ML20214K706
| ML20214K706 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/07/1986 |
| From: | Munro J, David Nelson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214K700 | List: |
| References | |
| 50-400-OL-86-02, 50-400-OL-86-2, NUDOCS 8612020425 | |
| Download: ML20214K706 (160) | |
Text
P ENCLOSURE 1 EXAMINATION REPORT 400/0L-86-02 Facility Licensee:
Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Facility Name:
Shearon Harris Nuclear Power Plant Facility Docket No.:
50-400 Written examinations were administered at McKimmon Center, Raleigh, North Carolina.
Simulator and oral examinations were administered at the Shearon Harris Energy and Environmental Center near New Hill, North Carolina.
/f t/- 1-rb Chief Examiner:
DavidJ.Neyson Date Signed o
N
- C. /
- /7 /74 Approved b.
~
Munr @ ct) (g Section Chief Date Signed dohn Summary:
Written examinations were administered on February 24, 1986, to 36 candidates, 35 of whom passed.
Simulator and Oral examinations were administered the weeks of May 19 and May 26, 1986, to 34 candidates, 23 of whom passed. One candidate was administered an oral re-examination and simulator exam of which he passed the oral and failed the simulator.
Two candidates withdrew after passing the written exam.
One candidate who did not pass the written exam did not take the cral and simulator exams.
The utility requested that 43 questions on the written examination be adjusted /
modified. NRC resolution of the utility comments showed that 13 (30.2%) of these resolutions were needed due to inaccurate / incomplete utility training materials.
Based on the results described above, 9 of 11 R0s passed and 14 of 24 SR0s passed.
Eh1202042S 861112 PI;p ADOCK 0500 O
V
-0 REPORT DETAILS 1.
Facility Employees Contacted:
~
- J. Harness, Assistant Plant Manager
- J. Willis, Plant General Manager
- J. Collins, Manager, Operations
- A. Powell, Manager of Training
- C. Olexic, Director of Nuclear and Simulator. Training
- W. Giess, Project Specialist, Training
- M. Wallace, Regulatory Compliance
- G. Blinde, Senior Training Specialist
- Attended Exit Meeting-2.
Examiners:
- D. J. Nelson, NRC J.-A. Arildsen, NRC
- C. A. Casto, NRC A. J. Vinnola,'EG&G.
T. Rogers, NRC W. Hemming, EG&G W. M. Dean, NRC F. Jaggar, EG&G L. J. Defferding, PNL R. E. Schreiber, PNL
- Chief Examiner, Written
- Chief Examiner, Oral and Simulator 3.
Examination Review Meeting At the conclusion of the written examinations, the' examiners provided a copy of the written examination and answer key.for review. The comments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.
4.
.NRC Written Examination, Comment Resolutions a.
R0 Exam (Applicable SRO Exam Questions are in parenthesis)
(1) Question 1.05:
Only " Redistribution" will be accepted in (5.16) place of " Void".
" Pressure" coefficient is identified in the referenced material as being
"... neglected because pressure changes cause insignificant changes in moderator density."
Additionally, the question addresses the
" power coefficient"; whereas the " pressure coefficient" contribution to the moderator i
coefficient is not power dependent.
1
N f
2
'(2) Question 1.15:
Accepted " Checkerboard-loading pattern for (5.15) lower enrichment fuel towards center of core" as an acceptable' replacement for answer number two based on the additional reference material provided by the facility.
(3). Question 1.21:
Accepted " rod misalignment" for half credit.as (5.20) replacement for answer number one based on the.
reference identified by the facility.
An ejected rod constitutes a. misalignment; i
l however, all rod misalignments will not be violations of rod-insertion limits.
Therefore, l
only half credit will be given.
(4) Question 1.22:
Accepted " Samarium" as a correct response based on additional reference-material provided by the facility.
(5) Question 2.2:
Accepted " loss of power to associated safety-bus" or " Blackout" as additional correct responses for answer number two based on the-i facility comment and additianal reference l
material provided.
l l
(6) Question 1!. 6:
The question specifically addresses "uninter-ruptible" power supplies.
The alignment described in the facility comment is not considered part of the uninterruptible portion of the system because mechanically interlocked l_
breakers must be re positioned.
l l
A schematic equivalent to the verbal descrip-tion in the answer key is an acceptable response.
l (7) Question 2.8b:
Accepted based on facility comment and updated reference material provided.
(8) Question 2.10:
Accepted "non-emergency" conditions if this assumption was stated as part of the response,
~dditional reference material based on a
identified by the facility. Setpoints are not-required for full credit.
(
(9) Question 2.15c:
Question deleted based on the facility i
comment.
The sectional point value is adjusted accordingly.
i-t I-
r-n-
3
_(10) Question 2.17:
Due to. the ambiguous wording o'f the question,
- the first half of the answer is not required; however, the point value for _the second half remains at one point, thereby reducing the total point value of the question.
The sectional point value is adjusted accordingly.-
(11) Question 2.19:
Question deleted based -on the facility comment.
The sectional point value is adjusted accordingly.
(12) Question 3.04:
Set points changed based on updated reference material provided by the. facility.
(13) Question 3.08a:
Accepted facility comment based. on further
- review of reference material.
(14) Question 3.09:
Setpoint changed based on updated reference (6.06)-
material provided by facility.
'(15) Question 3.11:
Question deleted based on facility comment (6.05) and confusing implication in the question wording.
The sectional point value is adjusted accordingly.
(16) Question 3.16:
Setpoints are not solicited or required for full credit.
(17) Question 4.04:
Accepted based on further review of the reference caterial.
(18) Question 4.18:
Either temperature is accepted based on the (7.12) facility comment and the close proximity of the two values.
b.
SRO Exam (1) Question 5.03:
Further review of the question choices indicates both "a"
and "b"
are correct responses.
(2) Question 5.08:
The following are accepted as correct responses based on the reference material-identified by the facility:
core cooling; thermal stress; thermal shock; and decay heat removal.
(3) Question 5.10:
Both "a"
and "b"
are accepted as correct responses based on the error in the reference material identified by the facility.
~'
4 (4) Question 5.13:
The assumed operational mode was not specified in the question; therefore, the mode dependent variations in answers-are accepted on the condition that the "Tavg" and " accident" choices are consistent with respect to modes.
This is based on the non-specific wording of the question.
(5) Question 6.02a:
Either "N44 Nuclear Power" or " Anticipated Steam Demand" is accepted for full credit based on further review of reference material.
(6) Question 6.02b:
Based on the NRC review, the answer key has been changed to accept " Steam Pressure" for
[+0.33] and " Flow Error" for [+0.67].
" Steam Flow" and " Feed Flow" are acceptable replace-ments at [+0.33] each for " Flow Error".
" Density Compensation" is not required for full credit.
(7) Question 6.09:
Accepted the additional answers based on additional reference material provided by the facility.
(8) Question 6.13:
The NRC considers that this question does not challenge the candidates' understanding of the Post Accident Hydrogen System, and therefore is deleted.
The sectional point value is adjusted accordingly.
(9) Question 6.20:
Accepted additional answer based on reference material identified by facility.
_(10) Question 7.03:
The NRC considers that these actions steps could likely take place simultaneously; therefore, this question has been deleted.
The sectional point value has been adjusted accordingly.
(11) Question 7.05:
The facility comments about pressurizer level is irrelevant to the question.
The NRC considers that this question tests conceptual knowledge of procedures, not step memoriza-tion; hence, the question and answer remain unchanged.
(12) Question 7.07:
The NRC considers that this question tests conceptual knowledge of procedures, not step memorization; hence, the question and answer remain unchanged.
5
- (13)' Question 7.10:
Based on' the non-specific. nature of the question, " Manual From the Board," "BTRS," and "RWST" are additional acceptable answers for
-[0.5] points each.
(14) Question 7.12:
Accepted recommended answers; however, - the question's -point value was reduced to 0.5 points based on the reduced detail required for full credit.
The sectional point value has been adjusted accordingly.
(15) Question 7.15:
Based on further NRC review, " Seal failure" is -
acceptable -for full-credit; however, the intent of ~ the question was to have the candidates explain
(" diagnose") the given information.
Therefore, the question point value is reduced to [0.5].to be consistent with other short answer questions.
The sectional point 'value has been adjusted accordingly.
(16) Question 7.16:
-Based on NRC review other appro'priate responses for " Containment symptoms" are accepted as correct.
(17) -Question 7.19c:
Accepted based on further review of reference material.
i (18) Question 7.22:
Repeat question has been deleted.
(19) Question 7.24:
Based on NRC review, the portion of the answers pertaining to the " upper head" being
" full" is deleted and question point value reduced accordingly to [1.0] points.
Half credit is awarded to each of "PZR level" and "subcooling".
The sectional point value has been adjusted accordingly.
(20) Question 8.03c:
Numerical value changed due to calculational l
error.
l (21) Question 8.06:
Accepted based on further NRC review of referenced material.
(22) Question 8.13:
Accepted based on additional reference i
material provided by the facility.
($3) Question 8.16:
The limit values have been deleted from the question; however, the " upper" or " lower" l
limit response is still required.
The l
question's point value was reduced to [1.0]
l point to reflect the deletions. The sectional point value has been adjusted accordingly.
t l
l
_____.,..J
6 (24) Question 8.19:
Accepted based on the facility reference material in error referencing an outdated paragraph in 10 CFR 50.
(25) Question 8.20:
An additional response pertaining to HP or Security involvement with " routine" entry is acceptable based on the vagueness of the question.
5.
Additional Changes to Written Exam / Answer Key Made as a Result of NRC Review Process a.
R0 Exam (1) Question 1.09c:
Due to confused wording, question reworded (5.09c) by the proctor during the exam.
(2) Question 1.19:
" Fuel Temperature Coefficient" was accepted as a replacement for " Doppler Coefficient."
(3) Question 4.10:
Accepted " reduction of power" as a replacement for answer number two.
(4) Question 4.13:
Typos corrected:
" sump" vs. " pump".
b.
SRO Exam Question 6.18:
Due to the vagueness of the question the Auxiliary Feedwater System automatic start conditions were also accepted for full credit.
6.
Exit Meeting At the conclusion of the oral and simulator exams, the examiners met with representatives of the plant staff. The following generic weaknesses were discussed:
The SR0 candidates' apparent lack of simulator training time.
The candidates' infamiliarity with logic diagrams.
The candidates' knowledge level of the AFW speed control system.
Also discussed was the absence of Annunciator Panel Procedures for the
" Axial Flux Alarm", "RHR Trouble," and "AFW Trouble" annunciators.
The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
l
(
I
I O
A:1 V
3 U. S. NUCLEAR REGULATORY C0f441SSION l
REACTOR OPERATOR LICENSE EXAMINATION J
Facility:
Harris 1 t
3 Reactor Type:
Westinghouse - PWR Date Administered: February 24, 1986 a
~
Examiner:
R. E. Schreiber e:
Candidate:
ANSWER KEY t
t INSTRUCTIONS TO CANDIDATE:
Use. separate paper for the answers.
Write answers on one side only..
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses af ter the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours af ter the examination starts.
Category
% of Candidate's
% of Value Total Score Cat. Value Category 30 25,3
- 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow R !.1' LY. I W
_.28
- 2. Plant Design Including Safety and Emergency Systems 30 d5.5
- 3. Instruments and Controls 30 25.5
- 4. Procedures: Normal, Abnormal, Emergency, and Radiological Control ItC.T
~120 TOTALS Final Grade All work done on this examination is my own; I have neither given nor received aid.
Candidate's Signature s
i e
L
(
.J
.o Page 1 Harris 1 February 24, 1986 Points Available i
1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERM 0DYNAMICL.
j HEAT TRANSFER AND FLUID FLOW (30.0)
OUESTION 1.01 f
Describe how (and why) the following will respond to a loss of natural circulation flow following a reactor trip from 100%
power equilibrium conditions.
j a.
RCS wide range temperature difference (AT)
(1.0) b.
Relationship between Tcold and P (1.0) steam ANSWER 1.01 a.
AT temperature increases [+0.7] as T increases (as hot boiling occurs in core) and T remains relatively cold constant [+0.3]. Ing 4. Told ld *kd'# i0 # > C0*#$) "A*4*
b.
T will n t follow Psteam [+0.7].
P will decrease cold steam (as boiloff occurs in S/G) while T remains relatively cold constant [+0.3].
Reference (s) 1.01 1.
General Physics, HT&FF, pp. 356-357.
2.
SHNPP FF-H0-1.3, p. 5.
I i
i
-Section 1.0 Continued on Next Page-l
{
l i
(
J
-+'
o Page 2 Harris 1 February 24, 1986 Points Available OUESTION 1.02 The plant is operating at 30% power, rod control in manual, turbine control in automatic, when loop 1 RCP trips. Assuming no reactor trip or operator actions, indicate whether the following parameters will be HIGHER or LOWER at the end of the transient compared to their initial values.
a.
f2 S/G pressure (0.5) b.
- 3 RCS loop flow (0.5) c.
T in loop 1 (0.5) c d.
Th in lo p 2 (0.5)
ANSWER 1.02 a.
LOWER [+0.5]
b..
HIGHER [+0.5]
c.
LOWER [+0.5]
d.
HIGHER [+0.5]
Reference (s) 1.02 1.
SHNPP, Heat Transfer and Fluid Flow Lesson Plans (Integrated Knowledge).
-Section 1.0 Continued on Next Page-
o Page 3 Harris 1 February 24, 1986 Points Available OUESTION 1.03 What effect (INCREASE, DECREASE, REMAIN THE SAME) would each of the actions below have on the Available Net Positive Suction Head to a centrifugal pump?
a.
Raising the pump elevation to be closer to the surge tank that feeds it.
(0.5) b.
Decreasing the inlet pipe diameter.
(0.5) c.
Placing a heat exchanger upstream of the pump to cool the fluid.
(0.5) d.
Increasing the discharge head by use of a downstream throttle valve.
(0.5)
ANSWER 1.03 a.
DECREASE.
[+0.5]
b.
DECREASE.
[+0.5]
c.
INCREASE.
[+0.5]
d.
INCREASE.
[+0.5]
Reference (s) 1.03 1.
SHNPP FF-H0-1,1, Fluid Mechanics in Pumps and Piping, Revision 2, pp. 39-40 of 65.
-Section 1.0 Continued on Next Page-
7 e
e Page 4 Harris 1 February 24, 1986 Points Available OUESTION 1.04 The reactor is operating at 50% power with rod control in manual and turbine control in automatic when a S/G PORV fails open.
Assuming no reactor trip or operator actions, which of the following best describes the resulting steady-state conditions.
(1.0)
'(a.) Final Tavg > Initial Tavg, Final Power = Initial Power (b.) Final Tavg ) Initial T,yg, Final Power > Initial Power (c.) Final T,yg ( Initial Tavg' Final Power = Initial Power (d.) Final Tavg ( Initial Tavg' Final Power > Initial Power (e.) Final T
= Initial Tavg, Final Power > Initial Power avg ANSWER 1.04 (d.)
[+1.0]
Reference (s) 1.04 1.
SHNPP, Heat Transfer and Reactor Theory Lesson Plans (Integrated Knowledge).
-Section 1.0 Continued on Next Page-2
e Page 5 Harris 1 February 24, 1986 Points Available 00ESTION 1.05 List the three (3) components of the power coefficient in order of decreasing reactivity contribution. Asiume BOL.
(1.5)
ANSWER 1.05 1.
Doppler (fuel temperature) 2.
Moderator temperature 3.
Void. Me. <>/S : (f dW*r) P e d.h N s' E uY t'0" a
[+0.3] forgcomponent. [+0.2] for order, e4d Reference (s) 1.05 0
1.
SHNPP RT-LP-1.10, pp.3 -15 and RT-TP-197.0.
14
-Section 1.0 Continued on Next Page-i l
Page 6-Harris 1 February 24, 1986 Points Available OUESTION 1.06 What is the correct order of the boiling phases listed below as they would occur in a coolant channel with normal flow and high heat flux?
(1.0) 1.
transition boiling 2.
bulk boiling 3.
film boiling 4.
sub-cooled nucleate boiling a.
2,4,3,1 b.
2, 4, 1, 3 c.
4,2,3,I d.
4, 2, 1, 3 ANSWER 1.06 d.
[+1.0]
Reference (s) 1.06 1.
SHNPP HT-LP-1.0, pp. 27-28 and Figures Hi-TP-56.0, Thermo-TP-60.0, and Thermo-TP-52.0.
-Section 1.0 Continued on Next Page-
r 8
Page 7 Harris 1 February 24, 1986 Points Available 00ESTION 1.07 In the sketch (below) of a simplified Ideal. Rankine Cycle, mark or. identify the line segments that represent major heat in, major heat out, and major work out.
(2.0) t 3
3 T
2 E
q ANSWER 1.07 2-3-4 Major heat in
[+0.5 each]
4-5 Major work out [+0.5]
5-1 Major heat out
[+ 0.5]
.(1-2 Minorworkin)
Reference (s) 1.07 1.
SHNPP Thermo-H0-1.4 Turbine and Rankine Cycle, Figure Thermo-TP-46.0.
-Section 1.0 Continued on Next Page-t
Page 8 Harris 1 February 24, 1986 Points Available OUESTION 1.08 Exolain why excessive condensate depression is to be avoided.
(0.5)
ANSWER 1.08 Subcooling of the condensate below the saturation temperature reduces niant ef ficiency.
[+0.5]
Reference (s) 1.08 1.
SHNPP Thermo-H0-1.5, Condenser Thermodynamics, Revision 1, p. 15 of 39.
QUESTION 1.09 The nart of the core mainly affected by control bank motion is (1.0)
(a.) the small volume around the tips of the rods.
(b.) the region along the length of each rod.
k+/i (c.) the upper part of the core d=:r. the sud U gr.
(d.) the whole core, including areas away from the rods.
ANSWER 1.09 (a.)
[+1.0]
Reference (s) 1.09 1.
SHNPP RT-H0-1.5, Reactivity Variations, Figure RT-TP-88.0, Revision 1.
-Section 1.0 Continued on Next Page-
F Page 9' Harris 1 February 24, 1986 Points Available OUESTION 1.10 The reactor is initially at 8 x 108 amps. Positive reactivity is introduced to put the reactor on a constant SUR of 0.35 DPM. The time it takes to. reach 1.2 x 108 amps falls in the range:
(1.0)
(a.) 5'to 15 seconds (b.) 15 to 20 seconds (c.) 20 to 50 seconds (d.) 50 to 75 seconds ANSWER 1.10 (c.)
[+1.0]
Reference (s) 1.10 1.
SHNPP RT-H0-1.6, Neutron Kinetics, Revision 2, pp.10-11 of 65.
-Section 1.0 Continued on Next Page-
s e
Page 10~
Harris 1 February 24, 1986 Points Available 00ESTION 1.11 Will the insertion of a g(iven amount of reactivity to a critical reactor at E0L produce a LARGER, SMALLER, or THE SAME) startup rate than at BOL.
Exnlain.
(1.0)
ANSWER 1.11 LARGER [+0.5].
The value of the effective delayed neutron fraction is smaller at E0L. A smaller Beta results in a larger SUR for a given reactivity change.
[+0.5].
Reference (sl 1.11 1.
SHNPP RT-H0-1.6, Neutron Kinetics, Revision 2, pp. 28-29 of 65.
i i
i T
t
-Section 1.0 Continued on Next Page-t
-m
+, - -, -
m,
-e-e
,e
,-,,,,--~r
-ne e
---e m
Page 11 Harris 1 February 24, 1986 Points Available OUESTION 1.12 For the following conditions, indicate whether the highest tensile stress occurs at the INSIDE WALL, CENTERLINE, or OUTSIDE WALL of the pressure vessel.
a.
Cooldown at 50 F/hr (0.5) b.
Increasing pressure 100 psig with 0 F/hr heatup rate.
(0.5) c.
Heatup at 100 F/hr.
(0.5)
ANSWER 1.12 a.
INSIDE WALL [+0.5]
b.
INSIDE WALL [+0.5]
c.
OUTSIDE WALL [+0.5]
Reference (s) 1.12 1.
SHNPP PTS-LP-1.0, p. 10 and Figures 6, 7, and 8.
QUESTION 1.13 What is the reason for having Primary Neutron Sources in the core?
(1.0)
ANSWER 1.13 To assure that the Excore Source Range Detectors are on scale.
(This prevents any unmonitored increase in flux during the first stages of startup, as well as control over the core while shutdown.)
[+1.0]
Reference (s) 1.13 1.
SHNPP RT-H0-1.7, Subcritical Reactor Theory, pp. 4-5 of 52.
-Section 1.0 Continued on Next Page-
r-Page 12 Harris 1 February 24, 1986 Points Available OUESTION'1.14 The reactor is initially shutdown with a count rate of 3 cps and keff of 0.920. A positive reactivity step is inserted that brings the steady-state count rate to 75 cps. Determine the new keff to 3 decimal places.
(1.0)
ANSWER 1.14 cry 1-k g 2
1 - kg CR O
\\
\\
1 k2"I-CR/(1-k)j t
2
[+1.0]
k2=1-(1-0.920)
= 0.997 Reference (s) 1.14 1.
SHNPP RT-H0-1.7, Subcritical Reactor Theory, Revision 2, pp. 29-31 of 52.
-Section 1.0 Continued on Next Page-
t
_;t p
Page 13 Harris 1 February 24, 1986 Points Available OUESTION 1.15 Giya two (2) means for flattening the radial flux distribution in subsequent core loadings.
(1.0)
ANSWER 1.15 1.
The distribution of bdnable poison rods.
[+0.5]
2.
The loading of fuel with hi her enrichment toward the
- Aubfg['
outside of the core.
[+0. 5
/h/f u ftf kr c-ht( cf 4 a'f w L, s e.,
c_
y Reference (s) 1.15 1.
SHNPP RT-H0-1.10, MTC and Total Power Defect, Revision 2,
- p. 19 of 41.
2.
SHNPP RT-H0-1.8, Core Construction, Revision 2, p. 21 of 36.
-Section 1.0 Continued on Next Page-
t..,-
~
Page 14 Harris 1 February 24, 1986 Points Available OUESTION 1.16' During the Harris startup there is boron in the burnable poison rods and there-is boric acid dissolved in the coolant.
It is necessary for boron to be in both places because:
(1.0)
(a.) The burnup rate for boron must be optimized.
(b.) The burnable poison rods corapensate for equilibrium xenon.
(c.) Moderator temperature coefficient must not be positive.
(d.) The soluble boron contributes to the Doppler broadening effect.
-ANSWER 1.16 (c.)
[+1.0]
Reference (s) 1.16 1.
' SHNPP RT-H0-1.9, Temperature Effects, Revision 2, p. 9 of 42.
[
l I
-Section 1.0 Continued on Next Page-
s Page 15 Harris 1 February 24, 1986 Points Available OUESTION 1.17 For the Moderator Temperature Coefficient (MTC), match the parameter change in Column A to the direction it will change the MTC in column B.
(2.0)
Column A Column B 1.
Moderator temperature increases a.
More negative 2.
Boron concentration increases b.
Less negative 3.
All rods in versus all rods out c.
No effect 4.
Flux shifting towards edge of core ANSWER 1.17 1.
a.
[+0.5]
2.
b.
[+0.5]
3.
a.
[+0.5]
4.
a.
[+0.5]
Reference (s) 1.17 1.
SHNPP RT-H0-1.10, pp. 20-22.
I
-Section 1.0 Continued on Next Page-j
_.,_.___m
.____,.._,m_.__
~
Page 16 Harris 1 February 24, 1986 Points Available OUESTION 1.18 Answer TRUE or FALSE. Control rods are more effective neutron absorbers at low moderator temperatures than at high moderator temperatures.
(0.5)
ANSWER 1.18
~
FALSE.
[+0.5]
Reference (s) 1.18 1.
SHNPP RT-H0-1.10, MTC and Total Power Defect, Revision 2, pp. 17-19 of 41; Figure RT-TP-193.0, Revision 0.
QUESTION 1.19 In the event of a Rod Ejection Accident, what will be the first reactivity coefficient to insert negative reactivity? Explain your answer.
(1.0) p*.) h4. cs 6.
ANSWER 1.19 j],lSoy The Doppler coefficient [+0.5].
It senses the effect of positive reactivity insertion first because the fuel increases in temperature before the coolant [+0.5].
Reference (s) 1.19 1.
SHNPP RT-H0-1.10, MTC and Total Power Defect, Revision 2,
- p. 6 of 41.
-Section 1.0 Continued on Next Page-
7 Page 17 Harris 1 February 24, 1986 Points Available OUESTION 1.20 Which of the following(at power) xenon and peak (after shutdown) statements concerning the reactivity values of equilibrium xenon is correct? Assume shutdown occurs from equilibrium conditions.
(1.0)
(a.) Equilibrium xenon is INDEPENDENT of power level; peak xenon is INDEPENDENT of power level.
(b.) Equilibrium xenon is INDEPENDENT of power level; peak xenon is DEPENDENT on power level.
(c.) Equilibrium xenon is DEPENDENT on power level; peak xenon is INDEPENDENT of power level.
(d.) Equilibrium xenon is DEPENDENT on power level; peak xenon is DEPENDENT on power. level.
ANSWER 1.20 (d.)
[+1.0]
Reference (s) 1.20 1.
SHNPP RT-LP-1.11, pp. 11, 13.
-Section 1.0 Continued on Next Page-
Page 18
. Harris 1 February 24, 1986 Points-Available OUESTION 1.21 What are the three (3) purposes of establishing Control Rod Insertion Limits?
(1.5)
ANSWER 1.21 4
'1.
To minimize the consequences of A rod ejection accident.
[+0.5]
h t4 cssd s'i -%
- t+d ats's n.l ta e, w s d".
2.
To guarantee sufficient shutdown margin.
[+0.5]
3.
To provide suitable axial flux distribution.
[+0.5]
Reference (s) 1.21 1.
SHNPP RT-H0-1.13, Control Rod Reactivity Effects, p. 20 li of 27.
l
?
P r
1 f
1
/
1 r
r
-Section 1.0 Continued on Next Page-1
?
..-.., ~.
Page 19 Harris 1 February 24, 1986 Points Available OUESTION 1.22 List five (5) items whose reactivity changes are considered in determining Shutdown Margin.
(2.5)
ANSWER 1.22 1.
Xenon 2.
Boron concentration 3.
Moderator temperature 4.
Control rod worth
(" rod worth" by itself only counts as 0.5) 5.
Shutdown bank worth 6.
Power defect 7.
Most reactive rod stuck out 17. S: <<
Any 5 [+0.5] each Reference (s) 1.22 1.
SHNPP GP-006, Mode 1 to 3, p. 4.
2.
SHNPP OST-1036, SDM, Attach. II, p 1.
-Section 1.0 Continued on Next Page-s -.
^ :_t
.\\
5 3
, s o
+.
4, Page 20 Harris 1 s
February 24,g 1986 s.3
- 3
' Points s
-t 4
\\
Available y
r_
\\\\
f t
OUESTION 1.23 4
l\\,
,\\
N 3
s Rod withdrawal or boron dilution may be done during Xe de' cay.
Why doesn't this violate the rule of not adding-. positive s
.s reactivity by more than one method at a time?'
(1.0).
s 3
w s
-z, 1-s ANSWER 1.23 y
1
-+4-n Xe decay effect is very slow, compared to rod or boron ~ changes.
(It is allowed in the precautions of GP-004.), [+1.0]
s Reference (s) 1.23.
s 1.
SHNPP GP-004, Reactor Startup from Hot _ Standby, p. 5.
s
-s 1
-End of Section 1.0-s s
'(
T s
F A
\\
f P
m g
\\
\\
s 3.,
1 N
} N
- s. s.
4
- - - - - - - ' - - - - - - - - - - ~ - -
~
Page 21 Harris 1 February 24, 1986 Points Available 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (30.0)
OUESTION 2.01 Select the reason why there are eight (8) decay tanks in the Gaseous Waste Processing System, rather than one (1) of the same total volume.
(1.0)
(a.) to reduce the site boundary dose in the event of a tank rupture.
(b.) to comply with ASME Pressure Vessel Code requirements for stress.
(c.) to reduce the radiation dose to personnel working with the tanks.
(d.) to make installation and removal easier in the confined space.
ANSWER 2.01 (a.)
[+1.0]
Reference (s) 2.01 1.
SHNPP SD-120.7, Gaseous Waste Processing System, p. 5.
-Section 2.0 Continued on Next Page-
,n,.
1 r~---
s Page 22 Harris 1 February 24, 1986 Points Available OUESTION 2.02 Giya the four (4) conditions for which the Motor-Driven Pumps of the Auxiliary Feedwater System will automatically start.
(2.0)
ANSWER 2.02 1.
1rsw-low level in any one (1) S/G [+0.5]
2.
loss of off-site power [+0.5]e Al50 AccepY: N O ( fRu Sg$sf3 3.
trip of all main feed pumps
[+0.5]
"#) bD 8" /058);
b C
4.
SI signal
[+0.5]
Reference (s) 2.02 1.
SHNPP SD-137, AFW, p. 11.
4
-Section 2.0 Continued on Next Page-
o Page 23 Harris 1 February 24, 1986 Points Available OUESTION 2.03 Describe the flow alignment after switchover to recirculation phase by indicating the sequence of flow from the sump to the reactor.
Include major pumps, omit valves, state whether flow enters hot or cold legs. Assume all automatic and manual switchover actions are complete.
(2.0)
ANSWER 2.03 (After the sump valves are open and the RWST is isolated), the RHR pumps take suction from the sump [+0.5] and send part to the CSIPs [+0.5] and part directly to the reactor [+0.5].
Flow from both the RHR pumps and the CSIPs enter the cold legs
[+0.5].
Reference (s) 2.03 1.
SHNPP SD-110, SIS, p. 16.
-Section 2.0 Continued on Next Page-
Page 24 Harris 1-
- February 24, 1986 Points Available OUESTION 2.04 Answer each item TRUE or FALSE with regard to the Steam Dump System.
a.
The purpose is to enable the NSSS to operate with turbine load rejections up to 10% step.
(0,5) b.
The combination of atmospheric and condenser dump valves are not part of plant safeguards.
(0.5) c.-
Below 15% power the steam dump system is used to maintain RCS temperatures at a constant no-load value.
(0.5)
ANSWER 2.04 a.
FALSE.
[+0.5]
b.
TRUE.
[+0.5]
c.
TRUE.
[+0.5]
Reference (s) 2.04 1.
SHNPP SE-126, Steam Dump System, pp. 9-10.
9 i
-Section 2.0 Continued on Next Page-
Page 25 Harris 1 February 24, 1986 Points Available 00ESTION'2.05 State the reasons for the cold leg accumulator minimum and maximum water volume.
(1.0)
ANSWER 2.Q5 1.
Minimum volume insures sufficient water from 2/3 CLA is available to reflood core.
[+0.5]
2.
Maximum volume is limited to insure sufficient nitrogen is present to force CLA contents into RCS.
[+0.5]
Reference (s) 2.05 1.
SHNPP CLA-LP-1.0, p. 8.
QUESTION 2.06 Describe the source and path of power through the "uninterruptible" 120 V ESF AC System for both normal and backup power.
(1.5)
ANSWER 2.06 The 480 V supply is transformed and rectified to 125 V DC
- +0.5;.
It then passes through an inverter to give 120 V AC
_ 0. 5.
If the 480 V supply is lost, the inverter is auto-
+
matically fed by a 125 V DC ESF battery.
[+0.5]
h c G. Q'l 9 W e) c}1 jn ft.k. 2.
f /g _,
LL[^, hey ju l2 2 lli:
~
Reference (s) 2.!2 Y#^
'I C 'O '
1.
SHNPP SD-156, Plant Electrical System, p. 7.
hwgj/LOtl LA lb 5 - T P ~ I. D, t s v. t > /0//4/fy 2,
'o 4
-Section 2.0 Continued on Next Page-
o Page 26 Harris 1 February 24, 1986 Points Available OUESTION 2.07 Describe the three (3) purposes of the piping connected to the Main Steam Lines in Containment between the S/Gs and the containment penetration.
(1.5)
ANSWER 2.07 The purposes of one tap are.to measure flow [+0.5] and pressure
[+0.5]; the purpose of the other tap is to vent the system [+0.5]
prior to heatup.
Reference (s) 2.07 1.
SHNPP SD-126,. Main Steam System, p. 7.
4
-Section 2.0 Continued on Next Page-
Page 27 Harris 1 February 24, 1986 Points Available OUESTION 2.08 a.
Idhat are the two (2) effects that contribute to shrink in the S/G during a decrease in power?
(1.0) b.
At low power, what design feature allows " tempering" of the S/G?
(0.5)
ANSWER 2.08 a.
A decrease in power results in a pressure increase [+0.5]
that decreases the apparent level. When the steaming and feed rates decrease, the water in the downcomer drops and its momentum [+0.5] causes the apparent level to oscillate.
b.
(The flow is directed through the) bvoass line (into the l,J.e M auxiliary feed nozzle initially. This allows mixing and minimal water hammer to the system.)
[+0.5]
((apy es $N*
Reference (21 2.08 1.
SHNPP SD-126, Steam Generator System, pp. 52-53 and Figure 7.34.
t.
S H-bltP L w s eP w -H - 6. o, t.o. ), //n /ty CPW -Lf'-l.0,Jef VT-tT 5.
=
" *E**
4-( f. P us,. a : h&u cd N* f ** E5 jg p, g4 (A +. +. %w ~%'% s.)
Adk n l A-v, h _
hhNYu myrMg3 f ern h gly e, wa.s d ws sn MI w ait of- <= 4 Slw p - J ua. n M k a. volv is m m e
VLJds he-et f[ HR t%-W)
WG. }0 ^t bM & -
J Section 2.0 Continued on Next Page-i-
.---m p.-.
-3,-
-e.7
-y y- - - - -t+y-y s-
-+.----rw-..-n.--
--y-9-
,.,g y
.w qer
- y y
,r-w+ywwvw-y
b Page 28 Harris 1 February 24, 1986 Points Available OUESTION 2.09 Eire the three (3) subsystems that comprise the Post Accident Hydrogen System.
(1.5) 1 ANSWER 2.09 1.
Electric Hydrogen Recombiners
[+0.5]
l 2.
Contair. ment Hydrogen Purge [+0.5]
i-3.
Containment Hydrogen Monitoring [+0.5]
Reference (s) 2.09 1.
SHNPP SD-125, Post Accident Hydrogen System, pp. 3-4.
QUESTION 2.10 Eire AbE four (4) conditions that must be satisfied before the Emergency Diesel Generator circuit breakers will close on the emergency buses.
(2.0)
%s q ex.c3 S WY *%p kl. ha e s.4.+ g.n,>,,' +
ANSWER 2.10 w w ph Spec.s'Se'e> ncnt se g g. $ jg 3 44w /,g g i
1.
The emergency bus is not energized nr Preferred power source breakgr open. [+0.5 for eitherl,,
ed 2 py17.
j A ccep t' ' Ys (A U sQs %'p t et/
2.
The diesel generator frequency is(54 Hz)or greater. [+0.5] 4cc. / "sp e=['
j 3.
The generator voltage is(90%)of rated or greater.
[+0.5]
4.
The diesel generator lockout relay in the diesel generator local panel is reset.
[+0.5]
c>p eM, P6 e%
A K,4 a.p e 4 y &s ifo'n, d jo es / / %
S.
S e l e c.f 6 r s e N-cA i%
ah Reference (fs_r.+ ken 6
o%
2.10
{
1.
SHNPP SD-155, Emergency Diesel Generator System, p. 19.
%.e.P - Ho-l.0, pp K M7 7..
't
-Section 2.0 Continued on Next Page-2 E
e.
.m
,.e.
- ,y,_,_y.
_,-r w
c-,
.r,..-,.-.,yey3.-,.--i-
,,.,-,--..9--
,w.-.--,.
ww ep
-y..,
.-r,.
.,--,-3 w
y - -
c
.y Page 29 Harris 1 February 24, 1986 Points Available OUESTION 2.11 Under what two (2) circumstances (plant conditions) is the Thermal Barrier Heat Exchanger relied upon to cool the RCP internals?
(1.0)
ANSWER 2.11 1.
Seal injection is' lost
[+0.5]
2.
No. I seal failure [+0.5]
Evaluate other answers on case by case basis.
Reference (s) 2.11 1.
SHNPP SD-100, RCS, p. 23.
-Section 2.0 Continued on Next Page-
n +=
- ,c Page 30 Harris 1 February 24, 1986 Points-Available OUESTION 2.12 Which of the following situations will cause the nonessential loop of the CCW system to isolate automatically?
(1.0)
(a.) Reactor trip (b.) Safet.y injection (c.) ATWS event (d.) Seismic event ANSWER 2.12 (b.)
[+1.0]
Reference (s) 2.12 1.
SHNPP SD-145, CCW System, p. 5.
-Section 2.0 Continued on Next Page-
Page 31 Harris 1 February 24, 1986 Points Available OUESTION 2.13 a.
Gire two (2) reasons why a small continuous flow is maintained through both PZR spray lines.
(1.0) b.
Why are loop seals maintained between the PZR and the main control valve for the sprays, as well as between the PZR and the safeties? The reason is not necessarily the same for both.
(1.0)
ANSWER 2.13 a.
To reduce thermal stress when main spray valves are opened
[+0.5], and to keep the coolant within the PZR from differing in chemical concentration from the main coolant
[+0.5]. M f : p w = d iAoCitte lu.erm er 10. f 3 sk wnwu Mn. we. % at wq t o.z.O b.
The layout (effectively a loop sealT of the spray line is to prevent steam building up all the way back to the control valves [+0.5].
(This is not a leakage problem because the pressure upstream of the valve is somewhat higher than the PZR pressure. CAE, with a continuous flow from the spray bypass, wouldn't this prevent steam buildup?F 4s.se g /af The loop seal in each of the safety lines is to prevent steam and hydrogen leaking through the valves (because the downstream pressure is much less than PZR pressure).
[+0.5]
Reference (s) 2.13 1.
SHNPP SD-100, RCS, pp. 34 and 37.
l
-Section 2.0 Continued on Next Page-
)
- Page 32 Harris 1 February 24, 1986 Points evailable OUESTION 2.14 Toward end of core life, Why is the BTRS preferred over normal dilution for removing boron from the coolant?
(1.0)
. ANSWER 2.14 Less water for the boron recycle system to process (by evaporation).
[+1.0]
Reference (sl 2.14 1.
SHNPP SD-108, Boron Thermal Regeneration System, p. 4.
4 3
-Section 2.0 Continued on Next Page-
Page 33 Harris 1 February 24, 1986 Points Available OUESTION 2.15 Answer each item TRUE or FALSE with regard to the Containment Spray System.
a.
Sodium tetraborate is added to the spray water to help remove iodine from the containment atmosphere and keep boron concentration up.
(0.5) b.
Cavitating venturis prevent damage to the containment spray pumps, due to runout, during the injection phase.
(0,5) onvaJ4esare has araty ggds c..
he ci dulati isol arepentop/econfinment pr ec ve ch ber, mosp ere.
.)
d.
Containment spray pump suction is automatically transferred from RWST to containment sump on RWST low-low level.
(0.5)
ANSWER 2.15 a.
FALSE.
[+0.5]
b.
TRUE.
[+0.5]
W ALSE. [@ Ab*WA M
El' O.-$p.1 k h c. N b gi-V\\ e.
d.
TRUE.
[+ 0.5]
Re ference(s) _. 2.15 l
1.
SHNPP SD-112, Containment Spray, pp. 4-6.
I l
l l
-Sectica 2.0 Continued on Next Page-L
Page 34 Harris 1 February 24, 1986 Points Available QUESTION 2.16 a.
Exolain why the mixed-bed demineralizers in the CVCS are first loaded with borate before being placed in service.
(1.0) l b.
What is the purpose of the cation demineralizer that is downstream of the mixed-bed demineralizers in the CVCS?
(1.0)
ANSWER 2.16 a.
To prevent inadvertent dilution by boron removal.
[+1.0]
b.
To be used occasionally to adjust lithium for pH control, or to maintain cesium concentration (below 1 pC1/cc with 1% defective fuel).
[+1.0]
Reference (s) 2.16 2
1.
SHNPP S0-107, CVCS, pp. 18-19.
QUESTION 2.17 Explain how RCS pressure is controlled during a cooldown on RHR and when the PZR is solid.
W /O MSWER 2.17 The huhhlg_ig,,th piiaaurizer, Dy meciia vi ^Ed b?'t 0M oiid-sways entra! ApresYumt.an1Li.1--t4w fZn is sMid W F^
M r
-t wm some flow is diverted through the normal letdown lines, allowing pressure control by (valve PCV-145 for letdown pressure control and) the charging pumps @E4].
Cr$3 1:o Reference (s)_2.1Z 1.
SHNPP SD-111, RHR, p. 15.
-Section 2.0 Continued on Next Page-
~
Page 35 Harris 1 February 24, 1986 Points Available 00ESTION 2.18 4
Answer each item TRUE or FALSE with regard to the Radiation Monitoring System.
a.
Provides historical as well as current mreasurements of radiological conditions.
(0.5) b.
Provides warning of leakage from primary to sccondary plant systems.
(0.5) c.
Provides an input to the Reactor Protection System in the event of radioactive release.
(0.5) d.
Provides for grab samples as well as continuous process monitoring.
(0.5)
ANSWER 2.18 a.
TRUE.
[+0.5]
b.
TRUE.
[+0.5]
c.
FALSE.
[+0.5]
.i d.
TRUE.
[+0.5]
Reference (s) 2.18 1.
SHNPP SD-118, Radiation Monitoring System, pp. 6-7.
-Section 2.0 Continued on Next Page-
,m_
,,__----4,
-,e..
y-.,
.-..,, -, -., - - -, -, ~., + -. - - -,.,.. - - -.
..e.-,
Page 36 Harris 1 February 24, 1986 Points Available OUES ON 2.19 five (5) different RCS penetrations that use t ermal slee es.-
(1.0)
ANSWER 2.19 1.
Normal arging connection C! eld t 2.
Alternate c arging connecti 3.
Return lines m RHRS/ S b E EM 4.
One end of PZR su line 5.
Other end of P surg line
-eA 6.
Spray lin connection to e PZR
.Any 5 at
.2] points each.
Referen s) 2.19 RCS-H0-1.0, p. 10.
QUESTION 2.20 What does the RCP 011 Lift System lift?
(0.5)
ANSWER 2.20 The thrust shoes (away from the thrust runner).
[+0.5]
Reference (s) 2.20 1.
RCP-H0-1.0, p. 18.
-End of Section 2.0-i i
e l
j
,?
^ -
Page 37 Harris 1 February 24, 1986 Points Available 3.0 INSTRUMENTS AND CONTROLS (30.0)
OUESTION 3.01 Select the correct statement'about the PORVs.
(1.0)
(a.) The rate of opening for all PORVs is related to the rate of pressure rise in PZR.
(b.) Control power is 125 VAC safety-related power.
(c.) PORVs automatically unblock as pressure goes above P-11 setpoint.
(d.) Loss of air to the PORVs causes them to fail as is, if open.
ANSWER 3.01 (c.)
[+1.0]
Reference (s) 3.01 1.
SHNPP PZRPC-LP-1,0, Revision 1, p. 14.
-Section 3.0 Continued on Next Page-
'N
. B
'Page 38 Harris 1 February 24, 1986 Points Available.
QUESTION 3.02-Why are Isolation Amplifiers used in control and protection circuitry?
(1.0)
ANSWER 3.02 To prevent feedback from control or monitoring / recording circuits from-interfering with protection circuitry that uses the same input signal.
[+1.0]
Reference (s)'3.02 1.
General Instrumentation Practice.
a 4
I
-Section 3.0 Continued on Next Page-
Page 39 Harris 1 February 24, 1986 Points Available OUESTION 3.03 Match the permissive interlocks in Column A with the statement in Column B about their. function in the Steam Dump System.
(2.5)
A B
J a.
Turbine trip, C-8 1.
Blocks operation of atmospheric dumps, b.
Condenser, C-9 2.
Prevents uncontrolled cooldown.
c.
Loss of load, C-7a 3.
Permits steam dump operation for cooldown below hot standby.
d.
Loss of load, C-7b 4.
Arms condenser dumps after 10% load rejection in ( 2
- minutes, e.
Low-Low Tavg, P-12 5.
Arms condenser after 50%
load rejection in ( 2 minutes.
6.
Protects Condenser from overpressurization.
7.
Arms atmosphere dumps after 50% load rejection in ( 2 minutes.
ANSWER 3.03
' 0. 5' a.
1
+
' 0.5' b.
6
+
' 0.5' c.
4
+
' 0.5' d.
7
+
e.
2 l+0.5l Reference (sl 3.03 1.
SHNPP SDCS-LP-1.0, Revision 2, pp. 6-7.
-Section 3.0 Continued on Next Page-
'Page 40 Harris 1 February 24, 1986 Points Available OUESTION 3.04 Comolete the table for SI:
(3.0)
Signal Auto Initiates At
_ Logic Coincidence Low PZR pressure psig
/
Low steamline pressure psig
/
in any one S/G Ctmt Hi I pressure psig
/
ANSWER 3.04 18$D HH3&t psig [+0.5]
2/3
[+0.5]
66/
-664-
[+0.5]
2/3
[+0.5]
5-
[+0.5]
2/3
[+0.5]
Reference (s) 3.04 1.
SHNPP ESFAS-LP-1.0, Revision 0, pp. 6-7.
-Section 3.0 Continued on Next Page-
.l Page 41 Harris 1 February 24, 1986 Points Available OUESTION 3.05 What are the first two-(2) loads sequenced onto the EDG after an undervoltage condition on an ESF line initiates the auto safeguards action?
(1.0)
ANSWER 3.05 1.
charging pumps
[+0.5]
2.
[+0.5]
Reference (s) 3.05 1.
SHNPP SEQ-LP-1.0, Revision 1, p. 5.
QUESTION 3.Q6 Select the correct statement for the PZR Pressure Control System.
(1.0)
(a.) The heaters and sprays overlap to provide positive control.
(b.) Heaters turn off, regardless of pressure, if PZR level drops below 17%.
(c.) The Master Controller controls two of the three PORVs.
(d.) PORVs are likely to open in transferring from Auto to Manual control of sprays, ANSWER.3.06 t
(b.)
[+1.0]
Reference (sl 3.06_
4 1.
SHNPP PZRPC-LP-1.0, Revision 1, pp. 11-14.
-Section 3.0 Continued on Next Page-
Page 42 Harris 1 February 24, 1986 Points Available OUESTION 3.07 Answer TRUE or FALSE. At normal operating temperature, the cold calibrated PZR level channel will indicate lesi than the hai calibrated channels.
(0.5)
ANSWER 3.07 TRUE.
[+0.5]
Reference (s) 3.07 1.
SHNPP PZRLC-LP-1.0, Revision 1, p. 10.
QUESTION 3.08 a.
What signal is generated by Low-T if coolant temperature drops below this limikyOfter reactor trip?
(0,5) b.
What is the purpose of setting Low-T above no-load avg Tavg?
(1.0)
MSWER 3.08 a.
Feedwater isolation.
[+0.5]A-ccN: Mo cl 63e, Mn&g b.
To prevent an excessive cooldown of the primary system d*
N (because the secondary is still steaming but the primary hasmainlyonlydecayheat).
[+1.0]
Reference (s) 3.08 1.
SHNPP RCTEMP-LP-1.0, Temperature Instrumentation System, pp. 12-13.
-Section 3.0 Continued on Next Page-
Page 43 Harris 1 February 24, 1986' I
Points Available OUESTION 3.09 i
comolete the following table of selected reactor trips:
(2.0)
Trio Initiated At Loaic coincidence SR H1 flux cps
/
Hi PZR pressure psig
/
Hi PZR level
/
Lo-Lo S/G level
% NR
/
l ANSWER 3.09 5
10
[+0.25]
1/2 [+0.25]
2385 [+0.25]
2/3
[+0.25]
92% [+0.25]
2/3
[+0.25]
M [+0.25]
2/3
[+0.25]
30.)
.L
_J U1
- f l
_2,
~. < ~
Reference (s) 3.01 1.
SHNPP RPS-LP-1.0, Revision 3, pp. 10, 14-16.
-Section 3.0 Continued on Next Page-
Page 44 Harris 1 February 24, 1986 Points Available OUESTION 3.10 Answer each item TRUE or FALSE.
a.
Turbine trip above 10% power causes reactor trip and vice versa.
(0.5) b.
SI causes reactor trip and vice versa.
(0.5) c.
Feed flow much less than steam flow will only cause reactor trip if it is to a low level S/G.
(0.5)
ANSWER 3.10 a.
TRUE.
[+0.5]
b.
FALSE.
[+0.5]
c.
TRUE.
[+0.5]
Reference (s) 3.10 1.
SHNPP RPS-LP-1.0, Revision 3, pp. 16-17.
-Section 3.0 Continued on Next Page-w,
. _ _ _ _ _ _ _ _, _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. - _ _ _ _ _ _ _ _ - _ - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^ ^ - - - - - - - - - - - - - - - - - - - - ~ ' - - - - - - - - - " ' ' - - ' ~ ~ ~ - - - - - - '
Page 45 Harris 1 February 24, 1986 Points Available OUESTION 3.11 tion would the turbine behave differently between a runback associated with OTAT-3%, and runback due to 'o s cf 0 " c' O
M NMe cd WAAt p W,y be'p-M h&
MSWER 3.11 Exceeding the limit OTAT-3% causes an intermittent runback
- It d k 'h p y (until RCS conditions drop below the limit) [+0.5].
Loss of
+eed causes a continuous runback (until the external load is matched and turbine systems stabilize) [+0.5].
(*200%/ min for 1.5 sec, 30 see hold, repeat.)
Reference (s) 3.11 1.
SHNPP RPS-LP-1.0, Revision 3, p. 21.
2.
SHNPP SD-131.5, Revision 1, p. 22.
l l
-Section 3.0 Continued on Next Page-j j
Page 46 Harris 1 February 24, 1986 Points Available OUESTION 3.12 Which is the correct statement about the reactor trip and bypass -
breakers?
(1.0)
(a.) Bypass breakers are interlocked to allow simultaneous closing.
(b.) RPS train A deenergizes trip breaker B and bypass breaker A.
(c.) When trip breaker is bypassed, associated RPS train is -
tripped.
-(d.) Logic signals control power to both undervoltage and shun,t trip coils.
1l ANSWER 3.12 f
(d.)
[+1.0]
Reference (s) 3.12
['
1.
SHNPP RPS-LP-1.0, Revision 3, p. 7.
.r,
A l-
-Section 3.0 Continued on Next Page-
.-._.-._.-_ _ _,._.-.~..--.__. __-
~ Page 47' Harris 1 February 24, 1986 Points Available OUESTION 3.13 Exolain how the Intermediate Range Flux Detector compensates for gammas.
(1.0)
ANSWER 3.13 The inner chamber detects gammas only, while the outer chamber detects both thermal neutrons and gammas. The ion current from the inner chamber is electrically subtracted from the ion current of the outer chamber, leaving a signal that is proportional to thermal neutron flux only.
[+1.0]
Referencefs) 3.13 1.
SHNPP NIS-LP-1.0, Revision 1, p. 9.
-Section 3.0 Continued on Next Page-
- s. s s
'y
- d.," 3
)
x%
%:q Page 48
'(.
~ Harris 1 February 24, 1986 t.
w v ;-
Points h
Available 4
3 s, x 4s t
s,,'
~,.
=
OUESTION 3.14,,
Answer e hi item TRUE or FALSE with regard to the Source Range s
Detector:n,
'e a.
Detects thermal neutrons and gammas.
(0.5) b.
Compensates for gammas.
,(0.5)
~
c.
Fast neutrons areLthermalized by the polythene cover.
(0.5) d.
Operates in the G-M plateau region.
'+
(0.5)
ANSWER 3.1[
3 a.
TRUE. ' [+0.5]
x b.
FALSE.
[+0.5],
,o c.
' TRUE.
[+0. 5] '
\\
- d. 7; F,ALSE.
[+05]~
N.
?
s Refhrence(s)3.14 s.
's 1..
SHNPP NIS-LP-1.0, Revision 1, p. 8.
s s
OUESTION 3.15 Answer TRUE or FALSE. A General Warning light on :the RPS system can mean a loss of one of the DC power supplie;.
(0.5)
~n ANSWER 3.15 kRUE. [+0.5]
s Refer'ence(s) 3.15 <
1.
SH,NPP RPS-LP-1,0, Revision 3, p. 8.
i
~
-Section 3.0 Continued on Next Page-
- a,
]
D'*
t>
-w
Page 49 Harris 1 February 24, 1986 Points Available QUESTION 3.16 Give 10 of the 11 control or protective functions supplied by the output from the Power Range Detectors.
(2.5)
ANSWER 3.16 1.
Overpower differential temperature (0 PAT) setpoint.
2.
Overtemperature differential temperature (OTAT) setpoint.
3.
Overpower Trip-Low Range (25%).
4.
P-10 nuclear at power permissive.
5.
Automatic rod control (N-44 only).
6.
Steam generator water level control.
7.
Overpower Trip-High Range (
%).
8.
C-2 high flux rod stop (103%).
9.
P-8 three loop flow permissive (( 49%).
- 10. Power range Positive Rate Trip.
- 11. Power range Negative Rate Trip.
Any 10 (+0.25) each.
Reference (s) 3.16 1.
SHNPP NIS-1.0, Revision 1, pp. 16-17.
-Section 3.0 Continued on Next Page-
Page 50 Harris 1 February 24,-1986 Points Available QUESTION 3.17 Waat indications should be checked if level transmitter LT-115
~
on the VCT has failed high?
(1.0)
ANSWER 3.17 LI-115 indicating full VCT high level alarm LI-112 (local) shows actual VCT level decreasing Auto M/U and auto emergency M/U actuation inoperable LCV-115A (ICS-120) diverted to BRS
[+0.2 each]
Reference (s) 3.17 1.
SHNPP A0P-003, Revision 1, Table I, p. 9.
-Section 3.0 Continued on Next Page-
Page 51 Harris 1 February 24, 1986 Points Available OUESTION 3.18 Exolain.how the data A and B channels of the DRPI system combine to give an accuracy of i 4 steps.
(1.0)
ANSWER 3.18 Channel A by itself has an accuracy of +4/-10 steps and channel B by itself has an accuracy of +10/-4 steps.
[+1.0]
Reference (s) 3.18 1.
SHNPP RODCS-LP-1.1, Revision 1, p. 10.
OVESTION 3.19 Narrow range TH and Tc are not directly displayed. Describe what RCS loop temperatures are displayed instead.
(1.5)
ANSWER 3 19 T
9, [+0.5] delta T [+0.5], and wide range temperatures [+0.5]
(on recorders 410, 420 and meters TI-410, 420, 413, 423).
Reference (s) 3.19 1.
SHNPP RCTEMP-LP-1.0, Revision 1, p. 8.
-Section 3.0 Continued on Next Page-
Page 52
. February 24, 1986 Harris 1 Points Available OUESTION 3.ZQ Select'the correct statement about Containment isolation.
(1.0)
(a.) Radioactivity above a certain limit in containment will cause a phase A isolation.
(b.) Manual actuation of containment spray will cause a phase B isolation.
(c.) Hi-1 containment pressure (2/3) will only cause containment ventilation isolation.
(d.) Manual actuation of phase A containment isolation will cause SI.
ANSWER 3.20 (b.)
[+1.0]
Reference (2) 3.20 1.
SHNPP ESFAS-LP-1.0, ESF Actuation System, Figures TP-2, TP-5, and TP-6.
1
-Section 3.0 Continued on Next Page-w-
Page 53 Harris 1 February 24,.1986 Points Available QUESTION 3.21 During a Design Basis Accident, the Emergency Diesel Generator will be shutdown automatically for what three (3) conditions?
(1.5)
ANSWER 3.21 1.
Engine or generator overspeed.
[+0.5]
2.
Generator differential relay action.
[+0.5]
3.
Generator bus fault.
[+0.5]
Reference (s) 3.21 1.
SHNPP SD-155, EDG System, pp. 21-22.
QUESTION 3.22 Indicate which of these detector types (ION CHAMBER, GM, or PROP 0RTIONAL COUNTER) would be best suited for the following applications.
(1.0) a.
Neutron survey b.
Personnel contamination survey ANSWER 3.22 a.
PROP 0RTIONAL COUNTER [+0.5]
b.
GM [+0.5]
Reference (s) 3.22 1.
SHNPP RP-LP-1.4, Radiation Protection Principles and Instruments, pp. 17 and 14, respectively.
-End of Section 3.0-
Page 54-Harris 1 February 24, 1986 Points Available PROCEDURES: NORMAL. ABNORMAL. EMERGENCY. AND RADIOLOGICAL' 4.0 CONTROL (30.0)
OUESTION 4.01 Excluding RCPs, state four (4) of six (6) components that must be assured of continued CCW flow after transfer from the RWST to recirculation after a LOCA.
(2.0)
ANSWER 4.01 1.
Excess letdown Hx 2.
RCDT Hx 3.
Letdown Hx 4.
Seal return Hx 5.
Recycle evaporator 6.
Spent fuel pool Hx Any four (4) (+0.5) each.
Reference (s) 4.01 1.
SHNPP E0P-EPP-10, Transfer to Cold Leg Recirculation,
- p. 3.
-Section 4.0 Continued on Next Page-
Page 55 Harris 1 February 24, 1986 Points Ayallable 00ESTION 4.02 What is the most visible evidence of voiding in the RCS during post LOCA cooldown and depressurization? Assume RCPs not running.
(1.0)
ANSWER 4.02 Rapidly increasing PZR level or decreasing RVLIS, if available.
[+1.0 for either]
Reference (s) 4.02 1.
SHNPP E0P-EPP-009, Post LOCA Cooldown and Depressurization,
- p. 7.
QUESTION 4.03 Liit the four (4) logic circuits that must be reset after a LOCA in order to stabilize the plant.
(2.0)
ANSWER 4.03 1.
[+0.5]
2.
phase A [+0.5]
3.
phase B [+0.5]
4.
FW isolation
[+0.5]
Reference (s) 4.03 1.
SHNPP E0P-EPP-008, SI Termination, pp. 3-4.
-Section 4.0 Continued on Next Page-
-Page 56 Harris 1 February 24, 1986 Points Available OUESTION 4.04 List the two (2) Reactor Coolant Pump trip criteria, as listed in foldout A of.the E0P for critical safety function status tree.
(1.0)
ANSWER 4.Qi (3fo
[/4 *d3 1.
RCS pressure less than F98F psig [t800 psig].
[+0.5]
2.
Charging /SI pumps, at least I running.
[+0.5]
Reference (s) 4.04 1.
SHNPP E0P-CSFST, Foldout A.
-Section 4.0 Continued on Next Page-
- ~
Page 57 Harris 1 February 24, 1986 Points Available
'00ESTION 4.05 Indicate whether each caution statement below, following a loss of AC power to busses IA-SA and 18-SB, is TRUE or FALSE.
a.
Critical safety functions status trees should be monitored for information only.
(0.5) b.
Verify that _EDGs are loaded immediately to prevent auto trip on overheating.
(0.5) c.
An ESW pump should be started as soon as an EDG restores bus power.
(0.5) d.
A S/G with a ruptured tube should not be used to supply steam to the SDAFW pump, if possible.
(0.5)
ANSWER 4.05 a.
TRUE.
[+0.5]
b.
FALSE.
[+0.5]
c.
TRUE.
[+0.5]
d.
TRUE.
[+0.5]
Reference (s) 4.05 1.
SHNPP E0P-EPP-001, Loss of AC Power to Busses IA-SA and IB-SB.
-Section 4.0 Continued on Next Page-l 1
Page 58 Harris 1 February 24, 1986 Points Available OUESTION 4.06 What are eight (8) of eleven (11) symptoms of a loss of RHR?
Do not include redundant systems.
(2.0)
ANSWER 4.06 1.
"RHR Loop A/B Discharge Low Flow" alarm.
2.
"RHR Pump A/B Trip or Close CKT Trouble" alarm.
3.
"RHR Pump A/B Auto Start Fail /0verride or Overload" alarm.
4.
Low RHR loop flow indication on FI-605 A.1/B.1 and possible alarm.
5.
Inability to open or control direct flow-path valves in RHR system.
6.
Inability to start either RHR pump or stopping of both pumps.
7.
Loss of CCW and possible "RHR HX A/B CCW Hi/ Low Flow" alarm (refer to A0P-14).
8.
"RHR HX A/B Out High Temp" alarm.
9.
Indication of break in RHR piping.
10.
Increase in PRT level (indication of lifted RHR suction reliefvalve).
11.
Increase in BRS RHT level (indication of lifted RHR discharge relief valve).
Any eight (8) (+0.25) each.
Reference (s) 4.06 1.
SHNPP A0P-020, Revision 1, Loss of Residual Heat Removal,
- p. 3.
-Section 4.0 Continued on Next Page-I n-
~
Page 59 Harris 1 February 24, 1986 Points Available OUESTION 4.07 State three (3) methods of post SGTR cooldown.
(1.5)
ANSWER 4.07 1.
backfill
[+0.5]
2.
blowdown
[+0.5]
3.
steam dump [+0.5]
Reference (s) 4.07 1.
SHNPP E0P-EPP-17, 18, 19, Post SGTR Cooldown Methods.
QUESTION 4.08 What two (2) immediate actions are required after a " Rod Control Urgent Alarm", or other symptoms of control bank response failure after a turbine load increase?
(1.0)
ANSWER 4.QS
/
t 1.
Manually position the c6ntrol-banks to restore equilibrium.
[+0.5]
2.
If bank can't move, adjust turbine load or RCS boron to restore equilibrium.
[+0.5]
Reference (s) 4.08 1.
SHNPP A0P-001, Failure of a Control Bank to Move, p. 7.
-Section 4.0 Continued on Next Page-
Page 60
. Harris 1 February 24, 1986 Points Available OUESTION 4.09 According to A0P-004, Ehat does Reactor Operator R0-1 do to transfer control to the ACP when the control room is inaccessible due to a non-fire situation?
(1.0)
ANSWER 4.09 Go to Switchgear Room A [+0.5], (open transfer panel A) and place both key-actuated security switches in transfer position [+0.5].
Reference (s) 4.09 1.
SHNPP A0P-004, Alternate Safe Shutdown, Fire or Inaccessible Control Room, p. 25.
QUESTION 4.10 List all the Immediate Actions for a dropped control rod.
(1.0)
ANSWER 4.10 1.
Transfer Rod Control to manual.
[+0.5]
-2.
Adjust turbine load to equalize T with T
[+0.5]
avg ref.
Acc epi : Fed eu. fin o f p rear 6 </w7 Reference (s) 4.10 1.
SHNPP A0P-001, p. 17.
-3ection 4.0 Continued on Next Page-
Page 61 Harris 1 February 24, 1986 Points Available OUESTION 4.11 Why is it necessary to use protective equipment and clothing when working near a CCW leak?
(0.5)
ANSWER 4.11 It contains toxic chemicals (chromites).
[+0.5]
Reference (s) 4.11 1.
SHNPP SOP-014, Loss of Component Cooling Water, p. 5.
QUESTION 4.12 List four (4) conditions which require Emergency Boration.
(2.0)
ANSWER 4.12 1.
Excessive control rod insertion.
2.
Uncontrolled cooldown.
3.
Unexplained or uncontrolled reactivity change.
4.
Two or more RPI fail to indicate rods inserted after a reactor trip.
5.
SDM ( 1770 pcm (2000 pcm in Mode 5)
Any four (4) [+0.5] each.
Reference (s) 4.12 1.
SHNPP A0P-002, rev. 2, p. 3 and 4.
-Section 4.0 Continued on Next Page-
r-Page 62 Harris 1 February 24, 1986 Points Available OUESTION 4.13 Glyg four (4) symptoms of RCS leakage to containment.
(2.0)
ANSWER 4.13 1.
Containment purge radiation monitor 2.
Containment atmosphere leak detection monitor 3.
Containment area radiation monitor 4.
Condensate collection system 5.
Dewpoint recorder 6.
Containment temperature 7.
Containment presssure 5
8.
Reactor vessel cavity pump level S
9.
Reactor vessel cavity jump flow rate sep
- 10. Reactor vessel cavity pump operation g
Any four (4) (+0.5) each.
Reference (s) 4.13 1.
SHNPP A0P-016, Excessive Primary Plant Leakage, p. 3.
-Section 4.0 Continued on Next Page-
~
Page 63 Harris 1 February 24, 1986 Points Available J
00ESTION 4.14 State the four (4) requirements for unescorted access to a Radiation Control Area.
(2.0)
ANSWER 4.14 1.
General employee training,- Level II
[+0.5]
2.
Whole body count [+0.5]
3.
Dosimetry [+0.5]
4.
Sign in on GRWP or SRWP [+0.5]
Reference (s) 4.14 1.
SHNPP AP-503, Entry into Radiological Areas, p. 11.
4 a
3
-Section 4.0 Continued on Next Page-
Page 64 Harris 1 February 24, 1986 Points Available OUESTION 4.15 Allowable radiation exposure limits given in 10 CFR 20 without an NRC Form 4 are as follows:
a.
Whole body.
rem / quarter (0.5) b.
Skin rem / quarter (0.5) c.
Extremities rem / quarter (0.5)
. ANSWER 4.15 a.
1.25
[+0.5]
b.
7.5
[+0.5]
c.
18.75
[+0.5]
Reference (s) 4.15 1.
SHNPP RP-LP-1.3, Radiation Protection, p. 9.
-Section 4.0 Continued on Next Page-
r
/.
Page 65 Harris 1 February 24, 1986 Points Available OUESTION 4.16 Answer each item TRUE or FALSE regarding procedures for placing a Hydrogen Recombiner in service after a LOCA.
a.
A power factor must be determined depending on the post-LOCA pressure in containment and the pre-LOCA temperature.
(0.5) b.
One Hydrogen Recombiner must be placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after LOCA or when hydrogen concentration reaches 3%.
(0.5)
ANSWER 4.16 a.
TRUE.
[+0.5]
b.
FALSE.
[+0.5]
Reference (s) 4.16 1.
SHNPP OP-125, Post Accident Hydrogen System, p. 5.
QUESTION ~4.17 Anmer TRUE or FALSE. While walking through the plant you discover a valve or switch out of position.
You are required to correct the situation and then report it to the Shift Foreman.
(0.5)
ANSWER 4.17 FALSE.
[+0.5]
Reference (s) 4.17 1.
SHNPP A0P-027, Response to Acts Against Plant Equipment,
- p. 3.
-Section 4.0 Continued on Next Page-
3.-
Page 66 Harris 1 February 24, 1986 Points Available OUESTION 4.18 If CCW is lost to the RCP motor coolers, what are the immediate operator actions?
(1.5)
ANSWER 4.18 1.
Trip the reactor (and turbine)
[+0.5]
2.
Stop all RCPs within 10 minutes or if any bearing g
temoerature reaches 190 F [+0.5]
0+ t o 3.
Maintain seal injection [+0.5]
Reference (s) 4.18 1.
SHNPP A0P-14, p. 4.
ACF-Cit, #GP A la d m d 0 s > w5e1 lqff 7.
f OUESTION 4.19 During a normal startup you are experiencing low AFW flow from the Condensate Storage Tank. Should you switch to the Emergency Service Water supply? Exclain.
(1.0)
ANSWER 4.19 No.
[+0.5] Service water is to be used only in extreme emergencies because it contains chlorine (which is damaging to pressurized systems containing stainless steel).
[+0.5]
Reference (s) 4.19 1.
SHNPP SD-137, AFW System, p. 5.
-Section 4.0 Continued on Next Page-
-c
- d 2,
Page 67 Harris 1 February 24, 1986 Points Available OUESTION 4.20 After the plant is operating at a constant power level and has been in equilibrium for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, you receive a call from an Equipment Operator saying he is taking one of the Instrument Air Compressors off line. Will this interfer with your operation of the plant?
(0.5)
ANSWER 4.20 No.
(At no time after plant startup should the entire Instrument Air System be removed from service, but it is acceptable to -
secure one compressor.) [+0.5] (Based on caution in procedure.)
Reference (s) 4.20 1.
SHNPP OP-151, Compressed Air Operating Procedure, p. 10.
QUESTION 4.21 After a depressurization accident in which a steam void has formed in the reactor vessel head, you are unable to establish conditions permitting restart of an RCP. You are preparing to emergency borate in natural circulation when the supervisor reminds you to be certain that all three (3) S/Gs are steaming.
Exolain why steaming of all S/Gs is necessary in these circumstances.
(1.0)
ANSWER 4.21 Due to low RCS flow rates encountered during natural circulation, all available S/Gs must be steamed, prior to verifying RCS boron concentration, to allow for sufficient chemical mixing in the RCS.
[+1.0]
Reference (s) 4.21 1.
SHNPP E0P-EPP-6, Natural Circulation Cooldown with Steam Void in Vessel with RVLIS, Note, p. 5.
-Section 4.0 Continued on Next Page-
h
^-
Page 68 Harris 1 February 24, 1986 Points:
Available
'0JESTION 4.22 Answer TRUE or FALSE to the following.
(2.0) a.
If off-site frequency is unstable, the EDG's must be parallelled with off-site power to prevent losing a 6.9 kV emergency bus.
b.
Upon indication of a RCS leak, safety injection is initiated when leakage exceeds the capacity of one charging pump.
c.
One'of the immediate actions for a high VCT level is to secure letdown, d.
A temperature difference between symmetric thermocouples is an indication of a misaligned rod.
ANSWER 4.22 a.
FALSE.. [+0.5]
b.
FALSE.
[+0.5]
c.
FALSE.
[+0.5]
d.
TRUE.
[+0.5]
Reference (s) 4.22 1.
SHNPP, A0P-029, p. 5.
2.
SHNPP, A0P-016, p. 6.
3.
SHNPP, A0P-003, p. 4.
4.
SHNPP, A0P-001, p. 31.
-End of Section 4.0-
-End of Exam-L
e e-*C,
?
e-EQUATION SHEET
+
Where my = m2
'(density)1(veloci.ty)1(area)1 = (density)2(velocity)2(area)2 2
KE = mv PE = mgh PE +KE +P V 1 i = PE +KE +P Y22 where V = specific -
i i
2 2
"If volume P = Pressure Q = mc (Tout-Tin)
Q = UA (T
-Tstm)
Q = m(h -h )
p ave i 2 P = P 10(SUR)(t) p = p e /T SUR = 26.06 T = (B-p)t t
o o
T p
delta. K = (Kef f-1)
CR (1-Keffl) = CR (1-Keff2)
CR = S/(1-Keff) 1 2
M = (1-Keff1)
SDM = (1-Keff) x 100%
(1-Keff2)
Keff 1 = A e-(decay constant)x(t)
In (2) 0.693 A
decay constant
=
=
g t
t 1/2 1/2 Water Parameters Miscellaneous Conversions 10 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /hr 3
3 6
Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 f t-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec l
['
3
%j^
f)#
U. S. NUCLEAR REGULATORY COFNISSION SENIOR REACIOR OPERATOR LICENSE EXAMINATION f
Facility:
Harris 1 Reactor Type:
Westinghouse - PWR Date Administered: February 24, 1986 Examiner:
R. E. Schreiber/R. L. Gruel Candidate:
ANSWER KEY INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers.
Write answers on one side only.
Staple 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 each category and a final grade of at least 80%. Examination papers wi.11 be picked up six (6) hours after 'the examination starts.
Category
% of Candidate's
% of Value Total Score Cat. Value Category 26 3 30
_ -Ws-
- 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 29 2.54
_V 25
- 6. Plant System Design, Control and Instrumentation 26 2.2. 8
-s--
- 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control "iS 9 29 -
.PP
- 8. Administrative Procedures, Conditions, and Limitations M&
TOTALS Final Grade All work done on this examination is my own; I have neither given nor received
)
aid.
I Candidate's Signature t
V[
w
'a.e o
v Page 1 Harris 1 February 24, 1986 Points Available 5.0 THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS AND THERMODYNAMICS (30.0)
OUESTION 5.01 Describe how (and why) the following will respond to a loss of natural circulation flow following a reactor trip from 100%
equilibrium conditions.
a.
RCS wide range temperature difference (AT)
(1.0) b.
Relationship between Tcold and P (1.0) steam ANSWER 5.01 a.
AT temperature increases [+0.7] as T increases (as hot boiling occurs in core) and T remains relatively Tao, S No e loe luv cor, Ed.'r3 w.tr,
c constant [+0.3].
b.
T will n t follow Psteam [+0.7].
P will decrease cold steam (as boiloff occurs in S/G) while T remains relatively cold constant [+0.3].
Reference (s) 5.01 1.
General Physics, HT&FF, pp. 356-357.
2.
SHNPP FF-H0-1.3, p. 5.
-Section 5.0 Continued on Next Page-4
F
~
c' e
r Page 2 Harris 1 February 24, 1986 O
Points Available QUESTION 5.02 The plant is operating at 30% power, rod control in manual, turbine control in automatic, when loop 1 RCP trips. Assuming no reactor trip or operator actions, indicate whether the following parameters will be HIGHER or LOWER at the end of the transient compared to their initial values, a.
- 2 S/G pressure (0,5) b.
- 3 RCS loop flow (0.5) c.
T in lo p 1 (0.5) c d.
Th in loop 2 (0.5)
ANSWER 5.02 a.
LOWER [+0.5]
b.
HIGHER [+0.5]
c.
LOWER [+0.5]
d.
HIGHER [+0.5]
Reference (s) 5.02 1.
SHNPP, Heat Transfer and Fluid Flow Lesson Plans (Integrated Knowledge).
I
-Section 5.0 Continued on Next Page-J
F i.
e Page 3 Harris 1 February 24, 1986 Points Available OUESTION 5.03 Select the correct statement regarding Net Positive Suction Head for a centrifugal pump.
41 n)-
(a.) The available NPSH generally parallels the pump characteristic curve by decreasing as flow increases.
If,0)
L (b.) Minimim NPSH recuired for the pump decreases with decreasing flow rate.
(c.) If total head is just equal to fluid saturation pressure, the pump won't cavitate.
(d.) The dynamic pressure at the pump inlet increases faster with flow rate than the static pressure decreases at the same location.
ANSWER 5.03 (a.)
[+h07 oV b.
Reference (s) 5.03 1.
SHNPP FF-H0-1.1, Fluid Mech. in Pipes and Pumps, pp. 39-40 of 65; Figure FF-TP-60.0.
-Section 5.0 Continued on Next Page-m
'Page 4 Harris 1 February 24, 1986 Points Available 00ESTION 5.04 The reactor is operating at 50% power with rod control in manual and turbine control in automatic when a S/G PORV fails open.
Assuming no reactor trip or operator actions, which of the following best describes the resulting steady-state conditions.
(1.0)
(a.) Final Tavg ) Initial Tavg, Final Power = Initial Power (b.) Final Tavg ) Initial Tavg, Final Power > Initial Power (c.) Final Tavg ( Initial Tavg, Final Power = Initial Power (d.) Final Tavg ( Initial Tavg, Final Power > Initial Power (e.) Final T
= Initial Tavg, Final Power > Initial Power avg ANSWER 5.04 (d.)
[+1.0]
$^
Reference (s) 5.04
^
1.
SHNPP, Heat Transfer and Reactor Theory Lesson Plans (Integrated Knowledge).
4 f
f
-Section 5.0 Continued on Next Page-A
j Page.5 Harris 1 February 24, 1986 Points Available OUESTION 5.05 Letdown water (60 gpm)- at g56 F enters the regenerative heat exchanger and exits at 290 F.
Assumingnormal, steady-sgate, at power operation (total charging flow of 69 gpm at 115 F),
what is the temperature of the charging. water entering the RCS from the regenerative heat exchanger? Show all work and state all assumptions.
(1.0)
ANSWER 5.05
[+0.5]
letdown " charging M ATgg=mc c
AT g(6/A)
AT = AT c
g c m = 60 gpm g
M = 45 gpm (letdown less 24 gpm for RCP seals) c ATc = 266 F (60/45)
U
= 355 F Outlet temperature is therefore 115 + 355 = 470 F [+0.5]
Reference (sl 5.05 1.
SHNPP HT-H0-1.1, Heat Exchangers, pp. 4-5 of 47; HT-TP-49.0.
-Section 5.0 Continued on Next Page-
)
h 3,
e, 5
/
n-
- w
= >
s 1
3 s, s s s
y
=,,
Page 6 Harris 1 February 24, 1986 q
4 Points Available
+
s OUESTION 5.06 What is the correct order of-the boiling phases listed below as;they would occur in a coolant channel'with normal flow and high heat flux?
(1.0) 1..
transition boiling 2.
bulk boiling s
3.^ film boiling 4.
sub-cooled nucleate boiling 3
s_________________2__________________________________________
a.
2,4,3,1 s
s b.
2, 4, 1, 3 c.
4,2,3,1 4
d.
4, 2, 1, 3 ANSE'ER5.06 d'
[+1.0]-
s Reference (s) 5.06 s
1.
SHNPP HT-LP-1.0, pp. 27-28 and Figures HT-TP-56.0, Thermo-TP-60.0, and'Thermo-TP-52.0.
s
-Section 5.0 Continued on Next Page-
's v
r-
-.m.-
h w
I c
Page 7 Harris 1 February 24, 1986 Points Available OUESTION 5.07 In the sketch (below) of a simplified Ideal Rankine Cycle, mark or identify the line segments that represent major heat in, major heat out, and major work out.
(2.0)
I i
3
+-
T 2
G l
S ANSWER E 07 2-3-4 Major heat in
[+0.5 each]
4-5 Major work out [+0.5]
5-1 Major heat out [+0.5]
(1-2 Minorworkin)
Reference (s) E 07 1.
SHNPP Thermo-H0-1.4 Turbine and Rankine Cycle, Figure Thermo-TP-46.0.
-Section 5.0 Continued on Next Page-A
F
'c Page 8 Harris 1 February 24, 1986 Points Available OUESTION 5.08 Exclain why it is important to maintain RCP operation as long as possible after SI initiation, or be assured that natural circulation is occurring, while there is substantial pressure in the primary system.
(0.5)
ANSWER 5.08 Minimizes PTS
[+0.5]
homM SAS [dIl Coo
- C oc) C O Slu,ek {c),r}
o Reference (s) 5.08 bu m kvYAw M [d.f}
1.
SHNPP PTS-H0-1.0, PTS, p. 19.
T W ~I40-t.o,pp. tim T.
L.
n 00ESTION 5.09 The par _t of the core mainly affected by control bank motion is (1.0)
(a.) the small volume around the tips of the rods.
(b.) the region along the length of each rod.
(c.) the upper ha d pet of the core dcwn the-rod tiph (d.) the whole core, including areas away from the rods.
ANSWER 5.09 (a.)
[+1.0]
Reference (s) 5.09 1.
SHNPP RT-H0-1.5, Reactivity Variations, Figure RT-TP-88.0, Revision 1.
-Section 5.0 Continuud on Next Page-
l 3-Page 9-Harris 1 February 24, 1986 Points Available 00ESTION 5.10 Doppler fuel temperature coefficient becomes:
(4.-e)--
(a.) more negative with increasing burnup.
(b.) less negative with increasing fuel temperature.
(g,o)
(c.) more negative with build-in of Pu-239.
(d.)lessnegativewiththeburnoutofPu-240.
ANSWER 5.10 (b.)
{+t10 o e.
4.
Reference (s) 5.10 1.
SHNPP RT-H0-1.5, Reactivity Variations, Fbi ure RT-TP-96.0.
Nw % ; {; s, a & l, e f. a n t vu.
1.
QUESTION 5.11 Will the insertion of a given amount of reactivity to a critical reactor at E0L produce a (LARGER, SMALLER, or THE SAME) startup rate than at BOL.
Explain.
(1.0)
ANSWER S E LARGER [+0.5]. The value of the effective delayed neutron fraction is smaller at E0L. A smaller Beta results in a larger SUR for a given reactivity change.
[+0. 5].
Reference (s) 5.11 1.
SHNPP RT-H0-1.6, Neutron Kinetics, Revision 2, pp. 28-29 of 65.
-Section 5.0 Continued on Next Page-i
F.
~
Page 10 Harris'1 February 24, 1986 Points Available 00ESTION 5.12 For the following conditions, indicate whether the highest tensile stress occurs at the INSIDE WALL, CENTERLINE, or OUTSIDE WALL of the pressure vessel.
0 a.
Cooldown at 50 F/hr
-(0.5) b.
Increasing pressure 100 psig with 0 F/hr heatup rate.
(0.5) 0 c.
Heatup at 100 F/hr.
(0.5)
ANSWER 5.12 i
a.
INSIDE WALL [+0.5]
b.
INSIDE WALL [+0.5]
c.
OUTSIDE WALL [+0.5]
Reference (s) 5.12 1.
SHNPP PTS-LP-1.0, p. 10 and Figures 6, 7, and 8.
l l
L l
-Section 5.0 Continued on Next Page-i l
.e Page 11 Harris 1 February 24, 1986 Points Available OUESTION 5.13 For each of the following,' indicate the most restrictive condition for Shutdown Margin requirements.
a.
Time in core life (BOL, MOL, or E0L)
(0.5) b.
Tavg (COLD, NO LOAD, or FULL LOAD)
(0.5) c.
Accident (ROD EJECTION, STEAM BREAK, or DILUTION)
(0.5)
ANSWER 5.13 a.
EOL [+0.5]
N"
- b. bad b.
NO LOAD [+0.5]
/-y
/%4 f
//w.bh c.
STEAM BREAK [+0.5]
c..
Reference (s) 5.13 1.
SHNPP TS, p. B 3/4 1-1. Me3 /-f 2.
Am sed Ts Atde S'
- [
hM Y thak nob 5, Q[
- f L.
- cold, C :5 k;lEh p
grit &S * -
V
-Section 5.0 Continued on Next Page-
..n
Page 12~
Harris 1 February 24, 1986 Points Available OUESTION 5.14 With no fuel in the reactor, the count rate is 100 cps. After loading six fuel assemblies, the count rate is 800 cps. A new detector, farther from the core, is now used and it has a count rate of 500 cps. Four more fuel assemblies are loaded, increasing the count rate to 1000 cps. What is the value of k yf?
~(1.0) e ANSWER 5.14 Moving the detector changes the base count rate:
C4 (new)
Co (new) = Co Cj
= (100 cps) 80
= 62.5 cps
[+0.5]
ps After loading four more fuel assemblies:
C (new) keff = 1 -
Cj
=1-
= 0.9375
[+0.5]
Alternately:
The value of keff with six fuel elements is:
keff = 1 -
= 1 100 c $ = 0.875
[+0.5]
9 cps After loading four more fuel assemblies:
C keff2 = 1 -
(1 - keffi)
=1-(1 - 0.875)
= 0.9375
[+0.5]
Reference (s) 5.14 1.
SMNPP RT-H0-1.7, Subcritical Reactor Theory, Revision 2, pp. 29-31 of 52.
-Section 5.0 Continued on Next Page-
)
{
Page 13 Harris 1 February 24, 1986-Points Available OUESTION 5.15 Ei.y.e two (2) means for flattening the radial flux distribution in subsequent core loadings.
(1.0)
ANSWER 5.15 1.
The distribution of burnable poison rods.
[+0.5]
2.
The loading of fuel with higher enrichment toward the outside of the core.
[+0.5]
A-cqt Ckder laed si (CMvh pew, I M1' ~
tW
/- -
./--4
/ : -
~, ' 2, - -'
-A.;, ;_.,~ __ -
,M
+,
u.
i r.
..r_.
r_
v --
Reference (sf 5.'15
~
~
1.
SHNPP RT-H0-1.10, MTC and Total Power Defect, Revision 2, p.19 of 41.
2.,
SHNPP RT-H0-1.8, Core Construction, Revision 2, p. 21 of 36.
QUESTION 5.16 List the three (3) components of the power coefficient in order of decreasing reactivity contribution. Assume BOL.
(1.5)
ANSWER 5.16 1.
Doppler (fuel temperature) 2.
Moderator temperature 4 v W c.
di A
iv M 3.
Void
% rdkb
- p%
w
[+0.3] for component, [+0.2] for order.
Reference (s) 5.16 1.
SHNPP RT-LP-1.10, pp. 14-15 and RT-TP-197.0.
- p. b o
S.
-Section 5.0 Continued on Next Page-
Page 14 Harris 1 February 24, 1986 Points Available OUESTION 5.17 For the Moderator Temperature Coefficient (MTC), match the parameter change in Column A to the direction it will change the MTC in column 8.
(2.0)
Column A Column B 1.
Moderator temperature increases a.
More negative 2.
Boron concentration increases b.
Less negative 3.
All rods in versus all rods out c.
No-effect 4.
Flux shifting towards edge of core ANSWER 5.17 1.
a.
[+0.5]
2.
b.
[+0.5]
3.
a.
[+0.5]
4.
a.
[+0.5]
Reference (s) 5.17 1.
SHNPP RT-H0-1.10, pp. 20-22.
-Section 5.0 Continued on Next Page-
'n.
Page 15 Harris 1 February 24, 1986 Points Available OUESTION 5 18 How will each of the following change the Quadrant Power Tilt Ratio (INCREASE, DECREASE, or REMAIN THE SAME)? Consider each item separately and assume initial QPTR.is 1.00.
a.
An axial xenon oscillation.
(0.5) b.
Control rod misaligned 12 steps below bank.
(0.5) c.
Overfeeding one system generator.
(0.5)
ANSWER 5.18 a.
REMAINS THE SAME [+0.5]
b.
INCREASE [+0.5]
c.
INCREASE [+0.5]
Reference (s) 5.18 1.
SHNPP RT-H0-1.15, Nuclear Power Distribution, pp.12-20 of 34.
-Section 5.0 Continued on Next Page-
?
Page 16 Harris 1 February 24, 1986 Points Available OUESTION 5.19 The reactor is subcritical with D Control Bank at 72 steps.
An ECP has just been run that shows 450 pcm are needed to reach criticality and be on a modest ramp toward 108 amps. use the curve for Integral Worth to determine the required bank position. Assume no change in boron concentration or xenon.
(1.0)
C4-ta SE d dt m
( ht e, W ult p &lds Lt A %(.
ANSWER 5.19 At 72 steps the total pcm in the rods is 1050 [+0.4], subtracting 450 pcm gives 600 [+0.2], at which the D b~ank is at 112 steps (accept 108 to 116) [+0.4].
Reference (s) 5.19 1.
SHNPP RT-H0-1.13, Control Rod Reactivity Effects, Revision 2, p.14 of 27.
i
-Section 5.0 Continued on Next Page-
Page 17 Harris 1 February 24, 1986 Points Availabla QUESTION 5.19 (Continued) c y
! [
j i
i
! CURVE A B INTEGRAL WORTH
.000 CONTROL BANKS HZP,BOL. EQUILIBRIUM XENON 5500-I l
5000-4500 i
i l
4000-REV:0 13APR78 3500 i.
An U
f l
[n i ING A
E t
y 2'00-i, i
ll Jlj
!l i
l
{
}
h,lI l
I
, 'j l
-I 2000 i
l 118 I
(
l 1500-
{
l l
l I
I I
' l l
l l
.M' l
- .!!i.
6000 l
.I i
i
' l l
J O
I I
l l
[
5 '
l 0-
,7- "j 1.."q l..q I I BANK A*O 40 80 12 0 160, 200l 228 l'
!'s ! ! ls 8
8 a l: es e CONTROL BANK INTEGRAt.
e!
es lj:
80 120' 1601200 228 s l e
a g
a ROD WORTH e
DANK B - 0 40 a8
, ea R T. T P-232.1 iREV-O BANK C 40 8'O l
12 16 0 228 p -.
hn BANKD-*-0 40 80 12 0 1 200 228 K, CU A 2-B CONTROL BANK POSITION (STEPS)
APR AV OATE v/n/r/
'. ~
TRAINING USE ONLY d
-Section 5.0 Continued on Next Page-
Page 18 Harris 1 February 24, 1986 Points Available r
OUESTION 5.20 What are the three (3) purposes of establishing Control Rod Insertion Limits?
(1.5)
ANSWER 5.20 1.
To minimize the consequences of a rod ejection accident.
[+0.5]
A-cc gif- '. N d AliS 6-(Q n m eCf f.d. LF3 2.
To guarantee sufficient shutdown margin.
[+0.5]
3.
To provide suitable axial flux distribution.
[+0.5]
Reference (s) 5.20 1.
SHNPP RT-H0-1.13, Control Rod Reactivity Effects, p. 20 of 27.
QUESTION 5.21 a.
What is the maximum Quadrant Power Tilt Ratio (QPTR) for unrestricted plant operations?
(0.5) b.
A 2-hour time allowance for operation with a QPTR greater than a. above is allowed. What is the basis for this time allowance?
(1.0)
ANSWER 5.21 a.
1.02
[+0.5]
b.
Allow (identification and) correction of a dropped (or misaligned) control rod.
[+1.0]
Reference (s) 5.21 1.
SHNPP TS, pp. 3/4 2-11 and B 3/4 2-6.
-Section 5.0 Continued on Next Page-
Page 19 Harris 1 February 24, 1986 Points Available OUESTION 5.22 Rod withdrawal or boron dilution may be done during Xe decay.
Why. doesn't this violate the rule of not adding positive reactivity by more than one method at a time?
(1.0)
ANSWER 5.22 Xe decay effect is very slow, compared to rod or boron changes.
(It is allowed in the precautions of GP-004.)
[+1.0]
Reference (s) 5.22 1.
SHNPP GP-004, Reactor Startup from Hot Standby, p. 5.
-Section 5.0 Continued on Next Page-t
Page 20 Harris 1 February 24, 1986
' Points Available OUESTION 5.23 Which of the following statements concerning the reactivity-values of equilibrium (at power) samarium and peak (after shutdown) samarium is correct? Assume shutdown occurs from equilibrium conditions.
(1.0)
(a.) Equilibrium samarium is INDEPENDENT of power level; peak samarium is INDEPENDENT of power level.
(b.) Equilibrium samarium is INDEPENDENT of power level; peak samarium is DEPENDENT on power level.
(c.) Equilibrium samarium is DEPENDENT on power level; peak samarium is INDEPENDENT of power level.
(d.) Equilibrium samarium is DEPENDENT on power level; peak samarium-is DEPENDENT on power level.
ANSWER 5.23 (b.)
[+1.0]
Reference (s) 5.23 1.
SHNPP RT-LP-1.11, p. 20 and RT-TP-218.0.
-Section 5.0 Continued on Next Page-
~ '..
'i Page 21 Harris 1 Febiuary 24, 1986 Points Available OUESTION 5.24 Answer TRUE or FALSE. The time required to reach the peak value of xenon following a trip increases with increasing initial xenon concentration.
(0.5)
ANSWER 5.24 TRUE. [+0.5]
Reference (s) 5.24 1.-
SHNPP RT-LP-1.11, p. 14.
-End of Section 5.0-
,o
~
Page 22 Harris 1 February 24, 1986 Points Available 6.0 PLANT SYSTEM DESIGN. CONTROL AND INSTRUMENTATION (30.0)
OUESTION 6.01 Answer TRUE or FALSE. When the PZR level reaches 5% above i ~
reference, the backup heaters turn on.
(0.5)
ANSWER 6.01 TRUE.
[+0.5]
Reference (s) 6.01 1.
SHNPP PZRLC-LR-1.0, Revision 1, p. 14.
I
-Section 6.0 Continued on Next Page-a 4
.. ~ - -.
.~.-...-..~.,-,.,----,~,,--,-,--.------n-,,-,
n-
,=-,,,.
n.w-
,f Page 23 Harris 1 February 24, 1986 Points,
Available OUESTION 6.02 Regarding the S/G Level Control System:
a.
Idha.t primary or secondary system information is used by the controller for the feedwater bypass valve that is not used by the main feedwater valve controller?
(1.0) b.
lihal primary or secondary system information is used by the controller for the main feedwater valve that is ont used by the byoass feedwater valve controller?
(1.0) c.
When the bypass system has exclusive control below 15%
power, what two (2) valves must be shut.
(1.0).
ANSWER 6.02 l,0 of I.O N44 nuclear power (4GE),3 anticipated steam demand [4 FEE 5].
a.
Steam" pressure [k,(density { mpensation for stea b.
flow)[Witi; feed flow [+0.3];
0, 6 c.
The main control valve [+0.5] and the feedwater isolation valve [+0.5]; #f t
(,3 435. jp; g /g {0,Q Reference (s) 6.02 1.
SHNPP SGWLC-LP-1.0, Revision 1: Figure SGWLC-TP-1.0, Revision 2; and Figure SGWLC-TP-2.0, Revision 2.
[6,6)
[h (s
ok kee
[cw e.e t c r e
- s. w A D 4
- 4. 5L/S5s fiU3 hon **n {+.cof or
- d. fIm bD)
{J. FIrv (t.U]
-Section 6.0 Continued on Next Page-
v Page 24 Harris 1 February 24, 1986 Points Available OUESTION 6.03 Exolain how the power mismatch circuit of the Rod Control System responds to a difference between reactor power and turbine power.
(1.0)
ANSWER 6.03 The rate comparator has an output proportional to rate of change. But zero rate change, even if with a constant difference, would result in no response [+1.0].
(Therecould be a calibration error, the steam dumps or PORVs or S/G reliefs could be open, causing a real but constant difference between reactor power and turbine power.)
Reference (s) 6.03 1.
SHNPP RODCSLP-1.0, Revision 1, p. 17.
-Section 6.0 Continued on Next Page-
Page 25 Harris 1 February 24, 1986 Points Available OUESTION 6.04 Eire 10 of the 11 control or protective functions supplied by the output from the Power Range Detectors.
(2.5)
ANSWER 6.04 1.
Overpower differential temperature (0 PAT) setpoint.
2.
Overtemperature differential temperature (OTAT) setpoint.
3.
Overpower Trip-Low Range (25%).
4.
P-10 nuclear at power permissive.
5.
Automatic rod control (N-44 only).
6.
Steam generator water level control.
OverpowerTrip-HighRange(b).
7.
8.
C-2 high flux rod stop (103%).
9.
P-8 three loop flow permissive (( 49%).
10.
Power range Positive Rate Trip.
11.
Power range Negative Rate Trip.
Any 10 (+0.25) each.
Reference (s) 6.04 1.
SHNPP NIS-1.0, Revision 1, pp. 16-17.
-Section 6.0 Continued on Next Page-
g Page 26 Harris 1 February 24, 1986 Points Available OUESTION 6.05 tion would the turbine behave differently between a runback associated witp OTAT-5, and runback due.to loss of 1Ml4:Uf in-lowW ctppu f3 u.Wgs ?
(1.0)
ANSWER 6.05 Exceeding the limit OTAT-3% causes an intermittent runback [+0.5]
(200%/ min for 1.5 sec, 30 sec hold, repeat). - Loss of ined tue c[g pup causes _ a continuous runback [+0.5].
Reference (s) 6.05 1.
SHNPP RPS-LP-1.0, Revision 3, p. 21.
2.
SHNPP SD-131.5, Revision 1, p. 22.
i d
-Section 6.0 Continued on Next Page-J
+, -
--,--,~--.w----.w.--,,,.
{
.Page 27 Harris 1 February 24, 1986 Points-Available OUESTION 6.Dji comolete the following table of selected reactor trips:
(2.0)
Trin Initiated At logic coincidence SR Hi flux cps
/
Hi PIR pressure psig
/
Hi PZR level
/
Lo-Lo S/G level
% NR
/
ANSWER 6.06 5
10
-[+0.25]
1/2 [+0.25]
2385 [+0.25]
2/3
[+0.25]
92%
[+0.25]
2/3
[+0.25]
-WW [+0.25]
2/3
[+0.25]
2b 3 M 4 xn.c/ f Iod Reference (s) 6.06 1.
SHNPP RPS-LP-1.0, Revision 3, pp. 10, 14-16.
-Section 6.0 Continued on Next Page-
Page 28 Harris 1 February 24, 1986 Points Available OUESTION 6.07 List the six (6) functions of the safeguards devices that are
-NOT tested at power, but are given continuity checks instead.
(1.5)
ANSWER 6.07 1.
[+0.25]
2.
generator trip
[+0.25]
3.
RCP trip [+0.25]
4.
steam line isolation
[+0.25]
5.
feedwater isolation
[+0.25]
6.
Reference (s) 6.07 1.
SHNPP RPS-LP-1.0, Revision 3, p. 9.
-Section 6.0 Continued on Next Page-
W Page 29.
Harris 1 February 24, 1986 Points Available OUESTION 6.08-What indications should be checked if level transmitter, LT-115, on the VCT has failed high?
(1.0)
ANSWER 6.08 LI-115 indicating full, VCT high level alarm '
LI-112 (local) shows actual VCT level decreasing Auto M/U and auto emergency M/U actuation inoperable -
LCV-115A(ICS-120)'divertedtoBRS
[+0.2 each]
Reference (sl 6.08 1.
SHNPP A0P-003, Revision 1 Table I, p. 9.
QUESTION 6.09 What three (3) signals that will cause the Control Roofn Area HVAC to isolate?
(1.5)
ANSWER 6.09 1.
[+0.5]
2.
Radioactivity at air intakeg [+0.5]
3.
Chlorine at an intake or in storage area [+0.5]
4 S Nde Ab i%M, Reference (sl 6.09 1.
SHNPP SD-173, Control Room Area HVAC System, p. 3.
hknRd it%%LO, / Ofn{W
- 2.,
't Section 6.0 Continued on Next Page-i m -., - _,.-rv m._e-e---.,r
,,,.m r-
,...--w-we.,
,.y.
-n.
,,,,-,--,--.,,,m.
,-_.-r,
---_w,.
e Page 30 Harris 1 February 24, 1986 Points Available OUESTION 6.10 List the five (5) purposes of the containment Ventilation System.
(1.5)
ANSWER 6.10 1.
Remove airborne radioactivity and radioactive iodine by recirculating atmosphere through filters.
[+0.3]
2.
Maintain low concentration of radioactivity by low rate purging.
[+0.3]
3.
Reduce radioactivity concentration by high rate purging to allow personnel entry.
[+0.3]
4.
Maintain Egative pressure during all modes.
[+0.3]
5.
Relieve vacuum if it exceeds limits.
[+0.3]
Reference (s) 6.10 1.
SHNPP SD-168, Ctmt. Vent and Vac. Relief, p. 4.
-Section 6.0 Continued on Next Page-
o Page 31 Harris 1 February 24, 1986 Points Available OUESTION 6.11 Answer each item TRUE or FALSE with regard to the plant electrical system:
a.
During shutdown conditions, offsite power is supplied through the startup transformers.
(0.5) b.
Relays protect the generator against faults, overloads, grounds, and motoring.
(0.5) c.
The 6.9 KV buses supply all major equipment of 500 hp or greater.
(0.5) d.
After reactor trip, power source is changed from unit to startup transformers with auto fast dead bus 8 cycle delay transfer.
(0.5)
ANSWER 6.11 a.
TRUE.
[+0.5]
b.
TRUE.
[+0.5]
c.
TRUE.
[+0.5]
d.
TRUE.
[+0.5]
Reference (s) 6.11 1.
SHNPP SD-156, Plant Electrical System, pp. 3, 9,14.
-Section 6.0 Continued on Next Page-
~ - -
,o Page 32 Harris 1 February 24, 1986 Points Available OUESTION 6.12.
Choose the correct statement regarding the Emergency Diesel
' Generators.
(1.0)
(a.) One EDG is capable of supplying all plant emergency power requirements for both safety trains.
(b.) Sufficient fuel is available for 2 weeks full load operation.
(c.) The air start system has capacity for eight attempts before,
recharging of the air receiver is necessary.
(d.) EDGs reach rated speed / volts 10 seconds'after start and are fully loaded within 45 seconds.
ANSWER 6.12 a
(d.)
[+1.0]
Reference (s) 6.12 1.
SHNPP SD-155, EDG System, pp. 5, 7.
.4 s
J;
-Section 6.0 Continued on Next Page-9
.wis-.%m-pre-r
-r-e-*w--vw6
=-r-7d-a
-+-v1-
-= --
T
'o Page 33 Harris 1 February 24, 1986 Points Available OUESTION 6.13 Hydrogen limits in containment are in the range:
M (a.) 1% to 3%
g (b.) 4% to 6%
g (c.) 7% to 9%
(d.) 10% to 12%
ANSWER 6.13 (b.)
[+1.0]
0 9-d..
Reference (s) 6.13 1.
SHNPP SD-125, Post Accident Hydrogen System, p. 3.
-Section 6.0 Continued on Next Page-i l
.v i
t
(>
j EQUATION SHEET Where mi = m2 1
f (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2 f
i 2
PE +KE W V t i = PE +KE +P Y22 where V = specific KE = mv PE = mgh 2
2 i
1 7
volume P = Pressure Q = mc (Tout-Tin)
Q = UA (T
-Tstm)
Q = m(h -h I p
ave i 2 P = P 10(SUR)(t) p, p e /T SUR = 26.06 T = (B-p)t t
o c
T p
I CR (1-Keffl) = CR (1'Keff2)
CR = S/(1-Keff delta K =.(Keff-1) 1 2
i M = (1-Keffl)
SDM = (1-Keff) x 100%
(1-Keff2)
Keff 1 = A e & cay connan W W in (2) 0.693 A
decay cor.stant
=
=
g t
t1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps f
1 gallon = 3.78 liters 1 kg = 2.21 lbs 3
1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /hr 6
3 1 MW = 3.41 x 10 Btu /hr Density =62.4lbg/ft Density = 1 gm/cm 1 Btu = 778 f t-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degres.1 C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec Ao
,0 Page 34 Harris 1 February 24, 1986 Points Available OUESTION 6.14 a.
What two (2) features of the Containment Spray System assure adequate mixing of the sodium hydroxide with the spray water?
(1.0) b.
What is the purpose of the Cavitating Venturis in the CSS 7 (0.5) c.
Describe the design feature of the CSS that allows access to the sump isolation valves, yet gives them the protection of containment.
(0.5)
ANSWER 6.14 a.
The eductors [+0.5] on the recirculation lines take up the NaOH to produce some mixing and pass it through the CSS pumps [+0.5] where it is more thoroughly mixed, b.
These devices provide a backpressure on the CSS pumps to' prevent dama e due to runout in the initially unfilled line.
[+0.5 c.
The isolation valves are in separate containment, connected to the Containment building, but not open to its atmosphere.
[+0.5]
Reference (s) 6.14 1.
SHNPP SD-112, Containment Spray System, pp. 4-6 and Figure CSS-TP-1.0.
-Section 6.0 Continued on Next Page-i l
,e Page 35 Harris 1
' February 24, 1986 Points Available QUESIION 6.15 What is the principal objection to using Service Water rather than the condensate storage tank as a source of auxiliary feedwater?
(1.0)
ANSWER 6.15 The high chlorine content of service water can damage stainless steel systems.
[+1.0]
Reference (s) 6.15 1.
SHNPP SD-137, AFW, p. 5.
-Section 6.0 Continued on Next Page-
4 Page 36 Harris 1 February 24, 1986 Points Available OUESTION 6.16 Answer each item TRUE or FALSE with regard to the CCW System.
(a.) The pressure is greater than the Service Water pressure.
(0.5)
(b.) The internal divisions of the surge tank prevents loss of CCW in both loops if one loop ruptures.
(0.5)
(c.) Relief valves in the system are mounted downstream of components, such as Hxs, to guard against water hammer.
(0.5)
(d.) Leakage to the CCW drain tank may be pumped to a holdup tank or returned to the surge tank.
(0.5)
ANSWER 6.16 a.
TRUE.
[+0.5]
b.
TRUE.
[+0.5]
c.
FALSE [+0.5]
(Guard against expansion due to heatup.)
d.
TRUE.
[+0.5]
Reference (s) 6.16 1.
SHNPP SD-145, CCW System, pp. 5, 8, 9, 17.
l-
-Section 6.0 Continued on Next Page-t
.I,,
~Page 37 Harris 1 February 24, 1986 Points Available OUESTION 6.17 Select the reason why there are eight (8) decay tanks in the Gaseous Waste Processing System, rather than one (1) tank of the same total volume.
(1.0)
(a.) to reduce the site boundary dose in the event of a tank rupture.
(b.) to comply with ASME Pressure Vessel Code. requirements for stress.
(c.) to reduce the radiation dose to personnel working with the tanks.
(d.) to make installation and removal easier in the confined space.
ANSWER 6.17 (a.)
[+1.0]
s Reference (s) 6.17 1.
SHNPP SD-120.7, Gaseous Waste Processing System, p. 5.
-Section 6.0 Ccntinued on Next Page-I
Page 38 Harris 1 February 24, 1986 Points Available OUESTION 6.18 Ely_e the five (5) plant casualties that result in operation of the Auxiliary Feedwater System.
(1.5)
ANSWER 6.18 1.
Loss of Main Feedwater with or without off-site power available.
[+0.3]
M S _ ah de 2.
Main Feedwater piping rupture.
[+0.3]
I
,Il I SU"'
3.
Main Steam piping rupture.
[+0.3]
c.c. MI M.'/
4.
Loss of all AC power.
[+0.3]
(. 3 3 J bM 5.
Loss of Coolant Accident.
[+0.3]
( o.r 3 Yp M stk /
y Reference (s) 6.18 1.
SHNPP SD-137, Auxiliary Feedwater, p. 4.
QUESTION 6.19 Exolain why the isolation valve on the RCP No. 1 seal leakoff line must be shut when the RCS pressure is below 100 psig.
(1.0)
ANSWER 6.19 Closing the No.1 seal leakoff isolation valve prevents the backflush of contaminants into the pump seal chamber. [+1.0]
Reference (s) 6.19 l
1.
SHNPP RCP-LP-1.0, Reactor Coolant Pumps, p. 21.
l l
1
-Section 6.0 Continued on Next Page-f l/
l
Page 39 Harris 1 February 24, 1986 Points Available OUESTION 6.20 a.
Giya two (2) reasons why a small continuous flow is maintained through both PZR spray lines.
(1.0) b.
Why are loop seals maintained between the PZR and the main control valve for the sprays, as well as between the PZR and the safeties? The reason is not necessarily the same for both.
(1.0)
ANSWER 6.20 a.
To reduce thermal stress when main spray valves are opened
[+0.5], and to keep the coolant within the PZR from differing in chemical cor. centration from the main coolant 4/S]MuMNbee %' A44MW Cd#3
[+0.5. M + t /
Me,.% PB A L 9 othch3 Co.103 b.
The layout (effectively a loop seal) of the spray line is to prevent steam building up all the way back to the control valves [+0.5].
(This is not a leakage problem because the pressure upstream of the valve is somewhat higher than the PZR pressure. CAE, with a continuous flow from the spray bypass, wouldn't this prevent steam buildup?)- A ss mud [asI The loop seal in each of the safety lines is to prevent steam and hydrogen leaking through the valves (because the downstream pressure is much less than PZR pressure).
[+0.5]
Reference (s) 6.2Q 1.
SHNPP SD-100, RCS, pp. 34 and 37.
-End of Section 6.0-i
.c Page 40 Harris 1 February 24, 1986 Points Available 7.0 PROCEDURES'- NORMAL. ABNORMAL. EVERGENCY. AND RADIOLOGICAL CONTROL (30.0)
OUESTION 7.01 Select the correct Emergency Classification for which the following is an important part of the definition:
(1.0)
" Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant."
(a.) Unusual Event (b.) Alert (c.) Site Emergency (d.) General Emergency ANSWER 7.01 (b.)
[+1.0]
Reference (s) 7.01 1.
SHNPP SP-015, Emergency Plan Support, pp. 10-11.
-Section 7.0 Continued on Next Page-
.~_
l
-Page 41 Harris 1 February 24, 1986 Points Available OUESTION 7.02 Answer TRUE or FALSE.' Loss of one (1). power supply, SA or
.SB, during Mode 5 will make both RHR loops inoperable.
(0.5)
ANSWER 7.02 TRUE.
(Each power supply provides power to one RHR suction valve in nadi RHR loop.)
[+0.5]
Reference (s) 7.02 1.
SHNPP A0P-020, Revision 1, Loss of RHR, p. 8.
I 4
l 1
-Section 7.0 Continued on Next Page-4
Page 42 Harris 1 February 24, 1986 Points Available QUESTION 7.03 Arrange in the correct order the operator actions necessary I
after a loss of AC power to busses 1A-SA and IB-SB.
a.
Try to restore power to any AC emergency bus.
O b.
Verify turbine trip.
c.
Verify AFW flow.
d.
Verify reactor trip.
e.
Verify RCS isolated.
ANSWER 7.03 c)
F 1.
d.
[z G L] o, b uOdbMANM% L
- 2.
b.
M 0, b I *
,S.
3.
e.
-[* W T] o.t 4.
c.
[.5E;;I] o. I 5.
a.
[M o. t 4 lf.(frit-Reference (s) 7.03 1.
SHNPP E0P-EPP-001, Loss of AC Power to Buses 1A-SA and 1B-SB.
-Section 7.0 Continued on Next Page-I
Page 43.
Harris 1 February 24, 1986 Points Available OUESTION 7.04 What is the caution statement in the SI termination procedure regarding loss of offsite power after SI reset?
(1.0)
ANSWER 7.04-Manual action may be required to restart safeguards equipment.
[+1.0]
Reference (s) 7.04 1.
SHNPP E0P-EPP-008, SI Termination, pp. 3, 16.
QUESTION 7.05 Answer TRUE or FALSE.
If RCS subcooling is lost during depressurization after LOCA, RCS depressurization must be stopped until subcooling is restored.
(0.5)
ANSWER 7.05 FALSE., [+0.5]
(Cautionsaysdon'tstop.)
Reference (s) 7.05 1.
SHNPP E0P-EPP-9, Post LOCA Cooldown and Depressurization,
- p. 7.
-Section 7.0 Continued on Next Page-
e
~
Page 44 Harris 1 February 24, 1986 Points Available OUESTION 7.06 Which of the following is not an SI termination criteria, as given in foldout A of the E0P for Critical Safety Function Status Tree?
(1.0)
(a.) RCS subcooling less than 25 F [30 F].
(b.) Narrow range level in at least 1 intact S/G greater than 10% [40%].
(c.) RCS pressure stable or increasing.
(d.) PZR level greater than 10% [45%].
ANSWER 7.06 (a.)
[+1.0]
Reference (s) 7.06 1.
SHNPP E0P-CSFST, Foldout A.
-Section 7.0 Continued on Next Page-i
I:
e Page 45 Harris 1 February 24, 1986 Points Available OUESTION 7.07 What '
(47=RCS dilution paths or components must be isolated after an ATWS?
(2.0)
ANSWER 7.07 1.
Reactor makeup water pumps. -[.0.3]
{f,oj Q gg 2.
Reactor makeup water supply (valve) -[ 0.5])
nd ludL
(, o 3.
BTRS
[4W5]
y 4.
Boric acid batch tank [+0.5] Mi.
C o.f eac h 3 (P)
Reference (s) 7.07 1.
SHNPP E0P-FRP-5.1, Response to Nuclear Power Generation /ATWS, 6.
OUESTION 7.08 Of the three (3) methods of cooldown after SGTR; 1.e., backfill, blowdown, steam dump, state which is the most desirable and why.
(1.5)
ANSWER 7.08 Backfill [+0.5].
The spread of radioactivity to other systems is minimized [+1.0].
(Boron concentration adjustments may be necessary.)
Reference (s) 7.08 1.
SHNPP E0P-EPP-17, 18, 19, Cooldown after SGTR.
-Section 7.0 Continued on Next Page-i i
r a'
Page 46 Harris 1 February 24, 1986 Points Available OUESTION 7.09 Which of the following is NOT a symptom of-auto rod control bank failure after an increase in turbine load?
(1.0)
(a.) Increasing T and possible alarm.
avg (b.) Low pressurizer pressure.
(c.) Rod control urgent alarm.
(d.) Failure of rods to move on T
-T than dead band.
avg ref deviation greater ANSWER 7.09 (a.)
[+1.0]
Reference (s) 7.09 1.
_SHNPP A0P-001, Failure of a Control Bank to Move, p. 5.
-Section 7.0 Continued on Next Page-i
4 Page 47 Harris 1 February 24, 1986 Points Available OUESTION 7.10 Even if the Reactor Auto Makeup Control is inoperative, it is still possible and permitted to operate the plant.
Exolain how this is done.
(2.0)
ANSWER 7.10 The plant may remain in operation with the reactor makeup control inoperative if normal plant makeup requirements are satisfied by alternately injecting primary makeup water (ICS-274) and boric acid (ICS-203 and ICS-287) through the manual valves
[+1.0].
Emergency boric acid supply valve (ICS-278) may also be used but it will provide a large quantity of 4 wt% boric acid in a short time [+0.5].
Manual valve (ICS-526) may be used to supply water directly from the boric acid tank to the suction of the CSIPs via gravity feed if the boric acid pumps are not operable [+0.5]. higW Q% % kg {c.p}
S hl S C O,6,
NW 5T {0,y}
Reference (s) 7.10 1.
SHNPP A0P-003, Malfunction Reactor Makeup Control, p. 8.
J
-Section 7.0 Continued on Next Page-
o o
.o Page 48 Harris 1 February 24, 1986 Pofnts Available OUESTION 7.11 What two (2) additional immediate actions are required after announcing an accidental release of radioactive gas in the Waste Processing Building and ordering an evacuation of that area?
(1.0)
ANSWER 7.11 1.
Have radwaste operator start all filtered exhaust fans in the Waste Building.
[+0.5]
2.
Isolate leak or stop gas addition to ruptured system.
[+0.5]
Reference (s) 7.11 1.
SHNPP A0P-009, Accidental Release of Waste Gas, p. 3.
00ESTION 7.12 If CCW is lost to the RCP motor coolers, state the two (2) actions that must be taken to protect the pumps after the reactor and turbine are tripped.
(1.0)
ANSWER 7.12 1.
Stop all RCPs within 10 minutes or sooner if bearing g
temperature reaches 190 F.
[+0.5]
oe i e 5F 2.
Maintain seal injection flow.
[+0.5]
Reference (s) 7.12 1.
SHNPP A0P-014, Loss of CCW, p. 4.
AOP - ol5, k W M M-ts s.[ Op s, uses Iq5'R 3.
v
-Section 7.0 Continued on Next Page-i
Page 49 Harris 1 February 24, 1986 Points Available OUESTION 7.13 Gfre the four (4) pieces of equipr.ent or functions that must be verified after an RCS pressure increase transient.
(2.0)
ANSWER 7.13 1.
PORVs reseat when pressure drops.
[+0.5]
2.
PZR sprays operate to reduce pressure.
[+0.5]
3.
PZR heaters off.
[+0.5]
4.
PZR level within program.
[+0.5]
Reference (s) 7.13 1.
SHNPP A0P-019, Manfunctions of RCS Pressure Control, p. 5.
00ESTION 7.14 The control room has been evacuated in a non-fire situation.
The R0s are actuating the transfer switches at Panels A and B in the Switchgear Rooms A and B.
What is your initial action at the ACP7 (1.0)
ANSWER 7.14 Place all four (4) XFER ACTUATE switches in the XFER position and release.
[+1.0]
Reference (s) 7.14 1.
SHNPP A0P-004, Revision 1, p. 25.
-Section 7.0 Continued on Next Page-
0 Page 50 Harris 1 February 24, 1986 Points available OUESTION 7.15 During your shift as SCO you are told that the leakoff flows from No. I and 2 seals on RCP-B seem to be increasing.
Later
_[,
the radial bearing temperature begins increasing. Standpipe level stays high. Diagnose what may be happening.
-(+:5)-
4,[:
ANSWER 7.15 fi-est No. 2 seal began to fail [+0.5], -then No.1 seal failed
- I fed
'I" also [+0.5].MThe failure of the other seals causes substantial back leakage through No. 3 seal [+0.5]. S +(
- 2. pA= ceded "I *
- L 3# {j" 9 ve4 l [0, F3,ATALP-6 (W(s {0,f}
A.cf 8 S'd M" Reference (s) 7.15 1.
SHNPP A0P-018, RCP Abnormal Conditions, p. 15.
OUESTION 7.16 The reactor operator informs you that one channel of PZR level is drifting down and seems erratic. The other channels are holding program level but charging rate has increased.
Containment radiation levels are rising slowly, but nothing else seems out of the ordinary. What is indicated by these symptoms? In what areas would you look for further, evidence?
(1.5)
ANSWER 7.16 Id A(small LOCA)is apparently in progress [+0.5].
A leaking of the lower fitting of the level tap is most likely [+0.5].
Continued development of the leak would show other Containment symptoms, such as increased temperature, dewpoint, area or purge radiation monitors, condensate collection.
[+0.5]
&ce-Q V V N We4
k-S.
C f n1 s y crf%'s.
A Reference (s) 7.16 1.
SHNPP A0P-016, Excessive Primary Plant Leakage, p. 3.
I j
-Section 7.0 Continued on Next Page-l
.~..
Page 51 Harris 1 February 24, 1986 Points Available OUESTION 7.17 It is necessary to repair a piece of equipment in a radiaticn area.
It has been determined that the following manpower allocations will complete the task in the noted length of time:
1.
One worker - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> duration 2.
Two workers - 20 minute duration 3.
Three workers - 15 minute duration From ALARA considerations, how many workers (one, two, or three) should be assigned to the task?
(1.0)
ANSWER 7.17 2 (lower total dose than other options).
[+1.0]
Reference (s) 7.ll 1.
General ALARA considerations.
2.
SHNPP AP-514, ALARA Job Evaluations, is potential reference (notprovided).
-Section 7.0 Continued on Next Page-i
Page 52 Harris 1 February 24, 1986 Points Available OUESTION 7.18 The operator informs you that the Bank Low-Low Insertion Limit has alarmed. Before you can investigate, T Low-Low Alert av9 alarms. What is your immediate action before further investigation?
(0.5)
ANSWER 7.18 Order Immediate Boration.
[+0.5]
Reference (s) 7.18 1.
SHNPP A0P-002, Revision 2, p. 3.
4 i
J
-Section 7.0 Continued on Next Page-i 1
- - w
- =
Page 53 Harris 1 February 24, 1986 Points Available OUESTION 7.19 AnsEer TRUE or FALSE regarding Fuel Handling Operations.
a.
Spent fuel from other plants may be received and stored at Harris.
(0.5) b.
New fuel may be stored either wet or dry until it is ready for transfer into containment.
(0.5) c.
Cask dccontamination is done in a separate area from the cleaning of handling tools.
(0.5) d.
The short handling tool is mainly used to move spent fuel from the upender to the spent fuel racks.
(0.5)
ANSWER 7.19 a.
TRUE.
[+0.5]
b.
TRUE.
[+0.5]
My c.
h
[+0.5]
d.
FALSE.
[+0.5] (# w b d'ow(g)
' Reference (sl 7.19 1.
SHNPP SD-115, Fuel Handling System, pp. 4, 16.
-Section 7.0 Continued on Next Page-i I
r Page 54 Harris 1 February 24, 1986 Points Available OUESTION 7.20 After verifving a Gross Failed Fuel Alarm, your next immediate action is:
(1.0)
(a.) Contact Chemistry for RCS sample.
(b.) Notify Radiation Control to survey area.
(c.) Place cation bed in service to reduce activity.
(d.) Reduce reactor power by 25% at maximum rate.
ANSWER 7.20 (d.)
[+1.0]
Reference (s) 7.20 1.
SHNPP A0P-032, High RCS Activity, p. 3.
QUESTION 7.21 Exolain how the startup power ramp is limited after refueling.
(1.0)
ANSWER 7.21 Power ramp rate is restricted after core fuel movement to 3%/HR power level for > power level. (Operation at or above a given from 20% to 100%
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> removes this restriction for power levels up to and including this power level.)
[+1.0]
Reference (s) 7.21 1.
SHNPP GP-005, Power Operation, Mode 2 to Mode 1, p. 7.
-Section 7.0 Continued on Next Page-i l
r Page 55 Harris 1 February 24, 1986 Points Available OUEST 7.22 Wha.t wo (2) additional immediate actions are required after announ ng an accidental release of radioactive gas in the Waste Pr essing Building and ordering an evacuation of that area?
(1.0)
ANSWER 7.22 1.
Have radwaste op ator start all filtered exhaust fans in the Waste Building.
[+0.5]
2.
Isolate leak or stop g addition to ruptured system.
[+0.5]
Reference (s) 7.22 1.
SHNPP A0P-009, Accidental Release o Waste Gas, p. 3.
)Me 6
- 7. h
-Section 7.0 Continued on Next Page-i
r Page 56 Harris 1 February 24, 1986 Points Available QUESTION 7.23 List four (4) conditions which require Emergency Boration.
-(2.0)
ANSWER 7.23 1.
Excessive control rod insertion.
2.
Uncontrolled cooldown.
3.
Unexplained or uncontrolled reactivity change.
4.
Two or more RPI fail to indicate rods inserted after a reactor trip.
~
5.
SDM ( 1770 pcm (2000 pcm in Mode 5). /{_cq$: fd9vd e S@
Any four (4) at [+0.5] each.
Reference (s) 7.23 1.
SHNPP, A0P-002, rev. 2, p. 3 and 4.
'l i
l l
4
-Section 7.0 Continued on Next Page-
-Page 57 Harris 1 February 24, 1986 Points Available OUESTION 7.24 Describe how the initial steps (start RCP if possible) differ in natural circulation cooldown with steam void for RVLIS
/. o available and not available.
(1.E)
ANSWER 7.24 If RVLIS not avaflable, PZR level () 65%) [+0.5] and RCS subcooling () 50 F) [+0.5] are used rather than RVLIS to--
det r-ine that th; upps. Scad t e OM ' [+0.5]. -
Reference (s) 7.24 1.
SHNPP E0P-EPP-6, rev. O, p. 4.
2.
SHNPP E0P-EPP-7, rev. O, p. 4.
-End of Section 7.0-I
f' ig Page 58 Harris 1 February 24, 1986 Points Available 8.0 ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS (30.0)
OUESTION B E Answer TRUE or FALSE.
If the preparer of a Temporary Change
- to a procedure cannot find qualified personnel onsite, telephone approval by qualified personnel is acceptable.
(0.5)
ANSWER 8.01-TRUE.- [+0.5]
Reference (s) 8.01 1.
SHNPP AP-007, Temporary and Advance Changes to Plant -
Procedures, Revision 2, Section 5.2.6, p. 9 of 21.
-Section 8.0 Continued on Next Page-
r t
I
.j t f.
c e / < >,,;-
9:
Page 59
_ Harris 1,.,'
C' February 24, 1986 y'
?
Points '
'/.
Available
~
OUESTION 8.02 f
Which of the following describes the proper identification of
,,e RCAs from lowest to highest radiation levels?
- f
',(1,0) i:
(a.) Low Dose Area, Radiation Area, Neutron RadiatioE Area,
't.
e I"'
~
(b.) Radiation Area, High Radiation Area, Locked.High Radiation ie i Area, Restricted High Radiation Area (c.) Low Dose Area, Radiation Area, High Radiation Area, Neutrpn
/
b',,
Radiation Area l'
m
~
(d.) Radiation Area, High Radiation Area,rRestricted Htgh
./
~^
Radiation Area, Locked High Radiation Area
- j. - A 9
ANSWER 8.02 e
(b.)
[+1.0]
5 y
Reference (s) 8.02
,a 1.
SHNPP AP-503, Entry Into Radiological Areas, Revisionf0, Sections 4.3, 4.4, 4.13-4.17, pp. 5-9 of 25.
f' s
I
.1
)'
-Section 8.0 Continued on Next Page-i p.
r e
- ___m,
r:
~
Page 60 Harris 1 February 24, 1986 Points Available OUESTION 8.03 An individual with an approved Form NRC-4 receives a dose of 115 mrem during one shift.
Prior to the shift, the individual had an available dose of 100 mrem.
a.
The individual's Administrative Limit on whole body radiation is mrem /qtr.
(0.5) b.,
Has the individual exceeded the Administrative Limit?
(0.5) c.
The individual's actual dose at the end of the shift is mrem for the quarter.
(1.0)
ANSWER 8.03 f
a.
1250.
[+0.5]
b.
No (20% buffer between Administrative Limit and Available Dose).
[+0.5]
to tS' c.
FFFF(the1250 mrem /qtrAdministrativeLimityieldsan Available Dose of 1000 mrem /qtr; if the individual had an Available Dose of 100 mrem /qtr, his/her pre-shift dose was 900 mrem /qtr).
[+1.0]
(will give +0.5 for answer between1000and1200)
Reference (s) 8.03 1.
SHNPP Dosimetry Procedure DP-012, Quarterly Dose Limit Extension Authorization, Revision 1, Section 5.2, p. 4 of 10.
-Section 8.0 Continued on Next Page-
(-
.o
~
Page 61 Harris 1
(
February 24, 1986 Points Available OUESTION 8.04 Answer TRUE or FALSE. When preparing to hang Clearance tags.
on a liquid radwaste system, it is necessary to prepare a tag '
for both the manual valve and itt reach rod.
(0.5)
ANSWER 8.Qi TRUE. < [+0;5]-
4 Reference (s) 8.04 2.
1.
SHNPP AP-020, Revision 1, p. 7 of 28.
QUESTION 8.05 Answer TRUE or FALSE. The individual who actually hangs the tags on a piece of equipment should sign the " verifier" and
+
" verified by" blocks on the clearance / caution form (Form AP-20-1) if Tech. Spec. Equip. is affected.
(0.5)
ANSWER 8.05 FALSE (requires independent verification).
[+0.5]
Reference (s) 8.05 1.
SHNPP AP-020, Clearance Procedure, Revision 0, Section 5.1.3.3, p. 12a of 22.
-Section 8.0 Continued on Next Page-
(-
a s
Page 62 Harris 1 February 24, 1986 Points Available OUESTION 8.06 lihat are five (5) types of equipment that require clearance before work is performed on them?
(2.5)
ANSWER 8.06 1.
Radioactive fluid systems
[+0.5]
2.
System with > 140 F temperature [+0.5]
3.
System with > 100 psi pressure
[+0.5]
4.
Rotating machinery [+0.5]
5.
Electrical equipment,-;6:never shift forc;cn det;; it necessary [+0.5]
6.
Wkwow S/P d uw s nace>poq (y,S)
Reference (s) 8.06 1.
SHNPP AP-020, Revision 1, Clearance Procedure, p. 4 of 28.
00ESTION 8.07 Answer TRUE or FALSE. The Plant General Manager's authorization is required before the reactor can be taken critical following a trip.
(0.5)
ANSWER 8.07 TRUE.
[+0.5]
Reference (s) 8.07 1.
SHNPP OMM-001, Operations-Conduct of Operations, Revision 2, p. 7 of 47, Advance Cnange Form dated 8/14/85.
-Section 8.0 Continued on Next Page-1
Page 63 Harris 1 February 24, 1986 Points Available OUESTION 8.08 comolete each of the following with the MOST restrictive requirement.
a.
A licensed reactor operator (RO) shall be (where?)
at all times during the operation of the facility.
(0.5) b.
A licensed senior reactor operator (SRO) shall be (where?)
at all times, except when the unit is in a Cold Shutdown Condition.
(0.5) c.
During Mode 4 operation, (how many?)
SRO and (how many?)
R0 licensed individual (s) represent the minimum shift crew.
(1.0)
ANSWER 8.08 a.
at the controls [+0.5]
b.
in the control room [+0.5]
c.
2
[+0.5], 2 [+0.5]
Reference (s) 8.08 1.
SHNPP OMM-001, Operations-Conduct of Operations, Revision 2, Section 3.3, pp.17-20 of 47.
2.
SHNPP TS, Section 6.2.2, Table 6.2-1, p. 6-5.
-Section 8.0 Continued on Next Page-
r.
~
6':,, '
Page 64 Harris 1 February 24, 1986 Points Available OUESTION 8.09 Answer TRUE or FALSE.
If an on-coming shift crew member is late, the off-going shift crew member can leave if the applicable shift crew position will be unmanned for less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
(0.5).
1 ANSWER 8.09 FALSE.
[+0.5]
Reference (s) 8.09 1.
SHNPP TS, Section 6.2.2, Table 6.2.1, p. 6-5.
QUESTION 8.10 Answer TRUE or FALSE. When preventive maintenance is to be performed on a RHR pump during operation at -100% power, it is permissible to voluntarily enter a LC0 condition.
(0,5)-
ANSWER 8.10 TRUE.
[+0.5]
4 Reference (si 8.10 1.
SHNPP AP-019, Guidance for Voluntary LCOs, Revision 0, Section 1.0, p. 4 of 8.
1
-Section 8.0 Continued on Next Page-r
....e-,
.c.,--,
-r--
+,-m---,,
+m,-,
rm...
3,
_m,-,y,--..-
--y-x,w,~,,,y,.-
-r--+-
y
Page 65 Harris 1 February 24, 1986 Points Available g TION 8.11 Each of the following situations, except one (1), has an associated 1-hour Technical Specification action item.
Which is the' exception? Assume Mode 1 operation.
(1.0)
(a.) One shutdown rod is found to be partially inserted.
(b.) One of three Overpower AT indications has failed.
-(c.) One isolation valve on a RCS accumulator is found closed.
(d.) The R'JST solution temperature is 35 F.
. ANSWER 8.11 (c.)
(immediate)
[+1.0]
Reference (sl 8.11 '
1.
SHNPP TS 3.1.3.5, p. 3/4 1-20; 3.3.1, Table 3.3.1, pp. 3/4 3-3, 7; 3.5.1, p. 3/4 5-18; 3.5.4, p. 3/4 5-9.
QUESTION 8.12 At least one (1) RCS vent path shall be OPERABLE and closed from the Reactor Vessel Head in accordance with Technical Specification 3.4.11.
Why must this vent path be OPERABLE during power operation?
(1.0)
ANSWER 8.12 To exhaust noncondensible gases from the RCS [+0.5] that could inhibit natural circulation cooling [+0.5].
Reference (sl 8.12 1.
SHNPP TS BASES 3/4.4.11, p. B 3/4 4-16.
-Section 8.0 Continued on Next Page-
/
g E
- i.
- Page 66 Harris 1 February 24, 1986 Points Available OUESTION 8.13 Primary containment internal pressure shall be maintained between and (1.0)
ANSWER 8.13
-4.0 inch water gage [+0.5]
h
.l Qh I
1.9 psig
[+0.5].
g M <>o Sh 4 Reference (s) 8.13 1.
SHNPP TS 3.6.1.4, p. 3/4 6-7.
QUESTION 8.14 Wha.t are the requirements of an OPERABLE diesel generator according to Technical Specification 3.8.1.17 (2.5)
ANSWER 8.14 1.
A day tank [+0.5] containing a minimum of 2760 gallons (or 93%) of fuel [+0.5].
2.
A main fuel oil storage tank [+0.5] containing a minimum of 83,200 gallons (0:- CAi%) of fuel [+0.5].
1.44v 3.
A diesel fuel oil transfer pump [+0.5].
Reference (s) 8.14 1.
SHNPP TS 3.8.1.1, p. 3/4 3-1.
-Section 8.0 Continued on Next Page-t
,--r.
e
-v.
Page 67 Harris 1 February 24, 1986 Points Available OUESTION 8.15 During a plant cooldgwn, the temperature of one (1) RCS cold leg decreases to 250 F.
AccordingtoTechnicafSpecification3.5.3,ECCSSUBSYSTEMS a.
-T-avgLESSTHAN350F,whatmustbeaccompfishedbefore any cold leg temperature decreases below 225 F?
(1.0) b '.
Why must this action be accomplished?
(1.0)
ANSWER 8.15 a.
Two of three CSIPs [+0.5] shall be demonstrated inoperable
[+0.5]. A cc. =pt + 4-t( bd cme WAcked c) d, b.
To assure that a mass addition pressure transient can be relieved [+0.5] by the operation of a single PORV [+0.5].
Reference (s) 8.15 1.
SHNPP TS 3.5.3.a. p. 3/4 5-7, BASES 3/4.5.3, p. 3/4 5-2.
-Section 8.0 Continued on Next Page-i
/
Page 68 Harris 1 February 24, 1986 Points Available OUESTION 8.16 a.
According to Technical Specification 3.2.5 for Mode 1 operation, the following DNB-related parameters shall be maintained within what limits? Provide value and soecify whether it is an upper or lower limit.
1.
Indicated Reactor Coolant System T-avg.
%). 5 2.
Indicated Pressurizer Pressure (Jaff. 7 D b.
Idhat is the bases for the above parameter limits?
(0.5) c.
According to Technical Specification 3.2.5, the indicated pressurizer pressure limit is not applicable during certain conditions. What is the operational reason fer waiving this limit during increases in rated thermal power in excess of 5% per minute (ramp) or 10% (step)?
(0.5) pMM ANSWER 8.16 lH
%s 592.6 F).
5]; upper limit [+0.5] (analytical limit is a.
1.
gg m
2.
.5]; lower limit [+0.5] (analytical limit is 22205psig).
b.
To maintain a minimum DNBR(of 1.30)
[+0.5]
c.
During an increase in rated thermal (turbine / generator) power, RCS temperature, hence volume, decreases.
This causes pressure to decrease.
[+0.5]
Reference (s) 8.16 1.
SHNPP TS 3.2.5, p. 3/4 3-14; BASES 3/4.2.5, p. B 3/4 2-6.
-Section 8.0 Continued on Next Page-
)
s Page 69 Harris 1 February 24, 1986 Points Available OUESTION 8.17 You are the Shift Foreman on day-shift duty.
a.
Wha classifies an emergency?
(0.5) b.
Wbn initially assumes the position of site Emergency Coordinator?
(0.5) c.
Wha can subsequently assume the position of site Emergency Coordinator?
(0.5) d.
What two (2) conditions are necessary before the designated responsibilities can be transferred to the subsequent site Emergency Coordinator?
(1.0) e.
With appropriate approval, up to (how much?)
Rem whole body exposure can be received for equipment repair and up to (how much?)
Rem whole body exposure can be received for life saving rescue?
(1.0)
ANSWER 8.17 a.
Shift Foreman
[+0.5]
b.
Shift Foreman [+0.5]
c.
Plant General Manager [+0.4] or designated alternate [+0.1]
d.
TSC (or E0F) is ready to be activated [+0.5].
The subsequent site Emergency Coordinator has been briefed
[+0.5]. Acc9tt t-u u % ceSte,
,w Cd oo TSC
' 0. 5' e.
25
+
75 l+0. 5 Rulerence(s) 8.17 1.
SHNPP PLP-201, Emergency Plan, Sections 2.3, 2.4, and Table 4.6-1; pp. 2-2, 3, 4; 4-25.
-Section 8.0 Continued on Next Page-
Page 70 Harris 1 February 24, 1986 Points Available OUESTION 8.18 a.
Under wha.t conditions, during a plant site evacuation, should evacuated personnel be directed to the Site Evacuation Monitoring Area?
(1.0) b.
Where is this area located?
(1.0)
ANSWER 8.18 a.
To perform radiological monitoring of personnel.
[+1.0]
Ace wt 6 nivu.s c-
% k -~ Me b.
Near the meteorological tower.
[+1.0] '
dec-yt (
otu. (* >3 c+N d /whw u fwev M(e Reference (s) 8.18 1.
SHNPP PEP-381, Evacuation, Revision 0, Change 1, pp. 5-8 of 11.
-Section 8.0 Continued on Next Page-
C e
Page 71 Harris 1 February 24, 1986 Points Available OUESTION 8.19 Each item in the list below must be reported to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 10 CFR 50.72 (indicate TRUE or FALSE for each item).
a.-
An overpressure incident occurs in which RCS pressure reaches 2750 psig.
(0.5) b.
An automatic actuation of the Auxiliary Feedwater System occurs.
(0.5) c.
An accidental radioactive release occurs.
(0.5) d.
A worker is taken to the hospital as a result of a non-fatal injury, which occurred onsite.
(0.5) s e.
An individual receives a 50 Rem whole body exposure during a life saving rescue.
(0.5)
ANSWER 8.19 a.
TRUE b.
4 RUE-Fods e c.
TRUE d.
%K. F+l5 e e.
TRUE
[+0.5each]
Reference (s) 8.19 1.
10 CFR 50.72 (a) (2), (7), (8), (9), (11).
I &l ad.
t <{ P T e l,
2.
t
-Section 8.0 Continued on Next Page-i i
1,..,
Page 72 Harris 1 February 24, 1986 Points Available OUESTION 8.20 With regard to locked areas on the plant, how does routine entry on rounds differ from emergency access?
(1.0)
ANSWER 8.20 A routine entry on rounds is by duty key ring.
Emergency access is obtained with keys from Shift Foreman Key Locker.
[+1.0]
Ace ey'f for hm%w dkh '. cbch wth WM Cn o+ bca,T3 Reference (s) 8.20 1.
'SHNPP PG0-09, Key Control for Duty Personnel, p. 1.
-End of Section 8.0-
-End of Exam-l I
u
~
ENCLOSURE 3 NRC REACTOR OPERATOR EXAM REVIEW For Shearon Harris Nuclear Power Plant Exam Date February 24, 1985
. Exam Reviewers:
A. W. Powell P. Rubin J. B. Hudson D. Shaffer G. M. Blinde R. Garner W. A. Bufe QUESTION #
COMMENT BASIS SECTION 1
[
1.05 Accept any one of the Redistribution and Pressure following:
Coefficient are considered as
" Redistribution", " Voids" or components on page 6 of the referenced
" Pressure" for answer number 3 dgcument RT-LP-1.10 (See Attachment) 1.15 Accept the following as Core Design Report presented to additional possible response:
candidates during Prelicense Review
" Checkerboard loading pattern class considers " Checkerboard" concept for lower enrichment fuel separate from loading high enrichment towards center of core" fuel towards the outside of the core.
' M 9J.
Both contribute towards a flatter radial flux (See Attachment).
1.21 Also accept " Minimize the T.S. Basis B3/4.1.3 Item f(3) refers effects of rod misalignment" to rod ejection as rod misalignment.
as a substitute for answer #1 The theory text from which the answer (rod' ejection) was drawn and the T.S. basis provided are saying essentially the same thing.
(See Attachment) 1.22 Also accept " Samarium" as OST 1036 considers Samarium in the SDM -
response.
calculation.
(See Attachment).
i SECTION 2 2.2 Accept any one of the Loss of off-site power is a generic l
following for item #2 of the answer derived from the Westinghouse answer key.
logics for AFW start signals.
" Loss of off-site power" or Recently SHNPP control wiring
" Loss of power to associated diagrams were modified such that a safety bus" (IASA or 1BSB) or loss of the respective safety bus
" Blackout" (lASA for ' A' MDAFP and 1BSB for
'B' MDAFP) would cause MDAFP start. This information was presented to license candidates due to its importance. The term blackout should also be accepted due to its presence on the MDAFP startup functional diagram distributed to license candidates.
(See Attached) 1 of 9 l
R.
NRC REACTOR OPERTAOR EXAM REVIEW 2.6 Accept "480V Bus t'o 120VAC The 120V ESF AC buses have two backup power panel _(120/208VAC sources of power.
One of these is in
~
system) to 120V ESF AC the answer key and the other is as System" as an alternate noted above (from a 120VAC power panel
~
answer for " Backup" power in the 120/208VAC system).
(See supply.
attached diagram 120V UPS-TP-1.0)
Change enswer to "Preheater The tempering feed flow system 2.8b.
Tempering Permissive which referred to in the answer key has allows small amount of feed been completely removed.
Some feed flow around the main feed flow goes through the preheater bypass isolation valve to warm up valve during low' power operations the maid feed line and S/G while the MFIV is still shut.
Preheater bypass flow (not tempering)
- preheater"
'y goes into the S/G via the auxiliary teed line but has no tempering function. Preheater tempering refers to the small amount of feed flow around the MFIV used to warm up the main feed line and S/G preheater before the MFIV opens up admitting
,~-
full feed flow in the preheater.
(See Attachment) 2.10
- 1. Accept any four out f L.'Since question did not specify either of the following, which mode of startup waa involved
, g.
(emergency or normal) but did two lists:
specify listing 4 conditions / four If student assumes emergency start:
M.
of either of the lists pro'vided a) Generator up to speed.and should be accepted for full credit.
voltage
- 2. Question did not specify that b) Operational / Maintenance setpoints required and therefore selector switch in credit should not be taken off for c) 2/3 UV relays tripped on
^ not including setpoints.
(See
" Operational" position attached pages 16 & 17 of respective bus.
SACP-H0-1.0).
d) Generator lockout relay reset e) Breaker in " Operate" (racked in) l If student assumes normal start:
a) Master control selector switch in "MCB" position b) Synchronizer switch on "MCB" position c) Synchronizing relay energized d) Generator up to speed and i
i voltage e) Lockout relay reset f) Breaker in " operate"
[
(racked in) g) Operator goes to "close" on breaker control switch
- 2. Do not delete points for not including setpoints 2 of 9 u
NRC REACTOR OPERTAOR EXAM REVIEW 2.15e Delete Should be deleted because the valve operator is not exposed to CTMT atmosphere, but the valve itself is.
The valve internals are located inside the pipe which is open to containment atmosphere.
Question could have been answered either True or False depending how question was interpreted by the candidate. Wording of this question was vague and unclear as to what part of the valve was being discussed.
2.17 Full credit should also be Question as written could be read as awarded if only the second one situation instead of the two half of the answer (Pressure separate situations intended (e.g.
Control by PCV145)is given cooldown on RHR with PZR solid instead of cooldown on RHR as one scenario and PZR solid as the second scenario)
The word "AND" can be read to imply both specified conditions are in effect.
Three of five reviewers interpreted this as one condition.
2.19 Delete The thermal sleeves were removed at Harris Plant. The candidates were g
40.
informed of the removal and considered
.*P thermal sleeves a thing of the past and not required information.
Answering the question as "Where they used to be" is not relevent to a candidates knowledge.
(See attached answer key) f SECTION 3 3.04 Change answers to 1850 psig, As per April 1985, Technical 601 psig, 3.0 psig Specifications.
T.S. are used as our l
source document for setpoints.
(See Attachment) l 3.08a Also accept:
" Auto Closes The signal generated from this Main Feedwater Regulating condition is sometimes referred to as Valves" FW Isolation but in fact does not give a full FW Isolation signal.
In this condition it only closes the Main FW Regulating (Control) Valves.
I 3.09 Change answer to 38.3%.
As per April 1985 T.S..
T.S. are used as our source document for setpoints.
(See Attachment) l 3 of 9 I
NRC REACTOR OPERTAOR EXAM REVIEW 3.11 Delete Question is not site specific.
SHNPP does not have a runback due to a loss of 50% of load. When clarified during exam, it was still confusing. Many candidates did not understand intent of question.
3.16
- 1. Change answer #7 to 109%
- 1. Answer as per April 1985 T.S.
instead of 108%.
should be 109% (See attachment)
- 2. Setpoints not required
- 2. The question did not specify that setpoints were required in answer therefore credit should not be taken off for not including them.
SECTION 4 4.04 Accept < 1360 psig (1600 psig]
Referenced document. E0P-CSFST, foldout A, has < 1360 [1600 psig] as setpoint (See attachment).
4.18 Accept either 190*F or 195'F Since stem of question did not for answer #2 reference a specific procedure the candidate could have answered with either 190*F or 195'F. AOP-14, i
" Loss of CCW" uses 190*F and AOP-18,
, j, "RCP Abnormal Operations", uses,195*F for loss of CCW to the RCP. A 4
procedure change has been submitted to correct this conflict.
(See Attachment) 4 of 9
NRC SENIOR REACTOR OPERATOR EXAM REVIEW For Shearon Harris Nuclear Power Plant Exam Date February 24, 1986 Exam Reviewers:
'A. W. Powell P. Rubin J. B. Hudson D. Shaffer G. M. Blinde R. Garner W. A. Buie QUESTION #
COMMENT BASIS SECTION 5 5.03 Accept either "a" or "b".
Answer "b" is a true statement. The converse of the statement in FF-HO-1.0 is also true (decreasing flow rate causes minimum required NPSH to decrease as well).
(See Attachment) 5.08 Accept " maintain core cooling" Training material from Transient and as an additional answer.
Accident Analysis, T&AA-HO-1.16, show that maintenance of core flow at high pressure during SI with a Small. Break LOCA is an advantage in that it serves to cool the core and help prevent core uncovery.
(See attachment)
+
5.10 Accept either answer "a" or RT-LP-1.9 is in error when discussing "b" as correct.
the effect of "BURNUP" (i.e. time in core life effects).
It states that doppler fuel temperature coefficient gets less negative over core life due to U238 depletion (predominant over Pu240 production). All operators were taught using the text of a letter from CP&L fuels personnel on this subject.
The text of that letter states that Pu240 produc.f on is predominant. This e
has been incorporated into a later version of SHNPP theory training material.
(See Attachment) 5 of 9
NRC SENIOR REACTOR EXAM REVIEW 5.13 Accept either " COLD" or "NO Since questfon did not give specific LOAD" for Tavg condition plant conditions, the candidates could (answer b) and " DILUTION" or have interpreted the SDM requirements
" STEAM BREAK" for accident as for Modes 1-4 (1770 pcm) or Mode 5 condition (answer c)
(2000 pcm). Most recent revision of T.S. given to license candidates has a higher required SDM for Mode 5 (due to dilution accident) than for Modes 1-4 where "NO-LOAD" is the plant condition and " STEAM BREAK" is the accident.
Note that the wording in T.S. basis was made before the required Mode 5 SDM went from 1000 PCM to 2000 PCM.
(See Attachment) 5.15 Accept the following as Core Design Report presented to additional possible response:
candidates during Prelicense Review
" Checkerboard loading pattern class considers " Checkerboard" concept for lower enrichment fuel separate from loading high enrichment towards center of core" fuel towards the outside of the core Both contribute towards a flatter radial flux (See Attachment).
5.16 Accept any one of the Redistribution and Pressure following:
Coefficient are considered as
" Redistribution", "Vofds" or components on page 6 of the referenced
" Pressure" for answer number 3 document RT-LP-1.10 (See Attachment) 5.20 Also accept " Minimize the T.S. Basis B3/4.1.3 Item f(3) refers effects of rod misalignment" to rod ejection as rod misalignment.
as a substitute for answer il The theory text from which the answer (rod ejection) was drawn and the T.S. basis provided are saying essentfally the same thing.
(See Attachment)
SECTION 6 6.02a Accept either "N44 Nuclear N44 Nuclear Power input is the Power" or " Anticipated Steam Anticipated Steam Demand input.
Demand" Since they are the same, only one f
should be required for full credit.
l.
6.02b Accept as answer; " steam Question asks for " primary or pressure", " feed flow" and secondary information".
Density
" steam flow" compensation and flow error do not fall into this category.
[
Question is not site specific.
SHNPP 6.05 Delete N
does not have a runback due to a loss of 50% of load. When clarified during exam it was still confusing. Many candidates did not understand intent of question.
6 of 9
NRC SENIOR REACTOR EXAM REVIEW 6.06 Change Lo-Lo S/G Level Trip As per April 1985 Tech Specs.
T.S.
to 38.3%
are used as source document for setpoints.
(See Attachment) 6.09 Accept any 3 of the following This information was presented to the 5 answers:
candidates during recent Prelicense
- 1. SIAS Review Training as a recent plant
- 2. Hf radioactivity at air change.
(See Attachment) intake
- 3. Chlorine at an intake or in storage area
- 4. Smoke' Detectors 5.
Rad Monitor Power Failure 6.13 Delete A hydrogen Ifmit is not stated in referenced SD-125.
4-6% is considered an explosive mixture and therefore we would expect the candidates to answer the question as 1-3% since it is most correct.
6.20 Accept the following as a The additional answer for part (a) is third answer for part (a):
found on page 24 of the referenced
" helps maintain uniform water document SD 100 (See Attachment)
~ ~ ~
temperature in the pressurizer"
~ ~~~
.y.
SECTION SEVEN
GENERAL COMMENT
Question numbers 7.03, 7.05, 7.07 and 7.24 require memorization of steps out of the body of various Emergency Plant Procedures and Functional Restoration Procedures beyond what would be expected of the candidates to know. NUREG 1021 guidelines state that the candidates should have a conceptual knowledge of the EPP's and FRP's.
This does not imply that step memorization is required.
In addition, procedures were not referenced in the stem of the questions which led to varfous interpretations.
7.03 Delete Question requires memorization of sequence of steps (see General Comment).
7.05 Delete Question did not give sufficient information for the candidate to properly answer.
This statement is true as per a caution in EPP-9 while reffiling the PZR (Not mentfoned in question), post LOCA, but is false in step 19 of same referenced EPP-9 while depressurizing the RCS.
In addition, question requires memorization of an EPP (see General Comment).
7 of 9
+ - -. - _ - -,.
_.___--.m
__=7
.y.... - -. -.
NRC SENIOR REACTOR EXAM REVIEW
-7.07 Delete Question requires step memorf ration of FRP (see General Comment) 7.10
- 1. Accept single response for
- 1. Question does not ask for a full credit.
specific number of responses. One
- 2. Accept additional responses response would answer question and as follows:
should be allowed for full credit.
a) Manually control blended
- 2. The type of failure is not stated flow f rom control board and there could be other valid via flow control valve answers as shown.
in manual b) BTRS can be used to borate or dilute c) Borate via RWST suctions to Charging /SI pumps 7.12 Accept either 190*F or 195*F Since stem of question did not for answer il reference a specific procedure the candidate could have answered with either 190*F:or 195*F.
AOP-14, " Loss of CCW" uses 190*F and AOP-18 "RCP Abnormal Operations", usea 195'F.
A procedure change has been submitted to correct this confifet.
(See Attachment)
The wording of this question does not 7.15 Accept any one of the
~
'?
lead the candidate to answer,fa's following three answers
- 1. #1 and #2 seal failure specific manner.
Therefore, simply
- 2. #1 seal failure stating a RCP B seal failure should be
- 3. RCP B seal failure sufficient. Also, in addition to the scenario in the answer key, a #1 seal failure by itself would give you identical symptoms.
(See Attachment) t 7.16 Accept additional responses Stem of question does not specify a for "other areas to look for procedure, therefore other answers as further evidence."
1.e. -
shown are feasible.
sump levels, visual, CTMT Leakage Detection System 7.19c Change from " false" to "true" Referenced SD-115 states that the same
" system" is used for Decontamination of both the casks and tools.
- However, I
the decontamination is done in totally
" separate areas."
7.22 Delete Repeat question.
Same question as 7.11.
8 of 9 i
i
N NRC SENIOR REACTOR EXAM REVIEW 7.24 Delete Question wording was confusing to the candidates. The candidate could not adequately determine what was being asked. The question did not ask what fndications and values are used to determine ff upper head is full.
In addition, question requires
. ~
memorization of an FRP (see General
~
Comment).
SECTION 8 8.03c Change to'1015 Math error. 900 MR pre-shift dose +
~
8.06 Accept any 5 of the following
' Referenced document AP-020, Rev. 1, six respon'ses:
has this for answer (See Attachment)
- 1. Radioactive Fluid Systems
- 2. System with > 140*F temperature
- 3. System with > 100 psi pressure
- 4. Rotating Machinery
- 5. Electrical Equipment
- 6. Whenever SF deems necessary 8.13 Accept -4"instead of 4" fhh This is as per April 1985 Tech' Specs answer which;are used as source document for setpoints (See Attachment). Tech Specs provided as reference material contained typo.
8.16 Accept "Later" as answer As per April 1985 Tech Specs which are are used as source d.ocument for setpoints. This was noted during-1 examination and candidates were told to put in "Later" if that was what was in T.S. (See Attachment) 8.19 Accept b and d as false As per referenced document 10C FR50. 72
- b. False - AFW is part of the ESF system which is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification
- d. False - This is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification and then only if person is contaminated (See Attachment) 8.20 Delete Question wording is confusing and unclear as to candidate being able to determine response being requested.
During examination this question was clarified but not adequately enough for correct interpretation.
9 of 9
-.