ML20141N972
| ML20141N972 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/24/1986 |
| From: | Elin J, Johnston G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20141N970 | List: |
| References | |
| 50-312-OL-86-01, 50-312-OL-86-1, NUDOCS 8603180372 | |
| Download: ML20141N972 (119) | |
Text
f U. S. NUCLEAR REGULATORY COMMISSION REGION V Report No.: 50-312/0L-86-01 Facility Name: Rancho Seco Nuclear Generating Station Docket No.: 50-312 Examinations Administered at: Rancho Seco NGS, Clay Station, California, from February 4-6, 1986 Chief Examiner: .
R 1- 2k'N W. Johnston, Operator Licensing Examiner Date Signed Approved by: -
2-2i-E
/ Elin, Chief, Operations Section Date Signed Summary:
Examinations on February 4-6, 1986 Written examinations were administered to two Reactor Operator candidates and six Senior Operator candidates. Two Reactor Operator candidates passed the written examinations. One Reactor Operator candidate received a waiver of the operating examination and one Reactor Operator candidate passed the operating examination. Six Senior Reactor Operator candidates passed the written examination. Five Senior Reactor Operator candidates passed the operating examination, while one Senior Reactor Operator candidate failed the operating examination.
8603180372 860303 PDR V ADOCK 05000312 PDR
I REPORT DETAILS
- 1. Persons Examined There were two candidates for Operator License examinations, three candidates for a Senior Operator Upgrade License examination, and three candidates for a Senior Operator Instant License examination.
- 2. Examiners
- G. Johnston, RV W. Apley, PNL
- Chief Examiner
- 3. Persons Attending the Exit Meeting NRC G. Johnston, RV W. Appley, PNL SMUD G. Coward, SMUD, Plant Manager P. Turner, SMUD, Training Manager T. Hunter, SMUD, Training Coordinator J. Mau, SMUD, Training Superinte; dent - OPS W. Spencer, SMUD, Nuclear Operation Superintendent H. Canter, SMUD, QA Operations Surveillance Supervisor
- 4. Written Examination and Facility Review Written examinations were administered as follows:
2 RO exams - February 4,1986 6 SRO exams - February 4, 1986 At the conclusion of the exam, the facility staff was given copica of the two examinations. They were instructed by the Chief Examiner to review the exams and provide the examiners with the comments either prior to their departure or within five days af ter the completion of the examinations. This is to facilitate resolving their comments prior to grading the examinations.
- 5. Operating Examinations Oral exams and facility walkthroughs were conducted February 5-6, 1986.
One general weakness was identified by the examiners. This weakness was in identifying the person to contact if the Plant Manager was not available when notification was required due to an Unusual Event. All but one of the candidates were indicated as clear passes during the exit meeting.
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- 6. Written Examination ,
A concern arose from the' review of the results of the written
~ examinations'for the Senior Operator candidates in the area of the logic associated with the Safety Features. Actuation System. None of the candidates demonstrated a clear' knowledge of the logic associated with the Reactor Building Spray actuation. This system has a unique and obvious design feature providing diversity to prevent inadvertent spray actuation by providing separate bistables in each channel for starting the pumps and for opening the valves. The candidates did evidence knowledge of the five minute time-delay relay, which when an actual actuation occurs provides sufficient, time to prevent an inadvertent spray down of the Reactor Building.
- 7. Exit Meeting The Chief Examiner met with the facility representatives denoted in paragraph 3 above to discuss the examination process and to indicate those candidates who were clear passes of the Operating Examination. The area identified as a weakness was discussed as well as other topics pertinent to the examination process.
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Resolution of Facility Comments Reactor Operators Examination Question 1.09:
" Normal core Delta T at 100 percent is 48 F. The question shows 43 Delta T for 80 percent. If intent is to see if the candidate can utilize (Qdot) =
(Mdot) (Cp) (Delta T), answer (c.) should or could be acceptable for at least partial credit if any mark is shown."
Resolution:
The examiner sees no reason to change key. Partial credit cannot be given in this case. The information is sufficient to calculate the mass flow rate.
Question 1.12 b and c:
" Candidates may respond that the reactor is slightly super critical. Also, actual plant values show that 200 ppm boron is approximately 2 percent Delta K/K not 10 percent Delta K/K. Therefore, if a statement of this sort is made, no penalty should be assessed."
Resolution:
For part 'c' it is true that the Keff will likely be slightly over unity. But for all intent will be as the key says 'at or near critical'. Part 'b' was inadvertently left in the exam and was dropped after the section was graded.
Question 2.08 b.:
"By procedure A.15 section 7.2 step 10 addresses adding nitrogen to back flush to ensure capacity for blow down. The back flush tank can be blown down to the Radwaste Crud Tank with nitrogen."
Resolution:
Accepted, key changed.
Question 2.11:
" Add Pressurizer Relief Tank (PRT) cooling coils as a possible response."
Resolution:
Accepted, key changed.
Question 3.01:
"RPS cabinet 'A' has a jack which allows selection of the pressure signal to come from A or B RPS which are powered from Vital A and Vital B inverters.
The answer key is acceptable if the jack is selected for ' A' . If the candidates assume the jack is selected for 'B' and no effect, the answer should be considered as correct."
Resolution:
The examiner agrees that the case may exist where a candidate might make such a reference. If so, the examiner will apportion credit according to the response.
Question 3.03:
" Rod shadowing may not be addressed by candidate due to all the rods out
! operation."
Resolution:
The examiner agrees and will drop the response for Rod Shadowing from the key and the point associated.
Question 3.06:
"More responses than those on the key may be given by the candidates.
Procedure B.9 section 3.2 steps .1 through .6 address interlocks during startup. Section 3.3 steps .1 through .6 address interlocks greater than 15%."
Resolution:
The examiner considers the stated interlocks are most obvious, but will, if the candidate indicates other interlocks, consider them on a case by case basis.
Question 3.10 a:
" Answer key should read ' Indication goes to bottom of the scale' (due to no resistance)."
Resolution:
The examiner agrecs, key will be changed.
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Question 4.02:
" Answer for second part of Rule 6 should state RCS temperature less than 500'F, RCP's off,IIPI flow."
Resolution:
Accepted, key will be changed.
Question 4.05:
" Obscure wording of ' potential problems' may confuse the candidate.
Procedure B.3 pre-empts the unplanned power return at nominal rates with the statement of power maneuvering not being normally done. Rather than potential problems, the three ' disadvantages' are referred to as a cost. A 15 minute hold at 50 percent is minimal and would create little problem. Also, enclosure 10.5 coufuses the issue as there is no pertinence to the question.
This may mislead the candidate."
Resolution:
The examiner agrees that the question addresses nominal rates. The enclosure indicates the limitations of using rods, and implins that boration and dilution are the chief methods used to maneuver. The examiner sees no reason to change key on the basis that he views the question as being a poss;ble case and would expect the candidate to be able to respond.
Question 4.10:
"We do not hold Control Room Operators (COs) responsible for bases of Technical Specifications."
Resolution:
The examiner feels that TS 3.1.3 " Minimum Conditions for Criticality' is particularly important and this does meet the intent of the Examiners Standard ES-202, ' Scope of Written Examinations Administered to Reactor Operators -
Power Reactors'. The standard states in part "... Technical Specifications, may be included to the extent they are directly applicable to an operator and the safe operation of the facility." This requires judgement on the part of an examiner, and in this case he feels that the question is appropriate.
Question 4.13 c: i l
"Part (c.) of this question is in limits and precautions of A.64. This
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is a I very obscure limit and precaution and specifically limits to 10 minutes cumulative operation as resonance in the blading may occur."
Resolution:
It does not appear to be obscure to the examiner, and is certainly required knowledge for the operators. No change to key.
F Resolution of Facility Comments Senior Reactor Operators Examination Question 5.03.b:
" Key requires discussion of fuel rod rupture, however, limit has a direct basis of center line fuel melt. The ' failure mechanism' asked about in the question doesn't specity cladding."
Resolution:
The examiner feels that the candidates should be fully aware of a basic Safety Limit. Including what concern arises from exceeding the limit. It is true that fuel center line melting is the basis of the limit, but that basis is to prevent potential rupture of the clad. The mention of ' clad failure' in the question provides a strong lead that the examiner felt was unnecessary. No change to the key.
Question 5.04 a and b:
" Key requires method to be shown but question does not."
Resolution:
There are only two possible ways on the basis of the information given to determine Keff. One is by the method shown in the key, and the key also indicates that if the candidate used the thumb rule for doubling of counts, that also would be acceptable for full credit. For part b similar reasons exist, there is only one way on the basis of the information given to calculate the reactivity addition. No change to key.
Question 5.04 c:
" Key is incorrec'. Keff will be greater than 1 (1.0058844)."
Resolution:
It is true that the Keff will likely be slightly over unity. But for all intent will be as the key says 'at or near critical' . No change to key.
Question 5.06 b:
" Acceptable answer should also be ' water activation product' ." l l
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Accepted. Key will be changed.
I Question 6.02 a:
" Key is incorrect. Charger with hi8hest initial voltage will assume majority of load."
Resolution: ,
Systems Training Manual, ' Vital Electrical Distribution 4160V and Below',
page 43-80 states. in part '.'The standby battery charger will assume more of the load, sin'ce it has a higher capacity." The reviewer did not include a reference for the statement, therefore the key will not be changed.
Question 6.02 b:
"The system procedure (A.61) provides for temporary paralleling of chargers and does not address adjusting voltage in parallel."
Resolution:
The key will be changed to account for a normal voltage adjustment.
Question 6.03:
"If pressure signal to heaters, EMOV, and spray valve is selected to 'B' RPS there will be no effect."
Resolution:
The examiner agrees that the case may exist where a candidate might make such a reference. If so, the examiner will apportion credit according to the response.
Question 6.06:
" Question requires students to be familiar with other plants in order to recognize the uniqueness of B&W relative to them."
Resolution:
The portion pertaining to ' Rod Shadowing' was dropped from the key. The examiner does feel however, that the effect of cold leg temperature on the nuclear instrumentation is relevant. The point associated with ' Rod Shadowing was also dropped from the examination.
Question 7.02:
"Second criteria should include RCP's off."
Resolution:
Agreed. Will change key.
F Question 7.03:
"' Lack of Heat Transfer' and ' Excessive Heat Transfer' have essentially the same priority so they could be listed vice versa."
Resolution:
No change to key. The question apportions points on the basis of (0.5) points for each item, and (0.5) for highest and lowest.
Question 7.06:
" Answer requires memorization of a non-emergency procedure."
Resolution:
There is no memorization required of the procedure, however, Examiner Standards NUREG-1021, ES-402 " Scope of Written Examination Administered to Senior Reactor Operators - Power Reactors" states in part "...The candidate should be able to describe generally the objectives and methods used in the normnal, of f-normal, and emergency procedures . . . . " Further ES-202 for the Reactor Operator candidates states that "... candidate is not expected to have normal procedures committed to memory but should be able to explain reasons, cautions, and limitations of normal operating procedures." The examiner therefore sees no need to change the key.
Question 7.07:
"15 minute time frame is not realistic to see these problems."
"Also, question implies the attached ' Enclosure 10.5' is to be used; however, it has nothing to do with the question."
Resolution:
The 15 minutes is for the period at 50 percent power. And the question addresses nominal rates. The enclosure indicates the limitations of using rods, and implies that boration and dilution are the chief methods used to maneuver. The examiner sees no reason to change key.
Question 7.08 c:
" Answer should allow student to have an HP person in fire brigade also.
Present Special Order has removed HP person, however, this S.O. was only issued Monday."
Resolution:
The examiner will accept as one of the persons on the brigade an HP technician.
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" Rule 3 also states that 'if cooldown required, then throttle AFW to limit cooldown rate to < or = 100 F/ Hour."
Resolution:
The examiner is only seeking what is in the key, but will accept the above as partial credit. The question pertains principally to natural circulation, it is highly unlikely that the cooldown limits will be exceeded.
Question 7.09 b:
'E.05, Excessive fleat Transfer also gives direction to stop 1.ny FW pumps if OTSG level is > or = 95 percent."
Resolution:
The examiner fails to see any connection with the comment and the question.
No change to key.
Question 8.02:
"This one question is 16 percent of the category value."
Resolution:
True, the question however is broke down into two parts each of which is two points and further broken down in the key. The examiner felt that the point value assigned is correct for the type of question. Question values are customarily limited to less than 20 percent of category value.
Question 8.04 a:
"---Affect--- should be: directly af fect the reactivity or power level of the rea c to r . "
Resolution:
Accepted.
Question 8.07 a:
" Answer #1 is only basis for 525 F. No mention of consistency with FSAR analysis. Answer #3 is a Spec in itself - not the bases for 525 F."
Resolution:
The bases for any Technical Specification Limiting Condition for Operation is consistency with analysis with the FSAR. Answer #3 will remain an answer in the key, to be added however, by review of the examiner is the temperature requirement assures that the saturation pressure does not go below 885 psia limiting the moderator pressure coefficient reactivity addition, per the bases of the TS 3.1.3.3.
Question 8.07 b:
"The answer only addresses the upper icvel. Ten inches level is based on the level being above minimum detectable."
Resolution:
Will add the minimum detectable icvel requirement and divide points, with (0.5) for each response.
Question 8.09 b:
"Also, the limit is 60 percent for the allowed, for the pump combination; i.e., if only three pumps are running the ICS will only reduce to 60 percent and the limit is 45 percent."
Resolution:
The candidate only has to relate the facts and not the numerical values.
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U.S. Nuclear Regulatory Commission Reactor Operator License Examination qQ y
Facility: RANCHO SECO Reactor Type: BABCOCK AND WILCOX ,
Date Administered: FEBRUARY 3. 1986
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Examiner: CARY JOHNSTON '
Candidate:
INSTRUCTIONS TO CANDIDATE: '
Use separate paper for the answers. Write answers on c.e side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The' passing grade requires at least 705 in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. ,
% of Category X of Candidate's Category ;
Value Total Score Value Category 25.0 25.1 1. Principles of Nuclear Power Plant Operation, Thermodynamics, :
Heat Transfer and Fluid Flow 2
25.0 25.t
- 2. Plant Design Including Safety and Emergency Systems .
.2f.o 2Y. 2 .
N N 3. Instruments and Controls 25.0 25.E 4. Procedures - Normal, Abnormal.
Emergency, and Radiological Control W.o
.Me-* .
TOTALS i
Final Grade X All work done on this examination is my own. I have neither given nor received aid.
Candidate's 51gnature 1
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EQUATION SHEET i
f = ma v = s/t w = ag 2 Cycle efficiency = Net k t) a = v,t + at E = aC a = (vg - y )/t ,
2 KE = mv vg = v, + a A = AN A = A,e C i PE = agh a = e/t A = In 2/t g = 0.693/tg W=vaP'
, (t )(t g) i AE-= 931Am (t + )
=[nCAT p I = I,e -I*
Q = UAAT ~M*
I=Ie Pwr = W g It
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I = I, 10 *!
P=P 10 SUR(t) TVL = 1.3/u p,p o ,t/T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) eff SUR = 26 CR = S/(1 - K,gfx)
(8-pj x T = (1,/p ) + [(8 - p)/Aegf] p CR y (1 - K,gg)g = CR (1 - K,gg)j 2
T = 1*/ 6 - M N " lf(l ~ Keff) = CR /CR g 0 I*I ~ D)! eff D M = (1 - K,gg)0 (1 - K,gg)g p = (K,gg-1)/K,gg = AK,gg/Keff SDM = (1 - K,gg)/K,gg a= [1*/Titygg ] + [I/(1 + A,ggT )]
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1* = 1 x 10 ' seconds P = E6V/(3 x 10 0) A,gg = 0.1 seconds
-I E = No Idgy=I22 l WATER PARAMETERS Id =Id32 g
I gal. = 8.345 lba (!
R/hr = (0.5 CE)/d (meters) i 1 gal. = 3.78 liters 3
R/hr = 6 CE/d (feet) ,
i ft = 7.48 gal.
MISCELLANEOUS CONVERSIONS Density = 62.4 lbm/ft 1 Curie = 3.7 x 1010 dps l Density = 1 ga/cm 1 kg = 2.21 lbm Heat of var orizations = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr i Heat of fusica = 144 Btu /lba 6 1 N = 3.41 x 10 Btu /hr I 1 Atm = 14.7 psi = 29.9 in. I's- 1 Btu = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in 1 inch = 2.54 cm '
F = 9/5 C + 32
- C = 5/9 (*F - 32)
SECTION 1 6
EBING1ELEE QE HQGLE8B EQWEB ELONI QEEBBI1QNs IHEBdQQ1Ned1GEi HEGI IBONEEEB GNQ Ebu1Q ELQWu 1.01 QUESTION (1.0)
The Xenon peak that occurs after a reactor trip from 100%
equilibrium xenon condition is greater than the peak for a trip from 50% power due to
- a. The fission yield for Xenon is higher at 100% power.
- b. There is more Iodine in the core at the time of a trip from 100% power.
- c. There are more thermal neutrons in the core at 100%
power.
- d. There are more delayed neutrons in the core at 100%
power.
1.01 ANSWER (1.0) b reference:
RT 16.3 - 16.7 i
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l 1.02 QUESTION (1.5) -
6 If reacto. power increases from 1000 cps to 5000 cps in 30 seconds, what is the Startup Rate (SUR) ?
1.02 Answer (1.5)
P = Po (10E:SUR(t)) (0.5) 5000 = 1000 (10E:EUR(.5))
SUR = (LOG 5)/(0.5) = (.699)/(.5) = 154 qed (1.0) l References t RT 10.1 - 10.4 l i i
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O 1.03 QUESTIDN (1.0) 6 Which of the following six factor formula terms increases on a power escalation to allow reactor power to match turbine power? (1.0)
- a. The fast fission factor.
- b. The thermal utilization factor.
- c. The reproduction factor.
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a g 1.04 QUESTION (1.0)
The most serious problem with reaching the critical heat flux (CHF) in a PWR reactor iu caused by: (1.0)
- a. the poor therreal conductivity of steam.
l b. the blockage of flow through the core when steam bubble formation becomes significant.
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- c. the displacement of boron from the core as steam bubble formation becomes significant.
- d. the high pressure surges in the reactor coolant ,
systems cau1 sed by steam bubble formation. ;
1.04 Answer (1.0)
- a. (notes talk to steam blanketing, clad burnout)
References HTFF Noten CH.4 Part e p.220-230 i
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1.05 GUESTION (1.0)
The Fuel Temperature coefficent increases (becomes more ,
negative) from BOL to EOL primarly due to (1.0)
- a. the reduction of fuel to clad gap distance.
- b. the reduction in the moderator's baron concentration.
- c. the increase in Pu-240 in the core.
- d. the increase in thermal neutron flux.
1.05 Answer (1.0) .
c
Reference:
RT-13.3 1
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s 1.06 QUESTION (1.O)
The reactivity worth of a control rod increases (1.0)
- a. as Tave increases from 150 degrees F to 500 degrees F.
- b. as reactor power is reduced from 100% to 50%.
- c. as a result of fission product buildup.
- d. when the soluble baron concentration increases.
1.06 Answer (1.0) a
Reference:
RT 14.1 -
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= 1.07 QUESTION (1.O)
Overall plant efficency will decrease if cire water toeporture increases. TRUE or FALSE 7 1.07 Answer (1.0)
True
Reference:
GP HT&FF notes I
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, 1.08 QUESTION (1.0)
Overall plant efficency will decrease if one train of Main Steam Reheat is isolated. TRUE or FALSE 7 1.08 Answer (1.0)
False
Reference:
SD CH.14 - Main Steam Sys.
and GP HT&FF notes Sect II Part B j
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, 1.09 QUESTION (1.0)
The reactor is at 80% power with a core delta T of 48 degrees F and a mass flow rate of 100%; when a STATION BLACKOUT occurs, Natural circulation is established and core delta T goes to 40 degrees F. If decay heat is assumed to be approximately 2% of full power, what is the mass flow rate (% of full flow) 7 (1.0)
- a. 1.9 %
- b. 2.1 %
- c. 2.4 %
- d. 3.0 %
1.09 Answer (1.0) d reference: GP HTE<FF Sect II part B l
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. 1.10 QUESTION (2.0)
Regarding the feedwater heaters:
- a. If the level in a feedwater heater is allowed to increase, what will happen to the temperature of the feedwater exiting the heater, increase, remain the same, or decrease? (1.0)
Why? (1.0) 1.10 Answer (2.0)
- a. Decrease 11 tQL. The level increase exposes less surface area in the heater tubes 1Qtgl. There:
is less latent hest removed from the steam, and therefore less heat 1transfered to the feedwater resulting in a lower exit temperature for the feedwater 19tgl.
Reference:
General Electric " Thermodynamics, Heat Transfer, and Fluid Flow!', Chapter 8.
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. 1.11 QUESTION (3.0)
Listed below are four factors that can affect Departure from Nucleate Boiling Ratio (DNBR). What will the effect on DNBR (increase or decrease) if the factors listed below are lagtggggd7
- a. Reactor Power. (0.75)
- b. Reactor Coolant System flow. (0.75)
- c. Reactor Coolant System temperature (average). (0.75)
! d. Reactor Coolant System pressure. (0.75) 1.11 Answer (3.0)
- a. DNBR decreases. (0.75)
- b. DNBR increases. (0.75)
- c. DNBR decreases. (0.75)
- d. DNBR increases. (0.75)
Reference:
Facility Manual, ' Heat Transfer and Thermodynamics',
pages 122 to 126, and 243 to 252.
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1.12 QUESTION (3.0)
During preparation for a normal startup, after calculating for an Estimated Critical Position it is determined that a 200 ppm boron dilution is necessitated. Before dilution begins the indication for the Source Range instrumentation shows one channel at 35 cps and the other at 40 cps. The dilution is started and after 100 ppm of dilution the count rates are 70 cps and 80 cps respectively.
Based on the above indications and assuming an initial Keff of 0.90 :
- a. What would Keff be after this 100 ppm dilution? (1.0)
- b. What reactivity was added by this 100 ppm (1.5) dilution?
- c. What would you expect Keff to be after the full (0.5) 200 ppm dilution?
1.12 Answer (3.0)
- a. CR1 (1-Keff1) = CR2 (1-Keff2) (0.5) 40 (1 - 0.9) = 00 (1 - Keff2)
Keff2 = 0.95 after 100 ppm dilution (0.5)
(Relation of doubling rule of thumb acceptable.)
- b. Rhoo = (Keffo - 1) / Keff0 =
= (0.9 - 1) / O.9 = -0.1111 (0.5)
Rhoi = (Keffi - A) / Koffi =
= (0.95 - 1) / ).95 = -0.05263 (0.5)
Reactivity added = RhoO - Rhoi = +0.05848 (0.5)
- c. The reactor will be at or near critical. (Keff (0.5)
= 1.0).
Reference:
Facility Manual, ' Reactor Theory', pages RT-6.2 to RT-8.11.
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. 1.13 QUESTION (2.5)
What would the most likely source (origin) be for the following isotopen if they were found in a Reactor Coolant Syst.em water sample 7
- a. Cobalt 60 (0.5)
- b. Nitrogen 16 (0.5)
- c. Argon 41 (0.5)
- d. Strontium 90 (0.5)
- e. Rubidium 08 (0.5) 1.13 Answer (2.5)
- a. Activated corrosion product. (0.5)
- b. Activation of Oxygen 16. (0.5)
- d. fission product. (0,5)
- e. Daughter product of Krypton 88 (fissicn (0.5) product).
Reference Glasstone and Sessonske, ' Nuclear Engineering'.
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l 1.14 QUESTION (1.5)
Three agents are typically added at various times to control R6 actor Coolant System chemistry. What agents are added to to the RCS to control the followings
- c. pH control (0.5) 1.14 Answer (1.5)
- c. Lithium Hyroxide - ( pH control ) (0.5)
Reference:
Systems Training Manual, Chapter 6, ' Reactor Coolant Chemical Addition and Sampling System *, page 6-2.
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, 1.15 . QUESTION (2.5)
There are two ways to describe flux distribution variations that can occur in the reactor core. These are typified as flux tilts.
- a. What is Axial flux tilt? (0.5)
- b. What is Radial flux tilt? (0.5)
- c. What are three possible conditions / events that (1.5) could cause a flux tilt to occur?
1.15 Answer (2.5)
- a. Variation of flun along the vertical axis (0.5)
(power in the top - power in the bottom).
- b. Uneven flun distibution across a hori: ental (0.5) plane of the core (or description of Guadrant Power Tilt formula).
- c. Any three JQ29) each.
- 1. Mispositioned rod.
- 2. Xenon oscillation.
- 3. Improper fuel load.
- 4. Flow imbalance, (3 RCP's etc.)
References Systems Training Manual, ' Nuclear Instrumentation *.
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, 1.16 QUESTION (1.0)
List three conditions that could cause the pressurizer level to remain constant while pressurizer pressure decreases. (1.0) i 1.16 Answer (1.0)
- a. Small steam space leak.
- b. lons of pressurizer heaters.
- c. partial or complete activation of pressuri:er coray.
reference: RS exam bank End of Section 1 16
i SECTION 2 l
Plant Design Including Safety and Emergency Systems 2.01 (2.0)
, Regarding an tmgrggnty start of the Emergency Diesel Generators:
- a. If the speed governors of the Emergency Diesel (2.0) i Generators were 103dyEC12D11y left in the scggg
! ggd3 of operation, what potential problems may l occur for the ggulament that tries to start on
- the 4160 volt safety related busses?
2.01 Answers-l
- a. The equipment breakers tripping from (1.0) l overcurrent or undervoltage.
Or, the equipment may not perform their safety (1.0) related functions / design objectives.
Reference:
Facility Systems Descriptions Chapter 45, ' Emergency Diesel Generators *. ;
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2.02 (3.0)
-Regarding the 125 volt DC systems
- a. When the standby and normal battery'. chargers (1.0) are aligned for parallel operation which unit will assume the highest load?
- b. When the chargers are operating in parallel how (1.0) is the voltage adjusted if needed?
- c. If AC power is lost for an extended period of (1.0) time, how long will the 125 volt.DC batteries i
carry their loads?
2.02 Answer:
I
- a. The standby unit. (Has highest capacity.) (1.0)
- b. By using the rheostats on the normal charger. (1.0)
I
- c. For two hours. (1.0)
References Systems Training Manual, ' Vital Electrical Distribution 4160 volt and Below', pages <
43-9 to 43-18, and 43-79, 43-80.
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2.03 (3.0)
Regarding the Reactor Coolant Pumps: f What is the purpose of the following RCP start interlocks?
I '
- a. Reactor pcwer must be less than 30% of full (1.0) power.
- b. Component Cooling Water flow must be greater (1.0) l than"307 gallons per minute.
- c. Reactor Coolant System temperature must be (1.0) greater than 500 deg. F. l 2.03 Answers
- a. Reduces the potential of a positive reactivity (1.0) insertion from cold water in an idle loop.
- b. Insures adequate flow of cooling water to seals (1.0) !
and motor.
[
[ c. Precludes core lift phenomena by insuring delta (1.0)
P across core is not sufficient to cause lift -
of fuel elements to occur.
Reference:
System Training Manual, ' Reactor Coolant System *,
pages 2-75 to 2-78.
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l 2.04 (1.0) l For the following core components indicate the composition i of the neutron absorbing materials
- a. Burnable poison rod. (0.5) l
- b. Control rod. (0.5) i 2.04 Answers
! a. Alumina with baron carbide. (0.5)
- b. Silver, Indium, Cadmium. (0.5) f References Systems Training Manual, ' Fuel Assembly and Control l
Components'.
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5
?
2.05 (1.0) t Which of the following parameters will not cause a trip of j
> the main turbine if the critical value is exceeded? ;
- a. High exhaust hood temperature. ,
r
- b. Low bearing oil pressure. ;
l- c. Excessive thrust bearing wear.
t
- d. Low hydraulic fluid pressure.
- ' 2.05 Answers I
- a. High exhaust hood temperature. (1.0) i
- References Systems Training Manual, ' Main Turbine', pages
+
166-116 to 16b-122.
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2.06 (2.5)
Regarding the Reactor Building Isolation Systems
- a. In the event of a failure of power or loss of (1.0) control air, what position will the valve assumn for the following?
- 1. Air operated isolation valve.
- 2. Motor operated valve.
- b. In general where would the valves, described (1.0) above, be located in relationship to the containment penetration (i . e. inboard or outboard) if they were on the same line?
- c. At e simple check valves used for outboard (0.5) automatic isolation valves?
2.06 Answers a.
- 1. Will fail closed. 50.5)
- 2. Will fail "as is". (0.5) b.
- 1. Outboard the penetration, (or if the motor (0.5) operated valve is outboard of the penetration it will be outboard the motor operated valve.)
- 2. Inboard the penetration, (or between the (0.5) penetration and the outboard air operated valve.)
- c. No. (0.5)
References Systems Training Manual, ' Reactor Building and Support Systems', pages 4a-44 to 4a-47.
6
e 2.07 (3.0)
Regarding the Emergency Core Cooling Systems
- a. What are the two conditions that would require (2.0) the use of the " Piggy-back" made following a l LOCA7 L
- b. What is the primary reason that necessitates (1.0) [
the use of the " Piggy-back" mode rather so;ne ,
other lineup? l r 2.07 Answers
- a. Whenever HPI is maintaining RCS inventory and a BWST low-low-level alarm is received J1291 Or to establish bcron dilution flow during long
- l term cooldown 11a.QL.
l l b. The HPI pumps cannot take a suction directly (1.0) from the Reactor Building emergency sump.
i l
Reference:
Systems Training Manual, ' Emergency Core Cooling System', page 27-17.
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l 2.08 (2.0) l Regarding the Makeup and Purification Systemt
- a. The outlets of the letdown coolers combine into (1.0) j a single header referred to as the letdown delay line. What is the design purpose of this line?
- b. After attempting to backflush the letdown (1.0) filter and not achieving satisfactory resultu, what action can be taken to facilitate an adequate flush of the f11ter?
2.08 Answers
- a. Allow the decay of Nitrogen 16 in the coolant. (1.0)
- b. Use of the compression release device to (1.0) relieve the compressive force on the disk l stack.
References Systems Training Manual, ' Makeup and Purification',
pages 5-8 and 5-26.
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l 2.09 (2.0)
The 'D' High Pressure Injection line is configured differently than the other three injection lines.
l
- a. This difference includes an extra motor (2.0) f operated valve. What function does this valve provide.
2.09 Answer
- a. Isolates this particular line 11tQL to provide HP! water to the pressuizer spray line as an l alternate source of water for pressuri=er spray 31A91 References Systems Training Manual, ' Emergency Core Cooling System', page 27-7.
9 i
2.10 (3.0)
Regarding the Main Generators
- a. What concern arinen if overencitation is (1.0) occuring in the main generator?
- b. What would an operator observo happoning to the following instrumentation if a large inductive i load was suddenly added to the nystem?
- 1. Power Factor. (0.5)
- 2. VARs (Volt-Ampu-Reactance). (0.5)
- 3. Generator Current. (0.5)
- 4. Generator load. (0.5) 2.10 Answurs
- a. Overencitation loads to enconnive current flow (1.0) in the gonorator stator and (the main trannformor). (Or stator high temperature.)
b.
I 1. Decomon moro lagging. (0.5)
- 2. Incruanos. (0.5)
- 3. Increases. (0.5)
- 4. No changu. (0,5)
Ruferences Syntems Training Manual, ' Main Generator', pages 40a-24 and 40b-44.
10
l l i
e l l i 2.11 . (1.0) .
.* l While the plant is at 100% power an alcem is received >
- indicating a high level condition exists in the Component [
l Cooling Water surge tank for train 'A*. Indication from the '
i radiation recorder reveals a steady increase in radiation levels for train 'A' of CCW.
l a. What are two possible sources for this apparent (1.0) i' j inleakage to the train of CCW7
~ ,
l l
2.11 Answers i a. Any two 11gQL
! I l 1. Letdown Coolers l
- 2. Reactor Coolant pump thermal shields. !
- 3. RCS sample cooler.
- 4. B.A. eva orators, and waste evaporators. I f", PQT Coe MIlb o i References Systems Training Manual, ' Component Cooling Water !
System', figure 7-1-22. i i
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2.12 (1.5)
Regarding the Auxiliary Feedwater pump P-310s
- a. Could the pump be operated if, after an (1.0) overspeed trip, the overspeed trip mechanism is not reset? EXPLAIN.
- b. Can the overspeed trip mechanism be reset from (0.5) the control room 7 0.12 Answers
- a. Yes 1QtQL. The steam admission valve is normally closed to operate the pump on the electric motor. The closure of the valve does not affect the operation of the motor IQtQL.
- b. No JQ3Dl.
References Systems Training Manual, ' Auxiliary Feodwater System',
pages 29-17 to 29-22.
End of Section 2 12
I SECTION 3 Instruments and Controls 3.01 (3.0)
Wugt g(( ggt will a failure of power to the Reactor ,
Protection System channel A pressure instruments have on the i following equipment andERD manual control of the function be
- taken in the control room 7
- s. Pressurizer heaters. (1.0)
I b. Spray valves. (1.0)
- c. EMOV. (1.0) 3.01 Answere
- a. Hesters will energize. Manual control by the (1.0) handswitch is availiable. ,
- b. The spray valves will fail closed. Manual (1.0) ,
control by the handswitch is availiable.
- c. The EMOV will fail cloned. Manual control by (1.0) l the handewitch is availietble.
1 l
Reference:
Systems Training Manual,
- Vital Electrical
! Distribution 4160 volt and Below ' , pages ,
l 43-02 anti 40-O s. -
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3.02 (4.0)
Using the enclosed diagram of the Integrated Control System explain the actions that would occur in the following control sections if an Assymetric Red condition occurred at 90% of full power. Include the actions that the control sections teke with respect to components and other interactions.
- a. Feedwater Control. (2.0)
! b. Reactor Control. (2.0) 3.02 Answers
- a. The feedwater demand c.ti cul a tor will be receiving a decreasing nignal when the l Assymetric Rod runbeck occurs 1QoDL. This will l cend a reduced domand signal to the feedwatur l demand units 1QtQL, which will sold a signal to the Feedwater Regulattng valves 1QtGL and the Feedwater Fump speed cnntrollars 192Dl.
- b. The Reactor Control section recieves the name decreasing nignal IQoQL. From the comparison bet +.een this signal and the Neutron Error the Caector Demand differenco unit will cond a signal that will call for the inmortion of rods J12Ol. The (rod bito at 90*/. may be weak) and a Reactor Cronslinit may occur limiting Feedwater flow 394D1 i Ref erences Systems Training Manual , ' Integrated Control Gystum',
l pages 32-77 to 32-93.
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Regarding the Excore Fower Range Nuclear Instrumentation: i
- a. Due to the design of the Dabco:k and Wilcox (2.0) t plant there are two plant parameters that affect the Encore Power Range calibration. What are those parametern and how do they a.ffect the detectors?
l 3.03 Answers !
- a. Rod shadewing - The Bt<W design has a more l l " rodded
- core, meanin0 the ragulating reds are f
- inserted much more than in other designs 1Qugl, l This would tend to 'chadow' the upper detectors !
l from seeing a representative flux in the ccre t
( 19AE1 Tc - Due to the location of the detectcra they [
ses relatively coldar water as the tamparature l from 157. to 100% for the cold tog drops 10SQL. I This is due to the Tavu being constant oVHF this range, nececeitating Tc decrutning over this rango JQigl. {
Reference Systems TrahnIng Menual, 'lluclear Instrumentaticn',
pages 2. l
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3.04 (2.0)
For the Safety Features Actuation Syntoms
- a. How does the SFAS circuitry prevent inadvertant (2.0) actuation of Reactor Building spray if two 30 psig bistables actuate unexpectedly?
3.04 Answers
- a. In sech analog cabinet there are two 30 psig bistables L Q)_. For a total of sin. With a two out of three logic there exists the possibility that an actuation will start either the pumps or open the valves, but not both 11tQL.
l l References Systures Training Manual, 'Safoty Features Actuation System', page 35-13.
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3.05 (1.5) 0 Regarding the Diamond Control panels
- a. What three conditions will cause an ' Auto (1.5)
Inhibit' to occur?
3.05 Answers
- a. Neutron error > or = 1%. (0.5)
Safety rods not withdrawn. (0.5)
Loss of power to the ICS. (0.5)
I
Reference:
Gystemn Training Manual, ' Control Rod Drivu Mechanisms and Control System', page 37b-61.
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l 3.06 (2.5) ;
i During feed and bleed operation of the Makeup and l Purification system certain interlocks are interposed by the ,
l Control Rod Drive system and the Nuclear Instrumentation. l What are those interlocks when reactor power is at the '
following levels?
I ,
i a. Reactor power in 10%. (1.0) ,
1 I b. Reactor power is 75%. (1.5) f 3.06 Answers
! a. Safety groups withdrawn. J. 5) i
! Regulating group 5 must be withdrawn 25%. .0.5) i
- b. Safety groups withdrawn. 'O.5) t Regulating group 5 must be withdrawn 25%. .0.5) [
Regulating group 7 must exceed nominal rod !
position margin. (0.5) t
Reference:
Operating Pracedure, D.9, ' Soluble Baron Concentre'.i on Control ' , page B. 9-3. ;
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3.07 (1.5)
Following a Reactor trip the Integrated Control System initiates a runback of feedwater flow via the Unit Load Demand runback at 20% per minute. However this is not as effective in the immediate case as two other control signals in the ICS.
- a. What are those two control signals and why are (1.5) they more limiting than the runback immydigigly following the reactor trip?
3.07 Answers ,
i m.
- 1. BTU limits due to hot leg temperature and (0.75) the OTGG pressure.
l
- 2. Cross limits from the reactor. (0.75)
References Systems Training Manual, ' Integrated Control System *.
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t 3.08 (3.0) l The Reactor Coolant Pump Power Monitor System senses electrical power to each RCP to determine whether the pump ,
i is operating.
i
l l b. What are four control functions that use the (2.0) output of the RCP Power Monitor as a parameter?
I 3.08 Answers
- a. Breaker position would not show a loss of flow (1.0)
- condition if there was a failure of the '
l shaft / impeller assembly (i . e. impeller becomes j separated from the shaft and the shaft continues to rotate with the breaker closed).
- b. Any four J92D1 each l
( RPS power / pumps trip.
I l ICS runback logic.
l ICS Aux. Feed. valve control.
Auto start of aux feed pumps.
Close permissive for CCW supply header containment isolation valve.
Reference:
Systems Training Manual, 'Ruactor Coolant System *,
' Reactor Protection System', ' Integrated Control ,.
System'.
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iy 3.09 (2.0)
's' Regarding the Nuclear Instrumentation:
- a. Indicate on Figure 3.1 the point where each of the following occurs:
- 1. Source range auto-turn off. (0.5)
- 2. Source range back-up auto-turn off. (0.5) 3.'Mantmum source range level you may-reach (0.5) without.intermidiate range response during a startup.
- 4. Minimum intermidiate range you may reach (0.5)
-without source range response during a shutdown.
3.09 Answer:
'As indicated on figure 3.1 IQtSL each.
t
Reference:
Systems Training Manual, ' Nuclear Instrumentation'.
4
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9
DETECTOR NEUTRON FLUX.NV I 1 1 I I I I I I I I I I I. I l- I I 5 5 5 5 5 5 5 5 5< 5 5 5 3
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3.10 (1.5)
What would happen to the indication in the circuit if the following instrumention failures occur as described?
- a. An short circuit occurs in the lead to a (0.5)
Resistance Temperature Detector (RTD).
- b. An open circuit occurs in the lead to a (0.5)
Thermocouple temperature detector (TC).
- c. The detector leg of a Pressurizer level (0.5)
Differential Pressure cell becomes plugged.
3.10 Answer:
- a. Indication goes GY to M scale boL 0IA (i . . T i . ite (0.5) resistance).
- b. Indication goes to zero on the scale. (0.5)
- c. Indication stays 'as is*. (0.5)
Reference:
Systems Training Manual, 'Non-nuclear Instrumentation
- 10
3.'11 (1.0)
Regarding the Fire Protection System:
- a. What will happen to the motor driven fire pump (1.0) as pressure in the fire main decreases from normal pressure in the header down to 90 psig and then returns to normal pressure?
3.11 Answer:
- a. The pump will start (93 + or 3psig) then run (1.0) for 7 minutes and stop.
Reference:
Systems Training Manual, ' Fire Protection System',
page 24-77.
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3.12 (1.O)
Regarding the Integrated Control Systems
3.12 Answer:
. a. The temperature compensation for the feedwater (1.0) temperature corrects for mass differences caused by temperature variations in the feedwater.
Reference:
Systems Training Manual, ' Integrated Control System".
End of Section 3 i
e j
! 12
Section 4 Procedures - Normal, Abnormal, Emergency, and Radiological Control 4.01 (3.0)
Regarding E.05 ' Excessive Heat Transfer's
- a. After termination of an overcooling transient (3.0) why is it important to maintain RCS temperature constant?
4.01 Answers
- a. Decay heat will soon heat the RCS up (1.0). The additional mass added from the HPI system J1 391 will exp and and potentially cause overpressuri z ati on or have the RCS go solid 11291
Reference:
E.05 ' Excessive Heat Transfer', page E.05-4.
1
4.02 (2. O) .
Regarding Rule 6 ' Reactor. ' Vessel Thermal Shock
. Considerations:
"a. What two criteria determine when Pressurized (2.0)
Thermal Shock is of concern?
^
4.'02 Answer:
- a. RCS temperature < 500 deg. F and cooldown ate (1.0)
.> 100 deg. F per hour.
Or RCS temperature < 500 deg. F and HPI Slow (1.0) exist =g R.c.Pb .F.
Reference:
Rule 6, ' Reactor. Vessel Thermal Shock. Considerations
- 2
4.03 (1.5)
Procedure AP.305-7, ' Area Definitions and Posting
- defines those areas of radiological concern which must be posted and established. What are the definitions for the following areas?
a.. Radiation Area (0.5)
- b. High Radiation Area (0.5)
- c. Secured High Radiation Area (0.5) 4.03 Answers
- a. > 2.5 mrem /hr but not > 100 mrem /hr. (0.5)
- b. > 100 mrem /hr but not > 1000 mrem /hr. (0.5)
- c. >1000 mrem /hr when measured at 18 inches. (0.5)
Reference AP.305-7, ' Area Definitions and Posting',
page AP.305-7.
h_______- __ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
'4.04 (2.0)
During a heatup to Hot Shutdown procedure B.2, ' Plant Heatup and Startup' requires in step 4.2.33 to verify that the Reactor Protection System channels are in shutdown bypass.
- a. What reactor trip protection is removed because (2.0) of the channels being in shutdown bypass?
4.04 Answers
- a. Poweh -
imbalance -
flow (0.5)
(flux / delta flux / flow) power -
pump (0.5)
Low reactor pressure (0.5)
Reactor pressure -
temperature (0.5)
Reference:
Procedure B.2, ' Reactor Coolant System Heatup to Hot Shutdown *, page B.2-10.
4
[--
4.05 (3.0)
Procedure B.3, ' Normal Operations', has as an option for power manuevering of an gaglaOOgd Egwgt Bgtura at (4gmlg31 Bates. . This allows maintaining the rods within the nominal position band to preserve return to power capability at nominal ramp rates. Enclosure 10.5 is included for reference.
- a. What potential problems will this present for (3.0)
.the . operators if a 100 100 (50% load change, staying at 50% power for 15 minutes) manuevcr is executed at nominal ramp rates?
4.05 Answers
- a. Large feed and bleed volumes for processing. (1.0)
Large negative core imbalance. (1.0)
Gross Xenon distribution imbalance. (1.0)
Reference:
Procedure B.3, ' Normal Operations *, page B.3-13.
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4 4.06~(2.0)
During the performance of a monthly surveillance test you notice that the previous test was conducted on December 1, 1985. Today is January 5, 1986. This is an interval of 36 days. The test was conducted previously on November 2, 1985 and October 1, 1985 (Calender on following page.)
- a. Does this. situation constitute non-compliance (2.0) with the Technical Specifications? EXPLAIN.
4.06 Answers
- a. No -11tQL. The .T.S. allow a surveillance interval to be extended by 25% 1Qtgl. With the "
interval for 3 consecutive surveillances not to exceed 3.25 times the interval IQuQl. .
Reference:
T.S., Definitions 1.9.
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o 4.07 (2.0)
Regarding AP.4A, ' Safe Clearance Procedures *:
- a. The plant is in Cold Shutdown and Maintenance personnel want to close out a clearance on the High Pressure Injection system. It's agreed that it is ready for a valve line-up. Is a dual verfication line-up required? (1.0)
EXPLAIN. (1.0) 4.07 Answers
- a. No 11cQL. It is not required until plant temperature exceeds 200 deg. F lluQL.
Reference:
Procedure AP.4A, ' Safe Clearance Practices'.
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. 4.08'(1.0)
Regarding 10 CFR 55 ' Operators' Licenses *: (
- a. What are '
Controls
- as defined by 10 CFR 557 (1.0) 4.08 Answers
- a. Controls are defined as aparatus and mechanisms the manipulation .Jg22_51 of which affect the reactor or reactor power J92251 Reference 10 CFR 55.4 and 55.10 B
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! o 1 , 4.09 (2.0)
Concerning Refueling activities:
- a. When must direct communications between the (1.0) refueling deck and the control room be established?
- b. During movement of fuel assemblies in the (0.5) containment, what condition must the airlocks and the equipment hatch be in?
- c. When may one train of Decay Heat Removal be (0.5) declared inoperable when moving fuel, without violating the Technical Specifications?
4.09 Answers
- a. "shall exist whenever changes in core geometry (1.0) are taking place."
- b. Airlocks must have at least one door closed. (0.25)
Equipment hatch must be secured by at least (0.25) four bolts.
- c. When the level in the transfer canal is at (0.5) least 37 feet.
Reference:
T.S. 5.8, ' Fuel Loading and Refueling *.
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, 4.10 (1.0)
What is the bases for the following statements
- a. The reactor shall be maintained suberitical (1.0) until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressuri=er.
4.10 Answer:
F
- a. Ensure RCS does not go solid in the eve nt of a (1.0) rod withdrawal accident gap,% [gdig.'wp
Reference:
T.S. 3.1.3, ' Minimum Conditions for Criticality *.
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. 4.11 (1.0)
Technical Specification 6.2 ' Organization's
- a. How many licensed operators must the shift crew (0.5) composition have during a reactor startup?
- b. How many hours may an individual work (0.5) continously before being relieved as a watchstander?
4.11 Answers
- b. 16 (Not including shift turnover time.) (0.5)
Reference:
T.S. 6.2, ' Organization'.
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4.12'(3.0) e During a power. escalation from 50% power the plant experiences a dropped rod at 75% power.
- a. What is the staximum load limit imposed by the (1.0) :
t ICS af ter the runback'? ;
B. Will the runback that occurred ensure that any Technical Specification limit was not exceeded 7 (1.0) ,
EXPLAIN. (1.0) 4.12 Answers ;
, a. An Asyeetric Rod will cause Unit Load Demand to (1.0) be limited to 60% of full power. ,
1 l b. No 11 g,QL. The runbace: goes to load limit and ;
not to reactor power, thus the possibility ;
l exists that a limit could bit encaeded without t
{ operater action 11tQ1 l
Reference:
A 71, ' Integrated Control System', pages A.71-12 to
~A.71-13.
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a i 4.13 (1.5)'
\o precautions and
}-
~
What are the reasons for the following
. limitations?
- a. There shall not be ' makeup' to the makeup tank (0.5) during the conduct of a heat balance calculation,
- b. Before RCS temperature decreases to less than (0.5) ;'
500 deg. F, at least one reactor coolant pump must ba stopped.
i
- c. The turbine generator must not be operated (0.5) below 59.5 Hertz.
4.'13 Answer: ,
- 4. Steady state conditions must be maintained for (0.5) ,
accurate calculation.
l b. Prevent core lift. (0.5) l l c. Vibration induced in last row of turbine (0.5) l blading.
References Operating Procedurec, A.46 ' Main Turbine System *, !
B,7 ' Heat Dalance Calculation', A.2 ' Reactor Coolant Pump System *.
I l End of Section 4 i
END OF EXAMINATION 1
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U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
. Facility: RANCHO SECO ,
Reactor Type: BABCOCK AND WILLUA Date Administerea: u nut'. spy 3. 1986 -
Examiner: mw inmisiny Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only. Staple r question sheet on top of the answer sheets. Points for each question are indi-cated in parentheses after the question. The passing grade requires at least ;
70% in each category ar.d a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts,
% of t i
Category % of Candidate's Category i
_Value Total Score _Value _
Category 6 1 '
25.0 _25.4 5. Theory of Nuclear -
- Power Plant Operation, Fluids, and Thermo-dynamics r
%.O 46,e-W M. b 6.
Plant Systems Design, Control, and Instrumentation N
M ;
~
. 7. Procedures - Normal, ,
Abncrmat Emergency, '
and Radiological ;
Control ;
5 25.0 25.t a. Administrative Fro- '
cedures, Conditions, and Limitations Totals i
~ Final Grace ~
i All work done on this examination is my own, I have neither given nor received aid. i l
Canotdate's Signature l
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_ - . - . . . .= . - . - . . . -
s EQUATION SHEET f = ma r = s/t w = mg 2
- a = v,t + at Cycle efficiency = --- "I E = aC a = (vg - v,)/t KE = leer vg = v, + at A = AN A=Aeg 4C .
PE = agh- e = e/t A = In 2/tg = 0.693/tg [
W = v&P '
(e, )(g, )
AE = 9314m
~
- % " "' ~ (t g+e) b 6=is,AT .
r . r .-Ex k=UAAT g , g ,-ux Fwr = Wg [n ~
I=I 10
- P=P 10 (*) TVI. = 1.3/u 1
P=P oe tT HVI. = 0.693/u l i SUR = 26.06/T '
+
T = 1.44 DT SCE. = $/(1 - K,gg) ;
IA*geh p SUR = 26 g, CR, = $/(1 - K,ggx)
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T=lt/p)+ [(f
- p)/A,f,p] 1 eff 1
- O II ~ eff)2 T ,= 1*/ (p "f) M " I/(I - Kegg) = CR /CR g g T = (3 - 0)/ A*gg o M = (1 - K,gg)CIII ~ Eeff)1 i
8 " ( eff~) eff * #efflEaff 303 . (I _ g,gg)fg.gg l p= [1*/TKygg ] + [B/(1 + A,ggt )] g* = 1 x 10 seconds
~ '
P = I(V/(3 x 10 0) ~I A,gg = 0.1 seconds Z = Ikr - I Idgy*I422 (
WATER PARAMETERS Id =Id22 g
1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (meters)
I sal. = 3.78 liters R/hr = 6 CE/d (feet)
I 1 fc3 = 7.48 gal. MISCEI.T.ANEOUS CONVERSIONS ,
Density = 62.4 lbm/fc 3 1 Curie = 3.7 x 10 dps 10 Density = 1 gn/cm 1 kg = 2.21 1ha [
Meat of vagoris'ation = 970 reu/lbm I hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lbs 1 Mw = 3.41 x 10 Etu/hr-6
! 1 Ata = 14,7 psi = 29.9 in, fg. 1 Etu = 778 ft-Ibf <
I 1 ft. H y0 = 0.4333 lbf/in 1 inch = 2.54 ca
'T = 9/5*C + 32 "C = 5/9 (*F - 32)
. SECTION 5 SENIOR GPERATORS EXAMINATION Theory cf Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.01 (2.0)
Regarding the feedwater heaters
- a. If the leavel in a feedwater heater is alloweo to increase, what will happen to the I temperature of the feedwater exiting the heater, increase, remain the same, or decrease? (1.0)
Why? (1.0) 5.01 Answers
- a. Decrease 11tQL. The level increase exposes less surface area in the heater tubes iQu@l. There is less latent heat renoved from the steam, and therefore less neat transfered to the feedwater resulting in a lower exit temperature for the feedwater iQunt.
Reference:
General Electric " Thermodynamics, Heat Transfer, and Fluid Fl ow" , Chapter G.
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5.02 (2.0)
Listed below are four factors that can affect the Departure from Nucleate Boiling Ratio (DNBR). What will the effect on DNBR be (increase or decrease) IF THE FACTORS LISTED BELOW ARE INCRgAggp?
- a. Reactor Power. (0.5)
- b. Reactor Coolant System flow. (0.5)
- c. Reactor Coolant System temperature (average). (0.5)
- d. Reactor Coolant System pressure. (0.5) 5.02 Answer:
- a. DNBR decreases. (0.5)
- b. DNBR increases. (0.5)
- c. DNBR decreases. (0.5)
- d. DNBR increases. (0.5)
Reference:
Facility Manual, ' Heat Transfer and Thermodynamics",
pages 122 to 126, and 243 to 252.
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5.03 (3.5)
The Tecnical Specifications discuss safety limits, such as DNBR in question 5.02 above. and Linear Heat Ratio (LHR).
- a. What is the maximum value and units of the (1.0)
Linear Heat Ratio as defined in the facility Technical Specifications?
- b. What failure mechanism would possibly occur if (1.5) the value cited for the LHR in (a.) above is exceeded?
- c. If DNBR limits were exceeded, would that mean (1.0) that LHR limits were also exceeded? Why?
5.03 Answers l a. 20.4 KW/ft. (1.0)
/
v' b . Exceeding the limit will lead to excessively ,
high fuel centerline temperatures 1QuGL. This could lead to fuel melting JQ351, and potentially fuel rod rupture from overpressurization (Burnout) IQt5t.
- c. No (0.5), the mechanisms are exclusive of one another (or not related) IQuQl.
Reference:
Facility Manual, ' Heat Transfer and Thermodynamics *,
pages 242 to 252.
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During preparation for a normal startup, af ter calcul ating for an Estimated Critical Position it is determined that a 200 ppm baron dilution is necessitated. Before dilution begins the indication for the Source Range instrumentation shows one channel at 35 cps and the other at 40 cps. The dilution is started and after 100 ppm of dilution the count rates are 70 cps and 80 cps respectively.
Based on the above indications and assuming an initial Koff of 0.90 :
- a. What would Keff be after this 100 ppm dilution? (1.0)
- b. What reactivity was added by this 100 ppm (1.5) dilution?
- c. What would you expect Keff to be after the full (0.5) 200 ppm dilution?
5.04 Answer:
- a. CR1 (1-Keff1) = CR2 (1-Keff2) (0.5) 40 (1 - 0.9) = 80 (1 - Keff2) i Keff2 = 0.95 after 100 ppm dilution (U.5)
(Relation of doubling rule of thumb acceptable.)
- b. Rhoo = (Keff0 - 1) / Keff0 =
= (0.9 - 1) / O.9 = -0.1111 (0,5)
Rhol = (Keffi - 1) / Keffi =
= (0.95 - 1) / ).95 = -0.0526; (0.5)
Reactivity added = RhoO - Rhoi = +0.05848 (0.5)
- c. The reactor will be at or near critical. (Keff (0.5)
= 1.0).
Reference:
Facility Menual, ' Reactor Theory', pages RT-6.2 to RT-8.11.
4 4
5.05 (3.0)
Refer to Figure 5.'1 which shows an instantaneous, positive reactivity insertion into an already critical reactor core (at time t=O), followed by a removal of this positive reactivity after a stable reactor period is reached (at time t=1) . Assuming no source neutrons:
- a. Show (on the figure provided) the resulting (1.0) reactor startup rate as a function of time for this reactivity change,
- b. Show (again on the figure provided) the reactor (1.0) power level as a function of time for this reactivity change.
- c. Explain the shape of the reactor power response (1.0)
(
at a time IMMEDIATELY AFTER time t=1.
l '
5.05 Answer:
- a. and b. Attached. (2.0)
- c. Prompt Jump in total neutron flux due to an (1.0) increase in prompt neutron production.
Reference:
Westinghouse Training Notes; Neutron Kinetics.
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5.06 (2.5)
What would the most likely source (origin) be for the following- isotopes if they were found in a Reactor Coolant System water sample?
- a. Cobalt 60 (0.5)
- b. Nitrogen 16 (0.5)
- c. Argon 41 (0.5)
- d. Strontium 90 (0.5)
- e. Rubidium 88 (0.5) 5.06 Answers
- a. Activated corrosion. product. (0.5)
- b. Activation of Oxygen 16. [6 der 4 ie - ( (0.5)
- d. Fission product. (0.5)
- e. Daughter product of Krypton 88 (f issi on (0.5) product).
References Glasstone and Sessonske, Nuclear Engineering.
6
5.07 (2.5)
Refer to Figure 5.2, a sketch of a centrifugal pump curve and a system charachteristic curve for a centrifugal auxillary feedwater pump system for the following questions:
- a. Show how the curve)s) will change as the PORV (0.75) opens and reduces the steam generator pressure by 50%.
- b. .Show the additional changes Efrom part (a.) (0.75) above3 to the curve (s) as the pump discharge valve is partially shut.
- c. Show the additional changes [from part (b.) (1.0) above] to the curve (s) as the pump speed is reduced by 25% (INDICATE MAGNITUDE OF CHANGES).
5.07 Answer .
Key attached.
Reference:
Facility Manual, Heat Transfer and Thermodynamics, pages 319 to 337.
7
FIGUPE 5.2 A
hPressur.S/0 .
f
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T E *a.
@e swem / T
& P 18eleM , ::sp iG ,g a si ...
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+ a r e. 4 Pump Head (h t) = Poischarge - Psuction I .
System Operating Curve 1 initial Operating Point j
%, /
"y Pump Operating i Curve h .
Volume Flow Rate (V) l
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6 f Pony 2
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' g, Point l ~ FordeW5 co.s)
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,,. Nu.p erre n 4 c9f6 .
r..e o soota 63)iS% Volume Flow Rate (U) l l
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i 5.08 (1.0)
Which reactivity coefficient *s behavior can best be explained by a discussion of the broadening of the U238 and Pu240 resonance absorption peaks?
Choose the best answers
- a. Moderator Pressure Coefficient
- b. Moderator Temperature Coefficient
- c. Moderator Void Coefficient
- d. Fuel Temperature Coefficient ,
5.08 Answers
- d. Fuel Temperature Coefficient (1.0)
Reference:
Facility Manual, Reactor Theory.
8
5.09 (3.0)
Three agents are typically added at various times to control Reactor Coolant System chemistry.
- a. What are those agents and what aspect of RCS (3.0) chemistry do they control?
5.09. Answer:
.a.
- 3. Lithium Hyroxide - pH control. (1.0)
Reference:
Systems Training Manual, Chapter 6, ' Reactor Coolant Chemical Addition and Sampling System *, page 6-2.
9
5.10 (2.5)
There are two ways to describe flux distribution variations that can occur in the reactor core. These are typified as flux tilts.
- a. What is Axial flux tilt? (0.5)
- b. What is Radial flux tilt? (0.5)
- c. What are three possible conditions / events that (1.5) could cause a flux tilt to occur?
8.10 Answers
- a. Variation of flux along the vertical axis (0.5)
(power in the top - power in the bottom).
- b. Uneven flux distibution across a horizontal (0.5) plane of the core (or description of Guadrant Power Tilt formula).
- c. Any three J93D) each.
- 1. Mispositioned rod.
- 2. Xenon oscillation.
- 3. Improper fuel load.
- 4. Flow imbalance, (3 RCP's etc.)
Reference Systems Training Manual, ' Nuclear Instrumentation'.
End of Section 5 10
SECTION 6 Plant Systems Design, Control and Instrumentation 6.01 (2.0)
Regarding an gmatqqngy start of the Emergency Diesel Generators:
- a. If the speed governors of the Emergency Diesel (2.0)
Generators were 1DadysctgDily left in the gcggg mgdg of operation, what potential problems may occur for the gqutgment that tries to start on the 4160 volt safety related busses?
i 6.01 Answers 4
- a. The equipment breakers tripping from (1.0) overcurrent or undervoltage.
I Or, the equipment may not perf,orm their safety (1.0) related functions / design objectives.
Reference:
Facility Systemu Descriptions Chapter 45, ' Emergency Diesel Generators *,
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6.02 (3.0)
Regarding the 125 volt DC systems
- a. When the standby and norr41 battery chargers (1.0) are aligned for parallel operation which unit will assume the highest load?
- b. When the chargers are operating in parallel how (1.0) is the voltage adjusted if needed?
- c. If AC power is lost f or an e:: tended period of (1.0) time, how long will the 125 volt DC batteries carry their loads?
6.02 Answers l
l a. The standby unit. (Has highest capacity.) (1.0)
- b. By using the rheost 9 t p on thq normal charger. (1.0) 6, a djusv:A mgw/ kwqh
- c. For two hours. (1.0)
Reference:
Systems Training Manual, ' Vital Electrical l Distribution 4160 volt and Dolow', pages 43-9 to 43-18, and 43-79, 43-80.
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e-6.03 (3.0) litigt gligqt will' a failure of power to the Reactor Protection System channel A pressure instruments have on the :
following equipment and EAD manual control of the function '
be taken in the control room?
I
/ a. Pressurizer heaters. (1.0) l l
( b. Spray valves. (1.0)
- c. EMOV. (1.0)
! -6.03 Answers
- a. Heaters . will energize. Manual control by the (1.0)
I handswitch is availiable.
l
- b. The spray valves will fail closed. Manual (1.0) control by the handswitch is availiable.
l
- c. The EMOV will fail closed. Manual control by (1.0) l the handswitch is availiable.
References Systems Training Manual, ' Vital Electrical Distribution 4160 volt and Below', pages 43-02 and 43-83.
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6.04 (3.0)
Regarding the Reactor Coolant Pumps:
What is the purpose of the following RCP start interlocks?
- a. Reactor power must be less than 30% of full (1.0) power.
- b. Component Cooling Water flow must be greater (1.0) than 307 gallons per minute. r
- c. Reactor Coolant System temperature must be (1.0) greater than 500 deg. F.
6.04 Answers
- a. Reduces the potential of a positive reactivity (1.0) insertion from cold water in an idle loop.
4
- b. Insures adequate flow of cooling water to seals (1.0) and motor.
~
- c. Precludes core lift phenomena by insuring delta (1.0)
P across core is not sufficient to cause lift of fuel elements to occur.
References System Training Manual, ' Reactor Coolant System',
pages 2-75 to 2-78.
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0 6.05 (4.0)
Using the enclosed diagram of the Integrated Control System explain the actions that would occur in the following control sections if an Assymetric Rod condition occurred at 90% of full power. Include the actions that the control sections take with respect to components and other interactions. l
- a. Feedwater Control. (2.0) i
- b. Reactor Control. (2.0) f 6.05 Answers
- a. The feedwater demand calculator will be ,
receiving a decreasing signal when the Assymetric Rod runback occurs 1QuGL. This will 1
send a reduced demand signal to the feedwater demand units 1QuQL, which will send a signal to j the Feedwater Regulating valves 1QuGL and the
- Feedwater Pump speed controllers J9tD1 I
- b. The Reactor Control section recieves the same
- decreasing utgnal IQuGl. From the comparison between this signal and the Neutron Error the Reactor Demand difference unit will send a I
signal that will call for the insertion of rods J1491 The (rod bite at 90% may be weak) and a 1 Reactor Crosslimit may occur limiting Feedwater flow j9291 j References Systems Training Manual, ' Integrated Control System',
pages 32-77 to 32-93. l I
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l .S) 6.06 W Regarding the Encore Power Range Nuclear Instrumentation:
- a. Due to the design of the Babcock and Wilcox - ;2. G F plant there are two plant design f,c) characteristics that affect the Excore Power Range calibration. What are those' design characteristics and how do they affect the detectors?
6.06 Answer gg I B&W design han a
- a. Rod shadowing - The more\
, " rodded" core, meaningtheregulatingrodsarej
[ inserted much more than in other designs 1QtQL.
This would tend to ' shadow
- the upper detectors from seeing a representative flux in the core I,
19101 --
Tc - Due to the location of the detectors they see relatively colder water as the temperature from 15% to 100% for the cold leg drops 1QtQl.
This is due to the Tave being constant over this range, necessitating Tc decreasing over this range JQ3DL.
Reference Systems Training Manual, ' Nuclear Instrumentation',
pages 2.
6
6.07 (1.0)
- For the following core components indicate the composition
- of the neutron absorbing materials
- a. Burnable poison rod. (0.5) i t
. b. Control rod. (0.5) ;
i j 6.07 Answer
)
- a. Alumina with boron carbide. (0.5)
Reference:
Systems Training Manual, ' Fuel Assembly and Control Components'.
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- 6.08 (2.0)
'For the Safety Features Actuation Systems
- a. How does the SFAS circuitry prevent inadvertant (2.0) actuation of Reactor Building Spray if two OO psig histables change stato?
6.08 Answers
- a. In each analog cabinet there are two OO psig bistabl es J,129),,. For a total of s i t: . Wi tti a two
- out of three logic there e>ists the possibility that an actuation will start either the pumpu l
or open the valves, but net both 11tQL.
Reference:
Systems Training Manual, ' Safety Featuren Actuation 1 System *, page 35-13.
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, 6.09 (1.0)
Which of the following parametern will not cause a trip of the main turbine if the critical valun is exceeded?
- a. High exhaust hood temperature.
- b. Low bearing oil pressure.
- c. Excessive thrust bearing wear.
- d. Low hydraulic fittid pressure.
6.09 Answers
- a. High exhaunt hood temperature. (1.0) '
Reference:
Systemu Training Manual, ' Main Turbine', pages i 16b-116 to 16b-122.
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, 6.10 (1.5)
Regarding the Dinmend Control panels t i ;
l a. What conditions wall cause an ' Auto Inhibit
- to (1.5) l occur?
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i 6.10 Answers .
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l a. Neutron Crror > or = 1L (0.5)
Safety rods not withdrawn. (0.5) ;
I Loss of power to the ICS. (0.5) j i
1 i References Systens Training Manual, ' Control Rod Drive Mechantums and Control Gyotom', page 37b-61.
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- l. 4.11 (2.5) [
1 I
l Regarding the Reactor Building Isolation System:
l l a. In ^the event of a fatture of power or loss of (1,0) ?
I r.ontrol air, what pcsition w111 the valve l l
assume f or the f ollowing? l
! t i 1.. Air operated isolation valve. i l !
l 2. Meter operated valve. l i
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- b. In general where would the valves, described (1.0) j 1 above, be l ocated in relationship to the j l containment penetration (i . e. inboard or l cutboard) if they were on the came line? L i
- c. Are simple check valves used for outboard (0.5) ;
autoe.atic isolation valves 7 j 6.11 Answers {
t
- a. [
- 1. Will fail closed. (0.5)
[
- 2. Will fail "As is". (0.5) {
- b. i
- 1. Outboard the penetraticn, (or af the motor (0.5) f
. operated valve is outboard of the penetration )
it will be outboard the motor operated valve.) (
- 2. Inboard the penetration, (or between the (0.5) {
l penetration and the outboard air operated l l valve.) !
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c, too. (0.5) 3 References Systems Training Manual, ' Reactor Building ano l Support Systems', pages 4a-44 to 4a-47. i h
i End of Section 6 t
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Section 7 Procedures - Nornal, Abncrmal, Emergency and Radiological Control 7.01 (3.0)
Regarding E.05 'E:tcessive Heat Transfer *:
- a. After termination of en overcooling tranutent (3.0)
Why is it important to maintain RCS temperatLre constant?
7.01 Answer:
- m. Decay heat will soon heat the RCG up 11 QL. The <
additional mass added from the HPI system fl291 will expand and potentially cause overpressurication or have the RCG go calid 11191 References E.05 'Excessi ve Heat Tranuf er ' , page C.05-4.
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7.02 (3.0)
Regarding Rule 6 ' Reactor Vescol Thermal Shock Ccneiderations:
- a. WNet two criteria determine when Fressurized (2.0)
Thornal Shock is of concern?
- 6. What exception can be taken to following the (1.0) guidelinen of Rule 67 6 7.02 Answer? ,
- a. RCG temperature < E00 dog. F and cooldown rato (1.0)
> 100 deg. F oer hour.
Or l RCS temperature < 500 deg. F and HPI flow (1.0) eriots)it, W S O T-l
- b. The gui d ul i r.o.5 of the rule should not be (3.0) folloneo sf an OTGG tube rupture has boon determined to exist.
References Rule 4 ' Reactor Ver.cel Thermal Chock Contadoraticns' l
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7.03 (3.0)
Procedure E.C2
- Vital Syntom Status Verfication' steps 15.0, 16.0, 17.0, and 18.0 list the 4:ur abnereal transient ,
syn.ptom dGtcrmanaticnn.
- a. What are those four abnormal synptom (3.0) determinations and 11ut tne n. in ordre of .
priority?
7.03 Answers a (0.E) fcr each syepten, and (0.5) foe hichaut and l owest priori ty. .
Loss of subcooling margin.
Lack of heat transicc. '
Excessive heat tran*sf er.
OTEG tube rupture.
Ref erences E. 02, ' Vit al Systnm Status Vn-Afication', cegen E.02-6 and E.02-7.
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7.04 (3.0)
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Pr ocedure AP.305-7, ' Area Definitions and Posting
- defines those areas of radiological concern which must be pouted and established. What are the definitions for the following areas?
- a. Radiation Area (1.0)
- b. Hl3h Radiation Area (1.0)
- c. Secured High Radiation Area (1,0) 7.04 Answers
- a. > 2.5 mrem /hr but not > 100 mrem /hr. (1.0)
- 6. > 100 mram/hr but not > 1000 meum/hr. (1.0)
- c. >1000 meem/hr when measured at 18 inches. (1.0)
Referenco AP.305-7, ' Area Definitions and Posting',
page AP.005-7.
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, 7.05 (2.0)
During a heatup to Hot Shutdown procedure B.2, ' Plant Heatuo and Startup' requires, in step 4.2.33, to verify that the Reactor Protection System channels are in shutdown bypass. ,
- a. What reactor trip protection is removed because (2.0) of the channels being in shutdown bypass? !
i 7.05 Answers i
- a. Power, -
imbalance -
flow (0.5)
(flux / delta flux / flow) power -
pump (0.5)
Low reactor pressure (0.5)
Reactor pressure -
temperature (0.5) ,
References Procedure B.2, ' Reactor Coolant System Heatup to Hot l Shutdown *, page B.2-10.
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7.06 (2.0)
Regarding procedure B.3, ' Normal Operations's
- a. If an axial Xenon oscillation is or has existed (2.0) the nominal core power level cannot be raised until what conditions have been met?
7.06 Answers
- a. Imbalance is within " Nominal" + or - 2% for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> J1291 and the controlling group is in the long term operating band J1291 References Procedure B.3, ' Normal Operations', page B.3-21.
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7.07 (3.0)
Procedure B.3, ' Normal Operations', has as an option for power manuevering of an 90DifDDEd E9WEC 8gtuCD et UgmlD31 Batgg. This allows maintaining the rods within the nominal position band to preserve return to power capability at nominal ramp rates. Enclosure 10.5 is included for reference.
- a. What potenti al problems will this present for (3.0) the operators if a 100 100 (50% load change, staying at 50% power for 15 minutes) manuever is executed at nominal camp rates?
7.07 Answers
- a. Large feed and bleed volumes for processing. (1.0)
Large negative core imbalance. (1.0)
Gross Xenon distribution imbalance. (1.0)
Reference Procedure B.3, ' Normal Operations', page B.3-13.
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7.08 (3.0)
While you are standing watch as the Shift Supervisor you recieve a call form a security guard notifying you of smoke exiting from a safety related switchgear room,
- a. Besides the location you were given, what other (1.0) information would you expect the security guard to convey to you?
- b. What conditions would compell you, as Emergency (1.0)
Coordinator, to report this event to the NRC Operations Center as an Unusual Event?
- c. Who are the members of the Fire Brigade? (1.0) 7.08 Answers
- a. Type and size of fire. (0.33)
Any injuries. (0.33)
Any other information pertinent to the fire. (0.33)
- b. Fire lasting more than 10 minutes which may (1.0) effect safety related equipment.
- c. Senior Control Operator (Fire Brigade Leader) (0.25)
Auxiliary Operator. (0.25)
Equipment Attendant / Power Plant Helper. (0.25)
Two security offficers. g ( gcp, d l g94 (O.25)
Reference:
AP.501, ' Recognition and Classification of Emergency',
page 29. AP.5, ' Fire', pages 2 and 3.
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'7. 0 9 (3.0)
During your watch as Shift Supervisor lightning strikes on the grid cause a loss of offsite power, concurrently the main feedwater pumps trip from an unknown cause. The reactor trips and normal system actuations occur.
- a. What conditions would have to occur to allow (1.0) throttling the auxiliary feedwater flow with the transient as described above?
- b. What are two indications that would require (1.0) that throttling of the auxiliary feedwater flow be terminated?
- c. What exception would require that auxiliary (1.0) feedwater flow be stggggd?
7.09 Answers
- a. Level reaches 95% on operate range. (0.5)
Natural circulation is verified. (0.5)
- b. Any two: 1Qu5) each.
Natural circulation has ceased.
AFW actuation was delayed after RCP's tripped.
- c. If excessive primary to secondary heat transfer (1.0) exists.
Reference:
Rule-3, 'Feedwater Throttling Guidelines'.
End of Section 7 9
Section 8 Administrative Procedures, Precautions arid Limitations T
8.01 (2.0)
During the performance of a monthly surveillance test you notice that the previous test was conducted on December 1, 1985. Today is January 5, 1986. This i s an interval of 36 days. The test was conducted previously on November 2, 1985 and October 1, 1985 (Calender on following page.)
- a. Does this situation constitute non-compliance (2.0) with the Technical Specifications? EXPLAIN.
8.01 Answers
- a. No 11t91 The T.S. allow a surveillance interval to be extended by 25% 19tQL. With the interval for 3 consecutive surveillances not to exceed 3.25 times the interval 19tQL.
Reference:
T.S., Definitions 1.9.
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8.02 ( 4. 0 ) .
Regarding AP.4A, 'Saf e Clearance Procedures:
- a. The plant is in Cold Shutdown and Maintenance
. personnel want to close out a clearance on the High Pressure Injection system. It's agreed that it is ready for a valve line-up. Is a dual verfication line-up required? (1.0)
EXPLAIN. (1.0)
- b. The Maintenance personnel want to conduct a (2.0) test of the HPI system by having a test tag hung while the clearance tag is still hung. Is this acceptable? EXPLAIN.
8.02 Answer:
- a. No 11tQL. It is not required until plant temperature exceeds 200 deg. F 11sQL.
- b. No J12Q1 The procedure allows only one tag to be hung at a time (1.0).
Reference Precedure Ap.4A, ' Saf e Clearance Practices' .
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8.03 (3.0)
Regarding procedure AP.23, ' Control Room Watchstanding*:
- a. You are the oncoming Shift Supervisor, what are (3.0) stil items of information that you would seek from the offgoing Shift Supervisor when you relieve him7 8.03 Answers
- a. Any six of the following IQu51 cacht
- 1. Status of safety-related systems.
- 2. Running equipment and safety system alignments.
- 3. Inoperable equipment, (including instrumen-tation, and LCOs, including surveillance requirements.)
- 4. Reasons for new annunciator alarms.
- 5. Tagged equipment, (including any surveillance / equipment work in progress at the time of shift relief.)
- 6. Unusual events that have occurred during the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 7. Procedure changes.
- 8. New standing orders.
- 9. LCOs in effect.
(Examiner will accept other items, procedure does state "should include, but are not limited to".)
Reference:
AP.23, ' Control Room Watchstanding', page AP.23-3.
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8.04 (3.0)
Regarding 10 CFR 55 ' Operators' Licenses *:
- a. What are ' Controls
- as defined.by 10 CFR 557 (1.5)
- b. An individual may manipulate the controls of a (1.5) facility without a license only under certain conditions. What are those conditions?
8.04 Answers
- a. Controls are defined as aparatus and mechanisms the manipulation JQ22g1 of which affect the reactgg or reactor power JQ2291
- b. The individual must be in training to qualify as an operator IQuZ@l. And must be under the direction of a licensed operator IQuZQL.
Reference 10 CFR 55.4 and 55.10 4
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8.05 (2.0)
_Regarding temporary. changes to procedures:
- a. What actions must be taken to make a temporary (2.0) change to a Process Standard (AP 100 - 199)7 8.05 Answers
- a. No change to intent. (0.5)
SS and SCO or Plant Engineer signatures. (0.5)
Manager of Nuc. Ops.( or NPS or NOS or RCS) (0.5) approval.
Reviewed and approved within seven days of (0.5) implementation.
Reference:
AP.2, ' Review, Approval and Maintenance of Procedures', page Ap.2-4.
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8.06 (3.0)
Concerning Refueling activities:
- a. When must direct communications between the (1.0) refueling deck and .the control room be established?
- b. During movement of fuel assemblies in the (1.0) containment, what condition must the airlocks and the equipment hatch be in?
- c. When may one train of Decay Heat Removal be (1.0) declared inoperable when moving fuel, without violating the Technical Specifications?
8.06 Answers
- a. "shall exist whenever changes in core geometry (1.0) are taking place."
- b. Airlocks must have at least one door closed. (0.5)
Equipment hatch must be secured by at least (0.5) four bolts.
- c. When the level in the transfer canal is at (1.0) least 37 feet.
Reference:
T.S.-3.8, ' Fuel Loading and Refueling *.
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18 . 0 7 (3.0)
What are the bases for the following statements:
- a. The reactor shall not be made critical if RCS (2.0) temperature is less than 525 deg. F. (Iug bases required response.)
- b. The reactor shall be maintained suberitical (1.0) until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressurizer.
8.07 Answers
- a. Any 2 J12.01 each.
- 1. Moderator temperature coefficient is within its analy=ed range.
- 2. Consistency with FSAR analysis.
- 3. Ensures DTT + 10 deg. F. /
- 4. Mavor Sa4. pm>n(rscSr.:s %n J puer, co*((k; tw e a
- b. Ensure RCS does not go spid in the even$ of a rod withdrawal accident cl M 'n, b .(cf 4 4/e b/ M4EEEf
,()f
Reference:
T.S. 3.1.3, ' Minimum Conditions for Criticality'.
7
o 8.08 (2.0)
Technical Specification 6.2 ' Organization':
- a. How many operators must be in the control room (0.5) a during reactor startup?
- b. How many licensed operators must the shift crew (0.5) composition have during a reactor startup?
- c. How many hours may an individual work (0.5) continously before being relieved as a watchstander7
- d. TRUE OR FALSE: (0.G)
During extended shutdown periods, the use of .
overtime can be considered on a staff basis and not necessarily on an individual basis.
8.08 Answers
- a. 2 licensed. (0.5)
- c. 16 (Not including shift turnover time.) (0.5)
- d. TRUE. (0.5)
Reference:
T.S. 6.2, 'Organi:ation'.
AP 23 '/o p fck<A~
8
o-8.09 (3.0) i During a power escalation from 50% power the plant experiences a dropped rod at 75% power.
- a. What is the maximum load limit imposed by the (1.0)
ICS after the runback?
B. Will the runback that occurred ensure that any (2.0)
Technical Specification limit was not exceeded?
EXPLAIN.
8.09 Answers
- a. An Asymetric Rod will cause Unit Load Demand to (1.0) be limited to 60% of full power.
- b. No 11cQL. The runback goes to load limit and not to reactor power, thus the possibility exists that a limit could be exceeded without operator action 11tQL.
Reference:
A.71, ' Integrated Control System *, pages A.71-12 to A.71-13.
End of Section 8 End of Examina ion k
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