ML20214R796
| ML20214R796 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/01/1987 |
| From: | Bjorgen J, Burdick T, Clark F, Hare E, Dave Hills, Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214R790 | List: |
| References | |
| 50-373-OL-87-01, 50-373-OL-87-1, NUDOCS 8706080320 | |
| Download: ML20214R796 (72) | |
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U.S. NUCLEAR PEGULATORY COMMISSION REGION III Report No. 50-373/0L-87-01 4
Docket Nos. 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station j Examination Administered At: Wilmington, IL and Marseilles, IL Examination Conducted: ' April 20-23, 1987 i Examiners: Nb E. D. Lanksbury 4
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. Operator Licensing Section i
Examination Summary Examination Administered on April 20-23, 1987 (Report No. 50-373/0L-87-01)
Written, oral, and simulator examinations were administered to three Senior Reactor Operator (SR0) candidates and a simulator retake to one Senior l Reactor Operator (SRO) candidate.
Results: Three SR0 candidates passed the oral examination, one passed the 4
simulator examination, and three passed the written examination. A total of one candidate passed overall.
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- REPORT DETAILS
- 1. Examiners
- R. D. Lanksbury, Region III F. E. Clark, Region III D. E. Hills, Region III J. C. Bjorgen, Region III E. A. Hare, Region III
- Chief Examiner
- 2. Examination Review Meeting Copies of the written examination and answer key were given to the facility personnel for review at the conclusion of the written examination. Facility personnel provided their comments to the examiners on April 23, 1987. Their comments as well as the resolution are enclosed as an attachment to this report.
- 3. Exit Meeting On April 23, 1987, an Exit Meeting was held. The following personnel were present at this meeting:
Commonwealth Edison Company S. Harmon, Operations Training R. Armitage, Operations Training Supervisor W. R. Huntington, Assistant Superintendent, Operations J. C. Renwick, Production Superintendent C. M. Allen, Nuclear Licensing Administrator S. L. Trubatch, Staff Attorney M. S. Turbak, Operating Plant Licensing Director G. J. Diederich, Station Manager NRC R. D. Lanksbury, Chief Examiner J. A. Malloy, Resident Inspector, LaSalle F. E. Clark, Examiner The following topics were discussed at the Exit Meeting:
- a. The examiners identified four generic deficiencies in the candidates' knowledge and abilities. The first deficiency was that the candidates were very weak on their ability to utilize the emergency operating procedures. The second deficiency was that the candidates were weak in their ability to locate and recognize proper keys needed for emergency operations. The third deficiency was the candidates' 2
ability to determine if equipment and components were safety related.
The fourth deficiency was the candidates' ability, while performing as Acting Station Director, to properly classify emergency events based on gaseous releases,
- b. Two additional items of concern were identified. The facility's Emergency Operating Procedure (E0P) format is not easy for the candidates to utilize during transient events. The E0Ps require the candidates to use multiple procedures concurrently. Since all procedures are contained in a single 8-1/2" x 11" binder, it is considered physically impossible to effectively use several procedures at once. The E0Ps were in a format to be removed from the binder; however, when removed from the binder the loose pages from several procedures were mixed, which made them impossible to use. The examiners also noted that the facility's Attachment (b) and (c) of LZP 1200-2, " Classification of Noble Gas Release", forms were cluttered with extra lines and information which made them easy to misread.
- c. During the Examination Review Meeting, which was held after the Exit Meeting, the Training Department committed to ensure that a March 1987 change to Procedure LOA-FW-01 was clarified. See Attachment 1, Comment 7.11.b.
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p ATTACHMENT 1 WRITTEN EXAMINATION COMMENTS AND RESPONSES SENIOR REACTOR OPERATOR EXAMINATION ADMINISTERED APRIL 20, 1987 Docket No. 50-373 Sheet 1 of 25 COMMENT: 5.01 a. Point A shows the initial pressure spike caused by the turbine valves closing on the turbine trip. This pressure spike is due to the turbine trip, the continuation of the curve at pressure in excess of 1100 psig could be related to bypass valve failure but should not be required to state " bypass valve failure" for full credit.
Reference the UFSAR, Chapter 15, Pages 15.2-8 and 9.
This sequence shows that even when bypass works properly pressure will spike to SRV lift setpoint. So a bypass valve failure is not necessary to see a large pressure spike like this.
- b. LaSalle only has one recorder that indicates steam flow.
This steam flow indication is measured at the restricting orifices and would not show SRV lift. (SRV's are upstream of orifices),
- c. Should also accept an answer that states "Recirc. pump downshift" for full credit. (Need not state E0C-RPT downshift).
- d. Should not be required to state ". . . and water in the annulus pumped into the core" for 1/2 credit. Full credit should be granted for saying the pressure spike causes void collapse which causes the indicated level decrease,
- e. The slope of the graph at Point E shows a level increase of approximately five to six inches per minute. The LaSalle Reactor vessel volume is approximately 200 gallons per inch which would mean level increased by approximately 1100 gpm. Please reference the LaSalle CRD Hydraulics lesson plan, Page 15, which states that maximum CRD flow would be approximately 200 gpm (assuming the scram is not reset). Therefore no credit should be taken off for stating the level increase could also be partially from feedwater input. (LaSalle Feed Reg. valves are notorious for leaking past the seat).
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Attachment 1 2 l
I f. Note the E0C-RPT is a downshift of the Recirc. pumps, so the pumps will still be running in slow speed. The forced flow is more significant than natural circulation, i therefore natural circulation need not be mentioned. Full credit should be given for simply stating that jet pump flow may increase due to the SRV's opening (and reducing back pressure).
RESPONSE: a. Coment accepted. Answer key changed to only require the turbine trip shutting the turbine stop and control valves,
- b. Comment accepted. Question 5.01 Part b. is deleted.
7 Answer key changed.
l c. Coment accepted. The reason for the Recirc pump 1 downshift is not specifically asked for; therefore, will l accept "Recirc flow decrease due to Recirc pump downshift."
i Will accept alternate wording for the E0C-RPT downshift.
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- d. Comment accepted. Answer key changed to:" Reactor water level decrease due to pressure spike collapsing the voids in the core."
- e. Comment accepted. Answer key changed to include level increase due to CRD pumps and feedwater flow,
- f. Coment accepted. Answer key changed to indicate increased jet pump flow due to steam being drawn off by lif ting SRVs.
COM ENT: 5.04 a. The assumed values of k and A should be any reasonable value given by the examinee and should not limited to the values given on the answer key. Note, the value given for A on the answer key (100 sec-1) does not agree with the value given on the equation sheet (0.1 sec-1). (Answer key is wrong.)
- c. The initial drop in power is primarily due to prompt neutrons but not only due to prompt neutrons. Need not imply "only" for full credit.
RESPONSE: a. Comment accepted. Will accept reasonable assumed values for y and d. Answer key had the incorrect value for ii . Answer key changed.
- b. Coment accepted. Answer key changed to read " primary" vice "only".
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Attachment 1 3 COMMENT: 5.04 General Comment - The answer key shows a total of points allowed as 3.55 points instead of the 3.0 points specified.
The problem seems to be in the point values assigned to E (0.05 pt.) and (0.5 point).
RESPONSE: Comment accepted. Typographical errors for point values removed from the answer key.
COMMENT: 5.05 c. The answer key is wrong. Increasing the speed of the Recirc. pumps will cause the required NPSH to increase.
This is one of the major reasons for having the Recirc.
pump downshifts from cavitation interlocks (i.e., less than 30% feed flow downshift). Please reference Page 70 of Heat Transfer & Fluid Flow lesson plan. Change answer key to state " Increase".
RESPONSE: c. Comment accepted. Answer key changed to " increase."
COMMENT: 5.06 Should also give credit as an acceptable answer the " cracking of the pellet" due to exposure. See reference, Page 41 of the Core Thermal Lesson Plan.
Also the discussion on cladding creepdown should not be required. This is not mentioned in the LaSalle Lesson Plan.
RESPONSE: Comment accepted. Answer key changed to remove " clad creepdown" and add " cracking of the pellet."
COMMENT: 5.07 c. The question is confusing in that it tries to tie the in critical power to the decreasing latent heat of vaporization and changing flux profile. In reality the rise in pressure is not tied to the flux profile change.
It is true that the rising flux profile is not shifted because of increase in pressure. These are independent of each other and are discussed as independent actions in the reference given for this answer. Suggest dropping part "C" of this question and reducing the point value of the question by 0.5.
RESPONSE: Comment accepted. Part c. of Question deleted. Answer key changed.
COMMENT: 5.08 a. Full credit should be given for stating that reduced flow would cause increased voiding and an increase in indicated level. The discussion on backpressure and decreased flow from the annulus should not be required for full credit.
- b. Should also accept a discussion on reduced core flow which causes reduced neutron moderation leading to reduced power.
Attachment 1 4
- c. Jet pump flow decrease is due to the Recirc. pump trip.
This should be adequate for full credit. The discussion on increased core voiding is not the cause for the reduction in flow and should be dropped from the answer key.
- e. If student assumes 3 element control on feedwater the answer is correct. If student assumes single element control on feedwater then he only need mention the increase in level causes feed flow to decrease.
- f. The question asks why does steam flow drop at point F.
This is simply due to the drop in power which produces a drop in steam production. This is adequate for full credit. The discussion on EHC should be dropped from the answer key since this describes why steam flow recovers at some reduced level and is not what the question asks.
RESPONSE: a. Comment partially accepted. The portion of the answer pertaining to, " reduced recirc flow from the annulus,"
will be removed from the answer key. The answer key will be changed to read, " Decreased recirc flow causes increased voiding in the core causing backpressure." The increased voiding causes a backpressure, which reduces the flow through the core and causes level in the annulus region to increase.
- b. Comment partially accepted. Alternate wording for
" negative reactivity insertion," will be accepted.
- c. Comment accepted. Answer key changed to accept "Recirc pump "B" tripped."
- e. Comment partially accepted. If the e didate states the assumption that feedwater control is .n single element, will accept, "due to level increase."
- f. Comment not accepted. The pressure change is due to the reactor pover decrease and the EHC system shutting the TCV's to control pressure. If the EHC system did not throttle the TCV's, the pressure change would be a gradual drop as reactor power decreased. Due-to the EHC system action, attempting to control pressure by shutting the TCV's throttle, the steam flow rate is rapidly reduced.
COMMENT: 5.09 d. See reference, Page 252 of the Rx Physics Lesson Plan. This discussion shows how rod worth follows the Keff curve over core age. Should also accept an answer that discusses that rod worth initially decreases, then increases, then decreases again.
RESPONSE: Comment accepted. Will accept wording that indicates rod worth follows the Keff curve. Answer key changed.
Attachment 1- 5 COMMENT: 5.11 This entire question should be deleted for the following reasons:
- a. This could be interpreted to mean as rods are withdrawn the Mod temp. would increase. As mod. temp. increases the MTC becomes more negative. See reference, page 252 of the.
Rx Physics Lesson Plan.
- b. This statement is False but the answer is given as True.
Fuel temperature would be the first thing to change on this transient and as. fuel temp increases this would add negative reactivity due to Doppler, again see reference, Page 252 of the Rx Physics Lesson Plan.
- c. Late in core life at a narrow temperature band MTC can be positive. See reference, Page 122 of Rx Physics Lesson Plan. Therefore, if this was assumed to be the case, the answer would be True not False.
- d. This question.is misleading since normally, when discussing the void coefficient, it is during saturated conditions. The facility feels that the wording on this question could intentionally lead the examinee into thinking voids only affect power during saturated conditions. The facility lesson plans do not discuss-the negative reactivity addition of voids during subcooled nucleate boiling since the affect is minute compared to MTC.
RESPONSE: a. Comment accepted. Part a. of Question 5.11 is deleted.
Answer key changed.
- b. Comment partially accepted. The statement in part b., as written, is false. Answer key changed to false.
- c. Comment partially accepted. If the candidate states the assumption that it is-late in core life and moderator temperature is between 150 - 200 F, then will accept true.
- d. Comment accepted. The answer key has been changed from False to True.
COMMENT: 5.12 Should accept an answer that discusses a decreased margin to the MCPR limit for full credit.
RESPONSE: Comment partially accepted. Will accept a discussion that states the operating MCPR is established to allow sufficient margin to prevent the MCPR limit from being exceeded for the j worst case. See reference, Technical Specifications, Page B 3/4 'l 2-5 and LaSalle Lesson Plan, Thermal Hydraulics, Page 34.
Attachment 1 6 COMMENT: 5.14 a. If the affected jet pump is a fully instrumented pump, indicated flow could go up due to reverse flow through the pump. (Flow is measured by ^P and doesn't matter which way flow is going.)
RESPONSE: Comment accepted. Will accept " increase" if the candidate states the assumption that the jet pump is fully instrumented.
COMMENT: 6.01 This question is not based on the learning objectives of the lesson plan. Since this is not an objective, the answer should be limited to a more basic discussion on the design of the orifices. See reference, Page 19, Chapter 21 of the lesson plans. Should accept for full credit a discussion on restricting inventory loss during a MSL break outside the containment and limiting the ^P across the Reactor internals.
Discussion on level above the TAF before MSIV isolation should not be required for full credit.
RESPONSE: Comment partially accepted. Answer key changed to " Limit Core D/P, Loss of level, or loss of coolant (0.5) following a steam line break outside of containment (0.5)."
COMMENT: 6.03 Also accept for max. time, " prevents release of fission products to the environment". See reference, Page 20, Chapter 1 21 of the Lesson Plans.
RESPONSE: Comment partially accepted. For minimum time, will also accept;
" slowly enough to assure Nuclear System Design limits are not exceeded (fuel vessel, etc...)
For maximum time, answer key changed to read " Limit release of reactor coolant during DBA assures the fuel barrier is protected against loss of cooling (0.25) and limit release of radioactivity to the environment (0.25). See reference, Section 5.4.5 of the UFSAR.
COMMENT: 6.05 a. Credit should be given for discussion on a recent LaSalle modification concerning a 9 minute timer which will bypass the high drywell pressure signal. If you have all the other initiation signals except 1.69 psig, for 9 minutes, ADS will initiate without the 1.69 psig. See reference, write up on the recent modification.
- b. Should also accept a discussion that states Operator actions are taken during this time to recover level.
(i.e. , Time delay allows for operator to recover level to normal bands.)
Attachment 1 7 RESPONSE: a. Comment accepted. If candidate stipulates recent modification does not require high drywell pressure, Will accept "9 min. time delay without high drywell pressure."
- b. Comment not accepted. See reference, LaSalle Lesson Plan 37, Page 6, which states the ADS logical time delay is long enough to ensure HPCS has had enough time to operate, yet not so long that LPCI/LPCS Systems would be able to cool the core.
COMMENT: 6.08 This question is misleading for the following reasons:
(1) The local start pushbutton is only a normal start for the D/G, and if started using this P. B. (at the local D/G Panel) all D/G trips are still in effect.
(2) The control room start pistol grip switch is also a normal start and all D/G trips are in effect.
(3) The control room divisional start push button will start the D/G, but also starts all the ECCS pumps in that particular division. If this push button is used then the only trips in effect are 3 listed on the answer key. Note the engine start failure is also bypassed during ECCS conditions.
Please reference Chap. 47 of the lesson plans and the LOP-DG procedures you already have.
Based on this, all D/G trips should be accepted for full credit. (Any 4) Reference LOP-DG-02, Page 4 and 5 for the following list:
Trips Never bypassed:
(1) Overspeed (2) Gen. diff. current (3) Emergency stop push button.
Trips bypassed during LOCA condition:
(1) Overcurrent (with voltage restraint)
(2) Gen. ground overvoltage.
(3) Underfrequency l (4) Loss of excitation '
(5) Reverse Power (6) Failure to start (7) High cooling water temp (8) Low oil pressure RESPONSE: Comments accepted. Question 6.08 deleted from the examination.
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Attachment 1 8 COMMENT: 6.12 Should also accept, for full credit, that pressure may increase.
- If the . leak is large enough to cause the N2 back-up bottle banks to discharge then. primary containment pressure may go i up. See reference, Figure 68-8 of Chapter 68 of Lesson Plans.
RESPONSE: Comment accepted. If candidate states the assumption that
- pressure drops to the N2 backup bottle discharge point; then,
. " increase" will be accepted.
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COMMENT: 6.14 Answer No. 1 should accept a discussion that says the mechanical vacuum pump only has a 1.75 min holdup volume. See reference, Page 28, Chapter 32 of the Lesson Plans.
Answer No. 2 is incorrect; if the mechanical vacuum pump is started during normal operation, vacuum will not be lost. The i
lesson plan only states that the mechanical vacuum pump will ,
j only pull vacuum to 23" hg. An acceptable answer is to prevent detonable H2 concentrations in the system, and should be
- accepted for full credit. See reference, LGP 1-1, page 17.
RESPONSE: 1. Comment partially accepted. Will accept; The MVP only has i a 1.75 minute holdup volume and discharges directly to the j stack. (0,5) This could cause a release to the environment. (0.5)
- 2. Comment accepted. Answer key changed to accept; "The MVP i should not be operated to prevent the buildup of detonable
- H2 and 02 concentrations in the MVP (1.0),
i j COMMENT: 7.02 a. The knowledge of specific components affected in bottom head region are not operationally significant. It is important that the operator understand in general, terms that bottom head components and recirc system components t I are subject to thermal shock when a cold, idle recirc loop i is started. Credit should be granted for a general, r
- conceptual explanation of the idle loop thermal shock ;
i phenomenon and not for the key words as indicated on the i answer key.
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, RESPONSE: Comment not accepted. Technical Specifications state the specific components for each limit. See reference, l 1 Technical Specifications, Page B 3/4 4-1.
l COMMENT: 7.03 Also accept statement as ADS being one of the methods - l this is the same as saying SRV's.
j RESPONSE: Comment accepted. Will accept SRV's or ADS. Answer key l changed.
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. Attachment 1 9 COMMENT: 7.04 a. Full credit should be given for saying less than or equal to 20% of rated power, mode switch position should not be weighted as half credit of this answer.
- b. The following answers should also be acceptable for full credit. (As part of the 4)
(1) Also initial the Rod Sequence package. See reference, LAP-100-13, Attachment A.
(2) Verify the rod motion announced by the NSO is correct. See reference, LGP 1-1, Page 10.
(3) Any words that indicate the second verifier is present at the panel and double checking the rod motions.
RESPONSE: a. Comment partially accepted. The LaSalle Technical Specifications state that conditions 1 and 2 and power less than 20% of rated thermal power for the RWM to be operable. Answer key changed to accept conditions "1 and 2" vice " mode switch in startup on run."
- b. Comment partially accepted. Will add "Second verifier to initial the rod sequence package" for response 6. Response
- 4 states, " verbally verify the rod motion with the unit operator prior to any rod motion" which is the same as,
" verify the rod motion announced by the NSO is correct."
Alternate wording will be accepted. Answer key changed.
COMMENT: 7.07 b. Should accept the first part of this answer for full credit. The H2 dilution is the major concern since a loss of SJAE's will cause a loss of vacuum anyway (without any ;
backflow occurring.) '
RESPONSE: Comment not accepted. LaSalle LOP-0G-07, Page 3, Precaution #4 specifically states the information in the l answer key and the question references the LOP precaution.
COMMENT: 7.08 Full credit should be given for the discussion that states; rapid condensing of the steam in the drywell could cause the negative design pressure to be exceeded. Mention of the vacuum breakers should not be required for partial credit as indicated.
RESPONSE: Comment not accepted. The capacity of the suppression pool and drywell vacuum breakers is the bases for the development of the Drywell Spray Initiation Pressure Limit. If the vacuum breakers were designed for a higher capacity, then this would not be a problem. Reference; General Electric, Emergency Operating Procedure Fundamentals.
Attachment 1 10 COMMENT: 7.09 There is no way for the operator to monitor 2200 F cladding temperature. The LGA's are written such that adequate core cooling is assured by either maintaining level above TAF (LGA-01) or by having spray cooling via HPCS/LPCS or using steam cooling as determined by the number of SRV's open and Rx pressure (LGA-04). Any discussion that deals with the concepts of maintaining Reactor level above the fuel (by any means) or use of core spray systems should be adequate for full credit.
Additionally, FSAR Accident analyses indicate that 2/3 core height will maintain adequate core cooling.
RESPONSE: Comment accepted. The following will be accepted as an alternate answer; (all three for full credit):
(1) Water level above fuel (minimum 2/3 core height)
(2) Spray cooling with HPCS/LPCS (3) Steam cooling with know flow (Rx pressure and SRV's open)
(First correct response (0.18) others at 0.16 each)
COMMENT: 7.10 b. Should also accept health physicist. See reference, LaSalle LRP-1000-1, Page 33.
RESPONSE: Comment accepted. Answer key changed to accept " Rad Chem.
1 Supervisor" or " Health Physicist."
COMMENT: 7.11 a. Answer should accept either the 45% stated or 49 x 10 6 lbs. feedwater flow - both are stated in LOA-FW-01.
Please reference your copy of this procedure,
- b. The discussion on LOA-FW-01 mentions 2 concerns - a reactor scram or entering the core instability region (see reference, LOA-FW-01). During the exam review this particular statement in the procedure was investigated.
This is a recent revision to the procedure (March of 87) and is not clear. The intent of this change was to point out, that during single loop operation a scram could be caused by a rapid reduction in Recirculation flow. This becomes a problem in single loop, since Technical Specifications require the flow biased scram setpoint to be reduced by 5.3% to (.66W + 45.7%) T. (In two loop operation this is not a problem and should be clarified in the procedure.)
The Training Staff will follow up on this procedure change to make it more accurately indicate the problem. Credit should thus be given for either of the answers given in the Key (scram) or for stating the limit is based on staying out of the " Instability region" as defined in Technical Specifications.
Attachment 1 11 i c. Full credit should be given for discussion of the core instability region and what could occur with LPRM oscillations being out of phase so that the APRM's do not sense the power oscillations. The question is not clear that it is asking for why reduced flow could cause a scram. If the examinee starts into the train of thought on part b of this question, that the examiner is
- questioning the instability region of reduced flow aspect, then it is logical he would continue this thought on part
, c. This would result in a discussion on what the instability region is and why it is a problem.
RESPONSE: a. Comment accepted. Answer key changed to also accept "49 x 106 lbm/hr feedwater flow."
- b. Comment accepted. Answer key changed to accept either
" Reactor scram" or " Core Flow Instabilities."
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- c. Comment accepted. Answer key changed to accept correct response to part b.
General Response: The training department commitment to more accurately indicate the problem as stated in part b will be added to the examination report.
COMMENT: 7.13 Should also accept a discussion that states; when placing a controller in MANUAL, conditions should be stable and under control, or when continued operation would worsen plant conditions. Action should be taken after review and approval of the SR0 immediately available. Also, accept a discussion on when a system is withdrawn from operation continued surveillance of the relevant parameters must be maintained by a licensed R0 . . . until it is no longer needed as prescribed by Technical Specifications or the system can be restored to normal. All of these areas are discussed in LAP-1600-2. See pages 5 and 6 of LAP-1600-2 and the comments and resolution to this same question asked on a previous NRC exam of 6-3-86. Enclosed are the question and answer key from this previous exam.
RESPONSE: Comment accepted. Answer key changed to accept alternate l wording consistent with LAP-1600-2, pages 5 and 6. I COMMENT: 7.15 b. Specific breakers should not have to be mentioned. Full credit should be given for stating that the bus trips and the D/G picks up on the bus.
RESPONSE: Comment partially accepted. Will accept alternate wording that indicates normal and alternate breakers are open.
Answer key changed.
Attachment 1 12 COMMENT: 8.02 c. Also, accept " scram the reactor", this is the same as mode switch to S/D. Should not be required to say operate one loop of pool cooling for full credit. Pool cooling was alreadyinitiatedigparta.ofthisquestionwhenpool temp went above 100 F.
RESPONSE: Comment accepted. Answer key changed to accept alternate wording that is consistent with answer key.
COMMENT: 8.06 This question has a couple of problems. First, the surveillance schedule is not something that the SR0 would figure out. The surveillance program is a computer based program that automatically calculates due dates and critical dates based on the last date the surveillance was done. The surveillance coordinator (Tech Staff) is responsible for setting dates on Tech Spec related surveillances.
Second, major credit should be given if the examinee discusses that quarterly means 92 days and the Tech Specs define compliance as 1.25 times the time interval and 3.25 times 3 consecutive surveillances. The examinee may miss the exact date due to c math error and should not docked failing credit for not having the exact date. In reality, these surveillance schedules are checked by numerous people, the facility does not rely on one individual for setting critical dates.
RESPONSE: The surveillance due dates and critical dates ar erequired knowledge per the KA catalog, "NUREG-1123." Comment partially accepted. Answer key changed to accept alternate answers that indicate the candidate has an understanding of the quarterly time frame (92 days) and the limits of extension, 1.25 times normal, not to exceed 3.25 times 3 consecutive surveillances.
COMMENT: 8.12 b. Full credit should be given for stating the Shift Engineer or SR0 immediately aveilable. It is not necessary Tor the examinee to state "if not sufficient time to locate the SE" as stated in the answer key. In reality, during this type of event either the Shift Engineer, SCRE or Shift Foreman will be in the control room and will be immediately consulted when conditions of this level of urgency arise. See reference, LAP-1600-2, Page 5.
RESPONSE: Comment not accepted. LaSalle LAP-1600-2, page 5, paragraph 2 discusses deviations from Technical Specifications. LAP-1600-2, Page 5, Paragraph 3, discusses withdrawing systems from operation, not deviating from Technical Specifications. Deviations from i
j Attachment 1 13 Technical Specifications requires approval of the Station Shift Engineer, or if not available, the available SRO.
Withdrawing operating equipment from operation only requires the SR0 immediately available.
- COMMENT
- 8.13 This answer should be changed. See reference, LAP-1600-2, The procedure states that in reactor conditions 4 Page 4.
or 5, the SE can authorize more than one surveillance or
, tests on the same system if the system is not required to
! be operable. Since none 07 the qualifying conditions are given, the question should be answered true to make it
- correct.
4 i
RESPONSE: Comment accepted. As stated, the question can be True or
, False. Question deleted from the examination. Answer key j changed.
- COMMENT
- 8.16 (2) and (3) Should also accept for one of these, the NSO.
l Although not specified in the Tech Specs, the NS0's at
, LaSalle are not fire brigade trained or considered part of the fire brigade.
RESPONSE: Comment not accepted. The Technical Specifications are very clear as to who is intended to be required, SCRE and Shift Supervisor (SS), the position of SCRE or SS cannot be filled with an NSO.
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f U. S. NUCLEAR REGULATORY COMMISSION
- SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _LASALLE_1p2_____________
REACTOR TYPE: _@WR-GEg_________________
DATE ADMINISTERED: _@Zf93f29________________
EXAMINER: , CLARK z_F3_______________
CANDIDATE: _________________________
INgIRgCIJOUS_IQ_CQNDJD9IE1 Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY V_ A_ L_ U E_ _ _ T O T A L_ _ _S_ C_ -O_R E_ _ - _ _V_ A_ L_U_ E_ _ _ - _ - _ _ - _ _ - _ _ - _ _C_ A_ T E_- G_ O_ R_ Y _ _ _ _ _ _ _ _ _ _ _ - -
(
_E E ; a _^ _ _M __ __ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS p
JEEass-- _?Er??~ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND IN.STRUMENTATION 7
_2Ez99 _ _2Ef7I ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 2.20 2 ql
_Ef_SO-_ _E@IE9 _____---__- ________ O. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS :
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_?tT99__ _____-_-___ ______-_% Totals Final Grade
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All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Dusing the administration of this examination the f ollowing rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil gnly to facilitate legible reproductions.
4 Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
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- 7. Print your name in the upper right-hand corner.of the first page of each section of the answer sheet.
B. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne si de of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least thtee lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly u' sed in facility l i t eta t u,te . ,
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
I 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE i OUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examinet only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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f 18, When you complete your ex ami nati on , you shall
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- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the ex ami n a t i on and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiter. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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- 2. 5" l QUESTION 5.01 0.^^'
Attached Figure epresents a transient that could occur at a BWR.
GIVEN: (ik The Turbine Generator tripped at T=30 seconds (2) No other operator actions occur (3) Recorder speed = 1 di vi si on = 30 seconds EXPLAIN the cause(s) of the f ollowing recorder indications:
- a. Reactor Pressure INCREASE n --+-- e+mm- r-(Point A)
(point n-ga. g/
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- c. Recirc Loop A and B Flow DECREASE (Point C)
- d. Reactor Water Level DECREASE (Point D)
- e. Reactor Water Level INCREASE (Point E)
- f. Reactor Jet Pump Flow INCREASE (Point F)
DUESTION 5.02 (1.50)
MATCH the appropriate Thermal Limit (a-c), to each FAILURE MECHANISM (F1-F3) AND to each LIMITING CONDITION (L1-L3) given below:
- a. Linear Heat Generation Rate (LHGR)
- b. Average Planar Linear Heat Generdtion Rate (APLHGR)
- c. Minimum Critical Power Ratio (MCPR)
FAILURE MECHANISM LIMITING CONDITION F1. Clad melting caused by L1. Coolant transition decay heat & stored heat boiling following a LOCA F2, Clad cracking from the surface L2. Clad plastic strain becoming vapor " blanketed" less than 1%
' Clad cracking caused by L3. Maximum clad temp-F3.
high stress from pellet erature of 2200 deg F s
expansion
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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
9t__IHEgBy_gE_ NUCLE @B_EgyEB_EL9NI_gEEB@l}gN2_ELUlgS _@NQ 1 PAGE 3
- IHESDggyN@DICS QUESTION 5.03 (1.00)
STATE for which condition the reactivity coefficient contribution would be MORE NEGATIVE. EXPLAIN your choice.
- 1. Moderator void coefficient for a 1% INCREASE in void fraction at 10% void fraction in the core,
-OR-
- 2. Moderator void coefficient for a 1% INCREASE in void fraction at 70% void fraction in the core.
2 OUESTION 5.04 (3.00)
- a. After notching out a rod with the reactor critical, you notice a 100 second period. HOW MUCH reactivity was added by the rod notch?
(ASSUME BOL, SHOW ALL CALCULATIONS)
- b. After a reactor scram from power the shortest STABLE period possi bl e is -80 seconds. EXPLAIN this statement.
- c. Is the INITIAL period IMMEDIATELY following the scram shorter than -80 seconds? EXPLAIN.
QUESTION 5.05 (2.00)
- a. Define Net Positive Suction Head (NPSH). -
- b. Opening the Recirc System Flow Control Val ve (FCV) will cause the available NPSH f or the Recirc Pumps to INCREASE, DECREASE or REMAIN THE SAME?
- c. The required NPSH f or the Recirc Pumps INCREASES, DECREASES, or REMAINS THE SAME, when the Recirc Pumps are shifted from slow speed j (15 Hz) to fast speed (60 Hz)?
QUESTION 5.06 (1.25)
The THRESHOLD power below which Pellet Clad Interaction (PCI) failures do not occur is known to DECREASE with fuel burnup. STATE three (3) of the four (4) reasons for this decrease in the PCI thr e shol d.
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. - IHERdODyN8DICS le b 1 QUESTION 5.07 1-1. C G ,
During power operations, core critical power will a._________
as reactor pressure increases. Two reasons this occurs, are the b._______ latent heat of vaporization in the moderator and the corresponding c.________ of the flux profile in the core.
(SELECT THE APPROPRIATE RESPONSES)
- a. (INCREASE, DECREASE or REMAINS THE SAME)
(INCREASED, DECREASED)
=
b.
- C: . ==m dese./,2 OUESTION 5.08 (3.00)
The attached Figure 5.08 represents a transient.that could occur-at a BWR.
Given: (1) Recirc Pump B Trips (2) No operator actions occur (3) Recorder Speed = 1 division = 30 seconds EXPLAIN the cause(s) of the following recorder indications:
- a. Reactor Water Level INCREASE (Point A)
- b. Reactor Power DECREASE (Point B)
- c. Jet Pump Flow DECREASE (Point C)
- d. Reactor Pressure DECREASE (Point D)
- e. Reactor Feedwater Flow DECREASE (Point E)
- f. Reactor Steam Flow DECREASE .( Point F)
GUESTION 5.09 (2.00)
State how each of the below listed conditions will affect control rod worth. (Limit the answer to INCREASE, DECREASE, or REMAINS THE SAME.)
- a. increasing moderator temperature
- b. increasing the percent voids i c. increasing the fuel temperature
- d. increasing core age l
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l OUESTION 5.10 (1.00)
Concerning control rod worths during a reactor startup from 100% PEAK XENON versus a startup under XENON-FREE conditions, which statement is MOST correct?
- a. BOTH central and peripheral control rod worth will be LOWER regardless of core XENON conditions,
- b. CENTRAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.
- c. BOTH central and peripheral control rod worth will be the SAME regardless of core Xenon conditions,
- d. PERIPHERAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.
/.5-QUESTION 5.11 '2^^"
A reactor startup is in progress with bulk coolant temperature less than saturation temperature. Excessive rod withdrawal causes reactor power to increase on a short period.
Indicate if the following statements, concerning the above transient, are TRUE or FALSE.
, n- +s- __ ,;. ;; -mac n-n w+ + h ar m. n thn -mMorator t emp er a tur e
.cse .. lent (MTC) is less 4eyoi. .c ;-d co 121 'r:L-r ? 2 t 'I d'O
' -' d +i -; i . O stent.
- b. Fuel Temperature coefficient will be the s.econd coefficient that affects the above transient because fuel temperature is the first to increase,
- c. MTC adds positive reactivity during this transient because the fuel time constant d el ays the temperature increase of the moderator,
- d. The void coefficient will not add negative reactivity to the reactor during this transient until the bulk coolant temperature increases above the saturation temperature.
QUESTION 5.12 (1.00)
Explain why the operating limit Minimum Critical Power Ratio (MCPR) must be modified when the plant operates at less than 100% rated flow.
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QUESTION 5.13 (1.50)
Heat transfer is primarily done by three methods:
- a. Name them.
- b. Heat transfer from the fuel cladding to the Reactor Coolant, during normal conditions, is primarily done by (1) _________;
whereas, the heat transfer from the fuel to the cladding is done by (2) _________.
QUESTION 5.14 (1.25)
INDICATE if the below listed parameters will INCREASE, DECREASE or REMAIN THE SAME, if the facility experiences a " Jet Pump Riser Failure":
- e. Failed Jet Pump Flow.
' b. Core Dif f erenti al Pressure. 4
- c. Reactor (APRM) Power.
- d. Indicated Core Flow.
- e. Actual Core Flow.
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6 __EL@NI_SySIEdp_gESJgN3_CgNI69Li_@Ng_JNSI6gdENI@IJgy PAGE 7 QUESTION 6.01 (1.00)
The main steam line restrictors are designed to prevent what event end during what accident conditions?
DUESTION 6.02 (1.00)
TRUE of FALSE?
A loss of power to Reactor Protective System (RPS) Bus A will initiate a start of both trains of the Standby Gas Treatment System (SEGTS),
but a loss of power to RPS Bus B will initiate a start of only one train of the SBGTS.
QUESTION 6.03 (1.00)
What are the bases for the minimum and maximum closing time of the Main Steam Isolation Valves? (Limit your answer to the nuclear fuel considerations.)
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bz__P6@N1_gygIEdp_gEglgN PAGE 8 1 _CgNIBg61_@yg_lyglBU[ENI@llgy 1 -
QUESTION 6.04 (2.00) j Unit 2 is operating at 100% rated thermal power, with recirc in i Master Manual (f l ux automatic). An operator inadvertently INCREASES the " pressure set" on EHC by 5 psig.
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! ASSUME: 1. No further operator actions.
! 2. All other EHC control settings are normal.
- 3. Starting parameters.
l o TCV's - 100% Steam flow position.
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o BPV's - 0% Steam flow position.
a o Power - 100% rated thermal power.
! o Pressure -
1005 psig.
T NOTE: FIGURE 26-3 IS ATTACHED FOR REFERENCE.
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[ Which of the following (a, b, c, or'd) most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different i parameters and components?
INITIAL RESPONSE a b c d o TCV's CLOSE ('83*/.) CLOSE ('83%) CLOSE ('83%) NO CHANGE i o BPV's NO CHANGE OPEN (*17%) NO CHANGE OPEN ('25%) L o POWER INCREASE NO CHANGE INCREASE DECREASE l
1 o PRESSURE INCREASE NO CHANGE INCREASE DECREASE i !
i FINAL STATUS
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'100%
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>100% <100%
I i o PRESSURE >1005 PSIG 1005 PSIG >1005 PSIG <1005 PSIG t
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i t DUESTION 6.C5 (2.00) 4 Answer the following concerning the Automatic Depressurization System ,
) (ADS). '
- a. LIST the four (4) ADS initiating signals. (setpoints not required)
- b. STATE the BASES for the length of the' time delay for ADS automatic initiation.
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i i l DUESTION 6.06 (2.25)
The APRM scram function actually consists of TWO separate setpoints:
- a. Fill in the Blanks: (Assume Two Recirc. Loop Operation) i Flow Biased Scram (1) ________
x W+ (2) _______%
i Fixed Neutron Flux-High (3) __________%
! b. LIST the component (s) and the sensing point (s) of the sensor (s) i
) which measure the variable "W".
l c. While operating at 100% power, one MSlV fails shut resulting in a brief (*1 second) flux spike to 121% power. State which of the two scram setpoints mentioned above (one or both) should initiate a reactor scram? JUSTIFY your answer.
) QUESTION 6.07 (3.00) 1
- With regard to the Unit 2 RCIC System
I a. Which of the following is the only normally CLOSED valve in the
) RCIC steam supply flow path in the STAND-BY lineup?
1 (1) Steam Supply Valve (F045) i (2) Outboard Steam Isolation Valve (FOOB)
(3) Trip Throttle Valve l 1
(4) Governor Valve ,
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] b. For each of the situations listed below, STATE whether final RCIC j injection into the reactor would CONTINUE AUTOMATICALLY, REINITIATE j AUTOMATICALLY, require CONTROL ROOM Operator Action, or require LOCAL j Operator action. Assume that RCIC had automatically initiated prior i to any of the following situations.
l
} (1) The Test Pypass Valve to the CST (FO22) FAILS OPEN.
l l (2) A 125% Overspeed Trip is received due to low control oil i pressure. Control oil pressure is then returned to normal.
(3) After decreasing to 50 psig, RCIC Steam Line Pressure increases to 150 psig, j (4) After increasing to > +60 inches, Reactor Vessel Water
- Level decreases to < -60 inches.
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e __P(9NI_@y@IEMS_DEgigN1_CQNIBQL1_9ND_lNSIBUMENI911gN PAGE 10
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DUESTIO g-. -g L:CT t'* four (4) Diesel Generator shutdown signals that remain OPERABLE if the Ulm_rL Ls_ started by using the local Emergency
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Start Pushbutton-NOTE: CONSIDER DIESEL GENERATOR 2A QUESTION 6.09 (2.00)
The MINIMUM injection time for the Standby Liquid Control System (SDLC) is fifty (50) minutes and the MAXIMUM injection time is one hundred twenty five (125) minutes.
- e. STATE the BASES for the MINIMUM SBLC injection time.
- b. STATE the BASES for the MAXIMUM SBLC injection time.
QUESTION 6.10 (1.00)
Concerning Reactor Vessel Level, indicate if the following statements cre TRUE or FALSE:
- a. At Level 4 (31.5") the Recirculation cyctem Flow Control valves runback to minimum position with two (2) turbine driven feed pumps in operation,
- b. Level 5 (55.5") trips the turbines and isolates High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) to prevent damage to the turbines and prevent hydroing, the reactor vessel.
QUESTION 6.11 (1.00) 4 Concerning the 250 volt battery:
List two (2) of the three (3) times when an equalizing charge is required.
1 QUESTION 6.12 (1.50)
The Drywell Instrument Air supply (IN) devel ops a line leak in the containment. Due to the leak, containment pressure will INCREASE, DECREASE, or REMAIN THE SAME? EXPLAIN your answer.
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6 t__P69N1_@y@IEU@_DE@lGNt_CQUIBO(t_@ND_lN@l8UDENI@llgN PAGE 11 l -
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i OUESTION 6.13 (1.00)
TRUE or FALSE?
4
- a. During refueling operations a ROD BLOCK will exist if the mode switch is in STARTUP and the Service Platform Holst is LOADED.
j b. During refueling operations a ROD BLOCK will NOT exist if the mode l
) switch is in REFUEL and the Service Platform Hoist is LOADED 1
4 OUESTION 6.14 (2.00)
! EXPLAIN why the Mechanical Vacuum Pump (MVP) is not run with the mode switch in RUN. TWO (2) reasons required for full credit.
1 QUESTION 6.15 ( .75)
TRUE or FALSE?
) With an ECCS initiation signal present, the "O" Diesel Generator j local shutdown pushbutton on panel ODGO3J is depressed. The "O" Diesel Generator will stop and the output breaker will open.
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It__EBggEDUSE@_;_NQ80@bt_@@ygBU@(t_EDESQENgy_@ND PAGE 12 BBDig6g@lg@L_QgNIBQL QUESTION 7.01 (1.00)
LOP-RR-04, " Recirculation System Operatons", .autions the operator NOT TO SHUT the Reactor Recirculation Pumps' discharge and suction valves prior to SHUTTING the seal purge isolation valve.
STATE the basis for this caution.
QUESTION 7.02 (2.00)
LOP-RR-04, "Startup of the Reactor Recirculation Pumps" has two Rocirculation pump starting limitations which are listed below.
Explain WHY each of the limitations is necessary.
- c. The pump in an idle recircul ation loop shall not be started unless the temperature of the coolant within.the idle and operating recirc loop are within 50 degrees F of each other.
- b. If the temperature, of the water, in the lower head is more than 145 degrees F below vessel saturation temperature, the recirculation pump shall not be started.
QUESTION 7.03 (2.00)
LIST six (6) systems which may be used RAPIDLY DEPRESSURIZE the Reactor Pressure Vessel during _. restoration.
lel GUESTION 7.04 (2.00)
In accordance with LGP 1-1, " Unit Startup to Hot Standby", and Technical Specifications.
- c. When must the Rod Worth Minimizer (RWM) be OPERADLE?
- b. Procedure LGP l-1, " Unit Startup to Hot Standby". requires a second technically qualified operator when the RWM is INOPERADLE.
LIST four (4) of the five (5) specific duties that the second verifier must perform.
QUESTION 7.05 (3.00)
Procedure LOA-RX-Ol, " Control Room Evacuation", lists THIRTEEN (13) immediate operator actions that should be performed prior to leaving the control room.
LIST S;A (6) of the operator immediate actions performed in the control room prior to evacuation.
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Z:__E59CEQUBEg_;_NQ80@(1_@@Nggd@61_EDE69ENCY_@Np PAGE 13 609196991C@t_CQNIBQL QUESTION 7.06 (1.00)
What FOUR items are required to be logged in the Unit log when the reactor is declared CRITICAL, per LGP l-1, " Unit Startup to Hot Standby".
QUESTION 7.07 (2.00)
A CAUTION in Procedure LOP-OG-07, "Startup of the Off Gas System",
otates:
Process gas should not be introduced into the Off Gas system until the IN62-F3OO A and B (2N62-F3OO A and D) Off Gas suction control switches are placed to the AUTO position.
- c. WHAT function is enabled by the Auto position?
- b. WHAT is the BASES for this precaution?
QUESTION 7.08 (2.00)
A LOCA has occured and a high temperature steam environment exists in the drywell. EXPLAIN why the drywell sprays must NOT be initiated if conteinment conditions exist such that the UNSAFE region of figure LGA-G3, " Drywell Spray Initiation Pressure Limit", is entered.
Figure LGA-G3, "Drywell Spray Initiation Pressure Limit," is attached.
QUESTION 7.09 ( .50)
Dafine ADEOUATE CORE COOLING, as used in the Emergency Procedures.
DUCSTION 7.10 (2.00)
According to the LaSalle Radiation Procedure. LRP-!OOO-1, " Radiation Pr ot ec t i on Standard", what MANACEMENT POSITION is the minimum level of authority that may grant an individual permission to exceed the following occupational dose limits.
- c. 50 mrem / day whole body but less than 100 meem/ day,
- b. 100 mrem / day whole body but lesb than 1250 mr em/qtr.
- c. 1250 mr em/ qtr but less than 5000 mrem /yr.
- d. 5000 mrem /yr but less than 7000 mrem /yr.
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2:__ESggEggSEg_;_NO@d@6t_@gNgSU961_EDEBgENgy,@Ng PAGE 14
- 69919L991C9L_CgNI696 QUESTION 7.11 (2.50)
One immediate action for Procedure LOA-FW-01, " Loss of Feed Water Hea t er (s) ", requires core flow be decreased approximately 5 x 10 E6 lbm/hr for each 10 degrees F decrease in feed water temperature.
The immediate action also states a minimum core flow limit.
- a. LIST the minimum core flow limit per LOA-FW-01.
- b. STATE what is prevented by maintaining core above the limit.
- c. EXPLAIN the reason for the minimum core flow limit.
QUESTION 7.12 (1.50)
A condition arises that requires entry into a HIGH RADIATION AREA.
The operator will receive an estimated whole body dose of 50 mrem.
You have the following data available Candidate 1 2 3 Age 27 38 24 Exposure Week 15 mrem 35 mrem 5 mrem Quarter 2954 mrem 1207 mrem 16 mrem Life 18000 mrem 45720 mrem 29995 mrem Remarks:
NRC Form 4 on file YES YES NO Authorized 5000 mrem /yr 1250 mrem /qtr 100 mrem / day
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Each candidate is technically competent and physically capable of performing the task. Emergency limits do not apply and time constraints do not permit authorization for an administrative exposure limit increase.
EXPLAIN the reasons for accepting or rejecting each candidate.
QUCSTION 7.13 (1.00)
According to Procedure LOA-GP, " General Precautions" WHAT precautions must be taken PRIOR TO placing an ECCG system in manual?
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Iz__EBgggpuggS_:_NQQd@(3_gBNQSdg(1_gd[BQgNQY_gNQ PAGE 15 BBD196991GBL_QQNISQL
- e QUESTION 7.14 ( .50)
TRUE or FALSE 7 During accident conditions when Drywell Pressure is greater than 63 psig, operability of Safety Relief Valves (SRVs) and Containment Vent and Purge Dampers (VD) cannot be assured.
QUESTION 7.15 (2.00)
Power is being supplied from the normal. source and 4 Kv Bus 141Y ennunciator A 314 " Degraded Voltage" alarm is received.
- a. What are the immediate actions per LOA-AP-09, "4 Kv ESS Bus Degraded Voltage"7 b.- If a Loss of Coolant Accident (LOCA) signal were present when the degraded voltage alarm were received, WHAT would happen to the Bus 141Y feeder breakers?
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.PAGE 16
! Et__09MINISIB@llyE_66gC[QU6ESz_CQNQlliQNSt ,9NQ,(lM11@llgN@
i QUESTION B.01 (1.00) l What is the basis for a more restrictive chloride limit f or the reactor c ool ant system during a reactor startup?
l ',
i DUESTION B.02 (1.50)
For the suppression chamber average water temperatures listed below, WHAT ACTIONS are required by the Technical Specifications with the unit in operational condition 1 or 2?
- a. 100 degrees F
- b. 107 degrees F during RCIC testing j c. 114 degrees F l
i 1
l OUESTION 8.03 (2.00)
I i STATE which Emergency Classification is appropriate for the following I definitions.
j a. Events are in progress or have occurred which involve actual.
or potential substantial degradation of the level of safety of l the plant.
- b. Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant.
j c. Events are in progress or have occurred which involve '
j actual or imminent substantial core degradation or melting '
+ with the potential for loss of contai nment' i ntegri ty.
i d. Events are in progrpss or have occurred which i nvol ve j an actual or likely major failure of plant functions needed j for protection of the public.
a i
3 QUESTION 0.04 (1.50) .
4 Unit 1 Technical Specifications define SHUTDOWN MARGIN as...
(
l
" Shutdown Margin shall be the amount of reac t i vi ty j by which the reactor is subcritical. or pould be j
subcritical assuming all control rods fu\.ly inserted except ...."
l LIST the three conditions which complete the definition of t SHUTDOWN MARGIN.
), (.*., zTe 4 [ .,. ,r .* ,
- f. i.u ,a 3. %d W ' "
I j ,g' { ["' ' i, *s, .,5 1 (***** CATEGORY OO CONTINUED ON NEXT PAGE *****)
4 5
Oz__0901012ISOIlyE_E60CEpuSE@z_CgNp]IlgN@z_@Np_Lidll@llgNg PAGE 17 QUESTION 8.05 (1.00)
TRUE or FALSE
- a. Equipment may be tested with the Master Out-of Service card still hung. Only the Personnel Protection Card and the Out-of Service Cards must be removed.
- b. Equipment Out-of Service Procedure, " LAP-900-4," is used to place equipment out of ser vi c e that is at the LaSalle Station but under the jurisdiction of the load dispatcher.
QUESTION 8.06 (2.00)
A Technical Specification Quarterly surveillance was performed on the Unit 1 High Pressure Core Spray system on January 1, 1987, and
$s eHecA b I, d67 eJ e h " f'i d ed "; " h li d (A w e i986Wnu+ h o. leef y ecw )
Apri1 16 1987.
ckys b e.+,c <c U.w I,e967
- a. WHAT is the last MONTH and DAV the next quarterly surveillance can be performed on?
- b. What is the consequence of failing to perform a surveillance in the specified time interval?
1997 Calendar attached.
QUESTION 8.07 (1.50)
In accordance with Technical Specification, Se.c t i on 6.2, and LaSalle Administrative procedure, LAP-620-4, " Temporary Procedure Changes",
temporary changes to procedures can be made if THREE conditions are met.
WHAT are the THREE conditions?
QUESTION O.00 (1.00)
Unit 1 Technical Specifications state that, "!RRADIATED FUEL shall not be handled in or above the reactor prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown." The reason for this is thats (Choose the MOST correct answer)
- c. The minimum time it takes to setup for fuel handling.
- b. The time required decay heat to be below the specified level,
- c. Technical Specification bases for fuel handling accidents.
- d. Must allow Xenon to decay for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
,',, / ,
,s .i / * * ,,p
(: t.m o s. 4.J '
f s su ,i
(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
9___0901NigI69IlyE_E69CEpuSE@g_CQNQlligN@2_9Np_LidlI@IlgN@ PAGE 18 QUESTION 8.09 (1.00)
LAP-900-12, " Caution Card Procedure", specifies five (5) personnel who can authorize clearance of a caution card if the requestor of the CAUTION CARD cannot be located and the card no longer has a valid function. List THREE of these personnel.
QUESTION B.10 ( .50)
TRUE or FALSE?
Station Quality Assurance personnel DO NOT require the Shift Engineers permission to enter the Control Room during abnormal operations.
QUESTION B.11 (2.50)
- e. What THREE " State and Federal" agencies are always notified when a GSEP classified event occurs?
- b. How often are updates provided to these agencies.
QUESTION 8.12 )' -
g, gua v- $W $
C rot S oudd
,Under certain circumstances it is reco n that operation /
outside the Technical Specificatic ay be required.
In accordance with LAP-1600-2 'onduct of Op,e r a t i on s " :
1
- a. WHAT three (3) ations permit you to violate Technical Specifica . .s ?
- b. WHO a approve this operation?
OUESTION 8.13 -
TRUE or FALSE?
The Shift Engine i authorize more than one surveillance or test on . same system to be conducted si mul taneousl y.
QUEGTION 8.14 (1.50)
A TEMPORARY SYSTEM CHANGC is required to be implemented on a weekend, what THREE conditionc must be met to comply with the ,
requirements of LAP-240-6, " em orgry e mm,ges" %p l y ' . . -a.'.
. h. ,f $a
- I /. - v,I '. L *** x.* ': *
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
i i
' PAGE 19 Br__0001NISIgellyE_PQQCEQUBES _CgNQlligNSg_@NQ_t Lid 11911QNS i
l' 3 OUESTION 8.15 (1.00)
LAP-100-13 " Control Rod Sequence Review and Implementation," requires J
1 Unit 1 rod 10-47 to be inserted and disarmed when reactor power is less
! than 20% of Rated Thermal Power.
i a. WHAT is the malfunction with rod 10-47 which requires the
! disarming?
j b. In accordance with the Technical Specification Bases, why must rod 10-47 be inserted and disarmed when reactor power is less than 20%
j of Rated Thermal Power? '
1
.;~.'
GUESTION 8.16 (2.00) 1 7
! (Fill in the blanks) j A site Fire Brigade of at least (1) _________ members shall be maintained
- onsite at all times. The Fire Brigade shall not include the (2) ________ ,
end (3) _________ and (4) ________ other members of the operating crew, i
l
! QUESTION 8.17 (1.50) l j
State whether each of the following statements is TRUE or FALSE: .
- a. When lifting leads or removing fuses as part of an equipment l
outage in accordance with the Equipment Out of Service Procedure i a " Temporary System Change" must also be processed.
' (LAP-900-4),
- b. If a "Tempor ary System Change" is a part of an approved j l procedure which returns the system to normal upon completion, ;
it is sufficient to " log out" the device.
' c. Leads lifted to meet Technical Specification Action Requirements need only the concurrence of two individuals holding active l
, SRO licenses.
I i
4 I
l 1
i !
i Ij i
i i i l
i
.f..
h Y y y u w wh i
6 %.+ 9 e
(***** END OF CATEGORY 00 *****)
1 (************* END OF EXAMINATION ***************)
I i
D __IUEQS!_QE_NyCLE96_EQWEB_EL@N1_QCEQ@IlgNt _ELylDS _@ND t PAGE 20 T_ H_ E_ R_ M_ O_ D_ Y_ N_ A_M_ _I C_ S_
ANSWERS -- LASALLE 162 -87/04/20-CLARK, F. i
/
.D ANSWER 5.01 )
- a. Pressure increase due to the Turb'ne renerator trip shutting the
" n-d m" h"'*=- '1 2-- 'n "
Turbine Stop and Control Valves -
,ml.m .c - ::: c ;; -
.t . La-m ne,-+,e Lsmom I; eccee.,en m.
-x,7 de.J[h dv
. -:- c ; r .~ . m ww.u ng tu y,m g;.Q f.
- c. Recirc Flow decrease due to "'" "P r"*' 7ECIRC PUMP 344C% (0.5)
- d. Reactor Water Level decrease due to pressure spike collapsing the
- voids in the core ( . 1 and water 4" t .m onnulus p w . o r s. - . - t - 2.'
em-- 'O.227 e
- e. Reactor Water Level increasq due to the CRD Pump inj ctingjnto he Reactor Vessel. w et. [s.ed W A.T W l6W LO* 2I Reactor Jet Pump (QAT)[OnlT) flow increase due to steam being drawn off by t e f.
SRV's lifting em'-'n; a=+_- _
;,mm. .t.;- t' :::" ' ' ' c n e n. (0.5)
REFERENCE LaSalle Requal Lesson Plan, Thermodynamics LaSalle Requal Lesson Plan, Reactor Physics LaSalle Requal Lesson Plan, Heat Transfer and Fluid Flow LaSalle License System Descriptions, Reactor Protective System, Main Turbine and Auxiliaries, EHC Electrical, Rx Feedwater, and Main Steam LaSalle Requal Lesson Plan, Thermal Hydraulics +
LaSalle LO4-TG-06 241000K101 241000K102 241000K103 241000K123 ...(KA'S)
ANSWER 5.02 (1.50)
F1. b F2. c F3. a L1. c L2. a L3. b (0.25 each)
REFERENCC i' LaSalle Requal Lesson Plan, Thermal Hydraulics, Page 40.
20!Oo9 ell 2 293OO9K107 293OO9V100 293OO9K119 293OO9K120
...tkA'S) 1 j rh S WPf , {,,.# 4 -
[ .
(ke{.3 h b %> *# d
5 t__ISE98Y_9E_NgGLE@S_EQWEB_E68NI_gEEQ@IlgN _E6y1Q@t_@NQ g PAGE 21 IbEBU99YN@dlGQ
-87/04/20-CLARK, F.
'. ANSWERS -- LASALLE 1&2 ANSWER 5.03 (1.00)
- 2. 70% void fraction in the core (0.5)
There is a larger % change in water volume for the same increase in voids (3.45% vs 1.1 */. ) (0.5) -OR- The voids produced at 70% VF have a larger effect on core reactivity since they are in an area of higher neutron flux. (0.5)
REFERENCE LaSalle Requal Lesson Plan, Reactor Theory, Pages 136-146.
292OO4K111 ...(KA*S)
ANSWER 5.04 (3.00)
- c. T=B-p/ p p = B / (T + 1)
.1 Assume B = 0.0072 (BOL);;. 7 and = 4G4 set '.5:
p = 0. 0072/ [ ( 100) (0.1 ) + 13 = 6.545 X 10 E-4 DK/KCO.93 (1.0 pts)
- b. = Ln 2/t1/2 = 0.693/55.6 = 0.0125 sec-1 T = 1/- = 1 / -0.0125 = -00 sec.
After the initial prompt drop, power cannot, decrease faster than the longest lived delayed neutron appears. (1.0 pts)
(Calculation not required for full credit.)
%$ FAN
- c. Yes,[0.53 the initial drop in power will em45 be due to prompt neutrons.CO.53 (and could be calculated by T = 1* / p) ^ - t r)
REFERENCE LaSalle Requal Lessen Plan, Reactor Theory, Pages92-110.
292OO3K108 ...(KA*S)
At45WER 5.05 (2.00)
- a. NP5H, the difference between the suction pressure and the pressure at which boiling for the existing temperature of the fluid entering the pump. (1.0)
- b. Decrease (0,5)
- c. uw, . -. 0.5)
.Z~e C, e f.o. $(.
.[
[. b l.. A
. (, t.x' ~
f , '.+*. . ** . ~ ' ' '
.Ut__IUEQ8y_gE_NyCLESB_EQWEE E(@N1_QEEQ@llgNg_E(Qlg@t_gNQ PAGE 22 l
- IUEQUQQyN@dlQS
- ANSWERS -- LASALLE 162 -87/04/20-CLARK, F.
- REFERENCE LaSalle Requal Lesson, Heat Transfer and Fluid Flow, Pages 64-75.
293OO6K103 293OO6K110 ...(KA'S)
- ANSWER 5.06 (1.25)
- 1. Neutron embrittlement of the cladding.
- 2. Thermally induced pellet growth. ep 8/44 l 3, 1-2ra mn+5nn ns +u risaatng -! ! : <; ::;er ', d g,4/bJ M A
} 4. Fission Product (Chemical) embrittlement of the l cladding (from Cd/I).
(3 @ O.42 each) i
- REFERENCE LoSalle Requal Lesson Plan, Core Thermal Hydraulics, Pages 41 and 42.
GE BWR ACAD. SERIES, MATERIAL SCIENCE, Chapter 5, Pages 21 - 26.
293OO9K131 293OO9K132 293OO9K135 ...(KA'S) o a ANSWER 5.07 ( )
- a. decrease (0.5)
- b. decreased (0.5)
. -._i,
^
- u. ,; 04
_j (OgEg j
REFERENCE :
! LcSalle Requal Lesson Plan, Core Thermal Hydraulics, Pages 29 and 30.
l 293OO9K124 293OO9K126 ...(KA'S) a i
l l
l k
1 I
f , ,.. .**'a'*"'" '
LA a.)
f, l
1
- . i./ Q ! L.
l i
a
St__IUEQQY Q[,Nyg(EQB_EQWE6_E69NI_QEEQ811QNt_E(Q1QQt,8NQ PAGE 23
. IbEQQQQyN9digQ
. ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER 5.00 (3.00)
DAMSO
- a. Decreased Rectre Flow ' - **= anni>>e .T"' -4 increased voiding in the core causing backpressure. "MS- (Ce d Decrease in Recire Flow (0.25) which (1,np/LL reas s core voidino b.
adding negative reactivit Recirc Pump B Tripped (6 [y (0.25) 2 : . . ~ . ;
W r-g' 1 -- -
%((g7[Njf7 [N
.-.m ., ..L.
c.
- d. Decreasing Reactor Power (0.5)
- o. FWCS response go Decreasino Steam Flow (0.25) and increasing s,su.m e r es ea *F8 t p lu. A c c. U T- ht To (.tv g*t TNcAE level (O*'*5) S i AK, t.K EcsA+ter.r C.e As7 L- J5 M,c .5 7Mipe Reactor power decreasing causes less steem to be pro uced; (0.25) 1 f.
! therefore, the EHC controls Reactor Pressure (with TCV's) (0.25)*
REFERENCE LaSalle Requal Lesson Plan, Thermodynamics LaSalle Requal Lesson Plan, Reactor Physics LaSal1e Requal Lesson P1an, Heat Transfer and Fluid F1ow LeSalle Requal Lesson Plan, Thermal Hydraulics 4
202OO2K301 202OO2K302 202OO2K304 202OO2K305 202OO2K306
...(hA'S) 1 ANSWER 5.09 (2.00)
- a. Increase
- b. Decrease
- T~>/6*D E* Alt M AA U /
W
- c. Rematns the Same og ROO WCW ycgMb/ Wg RcArts
~
- d. Decrease (4 & O.5 each)
- REFERENCF LaSalle Depual Lesson Plan, Reactor Theory, Pages 176, 200-202.
292OO5VIO9 ...(KA'S)
I i
l ANSLER 5.1O (l.00)
! d.
REFCRCNCC LaSalle Requal Lesson Plan, Reactor Theory, Page 220.
- 292OODklO9 ... (KA'S) e, #
i s ,e
.. ~
t
', m. -
I
Dz__ISEQBy_QE_UyGLEQQ EQWEQ_E(ONI_QEEQQIlgyt_E(ylpgt_QUp PAGE 24
- IdEQUgg!NQUlQg ANSWERS -- LASALLE 162 -87/04/20-CLARK, F.
/. S ANSWER 5.11 C . 'O P
- c. T_ *n cs 84.fG/2/
b . 4. we /~4 t g ( 0. 5 )
- c. False (0.5)
- d. : : 11 7'AUI (O.5)
RCFERENCE LaSall e Requal Lesson Plan, Reactor Theory, Page 200 i LoSalle Requal Lesson Plan, Reactor Theory, Pages 120-176 ;
292OO4H102 292OO4K100 292OO4K114 ...(KA'S)
ANSWER 5.12 (1.00)
To protect the fuel (core) from an inadvertant core flow increase (0.5) such that the safety limit MCPR requirement is not exce ed. ..
I (0.5). 3p -3 oc.c me M N habh 44 T
- I'MSI w * %'if
- m 40 ,b*dk WC#A. M 6 a..m *M W IIE
- e. $ Pvt.
REFERENCE T' Y '#' " *4h **~^ O '"' '
LaSalle Technical Specifications, Page D 3/4 2-5.
LcSalle Requal Lesson Plan, Thermal Hydraulics, Pages 35-37. l 29;OO9K127 ...(KA'S) 5.13 (1.50) j ANSWER
- c. Conduction ,
Convection
, Radiation j (first correct answer O.34, others 0.33)
- b. (1) Convection (0.25)
] (2) Conduction (0.25)
REFERCNCE J
LaSalle Requal Lesson Pl<in, Heat Transfer and Fluid Flow, Pages 76-70.
j 293OO7N101 ...(KA'G) 1 1
- s .v o.) s u,. a u as w a
L.- IdE96Y-QE UWGLE06_E9 WEB _ELOUI'-QEEB0IlQUt_ELUIDSm._00Q PAGE 25 ISESD9910001GS
- ANSWERS -- LASALLE IL2 -87/04/20-CLARK, F.
ANSWER 5.14 (1.25)
- 6. Decrease OA .3'AleAfht(Z5 C A Albf D M Yg? Sf*AM a NLLY D3% e At4% ,
- b. Decrease . Tera 4,ees)
- c. Decrease
- d. Increase
- o. Decrease (5 0 0.25 each) i l
REFERENCE LaSalle License System Description, Chapter 2.
20200!K301 202OO1K303 202OO1K601 ...(KA'S) t 0
i t
i d
/ ',,
+, t, ....',, .. As w. 4
) , ,s
tt__EL991_SYSIEdS_gEgight_CQNI69Lt_969_ldSlQydE@l@llQG PAGE 26 ANSWCRS -- LASALLC 1&2 -07/04/20-CLARK, I.
j ANSWER 6.01 (1.00) h=1~ the t_be mn+1no the* reactor veuel water level fenm 'nitinn 7g ; :;;;;; g gg tn- mt +ne c r r r, r, 4
- , ,;; ,, ; , e ny igne r.
ne'er- ernt_.--; :0 ?? , Mfgr th; c.- s t e a m, 12. ,a .;simi an .
t.'=
4 Im s'h d oM D/p,foss op gg d4 and d so c4 A/T 3
ul : ; cic c , .?':
)) /NVfN79 Aj/ (d.$) fell.cs.,I /A% /4 2 71r*AM : : 2 l/ W
\
RCFCRCNCC 6stfM Curs spg demis+sa.s4fM7 (do$
LaSalle License System Description, Chapter.P; Page 19 28 239001M404 ...(VA'S) i ANSWER 6.02 (1.00)
Falso t
REFERENCE LaSalle License System Description, Chapter 49.
I 261000K601 ...(KA'S)
ANSWER 6.03 (1.00)
Min = ensures the pressure r i se wi thin the reactor will not cause a significant increase in fuel cladding temperatures. (Limit power increases due to pressure s tke.) (0.5) e /f.
u ot s' o sty,eL4est .tV.It.c 7%CM AJ*#dou6 s !G Md A70 th .!"5 4
e Nor 6.wC Efed'a fucL YESS L L.s C TC~~)
q Max = assures the fuel barrier is pro (tected against I ss o/ cooling if the M51V closure takes the man. time (i.e., the steaming rate greater than the feed rate and could uncover core). (0.5) # gggy
~ Dst A*Ss/SM& Odd u, do'mit /ttlust en AgM rs/t f~oo gw*~6mouss' nerencNCc sAMset u,,,, s a-isinerectro o
ne a.cs s e es n".ecoes e . **.r v r' w e W*M y'v j,A*f"* fe.t 5)
- a s
- rorCo.er) l LaSalle License Syt.tum Det.cr i p t i on . Chapter A Page 21 22900lr407 .. 0:A'S) geffg yg.iklj gg 4
i ANSWCn 6.04 (2.00) I a
RErtntNCE i LaSalle License System testription. Chapter 26, PAGCS 13-16 and r6we 26-7.
24 SOL.)OA39D 245000t100 ...IkA'S) l ,
.DVt
- p.; ; / 3.,,J: ; u,o t sv k i.
bz__EL9NJ_Sy@Igdg_pESJGN _CQNI@g62,9NQ_JNSJ@gdgNI@IJQ@ PAGE 27 1
ANSWCRS -- LASALLE 162 -87/04/20-CLARK, F.
ANSWER 6.05 (2.00)
- a. Reactor Vessel low l evel (level 1, -129") O!b E )
High Drywel1 pressure (1.69 psig) [ MOF MU Confirmatory low low vessel level (level 3, 12.5")
Low pressure core cooling pump di scharge pressur e (RHR 119 psig, LPCS 146 psig)
(4 & O.25 each)
- b. Time set long enough to ensure HPCS has had enough time to operate ( 0. 5 ) and not so long that the LPC1/LPCS would be unable to cool the core (0.5)
REFERCNCE LeSalle License System Description, Chapter 37 Page 6.
210000k403 210000K501 ...(KA'S)
ANSWER 6.06 (2.25)
- c. 1. 0.66 (0.25)
- 2. S M'. (0.25)
- 3. 110% (0.25)
- b. Recirculation loop flow transmitters. (0.5)
- c. 110% fixed scram (0.5). This is because the flow biased scram incorporates a time delay into its actuation ('6 seconds) reprosentative of the fuel thermal time constant. (0.5)
REFERCNCE LaSalle License System Description, Chapter 14, APRM, pgs. 5, 16, 53 215005E109 215005K402 215005k407 ...(KA'S)
ANSWER 6.07 (3.00)
- a. (1) (1.0)
- b. (1) (Continue) AUTOMATICALLY (2) LOCAL Operator Action (2) CONTROL ROOM Oper ator A t;cn (4) (Reinitiates AUlOMATICALLY (4 & O.5 cath)
REFERCUCE LaSalle Licenne System Description, Chapter 41 2170004701 217000A401 217000A407 ...(KA'S)
. ~~ - m ,1*
, , g p .
W
. , , /'
- 3 e
! l ; ,,' v( L i .. b .. a.a f N .'. . k I-
! 6 t__E6eNI_gy@lEDS_QEglgyt_CgglggLt,@Ng,1NgI59dEUISIlgN PAGE 20 ,
1 ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
i
?
4 i l j ANSWER (1 l
! 1) Engine Overspeed i 2) rgency Sto ushbutton i
- 3) Engi Failure
- 4) D ren Current Lockout j w 0.5 each) i j REFERENCE i LaSalle License System Description, Chapter 47.
j 264000K401 264000K407 ...(KA'S) i i ,
i 4
ANSWER 6.09 (2.00)
- e. Too rapid an injection results in improper mixing (0.5) and reactivity " chugging" (reactor power oscilliations). (0.5)
J b. Fast enough to overcome the reactivi ty due to the cooldown following a xenon peak. (1.0) i j REFERENCE
- LaSalle License System Description, Chapter 10, Page 10.
l 211000G010 211000K502 211000K503 ...(KA'S) i l
ANSWER 6.10 (1.00) ,
i i
- a. False (0.5)
- b. True (0.5) {
- REFERENCE LaSalle License System Description, Chapter 3.
, 216000K104 216000h110 216000K113 216000K'16 ...(KA'S)
ANSWER 6.11 (1.00)
,i
- 1. Guarterly
~
- 2. When the individual cell voltage is below its limit.
i 3. When large discharges have been made.
fany 2 O O.5 each)
- REFERENCE i LaSalle License System Description, Chapter 43, Pages 7 and O.
- 263OOOA101 263OOOG010 ...(KA'S) a, ! .a e ..a - % w. t s
b.__PL@UI_SYSIEUS_DEg1 gut _CQUlBg6t_@ND_lNglBydEgl@llOU PAGE 29 ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER 6.12 (1.50)
Containment pressure will remain the same. (0.5)
The drywell instrument air systems suction is from the containment; (0.5) therefore, any leak from the system inside containment will not change containment pressure. (0. 5) gg(, AC REFERENCE / 8SM ME fM c"
LaSalle License System Description, Chapter 60, Page 21.
223OOlK118 ... (hA'S)
ANSWER 6.13 (1.00) 1
- a. True (0.5)
- b. False (0,5)
REFERENCE LaSalle License System Description, Chapter 67, Page 25.
201002K402 ... (KA'S) # g ha/d W
( h6 a I' AerpY.pWW.No(U- o s ~(% A, Fh Rf
(.o ANSWER 6.14 (2.00) No
- 1. The MVP bypasses the off gas filters and discharges directly to the stack (0.5) This could cause a release to the environment (0.5).
ML : a ,- - rnnHnnene e-
..,c iw can maintaan as L ., i.,1 : . 5 :-
-the M ve sk. dw. R v . a , 6 '- o(' gfg y $ D4 &, ck
.= : .7 %nsor u'- .. saasu a .w st g not b REFERENCE p # (D.53 % 4 (>MQs bia.id V of dd W Q (C 6 LaSalle License Eystem Description, Chapter 32, Page 28 ggp had1 OMS 2 271000k101 271000K102 271000K110 ...(KA'S) c. I hI* $. A
-LGf H g pWt* (o5 ANSWER 6.15 ( .75)
False REFERENCE LaSal1o LOP-DG-09, Pages 4 and 5.
LaSalle License System Description, Chapter 47 264000E101 264000k402 ...(KA'S) 0 1 - .
. t >" -
J /. #,4b, {) l,Y* y p ,"$ j Di[
-3 ,
. _
- c m 2- _ . .-s.. . . _ _ . . _ _ _ _ .
_4 &- . . - _ -._ . . _ _ ._
Zr__E599EggggS_;_UQBM8(g_8BNQ509Lg_EMESGgEgy,@Np PAGE 30 60919LQGig@L_gQN16g(
- ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
s i
i ANSWER 7.01 (1.00) 1 To prevent hydroing the Recirculation Pump and piping to CRD System pressure. (1.0)
REFERENCE LOP-RR-04, Page 3.
LaSalle License System Description, Chapter 5.
202000G001 202OO1K404 202OO1K405 ...(KA'S) 1
, ANSWER 7.02 (2.00) 2
- a. Prevents undue stress on the vessel nozzles (0.34), bottom head regi on (0.33) and recirc loops and pumps. (0.33)
- b. Limits undue stress on the vessel. (1.0)
REFERENCE Unit 1 Technical Specifications, Page B3/4 4-1.
202OO1K103 202OO1K117 202OO1K410 ...(KA'S)
ANSWER 7.03 (2.00)
(Any six of the f ollowing, first 2 correct O . 3,4 each, others 0.33 each.)
- 1. Main Turbine Bypass valves
- 2. RCIC
- 3. NHR (steam condensing mode)-
- 5. Turbine Driven Reactor Feed Pumps
- 6. Rad Waste Boiler
- 7. Offgas preheaters B. Gland Seal Steam Reboiler '
- 9. Main Condenser Deaerating Steam
- 10. Main Steam line drains
- 11. RPV head vent
- 12. SRVs ad. A05 REFERENCE LaSalle, LGA-04, Level Restoration Procedure.
295031E301 295031E302 . . .(KA'S) b .k !I.
h'
I Zs__BBOCEDUBES_;,0QSU@bz_@@N960@L't_EdgGGENC1_ANp PAGE 31 )
8891969GICe6_C9NIBQ6 ANSWERS -- LASALLE IL2 -87/04/20-CLARK, F.
ANSWER 7.04 (2.00) g
- a. Whenever the reactor is in the b7ARTUP or RUNI mode (0.25) and less than or equal to 20% of rated power (0.25).
- b. 1. Have no other duties
- 2. Be present during all rod movement
- 3. Check that the current sequence is correct prior to any further rod movement.
4 Verbally verify the rod motion with the unit operator prior to any rod motion.
- 5. Must make a log entry into the unit operating log stating the rod pattern is verified.
(4 & O.375 each) ( m g y gg Rop SteusessC ~PA LK Aff REFERENCE [ NOM
- M8M #
LaSalle LGP 1-1 and Technical Specification, Page 3/4 1-6.
201000 GOO 1 201000 GOO 5 ...(KA'S)
ANSWER 7.05 (3.00)
- 1. Announce control room evacuation and why.
- 2. Manually scram the reactor.
3., Place the mode switch in shutdown.
- 4. Verify that power is decreasing and that all control rods are inserted.
- 5. Start the Motor Suction and Turning Gear oil Pumps
- 6. Trip the Main Turbine.
- 7. Trip the retirc pumps.
- 8. Stop reject of reactor cool ant if in progress.
- 9. Confirm no LOCA indications.
- 10. Verify Bus 141Y (241Y) and 142Y (242Y) are energized.
- 12. Verify Bus 143 energized.
- 13. Place the HPCS diesel select switch in the local position.
(6 required at 0.5 each)
REFERENCE LaSalle Procedure LDA-RX-01.
295000G010 ...(KA*S) t i 4pTC fx PW% T[
}gp.g j e L,D E
s\ v %# t
?s__EBgCEggBES_;_Ng@d@(t_@@NQBd@(t_EDEB@ENCy_9NQ PAGE 32
. BB9196991C@6_CgNIBg(
ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
1 ANSWER 7.06 (1.00)
- 1. Time
- 2. Rod Position
- 3. Coolant temperature
- 4. Reactor period (4 @ O.25 each)
(Will accept; time Rad / Chem notified -or person in Rad / Chem notified)
REFERENCE LcSalle LGP l-1, Page 14 201000G001 201000G010 ...(KA'S)
ANSWER 7.07 (2.00) r
- a. The auto position enables a trip of the suction valves on low second stage SJAE steam flow. (1.0)
- b. This prevents hydrogen from entering the Off Gas system without i
adequate dilution flow (0.5) and prevents a loss of condenser vacuum from reverse flow, should the SJ AE be l ost. (0.5)
REFERENCE LaSal l e LOP-OG-07, kSLje. j 4-271000K301 271000K404 ...(KA*S)
ANSWER 7.08 (2.00)
Spray initiation above this limit may result in a containment depressurization rate which exceeds the relief capacity of the suppression pool to drywell vacuum breakers. (1.0) As a result, the containment negative design pressure may be exceeded, leading to containment failure. (1.0)
REFERENCE General Electric EPG - Emergency Operating Procedure Fundamentals.
205000 GOO! 275000 GOO 7 ...(VA'S) i.Q
,v.
A
~'"
f t y%) q iLA A p Cw**** 'L't .e%.
e-g a
7 __P69gEgySES_;_Nggd@(z_@@Ng60@(t_EDEEGENCY_@Ng PAGE 33
. 6891969GIC86_CgNIBg6
. ANSWERS -- LAGALLE IL2 -87/04/20-CLARK, F.
gA b (egs.\ AbestM (min 2/3d,<cSa.
A s e* q coot.q m % t Pe.s./ g pc.5 g g}.% a.or.3 e A k W ia Floc ANSWER 7.09 ( .50) C A P u u ( S M dt* M
' [ f.Itsr e.md <=sps< e.tg , o%ws, o.(L eacl)
Heat removal from +he rc; t e. au i i s u i m,L Lv r estor e ana maintain the pri ' u r! M ;;m ...g i-ew u tur e cetow avv vegr ees r W.DJ.
REFERENCE General Electric EPG - Emergency Operating Procedure Fundamentals.
295031K101 ...(KA'S)
ANSWER 7.10 (2.00) a..The individual's supervisor
- b. Rad-Chem Supervisor 0 4- NM 7h E[d #d /'Ip
- c. Administrative and Support Services Asst. Superindendent (Administrative Superintendent)
- d. Station Superintendent (P1 ant Manager)
REFERENCE LaSalle LRP-1000-1 294001K103 ...(KA*S)
ANSWER 7.11 (2.50)
- 4. Ib w FO W
- a. Limit of 457. of rated core flow. [O.53 04 p h
- b. The 457. of rated flow limit is to avoid a reactor scram. [O.53 4 gb.WI'
- c. Rapid flow biased setpoint decreases and/or core flow instability (0.5) plus the APRM signal input to the thermal power monitor being time displayed, (0.5) reduces the margin to APRM scrams during core flow reductions. (O.D)
REFERENCE LaSalle LOA-FW-01.
215005K407 ...(KA'S) f 7
.t.,.
i . ..g A P T- @
. A n ,J 2 w
- .v u e t h,$,.
a y . .
^
7..__ESQCEQgBES_;_NOBd@(t_8BNO$d@(t_EDEBQENCy_@NQ PAGE 34
. 60919L991C66_CQNIBQ(
. ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER 7.12 (1.50)
Candidate #1 - Reject, would exceed the 10 CFR 20 quarterly limit Candidate #2 - Rej ect , would exceed the admi ni str ati ve quarterly limit Candidate #3 - Accepted, even though 5(N-18) exceeded the limit only applies if going to 3000 mrem /qtr. Candidate will not exceed any admin or 10 CFR 20 limi ts.
REFERENCE LaSalle LRP-1000-1 10 CFR 20 294001K103 ...(KA'S)
ANSWER 7.13 (1.00)
Do not secure or place an ECCS in manual mode unless, by at least two independent indications, (0.5)
- 1. Mi sop er at i on in the AUTOMATIC Mode is confirmed. (0.25)
- 2. Adequate core cooling is assured. (0.25)
$ g.~fgtf.Q T'E yJO(Lt>idCo. 0,b N$ s S76 AT* O W l. Y $
REFERENCE LGA-GP, page 2. 401gg 203OOOK1 209001K1 209002K1 ...(KA'S)
ANSWER 7.14 ( .50)
True.
REFERENCE LaSalle LOA-VP-03, Page 4 223OO1K10 ...(KA*S)
!. A 3
- e. =
/ CTCO f)in%[
!> i -
i
- . _ . ~ . _ _ _ , . - - ,
PAGE 35 Z __P6QQEQUBES_ _UQBd@6t_@@NQBd@bt_EdESQENQY_@ND 60919LQQ1G66_G901BQL ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER 7.15 (2.00)
- a. 1. Check system voltage not degraded [#A
- 2. If degraded voltage not due to low system voltage, shed non-essential loads.
- 3. If the degraded voltage condition cannot be corrected (within five minutes) refer to procedure for Loss of 4 Kv ESS Bus (LOA-AP-03).
(First correct response O.34, others O.33 each)
- b. Bus 141Y Normal and Alternate feeder breakers immediately open (0.5) and the emergency diesel feeder breaker closes (0.5).
REFERENCE klQggg pg4 gg ggg #
LaSalle LOA-AP-09 262OO1K406 ...(KA'S) N g . @ gg.6s1*E W E. T u MC_
262OOOGO10 p gypkBLS-i 1
M.cA.g# rFP
..r h ~r v COP '
B:__09DINigIB@IlyE_PBgCEQUBEg2_CQNQlligNg2_@NQ_L]DlI@IlgN@ PAGE 36 ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER 8.01 (1.00)
The dissolved oxygen content of the reactor coolant may be higher than during normal power operations (0.5). This increases the influence of oxygen on potential chloride stress corrosion cracking (0.5).
REFERENCE LaSalle License System Description, Chapter 9, Page 38.
LaSalle Technical Specifications, Section B 3/4 4.4, Page B 3/4 4-2 205000GOlO ...(KA*S)
ANSWER 8.02 (1.50)
- a. Initiate suppression pool cooling or restore the average temperature to less than 100 degrees F (0.5) (If temperature cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> be in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - not required for full credit.)
- b. Stop RCIC testing (0.25) and initiate cooling or restore temperature to less than 100 degrees F. (0.25).
- c. Place the mode switch in shutdown (0.25) and operate at least one RHR loop in the suppression pool cooling mode (0.25).
REFERENCE [l- M A/J4 % W 6fL) [vks- %(5-TEA 37 40tMSN
[M g Unit 1 Technical Specifications, Page 3/4 6-16. MCT /g%
223000G005 ...(KA'5)
ANSWER B.03 (2.00)
- a. Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency (4 0 0.5 each)
REFERENCE LaSalle L2P-12OO-1.
295000 GOO 2 ...(K4'C) l I
l 1
I
-Nef:-m/' h f." Yyf k.\
- [y e/. u , b u'f
- ' o m. e e I
-)
B___0901NIS16@llyE_PBQCEQUBES t _CQhQlligNSt _@NQ_(lg11911gNS PAGE 37
' ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER B.04 (1.50)
(1) Highest worth rod (0.25) fully withdrawn (0.25)
(2) Xenon free (0.5)
(3) Cold (68 deg F) (0.5)
REFERENCE LaSalle Unit 1 Technical Specifications, Page 1-6.
292OO2K110 ...(KA'S)
ANSWER B.05 (1.00)
False (0.5)
False (0.5)
REFERENCE LaSalle LAP-900-4 1Ngg-N 294001K102 ...(KA'5)
/*[//
ANSWER B.06 (2.00) 4>
- a. rtr + n s - - I" (1.0) -7^ -
'j U rli b j%
Failure to meet the time interval for a surveillance constitutes g h [v9
- b. ,j W a failure to meet the operability requirement of the LCO. (1.0) r kc[
REFERENCE LaSalle Unit Technical Specifications, Page 3/4 0-2 206000 GOO 5 ...(KA'S)
ANSWER B.07 (1.50) 1 - Intent of the original procedure is not changed. (0.5) 2 - One of the required SRO signatures must be the Shift Supervisor licensed on the affected unit. (0.5) 3 - Must be reviewed by the Station Manager within 14 days (0,5)
REFERENCE LaSalle Technical Specifications, Section 6.2 LaSalle LAP-820-4 218000G001 ...(KA'S) t h f S d ..
Di [ I;,
- y. . ' h r*ef Q
$4 (# kw b b %$
0 __090lNISIB@llyE_P69CEQUBE@t_CgNQlligN@i_@NQ,Ligi1@llgNS- PAGE 38 ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
ANSWER B.08 (1.00) a C.
REFERENCE LaSalle Technical Specifications, Page B 3/4 9-1 234000 GOO 1 234000 GOO 5 ...(KA'S) i ANSWER B.09 (1.00)
Requestor's Immedi ate Supervi sor Requestor's Department Supervisor 4
Shift Engineer Operating Engineer Production Superintendent (Any three @ .33 each)'
REFERENCE LaSalle LAP-900-12 294001K102 ...(KA'S)
ANSWER 8.10 ( .50)
TRUE .
REFERENCE LaSal l e LAP-1100-12
- 294001K105 ...(KA'S)
ANSWER 8.11 (2.50) 4
- a. Illinois Emerg. Services and Disaster Agency (ESDA)
Illinois Dept. of Nuclear Safety (DNS)
Nuclear Regulatory Commi ssion 1
(3-0 0.5 each)
- b. When conditions / classification of the. event change (0.5) i
.or at least every hour (0.5)
REFERENCE l LaSal l e LZP-1110-1 295000G002 ...(KA'S) e C' V' .
., c ;, .
g, ,
. i !.
i% 7. ( '.Q %gy y is. k *~
3 5 *-
1 i
->-,w Sz__eDd1N1QIB@IlyE_PBQQEQQBEQt_QQNQlligNQt_6NQ_LidlI@IlQNQ PAGE 39 ANSWERS -- LASALLE IL2 -87/04/20-CLARK, F.
ANSWER 8.12 (2.00)
- a. 1. Prevent injury to the public or company personnel. (0.5)
- 2. Prevent releases off-site above tech spec limits. (0.5)
- 3. Prevent damage to equipment, (0.25) if damage could adversely effect public health and safety. (0.25)
- b. Station Shift Engineer (0.5) (will accept SRO if candidate states that there is NOT sufficient time to locate the Station Shift Engineer)
REFERENCE LaSalle LAP-1600-2, Page 5 29400lK116 ...(KA*S)
ANSWER f'. 007 h re' __ s . .;;-
REFERENCE LaSalle LAP-1600-2, Page 3 201000G001 ...(KA*S)
ANSWER 8.14 (1.50) 1 - Safety evaluation completed and reviewed by two SRO's, (0.25) one of which must have an engineering degree or equivalent. (0.25) 2 - Change must be authorized by the Shift Engineer (0.5) 3 - Onsite review should be conducted promptly following the change. (normally the next day) (0.5)
REFERENCE LaSalle LAP-240-6, Page 3 223OOOGOO1 ...(KA'S)
A C J %. f.L ,.. p\/i .
'# t ; ~ ve :. o y
---r - -
8:__e90161516611yE_PBQCEQUBE@t_CQNQlllQN@t_@NQ_Lldll@IlgNE PAGE 40 ANSWERS -- LASALLE 1&2 -87/04/20-CLARK, F.
e i r ANSWER 8.15 (1.00) i
- b. To comply with the Technical Specification for a rod drop i accident. (0.5)
REFERENCE
- - LAP-100-13, " Control Rod Sequence Review and Implementation" Technical Specification Bases 201003A202 201003K401 201003K402 . ..(KA'S)
ANSWER 8.16 (2.00)
- 1. 5
- 2. Shift Supervisor
- 3. Shift Control Room Engineer (SCRE)
- 4. Two (2) 3 (4 at 0.5 pts each)
REFERENCE LaSalle Technical Speci f i cati ons [
294001K116 ...(KA'S)
ANSWER 8.17 (1.50)
! a. False
- b. True
- c. True (3 @ O.5 pts each)
I REFERENCE i LaSalle LAP-240-6 294001K102 ...(KA'S) 1 a f I
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F = (tov)/(3 x 1010) R/hr = (0.5 CE)/d (meters) 2 l 3 = .N R/hr = 6 CE/o r (f.et)
- Miscellaneous Conversions Water Parameters 1 curie = 3.7 a 1010dps l 1 gal. 8.345 lim. 1 kg = 2.21 Its 3 stu/hr i
l
- 1 as). = 3.78 liters 1 Inp = 2.54 x 10 1 ft3 = 7.48 galt - 6 1 er = 3.41 a 10 stu/hr Density = 62.4 lbs/ft 1 in = 2.54 cm tensity = 1 en/cm3 'T = 9/5*C + 32 Itest of vaporization = 970 Stu/lba *C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lba 1 STU = 778 ft-1bf 1 Ata = 14.7 psi = 29.9 in. Ng.
1 ft H 2O = 0.433 lbf/in?
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