ML20107H267
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MADISON AVENUE AT PUNCH BOWL ROAD e MORRISTOWN, N.J.07960
- 201-539-6111
.... w w g,.,'.),,'l, pubne utii '.e corporation oeneret
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May 23, 1975 9
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NSM7ggv A f1'.
Mr. A. Giembusso, Director h
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Division of Reactor Licensing 7
U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Q., s kJ/yl,*>
Dear Mr. Giambusso:
SUBJECT:
OYSTER CREEK NUCLEAR GENERATING STATION DOCKET No. 50-219 CYCLE 5 RELOAD - ADDITIONAL INFORMATION In response to verbal concerns expressed by nembers of your Staff during the course of the review of the subject reload, we are submitting tha vaennneaa en nnaaein,e ennravnino cha tinane home nana r., r i nn vara which is experienced during the rod withdraual error transient and the fuel-misloading error and its relationship to fuel cladding strain limits dis-cussed its our reload information submittal. The linear heat generation rate correspending to center 11n?. melt as a function of burnup is also included.
In addition, responses to questions concerning the derivation of the overpower ratio with respect to the fuc1 cladding integrity safety limit, details of-single channel MCHFR/MCPR calculations for various power levels and a modification of our response to question 65 regarding additional pipe break locations are included as well.
Very truly yours,
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Ivan R. Fi frgdk, Jr.
Vice President i
Enclosures 4
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42 i 960213 DEKOK95-258 PDR 5713 Il
Arsplasticstrain10:sc:ceeded'intheroduithdru )V error and fuel
. ndoloading error?
'&.at is the LHGR associated uith tncse occurrences at their vorst?
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Answer Figu:e D-5 (page D-7) of Supplement No.1 to Amendment No. 76 shows the peak LHGR and relative. core power versus transient control rod positions. A rod bl5ck would occur at. notch 14 (3.5 feet out). The peak LHGR (KW/ft) at this rod position is approximately 21.5 for the peak 7 x 7 bundle and 19.0 for the peak 8 x 8 bundle.
If no rod block protection is assumed, the peak LHGR is approximately 22.0 KW/ft for the peak 7 x 7 bundle and 20.5 for the peak 8 x 8 bundle.
These transient LHGR values in Figure D-5 would occur for fuel at ~an exposure of 3 to 6 GMD/MTM.
Higher or lcwer exposures would result in a reduced LHGR peak for each fuel type.
The above results can be compared with the LHGR as a function of burnup for 0.75% clad strain for Oyster Creek fuel as shown in the following table:
Achieved Burnuo (GWD/MTM)
Steady State LHGR 0 0.75% Strain (KW/ft) 8x8 7x7 30 17.5 21.5 28 18 22 22 20 24 18 21 25 15 22 26 12 23 27 8
It can be seen that for exposures of 18 GWD/MTM or less, when compared to the worst LHGR values for the Rod Withdrawal Error Transient (Figure D-5), the 1
results are below the 0.75% strain values for both 7 x 7 and 8 x 8 fuel.
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The fuel misloading error 'for the worst case misloading results in a 17%
increase in the LHGR for 8 x 8 fuel and a 21% increase for 7 x 7 fuel.
Assuming the fuel were operating at the limiting LHGR of 17.2 KW/ft for 7 x 7 and 14.5 KW/ft for 8 x 8, the resulting 'LHGR for the fuel misloading error would be approximately 20.8 for 7 x' '7' and 17.0 for 8 x 8.
This would not result in clad i
l strain limits being exceeded as can be seen from the table above.
4
~
c What te 'the UlGR at centerline melt ca a fur.ction of burr.up?
Answer The:LHGR at centerline melt.as'a. function of burnup for both 7 x'7.and-
- 8 x18 fuel is presented in the following table:
LHGR 0 Centerline Melt Exposure (GWD/MTM)
(KW/ft) 8x8 0
25 1
25 2
25 4-24 10 23 15 22 25 21 7x7 1
26 2
26 4
25 1
10 24 14
'/a 20 23
-l 27.5 22 4
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QUESTION
. Discuss the derivation of the overpower ratio of 1.236 and justify the values.
RESPONSE
As-discussed in che Oyster Creek Station Technical Specifications. -
page 2.3-3, the APRM high neutron flux scram setting has been set to assure never reaching the fuel cladding integrity' safety it=it.
The system responds to neutron flux and is sec at 120% of rated po.'er to provide the protection
- while providing enough cargin to rated power to avoid spurious trips.
(See T.S. basis page 2.3-3, last paragraph.)
When power increases to 1690 and 1930 kWt were authorized, additional scram functions (turbine trip and generator load rejection) were added to the protective system to provide earlier response to anticipated trancients which result in rapid neutron flux increases.
The APRM high neutron flux scram was retained for protection against transients that result in slow power rises.
For these slow maneuvers or transients, core thermal power, surface heat flux and power transferred to the coolant follow the neutron flux so a scram occurring at a neutron flux-of 120% will assure thermal pcwcr has not exceeded 120% of rated thermal power.
Therefore, a neutron ilux scram at the safety li=it would be adequate to prevent violation of the safety licit.
A 3% margin between the ceram setpoint and the safety licit is maintained to account for any uncertainties that might exist. This is not a derived a wo 3 a lauludca fut cuu ui acium.
iuaLIuacui uncutialuty Luc tachet 1
Taking this information into account, a maximum steady state operating power icyc1 r be derived.
Operation at no greater than this power icvel will assure the safety 31 cit is not violated for these slow power Icyc1 increase tran _ents.
This power icvel is derived as follows:
The critical power for the limiting steady state power shape is calculated.
The safety limit cust then be established at a thernal power which is a ratio of 1.4 below that critical power. The safety limit pcwer is then reduced by a facter of 1.03 to provide a cargin of conservatise.
The resultant power is further reduced by a factor of 1.2 as discussed above in order to achieve the steady state operating power icvel. The ratio of the saf ety limit power to operating power is 1.236.
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DETAILS FOR SINGLE CEA'5EL
' MCHFR/MCPR CALCUU.TIONS FOR OYSTER CREEK 8x8 Fl'EL DISCUSSION The results of a single channel MCHFR calculation for various core power levels, axial power distributions and power factors are provided herein., Four power level cases were considered.
Cases 1 and II, 1765 MWt (MCPR = 1.73) and 3055,W.it (MCPR = 1.0), reflect radial, axial and local pcwcr f actors and an axial power distribution which are characteristic of steady state pcwer operation.
Cases Ill and IV, 1900. Tit (MCPR = 1.40) and 2660.m!c (MCPR = 1.0),
reflect the power factersand axial power distribution assured in the evaluation of the limiting transient with respect to thermal-hydraulic limits (Rod Withdrawal ErrorTransient).
The radial, axial and local power factors for each of the respective cases are provided in Table 1 and the pouer distributions are provided in Table 2..
The information presented in Tables 3 through 6 provides the pressure,enthalpy,(cassflo,w, quality,massvelocity,XN-2CHF,XN rod heat flux, CHFR {2 and the F-f actcr as a function of length.
The units for each variable are as follows:
VARIABLE ENGINEERING 15ITS Pressure PS1 Enthalpy BTU /lbo Macs Flow lbm/hr Quality Mass Velocity 106 lbm/hr-ft2 XN-2 CHF 106 BTU /hr-fc2
~
XN-1 CHF 106 BTU /hr-ft2 Rod Heat Flux 106 ETU/hr-ft2 CHFR F-factor Length Inchec The XN-1 CHF heat flux is the heat flux calculated using the X-l CHF correlation with all ccrrectors, i.e. spacer, local peaking and pressure correctors applied to the base XN-1 correlation. The XN-2 CHF heat flux is equal to the XN-1 CHF heat flux divided by the F-factor.
The F-factor is defined by the following expression:
ACHF F-factor =
q" (z)c -c(iCHF-Z) dZ ci q" CHF (1-e ChT)
O In the evaluatien cf the F-factor, the value of q" CHF is the rod heat flux
Q
_)
Cases III and IV consider the values of the above variables at critical power (Pc), and at the power Pc/1.4 for the Rod Withdrawal Error Transient.
Cases I and II consider the variables at the steady state operating power }FC and the critical power, Pc, corresponding to the steady state power factor.<3 and axial power distribution given in Tables 1 and 2.
The value of the pressure for Cases I and II is conservatively higher than normal operating pressure.
The pressure for Cases III and IV reflect l
the normal operating pressure at which the transient is initiated.
The assembly mass flow presented for each case includes allowance for 10% bypass flow and an engineering factor of 1.043 as discussed in Secti'ons III.C.5 and III. C. 6 of XN-74-32 (Rev. 3) uin Amendment No. 76.
The peaking factors used in Cases I and II are discussed in Appendix C of Supplement 3 to Amendment No. 76 and are conservative yet realistic representations of actual plant operating experience.
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v TABI.E l' POWER TACTORS In-II**.
Local' Peaking Factor
' 1.25
-1.26 Radial Peaking Factor 1.485 1.68 Axial Peaking Factor.
1.50,
1.60
' Heat Generation in Rod
.967 967
' Cases 1-and II'
- Cases'III and IV O
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TABLE 2 AXIAL POWER DISTRIB1*TIONS_
i 1
PEAK TO AVERAGE ACTIVE LENGTil, INC11ES 11=*10.'_.V NODE I * %,
r 138 0.5 0.818 12 (top) 126 0.7 1.403 11 114 0.92 1.601
/
- 10 102 1.12 1.454 9
.8' 90 1.32 1.344 78 1.47 1.241 7
1.47 1.042 66 6
54 1.32 0.955 5
42 1.12 0.866 4
30 0.92 0.619
)
3 18 0.70 0.420 2
f 6
0.50 0.238 1
1 (bottom) 4 i
L Cases I and II
- Cases III and IV i -
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TAPI.E 3 CASE I 1765 P't. FCTR = 1.73 XN-2 IN-2 IN-1 Red Flux F-F sr
'Fo rition Pressure Enthalpy Maes Flow Quality Mass Velo. h CEF CHF Rod Flux CFFR Axial Cerrectic 11.41 1025.527 517.95 97915.28
.04407
.9?500 1.03934
.93158
.05211 19.94467
.89632 17.12 1025.339 519.76
^
.04123
/\\
1.00943' n
.07451 13.54729
.92288 22.83 1025.151 522.25
.03733
.99004
.09610 10.10169
.94095 28.53 1024.840 525.42
.03234
.93118
.11744 8.35466
.94945 34.24 1024.652 529.32
.02625
.97956
.14000 6.96692
.95102 39.93 1024.465 533.97
.01901 l
.98091
.16735 5.66125
.94971 45.65 1024.154 539.48
.01039 I
.9S038
.19931 4.90644
.95022 51.36 3023.968 545.93
.00038
.96296
.22105 4.31729
.96741 57.07 1023.749 552.94
.01054
.94840
.21464 4.44511
.98226
-62.77 1023.519 560.29
.02197
.94329 w,
.24120 3.90993
.98758 68.45 1023.103 567.98
.03396
.94722
.93158
.24965 3.79419
.98349 74.19 1022.863 575.98
.04640
.94472
.91678
.26427 3.57486
.97042 79.59 1022.612 584.49
.05960
.93843
.8 ? S 76
.28823 3.25581
.95453 T;9.60 1022.037 593.72
.07402
.91686
.87209
.31200 2.93364
.95204 91.31 1021.783 603.60
.08938
.88062
.34?.44
.32627 2.69905
.96346 97.01 1021.483 613.87
.10332
.84919
.82364
.33064 2.52253
.96920 102.72 1021.187 624.53
.12187
.82393
.79666
.34751 2.37108
.96684 100.63 1020.447 635.60
.13914
.80430 76927
.36202 2.22167 5
114.13 1020.108 647.11
.15700
.79091
.740S0
.38345 2.06268
.L4 119.84 1019.751 659.15
.17570
.76257 71100 s--.40077 1.90278
.93237 125.55 1018.808 671.60
.19511
.70342
.68020
.39672 1.78568
.96017 131.25 1018.427 683.48
.21355
.63497
.65080
.36878' 1.72160 1.02%93 136.95 1018.040 6?4.53
.23070
.54219
.62345
.31658' 1.71264 1.14988 y
142.67 1016.970 703.67
.24493 Y
43184
.60]S6
.24416 1.76871 1.39138-143.37 1016.592 710.60 97915.28
.25577
.93500
.30581
.58369
.15993 1.90591 1.91495
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ai TABLE 4 C/SF. IV, 3055 r.'t, E t - 1.0 IN-2 IN-2 XN-1 itinn Preswure*
Enthalpy Mgss Flew Quality thss Velocitt CHF CHF Rod F';T F-Factor Rod Flux CHFR Axial Cerrection s.41 1025.527 518.?1 97915.28
.04367
.93500 1.03968
.93158
'.12 1025.339 520.74
.072?6 14.25089
.89602
.03970.
/t 1.01041 1.83 1025.151 524.23
.03526 l
.99153
.10432 9.68604
.92198 t.53 1024.840 529.67
.02731 8
,13455 7.36943
.93954
.98323
. 24 1024.652 334.13
.01881
.c8239
.16442 5.98009
.94747
). 95 1024.465 540.63
.00869
.98508
.19684 4.99079
.94S27 l
i.65 1024.149 548.36
.23430 4.20441
.94569
.00335
.98640
. 36 1023.905 557.38
.01737
.96907
.27974 3.52612
.94443
.07 1023.657 567.20
.C3263
.95309
.31227 3.10334
- .96131
.77 1023.401 577.49
.04062
.92814
.91304
.33776 2.74795
.98373
.32822 2.90382
.97743
'.4 8 1022.883 588.25
.06538
.90552
.88643
.34951 2.59082
.97892
.19 1022.601 599.45
.08279
.89206
.85370
.36997 2.41114
.96260
.89 1022.304 611.35
.10126
.P8262
.82927
.40353 2.18728
.93955 8.60 1021.592 624.28
.12142
.85757
.79726
.43680 1.95329
.92968
.31 1021.255 638.12
.14289
.81469
.76303
.45678 1.78856
.93659
.01 1020.394 652.50
.16520
.77486
.72746
.47130 1.64410
.93 SP '
.72 1020.513 667.42
.18335 74185
.69054 48652 1.52482
.930t
.43 1019.531 682.92
.21249 l
.715S8
.65219
.50683 1.41444
.90971E"
.13 1019.113 699.03
.23747 70217
.61233
.53683 1.30801
.87205
=
.84.
1018.674 715.89
.26361 l
.67820
.57061
.56108 1.20875
.84136
.33 1017.420 733.32
.29073
.62727
.52749
,35541 1,1;937
.84094
.25 1016.935 749.95
.31651
.55244
.4r633
.51629 1.07001
.88034
.?6.
1016.432 765.43
.34049 it 45475 44805 44321 1.02602
.93527 sf
.67 1014.971 778.21
.36043
.34139 41641
.34182
.99874 1.21975
.37 1014.471 787.93 97915.28
.37349
.93500
.22165
.39238
.22390
.98996 1.77028 h
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TAB 1E 5 7
CASE III, 1900 net, MCFR = 1.40l
~'
XN-2 F-factor -
XN-2
.IK-1 Rod flux
-Axial Position ~ Pressure Enthalpy Mass Flow Quality Nbss Velo CHF CRF una Flux CHFR' Corrector-
~
a.
-11.41 1240.292 518.69 95205.06-
.09665
.93912
.97466-
.93466 10070
- 9.67929
.95696 17.12 1240.104 521.97
.09t38
.95790
.93466 11991 7.98901
.97570 22.83 1239.417 525.92
.C8481
.95446
.93466 13580
.7,02852'
.97926 28.53 12'39.598 530.51-
.07715
.95563-
.93466
' 15286
- 6.25173
.97806 34'.24 1239.411
~535.71
.06353
.95756
.93466 17267 5.54508
.97609-39.45
'1239.224 541.53
.05388
.95838
.93466 19470 4.92231.
.97526 45.65 1238.904 508.04
.04:306
.95751-93466
.21710 4.00923.
.97614 51.36 1238.721 555.26
.03510
.95301
. 93466 23718 4.01875
.98075 57.07 1238.536 563.08
.02315
.94604
.93466 25010 3.79062
- .98589 62.77 1238.351~
571.33
.00949-
.94516
.93144 26207 3.60651
.95868
([y 66.48
~1238.036 580.01
.00490
.92983
.91039 27425 3.39043
.97909 74.19 1237.810 584.12
.01*490
.91146
.88827 29407 3.09945
.97455
.87948
.86491 30669 2.86768'
.98343
'79.89 1237.573 598.75
.03392
.83610
.84093 30267 2.76241 1.00582 89.60 1237.118 608.61
.05229 91.31 1236.871 618.13
.06505
.79641
.81786 28814 2.76396 1.026931 27061 2.82707
-1.03998 97.01 1236.617 627.31
.08323
.76502
-.79561 102.72 1236.359 635.45
.09753
.73666
.77164 25503 2.91131
.1.05155-108.43 1235.792 644.01
.11094
.70650
.75507 23404 3.01823'
- 1.06876 114.13 1235.529 651.46
.12326
.67390
.73701 21214
-3.17665 1.09364-119.84
.1235.266-658.19
.13i41
.64379
.72068 19004 3.38767 1.11900 125 55 1234.626 654.20
.]4144
.61707
.70608 18910 3.68427 1.14423 131.25 1234.366 669.59
.15337
.59414
.69301 15008 3.95875 1.16640 136.96 1234.103 674.37
.16129
.57320
.68141 13279 4.31600 1.18877 142.67 1235.419 676.53
.36427
.54845
.67132 11585 4.74243 1.22403 148.37 1233.154' 682.03 95205.06
.17.iO6
.93912
.50636
.66243' 09531 5.31292-1.30900 -
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TABLE 6, CASE IV, 2660 MWt.
MCFR = 1.0 XN-2 F-factor XN-2 XN-1 Rod Flux-AXIAL FOSITIO?!
PRES!URE E?rritALFY MMS FLUX
,qqALIU.,_
_F_X_ M "T.O CHF E
ROD m'r cpFR C01.1.ECTOR 11.41 1240.293 519.71 95205.06
.09517
.50942
.97522 43460
.17420 5.59816
.95842 i
17.12 1240.105 555.34
.09575
.95900 43460
.20740 4.62303
.97462 l
22.43 1239.917 532.22
.07942
.95600
.43460
.23493 4.06928
.97768 28.53 1239.598 540.16
.06123
.95420 43460
.26445 3.62358
.97540 34.24 1239.411 549.15
.0^634
.46164 43460
.29872 3.21441
.97140 39.95 1237.224 554.21
.02449
.46334
.43460
.33643 2.44297
.96423 45.65 1238.903 570.47
.01103
'46417 41352
.37564 2.56642
.96521 i
51.36 1238.713 542.97
.00703
.42991 40320 41029 2.26607
.97120
^
57.07 1238.453 596.50
.03242
.47370 47037
.43264 2.45904
.97645 62.77 1238.1 4 610.77 05)62 45796 43574 45330 1.59937
.97512 68.4%
1237.6G1 625.78
.03350
.82772
.79931
.47445 1.74454
.96568 74.19 1237.364 641.55
.10557
.80081
.76105
.50870 1.57404
.99035 79.89 1237.054 658.21
.13i11
.75680
.72063
.53057 1.42549
.992:4 85 60 1236.323 675.27
.16!34
.69300
.67922
.52362 1.32348
.98011 91.3L 1235.790 691.74
.18161
.62887
.63425
.49849 1.26076 1.01716 97.01 1235.640 707.61
.21583
.47070
.60075 46815 1.21426 1.05248 102.72 1235.296 722.56
.24154
.51A27
.56447
.43773 1.14422 1.03705 108.43 1234.366 726.51 25362 46434
.53062 40493 1.15784 1.13166 114.13 1234.003 744.34
.24594 41733 49418
.36701 1.13614 1.14741 119,84 1233.036 761.03
.34418
.36782 47112
.32871 1.1167%
1.28087 125.35 1232.?64 771.44
.32123
.32331 44586
.29251 1.10922 1.37*12 131.25 1232.181 780.76
.31i47
.38400
.42325 24462 1.09362 1.49830 ft 136.96 1231.794 789.03
.35054
.24400 40318
.22973 1.09386 1.61926 142.47 1230.531 798.23
.36252
.21710
.38571
.20007 1.08515 1.77668 143.37 1230.238 802.28 95205.06
.37651
.50412
.17778
.37104
.18428 1.07820 2.08710'
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QUESTION 65.
The spectrun of breaks submitted does not meet r ga yirements of Appendix K to 10 CFR Part 50, which requires t6at dhe*Mo.-dy multiplier i
range from 0.6 to 1.0 be spanned.
It is our position that the spectrum of breaks analyzed must include recirculation line breaks with approxi-mate areas (in square feet of 0.75, 2.0 and 4.0).
In addition as per our discussions, the following break locations must be analyzed:
feedwater line, cain steam line and core spray line. These analyses must be performed for both GE and Exxon fuel.
RESP 0NS,E Theresultsof,theblevdownandgeatupanalysesforGEfuelinOyster-Creek for the 0.75 ft', 2.0 ft2, 4.0.ft, main steamline, feedwater line and core spray line breaks are included in the following figures.
In addition, we have included the peak cladding tecperature and heat transfer coefficients versus time for the 0.02 and 0.05 ft' breaks which were emitted from our April 24, 1975 submittal because of their proprietary nature.
The core spray system at the Oyster Crcok Nuclear Generating Station consists of tuo identical loops either capable of supplying rated core spray flow.
The r.ctive components in each loop are redundant, each loop is powered by a separate diesel generator and the two loops are completely separated frem cach other so that effects on one loop frem missiles or' rupture will not affact esn neune snnn run ev.~nn eu...en-n ane<n-na en an,1ve-
,,,,a nn-n enr,7 flow to the reactor vessel in the event of an unlikely loss-cf-coolant accident even if there is no off-site pcwor available and emergency power is not available to one of the core spray systc=s due to a fault en one of the emergency diesel buses.
This is the case with all postulated pipe breaks within the reactor coolant system pressure boundary except for a break in one. of the core spray lines between the reactor vessel and the ecre spray check valves, a run of approximately 28 ft. of 6 inch ID pipe in each of the two core spray loops.
Should the loss-of-coolant accident result.frcm a break in this portion of a core spray loop, and no off-site power is available and the' diesel generator bus which powers the other core spray loop is assured to be the single pressure failure, core spray flew would not automatically reach the core.
However, the system is designed to indicate this event to the operator end sufficier.t time is available for the operator to supply water to the corc to permit cooling sufficient to mact the NRC's Final Acceptancc Criteria for ECCS.
The reliance on operator action in the emergency core cooling sequence in this case is justified for several reasons:
1.
The specific event probability is very small.
The probability of any loss of coolant accident is small in itself and the probability of a rupture in a specific run of pipe the length of which is small compared to the total pipe in the reactcr coolant pressure boundary is even smaller.
In all likelihood, were a LOCA to occur, cmergency 65-1
r Ic4Js including core spray would be'ponered from the highly reliable t
cif-site power network. The low probability of its unavailability and the simultaneous failure of the diesel generator associated with the core spray system which is,still intact when coupled with the very low probability of the LOCA in the core spray line reeults in a very unlikely event.
2.
Tic system is designed to alert the operator to this specific event.
In addition to all of the normal indications of a LOCA (i.e.,
reactor scram, high drywell pressure, decreasing water icvel, containment isolation, etc.) the operator is provided with a specific visual and audible alarm (individual one for each core spray loop) which reads " Core Spray System I Pipe Break" and " Core Spray
- System II Pipe ~ Break"..These alarms are initiated by differential pressure detectors on each core spray loop which compare the pressure in the bottom plenum of the reactor vessel with that in the core spray line just upstreau of the reactor vessel nozzle.
If-the pressure in the core spray line drops 18 psi below that in the reactor vessel lower plenum, the alarm is initiated signaling a depressur-ization of the core spray line, that is, a rupture.
Regardless of the location of the break in the non-isolable portiens of the core spray system these sensors would sense the depressurization.
The differential pressure detectors thecselves are outside the pri-mary containment and the sensing lines within containment are routed so as to preclude datage from a rupture in the other cort spray system.
3.
Concise procedures are available and operater action is cuickly accomplished.
Emergency Frecedures call for verificatien of both core spray systems running af ter LOCA indications.
Furth'er-more, indication of a " Core Spray System Pipe Break" calls for veritication that the other core spray system is operable.
11 the second core spray system diesci is not operating.then procedures call'for reduction of load on the operating diesel, connecting the two emergency buses together and once this is. accomplished the emergency safeguards loads on that bus, including core spray will automatically sequence on.
If the failure is due to a fault on the bus, which powers the operable core spray system, the operator will not be able to close the control room breaker which inter-connects the two emergency buses.
In order to supply adequate cooling to the core the operator must take action to start pumping water to the core by means of a condensate pump. There exists adequate water supply for the condensate pump for at least ten minut'es (45,000 gal.) frem the condenser hot wells.
In less than three (3) minutes the core would be covered.
After seven minutes of operatiens the condensate pump-could be secured due to the fact that the event has been turned around and the reactor pressure has been reduced to a level such that the fire pond pumps can supply adequate water through the core spray header to maintain the reactor water level above the active fuel.
65-2 Y
^
&^
, )
On May 21, 1975 the procedure to initiate the condensate pu:P
. water to the core was walked through at Oyster Creek and the operator was able to effect the necessary actions in less than' l
six (6) minutes.
Even recognizing the pressure of the emergency situation, this demonstration, the availability of clear indication of the probica, and concise action requiremencs, and the fact that all immediate actions required to establish adequate cooling flow can be~acceeplished provides reasonable assurance that these actions can be accomplished in time to provide adequate cooling required to meet the ECCS Final Acceptance Criteria.
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. RESULTS OF CORE SPRAY LINE EREAK LOSS-OF-COOLANT ACCIDEST EVALUATION ACCIDENT:- Break in one loop of core spray system.
Fault in.bes which provides power to' unbroken core spray loop.
SINGLE FAILURE:
Condensate pumps initiated 562 seconds into event.
MITIGATION ASSUMPTIONS:
HEATUP RESULTS (EXXON NUCLEAR FUEL)
Time (Sec)_
PCT (OF)
HTC_
0 503 19,306
% Local Metal-water 2.2 571 19,306 Reaction = 2.23%
2.3 574 2,467 50 926 46 44 70 1028 90 1087 35 13.4 1136 34 20.5 1141 34 SENSITIVITY TO Toumo 27.0 1130 34 35.5 1113 34 Tpump 47.0 1089 34 (sec)
PCT ( F)
% MWR 62.5 1053 34 i
42 1:00 14R1 0.16 82.5 968 936 42 500 1824 0.74 125.5 887 38-562 2139 2.28 19 5 161.5 863 32 600 2425 4.91 191.0 068 25 219.5 887 20 252.5 919 16 290.5 974 12 310.0 990 13 310.5 996 0
0 384.0 1347 510.0 1804 0
590.0 2139 0
595.0 2021 20
'675.0 1221 20 775.0 842 20 890.0 687 20 4
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