ML20210B344
| ML20210B344 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/22/1987 |
| From: | Elin J, Morrill P, Obrien J, Royack M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20210B313 | List: |
| References | |
| 50-275-OL-86-02, 50-275-OL-86-2, NUDOCS 8702090137 | |
| Download: ML20210B344 (72) | |
Text
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U. S. NUCLEAR REGULATORY COMMISSION t
REGION V Report No. 50-275/0L-86-02 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 1 and 2
' Examinations at: Avila Beach, California
~
Examination conducted: December 2-4, 1986
' ' Md -[
Examiners:
P. Morvill Date Signed W
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J
'Brien Date Signed
/
D MA.Ropck Date Si'gned U
Approved by:
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J E11%, Chief, Operations Section L$eSitned Sumary:
Examinations were conducted December 2 through 4, 1986. The written examination was administered on December 2,1986 to five senior reactor operator' candidates (SRO). The oral and simulator examinations were administered to the candidates during the period of December 3 and 4, 1986.
All candidates passed the oral and written examinations.
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PDR V
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4 I-REPORT DETAILS 3,
1.
Examiners:
P. Morrill, Chief Examiner, Region V J. O'Brien M. Royack J. Elin 2.
Persons Attending the Exit Meeting December 4, 1986 NRC P. Morrill, Region V
'J. O'Brien, Region V M. Royack, Region V Pacific Gas and Electric Company J. Sexton, Plant Superintendent L. Womack, Operations Manager T. Martin, Training Manager r
J. Becerra, Senior Training Instructor B. Terrell, Senior Training Instructor J. Welsch, Senior Training Instructor C. Leach, Senior Training Instructor J. Molden, Operations Training Supervisor R. Jett, Simulator Supervisor 3.
Written Examination Review 4
The written examination was administered on December 2, 1986. At the conclusion of the examination a copy of the examination key was provided to Mr. J. Molden, of the licensee's Training Department,.
for review. The written examination key was reviewed by the licensee's Messrs. Terrell, Grahm, Becerra, Somers, Dressler, Buckley and Molden.
The licensee's review of the written examination resulted in comments on eighteen examination questions and/or responses, which are included in Attachment A of this report. Licensee examination comments were I.
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37 1
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-reviewed with the licensee's Mr. J. Moldin, Operations Training Supervisor, and-the Nuclear Regulatory Commission's Mr. J. Elin,
.o ti ',
Operations Section, Chief, and Mr. M. Royack, Licensing ' Examiner.
';f The-written examination master key was revised, as annotated in the 4
response..section of Attachment A to this report, prior to grading m:
.the candidates responses.
P, 4.
Operating Examinations Simulator and operating examinations were conducted on December 3 and 4, 1986. During the simulator examination a' generic problem was noted in the candidates use of Diablo Canyon Emergency Operating Procedure (EOP)
EP ECA-0.3, " Restore Vital Bus".
Step 8oftheproceduredirectstheoperatorto:vehifyopenpower supply breakers to vital busses prior to re-energizing these buses..
In the simulator scenario which was performed,_some vital busses remained.
energized. Tripping'of the power supply to these energized ~ vital busses appears unnecessary as a step toward restoring power.
t.
Step 8 of E0P EP ECA-0.3 was discussed with the licensee.- The licensee agreed to further review and revise _EP ECA-0.3 as necessary.
5.
Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the examination. The examiners noted that there was a generic weakness found in the use of emergency operating procedures specifically EP ECA-0.3, " Restore Vital Bus".
The licensee committed to review and revise the procedure.
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ATTACHMENT A TO REPORT 50-275/0L-86-02 DIABLO CANYON UNITS 1 AND 2
)
NRC SRO UPGRADE EXAM (12-2-86) REVIEW Exam review conducted by the following:
Section 5: Terrell, Graham, Becerra Section 6: Terrell, Somers, Katz Section 7: Terrell, Buckley, Dressler Section 8: Terrell, Buckley, Katz Question'5.06:
Values given by key for equilibrium and peak Xenon are from the_old Operator Information Manual. New NIX tables and Volume 9 Table R17-T-2 gives values for peak Xenon up to -6800 pcm at BOL and for equilibrium of up to -3220 pcm at BOL. Request that the allowable tolerances on the key be changed as follows:
Peak
-200, +1000 pcm a
Equilib
-100, +450 pcm Response 5.06 Values for peak Xenon of up to -6800 and -3220 pcm for equilibrium were incorporated into answer key.
Question 5.08:
Answer key differs from the exact wording in the Tech Spec B 3/4 1-3:
Answer 1--accept " Potential effects of rod misalignment on associated accident analysis are limited" Answer 3--accept " Acceptable power distribution is maintained" Response 5.08:
Answer key revised to incorporate additional responses for answers 1 and 3.
Question 6.03:
Answer #3 should read low steam generator level coincident with steam flow greater than feed flow.
Response 6.03:
Answer #3 in key revised to read:
" low steam generator level coincident with steam flow greater than feed flow."
Question 6.05:
' Request that the key be modified to include individual trips that encompass the solenoid trip device which are:
Safety Injection P-14 Low Reservoir level Thrust bearing wear Control Room manual trip switch Response 6.05:
Answer key has been revised to include:
SIS Low lube oil reservoir level Excessive thrust bearing wear j.
High discharge header pressure Question 6.09:
Answer #4 request grader accept either Radiation instrument numbers, as listed below or word descriptions as is now in the key.
a.
RE-ll or RE-12 b.
RE-28A or 28B c.
RE-14A or 14B Response 6.09:
Answer #4 is revised to accept:
a.
RE-11 or RE-12 b.
RE-28A or RE-28B c.
RE-14A or RE-148 as acceptable responses Question 6.10:
Transformer numbers or Diesel numbers are not required by the question, acceptable answer should be:
1.
Auxiliary bus 2.
Startup bus (0.25) or Aux bus (0.25) is acceptable, on a long term basis the plant is normally backfed from the 500 kv yard through the Aux bus.
3.
Diesel Generator Response 6.10:
Comments 1 and 3 were not accepted, key remained unchanged.
Answer 2 is acceptable, answer key is revised to include comment 2.
Question 7.01:
Request that grader accept 300 gallons of Boric Acid as well'as 100 ppm since they are equivalent. Part 2
.should therefore be 100 ppm per rod or 600 gallons,:
Part 5 should be 100 ppm or.300, gallons.
~
Response 7.01:
Comments acceptable, answer keyiis revised to reflect 300 gallons of boric acid as equivalent to 100 ppm.
Question 7.02:
Request that grader accept TSC for part 6 (STA), even though not specifically called out for in the procedure.
The event notifications are normally done"by the STA in-this type of an event, the STA is responsible for the monitoring of plant information on the SPDS in the TSC.
Response 7.02:
Comment accepted, answer key has been revised to reflect comment.
Question 7.03:
Question is ambiguous in that it asks for a definition of the limits and this can be taken as strictly define or give limits. Consider this when grading.
Response 7.03:
Answer key has been revised to include limits as acceptable.
Question 7.04:
Part A does not call for definitions of Whole Body or Extremities, request full credit be given for the actual limits.
Response 7.04:
Whole body, extremities and skin dose have been added as acceptable definitions of radiation dose standards.
Question 7.06:
Tank capacity and GPM of pumps not requested.
Sea Water Evaporator is presently abandoned in place, even though the procedure has not been updated. The Lesson Guide LF-2 only lists Makeup water as a source.
Request that answer 3 be eliminated and that full credit be given for answers 1 and 2 only.
Response 7.06:
Tank capacities and GPM of pumps were not required as part of the acceptable response. Sea water evaporator has been deleted from the answer key.
Question 7.07:
Answer 2; request full credit be given for " Chemistry and Radiation Protection Supervisor", this is the title stated in the Chem Procedure CAP-5 page 11 of 66 paragraph b.
Response 7.07:
Definition of " CARP" is acceptable, answer key has been revised to accept Chemical and Radiation Protection Supervisor.
Question 7.08:
Answer c; correct quote from Tech Spec 3.11.2.6 is as follows:
"If the limit is exceeded, all additions to the tank are to be suspended and tank contents reduced to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />." Referenced lesson plan has been revised since mailing to you.
Response 7.08:
Corrected quotation has been added to answer key.
Question 7.12:
Add to the end of the answer of part b "if needed".
Response 7.12:
"If needed"-has been added to response part "b" answer key.
Question 8.02:
Reference should be 3/4.9.4 for:part a and 3/4.9.5 for part b.
Answer a; the Tech Spec only mentions:
" Core alterations",
request the deletion of the second portion of the answer "or movement of irradiated fuel inside containment".
Response 8.02:
Answer key revised to show correct references. Answer "a" revised on answer key to reflect coments.
Question 8.04:
Answer #6; Request the replacement of SS, S0L and SCO, with "SFM, SCO, and C0" this is the exact wording for DCPP.
Response 8.04:
Answer key revised to accept "SFM, SCO and C0".
~
' Question 8.08:
The question from 10 CFR 50.59 states "... temporary modification...", the actual wording should have been
"... proposed change test or experiment.." or paraphrasing to Unreviewed safety questions. Due to the fact that this is not considered part of the operators job, request that the grader use leniency in
'the grading process for this question.
Response 8.08:
Responses to question 8.08 are to be reviewed with respect to licensee's comments.
Question 8.09:
Part b to the answer is incorrect. December 5, 1986, at 1900 will be the 39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> point for the last. 3 intervals.
Part c to the answer is also incorrect. Tech Spec 3/4.5.1 states that the condition must be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HSB within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HSB within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Response 8.09:
Answer key has been revised to indicate correct responses.
EXAM KEY U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: Diablo Canyon Units 1 and 2 Reactor Type: Westinghouse PWR Date Administered: December 2, 1986 Examiner:
M. J. Royack Candidate:
Il[STRUCTIONSTOCANDIDATE:
Use separate paper for the answers. Write answers on one side only. Staple j
question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in eac,h category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
t
% of c
Category
% of Candidate's Category Value Total Score Value Category 25 25 5.
Theory of Nuclear Power Plant Operation, Fluids, and Thermo-dynamics 25 25 6.
Plant Systems, Design, Control, and
,j Instrumentation 25 25 7.
Procedures - Normal, Abnormal, Emergency, and Radiological Control 25 25 8.
Administrative Procedures, Conditions, and Limitations 100 Totals Final Grade All work done on this examination is my own, I have neither given nor received aid.
Candidate's Signature
-- ---m----- ---
/
ES-201-2 1
ATTACHMENT 2 REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS 1.
A single room shall be provided for completing the written examina-tion.
The location of this room and supporting restroom facilities shall be such as to prevent contact with all other facility and/or contractor personnel during the duration of the written examination.
If necessary, the facility should make arrangements for the use of a suitable room at a local school, motel, or other building.
Ob-
.taining this room is the responsibility of the licensee.
2.
Minimum spacing is required to ensure examination integrity as determined by the chief examiner.
Minimum spacing should be one candidate per table, with a 3-ft space between tables.
No wall 4
charts, models, and/or other training materials shall be present in the examination room.
3.
Suitable arrangements shall be made by the facility if the candi-dates are to have lunch, coffee, or other refreshments. These a
arrangements shall comply with Item 1 above. These arrangements shall be reviewed by the examiner and/or proctor.
4.
The facility staff shall be provided a copy of the written examination and answer key after the last candidate has completed and handed in his written examination. The facility staff shall then' have five working i
days to provide formal written comments with supporting documentation on the examination and answer key to the chief examiner or to the regional office section chief.
5.
The licensee shall provide pads of 8-1/2 by 11 in. lined paper in unopened packages for each candidate's use in completing the exam-ination.
The examiner shall distribute these pads to the candidates.
All reference. material needed to complete the examination shall be furnished by the examiner.
Candidates can bring pens, pencils, calculators, or slide rules into the examination room, and no other equipment or reference material shall be allowed.
6.
Only black ink or dark pencils should be used for writing answers to questions.
l Examiner Standards 11 of 18
ES-201-2 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
/
Print your name in the blank provided on the cover sheet of the
- 4.. examination.
L 5.
Fill in the date on the cover sheet of the examination (if necessary).
t 6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category
" as appropriate, start each category on a new page, write only one Ede of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10.
Skip at least three lines between each answer.
- 11. Separate answer sheets from pad.and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13.
The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14.
Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15.
Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION l
AND DO NOT LEAVE ANY ANSWER BLANK.
l l
16.
If parts of the examination are not clear as to intent, ask questions of
[
the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has l
been completed.
(
Examiner Standards 12 of 18
- 18. When you complete your examination, you shall:
a.
Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
I (3) Answer pages including figures which are a part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did T
not use for answering the questions. -
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
i t
s_
l l
Examiner Standards 13 of 18 i
s' EQUATION SHEET e
f = ma v = s/t 2
t) w = ag a = v,t +
at Cycle efficiency =
E = mC a = (vg - v )/t 9
t f=v + at A = AN A = A',e KE = my v
PE = agh a = 8/t A = in 2/tg = 0.693/tg W = vaP (g )(t )
h AE = 931Am h*
~
(tg+t) 3
= $C AT y,y,-Ix p
Q = UAAT y,y,-ux Pwr = W It g
I=I 10 *! M
~
SUR(t)"
P=P 10 g, y,3fp O
P = P, e /T t
HVL = 0.693/u SUR = 26.06/T l
T = 1.44 DT SCR = S/(1 - K,gg)
(A o\\
- ff
= S/(1 - K,ggx)
SUR = 26 CR g,
x C ~ eff}1 "
2(I ~~ Keff)2 T = D */p ) + [(s l p)/A,gf]
I p
T = 1*/ (p E
M = 1/(1 - K,gg) = CR /CR y
0
"( ~ 8)! effD M = (1 - K gg)0 (I ~ eff)1
- " ( eff~I)! eff = AK,ff/K aff SDM = (1 - K,gg)/K,gg
[L*/TK,'gg ] + [B/(1 + A,gg )]
T 1* = 1 x 10 seconds
~
p=
P = I$V/(3 x 1010)
-I x
= 0.1 seconds eff I = No Id =Id yy 22 WATER PARAMETERS Id =Id g
3 1 gal. = 8.345 lba R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters R/hr = 6 CE/d (feet)
I ft = 7.48 gal.
MISCELLANEOUS CONVERSIONS 3
10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm i kg = 2.21 lba 3
Heat of varorization = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr 0
Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in. I'g.
1 Btu = 778 ft-lbf I ft. H 0 = 0,4333 lbf/in 1 inch = 2.54 cm y
F = 9/5 C + 32 C = 5/9 ( F - 32)
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F Properties of Saturated Steam and Saturated Water
- Absolute Pressure Vacuum Temper-Heat of Latent Heat Total Heat Specific Volume Inches Inches ature the of of Steam y
Lbs. E.r Sq.
of Hg of Hg Liquid Evaporation t
Ile Water becam P'
o.a, v.
s.ofis. _
s.u ns.
seuris.
cu.r..
is.
cu.re.,,,is.
0.0087 0.02 29.90 32.018 0.0003 1075.5 1075.5 0.016022 3302.4 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 1081.4 0.016020 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 27.382 1060.1 1087.4 0.016032 1235.5 4
0.30 0.61 29.31 64.484 32.541 1057.1 1089.7 0.016040 1039.7 0.35 0.71 29.21 68.939 36.992 1054.6 1091.6 0.016048 898.6 0.40 0.81 29.11 72.869 40.917 1052.4 1093.3 0.016056 792.1 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 1048.6 10 %.3 0.016071 641.5 0.60 1.22 28.70 85.218 53.245 1045.5 1098.7 0.016085 540.1 0.70 1.43 28.49 90.09 58.10 1042.7 1100.8 0.016099 4 %.94
. 0.80
' l.63 28.29 94.38 62.39 1040.3 1102.6 0.016112 411.69 0.90 1.83 28.09 98.24
%.24 1038.1 1104.3 0.016124 368.43 1.0 2.04 27.88 101.74 69.73 1036.1 1105.8 0.016136 333.60 1.2 2.44 27.48 107.91 75.90 1032.6 1106 5 0.016158 280.%
1.4 2.85 27.07 113.26 81.23 1029.5 1110.7 0.016178 243.02 1.6 3.26 26.%
117.98 85.95 1026.8 1112.7 0.0161 %
214.33 7
1.8 3.%
26.26 122.22 90.18 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.85 126.0/
94.03 1022.1 1116.2 0.016230 173.76
[
2.2 4.48 -
25.44 129.61 97.57 1020.1 1117.6 0.016245 158.87 2.4 4.89 25.03 132.88 100.84 1018.2 1119.0 0.016260 146.40 l
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 0.016287 126.67 e
3.0 6.11 23.81 141.47 109.42 1013.2 1122.6 0.016300 118.73 3.5 7.13 22.79 147.56 115.51 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 20.76 157.82 125.77 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 1% 29 134.26 998.5 1132.7 0.016430 67.219 6.0 12.22 17.70 170.05 138.03 996.2 1134.2 0.016451 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 0.016441 53.650 7.5 15.27 14.65 179.93 147.93 990.2 1138.2 0.016510 50.294 8.0 16.29 13.63 182.86 150.87 988.5 1139.3 0.016527 47.345 8.5 17.31 12.61 185.63 153s65 986.8 1140.4 0.016545 44.733 l
9.0 18.32 11.60 188.27 I56.30 985.I 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 20.36 9.56 193.21 161.26 982.1 1143.3 0.016592 38.420 l
11.0 22.40 7.52 197.75 165.82 979.3 1145.1 0.016622 35.142 I
12.0 24.43 5.49 201.%
170.05 976.6 1146.7 0.016650 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1148.2 0.01 % 76 30.057 14.0 38.50 1.42, 209.56 177.71 971.9 1149.6 0.016702 28.043 w
L m
Pressure Temper-Heat of Latent Heat Total Heat Specific Volume l
Lbs. per Sq. In.
ature the of of Steam y
Absolute Gage Liquid Evaporation l
t he Water Steam P'
P o.
r.
s uns.
s und.
a.u ns.
cu.t.,.,in.
cu. re. p., is.
14.6 %
0.0 212.00 180.17 970.3
!!50.5 0.016719 26.799 15.0 0.3 213.03 181.21
%9.7 1150.9 0.016726 26.290 16.0 1.3 216.32 184.52
%7.6 1152.1 0.016749 24.750 17.0 2.3 219.44 187.%
%5.6 1153.2 0.016771 23.385 18.0 3.3 222.41 190.%
%3.7 -
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.%
1%.27 960.1
!!56.3 0.016834 20.087 21.0 6.3 230.57 198.90 958.4 1157.3 0.016854 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 10.3 240.07 208.52 952.1 1160.6 0.016927 16.301 26.0 11.3 242.25 210.7 950.6 1161.4 0.016944 15.7138 27.0 12.3 244.36 212.9 949.2 1162.1 0.016 % I 15.1684 28.0 13.3 246.41 214.9 947.9 1162.8 0.016977 14.6607 29.0 14.3 248.40 217.0 946.5 1163.5 0.016993 14.1869 30.0 15.3 250.34 218.9 945.2
!!64.1 0.017009 13.7436 31.0 16.3 252.22 220.8 943.9 1164.8 0.017024 13.3280 32.0 17.3 254.05 222.7 942.7 1165.4 0.017039 12.9376 33.0 18.3 255.84 224.5 941.5 11 %.0 0.017054 12.5700 34.0 19.3 257.58 226.3 940.3 11 %.6 0.017069 12.2234
P perties 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
Abeol,ute Gage Liquid Evaporation P
P n.ar r.
mris.
siurie.
meis.
cu. ri.,., is.
cu.ri.p.,in.
35.0 20.3 259.29 228.0 939.1 1167.1 0.017083 11.8959 36.0 21.3 260.95 229.7 938.0 1167.7 0.017097 11.5860 37.0 22.3 262.5a 231.4 936.9 1168.2 0.017111 11.2923 38.0 23.3 264.17 233.0 935.8 1168.8 0.017124 11.0136 39.0 24.3 265.72 234.6 934.7 1169.3 0.017138 10.7487 40.0 25.3 267.25 136.1 933.6 1169.8 0.017151 10.4 % 5 41.0 26.3 268.74 237.7 932.6 1170.2 0.017164 10.2563 42.0 27.3 270,21 239.2 931.5 1170.7 0.017177 10.0272 43.0 28.3 271.65 240.6 930.5 1171.2 0.017189 9.8083 44.0 29.3 273.06 242.1 929.5 1171.6 0.017202 9.5991 45.0 30.3 274.44 243.5 928.6 1172.0 0.017214 9.3%8 M.0 31.3 '
275.80 244.9 927.6 1172.5 0.017226 9.2070 47.0 32.3 277.14 2#.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 34.3 279.74 248.9 -
924.8 1173.7 0.017262 8.6770 50.0 35.3 281.02 250.2 923.9 1174.1 0.017274 8.5140 51.0 36.3 282.27 251.5 923.0
!!74.5 0.017285 8.3571 52.0 37.3 283.50 252.8 922.1 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 0.017351 7.5280 58.0 43.3 290.50 259.9 917.0 1177.0 =
0.017362 7.4059 59.0 44.3
- 291.62 261.1 916.2 1177.3 Q.017372 7.2879 60.0 45.3 292.71 262.2 915.4 1177.6 0.017383 7.1736 61.0 M.3 293.79 263.3 914.6 1177.9 0.017393 7.0630 62.0 47.3 294.86 264.4 913.8 1178.2 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 66.0 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 56.3 303.89 273.7 407.1 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
%5.0 1181.6 0.017519 5.8885 75.0 60.3 307.61 277.6 904.3 1181.9 0.017529 5.8144 76.0 61.3 308.51 278.5
%3.6 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 f
84.0 69.3 315.43 285.7 898.3 1184.0 0.017608 5.2249 85.0 70.3 316.26 286.5 897.7 1184.2 0.017617 5.1669 86.0 73.3 317.08 287.4 897.0 1184.4 0.017625 5.1101 87.0 72.3 317.89 288.2 8%.4 1184.6 0.017634 5.0546 88.0 73.3 318.69 289.0 895.8 1184.8 0.017642 5.0004 89.0 74.3 319.r 289.9 895.2 1185.0 0.017651 4.9473 90.0 75.3 320.a 290.7 894.6 1885.3 0.017659 4.s953 91.0 76.3 321.06 191.5 893.9 1185.5 0.017667 4.8445 1
92.0 77.3 32I.84 292.3 893.3 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 1886.0 0.017692 4.6952 95.0 80.3 324.13 294.7 591.5 1886.2 0.017700 4.6584
[
%.0 81.3 324.M8 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 0.0t7724 4.5166 99.0 84.3 327.10 297.8 889.2 1187.0 0.017732 4.4734 l
100.0 h5.3 317.81 298.5 888.6 lis7.2 0.017740 4.4380 10l.0 86.3 J28.54 299.3 888.1 1187.3 0.01775 4.3895 102.0 M7.3 329.26 300.0 887.5 1887.5
'O.01776 4.3487 103.0 88.J 329.97 300.8 886.9 1187.7 0.01776 4.3087 104.0 M9.3 330.67 301.5 886.4 1887.9 0.01777 4.2695 l
105.0 90.3 331.37 302.2 885.8 1188.0 0.01778 4.2309
~
i 106.0 91.3 332.06 303.0 885.2 1888.2 0.01779 4.1931 107.0 42.3 332.75 303.7 884.7 1188.4 0.01779 4.1560 10M.0 93.3 333.44 304.4 884.1 1188.5 0.01780 4.1195
__ I% 0 94.3 334.11 305.1 883.6 1188.7 0.01781 4.0837
_ Properties of Saturated Steam and Saturated Water-continued Pressure Temper-Heat of I atent Heat : Total Heat Specific Volume Lbs. per Sq. In.
ature the of of Steam p
Liquid Evaporation Absolute Ga8e 1
Water Steam P'
P o.s,. r.
sions.
aions.
nions.
cuan,is.
c. ri. n, is.
110.0 95.3 334.79 305.8 883.1 1188.9 0.01782 4.04M4 111.0
%.3 335.%
306.5 882.5 1189.0 0.01782 4.0138 l
112.0 97.3 336.12 307.2 882.0 1189.2 0.01783 3.9798 113.0 98.3 336.78 307.9 881.4 1189.3 fl01784 3.9464 114.0 99.3 337.43 308.6 880.9 II89.5 0.01785 3.9136 115.0 100.3 338.08 309.3 880.4 1189.6 0.01785 3.8813 116.0 101.3 338.73 309.9 879.9 1189.8 0.01786 3.8495 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 120.0 105.3 341.27 312.6 877.8 1190.4 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 3.6695 343.13 314.5 876.3 1190.8 0.01791 3.6411 i
123.0 108.3 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132 I
125.0 110.3 j 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 3 %.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 F68.5 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 146.0 131.3 356.31 328.4 865.2 1193.6 0.01806 3.0927 147.0 132.3 356.84 329.0 864.8 1193.8 0.01807 3.0726 148.0 133.3 357.38 329.5 864.3 1193.9 0.01808 3.0528 149.0 134.3 357.91 330.I' 863.9 1194.0 0.01808 3.0332 i
150.0 135.3 358.43 330.6 863.4 1194.1 0.01809 3.0139 l
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 160.0 145.3 363.55 336.1 859.0 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 166.0 151.3 366.50 339.2 856.5 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.J 370.31 343.2 853.1 11 %.4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 %.5 0.01825 2.5864 178.0 163.3 372.16 345.2 851.5 11 %.7 0.01826 2.5585 180.0 165.3 373.08 346.2 850.7 1196.9 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 186.0 171.3 375.77 349.1 848.3 1197.3 0.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 846.7 1197.6 0.01833 2.4030 i
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 0.01836 2.3322 198.0 183.3 380.%
354.6 843.6 1198.2 0.01838 2.3095 j
200.0 185.3 381.80 355.5 842.8 1198.3 0.01839 2.28728
(
l 205.0 190.3 383.88 357.7 840.9 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 220.0 205.3 389.88 364.2 835.4 1199.6 0.01850 2.08629 225.0 210.3 391.80 366.2 833.6 1899.9 0.01852 2.04143 230.0 215.3 393.70 368.3 831.8 1200.1 0.01855 1.99846 235.0 220.3 395.56 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3 397.39 372.3 828.4 1200.6 0.01860 1.91769 245.0 230.3 399.19 374.2 826.6 1200.9 0.01863 1.87970 i
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p
+%
5 Properties of Saturated Steam and Saturated Water-concluded Pressure Temper-Heat of Latent Heat Total Heat Spedfic Volume Lbs. per Sq. In.
sture the of of Steam I'
b 'P'
p Absolute GaSe t
(
P' P
o F.
neutie.
s urin.
niofis.
co. ei.,., is.
co. re. p.,is.
Water Steam 250.0 235.3 400.97 376.1 825.0 1201.1 0.01865 1.84317 255.0 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80802
/
260.0 245.3 404.44 379.9 821.6 1201.5 0.0I870 1.774I8 v
265.0 250.3 406.13 381.7 820.0 1201.7 0.01873 1.74157 270.0 255.3 407.80 383.6 818.3 1201.9 0.01875 1.71013 s
275.0 260.3 409.45 385.4 816.7 1202.1 0.01878 1.67978 i
280.0 265.3 411.07 387.1 815.1 1202.3 0.01880 1.65049 l
285.0 270.3 412.67 388.9 813.6 1202.4 0.01882 1.62213 I
290.0 275.3 414.25 3%.6 812.0 1202.6 0.01885 1.59482 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 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 N
400.0 385.3 444.60 424.2 780.4 1204.6 0.01934 1.16095 420.0 405.3 449.40 419.6 775.2 1204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535
.4 460.0 445.3 458.50 439.8 765.0 1204.8 0.01959 1.00921
'[
480.0
- 5.3 462.82 444.7 760.0 1204.8 0.01 % 7 0.96677 f
500.0 485.3
% 7.01 449.5 755.1 12Q4.7 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 500.0 565.3 482.57
- 7.5 736.5 1203.9 0.02006 0.79712 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
%0.0 645.3 4 %.57 483.9 718.8 1202.7 0.02036 0.69724 680.0 MS.3 499.86 487.8 714.5 1202.3 0.02043 0.67581 700.0 685.3 503.08 491.6 710.2 1201.8 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 760.0 745.3 512.34 502.7 697.7 1200.4 0.02072 0.60097
~
780.0 765.3 515.30 506.3 693.6 1199.9 0.02080 0.58457 800.0 785.3 518.21 509.8 689.6 1199.4 0.02087 0.568 %
820.0 805.3 521.%
513.3 685.5 1198.8 0.02094 0.55409 840.0 825.3 523.86 516.7 681.5 1198.2 0.02101 0.53984 1
860.0 845.3 526.60 520.1 677.6 1197.7 0.02109 0.52631
.a 880.0 865.3 529.30 523.4 673.6 1197.0 0.02116 0.51333 900.0 885.3 531.95 526.7 669.7 11 %.4 0.02123 0.50091 i
l.
920.0 905.3 534.56 530.0
%5.8 1195.7 0.02130 0.48901 940.0 925.3 537.13 533.2 661.9 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 0.45609 1000.0 985.3 544.58 542.6 650.4 1192.9 0.02159 0.445 %
r 1050.0 1035.3 550.53 550.1 640.9 1191.0 0.02177 0.42224 i
~
1100.0 1085.3 556.28 557.5 631.5 1189.1 0.02195 0.40058 1150.0 1135.3 561.82 564.8 622.2 1187.0 0.02214 0.38073 1200.0 1185.3 567.19 571.9 613.0 1184.8 0.02232 0.36245 1250.0 1235.3 572.38 578.8 603.8 1182.6 0.02250 0.34556 l!-
1300.0 1285.3 577.42 585.6 594.6 1180.2 0.02269 0.32991 1
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 0.28909 1500.0 1485.3 5 %.20 611.7 558.4 1170.1 0.02346 0.27719 1600.0 1585.3 604.87 614.2 540.3 1164.5 0.02387 0.25545 1
1700.0 1685.3 613.13 636.5 522.2 1158.6 0.02428 0.23607 t 800.0 1785.3 621.02 648.5 503.8 1152.3 0.02472 0.21861 1900.0 1885.3 628.56 M0.4 485.2 1145.6 0.02517 0.20278 i
2000.0 19NS.3 635.80 672.1 466.2 II38.3 0.02565 0.15831 2100.0 2085.3 642.76 683.8 4%.7 1130 5 0.02615 0.17501 j
2200.0 2185.3 649.45 695.5 416.7 1122.2 0.02 % 9 0.16272 2300.0 2285.3 655.89 707.2 406.0 1113.2 0.02727 0.15I33 i
2400.0 23AS.3
% 2.11 719.0 384.8 1103.7 0.02790 0.14076 2500.0 2445.3 h68.11 731.7 361.6 1093.3 0.02859 0.13065 2600.0 2585.3 673.91 744.5 337.6 1082.0 0.02938 0.12110 l
2700.0 2685.3 679.53 757.3 312.3 1069.7 0.03029 0.11194 2000.0 2785.3 624. %
770.7 285.1 1055.8 0.03I34 0.10305 2900.0 2845.3 690.22 745.1 254.7 1039.8 0.03262 0.09420 l
Juno.0 1945.3 695.33 sol.s 218.4 1020.3 0.03428 0.04500 3100.0 3085.3 700.28 824.0 169.3 993.3 0.03681 0.07452 1
J200.0 JINS.3 705.08 875.5 56.1 931.6
' O.04472 0.05 % 3 3208.2 3193.5 705.47 906.0 0.0 906.0 0.05078 0.05078 1
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1 SECTION 5 THEORY OF NUCLEAR POWER OPERATION, FLUIDS,AND THERMODYNAMICS QUESTION 5.01 (3.0)
The plant is in hot standby with one of the steam generators at 905 PSIG.
Une the attached steam tables or mollier diagram to answer the following questions.
(c) What is the steam temperaure in the steam generator?
(1.0)
(b) If an atmospheric relief valve is opened what will be the temperature of the discharge steam? (Assume discharge pressure is (1.0) atmospheric.)
(c) Will the discharge steam be superheated, saturated,or a mixture of saturated steam and water? Briefly explain your answer.
(Assume discharge pressure is atmospheric.)
(1.0) 4
'i ~ ANSWER 5. 01 (c) 534 F
+ or - 5 F.
(1.0)
(b) 306.5 F
+ or - 10 F.
(1.0)
,y (c) Superheated. Steam temperature is above saturation temperature (212 F) at atmospheric pressure.
(1.0)
REFERENCE:
Thermal-Hydraulic Principles and Applications to the PWR, Part II, Chapter 7.
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9 1
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QUESTION 5.02 (3.0)
As a senior reactor operator for unit 2 you are supervising a change in plant status from hot standby to 20*/. power using Operating Procedure OP L-2 (Hot Standby to Minimum Load).
For any calculations used show assumptions and all work.
(a) From the nuclear instrumention, what indication do you expect to see that shows that the reactor is critical?
(1.0)
(b) With all control bank rods inserted the source range instrumentation indicates 20 counts (assume K(eff)=0.935). When control bank D is at zero steps the count rate is observed to be 400 counts. What is K(ef f)
[
when the count rate is 400?
(1.0)
E E
(c) The procedures require that a start up rate of +0.75 DPM be established. How much reactivity is required?
(Assume Beta (
)=0.007 and Lamda (
)=0.1)
(1.0)
E 4
ANSWER 5.02 (a) The reactor is critical (or slightly supercritical) when source range instruments indicate a constant positive start up rate and a steadily increasing count rate with no reactivity being added to the core (no rod motion).
(1.0)
(b) CR1(1-K(eff))=CR2(1-K(eff))2 2O(1-0.935)=400(1-K(EFF))
1.3=400-400xK(eff) 398.7=400xK(eff)
K(eff)=398.7/400 K(eff)=0.9967 (1.0)
(c) SUR=26.06/T since SUR=+0.75 then T=34.75 Sec.
T=(((
/)}gg F
so 34.75= (0.007-/8 ) /O.1/2 :/83.475=0.007 -/
4.5 /8=0.OO7=700PCM d? = 156 PCM (1.0)
THEREFORE /
REFERENCE:
Operating Procedure OP O-2, Hot Standby to Minimum Load Westinghouse,
. Reactor Core Control For Large Pressurized Water Reactors, Pgs. P 9-15 9-17.
QUESTION 5.03 (2.0)
Figure 5.03-1 shows a rapid positive reactivity insertion into an already critical reactor (10-8 amps) at time t=0.
After a stable reator period is reached, an equal and opposite negative reactivity insertion is made, at t=5 minutes, leaving the reactor at the just critical point again. Assume that source neutrons are not significant in the transient, and that the reactivity from all sources is shown.
(a) Show the resulting start up rate as a function of time for the reactivity changes shown.
(1.0)
E^
(b) Show the reactor power level as a function of time for the reactivity changes shown.
(1.0) h-ANSWER 5.03 (a) See attached graph (b) See attached graph
REFERENCE:
Westinghouse, Fundamentals of Nuclear Physics, Chapter 7 P
W r
i i kI
+
6 1
5.03-1 KEY Y
m a-Y t'
- S rime
,e g(Prempt 3kmp)0.25 (stae biod)o.z5 j
ocpo (Stable,Su.R = 0)0.2 (Prom # Dror)o.2s b
(Prompt Drop)o.2.5 3
[
yeSSS10.15
(%ble. Puer)0.25 t
- i W
P
,g.
(&ompt Jump) 0.25 6
i t:6 t5 1ime
#-r-
- - -. - - -,,,g,-+ew a
w
- w---a
.-ww-mw---~e-*e*w-----w
-t-+7w---r-+*--m'nw e=+*
--a7--
m m m -w-e-
--r--
w
- o..
QUESTION 5.04 (2.0)
Xenon and samarium are two isotopes which add negative reactivity to the reactor core.
(a) What are the two production and two removal mechanisms for xenon? (1.0)
(b) What are the two production and one removal mechanism (s) for samarium?
(1.0)
/
ANSWER 5.04
'j' (a)
Xenon i:
k Production (1) Iodine decay (fission daughter product). (0.25)
- t
-1 (2) Direct yield from fission.
(0.25)
Removal (1) Burnout by neutron absorption.
(0.25)
(2) Decay.
(0.25)
(b)
Samarium Production (1) Promethium decay (fission daughter product).
(0.25)
(2) Direct yield from fission.
(0.25)
'?
Removal (1) Burnout by neutron absorption.
(0.5)
.p
[
(no decay)
REFERENCE:
Westinghouse Reactor Core Control For Large Pressurized Water Reactors, Pgs.4-11 thru 4-27 and 4-30 thru 4-34.
I I
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1 L
7 OUESTION 5.05 (1.5)
A station blackout occurs when the reactor is operating at 80% power with a core DELTA T of 48 degrees F and a mass flow rate of 100%. Subsequently natural circulation is established with a core DELTA T of 40 degrees F.
If decay heat is approximatly 2% of full power, what is the mass flow rate (in % of full flow)? Show calculations.
ANSWER 5.05 Q=mCp (delta T)
(Correct formula used)
(0.5) e-2%=$Cp (40) and 80%=100%Cp(48)
Cp=2%/$(40)
Cp=80%/(48)(100) 2%/m(40) 80%/(48)(100)
=
5 9600/3200
=
3%
(1.0)
=
REFERENCE:
Thermal-Hydraulic Principles and Applications to the Freeurized Water.
Reactor, Volume I, Pg. 2-41.
ik Y
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QUESTION 5.06 (2.0)
Figure 5.06-1 is a sketch of Reactor Power vs Time in hours. At time t=0 hours a reactor start up from xenon free conditions to 100% power occurs.
At time t=50 hours a reactor trip occurs followed by a reactor start up to 100% power at time t=65 hours.
(a) Sketch the Xenon reactivity response in the core as a result of this transient, from time t=0 through t=150. State the peak and final steady state xenon negative reactivity. (Show approximate magnitude and duration of each transient.)
(1.0)
(b) Indicate the time in hours that the maximum negative reactivity will 4
be inserted by Xenon.
(0.5)
(c) Indicate the time in hours that the maximum rate of rod insertion will have to occur in order to overcome the Xenon transient.
(Assume a constant Tave and no boration or dilution.)
(0.5)
ANSWER 5.06 (a), (b), (c) see attached Figure 5.06-1
REFERENCE:
Reactor Core Control For Large Pressurized Water Reactors, Pgs. 4-11 Thru 4-27.
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1
- f..%
i 5.06-1 I1TY 100--
n u
M 50 '
2O k
n D
50' 65 100 l'50 Time (Hours)
A Peak Xenon
, Time : 59 to 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (6300 pcm + or - 500 pcm)
Added burnout removal
- Maximum rate of rod insertion
=
(~) "
Time : 65 - 70 Hours nEo b*
D-s i
Equilibrium. Xenon H
.At 10I)% Power s
(-3220 pcm + or - 250 pen) w
.i N4:
M
"(+).
.,- n' 0
l T.
M a
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0 50 65 100 150 Time (Hours)
=
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,n. - - - - -,. - - -,,
QUESTION 5.07 (3.0)
The following is a list of data taken from the pump characteristic curve Figure 5.07-1 which is attached.
DATA (Original N= 1450 RPM)
POINT HEAD FLOW RATE HP (f eet)
(cubic ft/sec) 2.
30 1000 9800 From this data an'd the curve provided generate; 3
(a) volumetric flow rate (1.0)
(
(b) head (1.0)
L (c) break horsepower, (1.0) for the same pump operating at the same efficiency but at a new speed of 1750 RPM for point 2 Show calculations.
ANSWER 5.07 (0.5 pts. for correct formula, 0.5 pts. for correct answer)
N(1) / N(2)= V(1)/V(2) : 1750/1450x1000 = 1207 CFS + or - 30 CFS (1.0) r EN(2)/N(1)]
= H(2)/H(1) : [1750/14503 x30 = 43.7 ft. + or - 2 ft.
(1.0)
_ [N(2)/N(1)3 BHP (2)/ BHP (1) : [1750/14503 x9800 = 17228 Hp + or - 100 Hp
=
(1.0)
(ALTERNATE FORMAT)
Head Flow Rate Hp Point (feet)
(cubic ft/sec) 2.
43.6 1210 17228 (0.5)ea.
Formulas (0.5)ea.
REFERENCE:
Wastinghouse Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor, Pgs. 10-36 thru 10-40.
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6
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5.07-1 s
l e
9 100- -
{.-
80-L t
o 60--
g w
40- -
2 20--
O 4b0 8b0 1200 1600 2000 VOLUMETRIC FLOW RATE p.
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as e
o em o e ma ee o eem m
e~--,--,,,,....nn..,-.----,----.,,-m.--
,,-,.,n-_,.-,,,-.-,
..,-,,-m-
.m.m--n,.,.
,,,,-,.,,---r-
.--,,...,c.--.
QUESTION 5.08 (1.0)
Although the control rods may be positioned axially anywhere in the core, the rods must be above a specified height during reactor operations. This height is refered to as the rod insertion limit. Control rod are required to be maintained above this specified height during reactor operations.
State the three reasons for the control rod insertion limits.
/
ANSWER 5.08 (0.33) pts. each f
1.
Minimi e the consequences of a rod ejection accident, or potential effects of rod misalignment on associated accident analysis are limited.
2.
Quarantee a sufficient shutdown margin from a given power level.
4
- 3. Produce en axial flux distrbution which prevents high local peak power levels, or acceptable power distribution is maintained. (If the control rods are inserted to far into toe core the power production in the core is suppresed in it;p top of the core raising the power production in the bottom of the core. The higher power in the bottom of the core could cause abnormally high fuel temper etures and melt the f uel. )
REFERENCE:
Westinghouse Reactor Core Control for Large Pressured Water Reactors, i
Pgs. 6-29 & 6-30.
J
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QUESTION 5.09 (1.0)
According to the Bases in the Technical Specifcations what four reactor plant parameters are monitored to confirm that departure from nucleate boiling is not occuring?
ANSWER 5.09 RCS PRESSURE (0.25)
RCS TEMPERATURE (0.25)
RCS COOLANT FLOW (0.25)
NUCLEAR POWER (0.25)
REFERENCE:
TECH SPECS, Pgs. 2-6 & 2-7.
1 4
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OUESTION 5.10 (2.0)
The reactor is operating at 100% power (3425 MWit)) as indicated on nuclear power instrumentation.The following data is recorded:
Feedwater Temperature 439 deg F 6
Total Feedwater Flow 15.05x10 LBM/HR Steam Pressure 992 psig Feedwater Enthalpy(hj) 418.5 BTU /LBM I" '
Other System Losses and Gains
-O-From the information provided determine the actual percentage power output of the primary system.
ANSWER 5.10 0 = m delta h SYSTEM ENTHALPY FW ENTHALPY ENTHALPY RISE POWER (BTU /LBM)
(BTU /LBM)
(BTU /LBM) x Total FWFlow h'9 h
h h
BTU /HR 9
f f9 1193 418.5 774.5 11.656x10 (1.0) 1 6
11.656x10 BTU /HR X 1MW /3.413X10 BTU /HR = 0415.5 MW (0.5) 3415.5 MW/3425 MW = 99.7%
(0.5)
REFERENCE:
Westinghouse Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor, Vol.
I, Pg. 6-47.
i i
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m
+
QUESTION 5.11 (1.0)
During power operation hydrogen is added to the primary system and is maintained at 25 - 35 cc/kg of water per Tech Specs. What are two reasons per the design basis f or adding hydrogen to the primary system?
ANSWER 5.11 The design basis for adding hydrogen to the primary system during power 3
operations is: (Any two of the following for full credit) 1.
Remove oxygen from the primary coolant to prevent accelerated corrosion
).
of plant materials. Hydrogen reduces the dissociation of water thus reducing the formation of oxygen.
(0.5) o f
f 2.
Excess hydrogen prevents the formation of nitric acid in the primary coolant if nitrogen is present in the system during operation.
(0.5) i 3.
CAF
REFERENCE:
Rcdiation Chemistry and Corrosion Considerations for Nuclear Power Plant Application Pgs. 7-16 & 7-17.
NOTE: "per Tech. Specs." should be removed from future exams.
t 7
g.
QUESTION 5.12 (1.5)
Lithium hydroxide is added to the primary system for pH control.The Tech Specs indicate that the pH of the primary system be maintained between 4.2-10.5 at 25 C
(a) What are the two reasons per the Design Basis for maintaining the primary coolant system at a pH between4.2-10.5 at 25 C?
(1.0)
(b) What is the reason for adding lithium hydroxide enriched in lithium-7?
(0.5)
ANSWER 5.12 (a)1. Corrosion of metals is reduced in a neutral to slightly higher pH environment.
(0.5)
.g, 4
?
2.
The pH range of 4.2-10.5 tends to minimize the deposition of crud
{-
which fouls heat transfer surfaces and clogs mechanical components.
(0.5)
(b)
Lithium hydroxide enriched with lithium-7 is necessary to avoid tritium production from the interaction of neutrons with the i
lithium-6 isotope.
(0.5)
REFERENCE:
f Westinghouse Radiation, Chemistry and Corrosion Considerations for Nuclear f
Power Plant Application Pgs. 7-16 & 7-17.
u 3
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NOTE: Reference to Tech. Specs. should be omitted from future exams.
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QUESTION 5.13 (2.0)
During a reactor start up near the end of core life, will' the actual critical position (ACP) be HIGHER, LOWER, or THE SAME as the estimated critical position (ECP) if the following conditions occur? (Higher means further out of the reactor core.)
IMPORTANT : Consider each condition separately.
Briefly explain each answer.
?
(a) Actual boron concentration was 75 ppm lower than the value used for the ECP.
(1.0)
(b) The steam dump control pressure set point is raised by 50 psi about twenty minutes prior to criticality.
(1.0) 1 ANSWER 5.13 (a) ACP LOWER ; Boron contributing less negative reativity.
(1.0)
(b) ACP HIGHER p - Raising steam generator pressure raises Tave. MTC and FTC add more negative reactivity.
(1.0)
REFERENCE:
Westinghouse Reactor Core Control for Large Pressurized Water Reactors s
t
,(
5 END OF SECTION FIVE GO ON TO SECTION SIX
SECTION 6 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION OUESTION 6.01 (1.0)
/
The main steam system has check valves as a portion of the main oteam isolation system :
What is the purpose of the main steam line check valves and how do they
{
cccomplish this?
(1.0)
ANSWER 6.01 The main steam line check valves are designed to prevent backflow from the other three steam generators in the event of a main steam line rupture upstream of a main steam isolation valve.
(1.0)
REFERENCE:
Lesson plan C-2A, Main Steam Piping System,.Pg. 13 of 26.
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a OUESTION 6.02 (2.0)
There are two automatic closure signals provided to the main steam isolation valves.
What are the two signals and associated coincidence logic?
ANSWER 6.02
- 1. Hi-Hi containment pressure in 2 of 4 channels (0.5)
?
(22 psig) 2.
Hi steam flow in 2 of 4 steam lines coincident with either J.-
low-low Tave on 2 of 4 channels (less than or equal to 543 F) or low steam line pressure on 2 of 4 channels.
(less than or equal to 600 psig)
(1.5)
REFERENCE:
t Lesson Plan C-2a,pg. 13 of 26.
I
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OUESTION 6.03 (2.0)
Regarding the Main Steam system:
Besides the steam flow rate indications and pressure recorders main steam line flow signals are supplied to FOUR other systems or trips.
What are the four other systems or trips the main steam line flow signal is provided to?
ANSWER 6.03 j.
1.
Safety Injection and steam line isolation (due to high steam line j
flow coincident with low steam line pressure or low-low Tave).
(0.5) 2.
Steam Generator Water Level Control (SGWLC) system.
(0.5) 3.
Low steam generator level coincident with steam flow greater than feed flow.
(0.5) 4.
Main Feedwater pump speed control system.
(0.5)
REFERENCE:
Lesson plan C-2a, pg. 9 of 26.
F-8 e
QUESTION 6.04 (3.0)
The reactor protection system generates reactor trips which protect the reactor from various events. Technical specifications provide the bases for these trips.
For the following reactor trips BRIEFLY list each event the trip is designed to protect the reactor against in accordance with Technical
' Specification Bases.
(a) Power Range Neutron Flux Trip:
High (0.5) i Low (0.5)
(b) Overpower Delta T Trip (0.5) g (c) Pressurizer Pressure Trip:
High (0.5)
Low (0.5) d) Steam Generator Water Level Trip:
Low-Low (0.5)
ANSWER 6.04 (a) Power range neutron flux:
High - Provides protection during power operations to mitigate the consequences of,a reactivity excursion.
(0.5)
Low-Provides protection during subcritical and low power operations to mitigate the consequences of a power excursion.(0.5)
~
i (b) Overpower Delta T:
Limits rod heat flux (kw/ft) (fuel rod integrity under all overpower conditions).
(0.5)
(c) Pressurizer Pressure:
High - Protect tra reactor coolant system boundry from overpressurization.
(0.5)
Low - DNB (protects against low pressure which could lead to boiling or DNB in the core).
(0.5)
(d) The steam generator water level low-low trip protects the reactor from loss of heat sink.
(0.5)
REFERENCE:
Technical Specifications, Pgs. B 2-3 thru B 2-9.
i
QUESTION 6.05-(2.0)
In the main'Feedwater System'there are six main feedwater pump TURBINE trip DEVICES.
4 What are four of the sin DEVICES used to trip the main feedwater pump TURBINE.
ANSWER 6.05 (0.5) pts each Maximum 2.0 points total l
Main Feedwater Pump Turbine Trip devices f
1.
Low vacuum trip device t
- 2. Overspeed trip device
- f
- 3.. Manual trip device v
4.
Stop valve servamotor trip valves 5.
Turbine trip solenoid trip device i
6.
Low bearing oil pressure trip device 7.
Low-Low lube oil reservoir level e
h B.
High Feedwater Header pressure i
- 9. Safety Injection Signal e-
[.
[
-10. Excessive thrust bearing wear 1-
REFERENCE:
Lesson plan C-8c, Pgs. 34-39 of 56
QUESTION 6.06 (4.0)
There are five Emergency Core Cooling System initiation signals which can cetuate the Safety Injection System, one of which is a manual signal using twitch (VB-1 or CC-2). The remaining FOUR are automatic actuation signals.
What are the FOUR automatic Safety Injection signals? Include their logic cnd set point.
ANSWER 6.06 (4.0)
"[
1.
Low pressurizer pressure, 2 of 4 less than 1850 psig.
(1.0) 2.
High containment pressure, 2 of 3 greater than 3.0 psig.
(1.0)
J.
3.
High steam line differential pressure, 2 of 3 detectors on one steam line 100 psi below 2 of the other 3 steam lines.
(1.0) 4.
High steam flow, greater than the programmed set point (40-110%)
on 2 of 4 steam lines, coincident with low-low Tave 543 F or low steam line pressure 600 psig.
(1.0)
REFERENCE:
Lesson plan LB-3 pg. 16 of 66.
F e
1 QUESTION 6.07 (2.0)
The Vibration and Loose Parts Monitoring System is deigned to detect loose parts in the reactor coolant system and provide early warning of impending cechanical failure.
For items a thru d indicate by a true or false whether or not that item is monitored by the vibration and loose parts monitoring system.
Steam Generators /(all four)
(0.5) a.
^
b.
Reactor Coolant Pumps and Motors (0.5) s
- T c.
Pressurizer Internals (0.5) d.
Upper Reactor Vessel (0.5) y ANSWER 6.07 (0.5) each
- c. Steam Generators (all four)
TRUE I
b.
Reactor Coolant Pumps and Motors FALSE 4.
c.
Pressurizer Internals FALSE d.
Upper Reactor Vessel TRUE
- h
REFERENCE:
Lesson Plan LA-2e, pg 7 of 22.
4 1
I l
a 1
i QUESTION 6.08 (2.0) l The pressurizer relief tank receives, contains, and controls radioactive discharges from the reactor coolant system and supporting systems.
(a) Where is the pressurizer relief tank drained to?
(0,5)
I (b) What is the cooling medium for the pressurizer relief tank (if any)?
4 (0.5)
(c) What protects the pressurizer relief tank'from over pressure and where does it relieve to?
(1.0) r ANSWER 6.08 (a) The PRT is drained to the reactor coolant drain tank.
(0.5)
(b) The cooling medium for the PRT is Primary Make Up Water (which is I
sprayed into the PRT).
(0.5) l (c) Rupture Discs protect the PRT from overpressure (0.5), they relieve l
directly to the containment (0.5).
4
REFERENCE:
er Lesson Plan A-4b, Pgs.
6, 7 & B of 22.
$n-w QW N'
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1 i
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OUESTION 6.09 (3.0)
Containment Isolation is part of the engineered safety features of the plant, it is designed to prevent the release of radioactivity to the cnvironment.
What are the six actuating signals for Containment Ventilation Isolation?
ANSWER 6.09 (0.5) each 1.
SI signal (0.5) t p.
2.
Containment Isolation phase A manual actuation.
(0.5)
- 3. Containment Isolation phase B manual actuation.
(0.5)
- 4. High radioactivity above alarm point N
a.
containment air particulate, or RE-11 or RE-12.
(0.5)
- b. plant vent air particulate, or RE-28A or RE-288.
(0.5) c.
plant vent radiogas, or RE-14A or RE-14B.
(0.5)
REFCRENCE:
Lesson Outline LB-6c, pgs.7 & B of 21.
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QUESTION 6.10 (2.0)
The 4BO-volt vital load centers are powered from 4-kV busses 1F, 1G & 1H.
What are the normal sources of power to each of the 4-kV busses supplying the 4BO-volt load centers 1F, 1G & 1H during 1.
Normal Plant Operation 2.
Plant Shutdown
[
- 3. Loss of Offsite Power s.
ANSWER 6.10 f
I
- 1. 25-kV to 4-kV auxiliary transformer 12 (normal source)
(0.5)
E 2.
12-kV to 4-kV startup transformer 12, or auxiliary bus (on a long term basis the plant is normally backfed from the 500 Kv yard through the auxiliary bus).
(0.5)
- 3. Emergency diesel Generators (0.333) a.
No. l1 for bus 1H (0.22)
- b. No. 12 for bus 1G (0.22) j
- c. No. 13 for bus 1F (0.22) f
REFERENCE:
e h
L:sson No. LJ-7, pg. 7 of 28.
NOTE: Future questions should include a request for transformer and or diesel generator numbers.
J
_, -. _. - - - -.,.. - -,. - _ ~,
w.,,-y..-
QUESTION 6.11 (2.0)
Instrument AC distribution panels (PY11, PY11A, PY12, PY14, PY13, & PY13A) cre capable of having power supplied to them from three sources.
Indicate the three power sources avalable to the instrument AC distributio panels. DO NOT list specific component numbers.
ANSWER 6.11 c.
Inverter supplied power from (1.0)
(
h 1.
480-volt vital bus.
L 2.
125-volt dc (distribution panels)
- b. Transformer Voltage Regulator, backup power supplied fromt (1.0) 3.
400-volt vital bus.
REFERENCE:
A Lesson No. LJ-10 pgs.7 & 13 of 29, & Electrical Distribution Overview drawing J-1-1.
(
END OF SECTION SIX i
GO ON TO SECTION SEVEN I
r-4 '
o SECTION 7 s
PROCEDURES NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL r
OUESTION 7.01 (2.5)
Regarding Emergency Boation.
Emergency boration is defined as a flow greater than or equal to 10 gpm'of 20,000 to 22,500 PPM baron or equivalent. Emergency boration is requ, ired for several plant conditions. The conditions listed below require emergency 3j boration.
What is the condition at which emergency boration can be stopped for each of the plant conditions listed?
1.
Control rods are inserted below the low-low insertion limit when critical.
(0,5) 2.
Failure of any two control rods to fully insert following a reator trip, as indicated by rod position indication and rod bottom lights.
(0.5) 3.
Uncontrolled or unexplained reactivity increase or uncontrolled cooldown.
(0.5) 4.
Shutdown margin less than acceptable minimum limits per Tech. Specs.
,l 3.1.1.1, 3.1.1.2, and 3.9.1.
(0.5) i 5.
Inadvertent criticality below rod insertion limits.
(,0. 5 )
~
ANSWER 7.01 1.
Borate until rods are above rod insertion limits.
(0.5) 2.
Borate 100 PPM or 300 gallons baron per stuck rod.
(0.5) 3.
Borate until control regained.
( 0. 5) '
4.
Borate until adequate shutdown margin reached.
(0.5) 5.
Borate 100 PPM or 300 gallons boron.
(0.5)
REFERENCE:
Diablo Canyon 1 & 2 Abnormal Operating Procedure OP AP-6, pgs.
1, 2 & 4.
i
'.,q iI QUESTION 7.02 (1.5) t s
Units 1 & 2 are' operating at 100% of licensed power when the control room is declared uninhabitable b the Shift Supervisor due to fire and smoke.
Only the minimum crew referenced in the Technical Specifications is cvailab1'e to respond at the time of the event.
Whht are the responsibilities of each of the plant personnel listed below?
f 1.
Shift Supervisor (SS)
(0.25) 2.
Senior Licensed Operator (SCO)
(0.25) i 3.
Licensed Operator (CO Unit 1)
(0.25)
-4 3
4.
Licensed Operator (CO Unit 2)
(0.25) 7 5.
Licensed Operator (ACO Unit 1)
(0.25) 6.
Shift Technical Advisor (0.25)
ANSWER 7.02
' 1.
Shift Supervi+3or - supervisor of unit i shutdown (0.25) 2.
Senior Licensed Operator - supervisor of unit 2 shutdown (0.25)
- 3. Licensed Operator - hot shutdown panel operator unit 1 (iO. 25 )
I "4
Licensed Operator - hot shutdown panel operator unit 2 ~
(0.25) 5.
Licensed Operator (ACO) - turbine building watch unit 1 (0.25) 4 6.
Shit Technical Advisor - Conduct initial emergency notification and provide assistance to either unit as needed.
(May be performed at the TSC.)
(0.25)
REFERENCE:
Diablo Canyon 1 & 2 Operating Procedure OP AP-8, pgs. 1 & 2.
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OUEST10N 7.03 (2.5)
Technical Specifications state that Reactor Coolant leakage shall be limited to specific volumetric flow rates.
Define the following reactor coolant system leakage limits.
(a) Unidentified Leakage (0.5)
(b) Reactor to Secondary leakage (0.5)
(c) Identified leakage (1.0)
(d) Controlled leakage (0.5) i.
ANSWER 7.03 (a) All leakage which is not controlled or identified leakage, or 1 GPM.
(0.5)
(b) Reactor to Secondary leakage is any leakage from the reactor coolant system to the secondary cooling system through the steam generators, or, 1 GPM through all S/G's or 500 GPD through 1 S/G.
(0.5)
(c)
- 1. leakage, except controlled leakage, into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collectig tank, or (0.33)
- 2. leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakag'e detection systems or not to be PRESSURE BOUNDRY leakage, or (0.33)
- 3. reactor coolant system leakage through a steam generator to the secondary coolant system.
(0.33) 10 GPH for (c) 1.,
2.,
3.
(d) Controlled leakage is that seal water flow supplied to the reactor coolant pump seals, 40 GPM.
(0.5)
REFERENCE:
Tcchnical Specification : Reactor Coolant System Operational Leakage pg.
3/4 4-19.
QUESTION 7.04 (3.0)
Standards For Protection Against Radiation 10 CFR 20 specifies radiation dose standards for individuals working in radiation areas.
(a) What are the radiation dose standards for a Calendar Quarter?
(1.5)
(b) There are three requirements that must be met if the whole body limits for a calendar quarter are to be exceeded.
What are the three requirements?
(1.5)
ANSWER 7.04 (a)
(0.5) points for each of the following:
1.
1.25 REM - Whole Body (head and trunk; active blood forming organs; i
lens of the eyes; or gonads).
J 2.
18.75 REM - Extremities (hands and forearms; feet and ankles).
3.
7.5 REM - Skin Dose (of whole body).
(b) (0.5) points for each of the following:
1.
3 REM per calendar quarter.
2.
5(N - 18) total accumulated dose to the whole body where N is the I
individual's age in years at last birthday.
E f-3.
Form NRC - 4, or equivalent.
! p
REFERENCE:
-. =
QUESTION 7.05 (2.0)
In accordance with 10 CFR 20, Standards For Protection Against Radiation.
(a) What is the definition of a radiation area?
(1.0)
(b) What is the definition of a high radiation area?
(1.0)
ANSWER 7.05 (a) Radiation Area - Area (accessible to personnel) where the major part of
~
the body could receives i
5 mrem in one hour (0.5)
'[
100 mrem in 5 days (0.5) 7 (b) Area (accessible to personnel) where major parts of the body could receive :
100 mrem in one hour (1.0)
REFERENCE:
10 CFR 20 g.
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QUESTION 7.06 (1.5)
The Component Cooling Water System will normally be in continuous service.
In order to assure this, the component cooling water system has been designed with two direct sources of makeup water and one indirect.
What are the sources of makeup water to the component cooling water system?
ANSWER 7.06 (0.75) each t
j 1.
Transfer Tank (via two 250 gpm transfer pumps (tank 150,000 gal))
2.
Primary Water Storage tank (via two 150 gpm pumps (tank 2000,000 gal))
(v t,
REFERENCE:
Diablo Canyon 1 & 2 Operating Procedure OP F-2:1, pg. 2 of 5.
NOTE: Future questions should only request two direct sources of makeup water since the indirect source, seawater evaporator, is not nor ever has been a working component of the power plant.
2 2
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QUESTION 7.07 (1.0)
There are TWO individuals whose signature must be on a discharge permit in order to have an authorized discharge from the plant.
Who are the TWO individuals (by title) who must sign the discharge permit?
ANSWER 7.07 1.
Shift Foreman (SFM)
(0.5) 2.
Chemical and Radiation Department Engineer or chemical and radiation p
protection supervisor (CARP).
(0.5)
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REFERENCE:
i e
Diablo Canyon Lesson LG-2, pgs. 52 & 53 of 61.
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QUESTION 7.08 (3.0)
Technical Specifications limit the the radioactivity level i n each Gaseous Radwaste system gas decay tank.
(a) What is the noble gas limit for radioactivity level in the gas decay tank as stated i n the Technical Specifications?
(1.0)
(b) When does this limit apply?
(1.0)
/
(c) What steps are taken if the limit is exceeded?
(1.0)
ANSWER 7.08 (a) Radioactivity in each gas decay tank i s limited to 10 curies noble gasses.
(1.0)
(b) The limit applies at all times.
(1.0)
(c) If the limit is exceeded, all additions to the gaseous radwaste tank are to be suspended and tank contents reduced to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
(1.0)
REFERENCE:
Technical Specification 3.11.2.6 & Lesson No. LG-2 pg.56 of 61.
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QUESTION 7.09 (2.0)
Unit 2 is in mode 5 with spent fuel being moved in the spent fuel pool. As the senior reactor operator you are informed that due to a malfunction in the Fuel Handling Building Ventilation System (FHBVS) only one FHBVS system is operable.
(a) Can fuel movement within the spent fuel pool or crane operations with loads over the spent fuel pool continue?
(0.25)
(b) If it can continue, what provision must be met? If not provide justification.
(1.75) j.
ANSWER 7.09 (a) YES (with only one FHBVS operable movement of spent fuel within the spent fuel pool and crane operations with loads over the spent fuel pool may continue).
(0.25)
(b). Fuel handling within the spent fuel pool or crane operation with loads over the spent fuel pool may proceed provided the cperable FHBVS is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.
(1.75)
REFERENCE:
[
Lesson LG-3, pg. 22 of 27 & Technical Specification 3/4.9.12
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OUESTION 7.10 (2.0)
The Component Cooling Water system loop A provides cooling water to seven vital loads.(List for unit 2 only.)
List FOUR of the seven vital loads on loop A of the component cooling water cystem.(List for unit 2 only.)
ANSWER 7.10 (0.5 points for each complete answer up to 2.0) 1.
CCW pumps 2-2 & 2-3 stuffing box and lube oil cooler.
2.
Residual Heat Removal heat exchanger 2-1.
3.
Residual Heat Removal pump 2-1 seal water cooler.
4.
Centrifigal Charging pump 2-1 seal lube oil and gear coolers.
5.
Safety Injection pump 2-2 seal water and lube oil coolers.
6.
Containment Fan Cooler units 2-3 & 2-4.
7.
Post-LOCA Sampling coolers.
REFERENCE:
Lesson LF-2 pg. 21 of 40.
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QUESTION 7.11 (3.0)
There are three Technical Specification conditions for considering the Fire Water System Operable.
What are the three conditions?
ANSWER 7.11 1.
Both (high pressure) fire water pumps must be operable - each with a capacity of 2250 gpm.
(1.0) 2.
Suction from fire water tank containing a minimum usable volume of 270,000 gallons.
(1.0) 3.
Pump discharge aligned to the fire suppression header.
(1.0)
REFERENCE:
Lesson LK-2c
, pg. 19 of 23.
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QUESTION 7.12 (1.0)
~
There are five diesel generator sets designed to supply power to vital and non-vital safety related loads in the event of a loss of offsite power. Onq of the diesel generator sets (DG 1-3) is a shared unit which can supply power to either unit 1 or 2.
(a) If DG 1-3 were in service on unit 2's Bus F, but not as a result of a Safety Injection signal on unit 2, what would happen to DG 1-3's output breaker s in the event of a saf ety injection signal on unit 17 (0.5)
F (b) If DG 1-3 was responding to a safety injection signal on unit 2 and a safety injection signal was received on unit 1 3
&-f What must be done-in order for DG 1-3 to shift its power to unit 17
[
(0.5)
E ANSWER 7.12 (a) The DG 1-3 breaker to unit 2 would trip to allow unit 1 breaker to close.
(0.5)
(b) DG 1-3 would remain commited to unit 2, until the unit 2 safety J
injection signal is cleared (reset) then the DG 1-3 loading would be 4
shifted to unit 1,
if needed.
(0.5)
!Yj(,
REFERENCE:
't lf
, Diablo Canyon Lesson No. LJ-6b, pg. 41 of 51.
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END OF SECTION SEVEN l
GO ON TO SECTION EIGHT 6
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SECTION 8 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.01 OUESTION (2.0)
R2garding sealed valves and the associated Nuclear Plant Administrative Procedures requirements:
(a)
During an initial verification how does an operator verify a sealed valve shut ?
(0.5) i (b)
Both caution tags and seals are used to arsure non-operation of manual valves in operating systems. When must seals be used ? (as opposed to caution tags)
(0.5)
(c)
Normally valve position is independently verified by two qualified operators who do not accompany each other. Describe two situations where it is desirable for the two operators checking a valve position to accompany each other.
(1.0)
ANSWERS:
(c)
Attempt to move the handwheel or operator in the close direction. If the valve is in the correct position, no motion will occur.
(0.5)
/
(b)
For manual valves that are important to the proper functioning of safeguards systems, a seal shall be used rather than a caution tag to assure non-operation.
10.5) i y _ (c) 1.
When such action will result in a significant overall reduction in personnel radiation exposure.
(0.5) 2.
When the benifits of having the verifier accompany the worker substantially outweigh the reduction in independence; le. where mis-positioning could cause a safety injection or SCRAM.
(0.5)
REFERENCE; Nuclear Plant Administrative Procedures, NPAP C-104, 4.6.2; C-9, 1.2; and C-9S1, 5.
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8.02 OUESTION (2.0)
RIgarding Technical Specification requirements for refueling operations:
(c)
When must direct communications be maintained between the control room and personnel at the refueling station ?
(0.5)
(b)
During movement of irradiated fuel inside containment, what are the three required states for containment penetrations ?
(1.5)
/
ANSWERS:
b['
(a) During core alterations.
(0.5)
E.
(b) 1.
The equipment door is closed and held in place by a minimum of four bolts.
(0.5)
\\
2.
A minimum of one door in each airlock is closed, (0.5)
[
3.
Each penetration providing direct access to the outside atmosphere shall be either closed, or capable of being closed by an automatic operable containment ventilation isolation valve.
(0.5)
REFERENCE:
Technical Specifications 3/4.9.4 and 3/4.9.5 m.
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L 8.03 OUES TION (3.0)
In accordance with the Technical Specifications for Administrative Controls, Section 6.2.2 " Plant Staff":
(a)
What is the composition of the minimum Shift Crew with Unit 1 in mode 6 (refueling being conducted) and Unit 2 in mode 1 7 (Include all operations and support personnel required by the Technical Specifications.)
(1.0)
(b)
When can the STA position be unmanned 7 (Assume Unit 1 is in mode
(
one and Unit 2 is in mode 5.)
(0.5)
(c)
What allowance is made in Shift Crew composition to accommodate unexpected absence of on duty Shift Crew members ?
(1.0)
(d)
Who may not be assigned to the Site Fire Brigade ?
(0.5)
ANSWERS:
(A)
(0.125 each correct item) 1 SS, 1 SOL, 3 OL, 3 AD, 1 STA (unless SS or SQL meets STA Rqmts.)
1 SQL (observing fuel movements and core alterations) 1 H.P.
Technician l
5 members of the Site Fire Brigade.
l (b)
The STA position need not be manned when the person filling the SS or
{
SOL position meets the NRC STA qualification requirements.
(0.5)
- (c)
The Shift Crew composition may be less than the minimum requirements ig for a period of time not to exceed two hours in order to accomodate unexpected absence of on duty shift personnel, provided immediate action is taken to restore the shift composition to the minimum requirements.
(1.0)
(d)
The Fire Brigade shall not include the Shift Supervisor and the two other members of the minimum Shift Crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.
(0.5)
REFERENCE:
TGehnical Specification 6.2.2
/
8.04 OUESTION (2.0)
After a reactor trip, the Shift Supervisor may initiate a reactor startup cnd return to power if a reactor safety limit was not exceeded, there is no indication of core damage, and: (list f our additional requirements).
ANSWERS:
(0.5 each, up to 2.0) 1.
The cause of the reactor trip is adequately known, 2.
The reactor trip review sheet is reviewed and evaluated by the Shift Forman, STA, and plant engineer 3.
The Shift Foreman feels the reactor can safely be returned to power 4.
The verbal approval of the Plant Superintendent (or his delegate) is given.
5.
The reactor trip review sheet is completed 6.
Personal statements are obtained from the SS(SFM), SOL (SCO)and SCO (CO)on watch during the trip
REFERENCE:
Administrative Procedure, AP A-100S1 hI l
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m 0.05 QUESTION (2.0)
R2garding Surveillance Testing:
(a)
According to Technical Specifications what is a " STAGGERED TEST BASIS" ?
(1,o)
(b)
If a required surveillance test is not completed within the specified time interval, what extension of time is allowed to conihlete the test, and what additional limitation is placed on subsequent tests 7 (1.0)
.p.
ANSWERS:
(a)
A staggered test basis consists of 1.
A test schedule for n systems or components obtained by dividing the specified test interval into n equal subdivisions.
2.
The testing of one system or component at the beginning of each subinterval.
(1.0)
(b)
The extension is 25 % of the surveillance interval, but the combined time interval between any three consecutive tests may not exceed 3.25 times the surveillance interval.
(1.0)
REFERENCE:
Technical Specifications, Definitions 1.36, and LCO 4.0.2 4
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8.06 OUESTION (3.0)
The plant is in mode 3, preparing for a routine plant startup. Consider each of the following events one at a time.
(a)
A portion of the fire supression spray / sprinkler system is found to be inoperable for a portion of the system that protects redundant safety realated equipment. What actions are required ?
(1.5)
(b)
One centrifugal charging pump is declared inoperable, however the parts are on-site and the maintenance supervisor informs you that repairs can be completed in less thant 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Can the startup
)
proceed ? Explain your answer. (Technical Specification attached)
['
(1.5)
ANSWERS:
(a)
A continuous fire watch with backup fire supression equipment must be in the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
(1.5)
(b)
A charging pump out of service places you in an action statement. The startup cannot proceed unless the LCO is met. Entry into an operating mode is not allowed if an action statement must be relied on to do so (per T.S.
3.0.4)
(1.5)
REFERENCE:
[
Tcchnical Specifications, LCO 3.0.4, and 3/4.5.2 and 3/4.7.9 s
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EMERGENCY CORE COOLING SYSTEMS k
3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350'F avo LIMITING CONDITION FOR OPERATION 3.5.2 Two Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
..0ne OPERABLE centrifugal charging pump, a.
b.,
One OPERABLE Sa'ety Injection pump,'
One OPERABLE desidual Heat Removal heat exchanger, c.
d.
One OPERABLE Residual Heat Removal pump, and An OPERABLE flow path capable of taking suction from the Refueling e.
Water Storage Tank on a Safety Injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.
' APPLICAEILITY:
MODES 1, 2, and 3.
(
ACTION:
With one ECCS subsyst'em inoperable, restore the inoperable subsystem a.
L-to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least NOT. SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances 'of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
e
.L DIABLO CANYON - UNITS 1 & 2 3/4 5-3 t
EMERGENCY CORE COOLING SYSTEMS t
$URVEILLANCE REQUIREMENTS
(
-)
4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
At least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves a.
are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position 8703
[
RHR to RCS Hot Legs Closed 8802A Safety Ir.jection Closed 88028 to RCS Hot Legs Safety Injection Closed 8809A to RCS Hot Legs RHR to RCS Cold Open Legs 88098
't RHR to RCS Cold Open Legs 8835 Safety Injection Open 8974A to RCS Cold Legs Safety Injection Open Pump Recir. to RWST 8974B Safety Injection Open kj Pump Recir. to RWST -
- i 8976 RWST to Safety Open Injection Pumps 8980 RWST to RHR Pumps Open 8982A-Containment Sump to Closed
=
RHR 89828 Containment Sump to Closed RHR 8992 5 pray Additive Tank Open
At least onca per 31 days by:
1)
Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2)
Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
)
DIABLO CANYON - UNITS 1 & 2 3/4 5-4
e EMERGENCY CORE COOLING SYSTEMS
-SURVEILLANCE REOUIREMENTS (Continued)
By a visual inspection which verifies that no loose debris (rags
.c.
trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.
This visual inspection shall be performed:
1)
CONTAINMENT INTEGRITY, andFor all accessible area
..- 2)
Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is esta d.
At least once per 18 months by a visual inspection of the containme sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, e show no evidence of structural distress or corrosion; At least once per 18 months by:
e.
1)
. Verifying that each automatic valve in the flow path actuates
.f signal.to its correct position on a Safety Injection actuation test 2)
Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a)
Centrifugal charging pump, b)
Safety Injection pump, and c)
Residual. Heat Removal pump.
l' 1
f.
differential pressure on recirculation flow when te Specification 4.0.5:
1)
Centrifugal charging pump 1 2400 psid, 2)
Safety Injection pump 1 1455 psid, and 3)
Residual Heat Removal pump 1 165 psid.
L DIABLO CANYON - UNITS 1 & 2 3/4 5-5 1
EMERGENCY CORE Cf LING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
\\
By verifying the correct position of each electrical and/or mechanical g.
position stop for the following ECCS throttle valves:
1) l Within 4 hou:s following completion of each valve stroking opera-tion or maintenance on the valve when the ECCS subsystems are re-i quired to be OPERABLE, and
,2)
At least once per 18 months.
j
..s..
BoronInjection i,
. Throttle Valves Safety Injection y
Throttle Valves 8810A 8822A 88108 8822B 8810C 8822C 88100 8822D h.
By performing a flow balance test, during shutdown, following comple-tion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
1)
For centrifugal charging pump lines,'with a. single pump running:
a)
The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 346 gpm, and b)
The total pump flow rate is less than or equal to 550 gpm.
2)
For safety injection pump lines, with a single pump running:
~
a)
The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 463 gpm, and b)
The total pump flow rate is less than or equal to 650 gpm.
1.
By performing a flow test, during shutdown, following completion of modifications to the RHR system that alter the system flow character-istics, and verifying that with a single pump running, and delivering to all four cold legs, a total flow rate greater than or equal to 3976 gps.
3 DIABLO CANYON - UNITS 1 & 2 3/4 5-6
{I
8.07 OUESTION (1.5)
What are the three Technical Specification requirements for making tcmporary changes to procedures ?
ANSWER:
(0.5 each up to 1.5 total)
- 1. The intent of the original porcedure is not altered.
2.
The change is approved by two members of the plant management staff, at
[
Icast one of whom holds a SRO license on the affected unit (s).
- 3. The procedure change is documented, reviewed by the PSRC, and approved 1
by the plant manager within 14 days of implementation.
REFERENCE:
TGchnical Specification 6.8.3 b
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8.08 QUESTION (1.5)
According to 10 CFR 50.59 under what three circumstances would a temporary codification involve an unreviewed safety question ?
ANSWER:
(0.5 each) 1.
If the probability of occurance or the consequences of an accident or g
the malfunction of equipment important to safety previously evaluated in i
the Safety analysis Report may be increased. or Y
I, 2.
If the possibility of an accident or malfunction of a different type
{
than any evaluated previously in the Safety Analysis Report may be created.
or 3.
If the margin of safety as defined in the basis for any Technical Specification is reduced.
REFERENCE:
NOTE: The question states " temporary modification" the wording should have l,'
been " proposed change, test or experiment" or a paraphrasing to
[
unreviewed saf ety questions. These items are not considered as part 5
of the operators job. Therefore, the responses should be reviewed I
with the job responsibility as a consideration, t
.t-e:
s 8.09 OUESTION (3.0)
Ecch accumulator is' required by Technical Specification 3/4.5.1 to be demonstrated operable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying borated water volume, nitrogen cover-pressure, and accumulator isolation valve position.
Some extensions of the basic intervals are allowed by the Technical Specifications.
Records show that verifications were completed on:
December 3 at 0000 December 3 at 1500 7
December 4 at 0400 December 4 at 1600 December 5 at 0500 (a)
Identify any violations which may have occured.
(1.0)
("
(b)
When is the latest that the next surveillance can be done 7 (after Dec. 5 0500)
(1.0)
(c)
If the pressure is found to be below the specified limits and the nitrogen fill valve will not operate, what operator action is required 7 (1.0)
ANSWERS:
(a)
On December 4 at 1500 there were 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> (3.25 x 12) for the last three intervals. The 1600 verification was one hour late.
(1.0)
I (b)
December 5 at 1900 (1.0)
(c)
Commence shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
(1.0)
REFERENCE:
TGehnical Specifications, LCO 4.0.2, and 3/4.5.1.
(
8.10 QUESTION (2.0)
The Code of Federal Regulations, 10 CFR 55, defines the general provisions of Operator's licenses. In accordance with these regulations.
(a)
What are the " Controls" of a nuclear facility ?
(1.0)
(b)
Who may operate the " Controls" of a nuclear facility without an Operator'slicense7 (1.0)
ANSWERS:
s-(a)
Controls are apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.
(1.0)
.5; (b)
An individual may manipulate the controls as part of his training to qualify for an operator license under the direction and in the presence of a licensed operator or senior operator.
(1.0)
REFERENCE:
e b _.
8.11 OUESTION (3.0)
The CODE OF FEDERAL REGULATIONS 10 CFR 50.72 AND 10 CFR 20.503 requires icmediate notification (within a period of one hour) to the NRC Operations Csnter via the Emergency Notification System for various events.
Which of the events listed below require immediate notification ? (0.5 ccch)
~
(c)
The facility receives warning of a tsunami wave that could damage the intake structure.
{
(b)
A worker receives an e>:posure of 250 Rem to his hand.
(c)
Damage to property on the site of $100,000.
(d)
Declaration of an Unusual Event at Unit 2.
(a)
A reactor trip followed by depressurization to 1700 Psig without safety injection caused by a stuck open PORV.
(f)
Due to high winds one meterological tower is lost. No other threat to the plant is anticapated.
ANSWERS:
(0.5 each)
'{
(c) yes
- (b) no (C) no (d) yes (e) yes (f) no
REFERENCE:
10 CFR 50.72 and 20.403 f~. ' '
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END OF SECTION EIGHT END OF EXAMINATION