ML20127M820
| ML20127M820 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/27/1985 |
| From: | Morrill P, Pate R, Shiraki C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20127M740 | List: |
| References | |
| 50-275-OL-85-02, 50-275-OL-85-2, NUDOCS 8507010361 | |
| Download: ML20127M820 (113) | |
Text
r 4
U. S. NUCLEAR REGULATORY COMMISSION REGION V
. Report No.
50-275/0L-85-02 Docket Nos.
50-275 and 50-323 Licensee:
Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106 Facility Name:
Diablo Canyon Units I and 2 Examinations at:
Avila' Beach, California Examination conducted:
May 21-23, and 29-30, 1985 b
Examiners:
C. Y( Shi aki Date Signed h
dil2 f
~
P. 6. Morrill Date Signed-Approved By:
[ M6
[/27 f
R.~J. Fate', Chief, Operations Section Bate ' Signed Summary:
Examinations were conducted from May 21 to 30, 1985. The written examination was administered on May 21, 1985 to seven senior reactor operator candidates (SRO), seven reactor operator (RO) candidates and one instructor candidate.
In addition, one instructor candidate took only one section of the written examination. One SRO candidate failed a section of the written examination and failed overall but passed the operating examination. 'Two SRO candidates failed the operating examination, but passed the written examination. The instructor candidate who took only one section of the written examination passed that section. This results in four SRO licenses and two instructor certifications being issued. All seven reactor operator candidates passed the written and operating examinations. This results in seven R0 licenses being issued.
L 8507010361 850627 PDR ADOCK 05000275 G
PM i
G:
REPORT DETAILS 1.
Examiners:
7 C.'Shiraki, Chief Examiner', Region V-P.,Morrill, Region'V L. Defferding, Pacific Northwest Laboratories R.'Schreiber,. Pacific Northwest-Laboratories s
2.
Persons' Attending the Exit Meetings:
May 23, 1985-NRC C. Shiraki, Region V P.-Morrill, Region V-L. Defferding, Pacific Northwest Laboratories M.'Mendonca, Senior Resident Inspector Pacific-Gas and Electric Company R; Thornberry, Plant Manager
-T. Martin, Training Manager R. Fisher, Senior Operations Engineer R. Graham, Senior-Training Instructor W. Kaefer, Assistant Plant Manager / Support Services J. Molden, Operations Training Supervisor May!30,'1985
-NRC C.-Shiraki, Region V' P. Morrill, Region V R. Schreiber, Pacific Northwest Laboratories M. Mendonca, Senior Resident Inspector
.T. Ross, Resident Inspector T. Polich,. Resident Inspector Pacific Gas and Electric Company R.. Patterson, Plant Superintendent :
J. Sexton, Operations Manager T. Martin, Training Manager J. Holden, Operations Training Supervisor b +
e l
I.:
rc
. 3.
Facility Review of the Written Examination At the conclusion of the written examinations, the examiners met with R. - Graham, W. Steinke, J. Molden and J. Tinlin of the Training Department to review the written examinations and answer keys.
Their comments and additional ones from a letter dated May 29, 1985 were incorporated in to the master examination keys prior to grading the candidates' responses.
4.
Exit Meetings Operating Examinations Simulator and operating examinations were conducted on May 22, 23, 29 and 30, 1985. Several candidates shared similar weak areas.
They were unfamiliar with an intersystem loss of coolant a.
accident in the residual heat removal system.
b.
They did not understand the response of a resistance temperature detector under open circuit and short circuit conditions, c.
They were not familiar with the trips associated with the diesel generator under different operating conditions.
In addition, the following suggestions were provided that might assist in the smoother opera; ion of the plant.
a.
It was felt that the merit of writing a procedure for oetection and isolation of a small secondary loss of coolant accident should be investigated.
b.
The method for calculating boron concentration could be simplified considerably by consolidating the different tables and instructions into one procedure.
c.
The entrance and exit points for the radiation control areas are in close enough proximity that contamination of an individual who is entering the area by one who is exiting it, is highly possible.
Clear Passes The names of the candidates who clearly passed both the simulator and operating examinations were provided.
i U. S. NUCLEAR REGULATORY COMMISSION K
o SENIOR REACTOR OPERATOR EXAMIHATION Facility: DIABLO CANYON UNITS 1 AND 2 Reactor Type: WESTINGHOUSE PVR Date Administered: MAY 21, 1985 Examiner:
P. HDRRILL Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheet.
Points for each question are indicated in parenthesis after the question.
The passing grade requires at least 70% in each category and a final grade of at leest 80%.
Examination papers will be picked up six (6) hours after the examination starts.
Category
% of Ca ndidate's
% of Value Total Score Cat. Value Category o?Y75%
-MA%5 4n
- 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 25 M
- 6. Plant Systems Design, Cdntrol
- 2725y, and Instrumentation 25
+52-
- 7. Procedures - Normal, Abnornal, Emergency and Radiological pij%
Control N
~
- 8. Administrative Procedures, Conditions, and Limitations
+ 00-- d 3 TOTALS Final Gra*de all work done on this examination is my own; I have neither given nor received aiC.
Cancicate's Signature
EQUA110:1 f,HEET
'f 'o ca '
v o s/t Cycle efficien:y - (.Networt
~
out)/(Energy in)
?
o = ag s - V,t + 1/2 'at 2
7 g,,c
~
KE = 1/2 av a = (Vg - V,)/t A = AN A = A,e
~
FE t 89h Vg = V, + a t w = 8/t i = an2/t1/2 = 0.693/t1/2 1/2'II
- EI*1DII*bN t
~
((t I
- II )]
1/2 D
aE = 931 ma g = aCpat I = I,e daUAaT I = I,e~"*
! = I,10'*M Pwr = W an g
TVL = t.3/u P = P 10 "'I*}
5 HVL = -0.693/v P = P e*
SUR = 26.06/T SCR = 5/(1 - K,7f)'
CR, = 5/(1 - K,7f,)
SUR = 26e/t* + (s - p)T CR (1 - K,773) = CR II - teff2) j 2
l T = (1*/6) + [(a - e)/Ap]
M'= 1/(1 - K,77) = CR /CR, j
T = s/(, - s)
M = (1 - K,7f,)/(1 - K,7f j)
T = (8
-e)/( Ap)
SDM = (1 - K,7f)/K,ff l
e = (K,ff-1)/K,ff = AK,ff/K,7f 1* = 10 seconds
~ T = 0.1 seconds ~I o = [(t*/(T K,ff)] + [s,ff (1 + AT))
/
Idjj=Id 2,2 2 l
P = (reV)/(3 x 1010)
Id
~
gd jy 22 2
I = cN R/hr = (0.5 CE)/d (meters)
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbs.
I curie = 3.7 x 1010dps 1 gag.=3.78 liters 1 kg = 2.21 lba 3
= 7.48 gal.
1 hp = 2.54 x 10 8tu/hr 1 ft Density = 62.4 lbs/ft3 1 m = 3.41 x 106 8tu/hr Density = 19a/cse' lin = 2.54 cm Heat of vaportration = 970 8tv/lem
- F = 9/5*C + 32 Heat of fusion = 144 8tv/lbe
- C = 5/9 (*F-32)
I Ata = 14.7 psi = 2g.9 in. Hg.
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l 3n Appenosa a-ensecAt poopennes or muses Aae now cuanacteenstics or vAtvas. mnwos. Aeso pipe CRANE Properties of Saturated Steam and Saturated Water
- i Abeolute Pressure Vacuum Temper.
Heat of Latent Heat Total Heat specific Volume I
laches laches ature the of of steam y
)
1 Lbe. 5" sq.
of Hg of Hg Liquid Evaporation t
Ile Water 5 team P'
n r.
o.o fi..
e fim.
e fi.
co. n.
, e,.
co.e..
in.
O.0007 8.02 29.90 32.018 0.0003 1075.5 1075.5 0.016022 3302.4 i
0.10 0.20 29.72 35.023 3.026 1073.8 1076.8 0.016020 2945.5 0.15 0.31 29.61 45.453 13.498 1067.9 1001.4 4.016020 2004.7 0.20 0.41 29.51 53.160 21.217 1053.5 1084.7 0.016025 -
1526.3 0.25 0.51 29.41 59.323 27.302 1060.1 1087.4 0.016032 1235.5 0.30 0.61 29.31 H.404 32.541 1057.1 1089.7 0.016040 1039.7 1
0.35 0.71 29.21 68.939 "
M.992 1054.6 1091.6 0.016048 898.6 3
0.40 0.81 29.11 72.8H 40.917 1052.4 1093.3 0.016056 792.1 1
0.45.
0.92 29.00 76.387 44.430 1050.5 1094.9 0.016063 708.8
{
0.50 1.02 28.90 79.586 47.623 1040.6 1996.3 0.016071 Hl.5 0.60 1.22 28.70 85.218 53.245 1045.5 1090.7 0.016085 540.1 0.70 1.43 28.49 90.09 58.10 1042.7 1100.8 0.016099 466.94 0.00 1.63 28.29 94.38 62.39 1040.3 I102.6 0.016112 411.69 i
0.90 1.83 28.09 98.24 M.24 1038.1 1104.3 0.016124 368.43 1.0 2.04 27.08 101.74 69.73 1036.1 1105.8 0.0161M 333.60 l
1.2 2.44 27.40 107.91 75.90 1032.6 1108.5 0.016154 280.%
1 1.4 2.85 27.07 113.26 81.23 1029.5 1110.7 0.016178 243.02 1.6 3.26 26.M 117.98 85.95 1026.8 1112.7 0.0161 %
214.33 1.8 3.M 26.26 122.22 90.18 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.85 126.07 94.03 1022.1 1116.2 0.0lt230 173.76 2.2 4.40 25.44 129.61 97.57 1020.1 1117.6 0.016245 158.87 2.4 4.89 25.03 132.80 100.84 1018.2 1119.0 0.016260 t e.40 2.6 5.29 24.63 135.93 103.88 1016.4 1120.3 0.016274 135.80 2.8 7.70 24.22 138.78 106.73 1014.7 1121.5 0.016257 126.67 3.0 6.11 23.81 141.47 109.42 1013.2 1122.6 0.016300 118.73 3.5 7.13 22.79 147.M 115.51 1009.6 1125.1 0.016331 102.74 4.0 8.14 21.78 152.%
120.92 1006.4 1127.3 0.016358 90.64 4.5 9.16 20.76 157.02 125.77 1003.5 1129.3 0.016384 83.03 5.0 10.18 19.74 162.24 130.20 1000.9 1131.1 0.016407 73.532 5.5 11.20 18.72 lu.29 134.26 998.5 1832.7 0.016430 67.249 6.0 12.22 17.70 170.05 138.03 996.2 1134.2 0.0lH51 61.984 6.5 13.23 16.69 173.56 141.54 994.1 1135.6 0.016472 57.506 7.0 14.25 15.67 176.84 144.83 992.1 1136.9 0.0lH91 53.650 7.5 15.27 14.65 179.93 147.93 990.2 1138.2 0.C
- o310 50.294 8.0 16.29 13.63 182.06 150.87 988.5 1139.3 0.016527 47.345 8.5 17.31 12.61 185.63 153.65 986.8 1140.4 0.016545 44.733 9.0 18.32 11.60 188.27 156.30 985.1 1141.4 0.016561 42.402 9.5 19.34 10.58 190.80 158.84 983.6 1142.4 0.016577 40.310 10.0 20.M 9.56 193.21 161.26 982.1 1143.3 0.016592 38.420 11.0 22.40 7.52 197.75 165.82 979.3 1145.1 0.0lM22 35.142 12.0 24.43 5.49 201.%
170.05 976.6 11 #.7 0.01M50 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1148.2 0.01M76 30.057 14.0 38.50 1.42 209.56 177.71 971.9 1149.6 0.016702 28.043 i
Pressure Temper.
Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In.
ature the of of Steam y
Absolute GaSe Liquid Evaporation t
Il Water Steam e
l P'
P o.,
r.
s.on.
n..
o.on.
co n,,i..
co. n, n..
l 14.6 %
0.0 212.00 180.17 970.3 1150.5 0.616719 26.799 15.0 0.3 213.03 181.21 969.7 1150.9 0.016726 26.290 l
16.0 1.3 216.32 184.52
%7.6 1152.1 0.016749 24.750 1
17.0 2.3 219.44 187.M 965.6 1153.2 0.016771 23.385 18.0 3.3 222.41 190.M 963.7 -
1154.3 0.016793 22.168 19.0 4.3 225.24 193.52
%I.8 1155.3 0.016814 21.074 20.0 5.3 227.%
196.27 960.1 I156.3 0.016834 20.087 21.0 6.3 230.57 195.90 958.4 1157.3 0.016454 19.190 22.0 7.3 233.07 201.44 956.7 1158.1 0.016873 18.373 23.0 8.3 235.49 203.88 955.1 1159.0 0.016891 17.624 24.0 9.3 237.82 206.24 953.6 1159.8 0.016909 16.936 25.0 10.3 240.07 200.52 952.1 1160.6 0.016927 16.301 26.0 11.3 242.25 210.7 950.6 1161.4 0.016944 15.7138 27.0
- 12.3 244.M 212.9 949.2 1162.1 0.016961 15.1684 28.0 13.3 2 #.41 214.9 947.9 1162.8 0.016977 14.6607 29.0 14.3 240.40 217.0 9#.5 1163.5 0.016993 14.1969 30.0 15.3 250.34 218.9 945.2 IIH.1 0.017009 13.74M 31.0 16.3 252.22 220.8 943.9 IIM.8 0.017024 13.3280 32.0 17.3 254.05 222.7 942.7 1165.4 0.017039 12.9376 33.0 18.3 255.04 224.5 941.5 Ilu.0 0.017054 12.5700 i
34.0 19.3 257.50 226.3 940.3 IIM.6 0.017069 12.2234
- Atntracted imm ASME Steam Tables (l%7), with permission of the publisher. The American
( m d e.
Society of Mechanical En8meers. 345 East 47th Screet. New York. New York 10017.
- m. a e pos.)
l l
A-13 CRANE
' Aprewois A-PwmcAt reoreeties ce nuos Awo now enamacteeisnes ce vams. rminos. Awo eine Properties of Saturated Steam and Saturated Water-w:entinued Pressure Temper-Heat of Latent Heat Total Heat Specific Volume Lbe. per Sq. In.
ature the of of Steam p
Liquid Evaporation
'i Wa i
Steam Absol,ute Gage I
co. ri..t.er
. is i
co.<a
,is.
P P
c_ r.
m ons.
m ons.
nio ns.
35.0 20.3 259.29 224.0 939.1 1167.1 0.017083 11.8959 36.0 21.3 260.95 229.7 938.0 1167.7 0.017097
!!.58%
37.0 22.3 262.58 231.4 936.9 1168.2 0.017111 11.2923 38.0 23.3 264.17 233.0 935.8 1168.8 0.017124
!!.0136 39.0 24.3 265.72 234.6 934.7 1169.3 0.017138 10.7487 40.0 25.3 267.25 2%. I 933.6 1169.8 0.017151 10.4 % 5 41.0 24.3 268.74 237.7 932.6 1170.2 0.017164 10.2563 42.0 27.3 270.21 239.2
- 931.5 1170.7 0.017177 10.0272 l
43.0 28.3 271.65 240.6 930.5 1171.2 0.017189 9.8083 m'
l 29.3 273.06 242.1 929.5 1171.6 0.017202 9.5991 i
44.0 30.3 274.44 243.5 928.6 1172.0 0.017214 9.3988 i
M.0 31.3 275.80 244.9 927.6 1172.5 0.017226 9.2070 45.0 47.0 32.3 277.14 246.2 926.6 1172.9 0.017238 9.0231 48.0 33.3 278.45 247.6 925.7 1173.3 0.017250 8.8465 49.0 34.3 279 /4 248.9 924.8 1173.7 0.017261 8.6770 50.0 35.3 281.02 250.2 923.9 1874.1 0.017274 8.5140 51.0 h.3 282.27 251.5 923.0 1174.5 0.017285 8.3571 52.0 37.3 28J.50 252.8 922.1 1174.9 0.0172 %
8.2061 53.0 38.3 284.71 254.0 921.2 1175.2 0.017307 8.0606 54.0 39.3 285.90 255.2 920.4 1175.6 0.017319 7.9203 55.0 40.3 287.08 256.4 919.5 1175.9 0.017329 7.7850 56.0 41.3 288.24 257.6 918.7 1176.3 0.017340 7.6543 57.0 42.3 289.38 258.8 917.8 1176.6 '
O.017351 7.5280 58.0 43.3 290.50 259.9 917.0 1177.0 0.017362 7.4059 510 44.3 291.62 261.1 916.2 1177.3 0.017372 7.2879 60.0 45.3 292JI 262.2 915.4 1177.6 0.017383 7.1736 61.0
%.3 293.79 263.3 914.6 1177.9 0.017393 7.0630 62.0 47.3 294.86 264.4 913.8 I178.2 0.017403 6.9558 63.0 48.3 295.91 265.5 913.0 1178.6 0.017413 6.8519 64.0 49.3 2 %.95 266 6 912.3 1178.9 0.017423 6.7511 65.0 50.3 297.98 267.6 911.5 1179.1 0.017433 6.6533
%.0 51.3 298.99 268.7 910.8 1179.4 0.017443 6.5584 67.0 52.3 299.99 269.7 910.0 1179.7 0.017453 6.4662 68.0 53.3 300.99 270.7 909.3 1180.0 0.017463 6.3767 69.0 54.3 301.%
271.7 908.5 1180.3 0.017472 6.28 %
70.0 55.3 302.93 272.7 907.8 1180.6 0.017482 6.2050 71.0 56.3 303.89 273.7 907.1 1180.8 0.017491 6.1226 72.0 57.3 304.83 274.7 906.4 1181.1 0.017501 6.0425 73.0 58.3 305.77 275.7 905.7 1181.4 0.017510 5.% 45 74.0 59.3 306.69 276.6 905.0 1181.6 0.017519 5.8885 75.0 60.3 307.61 277.6 904.3 1181.9 0.017529 5.8144 75.0 61.3 308.51 278.5 903.6 1182.1 0.017538 5.7423 77.0 62.3 309.41 279.4 902.9 1182.4 0.017547 5.6720 78.0 63.3 310.29 280.3 902.3 1182.6 0.017556 5.6034 79.0 64.3 311.17 281.3 901.6 1182.8 0.017565 5.5364 80.0 65.3 312.04 282.1 900.9 1183.1 0.017573 5.4711 81.0 66.3 312.90 283.0 900.3 1183.3 0.017582 5.4074 82.0 67.3 313.75 283.9 899.6 1183.5 0.017591 5.3451 83.0 68.3 314 60 284.8 899.0 1183.8 0.017600 5.2843 84.0 69.3 315 43 285.7 8W 3 1184.0 0.017608 5.2249 85.0 70.3 316.26 286.5 897.7 1184.2 0.017617 5.1669 0.017625 5.1101 86.0 71.3 317.08
- 287.4 897.0 1184.4 '-
87.0 72.3 317.89 288.2 8%.4 1184.6 O.017634 5.0546 88.0 73.3 318.69 289.0 895.8 1184.8 0.017642 5.0004 89.0 74.3 319.49 289.9 895.2 1185.0 0.017651 4.9473 90.0 75.3 320.28 290.7 894.6 1185.3 0.017659 4.8953 91.0 76.3 321.06 291.5 893.9 1185.5 0.017667 4.8445 92.0 77.3 Mt.Se 292.3 893.3 1185.7 0.017675 4.7947 93.0 78.3 322.61 293.1 892.7 1185.9 0.017684 4.7459 94.0 79.3 323.37 293.9 892.1 1186.0 0.C17692 4.6982 95.0 80.3 324.13 294.7 891.5 1186.2 0.017700 4.6514
%.0 81.3 324.88 295.5 891.0 1186.4 0.017708 4.6055 97.0 82.3 325.63 2%.3 890.4 1186.6 0.017716 4.5606 98.0 83.3 326.36 297.0 889.8 1186.8 0.017724 4.5166 99.0
- B4.3 327.10 297.8 889.2 1187.0 0.017732 4.4734 100.0 85.3 327.83 298.5 888.6 1187.2 0.017740 4.4310 101.0 86.3 328.54 299.3 888.1 1187.3 0.01775 4.3895 102.0 87.3 329.26 300.0 887.5 1187.5 0.01776 4.3487 103.0 88.3 329.97 300.8
' 886.9 1187.7 0.01776 4.3087 104.0 89.3 330.67 301.5 886.4 1187.9 0.01777 4.2695 105.0 90.3 331.37 302.2 885.8 1188.0 0.01778 4.2309 106.0 91.3 332.%
303.0 885.2 1188.2 0.01779 4.1931 107.0 92.3 332.75 303.7 884.7 1188.4 0.01779 4.1560 108.0 93.3 333.44 304.4 884.1 1188.5 0.01780 4.1195 109.0 94.3 334.11 305.1 883.6 1188.7 0.01781 4.0837
,A-14 Artenma a-rwysicat peoceanes or nues Ano now cuAnacteeisncs or vnves. remnos a e pipe CRANE Properties of Saturated Steam and Saturated Water-continued Pressure Temper-Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In.
ature the of of Steam p
WW b eparat h h,
Absolute Ga8e t
St Wate co. e.,.,r co. e eam P'
P o.or F.
n im a.orm.
an.im.
m.
.,., m.
110.0 96.3 334.79 306.8 883.1 1188.9 0.01782 4.0484 111.0
%.3 335.#
306.5 882.5 1189.0 0.01782 4.0138 112.0 97.3 3M.12 307.2 882.0 1189.2 0.01783 3.9798 113.0 98.3 3M.78 307.9 881.4 1189.3 0.01784 3.9464 114.0 99.3 337.43 308.6 880.9 1889.5 0.01785 3.9136 115.0 100.3 338.08 309.3 880.4 1189.6 0.01785 3.8495 3.8813 116.0 101.3 338.73 309.9 879.9 1189.8 0.01786 117.0 102.3 339.37 310.6 879.3 1189.9 0.01787 3.8183 118.0 103.3 340.01
,311.3 878.8 1190.1 0.01787 3.7875 119.0 104.3 340.64 311.9 878.3-1190.2 0.01788 3.7573 120.0 105.3 341.17 312.6 877.8 1890.4 0.01789 3.7275 121.0 106.3 341.89 313.2 877.3 1190.5 0.01790 3.6983 122.0 107.3 342.51 313.9 876.8 1190.7 0.01790 3.u 95 123.0 108.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132 115.0 110.3 344.35 315.8 875.3 1191.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1191.2 0.01793 3.5586 127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 346.15 317.7 873.8 1191.5 0.01794 3.5057 129.0 114.3 346.74 318.3 873.3 1191.6 0.01795 3.4799 130.0 115.3 347.33 319.0 872.8 1191.7 0.017 %
3.4544 131.0 116.3 347.92 319.6 872.3 1191.9 0.01797 3.4293 132.0 117.3 348.50 320.2 871.8 1192.0 0.01797 3.4046 133.0 118.3 349.08 320.8 871.3 1192.1 0.01798 3.3802 134.0 119.3 349.65 321.4 870.8 1192.2 0.01799 3.3562 135.0 120.3 350.23 322.0 870.4 1192.4 0.01799 3.3325 136.0 121.3 350.79 322.6 869.9 1192.5 0.01800 3.3091 137.0 122.3 351.36 323.2 869.4 1192.6 0.01801 3.2861 138.0 123.3 351.92 323.8 868.9 1192.7 0.01801 3.2634 139.0 124.3 352.48 324.4 M8.5 1192.8 0.01802 3.2411 140.0 125.3 353.04 325.0 868.0 1193.0 0.01803 3.2190 141.0 126.3 353.59 325.5 867.5 1193.1 0.01803 3.1972 142.0 127.3 354.14 326.1 867.1 1193.2 0.01804 3.1757 143.0 128.3 354.69 326.7 866.6 1193.3 0.01805 3.1546 144.0 129.3 355.23 327.3 866.2 1193.4 0.01805 3.1337 145.0 130.3 355.77 327.8 865.7 1193.5 0.01806 3.1130 lu.0 131.3 3M.31 328.4 E5.2 1193.6 0.01806 3.0927 147.0 132.3 356.84 329.0 M4.8 1193.8 0.01807 3.0726 148.0 133.3 357.38 329.5
%4.3 1193.9 0.01808 3.0528 149.0 134.3 357.91 330.1 863.9 1194.0 0.01808 3.0332 150.0 135.3 358.43 330.6 863.4 1194.1 0.01809 3.0139 152.0 137.3 359.48 331.8 862.5 1194.3 0.01810 2.9760 154.0 139.3 360.51 332.8 861.6 1194.5 0.01812 2.9391 156.0 141.3 361.53 333.9 860.8 1194.7 0.01813 2.9031 158.0 143.3 362.55 335.0 859.9 1194.9 0.01814 2.8679 160.0 145.3 363.55 3%.I 859.0 1195.1 0.01815 2.8336 162.0 147.3 364.54 337.1 858.2 1195.3 0.01817 2.8001 164.0 149.3 365.53 338.2 857.3 1195.5 0.01818 2.7674 1%.0 151.3 366.50 339.2 856.5 1195.7 0.01819 2.7355 168.0 153.3 367.47 340.2 855.6 1195.8 0.01820 2.7043 170.0 155.3 368.42 341.2 854.8 11 %.0 0.01821 2.6738 172.0 157.3 369.37 342.2 853.9 11 %.2 0.01823 2.6440 174.0 159.3 370.31 343.2 853.1 11 %.4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 %.5 0.01815 2.5864 178.0 163.3 372.16 345.2 851.5 11 %.7 0.01826 2.5585 180.0 165.3 373.00 346.2 850.7 11 %.9 0.01827 2.5312 182.0 167.3 373.98 347.2 849.9 1197.0 0.01828 2.5045 184.0 169.3 374.88 348.1 849.1 1197.2 0.01830 2.4783 1%.0 171.3 375.77 349.1 848.3 1197.3 0.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 8#.7 1197.6 0.01833 2.4030 192.0 177.3 378.40 351.9 845.9 1197.8 0.01834 2.3790 194.0 179.3 379.26 352.8 845.1 1197.9 0.01835 2.3554 1%.0 181.3 380.12 353.7 844.4 1198.1 0.018M 2.3322 190.0 183.3 380.%
354.6 843.6 1198.2 0.01838 2.3095 200.0 185.3 381.80 355.5 842.8 1198.3 0.01839 2.28728 205.0 190.3 383.38 357.7 840.9 1198.7 0.01841 2.23349 210.0 195.3 385.91 359.9 839.1 1199.0 0.01844
'2.18217 215.0 200.3 387.91 362.1 837.2 1199.3 0.01847 2.13315 220.0 205.3 38*.88 M4.2 835.4 1199.6 0.01850 2.08629 225.0 210.3 391.00 366.2 833.6 1199.9 0.01852 2.04143 230.0 215.3 393.70 368.3 831.8 1200.1 0.01855 1.99846 235.0 220.3 395.M 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3 397.39 372.3 828.4 1200.6 0.01860 1.91769 1
248.0 2J0.3 399.19 374.2 826.6 1200.9 0.01863 1.87970 l
A-15
', t Awesem A-macAt reoreenas o' nosos am now enamactearrics or vAtvts, empeos. Ano met CRANI Propertie8 of Saturated Steam and Satur.;.d Water-<onduded Pressure Temper.
Heat of Latent Heat Total Heat Specific Volume the of of Steam p
Lbe. per Sq. In.
sture WW beporet h h,
Water Steam Absolute GeSe p,
p Depoos F.
Seu/lb.
Seu/m heu nb.
Cu ft.perIb Cu. fa pre 4 250.0 235.3 400.97 3f6.1
~815.0 1201.1 0.01865 1.84317 255.0 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80802 360.0 245.3 404.44 379.9 821.6 1201.5 0.01870 1.77418 365.0 250.3 406.13 381.7 820.0 1201.7 0.01873 I.74157 270.0 255.3 407.80 383.6 818.3 1201.9 0.01875 1.71013 275.0 260.3 409.4 385.4 816.7 1202.1 0.01878 1.67978 J
280.0 265.3 411.07 387.1 815.1 1202.3 0.01880 1.65049 285.0 270.3 412.67
" 388.9 813.6 1202.4 0.01882 1.62218 g
290.0 275.3 414.25 390.6 412.0 1202.6 0.01885
't.59482 "295.0 280.3 415.81 392.3 810.4 1202.7 0.01887 1.56835 300.0 285.3 417.35 394.0 808.9 1202.9 0.01889 1.54274 320.0 305.3 423.31 400.5 802.9 1203.4 0.01899 1.44801 j
340.0 325.3 428.99 406.8 797.0 1203.8 0.01908 1.36405 360.0 345.3 434.41 412.8 791.3 1204.1 0.01917 1.28910 380.0 365.3 439.61 418.6 785.8 1204.4 0.01925 1.22177 400.0 385.3 444.60 424.2 780.4 1204.6 0.01934 1.16095 420.0 405.3 449.40 429.6 775.2 1204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535
%0.0 445.3 458.50 439.8 765.0 1204.8 0.01959 1.00921 480.0
%5.3 462.82 444.7 760.0 1204.8 0.01 % 7 0.96677 500.0 485.3 M 7.01 449.5 755.1 1204.7 0.01975 0.92762 520.0 505.3 471.07 454.2 750.4 1204.5 0.01982 0.89137 540.0 525.3 475.01 458.7 745.7 1204.4 0.01990 0.85771 560.0 545.3 478.84
%3.1 741.0 1204.2 0.01998 0.82637 0.79712 540.0 565.3 482.57 467.5 736.5 1203.9 0.02006 600.0 585.3 486.20 471.7 732.0 1203.7 0.02013 0.76975 620.0 605.3 489.74 475.8 727.5 1203.4 0.02021 0.74408 640.0 625.3 493.19 479.9 723.1 1203.0 0.02028 0.71995 M0.0 645.3 4 %.57 483.9 718.8 1202.7 0.02036 0.69724 680.0
%5.3 499.86 487.8 714.5 1202.3 0.02043 0.67581 700.0 685.3 503.08 491.6 710.2 1203.8 0.02050 0.65556 720.0 705.3 506.23 495.4 706.0 1201.4 0.02058 0.63639 740.0 725.3 509.32 499.1 701.9 1200.9 0.02065 0.61822 700.0 745.3 512.34 502.7 697.7 1200.4 0.02072 0.60047 780.0 765.3 515.30 506.3 693.6 1199.9 0.02080 0.58457 800.0 785.3 518.21 509.8 (A9.6 1899.4 0.02087 0.568 %
820.0 805.3 521.06 513.3 685.5 1198 8 0.02044 0.55408 840.0 825.3 523.86 516.7 681.5 1198.2 0.02101 0.53988 860.0 845.3 526.60 520.1 677.6 1197.7 0.02109 0.52631 0.51333 880.0 865.3 529.30 523.4 673.6 1197.0 0.02116 900.0 885.3 531.95 526.7 M 9.7 11 %.4 0.02123 0.50091 920.0 905.3 534.56 530.0
%5.8 1195.7 0.02130 0.48901 940.0 925.3 537.13 533.2 M t.9 1195.1 0.02137 0.47759 960 0 945.3 539.65 536.3 658.0 1194.4 0.02145 0.46662 980.0 965.3 542.14 539.5 654.2 1193.7 0.02152 0.45604 1000.0 985.3 544.58 542.6 650.4 1192.9 0.02159 0.445 %
1050.0 1035.3 550.53 550.1 640.9 1191.0 0.02177 0.42224 1100.0 1085.3 556.28 557.5 631.5 1189.1 0.02195 0.40058 I
1150.0 1135.3 561.82 M4.8 622.2 1187.0 0.02214 0.38073 i
1200.0 1I85.3 567.19 571.9 613.0 1184.8 0.02232 0.36245 l
1250.0 1235.3 572,38 578.8 603.8 1182.6 0.02250 0.3455e 1300.0 1285.3 577.42 585.6 594.6 1180.2 0.02269 0.32991 l
1350.0 1335.3 582.32 592.2 585.6 1177.8 0.02288 0.31536 1400.0 1385.3 587.07 598.8 567.5 1175.3 0.02307 0.30178 1450.0 1435.3 591.70 605.3 567.6
__1172.9 0.02327 0.28904 l
1500.0 1485.3 5 %.20 611.7 558.4 1170.1 0.02346 0 27714 I
1600.0 1585.3 404.87 624.2 540.3 1164.5 0.02387 0.25545 j
1700.0 1685.3 613.13 636.5 522.2 1158.6 0.0208 0.23607 1800.0 1785.3 621.02 648.5 503.8 1152.3 0.02472 0.21861 1900.0 1885.3 628.56 M O.4 485.2 1145.6 0.02517 0.20278 l
2000.0 1985.3 635.80 672.1 4%.2 1138.3 0.02565 0.18831 2100.0 2085.3 642.76 653.8 446.7 1130.5 0.02615 0.17501 0.16272 2200.0 2185.3 649.45 695.5 426.7 1122.2 0.02669 2300.0 2285.3 655.89 707.2 406.0 1113.2 0.02727 0.15133 3400.0 2385.3 662.11 719.0 384.8 1103.7 0.02790 0.14076 1
l 2500.0 2485.3 M8.11 731.7 361.6 1093.3 0.02859 0.13006 I
2600.0 2585.3 673.91 744.5 337.6 1082.0 0.02938 0.12110 2700.0 3685.3 679.53 757.3 312.3 1069.7 0.03029 0.11194 2800.0 2785.3 654. %
770.7 285.1 1055.8 0.03134 0.10305 2900.0 2885.3 690.22 785.1 254.7 1039.8 0.032f 2 0.09420 l
3000.0 2985.3 695.33 801.8 218.4 1020.3 0.03428 0.08500 3100.0 3085.3 700.28 824.0 169.3 993.3 0.03681 0.07452 l
3200.0 3185.3 705.08 375.5 56.1 931.6 0.04472 0.05663 3208.2 3193.5 705.47 906.0 0.0 906.0 0.05078 0.05078 l
l l
- u. : -..... --..
.. A. n Aseawois a-mmicAt reoreenes ce nuios Amo now enAaAcreassnes or vAtves. nmwos. Amo met CRANI i'
Properties of Superheated 5 foam' V-spectne coium., cubic r e p.e pound h,= sotel hoot of steem, Bev p.r pound Pressure Sat.
Lbs. fn.
Temp.
Total Temperature-Degrees Fahrenheit (#)
sq.
Abs.
Case 33g.
400*
500*
600*
700*
808*
900*
100F 1100*
1300*
1500*
P' P
t 15.0 0.3 213.03 V 31.939 33. % 3 37.985 41.986 45.978 49.964 53.9 % 57.926 61.905 H.858 77.807 A,
1216.2 1239.9 1287.3 1335.2 1383.8 I433.2 1483.4 1534.5 15M.5 1693.2 1803.4 20.0 5.3 227.%
V 23.900 25.428 "24.457 31. # 6 34.#5 37.458 40.447 43.435 M.420 52.388 58.352 A,
1215.4 1239.2 12%.9 1334.9 1383.5 1432.9 I483.2 1534.3 15M.3 1693.1 1803.3 30.0 15.3 250.34 F
15.859 16.892 18.929 20.945 22.951 24.952 26.949 28.943 30.936 34.918 38.8 %
A, 1213.6 1137.8 12M.0 1334.2 1383.0 1432.5 1482.8 1534.0 1586.1 1692.9 1803.2 40.0 25.3 267.25 7
11.838 12.624 14.165 15.685 17.195 18.699 20.199 21.H7 23.194 26.183 29.168 A,
1211.7 1236.4 1245.0 1333.6 1382.5 1432.1 1442.5 1533.7 1585.8 1692.7 1803.0 50.0 35.3 241.02 P
9.424 10.062 11.306 12.529 13.741 14.947 16.150 17.350 18.549 20.942 23.332 A,
1209.9 1234.9 1284.1 1332.9 1382.0 1431.7 1482.2 1533.4 1545.6 1692.5 1802.9 60.0 45.3 292.71 P
7.815 8.354 9.400 10.425 11.438 12.4 #
13.450 14.452 15.452 17.448 19.441 A,
1208.0 1233.5 1283.2 1332.3 1381.5 1431.3 1481.8 1533.2 1585.3 IH2.4 1802.P 70.0 55.3 302.93 P
6.664 7.133 8.039 8.922 9.793 10.659 11.522 12.382 13.240 14.952 l'.o61 A,
1206.0 1232.0 1282.2 1331.6 1381.0 1430.9 1481.5 1532.9 1585.1 1692.2 a 802.6 80.0 65.3 312.04 P
5.801 6.218 7.018 7.794 8.560 9.319 10.075 10.829 11.581 13.081 14.577 A,
1204.0 1230.5 1281.3 1330.9 1380.5 1430.5 1481.1 1532.6 1584.9 1692.0 1802.5 90.0 75.3 320.28 E
5.128 5.505 6.223 6.917 7.600 8.277 8.950 9.621 10.290 11.625 12.956 A,
1202.0 1228.9 1280.3 1330.2 1380.0 1430.1 1480.8 1532.3 1584.6 1691.8 1802.4 100.0 85.3 327.82 C
4.590 4.935 5.588 6.216 6.833 7.443 8.050 8.655 9.258 10.460 11.659 A,
1199.9 1227.4 1279.3 1329.6 1379.5 1429.7 1440.4 1532.0 1584.4 1691.6 1802.2 120.0 105.3 341.27 P
3.7815 4.0786 4.6341 5.1637 5.6813 6.1928 6.7006 7.2060 7.7096 8.7130 9.7130 A,
1195.6 1214.1 1277.4 1328.2 1378.4 1428.8 1479.8 1531.4 1583.9 1691.3 1802.0 140.0 125.3 353.04 7
3.4661 3.9526 4.4119 4.8588 5.2995 5.7364 6.1709 6.6036 7.#52 8.3233 A,
1220.8 1275.3 1326.8 1377.I I428.0 1479.1 1530.8 1583.4 1690.9 1801.7 160.0 145.3 363.55 9
3.0060 3.4413 3.8440 4.2420 4.6295 5.0132 5.3945 5.7741 6.5293 7.2811 A,
1217.4 1273.3 1325.4 1376.4 1427.2 1478.4 1530.3 1582.9 1690 5 1801.4 180.0 165.3 373.08 9
2.6474 3.0433 3.4093 3.7621 4.1084 4.4500 4.7907 5.1289 5.8014 f.4704 A,
1213.8 1271.2 1324.0 1375.3 1426.3 1477.7 1529.7 1582.4 1690.2 1801.2 200.0 185.3 381.80 V
2.3598 2.7247 3.0583 3.3783 3.6915 4.0008 4.3077 4.6128 5.2191 5.8219 A,
1210.1 12H.0 1322.6 1374.3 1425.5 1477.0 1529.1 1581.9 1689 3 1800.9 220.0 205.3 389.04 V
2.1240 2.M38 2.7710 3.0642 3.3504 3.6327 3.9125 4.1905 4.7426 5.2913 A,
1206.3 12M.9 1321.2 1373.2 1424.7 1476.3 1528.5 1581.4 1649.4 1800.6 240.0 225.3 397.3%
7 1.9268 2.2#2 2.5316 2.8024 3.0Mt 3.3259 3.5831 3.8385 4.3456 4.8492 A,
1102.4 1264.6 1319.7 1372.1 1423.8 1475.6 1527.9 1580.9 1689.1 1800.4 260.0 245.3 404.44 V
2.0619 2.3289 2.5008 2.8256 3.0M3 3.3044 3.5408 4.0097 4.4750 i
A, 1262.4 1318.2 1371.1 1423.0 1474.9 1527.3 1580.4 1688.7 1800.1 280.0 265.3 411.07 P
1.9037 2.1551 2.3909 2.6194 2.8437 3.0655 3.2855 3.7217 4.1543 1
1 A,
1260.0 1316.8 1370.0 1422.1 1474.2 1526.8 1579.9 1684.4 1799.8 300.0 285.3 417.35 V
1.7M5 2.0044 2.2263 2.4407 2.6509 2.8585 3.0643 3.4721 3.8764
\\
A, 1257.7 1315.2 1368.9 1421.3 1473.6 1526.2 1579.4 1688.0 1799.6 s
320.0 305.3 423.31 9
1.6462 1.8725 2.0823 2.2843 2.4821 2.6774 2.8708 3.2538 3.6332 A,
1255.2 1313.7 1367.8 1420.5 1472.9 1525.6 1578.9 1687.6 1799.3
{
340.0' 325.3 428.99 9
1.5399 1.7561 1.9551 2.1M3 2.3333 2.5175 2.7000 3.0611 3.4186 A,
1152.8 1312.2 13M.7 1419.6 1472.2 1525.0 1578.4 1687.3 1799.0 360.0 345.3 434.41 F
1.4454 1.6525 I 8421 2.0237 2.2009 2.3755 2.5482 2.8898 3.2279 A,
1250.3 1310.6 1365.6 1418.7 1471.5 1542.4 1577.9 1686.9 1798.8
- \\bstracted from ASME Secam Tables (1967) with permissen of the publisher. the f=*==d aa American Society of Mecherucal En8meers. 345 East 47th Street. New York. N.Y.10017.
- a**' P*e*)
.. z. m=.:
-~
4
', e CEANE aseewom a-owseca esoseenes or nues Ase now cuanacreensncs ce vuves. mnwos. amo me A.17 Propertie5 of Superheated Steam continued V-sp. cme e we f e p., p w A,-ioe.i h e et se., seu p., p w t
Pressure Set.
Lbe.fn.
Temp.
Total Temperature-Dagrees Fahrenheit (e) sq.
Abs. W 800*
600*
700*
800*
980*
1000*
1100*
1200' 1300* ! 14 & }' 1500' P'
P r
-l 380.0 365.3 439.61 F
1.3606 1.5598 IJ410 1.9139 2.0025 2.2484 2.4124' 2.5750' 2.73 % 2.8973 3.0572 A,
12477 1309.0 1364.5 1417.9 1470.8 1823.8 1877.4 1631.6 1686.5 1742.2 1796.5
?
400.0 385.3 444.60 F
1.2841 1.4763 1.6499 1.018I 1.9759 2.1339 2.2901 2.4450 2.5987 2.7515 2.9037 A,
1245.1 1307.4 1363.4 1417.0 1470.1 1823.3 1876.9 1631.2' 16 #.2, 1741.9 1798.2 420.0 405.3 449.40 F
I.2148 1.4007 1.M76 1.7258 1.8795 2.0304 2.1795 2.3273 2.4739 2.61% 2.7647 A,
1242.4 1305.8 IM2.3 1436.2 l#9.4 1522.7 1876.4 1630.8 1685.8 1741.6 17 %.0 440.0 425.3 454.03 P
1.1517 1.3319 1.4926 I.6445 1.7918 1.9363 2.0790 2.2203 2.3605 ' 2.4998 2.6384 A,
1239.7 1304.1 1361.1 1415.3 l#8.7 1522.1 1875.9 1630.4 1685.5 1741.2 1797.7 460.0 445.3 458.50 F
1.0939 1.M91 1.4242 1.5703 1.7117 1.8504 1.9872 2.1226; 2.2569 2.3903 2.5230 A,
1236.9 1302.5 1360.0 1414.4 1#8.0 1521.5 1575.4 1629.9 1685.1 1740.9 1797.4 480.0
- 5.3
- 2.82 F
1.0409 1.2115 1.3615 1.5023 1.6384 1.7716 1.9030 2.0330 2.1619 ' 2.2900 2.4173 A,
1234.1 1300.8 1358.8 1413.6 1467.3 1520.9 1874.9 1629.5 1684.7, 1740.6 1797.2 500.0 485.3
- 7.01 F
0.9919 1.1584 1.3037 1.4397 I.5708 1.6992 1.8256 1.9507 2.07 # 2.1977 2.3200 A,
1231.2 1299.1 1357.7 1412.7 14 %.6 1820.3 1574.4 1629.1 1684.4 1740.3 17 %.9 520.0 505.3 471.07 F
0.94 %
1.1094 1.2804 1.3819 1.5085 1.6323 1.7542 1.87# 1.9940 2.1125 2.2302 A,
1228.3 1297.4 1356.5 1411.8 165.9 1519.7 1573.9 1628.7 1684.0. 1740.0 17 %.7 l
540.0 525.3 475.01 F
0.9045 1.0640 1.2010 1.3284 1.4508 1.8704 1.6880 1.8042 1.9193 2.0336 2.1471 A,
1225.3 1295.7 1355.3 1410.9 l#5.1 1519.1 1573.4 1628.2 1683.6 1739.7, 17 %.4 560.0 545.3 478.84 F
0.8653 1.0217 1.1552 1.2787 1.3972 1.5129 1.62 %
1.7388 1.8500 : 1.9603 2.0699 A,,
1222.2 1293.9 1354.2 1410.0 1#4.4 1818.6 1572.9 1627.8 1%3.3 1739.4 17 %.I 500.0 M5.3 482.57 7
0.8287 0.9824 1.1125 1.2324 1.3473 1.4593 1.5693 1.6780 1.7855 1.8921 1.9980 1
A, 1219.1 1292.1 1353.0 1409.2 163.7 1518.0 1572.4 1627.4 1682.9 1739.1 1795.9 600.0 585.3 486.20 F
0.7944 0.9456 1.0726 I.1892 1.3008 1.4093 1.5160, 1.6211 1.7252 l 1.8284 1.9309 A,
1215.9 1290.3 1351.8 1408.3 ; 1 # 3.0 1517.4 1571.9 1627.0 1682.6, 1738.8 1795.6 650.0 635.3 494.89 F
0.7173 0.8634 0.9835 1.0929 1.1%9 1.2979 1.3 % 9 I.4944 1.5909 1.6864 1.7813 A,
1207.6 1285.7 1348.7 1406.0 l#1.2 1515.9 1570.7 ' 1625.9 1681.6 1738.0 1794.9 700.0 M5.3 503.08 7
0.7928 0.9072 I.0102 1.1078 1.2023 1.2948 1.3858 1.4757 ' I.5647 1.6530 l
A, 1281.0 1345.6 1403.7 1459.4 1514.4 1569.4 1624.8 1680.7 ' 1737.2 1794.3 750.0 735.3 510.84 V
0.7313 0.8409 0.9386 1.0306 1.1195 1.2063 1.2916 1.3759 1.4592 1.5419 A,
1276.1 1342.5 1401.5 1457.6 1512.9 1568.2 1623.8 1679.8 1736.4 1793.6 000.0 785.3 518.21 F
0.6774 0.7828 0.8759 0.9631 1.0470 1.1289 1.2093 1.2885 1.3%9 1.4446 A,
1271.1 1339.3 1394.1 1455.8 1511.4,15%.9 1622.7 1678.9 1735.7 1792.9 850.0 835.3 525.24 V
0.62 %
0.7315 0.8205 0.9034 0.9830 1.0606 1.1366 1.2115 ~ 1.2855 1.3568 A,
1265.9 1336.0 1396.8 1454.0 1510.0 1565.7 1621.6; 1678.0 1734.9 1792.3 980.0 885.3 53I.95 F
0.5869 0.6858 0.7713 0.8504 0.9262 0.9998 1.0720!
(
A, 1260.6 1332.7 1394.4 1452.2 1508.5 1564.4 1620.6 1677.1
- 1734.1 ;1791.6 I.1430 1.2131 1.2825 1
980.0 935.3 538.39 F
0.5485 0.6449 0.7272 0.0030 0.8753 0.9455 1.0142 1.0817 ! 1.1484 'I.2143 l
A, 1155.1 1329.3 1392.0 1450.3 1507.0 1563.2 1619.5 1676.2 1733.3 1791.0 1880.0 985.3 544.50 F
0.5137 0.6000 0.6875 0.7603' O.8295 0.8966 0.9622 1,02 %
1.0901.1.1529 A,
1249.3 1325.9 1389.6 1448.5 1505.4 1561.9 1618.4 1675.3 1732.5 1790.3
]
1950.0 1935.3 550.53 P
4.4821 0.5745 0.6515 0.7216 0.7881 0.8524 0.9151 0.9767 1.0373 1.0973 A,
1243.4 1322.4 1387.2 14 #.6 1503.9 1860.7 1617.4 1674.4 1731.8 1789.6 1100.0 1885.3 556.28 F
0.453I 0.5440 0.6188 0.6865 0.7505 0.8121 0.8723 0.9313 0.9894 1.0#8 i
A, 1237.3 1318.8 1384.7 1444.7 1502.4 1859.4 1616.3 1673.5 1731.0 1789.0
)
1950.0 1135.3 561.82 V
0.4263 0.5162 0.5889 0.6544 0.7161 0.7754 0.8332 0.8899 0.9456 1.0007 A,
1130.9 1315.2 1382.2 1447*
1500.9 1858.1 I615.2 1672.6, 1730.2 1788.3
c A '18 A* mein A-nnsscat poorumes or muses Ase now enasacteessnes ce vatwes nmmes A*e mee CRANE Properties of Superheated stoom concluded V-sp.c.ac su, cowc f e p., pound h,=lotol h.ot of st.om, teu p.c pound Pressure Set Lbs. per Temp.
sq.In.
Total Temperature-Degrees Fahrenheit O)
Abs.
Gage 650' 700 750 000*
900*
1000 1100 1200 1300 I400 1500*
P' P
t r
1200.0 1195.3 M7.19 F 0.4497 0.4905 0.5273 0.5615 0.6250 0.6845 0.7418 0.7974 0.851 A,
1271.8 1311.5 13%.9 1379.7 1440.9 1499.4 15%.9 1614.2 1671.6 17 1300.0 1285.3 577.42 F
0.4052 0.4451 0.4804 0.5129 0.5729 0.6287 0.6822 0.7341 0.7847 0.8 A,
1261.9 1303.9 1340.8 1374.6 1437.1 1496.3 1554.3 1612.0 1%9.8 1727.9 17 %.3 1400.0 1385.3 587.07 F
0.3M7 0.4059 0.4400 0.4712 0.5242 0.5009 0.6311 0.6798 0.7272 0.7737 0.8195 A,
1251.4 12%.I 1334.5 1369.3 1433.2 1493.2 1551.8 1609.9 1%8.0 1726.3 1785.0 1500.0 1485.3 5 %.20 F
0.3328 0.3717 0.4049 0.4350 0.4894 0.5394 0.5869 0.6327 0.6773 0.7210 0.7639 A,
1240.2 1187 9 1328.0 1364.0 1429.2 1490.1 1549.2 1607.7 1%6.2 1724.8 1783.7 1600.0 1585.3 604.87 I
0.3026 0.3415 0.3741 0.4032 0.4555 0.5031 0.5482 0.5915 0.63 % 0.6748 0.7153 A,
!!28.3 1279.4 1321.4 1358.5 1425.2 1486.9 15%.6 1605.6 1%4.3 1723.2 1782.3 1700.0 1685.3 613.13 C
0.2754 0.3147 0.3468 0.3751 0.4255 0.4711 0.5140 0.5552 0.5951 0.6341 0.6724 A,
1215.3 1270.5 1314.5 1352.9 1421.2 1483.8 1544.0 1603.4 1 % 2.5 1721.7 1781.0 1800.0 1785.3 621.02 F
0.2505 0.2906 0.3223 0.3500 0.3988 0.4426 0.48%
0.5229 0.5609 0.5980 0.6343 A,
1201.2 1261.1 1307.4 1347.2 1417.1 1480.6 1541.4 1601.2 1%0.7 1720.1 1779.7 1900.0 1885.3 628.%
T 0.2274 0.2687 0.3004 0.3275 0.3749 0.4171 0.4565 0.4940 0.5303 0.5656 0.6002 A,
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dc5 key SECTION 5 THEORY OF NUCLEAR POWER PLANT DPERATION THERMODYNAMICS, HEAT TRANSFER, AND FLUID FLOW 5.01 QUESTION (4.0)
(a)
For Diablo Canyon Unit 1 on attached Figure 5.1, draw the curves for T(hot),
T(cold),
T(Ave.), and T(steam) as a function of power level.
State the zero power and 100% power temperature for each curve.
(0.75 each curve, 3.0 total)
(b) At Diablo Canyon what is the zero power steam pressure in the steam generators ?
(0.5)
(c)
If the steam generator steam temperature was 570 degrees F,
what would be the steam pressure? (0.5) 5.01 ANSWER (4.0)
(a) See attached figure 5.1 (3.0)
(b) 547 F.
saturated steam is at 1020 PSIA (0.5)
(c) 570 F.
saturated steam is at 1230 PSIA (0.5)
REFERENCE Operator Information Manual, page T-1-1 Thermodynamics (0217C), pages 4-62 & 4-63 Steam Tables (f or pressure of saturated steam 1
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5.02 QUESTIDN (2.0) l A diesel generator is supplying power to various equipment loads. l The diesel generator output is 1200 KW at 4200 volts, 500 Volt-I Amps reactive load, 310 amps, and 60.2 hertz. l (a) If a 200 KW motor is added to the original load how will the reactive load change ? (increase, decrease, stay the same) (O.5) (b) If a 200 KW heater load is added to the original load how will the reactive load change ? (increase,
- decrease, stay the sarae) (0.5)
(c) The diesel generator, at its' original voltage and frequency I but in the droop mode, is paralleled with an offsite power supply l at 4100 volts and 60 hertz. How will the reactive load carried by the diesel generator change ? (increase,
- decrease, stay the i
same) (0. 5) (d) With the same conditions as (c) above. How will the output KW change ? (increase, decrease, stay the same) (O.5) I 5.02 ANSWER (2.0) (a) The reactive load will increase. (0.5) (b) The reactive load will remain the same. (0.5) (c) The reactive load carried by the di esel generator will increase (or become more lagging). (0.5) (d) The diesel generator output KW will increase. (0.5) l l REFERENCE Standard Handbook for Electrical Engineers, Fink & Carrol pages 1-17 & 1-18 Operation procedure OP-J-6B:II Lesson
- Plans, J4a, Main Generator and 25 KW;
- J6b, Diesel Generators 2
5.03 QUESTION (2.0) \\ As plant temperature decreases the RCS coolant density will increase. At BOL with RCS temperature decreasing provide two reasons why the nuclear instrumentation count rate will decrease. ( Assume no rod motion or RCS boron concentration changes. ) i 5.03 ANSWER (2.0) 0 (1.0 each, for any two of the following) (a) Neutrons travel less distance, less leakage. (b) Higher boron density, more absorption by baron. (c) K(eff) could decrease due to positive moderator temperature coefficient. f i REFERENCE Westinghouse, Pressurized Water Reactor Systems Manual 3
5.04 QUESTIDN (2.0) A nonregenerative parallel-flow heat exchanger has a flow rate of 30,000 pounds per hour of primary coolant and 200,000 pounds per hour of component cooling water. The primary inlet temperature is at 290 F, the primary outlet is at 115 F, and the component cooling water (CCW) inlet is at 64 F. (a) What is the cooling water outlet temperature ? (1.0) (b) During an outage the heat exchanger piping is modified to make it a counter-flow heat exchanger. The primary coolant flow, primary inlet temperature, and CCW inlet temperature remain the same. In order to maintain the same primary outlet temperature will more or less CCW flow be required ? (0.5) (c) What is an advantage of the counter-flow heat exchanger over the parallel flow heat exchanger. (0.5) 5.04 ANSWER (2.0) = mass flow rate x specific heat x change in (a) (i) O temperature (0.25, state correct formula) (ii) 0, primary side = 0, component cooling water side (0.25, set heat lost on one side to heat gained on other) (iii) 30,000 x (290 - 115) = 200,000 x ( CCW outlet Temp. 64) (0.25, insert temperatures and flows into formula corectly) (iv) 5,250,000 = 200,000 x (CCW outlet Temp. -64) 26.25 + 64 = 90.25 F. CCW outlet Temp. (0.25, work out math and state answer correctly) (b) less CCW flow will be required (0.5) (The counter flow heat exchanger is more efficient) (c) Any one of the following for full credit. (0.5) (i) The more uniform temperature difference between the two fluids minimizes the thermal stresses thraucught the heat exchanger. (ii) The more uniform temperature difference results in a more uniform rate of heat transfer throughout the heat exchanger. (iii) The outlet temperature of the cold fluid approaches the inlet ' temperature of the hot fluid. REFERENCE Thermodynamics Text, Chapter 5 4
5.05 QUESTION (2.5) (a) On the attached figure 5.2 draw the profile of reactor axial power distribution at BOL hot full power. (0.5) (b) Dn the attached figure 5.3 draw the profile of reactor axial power distribution at EDL hot full power. (0.5) (c) Explain the difference between the BOL and EDL hot full power distributions. (1.0) (d) At equilibrum conditions with all control rods out how is the xenon concentration related to power distribution ? (0.5) 5.05 ANSWER (2.5) (a) Must show smooth convex curve with power peaking in the lower half of the core (ie. at 25 to 45 % of core height). (0.5) (b) Must show smooth curve which is flat on top or has two humps, one in bottom half of core and one in top half of core. (0.5) (c) As the core depletes (0.25), the higher production in the center of the core (0.25) results in non-uniform axial burn-up (0.25) which in turn flattens the core power distribution. (0.25) (1.0, total) (d) Xenon concentration is proportional to power. (0.5) i REFERENCE WCAP-8408, Reactor Theory Suppliment, pages 3.3 and 3.22 Thermodynamics, 0349C, page 13-47 Technical Specifications 3/4.2.1 5
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5.06 QUESTION (2.0) (a) Why is the control of dissolved oxygen in the reactor coolant system important 7 (0.5) (b) What are two sources of oxygen in the reactor coolant system ? (0.5) (c) Explain what is done to remove oxygen from the reactor coolant system at normal operating temperatures 7 (0.5) (d) Explain what is done to remove oxygen from the reactor coolant system during cold shutdown 7 (0.5) 5.06 ANSWER (2.0) (a) To minimize corrosion of the RCS. (0.5) (b) 0.25 for any two of the following (i) Oxygen is dissolved in the RCS during refueling. (ii) Oxygen is dissolved in the primary make-up water. (iii) Oxygen is generated by the radiolytic decomposition of water. (c) Hydrogen is added to the primary coolant system (through the VCT and make-up pumps -- CVCS) to combine with any oxygen present in the RCS. (0.5) (d) Hydrazine is added to the primary coolant system to remove oxygen during cold shutdown. (0.5) REFERENCE Technical Specification, 3/4.4.7 6
5.07 QUESTION (1.0) Refer to Figure 5.4. Early in core life, why does critical boron concentration drop quickly from 1200 to 850 ppm 7 (1.0) 5.07 ANSWER (1.0) Critical boron concentration drops quickley early in core life due to the build up of (1) xenon and samarium or (2) fission product poisons. (1.0, either i or 2 for full credit) REFERENCE WCAP-8408, Reactor Theor y Suppliment, page 3.9 Operator Information Manual, page R-7-1 and R-5-1 7
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5.00 QUESTION (2.25) Positive reactivity is inserted into an already critical reactor core at time t = 0, followed by an equal negative reactivity inserti on at time t = 10 minutes. During that time power increases from 10-8 Amps to 10-4 Amps. What percent millirem (PCM) was the reactivity inserition 7 5.08 ANSWER (2.25) (t) (i) P = Po x 10 x SUR Correct formula (0.25) (t) (ii) 10-4 = 10-8 x 10 x SUR Insert correct numbers (0.25) (iii) SUR = 4/10 DPM Solve equation (0.25) (iv) SUR = 26/T Correct equation (0.25) (v) T = 65 Sec Period (0.25) (vi) (a) T= (b p) / (ap) or Correct formula (0.25) (b) p = b/(1 + aT) (vii) (a) 65 = (0.0070 - p)/(0.1 x p) b= .0068 to.0072 (b) p = 0.0070/(1 + 0.1 x 65) Insert numbers correctly (0.25) (viii) (a) 6.5 x p = 0.007 - p: 7.5 x p = 0.007: p = 0.0009 (b) p = 0.007/(1 + 6.5): p = 0.0009 Find reactivity (0.25) (ix) 0.0009 delta K = 90 PCM (80 to 100 for full credit) Correct conversion to PCM (0.25) REFERENCE Operator Information Manual, page R-1-1 and R-3-1 8
5.09 QUESTION (1.O) Why are burnable poison rods designed into the reactor core rather than simply increasing the boric acid concentration in the p coolant 7 [ i 5.09 ANSWER (1.0) Burnable poison rods are used to reduce the soluble boron B concentration to ensure that the moderator temperature coefficient is negative for power operations. (1.0) t e REFERENCE WCAP-8408, Reactor Theory Suppliment, page 3.1 Lesson Plan A-2b, page 10 9
5.10 QUESTIDN (2.25) (a) While the reactor system is operating normally at 2250 PSIA one of the pressurizer relief valves lift. Assuming that the relief valve relieves to atmospheric pressure (14.7 PSIA), what is the discharge temperature and state of the steam (subcooled, mixture of water and steam, saturated
- steam, or superheated steam) 7 (1.25)
(b) If the reactor pressure were 1200 PSIG and the plant in hot standby when this event
- occured, what is the discharge temperature and state of the steam 7 (1.0) 5.10 ANSWER (2.25)
(a) (1.25) Full credit for: 212 degees F and mixture of steam and water, or partial credit as follows. (i) The enthalpy of the steam in the pressurizer at 653 F. is about 1116 BTU /lbm. (0.2) (ii) When steam expands to atmospheric pressure enthalpy remains constant. (0.2) (iii) The enthalpy of saturated steam at 15 PSIA is 1150 BTU /lbm. (0.2) (iv) This indicates that some steam must condense to water at a lower enthalpy than the steam. (0.2) (v) The relief valve discharge is a mixture of water and steam. (0.2) (vi) 212 F. (0.25) (b) (1.0) Full credit for: 280 degrees F and superheated, or partial credit as follows. (i) The enthalpy of saturated steam at 1200 PSIG is 1183 BTU /lbm. (0.2) (ii) When this steam expands to 1t.7 PSIA the enthalphy will remain the same. (0.2) (iii) The enthalpy of steam at atmospheric presseure is 1150 BTU /Lbm. (0.2) (iv) The steam will be superheated. (0.2) (v) 270 degrees F. (0.2) REFERENCE Standard Thermodynamics 10
5.11 QUESTIDN (1.0) At 100% power the Unit 1 rod insertion limit is 180 steps. The Unit 2 rod insertion limit at 100% power is about 189 steps. Explain why the RILs are different. (1.0) 5.11 ANSWER (1.0) Unit 2 operates at greater power and therefore a greater power defect. More rod worth must be inserted to meet shutdown margin esquirements. (1.0) M REFdRENCE Unit Differences, OPOO59, Technical Specifications Item 6 11 y+m - ~ - - - 9 --m.- - - ry p, m-w m r
e
5.12 QUESTION (1.5) Refer to Figure 5.5. Both the mechanical and centrifugal pumps are operating and valves A, B, and C are fully open. If valve C is inadvertently shut how will the following parameters be affected 7 (a) Pressure in the pipe (P2) between the centrifugal charging pump and check valve "B" (0.5) (a) Centrifugal pump motor amps (0.5) (b) Mechanical pump motor amps (0.5) 5.12 ANSWER (1.5) (a) The pressure will increase to the shut-off head of the centrifugal pump P (s). (0. 5) (b) Motor amps will decrease. (0.5) (Since the relief valve provides the only flow path and its' set point is above the shut-off head of the centrifugal pump, there will be no flow through the centrifugal pump. Since power is proportional to flow times pressure the only power used will be frictional losses in the motor and pump. Likewise power is the product of voltage and
- current, since motor voltage is essentially
- constant, the current drawn must decrease.)
(c) Motor amps will increase. (0.5) (Since the relief valve is the only flow path the pressure will rise to the relief valve set point. The mechanical charging pump is a positive displacement device with nearly constant flow. Flow remains the same while pressure rises indicates that power used must increase, thus motor amps must increase.) REFERENCE Operator Information Manual, page T-5-1 throug T-5-5 12
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5.13 QUESTION (1.0) Unit 1 control bank A is comprised of 8 CRDMs while Unit 2 control bank A is comprised of 4 CRDMs. Unit 1 control bank B is comprised of 4 CRDMs while Unit 2 control bank B has B CRDMs. (a) Why are these control banks different for the two plants ? (0.5) (b) Why dosen't this make any difference to normal plant operations 7 (0.5) 5.13 ANSWER (1.0) (a) Unit 2 is designed f or the plutonium recycle core and Unit 1 is not. (0.5) (b) During normal operations both of these banks are fully withdrawn. (0.5) REFERENCE Unit Differencen, Plant Design, Item 4 THIS IS THE END DF SECTION 5 13
dc6 key SECTION 6 PLANT SYSTEMS: DESIGN, CONTROL, AND INSTRUMENTATIDN 6.01 QUESTION (3.0) (a) List four of the five Safety Injection actuations. (Logic and setpoints not required) (1.0) (b) For a DBA LOCA, list the ECCS components that will inject water into the reactor coolant system. Include the operating pressures of each component. (2.0) 6.01 ANSWER (3.0) (a) 0.25 each for any four of the following: (i) Low pressurizer pressure (ii) High containment pressure (iii) High steamline flow with low-low T(Avg) or low steamline pressure (iv) High steamline differential pressure (v) Manual (b) 0.5 for each of the following: (+ or - 10 % for full credit) (i) Centrifugal charging pump (s) 2500 PSIG (ii) Safety injection pump (s) 1520 PSIG (iii) Accumulator (s) 650 PSIG (iv) RHR Pump (s) 175 PSIG REFERENCE FSAR Lesson Plan, B-6a, Reactor Protection Operator Information Manual, page 1-4 J l 1 g .,-m, a g y
6.02 QUESTION (2.0) Regarding the differences between the two reactors (a) Unit 2 does not have any incore thermocouples in the upper head area. What operational consideration does this have f or the operator during natural circulatior 7 (1.0) (b) With both units at 100 % powir which unit will have the highest pressurizer level 7 (1.0) 6.02 ANSWER (2.0) (a) Soak times on natural circulation will have to be strictly adhered to. (1.0) (b) The Unit 2 pressurizer level is higher. (1.0) REFERENCE Unit Differences Book,
- OpO59, Plant Design page 3 and Technical Specifications pages 2, 3,
and 8. 2
6.03 QUESTION (4.0) During normal full power operation one channel T(hot) fails high. The primary and secondary control systems are in automatic and/or normal line up. (a) Name f our immediate alarms you would expect 7 (1.0) (b) How are the rod control and steam dump systems affected 7 (1.0) (c) What would be an operator's immediate action (s) 7 (2.0) 6.03 ANSWER (4.0) (a) 0.25 each for any four of the following (i) T(ave) deviation from auctioneered high T(ave) (ii) Auctioneered high T(ave) deviation f rom T(ref) (iii) Delta-T deviation (vi) High T(ave) alarm (v) Overtemperature delta-T channel activated (vi) Overpower delta-T channel activated (vii) Protection channel activated (b) (i) The rod control system will drive rods in. (to counteract the apparent high T(ave)) (0.5) (ii) No effect. (In T(ave) mode (above 15% power) the steam dump system will have a demand signal but since it is not armed no bypass or dump will occur.) (0.5) (c) (i) If necessary, take manual control of affected systems. (1.0) (ii) Return the affected parameters to normal. (1.0) REFERENCE Emergency Procedure, OP AP-5, Malfunction of Protection or Control Channel Lesson Plans, A-2c 8< A-3a 3
6.04 QUESTIDN (2.0) You are conducting a recovery from a reactor trip during which all reactor coolant pumps tripped. Conditions have been restored to the point where reactor coolant pumps can be started. (a) Which pump will you start first 7 (1.0) (b) Explain your reasoning 7 (1.0) 6.04 ANSWER (2.0) (a) RCP # 2 (1.0) or RCP # 1 (0. 5, half credit) If applicant explains that RCP # 1 can be started resulting in
- reduced, but adequate, spray flow available; then full credit should be given. (1.0)
(b) This will restore normal spray flow. (Other RCPs will not provide adequate spray flow.) (1.0) REFERENCE Lesson Plan A-4a, page 13 EP OP-1.1, page 12 of 20, Rev. O 4
6.05 QUESTIDN (2.0) The Unit 1 or 2 4KV "F" Bus can be supplied by diesel generator 1-3. (a) With diesel generator 1-3 in service on Unit 2's Bus "F" for a load test, what would happen if a SI signal came in on Unit 1 ? (1.0) (b) If diesel generator 1-3 was responding to a SI signal from Unit 1 and a SI signal came in from Unit 2, what would happen (before and after the Unit 1 SI was reset) 7 (1.0) 6.05 ANSWER (2.0) (a) The SI signal would trip the feeder breaker to Unit 2 (0.5), and cause the diesel generator to be available for Unit 1 (0.5). (b) The diesel generator would remain committed to Unit 1 (first SI) (0.5), and transfer only when the SI on Unit 1 was reset. (0.5). REFERENCE Lesson Plan J-6b, Di esel Generators 5
6.06 QUESTIDN (4.0) For the Residual Heat Removal System: (a) Describe three interlocks or safety devices which prevent overpressurization of the RHR system when placing it in service or conducting a plant cooldown 7 (1.5) (b) What three RCS conditions should be verified prior to placing the RHR system in the plant cooldown line-up 7 (1.5) (c) What measure is taken to prevent the inadvertent isolation of the RHR suction from the RCS 7 (0.5) (d) What are the two purposes of the mini-flow bypass lines 7 (0.5) 6.06 ANSWER (4.0) (a) 0.5 each for any three of the following (i) The RHR suction from loop four isolation valve (8701) can only be opened if RCS pressure is below 390 PSIG and pressuri=er vapor space temperature is below 475 degrees F. (ii) The other RHR suction from loop 4 isolation valve (8702) can only be opened if RCS pressure is below 390 PSIG and will automatically close if RCS pressure exceeds 700 PSIG. (iii) Relief valves are located on the RHR pump suction, the cold leg discharge header, and the hot leg discharge header. (iv) Two PORVs are reset to less than 450 PSIG. (b) (i) RCS temperature less than 350 degrees F. (0.5) (ii) RCS pressure less than 390 PSIG (0.5) (iii) Boron concentration in the RHR system equal to or greater than that in the RCS. (0.5) (c) The breakers for the suction valves (MOVs 8701 and 8702) are maintained in the open position (except when it is necessary to stroke the valves). (0.5) Full credit if the applicant explains that this is no longer done because of a new procedure.(0.5) (d) The mini-flow bypass lines provide protection from pump over heating (0.25) and loss of suction (0.25). REFERENCES Lesson Plan, D-2, RHR System Operating Precedure OP B-2:V, RHR Place in Service During Plant Cooldown 6
6.07 QUESTION (3.5) For the Auxiliary Salt Water Systems (a) Why are vacuum breakers necessary in the piping between the ASW pumps and CCW nsat exchangers ? (0.5) (b) What automatic sigtals will start an ASW pump 7 (2.0) (note Control room standby selector switch in " Auto", Hot Shut Down Pannel switch in " Control Room", and control room control switch in " neutral") (c) From which two locations can the ASW pumps be manually started 7 (1.0) 6.07 ANSWER (3.5) (a) Vacuum breakers prevent water hammer. (0.5) (b) (i) Low (l ess than 40 PSIG) discharge pressure, or low voltage on opposite bus (no SIS or transfer to diesel generator). (0.5) (ii) On an automatic bus transfer to the start-up transformer (no SIS present, after a 10 second time delay) (0.5) (iii) On a transfer to the diesel generators (no SIS present, running pumps trip and then restart after 10 seconds) ( 0. 5 )- (iv) On a SIS pumps will receive a start signal (22 seconds after the bus is energized) (0.5) (c) (0.5 each, for any two of the f ollowing) (1) The ASW pumps can be manually started from the control room (ii) from the hot shutdown pannel (iii) or from the 4 KV switchgear. REFERENCE Lesson Plan, E-5, Auxiliary Saltwater System 7 l
6.OB QUESTION (1.O) Nasie f our of the five trips which lock out the diesel generator. 6.08 ANSWER (1.0) 0.25 each for any three of the following (i) Generator differential relay trips (ii) Low lube oil pressure while running (iii) Diesel engine overspeed occurs (iv) The emergency shutdown device is actuated (v) In local control (on manual), high Jacket water temperature e REFERENCE Lesson Plan, J-66, Diesel Generator and Auxiliary System 8 . -. ~, _ _ _. - _ _ - - -. - _
6.09 QUESTIDN (3.5) a For the Nuclear Instrumentation Systems (a) List all reactor trips and set points associated with the power range Nuclear Instrumentation System. (1.5) i (b) List two control functions (not trips) of the Rod Control System which depend upon signals supplied by the power range Nuclear Instrumentation System. (1.0) (c) List all other reactor trips and rod control system control functions associated with the Nuclear InstrL >ntation system. Include setpoints and the nuclear instrument used. (1.0) 6.09 ANSWER (3.5) (a) 0.25 for each of the following (i) High flux, low setpoint 25% (ii) High flux, high setpoint 109% (iii) Positive high SUR +5%in 2sec (iv) Negative high SUR -5%in 2sec (v) OT delta T trip variable (vi) OP delta T trip variable (b) (i) Rod Control System (f or power mismatch signal) (0.25) (ii) Rod withdrawl stop (1/4 103%) (0.25) (c) 0.33 for each of the following (i) Intermediate range high power trip 25% (ii) Source range high power trip 10-5 CPS (ii) Intermediate range rod stop 20% REFERENCE ~ Lesson Plan, B-4, Excore Nuclear Instrumentation System Westinghouse, Pressurized Water Reactor Systems Manual THIS IS THE END DF SECTION 6 1 l N l l
dc7 key SECTION 7 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIDLOGICAL CONTROL 7.01 QUESTION (3.O) Following a stuck open PORV a reactor SCRAM and Safety Injection (SI) occur. The PDRV backup valve is subsequently closed. (a) List the four SI termination criteria. (1.0) (b) State two of the three situations which require reinitiation of SI 7 (1.0) (c) How does the SI termination criteria change following a SI reinitiation 7 (0.5) (d) If the RCS subcooling meter is inoperable how would you determine RCS subcooling 7 (0.5) 7.01 ANSWER (3.0) (a) (i) Wide range RCS pressure stable or increasing, (0.25) (ii) and pressurizer level greater than 3 [20] % (0.25) (iii) and RCS indicated subcooling greater than 20 degrees F, .( 0. 2 5 ) (iv) and total AFW elow to steam generators greater than 460 GPM or steam generator narrow range level in at least one steam generator greater than 4 E20] % (0.25) (b) (i) Pressurizer water level drops below 3 [20] % (0.5) (ii) RCS indicated subcooling drops below 20 degrees F (0.5) (c) It'does not change (0.5) (d) Use wide range RTDs (or core exit thermocouples) in conjunction with wide range RCS pressure indication and a saturation curve. (0.5) REFERENCE Emergency Procedures, EP OP-0, Reactor Trip with Safety Injection, page 14, dated 7/9/84 EP OP-1, Loss of Coolant Accident, pages 4 and 6 1
QUESTION 7.02 (3.O) The alarm for CCW surge tank level high is received followed shortly therafter by a high radiation alarm on the CCW system (RE-17A or B). (a) What are four potential sources of inleakage to the CCW system 7 (2.0) (b) If the RCP thermal barrier return line (FCV-357) closes on high flow, what two indications should be monitored 7 (1.0) ANSWER 7.02 (3.0) (a) 0.5 each for any four of the following (i) Letdown heat exchanger (ii) Thermal barrier heat exchanger (iii) E : cess letdown heat exchanger (iv) RHR heat exchanger (v) NSSS sample coolers (vi) Surge tank make-up valve leakage (vii) GFPD sample cooler (viii) Seal water heat exchanger (b) RCP seal injection flows (0.5) and RCP radial bearing temperatures (0.5) should be checked. s REFERENCE Emergency Procedure, EP OP-11, Loss of Component Cooling Water, pages 4 and 5 2
~. -. QUESTION 7.03 (4.0) State and define the following RCS Technical Specificaiton operational leakage limits. (a) Unidentified leakage (1.0) (b) Primary to secondary leakage (1.0) (c) Identified leakage (1.0) (d) Controlled leakage (1.0) l ANSWER 7.03 (4.0) l (a) 1 GPM unidentified leakage (0.5). All leakage which is not controlled leakage or identified leakage. (0.5) (b) 1 GPM total through all steam generators and 500 GPD through any one steam generator primary to secondary leakage (0.5). Primary to secondary leakage is RCS leakage through a steam generator to the Secondary Cooling system. (0,5) (c) 10 GPM identified leakage (0.5).
- Leakage, except controlled l
- 1eakage, (i) into closed systems or (ii) leakage into the containment atmosphere which is specifically located and known either not to interfere with with the operation of leakage detection systems or not to be pressure boundary leakage or (iii)
RCS leakage through a steam generator to the Secondary Coolant System. (0.5) l (d) 40 GPM controlled leakage (at an RCS pressure of 2230 + or - 20 PSIG) (0.5). Controlled leakage is that seal water flow supplied to the reactor coolant pump seals. (0.5) l e REFERENCE Technical Specifications, 3/4.4.6, Reactor Coolant System Leakage 3
QUESTION 7.04 (2.0) The Technical Specifications set an upper and a lower limit for T(ave) while the reactor is critical. (a) What are these limits 7 (1.0) (b) Explain the reason (s) why these limits are imposed 7 (One reason for the upper limit, and three reasons for the lower limit) (1.0) ANSWER 7.04 (2.0) (a) (i) Upper 581 degrees F or 582 degrees F for Unit 2 (0.5) (ii) Lower 541 degrees F (0.5) (b) (i) Upper limit --- to maintain minimum DNBR (0.25) (ii) 0.25 each for any three of the following 1. Moderator temperature is within its analyzed temperature range. 2. Protective instrumentation is within its normal operating range. 3. P-12 interlock is above its setpoint. 4. Pressurizer is capable of being in an operable status with a steam bubble. 5. The reactor pressure vessel is above its minium RT(ndt) temperature. REFERENCE Technical Specifications, Bases 3/4.1.1.4, LCOs pages 3/4 2-17 L 3/4 1-6 4
QUESTION 7.05 (4.0) List. four situations which will require emergency borati on. Describe how each of these four situations is normally determined. Include any alarms or permissives which may be energized or de-energized. ANSWER 7.05 (4.0) 1.0 each for any four of the following (a) Rods inserted below the lou-low insertion limits when critical (0.5). Indicated by rod bank low-low insertion limit alarm or by comparing plant parameters with the Technical Specification RIL (0.5). (b) Failure of any two rods to insert after a reactor trip (0.5). Indicated by rod position indication system, indicating rods not fully inserted. (0.5) (c) Uncontrolled reactor cooldown following a reactor trip with no ESF action (0.5). Indicated by low-low T(ave) alarm (0.1), P-12 permissive (0.1), continuous or unexpected decrease in T(ave) below no load value (0.1), or abnormal decreasing pressurizer level (0.1) or decreasing pressure or RCS temperature (0.1). (d) Uncontrolled or unexplained reactivity increase (0.5). Indicated by unexplained control rod insertion (0.125), unexplained increasing T(ave) (0.125), increasing nuclear power with no increase in load demand (0.125), or unexpected uncreasing count rate when shutdown (0.125). (e) When boration is required and normal boration through the VCT make-up system is not possible (0.5). Generally indicated by equipment inoperability (0.5). (f) When the shutdown margin is less than acceptable Technical Specification minimums (0.5). Calculated margin less than - 1.6% in modes 1, 2, 3, and 4 (0.25) or less than - 1.0% in mode 5 (0.25). REFERENCE Abnormal Operating Procedure, OP AP-6, Emergency Boration 5
., = DUESTIDN 7.06 (3.0) In accordance with 10CFR20, " Standards for Protection Against Radiation": (a) What are the Radiation Dose Standards for individuals in restricted areas per Calender Quarter 7 (b) What are three requirements that must be met if the Whole Body limits for a Calendar Quarter are to be exceeded 7 ANSWER 7.06 (3.0) (a) (0.5 for each of the following) (i) 1.25 Rem - Whole Body; head and trunk; active blood f or ming organs; lens of eyes; or gonads (ii) 18.75 Rem - Hands and forearms; feet and ankles ('ii) 7.5 Rem - Skin of the whole body (b) (0.5 for each of the following) (i) 3 Rem per Calendar Quarter (ii) 5{N - 18} total accumulated dose to the whole body where N is the individual's age in years at last birthday. (iii) Form NRC - 4, or equivalent I i ~ REFERENCE 10CFR20.101 6
DUESTIDN 7.07 (2.0) Refer to Figure 7.1, a table of geometry correction f actors f rom the Radiation Control Procedures for Radiation surveys (RCP-7). (a) A spot of water on the floor appears to have come from the outlet line flange of the charging pump. The diamater of the spot of water is approximatly 2 inches. An RD-3 survey meter at one inch reads 0.9 R/Hr. with the window closed and 1.0 R/Hr with the window open. What are the "true" Beta and Gamma readings 7 (b) A four foot section of 1 inch piping reads 200 mR/Hr with the window open on an RO-3 survey meter in contact (less than 1/2 inch away from) with the pipe.With the window closed the reading is 60 mR/Hr. What are the "true" Beta and Gamma readings ? ANSWER 7.07 (2.0) (a) Beta (1.0 - 0.9) R/Hr x 2 = 200 mR/Hr. W(o) - W(c) = 0.25 Gamma O.9 x 1.4 = 1.26 R/Hr W(c) 0.25 = Cf = 0.25 each (b) Beta (200 - 60)mR/Hr x 10 = 1.4 R/Hr W(o) - W(c) = 0.25 ~ Gamma 60 mR/Hr x 4 = 240 mR/Hr W(c) =0.25 Cf = 0.25 each REFERENCE RC.P G-7 7
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40 u 3 e M* 4 3 2M 2 IM in 1 7 6 5 4 3 2 2 1M eistamct u g 20 t g I 1 3 on Less 13 - 10 m a g ja 1, 3 2 IN 2-3 I3 J 3 5 .3u 1 4-4 18 3 2 I IN g k
7.08 QUESTION (2.0) Regarding the normal discharge of a waste gas decay tanks (a) Who must sign a " Gas Decay Tank Discharge Form" ? (0.5) (b) Aside from checking the Technical Specifications, state two precautions and limitations which must be met prior or during a tank discharge. (0.5) (c) What actions would you take if the concentration of oxygen in the Gaseous RADWASTE System were found to be slightly over 4% ( the Technical Specification limit) ? (0.5) (d) What is the basis for this requirement ? (0.5) 7.08 ANSWER (2.0) (a) A Radiation and Protection Department Supervisor. (0.5) (b) (0.25 for any two of the following) (i) Only one tank can be discharged at a time (ii) Only a tank not selected for fill or standby can be vented (iii) Discharge should not be made if the indicated tank pressur e is dif f erent form the pressure stated on the Gas Decay Tank Discharge Form. (c) Immediatly suspend all additions of waste gases to the system (0.25) and reduce the concentration of oxygen to less than 4 within 1 hour. (0.25) (d) Maintain the concentration of potentially explosive gas mixtures in the waste gas system below the flammability limits of oxygen and hydrogen. (0.5) REFERENCE Operating Procedure, G-2:V, Gaseous Radwaste System - Gas Decay Tank Discharge Technical Specification pages, B 3/4 11-4 and 3/4 11-15 8
7.09 QUESTION (1.O) What are the Technical Specification requirements and personnel exclusions for the site Fire Brigade 7 7.09 ANSWER (1.0) (a) A site Fire Brigade of at least five persons will be on site at all times. (0.5) (b) The site Fire Brigade shall not include the Shift Supervisor or the (two) other members of the minimum shift crew necessary for safe shutdown of the units. (0.5) REFERENCE Technical Specifications, page 6-1 9
7.10 QUESTION (1.0) As a rule of thumb, if a G-M instrument (i e: E-140, RM-15, or PRN-6), is suddenly exposed to a source more than ten times as intense as the upper limit of the least sensitive
- scale, the instrument will act erraticaly and cannot be depended upon to warn the user.
State three actions which you can take to allow you to approach an unknown source without the risk of losing the dependability of a G-M instrument. 7.10 ANSWER (1.0) (a) Turn on the meter well before approaching any object (or area) under suspicion. (0.33) (b) Approach the object (or area) slowly. (0.33) (c) Observe the meter continuously as the object (or area) is approached. (0.33) REFERENCE RCP G-7, Radiation Surveys, page 25 THIS IS THE END OF SECTIDN 7 10
dcZkcy SECTION 8 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.01 QUESTION (1.5) Technical Specifications allow f or making temporary changes to procedures. State the three conditions which temporary changes to prcedures must meet. 8.01 ANSWER (1.5) (a) The intent of the original procedure is not altered (0.5) (b) The change is approved by two members of the plant management staff at least one of whom holds a Senior Reactor Operator's license on the unit affected. (0.5) (c) The change is documented, reviewed by the PSRC and approved by the plant manager within 14 days of implementation. (0.5) REFERENCE Technical Specifications, page 6-12 Administrative Procedure, E-4S4, Issuance and Approval of Temporary Procedure Changes 1
e 8.02 DUESTION (3.0) (a) Based on the Technical Specifications f or refueling operations state five situations which require immediate termination of operations involving fuel movements in the reactor. (2.0) (b) State three conditions which require immediately stopping dilution during ref ueling. (1.0) 8.02 ANSWER (3.0) (a) (0.4 each for any five of the following) (i) Loss of one or more source range neutron flux monitors. (ii) Reactor found to be subtritical for less than 100 hours. (iii) Containment penetrations open. (iv) Direct communication lost between the control room and the Refueling Station (v) Manipulator crane or hoist not operable. (vi) Less than 23 feet of water above the reactor pressure vessel flange (vii) K(eff) more than 0.95 or boron concentration less than 2000 ppm (viii)AC power supply less than one diesel generator and one off-site power supply (ix) No borated water source available (b) (0.33 each for any three of the following) (i) Loss of one or more source range neutron flux monitors. (ii) No RHR train operating (or operable) (iii) K(eff) more than 0.95 or boron concentration less than 2000 ppm (iv) AC power supply less than one diesel generator and one off-site power supply (v) No borated water source available REFERENCE Technical Specifications, pages 3/4 9-1 through 9-11, 3/4 1-11, and 3/4 8-10 2
8.03 QUESTIDN (4.0) Regarding the Technical Specification minimum shift crew composition (including both licensed and unlicensed operations personnel): (a) What is the composition of the minimum shift crew with Unit 1 in mode 3 and Unit 2 in mode 5 7 (1.0) (b)-What is the minimum shift crew composition with both Units in mode 6 7 (1.0) (c) What is the minimum shift crew composition with both units operating in mode 1 7 (1.0) (d) When can the STA position be unmanned 7 (1.0) 8.03 ANSWER (4.0) (a) shift supervisor (SRO) (0.2), one SRO (0.2), three ROs (0.2), three Aus (0.2), and the STA (0.2). (b) Shift Supervisor (SRO) (0.33), two ROs (0.33), and three ads (0.33). (c) Shift Supervisor (SRO) (0.2), one SRO (0.2), three ROs (0.2), three AOs (0.2), and the STA (0.2). (d) The STA position may be unmanned with both units in modes 5 or 6 (0.5) or if the Shift Supervisor or the individual with the Senior Operator license meets the qualifications of the STA. (0.5) REFERENCE Unit 2 Technical Specifications, page 6-4 ' ' ~ ~
e 8.04 GUESTIDN (2.O) Using the accident classifications listed below, classify the following situations. Initially, Unit 1 is at 100% power with all systems normal and Unit 2 is in its' first refueling. 1. None of the following classifications apply 2. Unusual event
- 3. Al ert 4.
Site Area Emergency 5. General Emergency (a) A fire is detected in diesel generator room 2-1 and cannot be brought under control by the fire fighting team in 10 minutes. (b) Unit 1 primary to secondary leakage increases from O.2 GPM to 200 GPM, a reactor trip and SI f ollow. (c) A new fuel eleront is dropped and damaged in the Unit 2 spent fuel pool. deleted (d) The seal injection line to RCP 1-3 breaks and cannot be isolated. dCICtCd (e) A waste gas decay tank ruptures and releases its contents. No radiological protective action is required. (f) One of the used fuel elements in Unit 2 is inadvertently loaded into the incorrect position in the Unit 2 core. The situation is corrected by the next shift. 8.04 ANSWER (2.0) 2 (a) 3 (b) 1 (c) deleted (d) deleted (e) 1 (f) l REFERENLt ) Emergency Operating Procedures, EP G-1, Accident Classification and Emergency Plan Activiation EP OP-1, Loss of Coolant Accident EP OP-3A, Steam Generator Tube Rupture EP M-6, Nonradiological Fire EP OP-30, Inadvertent Loading of a Fuel Assembly into an Improper Position EP OP-25, Tank Ruptures 4 l + - -, - --n --,w
J 8.05 QUESTIDN (2.O) (a) In what capacity are individuals permitted (by 10CFR55) to operate the controls that directly affect reactivity 7 (1.0) (b) When is an individual deemed to be operating the controls of a nuclear facility (as defined by 10CFR55) 7 8.05 ANSWER (2.0) (a) Licensed operators as part of their official duties (0.5) and non-licensed operators as part of their training program to qualify as a licensed operator or licensed senior operator (0.5) -(b) An individual is deemed to operate the controls of a nuclear facility if he directly manipulates the controls or directs another to manipulate the controls. (1.0) REFERENCE Administrative Procedure, NPAP A-100, General Authorities and Responsibilities of Nuclear Plant Operators 10CFR55 5
s-8.06 GKJESTION (2.O) Overtime and emergency relief restrictions are described in Administrative Procedure NPAP A-8. State whether the conditions listed below would meet the restrictions. (a) A reactor operator has been working in the auxiliary building for six hours. He is subsequently scheduled to work might hours to assist fueling in Unit 2. (b) The person described above is rescheduled to have eight hours off prior to assisting the refueling in Unit 2. (c) Your shift is scheduled to work for twelve days followed by one day off. (d) Travel time and meal time do not count for shift
- time, but turnover time does.
8.06 ANSWER (2.0) (a) Do not meet (0.5) (b) Do not meet (0.5) (c) Meet requirements (0.5) (d) Do not meet (0.5) REFERENCE Administrative Procedures, NPAP A-8, Overtime and Emergency Relief Restrictions 1 6
8.07 QUESTION (2.0) Refer to Technical Specification pages 3/4 1-1 and 3/4 1-2 (attached). While in modes 1, 2, 3, or 4 the shutdown margin (SDM) must be verified to be greater than or equal to 1.6% delta K. A review of the appropriate records indicate the following sequence of events. 1. March 15 at 1200, reactor in mode 4
- 2. March 15 at 1600, SDM verification
- 3. March 15 at 2300, reactor entered mode 3 4.
March 16 at 1800, SDM verification
- 5. March 16 at 2000, reactor entered mode 2 6.
March 16 at 2100, SDM verification
- 7. March 17 at 0200, reactor went critical
- 8. March 17 at 1000, SDM verification
- 9. March 17 at 2300, SDM verification (a) Which, if any, surveillance requirements have been violated ?
(b) When is the latest the next SDM verification can be done ? 8.07 ANSWER (2.0) (a) No violations. (The stated intervals can be exceeded by up to 25% as long as three consecutive intervals do not exceed 3.25 x the stated interval.) (1.0) (b) March 18 at 1200 is the latest. (March 16 2100 to March 18 at 1200 is 39 hours which is 3.25 x 12 hours) (1.0) l REFERENCE Technical Specifications, 4.1.1.1.1 and 4.0.2 7 1-.
l o 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CCNTROL i MARGIN - T,,, Greater Than 200*F v LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k. APPLICA8ILITY: M00ES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppe baron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k. a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperabic. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s). b. When in MODES 1 or 2 with K,ff greater than or equal to 1.0, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6. When in MODE 2 with K,ff less than 1.0', within 4 hours prior to c. achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6. d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6. "See Special Test Exception 3.10.1. I i DIABLO CANYON - UNIT 1 3/4 1-1 7A
a 'O REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) ~ When in M00ES 3 or 4, at least once per 24 hours by consideration of e. the following factd'rs: ~ 1. Reactor coolant system boron concentration, 2. Control rod position, 3. Reector coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. 0 e DIA8LO CANYON - UNIT 1 3/4 1-2 7b
e o 8.00 QUESTIDN (3.5) Electrical maintenance has just been found necessary on the 2-2 RHR pump motor which is in a controlled area. Unit two is heating up and presently at 300 degrees F. An electrical f oreman submits to you a " Clearance Request and Job Assignment Sheet" for your immediate approval. (a) As shift supervisor what two checks are you required to make to ensure Technical Specification Requirements are met 7 (1.0) t (b) Why would or would not a "Special Work Permit" he required 7 (0.5) 1 (c) What two groups fill out the "Special Work Permit" and what do they record on the SWP 7 (1.0) (d) What instructions (and to whom) does the Shift Foreman give which fix the time when the RHR pump is declared inoperable 7 (1.0) 8.08 ANSWER (3.5) (a) (i) Determine if a Technical Specification time limit is involved (0,5) (ii) Verification that the redundant equipment is oper able (0.5). (b) A SWP would be required (0.25)since the work is in a controlled area (0.25) (c) (i ) Radiation Protection Supervision (0.25) fills out the types of surveys and air sampling needed (0.25) (ii) Radiation Protection Personnel (0.16) perform the surveys (0.16) and indicate required protective measures. (0.16) (d) When the Control Operator is told by the Shift Foreman (0.25) to clear and hang the " Man on Line" tags on all clearance points (0.25) for the Shift Foreman (0.25) and each person reporting on (0.25). (1.0, total) REFERENCE Administrative Procedure, NPAP C-6S1, Clearance
- request, Jot Assignment, Special Work Permit Request Procedure 8
- T
NDik k3 U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION Facility: DIABLO CANTDN l'& 2 Reactor Type: PWR - Date Administered: MAY 21, 1985 Examiner: C. shiraki Candidate: INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the question. The passing grade requires at least 707, in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. Category 5 of Candidate's % of Value Total _ Score Cat. Value Category 25 25
- 1. Principles of Nuclear Power Plant Operation, Thermo-dynamics, Heat Transfer and Fluid Flow 25 25
- 2. Plant Design Including Safety and Emergency Systems 25 25
- 3. Instruments and Controls X 27.5 25
- 4. Procedures:
Normal, Abnormal. Emergency, and Radiological Control 100 % TOTALS Final Grade All work done on this examination is my own; I have neither given nor received aid. l Candidate's Signature ~- - ----- ~' - - - ~ ~ ~ ~ ~ ~ ~~
EQUATION SiiEIT f = ma v = s/t + Cycle efficiency = (Networv out)/(Energy in) 2 w = 29 .s = V, + 1/2 ac [ s ac 2 KE = 1/2 av a=(Vf - V,)/t A. At; a,g,-6t g PE = agn Vf = V, + a t w = e/t t = in2/t1/2 = 0'.693/t1/2 ~ ~ y, y 3p t1/2'## " ESEUII*5)3 [(t1/2) * (I )3 b aE = 931 an I = I e'
- o Q = YCpat Q = UAan I = I,e~"*
I = I,10-*/U L Pwr = W an f TVL = 1.3/u sur(t) P = P 10 HVL = -0.693/u P = P e /T t o suR = 26.06/T SCR = s/(1 - K,ff) CR = 5/(1 - Keffx} x SUR = 26a/t= + (a - o)T CR (1 - K,ffj) = CR (I ~ "eff2) j 2 T=(**/s)+[(s-a)/Io] M = 1/(1 - K,ff) = CR /CR, j T = 1/(o - s) M = (1 - K,ff,)/(1 - K,ffj) T = (a - o)/(ao) sDM = (1 - K,ff)/K,ff a = (X,ff-1)/K,ff = AK,ff/K,ff t' = 10 seconds T = 0.1 seconds o = [(**/(T K,ff)] + [T,ff (1 + ST)] / 4 I dj = I d j P = (r4V)/(3 x 1010) I d) 2,2 2 7d j 22 2 I = oN R/hr = (0.5 CE)/d (=eters) R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions ) gal. = 8.345 lbs. 1 curie = 3.7 x 1010dps 1 ga;. = 3.78 If ters 1 kg = 2.21 lom 1 fte = 7.48 dal. I hp = 2.54 x 103 Stu/hr Density = 62.4' 14/ft3 1 m = 3.41 x 100 Stu/hr Density = 1 ge/cW lin = 2:54 cm Heat of vaporization = 970 Stu/lom =F = 9/5'C + 32 Heat of fusion = 144 8tu/lbe =C = S/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in Hg. 1 BTU = 778 ft-lbf I ft. H O = 0.4335 lbf/in. 2
SECTION 1 - PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER, AND FLUID FLOW 1.1 (4.0 ) um Refer to FIGURE 1.13 a sketch of " delta t' across a flJm Layer of water wi th a subcooled bulk temperature in contact with a heated surface, and the heat flux across that surface. (a) What is the mec"hanism for heat transfer in the (0.5) m region from A to B ? (b) What is the mechanism for heat transfer in the (0.5) region from B to C ? (c) If the delta-t of the film is 50 degrees F for the (1.0) heat flux at point C, what would happen to the delta-t if the heat flux was increased slighly above the value at point C ? (d) How would you expect the change in delta-t in part (1.0) (c) above, to vary the temperature of the wall of the heated surface ? (e) How would you expect the change in delta-t in part (1. 0) (c) above, to vary the bulk temperature of the liquid (water) ? M[M convection (of liquid within the film). (a) (b) Nucleate boiling (within the film layer). 1 (c) Large increase in the delta-t (due to transition to Raolant heat transfer). (d) Wal l temperature of the heated surface must increase gr e a t l y (in order to increase the heat transfer rate). (e) Bulk temperature of the liquid will not greatly increase OK)( the heat transfer rate into the liquid is not increased greatlyM 6/(, $t4(,fc fgp3d p LL I N c if.Erk i E' to T*p Ref: Westinghouse training notes Boiling heat transf er (Sec. 3.3) 1
1.2 ~ FIGURE 1.1 U.S. NUCLEAR REGULATORY COMMISSION VARIATION OF HEAT FLUX WITH FILM TEMPERATURE DIFFERENCE J E 10' s k 10'
- C 22x D
D d i-<w I 10* 8 B A i 1 I 0.1 1.0 10.0 100 1000 e AT FILM (*F)
- Section I continued on next page ***
m ---_,__,,----e_ ___..,______.__g ,,--,--...,-,_,,.--4, ..,_-_...e,
I 1.2 (3.0) .I Refer to FIGURE 1.23 a skctch of a system of three cen tr i f ugal pumps recirculating water to a tank. ~ If pump "A" is secured and isolated by valve "D-A", - what happens to the following parameters ? a EXPLA!!N EACH ANSWER i (1.5) (a) Tank delta pressure. (1.5) (b) Loop B flow. (a) DECREASES (0.5) Total Vol ume tr i c f l ow is the sum of the fl ow f rom all pumps. (0.5) The tank differential pressure is a function of total vol ume tri c fl ow squared (delta-p = KV ) As total flow is reduced (only two pumps) delta-p is reduced. (0.5) i (b) INCREASES (0.5) i As the tank delta-p decreases, the delta-p in the pump "B" line must increase to account for the l pump discharge pressure. (0.5) l In order to increase the pump line delta-p, the vol ume tr i c fl ow rate must increase. (0.5) l WESTINGHOUSE TRAINING NOTES PG&E CHAPT 10 i j 2
1.3 FIGURE 1.2 D-A s-A g.__ g x 2. D.g g s*nv em q gg
- 5-B I
l W 6N 6MaY b 9-.a us.- g = = - =. J c:x D4 p.c, u~s i ( e-..a s-e sotrry l AP v k 1 4 h )e% R h P. 'I p F PW g/cgygg79* C f
- Section I continued on next page ***
I 1.3 (2.0) se Refer to FIGURE 1.23 a sketch of a system of three -cen tr i f ugal pumps recirculating water to a tank. How would each of the following operations effect the available NPSH for the operating pumps (increase,
- decrease, or remain the same) ?
Consider each operation separately. (a) Increase the temperature setpoint of the heating (0.5) system supply valve. (b) Increase the level in the tank. (0.5) (0.5) (c) Secure two of the three operating pumps. (d) Decrease the setpiont of the N pressure (0.5) g controller. (a) decrease (b) increase (c) increase (d) decrease STD 3
Points Available i 1.4 (1.0) Which one of the following is a definition of quadrant power tilt ratio (QPTR)? (1.0) I' a. Minimum upper detector output divided by average upoer detector output. b. Maximum upper detector output divided by average upper detector output. c. Minimum upper detector output divided by average lower detector output. d. Maximum upper detector output divided by average lower detector output. ANSWER b. Reference (s) DCPP Tech Spec. 3.2.4.1. Westinghouse Thermal, Chapter 13, pp. 45-46. e e e w, s - -, - - m e - ~ - - --
O e f n ~ 1.5 00) The reactor is at 80% power at a core AT of 48'F and a mass flow rate of 100% when a station blackout occurs. Natural circulation is established and core IT goes to 40*F. If decay heat is approxi-n .mately 2% of full power, what is the mass flow rate (% of full flow)? (1.0) a. 1.9% l b. 2.1% c. 2.4% d. 3.0% ANSWER i ..? k_..C_. g bT 100 7o AT=60F + -m H p p.___M%* OC P 40 C,- ,7,,_4o i f o% : too) C.f 48 Cr-1 4, n,a )- OO 2 4 8 usco W v40 4 ggg D 31,4& N = 3% M L d6* K
- __b
REFERENCE:
Westinghouse Properties of Water
? s= 1.6 (1.0) ~ The reactor is at 80% power, rods in manual, when an inadvertent boration occurs which increases RCS (NC) boron concentration by .10 ppm. Assume a boron coefficient (differential boron worth) of -10 pcm/ ppm. If power changes by 5% (assume no rod motion), what a is the approximate value of the power coefficient (pcm/t power)? l . Show calculations _. (1.0) i ANSWER 20 pcm/% power. c,,7,10 ppm x -10 pcm/ ppm, -100 pcm 5% power 5% power Reference (s) General Reactor Theory. i l l e e i
1.7 (3.0) (1.5) Af ter operating at 50% power for several days, reactor power a. is increased to 1001. Describe the initial reactivity changes over the next I to 2 hours, and explain why this reactivity change occurs. (1.5) After operating at 100% power for 50 days reactor power is_ b. reduced to 25% due to a problem with the turbine governor. Describe the initial nasctivity changes over the next I to E 2 hours, and explain why this reactivity change occurs. ANSWER Positive reactivity will be inserted (0.5) due to the faster burnup of Zenon at the higher flux levels. (1.0) (Initially a. the production from Iodine decay remains constant, and removal Although the yield from by Decay of Zenon remains constant. fission increases the removal by burnout increases faster.) Negative reactivity will be inserted (.5) due to the reduced b. Xenon burnup at the lower power. level while production from Iodine decay remains at the higher levels of 100% power equilibrium initially. (1.0)
Reference:
Reactor Theory Supplement l t i 1 G -l ,-.n. ,v,.,- ,, -,, ~ - - - -r,-,-m--- e n,aw, r-s--,-- ,-,,-- nwr,. -~e,,- -,-,------,n
a 1.8 (1.0) Which one of the following methods of energy release from a fis-(1.0) sion is by far the largest? a. Kinetic energy of fission fragments ~ b. Kinetic energy of fission neutrons Beta emission from fission products c. d. Gama emission from fission products ANSWER a. Reference (s) Vestinghouse Nuclear Physics, p.1-1.47. e ee e e -r ,-r, .-.-,--,.----..e,-,, =
1.9 (3.0) Consider a subcritical reactor with a source strength of 20 cps. At one point during a rod withdrawal, the count rate levels off at 100 cps. After further withdrawal, the count rate lev ^ tis off at 200 cps. Show calculations. What is the final" value of K,ff? (1.5) a. b. The final shut-down margin (SDM) is approximately what percent (1.5) of the old SDM? m Calculations for Answer 1.9 SCR = (1-K,ff) S 1-K,f f = gg -X,ff=SfR ~I 20 -K,ff = 200 -I -X,ff = -0.9 K,ff = 0.9 (final) K,ff initial = - 0*I
0.8 SDM
eff 0.2 3 = 1-0.8 " U F = 0.25 SDM 0.8 9=
0.11 SDM2
DM 11 2 = 3 ~50% 3,3 1 ANSWER ~ a. 0.9 b. 50% Reference (s) Westinghouse Subcritical Multiplication, p. I-4.14.
l 1.10 (1.0) v s Three of the following four expressions represent the same value of reactivity. Which one is different? ' (1.0) a. 1500% pcm b. 15 pcm c. 0.0015% AK/K d. 0.00015 AK/K ANSWER c. Reference (s) Westinghouse Neutron Kinetics, p.1-3.2. e ___v._~.,,--_w._ ..,e m_..m ,n.- _.,, -,, w_,
m_
1.11 (2.0) The reactor is operating at 65% power when a S/G PORY fails open. a. If rod control was in automatic during this transient, which one of the following describes the behavior of controlling T-ave? (1.0) (a) Increases and remains there. (b) Increases, then returns to original value. (c) Decreases, then returns to original value. (d) Decreases and remains there. b. If rod control was in manual during this transient, which one of the following describes the behavior of controlling T-ave? (1.0) (a) Increases and remains there. (b) Increases, then returns to original value. (c) Decreases, then returns to original value. (d) Decreases and remains there. ANSWER a. (c) b. (d) Reference (s) General Reactor Theory. e e O t ,e. .,- -~
l 1.12 -(1.0) Which one of the following does NOT significantly affect K,ff? (1.0) a. Moderator pressure b. Fuel. temperature c. Source, strength d. Core age ANSWER i c. Reference (s) Westinghouse Core Physics. b 0 a e 9 8 e 0 e
1.13 (2.0) To increase power level from 5 x 10-10 amps to 10-8 amps in two (2) minutes, the start-up rate (SUR) should be maintained at decades per minute (dpm). Show calculations. (2.0) ANSWER 50R( t) h = 10 a 10- P SUR(2) 10-10 = lo 5x SUR(2) ~ 20 = lo 1.3 = SUR x 2 SUR = 1.3 = 0.65 Reference ( s) Westinghouse Neutron Kinetics
- PP I-3.15 and 3.16.
e e e r ,-,e, -en-, ,. ~ ~ -- ,,-.<w,.,m.--,--,,-v------.--- -,v--e em... w- ,,----r. - - - - - -,, - - - -, -,,w-,
f SECTION 2 - PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS i 2.1 (3.0) l The following questions pertain to the Safety Injection (SI) System. a. List four of the five Safety Injection actuations. (Logic and setpoints are not required.) (1.0) b. For the design basis loss of coolant accident, list the emergency core cooling system components that will inject water into the reactor coolant system. Include the approximate reactor cool an t system pressure at which injection will take place. (2.0) a. Any four required for full credit, 0.25 each. 1. Low pressurizer pressure 2. High containment pressure 3. High steam line flow coincident with low-low Tavg or l ow steam line pressure. 4. High steam line differential pressure 5. Manual b. 1. Centrifugal charging pumps (0.25) 2500-2650 psig (0.25) 2. Safety injection pumps (0.25) 1475-1520 psig (0.25) 3. Accumulators (0.25) 595-647 psig (0.25) 4. Residual heat removal pumps (0.25) 175 psig + up to 45 pounds NPSH depending on RWST level (0.25)
Reference:
FSAR Lesson Plan, B-6a, Reactor Protection Operator Information Manual, pages 1-4 1
2.2 (1.5) A recovery is being conducted from a reactor trip in which all reactor coolant pumps tripped. Conditions have been restored to the point where reactor coolant pumps may be started. a. Which pump should be started first? (1.0) b. What is the reason for this? (0.5) a. Reactor coolant pumps No. 1 or 2. (1.0) b. This will restore normal pressurizer spray flow. (0.5)
Reference:
Lesson Plan A-4a, page 13 s 2 l
2.3 (2.0) On Unit 1 or Unit 2, the 4KV "F" Bus can be supplied by diesel generator 1-3. a. If diesel generator 1-3 were in service on Unit 2's "F" Bus for a load
- test, and a safety injection signal came in on Unit 1, would the feeder breaker to Uni t 2 remain closed or trip?
(0.5) 6.Would the diesel generator transfer on part (a) above to Unit 1 or remain committed to Unit 2? (0.5) c. With diesel generator 1-3 running in response to a safety injection signal from Unit 1, a safety injection signal comes in from Unit 2. Prior to resetting the Unit 1 safety injection, which unit would diesel generator 1-3 supply? (0.5) d. What would result after resetting the Unit 1 safety injection? (0.5) a. The feeder breaker to Unit 2 would trip. (0.5) b. The diesel generator would be available to Unit 1. (0.5) c. The diesel generator would remain committed to Unit 1. (0.5) d. The diesel generator would be available to Unit 2. (0.5)
Reference:
Lesson Plan J-66, Diesel Generators 3
2.4 (3.5) l For the Auxiliary Sal t Water System: a. Why are vacuum breakers necessary in the piping between the ASW pumps and the component cool ing water heat exchangers? (0.5)
- b. With the Control Room standby selector switch in " AUTO",
the Hot Shutdown Panel switch in " CONTROL ROOM" and the Control Room control switch in " NEUTRAL", what four automatic signals will start an ASW pump? (2.0) c. From what two locations can the ASW pumps be manually started? (1.0) a. They prevent water hammer (0.5) (due to vacuum flashing in ASW and heat exchanger line.) b. 1. Low (less than 40 psig) discharge pressure 2. Low voltage on opposite bus. (no SIS or transfer to diesel generator.) (0.5) 3. An automatic bus transfer to the start-up transformer. (No SIS present, after a 10 second time delay.) (0.5) 4. On a transfer to the diesel generators. (No SIS present, running pumps trip; and then restart after 10 seconds.) (0,5) 5. On an SIS, the pumps will receive a start signal (22 seconds after the bus is energized.) (0.5) c. Any two of the following for full credit. 0.5 points each. The ASW pumps can be manually started from The control room. The hot shutdown panel. The 4 KV switchgear room.
Reference:
CAF for 2.4.a. Lesson Plan E-5, Auxiliary Saltwater System e 4
I i 2.5 (3.5) In reference to the Main Steam Isolation and Check Valves: a. What two signals will automatically shut the MSIV's? (2.0) 6. What purpose does the main steam check valve on the affected steam generator serve in the event of a steam line rupture inside the containment? (0.5) c. How is the Reactor Coolant System affected if there is a main steam line rupture downstream of the main steam isolation and check valves and the MSIV's fail to shut? (1.0) a. 1. High steamline f1pw (0.5) coincident with lo-lo Tavg (0.5) or low steam line pressure (0.5). 2. High-high containment pressure or Phase B isolation-(0.5) b. Prevents the unaffected steam generators from blowing down through the break. (0.5) c. Overcooling of the RCS (1.0).
Reference:
Lesson Guide C-2a, Main Steam Piping 5
w 2.6 (2.5) Fce each channel of the Area or Process Radiation Monitoring briefly describe the automatic action, if
- any, Systems
- below, that is ini tiated upon receipt of an alarm.
(0.5) Spent Fuel Pool Area (0.5) a. b. Plant Vent Gas Monitor (0.5) Gas Decay Tank Discherge Gas Moni tor c. d. RHR heat exchanger compartment exhaust h (0.5) duct air particulate monitor. (0.5) Containment Radior.ctive Gas Monitor e. iodine removal a. Fuel Handling Building ventils.tien shif ts to modei. b. Con t a i nmen t Ven t i l at i or. I sol at i on c. Isol ate the discharge from Gas Decay Tanks d. None Containment Ventil ation Isol at i on e.
Reference:
Lesson Guide G-4 6
2.7 (1.5) In reference to the containment spray system: a. What signal will cause an automatic startup of the containment spray system? (0.5) i b. During the injection phase of operation, what is the source of water to the containment spray pumps? (0.5) c. During the recirculation phase of operation, what is the source of water to the containment spray pumps? (0.5) a. Containment high-high pressure (phase B isolation) (0.25) and Safety Injection (0.25) b. RWST (0,5) c. The containment spray pumps run from 33% to 0% RWST level then shutdman.(0.25) Then the RHR pumps are used to spray down the containment if additional pressure reduction is necessary.<0.25)
Reference:
Lesson Plan B-2 7
2.8 (3.5) , The hotwell level control system is actually comprised of two independent control systems, one controlling the hotwell makeup valve and the other controlling the hotwell rejection valve. j a. From what source do the two systems receive a load index? (0.5) b. How is hotwell level affected by load? (0.5) c. What two condi tions would cause the hotwell level control valve, LCV-8, to close? (1.0) d. Hotwell rejection valve, LCV-12, will modulate automatically provided that con trol power is available to SV-461 and control switch LCV-12 is in the CONTROL ONLY
- position, or if three other conditions are fulfilled.
What are these three conditions? (1.5) a. Turbine first stage shell pressure (PT 506). (0.5) b. High load - High hotwell level (0.25) Low load - Low hotwell level (0.25) c. Any two of the following. 0.5 points each. Condensate Storage Tank Level less than 50%. (0.5) Control power for SU-459 is lost. (0.5) High hotwell level (0.5), d. 1. There is no Load Rejection Signal (LRS) (also called Load Transient Bypass) (0.5) 2. The condensate discharge pressure is greater than 145 psig. (0.5) 3. Control switch LCV-12 is in AUTO. (0.5)
Reference:
Lesson Guide C-7a l 8
2.9 (2.5) Several plant modifications in the form of instrumentation and controls have been installed since the TMI-2 event. a. What positive indication of valve ~ ~ position or flow is available in the I control room for the reactor coolant system relief and safety valves? Differentiate between indications that are only on relief valves, only on the I
- safeties, or provided for both.
(1.5) b. In addition to the subcooling margin meter, what additional instrumentation is avai l abl e to provide indication of inadequate core cooling? (1.0) a. 1. The PORV's have limit switches operated by the valve stem.(0.5) 2. Acoustic monitors give direct indication of sat'ety valve flow. (0.5) 3. Discharge pipe temperatures are provided for the relief valves and the safeties. (0.5) b. 1. Reactor vessel level instrumentation system. (0.5) 3 2. Upr.ded thermocouple system. (0.5)
Reference:
Lesson Guide B-10 e 9
2.10 (1.5) Consider each of the following questions separately. a. If the Th RTD which provides input to the OT delta T network developed an open
- circuit, how would this affect the OT delta T setpoint?
(0.5) P If 2/4 Tc RTD's developed short .its, how would this affect the te._Jater system? (0.5) c. If the Th RTD which is an input to the auctioneered high Tavg circuit developed an open circuit, how would this affect the Rod Control System? (Assume the rod control mode selector switch is in AUTO.) (0.5) a. Setpoint is lowered. (0.5) 6. Would get a feedwater isolation if a reactor trip signal is present; otherwise, no immediate effect. (0.5) c. Rods woul d dr ive in. (0.5)
Reference:
Lesson Guide, A-Sa 10
III. INSTRUMENTS AND CONTROLS ( 25 POINTS ) . 3.1 (2.0) Refer to Figure 3-1, the intermediate range detector response following a reactor trip. A. Identify which curve represents: Correct compensation, (1.0) Under-compensation and Over-compensation. What is the dotted line (# 4 ) representative of 7 D. Why, at higher power levels ( > 10~ amperes), are (O.5) the effects of under or over compensation less evident ? C. After a reactor trip, or during a shutdown, how would (0.5) under-compensation affect the source range instrumentation 7 ANSWER: A. Curve # 7 is perfectly compensated (correctly) (0.25). Curve # 5 is under compensated (0.25). Curve # 6 is over compensated (0.25). The dotted line represents delayed neutrons ( O.25). B. Th2 effects of over and under compensation at higher power levels is not, evident due to the majority of the output current coming from Neutrons which vastely outnumber the Gemma's at power levels. C. P-6 permissive would block source range trips and keep high voltage from the source range detectors because intygmediate range detector current would not drop below 10 amps after reactor shutdown.
REFERENCE:
Lesson b-4,Pg. figure +
3.2 (3.C) - The steam generator feedwater control systems e A. How do the feedwater pumps discharge pressure and steam (1.0) header pressure affect feed pump speed ? Give two ( 2) reasons why speed control is desirable in addition to flow valve control. O. How are steam line pressure signals used in the S/G (1.0) feedwater control system ? Why is this necessary ? C. Give two ( 2) of the bases for the steam generator (1.0) " programmed level" ? ANSWER : A. 1. A programmed pressure differential is maintained between the feedwater pumps discharge header pressure and steam header pressure (0.5). 2. Speed control: (any two of the following three) Maintains control valves in linear range for better throttling characteristics (0.25), Reduces pump power requirements at part load (0.25), Reduces possibility of valve plug erosion due to excessive closure at part load (0.25). B. Steam line pressure signals are used to pressure compensate tha steam flow signal (0.5). Steam flow rate varies with density which varies with pressure (0.5) C. Basis for programmed level: (any two of the following four) 1. Limit containment pressure on steam break inside of containment (0.5). ( the larger the initial mass the higher the pressure after a break.) 2. Limit cooldown on the RCS during a steam break (0.5). ( the amount of cooldown is related to the initial mass.) 3. Minimize the effects of shrink and swell during operational transients (0.5). 4. Increasing power results in increasing steam volume within the S/G tube bundle decreasing the mass of water in the bundle relative to the downcomer (0.5). Therefore at low power levels a steam break accident is more severe.
- REFERENCE :
- A,B,C, lessons No. C-8b Pg.9,4,3.
3.3 (3.0) Rod Contrcl and inntrumentction cyctem : C. Why is the" Variable Gain Unit" necessary ? Will gain (1.0) (output) increase or decrease with increasing load 7 D. List four (4) of the rod withdrawal interlocks / rod (2.0) stops, which are in ef f ect while in Automatic ". Identif y the associated interlock / rod stop setpoint. ANSWER : A. The variable gain unit is necessary to control the nonlinear response of power to control rod movement and its purpose is to prevent overshooting at higher power levels (0.5). Decreasing gain as load increases (0.5) B. Any four of the following :
- 1. Power range nuclear overpower 103%
(0.5)
- 2. Intermediate range nuclear overpower 20%
(0.5) 3.over temperature Delta T 3% below setpoint (0.5)
- 4. overpower Delta T 3% below setpoint (0.5) 5.P-impluse < 15%
(0.5)
- 6. Control bank D withdrawn, 220 steps (0.5)
REFERENCE : lesson No. A-3a,Pg. 17,19,16.
3.4 (2. 05. Pressurizer spray A. What is the purpose of the temperature element located (0.5) upstream of each pressurizer main spray valve ? D. What is the temperature alarm setpoint ? (0.5) C. Technical specification 3/4.9.2 states that the (0.5) spray water temperature shall not exceed HOWpMANY Degrees F differential ? D. Does the auxiliary spray have temperature indication (0.5) upstream of its spray valve ? Is throttled bypass flow maintained around the auxiliary spray valve ? . ANSWER : lA. The temperature element is used to monitor temperature of the bypass flow around the spray valves (0.5). B. At 500 F the temperature elements alarm (0.5) C. There is a technical specification limit that spray water temperature shall not exceed 560 F differential (0.5) D. Yes (0.25) <on the outlet of the regenerative heat exchanger,which is to be monitored once per 12 hours when auxiliary spray is in use>. No (0.25). REFERENCE : Lesson A-4a,Pg. 4 & 5 and Teck Speck 3/4.9.2
m 3.5 (3.0) Emergency Diesel Generator : A. List all Auto Start's of the EDG. (0.75) D. What conditions will block Auto Transfer of the diesel (0.75) generator to the bum 7 C. What conditions will automatically trip the Diesel (0.75) Enoine following an Auto start ? O. What conditions will normally trip the Diesel (0.75) Generator ACB open ? ANSWER : 'A. 1. No potential on the 4 KV startup feeder ACB (0.25)
- 2. No potential on 4 KV bus (.8 sec. time delay )
(0.25) 3. Safety Injection signal. (0.25) B. 1. Auxiliary feeder ACB tripped on overcurrent (0.25) 2. Startup feeder ACB tripped on overcurrent (0.25) 3. 4 KV bus has tripped by differential relay action (0.25) ( ANY THREE OF THE FOLLOWING ) C.
- 1. Generator differential relay trips (0.25) 2.
Low lube oil pressure while running (0.25) 3 Diesel engine overspeed occurs (0.25) 4. Engine cranks for ten seconds and doesn't start. (0.25) D. 1. Diesel engine shutdown relay operates (0.25) 2. Di esel generator differential relay operates (0.25) 3. Differential relay on associated 4 KV bus (0.25) operates. REFERENCE : lesson J-6b,Pg.22,23,24. M
3.0 (2.C) . For a 100% load rejection, list all conditions, interlocks (2.0) switch positions, or setpoints that must exist to actuate the steam dump system while in the T,y, mode. ANSWER : ( MUST ASSUME EITHER TURBINE TRIP OR NO TURD. TRIP ) FOR NO TURBINE TRIP : a)-Adequate condenser vacuum established (0.2) at >20 in Hg (0.2). b) At least one circulating water pump running ( .4 ). c) A 10% or greater step reduction in load (0.2) or decreasing at greater than 5% / min.(0.2).
- C-7
- d) T greater than Lo-Lo T,, (0.2) set point 543 F (0.2).
- P-12*
e c) T -T > 4F (0.4). ave ref FOR A TURBINE TRIP : a) Turbine trip (.4) b) Adequate condenser vacuum established (0.2) at >20 in Hg (0.2). c) T > than Lo-Lo T (0.2) setpoint 543 F (0.2).
- P-12*
e ave d) At least one circulating water pump running. (.4) o) T VS T c ntrolling (0.4) e ad REFERENCE : lesson C-2b,Pg's 4 thru 10
3.7 (2.5) Component cooling water liquid radiation monitoring s A. What type of Scintillation detector is used in the CCW (0.5) liquid rad. monitor system ? How many detectors are used ? D. What type (s) of radiation is the CCW liquid radiation (0.5) monitoring system sensitive to (alpha, beta, gamma). C. Besides the non-regenerative heat exchanger and the (1.0) reactor coolant pump thermal barrier heat exchanger, list four (4) additional possible sources of radioactive contamination in the CCW system. D. What automatic acgions take place at the high activity (O.5) setpoint 3.6 x 10 cpm ? ANSWER : A. NaI detectors (0.25). Two (0.25) B. Gamma only C. Letdown non-regen heat exchanger Reactor coolant pump thermal barrier heat exchanger PLUS: any four of the following at 0.25 each:
- 1) RHR HT. EX. (if in service)
- 2) Excess letdown HT. EX.
- 3) Spent fuel pit HT. EX.
- 4) Gross failed fuel detector sample cooler
- 5) Primary sample coolers.
- 6) Seal water HT. EX.
D. Surge tank vent closes automatically (0.5) REFERENCE : lesson G-4 'A.Pg.12 .B.Pg.3 'C.Pg.12
- D.Pg.12 i
3.8 (2.0) During turbine load changes, steam generator steam flow and S/G 1evel change simultaneously, and momentarily produce opposing corrective signals to the S/G feedwater control valve due to " SHRINK" or " SWELL" effects on the S/G 1evel { d;tectors during these steam flow changes.A. Explain how l thn S/G feedwater control system manages the false level cignals due to " SHRINK" or " SWELL" effects during load changes. I ANSWER : t A. The feedwater flow rate is subtracted from the steam flow rate and the resultant signal is added to the I tivel error in the controller (0.5) (porportional/ integral) This controller has a time constant (reset) (0.5) This reset action causes the system to respond slowly (slugglishly ) to rapid changes (1.0). { REFERENCE
- lesson C-8b,Pg.5 of 11 l
l l l 0 m.
3.9 (2.0) .- Refer to figure 3-9 : l A. During normal operations, would a loss of DC input (0.5) to the instrument AC inverter be quickly detected 7. D. What is the function of the pre-charge push button 7 (0.5) C. Would you expect the voltage at V to be HIGHER, (0.5) LOWER,or the SAME as V ? 2 D. On a loss of AC power, with the line controller in (0.5) i AUTD, 480 Volt AC input power to the AC inverter is locked out for 40 seconds. Why 7 ANSWER : A. Loss of DC input would not be quickly noticed (0.5). (except by a curveillance procedure done every seven days.) B. To limit inverter current (while charging the large filter capacitors in the inverters) when closing the DC input breaker. C. SAME lD. As diesel load sequencer applies loads, voltage will fluctuate which could cause the AC input breaker to l trip on over voltage (0.5). l REFERENCE : lesson J-10,Pg.5 and figure. l l l e
3.10 (2.0) Main turbine generator overspeed protection in AUTO, 500 HW,a A. List two (2) conditions for overspeed protection (0.5) actuation due to a partial loss of load. D. Specifically, what does over speed protection do (0.5) if actuated for a partial loss of load ? C. How can the overspeed protection feature be (1.0) bypassed 7 Why would overspeed protection be bypassed 7 ANSWER: ( ANY TWO.25 EACH ) A. Turbine load > 22%, sensed by PT 9 One or both PCB's closed Turbine load > Generator load (as MW) by 30%
- ANY TWO OF THREE (0.25) APIECE ***
B. If actuated, it will close the intercept valves for .4 to 1.6 seconds (This will help reduce turbine load to equal generator load.) C. By going to Turbine Overspeed Test Permissive the DEH panel (0.5). Bypassed to test the trip setpoint(0.5). on REFERENCE : lesson C-3c,Pg.14,15 l l
3.11 (1.5) Over Power / Over tempsreturo Delto T protcction A. For a decrease in primary system pressure, will the (0.5) over temperature Delta T trip setpoint increase, decrease, or remain the same ? B. For a decrease in primary system pressure will the (0.5) over power Delta T trip setpoint increase, decrease, or remain the same ? C. Which of the following will cause the DP Delta T (0.5) calculator to reduce its setpoint ? 1. T above rated ave
- 2. primary pressure below 2235 psig
- 3. rate of change of T in a decreasing direction
- 4. Delta flux exceeding"Ihe deadband ANSWER :
A.. Decrease B. Remain the name C. 1. T above rated ave REFERENCE : Lessons B-6a,Pg. 3,4 m W e -.,n
Points-Available 4.1 (3.0) ~ Some of the emergency procedures contain two values for a level ~ indication such as "SG NR level in at least one SG - greater than 4[20]%." ~ a. Why do you have two val'ues? (1.0) b. What are the conditions which require you to use the unbracketed value? (1.0) What are the conditions which require you to use the c. bracketed "[ ]" value? (1.0) ANSWER One value is for normal containment atmosphere; the other for (1.0) a. adverse containment conditions. b. Normal containment. [CAF] (1.0) c. Adverse containment. [CAF] (1.0) Reference Energency Procedures OP-0, Reactor Trip or Safety Injection CAF. e
Q GUESTION (3.0) Following a stuck open PORV a reactor SCRAM and Safety Injection occur., The PORV backup valve is subsequently closed. (SI) (a) Lir,t the f our SI termination criteria. (1.0) (b) State two of the three situations which require reinitiation of SI ? (1.0) (c) How does the SI termination criteria change following a SI reinitiation 7 (0.5) (d) If the RCS subcooling meter is i'noperable how would you determine RCS subcooling ? (0.5) ANSWER (3.0) (a) (i) Wide range RCS pressure stable or increasing, (0.25) (ii) and pressurizer level greater than 3 [203 % (0.25) (iii) and RCS indicated subcooling greater than 20 degrees F, (0.25) (iv) and total AFW flow to steam generators greater than 460 GPM or steam generator narrow range level in at least one steam generator greater than 4 [203 % (0.25) (b) (i) Pressurizer water level drops below 3 C203 % (0.5) (ii) RCS indicated subcooling drops below 20 degrees F (0.5) (c) It does not change (0.5) (d) Use wide range RTDs (or core exit thermocouples) in conjunction with wide range RCS pressure indication and a saturation curve. (0.5) REFERENCE Emergency Procedures, EP OP-0, Reactor Trip with Safet Injection, page 14, dated 7/9/84 EP OP-1, Loss of Coolant Accident, pages 4 a,nd 6 e G ee 5 ,--r,.,-.c.,
.----,--,--,-e,e.
,c--., c--
Points Available as 4.3 (2.0) (2.0) The reactor has been operating at full power when annunciators - FK 04-14 " Reactor Trip" and PK 08-21" Safety Injection" alarm. List Several insiediate operator actions are listed below. an-six additional immediate accions. 1. Verify reactor trip. ~ 2. Verify turbine trip. 3. Check if SI actuated. 4. Ensure ECCS pump status. Check ECCS status. Verify monitor light B or C status (red light on, white lights 5. 6. off). Check if main steam line isolation has occurred. 7. 8. Verify AFW status. 9. Check D/G Status. (0 ea) ANSWER h* Verifyvital4KVhrsstatus, 1. 2. Verify flow through BIT. .3. Verify containment isolation. CL::h 1: Th. '; :- 1.aluu 1. ::;&=!." L. Ensure both main FWP turbines tripped. 5. 6. Verify ventilation modes. Reference DCPP EP OP-0. t+ f g Shr%.s /*~ Q ac W e $ ho (V' WW'( /kw[reL Lt cwex ir t%st a isw~ is rine wra ss ls /ZrQuNo in c essSro o u r A" "" ,,--,-.,-.-,_,,.n,,---
^ Points Available i 4.4 (1.5) List the three symptoms that would cause you to exit emergency procedure EP OP-1.2 and enter EP OP-3 S/G tube rupture. (1.5) I' ANSWER a.' Steam line radiation high alarm (PK 11-18.) (0.5) b. SJAE radiation alarm (RE-15). (0.5) c. S/G blowdown radiation alarm (RE-19). (0.5) LEVEL IN ckc%5 5" & D. SlG Reference (s) DCPP EP OP-3. E fOf* <l t Y$ O e.
l j Available 4.5 (3.5) i e' Under what four conditions is emergency boration required? (2.0) a. 1 b. List the three alternate methods of boration that are avail-able if valve CVCS-8104 emergency boration valve is not operable. (1.5) I . ANSWER g (-cccow [N6-b{
- a. @(M Control mds inserted below the low-low insertion limit when critical.
ROD BANK LO LO INSERTION LIMIT (PK 03-14) hfi(1 Failure of any 2 control rods to fully insert following a reactor trip as indicated by rod position indication and rod bottom lights. 3 (#) uncontroiied reactor cooiant system cooidown foiiowing a reactor trip with no ESF. ~ q 5 kgrg gygr 4 x lc mw Ybt,H Sfyt.CJ [ W Uncontrolled or unexplained reactivity increase. b tvoA M &% &* 4-trTS*> C f S 7~c% T'ht2.oM KT~ lS W 'W b. (i) Swap charging pump suction from the VCT to the RWST. (ii) Emergency borate via manual emergency borate valve, CVCS-8417. (iii Inject the BORON INJECTION TANK. 10 0(ke u 6 op.4-T~t b J, th mce Cn i?lan-sh Mmg Q j o (,,,g Reference (s) DCPP DP AP-6, pp. 1, 2 and 4. I i G i e.
0.6 (1.5) Comp 1cte the following Precautions and limitations of DP L-5 ( Plant cool down for minimum load to Cold Shutdown).
- 1. Tha maximum cool down rate for the RCS must not Exceed (0.5) in any one hour period per tech spec 's.
- 2. Do not exceed a primary to secondary differential pressure (0.5) of _
3. If Steam Generator 10 % atmospheric PORV's are used for (O.5) cool down, caution must be used to avoid safety injection dun to ANSWER : 1. 100 F (0.5) 2. 1600 psid (0.5) 3. High steamline differential (0.5) REFERENCES : Operating Procedure L-5, Pg. 5 and Tech. Specks. 3/4.4.9.
O ~ 1 O.7 (1.5) O safcty injection has occurred and is being diagnosed per,EDP-0. A, lict the four (4) containment building parameters (1.0) thtt must be normal in order to terminate the Saf ety Injection. D. Whst value of sub-cooling is required for termination 7 (0.5) ANSWER : ( EITHER THE OLD CMR NEW PROCEDURE ) 'OLD PROCEDURE : + ' Pressure (0.25) Temperature (0.25) Recirc. sump level (0.25) Area radiation. monitors (0.25) NEW PROCEDURE : N2w EOP does not require any (1.0) B. 20 F (0.5)
- REFERENCE :
- Dicblo EP OP-0, Reactor trip or Safety Injection 3
,,-,-.-+,v..--- w,,- - -. - - -
0,. O (2.5) Clas2ify th3 fc11owing citustiona pper The Emergency Plan Criteria A. Release or a gas decay tank (0.5) exceeded Technical Specificat on D. For-the last 15 minutes, a ir fighting team has (0.5) been trying to control af 'n the Auxiliary building. C. An acetylene bottle has ee damaged and released (0.5) an unknown quantity of f able gas in the diesel generator room. D. Complete loss of Auxil ' ry Saltwater System. (0.5) E. S/G tube rupture k/70gpm (O.5) ANSWER A. Unusual Event (0.5) B. Alert (0.5) C. Site Area ergency (0.5) D. General ergency (0.5) E. Alert (0.5) REFERENCE A.Page (2) B.Pcg2.7 s13) C.Pcge 39 (C) D.Page 50 E.Page 22 (3) -.c.,_.
,-.---..---v----
0,.9 .(1.C) f';. How does Technicc1 Specificctions define
- Controlled (0.5)
A Leakage" 7 o D. What is the Technical Specification limit for controlled (0.5) leakage 7 ANSWER : A. Controlled leakage shall be that seal water flow cupplied to the reactor coolant pump seals (0.5). B. 40 gpm (at RCS pressure of 2230 +/- 20 psig) (0.5) REFFERENCE : Tcchnical Specifications Deff. Pg. 1-2, and 3/4.4.6
O.10' (1.5) o A. When in thn C:1d Shutdown mod], whnt in the (0.5) required Shutdown Margin ? C. What is the required Shutdown Margin when (0.5) above 200 F? C. Why does the required Cold Shutdown Margin value (0.5) differ fgom the shutdown margin value at greater than 200 F? ANSWER : A. 1.0 delta K/K B. 1.6 delta K/K C'. Because the effect of the moderator coefficient is v=ry small at the lower temperature.
REFERENCE:
Technical Specifications 3/4.1.1, 3/4.1.2, 3/4.9.1 and basns page B 3/4 1-1
O. 4.11 (1.0) Rzdiction protection o Administratively, at what dose level must a radiation area (1.0) be posted ife A. Radiation area is within the radiation control area ? B. Radiation area is in an unrestricted area ? ANSWER : A. 2.5 MREM per hour (0.5) B. 0.6 mrem / hour (0.5) REFERENCE : A & B. RCP G-5,Pg. 3 I eo l . _.. _ _ _ _.. _. _ _ _ _ _ _ _ _ _. _, _ _ _ _. ~, _. _ _ _ _ _ _. _ _. _. _ _ _
s O O.12 (1.5) O Wh t are your quarterly administrative limits for (1.0) radiation doses to the : Whole Body ( " Maximum " and " Normal"/ quarter), Hands, SKIN of whole body ? ANSWER : Whole body = Maximum 3000 mrem /QTR (0.25) & Normal 1250 mrem /DTR (0.25) H:nds 18.75 REM /QTR (0.25) = SKIN of whole body = 7.5 REM /QTR (0.25) REFERENCE : RCS-1,Pg. 2&3 .}}