ML20136D596
| ML20136D596 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/31/1985 |
| From: | Gore B, Huenefeld J, Mcmillen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20136D593 | List: |
| References | |
| 50-346-OL-85-02, 50-346-OL-85-2, NUDOCS 8601060216 | |
| Download: ML20136D596 (110) | |
Text
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.U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-346/0L-85-02 Docket No. 50-346 License No. NFP-3 Licensee:
Toledo Edison Edison Plaza 1
300 Madison Avenue Toledo, OH 43652 Facility Name:
Davis Besse t
Examination Administered At:
Davis Besse Examination Conducted:
November 19, 20, 21, 1985
/J/J//86 Examiners:
ne d
Datd
/
h h
l//[D s
li. F. Gord I
Dat'e /
h Approved By:
)I.McMillen, Chief
/A/// Ib
'perating Licensing Section Dcyte /
Examination Summary Examination administered on November 19, 20, 21, 1985 (Report No. 50-346/0L-85-02) to one reactor operator and eight senior reactor operator candidates.
Results:
Four senior reactor operators candidates passed the examination.
10 2
6 0
-.- _ -- - - - i
REPORT DETAILS 1.
Examiners J. C. Huenefeld, PNL B. F. Gore, PNL 2.
Examination Review Meeting The examination review meeting was conducted in conjunction with the exit meeting.
3.
Exit Meeting At the conclusion of the examinations, an exit meeting was held with the licensee.
Persons in attendance were:
Examiners J. C. Huenefeld B. F. Gore Toledo Edison Personnel T. Bergner, Operations Training Program Supervisor B. O'Connor, Assistant Plant Manager, OPS R. Simpkins, Operations Training Manager M. Steward, Training Director The facility was informed of three candidates who were not clear passes.
In addition, the facility comments concerning the written examinations were discussed.
Those comments (Attachment 1) and their resolutions (Attachment 2) are attached to this report.
Note:
Prior to exam administration, question 1.2 on the reactor operator examination was deleted.
This action was taken by the chief examiner because it was realized that that question was disputed on the last examination given
- at this facility, and there were ample reactor theory questions in the category.
Point values were adjusted accordingly.
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ATTACHMENT 1 EXAMINATION REVIEW COMMENTS SR0 EXAMINATION CATEGORY 5 Q 5.09 The wording of this question does not make it clear if this is secondary or primary water addition.
Reference (AB 1203.03) describes the primary water that must be added for the contraction volume of the RCS. The answer key indicated (and a few of the candidates were told) that the question was asking for the number of gallons of feedwater that would have to be added to cool the RCS to 546 F.
An acceptable answer (b) would be given by performing a rough calculation based on known values of the Primary System:
(36 2 T) x 5"
x 24 gallons
= 4320 gallons 1
1" F 1" level Suggest full credit for performing prirnary side calculation.
For candidates performing secondary side water addition calculations, suggest question be deleted if this question makes a difference in pass / fail decision.
(Procedural references attached)
Q 5.10 Agree with the answer as a final condition, however, in order for a bubble to form in the candy can, the pressurizer must empty.
Suggest accepting "yes" (instead of "no") as long as the final result is the same, i.e., pressurizer level and loop level equal.
2 CATEGORY 6 Q 6.01 Agree with answer, however, question wording could elicit a pressure setpoint response.
Suggest accepting pressure setpoint response for full credit.
(Procedural reference attached)
Q 6.03.b SFRCS Steam Generator Level instrumentation, located at the SFRCS cabinets, are essentially powered.
Suggest accepting SFRCS Level Instruments for full credit.
(Procedural reference attached)
Q 6.05 Agree with answer, however, question wording could elicit a correct-response different that that identified on the answer key.
Suggest accepting the following response for full credit:
(1) Pressurizer heaters in auto and (2) Low pressurizer level of 40" reached (Procedural reference attached)
Q 6.08 Procedure lists eight (8) (instead of 7) RCP starting interlocks.
Seal return valve interlock was omitted from answer key.
Also, CCW interlocks are divided into four (4).
Suggest accepting Seal return valve open interlock as a correct response.
Suggest partial credit for individual CCW interlock response.
(Procedural reference attached)
Q 6.11 The answer to this question may be different due to weather conditions.
Suggest accepting cooling tower makeup as regular and service water returns as supplemerdal during cold weather operations.
(Procedural reference attached)
Q 6.15 Answer and reference is in error the second interlock discussed in answer is administratively controlled.
Suggest accepting response listed in procedural references for full credit.
(Procedural references attached)
Q 6.18 Suggest 8' as acceptable answer for SFAS Level 5 setpoint.
Also CTMT Pressure Setpoints should be PSIA instead of PSIG.
(Procedural references attached)
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3 CATEGORY 7 Q 7.04 The answer is in error.
Proper response is if seal return temperature increases to 200 F, indicating a failed seal is leaking excessively.
Suggest accepting above response for full credit.
(Procedural references attached)
Q 7.08 Wording of question may elicit response to problems with the trip throttle valve linkage and overspeed trip tappet sticking.
Suggest accepting above answer for full credit.
(Reference NUREG 1154) l l
Q 7.16 Unfortunately, there are five (5) major Supplementary Actions remaining in EP 1202.01 following the SFAS guidelines.
It is l
perceived a candidate may be confused and elicit the remaining Supplementary Actions including SFRCS Response, Subcooling, I
Overheating, Overcooling and 0TSG Tube Rupture Checks. While l
there is no concern of the key's accuracy, the question does not indicate completely to the candidate that the question is looking specifically for SFAS follow-up items.
Suggest allowance be provided if this proves to be the case.
1
4 CATEGORY 8 Q 8.10 Question does not elicit an explanation for each part (a, b, c).
Suggest c cepting answer "c" for full credit.
Q 8.13 Question wording may elicit other correct answers.
Suggest accepting Assistant Shift Supervisor (qualified individual) and Emergency Duty Officer on call for full credit.
(Procedural reference attached)
Q 8.15 Agree with answer, however, normal titles are different than Emergency Plan titles.
Suggest accepting any combination of the following for full credit:
EMERGENCY PLAN TITLE NORMAL TITLE INDIVIDUAL 1.
Station Operations Manager Plant Manager Lou Storz 2.
Radcon Operations Manager Chem & Health Physicist D. Briden 3.
Emergency Duty Officer As assigned As assigned 4.
Emergency Planning Supervisor Same J. Scott-Wasilk 5.
Emergency Operations Manager Nuc. Safety & Lic. Director T. Myers 6.
JPIC Manager Media Relations Manager R. Kelly 7.
Nuclear Security Manager Same C. DeTray 8.
Technical Engineer Same J. Lingefelter 9.
OSC Manager Assistant Plant Mgr., Maint.
S. Smith
- 10. Operation Engineer
- Asst. Plant Mgr., Operations W. 0'Connor
- 11. TSC Manager Nuclear Plant Systems Dir.
J. Wood
- Due to recent title changes, may also see Operations Superintendant.
~
5 EXAMINATION REVIEW COMMENTS R0 EXAMINATION The items identified in SR0 concerns are also applicable to the R0 exam.
These include item numbers as follows.
1.
R0 Question 1.15 SR0 Question 5.09
=
2.
R0 Question 2.16 SR0 Question 6.11
=
3.
R0 Question 3.8 SR0 Question 6.03
=
4.
R0 Question 3.11 SR0 Question 6.08
=
5.
RO Question 3.14 SR0 Question 6.15
=
6.
R0 Question 3.17 SR0 Question 6.18
=
7.
R0 Question 4.07 SR0 Question 7.04
=
8.
R0 Question 4.17 SRO Question 8.13
=
9.
R0 Question 4.20 SR0 Question 7.07
=
Q 2.12 Please find below other procedural guidance concerning Piggyback operations and use applicability.
Suggest acceptance of these as alternatives.
AB 1203.18 Loss of Makeup EP 1202.01 Specific Rule 2.4 AB 1203.26 Serious Control Room Fire And Additional Portions of EP 1202.01 (Please see your reference copy) r
ATTACHMENT 2 Examiner Response to Facility Comments:
Senior Reactor Operator Examination:
CATEGORY 5 Q 5.09 The facility comment is a valid one, and out of fairness to all candidates this question was deleted from the examination.
The point value for this question was 2.0.
Upon deletion, the total point value for Sections 1.0 and 5.0 became twenty-three (23.0).
Q 5.10 No candidate said both "Yes" and that pressurizer and loop level would be equal, so the facility comment did not come up as a point of contention.
However, it should be noted that pressurizer would not necessarily have to empty for vapor to form in the " candy-cane".
As soon as the water in the top of the candy-cane reached saturation pressure, levels would start to equalize whether the pressurizer was empty or not.
The question was graded as per the answer key.
d 2
CATEGORY 6 Q 6.01 Agree with the facility comment.
Q 6.03.b Agree with the facility comment.
Q 6.05 Agree with the facility comment.
Q 6.08 Agree with the facility comment; however, only partial credit was given for listing more than one of the CCW interlocks.
The answer key was amended to include the Seal Return valve being open as an-interlock.
Q 6.11 Candidates were given full credit regardless of which source that they labeled as supplemental and which was labeled as regular.
Q 6.15 Only one of the interlocks was required for full credit, provided the candidates answer was consistent with the facility reference (i.e., no interlock controlling how many pumps are loaded into one diesel bus).
Q 6.18 Agree with the facility comment.
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3 CATEGORY 7 Q 7.04 Agree with the facility comment; however, the candidates were s expected to discuss the affect of loss of the auxiliary impeller recirculation as it affects the ability to maintain seal return less than 200 Deg.
Q 7.07 Agree with the facility comment.
Q 7.16 Partial credit was given for this response.
It should be pointed out that in the question, one of the actions was stated to help clue the candidate into the expected response.
Given this fact, no more than 1/2 credit was given for the facility's anticipated response.
+
4 CATEGORY 8 Q 8.10 Agree with the facility comment.
Q 8.13 Agree with the facility comment.
Q 8.15 Agree with the facility comment.
Reactor Operator Examination:
All of those comments on the SR0 exam that apply toward the R0 examination were transferred to the R0 key and taken into account.
The only additional comment was on Q 2.12, and we agree with the facility comment.
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U. S. NUCLEAR REGULATORY COMilSS10N REACTOR OPERATOR LICENSE EXAMINATION Facility:
Davis Besse 1 Reactor Type:
PWR Date Administered: November 19, 1985 Examiner: U.J. Apley /J.C. Huenefeld Candidate:
Answer Key INSIRUCTIONS TO CANDIDATE:,
f 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 801. Examination papers will be picked up six (6) hours af ter the examination starts.
Category 1 of Candidate's 1 of Value Total Score Cat. Value Category 25
_.. 2 5
- 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 25 25
- 2. Plant Design including Safety and Emergency Systems 25 J
- 3. Instruments and Controls 25 25
- 4. Procedures: Normal, Abnormal.
Emergency, and Radiological Control 100 TOTALS Final Grade 1
All work done on this examination is my own; I have neither given nor received aid.
Candidate's 51gnature
Page 1 Davis Besse November 19, 1985 Points Available 1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS.
(25.0)
HEAT TRANSFER AND FLUID FLOW OUESTION 1.01 During an RCS natural circulation cooldown, why is it important to control the Pressurizer level at a steady or slightly in-creasing trend?
(2.0)
ANSWER 1.01 To prevent any outsurge of hot water from the Pressurizer into the hot leg where it could flash to steam and cause a loss of NC.
(+2.0)
Reference (sl 1.01 PP 1102.'10.9, p. 37.
-Section 1.0 Continued on Next Page-
e2 Davis Besse November 19, 1985 Points Available OUESTION 1.02 I
lion does the bo n concentration affect the accuracy of critical-ity estimates pr ided by a 1/M plot? Exolain why.
(2.0)
ANSWER 1.02 The greater the boron con ntration, the poorer the 1/M plot criticality estimate. The istant placement of the detectors requires a relatively large st neutron flux to provide suffi-cient thermal neutrons to the tectors to show K-eff for the core.. This is because the react reflector contains borated water which prevents thermal neutr s from exiting the reactor to reach the detectors. When critic ity is imminent, the fast neutron flux has increased enough o provide a true picture of conditions in the core to the detecto The greater the boron concentration, the less themal neut s reach the detec-tors.
(+1.25 answer /+0.75 reason)
Reference (s) 1.02 Licensing Info. Manual, Vol. I, pp. 1-51 and I-53.
Oe,leted fe;07 8#
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- a. psvious tsn. si,,a w y
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b'wsti u to Ad[y$<
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-Section 1.0 Continued on Next Page-
l Page 3 Davis Besse November 19, 1985 Points Available OUESTION 1.01 If the system temperature difference that is driving natural circulation flow is exactly doubled, exolain why the subsequent he.st removal rate actually increases by greater than a factor of two (2).
(1.0)
ANSWER 1.03 Q = mCpAT.
If a + Cp stayed constant, doubling AT would double Q.
But the flow rate m is caused by the AT, and therefore, the flow rate will also increase (though not linearly).
(+1.0)
Reference (s) 1.03 Equation Sheet - Basic Thermodynamic Theory 00ESTION 1.04 If the pressure in a leaking pipe is reduced by 50%, why does the leak rate only decrease by about 25%.
(1.0)
MSWER 1.04 The leak rate is the flow through the hole - m = pAV, where p is the fluid density, A is the area of the leak, and V is the velocity of the fluid leaving. Therefore, m is directly propor-tienal to V.
Butbytheconservatjonofenergy-pressureis proportional to kinetic energy, mV /2. Pressure is proportional to velocity squared. So if pressure is halved, the effect on velocity is the square root of that change - 1/4, and since velocity is directly proportional to leak rate, the leak rate only decreases by 1/4.
(+1.0)
Reference (s) 1.04 Equation Sheet - Basic Fluid Theory, General Energy Equation, Continuity Equation, Bernoulli Equation.
-Section 1.0 Continued on Next Page-
Page 4 Davis Besse November 19, 1985 Points Available OUESTION 1.05 Exolain what indication you should expect to see on the source range count rate monitors as the core is voided by a LOCA.
(1.5)
't ANSWER 1.05 Count rate would significantly increase (x 100 to 1000, numbers not important), and would be erratic as voiding oscillations took place.
(+1.5)
Reference (s) 1.05 B&W TMI-II Accident Lesson Plan, 1981.
QUESTION 1.06 Why does Xenon peak at a later time following a shutdown from high power than it does when following a shutdown from a lower
' SI power level?
MN ANSWER 1.06 The iodine to Xenon ratio is higher at high power levels.
When I-135 decays, it decays to Xe-135, thus reducing the Ng/
Nx ratio. This reduction in ratio continues, but once a ratto of,0.78 is reached, the peak has been reached.
(+1.0)
Reference (s) 1.06 Licensing Info. Manual, Vol. I, pp. I-65.
-Section 1.0 Continued on Next Page-
Page 5 Davis Besse November 19, 1985 Points Available 00ESTION 1.07 idhy aren't the sum of individual control rod assembly (CRA) s.(
rod worths always equal to the group rod worth?
M ANSWER 1.07 Rod worth is proportional to the square of (local thermal flux divided by average thermal flux in the core). Thus, the position of other rods within the group can effect the worth of one rod, so that individual rod worths will vary depending on se-quencing, position, etc.
(+1.0)
Reference (s) 1.07 Licensing Info. Manual, Vol. I, p. I-71 through 73.
QUESTION 1.05 The reactor has tripged and RCS temperature is being maintained at approximately 550 F.
There is a precaution in Davis-Besse procedures which says that if the cooldown is HQI started within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, Group I rods should be pulled before commencing the cooldown. idhy are Group I rods needed after 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />?
(1.5)
ANSWER 1.08 The 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is to assure that Xenon worth is sufficient to provide the necessary shutdown margin. After 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> Xenon decay is inserting positive reactivity.
(+1.5) l i
Reference (s) 1.08 PP 1102.03, p. 2 (trip recovery).
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Page 6 Davis Besse November 19, 1985 Points Available OUESTION 1.09 State how each of the following change from 00L (Beginning-of-Life) to EOL (End-of-Life). (Note: exact values not required.)
- o. s-Differential Boron Worth (% AK K) gg,99 a.
p
- o. (
b.
Inserted Rod Worth (%AK/K)
(Gr&)
c.C c.
Temperature Coefficient of Reactivity (% AK/K/ F)
Gert) c.6" d.
RCS Boron Concentration (ppmB)
(&;t)
ANSWER 1.09 a.
becomes a larger negative value (+0.3) b.
becomes a larger negative value (+0.3) c.
becomes a larger negative value (+0.3) d.
becomes a lower value
(+0.1)
Reference (s) 1.09 Aeactor Operators Curve Book (PP 1101.02) a.
Figure 5A and 5B b.
Figure 9A, etc.
c.
Figure 12 d.
Figure 6B i,
1
-Section 1.0 Continued on Next Page-I e
Page 7 Davis Besse November 19, 1985 Points Available OUESTION 1.10 The figure below shows neutron population response to reactivity insertion. Label each of the three (3) curves as supercritical, exactly critical,)or subcritical.
(Assumebelowthepointof feedback effects.
(1.0)
A B
C
:---~~'-~
g,
,l b
Neutron P pulation NaN N%
N- ~~I 3
N. -
i i
I i
11 i
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At, gg At time in seconds 3
ANSWER 1.10 a.
supercritical b.
exactly critical c.
subcritical Reference M 1.10 LIM, Vol. I,Section I, p. I-16.
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-Section 1.0 Continued on Next Page-I
Page 8 Davis Besse November 19, 1985 Points Available OUESTION 1.11 Given a large vented tank 30 ft. in diameter and 60 ft. high with a centrifugal pump taking a suction from its base. The pump is located at a vertical elevation corresponding to the bottom of the tagk. The tank is entirely full of water and is maintained at 60 F,by heaters. Assume the vent becomes totally clogged while the pump is in operation. Atmospheric pressure is 14.7 psia. Answer the following questions:
a.
What is the maximum differential pressure that could occur between the inside and outside of the tank? Exolain.
(1.5) b.
Exclain why the pump may begin to cavitate at a higher tank level than with the vent open.
(1.0)
ANSWER 1.11 Thelowestpressurethatthetgnkcoulddroptowouldbe a.
the saturation pressure for 60 F which is 0.256 psia. So maximum AP = 14.7 - 0.256 = 14.444 psia (full credit).
(1.4 pts if simply state 14.7 psia.)
b.
Without atmospheric pressure on top of the water, the equivalent head is lost putting the pump closer to its minimum NPSH limit.
Reference (s) 1.11 Centrifugal Pumos and System Hydraulics, Igor J. Karassik.
-Section 1.0 Continued on Next Page-
Page 9 Davis Besse November 19, 1985 Points Available OUESTION 1.12 The reactor is at 100% FP equilibrium xenon conditions when a reactor trip occurs. Three (3) hours later the reactor is restarted.
-8 a.
When at 10 amps, which direction will rods have to go to hold power ' constant? Exolain.
(1.0) b..
Sketch the xenon concentration through to equilibrium assuming the reactor is raised to 92% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the trip).
(1.5)
ANSWER 1.12 at10gilhavetobewithdrawnbecausetheburnoutterm Rods w a.
amps is not significant.
b.
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$9 Reference (s) 1.12 LIM, Vol. I, Section 1.
-Section 1.0 Continued on Next Page-
Page 10 Davis Besse November 19, 1985 Points Available OUESTION 1.13 a.
Define shutdown margin. Include any assumptions about axial power shaping rod or control rod positioning.
(2.0) b.
During a plant heatup in accordance with the Plant Startup Procedure, PP 1102.02, the reactor must be shutdown by 21% Ak/k. If Group 1 rods are withdrawn, does their worth count as part of that 1% Ak/k?
(1.0)
ANSWER 1.13 a.
SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of re-activity by which the reactor is subcritical or would be subcritical from its present condition assuming:
No change in axial power shaping rod position, and All control rod assemblies (safety and regulating) are fully inserted except for the single rod assembly of highest reactivity worth which is assumed to be fully withdrawn.
b.
No.
Even if Group 1 rods are to be withdrawn, the shutdown value must be 2 1% Ak/k.
Reference (s) 1.13 a.
Technical Specifications, p.1-3.
b.
Plant Startup, PP 1102.02.20, p. 20.
-Section 1.0 Continued on Next Page-
Page 11 Davis Besse November 19, 1985 Points Available OUESTION 1.14 While drawing a bubble in the pressurizer, the vapor space is vented to the quench tank. Describe how conditions in the quench tank can be used for determining when a steam bubble exists in the pressurizer.
(2.0)
ANSWER 1.14 Early in the venting process most of the gas vented to the quench tank will be nitrogen. Manual operation (from Control Room) of RC 222 Quench Tank Vent to Containment vent header will be required to keep quench tank pressure below 50 psig.
As the venting process continues, level and temperature control of the quench tank will be required to handle the condensed steam.
Reference (si'1.14 Pressurizer Operation, SP 1103.05, p. 13.
-Section 1.0 Continued on Next Page-
i Page 12 Davis Besse November 19, 1985 Points Available QUESTIO Assume at the condenser is not available and the operator is manually ontrolling steam generator pressure at 995 psig using the atmosp ric vent valves. Estimatehowmanygglionsof water would required to cooldown the RCS from 582 F to 546 F.
Show assumptio
, (Select one.)
(2.0) a.
350 gallons b.
3500 gallons c.
35,000 gallons d.
73,000 gallons ANSWER 1.15 b.
(Actually3,446 gallons: answ does not require memoriza-tion.
Ifhekgows1galofst released will cool 1000 gal of water 1 F, then (Vol RCSI(582 - 5461 (8 000 (361 1000
= answer ::
- 3200 gal.)
Reference (s) 1,15 AB 1203.03, Cooldown W/0 BWST and No Off-Site Pow
-End of Section 1-dim
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Page 13 Davis Besse November 19, 1985 Points Available 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25.0)
OUESTION 2.01 Match the sprinkler system name (letter) with its respective definition (number. 1 (2.0)
Dry Pipe Preac' tion prinkler System a.
b.
Dry Pipe Fusible Link System c.
Wet Pipe Link and Level Sprinkler System d.
Dry Pipe Automatic Deluge Valve Sprinkler System 1.
When the rate-of-rise heat detectors activate the system, all the sprinkler heads and spray nozzles receive and spray water at the same time. They do not have a fusible link on the sprinkler which melts out at a given tempera-ture.
2.
A manual system where the ionization-type smoke detectors and rate-of-rise heat detectors only actuate alarms and do not open any valves. Only the sprinklers that have been exposed to heat intense enough to melt the fusible link in the sprinkler head will spray water when the valve is opened by the operator. The one exception to this is the nozzles at the generator bearing which have open spray nozzles.
3.
Uses the ionization-type smoke detectors and rate-of-rise heat detectors. Both detectors will actuate alarms but the rate-of-rise heat detector will open valves to charge lines up to the sprinkler heads. As the fusible link o
sprinklers are opened due to excessive heat (165 F), they will spray water to extinguish the fire.
1 4.
Will activate when the rise in room temperature causes the fusible link on the sprinkler to melt. When the fusible link melts, the water in the sprinkler header is released through the sprinkler, reducing the pressure in the sprink-1er header, which was holding the deluge valve closed.
When the pressure in the deluge valve is low enough, it will open to charge the sprinkler header to extinguish the fire. When the deluge valve opens, it will alarm in -
the control room.
-Section 2.0 Continued on Next Page-
Page 14 Davis Besse November 19, 1985 Points Available ANSWER 2.01 3
a.
b.
2 4
c.
d.
1 Reference (sl 2.01 Licensing Info. Manual, Vol. II, p. II-334.
QUESTION 2.02 List eight (8) of the thirteen (13) sources of liquids that can be collected in the Miscellaneous Waste Drain Tank.
(2.0)
ANSWER 2.02,
Any 8 of:
chemical addition system, condensate demineralizer holdup tanks, spent resin tank, containment normal sump, steam generator drains, demineralizer flush, makeup and purification system, spent fuel pool overflow, auxiliary building equipment vents and drains, chemistry lab and sample sink drains, PWST overflow and SWST.
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Reference (sl 2.02 Licensing Info. Manual, Vol. III, III-11.
-Section 2.0 Continued on Next Page-
Page 15 Davis Besse November 19, 1985 Points Available OUESTION 2.03 a.
Describe the consequences of a complete rupture of the main condenser circulating water expansion joint on the inlet side of the condenser.
(2.0) b.
What plant design feature prevents flooding above the 585 elevation?
(1.0)
ANSWER 2.03 a.
The two (2) circulating pumps on one (1) pipe will run out.
In addition, water from the cooling tower and piping will flow back through the rupture.
The east and west condenser pit will flood, flooding the main feed pump turbine, lube oil sump, lube oil pump motors and condensate pump motors resulting in a trip of these pumps. Assuming a relatively low flow rate from the east to west pit, water will flood the east side pit to E1. 585.
Water flowing over E1. 585 will flow back into the west side pit through the large grating areas. Above E1. 585, water should flow into the circulating pump house and through leakage paths in the railroad door and other external doors. Water may also enter the following essen-tial areas:
Service Water Tunnel access stairwell Auxiliary Feed Pump Room access stairwell in the turbine building Component Cooling Water Equipment Room High Voltage Switchgear Rooms through doorways
+1.0 - knock out feed and condensate)
+0.5 - flooding east and west pit and out at 585 level)
+0.5 - danger of entering essential areas)
Section 2.0 Continued on Next Page-
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Page 16 Davis Besse November 19, 1985 Points Available ANSWER 2.03 (cont) b.
In order to prevent flooding above E1. 585, a high level switch has been provided in the condenser pits to alarm and alert the operator (computer alarms) of a high condenser pit sump level. This switch will also start the Condenser Pit Flood Pump.
(+0.5-alarm)
(+0.5 - pump start) l Reference (s) 2.03 AB-1203.24, Circulating Water Pump Trip / Circulating Water System Rupture, p. 5.
QUESTION 2.04 List the three (3) sources of makeup water to the Component Cooling Water (CCW) System in the order in which they are in-tended to be used.
(1.5)
ANSWER 2.04 i
1.
demineralized water 2.
primary water
(+1.5) 3.
service water Reference (s) 2.04 AB 1203.31, Component Cooling Water System Malfunctions, p. 7.
-Section 2.0 Continued on Next Page-
Page 17 Davis Besse November 19, 1985 Points Available OUESTION 2.05 What are the two (2) reasons for why bypass flow is maintained around the pressurizer spray control valve?
(1.0)
ANSWER 2.05 1.
To eliminate abnormal buildup or dilution of boric acid within the pressurizer.
2.
To minimize cooldown of the coolant in the spray and surge lines.
Reference (s) 2.05 SP 1103.05, Pressurizer Operation, p. 7.
QUESTION 2.06 When condenser vacuum is being established, why should the level in the deaerator be closely monitored?
(1.0)
ANSWER 2.06 When establishing vacuum, the condensate flow will be through the condensate pump recirculation line to provide cooling water to the steam jet air ejector and steam packing exhauster. If deaerator level control valve CD420 and CD421 should leak I
through, the deaerator and deaerator storage tank could become pressurized beyond design limits of 1000 psig (f not important).
Reference (s) 2.06 SP 1106.16, Condensate System, p. 10.
(
-Section 2.0 Continued on Next Page-l i
Page 18 Davis Besse November 19, 1985 Points Available QUESTION 2.07 Which system is used to provide the head necessary to vent the air from the high points in the Spent Fuel Pool (SFP) cooling and cleanup system piping?
(0.5)
ANSWER 2.07 BWST Reference (sl 2.07 SP 1104.06, SFP Cooling and Cleanup, p. 11.
QUESTION 2.08 List two (2) plant conditions where it would be necessary to utilize ltwo (2) CCW pumps and heat exchangers to provide the necessary heat removal capability.
(1.0)
ANSWER 2.08 1.
The station is undergoing a reactor shutdown and cooldown.
2.
Operating both Diesel Generators at the same time or operat-ing the Diesel Generator cooled by the Essential Loop opposite that which is already operating.
Reference (sl 2.08 SP 1104.12, CCW System, p. 13.
-Section 2.0 Continued on Next Page-
Page 19 Davis Besse November 19, 1985 Points Available OUESTION 2.09 Why must a condensate pump be in operation prior to drawing a vacuum in the main condenser?
(1.0)
ANSWER 2.09 Cooling water to the SJAE and Steam Packing Exhauster
(+0.5 each)
Reference (s) 2.09 SP 1104.35, Vacuum Systems, p. 4.
QUESTION 2.10 The BWST heater is gized such that it can increase BWST tempera-ture fr6m 50 to 90 F in (Selectone.)
(1.0) a.
a matter of hours b.
a day c.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> d.
I week if weather conditions are not too severe.
ANSWER 2.10 a.
(very powerful heater warranting a caution)
(+1.0) l l
Reference (s) 2.10 SP 1104.66, BWST, p. 8.
l l
-Section 2.0 Continued on Next Page-
Page 20 Davis Besse November 19, 1985 Points Available 00ESTION 2.11 a.
What is the capacity of one (1) Auxiliary Feedwater Pump (AFP) based on?
(0.6) b.
List the three (3) sources of water to the Auxiliary Feedwater Pumps.
(1.5)
ANSWER 2.11 a.
Decay heat removal requirements (+ 40 seconds) after a reactor trip from full power.
(+0.6) b.
1.
Condensate Storage Tanks 2.
Service Water System
(+1.5) 3.
Fire Protection System Reference (s1.2.11 SP 1106.06, Auxiliary Feedwater, p. 1.
-Section 2.0 Continued on Next Page-
Page 21 Davis Besse November 19, 1985 Points Available OUESTION 2.12 Describe the two (2) conditions where " piggyback" (suction to the HPI pumps from the DH pump discharge) operations would be required.
(2.0)
ANSWER 2.12 1.
If makeup to the RCS is required at a pressure higher than HPI discharge pressure, the additional head of the DH pumps can raise the injection pressure, or 2.
If, in the event of a small RCS leak, the RCS pressure is greater than the discharge pressure of the DH pumps such that HPI is required, and the BWST is nearing its low level setpoint, (8 ft), HPI suction must be supplied from
~
the emergency sump through the DH pumps. The lineup for both of these operations is the same except that the first will use water from the BWST while the second will start with BWST water and switch to the containment emergency sump.
r ea.. <. k,
- nwaf g
5se se ac Reference (s) 2.12 g t,4 SP 1104.04, DH System, p. 29.
00ESTION 2.13 i
i IRUE or FALSE. A mechanical stop has been added to the pres-i surizer spray valve so that the maximum that it can be open, in manual or automatic, is 40%.
(0.5)
ANSWER 2.13 FALSE l
Reference (s) 2.13 LIM, Vol. I, p. II-36.
-Section 2.0 Continued on Next Page-i i
I Page 22 Davis Besse November 19, 1985 Points Available 00ESTION 2.14 Why is the "011 Lift" System for reactor coolant pumps needed?
(1.0)
ANSWER 2.14 Since pressure in t'he primary coolant loops act on the reactor coolant pumps to produce a high upthrust at standstill, the thrust bearing friction (or breakaway torque) is considerably higher than that which the driving motor can overcome.
Reference (s) 2.14 LIM, Vol. I, p. II-45.
OUESTION 2.15 What is the difference between the " operational" and " spare" purification demineralizer?
(1.0) s e
I ANSWER 2.15 The operational demineralizer is lithium saturated, whereas the spare demineralizer is not.
Reference (s) 2.15 LIM, Vol. I, p. II-53.
-Section 2.0 Continued on Next Page-
Page 23 Davis Besse November 19, 1985 Points Available OUESTION 2.16 Identify the regular and supplemental makeup supplies to the cooling tower.
(1.0)
ANSWER 2.16 2 " ' e.'
+ 5 v >
Service water - regular 7,
"C Cooling tower makeup pumps - supplemental)
' > f f ' ' - -- <t.t / v rtc3 h e u. d;.
up w e.& v
'a ^sI'l i F"' t -
Reference (sl 2.16 LIM, Vol. I, p. II-91.
QUESTION 2.17 Brieflyl describe how H2 gas is removed from the generator (H )
2 side hydrogen seal oil. Eire the tasic flow path and ultimate release point for the hydrogen.
(1.4) l ANSWER 2.17 The oil drains to the H2 detraining tank. There the bulk of the H2 escapes. Then it is combined with the air side oil and sent to a vacuum tank. The vacuum is maintained by a seal cii vacuum pump which discharges to the roof.
Reference (sl 2.17 LIM, Vol. I, p. II-128.
I l
-Section 2.0 Continued on Next Page-
Page 24 Davis Besse November 19, 1985 Points Available' OUESTION 2.18 QIAM a basic one-line diagram of the 4160V system from the "A" and "B" bus to the "C1" and "D1" bus. Include the diesel generators and major breakers in your drawing. You do not need to show electrical loads.
(2.0)
ANSWER 2.18 See attachment.
Reference (sI 2.18 4160V System Switching Procedure, SP 1107.05, Attachment 1.
I l
1 Y
-Section 2.0 Continued on Next Page-l e
S
Page 25 Davis Besse November 19, 1985 Points Available ANSWER 2.18
.Ag IfttA AC (THE 80 t
i A AC C2 A BO D2 DIESEL SUS D2 SUS C2 DIESEL CEhEN Afut i GENEA TOR 2 Start up Feeduster rump Condensate Fuey 4
-4 Condensate Pump 2 Condenaste Fu.e 3 Aug. Beller Forced Draft Fee Coelles Tever etake-up Puey 1 Coeling Tever flake-up Pump 2 Coeling Tever Puer i Mester Drain rump 2 Coeling Water Pump 3 Suttchyear Feeder 2 lleater Drain Pump i Lighting Substa. Trans. DES Suitchyard Freder 1 Cee.llag Water rump 2 Lighting Substa. Trans. CF5 Ser. Sids. Substa. Treas. OF6 Backup Service Water Pump 1 ffri via DSC205 l)AClIOl)AClto)A80Cl A AC Di A 01 O n A DB 10 l
BUS Cl l~
SUS Di High Pressure lajection Pump 1 Nigh Pressure lajectise Pump 2 Decay liest Puep I Decay Nest Pump 2 Co.penent Ceeltog rue.p 1 C.=pement Ceeltag Pump 2 Service Water Pump l Service Water Pump 2 Ilakeup Fump i llakeup Fump 2 Transfesner CEl l Tranetermer DFl-l yreasiereerCEl2 Traestormer OFl-2 Cl feed to Service Water rim.1 Di Feed to Service Water Puey 3 C ree4 s. Cee,eneus C..ii g re-, 2 mi r.e4 i. C, eet C.eii.g r, 2
-End of Section 2-
Page 26 Davis Besse November 19, 1985 Points Available 3.0 INSTRUMENTS AND CONTROLS (25.0) 00ESTION 3.01 A LOCA is occurring. HPI pumps are keeping up with the leak, o
and RCS pressure has stabilized around 1500 psi (55 F saturation margin). As reactor operator you note that core flood tank pressure is the same as before, but that level has increased 4 inches. Exolain why.
(1.0)
ANSWER 3.01 Elevated reference leg temperature; lack of pressure change means actual level same as before.
Reference (sl 3.01 SP 1104f01.13 I
-Section 3.0 Continued on Next Page-I
Page 27 Davis Besse November 19, 1985 Points Available OUESTION 3.02 Liit the three (3) conditions that must exist to enable the ICS feed and bleed permissive.
(1.5)
ANSWER 3.02 1.
Control rod groups 1, 2, 3, and 4 must be 100% withdrawn for any feed and bleed operations.
2.
Control rod group 5 must be greater than 25% withdrawn.
3.
Reactor power determines the necessity of a third require-ment.
If the reactor power is greater than 15% full power, the actual position of group 7 must be more than 5% above or below the nominal position on the transient rod position band (see PP 1102.04, Power Operations for the figure showing transient band nominal position).
If reactor pow'er is less than 15% full power, group 7 position has no'effect on the permissive.
Reference (s) 3.02 SP 1103.04, Boron Concentration Control, p. 12.
-Section 3.0 Continued on Next Page-
. _ _ _ ~.
Page 28 Davis Besse November 19, 1985 Points Available OUESTION 3.03 In the event that the Steam and Feedwater Rupture Control System (SFRCS) does not actuate automatically, describe the two (2) manual ways to actuate the system.
(2.4)
ANSWER 3.03 1.
Each input device (PS, PDS, LSLL, etc.) has a test button on the input buffer card in the SFRCS cabinets. Pushing one of these buttons will have the same effect as the input device reaching its setpoint.
2.
There are ten (10) manual actuation switches on the panel C-5721 in the control room. Each switch is an HIS type with a red " TRIPPED" light and " Trip /Off" pushbuttons.
3, Atu,.A vl. s :J J cc-m c a wT rut;m.
Reference (s1 3.03 SP 1105.16, SFRCS Procedure, p. 7.
QUESTION 3.04 If the turbine bypass valves H/A selector stations are in " HAND",
and S/G exit pressure exceeds 1050 psig, what will happen?
(1.0)
ANSWER 3.04 The turbine bypass valves will open.
(+1.0)
Reference (s1 3.04 SP-1106.24, Main Steam System, p. 3.
i 1
1
-Section 3.0 Continued on Next Page-l f
r
I Page 29 Davis Besse November 19, 1985 Points Available OUESTION 3.05 Containment Recirculation Fans 1-1 and 1-2 are normally con-trolled from the Control Room by the manual control switch.
If the Containment Recirculation Fan Emergency Control Transfer Switch is turned to the LOCAL position, where does control transfer to?
(1.0)
ANSWER 3.05 Motor Control Center (MCC) Pushbutton (+1.0)
Reference (s) 3.05 SP 1104.54, Containment Recirculation, p.1.
QUESTION 3.06' Briefly state the difference between control of pressurizer heater bank 1 and the control of pressurizer heater banks 2, 3, and 4.
(1.0)
ANSWER 3.06 Heater bank 1 is SCR controlled and, therefore, has a variable output; whereas, heater banks 2, 3, and 4 are either on or off.
Reference (s) 3.06 Pressurizer Operation, SP 1103.05, p. 2.
-Section 3.0 Continued on Next Page-
l Page 30 Davis Besse November 19, 1985 Points ayailable OUESTIOfLimQZ IRUE or FALSE. The auto-opening feature of the power operated relief valve may be overridden by pushing the operating switch "IN'.
(0.5)
ANSWER 3.07 FALSE Reference (s) 3.07 Pressurizer Operation, SP 1103.05, p. 6.
QUESTION 3.08 a.
Which S/G 1evel instrumentation is temperature compensated? (1.0) b.
Which S/G level instrumentation is essentially powered?
(1.0) c.
Where can startup S/G levels above 250 inches be read?
(1.0)
ANSWER 3.08 a.
Operate range b.
Startup level, 5/7tch * *LI c.
Steam and Feedwater Rupture Control System (SFRCS) S/G level cabinets at the bar' af the control room.
Reference (s) 3.08 AB 1203.34, Steam Generator Overfill, p. 4.
-Section 3.0 Continued on Next Page-
i Page 31 Davis Besse November 19, 1985 Points Available OUESTION 3.09 WhatmustbedonetoarmtheRapidFeedwaterReduction(RFR) system? (Three (3) items required.)
(1.5)
ANSWER 3.09 1.
The defeat switch in the ICS is on.
2.
A MFPT is reset.
3.
All four (4) FW control valves are in auto.
Reference (s) 3.09 Trip Reccvery, PP 1102.03.13, p. 4.
QUESTION 3.10 The pressurizer heaters are interlocked in such a way that the heaters will not be energized when what two (2) conditions are simultaneously present?
(1.0)
ANSWER 3.10 DH-11 or DH-12 is open, &HD RCS pressure is > 301 psig.
Reference (s) 3.10 Licensing Information Manual, Volume 1, Section IIA, p. II-36.
-Section 3.0 Continued on Next Page-n-,
Page 32 Davis Besse November 19, 1985 Points Available OUESTION 3.11 List four (4) of the seven (7) reactor coolant pump starting interlocks.
(Numerical values not required.)
(1.6)
ANSWER 3.11 Voltage Oil lift system pressure Upper and lower bearing oil reservoir levels Component cooling water flow Seal injection flow Reactor power Cold leg temperature Reference (s) 3.11 LIM, Vol. I, p. II-48.
QUESTION 3.12 What happens if MU 3971 does not transfer to provide suction to the MVPs from the BWST upon reaching 10 inches in the Makeup Tank?
(1.0)
ANSWER 3.12 A 45 second (exact value not important) time delay will auto-matically trip the makeup pumps.
Reference (s) 3.12 LIM, Vol. I, p. II-51.
l
-Section 3.0 Continued on Next Page-
Page 33 Davis Besse November 19, 1985 Points Available 00ESTION 3.13 Describe the radiological interlock associated with the CCW surge tank.
(1.5)
ANSWER 3.13 The surge tank normally vents to the atmosphere. Should CCW activity be detected above limit, the vent path is automatically diverted to the miscellaneous waste drain tank via a three-way valve.
Reference (s) 3.13 LIM, Vol. I, p. II-105.
OUESTION 3.14 Describe the two (2) interlocks associated with the " third" service water and component cooling water pumps.
(2.0)
ANSWER 3.14 The third pump is mechanically interlocked such that only one (1) breaker supply may be closed at one time, therefore prevent-ing parallel operation of the two (2) safety busses. Addition-ally, the pumps are interlocked such that more than one (1) component cooling pump and more than one (1) service water pump may not be connected to either diesel generator at the same time.
Reference (s) 3.14 LIM, Vol. I, p. II-143.
-Section 3.0 Continued on Next Page-
Page 34 Davis Besse November 19, 1985 Points Available OUESTION 3.15 Draw a top view of the reactor showing the location of the out-of-core neutron detectors labeling each detector and stating its type.
(2.0)
ANSWER 3.15 X
PC Cic (NI.3)
(NI 1) t g
I *\\
N-)
Ni )
h
@p#
b"ilh.a.-M 4iin
,n y'0
- . l?-y Mi*!
W-
- e Hi@M D
N*'
o o
o/
'qip
.- ' ID, uCiC
.+.E
- j y4'. "
(N16) uCi Nsi CIC PC (NIO) 1 (NI-2) q Z
, Legend PC Proportional Counter. Source Range Detector CIC Compensated lon Chamber. Intermediate Range ostector UCIC Uncompensated son Chamber. Power Range Detector Reference (s) 3.15 LIM, Vol. 2, p. II-187.
-Section 3.0 Continued on Next Page-i
Page 35 Davis Besse November 19, 1985 Points Available OUESTION 3.16 IRUE or FALSE. The scale on the intermediate range instrument startup rate meter is sufficient to observe the stable negative startup rate occurring after a reactor trip.
(0.5)
ANSWER 3.16 TRUE Reference (s) 3.16 LIM, Vol. 2, p. 190.
- 4 4
-Section 3.0 Continued on Next Page-4
Page 36 Davis Besse November 19, 1985 Points Available OUESTION 3.17 For the five (5) SFAS incident levels, lisi the trip setpoints and glya a general [one (1) phrase] description of what action results upon trip of each incident level.
(2.5)
ANSWER 3.17 Level Trio Setooint Action 1
High containment radiation (2 x back-Containment ground at rated thermal power) c isolation High containment pressure (18.4 ps Low RCS pressure (1650 psig) c 2
High containment pressure (18.4 psig)
Initiate high pressure Low RCS pressure (1650 psic) c injection system 3
High containment pressure (18.4 psig)
Initiate low pressure Low low RCS pressure (450 psig)
(
injection system 4
High high containment pressure Initiate containment (38.4 psig)'
spray and cooling A
system 5
Low low BWST level (2 91.5 in and Containment emergency 5 100.5) sump recirculation E, 0 4, M interlock Reference (s) 3.17 LIH, Vol. 2, p. II-206.
-End of Section 3-O O
Page 37 Davis Besse November 19, 1985 Points Available 4.0 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY. AND RADILOGICAL (25.0)
CONTROL OUESTION 4.01 Containment vessel pressure increases to 18.4 psia. The Con-tainment Spray Automatic Control Valves do not operate correctly and have to be opened manually from panel C-5716. Assuming containment pressure increases to above 38.4 psia, What must be done with the control valves when suction transfer from the BWST to the Emergency Sump is made?
(1.0)
ANSWER 4.01 The control valves will have to be manually throttled.
(+1.0)
Reference (s) 4.01 SP 1104'.05.15, p. 2.
OUESTION 4.02 A Core Flood (CF) line breaks downstream of the LPI tie-in.
This renders one (1) LPI string inoperable. The Diesel supplying the second LPI pumo fails. The operator opens the motor-operated LPI crossover valve (DH830 or DH831), but forgets to close the suction valve on the disabled pump. What will happen?
(1.0)
ANSWER 4.02 Pressurize pump suction piping and lift emergency sump relief valve.
(+1.0)
Reference (s) 4.02 SP 1104.04.19, p. 4.
-Section 4.0 Continued on Next Page-4
Page 38 Davis Besse November 19, 1985 Points Available 00ESTION 4.03 Describe in general how to position Group 8 rods. Assume they are fully inserted and it is desired to withdraw them to 28%.
(1.0)
ANSWER 4.03 Group 8 is permanently latched and electrically inhibited from going below the in-limit position. Simply reset the PIs, Group set to 8, withdraw rods.
(+0.25 each including not latching)
Reference (s) 4.03 SP 1105.09, CRD System Operation, p. 17.
QUESTION 4.04 a.
What S/G level is considered " dry"?
(0.5) b.
How is minimum level re-established once it has been lost?
(0.5)
ANSWER 4.04 1.
5 8 inches on startup range level indication (+0.5) 2.
manual actuation of auxiliary feedwater through the auxiliary nozzles (+0.5)
Reference (s) 4.04 SP 1106.07, Main Feedwater System, p. 8.
-Section 4.0 Continued on Next Page-
,m...,
~
Page 39 Davis Besse November 19, 1985 Points Available OUESTION 4.05 The plant is initially at 100% full power. An automatic reduc-
~
tion to 25% indicated full power occurs due to a load rejection.
continues to be ramped After low level limits are reached, T*Eling lowered and what down by the ICS? Exolain why T is must be done to avoid a low prell8re trip.
(1.6)
ANSWER 4.05 The power range NIs will be reading significantly higher than actual power due to the power decrease causing a hotter Tc and less neutron shielding.
If the plant was initially at 100% of full power, power range NIs could be reading as much as 7 to 10% higher than actual at the end of the power reduction.
This could cause the automatic power reduction to 25% indicated To avoid. a low pressure tfT$,after low level limits are reached.
full power to ramp down T the reactor may have to be placed into manual when T,y, begins to drop.
(+0.8 - lower power than indicated;
+0.8 - manual reactor control)
Reference (s) 4.05 AB 1203.27, Load Rejection, p. 4.
QUESTION 4.06 If the actual and the estimated critical control rod positions differ by more than 0.5% Ak/k, then what must be done?
(1.0)
ANSWER 4.06 The regulating rods must be inserted until the problem is analyzed and resolved.
Reference (s) 4.06 Approach to Criticality, PP 103.08.6, p.2.
-Section 4.0 Continued on Next Page-l
Page 40 Davis Besse November 19, 1985 Points Available 00ESTION 4.07 A reactor coolant pump is idle, and seal injection to it has been lost. What must be done with controlled bleedoff? Why?
(1.5)
ANSWER 4.07 bcn $ *I ecturn B. Zooleg<c(
The controlled bleedoff valve should be shut because without sec 4 4 ull/6 seal injection and with no auxiliary impelle? to force water 1
through the seal cooler, there will be no means to cool the
-#.-- n e,,f,
seals.
O Reference (s) 4.07 Nuclear Steam System Limits and Precautions PP 1101.01.7, p. 21.
QUESTION 4.08 During a plant heatup along the 100 F/per hour heatup line, the Plant Startup Procedure, PP 1102.02, specifies an instantaneous limitation on heatup rate. What is that heatup rate limit?
(1.0)
ANSWER 4.08 1.67 F/ minute.
If the temperature deviates by more than 15 F fromthetemperaturewhichwouldoccuratthatgointintime, assuming the heatup rate was maintained at 1.67 F per minute, theheatupratemustgeadjustedsuchthattemperatureismain-tained within this 15 F limit.
Reference (s) 4mQB Plant Startup, PP 1102.02, p. 24.
l
-Section 4.0 Continued on Next Page-
Page 41 Davis Besse November 19, 1985 Points Available
. QUESTION 4.09 A reactor coolant pump must be tripped within what maximum interval if both seal injection and CCW are lost? (Select one.)
(1.0) a.
5 10 seconds b.
5 30 seconds c.
s 60 seconds d.
s 90 seconds.
ANSWER 4.09 d.
Reference (s) 4.09 Reactor Coolant Pump Operating Procedure, SP 1103.06.9, p. 7.
v i
l.
-Section 4.0 Continued on Next Page-
Page 42 Davis Besse November 19, 1985 Points Available 00ESTION 4.10 a.
What approximate power level on the intermediate range instrument corresponds to the point of adding heat?
(0.5) b.
What value should startup rate (SUR) be below prior to the point of adding heat? (Select one and explain why.)
(1.0) 1.
No limit 2.
1 3.
0.5 4.
0.1 ANSWER 4.10 Between 5 x 10-8 and 5 x 10-7 a.
amps.
b.
Less than 0.1 DPM to ensure that the RCS does not heat up and surge into the pressurizer faster than the letdown system can reduce RCS inventory.
(Answer 4)
Reference (s) 4.10 Plant Startup, PP 1102.02.21, p. 42.
-Section 4.0 Continued on Next Page-1 0
7
Page 43 Davis Besse November 19, 1985 Points Available OUESTION 4.11
. State the sequence for starting a Makeup Pump.
(1.5)
ANSWER 4.11 1.
Start the AC oil pump (DC oil pump will auto start, run momentarily, and stop).
2.
Verify auxiliary gear oil pump auto starts.
3.
Start the MU pump.
Reference (s) 4.11 Emergency Procedure, EP 1202.01.0, p. 16.
QUESTION 4.12 An RCS 1650 psig trip of SFAS has occurred. The first action called for by the emergency procedure is to verify proper SFAS Incident Level 1 and 2 actuation. State three (3) of the remaining five (5) actions.
(1.5)
ANSWER 4.12 Close RC11, PORV block valve Close RC10, pressurizer spray block valve If subcooling margin is not adequate,
'ip all RCPs Ensure MVP sucti'an shifts to BWST at 10 in.
Verify HPI is operating IAW rules 1 and 2.
Reference (s) 4.12 Emergency Procedure, EP 1202.01.1, p. 26.
~
-Section 4.0 Continued on Next Page-
T.
I it,
Page 44 Davis Besse November 19, 1985 Points Available
\\
OUESTION 4.13 Liit the six (6) conditions that require implementation of the Emergency Procedure, EP 1202.01.
(2.4)
ANSWER 4.13 Reactor trip SFAS trip SFRCS trip SG tube rupture larger than MU capacity Optistor judgement When directed by another procedure.
l c
Reference (si 4.13 Emergency Procedure, EP 1202.01.0, p. 1.
QUESTION 4.14
\\
One Diesel Generator is out-of-service and it would be desirable to take a Decay Heat Pump on the opposite bus out-of-service for maintenance. Would this be permissible by Technical Speci-fications? Exolain.
(1.5)
~
ANSWER 4.14 No. The Decay Heat Pump on the bus effected by the inoperable Diese,1 would have to be declared inoperable if the redundant
}
Decay Heat Pump were to be rendered inoperable.
5 Reference (s) 4.14
)
Techical Specifications, 3.05, p. 3/4 0-1.
s
-Section 4.0 Continued on Next Page-
?
I
\\
(
I 4
Page 45 Davis Besse November 19, 1985 Points Available OUESTION 4.15 When using the figures in the Reactor Operators Curve Book (PP 1101.02), for which one of the parameters listed below, is it permissible to interpolate. (Select one.)
(1.0) a.
between different temperatures b.
between " power" and " critical" reference conditions c.
between BOL (Beginning-of-Life) and EOL (End-of-Life) conditions d.
never ANSWER 4.15
- c. Only (+1.0) s Reference (si 4.15 PP 1101.02.7, p. 1.
QUESION 4.16 List the three (3) chemicals used'at Davis Besse from which a major toxic g+.s hazard could come.
(1.5)
ANSWER 4.16 1.
ammonium hydroxide 2.
chlorine 3.
hydrazine Reference (s) 4.16 AB 1203.14, Toxic Gas and Liquid Release, p. 4.
-Section 4.0 Continued on Next Page-
r-Page 46 Davis Besse November 19, 1985 Points Available OUESTION 4.17 Whn relieves the shift supervisor as Interim Emergency Duty Officer?
(0.5)
ANSWER 4.17 Shift Technical Advisor Reference (s) 4.17 EI 1300.02, Emergency Plan Implementing Procedure, Unusual Event, p. 1.
QUESTION 4.18 According to your Radiation Protection Manual HP 1602.01, a pocket dosimeter:
(Selectone.)
(1.0) a.
should be zerced as often as possible.
b.
should be zeroed at least every time that RACA is entered.
c.
should be zerced only when a reading of 20% full scale is achieved.
d.
should be zerced only when a reading of 80% full scale is achieved.
ANSWER 4.18 c.
Reference (si 4.18 External Personnel Radiation Exposure Monitoring, HP 1602.01 T-8937.
-Section 4.0 Continued on Next Page-
i Page 47 Davis Besse November 19, 1985 Points Available OUESTION 4.19 A control rod is stuck. Attempts to free it should only be made in what speed? Exolain.
(1.5)
ANSWER 4.19 The CRDM should only be operated in "run" speed to keep from overloading the control rod drive spider assembly.
Reference (s) 4.19 NSSS Limits and Precautions, PP 1101.01.6, p. 27.
QUESTION 4.20 During the Loss of Feedwater event, the equipment operators had considerable difficulty resetting the trip throttle valves for the No.1 and 2 AFW pumps. What trouble were they having and han was it resolved?
(1.0)
ANSWER 4.20 The equipment operators had only removed the slack in attempting to open the valve. The valve, however, was still closed and the differential pressure on the wedge disk made it difficult to turn the handwheel after the slack was removed. The valve was opened by using a valve wrench.
Reference (s) 4.20 NUREG-1154, Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9. 1985 i
-End of Section 4-l
-End of Examination-l l
l
EQUATION SHEET Where my = m2 (density 1 (velocity 1 (area)1 = (density)2(velocity 1 (area)2 1
1 2
where Y = specific 2
PE +KE +P V i i = PE +KE +P Y22 KE = mv PE = mgh 2
2 i
i
'T volume P = Pressure l
Q = ic (Tout-Tin)
Q = UA (T
-Tstm)
Q = m(h -h I ave i 2 p
P = P 10(SUR)(t) p, p e /T SUR = 26.06 T = (B-p)t t
o o
T p
I CR (1-Keffi) = CR Il*Keff2)
CR = S/(1-Keff delta K = (K,ff-1) 1 2
M = (1-Kegft)
SDM = (1_Keff) x 100%
K (1-Keff2) eff 1 = A e-(decay constant)x(t) 0.693 A
In (2) decay constant
=
=
g t
t1/2 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 i
3 1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /tr 6
3 1 MW = 3.41 x 10 Btu /hr Density =62.4lbg/ft Density = 1 gm/cm i 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 centimeters 2
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec
U. S. NUCLEAR REGULATORY COFMISSION SENIOR REAC10R OPERATOR LICENSE EXAMINATION Facility:
Davis Besse 1 Reactor Type:
PWR Date Administered: November 19, 1985 Examiner: W.J. Apley /J.C. Huenefeld g
Candidate:
Answer Key I
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 wi.11 be picked up six (6) hours af ter 'the examination starts.
l f
Category
% of Candidate's
% of Value Total Score Cat. Value Category 25 25
- 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics
- 6. Plant System Design, Control 25 25 and Instrumentation
- 7. Procedures - Normal, Abnormfi, 25 25 Emergency, and Radiological Control 25 25
- 8. Administrative Procedures, Conditions, and Limitations 10TALS 100 Final Grade 1
All work done on this examination is my own; I have neither given nor received aid.
Candidate's Signature O
,ce
_-..--n
--,,y
--,,.,.,,,,--..n------.,-n
,,-.em-
T Davis Besse 1 Page 1 November 19, 1985 Points Available 5.0 THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS.
(25.0)
AND THERM 0DvNAMICS l
00ESTION 5.01 a.
Define shutdown margin.
Include any assumptions about axial power shaping rod or control rod positioning.
(2.0) b.
During a plant heatup in accordance with the Plant Startup Procedure, PP 1102.02, the reactor must be shutdown by 21% Ak/k. If Group 1 rods are withdrawn, does their worth count as part of that 1% Ak/k?
(1.0)
ANSWER 5.01 a.
SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of re-activity by which the reactor is subcritical or would be subcritical from its present condition assuming:
No change in axial power shaping rod position, and All control rod assemblies (safety and regulating) are fully inserted except for the single rod assembly of highest reactivity worth which is assumed to be fully withdrawn.
b.
No. Even if Group 1 rods are to be withdrawn, the shutdown value must be 2 1% Ak/k.
Reference (sl 5.01 a.
Technical Specifications, p.1-3.
b.
Plant Startup, PP 1102.02.20, p. 20.
-Section 5 Continued on Next Page-4 4
-~
--m..
.r
Page 2 Davis Besse 1 November 19, 1985 Points Available 00ESTION 5.02 While drawing a bubble in the pressurizer, the vapor space is vented to the quench tank. Describe how conditions in the quench tank can be used for determining when a steam bubble exists in the pressurizer.
(2.0)
ANSWER 5.02 Early in the venting process most of the gas vented to the quench tank will be nitrogen. Manual operation (from Control Room) of RC 222 Quench Tank Vent to Containment vent header will be required to keep quench tank pressure below 50 psig.
As the venting process continues, level and temperature control of the quench tank will be required to handle the condensed steam.
Reference (si 5.02 Pressurizer Operation, SP 1103.05, p. 13.
OUESTION 5.03 During a reactor plant startup with three reactor coolant pumps running, Elll Quadrant Power Tilt (QPT) be more limiting at low power or high power? Exclain.
(2.0)
ANSWER 5.03 Joetk. 0 5 Because of the way QPT is calculated and because Della_T vdll_
be worse at lower oower, the worst values of QPT should Be
> increases.,Qw power with progressively better values as power expected 3t
~~~'
wo t% I, O ArTh (5
r Reference (si 5.03 Plant Startup, PP 1102.02.21, p. 40.
-Section 5 Continued on Next Page-
O Page 3 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.04 Curve "A" on the trace below is a logarithmic plot of total neutron power versus time after a reactor trip. Curve "B" is a plot of the neutron power due to delayed neutrons alone versus time for the same trip. Exolain why total neutron power (i.e.,
curve A) does not drop all the way down to the delayed neutron level (i.e., to curve B).
(2.0)
A - TOTAL NEUTRON POWER B - DELAYED
% s,,"-
C NEUTRONS 0
U N
y R
A T'
E ill1E ANSWER 5.04 The magnitude of the prompt drop is dependent upon the amount of negative reactivity that is inserted. The countrate above and beyond that due to delayed neutrons alone is caused by subcritical multiplication of the delayed neutrons.
If an infinite amount of negative reactivity were inserted, then Curve A would fall immediately to Curve B upon Rx trip.
Reference (s) 5.04 LIM, Vol. I,Section I.
-Section 5 Continued on Next Page-
Page 4 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.05 The reactor is critical at 10-8 amps. A stable 1 DPM startup rate is achieved.
If rods are inserted continuously until startup rate drops to zero, and then the rod insertion is immediately stopped, will the reactor be critical, supercritical, or sub-critical? Exnlain.
(2.0)
ANSWER 5.05 The reactor will be supercritical. Because the reactor is
" prompt subcritical," the insertion of negative reactivity will cause prompt power to turn before causing delayed power to turn. The "subcritical effect" of the prompt neutrons is prevalent because prompt neutrons comprise about 99% of all neutrons in the reactor.
Reference (s) 5.05 LIM, Vol I,Section I.
-Section 5 Continued on Next Page-i
Page 5 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.06 The figure below shows neutron population response to reactivity insertion. Label each of the three (3) curves as supercritical, exactly critical, or subcritical. (Assume below the point of feed-backeffects.)
(1.0)
A B
C N,'
,l h
Neutron Population N, s N,%
. N,s ~~7( I N. -
j!
l
/I\\
At, gg, Ata tirne in seconds a.
supercritical b.
exactly critical c.
subcritical Reference (s) 5.06 LIM, Vol. I,Section I, p. I-16
-Section 5 Continued on Next Page-
Page 6 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.07 Given a large vented tank 30 ft. In diameter and 60 ft. high with a centrifugal pump taking a suction from its base. The pump is located at a vertical elevation corresponding to the bottomofthetagk, The tank is entirely full of water and is maintained at 60 F by heaters. Assume the vent becomes totally clogged while the pump is in operation. Atmospheric pressure is 14.7 psia. Answer the following questions:
a.
What is the maximum differential pressure that could occur between the inside and outside of the tank? Exolain.
(1.5) b.
Exolain why the pump may begin to cavitate at a higher tank level than with the vent open.
(1.0)
ANSWER 5.07 Thelowestpressurethatthetgnkcoulddroptowouldbe a.
the-saturation pressure for 60 F which is 0.256 psia. So maximum AP =.14.7 - 0.256 = 14.444 psia (full credit).
(1.4 pts if simply state 14.7 psia.)
b.
Without atmospheric pressure on top of the water, the equivalent head is lost putting the pump closer to its minimum NPSH limit.
Reference (s) 5.07 Centrifuaal Pumns and System Hydraulics, Igor J. Karassik.
-Section 5 Continued on Next Page-
1 Page 7 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.08 i
l When conducting a rapid plant shutdown at 25 MW per minute (Abnormal Procedure AB 1203.07), why is the operator required l
to add boric acid to the makeup tank as unit load decreases?
J)
ANSWER 5.08 To prevent an imbalance trip.
Reference (s) 5.08 AB 1203.07, Abnormal Procedure.
9 l
~
)
-Section 5 Continued on Next Page-
Page 8 Davis Besse 1 November 19, 1985 Points Available OUESTI 5.09 Assu that the condenser is not available and the operator is manual controlling steam generator pressure at 995 psig using the atmos eric vent valves. Estimatehowmanygg11onsof water woul e required to cooldown the RCS from 582 F to 546 F.
Show assumpt ns.
(2.0) a.
350 gallons b.
3500 gallons c.
35,000 gallons d.
73,000 gallons ANSWER 5.09 b.
(Actually 3,446 gallons; answ r does not require memoriza-tion.
If he kgows 1 gal of stm\\{eleased will cool 1000 gal of water 1 F, then N
(vol. Rc 82 - 546) = answer :
- 3
= 3200 gal.)
88 00 6)
Reference (s) 5.09 AB 1203.03, Cooldown W/0 BWST and No Off-Site Power.'\\
's 4,s Odd 3, N fud6 b i '7 g
voitof at ct vahti nb.
-Section 5 Continued on Next.' age-
Page 9 Davis Besse 1 November 19, 1985 Points Atailable OUESTION E.10 i Consider the system shown below (an 80 ft high loop of 3 ft diameter steel pipe and a gressurizer).
Initially the pregsurizer is saturated at 640 F and the loop is subcooled at 546 F.
Assume that the entire system, including the pressurizer is allowed cooldown. Thethermalcongractionoftheentire system due to this cooldown is 600 ft. Will the pressurizer be empty?
Funlain.
(1.5)
_c_
+
r
+
e.-
"I 1
Aff,R 1
- 4g
~
4 9
}
w ANS] DER 5.10 No, a bubble will fonn in the " candy cane" of the loop, and the loop / pressurizer will form a manometer with level in the loop matching the level in the pressurizer.
Reference (s) 5.10 Fundamentals of Classical Thermodynamics, VanWylen and Sonntag.
-Section 5 Continued on Next Page-l 1
Page 10 Davis Besse 1 November 19, 1985 Points Available OUESTION 5.11 The reactor is at 100% FP equilibrium xenon conditions when a reactor trip occurs. Three (3) hours later the reactor is restarted, When at 10-8 amps, which direction will rods have to go a.
to hold power constant? Exnlain.
(1.0) b.
Sketch the xenon concentration through to equilibrium assuming the reactor is raised to 92% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (4hoursafterthetrip).
(1.5)
ANSWER 5.11 Rodswjl1havetobewithdrawnbecausetheburnoutterm a.
at 10- amps is not significant.
b.
6, g..
t
..)
- neu, t~ w s
V, t% snittst
- ~ ~ * ~ ~ ~ ~
d c.e=<.a n tedio n,
Eo wX l
A e
Y hv Qf a
t:., e. -->
Referencefs) 5.11 LIM, Vol. I, Section 1.
-Section 5 Continued on Next Page-
Page 11 Davis Besse 1 November 19, 1985 Points Available 00ESTION 5.12 When borating from a lower to a higher boron concentration, why will the final boron concentration value always be 3 to 4%
lower than anticipated?
(1.0)
ANSWER 5.12 The letdown purification demineralizer will absorb boron as it becomes saturated at the new value.
Referencefs) 5.12 AB1203.16, Loss of Reactor Coolant System Boron.
QUESTION 5.13 Why are SFRCS tripped and AFW actuated immediately following a loss of NNI X DC power, rather than waiting for the plant transient to initiate an SFRCS actuation?
(1.5)
ANSWER 5.13 The transient from a loss of NNI X DC power would be very severe (loss of pressurizer level and possible hot leg voiding). To avoid this severe transient, SFRCS is tripped and AFW is actuated immediately to avoid " shot feeding" the S/Gs from approximately 26 inches to 40 inches. This will reduce the cooldown rate of the RCS until makeup flow can be increased. This will slow the transient to a more controilable rate under the limiting circumstances.
Reference (s) 5.13 AB 1203.41, Loss of NNI Power, p. 22.
End Section 5-
+
+e-,
--e..---
Page 12 Davis Besse 1 November 19, 1985 Points Available 6.0 PLANT SYSTEM DESIGN. CONTROL. AND INSTRUMENTATION (25.0)
OUESTION 6.01 Briefly state the difference between control of pressurizer heater bank 1 and the control of pressurizer heater banks 2, 3, and 4.
(1.0)
ANSWER 6.01 Heater bank 1 is SCR controlled and, therefore, has a variable output; whereas, heater banks 2, 3, and 4 are either on or off.
Reference (s) 6.01 Pressurizer Operation, SP 1103.05, p. 2.
QUESTION 6.02 1 RUE or EALSE. The auto-opening feature of the power operated relief valve may be overridden by pushing the operating switch "IN".
(0,5)
ANSWER.6,02 FALSE Referencefs) 6.02 Pressurizer Operation, SP 1103.05, p. 6.
-Section 6 Continued on Next Page-
Page 13 Davis 3 esse 1 November 19, 1985 Points Available OUESTION 6.03 a.
Which S/G level instrumentation is temperature compensated? (1.0) b.
Which S/G level instrumentation is essentially powered?
(1.0) c.
Where can startup S/G levels above 250 inches be read?
(1.0)
ANSWER 6.03 a.
Operate range b.
Startup level / 5 /AC5 W cr / m l in,t, s m.. t.. t lo n,
c.
Steam and Feedwater Rupture Control System (SFRCS) S/G (5 ' h ( -(
level cabinets at the back of the control room.
cc,.uu c, g[)
Reference (s) 6.03 AB 1203.'34, Steam Generator Overfill, p. 4.
OdESTION 6.04 WhatmustbedonetoarmtheRapidFeedwaterReduction(RFR) system?
(Three(3)itemsrequired.)
(1.5)
ANSWER 6.04 1.
The defeat switch in the ICS is on.
2.
A MFPT is reset.
3.
All four (4) FW control valves are in auto.
Referencels) 6.04 Trip Recovery, PP 1102.03.13, p. 4.
-Section 6 Continued on Next Page-
Page 14 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.05 The pressurizer heaters are interlocked in such a way that the heaters will not be energized when that two (2) conditions are simultaneously present?
(1.0)
ANSWER 6.05
,,. O. \\
I N
r DH-11orDH-12isopen,AliDRCSpressureis>301psig.
Referencefs) 6.05 Licensing Information Manual, Volume 1, Section IIA, p. II-36.
QUESTION 6.06 IRUE or FALSE. A mechanical stop has been added to the pres-surizer spray valve so that the maximum that it can be open, in manual or automatic, is 40%.
(0.5)
ANSWER.6,06 FALSE Reference (s) 6.06 LIM, Vol. I, p. II-36.
-Section 6 Continued on Next Page-
Page 15 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.07 Why is the "011 Lift" System for reactor coolant pumps needed?
(1.0)
&NSWER 6.07 Since pressure in the primary coolant loops act on the reactor coolant pumps to produce a high upthrust at standstill, the thrust bearing friction (or breakaway torque) is considerably higher than that which the driving motor can overcome.
Reference (si 6.07 LIM, Vol. I, p. II-45.
QUESTION 6.08 List four (4)(of the seven (7) reactor coolant pump starting interlocks.
Numerical values not required.)
(1.6)
ANSWER 6.08 Voltage Oil lift system pressure Upper and lower bearing oil reservoir levels Component cooling water flow g $4.c, v v,llh Q,rv.., s g Seal injection flow Reactor power i
Cold leg temperature A l 5,1f,4 w v e ba "U.ed/(k
[ g,m 4g, 4,(
v Reference (si 6.08
[
LIM, Vol. I, p. II-48.
-Section 6 Continued on Next Page-w-
Page 16 Davis Besse 1
]
November 19, 1985 Points Available
~
OUESTION 6.09 What happens if MU 3971 does not transfer to provide suction to the MUPs from the BWST upon reaching 10 inches in the Makeup Tank?
(1.0)
ANSWER 6.09 A 45 second (exact value not important) time delay will auto-natically trip the makeup pumps.
Reference (s) 6.09 LIM, Vol. I, p. II-51.
QUESTION 6.10 What is the difference between the " operational" and " spare" purification demineralizer?
(1.0)
&HSWER 6.10 The operational demineralizer is lithium saturated, whereas the spare demineralizer is not.
Reference (s) 6.10 LIM, Vol. I, p. II-53.
t i
-Section 6 Continued on Next Page-a
.I
___-_,_.._._.._,__c_
\\
l l
Page 17 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.11 Identify the regular and supplemental makeup supplies to the cooling tower.
(1.0)
ANSWER 6.11 v *fd b 6f A veM +s L
Service water - W d
Cooling tower makeup pumps - supplemental Reference (s) 6.11 LIM, Vol. I, p. II-91.
QUESTION 6.12 Describe the radiological interlock associated with the CCW surge ta,nk.
(1.5)
ANSWER 6.12 The surge tank normally vents to the atmosphere. Should CCW activity be detected above limit, the vent path is automatically diverted to the miscellantous waste drain tank via a three-way valve.
Reference (s) 6.12 LIM, Vol. I, p. II-105.
-Section 6 Continued on Next Page-
Page 18 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.13 Briefly describe how H2 gas is removed from the generator (H )
2 side hydrogen seal oil. Gire the basic flow path and ultimate release point for the hydrogen.
(1.4)
ANSWER 6.11 The oil drains to the H2 detraining tank. There the bulk of the H2 escapes. Then it is combined with the air side oil and sent to a vacuum tank. The vacuum is maintained by a seal oil vacuum pump which discharges to the roof.
Reference (s) 6.13 LIM, Vol. I, p. II-128.
QUESTION 6.14, DRAW a basic one-line diagram of the 4160V system from the "A" and "8" bus to the "C1" and "01" bus. Include the diesel generators and major breakers in your drawing. You do not need to show electrical loads.
(2.0)
&HSMER 6.14 See attachment.
Reference (s) 6.14 4160V System Switching Procedure, SP 1107.05, Attachment 1.
-Section 6 Continued on Next Page-l
l t
I l
l Page 19 Davis Besse 1 November 19, 1985 l
Points c
Available ANSWER 6.14 l
MA
-MS g
EA 5 8 Bt 3 fila AC trit 80 NY NY s
i A AC C2 A 60 D1 f
DitSEL Str5 Of tus C1 Olt$tL CEntuntus I SEMER 704 1 Stastowp feeJwater Pump Condensate rump i
=<
Condensate Puey 2 Condensate Fuse 3
-4 Aug. Beller Forced Oraf t f an Coeling Tevet flake **p Fuer i' Copling Tower 11abe up Pump 2 Coeling Tever rue, I Reeter Orates Fusp 1 Caeling Water Fue, 3 Svitthyest feeder 1 liester Drain Puer i Lightlag 5 bate. Trans. DES Switchyard feeder &
Cee.llag Watee Puer !
Lighting Substa. Trans. CTS Ser. Sids. Substa. Tasme. Of6 Ischup Service Watet Puer I Frf t via 03C105 '
l)ACl10l)AClto)AsuCl A AC DI A 91 el A 01 to l
BUS Cl l
SUS On 314th fressere Injecties rump i Nigh Psessere lajettlee Fuey 3 Decay liest Pump i Deesy Nest Pump 3 Ceepenent Coeling Pump i Campvoest feellug Puey 2 f.ervice Water fump I Servies Water Pump 1 Itateep rump I states, Pue, 3 Transfeseet Ctl.1 fremeformer pflet TransfereerCti3 Treaefeveee ef tel Cl feed to Servlee Water rie-p 1, 58 feed to Seselee lfeter Pump 3 Ci reed t. C.e,eee.t C.eii.3 ru-, 2 si r.e4 i. c.e,s.e.t C lles
, s r
1 i
i
-Section 6 Continued on Next Page-I i
l J
Page 20 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.15 Describe the two (2) interlocks associated with the " third" service water and component cooling water pumps.
(2.0)
ANSWER 6.15 The third pump is mechanically interlocked such that only one (1) breaker supply may be closed at one time, therefore prevent-ing parallel operation of the two (2) safety busses. Mdttion-
.allyht-are -interlocked-such -that more than.one.. (1).
component <ooling pump and more than one (1)-service water pump may not be connected to either diesel-generator at-the same-time, c
'e W v: LM s:m 6(
O r, o y w
- MOI ( l<
Reference (s) 6.15 s *p v
LIM, Vol. I, p. II-143.
-Section 6 Continued on Next Page-
- -,,,. - - - - - - - - ~ - -
Page 21 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.16 Draw a top view of the reactor showing the location of the out-of-core neutron detectors labeling each detector and stating its type.
(2.0)
ANSWER 6.16 X
PC CIC (NI 1) g- [;,2. \\
Y,,g (N! 3) r ;k l UCIC
...D tat UCic
- 4
_1 (NI 7)
(NI-5)
O G i-i q
m i sd,W_,
m "g
e e
o
. 4ID UCIC
- ,',[ fhy
,@g, (N16)
UCIC (N N Ni-2)
, Legend PC Proportional Counter Source Range Detector Cic Compensated lon Chamber intermediate Range Detector UCIC Uncompenwied son Chamber-Power Range Detector Referencefs) 6.16 LIM, Vol. 2, p. II-187.
-Section 6 Continted on Next Page-
Page 22 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.17 IRUE or FALSE. The scale on the intermediate range instrument startup rate meter is sufficient to observe the stable negative startup rate occurring after a reactor trip.
(0.5)
ANSWER 6.17 TRUE Reference (si 6.17 LIM, Vol. 2, p. 190.
I l
-Section 6 Continued on Next Page-e
,-w
-..,.. - ~
,,,,---.,..,--,_,.,,,,_.-,-,--,n.--w
9 r
. Page 23 Davis Besse 1 November 19, 1985 Points Available OUESTION 6.18 For the five (5) SFAS incident levels, list the trip setpoints and giye a general [one (1) phrase] description of what action results upon trip of each incident level.
(2.5)
ANSWER.6.18 Lerel Trio Setoofat Action 1
High containr.ent radiation (2 x back-Containment ground at rated thermal power) isolation High containment pressure (18.4 psig)
LowRCSpressure.(1650psig) e' 2
High containment pressure (18.4 psig)
Initiate high pressure Low RCS pressure (1650 psig)
A injection system 3
High containment pressure (18.4 psig)
Initiate low pressure Low low RCS pressure (450 psig) injection system 4
High high containment pressure
'nitiate containment i
(38.4 psi,g) spray and cooling OL system 5
Low low BWST level (2 91.5 in. and Containment emergency 5 100.5) g$
we bcil,(
interlock
/
sump recirculation UW MT Reference (s) 6.18 LIH, Vol. 2, p. II-206.
I h
-End of Section 6-t 6
L V
Page 24 Davis Besse 1 November 19, 1985 Points Available 7.0 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY. AND (25.0)
RADIOLOGICAL CONTROL OUESTION 7.01 The plant is initially at 100% full power. An automatic reduc-tion to 25% indicated full power occurs due to a load rejection.
After low level limits are reached, T continues to be ramped down by the ICSf Exolain why T isgIingloweredandwhat a
mustbedonetoavoidalowprefl6retrip.
(1.6)
ANSWER 7.01 The power range NIs will be reading significantly higher than actual power due to the power decrease causing a hotter Tc and less neutron shielding.
If the plant was initially at 100% of full power, power range NIs could be reading as much as 7 to 10% higher than actual at the end of the power reduction.
This could cause the automatic power reduction to 25% indicated To avoid a low pressure tfY$,after low level limits are reached.
full power to ramp down T the reactor may have to be placed into manual when T,y, begins to drop.
(0.0 -t = ;-r *-'Mic:tM; M.?
nr.ual mi.ur W' "i b'
-VT
. 5-J-o n c t f u G y s, h plQs[ c A0 W A eT L
%l(Y fCry Up m u". ' -/
Reference (s) 7.01
" 6. t' e 4
(( m / ([ n $ g ## #6 u. /
t..
AB 1203.27, Load Rejection, p. 4.
- 5 m,y y
{-
'jm,sv 7 " 4 >h 43
-Section 7 Continued on Next Page-
Page 25 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.02 IRHE or FALSE. An estimated critical rod position need not be perfomed if the actual boron concentration is sufficiently close (i 10 ppm) to the estimated critical boron concentration.
(0.5)
ANSWER 7.02 TRUE Reference (s) 7.02 Approach to Criticality, PP 1103.08, p.3.
QQESTION 7.03 If the actual and the estimated critical control rod positions differ by more than 0.5% Ak/k, then what must be done?
(1.0)
ANSWER 7.03 The regulating rods must be inserted until the problem is analyzed and resolved.
Reference (s) 7.03 Approach to Criticality, PP 103.08.6, p.2.
-Section 7 Continued on Next Page-
Page 26 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.04 A reactor coolant pump is idle, and seal injection to it has been lost. What must be done with controlled bleedoff? Why?
(1.5)
ANSWER 7.04
- G ss.x ( < <km y 2.c0 '
The controlled bleedoff valve should be shut because without s
seal injection and with no auxiliary impeller to force water through the seal cooler, there will be no means to cool the seals.
Reference (s) 7.04 Nuclear Steam System Limits and Precautions PP 1101.01.7, p. 21.
QUESTION 7.05 IRME or hALSE. The component cooling water pump providing RCP-seal cooling and the makeup pump providing RCP seal injection should be aligned to opposite electrical buses.
(0.5)
ANSWER 7.05 TRUE Reference (s) 7.05 NSSS Limits and Precautions, PP 1101.01.7, p. 21.
-Section 7 Continued on Next Page-
Page 27 Davis Besse 1 November 19, 1985 Points Aypfhble 00ESTION 7.06 The RCP Operating Procedure, SP 1103.06, states t?.it operation of RCPs below 400 psig should be minimized. Xid is this necessary?
,/
(1.5)
ANSWER 7.06 To minimize seal wear. Seal wear can be minimized by ensuring at least a minimal pressure per seal (130 psig) and that is attainable at approximately 400 psig.
Reference (sl 7.06 RCP Operating Procedure, SP 1103.05.5, p. 5.
QUESTION 7.07 During tihe June 9,1985 Loss of Feedwater event, the equipment operators had considerable difficulty resetting the trip throttle valves for the No.1 and 2 AFW pumps. What trouble were they having and how was it resolved?
(1.5)
ANSWER 7.07 The equipment operators had only removed the slack in attempting to open the valve. The valve, however, was still closed and the differential pressure on the wedge disk made it difficult to turn the handwheel after the slack was removed. The valve was opened by using a valve wrench.
f Reference (sl 7.07 l
l NUREG-1154, Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9. 1985
-Section 7 Continued on Next Page-
Page 28 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.08 During a plant heatup along the 100 F/per hour heatup line, the Plant Startup Procedure, PP 1102.02, specifies an instantaneous liriitation on heatup rate. What is that heatup rate limit?
(1.0)
ANSWER 7.08 0
1.67 F/ minute.
If the temperature deviates by more than 15 F fromthetemperaturewhichwouldoccuratthatgointintime, assuming the heatup rate was maintained at 1.67 F per minute, theheatupratemustgeadjustedsuchthattemperatureismain-
^
tained within this 15 F limit.
Reference (s1 7.08 Plant Startup, PP 1102.02, p. 24.
QUESTION 7.09 What actions in the secondary plant may be taken if, during a g
reactor plant heatup, the 500 F temperature cannot be attained using 3 RCPs?
(1.0)
ANSWER 7.09 Perform main steam line drains and traps isolation.
Reference (s1 7.09 Plant Startup, PP 1102.02.20, p. 33.
1 i
1
-Section 7 Continued on Next Page-
i
^
Page 29 Davis Besse 1 November 19, 1985 Points Available QUESTION 7.10 a.
What approximate power level on the intermediate range instrument corresponds to the point of adding heat?
(0.5) b.
What value should startup rate (SUR) be below prior to the point of adding heat? (Select one and explain why.) (1.0) 1.
No limit 2.
1 3.
0.5 4
0.1 ANSWER 7.10
-8 and 5 x 10-7 a.
Between 5 x 10 am s.
b.
L'ess than 0.1 DPM to ensure that the RCS does not heat up and surge into the pressurizer faster than the letdown system can reduce RCS inventory.
(Answer 4) e Reference (s) 7.10 Plant Startup, PP 1102.02.21, p. 42.
-Section 7 Continued on Next Page-
Page 30 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.11 A reactor coolant pump must be tripped within what maximum interval if both seal injection and CCW are lost?
(Select one.)
(1.0) a.
s 10 seconds b.
s 30 seconds c.
s 60 secondi d.
5 90 seconds.
ANSWER 7.11 d.
Reference (s) 7.11 Reactor Coolant Pump Operating Procedure, SP 1103.06.9, p. 7.
QUESTION 7.12 What must be done prior to breaching an oil system on an RCP7 (1.0)
ANSWER 7.12 The RCS must be cooled down to less than the flash point of the oil.
Reference (s) 7.12 RCP Operating Procedure, SP 1103.06.11, p. 9.
-Section 7 Continued on Next Page-
Page 31 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.13 If one or more Main Steam Safety Valves continue to relieve after a trip with OTSG pressure at or below 1015 psig, what should be done?
(1.5)
ANSWER 7.13 Lower the TBPV setpoint to lower steam header pressure while paying close attention to SG levels and pressurizer level.
Reference (s) 7.13 Trip Recovery, PP 1102.03.14, p. 7.
OUESTION 7.14 List the six (6) conditions that require implementation of the Emergency Procedure, EP 1202.01.
(2.4)
ANSWER 7.14 Reactor trip SFAS trip SFRCS trip SG tube rupture larger than MU capacity Operator judgement When directed by another procedure.
Reference (s) 7.14 Emergency Procedure, EP 1202.01.0, p. 1.
-Section 7 Continued on Next Page-
1 Page 32 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.15 State the sequence for starting a Makeup Pump.
(1.5)
ANSWER 7.15 1.
Start the AC o'11 pump (DC oil pump will' auto start, run momentarily, and stop).
2.
Verify auxiliary gear oil pump auto starts.
3.
Start the MU pump.
Reference (s) 7.15 Emergency Procedure, EP 1202.01.0, p. 16.
QUESTION 7.16 An %CS 1650 psig trip of SFAS has occurred. The first action called for by the emergency procedure is to verify proper SFAS Incident Level 1 and 2 actuation. State three (3) of the remaining five (5) actions.
(1.5)
ANSWER 7.16 Close RC11, PORV block valve Close RC10, pressurizer spray block valve If subcooling margin is not adequate, trip all RCPs Ensure MVP suction shifts to BWST at 10 in.
Verify HPI is operating IAW rules 1 and 2.
Reference (s) 7.16 Emergency Procedure, EP 1202.01.1, p. 26.
-Section 7 Continued on Next Page-
Page 33 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.17 In the Inadequate Core Cooling Procedure, EP 1202.01, with the RCS in " Region 4" the operator is directed to jumper the start interlocks and start RCPs.
If CCW is not made available to the RCP motors within 30 minutes, the affected RCPs should be tripped. Why is this action necessary?
(1.0)
ANSWER 7.17 To minimize the possibility of a fire in containment in con-junction with high hydrogen in the containment.
Reference (s) 7.17 Emergency Pro,cedure, EP 1202.01.1, p. 123.
QUESTION 7.18 IRME or FALSE. During operation, there is a neutron radiation area in the Main Steam Line rooms in the vicinity of the main steam line penetration.
(0.5)
ANSWER 7.18 TRUE Reference (s) 7.18 Generic Guidance Memorandum, SEP-6, April 12, 1985.
-Section 7 Continued on Next Page-
Page 34 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.19 According to your Radiation Protection Manual HP 1602.01, a pocket dosimeter:
(Selectone.)
(1.0) a.
should be zeroed as often as possible.
b.
should be zeroed at least every time that RACA is entered.
c.
should be zeroed only when a reading of 20% full scale is achieved.
d.
should be zeroed only when a reading of 80% full scale is achieved.
ANSWER 7.19 c.
Reference (s) 7.19 External Personnel Radiation Exposure Monitoring, HP 1602.01 T-8937.
QUESTION 7.20 State your administrative quarterly beta gamma radiation limit for the whole body.
(0.5)
ANSWER 7.20 1250 mrem Reference (s) 7.20 Guides and Limits for Exposure to Radiation, 1601.01, p. 2.
I
-Section 7 Continued on Next Page-1
Page 35 Davis Besse 1 November 19, 1985 Points Available OUESTION 7.21 IRUE or FALSE. The administrative radiation limit for neutron radiation is different than the administrative radiation limit for beta gamma radiation.
(0.5)
ANSWER 7.21 TRUE Reference (si 7.21 1.
Containment Engry HP 1603.00.
2.
Guides and Limits for Exposure to Radiation HP1601.01.
QUESTION 7.22 On a control room evacuation due to a serious control room fire, what one (1) action is the Shift Supervisor or Assistant Shift Supervisor expected to perform and where does he/she go when the control room is evacuated 7 (1.0)
ANSWER 7.22 Obtain the emergency key rings from the Locked Valve Cabinet and report to the fire emergency cabinet outside the control room.
(+0.5foreach)
Reference (si 7.22 AB 1203.26, Serious Control Room Fire, p. 2.
-End of Section 7-r
-r r
Page 36 Davis Besse 1 November 19, 1985 Points Available 8.0 ADMINISTRATIVE PROCEDURES. CONDITIONS AND LIMITATIONS (25.0)
OUESTION 8 Q1 The unit has achieved a condition where the generated load is equal to the load demand; however, control rod motion is needed due to changes in boron concentration and/or Xenon equilibrium changes. Is the plant in " Steady-State Operation"?
(0.5)
ANSWER 8.01 YES Reference (s) 8.01 PP 1102.04.15, Power Operations, p. 24.
QUESTION 8.02' IRUE or FALSE. A steam generator tube leak is considered to be PRESSURE BOUNDARY LEAKAGE.
(0.5)
ANSWER 8.02 FALSE Reference (s) 0 Q2 Technical Specifications, p. 1-3.
-Section 8 Continued on Next Page-
l Page 37 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.03 One Diesel Generator is out-of-service and it would be desirable to take a Decay Heat Pump on the opposite bus out-of-service for maintenance. Would this be permissible by Technical Spect-fications? Exclain.
(1.5)
ANSWER 8.03 No. The Decay Heat Pump on the hus effected by the inoperable Diesel would have to be declared inoperable if the redundant Decay Heat Pump were to be rendered inoperable.
Reference (s) 8.03 Techical Specifications, 3.05, p. 3/4 0-1.
OUESTION 8.04 When using the figures in the Reactor Operators Curve Book (PP 1101.02), for which one of the parameters listed below, is it permissible to interpolate.
(1.0) a.
between different temperatures b.
between " power" and " critical" reference conditions c.
between BOL (Beginning-of-Life) and E0L (End-of-Life) conditions d.
never ANSWER 8.04
- c. only (+1.0)
Reference (s) 8.04 PP 1101.02.7, p. 1.
~
-Section 8 Continued on Next Page-
I Page 38 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.05 According to Technical Specifications there are four (4) re-quirements for continued power operation with a dropped rod.
State two (2) of these requirements. Exact power levels and frequencies are not required for full credit.
(2.0)
ANSWER 8.05 a.
Analyze the potential ejected rod worth b.
Determine Shutdown Margin regularly c.
Verify hot channel factors regularly d.
Reduce power and lower High Flux trip or re-align rods in the operable group with the. inoperable rod.
[Any two. (2) for full credit.]
Reference (s) 8.05 Technical Specifications, 3.1.3.1 c.2., p. 3/4 1-19.
00ESTION 8.06 IRUE or FALSE. There are no Technical Specification " ACTION" items associated with an inoperable axial power shaping rod (APSR).
(0.5)
ANSWER 8.06 FALSE Reference (s) 8.06 Technical Specifications, 3.1.3.2, p. 3/41-21.
-Section 8 Continued on Next Page-
Page 39 Davis Bes 1
November 19, 19 Points
%5 Available 0UESTION 8.07 Should axial power imbalance exceed the Technical Specification limit you must: (Selectone.)
(1.0)
, a.
Restore the axial power imbalance to within its limits within 15 minutes.
b.
Restore the axial power shaping rod group to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, c.
Be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
d.
Be in hot standby within I hour.
ANSWER 8.07 a.
Reference (s) 8.07 Technical Specifications 3.2.1, 3/4 2-1.
-Section 8 Continued on Next Page-
~
Page 40 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.08 Technical Specifications lists seven (7) instruments that are classified as " Remote Shutdown Monitoring Instrumentation."
State four (4) of these instruments.
(2.0)
ANSWER 8.08 1.
Reactor Trip Breaker Indication 2.
Reactor Coolant Temperature-Hot. legs 3.
Reactor Coolant System Pressure 4.
Pressurizer Level 5.
Steam Generator Outlet Steam Pressure 6.
Steam Generator Startup Range Level 7.
Control Rod Position Limit Switches Reference (s1 8.08 Technical Specifications,.p. 3-45.
-Section 8 Continued on Next Page-
Page 41 Davis Besse 1 November 19, 1985 Points Available QUESTION 8.09 During refueling, the Director of Fuel Handling Operations tells you (the Shift Supervisor) that oaerations are going to be performed in the vicinity of the 1-2 hot leg nozzle. He says that both decay heat pumps will need to be stopped. You check Tech Specs and they say that both decay heat pumps may be stopped for up to I hour per 8-hour period. You check the logs and both pumps were stopped from 6:30 am to 6:47 am.
It is now 11:30 am, a.
For how long can you stop the pumps right now?
(0.5) b.
When is the earliest time you you could secure the pumps for an entire hour?
(0.5)
ANSWER 8.09 a.
43 minutes (1 hr - 17 minutes) b.
1:47 pm (1/2 credit for any time afterward)
Reference (s) 8.09 PP 1502.04, Fuel / Control Component Shuffle, p. 5.
-Section 8 Continued on Next Page-
i Page 42 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.10 Which of the following conditions would cause you concern and warrant followup, regarding new fuel shipping containers?
(1.5) a.
The humidity indicator on one container is blue.
b.
One shipping container is stacked on another.
c.
There is an arrow in the shock indicator view port located on the end of one container.
ANSWER 8.10 a.
Normal - pink indicates moisture.
b.
OK - stacking limit is ( 3.
c.
Bad.- NPE or B&W Engineer notified immediately.
Reference (s) 8.10 SP 1506.03, New Fuel Shipping Container Operating Procedure, pp. 2, 20.
00ESION 8.11 List the three (3) che.nicals used at Davis Besse from which a major toxic gas hazard could come.
(1.5)
ANSWER 8.11 1.
ammonium hydroxide 2.
chlorine 3.
hydrazine Reference (s) 8.11 AB 1203.14 Toxic Gas and Liquid Release, p. 4.
-Section 8 Continued on Next Page-
Page 43 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.12 MPCa (40 hr) = 1 E-5 for Xe-133 MPCa (40 hr) = 4 E-8 for Mn-54 The actual concentrations are:
Xe-133 = 0.5 E-5 Mn-54 = 2.0 E-8 Hng long a time at the combined contamination level given above would it be before the 40 MPC restricted nours limit be a
reached?
(1.0)
ANSWER 8.12 CONC CONC ) "I time = (MPCg1 + MPC pl 40 2/
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
=
Reference (s) 8.12 AB 1203.13, High Airborne Activity, p. 3: Equation given in procedure is wrong.
OUESTION 8.13 Wha relieves the shift superviser as Interim Emergency Duty Officer?
(0.5)
ANSWER 8.13 Shift Technical Advisor Reference (s) 8.13 EI 1300.02, Emergency Plan Implementing Procedure, Unusual Event, p. 1.
-Section 8 Continued on Next Page-
Page 44 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.14 Which emergency classification (s) require activation of the Operations Support Center?
(1.0)
ANSWER 8.14 Alert and greater Reference (s) 8.14 EI-1300.06, Operations Support Center Activation, p. 1.
QUESTION 8.15 There are eleven (11) Key Emergency Response Personnel; those Toledo Edison Company individuals who are contacted on the first wave of notifications (following the Plant Manager) after an emergency condition has been declared at the Davis-Besse Station. They are available on a 24-hour per day basis via the Emergency Call System. Llit seven (7) of those personnel.
(1.4)
ANSWER 8.15 1.
Station Operations Manager 2.
Radcon Operations Manager 3.
Emergency Duty Officer 4.
Emergency Planning Supervisor 5.
Emergency Operations Manager 6.
JPIC Manager 7.
Nuclear Security Manager 8.
Technical Engineer 9.
OSC Manager
- 10. Operation Engineer
- 11. TSC Manager Reference (s) 8.15 EI-1300,12, Administrative Controls, p. 3.
-Section 8 Continued on Next Page-
\\
~
Page 45 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.16 The RCS is to be depressurized and opened. The maximum RCS hydrogen concentration to avoid a fi--/ explosion hazard is:
(Selectone.)
(1.0) a.
4 std cc/KG b.
15 std cc/KG c.
55 std cc/KG d.
100 std cc/KG.
ANSWER 8.16 b.
Reference (s) 8.16 Nuclear ' Steam Supply System Limits and Precautions, PP1101.01.8,
- p. 4.
P
-Section 8 Continued on Next Page-
h'gh Page 46 Davis Besse 1 ph November 19, 1985
,@A./
Points A',q3
(
Available 7
OUESTION 8.17
/
Due to the numerous responsibilitter/
assigned to the Shift Supervisorjt the onset of an emergency, the Emergency Plan has set ar/ order of priorit 4the four (4) actions listed below. l'11t those actions to ghest priority first, as noted in the Emergency Plan (2.0) a.
Ensure that immediate notification requirements are met.
b.
Ensure the safe operation of the plant.
c.
Perfom additional emergency actions as time and conditions permit.
d.
Dispatch, in the event of radiological emergencies, RMTs to designated locations onsite.
ANSWER 8.17 b, a, d,'c Reference (s) 8.17 Emergency Plan, p. 6-3.
QUESTION 8.18 Which two (2) individuals are responsible for determining and declaring when an emergency situation is stable and the station is ready to enter the reentry and recovery phase?
(1.0)
ANSWER 8.18 Emergency Duty Officer and Station Operations Manager Reference (s) 8.18 Emergency Plan, p. 9-1.
-Section 8 Continued on Next Page-
Page 47 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.19 During the June 9,1985 Loss of All Feedwater event, several equipment abnormalities / malfunctions occurred that either com-pounded the event or added additional distraction. List seven (7) equipment abnormalities / malfunctions that occurred during that event.
(2.1)
ANSWER 8.19 One main feed pump controller in manual One main feed pump tripped MSIVs went shut (counts for 2)
AFPT No. I and 2 overspeed trip (counts for 2)
Upon reset of SFRCS, AF599 and 608 wouldn't open (counts for 2)
PORY stuck open SPDS was inoperable No. 1 AFW pump suction transferred to service water Source range instrument was inoperable Control l room ventilation tripped into emergency.
Reference (sl 8.19 NUREG-1154, Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9. 1985.
-Section 8 Continaed on Next Page-
Page 48 Davis Besse 1 November 19, 1985 Points Available OUESTION 8.20 Eigt the RCS chemistry limits (both normal and transient, if applicable) for the following:
(1.5) a dissolved oxygen, ppm b
chlorides, ppm c
hydrogen, std cc/Kg H O 2
ANSWER 8.20 a) 1 0.1 normal i 1.0 transient b) s 0.15 normal 5 1.5 transient c) 15 to 40 i
Reference (s) 8.20 i
LIM, Vol. I, p. II-63.
i OUESTION 8.21 IRUE or FALSE. The reactor may not remain critical if steam generator chemistry is out of specification.
(0.5) 1-ANSWER 8.21 FALSE Reference (s) 8.21 Plant Startup, PP 1102.02, p. 38.
End of Section 8-
-End of Examination-
~
......-r~
_,._m..
._-,..m..~.
__,,_ _., ___ _ _ __= -. _..,..., -., _. -
EQUATION SHEET Where mi = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2 2
1 i = PE +KE +P V22 where V = specific KE = mv PE = mgh PE +KE +P V i
i 2
2
~li 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 = (8-p)t t
o o
I p
CR (1-Keffi) = CR (1-Keff2)
CR = S/(1-Keff) delta K = (Kef f-1) 1 2
M = (1-Keffi)
SDM = (1-Keff) x 100%
(1-Keff2)
K eff 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 3
1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /hr 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 centimeters 2
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec v