed as associated cables beyond the isolation dtvice.
Summary of Evaluation:
At CPSES, Class 1E 1ssociated circuits are treated the same as the Class 1E circuits with which they are associated. Scme instrumentation and control circuits are routed as associated and remain associated af ter the isolation device (i.e. the isolation device is between the Class IE circuit (s) and associated circuit (s) rather than the associated circuit (s) and non Class 1E load (s)).
However, in cccordance with the requirements of IEEE-384-1974, Section 4.6.2, "Non Class IE circuits shall be separated from associated circuits by the minimum separation requirements specified in Sections 5.1.3, 5.1.4, or 5.6.2 or (1) the effects of lesser separation between non Class 1E circuits and Class 1E circuits shall be analyzed to demonstrate that Class 1E circuits are not degraded below an acceptable icvel or ...."
The CPSr5 associated' instrumentation and control circuits affected by
    ' the FS LR- revision have protective devices or current limiting features      '
of their power supplies. A fault or overload condition in the non-Class 1E portion of the circuit will be interrupted by the circuit protective device _or limited by the current carrying capability of the circuit, due to the current limiting feature of the power supply:
therefore will not degrade the associated circuit.
Class 1E control and instrumentation circuits are protected by isolation devices which ensures that a fault or overload condition in the non Class 1E circuit (s) does not degrade the Class 1E circult(s).
Therefore, although the associated circuits are not isolated from non-Class IE circuits lesser separation has been deemed adequete based on individual circuit protective devices or current limiting features of the_ power supplies.
This design philosophy has been previously applied to power circuits, as described in FSAR Section 8.3.1.2.1.7a.
I e                . , . - -          , - - , -          -
                                                              -e .
 
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==Title:==
Modification to the fire Protection methodology f or tracking Actual Fire Load (AFL) values during plant operations.
Description of Change (s):
l The FPR provides two (2) cumbustible loading values for each fire i                                                                                                                    area / fire zone. Specifically, an Actual Fire Load (AFL) value is given to represent the in-situ combustible materials present at the time the FPR was issued (ir, situ combustibles are materials designed into the plant includino equipment, permanent storage, exposed cable, etc.).                Also, a Haximum Permissible Fire Loading (HPFL) which can be expected to be contained within an area without compromising fire safe shutdown capability.
This activity involves ellowing fire lood values (BTV/ft2) in plant areas to exceed the values provided in the FPR.
At any given Line, the AFL value associated with given fire areas / fire zones may exceed " base line" AFL values stated in the FPR.
Required administrative / procedural controls for monitoring and documenting in situ coinbustible changes in the plant to assure HPFL values are not exceeded.
The requirement to process separate change, for each chance in type, quantity or location of in situ combastibles incurred during plant operation is eliminated.
Summary of Evaluation:
Implementation of this activity does not affect plant systems.
Administrative / procedural controls assure Haximum Permissable Fire Load values are not exceeded without appropriate compensatory measures.              This activity does not introduce potential ignition sources into the plant, and the types, quantities, location and configuration of added combustible materials are evaluated for potential impact on fire safe shutdown capability.              The consequences of a fire in any plant crea has been assessed and documented in the FPR.
 
Attachment to TXX 91090                                                                                                                    TV Electric Page 17 of 90                                                                                                                              Unit: 1XN Evaluation Number SE-90 037 Activity
 
==Title:==
Addition of interpose Relay Between Electro Thermal Links and Fire Detection Extinguishing Board Description of Change (s):
Isolation relays have been added between the Electro Thermal links (ETL) and Fire Detection / Extinguishing boards. The relay is located in close proximity of the ETL (closest junction box and the remainder of the ETL Circuit (pig tail) is run in a metallic conduit and in accordance with industry standards (NFPA Code 720 and approved specs / procedures). This prevents spurious operation of ETL during transient which could in turn actuate the dampers for UPS and Battery Rooms. Halon is the fire suppression system for these rooms. This arrangement ensures that the ETL operated dampers close in response to either holon release or upon detection of fire in the room.
Summary of Evaluation:
This precludes a potential for open circuits in unsupervised sections of the circuit. By ensuring the closing of dampers when required improves the effectiveness of the Halon fire suppression system.
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Attachment to TXX 91090                                                TV Electric Page 18 of 90                                                          Unit 1XN Evaluation Number 5E 90-038 Activity
 
==Title:==
Compliance with GDC 55 by isolation of root valve (151 8961) for pressure indicator 1 PI 0929 on Safety injection System test line.
Description of Change (s):
The normal position of the Safety Injection System test line pressure            ,
indicator's root isolation valve (151 8961) is revised from normally open to locked closed and placed under administrative controls.
The Safety injection System test line pressure indicator is required to monitor pressure changes in the line as the various check valves in the system are tested for leak tightness, therefore, the isolation valve will only be opened during testing.
This change eliminates an unnecessary open valve (non essential) for a GDC 55 penetration, which previously took credit for the use of a pretsure indicator on the Safety injection System test line as containment boundary rather than utilizing the pressure indicator's root isolation valve. This change has been incorporated in the FSAR as discussed (during FSAR review and prior to Low Power License) and agreed with the Nuclear Regulatory Commission (NRC) Project Manager.
Summary of Evaluation:
This change provides a containment isolation valve arrangement as specified by 10 CFR 50, Appendix A, GDC SS, in accordance with GDC 55, one of the acceptable provisions for containment penetration isolation is to have one automatic isolation valve inside and one locked closed valve outside containment.
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Attachment to TXX 91090                                                  10 Electric Page 19 of 90                                                            Unit  IXN Evaluation Number SE 90 039 Activity
 
==Title:==
U 1ve position changes in the Waste Processing and Boron Recycle Systems.
4                                                                                            i Description of Change (s):
The following discussion outlines the changes and their respective design bases.
: 1. Tank recirculation valves (XBR 8508, XWP-7247.-XWP 7250 XWP 7230 XWP 033.-XWP 7449A. XWP 74498.XWP 7416A.-XWP 0226 XWP 7402 and XWP.
7331) were originally locked in position to provide minimum pump-recirculation. - System _ operation requires that prior to discharge f rom        l their respective tanks, the contents of each tank must be recirculated. This process requires.that the recirculation valves in question be fully open._ After recirculation is complete the                      !
recirculation valves are throttled to a miniflow position for pump protection before discharging the contents of the tank'.
These. changes leaveithe subject valves in an open position and removes the locking device. These changes allow for more operator flexibility
,        .and eliminate some actions that may expose the operator to higher than
          'necessary radiation levels (i'.e'. limit surveillance requirements for locked valves. reduce period of time operator must remain-at subject              ,
volvesito verify or position valves). Leaving the valve open but unlocked provides- adequate recirculation f or pump' protection but increases the likelihood of inadvertent recirculation valve closure                ;
due1to operator crror. Likelihood of failure of the pump due to                ,
inadvertent valve closure is; increased but the pumps failure will not increase _the likelihood of radiation release or breaching of the waste processing. system.
s Thus the increased possibility of pump failure is outweighed by the
        -improvement in AtARA considerations and operator flexibility.
: 2. _
XWP 7201 is. changed'from_a-locked closed to normally closed condition. 'This change is-to facilitate easier switchover between waste processing system and boron recycle evaporator. Allowing this interchange is consistent with system. design and plant requirements
:because~the waste processing (WP) evaporator is1the backup to the                  '
boronfrecycle evaporator. There:is no system or regulatory L  _
_ requirement to__. keep these systems isolated, thus_ locking-XWP 7251 closed islnot'necessary.
3 '.      XWP-7248 is revised from a normally open to normally closed position. This change.is.being performed to isolate the WP-evaporator during normal plant operation. During normal' plant operation the WP-evaporator will not be in operation. Processing of waste holdup tank water will;be through the-filter demineralizer portion of the WP.
4        Remove lucking requirement from valve XWP 0093 to facilitate I                  -
 
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easier alignment of valves that feed the filter deminerlizer system.
No locking requirements exist.
: 5.        Remove locks on valves XWP 0154 and XWP 7412 to facilitate daily operation of wet wash water processing through the LHST filter and strainer to the waste Monitor Tank (WH1).      No locking requirements exits.
: 6.      Remove lock closed position of valve XWP 0121 and change it to normally open due to its inaccessibily.      Valve XWP 0113 is changed to lock closed and XWP 0120 is changed to normally closed to provide adequate isolation requirements.      Changes improve overall ALARA considerations.
: 7.        Per WP operational procedures, valve XWP 743BA, the inlet valve to Waste Honitor lank No. 1 should be normally open.
B.        Close volve XWP 8033 to prevent spillage. No locking is required.
: 9.        Remove lock closed requirements on valves XWP 7469 and XWP 198. No  locking requirements between drain channels "A" and "B" are requii d.
: 10.      Close valves XWP 0001 and XWP 0002 because during normal operation the floor drain evaporator is not used.      The floor drain evaporator is only used when evaporation is the only means to process contaminated water. Closed volves prevent inadvertent admission of wastes to or from the evaporator.
: 11.      Close valve XWP 0221 because the FDT #1 outlet isolation needs to be closed for initial operation.      The contents of FDT #1 must be recycled before discheroe per WP procedures.                                !
: 12.      Close valve XWP 0307 which was originally supplied to provide flush water to XRE 5251 which has been abandoned.
: 13.      Close valve XWP 0102 because the Peverse Osmosis Concentrate Tank Fump is no longer used. The Reverse Osmosis Unit has been abandoned in place.
14      Lock close valves XWP 2271 XWP 0063 and XWP-0064 to prevent inadvertent contamination of the abandoned reverse osmosis portion of the WP system.
: 15.      Remove locks f rom valves XWP 7300 (Normally Open. Spent Resin Storage Tank TDX WPATRS-01 inlet isolation) and XWP-7347 (SRST normally closed drain valve) to facilitate easier operation.        No locking requirements exist.
: 16.      Close valves XWH 0096 and XWM 0097 to bypass the reagent bulk feeder during normal operation.
 
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: 17.      Close valve X HV WH161 to prevent inadvertent discharge to the environment . Normal flow is through the oil / water separator unit prior to discharge to the lake.
Summary of Evaluation:
Based on the discussion in the summary it may be concluded that:
The impact of radiological consequences of accidents considered to be impacted by implementation of this activity has not been increased, The new design of the system (i.e. only normal valve position changes) has not altered any potential radiological consequences that were previously evaluated.                      ALARA and Radiological Environmental Impact reviews have been completed with satisf actory results.
The subject _ changes do not breach the WP system as currently designed nor decrease the likelihood of proper equipment integrity.                        The only increased failure mode is regarding the possibility of WP tank recirculation or discharge pump failures due to changes described in item 1. But since these pumps are not required to provide any safety related active function. no increase in accident likelihood or consequences exists.          (Ref FSAR Section 11.2.2.7)
No new failure modes have been introduced by this change.
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Attachaent to TXX 91090                                                                        TV Electric Page 22 of 90                                                                                  Unitt IXN Evaluation Number SE 90 040 Activity
 
==Title:==
Non use of a 7,5 kilowatt electric unit heater in Room 104 Description of Change (s):
Added a note to Figure 9.4 4 (sh 1 of 2 ) to designate a 7.5 Kw electrical heater in Room 104 of the Safeguards Building as a spare.
This heater was not electrically connected and the associated switches were not installed. The heater is being lef t in place as a spare to reflect the as built installation of the plant.
Summary of Evaluation:
The 7.5 Kw heater is not safety related and there are no safety systems affected by the non functioning of this equipment. The original function of this heater was to maintain the temperature in the Room 104 above 40 degrees F.        A note was added to the FSAR figure to indicate that the installed heater is not connected and may be used as a spare equipment. This note is added to reflect as installed conditions of the plant, 1
y            -e                      -
 
Attachnent to TXX 91090                                                    TU Electric Page 23 of 90                                                              Unit 1XN Evaluation Number SE*90 041 Activity
 
==Title:==
Allowing additional aluminum and/or zine materials in containment during modes 1 4 to support maintenance and/or surveillance activities Description of Change (s):
The quantity of aluminum and/or zine material inside containment during modes 1 4 has been increased to facilitate the use of scaffolding and ladders necessary for maintenance and/or surveillance activities. This activity neither adds any hardware to nor modifies any components in systems triat are involved in the mitigation of Licensing Basis Accidents.
Summary of Evaluation:
The installation of scaffolding and ladders inside containment is administrative 1y controlled to ensure that they will not cause undesirable system interactions.          If these materials remain in containment during a design basis accident, this activity could result in an increased rate of post accident hydrogen production due to corrosion of aluminum and/or zinc. This increase in rate has no effect on the acceptance limit of 3.5 volume percent. This limit is below the concentration at which hydrogen in containment becomes flammable (4 volume percent).
This activity will reduce the time at which the hydrogen recombiners may need to be operated from approximately 12 days to approximately 8 days following a LOCA.        However, these additional sources do not affect the operability of the hydrogen recombiners since a-single hydrogen recombiner has the capacity to maintain hydrogen concentration well below the flammability limit if placed in operation when the concentration reaches 3.5 volume percent, i
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Attachment to TXX 91090                                                  TV Electric Page 24 of 90                                                            Unit: 1XN Evaluation Number SE 90 042 Revision 1 Activity
 
==Title:==
Removal of electrical interlocks between non engineered safety features (ESF) supply and ESF exhaust fans for the two of the PPVS.
Description of Change (s):
The Primary Plant Ventilation System (PPVS) interlocks between supply fans CPX VAFNAV 23 and 24 and exhaust fans CPX VAFNCB 21 and 22, were removed.
Exhaust fans (CPX VAFNCB 21 and 22) were upgraded to ESF status to provide additional capacity for the PPVS to maintain a slight negative pressure in the negative pressure envelope.      Their respective interlocks with non ESF supply fans CPX VAFNAV 23 and 24 were consequently removed.      The removal of the interlocks allows the supply fans to be utilized as standby supply fans.      If any supply fan is down for maintenance and additional cooling is required, these fans could be utilized as long as the negative pressure in the negative pressure enevelope is maintained.
Summary of Evaluation:
        -During a !oss of offsite power or an inadvertent black out signal (B05) (due to loss of a safety bus or other reasons) without a l'oas of off site power, all exhaust fans are automatically tripped.        If either supply fan CFX VAFNAV 23 or 24 continues to operate, the negetivo
        . pressure envelope could be affected. Even if the negative pressure were lost for a short duration, the operator would read an altrm annuniciator-and would take the appropriate action. To prevent this situation, a precaution is provided to the operators to manually t rip these fans in the event of a B05.
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==Title:==
Alternate discharge path for the Gas Decay Tank Description of Change (s):
This activity revises  the procedure for discharging water from the Gas Decoy Tank to provide  for an alternate discharge path in the event that the normal paths  are unavailable. The normal discharge path from the Decay Tenk is via  either the Volume Control Tank (VCT) or the Recycle Holdup Tank  (RHT). This alternate discharge path is via the Waste Gas Drain Pump to the Waste Holdup Tank.
Summary of Evaluation:
The primary difference between the normal discharge paths and the alternate path is that with the normal path tanks are vented back to the waste gas system whereas the alternate path Waste Holdup Tank is vented to the atmosphere. This does not create a different accident or increase consequences of an analyzed accident, since the accident described in the FSAR assumed the release of the Gas Decay Tank to the atmosphere with no particular leakage path specified. The probability of the accident is not increased since the alternate flow path is only used in the event of unavailability of the normal discharge paths and does not represent a change in operational philosophy, i
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==Title:==
Addition of second fume hood in the Chemistry Hot Lab (Room 30)
Description of Change (s):
This change adds a second fume hood to the chemistry hot laboratory located in the office and service areas. The addition of second fume hood provides additional capacity to support chemical analysis activities for Unit 1 operations. The fume hood exhaust is vented to the Primary Plant Ventilation System (PPVS).
Summary of Evaluation:
The addition of second fume hood in the chemistry laboratory will provide more equipment to support the chemical analyses and does not have any effect on the performance of any safety systems. The fume hood exhaust is filtered through the PPVS exhaust air filtration units which minimizes the potential airborne contamination.
 
Attochment to TAX-91090                                                TV Electric vspe 27 of 90                                                          Unit: 1XN Evaluation Number SE 90 U45 Activity
 
==Title:==
Relocate Auxiliary Steam High Energy Line Break (HELB) Pressure Taps Description of Change (s):
The Auxiliary Steam System has two headers which respectively supply steam to the Floor Drain Waste Evaporator Package and Waste Processing System (WPS) Waste Evaporator, and the Chemical Volume Control System (CVCS) Boric Acid Batching Tank and the Boron Recycle System (BRS)
Recycle Evaporator Package. Each header has two pressure sensors (located on a common top) for detection of a HELB. The sensors were located on the headers such that if the isolation valves for the WPS Waste Evaporator Package or the BR$ Recycle Evaporator Package were shut, then the sensors were isolated from the steam headers,      in this configuration, 11 the WPS Waste Evaporator Package or the BRS Recycle Evaporator Package were secured, then the CVCS Boric Acid Batching Tank or the Floor Drain Evaporator Package could not be operated separately without losing HELB detection capability.
The modification moved the sensor taps upstream of the isolation valves for the WPS Waste Evaporator Package and BRS Recycle Evaporator Package,    in that location, any combination of steam loads can be operated without isolating HELB detection capability.
Summary of Evaluation:
The modification has no impact on the operation of the Auxiliary Steam System or HELD detection capability because the difference in pressure readings between the old and new sensor tap locations is minimal. In addition, since the change was physically minor in nature, the change does not impact any accidents nor introduce any new failure modes.
 
Attachment to TXX 91090                                                TV Electric Page 28 of 90                                                          Unitt 1XN Evaluation Number SE 90 046 1
Activity
 
==Title:==
Addition of a Sample Sink to Secondary Sampling System Description of Change (s):
The Secondary Plant Sampling System is designed to monitor water chemistry in the secondary systems. The Condensate Storage Tonk upper and lower volumes are sampled for dissolved oxygen and cation conductivity. In the event a grab sample is desired, a valve on the sample line allows the drawing of such a sample.
A sample sink has been added to facilitate the drawing of these grab samples. The sink prevents grab sample fluids from spilling on the floor and provides controlled draining of samples. Sampling will be enhanced by having a confined-area for fillir.; sampic bottles.
Summary of. Evaluation:
No credible failures have been identified with this change. The sample sink has been designed to Seismic Category 11 critierie to prevent interaction with the surrounding safety class components.
Tubing failure has been analyzed. The maximum system flow rate from the Condensate Storage Tank (CST) via the Secondary Sampling System would not' affect the safety function of the CST. Additionally, the floor drains have sufficient capacity to prevent flooding in the event of a tube rupture.
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==Title:==
Revision to SWIS groundwater elevation described in FSAR 2.5.4.10.3.
Description of Change (s):
The Service Water Intake Structure (SWIS) was reanalyzed utilizing a more conservative design basis groundwater elevation of 793 feet, which is the probable maximum flood (PHF) level for the Safe Shutdown impoundment including wave run-up at the SWis. This envelopes the highest groundwater level expected f or the life of the plant.
Summary of Evaluation:
A calculation was performed which demonstrated that increased loads sn
        -the SWIS did not affect its stability and integrity for a groundwater elevation of 793 feet and the SWIS is still within the acceptance criteria as specified in FSAR sections 3.8.4.5 and 3.0,$.S.
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==Title:==
Revision to the Control Logic on AFW Pump Discharge Control Valves IPV 2463 ALB and IPV 2454 ALD Description of Change (s):
The control logic for the auxiliary feedwater pressure control valves is changed to correct a flow imbelance in the automatic mode.
Testing demonstrated that the signal / valve response times between the two parallel valves in each motor driven AFW pump discharge results in significantly trismatched flows to the steam generators.
In addition, the change corrects a control anomaly for an $1 initiated start of auxiliary feedwater pumps where the sequencer delayed start of the motor driven putops allowed the automatic pressure control to close the control valves fully before the putops started.
The automatic features of prest,ure control for both motor driven AFW pumps are deleted. The trip to " auto" functions of High 1 flow coincident with low discharge pressure and High-2 flow are removed.
Any automatic start of auxiliary feedwater will result in on automatic opening signal to each valve. Thereafter, pressure / flow control is accomplished manually.
Summary of Evaluation:
The automatic control featurec of these valves were initially installed to protect the AFW pumps from runout during heotup and cooldown transitions. During design modification post work functional testing of the AFW system, it was demonstrated that the AFW piping 6        provides sufficient inherent resistence for limiting pump runout and thus the automatic control features of the valves are not required.
The increased delivery of AFW flow to the S/Gs during a postulated Feedwater line break accident was also analyzed considering the auto opening feature of the valves.      It was coricluded that this does not adversely affect the accident analyses.
 
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4 Attachment to TXX 91090                                              TV Eloctric    l Page 31 of 90                                                        Unit: IXN      l Evaluation Number SE 90 050 Activity
 
==Title:==
I'rovide submersible pump for increased flow to 1 RE 5100 in the                I Turbine Building Sump No. 2 Devription of Change (s):                                                            l The previous radiation monitoring system for the Turbine Duilding Sump h 2 received flow from an orifice located ir the common discharge header from the sump pumps. This orifice supplied approximately 4 gpm when the high capacity sump pumps were operating (generally a short period). Due to the low sampling flow rate and the short duration of punp operation, the sampling line intertals (including a sample flow switch) were susceptible to clogging from dirty samLle water.
The sump sampling system was modified to provide for a seperate submersible pump to provide approximately 20 gpm for sampling. The piping configuration and flow switch were also modified to be more resistant to clogging failure. The higher flow rate from the continuously operated sump pump will keep sediment flushed a,td also            >
provide a more representative sample of the water being discharged.
Summary of Evaluation:
Faili.re modes for the revised sampling system were reviewed. Failure of this sampling system results in the same actions that were required under the previous system. However, failures are expected to be less frequent as a result of this modification.
The only effect of the implementation of this change is to supply more sample flow (i.e., from 4 gpm on a periodic basis to 20 gpm or, a continuous basis) to the Radiation Honitoring System. The setpoint of the radiation monitor, and therefore, the radiation levels allowed to be released have not changed. The flow switch function (i.e., on a low flow, to alarm in the control reora and to divert sump flow from the Low Volume Waste retention pond to the Co Current Waste Treatment System) has not changed.
Because the modification only changes the means of supplying samples to the Radiation Monioring System and does not affect the Radiation Monitoring system itself, there is no impact on any Technical Specification or ODCH requirement and no impact on the radiological consequences of any accident.
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==Title:==
Watertight Door Position Updates Description of Change (s):
                          'he philosophy used for determining " Watertight Door Positions" of Un!t 1 and Commun watertight door operation has been updated to provi d e flexibility of operation for the convenience and safety of operation personnel working in the rooms without affecting the ability to close and/or successfully latch the door for credible accidents.
The dooro S26 and $28 (DG Building) are security doors as well as fire protectio 1 doors. These doors are required to be shut and at least one latch engaged after every ingress or egress.      This ensures the room boundary is maintained for plant security as well as for the fire protection program. Other doc"s $2. 53. 54, 55, $7 and 59 (in ECCS pump rooms) and A13 (in waste holdup tank room) may be left open during plant surveillance round. This is based on the premise that accidents (line break) pertaining to these rooms are slow acting providing amole time for the operator to close these doors.
Summary of Evaluation:
The postulated accidents for the D/G rooms, the ECCS pump rooms, and the waste holdup tank room are either not severe enough to require the doors be watertight, or are slow enough in occurring so as to allow ample time for operator action to secure the door.
The postulated accidents for the main steam door area is a fast accident that would surely incapacitate personnel in the area: however the piping is considered in a break exclusion area and is not postulated for pipe breaks occurring. The analysis in the SAR    was required by the NRC for evaluation of environmental concerns for equipment in the pipe tunnel area. The hypothesis for this evaluation is that the probability of a licensing basis main steam or main or auxiliary feedwater line break in the break exclusion area simultaneously with a personnel entry where the door was closed and fastened with at least one latch, is sufficiently low as to not be credible,    it is not expected that this area will be a highly trafficked area and access will be for operator rounds and necessary maintenance, and that much of the maintenance will not be performed with the plant at Mode 4 or above. This conclusion is supported by a qualitative review of the SAR analysis for this area as a break exclusion area thus indicating the surely low probability of failure along with the limited time when the doors will be latched with less than all latches secured.
 
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Activity
 
==Title:==
Eff ect Of High Energy Line Break (HELB) On Control Room Habitability Description of Change (s):
This activity involves only the documentation that a HELD has no effect on control room habitability. The FSAR change incorporates the conclusions from an analysis of the effect of a postulated $ team Generator Blowdown High Energy Line Break (HELB) in the Electrical and Control Building on control room availability. The analysis shows that a HELB in the Electrical and Control Building will have no ef fect on control room habitability.
This supports the change to the response to NRC Action plan 111.D.3.4 and substantiates compliance with SRP section 6.4. paragraph 111.5.c(3).
Summary of Evaluation:
The analysis considered the eff ects of a postulated HELB and the ability of the plant design to limit the amount of steam capable of entering an adjacent stairwell that leads to the control room.      The analysis also verified that the f esulting pressure in the adjacent stairwell due to the postulated Steam Generator Blowdown HELB is limited to levels that would not compromise the control room doors.
 
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==Title:==
Provide vendor taps on cd current Weste System Description of Change (s):
This modification provides piping connections for air and water and power supplies to a temporary filter skid which is to be used to reduce the total suspended solids in the Co-Current Waste System.                                                          By limiting the suspended solids, the total quantity of. radiactivity contained in the outdoor storage tanks can be reduced.
Summary of Evaluation:
All equipment associated with this modification is non safety related and located in the turbine building. The Co Current Waste System handles only low level-radioactive waste. The ALARA and environmental reviews of this modification have been performed and determined to have no adverse impact on health and safety of the plant personnel and public. Periodic surveillance requirements for analyzing samples f ron.
the temporary skid ensure that there will be no uncontrolled releases exceeding those previously analyzed (i.t., FSAR Sections 9.2.8.2.2 and                                                                '
1*.S.3.2).
 
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Evaluation Number SE 90 055
                                                                                                  )
i Activity
 
==Title:==
Revise Process Sampling System Conditioned Sample Temperature and flow Description of Change (s):
The Process Sampling System is designed to provide sampling capability for monitoring chemistry parameters in various primary systems. To ensure representative samples, minimum times for purging sample lines, minimum or maximum sample fluid temperatures, and minimum sample flow rates are specified if the chemistry parameters are affected by these system operating parameters.
The sample temperature is specified in the Final Safety Analysis Report as 105 degrees plus or minus 10 degrees. The sample                                ,
temperature has been changed to "less than 120 degrees,"            In addition, the minimum flow requirements for sampling the influent and effluent of the Chemical Volume Control System (CYCS) Purification Demineralizer have been removed.
Summary of Evaluation:
The sample temperature was changed to "less that 120 degrees," The temperature limitation is for personnel safety and changing the upper limit from 115 to 120 is inconsequential. The lower limit of 95 was removed because none of the chemistry samples are measurably affected by the lowet,t expected sample temperature.
The removal of minimum requirements for sampling the influent and effluent of the CVCS Purification nemineralizer does not significantly affect the measured chemistry parameters, chloride and activity.
The change to the Process Sampling System did not require any hardware changes. The sampliig temperature and purge / sample flow rates do not affect any licensing basis accidents.
 
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Activity
 
==Title:==
Isolation of the Laundry Reverse Osmosis Chiller Package from the Turbine Plant Cooling Water System Description of Change (s):
The Turbine Plant Cooling Water (TPCW) System provides cooling for various secondary and auxiliary systems including the Laundry Reverse Osmosis Unit. The Laundry Reverse Osmosis Unit will not be used, but instead of being removed, has been abandoned in place. The supply and return isolation valves which provide TPCW to the Reverse Osmosis Chiller package have been changed from normally open to normally closed since cooling is no longer required.
Summary of Evaluation:
Turbine Plant Cooling Water is a non nuclear safety system and has no impact on any Itcensing basis accidents.        The only potential failure mode associated with this change involves a f ailure of the supply or return isolation valves that are being changed from normally open to normally closed.        If the valves failed open, then the system configuration would be the same as before this change, so this change does nat creete any new or unanalyzed conditions.
1
 
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==Title:==
Discharge from waste water hnid up tanks to low volume wa~te retention pond, Description of Change (s):
Revises the ODCH the FSAR and the implementing procedure STA 603 to allow water from the Waste Water Holdup Tanks (WWHis), of the Co-Current Weste System, to be discharged to the Low Volume Waste retention ponds for further environmei.tal water treatment and subsequent discharge, if the required sampling confirms that its radioactivity is less than the specified Lower Limit of Detection (LLD).
To perform this option, the Waste Water Holdup Tanks are drained into the Unit 2 Turbine Building Sump # 4 and then transferred to the Low Volume Weste Retention Pond, To provide assurance against inadvertent                        i discharge, sampling requirements have been added for Sump # 4 and if the activity is found to be above the LLD the water is pumped back to the WWHis, Summary of Evaluation:
The Low Volume Waste System is designed to handle normal wastes with activities less than LLD, The Co Current Waste System is designed to handle radioactive waste, The described changes allow water from the WWHTs, that is proven by sampling to have activity less than LLO. to be transferred to the Low Volume Waste System (Turbine Sump #4). The Turbine Sumps normally discharge to the Low Volume Weste Retention Pond. The Sumps are diverted back to the WWHis if known activity is present in the sumps. This transfer between systems is based on the condition of the waste being radioactive or non radioactive (activity less than LLD) and does not impact the operation or failure modes associated with these systems. The normal discharging or transferring of waste between systemt does not introduce any new failure modes because no new intreduction of radioactivity into systems not previously containing radioactivity occurs, nor is there any increase in the concentrations of existing radioactive liquids as a result of these changes, i
 
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Decrease of Pressure Differential Detween the RCS and the Charging Header Description of Change (s):
The differential pressure between charging header and the Reactor Coolant System (RCS) has been revised from approximately 190 psid to 125 psid to allow the Reactor Coolant Pump (RCP) seal injection isolation valves to be opened such that the total resistance in the seal injection flow can be reduced. This results in a reduction in the flow rate from the centrifugal charging pumps (CCP) into the cold leg during safety injection. Because the use of the lower differential pressure between the charging header and the RCS, the position of the RCP seal injection isolation valves can be more accurately set, thus allowing increated certainty that the CCP safety injection flow is within its analyzed value.
Summary of Evaluatian The reduction in the CCp safety injection flow rate has been evaluated for induced loads (FSAR Chapter 3), Containment Response (Chapter 6),
and all accident scenarios in FSAR Chapter 15 in which the ECCS is actuated and the results of the evaluation indicated that the proposed p duction in the ECCS performance did not result in any design or regulatory limit being exceeded, and there is no increase in the consequences of any of these events. Thus, it is concluded that the activity would not decrease the margin of safety as defined in the
    ' basis listed below for the Technical Specification:
* Hinimum DNP ratio greater than 1.30; Peak Cladding Temperature less than 2200 F:
Peak Containment prcssure less than 50 psig.
 
AttachGent to TXX+91090                                                TU Electric Pave 39 of.90                                                          Unit: 1XN Evaluation Number SE 90 059 Activity Title                                                                      -
Revision to RPI-312 R3 to remove fuses DescriPtion of Change (s):
Procedure RPI-312 addresses the changing of Iodine and Particulate filters.in the Containoent lodine, and Gas    (P!G) radiation monitor.
Procedure RPI-312 R3 tas changed to reqeire removal of a set of fuses in the Balance of Plant (00P) relay racks, Removal of the fuses in the DOP relay-rack 5 blocks the containment ventilation isolation signal from the PIG ionitor, Summary of. Evaluation:
This activity defeats or blocks the high radiation input to the containment ~ventilal on isolation circuitry during the filter
    .changeout maintenance. Removal of the fuses does not affect the ability of the Safety Iniection (SI) input to initiate containment ventilation isolation. The flG is declared out of service during maintenance, and administrative controls exist to ensure that the Particulate, lodine, Gas (PIG) monitor is restored to as-foind after filter replacement, Blocking of the containment ventilation isolation signal from an out of service monitor does not introduce any new              ,
accidents.
s            ..
 
      - . - ~                      .          .  .-            -_        -      -
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Page 40 of 90                                                          Unit: 1XN Evaluation Number SE 90 060 I
Activity
 
==Title:==
Revision of process sample system pressurized sample vessel or bomb, (FSAR Section 9.3.2.2.2 and Figure 9,3 4 Sheet 2)
Description of Change (s):
The redesign shortens the length of tubing associated with the vessel and deletes the vessel's pressure gauge, Also the redesign adds an isolation valve prior to the quick connect fittings for the vessel, Sample vessels or bombs are used for all gas sampling and reactor coolant, pressurizer and RHR system liquid samples. The vessels are of stainless steel construction designed to withstand the reactor coolant design pressure and temperature, Summary of Evaluation:
The sample vessel's redesign reduces the potential for leakage by eliminating some of the fittings and reducing the overall weight of the ve,sel assembly. The additional isolation valves reduce the poscibility the quick disconnects will fail in the uncoupled position.
The sample vessel and the associated tubing are hydrotested to 3000 psig to ensure component integrity. The fittings are seal welded to preclude leakage.
The pressure gage primarily indicated the presence of fluid in the vessel. Administrative controls will ensure the sample vessel is discharged before use and the samples are properly identified, The new design is more rigid and has less mass than the old design, therefore no new events / accidents are created. The affected portion Lof the process sample system is non safety related. This design modification does not adversely affect plant safety and is considered acceptable.
 
                      . Attachment to TXX 91090                                                                                                      70 Electric "Page ' 411 of190 :                                                                                                            Unit: lVH Evaluation Humber SE 90 061
:: Activity'
 
==Title:==
Correction-in.the maximum' allowable opening-of containment pressure relief. valves from 70 to 65 degrees.
Description of Change (s):
The' maximum allowable opening of two 18 inch purge isoittion valves (1 HV 5548 and 5549)'is corrected from 70 to 65 degrees.
                              'The-reason-forlthis change is to provide consistency between the the FSAR-(Figure 9.4 6) and the manufacturer recommendation and NUREG 0797 Supplement 23,isection 22 item II .E.4.2.                                                                                          .
Also.Ethe' locked closed requirements depicted in FSAR Figure 9.4-6 are changed.
The; purpose of the valves is to isolate the containment during and afteroat accident.-.0therwise, they are utilized to relieve pressure                                                            ,
                                ;that may,egist inside containment during normal operation.
                        -Summary of' Evaluation:-          .
By limiting the: maximum allowable opening of these valves from 70 to
                                -65Ldegrees ,the ability for the valves to close under LOCA conditions                                                            i is enhanced.'- In reality, since the valves are set below this allowable: maximum opening,ithe. change'willinot have any direct. impact on the operation of the pressure' relief system in which they are installed nor will it affect =their closure time after a LOCA conditioni; The: valve manufacturer =(Posi-Seal) provided a report-(Report-#34977SL-001) that indicates that the valve can--close within five (5). seconds against.a~ pressure dropioffftfty-(50) pounds per square inch at a-
                                ;settingJof 70 degrees. However, it is recommended that the setting nci exceed 65 degrees. :This will ensure 1 closure and provide a good-margin c of safety.
                                'The!effect this change will'have is negligible since the valves are
                                -set at 55; degrees' (for-1LHV 5549:CP1-VADPBC-9) and 53 degrees (for 11 HV-5548:CP1-VADPBC-10)-and were tested to close within five seconds.
Furthermore, these valves-were inadvertently shown'as " Locked Closed"'
                                  -iri FSAR F.igure:9.4a6 which was contrary to design requirement.
There'are noLsafety systems and systems important to safety' considered
                                              ~
that-are potentially affected by the implementation of this activity.
 
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==Title:==
Unique identification of tag numbers between electrical panels and motor control centers, Description of Change (s):
Technical Requirements Manual (TRM) Table 4,1-1 has been revised to reflect unique tog numbers between pressurizer heater panelboards (1EB1-1, IEB1 2, IEB2 1, 1EB2 2, IEB3 1, 1EB3-2, 1EB4 1, and 1EB4 2) and non safety related 480V motor control centers (MCC's),
Summary of Evaluation:
Prior to the Technical Requirements Manual (TRH) revision, the safety-related pressurizer heater panelboards and several 480V HCC's had identical tag numbers. The duplicate use of the same tog number for two different pieces of equipment created some confusion and could have potentially created operating errors during plant operations, Therefore, to preclude future confusion, the tag numbers for the panelboards were relabeled to 1EB1-1, 1EB1 2, IEB2 1, 1EB2 2, IEB3 1, IEB3 2, 1EB4 1, 1EB4 2,                        No change was made to the MCC tag numbers.
The TRM revision affects applicable plant documents, such as drawings, and the panelboard tag numbers, only, and do not affect the equipments' ability to perform their intended functions,
 
Attachment to TXX 91090                                                TU Electric
  'Page 43 of 90                                                          Unit: 1XN Evaluation Number SE 90 063 Activity
 
==Title:==
Procedural revision to delete prerequisite step requiring that all RCCAs be inserted prior to initial criticality.
Description of Change (s):
The prerequisite step to procedure ISU 022A, Revision 2. " Reactor Coolant System Leakage Rate Test" and FSAR Table 14.2 3, " Sheet 29.
Reactor Coolant Leal Test Summary", requiring that oli RCCAs be inserted with the reactor shutdow: orior to initial criticality was revised to delete the requirement    iat all RCCAS e inserted.
Summary of Evaluation:
      -The basis for the prerequisite that "the reactor is shutdown prior to initial criticality" is to ensure that the reactor leakage is within technical specification limits prior to increasing the radioactive cortent of the coolant by achieving initial criticality. The requirements for the reactor shutdown and for testing prior to initial criticality is unchanged._ The position of the RCCAs has no bearing to the validity of the test results. The position of the RCCAs are controlled by other initial startup procedures and are not necessarily fully inserted. This activity complies with RG 1,68 requirements that the sequence of startup tests be ordered so that safety is not dependent un untested structures, systems or components.
 
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Page 44 of 90                                                        Unit: 1XN Evaluation Number SE-90 064 Activity
 
==Title:==
Addition of a new sodium hypochlorite and sodium bromide system for the Station Service Water System.
Description of Change (s):
This design modification adds a new chlorination system for the treatment of organic bio fouling in the sevice water system. The new system consists of skid mounted pumps and storage tanks to feed sodium hypochlorite and sodium bromide into the existing non-safety service water dilution line. The equipment is located in the non-seismic sevice water chlorination building. The two tanks are vented to the outside of the building and a separate spill catchment is provided for each tank. Relief valves are provided on the discharge of the positive displacement pumps to protect against system over-pressurization Summary of Evaluation:
The implementation of this modification will preserve the Service Water System in that it is designed to control organic bio fouling.
This function was previously performed using a gaseous chlorination system requiring the on-site storage of chlorine gas. As such this change represents an improvement in safety as related to control room habitability considerations.
The equipment added by this modification to inject chemicals is not safety related nor does it adversely impact any existing equipment which 's safety related. Testing has been performed at CPSES and at other utilities which have determined that the hypochlorite and bromide are compatible with the materials used in the Service Water System.
 
I                  ' Attachment to TXX I91090-                                                                                                                                    TV Electric Page 45 of 90-                                                                                                                                                Unitt IXN
  ,1 Evaluation Number SE-90 065 LActivity
 
==Title:==
Operability requirement of one filtration unit per train                                                                                              Technical
                                . Interpretation No 010.
Description of Change (s):
This evaluation 11s performed for the Technical Interpretation No 10 Revision 0 for Technical Specification section 3.7.8 of 3/4.7.8. The interpretation provides that one unit per train is sufficient to meet the train operability-requirement of the Technicel Specification 3/4.7.8.1 .Following a LOCA, four safety related exhaust fans (two from Train A=end two from Train B) and ESF. filtration units are available for post-accident ECCS filtration for maintaining a slightly
                                . negative; pressure'in the Auxiliary, Safeguards and Fuel Buildings.
During testing of the' fire protection-deluge system, an accidental water _ spill-resulted in wetting of the portion of charcoal for Train A
                                '_ESF-filter. CPX VAFUPK-15.- _The. associated exhaust fan, CPX-VAFNCB 21 was manually tripped to prevent operation of the fan motor until the wet charcoal could be, removed. -With only the remaining-Train A
                                  . exhaust f ani CPX-VAFNCB 07.- and associated ESF filter, CPX-VAFUPK-01,
: operable _and assuming a. single failure (i.e-Train B), the-exhaust fan
                                  .and filter.would have-been the only ESF ventilation unit available.
                              . Summary of_ Evaluation:
CPSES was originally designed with only one safety-relattd exhaust fan
                                  -and ESF filter per-train. The minimum exhaust-flow of one fan is 15,000LcfmW nominally,-.-which isiconsidered; adequate to maintain the inegat'ive pressure boundary at slightly negative pressure conditions.
DTo provide additional 1 conservatism:in-the-design _and_ provide-additional-margin to' attain the desired negative pressure. both Trains were upgraded to' include a second safety-relsted exhaust f an and filter. EThe upgraded design ensures 1that potential ECCS leakage.-
post-LOCA is captured ar.f filtered' by ESF filters . prior to release.
The' current system design is _ verified by an 18 month-surveillance
                                  -wh'ich measures the pressure _within-the negative pressure envelope                                                                                                            .#
                                    ;using two exhaust fans =and filter units. -However..one1 exhaust fan and-filter'has.been_ tested per test procedure. EGT-TP 90A 15, and yielded satisfactory =results.---Therefore as a one: time only configuration.
                                  ? operating with a single exhaust fan .and filter unit' did not adversely-
                                    'af fect plant: operations. = However,n future operations in a similar configuration would require demonstration of the. single exhaust fan-and filter? unit's perf ormance, since - the current Lsurveillance - takes -
into consideration both filtration units per Train.
                                                                                                                                                                                                                  -l
 
                                                                                  }
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==Title:==
Identifying Unique Tag Numbers between a Switchboard Panel and Motor Control Center Description of Change (s):
FSAR Figures 8.3-1 and 8.3-7, Sheet 2 have been revised to reflect unique tag numbers between non safety related switchboard panel, 102 1, and non safety related 480V motor control center (HCC), 182 1.
-Summary of Evaluation:
Prior to the FSAR revision, the pressurizer switchboard panel and one of the turbine building 480V HCC's had identical tag numbers, 102-1.
The duplicate use of the same tag number for two different pieces of equipment created some confusion and could have potentially created operating errors during plant operations. Both pieces of equipment are non-safety related equipment. Therefore, to preclude future confusion, the tag number for the switchboard panel was relabeled to 182-1-1. No change was made to the HCC tag number.
The-FSAR revision affect applicable plant documents, such as drawings, and the switchboard panel tag, only, and does not affect the equipment's-ability to perform its intended function.
 
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==Title:==
Changes in the Operations Review Committee (ORC) Hembership for the Plant Manager and Director Quality Assurance (0A)
Description of Change (s):
Revised FSAR Sections 17.2.1.1,3 and 17.2.1.1.4 to remove the
  . requirements of mandantory membership to ORC for the Plant Manager and Director, GA. These changes reflect that the requirement to serve as a member of ORC is by oppointment only by the Executive Vice President. NEO.
Summary of Evaluation:
Prior to the FSAR revision, FSAR Sections 17.2.1.1.3 and 17.2.1.2.4 designated the Plant Manager and Director. 0A as having responsibilities of being ORC members. This was interpreted that membership to ORC was by job title.      The ORC performs independent reviews and audits of designated safety related activities and issues pertaining to CPSES operations, lhe nembership to the ORC is by appointment only by the Executive Vice President, NE0 and therefore membership by title need not be mandated.
This is an administrative change to clarify the requirement for ORC membership in the FSAR, such that it is consistent with the ORC Hanual and Techiical Specifications. Section 6.5.2,    In addition, no changes to the current ORC membership was mude based on the change. In fact, both the Plant Manager and Director. 0A are present members of ORC.
 
Attachment to TXX 91090                                                                                                                                  TV Electric Page 48 of 90                                                                                                                                            Unit: IXN Evaluation Number SE-90 068 Activity
 
==Title:==
Reanalysis of feedwater System Halfunction Accident, FSAR Section 15,1 Description of Change (s):
While preparing the response to Generic Letter 89-19, " Unresolved Safety Issue A 47, Safety implication of Control Systems in LWR Nuclear Power Plants." it was discovered that a postulated failure of either inverter 1-PC-1 or 1 PC-2 could cause two feedwater control valves to fully open, resulting in the overfeeding of two steam generators rather just one steam genetator as assumed in the Final Safety Analysis Report (FSAR), section 15,1,2, "Feedwater System Halfunctions that Result in an Increase Feedwater Flow.                                                                                      The feedwater malfunction accident was reanalyzed using the new scenario,                                                                                      This safety evaluation was performed to evaluate the new results and the change in analysis assumptions for the FSAR.
Summary of Evaluation:
The reanalysis of the feedwater malfunction accident determined that the DNB ratio was slightly closer to the limit of 1,30 when two feedwater control valves fully opened.                                                                                      However, since the ONB ratio was still maintained greater than the 1,30 minimum, there was no impact on the margin of safety or radiological consequences of the accident, The FSAR has been changed to include the overfeeding of two steam generator,-but since the safety criterion of this accident (minimum DNB ratio of 1,30) was not violated, the conclusions of the
                        -FSAR remain unchanged.
 
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==Title:==
Physical security program revision.
Description of Change (s):
The Physical Security Plan is revised to address pre-licensing inspection comments.      The changes address oraanization, key and lock control, vital area discription and controls, updated security facilities, drawing corrections, emergency access and program / design changes. The specific details consitute SAFEGUARDS INFORMATION, Summary of Evaluation:
The are no failure modes associated-with the implementation of the activities described above that were considered as credible since the changes were administrative in nature,          As a result, all modes of events considered within this evaluation constitute events which have been reviewed previously and represent no impact on the safety of the plant nor a reduction in the margin of safety of the Security program, i
 
          .                .      .        _    .              .      . _ . _ . _..-.m.      - . .~ _ . - -.            _ _ _ _
I
 
==Attachment:==
to TXX 91090                                                              TV Electric
                              -page-50iof 90-                                                                        Unit: 1XN                                    :
Evaluation Number SE-90-070                                                                                  ;
Activity Title
                                .Hodification of Feedwater Hammer Low Flow Interlocks
                            - Description of Change (s):
LThe Feedwater Isolation Valves (FWlV) have control grade interlocks for opening.which prevent bubble collapse water hammer. System parameters at-severalLlocations are monitored and, with timing circuits where. appropriate, permissive open signals are generated.
                                  -At Comanche Peak, the opening-of.the'FWIV can cause a momentary flow
                                  -instability. - This instability is of sufficient magnitude and duration to be sensed as a Low Flow condition. The interlock circuitry, sensing that feedwater-flow has stopped, closce the FWlV to. prevent a L                                feedwater surge ifsand when feedwater flow were to-be restored.                    This inadvertent' shutting of the FWIV immediately after opening the FWIV is a nuisance.
This change:addst relay _: logic to the feedwater interlocks so that FWIV closure does not occur-on-low-flow alone. With this logic added,                    the FWIV will shut when low = flow is' coincident with low feedwater
                                  ' temperature which provides assurance that the FWlV will not shut on momentary flow instabilities.
                  ,          . Summary.'of Evaluation:
                                  =Th'e interlocks lare designed to prevent bubble collapse water hammer Ewhichtis notsan: initiating-event for any-licensing basis accident.
lThe; plant isedesignedlfor such an occurence without-loss of safety -
afunctions''or. damage-to the steam generator pressure boundaries, The-rsteam generators are: designed for. ten bubble collapse water-hammer            -
Levents.t The probability of occurence is -very -low and :this- change does -
not increase that/ probability significantly with respect'to the design
                                  -criteria of ten events.
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                                                                            +    -        - - -                      -
 
Attachment to TXX 91090                                                TV Electric Page 51 of 90                                                          Unit: 1XN Evaluation Number SE 90 071 1
Activity
 
==Title:==
Provision of water and sewage services to new warehouse and office building.
Description of Change (s):
Adds connections to the potable water system and sewage water system for providing water supply and sanitary facility to the new warehouse and office building.
Summary of Evaluation:
Potable Water Supply to the building is used for water fountains, lavoratories. commodes and miscellaneous uses in the building and will not significantly-impact the overall capability of the Potable Water System. Drainage connections to the building is provided for the sanitary and drainage in the building and will not have any significant effect on the sewage treatment system. The Potable Water System and it's = users are non safety related, and the system is independent of the operation of both units. The Sewage Treatment System does not interface with operation of both units. There will be no adverse impact'on. plant systems, structures or components due to installation of these new pipeline connections; nor due to any maintenance or repairs performed on these lines,
 
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:: Attac'hment to TXX 91090                                                  TV Electric      :
              ?page'52'of=90o                                                              Unit: 1XN        {
Eva' '9' i on Number                                    i Si 90 072 i
ActivityLTitle:
                            ~
                  = Addito Existing Utilities- (Fire Prote'; tion Piping) to Support -                    !
Warehouse and AAP Extension Description'of; Change (s):
This activity' consists of adding a connection to the main loop to                    4 supply fire protection water to the new Haterial Handling and
                  -Inspection Buildings.
Implementation of this activity will have the following effects:
o        Adding a connection to the plant main loop for the new buildings' water-supply, i
o        Temporarily impairing a, portion of the plant main loop'          !
piping to-facilitate the work described above.
As with all- connections to- the. plant main loop _ for supplying -water to non plant _(i'.e., support) buildings, this connection has provision for supplying water _to safety related plant structures.is not affected or.
                                                                                                        -l
                  -degraded,- Additionally, since alternate. flow paths are available, the--            ~;
portion'of the main-loop: piping temporarily impaired during execution              '
of the work-is not required for safe shutdown of the plant in the event -:of a: fire, JSummaryL of . Evaluation:.                                                                  i This change involves-the addition of~ Utilities to the-plant main loop,_              l As withialliconnectl_ons.to the plant main loop for supplying water to.
non-plant (1-.e.i-Support) buildings, this- connection has provisions for          9 Lisolating the main loop such that,the capabilitylof supplying water to
                                                  ~
j
: safety related~plantLstructures is-not degraded,' ~ Additionally.Lsince'-
ci alternate flow paths _ are available;,the portion of the main piping                j (loop temporarly>: impaired during execution of the work is not requiredT                !
                  ;to;supplyiwater to Unit'1 fire supression systems' required for safe
: shutdown of th6' plant,insthe event of: 0 fire.
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l Attachcent to TXX 91090                                                                                              TV Electric Page 53 of 90                                                                                                        Unit: 1XN Evaluation Number SE 90-073
,          Activity
 
==Title:==
Revise / Enhance Technical Requirements Hanual on feedwater Isolation Valve Brittle Fracture Prevention /Honitoring Description of Change (s):
In early 1990. TU Electric reported to the Nuclear Regulatory Commission (NRC). in accordance with 10CFR50.55(e). that the ferritic material of which the feedwater Isolation Valves (FWlV) were manufactured had not had been impact resistance tested as stated in the Final Safety Analysis Report (FSAR).                                            Upon further testing and analysis. TV Electric established a lower service temperature for the FWlV of 90 degrees Farenheit when system pressure exceeded 675 psig.
These pressure and temperature restrictions were placed in the Technical Requirements Hanual (TRM).
This change made several modifications to the TRH. The changes were required to better define the temperature and pressure restrictions and to provide operating restrictions that were consistent with the testing and engineering analysis without being overly conservative or restrictive.
Surveillances were added in Hode 1 when the FWIV are shut,                                                    in that condition. the feedwater cools down and the FWIV may not remain above the 90 degree limit.
The Applicability was changed to include pressure testing of the steam generator or main feedwater which could pressurize the FWlV when they are below 90 degrees in Modes 4. 5 or 6.
The ability to restore FWIV temperature above 90 degrees was added as a Compensatory Heasure. This is consistent with the previous Operability Criteria since restoration above 90 degrees would not require completion of Compensatory Heasures.
The wording of the Compensatory Heasure has been clarified to convey that engineering evaluation is required even if the Operability Criteria are restored to within limits.                                          This was was the original intent. but it wasn't as clear . Additionally, the engineering evaluation is now required to be completed within 72 hours.
Summary of Evaluation:
In general. changes have been made to the TRH to clarify and better convey the intent of the Technical Requirement as originally written.
Additional restrictions and requirements have been added to control areas which were not previously addressed.
The probability of a feedwater line break has not been increased by these changes. The only change added that is not more conservative is                                                                                                            t the 72 hour limit for the engineering evaluation, which was not                                                                                                                  I
 
  . Attachment to TXX-91090                                            TV Electric Page 54 of 90                                                      Unit: 1XN Evaluation Humber SE-90 073 previously addressed.
The172 hour time limit for engineering evaluation ensures that the evaluation is completed in a reasonable amount of time. The time limit is_not considered excessive because if brittle fracture had not occurred while outside the Operability Criteria, it is highly improbable that a brittle fracture would occur during the 72 hours while meeting the pressure and temperature criteria.
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Attachment to TXX 91090                                              TU Electric Page 55 of 90                                                        Unit: 1XN Evaluation Humber SE 90-075 Activity
 
==Title:==
Deletion of alarm circuits and alarm windows for control room dampers CPX-VADPOV 46,47.48 and 49.
Description of Change (s):
This change involYes adding a note to FSAR Figure 9.4-1 to explain that control room ventilation system air balance dampers CPX-VADP00-46, 47. 48 and 49 are locked by mechanical means in their full open position. This position was determined by air balance testing. As part of this change, associated alarm circuits X ALB-134, and wind?ws 3.7 and 4.7 are deleted, since dampers will not require any modulation or remote opening and closing.
Summary of Evaluation:
No new accident or increased consequences of a previously analyzed accident is introduced by this change nor does the change affect the function or performance of the control room ventilation system. The full open position is the fail safe position for these dampers, l'
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        .    . . -                  m. _        - . . - . . _ _ . - _ . __                        _ _ _ - _ _ . .              . _ . _ . _ _ _ _ . . .
Attachment to TXX+91090-                                                                            TV Electric Page 56Lof _90                                                                                      Unit: 1XN Evaluation Number-SE-90-076 Activity _
 
==Title:==
Revised matrix of CPSES management affecting OA organization.
Description of Change (s):
As a result of Comanche Peak Unit 1 becoming operational, there has been organizational changes which have affected'the Quality Assurance                                                  .
Program, but has not reduced the commitrents in the program                                                          .I description.- The organization changes reflect the changeover to an                                                    ;
                                    " operating philosophy" as well as the' downsizing of the construction
  ;                                  workforce.
The most notable-of the changes include:
                                    .1he Quality Control organization has been divided into two groups, Operations-0C and. Construction 00.                        A Manager is In charge of each group. This1 split gives each group better focus.
U
                                  -The position ;of Deputy Director. 0A- has been- deleted.                            This position served as an' assistant to and in~the absence of the Director. 0A.
i The positionLof Procurement 0A Hanager was created to place more management emphasis on procurement documentation, vendor controls and
                                  - receipt inspections.
                                  <The Vice President, Nuclear Operations. reports directly to the
                                  ~ Executive Vice President, Nuclear Engineering:and 0perations on zmatters concern i ng Un i t I rather than through-the Senior Vice President"as!before. However ;forimatters/ pertaining to Unit 2, the
                                  ;Vice1 President Nuclear Operations sti_11 reports to the Senior Vice President.
Deficiency $reportsJ(DR)-and nonconformance' reports (NCR) for Unit 4
2:-have1been' replaced. Both have been replaced by the_TV1 Evaluation
: Form:(TVE Form).- This-change parallels the. consolidation of                                                          l 3:                                -deficiency reporting forms for Unit 111nto_ Operations Notification-and.
                                  - Evaluation Forms-(ONE Forms).
Summary of Evaluation s.
                                  - These>changesihave .little tor. no impact because' they are administrative
                                              ~
Lin nature :end ado .not affect the implementation of the QA Program.
                                  -Similar1 organizational controls are in' place and the same or enhanced implementing procedures 'are in place for the new organizations so the inew organizations _are:at least as effective as the old organizations.
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Attachnent to TXX-91090                                                                                                        TU Electric Page 57 of 90                                                                                                                  Unit: 1XN Evaluation Number SE-90 077 Activity
 
==Title:==
1 Simplification of CPSES Cable Separation Criteria Description of Change (s):
FSAR Section 0.3.1.4.5 was revised to include additional cable separation criteria between redundant cable trays, cables, and conduits outside of equipment. The change states that for power, control and instrumentation circuits, one inch and one barrier provides adequate protection to ensure electrical independence,                                                            in addition, where plant conditions preclude maintaining these distances, lesser separation may be acceptable on a case-by case basis when supported by test and analysis for the specific configuration.                                                            The change applies to linit i new installations only. The original separation criteria apilies to rework activities.
Summary of Evaluation:
The change to FSAR Section 8.3.1.4.5 states that for power, control and instrumentation circuits, one inch and one barrier provides adequate protection to ensure electrical independence,                                                          in addition, where plant conditions preclude maintaining these distances, lesser separation is acceptable on a case-by-case basis if supported by test and analysis for specific configurations.
The one inch and one barrier separation criterion between conduit to cable / tray; cable to cable with one cable wrapped in Siltemp: and solid / ventilated covered tray to tray or cable in free air to tray has been demonstrated to provide adequate separation based on Wyle Test Reports #17666-02 and #47906-02.                                                        Throughout the test reports, lesser separation was demonstrated to be adequate for the same or similar CPSES configurations. However, in these cases, except as noted below, TV Electric conservatively elected to continue to use one inch and one barrier separation criteria.
                                                  ~TU Electric has limited the scope where lesser separatlon is considered acceptable.                                                  Zero inch separation can be used between redundant instrument / control conduit to condult; non Class 1E instrument / control and Class IE conduit to conduit; and conduit to cable / tray.                      The zero inch separation her, been demonstrated to provide adequate separation in Wyle Test Report #17666-02. In addition, one quarter inch separation can be used between power conduits.                                                          One quarter inch separation has been demonstrated to provide adequate separation in Wyle Test Report #47906-02.
This change is intended to simplify the existing CPSES separation criteria by specifying the minimum separation required, such that adequate circuit protection and electrical independence is maintained.
TV Electric still intends to use conservatism, wherever pratical, by maintaining maximum separation.
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Attachment to TXX 91090                                                  TV Electric Page 58 of 90                                                            Unit: IXN
                                                        ~
Evaluation Number SE 90 079 Activity
 
==Title:==
Change the delta T between the RTD's installed outside containment and RTD's mounted at a low point in the FDWTR lines from 5 to 10 deg F Description of Change (s):
Westinghouse document " Minimization of Split-Flow Preheat Steam Generator preheater Pressure Transients" (WPT 2225) describes the design requirements to minimize the potential for occurrence of pressure transients by preventing the introduction of cold water to the steam generator through the main feed nozzle at any time when significant voids may be present.      Temperature measurement is to be provided in the feedwater piping between the feedwater isolation valve and the steam generator at points where cold water may tend to resist displacement during tne purging operation. The feedwater temperature at all measuring points must be equal to or above the setpoint temperature (250 F) before the Fly is opened.        Appendix 'D',"
instrumentation range and trip accuracies," also specifies instrument trip accuracy of +5 F.
To conform to the temperature requirements, the existing design measures FW temperature at two locations, one in the FW line outside containment downstream of the main FIV and the other in the piping low point inside containment near the main FW nozzle.      The FIV opening permissives require that the FW temperature near the FW nozzle must be at least 250 F and, in addition, the temperature difference between the two elements must not exceed 5 F indicating that the line has been adequately purged of cold water. This 5 F temperature difference encroaches on the trip accuracy of the instruments.        The
                        +5 F Westinghouse reported accuracy of each instrument results in a " Root of-Sum of the Squares" 7 F when combining these two instruments. This temperature difference was increased to 10 F which should also prevent inadvertent alarm annunciation caused by the difference in ambient temperatures inside and outside the containment.
Summary of Evaluation:
This change to the interlock setpoint does not reduce the conservatism in design to minimize-pressure transients. The time delays and the temperature interlocks in the FIV opening permissives are interrelated. The time delays are governed by two timers, one initiated by the opening of the feedwater isolation bypass valve to admit the purging flow for 50 minutes and the other one initiated by signal from the temperature element inside containment after FW temperature has reached 250 F. The latter timer ensures that the portion of FW line between the temperature element and the steam generator nozzle is also purged of water below 250 F and is set at 10 minutes. The time delays and the temperature interlocks, therefore, adequately assure that the temperature requirement for FIV opening permissives is met during forward flushing operation.
 
9 Attachment to.TXX 91090:                                                                TV Electric Page 59Lof 90                                                                            Unit: 1XN Evaluation Number SE 90 080 Revision 3 Activity.
 
==Title:==
              . Changes to security system to accomodate potential area material staging building and auxiliary access point extention, Description of Change (s):
This change expanded the Alternate Protected Area Access Point, added the Protected Area Warehouse, changed the Protected Area boundary (bring the warehouse into the PA), and increased the diesel generator loading.
Summary-of. Evaluation:
The activities involved were the addition-of security fence, grading,
            - new cameras, microwave units - lighting, a multiplexer, relocation of equipment-inside the Alternate Access Point, and the process of interconnecting.this equipment.      Sin'ce no-tie-ins are being made to any other system (no safety systems) and there is no electrical ductbank or piping in the-area where the new ductbanks and ground 11 oops are being added'and the cables-are-installedLin security system ductbanks'which carry no other system cables, this activity does not impact any plant safety systems or parameters.
Security lighting and multiplexer power has access to the plant
              .cmergency-diesel generators.      Changes to the lighting system has not
"            -affected;the diesel. generator loading.. The addition of 5.3~kVA load Jon the diesel = generator due to-the additional muxiplexer load of                                            ;
0'.44 kVA and other lead additions have been evaluated and found to be of no' impaction _ the diesel generator capability-to provide the
                  ~
required. motor' starting capabilityfand to meet the licensing
              --c ommi tments .
            -When-the work was performed on the security system, those security functions taken out of_ service were compensated by se^urity officers.
              .N0TE: -The warehouse and related Protected Area changes-were fully                                            .
impiemented.in 1991.
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Attachment to TXX 91090                                                TV Electric page 60 of 90                                                          Unit: IXN Evaluation Number SE 90 081 Activity
 
==Title:==
Modification of Steam Generator Atmospheric Relief Valve Capacity Description of Change (s):
The Steam Generator Atmosheric Relief Valves (ARVs) provide a means for removing heat from the HSSS to the atmosphere during periods when the condenser is not in service. They are sized to remove the reactor decay heat generated following a reactor shutdown. In addition, the ARVs are used to avoid, whenever possible, the opening of the lowest set main steam safety valve.
In July 1990, the ARVs were stroke tested and it was discovered that
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the ARVs had drifted out of calibration. Three of the four ARVs did not appear to provide the minimum required steam flow capacity to support the analysis for cooling down after a Steam Generator Tube Rupture (SGTR). As a result, the stroke lengths for the ARVs have been changed so that drift will not place the ARVs capacity outside of the design basis,  in doing so, reanalysis was required to show that the increase in stroke length did not create an unreviewed safety question.
Summary of Evaluation:
The increased capacity of the ARVs affects the inadvertent Opening of a Steam Generator Relief or Safety Valve Accident analysis and the SGTR Accident analysis. No other accident analyses credit the operation of the ARVs.
For_the Inadvertent Opening of a Steam Generator Relief or Safety Valve Accident analysis, the assumed flow rate through the open valve is 968.400 lbm/hr referenced to 1200 psia. The revised maximum nominal ARV capacity is approximately 897,000 at 1200 psia. Since the revised capacity is enveloped by the absumed flow in the analysis, the analysis is not affected by the increased ARV capacity.
For the SGTR Accident, the calculated offsite doses must be within the guidelines of 10CFR100. For the reanalysis, the relief capacity of the ARV on the ruptured steam generator was assumed to be 968,400 lbm/hr referenced to 1200 psia. (The previous analysis assumed a nominal ARV capacity of 779,000 lbm/hr at 1200 psia.) The reanalysis showed that the dose rates were within the values stated in the FSAR.
 
1 Attachoent to TXX-91090                                              TV Electric Page 61 of 90                                                        Unit: 1XN Evaluation Number SE 90 084 Activity
 
==Title:==
Additon of-ventilation systems to the chemistry hot laboratory.
Descriptica of Change (s):
Adds make up and Exhaust Air System to the Chemistry Hot Laboratory to accomodate the operation of a new atomic absorption equipment. The atomic absorption unit is used to chemically analyze selected elements in the primary water.
Summary of Evaluation:
Addition of a ventilation (make-up & Exhaust) System is required when atomic absorption unit is in operation. The exhaust air is vented to the Primary Plant Exhaust System by means of the new ventilation systems, which ensures that no. unfiltered gas is released from the plant to the_ atmosphere.
Their is no safety systems or systems important to safety considered that are potentia'lly affected by implementation of the activity.
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AttachCent to TXX 91090                                                                                                TU Electric Page 62 of 90                                                                                                          Unit: 1XN Evaluation Number SE 90-089 Activity
 
==Title:==
Revise FSAR Figure to Reflect As-installed Auxiliary Steam System Description of Change (s):
1he Auxiliary Steam System is designed to supply low pressure steam to and collect condensate from miscellanceous warmup and process services in the Turbine and Auxiliary buildings.                                          The system is non nuclear safety-related and is not seismically designed.
This change consisted of addition of blowdown valves for the Auxiliary Steam Header Drip Pot traps, addition of vent and drain valves and a 1 relief valve on the Economizer section of the Auxiliary Boiler, clarification of instrument line configuration on the Auxiliary Boiler Steam Drum, and addition of valves and changes to piping configuration on the Auxiliary Boiler burner centrol.
None of these changes involved physical modifications. This evaluation documents the char.ges to the drawing to match the as-built design.
Summary of Evaluation:
                                  ' Al l of the changes made are minor in nature.                                      The operation of the Auxii.ary Steam System is not affected by any of these changes.                                          No licensing bases accidents are affected by these changes.
 
AttachQent to TXX 91090                                              TV Electric Page 63 of 90                                                        Unit: 1XN Evaluation Number SE 90 090 Activity
 
==Title:==
Addition of double isolation valves to the Surface Water Treatment System Line, Description of Change (s):
Adds double isolation valves to the surface water treatment line 4WT X 933 which connects degasified water clearwell to the reverse osmosis system cleaning tank. Also adds two drain valves to drain the line Summary of Evaluation:
Addition of double isolation valves to the surface water system line 4WT X 933 provides the separation required for protection of possible contamination source connecting to the potable water systems. This change also meets the separation criteria requirements of Texas Department of Health.
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Attachaent to TXX-91090                                              TU Electric Page 64 of 90                                                        Unit: 1XN I
Evaluation Number                                  i SE 90 091 Activity
 
==Title:==
Change Normal Position of Heater Drain Discharge Sample Valve Description of Change (s):
    -The Heater Drains System is designed to regeneratively heat feedwater by cascading the higher energy drains through successively lower energy stages of feedwater heaters. Since the system contains steam generator feedwater, the water quality of the Heater Drains System needs to be contiuously monitored via the Secondary Sampling System.
This change opens an isolation valve between the Heater Drains System and the Secondary Plant Sampling System. As discussed above, the ability to continuously monitor the water quality of steam generator feedwater is required to er.sure chemistry control of vital components.
Summary of Evaluation:
The Heater Drains System is a non-safety system and is not required for safe shutdown of the plant or control of radioactive releases.
This change does not affect any licensing basis accident, i
 
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==Title:==
Organizational and administrative changes within the startup department.
Description of Change (s):
This activity involves organizational and administrative changes within the Startup Department as a result of a reorganization at
  -CPSES,    it clarifies responsibilities of the Startup Department as distinct from the Operations Department for Unit 1 operations. This activity also-includes changes for Unit 2 organization for Unit 2 projects and of Unit 2 startup thus illustrating the overall interface between operation and construction.
Summary of Evaluation:
The organizational changes reflected within this activity do not affect _ plant-systems, components, nor parameters. The changes do not affect plant safety and, therefore, are considered acceptable.
 
    . . - -.i
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                                                                - Evaluation Number                                                            .
SE-90 094 Activity
 
==Title:==
Addition Of New. Steam Generator Blowdown <1cinup (SGB) System HELB Locations To FSAR Section'3.00 Figures Description of Change (s):.
The subject; safety-evaluation involves a revison_to FSAR figure                                              <
3,60 43 to reflect three new High Energy Line Breaks (HELB's) postulated.at welded attachments in the SGB system within the Turbine.                                        !
Building.
Summary of Evaluation:-
The revision to FSAR-figure 3,6B 43 to add three new postulated
                                -HELD locations in the SGB-system, results in only one out of the three
                              - ne -postulated HELB' locations presenting- the potential- for effecting an essential system' credited with mitigating _the consequences of a
: HELB in_that system. The effected essential system is only required to'mitigh'q.the environmental effects of a break in'this system. .The effected essential: system components-are not' required for these SGB H E LB ' s ~. This new'HELB is locatedsin the same area as existing postulated,HELB's-that have been previously~enalyzed for HELB effects, q-          so no. unacceptable' consequences are presented by this-new-HELB, The
                                ' remaining two new postulated HELB's are. located outside of areas
                              ; containing essential' components required for-safe' shutdown or performancolof a safety function.
                    ^                                '                        '
    . s.                  .
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                          *.*J                                        -i,w+v -
                                                                                  -+ - .. -,een t: y ye+-'p  +y                    -        Y
 
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Activity-litle:
Revision of process sample system flow diagrams to agree with system operation (figure 9.3 4. Sheets 1, 2. 3, & 3A).
,                        Description of Change (s):
This change revises the process sompling flow diagrams to depict the normal system operation of equipment as installed. There are no changes to equipment in the plant.
Summary _of Evaluation:
This change did hot create the possibility of a new unanalyzed event nor is the probability of en accident increased. This change only affects the system's operation but not the system's function. For additional information. Safety Evaluation's 90 060 and 90-100 discuss changes which are related to this changt.- The change has no effect the safety function of the system.
 
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==Title:==
Revision to the process control program to permit solidification of liquid and wet radioactive wastes.
Description of Change (s):
Revises the Process Control Program to permit solidifcation of liquid and wet radioective wastes. This revision allows the processing of evaporator wastes and other non dewaterable weste should these wastes be generated at CPSES.
Summary of Evaluation:
The system modification is in compliance with Regulatory Guide 1.143.
1.21 and Standard Review Plan 11.5. The vendor topical reports regarding the equipment and the process Control Program has been approved by the NRC.        The equipment added to handle the radioactive waste is non sa'ety related and is located in the fuel Handling Building where there are no other systems important to safety.
Neither the Activity nor the failure modes associated with this activity could have any potential iroact en the probability of failure of a safety system.
 
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Page 69 of 90                                                                                                                                              Unit: IXN                                          !
Evaluation Number SE 90 099 Activity Title
                                                                                                                                                                                                                                )
Emergency low pressure heater bypass control modification                                                                                                                                                  !
Description of Change (s):
The Emergency Low Pressure Heater Bypass valve actuation logic. (by                                                                                                                                          l the Steam Condenser High Differential Pressure Switch and feed Pump low suction pressure switches.) is (hanged in this Delance of Plant non safety system to minimize unnecessary reactor trips from a one-out-of one to a two out of three logic.                                                                                                                                                                      -
l This activity attempts to reduce unnecessary reactor trips by eliminating a possiblity of an uni,tentional trip during the time the plant is transitioning to high powei Summary of Evaluations                                                                                                                                                                                          i The modification was made in a non safety related system (Condensate) and does not impact the safety analysis nor does it have any radiological consequence.
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o                                                                                                                                                                                                                              ;
, -                                                                                                                                                                                                                            1 5
 
_    __    __ _ _ _ _ . . - ~ . _ _ _ _ > -
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                                                                                  $E 90 100 l
Activity
 
==Title:==
Isolation of the sample cylinders in the grab sample hood f or the primary sample system during operation.
I Description of Change (s):
The activity changes the normally open isolation valves for the sample cylinders in the Grab sample Hood to normally closed valves. The tag numbers for the sample cylinders have been deleted from the flow diagrams for the primary $ ample $ystem.. There is no physical change to the plant, and involves only the valve lineup and the flow diagram for the system.
Summary of Evaluation:
The pressurized sample cylinders are removable, stored by Chemistry in a remote (non installed) location, and connect in the system via quick disconnects. Cnanging the designation of these valves from a normally open to a normally closed and removal of the sample cylinder tag designators in the flow diagram improves the flexibility of system operation by allowing cylinders to be used in more than the location designated by a tag, reducing pressure on the disconnects and reducing the number of valve manipulations needed for grab samples.
The activity affects the valve lineup f or taking a chemistry sample and does not physically modify the plant. The changes to this non-safety system has no effect on any safety systems or systems considered important to safety.
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Activity
 
==Title:==
PCN ALM 0081 R2 12 Alarm Procedure 1 ALB 8A                                                            >
S/G 1 (2,3,4) FW NZL FLO H1 Description of Change (s):
The resolution of the 04/05 steam generator tube vibration concern, included reducing the flow through the preheater section of the steam generator to approximately 90% flow at 100% reactor power (the retnaining 10% to bypass the preheater via the Aux feed nozzle). CPSES Alarm procedure 1 AL8*8A set the high flow alarm for feedwater flow (through preheater) at 3.4x106 lbm/hr.
The purpose of this safety evaluation is to allow an interim increase of the high flow alarm setpoint to 3.64x106 lbm/hr (approximately 96% flow). This higher set point includes the instrument accuracy of +2.6%. The time for this interim increase is approximately three weeks to allow Westinghouse to complete an evaluation of a permanent revision to the flow alarm setpoint. This permanent revision to be evaluated under a separate safety evaluation.
    . Summary of Evaluation Westinghouse was requested to evaluate the setpoint change.
Westinghouse agreed that no S/G tubes would experience wall thinning to an extent that would require tube plugging during the first refueling cycle and that operation at 100% power could continue for the.approximately three weeks at the higher flow alarm setpoint.
Based on the.relatively short duration no increase in the probability for a S/G tube rupture accident is expected, i
1
 
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==Title:==
Provision for constant nitrogen supply to pressurizer PORV's and SI Accumulator Tanks.
Description of Change (s):
The position of Yalves, 1$1 0132, 151 0154, 1 8893 and 1-8880, in the flowpath between the plant nitrogen supply and the Safety Injection (SI) accumulator tanks and the pressurizer PORV's is revised from closed to open during normal plant operations.
This change provides a constant source of nitrogen to the Si accumulators and PORV's with minimal operator interf ace.
Summary of Evaluation:
As described in FSAR section 6.3.2,2.1, the nitrogen gas supply to the accumulators is normally isolated but can be adjusted as required during normal plant operation. Gas relief valves on the accumulators protect them from pressures in excess of design pressure and accumulator gas pressure is monitored by indicators and alarms.
The supply header for the accumulators also supplies nitrogen gas to the pressurizer PORV's as shown in FSAR Figure 6.3-1 sheet 2.
The Si accumulators are normally isolated from the nitrogen supply by valves 1 8875A-D and the 51 secumulators and PORV's will still be protected from potential nitrogen overpressurization by relief valves.
Leaving containment isolation valve 18880 open during norm"al plant operation provides an acceptable arrangement as specified by 10 CFR 50, Appendix A, GDC 56 and does not increase the likelihood for failure of this component or invalidate the existing design criteria.
This change has not af fected existing plant testing programs (i.e.,
ASHE XI IST program) therefore acceptable methods for detecting leakage within the SI system are available.
 
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Page 73 of 90                                                                  Unit: 1XN Evaluation Number SE 90 100 Activity
 
==Title:==
Closure of procedure ISU 210A, Rev. 3, Process and Ef fluent Monitoring Performance through the deferral of testing for six radiation monitors Description of Change (s):
Defer testing of process radiation monitors 1 RE-5179 and X RE 3230 until modifications are complete to increase flow rates past each monitors. Defer testing monitors X RE 4180, X RE 4181, X RE 4863 and X RE 4064 until the fuel pool can be filled.
Summary of Evaluation:
Radiation monitors 1 RE 5179 and X RE 3230 monitor the Steam Generator Blowdown Sampling System and Auxiliary Steam Condensate System.
* respectively. Radiation monitors X RE 4180, X RE 4181, X RE 4863 and X RE 4864 monitor the Spent fuel Pool.              These radiation monitors are not Technical Specification nor ODCH monitors and they are not used for effluent monitoring. The deferral of testing does not impact the probability of failure of the system since the testing of these monitors during initial startup would not provide any additional data other than what was achieved during their calibration as described in FSAR 11,5.2.11. During calibration, the monitors were exposed to sources of known activity which provides sufficient data to ensure proper monitor operation. The implementation of this activity does not impact the accide,  analysis described in the License Basis Documents and has no adverse effect on any safety related systems.
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==Title:==
Setpoint Change No. 90 025, Feedwater Flow to Main Feedwater Nozzle Description of Change (s):
The Westinghouse Model 04 steam generators at Comanche Peak Unit I were previously modified to minimize the potential for wear of the tubes at the preheater baffle plates due to flow induced vibration.
The modification involved tube expansions of selected tubes at the B and D preheater baffle plates and the diversion of a specific percentage of the feedwater flow though the auxiliary nozzle which is installed in the upper shell of the steam generator. The design flow split for Comanche Peak Unit 1 is approximately 90/10.
When feedwater flow exceeds the alarm setpoint of 89.5%, an alarm alerts the operators to the condition,      Recently, the alarm has been annuciating frequently. This safety evaluation has been performed to raise the setpoint of this alarm until action can be taken to correct the cause of the alarm.
The alarm setpoint has been changed from 89.5% to 91.4% of full flow through the main feedwater nozzle. When instrumentation uncertainty is considered (2.6%). the maximum flow into the preheater from the main feedwater line may be as high as 94% of full flow when the alarm annuciates.
Summary of Evaluation:
The evaluation considered 1) the increased potential for tube wear and
: 2) the effect of additional stress and fatigue on the expanded tube regian as a result of a potential increase in flow of 2% through the main nozzle.
The tube wear evaluation which was based on a comparative study of a similar plant, showed that only a small number of additional steam generator tubes may experience a reduction in tube wall thickness due to wear at the preheater baf fle plates equivalent to an allowable wall loss of 65%. This fact coupled with the eddy current inservice inspection sampling plan should preclude the occurrence of a single tube rupture event <
_The analysis of stress and fatigue of the expanded tube region showed that the analysis for the 90/10 flow split bounds the 94/6 flow split, in order to minimize the potential for tube wear beyond allowable limits during-subsequent operation due to a potential 2% increase in the main nozzle flow, the following remedial actions are planned:
: 1.        At the first refueling outage, complete an eddy current inspection of all tubes in the first six rows of the preheater, the tubes adjacent to the T-slot, and tubes on the l
t
 
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                                                          $E 90 201 periphery of the bundle between rows 30 and 49 in each steam generator.
: 2.      At subsequent in$pections, a sampling plan will be selected to assure that the tubes identified above are inspected in intervals of 5 to 6 years of operation, e
il k'
e
            .i.......
 
i Attachoont to TXX 91090                                                    TV Electric Page 76 of 90                                                              Unit: 1XN Evaluation Number SE 90-202 Activity
 
==Title:==
Changes to the Fire Protection Report to Reflect Existing Design Description of Ctange(s):
This activity consists of revising the FPR to incorporate editorial changes, revise Actual Fire Load (AFL) values, and Fire Duration values. Additional changes are incurred due to the revision of the ecleulations which support the material in the FPR. This activity has no impact on the fire protection program of Unit 1 of CPSES.          As part of these revisions, implementation of this Activity will have the following specific effects:
: a.      AFL values will be rounded off to the nearest 100 Btu per square foot for conservatism,
: b.      The AFL values for many of the fire zones have increased, but in no new case has the AFL value exceeded the fire zones Maximum permissible Fire Load (HPFL) value,
: c.      A11 HPFL values have remained unchanged with the exception of the HPFL for Fire Zone $D2x which has increase from 37,300 to 80,000 Dtu per square foot. This increased HPFL resulted frum the decreased AFL for the fire zone (from 10,000 to 4,300 Btu per square foot).        This was due to selection of more representative curve for generation of HPFL.
The changes are editorial, and some changes in AFL values as well as                '
Fire duration values. But these changes remain will within HPFL values. Hence, these changes do not affect plant safety.
Summary of Evaluation:
This activity is primarily technical changes and some editorial changes to-the fire protection program as described in the FPR.          This change serves to document the AFland Fire Duration valuesand other fire protection system descriptions.
These changes have no impact on the administrative and procedural controls which are already in place which assure that the HPFLs in the plant are not exceeded without appropriate compensatory measures.
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==Title:==
Temporary replacement of the cartridge for Reactor Coolant Pump (RCP) seal injection filter with a finer mesh and using a different filter.
Description of Change (s):
One of the cartridge for the RCP seal injection filter housing (TBX CSFLS! 02, Su 98% removal, 23u 100% removal) has been replaced with a finer mesh using a dif ferent filter medium (.6u 99% removal, lu 100% removal).
This temporary modification is intended to improve filtration of the RCP seal injection water for the purpose of a reduction in seal degradation and reduce the activity in the CVCS overall.
As discussed in FSAR section 9.3.4.1.2.5, two seal water injection filters are located in parallel in a common line to the RCP seals:
they collect particulate matter that could be harmful to the seal faces. Each filter is sized to accept flow in excess of the normal seal water flow requirements < A differential pressure indicator monitors the pressure drop across each seal water injection filter and gives local indication with high differential pressure alarm on the main control board.
Summary of Evaluation:
The replacement cartridge has the same dimensions and nominal weight, has the same maximum dP, conforms to the design flow rate, has a clean filter dP estimated-to be 0.3 psid higher at 32 gpm and 0.9 psid higher at 80 gpm than that of the old cartridge.
RCP seal damage resulting from loss of flow, caused by initial rapid seal injection filter clogging, is prevented by the following factors:
When the lu cartridge reaches high dP limits, seal injection flow will be shifted to the other filter, which will not be subject to rapid elogging due to its coarser mesh, thus, seal injecti 1 flow will be maintained while the cartridge is replaced.
The current replacement criteria (dP limit of i0 psid) will continue to be utilized with augmented dP monitoring.
If seal injection flow is terminated extended RCP operation is still possible when component cooling water flow is maintained to the thermal barrier heat exchanger, The difference of slightly higher clean filter dP is conservative relative to the seal injection header resistance and thus does not-negatively impact analyzed ECCS performance.
 
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==Title:==
Procedural revision for valve lineups and position changes for the diesel generator system.
Description of Change (s):
Station operating procedure 50P 609A R7 6 has been revised to add valves 100 0476, 0477, 0478 and 0479 to the valve lineup. Also, the position of these valves, as shown in FSAR Figures 9.5 55 and 9.5 56 will be changed from "normally open" to "normally closed" to prevent an inadvertent slow depressurization of the starting air receivers for the emergency diesel generators. These instrument valves are used for testing and calibration: dnd are located of f the instrument line for pressure indicator on the starting air receiver. Changes in the valve line up for the lube oil system will result in reducing lube oil fumes accumulating in the diesel generator rooms.
Summary of Evaluation:
These valves need to be 'normally closed
* during the operation of the system to prevent bleeding of the air receivers which depressurizes the startup air system          This will also prevent lube oil fumes from accumulating in the Diesel Generator Rooms.
Therefore this change clarifies the procedure and drawings with respect to the proper operation of the system to prevent the bleeding down of the air receivers.
I~                                                                                                                                      ,
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Page 79 of 90                                                                                        Unit: 1XN Evaluation Number SE 90 206 Activity
 
==Title:==
Setpoint Change No. 90 041, Feedwater Flow to Main reedwater Nozzle Description of Change (s):
This setpoint change is an extension of setpoint change 90 025. which is evaluated in Safety Evaluation 5E-90 201                                  However, for this setpuint change, the focus was the potenti.                                  for thru wall tube wear during the first cycle of operation,                                  for background information, refer to SE 90 201.
Setpoint change 90 025 raised the alarm setpoint for high feedwater flow from 89.5% to 91.4%. This setpoint change (90 041) raises the alarm setpoint from 91.41 to 93.9%. When instrumentation uncertainty is considered, the maximum flow into the preheater from the main feedwater line may be as high as 96.5% of full flow when the alarm annuciates.
Summary of Evaluation:
As is the previous evaluation, the increased potential for tube wear and the effect of additional stress and fatigue on the expanded tube region due to increased flow were considered.                                  For this flow increase, the. tube wear rate could increase as much as 47%. This increased wear rate was not expected to result in tube wear in excess of the 65%
allowable limit during one cycle of operation. The additional stress end fatigue for the 90.5/3.t flow split were judged to be bounded by the analysis for the 90/10 flow split.
However, the remedial actions taken to minimize the potential for tube wear beyond allowable limits during subsequent operation have been                                            ,
changed to the following*
: 1.      Conduct an eddy current inspection at an outage following an initial 6 to IE effective full power months of plant operation                                        ,
where the sampling plan includes all tubes in the first six rows of the preheater, the tubes adjacent to the T slot, and tubes on the periphery of the bundle between rows 30 and 49 in each steam generator.
: 2.      At subsequent refueling outages, the tubes identified above will be inspected at time intervals which are established based on average wear rates obtained from measured tube wear such that the 65% allowable wall loss limit will not be projected to be exceeded during the designated period of plant                                          '
operation.                                The secnnd eddy current inspection will be conducted no later than 4 effective full power years following initial plant operation, assuming that an approximate 90/10 flow split is reestablished in future cycles.
 
Attachment to TXX 91090                                                                                      TV Electric Page 80 of 90                                                                                                Unit 1XN Evaluation Number SL 90 208 Activity Title                                                                                                                    1 l
Temporary modification to defeat alarm contacts from Non class 1E                                                              l switchboard 102 ground detection relays,                                                                                      l 1
Description of Change (s):                                                                                                        l Jumpers were installed across the ground detection relay alarm contacts. The jumpers defeated the ground current remote annunciation for twitchboard 102. Control room annunciation for overvoltage.
undervoltage, and ground current is common for non 1E de swbd ID1 and 102. Remote control room annunciation for the remaining functions is not affected. The temporary modification allows remote monitoring
    -of the remaining parameters while isolation and correction of the ground problem on switchboard 1D2 is performed.
Summary of Evaluation:
Installation of jumpers across the ground fault relay alarm contacts disables the ground fault remote annunciation.                                          Remote control room annunciation for overvoltage, undervoltage, and ground current is common for non 1E de swbd 101 and 102.                                        Defeat of the ground current remote annunciation function, in effect, enables remote annunciation and monitoring of overvoltage and undervoltage for the switchboard 102, and overvoltage.-undervoltage, and ground deterction of swbd 101.
The jumpers across the ground fault relay alarm contacts do not introduce any failure mechanisms for this non safety related switchboard.              The modification has no impact on plant operations.
-      , - - .  -_- . , - ._- .            ~  . - - - - - - _ _ _ - - _ _ _ . . - _ , _ _                        . , . . -. - ,-
 
Attach 2ent to TXX 91090                                                TV Electric ,
Page 81 of 90                                                          Unit: 1XN Evaluation Number
                                      $E 90 209 Activity
 
==Title:==
Deletion of Procedure NE0 9.16. Revision 1. " Engineering Functional Evaluation Transition and Completion Program."
Description of Change (s):
The Engineering Functional Evaluation (EFE) started in May 1987 as part of the Corrective Action Program. The EFE employed a vertical slice approach in the evaluation of the Containment Spray System, Control Room HVAC System and Safeguards Building.      This approach resulted in indepth evaluations of significant portions of these subjects by a team of highly qualified and experienced engineers,      in addition to the overview of design validation activities, the EFE evaluated the utilization of design data in the testing, maintenance and operations programs, and evaluated the controls that will assure that-the design is maintained throughout the life of the plant.
The EFE has concluded the activities listed above and the EFE engineers have been assigned other tasks. Because this program has been completed, the procedure which controlled this activity has been deleted.
Summary of Evaluation:
The deletion of this procedure does affect any accidents in the Licensing Basis Documents because the program was completed and findings or weaknesses identified were corrected. With the program completed, deletion of the procedure has no affect on continued plant operation.
 
Attachaont to TXX 91090                                                            TV Electric Page 82 of 90                                                                      Unit: 1XN Evaluation Number SE 90 211 Activity
 
==Title:==
Procedural revisions to addrest the requirements of the hydraulic valves in the Personnel Airlock (PAL) as containment isolation valves.
Description of Change (s):
Plant procedures (50P 907A k3, ODA 403 02 R1, OPT 218A 1 R3 and OPT-219A 1 R3) have been revised to assure that PAL hydraulic system containment isolation valves (1BS 0016, 0017, 0037, 0038, 0039, and 0040) are tested along with the personnel hatch and to maintain valves 185 0016 and 0017 locked closed when the hatch is not in use.
This change was an interim change in response to the immediate corrective actions as described in Licensee Event Report 90 032 00 submitted in TU Electric letter logged TXX 90356, dated October 29, 1990.
Summary of Evaluation:
4                                    Designating these valves as containment isolation valves and revising the applicable administrative procedures will bring these valves into compliance with the containment local leak rate testing program, thereby providing an early indication of seal degradation.
Controlling the valve position during modes 1.2.3.4 will bring the plant into compliance with Technical Specification 3/4.6.3.
Failure of these valves is not considered credible since the valves are rated in excess of the LOCA pressure and they do not have to move to perform their safety function, which is to prevent excess leakage from containment.
See SE 90 222 for the permanent change.
 
Attachasnt to TXX 91090                                                                      TV Electric Page B3 of 90                                                                                Unit 1XH Evaluation Number SE 90 221 Activity
 
==Title:==
Addition of manual loading regulator and air supply from the instrument Air System.
Description of Change (s):
Adds a loading regulator and a root valve to permit manual control of 1 LV 2592 during startup. This additional air supply path provides the required pressure specified by the valve manufacturer for its operation.
Summary of Evaluation:
The addition of instrument air root valve 1 LV 2592 AS2 is required for addition of loading regulator 1 LK-2592. A steady state demand of 21 scfh and supply demand of 400 scfh for the I/P being supplied to                                        .
the instrument air root valve do not impose excessive loads on the                                        '
Instrument-Air System. The Instrument Air System is highly reliable and impact on the system is minimal since the loading regulator does not add any consumption or impose excessive load.
D 1
l
 
Attachaent to TXX 91090                                                                                                          TO Electric page 84 of 90                                                                                                                    Unit 1XN Evaluation Number SE 90 222 Activity
 
==Title:==
Upgrade documentation for airlocks and modify the Personnel Airlock Hydraulic System Description of Change (s):
The upgrade of documentation in the Licensing and Design Basis Documents for CPSES were made to properly reflect the containment isolation functions and required administrative controls for the personnel airlock and the emergency airlock, Also included in the documentation was a personnel airlock design modification to add an additional valve in series with each three way hydraulic diverter valve inside containment (four total). The addition of these valves enhances the leak tightness of the hydraulic system as a containment isolation boundary.
The change allowed the personnel airlock to be returned to its normal automatic mode of operation.
These changes are in response to the corrective actions described in Licensee Event Report 90 032 and 90 033 submitted via TXX-90356, dated October 29, 1990 and TXX 90357, dated October 26, 1990 respectively.
Summary of Evaluation:
The documentation changes added the airlocks' appurtenances to the licensing and design documents in accordance with GDCs 56 and 57. The operation of the system.. including administrative controls, are now in accordance with GDC-50 GDC 52, GDC 53, and 10CFR50, Appendix J.
The design modification included test requirements to ensure 10CFR50, Appendix J Section IV.A requirements for Containment Hodification were satisfied.                  The modification included design validation of the personnel airlock appurtenances for containment isolation.
See TXX 90356 dated October 29, 1990 for related information,
 
Attachaent to TXX 91090                                                                                                  10 Eltetric Page 85 of 90                                                                                                            Unitt IXH Evaluation Number SE 90 223 Activity
 
==Title:==
Adjust level instrumentation and setpoints for feedwater Heaters 6A&B.
Description of Change (s):                                                                                                              ,
This change invoves the following activities:
a) Raises the setpoint for actuation of the alternate drain path valve to prevent excessive cycling.
b) Changes the mechanism for controlling level from a level controller to a level switch, c) Deletes Hi level alarm switches LS 2542 and 2545 since insufficient time exists following the alarm for operator 3ction.
The alarm function is performed by the drain valve actuation alarm, d) Isolates the instrument air previously supplied to the level controller.
Summary of Evaluation:
The systems involved are non safety related and the modifications do not affect any safety related systems or functions. No additional failure modes potentially associated with this activity are considered credible. The Hi Hi level trip setpoint and associated actuations are unchanged and will preclude turbine water induction (which could cause-a plant trip).
 
Attachoont to VXX 91090                                                    TV Electric Page 86 of 90                                                              Unit: 1XN Evaluation Number 5t 90 226                                                                                                        '
Activity Title
,.                C$tablishment of the Nuclear Overview Department Description of Change (s):                                                                                                                                !
This activity involves organizational changes due to the creation of the Nuclear Overview Department to improve effectiveness and better utilitettorv of resources within the Ouelity Assurance Organization (OA). Plont Evaluations and the Independent safety Engineering Group                                                                                  '
(l$tG).- These changes are reflected in Chapters 13 and 17 of the i                  FSAR.                                                                                                                                                  =
Summary of Evaluation:
These organizational changes do not effect any plant system or parameters. There is no impact on plant operations or plant safety                                                                                    '
and does not result any reduction to existing quality assurance commitments.
                                                                                      ,  . , . . ~ . - - - . - - . - - - . . - . , - - . . - - . - - - - - . -
 
Attach ent to TXX 91090                                                                                                                                                                    TV Electric Page 87 of 90                                                                                                                                                                              Unit: IXN Evaluation Number SE 90 228 Activity
 
==Title:==
Temporary modification to isolate steam leak on Main Steam / Secondary Sampling 3/8" tubing.
Description of Change (s):
This modification involves the installation of an isolation valve in the secondary sampling system to isolete a steam leak due to a rupture on a 3/8" tubing coming from the Main Steam System.
Summary of Evaluation:
The systems involved are not safety related.                                  This change does not result in any new accidents or increased consequences of previously
              -analyzed accidents, e
Juidus mi                                    -
                                                          -_____.._.-._-_____..---.___.__m_.        - . _ _ . - _ _ _ _ . _ . _ _ _ _ . _ _ . - _ _ _ _ . . _ _ _ - . _ _ . _ _ _ _ - _ _ _ _ _ _ _ _            -._____m._.____.-  _ _ _ _ . _ -
 
Attach ent to TXX 91090                                              TV Electric Page 88 of 90                                                        Unit 1XN Evaluation Number SE 90 229 Activity
 
==Title:==
Re alignment of normal source to Reactor Protection System (RPS) inverters using the Class IE 125 Voc system.
Description of Change (s):
The activity was to open the class IE 480V ac power supply to the RPS inverters thus allowing the class IE 125 Vdc buss to be the normal supply. The change provides a regulated source of power for the inverters and isolates them from any voltage fluctuations and
  -transients on the 480 Vac system and inverter AC input circuitry.
Summary of Evaluation:
Justification Plant configuration is such that there are two independent and redundant class IE 480V sources which provide two Reactor Protective System inverters each. Similarly there are two DC Busses which serve two inverter each. Each DC buss is powered by a battery that has a designated battery charger plus a spare. The chargers are powered by their respective train class IE 480V HCC. The loss of one of the DC busses like the loss of one of the 480V sources will affect                ,
only two inverters and does not compromise the redundancy or                  '
independence of the inverters' power sources.
This change is within the rating of the battery chargers.
The effects of this change on Fire Safe Shutdown analysis and system interaction were considered and determined to be acceptable.
 
Attach 3ent to TXX 91090                                                TU Electric Page 89 of 90                                                          Unit: 1XN Evaluation Number SE 90 230 Activity
 
==Title:==
Utilization of Equipment History and Trending programs to identify Potential Equipment Ouellfication Degradation and failures Description of Change (s):
FSAR Appendix 3A, Section 2.4 was revised to better describe the CPSES Equipment History and Trending Programs as they relate to Equipment Qualification.
Summary of Evaluation:
The revised FSAR takes credit for existing CPSES Equipment History and Trending Programs to review potential equipment degradation of E0 equipment, rather than periodic surveillance.      The revised FSAR reflects that if a common mode failure is identified, the appropriate            '
corrective actions will be taken to prevent recurrence, and not just a revision to the E0 report (s)_ and E0 package (s).
j l
 
Attachaent to TXX 91090                                                                                    TV Electric Page 90 of 90                                                                                              Unit: 1XN Evaluation Number SE 90 E36                                                                                i Activity
 
==Title:==
4 Update of check valve position resulting from the provisions for con-stant nitrogen supply to pressurizer PORV's and St Accumulator Tanks, Description of Change (s):
The position of containment isolation check valve 1$1-8968 is revised to open since it may be open as a result of having valve 1 8880 normally open for the new nitrogen supply flowpath configuration to the pressurizer PORV's and St Accumulator Tanks described in Evaluation SE 90 107.
Summary of Evaluation:
              -Updating the status of containment isolation check valve 151 8960 during normal plant operation is an acceptable arrangement as specified by 10 CFR 50, Appendix A. GDC 56 and does not increase the
              -likelihood for failure of this component or invalidate the existing design criteria.
I i
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